click to view/download abstract book

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The organizing committee cordially welcomes you to the international conference on the Physics of Reactors – PHYSOR2010! The 2010 edition of this meeting is the first to take place following orders for new nuclear power plants in the U.S. The theme chosen for this conference is “Advances in Reactor Physics to Power the Nuclear Renaissance”. Indeed, the techni-cal contents and the quality of the papers submitted reflect the worldwide nuclear renaissance.

Our opening plenary session speakers are well known leaders in the nuclear industry: • Dr. Aris Candris, President and CEO, Westinghouse Electric Company, • Dan Pace, Senior Vice-President, First Energy Nuclear Operating Company, • Dr. Bertrand Barré, Scientific Advisor, AREVA, and • Dr. Larry Foulke, Director Nuclear Program, University of Pittsburgh.Their perspectives on the nuclear renaissance and the technological advancements needed in reactor technology are sure to be of interest to all of us.

We also planned two conference lunches and a dinner banquet with distinguished keynote speakers who will address other aspects of the nuclear industry. • At the Tuesday conference lunch, Dr. Dei, Chief Scientist, U.S. Department of Energy, will talk about the evolution of reactor physics in the naval reactors program. • At the Wednesday dinner banquet, Robert Buechel, Vice-President Fuel Engineering, Westinghouse Electric Company, will describe the real-world applications of computational fluid dynamics technology in improving fuel performance. • At the Thursday conference lunch, Dr. Everett Redmond II, Senior Project Manager, Nuclear Energy Institute, will present a past, present and future view of nuclear fuel management.

The conference has attracted an outstanding number of technical papers from nuclear scientists and engineers all over the world with a good mix of national distributions. The 292 papers accepted will be presented in 67 oral sessions. A broad range of topics are covered by the technical sessions ranging from the traditional subjects such as nuclear data, deterministic and stochastic transport theory, reactor analysis to more recent research concepts of multi-physics simulations, validation and verification, and radiation applications.

The technical program also includes a panel session on education on Monday afternoon. The panelists from different re-gions of the world are gathered to discuss the global education and training needs in nuclear engineering.

A total of 9 workshops are offered to the participants to enrich their knowledge in their areas of interest. Three workshops on AP1000, ABWR, and PBMR new reactor technology are offered, in addition to workshops on new physics and numerical methods.

Finally, thank you for coming and spending your valuable time with us. We appreciate your willingness to participate and make the meeting a success. We sincerely hope that you will have an excellent stay in Pittsburgh.

Mohamed Ouisloumen Baard JohansenTechnical Program Chair General Chair

PHYSOR 2010 - Advances in Reactor to Power the Nuclear RenaissanceMay 9 - 14, 2010

Foreword

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The organizing committee would like to thank:

• Our financial sponsors for their generous contributions to the success of the meeting. Please take the time to thank them for their commitment.

• Our co-sponsoring technical societies and ANS technical divisions for helping us to publicize the meeting.

• The technical program committee both local and international for soliciting papers, organizing and performing the paper reviews, and chairing the technical sessions.

• Ms. Hanna Shapira (TICS) for management of the website, program book and CD designs.

• Our plenary session speakers, banquet speaker and conference lunches speakers who were willing to devote their precious time to address this conference.

• Finally, we would personally like to thank the organizing committees, who have done an outstanding job of coordinating the program and the meeting arrangements. The hours you spent ensuring a successful meeting are very much appreciated; the conference simply would not have happened without your dedication and professional work for almost two years.

Mohamed Ouisloumen Baard JohansenTechnical Program Chair General Chair


Acknowledgement

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Organizing CommitteeHonorary Chairs Aris Candris President & CEO, Westinghouse Electric Company Bertrand Barré Scientific Advisor, AREVA Larry Foulke Director Nuclear Programs, University of PittsburghGeneral Chair Baard Johansen Westinghouse Electric CompanyGeneral Co-chairs Robb Borland FirstEnergy Nuclear Operating Company Jack Brenizer Pennsylvania State University Jim Colletti Bettis Atomic Power Laboratory Kenneth Petersen Exelon Nuclear Gene Piplica Westinghouse Electric Company Johan Slabber Pebble Bed Modular Reactor (Pty) LtdTechnical Program Chair Mohamed Ouisloumen Westinghouse Electric Company Co-Chairs Abdul Dulloo Westinghouse Electric Company Fausto Franceschini Westinghouse Electric Company Kostadin Ivanov Pennsylvania State University Moussa Mahgerefteh Exelon Nuclear Frederik Reitsma Pebble Bed Modular Reactor (Pty) Ltd Michael Zerkle Bettis Atomic Power Laboratory Advisors Jeffrey Bradfute Westinghouse Electric Company Kurshad Muftuoglu GE Hitachi Nuclear EnergyFinance Chair Melissa Hunter Westinghouse Electric Company Member George Guzik Westinghouse Electric CompanyRegistration Chair Judy Pavlecic Westinghouse Electric Company Members Melissa Hunter Westinghouse Electric Company George Guzik Westinghouse Electric Company Mohamed Ouisloumen Westinghouse Electric Company Hanna Shapira TICSIT Support Webmaster Hanna Shapira TICS Advisors Charles Fuller Westinghouse Electric Company Vincent Penkrot Westinghouse electric CompanyArrangements Chair Lori Piplica Westinghouse Electric Company Members Bill Bordogna Westinghouse Electric Company Judy Pavlecic Westinghouse Electric CompanyPublicity & Publications Chair Vefa Kucukboyaci Westinghouse Electric Company Members Cenk Guler Westinghouse Electric Company Dimitry Paramonov Westinghouse Electric Company Judy Pavlecic Westinghouse Electric CompanyGuest Program Chair Reanea Hunter Westinghouse Electric Company Member Lori Piplica Westinghouse Electric CompanyTours Program Robb Borland FirstEnergy Nuclear Operating Company Judy Pavlecic Westinghouse Electric Company Art Wharton Westinghouse Electric CompanyWorkshops Cenk Guler Westinghouse Electric Company Moussa Mahgerefteh Exelon Nuclear Mohamed Ouisloumen Westinghouse Electric Company Art Wharton Westinghouse Electric Company

Exhibits George Guzik Westinghouse Electric Company Judy Pavlecic Westinghouse Electric Company Lori Piplica Westinghouse Electric CompanyStudent Program Kostadin Ivanov The Pennsylvania State University Hans Gougar Idaho National Laboratory Melissa Hunter Westinghouse Electric Company Michael Zerkle Bettis Atomic Power Laboratory


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Technical Program Committee

1. Nuclear Data Michael Zerkle USA - Lead Michal Herman USA Yolanda Rugama France Keiichi Shibata Japan

2. Deterministic Transport Theory Mohamed Ouisloumen USA - Lead Richard Sanchez France Yousry Azmy USA Akio Yamamoto Japan Gianluca Longoni USA

3. Monte Carlo Methods Michael Zerkle USA - Lead Thomas M. Sutton USA Joachim Miss France Nam Zin Cho South Korea

4. Reactor Analysis and Optimization Fausto Franceschini USA - Lead Bojan Petrovic USA Chao, Yung-An China Rakesh Chawla Switzerland Cenk Guler USA Taek K Kim USA Tomasz Kozlowski Sweden Charles A Wemple USA Glenn E Sjoden USA

5. Reactor Design and Operation Moussa Mahgerefteh USA - Lead Hans-Dieter Berger Germany Joe Miller USA Tanguy Courau France Alireza Haghighat USA

6. Nuclear Fuel Cycle Moussa Mahgerefteh USA - Lead Pavel Tsvetkov USA Robert St Clair USA Dieter PORSCH Germany Hirosi SEKIMOTO Japan

7. Nuclear Criticality Safety Vefa N Kucukboyaci USA – Lead Charles T Rombough USA Dennis Mennerdahl Sweden Sedat Goluoglu USA

Technical Tracks Organizers:8. Transient, Safety Analysis and Thermal Hydraulics Kostadin Ivanov USA - Lead Thomas Downar USA Martin A Zimmermann Switzerland Serhat Lider USA

9. Research Reactors and Spallation Sources Frederik Reitsma South Africa – Lead David W Nigg USA Jim Kuijper The Netherlands Sooyoul Oh Korea Greg J Storr Australia

10. Integral Experiments & Facilities for Safety Research Frederik Reitsma South Africa – Lead J. Blair Briggs USA Pierre Joseph D’hondt Belgium Toshikazu Takeda Japan

11. Verification, Validation and Uncertainty Analysis Kostadin Ivanov USA - Lead Kurshad Muftuoglu USA Carlo Parisi Italy Yassin Hassan USA

12. Radiation Applications and Nuclear Safeguards Abdul Dullo USA – Lead Staffan J Svärd Sweden Frank H Ruddy USA

13. Nuclear Power and Sustainable Development Fausto Franceschini USA - Lead Kevin W Hesketh UK Sudhinder Thakur India Jess C. Gehin USA Stefano Monti Italy

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Technical Program Committee

Armenia Surik Bznuni

Australia Greg Storr

Austria Bismark Tyobeka

Bangladesh Md. Muslehuddin Sarker

Belgium Roland Carchon Silva Kalcheva Edgar Koonen Klaas van der Meer

Brazil Ricardo Barros Adimir Santos

Canada Blair Bromley Mohamed Dahmani Jim Donnelly Alain Hebert Doddy Kastanya Tao Feng Liang Guy Marleau Robert Rulko Wei Shen Mohamed Tahar Sissaoui

China Yung-An Chao

France Patrick Blaise Tanguy Courau Aldo Dall’Osso Cheikh Diop Michel Doucet Eric Dumonteil Tatiana Ivanova Emiliano Masiello Joachim Miss Stefano Panebianco Eric Royer Yolanda Rugama Richard Sanchez Simone Santandrea Enrico Sartori Denis Verrier Igor Zmijarevic

Germany Hans-Dieter Berger Ron Dagan Axel Hoefer Yann Périn Rene van Geemert

India Sudhinder Thakur Italy Sandra Dulla Carlo Parisi Carlo Petrovich Piero Ravetto

Technical Reviewers Committee:Continued

Japan Tomohiro Endo Takanori Kitada Teruhiko Kugo Hiroshi Sekimoto Keiichi Shibata Toshikazu Takeda Masahiro Tatsumi Akio Yamamoto

Mexico Juan Francois Cecilia Martin-del-Campo

Republic of Korea Nam Zin Cho Han Joo Hyung-Jin Shim

Slovenia Ivan Kodeli

South Africa Mohamed Belal Pavel Bokov Alex Matev Rian Prinsloo Frederik Reitsma

Spain Nuria García-Herranz

Sweden Hongwu Cheng Tomasz Kozlowski Dennis Mennerdahl Frigyes Reisch Staffan Svärd

Switzerland Rakesh Chawla Annalisa Manera Martin Zimmermann

The Netherlands Arjan Koning Jim Kuijper

UK Kevin Hesketh

USA Arzu Alpan Maria Avramova Yousry Azmy Bernard Bandini Kaushik Banerjee James Banfield Troy Becker John Bess Thomas Booth David Boyle Forrest Brown Damon Bryson Jae Chang Richard Chang Jianwei Chen Ren-Tai Chiang David Chichester Gregory Core

Gabriel Cuevas Vivas Virginia Dean K. Russell DePriest David Diamond Thomas Downar Clif Drumm Abdul Dulloo Jose Duo A Nichole Ellis Robert Flammang Benoit Forget Amy Francisco Massimiliano Fratoni Sedat Goluoglu Hans Gougar Reza Gouw Gerardo Grandi David Griesheimer Michal Herman Hermilo Hernandez-Noyola Hikaru Hiruta Dan Ilas Germina Ilas Boyan Ivanov Kostadin Ivanov Atul Karve Taek Kim Szelim Kong Vefa Kucukboyaci Richard Lell Jaakko Leppänen Samuel Levine Elmer Lewis Serhat Lider Gianluca Longoni Moussa Mahgerefteh Larry Mayhue Richard McKnight Walid Metwally Russell Mosteller Brian Nease Adam Nelson David Nigg Mohamed Ouisloumen Bojan Petrovic Paul Romano Charles Rombough Frank Ruddy Zeev Shayer Eugene Shwageraus Glenn Sjoden Thomas Sutton Temitope Taiwo Alberto Talamo Pavel Tsvetkov Don Williamson Emily Wolters Zhiwen Xu Serkan Yilmaz Michael Zerkle Zhaopeng Zhong

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Sponsors

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Registration

Registration is required for all attendees and presenters. Badges are required for admission to all events.

Full registration includes: All technical sessions, book of abstracts, CD proceedings, welcome reception, dinner cruise, hosted lunch, and banquet.

1-Day registration includes: All technical sessions for the registered day and a book of abstracts, and CD proceed-ings.

Student registration includes: All technical sessions, book of abstracts, CD proceedings, and welcome reception.

Guest registration includes: Welcome reception, dinner cruise, and banquet.

Conference ProceedingsConference Proceedings, in CD-ROM format, are included with the program book. Please check the vinyl pocket inside the back cover of the program book.

Guidelines for Speakers

There will be five parallel sessions. Each presentation will last 20 minutes at the maximum, followed by 5 minutes for questions. In order to allow conference participants to attend the presentation of papers in different sessions in a timely manner, we, as organizers, will request the chairpersons to comply with the time schedule rigorously. In view of the given time constraints, please make sure that your presentation fits within the prescribed 20 minute limit leaving adequate time for questions from the audience.

The laptops going to be used in sessions will have the follow-ing programs installed: • Microsoft Office Word 2007 • Microsoft Office Excel 2007 • Microsoft Office PowerPoint 2007 • Adobe Reader 9

The laptops will be using Microsoft Windows XP operating system with the Service Pack 3 installed. Default Windows font set will be available on these machines.

You will need to bring your presentation to the Session Chair at least 10 minutes before the session starts. Session Chair will upload the presentation to the laptops. It is highly encouraged to test the presentation (especially if you have animation) at the lobby area where two computers with the same settings as that in the session room will be provided.

We highly recommend that you create a PDF version of the presentation so that you can switch to the PDF in case of a problem with the PowerPoint.

You are kindly requested to attend the Speakers’ Breakfast between 06:30 and 08:00 am on the day of your presentation. Please proceed to the table bearing the number of your ses-sion. You will have the opportunity to meet your Chairpersons and your fellow presenters, and to discuss any specific details pertaining to your presentation. Please prepare, in advance, a brief introduction (no more than a few lines) for the Session Chair to introduce you to the audience.

A microphone will be used for the presentation, please make sure that you keep close to the microphone during your talk.

We also request the workshop speakers to adhere to the same rules if they plan to use our computers.


General Information

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General Events

Welcome Reception

Reflections RoomSunday May 9, 2010 - 6:30 PM

Banquet

Grand Station BallroomWednesday May 12, 2010 - 7:00 PMGuest Speaker: Robert J. Buechel, Fuel Engineering Vice President, Westinghouse Electric Company

Conference Lunches

Tuesday May 11, 2010 - 12:00 PMGuest Speaker: Dr. Donald E. Dei, Chief Physicist, Naval Reactors, USA Department of Energy

Thursday May 13, 2010 - 12:00 PMGuest Speaker: Dr. Everett Redmond II, Senior Project Manager, Used Fuel Storage and Transportation, Nuclear Energy Institute

River Cruise

Monday, May 10 2010 - Boarding Starts at 6:30 PM

The Gateway Clipper Fleet presents an experience that truly captures the personality of a great river city - Pittsburgh. Join us on a sightseeing cruise and view many points of historic and current interest on the Monongahela, Allegheny and Ohio Rivers. Come aboard and enjoy true river hospitality on a din-ner cruise from the Gateway Clipper Station Square Dock.

Technical Tours

Westinghouse Waltz Mill & the AP1000 Control Room Mock-UpFriday, May 14 2010

Westinghouse Electric Company, a group company of Toshiba Corporation, is the world’s pioneering nuclear power com-pany and is a leading supplier of nuclear plant products and technologies to utilities throughout the world. Westinghouse supplied the world’s first PWR in 1957 in Shippingport, PA. To-day, Westinghouse technology is the basis for approximately one-half of the world’s operating nuclear plants, including 60 percent of those in the United States. Tour participants will visit C Bay & D Bay, which house the Pump & Motor Mock-Ups, Steam Generator Mock-Ups, Reactor Vessel Head Mock-Up, and Refueling training equipment. These are the facilities where personnel train on providing heavy equipment services for the operating nuclear power industry feet. Lunch will be provided as a part of this tour.

The AP1000 Control Room Mock-Up is contained at the new Westinghouse Headquarters building in Cranberry Township, PA. Participants will be taken to the Westinghouse Headquarters building to view the AP1000 Control room which will be the first all-digital control room in a Westinghouse nuclear power plant. Departure from the hotel lobby is at 09:00AM.

Beaver Valley Power Plant TourFriday, May 14 2010

The Beaver Valley Power Station is operated by First Energy Nuclear Operating Company (FENOC). The site houses two Westinghouse 3-loop Pressurized Water Reactors pro-viding 1,738 MWe of power to Western Pennsylvania and Eastern Ohio. Participants will tour the Beaver Valley site and learn about the operations and maintenance activities which enable FENOC to provide clean, safe nuclear energy to the public. Lunch will be provided as a part of this tour. Departure from the hotel lobby is at 09:00AM.

Please note that according to the U.S. Department of Energy (DOE) regulations, if you are a citizen of a country listed in 10 CFR 810.8 (a) then you cannot be provided with non public information regarding the operation of a nuclear plant without DOE’s prior approval. Therefore, we cannot allow individuals from these countries to participate in our tour of the Beaver Valley Nuclear Power Station.


Special Events

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Meeting Rooms

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PHYSOR 2010 - Advances in Reactor to Power the Nuclear RenaissanceMonday May 10, 2010 - 08:30 AM - Grand Station Ballroom

Plenary Session

8:30 AM

Dr. Aris CandrisPresident & CEO, Westinghouse Electric Company LLC (USA)

The Nuclear Renaissance – Is it real?

09:20 AM

Dan PaceSenior Vice President, FirstEnergy Nuclear Operating Company (USA)

Future Advancements Needed in Reactor Technology

10:05 - 10:30 Coffee Break

10:30 AM

Bertrand BarréScientific Advisor, AREVA (France)

Key Issues in the Future of Nuclear Power

11:15

Dr. Larry FoulkeDirector of Nuclear Programs, University of Pittsburgh (USA)

Most Significant Historical Advances in Reactor Physics since CP-1

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PHYSOR 2010 - Advances in Reactor to Power the Nuclear RenaissanceMonday May 10, 2010 - 08:30 AM - Grand Station Ballroom 1

Panel

ChinaYung-An ChaoSchool of Nuclear Science and Engineering, Shanghai Jiao Tong University, Shang-hai, China and Westinghouse Electric Company, Pittsburgh, Pennsylvania, USA

In contrast to the background of the extremely rapid growth of nuclear power industry in China, the general situation of its nuclear engineering education will be presented. University and industry relation will be discussed. Problems of graduate programs will be emphasized, in particular the issues existing in the reactor physics area.

JapanAkio Yamamoto, Ph.D., Associate ProfessorDepartment of Material, Physics and Energy Engineering, Graduate School of Engi-neering, Nagoya University, Japan

After the “lost decade” in Japan, which is the period of stagnation on nuclear power, we are now facing the gap between the demand for engineers from industries and the supply from universities. In order to address this issue, several programs promoting the nuclear engineering education have been launched in Japan in the last five years. Current situation of nuclear engineering education in Japan, focusing on the reactor physics, will be briefly described in this presentation.

Global Education and Training Needs in Nuclear Engineering and Reactor Physics

Enhancing Nuclear Knowledge for New Infrastructures Development in Energy EngineeringOum Keltoum BouhelalNational School of Mineral Industry, Agdal Rabat, Morocco

If today the nuclear power option is subject to a close attention in several developing countries, decision making needs to carry up early a teaching and training federator program joining the university , the industry and the economic sector; progress brought by the actual reforms for upgrading academic background needs also to target new curricula to become familiar with nuclear engineering tools and build capacity able to investigate the viability of the nuclear power option: NPP grid integration regarding its capacity size, safety culture, public acceptance, as well as the conditions that lead to an affordable and profitable NPP project. Such challenge means to implement early coordinated packages covering multidisciplinary topics in nuclear sciences and engineering: three blocks of courses are considered that can help improve the level of nuclear knowledge delivered in the current programs and identify capacity building areas to be promoted in partnership with the industry and the economic sector.

Nuclear Engineering Education at Penn StateJack Brenizer, Jr.J. ‘Lee’ Everett Professor of Mechanical and Nuclear Engineering and Chair, Nuclear Engineering Program, Pennsylvania State University

This presentation will describe the The Pennsylvania State University’s nuclear engi-neering education at the undergraduate and graduate levels, including a brief descrip-tion of the courses pertaining to reactor physics. The Bachelor of Science in Nuclear Engineering curriculum includes a sequence of three required lecture courses and one experimental course in reactor physics. Graduate students may also enroll in the two 400-level courses, and in upper level courses in transport theory, reactor kinetics and fuel management are offered. Several of these courses are available as distance education course offerings.

The worldwide renaissance of nuclear power has provided both opportunities and challenges for education and training in both nuclear engineering in general, and in reactor physics in particular. Panelists representing different parts of the world will discuss the status and recent developments in gradu-ate and undergraduate education and training programs. The panel will report efforts to align programs with the expanding and changing needs of the nuclear industry in different parts of the globe.

Panelists Region CoveredCarla Notari ArgentinaYung-An Chao ChinaBertrand Barre FranceAkio Yamamoto JapanNam-Zin Cho Republic of KoreaJoseph DeThomas Middle EastOum Keltoum Bouhelal MoroccoFrederik Reitsma South AfricaJack Brenizer United StatesMichael Robinson (Moderator) United States

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PHYSOR 2010 - Advances in Reactor to Power the Nuclear RenaissanceMonday May 10, 2010 - 01:30 PM - Grand Station Ballroom 5

1A - Nuclear Data Measurements and EvaluationssSession Chair: Mike Zerkle (Bettis), Larry Mayhue (Westinghouse)1:30 PMSigma Web Interface for Reactor Data Applications B. Pritychenko and A.A. Sonzogni Brookhaven National Laboratory, NY, USA

We present Sigma Web interface which provides user-friendly access for online analy-sis and plotting of the evaluated and experimental nuclear reaction data stored in the ENDF-6 and EXFOR formats. The interface includes advanced browsing and search capabilities, interactive plots of cross sections, angular distributions and spec-tra, nubars, comparisons between evaluated and experimental data, computations for cross section data sets, pre-calculated integral quantities, neutron cross section un-certainties plots and visualization of covariance matrices. Sigma is publicly available at the National Nuclear Data Center website at http://www.nndc.bnl.gov/sigma.

1:55 PM235U(n,F) Prompt Fission Neutron SpectraVladimir M. Maslov, et.al,Joint Institute of Nuclear and Energy Research, 220109, Minsk-Sosny, Belarus

A new prompt fission neutron spectrum matrix for the n+235U system is proposed. The thermal neutron spectrum is this matrix describes the data by Hambsch et al. (2009). The longstanding problem of inconsistency of integral thermal data testing and differential prompt fission neutron spectra data (PFNS) seems to be solved. It was mostly due to rather poor fits of differential PFNS data in major data libraries. The older measured database is updated here using spontaneous fission neutron spectra of 252Cf(sf). That largely removes the inconsistency of older thermal PFNS data with the newest data. A phenomenological approach, developed by Kornilov et al. (1999), for the first-chance fission and extended for the emissive fission domain by Maslov et al. (2005) is normalized at Eth to predict both the PFNS average energy and PFNS shape. In the first-chance and emissive fission domain evaluated PFNS are consistent with the data by Ethvignot et al. (2005). A compiled ENDF/B-formatted file of the 235U(n,F) PFNS largely removes the inconsistencies of the evaluated differential PFNS with integral data benchmarks. Fast integral critical experiments like GODIVA or Flattop benchmarks are reproduced with the same accuracy as with the PFNS of the major data libraries. That reveals a rather delicate compensation effect, since present and previous PFNS shapes are drastically different. Thermal assembly benchmarking reveals positive biases in keff, which might be attributed to the influence of soft energy tail of the present PFNS. For some of Valduc’s LCT benchmarks biases in keff are less than 20 pcm.

2:20 PMCommissioning of the n_TOF-Ph2 FacilityS. Andriamonje, et.al,European Organization for Nuclear Research (CERN), Geneva, Switzerland

The white spectrum neutron time-of-flight facility n_TOF is operating at CERN since 2001. The neutron beam has a very high instantaneous flux and high resolution in energy. The long distance of 187 m between the spallation target and the experimen-tal area implies a favorable signal to background ratio for neutron capture and fission studies on radioactive isotopes, thus making the facility well suited for accurate cross-sections measurements. This is especially true for highly radioactive targets which are of major importance in new nuclear energy systems such as Gen-IV reactors, especially for those with a fast spectrum. Combined with state-of-the-art detectors and with advanced data acquisition systems, the innovative characteristics of the n_TOF neutron beam allows one to collect data on a variety of stable and radioactive isotopes of interest for nuclear astrophysics and for applications in advanced reactor technolo-gies. The n_TOF facility resumed operation in 2008 after the old spallation target had been replaced.

2:45 PMProposal For Improvement Of The Resonance Cross Section DescriptionN.Koyumdjieva, N.Janeva and A.A.LukyanovInstitute for Nuclear Research and Nuclear Energy - Bulgarian Academy of Sciences, Sofia, Bulgaria

The resonance region is very important for the innovative reactor technologies. The resonance part of evaluated nuclear data files is an object of a continuous re-evalu-ation on the base of old and new of high quality experimental data. In particular the investigations are focused on more accurate estimation of the neutron cross section, especially in the cross section minimum. This activity requires new models for the cross section representation, mainly in the region of experimentally unresolved reso-nances, as well as experimental data of high quality in the whole resonance region.The transmission measurements performed with a thick sample in the whole reso-nance region, in a combination with a unified resonance analysis in the both resolved and unresolved region could supply new top quality resonance data. A possible way for estimation of the optimal sample thickness in transmission experiment so to reach to the asymptotic part of the transmission function is discussed. A new statistical model with limited number of “fixed” resonances has been used for the evaluation of 232Th. This model provides a clear, unified approach in the cross section calculation, and also their functional for both the resolved and unresolved resonance regions.

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1:30 PMTENDL-2009: Comprehensive Nuclear Data Library with Covariances D. Rochman and A.J. Koning Nuclear Research and Consultancy Group NRG, Petten, The Netherlands

This paper describes the status of the new TENDL-2009 library (TALYS-based Evalu-ated Nuclear Data Library) developed at NRG. This library consists of a large set of ENDF and tabulated data les for neutrons, protons, gammas, deuterons, tritons, he-lions and alpha particles, ACE les, plot les and processed covariance les for isotopes from (19)F to (273)Ds, stable, long-lived and short-lived nuclides with half-live larger than 1 sec, completely and consistently evaluated using the TALYS-1.2 nuclear reac-tion code package. The result is a nuclear data library with mutually consistent reac-tion and covariance information for all isotopes, i.e. it is complete in nuclide coverage and reaction description. The more important a nuclide, the more effort is invested in reproduction of experimental data. In the resonance range, covariance information is given for resonance parameters. In the fast neutron range, covariance les for cross sections are obtained by means of Monte Carlo calculations. Finally, the validation of TENDL-2009 is performed for 686 criticality-safety benchmarks, showing that it outperforms other main libraries.

1:55 PMParagon Library With 4.0eV Upscattering Cut-OffVefa N. Kucukboyaci, Harish C. Huria, and Mohamed OuisloumenWestinghouse Electric Co.,Pittsburgh, Pa, USA

A cross-section library has been generated using a 4.0eV upper bound for upscatter-ing effects and ENDF/B-6.3 data for the Westinghouse lattice physics code PARA-GON. The new library has been used to analyze standard benchmarks and also to compare the measured and predicted parameters for different types of Westinghouse and CE type operating reactor cores. Test cases have confirmed that this new library provides more accurate predictions for MOX fueled plants in terms of end-of-cycle and Hot Zero Power critical boron concentrations and has relatively little impact on UO2 fueled plant calculations.

2:20 PMENDF/B-VII.0 Based Library for PARAGONHarish C. Huria, Vefa N. Kucukboyaci, and Mohamed OuisloumenWestinghouse Electric Co., Pittsburgh, Pa, USA

A new 70-group library has been generated for the Westinghouse lattice physics code PARAGON using the ENDF/B-VII.0 nuclear data files. The new library retains the major features of the current library, including the number of energy groups and the reduction in the U-238 resonance integral. The upper bound for the upscattering effects in the new library, however, has been moved to 4.0 eV from 2.1eV for better MOX fuel predictions. The new library has been used to analyze standard benchmarks and also to compare the measured and predicted parameters for different types of Westinghouse and CE type operating reactor cores. Results indicate that the new library will not impact the reactivity, power distribution and the temperature coefficient predictions over a wide range of physics design parameters, however, will improve the MOX core predictions. In other words, the ENDF/B-VI.3 and ENDF/B-VII.0 produce similar results for reactor core calculations.

2:45 PMGeneration of 69 Group Cross Section Library Based on JEF Data for TRIGA Reactor Calculations and its Vali-dation by Analyzing the Benchmark Lattices of Thermal ReactorsS.M.T. Islam(1), M.M. Sarker(2), M.J.H. Khan(2), and S.M.A. Islam(1)1 Department of Physics, Jahangirnagar University, Dhaka, Bangladesh, 2 Institute of Nuclear Science &Technology, Atomic Energy Research Establishment, Savar, Dhaka, Bangladesh

A new executable, identified as NJOY99.0 has been created to generate the 69-group cross-section library for the reactor lattice transport code WIMS. The new code incor-porates modifications in the WIMSR module of NJOY to generate the 69-group library, which will be used for TRIGA reactor calculations. The basic evaluated nuclear data file JEF-2.2 was used to generate the 69-group cross-section library in WIMS format. The results for TRX-1, TRX-2, BAPL-1, BAPL-2, and BAPL-3 benchmarks obtained by using the generated 69-group cross-section library from JEF-2.2 were analyzed. The following integral parameters were considered for the validation of the 69-group library: finite medium effective multiplication factor (keff), Ratio of epithermal to thermal 238U captures (ρ28), Ratio of epithermal to thermal 235U fission (δ25), Ratio of 238U fission to 235U fission (δ28) and Ratio of 238U captures to 235U fissions (C*).The TRX and BAPL benchmark lattices were modeled with optimized inputs, which were suggested in the final report of the WIMS Library Update Project (WLUP) Stage-I by Ravnik. The calculated results of the integral parameters of TRX and BAPL Bench-mark Lattices obtained by using the new version of code WIMSD-5B were found to be in good agreement with the experimental values. Besides, The TRX and BAPL calculation results showed that JEF-2.2 is reliable for thermal reactor calculations through benchmarking the integral parameters of TRX and validation the 69-group library, which will be used for the neutronic calculation of the TRIGA Mark-II research reactor at AERE, Savar, Dhaka, Bangladesh.

PHYSOR 2010 - Advances in Reactor to Power the Nuclear RenaissanceTuesday May 11, 2010 - 1:30 PM - Grand Station Ballroom 4

1B - Nuclear Data LibrariesSession Chairs: Dimitri Rochman (NRG), Mike Zerkle (Bettis)

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10:20 AMENDF/B-VII.0 Versus ENDF/B-VI.8 in CANDU® Calcula-tions D. Altiparmakov Atomic Energy of Canada Limited, Chalk River Laboratories, ON, Canada

Following the release of ENDF/B-VII.0, an assessment of the impact of the new evalu-ated data on CANDU-type reactor calculations has been carried out at AECL. A series of full-core MCNP simulations of ZED-2 reactor experiments was performed in order to assess the ability of the new nuclear data in the modeling of the real physical pro-cesses. Another series of calculations was carried out using an incremental replace-ment of ENDF/B-VI.8 data with the corresponding ENDF/B-VII.0 in order to identify the major contributors to the changes in the simulation results. The calculations were car-ried out with two continuous-energy multi-temperature libraries created at AECL, one based on ENDF/B-VI.8 and the other on ENDF/B-VII.0 data. This paper shows that the neutron multiplication factor values of the ZED-2 experiments calculated with the new data are well clustered around unity, while an under-prediction of about 0.4 – 0.6% is present in the results obtained with ENDF/B-VI.8 data. This finding is in good agree-ment with the conclusion of the ENDF/B-VII.0 benchmarking that accompanied the official release of the new data, which states that, for the category of thermal spectrum low-enriched compound-fuel reactors, to which the CANDU-type reactors belong, the new data improves the prediction of the neutron multiplication factor by about 0.5%. The same approach of incremental replacement of ENDF/B-VI.8 with ENDF/B-VII.0 data was applied to infinite CANDU lattice calculations at several burnup stages. The results show that the changes in the simulation results are mainly due to new data for 238U, 90Zr, 91Zr and 16O.

10:45 AMImpact of Difference of Neutron Cross Section Libraries to Isotopic Concentration of ActinidesShogo Matsuoka, Takuya ItoNuclear Fuel Industries, LTD., Muramatsu, Tokai-mura, Naka-gun, Ibaraki, Japan

Calculated isotopic compositions with several neutron cross section libraries are com-pared with radiochemical analysis results of MALIBU international program, and the impact of difference of cross section libraries to isotopic compositions are measured. Continuous energy Monte-Carlo burnup code, “MVP-BURN” is used in the computa-tional calculations to eliminate the approximations both of geometry and of evaluation of effective resonance cross section, and JENDL-3.2, JENDL/AC-2008, ENDF/B-VII.0 and JEFF-3.1 are used in the calculation as incident neutron cross section libraries. In computational calculations, the detail irradiation conditions are considered.For major actinides, uranium and plutonium, the agreement between MALIBU mea-surement results and calculated results in this work is excellent. The differences are several percent for most of them. For minor actinides, relatively large uncertainty of measurements than for major actinides are given, and for some of them, the difference between MALIBU measurement results and calculated results in this work are several tens percent. Finally, the nuclides with large impact of difference of neutron cross sec-tion libraries are discussed for major and minor actinides.

11:10 AMThermal Spectrum Critical Experiments Using ENDF/B-VII.0, JEFF 3.1.1 And JENDL 3.3Harish C. Huria, Vefa N. Kucukboyaci, and Mohamed OuisloumenWestinghouse Electric Co., Pittsburgh, Pa, USA

Continuous energy cross-sections for MCNP were generated using the nuclear data files from ENDF/B-VII.0, JEFF 3.1.1, and JENDL 3.3 using NJOY99.259. A number of thermal spectrum critical experiments from the ICSBEP handbook were analyzed using MCNP5. The selected benchmarks covered both UO2 and MOX critical assem-blies with an extensive range of enrichments and temperatures. This paper presents a comparison of the calculated eigenvalues using three different sources of basic nucle-ar data. The main objective was to bring out some salient differences in the basic data of these libraries. Besides the comparison of the eigenvalues, reaction rates for the isotopes of uranium and plutonium were also examined to determine which isotopes could be responsible for the observed differences in the three sets.

11:35 AMReactivity and Flux Calculations Using MCNP for Heavy-Water Experiments in the ZED-2 Critical Facility Using Low-Enriched Uranium Fuel BundlesF. Choy Wong, Jeremy Pencer, Blair P. Bromley, Julian Atfield and Mike ZellerAtomic Energy of Canada Limited (AECL), Chalk River Laboratories, Chalk River, On-tario, Canada

Experiments were performed in the ZED-2 (Zero Energy Deuterium) critical facility at AECL Chalk River Laboratories using a 24-cm pitch square lattice of 52 channels of heavy-water-moderated ACR-LEU and CANFLEX-LEU 43-element low-enriched uranium fuel bundles. These fuel bundles were cooled with H2O, air (void) or a check-erboard pattern of H2O and air coolants. MCNP was used to calculate the reactiv-ity and global flux distributions, which were compared with experimental measure-ments. Calculations demonstrated a small positive bias in the MCNP predictions of coolant void reactivity. MCNP predictions of axial and radial flux distributions agreed with experimental measurements within 2%. This study helps quantify the agreement between calculation and measurement for MCNP, which is one of the codes used in the analysis of the ACR-1000 reactor. Results demonstrate good agreement in code predictions for flux distributions and reactivity calculations. This is one component in the validation of MCNP. The extension of these MCNP results, compared against ZED-2 experiments, to the design and operating conditions of the ACR-1000 will require the use of extension methods.

PHYSOR 2010 - Advances in Reactor to Power the Nuclear RenaissanceWednesday May 12, 2010 - 10:20 AM - Grand Station Ballroom 4

1C - Nuclear Data Testing and Validation ISession Chairs: Jeremy Pencer (AECL), Mike Zerkle (Bettis)

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3:30 PMMeasurements and Analysis of Reactivity Worth of Rare-Earth Elements at Kyoto University Critical Assembly Kosuke Shimozato(1), Koichi Ieyama(1), Takanori Kitada(1), Hidemasa Okochi(2), Akio Yamamoto(2), Hironobu Unesaki(3) 1) Osaka University, Osaka, Japan, 2) Nagoya University, Nagoya, Japan, 3) Kyoto University Research Reactor Institute, Osaka, Japan

Critical experiments loaded with rare-earth elements (Dy, Ho, Er and Tm) were carried out and their reactivity worth were measured at the solid moderated core of Kyoto Uni-versity Critical As-sembly (KUCA). The measured reactivity worth was compared with calculation results by the continuous energy Monte-Carlo code MVP using various cross section. The calculation results of the reactivity worth agree with the measured value approximately within statistical uncertainty (one standard deviation), although the standard deviation and the experimental error are about 3% in relative value, re-spectively. The present preliminary analysis suggests the validity of the cross section data of the rare-earth elements used in the Monte-Carlo calculations.

3:55 PMValidation of the JEFF 3.1.1 Library for the Calculation of the Physical Measurements in the N4 Chooz-B1 PWR Using the Continuous Energy Monte Carlo Code TRIP-OLI-4P. Leconte and J-F. Vidal(1), N. Kerkar(2)1) CEA Cadarache, DEN/DER/SPRC/LEPh, St Paul Lez Durance, France, 2) EDF/SEPTEN, 69100 Villeurbanne, France

Since the release of the JEF-2 European File in 1992, constant improvement of this library has been ongoing. The incremental method sustained by the CEA allowed new evaluations only if JEF-2 nuclear data were found in disagreement with differential measurements and if calculation/measurement biases in targeted integral experiments were unacceptable. The release of JEFF-3.1.1 in 2008 benefit from the feedback of experimental validation, from critical facilities, fast breeder reactors, fuel cycle, critical-ity safety and burn-up credit.In this context, we propose to demonstrate the accuracy of the recommended JEFF3.1.1 library for LWR calculations through the analysis of physical measurements made in the French Chooz-B1 Pressurised Water Reactor. (PWR). Parameters of in-terest are the boron efficiency, the isothermal reactivity temperature coefficient (RTC) and the reactivity at zero power. Calculations are performed with the reference proba-bilistic code TRIPOLI-4, using a full heterogeneous description of the core and using continuous energy cross sections. Calculation results show good agreements with the experiment, confirming the ability of the JEFF3-1.1. library to accurately predict operating parameters:* 0.1 pcm/ppm on the boron efficiency,* 2 pcm/°C on the isothermal RTC,* 260 pcm on the reactivity at zero power. Moreover, the interest of this work is to demonstrate the ability to perform full core calculations with probabilistic codes for the experimental validation of nuclear data.

PHYSOR 2010 - Advances in Reactor to Power the Nuclear RenaissanceThursday May 12, 2010 - 3:30 PM - Grand Station Ballroom 4

1D - Nuclear Data Testing and Validation IISession Chair: Russell D. Mosteller (LANL) and Mike Zerkle (Bettis)

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1:30 PMCalculation of k-Eigenvalues and Multi-Group Eigen-functions Using the Hybrid “Functional Monte Carlo” Method Jinan Yang and Edward W. Larsen Department of Nuclear Engineering & Radiological Sciences, University of Michigan, Ann Arbor, MI, USA

The Functional Monte Carlo (FMC) method is a recent hybrid technique for neutron transport simulations in which Monte Carlo (MC) methods do not directly estimate the neutron flux. Instead, MC is used to estimate nonlinear functionals, which oc-cur in low-order equations that are derived without approximation from the high-order Boltzmann transport equation. (The deterministic solution of the low-order equations yields the FMC estimates of the eigenvalue and eigenfunction.) Because the low-order equations are derived without approximation from the Boltzmann transport equation, the FMC solution contains only statistical errors. Also, because the MC-estimated non-linear functionals have much smaller statistical errors than the MC-estimated fluxes, the resulting FMC solution has much smaller statistical errors than the standard MC solution. In this paper, we generalize the FMC method to account for a high-order continuous-energy transport problem and a low-order spatially-discrete multigroup dif-fusion problem. (In our previous work, the low-order problem was energy-integrated.) We present numerical results showing that the resulting multigroup FMC fluxes and eigenvalue are much more accurate than the standard MC fluxes and eigenvalue.

1:55 PMA Raviart-Thomas-Schneider Implementation of the Simplified Pn Method in 3-D Hexagonal GeometryA. Hébert École Polytechnique de Montréal, QC, Canada

We present an implementation of the Raviart-Thomas-Schneider finite element method for solving the simplified Pn (SPN) approximation of the transport equation in 3-D hexagonal geometry. The SPN method is a practical approach to solve transport problems in physical systems that can be considered as nearly plane and where the neutrons travel along directions orthogonal to planes. The Raviart-Thomas-Schneider method is based on a dual variational formulation defined over lozenge elements with a Piola transformation of the polynomial basis. An efficient ADI numerical technique was set up to solve the resulting matrix system. Our implementation of the Raviart-Thomas method is based on existing numerical solutions of the diffusion equation, available in the TRIVAC code. They use symmetric variational acceleration (SVAT) and alternating direction implicit (ADI) strategies. TRIVAC is a component of the Ver-sion4 reactor physics code distribution and is proposed as a production tool to reactor designers and operators. Version4 includes DONJON, the latest release of our full-core simulation tool. Another motivation for this work is the need of an open-source implementation of the Raviart-Thomas method, released under the Lesser General Public License (LGPL). Validation results are given for selected benchmarks in 2- and 3-D hexagonal geometry. Further investigations are required to validate the multigroup treatment of TRIVAC and the simulation capabilities of DONJON in the field of fast reactor physics.

2:20 PMA Low-order Quasidiffusion Discretization via Linear-Continuous Finite Elements on Unstructured Triangular MeshesW.A.WieselquistPaul Scherrer Institute (PSI), Villigen, Switzerland

A new finite element discretization is presented for the low-order quasidiffusion (LOQD) equations, based on a linear-continuous finite elements (LCFE) discretization of a second-order form of the LOQD equations with the scalar flux as the only unknown. The high-order (transport) equation of the proposed QD method utilizes a standard linear-discontinuous finite element (LDFE) discretization [12] plus a hybrid-collocation Galerkin-SN angular treatment [7]. The result is a completely finite element-based QD discretization. Numerical results demonstrate the expected first-order spatial conver-gence in a simple test and accuracy comparable to the pure LDFE transport discreti-zation in a more complex one. Possible improvements to the method are discussed.

2:45 PMThe Wavelet Function Expansion Method for Solving the Neutron Transport EquationHongchun Wu, Liangzhi Cao, Youqi Zheng, Weiyan YangSchool of Nuclear Science and Technology, Xi’an Jiaotong University, Shaanxi, China

A coupled method of using discrete ordinate discretization for polar angle and wavelet expansion for azimuthal angle in calculating neutron transport equation is developed. And another synthetic method of using multi-group discretization for non-resonance energy and wavelet expansion for resonance energy is also given. Based on these models, a wavelet-based neutron transport calculation code WAVTRAN and a reso-nance neutron spectrum calculation code WAVSON are encoded. Numerical results demonstrate that these methods are effective. The wavelet-based angular and energy discretization schemes are more powerful than the traditional ones for some prob-lems.


2A - Numerical Methods for Deterministic and Stochastic TransportSession Chairs: Igor Zmijarevic (CEA), Dmitriy Anistratov (NCSU)

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1:30 PMNonlinear Acceleration Methods for Even-Parity Neutron Transport W. J. Martin and C.R.E. de Oliveira(1), H. Park(2)1) Department of Chemical and Nuclear Engineering, University of New Mexico, Al-buquerque, NM. 2) Multiphysics Methods Group, Idaho National Laboratory, Idaho Falls, ID, USA

Convergence acceleration methods for even-parity transport were developed that have the potential to speed up transport calculations and provide a natural avenue for an implicitly coupled multiphysics code. An investigation was performed into the acceleration properties of the introduction of a nonlinear quasi-diffusion-like tensor in linear and nonlinear solution schemes. Using the tensor reduced matrix as a precon-ditioner for the conjugate gradients method proves highly efficient and effective. The results for the linear and nonlinear case serve as the basis for further research into the application in a full three-dimensional spherical-harmonics even-parity transport code. Once moved into the nonlinear solution scheme, the implicit coupling of the convergence accelerated transport method into codes for other physics can be done seamlessly, providing an efficient, fully implicitly coupled multiphysics code with high order transport.

1:55 PMAcceleration Technique Using Krylov Subspace Meth-ods for 2d Arbitrary Geometry Characteristics SolverHongbo Zhang, Hongchun Wu, Liangzhi CaoSchool of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an Shaanxi, China

The Generalized Minimal RESidual (GMRES) method, which is a widely-used version of Krylov subspace methods for solving large sparse non-symmetric linear systems, is adopted to accelerate the 2D arbitrary geometry characteristics solver AutoMOC. In this technique, a form of linear algebraic equation system for angular flux moments and boundary fluxes is derived as an alternative to traditional characteristics sweep (i.e. inner iteration) formalism, and then the GMRES method is implemented as an efficient linear system solver. To some degree, this is quite a favorable acceleration technique for the Method Of Characteristics (MOC) with the following advantages: firstly, it has great geometric flexibility by nature, thus, can theoretically be applied to accelerate absolutely arbitrary geometry MOC solver just like AutoMOC; secondly, it simply solves systems of linear equations and doesn’t require any additional calcula-tion, so it is convenient to implement. In our investigation, angular flux moments and incident angular fluxes from outer boundaries are involved in the linear systems and solved simultaneously, thus we can deal with boundary conditions and anisotropic scattering more accurately than linear systems which only involves scalar fluxes and incident currents. Several numerical results demonstrate that the acceleration tech-nique based on Krylov subspace methods can be applied to arbitrary geometry MOC solver successfully, and may obtain higher efficiency than the original characteristics solver does because of its spectacular effect on reducing both the number of outer iterations and the total computing time. The results could be improved by Lyusternik-Wagner extrapolation technique in some cases.

2:20 PMGeneralized Coarse-Mesh Finite Difference Acceleration for the Method of CharacteristicsXiaoming Chai and Dong Yao(1), Kan Wang(2)1) National Key Laboratory of Science and Technology on Reactor System Design Technology in China, Chengdu, China, 2) Department of Engineering PhysicsTsinghua University in China, Tsinghua University, Beijing, China

Based on generalized coarse mesh rebalance (GCMR) method, this paper proposes a new acceleration method for the method of characteristics (MOC) in unstructured meshes: the generalized coarse-mesh finite difference (GCMFD) method. The GC-MFD method, which applies equivalent width of coarse mesh to establish the finite difference discretization, can use unstructured coarse meshes composed of adjacent fine meshes to speed up the MOC iteration. The convergence property of the GCMFD method is controlled by width factor. However, the optimal width factor cannot be given a priori. Method of adjusting width factor automatically is proposed in this paper. Fi-nally, the GCMFD method is adopted in the 3-D neutron transport MOC code TCM. Numerical tests show that the GCMFD, using generalized-geometry coarse meshes, can accelerate the MOC iteration with good speedup.

2:45 PM3D Coarse Mesh NEM Embedded with 2D Fine Mesh NDOM for PWR Core AnalysisShengyi SiShanghai Nuclear Engineering Research & Design Institute, Shanghai, China

Method (NDOM). The main idea of the algorithm is to substitute the radial part of 3D NEM inner iteration, which is based on coarse mesh diffusion theory, with 2D NDOM inner iteration, which is based on transport theory. Taking advantages of both NEM and NDOM, the algorithm efficiently models the heterogeneous pin-by-pin layout in radial planes of PWR core and overcomes the challenges of computer memory and compu-tation time, which are inherent bottlenecks of 3D fine mesh discrete ordinate method (Sn). 2 prototype codes MGNSNM and MGNEM are developed, which are based on 2D multi-group NDOM and 3D multi-group NEM respectively; the final computer code HANWIND is integrated based on the above 2 prototype codes. Numerical ex-periments on benchmark problem OECD/NEA-2D C5G7MOX and a self-established benchmark problem of 2-loop PWR 3D core are summarized in the paper.


2B - Acceleration Techniques for Deterministic Transport MethodsSession Chairs: Richard Sanchez (CEA), Bill Martin (Univ. Michigan)

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10:20 AMPin Cell Discontinuity Factors in the Transient 3-D Dis-crete Ordinates Code TORT-TD A. Seubert Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH Forschungsinstitute, Garching, Germany

This paper describes the application of generalized equivalence theory to the time-de-pendent 3-D discrete ordinates neutron transport code TORT-TD. The introduction of pin cell discontinuity factors into the discrete ordinates transport equation is described by assuming a linear dependence of the homogenized neutron angular flux within a pin cell which may be discontinuous at the interfaces to adjacent cells. The homog-enized flux discontinuity at cell interfaces is expressed by pin cell discontinuity factors which in turn are determined from fuel assembly lattice calculations using HELIOS. Application of TORT-TD to the all rods in state of the PWR MOX/UO2 Core Transient Benchmark with pin cell homogenized nuclear cross sections demonstrate the poten-tial of pin cell discontinuity factors to reduce pin cell homogenization errors.

10:45 AMLinear Discontinuous (LD) Coefficients in the Slice Bal-ance Approach (SBA) Mathematical Framework for the Discrete Ordinates Code JaguarR.A. Kennedy(1), A.M. Watson and R.E. Grove(2) 1) The Ohio State University, Columbus, Ohio, USA. 2) Knolls Atomic Power Labora-tory, Bechtel Marine Propulsion Corporation, Schenectady, New York, USA

This discussion presents the derivations, approximations, and solution to one possible application of Linear Discontinuous (LD) Finite Element Methods (FEMs) on a general geometry discrete ordinates transport problem. The Linear Discontinuous Slice Bal-ance Approach (LD-SBA) coefcients are derived through the application of a set of approximations to describe the slice-averaged spatial ux moments in terms of the incoming and outgoing spatial ux moments. A test problem for comparison to accepted one-dimensional LD results is also presented.

11:10 AMUtilization of Discontinuity Factor in Integrodifferential Type of Boltzmann Transport EquationAkio YamamotoGraduate School of Engineering, Nagoya University, Japan

An approach incorporating the discontinuity factor in transport calculations based on the integrodifferential transport equation, e.g., the discrete-ordinate method, the method of characteristics, and the Monte-Carlo method, is proposed. In the pres-ent approach, effect of the discontinuity factor is incorporated by the corrections on cross sections (absorption, production and scattering cross sections are divided by the discontinuity factor) and the anisotropic scattering cross sections of odd-order are corrected with the discontinuity factor and the total cross section. Validity of the pres-ent method is confirmed through simple benchmark calculations using the method of characteristics. The present method would be a candidate of a mitigation method of errors associated with approximations, e.g., the energy condensation, the spatial homogenization, or the coarse discretization, in transport calculations.

11:35 AMAngular Flux Discontinuity Factors for Reactor Core Transport CalculationXu Hong(1), Hu Yongming(1,2), Zhou Zhiwei(1,2)1) State Nuclear Power Technology R&D Center, Beijing, China, 2) Tsinghua Univer-sity, Beijing, China

Assembly uniformity is needed to lessen computation time in the transport calculation of reactor core, specially, when the heterogeneous absorber assemblies exist in the reactor core. This paper studies the uniformity of assemblies that have heterogeneous control rod in the research reactor core. We upgrade the two dimension Sn code so as to have the function of calculating the angular discontinuous factors which can be used in other Sn code. This paper studies the Japanese Research Reactor, JRR3M, using the super cell model to calculate the angular discontinuous factors of the control rod assemblies (iterative method), and then calculates the core with them. The study shows that the Keff of the transport calculation of core with angular discontinuous fac-tors is approximately the same as that of the heterogeneous calculation of core. So do the angular flux distributions of the assemblies and the neutron importance of absorb-ers. This paper indicates the necessity and accurateness of the angular discontinuous factors in the transport theory.

PHYSOR 2010 - Advances in Reactor to Power the Nuclear RenaissanceTuesday May 11, 2010 - 10:20 AM - Grand Station Ballroom 5

2C - Applications of Discontinuity FactorsSession Chairs: Akio Yamamoto (Nagoya Univ.), Kord Smith (Studsvik)

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10:20 AMEvolution of Computer Codes for CANDU Analysis D. Altiparmakov(1), W. Shen(2), G. Marleau(3), B. Rouben(4)1) Atomic Energy of Canada Limited, Chalk River Laboratories, Station Keys, Chalk River, ON, Canada. 2) Atomic Energy of Canada Limited, Reactor Core Physics, Mississauga, ON, Canada. 3) École Polytechnique de Montréal, Montreal, Quebec, Canada. 4) 12 & 1 Consulting, Toronto, ON, Canada

From the outset of the development of the CANDU®1 reactor design, the reactor physics analysis of the core has relied on computer programs developed in Canada and international codes that have been modified and improved. Deterministic CANDU analysis is done in three stages: calculation of basic lattice properties, computation of the incremental effects of reactivity devices, and finally integration in finite-core calculations. Computer codes have evolved to account for the unique characteristics of CANDU reactors, such as the use of a cluster of fuel pins in rings, surrounded by a relatively large volume of heavy-water moderator, the three-dimensional arrangement of reactivity devices perpendicular to the fuel channels, and the on-power-refuelling feature of CANDU in channels with bi-directional coolant flow and bi-directional re-fuelling. The specific physics toolset methodologies that have evolved in the lattice code (WIMS-AECL), the reactivity-device code (DRAGON), and the finite-core code (RFSP) are reviewed. Examples of improved methods that have been developed more recently include, for instance, multi-cell capabilities in core-reflector interface models, transport calculations of the effect of interstitial reactivity devices, the time-average model for an equilibrium CANDU core and history-based local-parameter methods for snapshot core calculations. Coupled finite-core neutronic-thermalhydraulic calcula-tions in models with very detailed three-dimensional coolant-density distributions have also evolved.

10:45 AMComparison of WIMS-AECL / DRAGON / RFSP and MCNP Results With ZED-2 Measurements for Control Device Worth and Reactor KineticsJeremy Pencer, F. Choy Wong, Blair P. Bromley, Julian Atfield and Mike ZellerAtomic Energy of Canada Limited (AECL), Reactor and Radiation Physics Branch, Chalk River Laboratories, Chalk River, ON, Canada

This paper summarizes comparisons between MCNP5 and WIMS-AECL / DRAGON / RFSP calculations and experimental results obtained from the Zero Energy Deuterium (ZED-2) critical facility at AECL Chalk River Laboratories. MCNP5 and WIMS-AECL / DRAGON / RFSP were used to calculate reactivity worths for two reactivity devices, a mechanical zone controller (MZC) and shut-off rod (SOR) in a lattice similar to that of the ACR-1000®. WIMS-AECL / DRAGON / RFSP was also used to obtain kinetics pa-rameters for a transient based on a rod drop of a ZED-2 standby absorber rod (SAR). ZED-2 experiments were performed using 43-element ACR Low Enriched Uranium (ACR-LEU) fuel bundles with H2O- or air-cooled fuel bundles arranged in a 24-cm pitch square lattice. Calculations with MCNP5 gave biases in device worths that were within 0.2 mk of measured values, while WIMS-AECL / DRAGON / RFSP gave values that were within 0.3 mk of measured values. Transient analyses using the CERBERUS module within RFSP yielded a total delayed neutron fraction (β) that was within 4% of the value derived by point kinetics analysis of experimental data. The corresponding delayed photo-neutron fraction (βphoto-neutron) from CERBERUS was within 5% of that derived by point kinetics. This study has helped quantify the agreement between calculation and measurement for codes that are used in the safety analysis of the ACR-1000 reactor. Results demonstrate good agreement in code predictions.

11:10 AMApplication of 3D Collision Probability Method to VHTR Spherical GeometriesLajoie M-A., Marleau G., Martin. N., Hébert A.École Polytechnique de Montréal, Institut de Génie Nucléaire, Montréal, Québec, Canada

The analysis of VHTR fuel tends to be difficult, when using deterministic methods cur-rently employed in lattice codes, notably because of limitations on geometry represen-tation, and the stochastic positionning of spherical elements. The method proposed here and implemented in the lattice code DRAGON is to generate the positions of multi-layered spheres using random sequential addition, and to analyze the resulting geometry using a full three-dimentionnal spherical collision probability method. The preliminary validation runs are consistent with results obtained using the Monte-Carlo method. Comparisons have also been made with two double-heterogeneity models already available in DRAGON.

11:35 AMCOMET Calculations of a Postulated Stylized Small Re-gion of a CANDU CoreDingkang Zhang and Farzad Rahnema(1), Dumitru Serghiuta(2)1) Georgia Institute of Technology, Atlanta, GA USA. 2) Canadian Nuclear Safety Commission, Ottawa, Ontario, Canada

In this paper, three states of a postulated stylized 4x4 fuel channels region of a CANDU core were developed to test the coarse mesh transport code COMET. The benchmark problem consists of 192 fuel bundles in 16 fuel channels with 8 different burnups. The model includes a total of 6 adjuster rods of the same type perpendicular to the fuel channels and located on three axial locations. COMET calculations were performed based a pre-computed two-group response function library, which was generated us-ing a modified version of the Monte Carlo cod e MCNP. In these calculations, the incoming/outgoing angular currents on the bounding surfaces of each coarse mesh were expanded in terms of tensor products of Legendre polynomials. The orders of expansion in the two spatial variables and the polar and azimuth angles were 4, 4, 2 and 2, respectively. Reference solutions were computed by using MCNP5 and a two group cross section library generated with lattice depletion transport code HELIOS. The comparison showed the global eigenvalue predicated by COMET agrees very well with the MCNP reference solution (within 50 pmc). For the pin-fission density distribution of 7104 fuel pins, the maximum and average relative difference are about 2% and 0.9%, respectively.

PHYSOR 2010 - Advances in Reactor to Power the Nuclear RenaissanceThursday May 13, 2010 - 10:20 AM - Grand Station Ballroom 5

2D - Validation of Computer CodesSession Chairs: Alain Hebert (Ecole Polytechnique), Masahiro Tatsumi (NFI)

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08:00 AMDeriving a Modified Asymptotic Telegrapher’s Equation (P1) Approximation Shay I. Heizler(1,2) 1) Department of Physics, Bar-Ilan University, Ramat-Gan, ISRAEL, 2) Department of Physics, Nuclear Research Center-Negev, Beer Sheva, ISRAEL

The well known asymptotic diffusion approximation was first developed in the 50’s by Frankel and Nelson, and expanded by Case et al. and by Davison, to handle the as-ymptotic steady-state behavior. But, in time-dependent problems, the parabolic nature of the diffusion equation predicts that particles will have an infinite velocity; particles at the tail of the distribution function will show up at infinite distance from a source in time t = 0+. The classical P1 approximation (or the equivalent Telegrapher’s equa-tion) has a finite particle velocity, but with the wrong value, namely v/p3. In this work we develop a new approximation from the asymptotic solution of the time-dependent Boltzmann equation, which includes the correct eigenvalue of the asymptotic diffusion approximation and the (almost) correct time behavior (such as the particle velocity), for a general medium. The resulting scalar flux from the new approximation shows a good agreement with the exact solution of the Boltzmann equation.

08:25 AMSpatial and Time-Dependent Reactor Kinetics Method-ology Based on The Method of CharacteristicsXue Yang(1), Tatjana Jevremovic(2) 1) School of Nuclear Engineering, Purdue University, West Lafayette, IN, USA. 2) Utah Nuclear Engineering Program, The University of Utah, Salt Lake City, UT, USA

A new approach based on the method of characteristics (MOC) and Rosenbrock meth-od is developed to solve the time-dependent transport equation in one-dimensional (1D) geometry considering delayed neutron without any approximation. Using the MOC and newly developed exponential correlation, the 1D time-dependent transport equation is decomposed into a series of locally coupled ordinary differential equations (ODE). In the set of ODE, the angular flux is coupled with all other angular flux and delayed neutron precursor densities in the same zone and in the previous zone along the neutron trajectories. Rosenbrock method is used to solve the system of ODEs; it is a fourth order explicit method with automatic step size control feature developed for stiff ODEs. The FORTRAN90 numerical program was developed to “translate” this complex and detailed kinetics model in solving the time-dependent transport equation considering delayed neutrons in 1D geometry with both vacuum and reflective bound-ary conditions. The step and ramp perturbations are firstly selected to benchmark the new methodology. Based on two comprehensive benchmarks (a fast reactor with step perturbation and a thermal reactor with ramp perturbation), the developed numerical methodology was verified showing excellent accuracy and computational efficiency. This methodology is being developed for multi-dimensional arbitrary geometries. This paper presents the first results toward 3D kinetics methodology to be applied to pebble bed reactor (PBR) core.

08:50 AMA Hybrid Transport-Point Kinetic Method for the Analy-sis of Source Transients in Subcritical SystemsPaolo Picca, Barry D. Ganapol, Roberto FurfaroAerospace and Mechanical Engineering, University of Arizona, Tucson, AZ

The paper describes a novel technique for approximating the time-dependent neutron transport equation in the simulation of source-driven transients. The methodology, based on a physical intuition, adopts a multi-generational approach to improve the accuracy of lumped parameter approximations and better account for the propagation effects typical of a source switch-on. More precisely, the first few neutron generations are simulated collision by collision through the Boltzmann equation while a simpli-fied lumped parameter model is adopted for the multiply collided neutrons. Numerical simulations prove that a significant gain in accuracy is obtained compared to classical point kinetics, especially for detector signal analysis. In terms of computational costs, the technique is more time consuming than point kinetics but, at low orders, remains sensibly faster than full time dependent transport solution or quasi-static approxima-tions.

09:15 AMApplication of Artificial Neural Networks to Infer Sub-criticality Level Through Kinetic ModelsPaolo Picca, Roberto Furfaro, Barry D. Ganapol(1), Sandra Dulla, Piero Ravetto(2)1) Aerospace and Mechanical Engineering, University of Arizona, Tucson AZ. 2) Po-litecnico di Torino, Torino, Italy

The paper presents some recent advances in the study of the inverse kinetics for subcritical systems. A neural-based approach is adopted to predict the reactivity of the multiplying medium through the analysis of the reactor response to a source pulse. An artificial neural network is designed to infer the subcriticality level through the analysis of power evolution. The training set is computed using an approximate model and its performances are then tested directly on experimental measures, here simulated through a detailed space-energy kinetic model. In order to improve the accuracy of the reactivity estimation, various strategies are proposed and compared, including a multi-transient inversion and the use of different kinetic models for the training. The issue of robustness of the inversion scheme to experimental noise is also addressed.

09:40 AMAdjoint Based Error Measures for Functional Defect Cor-rection in Deterministic Neutron Transport ApplicationsS. R. Merton and C. C. PainDepartment of Earth Science and Engineering, Imperial College London, London, United Kingdom

In this paper an adjoint (or sensitivity) based error estimate is formulated which mea-sures the defect contribution of the solution variables to a bulk (volume integrated) functional in source-detector calculations. Bulk functionals can represent a wide range of parameters, such as the Keff eigenvalue in nuclear criticality assessments, or lift and drag past an interface in flow problems such as aeordynamics simulations. Ob-taining an estimate of the error within these quantities allows the required accuracy to be attained with significant reductions in computational cost. The error estimate may alternatively be used as a bound to assess reliability of a numerical scheme. The a posteriori error measure developed in the current work involves the solution of a primal and its associated adjoint problem. Both solutions are computed in-line, with the ad-joint variables used to drive a defect iteration that improves fidelity of the functional in-tegral. This allows the estimate of the functional to be improved and it is shown that the algebraic convergence with increased numerical resolution is improved dramatically with this new method. The improved functional is found to be superconvergent, due to the error within it diminishing faster than the error in the underlying discretisation scheme. In the current work, this is demonstrated using linear discontinuous finite ele-ments on the spatial domain and discrete ordinates and spherical harmonics in angle. In addition to improving functional convergence, the a posteriori approach provides error norms with which ultimately automatic mesh adaptivity methods can be applied. The method also indicates where in functional phase space the errors are large. This offers guidance on which phase variable requires adaption. However, even without mesh adaptivity the method indicates which regions of the solution domain most need to be mesh refined in order to improve functional integrals.


2E - Kinetic Transport Methods and Sensitivity Analysis ApplicationsSession Chairs: Robert Roy (Ecole Polyechnique de Montreal), Piero Ravetto (Politecnici Di Torino)

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3:30 PMOn the Order of Spherical Harmonics Approximation for the Neutron Transport Equation Haileyesus Tsige-Tamirat European Commission, Joint Research Centre, Institute for Energy Westerduinweg 3, LE Petten, Netherlands

The order of spherical harmonics approximation required to reproduce the asymptotic solution for one-dimensional neutron transport equation in plane geometry is analyzed using the eigenvalues of the characteristic equation. For isotropic scattering, it de-pends in nonlinear manner on the multiplication factor c as it varies within the physical limit below unity. It is shown that the dependence of the approximation order increases rather drastically for highly absorbing problem. On the other hand, for moderately ab-sorbing problem c 0.7 already a first order approximation is sufficient. Furthermore, an approximation order of P35 reproduces the diffusion length for any one-dimensional neutron transport problem with isotropic scattering in plane geometry.

3:55 PMAn Analytical Spatial Reconstruction Scheme for the Spectral Diamond – Spectral Green’s Function – Con-stant Nodal Method for One-Speed X,Y-Geometry Sn Eigenvalue ProblemsWelton Alves Menezes, Hermes Alves Filho and Ricardo C. Barros(1), Dany S. Dominguez(2) 1) Departamento de Modelagem Computacional, Instituto Politécnico, Universidade do Estado do Rio de Janeiro, RJ, Brazil. 2) Laboratório de Computação Científica, Universidade Estadual de Santa Cruz, BA, Brazil

In this paper the X,Y-geometry spectral diamond – spectral Green’s function - constant nodal (SD-SGF-CN) method is used to determine the one-speed node-edge average angular fluxes in neutron fission reacting media. This hybrid spectral nodal method uses the spectral diamond (SD) auxiliary equations for the multiplying regions and the spectral Green’s function (SGF) auxiliary equations for non-multiplying regions of the domain. Moreover, we consider constant approximations for the transverse-leakage terms in the transverse integrated SN nodal equations. We solve the SD-SGF-CN equations using the one-node block inversion (NBI) iterative scheme, which uses the most recent estimates available for the node-entering fluxes to evaluate the node-exiting fluxes in the directions that constitute the incoming fluxes for the adjacent node. Using these results, we offer an algorithm for analytical reconstruction of the coarse-mesh nodal solution within each spatial node, as localized numerical solutions are not generated by usual accurate nodal methods. Numerical results are presented to illustrate the accuracy of the offered algorithm.

4:20 PMAnalytic Basis Function Expansion Nodal Method for Neutron Diffusion Equations in Triangular GeometryKunpeng Wang, Hongchun Wu, Liangzhi CaoSchool of Nuclear Science and Technology, Xi’an Jiaotong University, China

An analytic basis function expansion nodal method for directly solving the two-group neutron diffusion equation in the triangular geometry is proposed in the present paper. In this method, the distribution of neutron flux is expanded by a set of analytic basis functions. The diffusion equation is satisfied at any point in a triangular node for each group assuming that the flux within a node is flat. No transverse integration is needed. To improve the nodal coupling relations and computation accuracy, nodes are coupled with each other fulfilling both the zero- and first-order partial neutron current moments across all the three interface of the triangle mesh at the same time. Coordinate con-version is used to transform arbitrary triangle into regular triangle in order to simplify the derivation. A new sweeping scheme is developed for the triangular mesh and the response matrix technique was used to solve the nodal diffusion equation iteratively. Based on the proposed model, the code ABFEM-T is developed with the L-W extrapo-lation for acceleration. Validation of code for accuracy and efficiency are carried out by calculating both rectangular and hexagonal assembly benchmark problems. Numeri-cal results for the series of benchmark problems show that both the multiplication fac-tor and nodal power distribution are predicted accurately. Therefore this method can be used for solving neutron diffusion problems in complex unstructured geometry.

4:45 PMA 2D Triangular NODAL-SP3 Method for Solving Neu-tron Transport EquationYunzhao Li, Hongchun Wu, Liangzhi Cao, Qichang ChenSchool of Nuclear Science and Technology, Xi’an Jiaotong University, Shaanxi, P.R. China

In this triangular nodal-SP3 method, neutron transport equation is transformed by employing an isotropic SP3 method into two coupled equations that are both in the same mathematic form with diffusion equation, and then a triangular nodal method is proposed to solve the two coupled equations. In the triangular nodal method, adjacent nodes are coupled through partial currents. Since transverse integral technique which is widely used in regular nodal method can not be used in triangular geometry because of mathematical singularity, a nodal response matrix between incoming and outgoing currents is obtained by expanding detailed nodal flux distribution into a sum of expo-nential functions. Numerical results demonstrate that keff and power distribution agree well with other codes, and the triangular nodal-SP3 method appears faster.

5:10 PMAnalytic Basis Function Expansion Method for Neutron Diffusion Calculation in the Arbitrary Triangular NodesHaoliang Lu, Huang Hao(1), Hongchun Wu(2)1) China Nuclear Power Technology Research Institute, Shenzhen, Guangdong, PR China. 2) School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shanxi, PR China

A new nodal method for directly solving the neutron diffusion equation in the arbitrary triangular nodes was proposed. The neutron flux distributions within a node were ex-panded in a series of analytic basis functions for each group. Nodes were coupled each other with both the zero- and first-order partial neutron current moments simul-taneously. The relationships between a regular triangle and an arbitrary triangle are utilized to describe the neutron current moments. With a new sweeping scheme devel-oped for arbitrary triangular node, the response matrix technique was used to solve the nodal diffusion equations iteratively. Based on the proposed model, the code ABFEMT was developed. The numerical results for a series of benchmark problems show that the core multiplication factor and nodal powers are predicted accurately using this model for unstructured neutron diffusion problems.


2F - Analytical Transport and Numerical Methods for The Diffusion EquationSession Chairs: Farzad Rahnema (GATech), Ricardo Carvalho de Barros (Univ. Rio de Janeiro)

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1:30 PMA Parallel Algorithm for Solving the Multidimensional Within-Group Discrete Ordinates Equations with Spatial Domain Decomposition R. Joseph Zerr(1), Yousry Y. Azmy(2) 1) Department of Mechanical and Nuclear Engineering, The Pennsylvania State Uni-versity, University Park, PA, USA. 2) Department of Nuclear Engineering, North Caro-lina State University, Raleigh, NC, USA

A spatial domain decomposition with a parallel block Jacobi solution algorithm has been developed based on the integral transport matrix formulation of the discrete or-dinates approximation for solving the within-group transport equation. The new meth-odology abandons the typical source iteration scheme and solves directly for the fully converged scalar flux. Four matrix operators are constructed based upon the integral form of the discrete ordinates equations. A single differential mesh sweep is performed to construct these operators. The method is parallelized by decomposing the problem domain into several smaller sub-domains, each treated as an independent problem. The scalar flux of each sub-domain is solved exactly given incoming angular flux boundary conditions. Sub-domain boundary conditions are updated iteratively, and convergence is achieved when the scalar flux error in all cells meets a pre-specified convergence criterion. The method has been implemented in a computer code that was then employed for strong scaling studies of the algorithm’s parallel performance via a fixed-size problem in tests ranging from one domain up to one cell per sub-domain. Results indicate that the best parallel performance compared to source it-erations occurs for optically thick, highly scattering problems, the variety that is most difficult for the traditional SI scheme to solve. Moreover, the minimum execution time occurs when each sub-domain contains a total of four cells.

1:55 PMFPGA Hardware Acceleration for High Performance Neutron Transport Computation Based on Agent Meth-odologyShanjie Xiao and Tatjana Jevremovic Utah Nuclear Engineering Program, The University of Utah, Salt Lake City, UT

The accurate, detailed and 3D neutron transport analysis for Gen-IV reactors is still time-consuming regardless of advanced computational hardware available in devel-oped countries. This paper introduces a new concept in addressing the computational time while persevering the detailed and accurate modeling; a specifically designed FPGA co-processor accelerates robust AGENT methodology for complex reactor ge-ometries. For the first time this approach is applied to accelerate the neutronics analy-sis. The AGENT methodology solves neutron transport equation using the method of characteristics. The AGENT methodology performance was carefully analyzed before the hardware design based on the FPGA co-processor was adopted. The most time-consuming kernel part is then transplanted into the FPGA co-processor. The FPGA co-processor is designed with dataflow-driven non von-Neumann architecture and has much higher efficiency than the conventional computer architecture. Details of the FPGA co-processor design are introduced and the design is benchmarked using two different examples. The advanced chip architecture helps the FPGA co-processor obtaining more than 20 times speed up with its working frequency much lower than the CPU frequency.

2:20 PMAnalysis of the Pool Critical Assembly Benchmark Us-ing RAPTOR-M3G, a Parallel Deterministic Radiation Transport CodeGreg A. FischerWestinghouse Electric Company LLC, Monroeville, PA, USA

The PCA Benchmark is analyzed using RAPTOR-M3G, a parallel SN radiation trans-port code. A variety of mesh structures, angular quadrature sets, cross section treat-ments, and reactor dosimetry cross sections are presented. The results show that RAPTOR-M3G is generally suitable for PWR neutron dosimetry applications.

2:45 PMApplication of Raptor-M3G to Reactor Dosimetry Prob-lems on Massively Parallel ArchitecturesGianluca LongoniWestinghouse Electric Company, Monroeville, PA, USA

The solution of complex 3-D radiation transport problems requires significant resourc-es both in terms of computation time and memory availability. Therefore, parallel al-gorithms and multi-processor architectures are required to solve efficiently large 3-D radiation transport problems. This paper presents the application of RAPTOR-M3G (RApid Parallel Transport Of Radiation – Multiple 3D Geometries) to reactor dosim-etry problems. RAPTOR-M3G is a newly developed parallel computer code designed to solve the discrete ordinates (SN) equations on multi-processor computer architec-tures. This paper presents the results for a reactor dosimetry problem using a 3-D model of a commercial 2-loop pressurized water reactor (PWR). The accuracy and performance of RAPTOR-M3G will be analyzed and the numerical results obtained from the calculation will be compared directly to measurements of the neutron field in the reactor cavity air gap. The parallel performance of RAPTOR-M3G on massively parallel architectures, where the number of computing nodes is in the order of hun-dreds, will be analyzed up to four hundred processors. The performance results will be presented based on two supercomputing architectures: the POPLE supercomputer operated by the Pittsburgh Supercomputing Center and the Westinghouse computer cluster. The Westinghouse computer cluster is equipped with a standard Ethernet net-work connection and an InfiniBand® interconnects capable of a bandwidth in excess of 20 GBit/sec. Therefore, the impact of the network architecture on RAPTOR-M3G performance will be analyzed as well.

PHYSOR 2010 - Advances in Reactor to Power the Nuclear RenaissanceWednesday May 12, 2010 - 1:30 PM - Grand Station Ballroom 4

2G - Parallel Algorithms and ApplicationsSession Chairs: Yousry Azmy (NCSU), Gianluca Longoni (Westinghouse)

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1:30 PMComputationally Optimized Multi-group Cross Section Data Collapsing Using the YGROUP Code C. Yi and G. Sjoden(1), J. Mattingly(2), T. Courau(3) 1) Department of Nuclear and Radiological Engineering, University of Florida, FL, USA. 2) Sandia National Laboratories, Albuquerque, NM, USA. 3) Electricité de France, France

A multi-group cross section collapsing code, YGROUP, has been developed to speed up deterministic particle transport simulations by reducing the number of discrete en-ergy groups while maintaining computational transport accuracy. The YGROUP code leverages previous studies based on the “contributon” approach to automate group selection. First, forward and adjoint deterministic transport calculations are performed on a smaller problem model, or on one section of a large problem model representa-tive of problem physics using a fine group structure. Then, the calculated forward flux and adjoint function moments are used by YGROUP to collapse the fine group cross section library and generate a problem-dependent broad group cross section library. Finally, the broad group library is used for new transport calculations on the full scale/ refined problem model. YGROUP provides several weighting options to collapse the cross section library, including flat, flux, and contributon (the product of forward flux and scalar adjoint moments). Users can also specify fine groups in specific energy ranges of interest to be reserved after collapsing. YGROUP also can be used to evalu-ate the Feynman-Y asymptote characterizing neutron multiplicity.

1:55 PMAnalysis of Cross-Section Condensation in PWR Simu-lationsSteven Douglass and Farzad Rahnema Georgia Institute of Technology, Atlanta, GA, USA

As computer power has improved in recent years, the nuclear industry has more pow-erful tools to solve large-scale reactor problems more effectively. Historically, due to limited computing power, has relied on multi-group nodal diffusion methods to solve whole-core problems, which requires cross sections to be homogenized over assem-blies and collapsed into a coarse group structure (2-4 groups). However, increased computing power has generated an increased interest in the development of direct transport methods for whole-core analysis. Using continuous-energy or fine-group cross sections in whole-core transport is still beyond the capability of available com-puters, and in the development of whole-core transport methods, it is desirable to know the degree to which the collapsing of cross sections affects the solution. In this paper, an analysis of the effect of the collapsing procedure is presented for PWR assemblies, with the goal of determining the number of coarse groups necessary to effectively solve whole-core transport problems.

2:20 PMElements of Validation of the EDF R&D Sn and SPn Solv-ers Vs. MCNP Multigroup Assembly CalculationsPonçot Angélique and Courau TanguyEDF R&D, CLAMART CEDEX, FRANCE

EDF/R&D is developing a new calculation scheme based on a 3D transport-simplified PN (SPN) approach. In this context and for validation purposes, a 3D SN solver is developed to provide reference solutions. This paper focuses on 2D PWR assem-bly pin-by-pin calculations relying on a 2 or 8-group crosssection library. It presents some elements of validation of the SN solver by comparing the results obtained to a MCNP multigroup calculation. Results compared are multiplication factor (keff ) and two-group flux. Comparing the Monte Carlo and SN solvers, discrepancies observed are less than 10 pcm for the keff and less than 0.5% for the flux. Based on the same cross-section libraries, it was also possible to compare the SN and SPN solvers for various orders N of the flux development. Results obtained show that discrepancies are less than 10 pcm for the keff and 0.8% for the flux if using SPN with N 9. Finally, this study confirms the benefit of a deterministic reference solution, provided here by the SN solver, to facilitate the industrial SPN solver optimization process.

2:45 PMAdvanced Method of Solution of Neutron Transport Equation in Nuclear Reactor CellTamara PoveschenkoRRC “Kurchatov Institute”, Moscow, Russia

Method of solution of neutron transport integral equation has been developed. It is aimed into calculation analysis of neutron flux in nuclear reactor cell with complicated geometry and different boundary conditions. On this stage of nuclear reactor calcula-tion it is important to take into account special futures of neutron flux behavior included anisotropy scattering. Modern computational strategy requires the ability to accurately solution of Boltzmann transport equation in the shortest possible time. This approach is based on neutron flux expansion with orthogonal polynomial system in every uni-form mesh of the cell. As result of this approximation the system of linear integral equa-tion is reduced to algebraic system with coefficients that are the six-fold integrals over the cell area in general case. In this paper formulae for calculation of these values are given. The algorithm of computer code for neutron flux calculation is described. The results obtained with general version of collision probabilities method code are given. The advantage of above described approach has been demonstrated.


2H - Cross Section Collapsing & Cell/Assembly CalculationsSession Chairs: Alireza Haghighat (Univ. Florida), Greg Fischer (Westinghouse)

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08:00 AMInvestigation of CMFD Accelerated Monte Carlo Eigen-value Calculation with Simplified Low Dimensional Mul-tigroup FormulationMin-Jae Lee, Han Gyu Joo(1), Deokjung Lee and Kord Smith(2)1) Seoul National University, Seoul, Korea. 2) Studsvik Scandpower, Inc., Idaho Falls, ID, USA

The coarse mesh finite difference (CMFD) formulation is applied to accelerate and stabilize Monte Carlo (MC) eigenvalue calculation which suffers from slow source con-vergence and underestimation of real variance of local properties for high dominance ratio problems. The CMFD linear system is directly obtained from the MC calculation by properly accumulating tallies needed for determining the CMFD homogenization parameters. The feedback of the CMFD solution to the MC calculation is incorporated by a weight adjustment scheme which is to force the MC fission source distribution to be the same as the one determined by the CMFD calculation. 1D and 2D multi-group problems with high dominance ratios are examined and it is shown that the CMFD scheme is quite effective in quickly converging the MC source distribution so that the number of inactive cycles can be greatly reduced. It is also demonstrated that con-siderably improved statistics is attainable in terms of smaller real variance due to the fission source weight adjustment scheme which effectively reflects the accumulated global fission source shape as well as the cycle-to-cycle fluctuation.

08:25 AMRefinement of Monte Carlo Anchoring Method Tested on Continuous-Energy Loosely-Coupled Fissile ProblemsSunghwan Yun and Nam Zin Cho KAIST, Daejeon, Korea

In our previous studies, the Monte Carlo anchoring method, which provides more ac-curate fission source distribution using a small number of histories per generation, was proposed. In this Monte Carlo anchoring method, the fission source term is decom-posed into a conventional Monte Carlo fission distribution and an “anchoring” distribu-tion. The anchoring distribution is provided by a low-order deterministic method and it plays a role of stabilizing the fission source distribution. By introducing anchoring factor α, the biasing effect owing to a potentially inaccurate deterministic solution is minimized. In this paper, the Monte Carlo anchoring method is further refined and implemented in the MCNP5 code, and tested on two-dimensional continuous-energy eigenvalue problems. The tests on symmetric and asymmetric loosely-coupled prob-lems provide encouraging results.

08:50 AMInvestigation of the Entropy-Based Convergence Diag-nostics in MCNP5 using the OECD/NEA Fuel Storage Pool Benchmark ProblemBo Shi and Bojan PetrovicNuclear and Radiological Engineering, Georgia Institute of Technology, Atlanta, GA, USA

Monte Carlo eigenvalue simulation of large systems is prone to slow convergence which makes the diagnostics very challenging and the estimate of the statistical vari-ance questionable. This paper investigated the entropy-based diagnostics in MCNP5 using the OECD/NEA fuel storage pool problem. The results indicated the improve-ment of non-convergence detection, but the slow converging problem remained chal-lenging and non-convergence was not always detected. As one possible approach, the entropy bounding was examined. To further understand the convergence process, we analyzed propagation of the flux distribution through the problem. It turned out to be a potential robust convergence diagnostics approach in the new future.

09:15 AMAnalysis of the Higher Eigenfunction Calculation Using Modified Power Iteration MethodBo Shi and Bojan PetrovicNuclear and Radiological Engineering, Georgia Institute of Technology, Atlanta, GA, USA

Monte Carlo method combined with power iterations is widely used in criticality simula-tions. By powering out higher modes, source distribution will converge to the funda-mental mode. A modified power iteration scheme was proposed by T. Booth to obtain higher eigenfunctions and eigenvalues simultaneously and accelerate the conver-gence rate. We used a simple matrix as well as a one-dimensional nuclear system as examples to illustrate the validity and drawbacks of this method. To overcome the drawbacks, an alternative approach is proposed. All of these efforts would make the method more robust for further applications.

09:40 AMA Hybrid Source-Driven Method to Compute Fast Neu-tron Fluence in Reactor Pressure VesselRen-Tai ChiangAREVA NP INC., San Jose, CA, USA

A hybrid source-driven method is developed to compute fast neutron fluence with neutron energy greater than 1 MeV in nuclear reactor pressure vessel (RPV). The method determines neutron flux by solving a steady-state neutron transport equation with hybrid neutron sources composed of peripheral fixed fission neutron sources and interior chain-reacted fission neutron sources. The relative rod-by-rod power distribu-tion of the peripheral assemblies in a nuclear reactor obtained from reactor core deple-tion calculations and subsequent rod-by-rod power reconstruction is employed as the relative rod-by-rod fixed fission neutron source distribution. All fissionable nuclides other than U238 (such as U234, U235, U236, Pu239 etc) are replaced with U238 to avoid counting the fission contribution twice and to preserve fast neutron attenuation for heavy nuclides in the peripheral assemblies. An example is provided to show the feasibility of the method. Since the interior fuels only have a marginal impact on RPV fluence results due to rapid attenuation of interior fast fission neutrons, a generic set or one of several generic sets of interior fuels can be used as the driver and only the neutron sources in the peripheral assemblies will be changed in subsequent hybrid source-driven fluence calculations. Consequently, this hybrid source-driven method can simplify and reduce cost for fast neutron fluence computations. This newly devel-oped hybrid source-driven method should be a useful and simplified tool for computing fast neutron fluence at selected locations of interest in RPV of contemporary nuclear power reactors.


3A - Monte Carlo Source Convergence Acceleration, Issues and ApplicationsSession Chairs: David Griesheimer (Bettis) and Russell D. Mosteller (LANL)

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3:30 PMStatistical Implications in Monte Carlo Depletions Zhiwen Xu, Joel Rhodes and Kord Smith Studsvik Scandpower, Inc., Idaho Falls, ID, USA

As a result of steady advances of computer power, continuous-energy Monte Carlo depletion analysis is attracting considerable attention for reactor burnup calculations. The typical Monte Carlo analysis is set up as a combination of a Monte Carlo neutron transport solver and a fuel burnup solver. Note that the burnup solver is a determinis-tic module. The statistical errors in Monte Carlo solutions are introduced into nuclide number densities and propagated along fuel burnup. This paper is towards the under-standing of the statistical implications in Monte Carlo depletions, including both statis-tical bias and statistical variations in depleted fuel number densities. The deterministic Studsvik lattice physics code, CASMO-5, is modified to model the Monte Carlo deple-tion. The statistical bias in depleted number densities is found to be negligible com-pared to its statistical variations, which, in turn, demonstrates the correctness of the Monte Carlo depletion method. Meanwhile, the statistical variation in number densities generally increases with burnup. Several possible ways of reducing the statistical er-rors are discussed: 1) to increase the number of individual Monte Carlo histories; 2) to increase the number of time steps; 3) to run additional independent Monte Carlo depletion cases. Finally, a new Monte Carlo depletion methodology, called the batch depletion method, is proposed, which consists of performing a set of independent Monte Carlo depletions and is thus capable of estimating the overall statistical errors including both the local statistical error and the propagated statistical error.

3:55 PMValidating the VESTA Monte Carlo Depletion Interface Using ARIANE Chemical Assay Data for Pressurized Water Reactor ApplicationsL. Cousin,W. Haeck, B. Cochet Institut de Radioprotection et de Sûreté Nucléaire, Fontenay aux Roses cedex, France

The validation of the first release of VESTA (a Monte Carlo depletion interface code that is currently under development at IRSN) has been initiated using the ARIANE chemical assay data from the SFCOMPO database. The Actinide Research In A Nu-clear Element (ARIANE) program examined irradiated MOX (Mixed OXide) and LEU (Low Enriched Uranium) fuel samples in both commercial PWR (Pressurized Water Reactor) and BWR (Boiling Water Reactors) power reactors. The chemical analysis of the studied fuel samples were performed at three independent laboratories, being ITU (Germany), SCK-CEN (Belgium) and PSI (Switzerland). The presented work is limited to 6 of the PWR fuel samples from this program (three MOX samples and three LEU samples). The relatively good agreement seen between the measurements from the ARIANE program and the values predicted by VESTA are very encouraging and are consistent with results obtained with other codes on the general tendencies in the results. This validation exercise has also revealed that a lot of effort is required in preparing these simulations due to the way the experimental results and irradiation conditions are given in the experimental reports. A lot of time could be saved in the validation process of depletion codes if this type of experimental data is transformed into a standardized format for benchmarking purposes.

4:20 PMIn-Line Xenon Convergence Algorithm for Monte Carlo Reactor CalculationsDavid P. GriesheimerBettis Atomic Power Laboratory, Bechtel Marine Propulsion Corporation, West Mifflin, PA, USA

This paper presents an integrated xenon feedback method for use in Monte Carlo reactor calculations. This method enables the calculation of the equilibrium xenon distribution for a reactor during the initial Monte Carlo fission source convergence process. A new extension to this capability allows the calculation of the peak xenon concentration within a reactor following shutdown from power operations. These algo-rithms have been implemented and tested in MC21, an in-house, continuous-energy Monte Carlo particle transport code. This paper provides a brief description of the so-lution algorithms used in the integrated xenon feedback method, along with numerical results for a pressurized water reactor (PWR) example problem.

4:45 PMThe Log Linear Rate Constant Power Depletion MethodDavid C. Carpenter and Joseph H. Wolf IIIBettis Atomic Power Laboratory, West Mifflin, PA USA

The effective delayed neutron fraction, βeff, and the prompt neutron generation time, Λ, in the point kinetics equation are adjoint-weighted quantities. There have been some approaches to estimating them based on the use of either the constant source adjoint function or the self-consistent adjoint function. In this paper, we derive the MC algorithms for computing βeff and Λ weighted by using two kinds of the adjoint func-tions that can be obtained from the MC forward eigenvalue calculations. The results of applications for the infinite homogeneous 2-group problems and some critical facilities are compared with analytic solutions and experimental measurements, respectively. It is demonstrated that βeff and Λ from the use of the self-consistent adjoint function as the weighting function agree better with measurements within a 3% error.

5:10 PMMC21 Monte Carlo Analysis of the Hoogenboom-Martin Full-Core PWR Benchmark ProblemDaniel J. Kelly, Thomas M. Sutton, Timothy H. Trumbull, and Peter S. Do-breffKnolls Atomic Power Laboratory, Bechtel Marine Propulsion Corporation, Schenectady, NY USA

At the 2009 American Nuclear Society Mathematics and Computation conference, Hoogenboom and Martin proposed a full-core PWR model to monitor the improve-ment of Monte Carlo codes to compute detailed power density distributions. This paper describes the application of the MC21 Monte Carlo code to the analysis of this bench-mark model. With the MC21 code, we obtained detailed power distributions over the entire core. The model consisted of 214 assemblies, each made up of a 17×17 array of pins. Each pin was subdivided into 100 axial nodes, thus resulting in over seven million tally regions. Various cases were run to assess the statistical convergence of the model. This included runs of 10 billion and 40 billion neutron histories, as well as ten independent runs of 4 billion neutron histories each. The 40 billion neutron-history calculation resulted in 43% of all regions having a 95% confidence level of 2% or less implying a relative standard deviation of 1%. Furthermore, 99.7% of regions having a relative power density of 1.0 or greater have a similar confidence level. We present timing results that assess the MC21 performance relative to the number of tallies requested. Source convergence was monitored by analyzing plots of the Shan-non entropy and eigenvalue versus active cycle. We also obtained an estimate of the dominance ratio. Additionally, we performed an analysis of the error in an attempt to ascertain the validity of the confidence intervals predicted by MC21. Finally, we look forward to the prospect of full core 3-D Monte Carlo depletion by scoping out the required problem size. This study provides an initial data point for the Hoogenboom-Martin benchmark model using a state-of-the-art Monte Carlo code.


3B - Monte Carlo Depletion and Propagation of UncertaintiesSession Chair: Joachim Miss (IRSN), Vefa Küçükboyaci (Westinghouse)

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10:20 AMCalculating Kinetics Parameters and Reactivity Changes with Continuous-Energy Monte Carlo Brian Kiedrowski and Forrest Brown(1), Paul Wilson(2)1) Los Alamos National Laboratory, Los Alamos, NM USA. 2) Department of Engineer-ing Physics, University of Wisconsin-Madison, Madison, WI USA

The iterated fission probability interpretation of the adjoint flux forms the basis for a method to perform adjoint weighting of tally scores in continuous-energy Monte Carlo k-eigenvalue calculations. Applying this approach, adjoint-weighted tallies are devel-oped for two applications: calculating point reactor kinetics parameters and estimating changes in reactivity from perturbations. Calculations are performed in the widely-used production code, MCNP, and the results of both applications are compared with discrete ordinates calculations, experimental measurements, and other Monte Carlo calculations.

10:45 AMCalculating the Kinetic Parameters in the Continuous Energy Monte-Carlo Code MORETA. Jinaphanh, J. Miss, Y. Richet(1), O. Jacquet(2)1) IRSN (French Institute for Radiological Protection and Nuclear Safety), Fontenay-aux-Roses, France. 2) Independent consultant

This article deals with the implementation of the calculation of kinetic parameters re-cently developed in the MORET Monte Carlo neutron simulator. The direct calcula-tion of the effective neutron lifetime and the effective delayed neutron fraction were implemented. These calculations were made using continuous energy representa-tion, which leads to some approximations especially concerning the adjoint flux. The used techniques are based on Nauchi and Kameyama [3], Meulekamp [4] and Haeck [5] works. The first validations were performed on comparisons with measures from benchmark based on CRAC, SILENE and CALIBAN experiments. The study shows that delayed fission spectra are needed in order to estimate precisely the effective delayed neutron fraction.

11:10 AMEstimation of Adjoint-Weighted Kinectics Parameters in Monte Carlo Forward CalculationsHyung Jin Shim and Chang Hyo Kim(1), Yonghee Kim(2)1) Department of Nuclear Engineering, Seoul National University, Seoul, Korea. 2) Interdisciplinary School of Green Energy, Ulsan National Institute of Science and Tech-nology, Ulsan, Korea

The effective delayed neutron fraction, βeff, and the prompt neutron generation time, Λ, in the point kinetics equation are adjoint-weighted quantities. There have been some approaches to estimating them based on the use of either the constant source adjoint function or the self-consistent adjoint function. In this paper, we derive the MC algorithms for computing βeff and Λ weighted by using two kinds of the adjoint func-tions that can be obtained from the MC forward eigenvalue calculations. The results of applications for the infinite homogeneous 2-group problems and some critical facilities are compared with analytic solutions and experimental measurements, respectively. It is demonstrated that βeff and Λ from the use of the self-consistent adjoint function as the weighting function agree better with measurements within a 3% error.

11:35 AMAn Experience of Applying Iterated Fission Probability Method to Calculation of Effective Kinetics Parameters and Keff Sensitivities with Monte CarloKirill Fedorovich Raskach and Aleksandr Aleksandrovich BlyskavkaInstitute for Physics and Power Engineering, Obninsk, Kaluga Region, Russia

This paper demonstrates use of the Hurwitz iterated fission probability method of es-timating the neutron importance function for calculating the prompt neutron lifetime, the effective delayed neutron fraction and the first-order derivatives (or sensitivities) of the effective multiplication factor with respect to neutron constants and material densities. Special techniques of calculating aforementioned quantities have been pro-posed by different authors. This paper is based on these earlier works but presents all techniques from one point of view. It is noted that in all cases random contribu-tions of adjoint-weighted values expressed through the integral importance of fission neutrons depending on the space variable only (rather than the differential neutron importance depending on the space, energy and angular variables) can be found. This importance is estimated by sampling auxiliary neutron histories. This approach can be implemented in Monte Carlo criticality codes without considerable modifications of calculation procedure. At the same time, it gives quite robust estimates of the adjoint-weighted quantities. While it has been used in multigroup Monte Carlo codes, it can be implemented in continuous-energy Monte Carlo codes without any visible difficulties. The approach is one of possible Monte Carlo interpretations of the Hurwitz iterated fission probability method while there are others. A comparative analysis of different interpretations is not done in this paper.


3C - Calculation of Kinetics Parameters in Monte Carlo CodesSession Chair: Forrest Brown (LANL) , Cenk Güler (Westinghouse)

27

08:00 AMAn Investigation on the Use of Probability Table Cross-Sections for Monte Carlo Lattice CalculationsN. Martin and A. HébertÉcole Polytechnique de Montréal, QC, Canada

Lattice calculation by Monte Carlo methods is expected to compete against deter-ministic tools such as collision probability or the method of characteristics, provided that its calculation time can be shortened. In this paper, we investigate the possibility of using cross-section probability tables conjointly with optimized energy meshes in a multigroup Monte Carlo algorithm. Regarding the Monte Carlo method, probability tables cross-sections are frequently used in the unresolved range. In deterministic lattice codes, this formalism can serve during the self-shielding stage, typically in a subgroup method. We propose using probability tables cross-sections on the entire energy range in a Monte Carlo algorithm, together with an adequate energy struc-ture. Numerical results are proposed using legacy benchmarks and comparison with multigroup, deterministic and pointwise, Monte Carlo solutions is discussed. We dem-onstrate that using probability tables in Monte Carlo multigroup calculations with only 295 or 361 groups can lead to accurate results, as far as the probability tables are consistently evaluated.

08:25 AMRiding Bare-Back on Unstructured Meshes for 21st Cen-tury Criticality CalculationsKaren C. Kelley, Roger L. Martz, and David L. Crane Los Alamos National Laboratory, Los Alamos, NM USA

MCNP has a new capability that permits tracking of neutrons and photons on an un-structured mesh which is embedded as a mesh universe within its legacy geometry capability. The mesh geometry is created through Abaqus/CAE using its solid model-ing capabilities. Transport results are calculated for mesh elements through a path length estimator while element to element tracking is performed on the mesh. The results from MCNP can be exported to Abaqus/CAE for visualization or other-physics analysis. The simple Godiva criticality benchmark problem was tested with this new mesh capability. Computer run time is proportional to the number of mesh elements used. Both first and second order polyhedrons are used. Models that used second or-der polyhedrons produced slightly better results without significantly increasing com-puter run time. Models that used first order hexahedrons had shorter runtimes than models that used first order tetrahedrons.

08:50 AMMoment-Conserving Histopolating B-Splines for Con-tinuous Tally EstimationJustin M. Pounders and Farzad RahnemaNuclear and Radiological Engineering/Medical Physics Programs, George W. Wood-ruff School, Georgia Institute of Technology, Atlanta, Georgia USA

A continuous B-spline representation of bin–averaged (histogram) data is derived. It is shown that this spline representation is a natural extension of the histogram to higher orders. The spline approximations are additionally constrained to conserve the first few algebraic moments of the solution. This work has application to the unfolding of Monte Carlo binned tally data.

09:15 AMA Monte Carlo Surface Source Method for Advanced Test Reactor Experiment PrototypingPaul P. H. Wilson and Patrick Snouffer(1), Erich A. Schneider and Joshua L. Peterson(2)1) University of Wisconsin - Madison, Madison, WI USA. 2) The University of Texas at Austin, Austin, TX USA

The Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) has made ir-radiation facilities available to outside experiments as a National Science User Facility (NSUF). Experimenters need to predict the performance of their designs to determine whether they meet their experimental objectives and whether they will operate safely in the core. Such an analysis requires tools that perform radiation transport simulations to calculate quantities such as displacements per atom (dpa), gas production, and nuclear heating. This paper describes a Monte Carlo analysis approach developed to support design iterations by experimenters. The approach utilizes particle flux tallies to create boundary conditions for use with small-scale models of the experiment position alone. The small-scale model may be created within the Monte Carlo simulation pack-age framework or imported from a computer-aided design model. Although the paper focuses upon ATR in-core experiments, the method is applicable to numerical analysis of a local perturbation to any large-scale system.

09:40 AMEstimation of Coincidence and Correlation in Non-Anal-ogous Monte Carlo Particle TransportMáté Szieberth(1), Jan Leen Kloosterman(2)1) Institute of Nuclear Techniques, Budapest University of Technology and Economics, Műegyetem rkp. 9., Hungary. 2) Department of Radiation, Radionuclides and Reac-tors, Delft University of Technology, Mekelweg 15, The Netherlands

The conventional non-analogous Monte Carlo methods are optimized to preserve the mean value of the distributions and therefore they are not suited for non-Boltzmann problems like the estimation of coincidences or correlations. This paper presents a general method called history splitting for the non-analogous estimation of such quan-tities. The basic principle of the method is that a non-analogous particle history can be interpreted as a collection of analogous histories with different weights according to the probability of their realization. Calculations with a simple Monte Carlo program for a pulse-height-type estimator prove that the method is feasible and provides unbiased estimation. Different variance reduction techniques have been tried with the method and Russian roulette turned out to be ineffective in high multiplicity systems. An alter-native history control method is applied instead. Simulation results of a Feynman-α measurement shows that even the reconstruction of the higher moments is possible with the history splitting method, which makes the simulation of neutron noise mea-surements feasible.


3D - Monte Carlo Methods - General ISession Chairs: Tom Sutton (KAPL) and Roger Martz (LANL)

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08:00 AMVerification of PWR In- and Ex- Vessel Neutron Fluence Calculations With MCNPX, Based on the “H.B. Robin-son-2” Dosimetry BenchmarkAlexander Vasiliev(1), Romain Pittarello(2), Hakim Ferroukhi(1), Edwin Kolbe(1)1) Paul Scherrer Institut, Villigen, Switzerland. 2) ETHZ/EPFL Nuclear Master Pro-gram Student

The use of Monte Carlo particle transport codes for reactor fluence/dosimetry calcula-tions is nowadays a state-of-the-art approach. At Paul Scherrer Institut the MCNPX code is used for studies related to LWR vessel and internals neutron exposure. In or-der to enlarge the in-house validation database as well as assess ability for ex-vessel dosimetry studies, the “H.B. ROBINSON-2” neutron dosimetry benchmark available from the OECD/NEA SINBAD Database has been analyzed recently. The first results obtained employing the JEFF-3.1 library, indicated a rather good performance of the Monte Carlo based solution. Especially, an improvement was observed for the fis-sion dosimeters located in the reactor cavity, versus the originally reported results obtained with the DORT code. However, the statistical precision of the preliminary MC-NPX-2.4.0 modeling, mostly for the (n,p) and (n,α) dosimeters, had to be improved. For that, optimization studies have been performed using some of the standard vari-ance reduction options available with MCNPX code, and the given paper presents the obtained estimates for the dosimeters specific activities C/E values. The work is currently continued with the focus on refinement of some modeling approximations involved in the performed analyses.

08:25 AMMonte Carlo Methods for Neutron Transport on Graph-ics Processing Units Using CUDAAdam G. Nelson(1), Kostadin N. Ivanov(2) 1) Naval Reactors, US Navy, Washington Navy Yard, DC, USA. 2) Department of Me-chanical and Nuclear Engineering, Penn State University, University Park, PA

This work examined the feasibility of utilizing Graphics Processing Units (GPUs) to accelerate Monte Carlo neutron transport simulations. First, a clean-sheet MC code was written in C++ for an x86 CPU and later ported to run on GPUs using NVIDIA’s CUDA programming language. After further optimization, the GPU ran 21 times faster than the CPU code when using singleprecision floating point math. This can be further increased with no additional effort if accuracy is sacrificed for speed: using a compiler flag, the speedup was increased to nearly 23x. Further, if double-precision floating point math is desired for neutron tracking through the geometry, a speedup of 11x was obtained. The GPUs have proven to be useful in this study, but the current genera-tion does have limitations: the maximum memory currently available on a single GPU is only 4GB; the GPU RAM does not provide error-checking and correction; and the optimization required for large speedups can lead to confusing code.

08:50 AMPET Image Reconstruction with On-The-Fly Monte Carlo Using GPUD. Légrády and Á. Cserkaszky(1), A. Wirth(2), B. Domonkos(3)1) Institute of Nuclear Techniques, Budapest University of Technology and Econom-ics, Budapest, Hungary. 2) Semmelweis University, Department of Diagnostic Radiol-ogy and Oncotherapy, Budapest, Hungary. 3) Mediso Medical Imaging Systems Ltd., Budapest, Hungary

Particle transport Monte Carlo is an inherently parallel algorithm and with all the further similarities to visible light ray-tracing it can easily be realized on current Graphics Pro-cessing Units (GPUs). This paper describes a GPU-based gamma photon transport code with application to the iterative image reconstruction steps of Positron Emission Tomography. The aim of the investigations here is the development of a Monte Carlo code capable of calculating the forward projection of each iteration step in the recon-struction with calculation times on the minute scale. Achieved simulation speed is in the order of 10(8) positrons per second.

09:15 AMTMCC: A Transient Three-Dimensional Neutron Trans-port Code by the Direct Simulation MethodHuayun Shen, Zeguang Li, Kan Wang, Ganglin YuDepartment of Engineering Physics Tsinghua University Beijing, China

A direct simulation method (DSM) is applied to solve the transient three-dimensional neutron transport problems. DSM is based on the Monte Carlo method, and can be considered as an application of the Monte Carlo method in the specific type of prob-lems. In this work, the transient neutronics problem is solved by simulating the dy-namic behaviors of neutrons and precursors of delayed neutrons during the transient process. DSM gets rid of various approximations which are always necessary to other methods, so it is precise and flexible in the requirement of geometric configurations, material compositions and energy spectrum. In this paper, the theory of DSM is intro-duced first, and the numerical results obtained with the new transient analysis code, named TMCC (Transient Monte Carlo Code), are presented.

09:40 AMMonte Carlo Simulations and Experimental Validations of α EigenvaluesZeguang Li, Xin Yang, Tianya Li, Ganglin Yu, Kan WangDepartment of Engineering Physics Tsinghua University Beijing, China

A Monte Carlo method called transient curve fitting method was developed to calculate α eigenvalues by first simulating the existing neutrons and precursors in the system, then calculating the eigen-distribution of neutron flux and calculating the α eigenvalues using the transient results based on the eigen-distribution by the code TMCC. The results of this method are tested by calculating Godiva Benchmark problems and they agree well with the benchmark results. Then the reasonable results of Subcritical Fa-cility in Tsinghua University are given by TMCC, and the results are compared with the experimental results measured by Rossi-α method. Even in the deep subcritical cases, the method can give results consistent with experimental results.


3E - Monte Carlo Methods - General IISession Chairs: Tom Sutton (KAPL) and Brian Kiedrowski (LANL)

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1:30 PMAnalysis of Photo-Neutron Physics during Moderator Event in a Deeply Sub-Critical CANDU Reactor J.V. Donnelly and A. Inglot AMEC-NSS, Toronto, Ontario Canada

In CANDU reactors while shutdown in deeply sub-critical configurations, photo-neu-trons are the main source of neutrons in the cores and result in significant differences in reactor characteristics from those in the normal operating configuration. During a moderator-chemistry event impacting on neutron poison in the moderator while in this shutdown configuration, MCNPX simulations provided key components of the understanding of the reactor state as well as assurance of the maintenance of the safe shutdown configuration of the reactor. Coupled neutron-photon calculations were required to represent the important physical processes in this core configuration, as conventional methods of core analyses simulating neutrons alone are inadequate for this application. This is due to the high poison concentrations and the consequential reduction in the applicability of diffusion theory in calculating ex-core detector signals outside of the heavily poisoned reflector. In addition, it is necessary to apply coupled neutron-photon calculations to analyze detector signals in CANDU configurations with high moderator poison concentrations as the distances travelled by photons have dif-ferent characteristics from those of neutrons.

1:55 PMCorrection Methods for Reactivity Monitoring Tech-niques in Pulsed Neutron Source (PNS) MeasurementsV. Bécares, D. Villamarín, M. Fernández-Ordóñez and E. M. González-Romero(1), Y. Fokov(2) 1) Nuclear Innovation Unit, CIEMAT, Madrid, Spain. 2) JIPNR - “Sosny”, Minsk, Be-larus

Reactivity monitoring of subcritical systems has been a major topic of study in the 5th and 6th Framework Programme of the EU (MUSE-4 and EUROTRANS) [1]. This inter-est has been motivated by the proposal of using Accelerator Driven Systems (ADS) to stabilize or reduce the high level nuclear waste inventory. However, licensing and building any industrial scale ADS will require, among other developments, that of a reactivity monitoring system. Among the existing reactivity determination techniques, those based on the point kinetic model are of the most interest due to its simplicity. Pulsed neutron source experiments (PNS) have been used widely in the past since they provide a simple way to apply the prompt decay constant and the area method technique, both based on the point kinetics model. However, MUSE-4 and YALINA-Booster experiments have shown that these techniques can be severely biased in reactors with large reflectors and heterogeneities, making necessary to apply cor-rections. In addition, the necessity to know the kinetic parameters, e and , to get the criticality constant ke forces the use of simulation codes. Hence, this paper proposes a methodology based on Monte Carlo calculations to provide an accurate value of the reactivity using the area method and prompt decay constant results. Furthermore, this methodology reduces the number of kinetic parameters that has to be calculated to just the delayed neutron fraction, which can be calculated with higher accuracy than the mean neutron generation lifetime.

2:20 PMDeterministic Calculation of the Effective Delayed Neu-tron Fraction without using the Adjoint Neutron FluxA. Talamo, Y. Gohar, G. Aliberti, Z. Zhong(1), V. Bournos, Y. Fokov, H. Ki-yavitskaya, C. Routkovskaya, I. Serafimovich(2)1) Argonne National Laboratory, Argonne, IL, USA. 2) Joint Institute for Power and Nuclear Research-Sosny, Minsk, Belarus

In 1997, Bretscher calculated the effective delayed neutron fraction by the k-ratio method. The Bretscher’s approach is based on calculating the multiplication factor of a nuclear reactor core with and without the contribution of delayed neutrons. The multiplication factor set by the delayed neutrons (the delayed multiplication factor) is obtained as the difference between the total and the prompt multiplication factors. Bretscher evaluated the effective delayed neutron fraction as the ratio between the delayed and total multiplication factors (therefore the method is often referred to as k-ratio method). In the present work, the k-ratio method is applied by deterministic nuclear codes. The ENDF/B nuclear data library of the fuel isotopes (238U and 238U) have been processed by the NJOY code with and without the delayed neutron data to prepare multigroup WIMSD nuclear data libraries for the DRAGON code. The DRAG-ON code has been used for preparing the PARTISN macroscopic cross sections. This calculation methodology has been applied to the YALINA-Thermal assembly of Be-larus. The assembly has been modeled and analyzed using PARTISN code with 69 energy groups and 60 different material zones. The deterministic and Monte Carlo results for the effective delayed neutron fraction obtained by the k-ratio method agree very well. The results also agree with the values obtained by using the adjoint flux.

2:45 PMNeutron Noise Calculations using the Analytical Nodal MethodViktor Larsson and Christophe DemaziéreChalmers University of Technology, Department of Nuclear Engineering, Gothenburg, Sweden

In this study, the neutron noise, i.e. the stationary fluctuations of the neutron flux around its mean value, is calculated in 2-group diffusion theory using the Analytical Nodal Method. A brief description of the calculation of the static flux is also included. The static solution is benchmarked against a reference solution in the case of a ho-mogeneous core. The same calculational scheme for the neutron noise as for the static flux is used. As a dynamical benchmark, the calculated neutron noise for a 2D fully homogeneous reactor is compared with the analytical solution, which can eas-ily be determined for homogeneous reactors, at different frequencies. The result of the benchmarks is that the numerical calculations using ANM accurately match the analytical solutions.


4A - Sub-critical System & Reactor Dynamic - ISession Chairs: Sandra Dulla (University of Torino), Imre Pázsit (Chalmers University of Technology)

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3:30 PMNumerical Solution of Reactor Kinetics Equation with Krylov Subspace Method for Matrix Exponential Yuichiro Ban, Akio Yamamoto, Yoshihiro Yamane(1), Tomohiro Endo(2) 1) Graduate School of Engineering, Nagoya University, Nagoya, Aichi, Japan. 2) Nu-clear Fuel Industries, Ltd, Sennan-gun, Osaka, Japan

The spatial discretization form of the space-dependent kinetics equation is a first-order simultaneous ordinary differential equation in time. Conventional numerical methods of the space-dependent kinetics equation, i.e., the Generalized Runge-Kutta method, the Backward Euler method and the Theta method, are based on the temporal differ-encing. However, the present study adapts the analytical solution of space-dependent kinetics equation expressed by the matrix exponential and no-temporal differencing is used. In this context, our present approach is classified as an explicit method in which no iteration calculation on space and energy is necessary. The Krylov subspace method is used to evaluate the matrix exponential appeared in the solution of the spatially-discretized space-dependent kinetics equation. The Krylov subspace method is implemented into a space-dependent kinetics solver. In order to examine the effec-tiveness of the Krylov subspace method, the TWIGL benchmark problem is analyzed as a validation calculation. The calculation results suggest that the Krylov subspace method is a good candidate for a space-dependent kinetics solver from the view point of accuracy and computation time.

3:55 PMInvestigation of the Space-Dependent Noise Induced by Propagating Density FluctuationsV. Dykin and I. Pázsit Chalmers University of Technology, Department of Nuclear Engineering, Göteborg, Sweden

The space-dependent behavior of neutron noise induced by a propagating perturba-tion, represented by the fluctuations of the absorbtion cross section propagating with the coolant of a PWR, is investigated in a one-dimensional one-group approach. The general space-frequency problem is solved for this specific noise source with the help of Greens function technique. All calculations are made in the frame of first-order perturbation theory. The solution is investigated for a different frequencies and system sizes. The limits of point-kinetic and space-dependent behaviour were investigated. An interesting interference phenomenon was found between the point kinetic and the pure space dependent components of the noise for certain combinations of the fre-quency and system size. The results bear a significance for the dynamics of Molten Salt Reactors (MSR), which will be reported on in a companion paper.

4:20 PMReactor Kinetics, Dynamic Response and Neutron Noise in Molten Salt ReactorsI. Pázsit and A. JonssonChalmers University of Technology, Department of Nuclear Engineering, Göteborg, Sweden

The dynamic space- and frequency dependent response of a Molten Salt Reactor (MSR) to stationary perturbations is investigated in a simple analytical model. The Green’s function of the system is investigated in the general case of arbitrary fuel re-circulation velocity and in the limiting case of infinite fuel velocity which permits closed form solutions both in the static and dynamic case. It is found that the amplitude of the induced noise is generally higher and the domain of the point kinetic behaviour valid up to higher frequencies than in a corresponding traditional system. This is due to the differing behaviour of the delayed neutron precursors as compared to the traditional case. The MSR equations are not self-adjoint and the adjoint equation and adjoint function have to be constructed, which is also done here. Finally the space-dependent neutron noise, induced by propagating perturbations of the absorption cross section is calculated. A number of interesting properties that are relevant to full size MSRs are found and interpreted. The results are consistent with those in traditional systems but the domains of various behaviour regimes (point kinetic, space dependent etc.) is shifted to higher frequencies or system sizes.

4:45 PMFast Breeder Reactor Kinetics. A Direct ProblemE.F.Seleznev, A.A.Belov and A.A.Mushkaterov(1), I.P.Matveenko, A.M. Zhukov and K.F.Raskach(2)1) Nuclear Safety Institute of the Russian Academy of Sciences (IBRAE), Moscow, Russia. 2) Institute for Physics and Power Engineering (IPPE), Obninsk, Kaluga re-gion, Russia

An analysis of solution of the spatial reactor kinetics equation using on fast breeder reactor is presented in this paper. To solve the spatial reactor kinetics equation in three-dimensional geometry in multi-energy group diffusion approximation, a program calculated module TIMER was created. The number of groups of the delayed neutron precursor concentrations varies from 6 to 8. In the solution of a transient problem two specific tasks are distinguished. The first of them is a direct problem wherein the neutron flux and its derivatives such as reactor power etc. are determined at each time step. The second (inverse) problem exists for the point-kinetics equation where reac-tor reactivity is calculated using the known dependence of reactor power on time. The paper presented focuses on solution of the first problem using experiment calculations for BFS-105 critical assembly.

5:10 PMFast Breeder Reactor Kinetics. An Inverse ProblemE.F.Seleznev, A.A.Belov and A.A.Mushkaterov(1), I.P.Matveenko, A.M. Zhukov and K.F.Raskach(2)1) Nuclear Safety Institute of the Russian Academy of Sciences (IBRAE), Moscow, Russia. 2) Institute for Physics and Power Engineering (IPPE), Obninsk, Kaluga re-gion, Russia

An analysis of solution of the spatial reactor kinetics neutron-transfer equation using a fast reactor as an example is presented in this paper. To solve the spatial reactor kinetics equation in three-dimensional geometry in multi-energy group-diffusion ap-proximation, a program calculated module TIMER was created. The number of groups of delayed neutron precursor concentrations varies from 6 to 8. When solving a tran-sient neutron-transfer problem, two specific tasks are distinguished. The first of them is a direct problem wherein the neutron flux density and its derivatives such as reactor power etc. are determined at each time step. The second (inverse) problem exists for the point-kinetics equation where the reactor reactivity is calculated using the known dependence of reactor power on time. The paper presented focuses on solution of the inverse problem using experiment calculations for BFS-105 critical assembly.


4B - Sub-critical System & Reactor Dynamic - IISession Chairs: Sandra Dulla (University of Torino), Imre Pázsit (Chalmers University of Technology)

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08:00 AMDr. Rudolf J.J. Stamm’ler - in MemoriamJuan J. Casal(1), Jan H. Blomstrand(2)1) Westinghouse Electric Sweden, Västerås, Sweden. 2) Professor Emeritus, Nuclear Reactor Engineering, KTH – Royal Institute of Technology, Stockholm, Sweden

In memoriam of Dr. Rudi Stamm’ler, who made significant contributions to the devel-opment of reactor physics, in particular in the area of neutron transport calculations at assembly level.

08:25 AMRudi Stamm’ler Contributions and DRAGONR. Roy, G. Marleau and A. Hébert École Polytechnique de Montréal, Montréal, Canada

The lattice code DRAGON has been in constant development over the last 25 years. During this period, the DRAGON development team has often been directly influenced by the excellent work of Rudi Stamm’ler. First, his book on reactor physics has inspired a large number of programming and calculation techniques that were implemented in DRAGON. Then, the work of Rudi and his collaborators on the lattice code HELIOS, has also prompted a friendly competition that lead us to continuously improve our code in such a way that it could match the performance achieved by HELIOS. This paper provides a description of some characteristics or technologies implemented in DRAGON that were influenced by the work of Rudi Stamm’ler. It also describes a CANDU simulation exercice where the capabilities of the HELIOS and DRAGON codes were combined.

08:50 AMAssembly Depletion with Critical Spectrum in McCARD Monte Carlo Calculations and Comparison with HE-LIOSHo Jin Park, Han Gyu Joo, Hyung Jin Shim and Chang Hyo KimSeoul National University, Gwanak-gu, Seoul, Korea

A method for Monte Carlo depletion with critical spectrum to generate assembly ho-mogenized few group cross sections is presented. It is based on the 0-dimensonal fine-group B1 equations. The groupwise ratios of the critical spectrum to the infinite medium spectrum are applied uniformly across the whole depletion region in the as-sembly in order to adjust the groupwise microscopic reaction rates at each depletion region considering the critical spectrum. Comparison of the critical spectrum based assembly depletion is made between the McCARD Monte Carlo depletion code and the HELIOS lattice transport code. The k-infinities and two-group cross sections ob-tained with and without the critical spectrum are compared. In addition, a Monte Carlo checkerboard depletion calculation is performed to verify that the critical spectrum based McCARD assembly depletion calculation indeed generates few group con-stants correctly.

09:15 AMA Resonance Calculation Method based on the Multi-Terms Rational Approximation for General Geometry with Gray Resonance AbsorbersHiroki Koike, Kazuya Yamaji, Daisuke Sato, Shinobu Tsubota and Hideki Matsumoto(1), Akio Yamamoto(2)1) Mitsubishi Heavy Industries, Ltd., Hyogo-ku, Kobe, Japan. 2) Nagoya University: Furo-cho, Chikusa-ku, Nagoya, Japan

A new resonance calculation scheme based on the multi-terms rational approximation is developed and its outline is presented in this paper. On the present method, coef-ficients appeared in the expression of escape probability can be evaluated including the gray resonance absorption effect. Neutron fluxes are calculated by one group fixed source transport calculations with the method of characteristics for different macro-scopic total cross sections of resonance materials. Then multi-terms coefficients of the escape probability are evaluated by fitting the results of the previous transport calculations on the wide range of the optical length. The present method can be used for general geometry and the escape probability is accurately reproduced both for black and gray absorbers. On the traditional resonance calculations based on the equivalence principle, the blackness assumption has been applied both for the single and multi-terms rational approximations, so the present method is an enhanced ap-proach for considering the resonance self-shielding effect. Verifications of the present method are carried out in some simple benchmark problems. As a result, the theoreti-cal validity of the present method is numerically confirmed. The present method has been implemented in the next generation lattice physics code GALAXY of Mitsubishi Heavy Industries, Ltd..

09:40 AMNeutronic Calculations for Steel-Reflected Fast Critical Systems with the Sub-Group Sn MethodGo Chiba and Teruhiko KugoJapan Atomic Energy Agency, Shirakata, Tokai, Ibaraki, Japan

In the present paper, we perform neutronic calculations for steel-reflected fast critical systems with the sub-group SN method. In order to extend the applicability of the sub-group SN method, we consider sub-group to sub-group transfer probabilities for in-group scattering sources. In addition, sub-group dependence of out-group scattering sources, which has been ignored in previous studies, is also taken into account. The present sub-group SN method is applied to neutronic calculations for several steel-reflected fast systems included in the ICSBEP handbook. It is shown that the present sub-group SN method reproduces quite well the reference Monte-Carlo solutions for effective multiplication factors and neutron flux spatial distributions above 0.1 MeV in reflector regions. This method, however, shows poor accuracy in neutron flux calcula-tions for specific energy groups in which large and wide resonances exist or in which the contribution of the inelastic scattering source is large. These are the limitations of the present sub-group SN method, and remedy for them is necessary if accurate neutron flux calculations are required for such energy groups.


4C - Lattice Physics Calculations: in Memory of Rudi Stamm’ler - ISession Chairs: Charles A. Wemple (Studsvik), Tomasz Kozlowski (Nuclear Power Safety - Sweden)

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10:20 AMAPOLLO2-A – AREVA’s New Generation Lattice Physics Code: Methodology and Validation E. Martinolli, T.C. Carter, F. Clément, P.M. Demy, M. Leclère, P. Magat, A. Marquis, V. Marotte(1), J. Marten, S. Misu(2), M. Schneider(2), S. Thareau(3), L. Villatte(1)1) AREVA NP, Plants Sector, Tour AREVA, Paris La Défense, France. 2) AREVA NP, Fuel Sector, Paul-Gossen-Strasse 100, 91052 Erlangen, Germany, 3) AREVA NP, Fuel Sector, 10, rue Juliette Récamier, Lyon, France

AREVA has developed the ARCADIA® reactor code system including the lattice phys-ics transport code APOLLO2-A. Based on the APOLLO2 kernel developed by CEA, APOLLO2-A features a state-of-the-art methodology designed by AREVA for Light Water Reactor industrial applications. The validation of the code is achieved through comparisons with a comprehensive experimental database and with Monte-Carlo ref-erence codes. In this paper, the main features of APOLLO2-A, the methodology and results from the validation base are presented.

10:45 AMQualification of APOLLO2.8/JEFF-3.1.1 Code Package for the Calculations of PWR Plutonium Recycling Using the EPICURE ExperimentsJ-F. Vidal, A. Calloo, P. BlaiseCEA-Cadarache, DEN/DER/SPRC/LEPh, 13108 Saint Paul lez Durance, France

The recent release of the JEFF-3.1.1/CEA2005v4 nuclear data library and the latest improvements of the APOLLO2.8 code (281-group SHEM energy mesh, mixture self-shielding treatment and performing 2D solver using the method of characteristics) have allowed the development of new powerful LWR calculation schemes : reference SHEM-MOC and optimized REL2005 for industrial applications. For three years, CEA has been carrying out an intensive qualification program of this new APOLLO2.8 neu-tronics package. This paper is devoted to the experimental validation of the pin-by-pin power distribution in the specific case of the Plutonium recycling in French 900 MWe PWRs. UH1.4 (UO2 17x17 mock-up) and UMZONE (mixed MOX-UO2 loading mock-up) experiments from the EPICURE program performed in the EOLE facility at Cadarache Centre have been analysed. Calculation results show good agreements with the experiments:- Reactivity is slightly overpredicted for the two configurations : +480 pcm ± 600 pcm (2 std)with SHEM-MOC and +510 pcm ± 600 pcm (2 std) with REL2005- The average power ratio PMOX/PUO2 is also slightly overpredicted but remains within 2std experimental uncertainty margins: +1.2% ± 2.4% (2 std) with SHEM-MOC and +1.8% ± 2.4% (2 std) with REL2005.- one pin excepted in each type of assembly, the results obtained on pin-by-pin fis-sion rates for UO2 and MOX assemblies are also very satisfactory, in particular near the interface: 1.4% ± 1.6% (2 std) with SHEM-MOC and ±1.5% ± 1.6% (2 std) with REL2005.These results demonstrate the ability of both APOLLO2.8 schemes and JEFF-3.1.1 library to predict major neutronics parameters of MOX-UO2 fuelled cores, with the same precision than standard UO2 ones.

11:10 AMAnalysis of the Isotopic Distributions During Burnup of UMOX- and ThMOX-Fuels on Unit Cell BasisB. Merk and S. SchollForschungszentrum Dresden-Rossendorf, Institut für Sicherheitsforschung, Dresden, Germany

A detailed analysis of the burning of Plutonium in a representative PWR fuel pin is per-formed for comparison of the performance of ThMOX and UMOX fuel. Special effort is spent in the analysis of changes in the spatial distribution of isotope concentrations during burnup in this representative LWR fuel pin. This unique analysis of the changes in the spatial nuclide densities gives new insight into the system behavior. The different ways of Plutonium breeding and reduction for the major isotopes of the two considered fuels are analyzed and discussed. Finally, the advantages and limits of the use of Tho-rium based MOX fuel for the burning of Plutonium are discussed. The calculations are performed with the licensing grade code module HELIOS 1.9.

11:35 AMAnalysis of the MISTRAL Experiment on the Reactivity Temperature Coefficient for UOX and MOX Lattices us-ing JEFF3.1.1 Nuclear Data LibraryLahoussine Erradi And Alain SantamarinaCEA, DEN, SPRC, F-13108 Saint-Paul lez Durance, France

Japanese Nuclear Power Engineering Corporation (NUPEC) and the French Commis-sariat à l’Energie Atomique (CEA) have jointly established the MISTRAL programme to carry out reactor physics experiments of 100% MOX cores in the EOLE facility. The parameter of interest in this work is the Reactivity Temperature Coefficient (RTC) which was measured in MISTRAL1 configuration for UOX lattices and in MISTRAL2 and MISTRAL3 configurations for MOX lattices in the temperature range from 10°C up to 80°C. The analysis of these experimental configurations was carried out using the APOLLO2 code with the most recent JEFF-3.1.1 nuclear data library. The discrep-ancies between calculation and experiment on the RTC for UO2 lattices is less than 0.9 pcm/°C in the whole temperature range (on average: C-E = +0.4 ± 0.3 pcm/°C). Concerning MOX lattices, the C-E bias on the average RTC is more significant: -0.9 ± 0.3 pcm/°C that is the same magnitude of the target-accuracy ±1 pcm/°C for LWR design calculations.


4D - Lattice Physics Calculations: in Memory of Rudi Stamm’ler - IISession Chairs: Tomasz Kozlowski (Nuclear Power Safety - Sweden) , Taek K. Kim (ANL)

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1:30 PMValidation of the Lattice Physics Code LANCER02 with ENDF/B-VII Library Ugur Mertyurek, Scott Palmtag, Godfree Gert, and Joshua Finch Global Nuclear Fuel, Wilmington, NC

The verification and validation of the LANCER02 lattice physics code is presented, consisting of a suite of test cases comparing LANCER02 to MCNP using a consistent ENDF/B-VII Rev. 0 cross section library. MCNP results using this data library have been validated against measured data from critical experiments. In addition, several comparisons of LANCER02 are made to MONTEBURNS to qualify the depletion ca-pability in LANCER02. These results show excellent agreement between LANCER02 and MCNP and provide a firm foundation for use in downstream applications.

1:55 PMNodal Expansion Method with Axially Varying Cross-sectionsBenjamin Collins, Andrew Ward, Yunlin Xu, Thomas Downar Department of Nuclear Engineering and Radiological Sciences, University of Michi-gan, Ann Arbor, MI USA

A new nodal expansion method, NEMZXS, has been developed to model the axial variation of the cross-sections. The flux is treated as a fourth order polynomial and the source term and cross-sections are second order polynomials. Legendre polynomials are used as the basis functions. Three weighted residual equations and flux and cur-rent continuity equations allow for the coefficients of the flux expansion to be solved at each node. The cross-sections are known from the node average value as well as the top and bottom boundaries of the node. In the limit where the cross-sections are con-stant inside a node, NEMZXS reduces to the original NEM solution. This method has the most straightforward application in the BWR where the density continuously var-ies in the axial direction but this method can be applied in other applications such as depletion where the exposure causes a continuous change in the cross-sections. This method is compared with the NEM with constant cross-sections for a BWR assembly with an average density distribution. The results show that the NEMZXS maintains the fine mesh solution even after the mesh is significantly reduced by a factor of four. The NEM solution could only be reduced by a factor of two to maintain the fine mesh solution. The ability to reduce the number of nodes can have a significant impact on the run time for both steady state and transient nodal calculations.

2:20 PMDevelopment of the HELIOS/CAPP Code System for the Analysis of Pebble Type VHTR CoresHyun chul Lee, Chang Keun Jo, and Jae Man NohKorea Atomic Energy Research Institute, Yuseong-gu, Daejeon, Republic of Korea

Korea Atomic Energy Research Institute is developing the HELIOS/CAPP code sys-tem for the analysis of a very high temperature gas cooled reactor. The HELIOS/CAPP code system adopted a two-step core analysis procedure. In the first step, the HELIOS code is used to generate few-group cross section table sets and, in the second step, the CAPP code is used to perform core physics analyses. Some specific methodolo-gies have been developed for the HELIOS/CAPP code system to treat some unique neutronic characteristics of the VHTR core such as the double heterogeneity effect, a strong fuel/reflector interaction, etc. This paper introduces the specific methodologies and presents the development status of the CAPP code. The CAPP code has a multi-group neutron diffusion equation solver based on the finite difference method and finite element method. The functionality of the CAPP code has been extended on top of the flux solver to be able to perform steady state core physics analysis and a core deple-tion analysis. A specific module was developed to treat the tabulated cross-sections as a function of state variables such as burnup, fuel temperature, moderator temperature. A simplified steady state thermo-fluid solver was developed for the thermal feedback calculation. A micro-depletion module was developed for the depletion analysis of a pebble type VHTR core. This paper also presents some verification results of the HE-LIOS/CAPP code system.

2:45 PMA Macro Reactivity-Equivalent Physical Transformation Method for Prismatic Gas-Cooled Reactors Utilizing Uranium Oxycarbide FuelBrian AdeSchool of Nuclear Engineering, Purdue University, West Lafayette, IN USA

Prismatic block nuclear reactor geometric homogenization is of interest in order to decrease the expense of full core calculations. The reactivity-equivalent physical transformation (RPT) method of homogenizing TRISO particle fuel regions has been proven for the rods of prismatic gas reactors as well as the pebbles of pebble bed reactor cores. After application of the RPT method, each fuel block of a prismatic gas reactor contains more than 900 different regions to explicitly model. A full core simula-tion of a problem of this size would simply be too computationally expensive to perform regularly. In order to perform needed calculations for these reactors, simplification of the models is a must. One option is to use a simple volume-weighted homogenization (VWH), which results in a highly underestimated reactivity for the fuel block. The goal of this research is explore methods of homogenization that will correctly predict key parameters of the problem while providing a significant decrease in the computational expense. This paper outlines a macro reactivity-equivalent transformation (MRPT) method applied to prismatic gas reactor fuel blocks that greatly decreases the expense of the calculation while maintaining accuracy of the key parameters of the problem.


4E - Lattice Physics Calculations: in Memory of Rudi Stamm’ler - IIISession Chairs: Taek K. Kim (ANL) , Charles A. Wemple (Studsvik)

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3:30 PMHTGR Fuel Element Depletion Benchmark: Stage Three Results Emil Fridman(1), Eugene Shwageraus(2) 1) Institute of Safety Research, The Forschungszentrum Dresden-Rossendorf, Dres-den, Germany. 2) Department of Nuclear Engineering, Ben-Gurion University of the Negev, Beer Sheva, Israel

Recently, a new numerical benchmark exercise for High Temperature Gas Cooled Re-actor (HTGR) fuel depletion was defined. The purpose of this benchmark is to provide a comparison basis for different codes and methods applied to the burnup analysis of HTGRs. The benchmark specifications include three different models: (1) an infi-nite lattice of tristructural isotropic (TRISO) fuel particles, (2) an infinite lattice of fuel pebbles, and (3) a prismatic fuel including fuel and coolant channels. In this paper, we present the results of the third stage of the benchmark obtained with MCNP based depletion code BGCore and deterministic lattice code HELIOS 1.9. The depletion cal-culations were performed for three-dimensional model of prismatic fuel with explicitly described TRISO particles as well as for two-dimensional model, in which double het-erogeneity of the TRISO particles was eliminated using reactivity equivalent physical transformation (RPT). Generally, good agreement in the results of the calculations obtained using different methods and codes was observed.

3:55 PMHTTR Fuel Block Simulations With SCALEDan Ilas, Jess Gehin Oak Ridge National Laboratory, Oak Ridge, TN USA

The SCALE code system is currently being updated to improve the methods and data to support High Temperature Gas Cooled reactor analysis. This paper presents the re-sults of a High Temperature Engineering Test Reactor (HTTR) fuel block analysis with SCALE6.0, which is included as one of the sources of validation data. Good agree-ment is reported between continuous energy and multigroup SCALE/KENO models of the HTTR block for eigenvalue and fission density distribution. The agreement of the SCALE/KENO results with the MCNP5 results is also very good. The boundary condi-tion is shown to have a large effect on both the calculated eigenvalue and the fission density distribution within the fuel block. In addition to the reported results, a method is proposed to alleviate the fuel mass non-conservation due to the clipping of the lattice of grains within the fuel element in the Monte Carlo continuous energy models.

4:20 PMOn SCALE Validation for PBR AnalysisGermina IlasOak Ridge National Laboratory, Oak Ridge, TN USA

Studies were performed to assess the capabilities of the SCALE code system to pro-vide accurate cross sections for analyses of pebble bed reactor configurations. The analyzed configurations are representative of fuel in the HTR-10 reactor in the first critical core and at full power operation conditions. Relevant parameters—multiplica-tion constant, spectral indices, few-group cross sections—are calculated with SCALE for the considered configurations. The results are compared to results obtained with corresponding consistent MCNP models. The code-to-code comparison shows good agreement at both room and operating temperatures, indicating a good performance of SCALE for analysis of doubly heterogeneous fuel configurations.

4:45 PMPEBBLES Mechanics Simulation SpeedupJoshua J. Cogliati and Abderrafi M. OugouagIdaho National Laboratory, Idaho Falls, ID USA

Pebble bed reactors contain large numbers of spherical fuel elements arranged ran-domly. Determining the motion and location of these fuel elements is useful for calcu-lating certain operating parameters of pebble bed reactors. These simulations involve determining the entire core motion as hundreds of thousands of pebbles are recircu-lated over months to years of physical time. Single processor algorithms are insuf-ficient for this simulation because they would take decades to centuries of wall-clock time. This paper describes the process of parallelizing and speeding up the PEBBLES mechanics simulation code. Shared memory programming with the Open Multipro-cessing Application Programming Interface and distributed memory programming with the Message Passing Interface are used simultaneously in this process. A new shared memory lock-less linear time collision detection algorithm is described, a method that allows faster detection of pebbles in contact than generic methods. This paper also provides an overview of the computationally expensive portions of computer codes and the different areas to be optimized. These improvements to PEBBLES combine to make full simulations of recirculations in German AVR-sized reactors possible in months of wall clock time.

5:10 PMPEBBLES Simulation of Static Friction and New Static Friction BenchmarkJoshua J. Cogliati and Abderrafi M. OugouagIdaho National Laboratory, Idaho Falls, ID USA

Pebble bed reactors contain large numbers of spherical fuel elements arranged ran-domly. Determining the motion and location of these fuel elements is useful for calcu-lating certain parameters of pebble bed reactor operation. Static friction affects both the arrangement of the pebbles and the forces they transmit. This paper documents the PEBBLES static friction model used to simulate the motion of pebbles in pebble bed reactors. The model uses a three-dimensional differential static friction approxima-tion extended from the two-dimensional Cundall and Strack model. The derivation of determining the rotational transformation of pebble-to-pebble static friction force is pro-vided. A new implementation for a differential rotation method for pebble-to-container static friction force has been created. Previous published methods are insufficient for pebble bed reactor geometries. A new analytical static friction benchmark is document-ed that can be used to verify key static friction simulation parameters. This benchmark is based on determining the exact pebble-to-pebble and pebble-to-container static fric-tion coefficients required to maintain a stable five-sphere pyramid.


4F - Computational Challanges of HTGRsSession Chairs: Germina Ilas (ORNL), Eugene Shwageraus (MIT)

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08:00 AMImproved Reflector Modeling for LWR AnalysisDavid Colameco(1), Mohamed Ouisloumen(2), Larry T. Mayhue(2), Kostadin N. Ivanov(1)1) Department of Mechanical and Nuclear Engineering, The Pennsylvania State Uni-versity. 2) Westinghouse Electric Company

Special treatment of the reflector in reactor analysis is required due to the drastic dif-ferences of neutronics properties between the core and reflector regions. The strong spectrum change observed at core/reflector boundaries combined with geometry complexity, material and structural heterogeneity of radial and axial reflectors make such treatment a challenging task. The correct modeling of the reflector response is important for accurate predictions of core power distribution especially in regions next to the reflector. For this reason special care is taken in generation of reflector homog-enized cross-sections and discontinuity factors (DFs). Historically, one-dimensional (1-D) color set problems are used for the reflector, which is different from the unit fuel assembly cross-section generation models. The investigations presented in this paper are further extension of studies performed elsewhere to achieve a more correct mod-eling by introducing improved color set models for reflector cross-section and DF gen-eration and parameterization. These color set models more accurately capture the 2-D effects that occur on reentrant surfaces. From the transport solution of the color set model, discontinuity factors are calculated, which in turn preserve the transport solu-tion in the nodal code calculations. Two sensitivity studies have been performed. The first study evaluates the effect of the size of the color-set model on reflector constants as compared to full 1/8 core sector of symmetry. The second sensitivity study is con-ducted with the aim of determining a parameterization for the discontinuity factors as function of the core conditions such as boron concentration, moderator temperature and density. In addition, the effects of the loading pattern next to the reflector region on the discontinuity factors are examined through the use of color sets that include Mixed Oxide (MOX) fuel and the traditional UO2 fuel. The prediction improvements that are achieved in both the global eigenvalue and power distribution from the selected op-timal 2-D color sets as compared to the 1-D models currently in use are discussed in this paper for two nuclear power plants. The newly calculated discontinuity factors show an improvement in predicting the global eigenvalue and power distribution and correcting the power tilt that was previously observed.

08:25 AMAnalysis of Cold Critical Experiments by a Pin-by-Pin Core Analysis Method Using a Three-Dimensional Di-rect Response MatrixTakeshi Mitsuyasu, Kazuya Ishii, Tetsushi Hino and Motoo Aoyama Energy and Environmental Systems Laboratory, Hitachi, Ltd., Hitachi-shi Ibaraki-ken, Japan

The core analysis method using the three-dimensional direct response matrix (DRM) method and its computing algorithm have been developed. As the sub-response ma-trices were obtained by pin-by-pin evaluation in Monte Carlo calculations, the DRM method can reproduce the pin-by-pin production rates. Considering sub-response ma-trices as 2-dimensional matrices, the response matrix R formed by polynomial terms was transformed to the inverse matrix. The computational time and memory usage were reduced without causing a difference of the eigenvalue due to the nodal sym-metry realized from using symmetric block matrices and the mathematical inverse matrix solution. In the quarter core of an ABWR, the computational time and memory usage were reduced 10% and 33%, respectively. For verification of the DRM method, cold critical experiments were analyzed. In commercial plants, the standard deviation was 0.07 %Δk for a BWR-5 and 0.11 %Δk for an ABWR. The k-effective values could be evaluated with relatively small variations for both plants against various withdrawal patterns of control rods. In the BASALA MOX critical experiments, several cases were analyzed. In the test region, the fission rate root-mean-square difference from 9 ex-perimental measurements was about 1.5% which was equivalent to the experimental error. Moreover, the DRM method also had the same accuracy as the continuous energy Monte Carlo code, MVP. The DRM method has a matrix computing algorithm suitable for parallel computing and it can accurately reflect the effects of intra- and inter-assembly heterogeneities in heterogeneous systems.

08:50 AMGroup Decoupled Multi-Group Pin Power Reconstruc-tion Utilizing Nodal Solution 1D Flux ProfilesLulin Yu(1), Dong Lu(1), Shaohong Zhang(1) and Yung-an Chao(1,2)1) School of Nuclear Science and Engineering, Shanghai Jiao Tong University, Shang-hai, China. 2) Westinghouse Electric Company, Pittsburgh, PA, USA

A group decoupled direct fitting method is developed for multi-group pin power re-construction, which avoids either the complication of constructing 2D analytic multi-group flux solution or any group coupled iteration. In addition to nodal volume and surface average fluxes and corner fluxes, transversely integrated 1D nodal solution flux profiles are also used. For each energy group, a two-dimensional expansion in nine polynomials and eight hyperbolic functions is used to perform a constrained least squared fit to the 1D intra-nodal flux solution profiles. The constraints are on the con-servation of nodal volume and surface average fluxes and corner fluxes. The corner fluxes are determined thru the condition of corner currents conservation. An optional slowing down source improvement method is also developed to further enhance the solution if needed. Two test examples are shown with very good results. One problem is a four-group BWR mini-core with all control blades inserted and the other problem is the OECD seven-group MOX benchmark, C5G7.

09:15 AMAccuracy Enhancements of the Coarse-Mesh Diffusion Core Model QUABBOX/CUBBOX for Highly Heteroge-neous Core ConfigurationsKlein M., Pasichnyk I., Pautz A., Velkov K., Zwermann W.GRS mbH, Garching, Germany

This paper discusses recently implemented improvements of the QUABOX/CUBBOX core model, in order to meet the requirements for achieving higher accuracy in the prediction of assembly- and pin-power distributions in LWR cores. The accuracy of the coarse mesh neutron diffusion code QUABOX/CUBBOX was enhanced by incor-porating the assembly discontinuity factors in its cross section libraries. Additionally the SPH method was successfully implemented for the prediction of pin-power dis-tributions. It is shown that even for highly heterogeneous UOX fuel assemblies with fully inserted control rods, pin-power distributions can be predicted within a very good accuracy by applying a hybrid diffusion-transport equivalence algorithm based on the SPH method.

09:40 AMAn Alternative Solver for the Nodal Expansion Method EquationsFernando Carvalho da Silva, Antonio Carlos Marques Alvim and Aquilino Senra MartinezPrograma de Engenharia Nuclear, COPPE/UFRJ, Rio de Janeiro, RJ

An automated procedure for nuclear reactor core design is accomplished by using a quick and accurate 3D nodal code, aiming at solving the diffusion equation, which describes the spatial neutron distribution in the reactor. This paper deals with an alter-native solver for nodal expansion method (NEM), with only two inner iterations (mesh sweeps) per outer iteration, thus having the potential to reduce the time required to cal-culate the power distribution in nuclear reactors, but with accuracy similar to the ones found in conventional NEM. The proposed solver was implemented into a computa-tional system which, besides solving the diffusion equation, also solves the burnup equations governing the gradual changes in material compositions of the core due to fuel depletion. Results confirm the effectiveness of the method for practical purposes.


4G - Core Physics Methods - ISession Chairs: Yung-An Chao (Westinghouse), Akio Yamamoto (Nagoya University)

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10:20 AMA New Robust Cross Section Representation Methodol-ogy for PWR Core Simulator Daisuke Sato, Shinobu Tsubota, Kazuya Yamaji, Hiroki Koike, Hideki Mat-sumoto(1), Akio Yamamoto(2)1) Mitsubishi Heavy Industries, Ltd.: 1-1 Wadasaki-cho 1-chome, Hyogo-ku, Japan. 2) Nagoya University: Furo-cho, Chikusa-ku, Nagoya, Japan

A new core design code system, GALAXY/COSMO-S, has been developed. GALAXY is a lattice physics code which is used to generate a cross section data for the PWR core simulator COSMO-S. This paper presents a new robust cross section represen-tation methodology for COSMO-S. The development of a cross section representation methodology has been carried out based on the theoretical and quality engineering approach. The concept of quality engineering to achieve high quality and high produc-tivity at the same time has been incorporated in the cross section representation meth-odology to establish the robustness and the efficiency in actual applications. In order to validate this cross section representation methodology, direct comparisons with a lattice physics code have been made on various calculation conditions. As a result, it is concluded that a new cross section representation methodology has enough robust-ness and accuracy to generate a cross section data for a PWR core simulator.

10:45 AMImproving Cross Sections Via Spectral Rehomogeniza-tionAldo Dall’Osso and Daniele Tomatis(1), Yun Du(2)1) AREVA NP, Tour AREVA, 92084 Paris La D´efense Cedex, France. 2) Ecole Nation-ale Sup´erieure des Mines de Paris, Paris, France

Cross sections used in nodal calculations come from infinite medium flux homog-enization. Their dependencies from physical working conditions are solved through interpolation on common multi-parameterized tables. If the difference between the neutron spectrum in the true environment and in the infinite medium is relevant, these averaged values do not permit to represent true reaction rates. For example, the er-ror induced is not negligible with neighboring assemblies with different compositions. This is the case of UO2 assemblies placed near MOX assemblies. The purpose of the method presented here is determining a homogenization correction to multi-group cross sections, corresponding to the difference between the environmental and infinite medium spectrum. An investigation of the feasibility of using the spectral rehomogeni-zation as method to model the dependence of the cross sections on local conditions, such as the insertion of a control rod bank, is presented.

11:10 AMInvestigation on Macroscopic Cross Section Model for BWR Pin-By-Pin Core AnalysisTatsuya Fujita, Kenichi Tada, Akio Yamamoto, Yoshihiro Yamane(1), Shinya Kosaka, Gou Hirano(2)Graduate School of Engineering, Nagoya University, Nagoya, Japan. 2) TEPCO SYS-TEMS CORPORATION, Koto-ku, Tokyo, Japan

A cross section model used in the pin-by-pin core analysis for BWR is investigated. In the pin-by-pin core calculation method, pin-cell averaged cross sections are calculated for many combinations of state and history variables that have influences on the cross section and are tabulated prior to the core calculations. Variation of a cross section in a core simulator is classified into two different types, i.e., the instantaneous effect and the history effect. The instantaneous effect is incorporated by the variation of cross section which is caused by the instantaneous change of state variables. For this ef-fect, the exposure, the void fraction, the fuel temperature, the moderator temperature and the control rod are used as indexes. The history effect is the cumulative effect of state variables. We treat this effect with a unified approach using the spectral history. To confirm accuracy of the cross section model, the pin-by-pin fission rate distribution and the k-infinity of fuel assembly which are obtained with the tabulated and the refer-ence cross sections are compared. For the instantaneous effect, the present cross section model well reproduces the reference results for all off-nominal conditions. For the history effect, however, considerable differences both on the pin-by-pin fission rate distribution and the k-infinity are observed at high exposure points.

11:35 AMHomogenization of Cross Sections and Related Perfor-mance Improvements in PWR Burnup using the PEN-BURN/PENTRAN Burnup SequenceK. Manalo, R. Cerge, D. Wegner, and G. SjodenDepartment of Nuclear and Radiological Engineering, University of Florida, FL USA

A standard homogenization/condensation approach for homogeneous macroscopic cross sections has been implemented with the new HMIX code, used in conjunction with the PENBURN/PENTRAN burnup sequence. PENBURN is a zone-based burnup/depletion code that uses PENTRAN, a 3-D parallel Sn multigroup code. The HMIX code permits the creation of modified homogeneous microscopic cross sections which are constructed from heterogeneous microscopic cross sections for utilization in PEN-BURN. In HMIX, reaction rate factors for collapsing heterogeneous macroscopic cross sections into homogeneous macroscopic cross sections are based on initial hetero-geneous transport simulations, where the resulting fine mesh scalar fluxes defined by detailed heterogeneous material specifications are required. Therefore, the procedure automatically enables proper burnup and depletion of isotopes even for highly homog-enized fuel zones. To demonstrate the methodology, we examined the procedure of homogenization in a standard PWR fuel pin, and the impact of homogenized cross sections, flux preconditioning, and predictor corrector methods for keff as a function of integrated fuel burnup. Overall, the PWR models demonstrate good consistency between the homogenized fuel-clad and heterogeneous models.


4H - Core Physics Methods - IISession Chairs: Ronald J. Ellis (Oak Ridge National Laboratory – ORNL), Rakesh Chawla (SFIT, Lausanne - EPFL)

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1:30 PMDevelopment of Iterative Transport - Diffusion Methodol-ogy for LWR Analysis Damon R. Roberts(1), Mohamed Ouisloumen(2), Vefa N. Kucukboyaci(2) and Kostadin N. Ivanov(1) 1) Department of Mechanical and Nuclear Engineering, The Pennsylvania State Uni-versity. 2) Westinghouse Electric Company

The use of two group nodal diffusion calculations utilizing assembly-homogenized cross-sections and assembly discontinuity factors is a major cause of uncertainty in Light Water Reactor (LWR) core analysis. This paper presents the work being done on the development of an Iterative Transport-Diffusion Methodology (ITDM) for Light Water Reactor (LWR) core analysis in order to reproduce high-fidelity solutions of neutron transport methods in an efficient manner. The feasibility of the ITDM has been demonstrated on a set of challenging Two-Dimensional (2-D) LWR mini-core bench-mark problems. The current efforts are focused on the analysis of 3-D LWR mini-core benchmark problems and the presented preliminary results for the 3-D BWR problem are very encouraging. The next steps include further improvements of the efficiency and accuracy of the ITDM and its application to realistic 3-D LWR (PWR and BWR) core models with depletion.

1:55 PMA Comparison between Nodal Expansion Method and Nodal Green’s Function MethodWang Deng-Ying, Li Fu, Hu Yong-Ming, Guo Jiong, Wei Jin-Feng, Zhang Jing-Yu Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing, China

This paper presents a unified formulation of the Nodal Expansion Method (NEM) and Nodal Green’s Function Method (NGFM) in Cartesian geometry although there is a significant difference between them. Both methods employ the same inner iterative scheme namely Row-Column iteration strategy to solve the interface current equation. It’s generally believed that the NEM is somewhat faster than the NGFM. However, calculations of IAEA3D benchmark problem carried out by newly implemented NGFM and NEM show that not only the accuracy but also the performance of the NGFM are better than that of the NEM in Cartesian geometry. Both the NGFM and NEM are extended to solve neutron diffusion equation in cylindrical geometry. Since the traditional transverse integration fails to produce a 1-D transverse integrated equation in θ-direction, a simple approach is introduced to obtain this equation in θ-direction. The 1-D transverse integrated equations in r-direction are solved by the NEM using the special polynomials and by the NGFM using Green’s function based on modi-fied Bessel function respectively. The same iterative scheme employed for Cartesian geometry can be readily applied to the cylindrical geometry case. The Cylindrical Nodal Expansion Method (CNEM) and the Cylindrical Nodal Green’s Function Method (CNGFM) codes are developed and applied to Dodd’s r-z benchmark problem. The results show that both the CNEM and CNGFM are capable of very high performance and accuracy in cylindrical geometry. Meanwhile this paper demonstrates that nodal methods have prominent advantages over traditional finite difference method in both Cartesian geometry and cylindrical geometry.

2:20 PMMixed Energy Reactor Simulations using the Discrete Generalized Multigroup MethodBenoit Forget and Lei ZhuDepartment of Nuclear Science and Engineering, Massachusetts Institute of Technol-ogy, Cambridge, MA, USA

An innovative approach to mixed energy simulations was proposed. This approach is based on the generalized multigroup theory using discrete orthogonal polynomials. The discrete expansion provides a link between a coarse group structure and a fine group structure, thus allowing for coarse group calculations to provide an approxima-tion of the within group spectral information. A mixed energy scheme can be adapted from this work by which certain regions are based on the coarse group energy struc-ture which is a subset of a finer group structure that can be used in certain regions. The higher moment equations of the coarse group calculations are used to provide fine group spectral information at the boundaries of the regions. Simple 1-D diffusion results indicate that fine group flux can be reproduced accurately in a mixed energy solution and that significant improvement of the eigenvalue estimate can be obtained when comparing with common coarse and fine multigroup solutions.

2:45 PMDevelopment of the Triangular Nodal Method TRIPEN for Prismatic HTR AnalysisBrendan Kochunas, Yunlin Xu and Thomas DownarDepartment of Nuclear Engineering, University of Michigan, Ann Arbor, MI USA

A newly developed triangular polynomial expansion nodal method (TRIPEN) for solv-ing the multi-group diffusion equation is presented. This new method improves on the previous triangle-based polynomial expansion nodal method (TPEN) for hexagonal core analysis by explicitly treating the six triangles within the node as the homogenized region. The TRIPEN method provides more versatility for the neutronics treatment of the hexagonal assembly of the prismatic HTR which has large off center circular control rod regions. The TRIPEN method results in a linear system of equations for the one node problem which is considerably simpler than TPEN.. The new method, TRIPEN, is compared to the performance of the previous method, TPEN using both hexagonal and triangular based cross sections. A reference transport solution for a 2-D seven hexagon (assembly) VHTR-like problem is solved using the HELIOS lat-tice physics code. Each method is used to predict the control rod worth of the system, where the central assembly contains the control rod. Cross sections for the nodal methods are generated based on single assembly and seven assembly models to bet-ter understand the source of errors. Overall TRIPEN shows very good agreement with the HELIOS transport solution, predicting a rod worth that is ~240 pcm Δρ different from the transport solution when using discontinuity factors, while TPEN predicts a rod worth 1145 pcm Δρ different from the transport solution.


4J - Core Physics Methods - IIISession Chairs: Bojan Petrovic (Georgia Technology University), Yung-An Chao (Westinghouse)

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3:30 PM3D Finite Element Method for the Lambda Modes Problem in Hexagonal-Z Geometry S. González-Pintor and G. Verdú(1), D. Ginestar(2) 1) Departamento de Ingeniería Química y Nuclear, Universidad Politécnica de Valen-cia, Valencia, Spain. 2) Departamento de Matemática Aplicada, Universidad Politéc-nica de Valencia, Valencia, Spain

A 3D Finite Element Method (FEM) has been developed to compute a set of dominant modes of a nuclear power reactor with Z-Hex geometry. This method is based on a mesh made of triangular prisms and assumes that the flux can be approximated by means of a truncated polynomial expansion on each one of the elements in which the domain defining the problem is discretized. The polynomial basis used is defined by means of a tensor product of polynomials in the z variable per polynomials in the XY plane. The performance of the method is evaluated computing the dominant lambda modes of the 3D benchmark problems VVER 1000 and VVER 440.

3:55 PMEffect of Void Dependent Reactivity Modeling Bias on BWR Axial Power TiltYu Han(1), Chuntao Tang(1), Shaohong Zhang(1) and Yung-an Chao(1,2) 1) School of Nuclear Science and Engineering, Shanghai Jiao Tong University, Shang-hai, China. 2) Westinghouse Electric Company, Pittsburgh, Pennsylvania, USA

Inspired by a Pennsylvania State University study reported in the PHYSOR08 con-ference related to BWR axial power tilt, an investigation is made as to if there could be any intrinsic deficiency in the conventional nodal diffusion theory leading to a tilt in BWR axial power prediction. It is found that the two-group nodal diffusion theory itself does not cause axial power tilt compared to multigroup deterministic 3D trans-port calculation. However, the usual practice of flat source approximation in 2D lattice codes based on the Collision Probability (CP) method or the Method of Characteristics (MOC) can cause a void dependent reactivity bias when used to generate the node homogenized cross-sections, if the same geometry meshing modeling is used for all the axial sections over which the void fraction varies severely. The reactivity bias can be removed by using lattice code models based on SN method or MOC method with linear instead of flat source approximation. In absence of the void dependent reactiv-ity bias, there is no axial power tilt between two-group nodal diffusion calculation and multi-group deterministic 3D transport calculation. Therefore there is no intrinsic defi-ciency in the nodal diffusion theory that causes BWR axial power tilt. But one has to be careful with using the flat source approximation and the generic-meshing modeling in generating homogenized nodal diffusion cross-sections for high void cases. As for any possible axial power tilt between multi-group deterministic 3D transport calculation and continuous energy Monte Carlo 3D simulation, that would be a different issue.

4:20 PMGeneralized Equivalence Theory for Checkerboard Con-figurations in Natural-Uranium CANDU LatticesEleodor NichitaFaculty of Energy Systems and Nuclear Science, University of Ontario Institute of Technology, Oshawa, ON, Canada

CANDU reactors are heavy-water moderated and cooled. They consist of a horizontal, cylindrical, non-pressurized calandria vessel which contains the heavy-water modera-tor and is penetrated axially by fuel channels. Fuel channels hold the fuel bundles, which are cooled by the flow of coolant at high temperature and pressure. The coolant flows in opposite directions in adjacent channels. For a CANDU 6 reactor, there are 380 fuel channels, each holding 12 37-element fuel bundles, approximately 10-cm in diameter and 50-cm long each. The distance between channels (lattice pitch) is 28.575 cm. CANDU reactors in current operation are fuelled exclusively with natu-ral uranium. Traditionally, neutronics calculations for natural-uranium CANDU lattices use standard homogenization and therefore do not employ discontinuity factors. It is therefore of interest to find out whether gains in accuracy are to be made by adopting Generalized Equivalence Theory (GET) and hence the use of discontinuity factors. Of particular interest are configurations in which neutronic properties of lattice cells alternate in a checkerboard pattern. Such patterns occur in CANDU reactors because of alternating fuelling directions in neighbouring channels and because neighbouring channels belong to different coolant loops or different passes within the same loop. The work presented in this paper assesses the accuracy of the standard homogeniza-tion method when applied to checkerboard configurations and investigates whether the use of GET yields more accurate results. The conclusion is that for checkerboard configurations in natural-uranium CANDU lattices use of GET achieves only modest accuracy gains due to the fact that discontinuity factors vary negligibly with local bur-nup and void conditions.

4:45 PMA Response Matrix Method for Improved Modeling of Neutron Streaming in Large Control Rod Holes of Pris-matic VHTR CoresYeon Sang Jung, Changho Lee, and Won Sik YangArgonne National Laboratory, Argonne, IL, USA

This paper presents a response matrix method developed for accurate modeling of neutron streaming through empty, large control rod holes in VHTRs. In this approach, a response matrix based on transport solution is derived for each control rod channel region and embedded in the whole-core transport solution scheme. Depending on the region geometry only, each element of the response matrix represents the outgoing partial current at a surface due to a unit incoming partial current at another surface. In order to improve the axial solution accuracy, this response matrix approach was incor-porated into the DeCART code that solves whole-core transport problems by coupling two-dimensional MOC and one-dimensional nodal solutions. Verification test results showed very good agreements in control rod worth and axial power distributions with MCNP5 solutions.

5:10 PMThe Possibility to Online Reconstruct the HTR In-Core Power DistributionFu Li, Xuhua Zhou, Jiong Guo, Dengying Wang, Jinfeng Wei, Jingyu Zhang, Chen HaoInstitute of Nuclear and New Energy Technology (INET), Tsinghua University, Beijing, China

The necessity, demand and status to online monitor the in-core power distribution of HTR are reviewed. The methodology to reconstruct in real-time the in-core power dis-tribution from the finite number of ex-core detector readings is introduced. The difficulty and key issues induced by particularity of HTR are discussed. The research progress, including high order harmonics calculation on cylindrical geometry, three-dimensional diffusion model with accurately modeled control rod region, three-dimensional ex-core detector spatial response function calculated by adjoint discrete ordinates transport method, are presented.


4K - Core Physics Methods - IVSession Chairs: Rakesh Chawla (SFIT, Lausanne - EPFL), Bojan Petrovic (Georgia Technology University)

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08:00 AMMultilevel Criticality Computations in AREVA NP’s Core Simulation Code ARTEMISRené van GeemertAREVA, AREVA NP GmbH, 91052 Erlangen, Germany

This paper discusses the multi-level critical boron iteration approach that is applied per default in AREVA NP’s whole-core neutronics and thermal hydraulics core simulation program ARTEMIS. This multi-level approach is characterized by the projection of variational boron concentration adjustments to the coarser mesh levels in a multi-level rebalancing hierarchy that is associated with the nodal flux equations to be solved in steady-state core simulation. At each individual rebalancing mesh level, optimized variational criticality tuning formulas are applied. The latter drive the core model to a numerically highly accurate self-sustaining state (i.e. with the neutronic eigenvalue being 1 up to a very high numerical precision) by continuous adjustment of the boron concentration as a system-wide scalar criticality parameter. Due to the default applica-tion of this approach in ARTEMIS reactor cycle simulations, an accuracy of all critical boron concentration estimates better than 0.001 ppm is enabled for all burnup time steps in a computationally efficient way. This high accuracy is relevant for precision optimization in industrial core simulation as well as for enabling accurate reactivity perturbation assessments. The developed approach is presented from a numerical methodology point of view with an emphasis on the multi-grid aspect of the concept. Furthermore, an application-relevant verification is presented in terms of achieved coupled iteration convergence efficiency for an application-representative industrial core cycle computation.

08:25 AMEvaluation of Gamma Scanning in Oskarshamn2 with SIMULATE-5Sten-Örjan Lindahl(1), Tamer Bahadir(2) 1) Studsvik Scandpower AB, Västerås, Sweden. 2) Studsvik Scandpower, Inc., New-ton, MA, USA

SIMULATE-5, Studsvik Scandpower’s next generation nodal code, has been bench-marked against recent gamma scan measurements of Oskarshamn2, an ABB/West-inghouse design BWR, and results are presented in this paper. Gamma scan mea-surements following the Cycle-32 operation includes 48 assembly gamma scans of advanced Optima2 and Atrium-10B fuels. Two Optima2 fuel assemblies were disas-sembled for more detailed gamma scanning of 49 fuel rods. SIMULATE-5 is a three dimensional multi-group analytical nodal code with microscopic depletion capability. Its state of the art neutronic modules have been developed based on first principals with considerations of today’s highly heterogeneous cores. SIMULATE-5 has a built in capability of tracking Ba140 isotopics in a node as well as fuel pin level which makes the evaluations of gamma scans straightforward. The assembly gamma scan comparisons reveal that SIMULATE-5 accurately predicts the nodal Ba140 densities, hence nodal/assembly powers, regardless of fuel assembly type, burnup or location of the fuel assembly in the core. Heterogeneities in the axial direction, due to enrich-ment/burnable absorber zoning, part length fuel and spacer/grids are handled well. Pin gamma scan evaluations show that SIMULATE-5 offers significant improvements in accuracy to calculate pin powers vs. existing methods, especially for the fuel as-semblies with control rod which exhibits strong radial power/flux gradients. The last part of this paper summarizes the benchmark results for core tracking calculations for the last 18 cycles of Oskarshamn2.

08:50 AMValidation of BWR Simulation Methods via Bundle and Pin-By-Pin Power Gamma Scan ResultsBrian R. Moore and John P. ReaGlobal Nuclear Fuel, Wilmington, NC USA

Gamma scan is a non-destructive method to determine the relative fission product inventory in nuclear fuel, and is used to validate the power distributions calculated by BWR lattice and 3-D simulator codes. A bundle gamma scan campaign was completed in 2002 at the Cofrentes Nuclear Power Plant. Fifty bundles were examined including multiple vendor 9x9 and 10x10 bundle designs. In 2005 an additional bundle gamma scan campaign was conducted, measuring fifty 10x10 bundles. In 2006, pin-by-pin

gamma scan measurements were made on two 10x10 Global Nuclear Fuel designed fuel assemblies. The agreement between the measurements and predictions using the TGBLA06 lattice physics code and the PANAC11 BWR core simulator is excellent, with radial RMS errors approximately 2% at the bundle level and less than 5% at a nodal level. The data validate the applicability of the assembly power uncertainty for mod-ern core and fuel designs and operational strategies. Additional observations include confirmation of the consistency between validation using operational TIP data and the validation resulting from a gamma scans.

09:15 AMParallel Computing in SCALEM. D. DeHart, M. L. Williams, and S. M. BowmanOak Ridge National Laboratory, Oak Ridge, TN USA

The SCALE computational architecture has remained basically the same since its in-ception 30 years ago, although constituent modules and capabilities have changed significantly. This SCALE concept was intended to provide a framework whereby independent codes can be linked to provide a more comprehensive capability than possible with the individual programs – allowing flexibility to address a wide variety of applications. However, the current system was designed originally for mainframe computers with a single CPU and with significantly less memory than today’s personal computers. It has been recognized that the present SCALE computation system could be restructured to take advantage of modern hardware and software capabilities, while retaining many of the modular features of the present system. Preliminary work is be-ing done to define specifications and capabilities for a more advanced computational architecture. This paper describes the state of current SCALE development activities and plans for future development. With the release of SCALE 6.1 in 2010, a new phase of evolutionary development will be available to SCALE users within the TRITON and NEWT modules.

09:40 AMBWR Modeling Capability and SCALE/TRITON Lattice-to-Core Integration of the NESTLE Nodal SimulatorJack Galloway, Hermilo Hernandez, G. Ivan Maldonado(1), Matt Jessee, Emilian Popov, Kevin Clarno(2)1) Nuclear Engineering Department, The University of Tennessee, Knoxville, Ten-nessee USA. 2) NSTD Reactor Physics Group, Oak Ridge National Laboratory, Oak Ridge, Tennessee USA

This article reports the status of recent and substantial enhancements made to the NESTLE nodal core simulator, a code originally developed at North Carolina State University (NCSU) of which version 5.2.1 has been available for several years through the Oak Ridge National Laboratory (ORNL) Radiation Safety Information Computa-tional Center (RSICC) software repository. In its released and available form, NESTLE is a seasoned, well-developed and extensively tested code system particularly useful to model PWRs. In collaboration with NCSU, University of Tennessee (UT) and ORNL researchers have recently developed new enhancements for the NESTLE code, in-cluding the implementation of a two-phase drift-flux thermal hydraulic and flow redis-tribution model to facilitate modeling of Boiling Water Reactors (BWRs) as well as the development of an integrated coupling of SCALE/TRITON lattice physics to NESTLE so to produce an end-to-end capability for reactor simulations. These latest advance-ments implemented into NESTLE as well as an update of other ongoing efforts of this project are herein reported.


4L - Advances and Validation of Reactor Physics Tools - ISession Chairs: Blair Bromley (AECL), Sten-Orjan Lindahl (Studsvik Scanpower AB)

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10:20 AMStatus of Updating the Cross Section Library of the Juelich Reactor Dynamic Codes TINTE/MGT to ENDF-B-VII K. Nünighoff, J. Li, C. Druska, H.-J. AlleleinForschungszentrum Juelich GmbH, Institut fuer Energieforschung (IEF), Juelich, Ger-many

The nuclear cross section library SPECLIB used in the Juelich reactordynamic codes TINTE and MGT are based on the 43 groups MUPO library. This paper reports on updating the cross section library to ENDF-B-VII data. The generation of transport, absorption, and fission cross section as well as the preparation of resonance tables will be described. Absorption cross sections of 232Th, 238U, 240Pu, and 242Pu in the resonance range as function of temperature and nuclide concentration were cal-culated with MCNP4C. First MGT simulations performed with the new ENDF-B-VII based SPECLIBN2009 libray and 43 energy groups will be presented and compared to the results obtained with the old MUPO based SPECLIBN-r04 library. First results obtained with the new library for a steady state as well as for selected transients will be presented. Deviations compared to results calculated with the old library will be discussed. It will be shown, that the new ENDF-B-VII based library influences strongly keff , but shows only negligible deviations in case of temperatures, power profiles, and neutron fluxes. With the new updated TISPEC library SPECLIBN2009 the Juelich core design program system VSOP and the Juelich reactordynamic codes TINTE/MGT are now based on the same evaluated nuclear data library ENDF-B-VII.

10:45 AMX2: a Coupled Neutronic - Thermal-Hydraulic Code Sys-tem for Compact Research Reactor CoresHarald Breitkreutz, Anton Röhrmoser, Winfried PetryForschungsneutronenquelle Heinz Maier-Leibnitz, FRM II, Technische Universität München, Garching, Germany

The neutronic Monte Carlo code MCNPX, the thermal-hydraulic CFD code CFX and a modified version of the MonteBurns burn-up code have been coupled to satisfy the increasing demand for 3-dimensional results with high spatial resolution in nuclear re-actor physics for compact research reactor cores. The code system, X2, was verified by a code-to-code comparison on calculated properties and by the correct prediction of measured properties of the compact core of the FRM II research reactor.

11:10 AMDevelopment and Verification of Quick and Precise Core Physics Analysis CodeYongming Hu, Zhihong Liu(1), Xiu’an Shi(2), Zhiwei Zhou(1)1) INET, Tsinghua University, Beijing, China. 2) China National Nuclear Corporation, Beijing, China

The study on core physics analysis code CPACT developed by Tsinghua University is presented. Some factors affecting the accuracy of CPACT calculation have been found. After these problems were solved with more suitable methods, the modified CPACT code achieves high accuracy in reactor core calculations for in-core fuel man-agement of Nuclear Power Plants.

11:35 AMA Fast and Flexible Reactor Physics Model for Simulat-ing Neutron Spectra and Depletion in Fast ReactorsGD Recktenwald, LA Bronk and MR DeinertDepartment of Mechanical Engineering, The University of Texas at Austin, Austin, TX USA

Determining the time dependent concentration of isotopes within a nuclear reactor core is central to the analysis of nuclear fuel cycles. We present a fast, flexible tool for determining the time dependent neutron spectrum within fast reactors. The code (VBUDS: visualization, burnup, depletion and spectra) uses a two region, multigroup collision probability model to simulate the energy dependent neutron flux and tracks the buildup and burnout of 24 actinides, as well as fission products. While originally developed for LWR simulations, the model is shown to produce fast reactor spectra that show high degree of fidelity to available fast reactor benchmarks.


4M - Advances and Validation of Reactor Physics Tools - IISession Chairs: Blair Bromley (AECL), Sten-Orjan Lindahl (Studsvik Scanpower AB)

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08:00 AMShielding Fuel Assemblies Used to Protect the Beltline Weld of the Reactor Pressure Vessel from Fast Neutron Radiation in Ringhals Unit 3 and 4U. Sandberg, H. Nylén, J. Roudén and P. Efsing(1), J. Marten(2)1) Ringhals AB, Väröbacka, Sweden. 2) AREVA, AREVA NP GmbH, Erlangen, Ger-many

The beltline weld on the reactor pressure vessel of Ringhals units 3 and 4 has a lifetime estimate of 40 years under the current operating conditions. In the event of power up rates and lifetime extension the irradiation embrittlement of the beltline weld may be a limiting condition. One way to solve this problem is to limit the fast neutron radiation on the reactor pressure vessel. This paper will focus on a solution with spe-cially designed assemblies for the positions that have the highest influence on the fast neutron dose for the beltline weld.

08:25 AMA Method for Channel-Bow Indication by Neutron Flux MeasurementsJohn Loberg, Michael Österlund(1), Klaes-Håkan Bejmer, Jan Blomgren(2) 1) Uppsala University, Department of physics and astronomy, Division of applied nu-clear physics, Uppsala, Sweden. 2) Vattenfall Nuclear Fuel AB, Stockholm, Sweden

Channel bow in boiling water reactors (BWR) can be detected by simultaneous mea-surement of the thermal and fast neutron fluxes. There is a robust linear correlation between the ratio of the thermal and fast neutron fluxes and the void fraction close to the detector through the core. The moderation will be modified by channel bow and this change in moderation will affect the thermal neutron flux strongly and the fast flux weakly, i.e., the thermal flux will act as an indicator that the moderation or power has changed but the fast flux will reveal if the change is due to difference in moderation or power. The thermal and fast fluxes are furthermore affected in opposite directions, thus enhancing the deviation from linearity of the ratio between the fluxes, caused by channel bow. By comparing the void fraction calculated by a core simulator with the void fraction predicted by the ratio of the thermal and fast neutron fluxes, channel bow can be detected as a deviation from the linear void correlation of 4% per mm bow. This implies that it is reasonable to detect channel bow 2 mm or larger, assuming that the uncertainty of the detectors does not exceed 5%.

08:50 AMDevelopment of an Efficient PENTRAN Model for Neu-tron Flux Simulation of the UFTR Shield TankAmrit D. Patel and Alireza HaghighatUniversity of Florida, Gainesville, FL USA

The motivation of this work is to support development of a burnup reconstruction de-vice which will provide 3-D characterization of spent nuclear fuel at the University of Florida Training Reactor (UFTR) within the shield tank experimental unit. In order to facilitate the design of such a device, a procedure is required to quickly and ac-curately determine radiation fields within the shield tank region. Utilizing both a 3-D Monte Carlo (MCNP5) and a 3-D deterministic particle transport code (PENTRAN), multi-group neutron flux distributions are calculated. The accuracy and efficiency of the PENTRAN code based on flux distributions throughout the reactor core, graphite reflector, and shield tank regions are assessed and further compared with MCNP5 results. It is demonstrated that the deterministic PENTRAN code package achieves accurate solutions at significantly reduced computational time compared to the analog Monte Carlo calculations.

09:15 AMPreliminary Feasibility Study of a Water Space Reactor with an Innovative Reactivity Control SystemVito Memoli, Andrea Bigoni, Antonio Cammi, Marco Colombo, Carlo Lom-bardi, Davide Papini, Marco E. RicottiPolitecnico di Milano – Department of Energy, Nuclear Division-CeSNEF, Milano, Italy

Power limitation represents a major issue within space applications aimed to human settlements on solar system planets. Among these planets, Mars is considered the most attractive because of its proximity to the Earth and the probable presence of min-erals which could be used by the settlers. In this frame, small size nuclear power plants can be an interesting solution to overcome the energy supply problem. This paper presents a preliminary feasibility study of a 100 kWe water space reactor, with the aim to design a system characterized by compactness, intrinsic safety and simplicity of the main reactor control components. To this end, an innovative reactivity control system, based on the control of the primary coolant mass flow rate, was proposed. The adop-tion of this system in the reactor design required a comprehensive core neutronics analysis in order to properly quantify the effect of the coolant on the reactor behavior also as a function of the fuel burn-up. The main results gained concerning neutron flux profiles and multiplication factors are discussed within the paper. Moreover, prelimi-nary results on long term reactivity control are presented, showing the possibility to operate the reactor for as long as 7 years with no need of human intervention.

09:40 AMAnalysis of Accuracy of Reactivity Balance Components in BN-600 Power Fast ReactorA.V. Moiseev, Yu.S. Khomyakov, M.Yu. Semenov, A.S. Seregin, A.M. Tsi-boulya(1), V.A. Zheltyshev, V.V. Maltsev(2), E.F. Seleznev, A.A. Belov(3), S.B. Belov, M.R. Farakshin, B.A. Vasiliev(4)1) State Scientific Center of the Russian Federation – Institute for Physics and Power Engineering named after А. Leypunsky (SSC RF - IPPE), Obninsk, Russia. 2) Branch of Rosenergoatom Concern – Beloyarsk NPP named after I. Kurchatov (BNPP), Za-rechny, Russia. 3) Nuclear Safety Institute of Russian Academy of Sciences (IBRAE), Moscow, Russia. 4) Experimental Design Bureau of Mechanical Engineering named after I. Аfrikantov (ОКBМ), Nizhny Novgorod, Russia

Reactivity balance is a key element for nuclear safety substantiation for reactor fa-cilities of any type. Commercial operation of BN-600 power fast reactor in Beloyarsk NPP stimulated work on finding optimal balance between conservative safety mar-gins and reactor performance. This paper is devoted to the studies on the possibil-ity of achieving high indices on the accuracy of reactivity balance justification. It is described calculation and experimental methods for determination of components of BN-600 reactivity balance. Results of comparison calculation and experimental data are presented. Taking into account experimental data and verification of codes facili-tates increase of accuracy of max excess reactivity forecast up to 0.2%Δk/k. Forecast of the BN-600 reactor core subcriticality after SR setting can be made to the accuracy of 0.2-0.4%Δk/k, accuracy of SHR system worth forecast is 7%, single SR and reac-tor safety system 7-8%. These values make it possible to significantly decrease (by a factor of 2) conservatism of evaluation of these parameters justifying nuclear safety of the BN-600 reactor core.


4N - Reactor Operation & DesignSession Chairs: Dan Ilas (ORNL), Cenk Güler (Westinghouse)

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1:30 PMLoading Pattern Optimization Cooperatively Using Two New Algorithms Zhaohu Gong(1,2), Kan Wang(1), and Dong Yao(2) 1) Department of Engineering Physics, Tsinghua University, Beijing, P.R.China. 2) Na-tional Key Laboratory of Science and Technology on Reactor System Design Technol-ogy, Nuclear Power Institute of China, Chengdu Sichuan, P.R.China

Loading pattern optimization (LPO) for a PWR in nuclear power plant contains three parts: fuel assembly location optimization, burnable poison placement optimization, and used fuel assembly orientation optimization. To solve the former two parts, this paper devises an innovative stochastic evolutionary algorithm—Interval Bound Algo-rithm (IBA), which can optimize fuel assembly location and burnable poison place-ment together. IBA just uses the fuel assembly’s infinite multiplication factor to get rid of unfavorable patterns and to explore new promising solution space. To solve the last part, this paper applies Estimation of Distribution Algorithms (EDAs), which also belong to evolutionary algorithms. These three parts depend on each other, so it is better not to solve them separately. In order to optimize these parts in a coupled way, we use Symbiotic Co-evolutionary Algorithm (SCA) to incorporate IBA and EDAs. This technique could reflect the real optimization process. Based on these algorithms, the corresponding LPO code of IBALPO is developed. To avoid search direction to offset for inconsistency between the LP search code and the design code, IBALPO directly adopts production core design code to evaluate LPs in a parallel computation environ-ment. Finally, this code system is used to solve a realistic reload problem to show its performance. Obtained results have illustrated that IBALPO is efficient and robust. It can find satisfying LPs in two days using 18 CPUs after evaluating about 10000 LPs for a core containing 157 assemblies.

1:55 PMStudy on the Mechanism and Efficiency of Simulated Annealing Using an LP Optimization Benchmark Prob-lemQianqian Li, Xiaofeng Jiang and Shaohong Zhang School of Nuclear Science and Engineering, Shanghai Jiao Tong University, Shang-hai, China

Simulated Annealing Algorithm (SAA) for solving combinatorial optimization problems is a popular method for loading pattern optimization. The main purpose of this paper is to understand the underlying search mechanism of SAA and to study its efficiency. In this study, a general SAA that employs random pair exchange of fuel assemblies to search for the optimum fuel Loading Pattern (LP) is applied to an exhaustively searched LP optimization benchmark problem. All the possible LPs of the benchmark problem have been enumerated and evaluated via the use of the very fast and accu-rate Hybrid Harmonics and Linear Perturbation (HHLP) method, such that the mecha-nism of SA for LP optimization can be explicitly analyzed and its search efficiency evaluated. The generic core geometry itself dictates that only a small number LPs can be generated by performing random single pair exchanges and that the LPs are necessarily mostly similar to the initial LP. This phase space effect turns out to be the basic mechanism in SAA that can explain its efficiency and good local search ability. A measure of search efficiency is introduced which shows that the stochastic nature of SAA greatly influences the variability of its search efficiency. It is also found that using fuel assembly k-infinity distribution as a technique to filter the LPs can significantly enhance the SAA search efficiency.

2:20 PMDirect Optimization of Loading Pattern Search Using the HHLP Theory and the MILP MethodWencong Wang(1), Xiaofeng Jiang(1) and Yung-An Chao(1,2)1) School of Nuclear Science and Engineering, Shanghai Jiaotong University, Shang-hai, China. 2) Westinghouse Electric Company, Pittsburgh, Pennsylvania, USA

A direct optimization algorithm based on the Hybrid Harmonics and Linear Perturbation (HHLP) theory and the Mixed Integer Linear Programming (MILP) method is proposed to solve the loading pattern optimization problem. Unlike search methods based on stochastic or heuristic shuffling, the MILP method covers the complete search space for a direct optimization. The HHLP theory developed at SJTU can provide very fast and accurate calculation for extremely large perturbations. Combining HHLP with MILP, the entire LP search process and the LP evaluation process are incorporated into a direct search mathematical programming model. Highly accurate optimum solu-tions can be generated in this direct search over the entire search space. Since HHLP still deals with an eigenvalue problem which can not be efficiently handled by an MILP algorithm, a process of iteratively building the HHLP based linear MILP model is de-veloped. The iteration process is very effective, although it can not guarantee reaching absolutely the global optimum. The HHLP based linear MILP model can be solved efficiently by the generic MILP software CPLEX (commercialized by IBM ILOG). Ap-plications to several test problems, including depletion and BA loading assignment, showed very good results.

2:45 PMDevelopment of an Automatic Tool for BWR Loading Pattern Optimization by Utilizing Genetic AlgorithmChien-Hsiang Chen(1), Cheng-Hsi Wu(2)1) Institute of Nuclear Energy Research, Taoyuan, Taiwan, R.O.C. 2) Taiwan Power Company, Taipei, Taiwan, R.O.C.

An automatic tool, INERLOAD, is developed to search for the optimal in-core fuel loading pattern of boiling water reactors by using genetic algorithm. Sub-pattern swap concept is applied for the crossover operator. Different fresh fuel check-board arrange-ments including two standard checkboard patterns and experience-based check-board patterns are implemented into the code. The objective function of a loading pattern consists of the keff at the end of cycle, the maximum power peaking factor, and the cold shutdown margin at the beginning of cycle with the corresponding weight-ing factors. In order to evaluate these parameters, the steady-state core simulator, SIMULATE-3, is utilized to perform the three-dimensional calculations. Haling axial power shape is applied to estimate keff at the end of cycle and maximum power peak-ing factor while using SIMULATE-3. The code is tested for both long and short cycle length cases which are based on an actual core of BWR-6 reactor, Kuosheng Nuclear Power Plant. Results show that this automatic tool is very competitive to well-trained engineer’s designs for both cases.


4P - In-Core Fuel Management and Optimization - ISession Chairs: Juan Luis Francois Lacouture (Universidad Nacional Autónoma de México-UNAM), Serkan Yilmaz (GE)

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3:30 PMBWR Bundle Optimization by Linear Perturbation and MILP Tao Wang, Huanfeng Ye, Shaohong Zhang School of Nuclear Science and Engineering, Shanghai Jiao Tong University, Shang-hai, China

A BWR bundle optimization method is presented, which uses the commercial Mixed Integer Linear Programming (MILP) code CPLEX to solve the linear perturbation equations constructed from a Sensitivity Matrix (S-matrix). The S-matrix represents the pin power distribution response throughout depletion to perturbation in enrichment and Gd weight percent at all pin locations. With any reasonable reference bundle de-sign chosen, the calculated S-matrix has generic applicability. Numerical tests show that the S-matrix linear model is very accurate and CPLEX can find the global optimum solution very fast. Application to a 3D bundle design example results in very signifi-cant gain on thermal margin throughout depletion, suggesting a great potential of the method for engineering application.

3:55 PMCore Analysis, Design and Optimization of a Deep-Burn Pebble Bed ReactorB. Boer and A.M Ougouag Idaho National Laboratory, Idaho Falls, ID, USA

Achieving a high fuel burnup in the Deep-Burn pebble bed reactor design, while remaining within prescribes safety limits for fuel temperature, power peaking and temperature reactivity feedback, is challenging. The high content of Pu and Minor Actinides in the Deep-Burn fuel significantly impacts the thermal neutron energy spec-trum as compared to a ’standard’ UO2 fueled core. Regions in the pebble bed core near the graphite reflectors experience power and temperature peaking that result from the local softer neutron energy spectrum. Furthermore, the interplay of the Pu resonances of the neutron absorption cross sections at low-lying energies can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. To investigate the aforementioned effects a new code system based on existing codes has been developed for neutronic, thermal-hydraulic and fuel depletion analysis of Deep-Burn pebble bed reactors. A core analysis of a Deep-Burn Pebble Bed Modular Reactor (400 MWth) design has been performed for two Deep-Burn fuel types and possible improvements of the design with regard to power peaking and temperature reactivity feedback are identified.

4:20 PMBreed-and-Burn Depleted Uranium in Fast Reactors without Actinide SeparationFlorent Heidet and Ehud GreenspanUniversity of California Department of Nuclear Engineering, Berkeley, CA, USA

Using multi-recycling without actinides separation it is possible to establish a breed-and-burn mode of operation in fast reactor cores that are fed with depleted uranium without violating any of the presently accepted radiation damage constraints. The maximum burnup attainable using this mode of operation in large fast reactors is found to be 55%. This corresponds to two orders of magnitude increase in the uranium ore utilization relative to that achievable in once-through LWRs. The energy value of the depleted uranium accumulated worldwide, when used in the proposed breed-and-burn energy system, is equivalent to 3000 years of the present global nuclear capacity. Relative to LWR operating with the once-through fuel cycle, the fuel discharged from the breed-and-burn fast reactor and fuel cycle hereby proposed features, per unit of electricity generated: (a) ~40% the amount of TRU and Pu; (b) ~10% the inventory of 237Np and its precursors; (c) ~12% of the decay heat from TRU one year following discharge; (d) ~28% of the radiotoxicity; (e) ~7% the neutron emission rate one year following discharge. The fraction of the fissile isotopes in the discharged plutonium is comparable but the decay heat and neutron emission rate per unit mass of discharged plutonium are nearly half as large. The proposed mode of operation is expected to improve the economics and the proliferation resistance and, hence, may justify sooner deployment of fast reactors. The deployment of the suggested fast reactor system will constitute a significant step forward towards sustainable nuclear energy.

4:45 PMNeutronic Evaluation of Breed-and-Burn Reactor Fuel Types Using an Infinite-Medium Depletion Approxima-tionRobert Petroski, Benoit Forget, Charles ForsbergDepartment of Nuclear Science and Engineering, Massachusetts Institute of Technol-ogy, Cambridge, MA USA

A new method of analyzing breed-and-burn reactors is developed which allows the reactor’s equilibrium-cycle reactivity to be stated in terms of the depletion history of the fertile feed fuel. The method is centered on calculating the total number of neutrons produced and absorbed by a volume of feed fuel over its depletion history. The net number of neutrons produced or absorbed over the feed fuel’s life is termed its neutron excess. Neutron excess is found to be insensitive to a number of factors, including reactor geometry, cycle length, discharge burnup distribution, and equilibrium-cycle shuffling path. This allows the reactivity-burnup relationship for a breed-andburn reac-tor to be accurately estimated using simple one-dimensional models. One method of approximating the feed-fuel depletion history is to deplete an infinite medium of feed material in its self-produced spectrum. Performing a neutron excess calculation on the infinite-depletion case allows a reactivity estimate to be made, which is found to agree closely with reactivities calculated in actual models of a breed-and-burn system. The infinite-medium approximation is applied to compare the neutronic performance of different metal and ceramic fuels, including thorium. The minimum burnup and fast fluence required to sustain breed-and-burn operation is reported for each fuel. Ura-nium-zirconium alloys are found to be the best performing metal fuels while U3Si2 and uranium nitride with enriched nitrogen-15 were found to be the best performing ceramic fuels.

5:10 PMSingle-cycle and Equilibrium-cycle Reloading Optimiza-tion on Real Nuclear Power PlantZhihong Liu, Yongming Hu, Zhiwei Zhou(1), Xiuan Shi(2)1) INET, Tsinghua University, Beijing, China. 2) China National Nuclear Corporation, Beijing, China

To make fuel management optimization code be more effectively to meet the engineer-ing requirements of nuclear power plants, some improvements were executed on the optimization code with CSA (Characteristic Statistic Algorithm) algorithm, which is de-veloped by our team. The improved fuel management code was used on single-cycle reloading and equilibrium-cycle reloading optimization of QINSHAN Nuclear Power Plant of China. The results show that this code is quite helpful for in-core fuel manage-ment of nuclear power plants.


4Q - In-Core Fuel Management and Optimization - IISession Chairs: Juan Luis Francois Lacouture (Universidad Nacional Autónoma de México-UNAM), Serkan Yilmaz (GE)

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08:00 AMVVER-440 Control Rod Follower Induced Local Power Peaking Computational Benchmark: MCNP and KARATE SolutionsE. Temesvári, Gy. Hegyi, G. Hordósy, A. Keresztúri, Cs. MaráczyHungarian Academy of Sciences, KFKI Atomic Energy Research Institute (AEKI), Bu-dapest, Hungary

With the original VVER-440 follower design the relatively large amount of water in the coupler between the absorber and fuel part of the control assembly can cause sharp power peaking in the fuel rods next to the coupler. The power peaking can be espe-cially high after control rod withdrawal when the coupler reaches a low burnup level region of the adjacent assembly. Though the modernized coupler has a Hf plate in the critical region to suppress the power peak, the complicated structure needs a refer-ence Monte Carlo calculation as a basis of engineering code validation. The coupler mathematical benchmark was solved by the KARATE code system using the same methods and approximations as in case of NPP applications and the results were compared to that of the reference MCNP. The need for treating the Hf burnout in the reflector region was also investigated.

08:25 AMImprovement of the Conversion Ratio in PWRF. Damian, S. Douce, A. Bergeron(1), B. Gastaldi, A. Conti(2), F. Moutin, D. Pialla(3) 1) C.E.A Saclay, DEN/DANS/DM2S, Gif-sur-Yvette Cedex, France. 2) C.E.A Cadarache, DEN/DER, Saint-Paul-lez-Durance Cedex, France. 3) C.E.A Grenoble, DEN/DER, Grenoble Cedex, France

For the sustainable energy supply, the nuclear power appears as a mandatory solution to the green-house effect and the global warming. The future of nuclear power re-quires a clear strategy for Plutonium management. Pressurized water reactors (PWR) and the associated fuel cycle facilities represent the largest part of the French power plants. Furthermore, Light Water Reactor will produce the major part of the nuclear electricity during the current century. Thus, the continued deployment of LWR with their low uranium utilization leads to design alternative LWR concepts. In order to reach higher conversion ratio (>0.8), a hard neutron spectrum is required and mod-eration ratios lower than unity are being considered, this led to low moderation PWR Fuel Assembly (FA) designs. 239Pu can be produced from fertile nuclide 238U in all proven reactor types. The main issues are the void reactivity coefficient, the moderator temperature coefficient, the critical heat flux margin and the pressure drop. Physical aspects of HCPWR concepts, which represent an innovative PWR fuelled with mixed oxide and having higher conversion ratio than the standard PWR are presented with regard to the neutronic and the thermal-hydraulic aspects.

08:50 AMInvestigation of RCCA Absorber Depletion in Normal Plant Operation. Application to the Shutdown Margin AnalysisMarcel Bouffier(1), Sylvie Ebalard(2)1) AREVA, AREVA NP, Lyon Cedex, France. 2) AREVA, AREVA NP, Paris La Défense Cedex, France

To ensure that nuclear power plants operate correctly, it must be guaranteed that all the functional requirements of the Rod Control Cluster Assemblies (RCCAs) will be met with passing service cycles. For many years, the improvements made to the design of the RCCAs, in particular the enhancement of the wear strength of the control rods in the HARMONITM RCCAs by the ion nitriding process have provided a signifi-cant extension in their lifetime. In parallel, the operating modes on the PWR have been improved to adjust quickly the demand and the production. Very often, it is realized with the help of RCCAs, their insertions are more frequent and they are more deeply inserted. The consequence is a depletion of the absorbers during the normal plant operation. Today, both the mechanical wear and the neutronic wear of the RCCAs limit their expected time life in reactor. Within the Safety Analysis Report an extra margin of 100 pcm to cover this neutronic wear is applied. Checking this neutronic margin is now mandatory, and becomes a requirement to design the RCCAs. The paper seeks to detail the method, supported by an illustration from shutdown margin analysis.

09:15 AMSCWR Assembly Designs with Peripheral Moderator ChannelsPeng Zhang, Kan Wang, Ganglin YuDepartment of Engineering Physics, Tsinghua University, Beijing, China

SCWR assembly designs with peripheral moderator channels have been studied in this work. Compared with the existing designs, this type of arrangement makes the fuel rods more uniformly moderated, thus the fuel assembly has a flatter power and burnup distribution. In addition, this design can reduce the need of structure materi-als and increase the volume fraction of fuel cells compared with the 3 by 3 water rod designs, and hence the neutron economics. The assembly design parameters would be optimized further by, considering the thermal hydraulics, core neutronics and me-chanics analysis.

09:40 AMReview of Worldwide Reactor-Physics Codes and Their Applicability to the Current and Next-Generation CAN-DU® ReactorsW. Shen and M. DahmaniAtomic Energy of Canada Limited, Mississauga, ON, Canada

WIMS-AECL, DRAGON and RFSP are the three Industry Standard Toolset (IST) codes currently used in the CANDU® industry. WIMS-AECL is a 2-D lattice-cell code used for transport calculation and cross-section condensation in CANDU lattices. DRAGON is a 3-D supercell code used for transport calculation and incremental cross-section calculation of CANDU reactivity devices, which are perpendicular to the fuel chan-nels. RFSP is a core-analysis code for CANDU 3-D static and dynamic analysis. To evaluate how other codes can be applied to current and future CANDU applications, the basic requirements of a reactor-physics toolset for CANDU analysis are outlined first. A high-level review of worldwide deterministic lattice-cell codes, deterministic core-analysis codes, and stochastic codes, are then performed and summarized with particular emphasis on those aspects that are relevant to CANDU analysis. The review concentrates on the main features of each code with respect to the reactor-physics method, numerical solution and programming aspects. Where relevant, the applicabil-ity of these codes to CANDU analysis is covered as well.


5A - Plant ImprovementsSession Chairs: C. Garzenne (EDF), Tanguy Courau (AREVA)

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08:00 AMFeasibility Study of Ultra-Long Life Fast Reactor Core ConceptT. K. Kim and T. A. TaiwoArgonne National Laboratory, Argonne IL USA

An ultra-long life core concept is proposed targeting capital and operational cost re-ductions and ultra-high discharge burnup in a fast reactor system. The core concept is achieved by derating the power density and adopting annular core geometry to maintain criticality for more than 40 years without refueling. The ultra-long life core has a specific power of ~10 MW/t and an average driver fuel discharge burnup of ~300 GWd/t. It is assumed such ultra-high burnup fuel can be developed within an advanced fuel cycle program. Several benefits are expected from the ultra-long life core concept such as capital and operational cost reductions, low proliferation risk, and effectively holding LWR spent fuel without disposal until technologies for a closed nuclear fuel cycle are developed and deployed. As future work, safety analysis, de-velopment of the advanced core cooling methods, and comparative cost analysis are expected.

08:25 AMStudies of Metal (U-Pu-Zr) Fuel in BOR-60 Reactor with Focus on Its Use in High-Power Sodium Fast ReactorsE.F.Mitenkova and N.V.Novikov(1), G.I.Gadzhiev and V.N.Syuzyov(2) 1) Nuclear Safety Institute of Russian Academy of Sciences, Moscow, Russia. 2) ATOMENERGOPROM, OJSC Joint Stock Company “State Scientific Center Re-search Institute of Atomic Reactors”, Ulyanovsk region, Russia

Metal (U-Pu-Zr) fuel compositions are considered as a perspective fuel for next gen-eration of sodium fast breeder reactors. Studies of irradiated eFA have shown high working capacity of the fuel elements with (U-10Zr) and (U- 18Pu-10Zr) fuel. Increase in FE diameter occurs basically due to cladding material swelling; there is practically no pressure upon cladding by the fuel. FE operational life is defined by dislocation per atom in constructional materials and gas cavity height for fission products to be not less than core height. Comparative analysis of experimental and calculated results of irradiated (U-Pu-Zr) fuel in BOR-60 reactor is discussed. Analysis results include neutron spectra, reaction rates, Kr and Xe accumulation in fuel. Estimated calculations of basic neutronic characteristics are provided for model of high-power sodium fast reactor with (U-Pu-Zr) fuel. Preliminary estimations of isothermal temperature coef-ficients of reactivity, sodium void worth, BR, etc. along with high negative fuel expan-sion reactivity coefficients provide the possibility for developing safe high power fast reactor with (U-Pu-Zr) fuel. Neutronic calculations are made using MONTEBURNS-MCNP-ORIGEN codes and nuclear data libraries being compiled from ENDF/B-VII and JENDL 3.3.

08:50 AMVoid Reactivity Decomposition for the Sodium-Cooled Fast Reactor in Equilibrium Closed Fuel CycleKaichao Sun(1,2), Jiri Krepel(1), Konstantin Mikityuk(1), Rakesh Chaw-la(1,2)1) Paul Scherrer Institut (PSI), Villigen, Switzerland. 2) Ecole Polytechnique Fédérale de Lausanne (EPFL), Lausanne, Switzerlan

The sodium-cooled fast reactor (SFR) is one of the most promising Gen-IV systems with many advantages, which however has a dominating neutronic drawback – a sig-nificantly large positive sodium void reactivity. The aim of this paper is to develop and apply a methodology, which should help better understand the causes for the sodium void effect and thereby lead to its further optimization. The study focuses mainly on the SFR equilibrium closed fuel cycle, along with its comparison to beginning-of-life conditions. Various voiding scenarios, corresponding to different spatial void distri-butions were analyzed, and the most conservative case (with voiding of just the in-ner and outer fuel zones) was selected as the reference scenario. The sodium void reactivity for this reference scenario was decomposed reaction-wise, isotope-wise, and energy-group-wise, the decompositions being based on neutron balance calcula-

tions using the ERANOS code. A strong correlation has been found between the flux distribution and the local void effect. The reaction-wise decomposition shows that the change in absorption is the main cause of the positive void reactivity, the changes in production and leakage making negative contributions. More specifically, the reduced 238U capture rate in the energy range below the 23Na elastic peak at 3 keV is a key cause for the positive void reactivity. In the same energy range, the 239Pu fission rate decreases, implying reduced production and hence improved void reactivity. The results for beginning of equilibrium closed fuel cycle show a 515 pcm (1 pcm = 10-5) void reactivity increase compared to the beginning-of-life condition, caused mainly by a less negative contribution of the production term.

09:15 AMConceptual Core Design of a Generation IV LFR Demon-stration PlantSara Bortot(1), Patrizio Console Camprini, Giacomo Grasso(2), Carlo Arti-oli, Stefano Monti(3)1) Politecnico di Milano, Department of Energy, Nuclear Engineering Division - CeS-NEF, Milano, Italy. 2) Università degli Studi di Bologna, Nuclear Engineering Labora-tory (LIN) of Montecuccolino, Bologna, Italy. 3) National Agency for the New Technolo-gies, Energy and Environment, Bologna, Italy

The Lead-cooled Fast Reactor (LFR) is one of the six concepts selected by the Gen-eration IV International Forum (GIF) as candidates for the next generation of nuclear power plants to be industrially deployed by 2040. Due to its significant technological innovations, the European Sustainable Nuclear Energy Technology Platform (SNETP) has recognized that GEN-IV LFR development requires -as a fundamental intermedi-ate step- the realization of a demonstration plant (DEMO) of some hundreds MWth. Accordingly, in the frame of an I-NERI between ENEA (Italy) and Argonne National Laboratory (USA) under the auspices of the EURATOM-DOE Agreement, the con-ceptual design of a pool-type LFR DEMO is being developed, assuming as reference industrial-size LFR the European Lead-cooled SYstem (ELSY) under development within the 6th and 7th EURATOM Framework Programmes. A demonstration reactor is expected to prove the viability of technology to be implemented in the First-of-a-Kind industrial power plant. Therefore, the first step towards DEMO conceptual design has been the definition of its specifications as a nuclear power facility able to validate ELSY main features and performances, and to qualify numerical codes and tools. Suitable design parameters have been set to meet the foremost objective of reaching a high fast neutron flux while respecting all technological constraints. A fuel cycle hypothesis has been adopted and burn-up calculations have been performed. Two different inde-pendent sets of absorbers have been introduced to guarantee the required reliability for safe reactor shut-down. Preliminary thermal-hydraulic analyses have been carried out to verify safety limits are not exceeded.

09:40 AMApplication of Method of Characteristics to Fast Reactor Core AnalysisToshikazu Takeda(1), Takanori Kitada(2), Hiroshi Nishi and Junichi Ishiba-shi(3)1) University of Fukui, Research Institute of Nuclear engineering, Fukui, Japan. 2) Osaka University, Division of Sustainable Energy and Environmental Engineering, Osaka, Japan. 3) FBR Plant Engineering Center, Japan Atomic Energy Agency, Fast Breeder Reactor Research and Development, Fukui-ken, Japan

The method of characteristics (MOC) has been applied to fast reactor core with hex-agonal geometry to accurately evaluate pin by pin powers. This enables one to obtain reaction rates in a particular test assembly which contains special fuels such as minor actinides or special foils. The accuracy of this calculation is confirmed by comparing Monte-Carlo results, and it is found that the present MOC calculation gives a satisfac-tory keff value and fluxes compared to the reference Monte-Carlo calculation. Further-more, it is seen that the Monte-Carlo calculation does not yield a smooth flux profile in the lower energy range even with 108 neutron histories. Thus the present MOC calculation is satisfactory to be used in detailed pin by pin flux calculations.


5B - Liquid Metal Fast Reactor Design & AnalysisSession Chairs: Denis Verrier (AREVA), Joe Miller (EDA, Inc)

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08:00 AMMacroscopic Cross-Sections of Neutron Radiation Cap-ture By Pb-208, U-238 and Tc-99 Nuclides in the Accelera-tor Driven Subcritical Core Cooled with Molten Pb-208Georgy L. Khorasanov and Anatoly I. BlokhinInstitute for Physics and Power Engineering named after A. I. Leypunsky, Obninsk, Kaluga region, Russian Federation

In the paper macroscopic cross sections <σ(n,g)> for several isotopes: 208Pb, 238U, 99Tc and natural mix of lead isotopes, natPb, averaged over neutron spectra of the accelerator driven subcritical core cooled with natPb or 208Pb are given. It is shown that macro cross sections for a coolant from 208Pb are by 6.2 times smaller than those for the coolant consisted from natPb. The economy of neutrons in the core cooled with molten 208Pb can be used for reducing initial fuel load, increasing plutonium breeding and enhancing transmutation of such long lived fission products as 99Tc. The values of macro cross sections calculated for 238U and 99Tc, equal to 0.6 and 0.8 barns, respectively, are comparable with the values of the same nuclide macro cross sections <σ(n,g)> for neutron spectrum of the fast reactor core cooled with sodium. Good neu-tron and physical features of molten 208Pb permit to assume it as perspective coolant for fast reactors and accelerator driven systems.

08:25 AMThe Research of Pb208 Coolant in Thorium-based Long Life ReactorGanglin Yu, Kan Wang Department of Engineering Physics, Tsinghua University, Beijing, China

This paper introduce the thorium fuel based long-life reactor concept and its prelimi-nary design. The Lead-Bismuth has been chosen as the coolant of the long-life reac-tor. Compare with the traditional heavy metal coolant such as Lead or Lead-Bismuth, the enriched Pb208 has some advantage in reactor physics and which will benefit the long-life core design. The research work on the pure Pb208 coolant and thorium fuel in fast spectrum has been done to verify the physical characteristic of Pb208 based on one fast reactor pin cell. The results show the application of Pb208 coolant will in-crease the system criticality and decrease the positive void reactivity coefficient in fast reactor design. We used the MCB and ORIGEN codes to perform calculation on the thorium-based long-life core model with Pb208 coolant, focus on the void coefficient in the center of the fast reactor. It can be concluded that the Pb208 will greatly improve the characteristics of the long-life core.

08:50 AMHigh Performance Fast Reactor Cores Employing Deu-teride ModeratorsTsugio Yokoyama(1), Toshio Wakabayashi(2)1) Toshiba Nuclear Engineering Services Corporation, Isogo-ku, Yokohama, Japan. 2) Department of Quantum Science & Energy Engineering, Graduate School of Engi-neering, Tohoku University, Sendai,Japan

The application of deuteride moderators for fast reactor cores is proposed to mitigate thermal spikes of fuel pin power in fuel assemblies and a decrease in breeding ratio that is often observed when hydrogen moderator is applied to fast spectrum cores. The effect of zirconium deuteride on a medium size fast reactor core applied in the form of pin arrays at the inner most rows of radial blanket assemblies is evaluated where it works as a reflector to flatten the radial power distribution of the core. Major core performances including breeding ratio and pin wise power distribution are ana-lyzed and compared with that of the core applied with zirconium hydride. The result indicates that the zirconium deuteride provides better power distributions relative to the hydide and the breeding ratio of the deuteride used core is about 10% larger than that of the hydride used core.

09:15 AMNeutronic Analysis of Water-Steam Injection Accidents for Generation IV Gas-Cooled Fast ReactorsGaëtan Girardin(1), Aaron Epiney(2), Konstantin Mikityuk(2), Rakesh Chawla(1)1) Ecole Polytechnique Fédérale de Lausanne (EPFL), Lausanne, Switzerland. 2) Paul Scherrer Institut (PSI), Villigen PSI, Switzerland

The present paper is addressing the static neutronic analysis, and a code-to-code vali-dation, of water-steam injection accidents for three different Gas-cooled Fast Reactor (GFR) core designs. It is assumed that this type of accident can occur as consequence of either the rupture of a pipe in the main steam generator or a leak in the decay heat removal heat exchanger. The analysis is focused on fast-spectrum, helium-cooled systems currently being developed and investigated in the context of the Genera-tion IV International Forum (GIF) and the 6th Framework Program of the European Union. More specifically, two 2400 MWth GFR cores and a small-size 50 MWth GFR demonstrator have been analyzed and systematically compared for a large range of water-steam densities from 0 (dry core) to 250 kg/m3 within the core. The neutronic analysis was performed using both the deterministic system code ERANOS-2.1 and the Monte Carlo method MCNPX-2.5 code, in association with modern nuclear data libraries, i.e. JEF-2.2 and ERALIB1 (adjusted library) for ERANOS, and JEF-2.2 for MCNPX. First, simulations were performed based on cell models and then with whole core representation, in order to ease the code-to-code comparison. Based on the core analysis, the keff-value for the two GFR core designs is seen to first increase with the water-steam density, and then, beyond 40-100 kg/m3, to decrease monotonically. On the contrary, for ETDR, the keff-value increases throughout the analyzed water-steam density range. Globally, a good agreement is obtained between the deterministic and stochastic results, the discrepancy being in the range of a few hundreds of pcm. Ad-ditional investigations have been conducted on effects such as the neutron spectrum softening, leakage reduction and the contribution of structural materials. It has been observed that non-conventional materials, such as tungsten, play a major role and help counteract the positive reactivity effect. Finally, the analysis has shown the suit-ability of ERANOS/ERALIB1 for the core analysis of the GFR cores, characterized by a softener neutron spectrum compared to Na-cooled systems.

09:40 AMCore Design Optimisation in Advanced Heavy Water Re-actor for Achieving Self-Sustenance in 233UNeelima Prasad, Arvind Kumar, P.D. Krishnani and R.K. SinhaBhabha Atomic Research Centre, Reactor Design and Development Group, Trombay, Mumbai, India

Advanced Heavy Water Reactor (AHWR) being designed for 920 MWth, is a vertical pressure tube type thorium-based reactor cooled by boiling light water and moderated by heavy water designed to maximise power production from thorium. This is a unique reactor designed for large scale commercial utilization of thorium and integrated tech-nological demonstration. India’s experience in design, operation and safety analyses/aspects of Pressurised Heavy water Reactors (PHWR) and Boiling Water Reactors (BWR) has been used to design this innovative reactor. The equilibrium fuel cycle is based on the conversion of naturally available thorium into fissile 233U driven by plu-tonium as external fissile feed. Plutonium is used as makeup fuel to achieve high dis-charge burnup and self-sustaining characteristics of Th-233U fuel cycle. The uranium in the spent fuel will be reprocessed and recycled back into the reactor. This paper gives the fuel cluster design and core design parameters for achieving self-sustenance in 233U in AHWR.


5C - Advanced Reactor DesignsSession Chairs: F. Damian (CEA), Guy Marleau (EPM)

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10:20 AMAssessment of Fission Gas Vented Fuel Impact with Advanced Burner ReactorT. K. Kim and C. GrandyArgonne National Laboratory, Argonne, IL USA

The fission-gas vented fuel was assessed with the reference ABR-1000 plant design. By adopting so-called diving-bell device for fission gas ventilation, the total assembly length could be reduced by more than 100 cm and the gas pressure in cladding, which is a primary barrier for high burnup caused by cladding breach, was also sig-nificantly decreased. As a result, major reductions in capital and operational costs are expected by adopting vented fuel in the ABR. The impact of the vented fuel on the core performance, radiation activity level in the reactor, and the required extra biological shield was assessed. The fission-gas vented fuel does not affect the core performance characteristics but does affect the radiation activity levels of the primary coolant and cover gas space. The radiation activity level of the cover gas space was increased significantly mainly due to the vented fission gases. However, additional biological shielding was not required to maintain the radiation dose level below the design constraint although it was conservatively assumed that the cover gas purge is not allowed during the reactor operation.

10:45 AMA Standard for the Determination of Thermal Energy De-position Rates in Nuclear ReactorsDimitrios CokinosBrookhaven National Laboratory, Upton, NY USA

The standard is intended to cover thermal energy deposition calculations for the entire nuclear industry; from fast to thermal reactors and from research to power reactors. The standard discussed here provides criteria for the (a) determination of the spatial energy deposition rates resulting from the interaction of neutrons with materials in the core (b) calculation of the spatial energy deposition rates resulting from the interac-tion of photons with matter; (c) determination of the energy allocation of among the principal particles and photons produced in fission (d) treatment of heavy charged particles and electrons slowing down in matter; and (e) presentations of the results of calculations, the verification of the accuracy of these results and the determination of uncertainty. The standard is of a general nature since many different types of calcula-tions are performed, each having its own requirements for accuracy and verification. Five major areas provide the essential requirements which must be met by the stan-dard: source distribution; selection of models and methods; Verification; evaluation of accuracy and documentation.

11:10 AMIncore Instrument Subcritical Verification (INCISV) – Core Design Verification MethodMichael C. Prible, Michael D. Heibel, Shannon L. Conner, Patrick J. Sebas-tiani and Daniel P. KistlerWestinghouse Electric Company, Monroeville, PA USA

According to the standard on reload startup physics testing, ANSI/ANS 19.6.1, a plant must verify that the constructed core behaves sufficiently close to the designed core to confirm that the various safety analyses bound the actual behavior of the plant. A large portion of this verification must occur before the reactor operates at power. The INCISV Core Design Verification Method uses the unique characteristics of a Westinghouse Electric Company fixed incore self powered detector design to perform core design verification after a core reload before power operation. A Vanadium self powered detector that spans the length of the active fuel region is capable of con-firming the required core characteristics prior to power ascension; reactivity balance, shutdown margin, temperature coefficient and power distribution. Using a detector element that spans the length of the active fuel region inside the core provides a signal of total integrated flux. Measuring the integrated flux distributions and changes at vari-ous rodded conditions and plant temperatures, and comparing them to predicted flux levels, validates all core necessary core design characteristics. INCISV eliminates the dependence on various corrections and assumptions between the excore detectors and the core for traditional physics testing programs. This program also eliminates the need for special rod maneuvers which are infrequently performed by plant operators during typical core design verification testing and allows for safer startup activities.

11:35 AMComparison of Experiments and Calculations of Void Fraction Distributions in Randomly Stacked Pebble BedsG. J. Auwerda, J. L. Kloosterman, A. J. M.Winkelman, J. Groen, V. J. J. van DijkDelft University of Technology, Department of Radiation, Radionuclides and Reactors Physics of Nuclear Reactors, Delft, Netherlands

In pebble bed reactors the fuel forms a randomly stacked pebble bed with non-uniform fuel densities, affecting neutronics (streaming) and thermodynamics (wall channeling). To investigate these effects, computational tools are needed capable of generating realistic pebble beds, and experimental results to validate these tools. Using gamma-ray scanning the absolute 0 and radial void fraction profile r(r) of a randomly stacked pebble bed was measured. Results were used to validate three different methods: Discrete Elements Method (DEM), Monte Carlo (MC) rejection method, and expanding system method. The bed consisted of 5457 acrylic pebbles with a diameter d = 12.7 mm in an acrylic cylinder with diameter D = 229 mm (D/d = 18.0), and had an average void fraction 0 = 0.395. The radial void fraction profile showed large, dampened oscil-lations near the wall extending up to five pebble diameters into the pebble bed, with a minimum void fraction of 0.22 half a pebble diameter away from the wall. The MC rejection method resulted in a 0 much higher than measured, and could not reproduce well the oscillations in r observed in the experiment. Both the DEM and expanding system method showed excellent agreement with the experiment for both 0 and r, with the expanding system method having the benefit of creating pebble beds with no overlapping pebbles, suitable for exact pebble bed models in other codes.


5D - Core MonitoringSession Chairs: Charles Rombough (CTR Technical Services), Walid Metwally (GNF)

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3:30 PMImplementation Strategy of Thorium Fuel Cycle Takashi Kamei Institute of Sustainability Science, Kyoto University, Gokasyo, Uji, Japan

Nuclear power is called again as a countermeasure of climate change recently. Nucle-ar power does not emit carbon dioxide (CO2) when it generates electricity. However there are still existing concerns such as the nuclear proliferation, long-term radioactive waste. Nuclear power was not included as a technical method of CDM (clean develop-ment mechanism) of Kyoto protocol. The use of the thorium is expected to overcome these concerns. Even though thorium utilization was known in the very early stage of nuclear application in 1940’s, thorium was not used as primary source due to its lack of fissile material. Plenty amount of plutonium stock in the spent nuclear fuel from more than 50 years operation of the uranium fuel cycle can be used as starter of thorium fuel cycle. Declaration of the “world without nuclear weapon” by the president Obama will also help to use weapon grade plutonium for starting thorium fuel cycle. In this paper, I will discuss how much amount of thorium cycle can be implemented triggered by the plutonium stock in spent nuclear fuel and by the weapon grade plutonium. Several implementation scenarios of thorium fuel cycle will be considered. Several types of molten-salt reactor were candidates of thorium nuclear power plant. The capacity of the thorium fuel cycle is estimated to be 450 GWe around at 2050. Some additional discussions on reducing carbon dioxide emission will be carried on rare-earth mining and electric vehicle in view of thorium utilization.

3:55 PMRoadmap Design for Thorium-Uranium Breeding Re-cycle in PWRShengyi Si Shanghai Nuclear Engineering Research & Design Institute, Shanghai, China

The paper was focused on designing a roadmap to finally approach sustainable Thorium-Uranium (232Th-233U) Breeding Recycle in current PWR, without any other change to the fuel lattice and the core internals, but substituting the UOX pellet with Thorium-based pellet. At first, the paper presented some insights to the inherence of Thorium-Uranium fuel conversion or breeding in PWR based on the neutronics theory and revealed the prerequisites for Thorium-Uranium fuel in PWR to achieve sustain-able Breeding Recycle; And then, various Thorium-based fuels were designed and examined, and the calculation results further validated the above theoretical deduc-tions; Based on the above theoretical analysis and calculation results, a roadmap for sustainable Thorium-Uranium breeding recycle in PWR was outlined finally.

4:20 PMThorium Fuel Utilization in the BWR: Lattice Physics Analysis of Reactivity CoefficientsAndrew M. Ward, Volkan Seker, Benjamin S. Collins, Thomas J. DownarNuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI, USA

Although there has been considerable research into the Thorium fuel cycle for Light Water Reactors, the reactivity coefficients have not been studied extensively. The work reported here is a systems level lattice analysis in order to provide an assessment of Thorium performance in a Boiling Water Reactor. Several Thorium fuel assemblies were designed in an oxide fuel matirx with one of three fissile materials: 20 wt% en-riched U-235 oxide, reactor grade Plutonium, and U-233, which was assumed to be reprocessed. Results indicate that it is possible to achieve comparable burnup be-havior and pin powers with Thorium fuel compared to Uranium oxide fuels. However, significant differences were observed in the reactivity coefficients for each fuel type. Each of the Thorium fuels had either too negative or slightly positive Coolant Void Reactivity (CVR) coefficients, and the control rod worth and the Doppler coefficients were more negative in all cases. A second stage of the analysis was performed which considered changes in the fuel geometry and some improvements were achievable in the reactivity coefficient performance.

4:45 PMFeasibility and Desirability of Employing the Thorium Fuel Cycle for Power GenerationBal Raj SehgalNuclear Power Safety, Royal Institute of Technology (KTH), AlbaNova University Cen-ter, Stockholm, Sweden

Thorium fuel cycle for nuclear power generation has been considered since the very start of the nuclear power era. In spite of a very large amount of research, experimen-tation, pilot scale and prototypic scale installations, the thorium fuel was not adopted for large scale power generation [1,2]. This paper reviews the developments over the years on the front and the back-end of the thorium fuel cycle and describes the pros and cons of employing the thorium fuel cycle for large generation of nuclear power. It examines the feasibility and desirability of employing the thorium fuel cycle in concert with the uranium fuel cycle for power generation.

5:10 PMPower Flattening for Sodium Cooled Metallic Fuel CAN-DLE Reactor by Adding Thorium in Inner CoreHiroshi Sekimoto, Sinsuke Nakayama, Hiroshi Taguchi And Tsuyoshi Oh-kawaResearch Laboratory for Nuclear Reactors, Tokyo Institute of Technology, Ookayama, Tokyo, Japan

CANDLE reactor shows excellent performances on the difficult problems inherent to nuclear energy concerning sustainability, safety, bomb and wastes. However, concern-ing economy we have still problems to be solved. Core height should be as short as possible from economy consideration. In this paper the power density profile is intend-ed flatten for a sodium-cooled metallic-fuel CANDLE fast reactor by adding thorium uniformly in the inner core region. The total power is increased by the factor of 1.28 under the constraint of maximum axially integrated power density. The core height can be also decreased.


6A - Thorium Fuel CycleSession Chairs: Moussa Mahgerefteh (Exelon Nuclear), Kevin Hesketh (UK National Laboratory)

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1:30 PMEnhancing Advanced CANDU Proliferation Resistance Fuel with Minor Actinides Gray S. Chang Idaho National Laboratory, Idaho Falls, Idaho USA

The advanced nuclear system will significantly advance the science and technology of nuclear energy systems and to enhance the spent fuel proliferation resistance. Minor actinides (MA) are viewed more as a resource to be recycled, and transmuted to less hazardous and possibly more useful forms, rather than simply disposed of as a waste stream in an expensive repository facility. MAs can play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. In this work, an Advanced CANDU Reactor (ACR) fuel unit lat-tice cell model with 43 UO2 fuel rods will be used to investigate the effectiveness of a Minor Actinide Reduction Approach (MARA) for enhancing proliferation resistance and improving the fuel cycle performance. The main MARA objective is to increase the 238Pu / Pu isotope ratio by using the transuranic nuclides (237Np and 241Am) in the high burnup fuel and thereby increase the proliferation resistance even for a very low fuel burnup. As a result, MARA is a very effective approach to enhance the proliferation resistance for the on power refueling ACR system nuclear fuel. The MA transmutation characteristics at different MA loadings were compared and their impact on neutronics criticality assessed. The concept of MARA, significantly increases the 238Pu/Pu ratio for proliferation resistance, as well as serves as a burnable absorber to hold-down the initial excess reactivity. It is believed that MARA can play an important role in atoms for peace and the intermediate term of nuclear energy reconnaissance.

1:55 PMThe Potential of Pressurized Water Reactors for Recycle of Americium-Curium G M Thomas, K W Hesketh(1), Stuart T Arm(2) 1) UK National Nuclear Laboratory (NNL), Preston Laboratory, Springfields, Preston, UK. 2) EnergySolutions LLC, Richland, WA USA

The UK National Nuclear Laboratory, in collaboration with EnergySolutions and other partners, has carried out studies related to americium-curium (Am-Cm) target recycle in a Pressurized Water Reactor PWR). This work, which was carried out in the context of Global Nuclear Energy Partnership (GNEP), was intended to establish whether PWRs could be used to irradiate worthwhile quantities of Am-Cm target rods and what transmutation rates could be achieved. 3-D core analysis methods were used to assess the impact on the nuclear design of an equilibrium core containing hetero-geneous Am-Cm target rods. The study shows that up to 100 kgHM of Am-Cm can be accommodated in every 18 month refueling cycle without significantly impacting the nuclear design limits. The study demonstrated that a high transmutation rate is achievable for americium, though not for curium. A complementary study examined the broader impact of Am- Cm target irradiation on the PWR fuel cycle and considered how it would fit with some high level strategic goals. This paper reports on the two studies and also discusses how the irradiation of Am- Cm targets might fit within the overall GNEP strategy.

2:20 PMRecycling of VVER Minor Actinides in a Gas-Cooled Fast ReactorZoltan Perko, Sandor Feher(1), Jan Leen Kloosterman, Stuart Christie(2)1) Institute of Nuclear Techniques, Budapest University of Technology, Budapest, Hungary. 2) Department of Radiation, Radionuclides and Reactors, Delft University of Technology, Delft, Netherlands

The Gas-Cooled Fast Reactor (GCFR) is one of the six Generation IV reactor designs, which has the potential to efficiently consume Minor Actinides (MAs). In this study, the possibility of using the reference GFR600 design as a MA burner was examined, as-suming an initial MA composition corresponding to that of VVER440 spent nuclear fuel (SNF). For the calculations the KENO-VI Monte Carlo code and the ORIGEN-S burnup code of the SCALE 5.1 system were used, applying a precise three-dimensional reac-tor model. Burnup studies were performed for cores with different initial MA contents. Besides the reactivity swing, the fuel inventory and the delayed neutron fraction were calculated as a function of burnup. Multiple recycling of MAs was also examined, as-suming two recycling strategies, furthermore a comparison was made between the results of using MAs and Pu from LWR and VVER spent fuel. It was concluded that the isotopic composition of the Pu has a strong effect on the reactivity. The loss of reactiv-ity is significantly larger when using Pu from VVER SNF, and the reactivity swing does not turn positive when MAs from VVER SNF are added to the fuel. The MA burning capability of the reactor is to a large extent determined by the composition of the MAs and the Pu. Mainly Am is destroyed when using MAs and Pu from LWR, and Np when using those from VVER spent fuel. However, the delayed neutron fraction shows only a slight dependence on the origin of Pu and MAs.

2:45 PMInvestigation of Direct Transmutation of Actinides by Spallation NeutronsIstván Rovni, Máté Szieberth and Sándor FehérBudapest University of Technology and Economics, Institute of Nuclear Techniques, Budapest, Hungary

Accelerator driven spallation sources are nowadays designed and built as high flux and high energy neutrons sources for different purposes. The aim of this study is to investigate the applicability of such a realistic spallation source for the direct transmu-tation of minor actinides (MAs). A deeply subcritical and relatively low power blanket containing MAs, which is easier to build and operate, is designed around the target. The fuel assembly of the gas cooled nuclear reactor concept has been used for the structure and material composition of the blanket. The code MCNPX has been used for the transport calculations and the determination of the reaction rates. The code had to be modified in order to estimate the isotope transformation rates in the energy range of the physical models. A special code has been developed to follow the burn-up in the blanket with the matrix exponential method. The MA consumption has been investigated with different initial MA concentration. The results show that in the spec-trum of the spallation neutrons the MAs can be transmuted more efficiently; however the flux of the fission neutrons can be higher in such extent which results in higher MA consumption.


6B - Minor Actinide ManagementSession Chairs: Ayodeji Alajo (Texas A&M), Bojan Petrovic (Georgia Tech)

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10:20 AMPreliminary Assessment of a Fusion-Fission Hybrid ReactorTerry KammashUniversity of Michigan NERS, Ann Arbor, MI USA

A great deal of interest has recently been shown in the fusion-fission hybrid reactor as a major potential answer to the world’s daunting and critical need for carbon-free energy. While much has been made of the technological problems associated with the fission and fusion components of the system, the fact remains that the advantages far outweigh the disadvantages particularly when seen from the viewpoint of dwin-dling mined uranium resources and concern about weapon grade proliferation issues, among others. In this paper, we attempt to assess the power-producing capability of a hybrid reactor whose fusion component serves only as a source of fast neutrons, thereby allowed to operate at or near “breakeven” condition. Our choice for this com-ponent is the gasdynamic mirror (GDM) magnetic confinement device which is linear cylindrically symmetric and whose plasma confinement properties are quite well un-derstood. It will be surrounded by a thorium-232 laden blanket where uranium-233 is bred and simultaneously burned to generate power. Assuming steady state operation, we calculate the power produced per unit length using a particle balance equation for the U233 density and a diffusion equation for the neutron flux. We find that a system of large aspect ratio with no cooling ducts in the blanket is capable of producing tens of gigawatts of thermal power per centimeter. Moreover, we show that the system is “subcritical” and can reach steady state in about 4 months if no “spiking” is initially employed.

10:45 AMDefinition of Breeding Gain for Molten Salt ReactorsK. Nagy, J. L. Kloosterman, D. Lathouwers and T. H. J. J. van der HagenDelft University of Technology, Department of Radiation, Radionuclides and Reactors, Delft, The Netherlands

The graphite-moderated Molten Salt Reactor (MSR) is a potential breeder reactor using the thorium fuel cycle. The MSR has unique properties due to the possibility of making changes to the salt composition during operation. Most important is the extrac-tion of protactinium, which separates the fissile uranium production into two volumes: the reactor core and the external stockpile. The paper focuses on the definition of breeding gain in such a system. The prospects of using breeding gain expressions defined for solid fuel reactors are investigated and new definitions are given which incorporate the processes occurring in the reactor core and the external stockpile. The difference of the growth rate of the mass of fissile material and breeding gain is pointed out. The new definitions are applied to an optimization study of the graphite-salt lattice of a breeder MSR.

11:10 AMFast Reactor Transmutation Performance and Conver-sion Ratio EffectsEdward A. Hoffman and Temitope A. TaiwoArgonne National Laboratory, Argonne, IL USA

The sometimes highly complex physics analyses performed to evaluate transmuta-tion systems often make interpretation of the results challenging. There are a num-ber of underlying relationships that have been developed to understand the general performance of transmutation systems. Those relationships and their implications are discussed. The assumed constraints imposed upon the design often dictate the per-formance. Key constraints and their impacts are analyzed. The fuel residence time dictates much of the key performance parameters for a given conversion ratio, such as burnup. The TRU consumption rate is independent of the burnup and is deter-mined solely by the conversion ratio as is the uranium resource utilization. However, minimizing transuranics in the waste requires minimizing the transuranics sent to re-processing, which is fairly independent of conversion ratio, but is a strong function of fuel residence time. The fraction of energy generated by fast reactors in a nuclear park can be evaluated in several ways. Integrated over time, it is determined by the conversion ratio. However at a given time for a dynamic system, it is a function of many additional variables. A simple approximation is provided to evaluate the dynamic equilibrium state of single-tier closed fuel cycle. The large amount of transuranic ma-terial required to deploy new fast reactors results in a greatly reduced fraction of the energy generated by fast reactors in a growing nuclear enterprise compared to the ultimate fraction of energy that will be generated by the fast reactors if the fuel cycle is run to completion.

11:35 AMTravelling Wave Reactor: Velocity Formation Mecha-nismsV. M. Khotyayintsev(1), V. M. Pavlovych and O. M. Khotyayintseva(2)1) Department of Physics, T. Shevchenko National University of Kyiv, Kyiv, Ukraine. 2) Institute for Nuclear Research of NAS of Ukraine, Kyiv, Ukraine

Travelling wave reactor (TWR) is a fast reactor in which nuclear burning propagates as a wave. This paper is aimed to understand a stationary nuclear burning wave (NBW) as a macroscopic physical process. One-group neutron diffusion equation and kinetic equations for nuclei densities are applied to a plane wave in an infinite reactor. We show that dimensionless wave velocity is small at realistic power densities enabling a perturbation approach. Numerical solution confirms and complements the analyti-cal results. In contrast to a single criticality condition for conventional fast reactors (CFRs), there are two balance conditions of the integral type in TWR. They determine velocity and final fluence of a stationary wave. Criticality concept alone fails to explain the changes of the wave parameters. Mechanisms determining the wave velocity are related to the second balance condition and have no analogs in the physics of CFRs at all. Normally they are related to the kinetics of the intermediate nuclide 239Np/233Pa for uranium/thorium cycle, correspondingly. Main velocity formation mechanism origi-nates in the process which is negligible in CFRs, namely, in the intermediate nuclide burn out, at least for uranium cycle. At very low velocities the effect of instability of 241Pu becomes crucial. Admittedly, it causes the existence of minimal NBW velocity estimated as 5 cm/year. For thorium cycle there is no such effect. Macroscopic physics of operation for TWRs substantially differs from the one for CFRs. Thus, theoretical basis for this new class of fast reactors should be reformulated and thoroughly taken into account.


6C - Advanced and Fast Reactor DesignsSession Chairs: Kevan Weaver (TerraPower), Robb Borland (First Energy)

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3:30 PMClosing the Nuclear Fuel Cycle in the 21st Century while Minimizing Proliferation Risk Stuart Arm and Christopher PhillipsEnergySolutions LLC, Richland, Washington, USA

Since September 2007, EnergySolutions, with its team drawn from North America and the UK, has performed conceptual design studies for the United States Department of Energy on the subject of closing the nuclear fuel cycle. Three major reports have been produced: the Business Plan; the Conceptual Design study; and the Technology Development Roadmap. Together, these reports examine the commercial, engineer-ing and technical aspects of closing the nuclear fuel cycle from the year 2025 through 2100. Management of plutonium is important to minimizing proliferation risk in pro-cessing Used Nuclear Fuel (UNF) because of the fissile plutonium-239 content of the ~1w% plutonium that is generated in a commercial light water reactor (LWR). Closing the nuclear fuel cycle requires the separation of the plutonium from the unused urani-um so that each can be recycled into new fuel. The proliferation risk from this process can be minimized by: (i) ensuring plutonium is always mixed with some uranium to minimize its attractiveness for weapons use and the EnergySolutions “NUEX” separa-tion process achieves this, (ii) recycling the plutonium as Mixed Oxide or Inert Matrix fuel in nuclear reactors to fission the plutonium-239 and so minimize the quantity of it available for potential weapons use, (iii) implementing Near Real Time Materials Accountancy to provide on-line tracking of plutonium so as to detect immediately any clandestine removal of plutonium, (iv) security measures to minimize the risk of unau-thorized entry to the recycling facilities and removal of fissile material.

3:55 PMIntroduction of MOTTO Cycle to CANDLE Fast ReactorHiroshi Sekimoto and Akito Nagata 1 Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, Meguro-ku, Japan

CANDLE reactor shows excellent performances on the inherent difficult problems con-cerning sustainability, safety, bomb and wastes. However, concerning feasibility and economy we have still problems to be solved. Cladding integrity should be considered from feasibility and core height should be as short as possible from economy consid-eration. In this paper the cladding integrity problem is solved by recladding, and the core height is tried to be short by making the axial power centroid position to the same height at any radial position by introducing MOTTO cycle. As a result, the core height becomes 1.6 m, where the reactivity swing during operation is 0.0007.

4:20 PMSolution of the Equilibrium Fuel Vector in Closed Fuel Cycles and Application to a Lead Fast ReactorC. Petrovich, C. Artioli(1), G. Grasso(2)1) ENEA, Bologna, Italy. 2) Nuclear Engineering Laboratory (LIN) of Montecuccolino, University of Bologna, Italy

An ideal strategy for the sustainability of nuclear energy is to implement closed fuel cycles in fast reactors, i.e. to recover all the actinides in the spent fuel and recycle them in the reactor itself. In this way input uranium feed and output long-term wastes are minimized: the fuel cycle can have as input only depleted (or natural) uranium and as final waste only the fission products and the losses due to fuel reprocessing. From the neutronic point of view, the feasibility of this approach depends on the equilib-rium fuel vector, whose solution therefore becomes crucial. The equilibrium vector is usually calculated by an approximate steady-state approach or by a direct irradiation simulation. Here a different approach has been devised, imposing a priori the equilib-rium requirement and taking into account the evolution of the fuel during irradiation, as well as its cooling time before being reprocessed and reloaded. The equilibrium for the transuranic elements becomes thus “cyclic”. The method is based on a recursive integration of the Bateman equation, and is validated by the codes FISPACT and MCNPX. The differences with respect to the steady-state approach are shown. The method is then applied to the European lead fast reactor ELSY. The fraction (on the total actinides) at equilibrium turns out to be in this case 0.9% for the MA and 17.2% for the Pu. The reduction of transuranic mass waste is calculated with respect to once-through fuel cycles (resulting lower by two orders of magnitude).

4:45 PMNuclear Fuel Cycle Sensitivity to the Variation of Physi-cal ParametersAnthony Scopatz(1), Jun Li(2), Man-Sung Yim(3), and Erich Schneider(1)1) Department of Mechanical Engineering, Nuclear Engineering Program, The Univer-sity of Texas at Austin, Austin, TX USA. 2) Institute for the Environment, The University of North Carolina – Chapel Hill, Chapel Hill, NC USA. 3) Department of Nuclear Engi-neering, North Carolina State University, Raleigh, NC USA

This study presents a development of a modeling platform and a supporting sensitivity study for identifying near-optimal scenarios of fuel cycle development. Modifications to our existing integrated fuel cycle systems analysis model are described to sup-port this study and results of sensitivity analysis are presented. Results indicated that separation efficiencies of key tansuranic (TRU) nuclides from reactors have the larg-est impact on the fuel cycle performance. Limitations of the present work and future directions are discussed.

5:10 PMConsideration of Reactor Systems for Uranium Re-source Extension and Waste MinimizationTemitope A. Taiwo, Taek K. Kim, Robert N. Hill(1) P. J. Finck(2)1) Argonne National Laboratory, Argonne, IL USA. 2) Idaho National Laboratory, Idaho Falls, ID USA

The resource utilization and waste management potentials of current and proposed advanced nuclear reactor systems have been evaluated and reported in this paper. Systems considered include light water reactors (once-through, limited-, and multi-recycle), gas-cooled very high temperature reactors, and fast reactors (single- and continuous-recycle). The resource utilization of the LWR system was found to be lim-ited to below 1%, while those in proposed once-through fast reactors system could be higher by a factor of two to three. The use of high burnup fuel in LWR was found not to help the resource utilization due to the high fuel enrichment required. Uranium utilization as high as 99% could be obtained with continuous recycle in advanced fast reactor systems. Collectively, high burnup can be used to reduce the quantity of waste arising in the repository, but the radiotoxicity of the material is not significantly reduced. By continuously recycling used nuclear fuel, the normalized radiotoxicity could be re-duced from that of geologic time scale (thousands to a million years) to an engineered times scale (hundreds of years).


6D - Alternative Fuel CyclesSession Chairs: Ayodeji Alajo (Texas A&M), Porsch Dieter (AREVA NP GmbH)

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1:30 PMDesign Studies for Full-MOX BWR Cores G. Bender, M. Rost and W.U. Timm AREVA, AREVA NP, Erlangen, Germany

Design studies for full-MOX cores in boiling water reactors (BWRs) were performed for a variety of bundle designs, e.g. ATRIUM™ 10, ATRIUM™ 11, high/low fissile Pu content, Pu of high/low quality. For all cases, the feasibility of full-MOX cores could be demonstrated. The number of Gd rods per assembly is not higher than in UO2 assemblies with comparable discharge burnup, especially assemblies with low Pu quality need only few Gd rods per assembly. The reactivity coefficients are acceptable: even for high Pu content and low Pu quality the void coefficient is clearly negative. Re-quirements for the liquid boron shutdown system are significantly higher than for UO2 cores. The “reactivity equivalence” to UO2 assemblies could be quantified for various Pu vectors. Mass balances for UO2 and MOX cores are compared.

1:55 PMTemperature Reactivity Coefficient for Plutonium Fuel in a High Temperature Reactor Christoph Pohl Institute of Energy Research, IEF-6: Safety Research and Reactor Technology, Forsc-hungszentrum Jülich GmbH, Jülich, Germany

In this paper the safety behavior with respect to the temperature reactivity coefficient of a pure plutonium oxide fuel for a high temperature helium cooled graphite moder-ated pebble bed reactor was investigated. In contrast to epithermal resonances of resonance absorber isotopes as Th-232 or U-238 the plutonium isotopes have reso-nances in the higher thermal energy region which are directly affected by a harder neutron flux spectrum. The total temperature coefficient for a fuel containing 2g heavy metal of first generation light water plutonium per pebble is positive for a cold zero power core condition and is mainly driven by the moderator coefficient. The modera-tor coefficient shows a maximum at a lower core temperatures which is influenced by the plutonium isotope composition. The results show a correlation of the height of the maximum with the fraction of Pu-241 while the temperature position of the maximum is influenced by the fractions of the isotopes Pu-240 and Pu-242. The thermal neutron leakage dominated by neutrons moderated and scattered back from the reflector to the core show also a maximum but at a higher isothermal core temperature than the moderator coefficient. Investigations of the total η value of a specific high power region of the core indicate a maximum at a higher thermal energy range than the maximum of the moderator coefficient. The found correlations offer possibilities to reduce the mod-erator coefficient without a high amount of additional resonance absorber materials.

2:20 PMThe Behaviour of a Preliminary Transuranic Mixed Ox-ide Fuel Design in A CANDU 6 ReactorA. C. Morreale, W. J. Garland and D. R. NovogDepartment of Engineering Physics, McMaster University, Ontario, Canada

The reprocessing of spent fuel such as the extraction of actinide materials for use in mixed oxide fuels is a key component of reducing the end waste from nuclear power plant operations. These reprocessed actinides can be burned for a limited amount of cycles in current thermal reactors before being sent to fast reactors resulting in a reduction of as much as 95% of spent fuel waste. The use of current thermal reactors as an intermediary step reduces the fast reactor infrastructure needed by a factor of almost 13.5 [1]. This paper examines the features of actinide mixed oxide fuel, TRU-MOX, in the CANDU 6® nuclear reactor. The actinide concentrations used were based on extraction from 30 year cooled spent fuel and mixed with natural uranium in 4.75% actinide MOX fuel. The WIMS-AECL model of the fuel was transferred to RFSP and a time average full core model was produced. The model was created with the standard CANDU 6® limits on bundle and channel powers and a burnup target of 45 MWd/kgHE. The TRUMOX fuel design achieved its goals and performed well under normal operations simulations. This effort demonstrated the feasibility of using the current fleet of CANDU 6® reactors as an intermediary step in burning reprocessed spent fuel. The recycling, reprocessing and reuse of spent fuels produces a much more sustain-able and efficient nuclear fuel cycle.

2:45 PMAttainable Burnup in a Life Engine Loaded with Deplet-ed UraniumMassimiliano Fratoni, Kevin J. Kramer, Jeffery F. Latkowski and Ryan P. Abbott(1), Jeffrey E. Seifried and Jeffrey J. Powers(2)1) Lawrence Livermore National Laboratory, Livermore CA, USA. 2) University of Cali-fornia, Berkeley Nuclear Engineering Department, Berkeley CA, USA

The Laser Inertial Fusion-based Energy (LIFE) system uses a laser-driven fusion source for electricity production. The (D,T) reaction, besides a pure fusion system, al-lows for the option of driving a sub-critical fission blanket in order to increase the total energy gain. In a typical fusion-fission LIFE engine the fission blanket is a spherical shell surrounding the fusion source, preceded by a beryllium pebble layer for neutron multiplication by means of (n,2n) reactions. The fuel is in the form of TRISO particles dispersed in carbon pebbles, cooled by the molten salt flibe (2LiF-BeF2). The optimal design features 80 cm thick blanket, 16 cm multiplier, and 20% TRISO packing factor. A blanket loaded with depleted uranium and depleted in a single batch with continuous mixing can achieve burnup as high as ~85% FIMA while generating 2,000 MW of total thermal power and producing enough tritium to be used for fusion. A multi-segment blanket with a central promotion shuffling scheme enhances burnup to ~90% FIMA, whereas a blanket operated with continuous refueling achieves only 82% FIMA under the same constraints of thermal power and tritium self-sufficiency. Both multi-segment and continuous refueling options eliminate the need for a fissile breeding phase.


6E - Transuranic ApplicationsSession Chairs: Robert St.Clair (Duke Power), Moussa Mahgerefteh (Exelon Nuclear)

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08:00 AMFCC-NG: the AREVA Safe Fresh Fuel Shipping Cask to Deliver the Future EPR™ ReactorsM. Doucet, P. Faye, Ch. Faignet, R. Babut, M. Landrieu(1), F. Marvaud, G. Romanet(2)1) AREVA, AREVA NP, Lyon France. 2) AREVA, AREVA NC, TN International, Saint Quentin en Yvelines Cedex

AREVA as a worldwide PWR fuel provider has to have a fleet of fresh UO2 shipping casks being agreed within a lot of countries and to accommodate foreseen EPR™ Nuclear Power Plants fuel buildings. To reach this target the AREVA NP Fuel Sector together with TN International (a subsidiary of AREVA NC) decided to develop an up to date shipping cask gathering experience feedback of the today fleet and an improved safety allowing the design to comply with international regulations (NRC and IAEA) and other local foreign Safety Authorities requirements. Based on pre design features a safety case was set up to highlight safety margins. Criticality hypothetical acciden-tal assumptions were defined: • Preferential flooding, • Fuel rod array expansion (so called “bird caging” effect), • Fuel sliding, • Neutron absorber penalty. The French criticality code package CRISTAL is used to check several configuration Keff and de-rived safety margins. The CRISTAL code package relies on a wide validation process helping defining uncertainties as required by regulations in force. The shipping cask, containing two fuel assemblies, is designed to maximize fuel isolation inside the cask and with neighboring ones even for large array configuration cases. Few and proven industrial products are used: • Stainless steel for the structural frame, • Balsa wood for impact limiters, • BORA® resin as neutron absorber. The cask is designed to handle mainly the EPRTM fuel assembly type and may be extended to other contents such as APWR fuel assembly type. After a brief presentation of the computer codes and the description of the shipping cask, the CRISTAL calculation results as well as the allow-ances for biases and uncertainties will be discussed.

08:25 AMOn the Criticality Safety of Transuranic Sodium Fast Re-actor Fuel Transport CasksSamuel Bays(1), Ayodeji Alajo(2) 1) Idaho National Laboratory, Idaho Falls, ID USA. 2) Department of Nuclear Engi-neering, Texas A&M, College Station, TX USA

This work addresses the neutronic performance and criticality safety issues of trans-port casks for fuel pertaining to low conversion ratio sodium cooled fast reactors, con-ventionally known as Advanced Burner Reactors. The criticality of a one, three, seven and 19-assembly cask capacity is presented. Both dry “helium” and flooded “water” filled casks are considered. No credit for fuel burnup or fission products was assumed. As many as possible of the conservatisms used in licensing light water reactor univer-sal transport casks were incorporated into this SFR cask criticality design and analy-sis. It was found that at 7-assemblies or more, adding moderator to the SFR cask increases criticality margin. Also, removal of MAs from the fuel increases criticality margin of dry casks and takes a slight amount of margin away for wet casks. Assum-ing credit for borated fuel tube liners, this design analysis suggests that as many as 19 assemblies can be loaded in a cask if limited purely by criticality safety. If no credit for boron is assumed, the cask could possibly hold seven assemblies if low conver-sion ratio fast reactor grade fuel and not breeder reactor grade fuel is assumed. The analysis showed that there is a need for new cask designs for fast reactors spent fuel transportation. There is a potential of modifying existing transportation cask design as the starting point for fast reactor spent fuel transportation.

08:50 AMISOCRIT: A Burnup Credit Tool for Spent Fuel Pool Stor-age CalculationsVefa N. Kucukboyaci and William J. MarshallWestinghouse Electric Co., Pittsburgh, Pa, USA

In order to conservatively apply burnup credit in spent fuel pool criticality safety analy-ses, Westinghouse has developed a software tool, ISOCRIT, for generating depletion isotopics. This tool is used to create isotopics data based on specific reactor input pa-rameters, such as design basis assembly type; bounding power/burnup profiles; reac-tor specific moderator temperature profiles; pellet percent theoretical density; burnable absorbers, axial blanket regions, and bounding ppm boron concentration. ISOCRIT generates burnup dependent isotopics using PARAGON; Westinghouse’s state-of-the-art and licensed lattice physics code. Generation of isotopics and passing the data to the subsequent 3D KENO calculations are performed in an automated fashion, thus reducing the chance for human error. Furthermore, ISOCRIT provides the means for responding to any customer request regarding re-analysis due to changed parameters (e.g., power uprate, exit temperature changes, etc) with a quick turnaround.

09:15 AMTGBLA Spent Fuel Isotopic Predictions and Their Effect on Criticality CalculationsWalid A. Metwally, Masatoshi Sugawara, and Vernon W. Mills(1), John C. Hannah(2)1) Global Nuclear Fuel (GNF), Wilmington NC USA. 2) GE Hitachi NuclearEnergy (GEH), Wilmington NC USA

TGBLA is the GNF lattice design computer program for conventional BWRs. Thirty-six spent fuel pin samples, obtained from the Quad Cities BWR, were used to assess the accuracy of uranium and plutonium spent fuel isotopic composition computed by the GNF lattice physics system TGBLA. The samples were obtained from different ra-dial and axial locations. The measured and TGBLA calculated isotopic concentrations were in good agreement. The isotopic uncertainty for the U235 ~ 238 and Pu239 ~ 242 isotopes was quantified and conservative correction factors were calculated for use in criticality calculations. A study of two spent fuel storage racks is used to investigate the effect of the isotopic concentration uncertainty on criticality calculations. Eigenvalue differences up to 0.014 in the racks were observed as a result of the spent fuel isotopic uncertainty.

09:40 AMMCNP5 Criticality Experiment Benchmark for the Dry Storage System of Chinshan Nuclear Power PlantYung-Hung Teng, Jau-Tyne Yeh and Chung-Hsing Hu(1), Cheng-Hsi Wu(2)1) Institute of Nuclear Energy Research, Lungtan, Taoyuan, Taiwan, R.O.C. 2) Taiwan Power Company, Taipei, Taiwan, R.O.C.

For the necessity of the criticality analysis of dry storage system for Chinshan Nuclear Power Plant, MCNP5 is benchmarked against 138 critical experiments. The code bias and uncertainty of MCNP5 are determined by the 95/95 one-side tolerance limit meth-od. The validity of applying this statistical methodology is checked by two statistical tests and trend analysis. The k-eff of 0.93973 will be taken as the Upper Subcritical Limit for INER’s dry storage system.


7A - Nuclear Criticality Safety - ISession Chairs: Dennis Mennerdahl (E Mennerdahl Systems), Tatiana Ivanova (IRSN)

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08:00 AMOECD/NEA Expert Group on Uncertainty Analysis for Criticality Safety Assessment: Current ActivitiesTatiana Ivanova, et al.Institut de Radioprotection et de Sûreté Nucléaire, Fontenay aux Roses, France

The expert group (EG) on Uncertainty Analysis for Criticality Safety Assessment (UACSA) was established within the OECD/NEA Working Party on Nuclear Critical-ity Safety in December 2007 to promote exchange of information on related topics; compare methods and software tools for uncertainty analysis; test their performance; and assist in selection/development of safe and efficient methodologies. At the cur-rent stage, the work of the group is focused on approaches for validation of criticality calculations. With the diversity of the approaches to validate criticality calculations, a thorough description of each approach and assessment of its performance is useful to the criticality safety community. Developers, existing and potential practitioners as well as reviewers of assessments using those approaches should benefit from this effort. Exercise Phase I was conducted in order to illustrate predictive capabilities of criticality validation approaches, which include similarity assessment, definition of keff bias and bias uncertainty, and selection of benchmarks. The approaches and results of the exercises will be thoroughly documented in a pending state-of-the-art report from the EG. This paper provides an overview of current and future activities for the EG, a summary of the participant-contributed validation approaches, and a synthesis of the results for the exercises.

08:25 AMKeff Sensitivity/Uncertainty Analysis of the Phase VII Burnup Credit BenchmarkB. Cabellos, O. Cabellos, N. García-Herranz Department of Nuclear Engineering, Universidad Politécnica de Madrid

The aim of this paper is to present the keff sensitivity/uncertainty analysis for the Phase-VII Burnup Credit Benchmark using the sensitivity profiles obtained with the perturbation option of MCNP and the variance/covariance cross section and decay data from different uncertainty data libraries. Firstly, results obtained for Phase-VII Benchmark using the nuclear data library JEFF- 3.1.1, and the computer programs ACAB (for the inventory prediction) and MCNP (for criticality calculation) are present-ed. Additional calculations have been performed to assess the importance of nuclear data and computational tools. Secondly, a sensitivity/uncertainty analysis permits to conclude that the uncertainties in decay data libraries are negligible in keff calculation; and crosssection data uncertainties and their impact in keff for long-term disposal have been assessed.

08:50 AMDetermination of a Depletion Uncertainty from Fuel Management ExperienceDale B. Lancaster and Charles T. RomboughNuclearConsultants.com, CTR Technical Services, Inc.

In the United States, spent fuel pool burnup credit has allowed a depletion uncertainty of 5% of the delta k of depletion. This uncertainty was based on engineering judgment and requires more documentation. It is shown that data from commercial power reac-tors can be used to validate burnup credit criticality calculations via simplified bench-marks produced by fuel management codes. The accuracy of these simplified bench-marks is established by comparison between the reactor measurements and the fuel management codes. In turn these simplified benchmarks are used to determine a bias between the criticality tools and the fuel management tools. This bias, when added to the bias and uncertainty of the fuel management tools, becomes a basis for a deple-tion uncertainty. Preliminary analysis shows that the depletion uncertainty for a typical criticality code set is probably less than 2% for the burnups of interest.

09:15 AMThird Order Time Correlation Method Applied to SILENE Absolute Criticality MeasurementsPhilippe HumbertCommissariat à l’Energie Atomique, Centre DAM-Ile-de-France, Bruyères-le-Châtel, ARPAJON, France

A previous time-list mode experiment performed with SILENE reactor is analyzed us-ing the thirdorder correlation technique. This method originally proposed by A. Furu-hashi is an extension to the well known second order Feynman-a technique. The inter-est of the method is that it allows for the absolute determination of the multiplication coefficient using second and third order moments of the counting number distribution as function of the time gate width. In the framework of the point reactor model the reactivity is related to the measured parameters using a simple formula. In the ana-lyzed SILENE experiment the reactivity given by the Furuhashi method is compatible with MCNP criticality calculation although the accuracy of the method is bad due to the statistical noise in the third order correlation parameter when the counting time is limited.

09:40 AMHierarchical Monte-Carlo Approach to Bias Estimation for Criticality Safety CalculationsOliver Buss, Axel Hoefer, Jens Christian Neuber, Michael SchmidAREVA NP GmbH, Mechanical Analyses NEEA-G, Offenbach am Main, Germany

We present a hierarchical Monte Carlo method capable of predicting the computation-al bias for the neutron multiplication factor evaluation within a criticality safety analysis. Bias predictions are based on evaluations of representative sets of criticality experi-ments taking into account their uncertainties and also correlations between uncertain-ties of different experiments. The presented procedure relates the keff evaluations for the experiments and their covariance matrix to a computational bias prediction for the application case using trending techniques taking the covariance matrix fully into account. Additionally, we present a method to determine proper sets of explanatory variables needed for the trending procedure.


7B - Nuclear Criticality Safety - IISession Chairs: Sedat Goluoglu (ORNL), Charles T. Rombough (CTR Technical Services)

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3:30 PMABWR Start-Up Test Analysis with Transient Code POLCA-T Kazuki Yano, Yamato Hayashi(1), Kenichi Harada(2)Toshiba Corporation, Isogo-ku, Yokohama, Japan. 2) Chubu Electric Power Co., Inc., Higashi-ku, Nagoya, Japan

ABWR (Advanced BWR) is an evolutionary design of a conventional BWR and the only design, among the third generation designs, with construction and operating ex-perience. Toshiba developed the ABWR design and plants have been constructed successfully in Japan with a generator output of 1,350 - 1,380MWe, and now several ABWR plants are in commercial operation. New ABWR plants are planned or con-sidered in the U.S. and other countries as well as in Japan. Comprehensive safety analyses have been performed for the ABWR in order to demonstrate safe and stable performance. Westinghouse analysis code POLCA-T, which is a BWR transient and stability analysis code, consists of a three-dimensional thermal-hydraulic model for the coolant loop of the reactor vessel and a three-dimensional neutron kinetics model, is now under the review by the NRC. Many validations of the POLCA-T code have been performed and good agreement with measured data has been obtained. In this paper, POLCA-T was validated for ABWR by comparison with the data collected from Hamaoka-5 ABWR startup test. Based on this comparison data, it is expected to be capable for the design and prediction of transient behavior for ABWR.

3:55 PMPeach Bottom 2 Turbine Test 2 Simulation by TRACE-S3K Coupled CodeKonstantin Nikitin(1), Jerry Judd, Gerardo M. Grandi(2), Annalisa Manera and Hakim Ferroukhi(3) 1) Paul Scherrer Institut, Villigen, Switzerland. 2) Studsvik, Idaho Falls ID USA. 3) Paul Scherrer Institut, Villigen, Switzerland

A coupling between the TRACE system thermal-hydraulics code and the SIMULATE-3K (S3K) three-dimensional reactor kinetics code has recently been developed in a collaboration between the Paul Scherrer Institut (PSI) and Studsvik. In order to verify the coupling scheme and the coupled code capabilities the NEA/OECD Turbine Trip benchmark was simulated. The core/plant system data were taken from the bench-mark specifications while the nuclear data were generated with Studsvik’s lattice code CASMO-4 and core analysis code SIMULATE-3. The comparison with the ex-perimental data shows that the TRACE/S3K code reproduces well the main transient parameters, namely, the pressure wave propagation, void collapsing and core power response.

4:20 PMModeling of VVER-1000 Partial Trip with Kinetic CodesA.I. Zhukov, A.M. Abdullayev, and S.V. MaryokhinNSC “Kharkov Institute for Physics and Technology”, Kharkov, Ukraine

A transient during fast power reduction of a VVER-1000 core is analyzed. Results obtained with point and 3D kinetics are presented. Results are compared with op-erational data and with some other kinetics codes. It is shown that the finite element kinetic code DiFis is acceptable for such transients in VVER-1000 cores.

4:45 PMBenchmarking of Transient Codes against Cycle 19 Sta-bility Measurements at Leibstadt Nuclear Power Plant (KKL)Carlos R Aguirre, et al.Kernkraftwerk Leibstadt AG, Leibstadt, Aargau, Switzerland

Coupled neutronics-thermal hydraulic codes are used by many utilities, research in-stitutes and regulatory authorities worldwide for performing BWR stability analysis. RAMONA-3 has been established in the industry for quite a long time as a reliable time-domain dynamic code with best performance for predictive calculations. Next generation of codes such as RAMONA-5, SIMULATE-3K and POLCA-T, with ad-vanced two-group neutronics and more detailed plant description and thermal hydrau-lics models have been introduced. The performance of these codes against the stabil-ity measurements performed in cycle 19 at the Swiss nuclear power plant Leibstadt (KKL), a BWR/6 from General Electric, is presented in this paper. Important suppliers of the nuclear industry such as Westinghouse Electric Sweden, AREVA NP Germany, Studsvik Scandpower Inc. USA, and the Swiss research institute PSI have participated in this work. The validation of calculation methods against the KKL stability measure-ments was considered important by the various organizations for different reasons. Amongst others, Studsvik Scandpower aimed at filling a gap in the SIMULATE-3K stability benchmark database to include a jet pumps driven plant, AREVA NP had to fulfill fuel licensing requirements, and Westinghouse planned to launch POLCA-T parallel to a validation of RAMONA-5 as a production code. PSI cooperated with KKL in stability issues from the very beginning and introduced the stability test project in the framework of NACUSP, a European consortium that aimed for a better understanding of the BWR stability problem. For that purpose, this validation provides an assessment of advanced stability codes for modern BWR core designs.

5:10 PMExperimental Orthogonal Functions for the Qualifica-tion of BWR Stability Events. Application To Peach Bot-tom NPPT. Barrachina, R. Miró, G. Verdú(1), D. Ginestar(2)1) Institute for Industrial, Radiophysical and Environmental Safety (ISIRYM), Univer-sitat Politècnica de València (UPV), València, Spain. 2) Department of Applied Math-ematics, Universitat Politècnica de València (UPV), Valencia, Spain

In this work, BWR stability analysis was performed on an operating point (PT_UPV) of Peach Bottom NPP which is inside the exclusion region. The simulation was made with the coupled code RELAP5-MOD3.3/PARCSv2.7. This point is achieved depart-ing from test point 3 by a control rod movement as it is usually performed in Nuclear Power Plants. The transient starts with this control rod movement. The purpose of this study is to qualify this coupled code against this kind of 3D complex accidents that take place inside the core. The calculated results show that point PT_UPV is an unstable point and the obtained relative axial power distribution shows a bottom-peaked profile, which is characteristic of unstable cores. After control rod movement, a limit cycle inphase oscillation on the total reactor power evolution is obtained, together with a coupled out-ofphase oscillation in the whole 3D power evolution. In order to complete the stability analysis, the local power range monitors (LPRMs) simulated signals ob-tained from the neutronic code has been analyzed using the singular system analysis. The results confirm the out-of-phase power oscillation in the core.


8A - Three-dimensional Transient Analysis Methods - ISession Chairs: Gerardo M Grandi (Studsvik) , Martin Zimmermann (PSI)

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3:30 PMBWR-Stability: Analysis of Cladding Temperature for High Amplitude Oscillations P. Pohl and F.WehleAREVA, AREVA NP, Erlangen, Germany

Power oscillations associated with density waves in boiling water reactors (BWRs) have been studied widely. Industrial research in this area is active since the invention of the first BWR. Stability measurements have been performed in various plants dur-ing commissioning phase but especially the magnitude and divergent nature of the oscillations during the LaSalle Unit 2 nuclear power plant event on March 9, 1988, renewed concern about the state of knowledge on BWR instabilities and possible consequences to fuel rod integrity. The objective of this paper is to present a simplified stability tool, applicable for stability analysis in the non-linear regime, which extends to high amplitude oscillations where inlet reverse flow occurs. In case of high amplitude oscillations a cyclical dryout and rewetting process at the fuel rod may take place, which leads in turn to rapid changes of the heat transfer from the fuel rod to the cool-ant. The application of this stability tool allows for a conservative determination of the fuel rod cladding temperature in case of high amplitude oscillations during the dryout / rewet phase. Moreover, it reveals in good agreement to experimental findings the stabilizing effect of the reverse bundle inlet flow, which might be obtained for large oscillation amplitudes.

3:55 PMValidation of ARCADIA® Transient Model with Rod Drop Test BenchmarkingS. Zheng(1), K. Segard(2) 1) AREVA , AREVA NP, Lyon Cedex, France. 2) AREVA , AREVA NP Inc., LYNCH-BURG VA, USA

A new generation of industrial code system ARCADIA® is developed by AREVA NP for LWR fuel and core design and safety analysis. The validation of this new code package is ongoing. This paper presents the validation of the ARCADIA® transient model against rod drop tests performed in French PQY 1300 MWe reactors. The full coupled neutronics/thermal-hydraulic core model was built to perform the transient calculations. The comparison of the predicted results with the measured results shows good consistency in the reactor excore detector responses on either the nuclear flux or the derivative of the nuclear flux, which is the important parameters used to define the reactor protection system in case of rod drop accident.

4:20 PMSensitivity Studies for the Main Steam Line Break in the Loviisa NPP With the HEXTRAN-SMABREElina Syrjälahti and Anitta HämäläinenVTT Technical Research Centre of Finland, VTT, Finland

The best-estimate modelling of nuclear power plants is part of the modern safety analysis. Because with three-dimensional neutron kinetics codes conservatism of all assumptions cannot be guaranteed, present trend is to use best-estimate analysis and to complete them with sensitivity and uncertainty methods. Especially statistical methods demand large amount of transient calculations. For that reason new sensi-tivity tool was programmed. The sensitivity analysis tool makes it possible to easily perform large amount of transient calculations with varied values of input parameters without remarkable amount of extra work. The main steam line break is one of the basic cases to be analysed as a part of the safety analysis of a pressurized water reactor. Its accurate modelling is very demanding task, because various complicated processes such as asymmetric power generation, flow mixing in the reactor vessel and various protection and conventional automation signals, contribute to the sce-nario. In this paper, sensitivity analysis tool has been utilised in the analysis of the main steam line break of the Loviisa VVER-440 power plant. Previous analysis has shown that the worst case might be reached with a smaller break size than double-ended guillotine break. In this paper, worst case has been searched by using several break sizes, by assuming failure of some protection systems and also starting from different initial power levels.

4:45 PMRod Ejection Accident 3D-Dynamic Analysis in Trillo NPP With RELAP5/PARCS V2.7T. Barrachina, R. Miró, G. Verdú(1), A. Ortego(2), J. C. Martínez-Murillo(3)1) Institute for Industrial, Radiophysical and Environmental Safety (ISIRYM), Univer-sitat Politècnica de València, Valencia, Spain. 2) IBERINCO. Avenida de Burgos, Ma-drid, Spain. 3) Almaraz-Trillo AIE, Av. Manoteras, Madrid, Spain

The Rod Ejection Accident (REA) belongs to the Reactivity- Initiated Accidents (RIA) category of accidents, and it is part of the licensing basis accident analyses required for pressurized water reactors (PWR). The REA consist of a rod ejection due to the failure of its driving mechanism. The evolution is driven by a continuous reactivity insertion. In the present work, we have analyzed this transient in Trillo NPP at differ-ent power conditions at the beginning of cycle (BOC) and at the end of cycle (EOC) using the coupled code RELAP5- MOD3.3/PARCSv2.7. The transient departs from an initially critical core, being the withdrawal speed of the control rod a typical bound-ing value. The simulation includes the SCRAM signal in order to compare our best-estimate results with the results from the conservative calculations. These analyses will allow knowing more accurately the PWR real plant phenomenology in the RIA most limiting conditions, in this way the conclusions will be realistic. The aim is to improve the understanding of these accidents using advanced methods.

5:10 PMFast Transient Evaluation of Rod Ejection in VVER-1000 CoreS.M.Aletaha and Ali PazirandehScience and Research Branch, Islamic Azad University, Tehran, Iran

The control rod ejection accident (REA) is categorized as a design basis accident and is analyzed at different reactor power levels. This accident is also analyzed coincident with loss of power to NPP in order to obtain the most conservative results. In this study, the HZP and HFP states at the BOC were analyzed. The control rod was always assumed to be ejected in 100ms from the fully inserted to the completely withdrawn position. The scram signal delay was taken 1s. The criterion for determining if any fuel damage has occurred, has traditionally been reaching the limiting DNBR. To simulate the accident, the PRORIA code, a computer code for analysis of fast reactivity initiated accident, was used. Since PRORIA code is designed for western PWR designs, some modifications implemented on PRORIA to match VVER-1000 hexagonal design and the resultant code is named PRORIA-IR. The results were compared with the FSAR of the reactor. In case of emergency protection system failure and loss of scram of the control rods in both HFP and HZP states, the integrity of fuel rods remain intact. Also, PRORIA-IR calculation is shown that the fuel enthalpy is below criteria from reach-ing DNB. The results of our calculations showed that power variations and total core reactivity are more or less similar to that of DINAMIKA, but, as compared with fuel temperature after 3s, there is a wide difference with DINAMIKA result. We assume that the difference is due to difference in thermal conductivity of the fuel material.


8B - Three-dimensional Transient Analysis Methods - IISession Chairs: Serhat Lider (Westinghouse), Soeren Kliem (Forschungszentrum Dresden-Rossendorf)

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10:20 AMAn Approach of SFR Safety Study through the Most Pe-nalizing Sodium Void ReactivityVincenzo Tiberi, Evgeny Ivanov and Sophie PignetIRSN, BP 17, Cedex, France

Sodium void reactivity effects can affect the plant safety significantly during accidental transients. Accordingly, they have to be accurately investigated for any new sodium cooled fast reactor concept, even if a fuel with a melting point lower than the sodium boiling temperature is adopted. Thus all new reactor concepts should be compared to each – others adopting the value of the maximal possible sodium void reactivity as a discrimination parameter. However, taking into account that the sodium void worth is spatially depended, it is not evident which volume could be voided in order to obtain the maximal reactivity increase. Typically the complete active core voiding (zones ini-tially loaded with 235U or 239Pu) is taken into account. This paper summarizes the extensive work carried-out in the IRSN to investigate the sodium-void reactivity spatial profiles of a fast sodium-cooled reactor core in the aim of defining a methodology to search for the area where the void contribution to the reactivity is strictly positive. Perturbation theory design approach available in the ERANOS 2.1 code has been adopted to evaluate the “area of positive void worth”. To do that, the code has been previously validated against experimental based benchmarks (IRPhEP) and reference calculations. The evaluation of the absolute values of reactivity profiles can be im-proved later-on adopting more sophisticated methodologies to perform more accurate calculations of the sample with the voided area determined adopting the rough pro-cedure described here. It has been demonstrated that even the non-reference way of ERANOS calculations could be used to provide the basis for different core concepts inter-comparison.

10:45 AMAnalysis of Sodium Void Coefficient for 500 MWe Metal ReactorC.P. Reddy, Kumar G Raghu and V. SathyamoorthyIndira Gandhi Centre for Atomic Research, Kalpakkam, INDIA

Metal fuelled sodium cooled fast reactors are known to have high breeding ratio and short doubling time. Due to these features, they play a very important role in the ener-gy scenario where higher power growth is required. Large sodium cooled fast reactors have positive sodium void coefficient which is considered to be undesirable feature, even though reactor safety can be established for all design based accidents like loss flow and transient over power accidents. Metal fuelled fast reactors which have harder neutron spectra are having higher sodium void coefficient compared to ceramic fuelled fast reactors. In many of the safety analysis the total sodium void is calculated and is used for the safety evaluation. However the sodium in the metal fuelled reactor has got three parts, namely bonding sodium, coolant sodium and the sodium in the inter space of subassembly hexagonal cans. In the reactor accident scenario the behavior of these three components of sodium will differ and also the sequence of the accident. In this paper reactor is modeled with all the pins and hexagonal cans fully modeled, in a detailed manner, in to Monte Carlo code and the sodium void coefficient is calculated for each of the component for two fuel zones. This study will be helpful in improving safety of the reactor and reducing the conservatism in the safety features.

11:10 AMCoupled 3-D Neutronics/Thermal-Hydraulic Core Analy-sis of a BWR Nuclear Heating TransientHakim Ferroukhi(1), Kurt Hofer(2)1) Paul Scherrer Institut, Villigen-PSI, Switzerland. 2) Axpo AG Kernenergie, Baden, Switzerland

Modern BWR core designs are characterized by an increased usage of Partial Length Rod fuel assemblies, which have been designed principally to enhance the shutdown margins and the stability performance. The increased moderator/fuel ratio induces less negative isothermal temperature coefficients (ITC) that can become positive at cold conditions. This can become a matter of concern during nuclear heating if critical-ity is reached late in the rod withdrawal sequence due to some non-negligible residual xenon poisoning or at end-of-cycle (EOC). This was recently experienced in a Swiss BWR when startup was made at EOC following a short shutdown period for main-tenance. A small uncontrolled power transient occurred just after that criticality had been reached due to a positive ITC combined with a deactivation of one subsystem of the plant residual heat removal system. Because the validation basis of modern three-dimensional (3-D) core neutronic/thermal-hydraulic transient codes at such con-ditions is rather scarce, it was considered as an appropriate case to analyze with the SIMULATE-3K code that is currently being established as principal 3-D kinetic solver at PSI. The results presented in this paper show that the qualitative behavior of the transient can be reasonably captured while the quantitative agreement with plant data is found to be highly sensitive upon the ITC magnitude. Small uncertainties in that coefficient are sufficient to affect completely the simulation quality and in that context, an adequate upstream steady-state 2-D lattice / 3-D reactor analysis methodology is shown to play a major role. During the transient, the thermal-hydraulic modeling be-comes in addition very important in order to adequately capture the formation of void, recalling the low pressure conditions, as this strongly affects the subsequent evolution of the reactivity coefficients and thereby, the core dynamical response.

11:35 AMValidation of ATUCHA-2 PHWR HELIOS and RELAP5-3D Model By Monte Carlo Cell and Core CalculationsC. Parisi, M. Pecchia and F. D’Auria(1), K. N. Ivanov(2), O. Mazzantini(3)1) San Piero a Grado Nuclear Research Group (GRNSPG), University of Pisa, Pisa, ITALY. 2) Reactor Dynamics and Fuel Management Group (RDFMG), The Pennsylva-nia State university, State College, PA, USA. 3) Nucleo-electrica Argentina Societad Anonima (NA-SA), Buenos Aires, ARGENTINA

Within the framework of the Second Agreement “Nucleoelectrica Argentina-SA – University of Pisa”, a complex three dimensional (3D) neutron kinetics (NK) coupled thermal-hydraulic (TH) RELAP5-3D model of the Atucha 2 PHWR has been developed and validated. Homogenized cross section database was produced by the lattice phys-ics code HELIOS. In order to increase the level of confidence on the results of such sophisticated models, an independent Monte Carlo code model, based on the MON-TEBURNS package (MCNP5 + ORIGEN), has been set up. The scope of this activity is to obtain a systematic check of the deterministic codes results. This necessity is particularly felt in the case of Atucha-2 reactor modeling, since its own peculiarities (e.g., oblique Control Rods, Positive Void Coefficient) and since, if approved by the Argentinean Safety Authority, the RELAP53D 3D NK TH model will constitute the first application of a neutronic thermal-hydraulics coupled code techniques to a reactor licensing project.


8C - Three-dimensional Transient Analysis Methods - IIISession Chairs: Bedirhan Akdeniz (Westinghouse), Gumersindo Verdu (Univ. Politecnica Valencia)

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10:20 AMHEMERA: A 3D Computational Tool for Analysis of Ac-cidental TransientsM. Clergeau, F. Dubois, B. Normand, A. SargeniIRSN, Fontenay-aux-Roses, FRANCE

HEMERA (Highly Evolutionary Methods for Extensive Reactor Analyses), based on the French SAPHYR code system, composed of APOLLO2, CRONOS2, FLICA4 and the system code CATHARE, is a fully coupled 3D computational chain developed jointly by IRSN and CEA. Multi-level and multi-dimensional models are developed to account for neutronics, core thermalhydraulics, fuel thermal analysis and system thermal-hydraulics, dedicated to best-estimate and conservative simulations and sen-sitivity analysis. Currently, the HEMERA chain is adopted to investigate Rod Ejection Accident (REA) and Main Steam Line Break (MSLB) accident in French PWRs. This paper presents two examples of HEMERA calculations: in the first one – concerning the REA - a comparison is made between a pin by pin approach and the more conven-tional scheme adopting four radial meshes by assembly; in the second one – concern-ing the MSLB – nuclear power distribution is penalized and the a comparison is made between the ‘photo-like’ approach (i.e., according to which the nuclear power is modi-fied externally in such a way that neutronics and thermal-hydraulics computations are decoupled) and a built-in methodology, allowing the modification of the power distribu-tion at each transient time-step, in such a way that the thermal-hydraulics calculations take continuously into account the power distortion.

10:45 AMEffect of CASMO-5 Cross-section Data and Doppler Temperature Definitions on LWR Reactivity Initiated Ac-cidentsGerardo Grandi, Kord Smith, Zhiwen Xu and Joel RhodesStudsvik Scandpower, Inc., Idaho Falls, ID, USA

During LWR Reactivity Initiated Accidents (RIA), the accurate evaluation of the Dop-pler reactivity feedback depends on the Doppler coefficient computed by the lattice physics code (e.g. CASMO-5), and on the effective Doppler temperature computed by the transient code (e.g. SIMULATE-3K) using the non-uniform intra-pellet tem-perature profile. CASMO-5 has many new features compared with its predecessor. Among them, the replacement of the L-library (based primarily on ENDF/B IV data) by the latest available nuclear data (ENDF/B VII.0), and the Monte Carlo based reso-nance elastic scattering model to overcome deficiencies in NJOY modeling have a significant impact on the fuel temperature coefficient, and hence on LWR RIA. The Doppler temperature effect in thermal reactors is driven by the 238U absorption. The different effective Doppler temperature definitions, available in the literature, try to capture the considerable self-shielding of the 238U absorption that occurs in the pellet surface by defining an appropriate fuel temperature to compute cross-sections. In this work, we investigate the effect of the nuclear data generated by CASMO-5 on RIA, as well as the impact of different effective Doppler temperature definitions, including one proposed by the authors. It is concluded: 1) LWR RIA evaluated using CASMO-5 cross section data will be milder because the energy released is ~10% smaller; 2) the prompt enthalpy rise is barely affected by the choice of the Doppler temperature definition; and 3) the peak fuel enthalpy is affected by the choice of the Doppler tem-perature definition, the under-prediction of the Doppler reactivity by the ‘NEA’ Doppler temperature results in a conservative estimate of the peak fuel enthalpy.

11:10 AMStress Analysis of Coated Particle Fuel in the Deep-Burn Pebble Bed Reactor DesignB. Boer and A.M OugouagIdaho National Laboratory, Idaho Falls, ID, USA

High fuel temperatures and resulting fuel particle coating stresses can be expected in a Pu and minor actinide fueled Pebble Bed Modular Reactor (400 MWth) design. Such stresses may be different from those expected to arise in ’standard’ UO2-fueled cores. The high discharge burnup aimed for in this Deep-Burn design results in increased power and temperature peaking in the pebble bed near the inner and outer reflectors. Furthermore, the pebble power in a multi-pass in-core pebble recycling scheme is relatively high for pebbles in their first pass through the core. These conditions might result in an increase in the incidence of mechanical failure of the TRISO fuel particle coatings, which serve as the containment of radioactive fission products in the pebble bed design. The PArticle STress Analysis (PASTA) is used to investigate the integrity of the particle fuel coatings as a function of the irradiation time (i.e. burnup), core position and conditions induced by a Loss Of Forced Cooling (LOFC) incident. For this study, the PASTA code has been coupled to the PEBBED code for neutronics, thermal-hy-draulics and depletion analysis of the core. Two deep-burn fuel types, plutonium (Pu) with and without initial minor actinides (MA) content have been investigated with the new code system for normal and transient conditions. The study also incorporates the effect of the statistical variation in the thickness of the coating layers.

11:35 AMConsistent Comparison of Full Core PWR Reactivity Ini-tiated Accident with the Method of Characteristic Code DeCART and the Coarse Mesh Nodal Code PARCSMathieu Hursin(1), Thomas J. Downar, Brendan Kochunas(2)1) University of California at Berkeley, Berkeley, CA, USA. 2) University of Michigan, Ann Arbor, MI, USA

The current state of the art in analysis of a control rod ejection event in a Pressurized Water Reactor (PWR) relies upon the assembly averaged power from a whole core nodal neutronics simulator and some type of pin power reconstruction within the fuel assembly. Recently, there has been interest in taking advantage of the DeCART code to perform a higher fidelity solution which could lead to more accurate pinpower results as well as provide intrapin power information during the transient. The work described in this paper is the comparison of PARCS and DeCART analysis of two Reactivity Initi-ated Accidents. The methods used in PARCS and DeCART are briefly described as well as the approach to generate the needed temperature feedbacks. The generation of the macroscopic cross sections and kinetic parameters for PARCS is detailed. The results of both scenarios are shown and the main differences of both approaches are discussed.


8D - Multi-Physics Reactor Simulations - ISession Chairs: Tom Downar (univ. Michigan), Carlo Parisi (Univ. Pisa)

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1:30 PMPrototype Coupling of the CFD Software ANSYS CFX With the 3D Neutron Kinetic Core Model DYN3DS. Kliem and U. Rohde(1), J. Schütze(2), Th. Frank(3) 1) Forschungszentrum Dresden-Rossendorf; Institute of Safety Research, Dresden, Germany. 2) ANSYS Germany GmbH, Darmstadt, Germany. 3) ANSYS Germany GmbH, Otterfing, Germany

The CFD code ANSYS CFX has been coupled with the neutron-kinetic core model DYN3D. ANSYS CFX calculates the fluid dynamics and related transport phenomena in the reactor’s coolant and provides the corresponding data to DYN3D. In the fluid flow simulation of the coolant, the core itself is modeled within the porous body ap-proach. DYN3D calculates the neutron kinetics and the fuel behavior including the heat transfer to the coolant. The physical data interface between the codes is the volumetric heat release rate into the coolant. In the prototype that is currently avail-able, the coupling is restricted to single-phase flow problems. In the time domain an explicit coupling of the codes has been implemented so far. Steady-state and transient verification calculations for a small-size test problem confirm the correctness of the implementation of the prototype coupling. This test problem was a mini-core consist-ing of nine real-size fuel assemblies. Comparison was performed with the DYN3D standalone code. In the steady state, the effective multiplication factor obtained by the ANSYS CFX/DYN3D codes shows a deviation of 9.8 pcm from the DYN3D stand-alone solution. This difference can be attributed to the use of different water property packages in the two codes. The transient test case simulated the withdrawal of the control rod from the central fuel assembly at hot zero power. Power increase during the introduction of positive reactivity and power reduction due to fuel temperature in-crease are calculated in the same manner by the coupled and the stand-alone codes. The maximum values reached during the power rise differ by about 1 MW at a power level of 50 MW. Beside the different water property packages, these differences are caused by the use of different flow solvers.

1:55 PMNew Developments of the MCNP/CTF/NEM/NJOY Code System – Monte Carlo Based Coupled Code for High Accuracy Modeling Federico Puente Espel, Maria N. Avramova, Kostadin N. Ivanov(1), Stefan Misu(2) 1) Department of Mechanical and Nuclear Engineering, The Pennsylvania State Uni-versity, University Park, PA, USA. 2) AREVA, AREVA NP GmbH, Erlangen, Germany

High accuracy code systems are necessary to model core environments with con-siderable geometry complexity and great material heterogeneity. These features are typical of current and innovative nuclear reactor core designs. Advanced methodolo-gies and state-of-the art coupled code systems must be put into practice in order to model with high accuracy these challenging core designs. The presented research comprises the development and implementation of the thermal-hydraulic feedback to the Monte Carlo method and of speed-up mechanisms to accelerate the Monte Carlo criticality calculation. Coupled Monte-Carlo calculations can serve as reference solutions for verifying high-fidelity coupled deterministic neutron transport methods with detailed and accurate thermal-hydraulic models. The development and verifica-tion of such reference highfidelity coupled multi-physics scheme is performed at the Pennsylvania State University (PSU) in cooperation with AREVA, AREVA NP GmbH in Erlangen, Germany, on the basis of MCNP5, NEM, NJOY and COBRA-TF (CTF) computer codes. This paper presents the latest studies and ameliorations developed to this coupled hybrid system, which includes a new methodology for generation and interpolation of Temperature-Dependent Thermal Scattering Cross Section Librar-ies for MCNP5, a comparison between sub-channel approaches, and acceleration schemes.

2:20 PMImplicit Time-Integration Method for Simultaneous Solu-tion of A Coupled Non-Linear SystemJustin K. Watson, Kostadin N. IvanovPenn State University, Applied Research Laboratory, State College, PA USA

Historically large physical problems have been divided into smaller problems based on the physics involved. This is no different in reactor safety analysis. The problem of analyzing a nuclear reactor for design basis accidents is performed by a hand-ful of computer codes each solving a portion of the problem. Traditionally each of these codes operates independently from each other using only the global results from one calculation as boundary conditions to another. Industry’s drive to uprate power for reactors has motivated analysts to move from a conservative approach to design basis accident towards a best estimate method. To achieve a best estimate calcula-tion efforts have been aimed at coupling the individual physics models to improve the accuracy of the analysis and reduce margins. The current coupling techniques are sequential in nature. During a calculation time-step data is passed between the two codes. The individual codes solve their portion of the calculation and converge to a solution before the calculation is allowed to proceed to the next time-step. This paper presents a fully implicit method of solving the neutron balance equations, heat conduc-tion equations and the constitutive fluid dynamics equations. The paper discusses the problems involved in coupling different physics phenomena within multi-physics codes and presents a solution to these problems. The paper also outlines the basic concepts behind the nodal balance equations, heat transfer equations and the thermal hydraulic equations, which are coupled to form a fully implicit nonlinear system of equations.

2:45 PMBond Graph Based Coupled Reactor SimulationsEugeny Sosnovsky, Benoit Forget(1), Christopher Newman(2)1) Massachusetts Institute of Technology, Department of Nuclear Science and Engi-neering, Cambridge, MA USA. 2) Idaho National Laboratory, Idaho Falls, ID USA

The bond graph formalism was first introduced to solve the multiphysics problem in electromechanical systems. Over the years, it has been used in many fields including nuclear engineering, but with limited scope due to its perceived impracticality in large systems. This paper introduces this not so well known formalism and presents results from an automated bond graph processing code that simplifies the solution of mult-iphysics problems. The bond graph of a fully coupled 1D heat transfer and neutron dif-fusion reactor is derived and the transient solution obtained from the proof-of-concept code is verified using a manufactured solution.


8E - Multi-Physics Reactor Simulations - IISession Chairs: Maria Avramova (PSU), Yann Perin (GRS)

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08:00 AMCOBRA-TF / QUABOX-CUBBOX: Code System for Cou-pled Core and Subchannel AnalysisPérin Y., Velkov K., Pautz A. Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH, Forschungsinstitute, Garching, Germany

An overview is given on the recent GRS developments and applications of a neutron kinetic/thermal-hydraulic (NK/TH) core model based on the coupling of the 3D-neutron kinetics model QUABOX-CUBBOX (coarse-mesh diffusion code) with the thermal-hydraulic sub-channel code COBRA-TF. The GRS global coupling approach and the features of both codes are briefly presented. A test case on the basis of the OECD/NEA/US-NRC MOX/UO2 core transient benchmark is discussed and used to check the new coupled code system. Relevant results obtained from this test case are sum-marized and compared with the results obtained with another GRS coupled code sys-tem ATHLET/QUABOX-CUBBOX. The practical applicability of the new coupled code system is discussed.

08:25 AMImprovements to the NEM/THERMIX Coupled Code Analysis of High Temperature ReactorsPeter T. Mkhabela and Kostadin N. Ivanov The Pennsylvania State University, University Park, PA, USA

The NEM-THERMIX code has been developed at Pennsylvania State University (PSU) for High Temperature Reactor (HTR) core design and safety analysis. This coupling methodology is based on the neutronics core simulator NEM and the HTR thermal hydraulic code THERMIX-DIRECT. This paper presents the results of the cur-rent developments, which enhance the coupled code capabilities of analyzing the HTR transients. The obtained results are from the analysis of PBMR- 400 reactor model described in the OECD/NEA PBMR-400 benchmark specifications. The DLOFC and PLOFC results show that about 23% of fuel will be above the 1600°C limit for the safe operation of the reactor.

08:50 AMDevelopment and Optimization of Coupling Interfaces Between Reactor Core Neutronics and Thermal-Hydrau-lic CodesIlyes Gouja, Maria Avramova, Adam RubinThe Pennsylvania State University (PSU), Department of Mechanical and Nuclear Engineering, University Park, PA USA

Coupling nuclear engineering codes has become a necessity as the nuclear industry increasingly needs a “best estimate” approach to design new reactors or to optimize the load and the cycle of the nuclear fuel in the existing reactors. This best estimate methodology based on a multi-physics capability, as an alternative of the conservative approach used so far, aims to produce more accurate safety margins and to utilize better the fuel potential and the core design. The main goal of this work is further development of multi-physics coupling methodologies for design and safety evalu-ations of current and innovative reactor systems by establishing and optimizing the coupling schemes between neutron physics and thermal hydraulic (T-H) simulation codes. The focus is on the development and benchmarking of the efficient coupled multi-physics code system CTF/NEM designed to provide a unique and novel platform for implementation of embedded multi-scale (global scale – fuel assembly level and local scale – fuel pin level) algorithms in space and time domains of non-linear coupled simulations. The two coupled codes NEM and CTF are maintained at the Pennsylva-nia State University (PSU). The testing of the interfaces for steady state and transient calculations reported in this paper is based on code-to-code comparisons using an appropriate benchmark (partially fueled PWR MOX core), which has been performed under the auspices of OECD/NEA. The ability of the coupled CTF/NEM code to simu-late steady state and transient of a PWR core has been demonstrated.

09:15 AMOn Delayed Neutron Effects at the Starting of Accelera-tor Driven Systems to the Nominal PowerRubens Souza dos Santos(1,2)1) Instituto de Engenharia Nuclear,Cidade Universitária, Rio de Janeiro, Brazil. 2) In-stituto Nacional de Ciência e Tecnologia em Reatores Inovadores CNPQ/MCT, Brazil

Accelerator Driven Systems (ADS) are sub-critical nuclear reactor cores driven by an external spallation neutron source. These promising devices must be used not only as dedicated burners of transuranic elements but also as energy producers. The spal-lation neutrons are provided by the bombardment of a heavy metal, when impinged by proton beam, from a high energy proton accelerator. Usually, most simulations of these systems are based on a steady state condition given by the power opera-tion. There is a common sensus that ADS are source dominated. That implies that delayed neutrons used to be neglected in most of transient simulations. In this paper, we carried out some transient calculations to assess the effect of delayed neutrons on the ADS starting up. For that, use was made of a kinetic reactor multigroup diffusion model, considering a train of proton beam pulses, localized in the middle of a 1D core. Two cases were analyzed: one disregarding the delayed neutron presence; another one, taking into account the delayed neutron fraction. Numerical results showed that, despite of source dominance, the power level of the ADS during the startup is strongly dependent on the presence of delayed neutrons. This dependence is accentuated if the sub criticality level is reduced. In addition, the results also lead to the necessity of managing the beam intensity with the pulse duration time to maintain ADS power in a stable condition, when a pulsed variation form nominal power is simulated.

09:40 AMCoupled TORT-TD/CTF Capability for High-Fidelity LWR Core CalculationsM. Christienne, M. Avramova(1), Y. Perin, A. Seubert(2)1) Department of Mechanical and Nuclear Engineering, The Pennsylvania State Uni-versity, University Park, PA USA. 2) Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH, Forschungsinstitute, Garching, Germany

This paper describes the developed coupling scheme between TORT-TD and CTF. TORT-TD is a time-dependent 3D discrete ordinates neutron transport code. TORT-TD is utilized for high-fidelity reactor core neutronics calculations while CTF is provid-ing the thermal-hydraulics feedback information. CTF is an improved version of the advanced thermal-hydraulic sub-channel code COBRA-TF, which is widely used for best-estimate evaluations of LWR safety margins. CTF is a transient code based on a separated flow representation of the two-phase flow. The coupled code TORT-TD/CTF allows 3D pin-by-pin analyses of transients in few energy groups and anisotropic scattering by solving the time-dependent transport equation using the unconditionally stable implicit method. Steady-state and transient test cases, based on the OECD/NRC PWR MOX/UO2 Core Transient Benchmark, have been calculated. The steady state cases are based on a quarter core model while the transient test case models a control rod ejection transient in a small PWR mini-core fuel assembly arrangement. The obtained results with TORT-TD/CTF are verified by a code-to-code comparison with the previously developed NEM/CTF and TORT-TD/ATHLET coupled code sys-tems. The performed comparative analysis indicates the applicability and high-fidelity potential of the TORT-TD/CTF coupling.


8F - Multi-Physics Reactor Simulations - IIISession Chairs: Brian Boer (INL), Antonio Sargeni (IRSN)

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08:00 AMEvaluation of Aleatory and Epistemic Uncertainties in Level 2 Probabilistic Safety Analysis by Means of Bayes-ian Monte Carlo MethodsAxel Hoefer(1), Gerben Dirksen, Jürgen Eyink, Eva-Maria Pauli(2) 1) AREVA NP GmbH Offenbach, Offenbach, Germany. 2) AREVA NP GmbH Erlan-gen, Erlangen, Germany

Uncertainties to be taken into account in a Level-2 PSA are due to the stochastic na-ture (aleatory uncertainty) of parameters characterizing accident progression and due to the limited knowledge (epistemic uncertainty) about the corresponding probability distributions and about relevant physical parameters defined by fixed values. Both kinds of uncertainty can be described consistently within Bayesian statistics. Using a Bayesian approach, it is shown how uncertainties in the basic parameters are trans-lated into branch probability distributions (bpd’s) corresponding to the questions in an accident progression event tree (APET), reflecting the degree of belief with respect to probability values assigned to the answers to the respective APET questions. Since an analytical evaluation of a bpd is not feasible, in general, we propose a Monte Carlo procedure for drawing random samples from a bpd which afterward can be estimated from the Monte Carlo data. This Monte Carlo approach also offers the possibility to directly include information in the form of empirical data. The obtained bpd’s of differ-ent questions can finally be combined according to the respective APET, with the aid of a suitable event tree program, which yields probability distributions reflecting the degree of belief assigned to probability values corresponding to the different release categories.

08:25 AMFast Sensitivity Ranking and Reliability Analysis of Pas-sive Thermal Hydraulic Safety System by Automatic Dif-ferentiationA. John Arul, P. Mohanakrishnan(1), N. Kannan Iyer and A. K. Verma(2) Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu, India. 2) Depart-ment of Mechanical Engineering, *Department of Electrical Engineering, Indian Insti-tute of Technology, Mumbai, India

A method for efficient estimation of passive safety system reliability and sensitivity ranking of input parameters has been developed. It is based on generating linear response surface about the design point using the technique of Automatic Differentia-tion, which is subsequently used for propagating input uncertainty through the best estimate thermal hydraulics code by importance sampling Markov Chain Monte Carlo simulation in the most probable failure region. Results on the application of the method to a simplified model of the passive residual heat transport system in fast reactors, is presented to demonstrate the gain in computational time and improved prediction of low values of failure probability.

08:50 AMApplication of SMAP Methodology to a 10% Power Up-Rate for Zion NPP SBO TransientF. Fouet, P. Probst, J.M. LanoreIRSN, Institut de Radioprotection et de Sûreté Nucléaire, Fontenay-aux-Roses, France

During the past years, nuclear power plants underwent some major changes in their design and operation mode to fulfill new objectives, such as power up-rate, life ex-tension and/or increased fuel burn up. While fulfilling all the regulatory requirements, these changes – not necessarily and/or completely accounted for in the original design – can challenge the plant safety margins and induce a potential increase of the risk. In order to assess the impact of such modifications on the safety margins, the Com-mittee on the Safety of Nuclear Installations (CSNI) approved in December 2003 a Safety Margins Action Plan (SMAP) and established an international Working Group aimed at developing a new methodology to address the problem, which has been successfully done [1]. Then, the CSNI mandated a SM2A (Safety Margin, Assessment and Application) group to apply the SMAP methodology to a real case. A 10% power up-rate for Zion PWR has been selected as NPP modification and the increase of Core Damage Frequency (CDF) was selected as a metric for change in safety margin ap-praisal. A Probabilistic Safety Analysis – based (PSA) investigation phase, shared by the organizations participating in the group, aimed at selecting the event trees and the main sequences, which the change could potentially affect in a significant way. It was agreed that, once the selection done, calculations had to be run to quantify the effects. The reference PSA for Zion NPP used in the exercise was performed in the framework of NUREG 1150. For the SM2A exercise it was agreed that the risk measure is CDF. IRSN was in charge of the Loss Of offSite Power (LOSP) event trees. In this family of events it appeared that 2 sequences correspond to the conditions (a potential effect with a non negligible probability). These 2 selected sequences were Station Black Out (SBO) scenarios: SBO with loss of Auxiliary Feed Water (AFW) and SBO with a seal Loss Of Coolant Accident (LOCA).

09:15 AMFunctional Reliability Analysis of Safety Grade Decay Heat Removal System of PFBR By Augmentation of Ap-proximate Linear Response SurfacesSajith Mathews T(1), A John Arul(2), U. Parthasarathy(2), K. V. Subba-iah(1)1) AERB-Safety Research Institute, Tamilnadu, India. 2) Indira Gandhi Centre for Atomic Research, Tamilnadu, India

Functional reliability analysis of passive safety systems is necessary to quantify the ability of the system to accomplish the intended function unaffected by the uncer-tainties pertained to the underlying physical processes. Functional reliability analysis includes Monte Carlo sampling of the uncertainties followed by computation of sys-tem response by a deterministic system code. For complex passive safety systems of small failure probability, Monte Carlo simulations using deterministic system codes are computationally expensive and often prohibitive. Functional failure analysis using computationally efficient approximate solutions like response surfaces has been quite popular. But in high dimensional problems, due to their approximate nature, response surfaces are not accurate enough. Nevertheless, approximate solutions provide a great source of insight for understanding system behavior. In this respect, the recently proposed Response Conditioning Method (RCM) based on Subset Simulation (SS) is considered in this paper to incorporate the knowledge obtained by approximate solutions in functional reliability analysis for obtaining consistent and computationally efficient reliability estimates. The method is applied to evaluate functional reliability of Passive Safety Grade Decay Heat removal system of Indian 500 MWe Prototype Fast Breeder Reactor (PFBR). The results are compared with direct Monte Carlo simula-tion and observed that the method considered is computationally very efficient and provides consistent reliability estimates.


8G - Developments in Probabilistic Risk AssessmentsSession Chairs: Cesare Frepoli (Westinghouse), Hongwu Cheng (Vattenfall)

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1:30 PMAnalysis of Steam Generator Tube Rupture Event in HTR-MODUL Reactor Alex Matev Pebble Bed Modular Reactor (Pty.) Ltd., Centurion, South Africa

The RELAP5-3D code was used to model the HTR-MODUL reactor and analyze an event of a single steam generator tube rupture (SGTR), proceeding with, or without operator intervention, or actuation of any reactor protection. Four cases have been analyzed to evaluate the impact from mitigation actions and from the location of the steam generator (SG) tube break on reactor power, primary coolant pressure, and mass of vapor in the core. The simulation results provide a basis for evaluating the magnitude of challenges to plant safety as reactor overpower, reactor coolant system (RCS) over-pressure, and release of radioactive nuclides from the primary system.

1:55 PMPressure Vessel SBLOCA Simulation with TRACE5. Ap-plication to LSTF (ROSA V) Abella V, Gallardo S and Verdú G Departamento de Ingeniería Química y Nuclear, Universidad Politécnica de Valencia, Valencia, Spain

In this work, an overview of the results obtained in the simulation of an Upper Head Small Break Loss-Of-Coolant-Accident (SBLOCA) under the assumption of total fail-ure of High Pressure Injection System (HPIS) in the Large Scale Test Facility (LSTF) is provided. In previous works, an SBLOCA located in the Pressure Vessel (PV) Lower Plenum was simulated with TRACE. In that case, an asymmetrical steam generator secondary-side depressurization was produced as an accident management action at the Steam Generator in loop without pressurizer after the generation of safety injec-tion signal to achieve a determined depressurization rate in the primary system. The new SBLOCA scenario has been simulated and results compared with experimental values, with the purpose of completing the analysis of PV SBLOCA. This study is developed in the frame of the OECD/NEA ROSA Project Test 6-1 (SB-PV-9 in JAEA). Finally, the present paper represents a contribution for the study of safety analysis of vessel SBLOCAs and the assessment of the predictability of thermal-hydraulic codes like TRACE.

2:20 PMApplication and Licensing of Westinghouse Realis-tic Large Break Loca Evaluation Model (ASTRUM) for Maanshan Units 1 and 2 Nuclear Power PlantCesare Frepoli(1), Julian S. Chiang(2), Jeffrey S. Petzold(1), Dewey C. Olinski(1)1) Westinghouse Electric Company, Pittsburgh, PA, USA. 2)Taiwan Power Company, Taipei, Taiwan

A complete best-estimate plus uncertainty analysis was performed to evaluate the large break lossof- coolant accident (Design Basis Accident) for Maanshan Units 1 and 2 using Westinghouse Evaluation Model - Automated Statistical Treatment of Uncer-tainty Method (ASTRUM). ASTRUM follows the steps in the Code Scaling Applicability and Uncertainty (CSAU) methodology, where the1 uncertainty analysis (Element 3 in the CSAU) is replaced by a technique based on order statistics. ASTRUM methodol-ogy received generic approval (Safety Evaluation Report) by the USNRC in 2004. ASTRUM methodology was here applied to Maanshan Units 1 and 2 and has been under review by the local Safety Authorities (Republic of China Atomic Energy Coun-cil – ROCAEC). The paper describes the development of the model, the execution of the analysis and the process utilize to assess the uncertainties. Based on the analysis assumptions, the results show compliance with the Licensing Limits for Large Break LOCAs as described in 10 CFR 50.46 (b)(1), (b)(2), and (b)(3).

2:45 PMOECD/NRC Benchmark Based on NUPEC PWR Sub-channel and Bundle Tests (PSBT)A. Rubin and M. Avramova(1), H. Utsuno(2)1) Nuclear Engineering Program, The Pennsylvania State University, University Park, PA, USA. 2) Japan Nuclear Energy Safety Organization, Minato-ku, Tokyo, Japan

The OECD/NRC PSBT Benchmark is organized to provide a test bed for assessing the capabilities of various thermal-hydraulic subchannel, system, and computational fluid dynamics (CFD) codes. The benchmark is designed to systematically assess and compare the participants’ numerical models for prediction of detailed subchannel void distribution and DNB, under steady-state and transient conditions, to full-scale experimental data. It consists of seven exercises grouped in two phases. In an attempt to determine the strengths and weaknesses of the models utilized in each code, code-to-code comparisons will be reformed in addition to the code-to-data assessment. The benchmark team is based on collaboration between the Pennsylvania State University (PSU) and the Japan Nuclear Energy Safety (JNES) organization including the partici-pation and support of the U.S. Nuclear Regulatory Commission (NRC) and the Nuclear Energy Agency (NEA), OECD. The PSU in-house code CTF, an improved version of the advanced thermal-hydraulic subchannel code COBRA-TF, is used for preliminary scoping calculations of selected benchmark exercises.


8H - Safety Analysis and Thermal-HydraulicsSession Chairs: Kurshad Muftuoglu (GE), Axel Hoefer (AREVA)

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1:30 PMResearch Activities in the ZED-2 Critical Facility A. M. Rauket and G. P. McPhee Atomic Energy of Canada Limited, Reactor and Radiation Physics Branch, Chalk River, Ontario, Canada

This paper will provide an updated overview of the capabilities and standard experi-ments undertaken in the ZED-2 (Zero Energy Deuterium) critical facility. ZED-2 is a versatile, heavy-water moderated, low-power research reactor located at Atomic Energy of Canada Limited’s Chalk River Laboratories (CRL). Its primary role is to provide reactor core physics data. This data has been historically used to support the fuel and reactor physics code development for AECL technologies (primarily CANDU® technologies) and is currently used in support of fuel development and reactor physics code validation supporting the development of AECL’s advanced reactor technologies. The reactor is a large, heavy-water moderated and graphite reflected tank type critical facility with a cylindrical calandria. The average thermal neutron flux at full power is approximately 5 x 108 n cm-2 s-1 with a peak of 1 x 109 n cm-2 s-1. Standard ex-periments that are performed included: • critical configurations, • global and local flux shape measurements, • reaction rate measurements (including inside and between fuel bundles), • kinetics measurements, • reactivity worth measurements of control devices, and • calibration of reactor instrumentation.The unique features of ZED-2 include the ability to readily change the position, spac-ing and coolant type of the fuel assemblies in the reactor (heavy water, light water and gas coolants are commonly used). In addition, various fuel compositions (e.g. natural uranium, low-enriched uranium, mixed-oxides and thorium-based fuels) are readily accommodated.

1:55 PMLarge-Scale Irradiated Fuel Experiments at PROTEUS Research Program M. F. Murphy, G. Perret, O. Köberl, K. A. Jordan, P. Grimm, H. Kröhnert, M. A. Zimmermann Paul Scherrer Institut, Villigen, Switzerland

The LIFE@PROTEUS joint experimental program between the Paul Scherrer Institut and the Swiss nuclear utilities aims at studying the interface between highly burnt and fresh fuel assemblies in pressurized and boiling water reactors using the PROTEUS zero power research reactor.In a first phase, compositions, burnups and reactivity worths of burnt (40 and 60 GWd/t) PWR fuel pins next to fresh ones will be measured. Reaction rates distri-butions across the fresh/burnt fuel pins interface will also be measured using newly developed techniques in PROTEUS. The techniques are based on delayed neutrons or high energy gamma-rays emitted by short lived fission products, both produced in pins re-irradiated in PROTEUS.LIFE@PROTEUS started in 2006 and will require a two year refurbishment of PRO-TEUS to operate with highly burnt fuels. Today, in parallel to the effort for the submit-tal of the construction license request to the safety authorities, future core loading configurations are predicted with MCNPX and CASMO-4E. Preliminary results show that a 6x6 burnt fuel bundle should adequately represent the interface between fresh and burnt PWR fuel assemblies, and that CASMO-4E appears suitable to analyze this complex experiment.

2:20 PMBenchmark Evaluation of the Start-Up Core Reactor Physics Measurements of the High Temperature Engi-neering Test ReactorJohn Darrell BessIdaho National Laboratory, Idaho Falls, ID USA

The benchmark evaluation of the start-up core reactor physics measurements per-formed with Japan’s High Temperature Engineering Test Reactor, in support of the Next Generation Nuclear Plant Project and Very High Temperature Reactor Program activities at the Idaho National Laboratory, has been completed. The evaluation was performed using MCNP5 with ENDF/B-VII.0 nuclear data libraries and according to guidelines provided for inclusion in the International Reactor Physics Experiment Eval-uation Project Handbook. Results provided include updated evaluation of the initial six cold critical core configurations (five annular and one fully-loaded). The calculated keff eigenvalues agree within ~3σ of the benchmark values and are approximately 2% greater than the benchmark values. Reactor physics measurements that were evaluated include reactivity effects measurements such as excess reactivity during the core loading process and shutdown margins for the fully-loaded core, six isothermal temperature reactivity coefficient measurements for the fully-loaded core, and axial re-action rate measurements in the instrumentation columns of three core configurations. The calculated values agree well with the benchmark experiment measurements. Fully subcritical and warm critical configurations of the fully-loaded core were also assessed. The calculated keff eigenvalues for these three configurations also agree within ~3σ of the benchmark values. Results are comparable with those obtained in Japanese evaluations. The reactor physics measurement data can be used in the validation and design development of future High Temperature Gas-cooled Reactor systems.

2:45 PMCalculation of the Core Parameters Measured During the Commissioning of the OPAL ReactorEduardo Villarino, Daniel Hergenreder(1), George Braoudakis and Tunay Ersez(2)1) INVAP SE, Nuclear Engineering Division, Bariloche, Argentina. 2) ANSTO, Nuclear Analysis, Sydney, Australia

The OPAL Research reactor is a multi-purpose open-pool type reactor. The nominal fission power of the reactor is 20 MW. It was commissioned during the second half of the year 2006. The reactor has several nuclear safety related design criteria that have to be experimentally verified during Stage B of the commissioning of the reactor. The present work presents the measurements carried out during the Stage B of the commissioning of the OPAL reactor, and the numerical verification of the calculated values using the design calculation methodology against these measured values. A brief description of the OPAL reactor, its commissioning plan, its nuclear safety related design criteria and the calculation and the experimental methodology are presented. The measured values and a comparison with the calculated is also given.


9A - Research Reactor Facilities and BenchmarkingSession Chairs: Frederik Reitsma (PBMR), John Bess (INL)

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3:30 PMSelected Studies of Past Operations At the ORNL High Flux Isotope ReactorDavid Chandler(1), R. T. Primm, III(2)1) Nuclear Engineering Department, The University of Tennessee, Knoxville, Tennes-see USA. 2) Oak Ridge National Laboratory, Research Reactors Division, Oak Ridge, Tennessee USA

In response to on-going programs at Oak Ridge National Laboratory, two topics re-lated to past operations of the High Flux Isotope Reactor (HFIR) are being reviewed and include determining whether HFIR fuel can be converted from high enriched ura-nium (HEU) to low enriched uranium (LEU) and determining whether HFIR beryllium reflectors are discharged as transuranic (TRU) waste. The LEU conversion and TRU waste studies are being performed in accordance with the Reduced Enrichment for Research and Test Reactors program and the Integrated Facility Disposition Project, respectively. While assessing data/analysis needs for LEU conversion such as the fuel cycle length and power needed to maintain the current level of reactor performance, a reduction of about 8% (~200 MWD) in the end-of-cycle exposure for HFIR fuel was observed over the lifetime of the reactor (43 years). Potential causes for the decline are examined. In a second study, the SCALE 6.0 computational system was used to evaluate discharged beryllium reflectors and it was discovered if the reflectors are procured according to the current HFIR standard, discharged reflectors would not be TRU waste, but the removable reflector (closest to core) would become TRU waste approximately 40 years after discharge. However, beryllium reflectors have been fab-ricated with a greater uranium content than that stipulated in the standard and these reflectors would be discharged as TRU waste.

3:55 PMWashington State University Reactor HEU to LEU Con-version: A Comparison of HEU vs. LEU Fuel BehaviorDon E. Wall

Presentation Only

4:20 PMLEU Conversion of the NIST Research Reactor: Main-taining Core Geometry to Avoid a Startup CoreRobert E. Williams, Wade J. Richards, Sean O’Kelly and J. Michael Rowe (Consultant)(1), David J. Diamond, Albert L. Hanson, Lap-Yan Cheng and Arantzazu Cuadra(2)1) NIST Center for Neutron Research, Gaithersburg, MD USA. 2) Energy Sciences and Technology Department, Brookhaven National Laboratory, NY USA

The research reactor, NBSR, at the National Institute of Standards Technology (NIST) is one of the five High Performance Reactors in the United States targeted for conver-sion to LEU fuel with the U-Mo fuel currently under development. Initial studies of the impact of conversion on the performance of the experimental facilities showed that we can expect losses of about 10% of the flux intensity available to the thermal and cold neutron scattering instruments. To mitigate this loss without raising the reactor power, many changes in the fuel element geometry have been studied. An axially compressed core with ~20% shorter fuel plates could restore nearly half of the performance loss that would accompany conversion to a LEU core with the existing geometry. Analysis indicates that the thermal-hydraulic margins will be adequate. Many other factors had to be considered, however, and as a result, a decision was made to keep the existing fuel element geometry and convert the NBSR over eight reactor cycles according to the existing fuel management scheme.

4:45 PMImproved Computational Characterization of the Ther-mal Neutron Source for Neutron Capture Therapy Re-search At the University of MissouriStuart R. Slattery(1), David W. Nigg(2), John D. Brockman(3), M. Frederick Hawthorne(4)1) University of Wisconsin, Madison, WI USA. 2) Idaho National Laboratory, Idaho Falls, ID USA. 3) University of Missouri Research Reactor Center, Columbia, MO USA. 4) University of Missouri, International Institute of Nano & Molecular Medicine, Colum-bia, MO USA

Parameter studies, design calculations and initial neutronic performance measure-ments have been completed for a new thermal neutron beamline to be used for neu-tron capture therapy cell and small-animal radiobiology studies at the University of Missouri Research Reactor. The beamline features the use of single-crystal silicon and bismuth sections for neutron filtering and for reduction of incident gamma radiation. The computational models used for the final beam design and performance evalua-tion are based on coupled discrete-ordinates and Monte Carlo techniques that permit detailed modeling of the neutron transmission properties of the filtering crystals with very few approximations. This is essential for detailed dosimetric studies required for the anticipated research program.

5:10 PMControl Rod Malfunction at the NRAD ReactorThomas L. Maddock and Ronald C. JohansenIdaho National Laboratory, Idaho Falls, ID USA

The neutron Radiography Reactor (NRAD) is a training, research, and isotope (TRI-GA) reactor located at the INL. The reactor is normally shut down by the insertion of three control rods that drop into the core when power is removed from electromagnets. During a routine shutdown, indicator lights on the console showed that one of the con-trol rods was not inserted. It was initially thought that the indicator lights were in error because of a limit switch that was out of adjustment. Through further testing, it was de-termined that the control rod did not drop when the scram switch was initially pressed. The control rod anomaly led to a six month shutdown of the reactor and an in depth investigation of the reactor protective system. The investigation looked into: scram switch operation, console modifications, and control rod drive mechanisms. A number of latent issues were discovered and corrected during the investigation. The cause of the control rod malfunction was found to be a buildup of corrosion in the control rod drive mechanism. The investigation resulted in modifications to equipment, changes to both operation and maintenance procedures, and additional training. No reoccurrenc-es of the problem have been observed since corrective actions were implemented.

5:35 PMThe AFRRI Triga Reactor: A Summary of Applications in Mouse StudiesG. David Ledney and T.B. ElliottCombined Injury Project, Scientific Research Department, Armed Forces Radiobiology Research Institute, Bethesda, MD USA

The AFRRI TRIGA reactor was used to simulate nuclear weapon mixed-field radiation injuries with and without additional tissue trauma. The severity of reactor-produced mixed-field radiations over that of γ-photon irradiation was evaluated in mice. Lethal doses (LDs) to 50% of groups of mice were determined for marrow cell (LD50/30, the dose required to kill 50% of the subjects within 30 days) and intestinal cell (LD50/6, the dose required to kill 50% of the subjects within 6 days) injury. As neutron (n) propor-tions in the total (t) radiation dose (Dn/Dt) increased LD values decreased. Relative bi-ological effectiveness (RBE) values for reactor-generated Dn/Dt used 60Co γ photons and 250-kVp x-rays as reference standards. RBEs for irradiated mice increased as Dn/Dt increased and was further increased by wound trauma. Compared to γ-photon irra-diation, mixed-field irradiation delayed wound closure times 25% to 50%. WR-151327 (200 mg/kg), a radioprotective chemical, injected i.p. into mice prior to either radiation quality alone or into combined injured mice increased 30-day survival and reduced susceptibility to challenge with Klebsiella pneumoniae. Protection against irradiation and resistance to bacterial challenge afforded by the WR compound was greater for γ-photon irradiation than for mixed-field irradiation. The TRIGA reactor can be used to simulate nuclear radiation-induced situations that include traumas or infections. Coun-termeasures for increasing survival after mixed-field irradiation may be more difficult than for γ-photon irradiated casualties.


9B - Special Session on the Test, Research & Training Reactor (TRTR) 2009 meetingSession Chairs: David Nigg (INL), Frederik Reitsma (PBMR)

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3:30 PMStartup Reactivity Accountability Attributed to Isotopic Transmutations in Irradiated BerylliumDavid Chandler and G. Ivan Maldonado(1), R. T. Primm, III(2)1) Nuclear Engineering Department, The University of Tennessee, Knoxville, TN USA, 2) Oak Ridge National Laboratory, Research Reactors Division, Oak Ridge, TN USA

The objective of this study is to develop a methodology to predict the reactivity impact as a function of outage time between cycles of 3He, 6Li, and other poisons in the High Flux Isotope Reactor’s (HFIR) beryllium reflector. The reactivity worth at startup of the HFIR has been incorrectly predicted in the past after the reactor has been shut-down for long periods of time. The incorrect prediction was postulated to be due to the erroneous calculation of 3He buildup in the beryllium reflector. It is necessary to de-velop a better estimate of the start-of-cycle symmetric critical control element positions since if the estimated and actual symmetrical critical control element positions differ by more than $1.55 in reactivity (approximately one-half inch in control element startup position), HFIR is to be shutdown and a technical evaluation is performed to resolve the discrepancy prior to restart. 3He is generated and depleted during operation, but during an outage, the depletion of 3He ceases because it is a stable isotope. 3He is born from the radioactive decay of tritium, and thus the concentration of 3He increases during shutdown. The computer program SCALE, specifically the TRITON and CSAS5 control modules including the KENO V.A, COUPLE, and ORIGEN functional modules were utilized in this study. An equation relating the down time (td) to the change in symmetric control element position was generated and validated against measure-ments for approximately 40 HFIR operating cycles. The newly-derived correlation was shown to improve accuracy of predictions for long periods of down time.

3:55 PMRefined Control Rod Worth Estimation by Kinetics Equa-tions With Included PhotoneutronsSilva Kalcheva and Edgar KoonenSCK•CEN – BR2 Reactor, Mol, Belgium

In this paper we show that in the BR2 reactor, which is reflected and moderated by a beryllium matrix, the reactivity worth of the control rods is strongly influenced by the delayed photoneutrons, emitted after ),(nγ– reactions of photons with beryllium. To analyze this reactivity worth, important to evaluate the correct value of the shut down margin, refined estimations of the control rod worth are performed using the reactor kinetics equations including delayed photoneutron groups. The delayed photoneu-tron parameters were experimentally determined in the BR02 mock-up reactor. The reactivity worth of the control rods are derived from nuclear measurements, which have been performed during several BR2 shutdowns in 2009. The experimental pro-gram included 30 rod-drop tests and asymptotic reactor period measurements. The importance of the photoneutrons for the reactivity worth determination depends on the chosen analysis method. The reactivity worth from the rod-drop tests is estimated using an approximate equation for the neutron density decay, the validity of which was verified by comparison with the numerical solutions, obtained by the transient code PARET/ANL V7.5. The contribution of the photoneutrons into the reactivity worth, ana-lyzed by this method for the BR2 reactor conditions is 2%. Neglecting the photoneu-trons contribution in the inhour equation leads to underestimation of the rod worth by 18%. Comparing the different experimental methods we conclude that the rod-drop test gives more reliable results for the absolute values of the rod worth vs. the reac-tor period measurement. Finally, a comparison with reactivity worths, calculated by MCNPX are presented.

4:20 PMDose Rate Assessment for the HFR RepairS.C. van der Marck, J. Oppe(1), J. Valkó(2)1) NRG, Petten, Netherlands. 2) Ikesol Consulting, Delft, the Netherlands

During 2010 part of the piping of the primary cooling system of the High Flux Reactor in Petten, the Netherlands, will be repaired. The materials used in this repair need to be radiation resistant. The dose that these materials will receive throughout the remaining life time of the reactor should therefore be quantified reliably. In this paper we report on shielding calculations from the reactor core to the repair location two meters away from the core. The calculations included both neutron and photon flux, attenuated over more than five orders of magnitude. The uncertainty in the dose at the repair location was analyzed. Finally, dose rate measurements were performed at locations close to the repair locations, in order to validate the calculations. The results of the calculations were always conservative with respect to the measurement results, but not more than a factor two, even though the calculations needed to cover more than five orders of magnitude.

4:45 PMStudy of RHF Conversion in LEU Fuel Neutronic Calcu-lations With MCNPX and VESTAA. Bergeron(1), Y.Calzavara(2), W.Haeck, B.Cochet(3)1) Argonne National Laboratory, Argonne Illinois USA. 2) Institut Laue-Langevin, Grenoble, France. 3) Institut de Radioprotection et de Sûreté Nucléaire, Fontenay aux Roses cedex, France

We carried out MCNPX and VESTA calculations in order to estimate the impact of the ILL reactor (the RHF) conversion in a LEU fuel on its performances. To do so, we evaluated the cycle length, neutron flux and heat flux variations of several configura-tions. For the switched LEU configuration, the performances are too much reduced to be acceptable. Nevertheless some promising ways still exist but would need to modify the whole RHF fuel.

5:10 PMStudies on the Transition to the Equilibrium Core of the Jules Horowitz ReactorC. D’Aletto, C. Döderlein, P. SirétaCEA/Cadarache, DEN/SPRC/LPN F-13108 Saint Paul lez Durance, France

The Jules Horowitz Reactor (JHR) is the French future European Material Testing Reactor (MTR). Its innovative design, based on an irregular assembly arrangement, led to the development of a new version of its calculation scheme HORUS3D/N. The use of HORUS3D/N has made it possible for the first time to demonstrate the possibil-ity to reach the equilibrium of the JHR core, starting with fresh fuel. In a first step, an intermediate equilibrium is reached by using U3Si2 with a mass enrichment in 235U of 19.75%. Five fixed hafnium rods, positioned in the rack between the fuel assemblies, are necessary to compensate the reactivity excess of the first cycle. The equilibrium is reached after about 8 cycles, using 56 assemblies to achieve it. The mean cycle length is up to 26 days. The core is reloaded per quarter and operated at a power of 70 MW. The burnup of the unloaded assemblies is below the maximal value for which the U3Si2-Al fuel is qualified. The safety criteria for the power factors (assembly and plate) and control rod efficiencies (for the 3 classes of rods: compensation, control and safety rods) are verified at the beginning of cycle, at the Xenon equilibrium, at the half-time cycle and at the end of cycle. In a second step, from the first equilibrium, the core may be reloaded with fresh fuel (27%) per quarter and the definitive equilibrium is reached after 4 cycles.


9C - Operational Support AnalysesSession Chairs: Jim Kuijper (NRG), Alireza Haghighat (UFL)

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3:30 PMPreliminary Analyses for Accelerator-Driven System with High-Energy Protons in Kyoto University Critical AssemblyJae-Yong Lim, Cheol Ho Pyeon, Tsuyoshi Misawa and Seiji ShiroyaResearch Reactor Institute, Kyoto University, Osaka, Japan

At the Kyoto University Research Reactor Institute, the first injection of spallation neutrons generated by the high-energy proton beams into a reactor core was ac-complished on 4th March, 2009. Using 3 detectors which located at near active core regions, the prompt and delayed neutron behaviors by proton injection are experi-mentally observed and the neutron beam characteristics at the beam duct are also watched by Gafchromic films. Under the subcritical condition with 0.76 %Δk/k, an In wire irradiation experiment is accomplished horizontally. The 115In(n, γ)116mIn reac-tion rate comparison is also performed by MCNPX simulation and its errors shows within the allowance of the experimental statistical errors. By numerical analysis, the feasibility of neutron shield and beam duct is verified and the performance change inside of critical assembly is investigated. Finally, it is confirmed that the effect of dif-ferent injected proton energy is not intensified because of well-thermalized KUCA core condition by sufficient polyethylene moderators and reflectors.

3:55 PMReactivity Determination Techniques in ADS Systems for the Incineration of Radioactive WastesM.Fernández-Ordóñez, D.Villamarín, V.B´ecares, E.González-Romero(1), C.Berglof(2)1) CIEMAT, Madrid Spain. 2) Department of Reactor Physics, KTH, Sweden

The Subcritical Accelerator Driven Systems (ADS) have been proposed as one of the strategies for reducing the inventory and radiotoxicity of the spent nuclear fuel. The reactivity monitoring of ADS has been found to play a key role in the development of this technology. Although the current-to-power technique has been proposed as the main reactivity monitoring technique, it is necessary to periodically calibrate it using additional techniques. In this work, the standard reactivity monitoring techniques using Pulsed Neutron Source experiments have been evaluated in the coupled fast-thermal subcritical assembly YALINA-Booster. In addition, a new method to measure the ab-solute value of the system reactivity is proposed by using micro-interruptions of the continuous external neutron source (beam-trips). This technique, used for the first time in a subcritical core, provided results compatible with the standard PNS methods and can be used in future power ADS. Even more, to allow instantaneous reactiv-ity checks, we have developed and tested the necessary electronic chains and data acquisition system to determine the system reactivity during a single beam trip, thus allowing the determination of the reactivity within a second.

4:20 PMSet-Up of a Deterministic Model for the Analysis of the Guinevere ExperienceGiancarlo Bianchini, Mario Carta, Fabrizio Pisacane(1), Manuela Frisoni, Vincenzo Peluso(2)1) ENEA,CR, Rome Italy. 2) ENEA Centro, Bologna Italy

The experimental European project GUINEVERE was launched in the frame of IP EUROTRANS/6th EU Framework Programme. The experience, mainly devoted to the issues concerning on-line reactivity monitoring in ADS, will be performed by using a modified lay-out of the VENUS critical facility located at the Belgium SCK•CEN Mol-site, coupling a subcritical core facility to a deuteron accelerator delivering, by a con-tinuous beam, 14 MeV neutrons by deuterium-tritium reactions. This report describes how a deterministic model (rectangular XYZ geometry) of the modified VENUS facility has been implemented by the ERANOS French code. A comparative analysis was carried out, partially by perturbation theory techniques, about JEFF 3.1 library lead cross sections used by MCNP and ECCO/ERANOS. Comparison between ERANOS and MCNP results are very encouraging, both for keff and some flux/reaction rates profiles.


9D - Accelerator and Spallation PhysicsSession Chairs: Pierre D’hondt (SCKCEN), Frederik Reitsma (PBMR)

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10:20 AMKAMINI Fast Neutron Beam Attenuation Measurements in Boron Carbide Shield and their Prediction by Monte Carlo ModellingD. Venkata Subramanian, E. Radha, Adish Haridas, M.M. Shanthi, C.P. Reddy and R.S. KeshavamurthyIndira Gandhi Centre for Atomic Research, Kalpakkam, India

Shields around core and blankets form major part of reactor assembly in fast reactors. The spectrum of neutrons leaking from the blankets is hard with negligible thermal component and has anisotropic angular distribution. Boron carbide and stainless steel have been the main choice shield materials for shields in fast reactors. Kalpakkam MINI (KAMINI) reactor beam was used to measure attenuation of the threshold in-elastic (n,n’) reaction rates on Cd-111, In-115 and Rh-103 in Boron carbide. Shield. KAMINI reactor including beam hole and shield was modeled and Monte Carlo calcu-lations carried out using the code (MCNP). The details of experimental measurements and MCNP calculations are presented. Comparison of MCNP predicted attenuation behaviour with measured attenuation pattern is made. The values are in fairly good agreement It is also observed that the predicted values are lower in all the cases.

10:45 AMDelayed Neutron Measurements of Induced Fission Rates in Burnt LWR Fuel Samples at the PROTEUS Ze-ro-Power Reactor FacilityK. A. Jordan, G. Perret, and M. F. MurphyPaul Scherrer Institut, Villigen-PSI, Switzerland

The LIFE@PROTEUS program at the Paul Scherrer Institut is being undertaken to characterize the interfaces between burnt and fresh fuel assemblies in modern LWRs. Techniques are being developed to measure fission rates in burnt fuel, following re-irradiation in the zero-power PROTEUS research reactor. In the presented approach, the fission rates are estimated by measuring delayed neutrons emitted by re-irradiated fuel. To demonstrate the feasibility of this technique, fresh and burnt fuel samples (with burnup varying from 36 to 64 GWd/MTU) were irradiated in the PROTEUS reactor, and their neutron outputs were recorded shortly after irradiation. Relative fission rates between different core lattice positions were derived for a fresh sample as well as for the three burnt samples. The measured fission rate ratios have 1-σ uncertainties between 2% and 3.5%, with the larger uncertainties corresponding to the more highly burnt fuel. Results obtained by Monte Carlo simulations agree with the experimentally determined values within these limits. With further development of the technique, the experimental uncertainties can be further reduced. Continuing effort is being directed towards accurate comparison of fission rates between fuel samples of different burn-up.

11:10 AMFission Rate Measurements in Spent Fuel via Gamma-Ray Spectrometry of Short-Lived Fission Products In-duced in a Zero Power ReactorHanna Kröhnert, Gregory Perret and Michael F. Murphy(1), Rakesh Chaw-la(2)1) Paul Scherrer Institut (PSI), Villigen, Switzerland. 2)Laboratory of Reactor Physics and Systems Behaviour, Ecole Polytechnique Fédérale de Lausanne (EPFL), Lau-sanne, Switzerland

A new measurement technique is being developed to determine fission rates in fresh and spent power reactor fuel following irradiation in a zero-power research reactor. The technique is required for the future experimental program LIFE@PROTEUS, one goal of the program being the investigation of power profiles across fresh and burnt fuel interfaces typical of a newly reloaded power reactor. In order to discriminate against the intrinsic activity of spent fuel, the approach described here uses high-energy γ-rays (above 2200 keV) emitted by freshly produced short-lived fission products. To dem-onstrate the feasibility of such a technique, fresh and spent UO2 fuel samples with nominal burn-ups of 0, 36, 46 and 64 GWd/t were irradiated in the PROTEUS reactor and their γ-ray activities were recorded directly after the irradiations. For the first time, following irradiation in a zero-power research reactor, it was possible to compare the freshly induced short-lived γ-ray activity from spent fuel samples having high intrinsic γ-ray backgrounds with corresponding activities induced in fresh fuel. In this paper, first results of derived fission rate ratios between a fresh and a 36 GWd/t spent sample based on four high-energy peaks (142La (2542 keV), 89Rb (2570 keV), 138Cs (2640 keV) and 95Y (3576 keV)) are presented. The measured fission rate ratios from the various fission products agree within 1-2 standard deviations, the 1σ uncertainties be-ing ~2.5 - 4.5%. At the current state of analysis, calculated and measured fission rate ratios agree within 1-2σ, but a bias of about 4% could be observed.

11:35 AMEvaluation of Gold Activation Detectors on the Epither-mal Neutron Beam at the LVR–15 ReactorZdena Lahodová, Ladislav Viererbl(1), Vít Klupák, Alexander Voljanskij(2)1) Research Center Řež Ltd., Řež, Czech Republic. 2) Nuclear Research Institute Řež plc, Řež, Czech Republic

The increasing demands on irradiation conditions require more precise knowledge of the neutron spectrum. The spectrum on the horizontal channel with epithermal neutron beam at the LVR-15 light water research reactor is measured by activation detectors (Au, In, Sc, W, La, Mn, Cu, and Ni). Especially gold is often used as a reference detec-tor thanks to its suitable physical and chemical characteristic. Exploration of two types of the Au detectors (the pure Au (100 %) detector and the Au (1 %) detector (Al alloy)) is described in this article. Three experiments on the epithermal beam were realized, one with the bare Au detectors, one with the cadmium covered Au detectors and one with a polyamide (PA) filter (moderator). Cadmium ratios were determined for the Au (100 %) and the Au (1 %) detectors. Although a method of activation detectors has been used for a long time, comparative measurements for each irradiation position are considered substantial requirement for achieving high accuracy of determination of neutron spectrum. It is also proposed to calculate the self-shielding factors and the cadmium ratios by the Monte Carlo neutron transport code (MCNP) and to compare the calculated results with the experimental measurements.


9E - Physics Tests, Measurements and EvaluationSession Chairs: Hans Gougar (INL), Jim Kuijper (NRG)

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1:30 PMBenchmark Evaluation of the Initial Isothermal Physics Measurements at the Fast Flux Test Facility John Darrell Bess Idaho National Laboratory, Idaho Falls, ID USA

The benchmark evaluation of the initial isothermal physics tests performed at the Fast Flux Test Facility, in support of Fuel Cycle Research and Development and Genera-tion-IV activities at the Idaho National Laboratory, has been completed. The evaluation was performed using MCNP5 with ENDF/B-VII.0 nuclear data libraries and according to guidelines provided for inclusion in the International Reactor Physics Experiment Evaluation Project Handbook. Results provided include evaluation of the initial fully-loaded core critical, two neutron spectra measurements near the axial core center, 32 reactivity effects measurements (21 control rod worths, two control rod bank worths, six differential control rod worths, two shutdown margins, and one excess reactivity), isothermal temperature coefficient, and low-energy electron and gamma spectra mea-surements at the core center. All measurements were performed at 400 ºF. There was good agreement between the calculated and benchmark values for the fully-loaded core critical eigenvalue, reactivity effects measurements, and isothermal temperature coefficient. General agreement between benchmark experiment measurements and calculated spectra for neutrons and low-energy gammas at the core midplane exists, but calculations of the neutron spectra below the core and the low-energy gamma spectra at core midplane did not agree well. Homogenization of core components may have had a significant impact upon computational assessment of these effects. Future work includes development of a fully-heterogeneous model for comprehensive evalu-ation. The reactor physics measurement data can be used in nuclear data adjustment and validation of computational methods for advanced fuel cycle and nuclear reactor systems using Liquid Metal Fast Reactor technology.

1:55 PMAnalysis of Core Physics Experiment on BWR 10x10 MOX Fuel Assemblies Toru Yamamoto and Yoshihira Ando(1), Leim Peng Hong(2) 1) Japan Nuclear Energy Safety Organization, Tokyo, Japan. 2) NAIS Co. Inc., Iba-raki-Ken, Japan

JNES has performed MOX core physics experiments FUBILA using the EOLE LWR critical facility of the CEA Cadarache center in collaboration with a French consortium (CEA and COGEMA). The experiments aim to obtain core physics data of operat-ing conditions of full MOX BWR cores consisting of high Pu-enriched BWR MOX as-semblies. One of the experimental cores is a full MOX core containing 10x10 MOX fuel assemblies which have an assembly average total Pu enrichment of 10.6 wt%. Measurement core parameters are critical mass and core fission rate distributions. Theoretical analysis of the experimental data has been carried out using deterministic codes based on diffusion and transport calculations and a continuous energy Monte Carlo calculation code with major nuclear data libraries. The critical keff’s are 0.997 for the diffusion calculation and from 1.001 to 1.002 for the transport calculations with JENDL-3.3-base group constants. Those of the Monte Carlo calculations are 1.001 for JENDL-3.3, 0.999 for ENDF/B-VI.8, 1.004 for -VII and 1.001 for JEFF-3.1. The root mean squares (RMS’s) of differences between the calculated and measured core radial fission rates are 2.3% for the diffusion calculation, 1.5% for the transport calcu-lation and 1.2% for the Monte Carlo calculation.

2:20 PMExperimental Validation of Reaction Rate Distributions in an SCWR-Like Fuel Lattice at PROTEUSDominik Rätz, Kelly A. Jordan, Michael F. Murphy, and Gregory Perret(1), Rakesh Chawla(2)1) Paul Scherrer Institut, Villigen, Switzerland. 2) Paul Scherrer Institut, Villigen, Swit-zerland and École Polytechnique Fédéral de Lausanne, Switzerland

High resolution gamma-ray spectroscopy experiments were performed on 61 pins of an SCWRlike fuel lattice at the PROTEUS zero-power research reactor at the Paul Scherrer Institut in Switzerland. The derived reaction rates were compared to calcu-lated results from full-core Monte Carlo simulations with MCNPX. Reaction rates (cap-tures in 238U and total fissions in 235U and 238U) were measured and mapped pin-wise on the lattice. Ratios of calculated to experimental values (C/E’s) were assessed for both reaction rates. These C/E’s show excellent agreement between the calcula-tions and the measurements. In the case of neutron captures in 238U all experimental results agree within 2σ while for the total fissions only 5 (out of 61) values are outside the 2σ confidence interval. For the 238U capture rate, the 1σ level in the comparisons corresponds to an uncertainty of ±2.5%, while for the total fission rate the 1σ level is below ±1%. To determine the reproducibility of these results, the measurements were performed twice, once in 2006 and again in 2009. The agreement between these two measurement sets is better than 1%.

2:45 PMTRIGA Reaction Rate BenchmarkL. Snoj(1,2), A. Trkov(1), G. Žerovnik(1), R. Jaćimović(3), P. Rogan(1), M. Ravnik(1)1) Reactor physics division, “Jožef Stefan” Institute, Ljubljana, Slovenia. 2) Reactor Infrastructure centre, “Jožef Stefan” Institute, Ljubljana, Slovenia. 3) Environmental Sciences Division, “Jožef Stefan” Institute, Ljubljana, Slovenia

In order to validate the computational model of the “Jožef Stefan” Institute TRIGA re-search reactor, a series of neutron activation experiments in irradiation channels has been performed. Activation of gold-aluminium foils irradiated at 6 different locations in the reactor core and 27 locations in the carrousel irradiation facility in the reflector was measured. Thermal and fast neutron flux distributions were estimated via (n,γ) reaction rates on 197Au and (n,α) reaction rates on 27Al, respectively. In addition the experi-mental uncertainties were evaluated using the most upto- date nuclear data libraries (ENDF/B-VII.0) and advanced Monte Carlo code MCNP5. The computational model is based on TRIGA criticality benchmark model but enhanced by adding to the model several reactor components which affect the neutron flux distribution, at irradiation po-sitions. The calculations agree well with experiments, indicating that the material and geometrical properties of the reactor core are properly modelled. Due to relatively low experimental uncertainties, the experiment can be used as the benchmark for reaction rate distribution in TRIGA reactors and other UZrH fuelled systems.


10A - Integral Experiments and Facilities for Safety Research - ISession Chairs: Pierre D’hondt (SCKCEN), Frederik Reitsma (PBMR)

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10:20 AMAn Improved Benchmark Model for the BIG TEN Critical AssemblyRussell D. MostellerLos Alamos National Laboratory, Los Alamos, NM, USA

A new benchmark specification is developed for the BIG TEN uranium critical assem-bly. The assembly has a fast spectrum, and its core contains approximately 10 wt.% enriched uranium. Detailed specifications for the benchmark are provided, and results from the MCNP5 Monte Carlo code using a variety of nuclear-data libraries are given for this benchmark and two others.

10:45 AMA Solution for the Telegrapher’s Equation with External Source: Application to YALINA - SC3A and SC3BB. Merk, V. Glivici-CotruŃă, and F. P. WeißForschungszentrum Dresden-Rossendorf, Institut für Sicherheitsforschung, Dresden, Germany

This work represents the detailed comparison of the analytical solutions for the space and time Telegrapher’s equations with the experimental results, obtained for the YALI-NA-Booster subcritical facility in 2008. The derivation of analytical solution for the Tele-grapher’s equation with a special temporal shape of the external source is described. The Green’s function method was applied. Qualitative results of the obtained solutions and the experimental results are analyzed. The special configuration of the YALINA-Booster facility is discussed.

11:10 AMSubcriticality Measurement by the Imaginary Source Multiplication Method with Multi-DetectorsShinichi Kawaguchi(1), Tsuyoshi Misawa, Cheol Ho Pyeon, and Seiji Shi-roya(2) 1) Department of Fundamental Energy Science, Graduate School of Energy Science, Kyoto University, Kyoto, Japan. 2) Nuclear Engineering Science Division, Research Reactor Institute, Kyoto University, Osaka, Japan

For the steady subcritical reactor, the subcriticality change by a perturbation can be estimated by the imaginary source multiplication method, to the extent that the pertur-bation is qualitatively known beforehand. For this method, the numerical calculations are executed only for ‘unperturbed’ systems, and not for the ‘perturbed’ system, whose macroscopic cross sections are usually unknown. The most important property of this method is that the assumption of completeness of the eigenfunctions is excluded from the derivation in order to maintain the theoretical consistency. By this method, the sub-criticality in the perturbed system is obtained from the combination of the numerical calculation and the experiment, even when the first-order approximation is applied to a large perturbation. The basic assumption employed in this method is that the perturba-tion is represented as a superposition of known operators. For the practical reactor, subcriticality can be estimated by numerical methods, except for some discrepancy between the calculation and the true value. Regarded as a perturbation, this discrep-ancy is assumed to be represented as a superposition of some known operators, in order to apply the imaginary source multiplication method. Although this assumption is seldom satisfied, the loss of accuracy is expected to be relaxed when sufficiently many neutron detectors are employed. Hence, the subcriticality in the perturbed system can be estimated accurately by the imaginary source multiplication method even when the position of the perturbation is not beforehand given. The effectiveness of using multiple detectors is numerically confirmed in this paper.

11:35 AMBenchmark-Experiments with HEU and Tungsten at FKBN-2 Facility performed in RFNC-VNIITFAndrey V. Dubina, Valeriy D. Lyutov, Andrey V. Serikov, Yury A. Sokolov, Vladimir M. ShmakovRFNC -VNIITF, Snezhinsk, Chelyabinsk Region, Russia

There is given a brief description of FKBN-2 facility presented along with the results of the criticality experiments with HEU and Tungsten performed at the RFNC-VNIITF. Be, BeO, CH2 have been used as reflectors and moderators to produce the variety of neutron spectra. Multiplication systems under study were cylindrically formed and composed of HEU disks and the materials mentioned hereinabove. HEU ROMB disks have been used as fissionable materials. Results of the experiments have been used for Tungsten neutron data’s verification. The experiments carried out in the framework of the International Criticality Safety Benchmarks Evaluations Project (ICSBEP).


10B - Integral Experiments and Facilities for Safety Research - IISession Chairs: Toshikazu Takeda (University of Fukui), Frederik Reitsma (PBMR)

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1:30 PMSpectral Indices of the IPEN/MB-01 Reactor: A Nuclear Data Validation Adimir dos Santos and Leda C. C. B. Fanaro Instituto de Pesquisas Energéticas e Nucleares, IPEN/CNEN-SP, São Paulo, Brasil

New experimental results for the spectral indices of the IPEN/MB-01 reactor are pre-sented in this work. The experimental approach considers a novice technique for the 28r case which does not require any sort of calculated correction factors. This aspect gave to the IPEN/MB-01 experiment an excellent quality and free of possible bias due to these calculated correction factors. The uncertainty analysis show, even consider-ing the uncertainties of the geometric and material data of the facility, that the final uncertainties are small enough and well understood for a benchmark problem. The theory/experiment comparison reveals that there is considerable progress in the 238U nuclear data for application in thermal reactors. The theory/experiment comparison reveals an excellent agreement for the spectral index 28r*; aspect never found be-fore in several other comparisons. The long term overprediction of the 238U neutron epithermal capture appears to be more related to the deficiencies in the previous experiments generally used to check the 238U data than to a nuclear data problem. The experimental performed at the IPEN/MB-01 reactor supports the changes in the 238U nuclear data incorporated in ENDF/B-VII.0 and in the other libraries studied in this work. The theory/experiment comparison of 25d* and (C8/F)ept. show that these spectral indices are in general slightly overpredicted, thus suggesting that the thermal fission cross section of 235U might be a little bit underestimated. The overall analysis of the theory/experiment comparisons show the excellent applicability of ENDF/B-VII.0 for thermal reactors fuelled with slightly enriched uranium.

1:55 PMPower Distribution Gradients in WWER-440 Standard and Low Leakage Type Cores with Possible Influence on Fuel Failure Ján M. Mikuš Research Centre Řež Ltd., Czech Republic

As known neutron flux non-uniformity and gradients of neutron current, resulting in corresponding power (fission rate) distribution changes can represent root causes of the fuel failure in the WWER-440 reactor cores. Such situation can be expected in vicinity of some core heterogeneities and construction materials. Since needed data cannot be obtained from nuclear power plants, results of some benchmark type ex-periments performed on light water, zero-power research reactor LR-0 were used for investigation of the above phenomenon. Attention was focused on determination of the: 1. Power distribution changes on the periphery of the WWER-440 standard and low leakage type cores, in fuel assemblies neighbouring the core blanket and dummy steel assembly simulators. For this purpose two WWER-440 type mock-ups were as-sembled on reactor LR-0 and the WWER-440 geometry sector ~ 60° from the core periphery area to the biological shielding modelled. The core, core blanket and basket were located inside LR-0 tank, and the reactor pressure vessel and biological shield-ing ones outside tank. 2. Power distribution changes in fuel assemblies neighbouring control rod absorbing part in WWER-440 type cores at two boron acid concentrations in moderator. For this purpose two WWER-440 type cores containing 19 fuel assem-blies with “authentic” models of the real WWER-440 control rod were assembled.Power distribution values in selected fuel pin positions were determined using gamma scanning of irradiated fuel pins measuring their (fission products) gamma radiation. Obtained experimental results can be used for code validation and subsequently for fuel pin failure occurrence investigation.

2:20 PMStatus of ERANOS-2 Code System Validation for Sodi-um Fast Reactor ApplicationsJ. Tommasi, J.F. Lebrat and P. Archier(1), J.M. Ruggieri(2)CEA, DEN, DER, Saint Paul lez Durance, France. 2) CEA, DEN, DTN, Saint Paul lez Durance, France

The analysis of several types of experiments with ERANOS-2.2 associated to the JEFF3.1.1 library is underway at CEA. This paper presents the current status of the analysis, except for sodium void, dealt with in a companion paper [1]. It shows the good predictability of these calculation tools so far for criticality calculations (C-E = 300 ± 250 pcm), beff (C/E = 1.03 ± 0.02), Doppler feedback (C/E = 1.00 ± 0.06), CR worth (C/E = 1.00 ± 0.02), capture rate prediction (depends on specific nuclide). From this validation work, some required improvements on nuclear data are highlighted, as well as the need for new specific integral experiments.

2:45 PMSodium Void Validation with ERANOS on Zero Power Facility ExperimentsJ. Tommasi, P. Archier(1), J.M. Ruggieri(2)1) CEA, DEN, DER, Saint Paul lez Durance, France. 2) CEA, DEN, DTN, Saint Paul lez Durance, France

Near sixty Na void experiments performed in the zero power reactors MASURCA (CEACadarache) and ZPPR (Argonne West – Idaho) have been analyzed using JEFF-3.1 nuclear data and the ERANOS-2.1 (deterministic) and TRIPOLI-4 (Monte-Carlo) codes. They have been selected to cover spectral conditions ranging from the harder flux in the outer zone of a small fast reactor to the softer flux in the inner zone of a large fast reactor. For in-fuel Na void patterns, there is a good agreement between ERANOS and TRIPOLI computations, while the deterministic calculations significantly underes-timate the leakage component for Na void patterns in fertile regions. The agreement between ERANOS-2.1+JEFF-3.1 predictions and experimental values is excellent for in-fuel Na void patterns in MASURCA (C/E » 0.98 ± 0.02), but a significant underes-timation of the leakage component occurs for in-fuel Na void patterns in ZPPR. For fertile Na void patterns, calculation underestimates the leakage component.


10C - Integral Experiments and Facilities for Safety Research - IIISession Chairs: David Nigg (INL), Jim Kuijper (NRG)

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10:20 AMBenchmarking TransLAT for Fission Product Genera-tionDavid J. Sweeney and William S. CharltonNuclear Security Science & Policy Institute, Texas A&M University, College Station, TX USA

The TransLAT lattice physics code was investigated to characterize its ability to sim-ulate fission product generation in a nuclear reactor. Pin cell models for fuel from H.B. Robinson Unit 2, Calvert Cliffs No. 1, and Mihama Unit 3 were developed using TransLAT. These models were used to simulate reactor operation corresponding to measured samples with isotopic data reported in available literature for the three reac-tors. For this study, TransLAT was run using the Method of Characteristics. The results of the TransLAT simulations were then compared to the measured data. The results show that the TransLAT could predict the absolute concentrations of most fission prod-ucts to within ±20% of the measured values and it could predict the ratios of fission products to within ±10% of the measured values. Thus TransLAT may be expected to produce reasonable estimates of fission product concentrations generated during reactor operation even using simple pin cell models.

10:45 AMDevelopment of a “Best Representativity” Method for Measurement Data Analysis – Application to Critical Ex-periment Simulating PWR at the Toshiba NCA FacilitiyTakuya Umano, Kenichi Yoshioka, Mitsuaki Yamaoka(1), Satoshi Suga-hara(2), Mohamed Ouisloumen(3)1) Toshiba Corporation, Power Systems Company, Kawasaki-shi, Japan. 2) Toshiba Corporation, Power Systems Company, Yokohama-shi, Japan. 3) Westinghouse Elec-tric Company, Pittsburgh, PA, USA

For the qualification and the validation of the nuclear calculation codes, the precise measurement data from the nuclear critical experiments are both necessary and prof-itable. At Toshiba, we have a tank type light water moderated nuclear critical facility named Toshiba Nuclear Critical Assembly (NCA). After the recent realization of the collaboration between Toshiba and Westinghouse, Toshiba NCA is becoming the facil-ity of choice utilized both for BWR and PWR experiments and research. In a PWR core, the boric acid (H3BO3) is employed to control the core reactivity. At the operating conditions, the light water moderator in the core is under high pressure (15.8 MPa) and at high temperature (~580 K), thereby the water density is about 0.7 g/cm3 which is much smaller than the density of room conditions. In order to closely simulate these specific operating PWR situations, polystyrene blocks containing the boron carbide (B4C) material have been developed. In addition to these preparations for the experi-ments, when we judge the value/worth of the experimental data, it is quite important to evaluate the experiment similarities to the objective of the actual reactor conditions or actual reactor equipment. Recently a concept of “the representativity factor” is often used in the nuclear critical experiment field. In accordance with this concept, a new numerical evaluation method and a calculation system have been developed to under-stand the representativity of NCA experiments to the target reactor conditions. In this paper this mathematical method and simplified numerical results are presented.

11:10 AMSCALE 6 Analysis of HTR-10 Pebble-Bed Reactor for Ini-tial Critical ConfigurationEva E. Sunny(1), Germina Ilas(2) Department of Nuclear Engineering and Radiological Sciences, University of Michi-gan, Ann Arbor, MI, USA. 2) Oak Ridge National Laboratory, Oak Ridge, TN, USA

HTR-10 is a high temperature pebble-bed reactor located in China, operated at 10 MW thermal power. The purpose of this study is to create an accurate model of HTR-10 for its initial critical configuration using the ORNL SCALE 6 code system to subsequently validate the methods used in SCALE 6 for treatment of doubly-heterogeneous fuel and the associated data libraries. KENO VI, a three-dimensional Monte Carlo transport code within SCALE 6, is used to create the computational model for HTR-10 based on the benchmark specifications provided in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEHandbook). The results from KENO VI are compared to results obtained with a consistent MCNP model of the same con-figuration, as provided in the IRPhE-Handbook. The comparison shows a difference in keff of 73 ± 34 pcm between MCNP and SCALE 6

11:35 AMQualification of the French APOLLO2.8/CEA2005V4 Code Package on Absorber Clusters in 17X17 PWR Type Lattices through the CAMELEON ProgramPatrick Blaise, Olivier Litaize, Jean-François Vidal(1), Alain Santamari-na(2)1) Commissariat à l’Energie Atomique, Centre de Cadarache, DEN-CAD/DER/SPRC/LEPh, Saint Paul-Lez-Durance, FRANCE. 2) Commissariat à l’Energie Atomique – Centre de Cadarache, DEN-CAD/DER/SPRC, Saint Paul-Lez-Durance, France

This paper details the experimental validation of B4C, Ag-In-Cd (AIC) and Hf-absorber clusters reactivity worths in 17x17 PWR lattices. This validation is performed through several configurations of the CAMELEON experimental program, using the newly qualified APOLLO2.8/CEA2005v4 French code package. After a general presentation of the CAMELEON program that took place in the EOLE critical Facility in Cadarache, one describes the new APOLLO2.8/CEA2005v4 code package relying on the deter-ministic transport code APOLLO2.8 with the method of characteristics (MOC), and its new CEA2005V4 library based on the latest JEFF-3.1.1 nuclear data evaluation, used in both reference 281group SHEM-MOC scheme and REL2005 optimized 26groups scheme. Comparisons with the reference TRIPOLI-4 Monte Carlo code are made for reactivity and absorber worth. For critical masses, the average Calculation-to-Experi-ment (C-E) discrepancies on the keff of unclustered cores are (240 ± 70) pcm and (370 ± 130) pcm for the reference 281 group MOC and optimized 26 group MOC schemes respectively. For pin-by-pin radial power distributions, the REL2005 results are close to the experimental values, with maximum (C-E)/E discrepancies of the order of 3%, i.e., in the experimental uncertainty limits. Absorber worths are overestimated by 1.2% to 3.6% with deterministic codes, if proper 3D correction, taking into account axial buckling variation, is applied on 2D keff. These results are consistent with the Monte Carlo analysis. The calculations demonstrate the good knowledge of B4C and AIC cross sections, and an overestimation of the natural Hf capture cross section by about 3% in the JEFF3.1.1 evaluated nuclear data library.


11A - Verification and Validation - ISession Chairs: Blair Bromley (AECL), Takanori Kitada (Osaka Univ.)

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1:30 PMMCNPX Burnup Simulation of the Isotope Correlation Experiment Yan Cao and Yousry Gohar(1), Cornelis H. M. Broeders(2) 1) Nuclear Engineering Division, Argonne, IL USA. 2) Forschungszentrum Karlsruhe, Institute for Neutron Physics and Reactor Technology, Karlsruhe, Germany

This paper presents the simulation work of the Isotope Correlation Experiment (ICE) using the MCNPX computer code package including CINDER90. The numerical simu-lation is based on a pin cell model for the fuel lattice. The temperature dependent JEFF3.1 library is utilized as the main library for the simulation. The comparison of the results from the Monte Carlo simulation and the experiment is up to a burnup of about 30 GWD/t. The MCNPX simulation results show an excellent agreement with the experimental values for predicting the depletion of the uranium fuel and the buildup of the plutonium isotopes in a PWR-type thermal reactor.

1:55 PMConfirmation of the Similarity between Critical Experi-ments and Real Cores by Using Representativity Factor in LWR Cell Takanori Kitada and Yousuke Inoue Osaka University, , Osaka, Japan

To be clear the influence of the different geometry between cylindrical and plate sys-tems for LWR cell, the new code was developed and applied it to check the simi-larity between two systems by evaluating representativity factor and sensitivity coef-ficients based on integral transport theory. Erbia worth was selected as the target neutronic property which has been measured at Kyoto University Critical Assembly, and checked the variation of the representativity factor by changing the unit-cell com-posed of the experimental core. It was found that the similarity can be improved by considering the H/U value, enrichment of 235U and the sample enrichment used in the critical experiment, and the degree of the improvement can be evaluated by using the representativity factor.

2:20 PMResults of the IAEA CRP5 - Benchmark Analysis Eelated to the PBMR-400, PBMM, GT-MHR, HTR-10 and the AS-TRA Critical FacilityBismark Tyobeka(1), Frederik Reitsma(2)1) International Atomic Energy Agency, Vienna, Austria. 2) Pebble Bed Modular Reac-tor (Pty) Ltd., Centurion, South Africa

The IAEA has facilitated an extensive programme that addresses the technical devel-opment of advanced gas cooled reactor technology. Included in this programme is the Coordinated Research Project on Evaluation of High Temperature Gas Cooled Reac-tor (HTGR) performance, which is the focus of this paper. This Coordinated Research Project (CRP) was established to foster the sharing of research and associated techni-cal information between participating Member States in the on-going development of the HTGR as a future source of nuclear energy. Within the CRP, computer codes and models are verified through actual test results from operating reactor facilities. A code-code validation is also performed as part of the CRP in cases where experimental data does not exist. The paper summarizes the outcomes of this CRP with the focus on the analysis of benchmark problems pertaining to the HTR-10 test reactor operation and safety tests and the proposed Pebble Bed Modular Reactor 400 MWth annular core design.

2:45 PMCoupling of WIMS-AECL and ORIGEN-S for Depletion CalculationsGeoffrey W.R. EdwardsAtomic Energy of Canada Limited (AECL), Chalk River Laboratories, Chalk River, On-tario, Canada

One of the more powerful tools for isotope depletion calculations in neutron-irradiated material is the SCALE (Standardized Computer Analyses for Licensing Evaluation) module ORIGEN-S, maintained and developed by Oak Ridge National Laboratory. ORIGEN-S takes as input, in addition to a material description, a problem-dependent cross section library in which relative reaction rates for each nuclear process have been pre-evaluated. Creating different libraries for different stages of burnup, and for different materials, allows the ‘point’ code phenomenology of ORIGEN-S to be ex-tended to more complicated geometries. To this end, AECL (Atomic Energy of Canada Limited) has coupled its successful 2-D neutron transport solver WIMS-AECL 2.5d to ORIGEN-S to create the coupled code ‘WOBI’ (WIMS-ORIGEN Burnup Integration). This code has been validated against PIE (post irradiation examination) results for CANDUTM reactors and for light-water reactors, and is extensively used at AECL to calculate exit compositions and decay heats for high and low enriched uranium fuels at the NRU (National Research Universal) research reactor located at the Chalk River Laboratories. In addition, because of the significantly expanded list of reactions avail-able in ORIGEN-S, WOBI is more useful for advanced fuel cycle studies than WIMS-AECL alone. This paper discusses the validation results, and verification of WOBI against simple WIMS-AECL and ORIGEN-S stand-alone models.


11B - Verification and Validation - IISession Chairs: Walid Metwally (GE), Patrick Blaise (CEA)

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3:30 PMHigh Order Finite Difference Approximations to the One-Group Neutron Diffusion Equation in 1D Heterogeneous Media Part I: Theory in Plane MediaBarry D. Ganapol(1), David W. Nigg(2)1) University of Arizona, Tucson, AZ. 2) Idaho National Laboratory, Idaho Falls, ID

Verification that a numerical method performs as intended is an integral part of code development. Semi-analytical benchmarks enable one such verification modality. Unfortunately, a semianalytical benchmark requires some degree of analytical fore-thought and treats only relatively idealized cases making it of limited diagnostic value. In the first part of our investigation (Part I), we establish the theory of a straightforward finite difference scheme for the 1D, monoenergetic neutron diffusion equation in plane media. We also demonstrate an analytically enhanced version that leads directly to the analytical solution. The second part of our presentation (Part II, in these proceedings) is concerned with numerical implementation and application of the finite difference solutions. There, we demonstrate how the numerical schemes themselves provide the semianalytical benchmark. With the analytical solution known, we therefore have a test for accuracy of the proposed finite difference algorithms designed for high order.

3:55 PMHigh Order Finite Difference Approximations to the One-Group Neutron Diffusion Equation in 1D Heterogeneous Media Part II: Implementation and ApplicationBarry D. Ganapol(1), David W. Nigg(2)1) University of Arizona, Tucson, AZ. 2) Idaho National Laboratory, Idaho Falls, ID

Verification that a numerical method performs as intended is an integral part of code development. Semi-analytical benchmarks enable one such verification modality. Unfortunately, a semi-analytical benchmark requires some degree of analytical fore-thought and treats only relatively idealized cases making it of limited diagnostic value. In the first part of our investigation (Part I, in these proceedings), we established the theory of a straightforward finite difference scheme for the 1D, monoenergetic neutron diffusion equation in plane media. We also demonstrated an analytically enhanced version that leads to the analytical solution. The second part of our presentation (Part II) concerns the numerical implementation and application of the finite difference solu-tions of Part I. Here, we demonstrate how the numerical schemes themselves provide the semi-analytical benchmark. With the analytical solution known, we therefore have a test for accuracy of the proposed finite difference algorithms designed for high or-der.

4:20 PMThe OECD/NEA/NSC PBMR 400 MW Coupled Neutron-ics Thermal Hydraulics Transient Benchmark: Transient ResultsGerhard Strydom and Frederik Reitsma(1), Prisca T Ngeleka and Kostadin N Ivanov(2)1) nPebble Bed Modular Reactor (Pty) Ltd., Centurion, South Africa. 2) The Pennsyl-vania State University, University Park, PA

The PBMR is a High-Temperature Gas-cooled Reactor (HTGR) concept developed to be built in South Africa. The analysis tools used for core neutronic design and core safety analysis need to be verified and validated, and code-to-code comparisons are an essential part of the V&V plans. As part of this plan the PBMR 400 MWth design and a representative set of transient exercises are defined as an OECD benchmark. The scope of the benchmark is to establish a series of welldefined multi-dimensional computational benchmark problems with a common given set of crosssections, to compare methods and tools in coupled neutronics and thermal hydraulics analysis with a specific focus on transient events. This paper describes the current status of the benchmark project and shows the results for the six transient exercises, consisting of three Loss of Cooling Accidents, two Control Rod Withdrawal transients, a power load-follow transient, and a Helium over-cooling Accident. The participants’ results are compared using a statistical method and possible areas of future code improvement are identified.

4:45 PMCurrent Status and Results of the PBMR “Pebble Box” Benchmark within the Framework of the IAEA CRP5Frederik Reitsma(1), Bismark Tyobeka(2)1) Pebble Bed Modular Reactor (Pty) Ltd., Centurion, South Africa. 2) International Atomic Energy Agency, Vienna, Austria

The verification and validation of computer codes used in the analysis of high tempera-ture gas cooled pebble bed reactor systems has not been an easy goal to achieve. A limited amount of tests and operating reactor measurements are available. Code-to-code comparisons for realistic pebble bed reactor designs often exhibit differences that are difficult to explain and are often blamed on the complexity of the core models or the variety of analysis methods and cross section data sets employed. For this rea-son, within the framework of the IAEA CRP5, the “Pebble Box” benchmark was formu-lated as a simple way to compare various treatments of neutronics phenomena. The problem is comprised of six test cases which were designed to investigate the treat-ments and effects of leakage and heterogeneity. This paper presents the preliminary results of the benchmark exercise as received during the CRP and suggests possible future steps towards the resolution of discrepancies between the results. Although few participants took part in the benchmarking exercise, the results presented here show that there is still a need for further evaluation and indepth understanding in order to build the confidence that all the different methods, codes and cross-section data sets have the capability to handle the various neutronics effects for such systems.


11C - Standards and BenchmarksSession Chairs: Hans Gougar (INL), Bismark Tyobeka (IAEA)

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10:20 AMSensitivity and Uncertainty of the Neutronic Parameters of BWR Ringhals-1 TRACE/PARCS Stability PredictionIvan Gajev and Tomasz Kozlowski(1), Yunlin Xu and Thomas Downar(2)1) Division of Nuclear Power Safety, Royal Institute of Technology, Alba Nova Univer-sity Centrum, Stockholm, Sweden. 2)Department of Nuclear Engineering and Radio-logical Sciences, University of MichiganAnn Arbor, MI USA

Unstable behavior of Boiling Water Reactors (BWRs) is known to occur during opera-tion at certain power and flow conditions. This paper reports the results of a prelimi-nary investigation of the impact of uncertainty neutronic parameters on the prediction of the stability of the BWR within the framework of OECD Ringhals-1 stability bench-mark. The time domain code TRACE/PARCS was used in the analysis and sensitivity calculations were performed for every neutronic parameter which was anticipated to affect the stability prediction of the reactor. The overall objective of the work here was to identify parameters most significantly affecting the stability phenomena in the Rin-ghals-1 stability benchmark. Using selected parameters, an uncertainty method based on the propagation of code input errors was used to obtain tolerance limits for the de-cay ratio and the frequency prediction. The scope of the uncertainty study reported in this paper was limited to the neutronic effects. Further studies will include other effects to include thermal-hydraulic parameters and modeling effects.

10:45 AMUncertainty Contribution to Final In-Rack K(95/95) from the In-Core Kinf Criterion Methodology for Spent Fuel Storage Rack Criticalty Safety AnalysesJohn C. Hannah(1), Walid A. Metwally and Vernon W. Mills(2)1) GE Hitachi Nuclear Energy (GEH), Wilmington NC USA. 2) Global Nuclear Fuel (GNF), Wilmington NC USA

The in-core k∞ criterion methodology is identified as an appropriate technique for ensuringcompliance with the 0.95 keff limit for spent fuel pool storage systems. Equa-tions to define theuncertainty associated with a linear fit of the in-core to in-rack reac-tivity are presented. A methodfor determining an appropriate lattice for use in bias and tolerance studies is also reported. Astudy of 12 BWR lattices is used to establish the typical total contribution to the maximumK(95/95) of the rack system from the uncer-tainty associated with the in-core k∞ criterionmethodology.

11:10 AMSensitivity Analyses for Small Fast Reactor Nuclear Characteristics with a Discrete Ordinate Transport Cal-culation MethodMasatoshi Kawashima(1), Yasushi Tsuboi and Akito Nagata(2), Mitsuaki Yamaoka(3) 1) Toshiba Nuclear Engineering Services Corporation, Yokohama, Japan. 2) Toshiba Corporation, Yokohama, Japan. 3) Toshiba Corporation, Kawasaki, Japan

The new sensitivity analysis tools, using a generalized perturbation approach with a discrete ordinate transport calculation method, have been applied to the 4S core in a two-dimensional RZ geometry model. Cross section sensitivity coefficients are compared with those from the conventional method. Through those comparisons, the results have shown that application of the new sensitivity calculation tools are efficient in order to reduce uncertainties in the sensitivity and uncertaintyS&U analyses of the 4S core.

11:35 AMSensitivity and Uncertainty Analysis with New Covari-ance Data of Minor Actinides for AFC DevelopmentsDo Heon Kim, Choong-Sup Gil, Hyeong Il Kim, and Young-Ouk LeeKorea Atomic Energy Research Institute, Yuseong, Daejeon, Korea

Nuclear data sensitivity and uncertainty analysis of keff has been performed with new covariance data of some minor actinides having been evaluated for AFC develop-ments. The DANTSYS/SUSD3D-based code system was utilized to estimate the keff uncertainties. The new covariance data of 237Np, 240Pu, and 244Cm have been assessed with appropriate benchmark problems taken from the ICSBEP benchmark specification. In addition, a fictitious critical system composed of a content of each nuclide to 100% was introduced to clarify the effects of the new covariance data. The covariance data from JENDL-3.3 and/or Low-Fidelity data were taken into consider-ation in the comparative analysis.


11D - Uncertainty Analysis in Modeling - ISession Chairs: Pino Palmiotti (INL), Ivan Kodeli (Jožef Stefan Institute)

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1:30 PMEvaluation and Uncertainty Analysis of the KRITZ-2 Criti-cal Benchmark Experiments I. Kodeli and L. Snoj Jožef Stefan Institute, Ljubljana, Slovenia

In order to validate new nuclear cross-section evaluations and computational meth-ods, a large number of benchmarks experiments were performed in the past. OECD/NEA launched several projects aiming to collect, preserve and disseminate the bench-mark data in a user-friendly format. Reactor physics benchmarks are covered by the IRPhE project. This paper presents the preparation of the IRPhE compilation for the KRITZ-2 critical experiments, consisting of altogether six configurations, both UO2 and MOX, measured at 2 different temperatures at STUDSVIK. These configurations were selected for the purpose of the OECD/NEA Uncertainty Analyses in Modeling (UAM) benchmark activities. Uncertainties due to input data uncertainties, modeling errors, and numerical approximations were studied, with particular emphasis on the uncertainties in the nuclear cross section data. SUSD3D sensitivity-uncertainty code with the SCALE-6.0 and JENDL-3.3 covariance data were used in this study. The consistency among the C/E values and the overall computational uncertainties is dis-cussed.

1:55 PMApplication of Global Sensitivity Analysis Approach to Exercise I-2 of the OECD LWR UAM Benchmark F. Puente Espel, S. Ghrayeb, K. Ivanov(1), S. Tarantola(2) 1) Department of Mechanical and Nuclear Engineering, The Pennsylvania State Uni-versity, University Park, PA, USA. 2) Joint Research Centre of the European Commis-sion, Ispra, Italy

This paper is three fold. In the first part, continuous energy Monte Carlo calculations are performed as reference solutions to the test problems of Exercise 2 of Phase 1 (I-2), “Lattice Physics”, of the OECD LWR UAM Benchmark. Using an automated tool for the generation of temperature-dependent continuous-energy cross section librar-ies, developed at the Pennsylvania State University (PSU), reference solutions are obtained using the different available Nuclear Data Libraries (NDL). These reference solutions provide the required data to the participants of the OECD UAM Benchmark. The second part of this paper focuses on the comparison between a new thermal scattering cross section library and the MCNP5 built-in thermal libraries. This gives the possibility of performing MCNP5 criticality calculations at the correct moderator tem-perature and improving the accuracy of the calculation. In the final part of the paper, global sensitivity analysis is applied to the modeling of nuclear reactor calculations for better model understanding. Specifically, it is investigated how much criticality condi-tions are affected by uncertainties in various inputs, including nuclear cross-sections, at different energies, from several isotopes in the fuel, the absorber and the modera-tor. The sensitivity analysis uses the Sobol’ and Jansen formulas, which allow us to estimate, for each uncertain input, its main effect, its total effect (i.e. the overall effect, which includes all the interactions, at any order, with all the other uncertain inputs), and all two-way interactions among all possible pairs of uncertain inputs. The sensi-tivity analysis consists of a number of model simulations, which are performed using MCNP5.

2:20 PMCross-Section Covariance Propagation for LWR Fuel Cells in One and Two DimensionsM. Ball and D. R. Novog(1), C. Parisi and F. D’Auria(2)1) McMaster University, Hamilton, ON, CANADA. 2) San Piero a Grado Nuclear Re-search Group (GRNSPG), University of Pisa, Pisa , ITALY

Within the framework of the Uncertainty Analysis in Modeling (UAM) for Design, Op-eration and Safety Analysis of LWRs Benchmark sponsored by the OECD/NEA, a tool has been developed for the propagation of covariance uncertainty through resonance self-shielding and other neutron kinetics calculations using a direct, cross-section gen-eration and substitution approach. The motivation behind the work described in this paper was to develop a portable uncertainty propagation tool that could be easily im-plemented with several neutron kinetics codes, without relying on detailed knowledge of the internal workings of those codes or access to adjoint solutions. Implemented ini-tially with the SCALE code package, “self-shielded” covariance matrices for common LWR fuel cells have been calculated, as well as contributions to Keff uncertainty by selected neutron cross-sections and processes in both one and two dimensions. The one dimensional results generated by the tool are compared against those obtained using the TSUNAMI-1D module of SCALE in order to verify the efficacy of the meth-odology. Onedimensional results show good agreement with TSUNAMI-1D, but there is also an indication that the loss of dimensionality corresponding to one-dimensional equivalent geometries of twodimensional fuel cells may lead to significant changes in the calculated uncertainty on Keff arising from particular neutron-nuclide reactions.

2:45 PMComparison of Uncertainty Quantification Methods for Fast Reactor NeutronicsC. Rabiti, G. Palmiotti, M. Assawaroongruengchot(1), B. M. Adams(2)1) Idaho National Laboratory, Idaho Falls, ID, US. 2) Sandia National Laboratories, Albuquerque, NM, US

This paper describes a comparative study of two Uncertainty Quantification (UQ) tech-niques for reactor neutronics analysis. Adjoint- and sampling-based uncertainty quan-tification approaches are applied to a simplified model of a sodium-cooled fast reactor to evaluate uncertainty in four integral parameters (void reactivity, Doppler reactivity, Keff, and control rod reactivity). Results obtained with the two UQ methodologies are compared and discussed. This preliminary work, performed in a well-studied field like neutronics, will inform sensitivity and uncertainty method selection for future calcula-tions in other disciplines like coupled neutronics/thermo-hydraulics.


11E - Uncertainty Analysis in Modeling - IISession Chairs: Maria Avramova (PSU), Carlo Parisi (Univ. Pisa)

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3:30 PMA Methodology for Determination of Detector Response for Inspection of a Spent Fuel PoolWilliam Walters, Alireza Haghighat and Michael Wenner(1), Shivakumar Sitaraman and Young Ham(2)1) Department of Nuclear and Radiological Engineering, University of Florida, Gaines-ville, FL USA. 2) Lawrence Livermore National Laboratory, Livermore, CA USA

In this paper we examine the predicted response of a fission chamber detector in the spent fuel pool at the Atucha-I reactor using the adjoint methodology with the aim of detecting proliferation. Burnup calculations to determine material composition and intrinsic neutron source are performed using the ORIGEN-ARP depletion code. Sub-critical multiplication is modeled using a simplified fission-matrix method. Fission-matrix coefficients are determined using MCNP for several burnups and decay times, which can be interpolated to arbitrary values. This method can quickly and accurately calculate the subcritical multiplication for a pool of any size, assembly burnup and cooling time. Adjoint function calculations for a fission chamber placed in the pool were performed using the PENTRAN Sn code. These results show that the detector field-of-view (FOV) is relatively insensitive to detector position within the pool, assembly burnup and cooling time. The adjoint results are coupled with the source calculations to predict the detector response in the spent fuel pool under both normal and prolifera-tion scenarios. Two spent fuel diversion scenarios are examined, including substitu-tion with dummy assemblies and attempted masking of a dummy assembly using a high-burnup assembly. Both assembly diversion scenarios show a predicted deviation from unperturbed response by at least 20%.

3:55 PMModeling of the Cherenkov Light Emission from Nuclear Fuel Assemblies with Partial DefectsS. Jacobsson Svärd, S. Grape(1), A. Hjalmarsson(2)1) Department of Physics and Astronomy, Division of Applied Nuclear Physics, Upp-sala University, Uppsala, Sweden. 2) The Svedberg Laboratory, Uppsala University, Uppsala, Sweden

The International Atomic Energy Agency poses requirements on the detection and verification of partial defects of nuclear fuel assemblies before being placed in dif-ficult-to-access storage. One instrument being considered for such detection is the Digital Cherenkov Viewing Device, with which images of the Cherenkov light from fuel assemblies in storage pools can be recorded and analyzed. This paper accounts for a software toolkit for simulating the Cherenkov photon distribution in the fuel us-ing GEANT4. The toolkit enables the user to access information on individual photon emission coordinates and their momentum vectors, as well as to take into account the expected rod-by-rod burnup distribution at different axial levels. An example of this modeling is demonstrated.

4:20 PMMANCINTAP: A Numerical Tool for the Analysis of Neu-tron Induced Activations and Their Radiological Effects in Complex 3D GeometriesM. Frignani, G. Firpo, S. FrambatiAnsaldo Nucleare S.p.A., Genova, Italy

Neutron activated materials are one of the major radiation source terms during shut-down conditions in nuclear power plants. Ex-core monitoring systems as well as main-tenance and decommissioning are strongly affected by the gamma radiation field due to the decay of activated materials. A good prediction of the radiological effects of neu-tron activations is of great importance in the design of nuclear power plants. A method aimed at solving this issue in a complex 3D environment was developed exploiting the capabilities of modern computer codes dedicated to neutron and photon Monte Carlo transport (MCNP) and to activation characterization of materials exposed to neutron sources (ANITA). A flexible and completely automated interface, MANCINTAP, allows the user to reduce the human effort, to avoid simplifying hypotheses and to increase the accuracy of results preserving a detailed representation of the environment. The tool was efficiently employed for the analysis of the time evolution of radiation levels inside the lower reactor cavity of Westinghouse AP1000 during shutdown conditions.

4:45 PMKrško Neutron Streaming Analysis and Measurement for Concrete Missile Shield RemovalJianwei Chen and Arnold H. FeroWestinghouse Electric Company LLC, Monroeville, PA, U.S.

An approach combining deterministic and Monte Carlo neutron transport methods is applied to analyze the neutron streaming problem inside the containment at the Krško Nuclear Power Plant. This problem has been raised due to the planned removal of the concrete missile shield blocks, which could potentially cause elevated neutron dose rates on the operating deck. In this analysis, the neutron source term leaking from the reactor vessel beltline is calculated using the DORT/DOMINO/MCNPBQ code sys-tem to account for the cycle-specific core power distribution. This source term is then coupled with a detailed MCNP model of the containment building in order to calculate the neutron dose rates and solid state track recorder (SSTR) results on the operating deck. The cell-based weight window and other variance reduction techniques have been used in the MCNP transport simulations to improve the calculation efficiency. The results show reasonable agreement between the calculation and the measure-ments with the concrete missile shield. The calculation performed without the concrete missile shield considered present shows an increase in neutron dose rates for most of the locations of interest. However, the analysis also shows that, with proper backfit shielding in place, the neutron dose rates on the operating deck can be reduced by a factor of up to 30 after removing the concrete missile shield, which is favorable from both regulatory and operational point of views.


12A - Radiation Applications and Nuclear Safeguards - ISession Chairs: Staffan Svärd (Uppsala Univ.), Rob Flammang (Westinghouse)

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3:30 PMVerification of a Hybrid Adjoint Methodology in TITAN for Single Photon Emission Computed TomographyKatherine Royston, Alireza Haghighat, and Ce YiNuclear and Radiological Engineering Department, University of Florida, Gainesville, FL USA

The hybrid deterministic transport code TITAN is being applied to a Single Photon Emission Computed Tomography (SPECT) simulation of a myocardial perfusion study. The TITAN code’s hybrid methodology allows the use of a discrete ordinates solver in the phantom region and a characteristics method solver in the collimator region. Cur-rently we seek to validate the adjoint methodology in TITAN for this application using a SPECT model that has been created in the MCNP5 Monte Carlo code. The TITAN methodology was examined based on the response of a single voxel detector placed in front of the heart with and without collimation. For the case without collimation, the TITAN response for single voxel-sized detector had a -9.96% difference relative to the MCNP5 response. To simulate collimation, the adjoint source was specified in directions located within the collimator acceptance angle. For a single collimator hole with a diameter matching the voxel dimension, a difference of -0.22% was observed. Comparisons to groupings of smaller collimator holes of two different sizes resulted in relative differences of 0.60% and 0.12%. The number of adjoint source directions within an acceptance angle was increased and showed no significant change in accu-racy. Our results indicate that the hybrid adjoint methodology of TITAN yields accurate solutions greater than a factor of two faster than MCNP5.

3:55 PMA Time-Dependent Calculational Model for Evaluation of Plutonium Proliferation ProtectionKulikov E.G., Shmelev A.N. and Apse V.A.(1), Kulikov G.G.(2)1) National Research Nuclear University “MEPhI”, Moscow, Russia. 2) International Science and Technology Center, Moscow, Russia

A calculational model is presented for evaluating fissionable materials protection against unauthorized proliferation. The model takes into consideration the main nu-clear physics and thermal processes in a crude (hypothetical) implosion-type nuclear explosive device (NED). It is proposed that plutonium is used as fissionable material in implosion-type NED. To protect fissionable material (plutonium) it is suggested to add some amount of radioactive isotope (238Pu), the alpha-decay heat of which could overheat implosion-type NED and render it non-functional. A mathematical model of non-stationary (time-dependent) warm-up of implosion-type NED has been developed in order to evaluate the rate of loss of its effectiveness for different isotopic composi-tions of plutonium and for different methods of heat removal. The importance of mea-sures for slowing down warm-up process of implosion-type NED was demonstrated. It was discovered that the time, during which implosion-type NED is functional, is a decisive factor in terms of evaluation of plutonium proliferation protection.

4:20 PMRecent Progress in the Practical Calculation Mainte-nance of Commercial Production of Cobalt-60 at Lenin-grad NPPA.V. Elshin, A.S. Ivanov(1), A.N. Pimenov, E.K. Gorbunov, R.V. Ikonnik-ov(2)1) FSUE “Alexandrov Research Institute of Technology”, Sosnovy Bor, Russia. 2) JSC “Concern Rosenergoatom”, Affliliate “Leningrad Nuclear Power Plant”, Sosnovy Bor, Russia

Leningrad NPP commercially produces Cobalt-60 in special assemblies inserted in RBMKreactor core. Calculation tools are used to determine cobalt activity and make decision on its timely extraction from the reactor. The paper describes results of cobalt activity calculation as compared to measurements, and influence of improved calcula-tion model on the calculation results.


12B - Radiation Applications and Nuclear Safeguards - IISession Chairs: Staffan Svärd (Uppsala Univ.), Rob Flammang (Westinghouse)

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3:30 PMA United Kingdom Nuclear Fuel Cycle Scenario Assess-mentRobert GreggThe United Kingdom National Nuclear Laboratory (NNL), Preston Laboratory, Spring-fields Works, Salwick, Preston, Lancashire, United Kingdom

With the advent of a substantial new nuclear build programme in the United Kingdom and the large amounts of plutonium already separated from MAGNOX and AGR spent fuel, it is important to develop fuel cycle scenarios in order to help guide policy mak-ing. Using in-house computer codes, the UK National Nuclear Laboratory (NNL) has assessed several potential UK fuel cycles with the aim of: • utilising the UK’s plutonium and depleted / reprocessed uranium stockpiles, • minimising the radiotoxic impact on a deep store repository, • maintaining current levels of nuclear power generation, • investigating potential reductions in uranium ore requirements assuming different re-use strategies, • highlighting any future revenue streams from fuel fabrication and reprocessing Using FCM, a new computer program developed by the NNL, the UK nuclear fuel cycle has been realistically modeled from 1950 in order to accurately determine the current situation and to prove the methods used by comparing against current levels of separated plutonium, depleted uranium and irradiated fuel. Six future (hypothetical) scenarios were then developed and the impact on fuel cycle logistics, plant sizing, material flow, radiotoxicity and decay heat determined. The work has focused on Pu and RepU (reprocessed uranium) re-use and the impact on fuel fabri-cation throughputs and an eventual waste repository. It should be noted that the fuel cycle scenario options and any of the conclusions are those of the author and the NNL and not necessarily those of the policy makers or UK Government.

3:55 PMMulti Criteria Analysis for the Long Term Planning of the Mexican Electrical System ExpansionCecilia Martin-del-Campo, Rodrigo Guadarrama and Juan Luis FrançoisDepartamento de Sistemas Energéticos, Facultad de Ingeniería, Universidad Nacio-nal Autónoma de México

A multi-criteria analysis was applied to the long term electricity expansion planning for Mexico for the 2008-2030 period. This methodology is based on a fuzzy logic inference system, which allows for the definition of a decision function that takes into account all the evaluation parameters. This function permits one to rank the alterna-tive expansion plans in order to determine the most attractive option. In this study four evaluation parameters were considered: (a) the total generating cost obtained from an optimization expansion using the WASP-IV model, (b) the economic risk associated with fuel prices increases, (c) the diversity of technologies in the mix, and (d) the exter-nal costs. The analysis was applied to a base case and to three additional expansion cases, which are very similar to the base case, but each of them excludes the addition of a certain type of candidate technology in the optimization planning. The base case is Plan A which has six candidate technologies available for the optimization planning. Plan B excludes coal; Plan C excludes oil, and Plan D excludes nuclear energy. After the decision analysis was made it was found that Plan B is best followed by Plan A, then Plan C and finally Plan D. The worst plan expansion was obtained when the nuclear candidate was excluded in the program of additions during the time period. The primary conclusion is that nuclear energy must participate in the mix of electricity generation. This result can be used to define the energy policy for electricity produc-tion in Mexico in the medium-long term scenario.

4:20 PMBrazilian Nuclear Renaissance in a Sustainable Devel-opment ScenarioM.S. Dias and J.R.L. MattosNuclear Technology Development Centre - CDTN, National Nuclear Energy Commis-sion - CNEN, Belo Horizonte, MG, Brazil

Brazil generated 326 TWh of hydroelectric power in 2005, accounting for ~81% of the electricity production for that year. The 2005 to 2030 projections for the Brazilian development indicate growths from 2,020 kWh to 4,380 kWh in the per capita electric-ity consumption and from US$4,300 to US$8,950 in the per capita GDP (in market exchange rate and 2005 US$). The consumption of electricity is to grow from 375 TWh in 2005 to 1,030 TWh in 2030. In simple view and without considerations of aspects related with energy efficiency, this growth means 1.8 times the generation capacity of the 20th century should be built along of 25 years of the 21st century. The Brazilian electricity generation will demand all primary sources to meet the foreseen growth of the electricity consumption. As economical, safe and clean primary energy source for electricity generation, the nuclear energy is one option capable of large-scale and

shortterm deployment in the Brazilian growth of the electricity consumption. The contri-bution of nuclear generation in the electricity consumption should evolve from the cur-rent 2.6% for amounts over 5% in 2030. The perspective of the Brazilian nuclear sector is evolving to be resumed and present an opportunity for pooling and rationalizing the available skills – technical, cultural and human. The role of business opportunities and the future demands in the value chain of nuclear activities are summarized in this document. Institutions of R&D and Brazilian universities play an important role for the formation of new demanded knowledge and human resources.

4:45 PMThe Status and Prospect of Nuclear Power Development in ChinaTianmin Xin And Daiyong SongChina Nuclear Power Engineering Co., Ltd (CNPE), haidian district, Beijing, China

It has been proved by practice that nuclear power is a safe, clean and reliable energy source for optimizing the energy supply structure and guaranteeing the national en-ergy safety, economic safety and environmental safety. Nuclear power has developed for over 20 years in China, to satisfy the increasing energy demand for rapid eco-nomic growth and optimize the energy supply structure for maintaining the sustainable development, China’s government has changed the strategy of developing nuclear power to “actively boosting nuclear power construction” and established a “Long- and Medium-term Development Plan of Nuclear Power (2005-2020)” which planned out the installed nuclear power capacity of 40 million kilowatts by 2020, about four percent of the country’s total. This lecture shows the current status and future prospect of the nuclear power development in China.

5:10 PMSustainable Development in India- A Case for Nuclear PowerSudhinder ThakurNuclear Power Corporation of India Limited. Mumbai, INDIA

India needs a sustained high economic growth to realize its objectives of poverty al-leviation and improving the standard of living of its population. Energy/ Electricity being the key driver for economic growth, there is a pressing need for large augmentation in generation capacity, infrastructure and enhancement of energy efficiency to ensure that there is equity amongst population as far as energy availability is concerned. India is not very energy rich and has limited resources of fossil fuels. India’s nuclear power resources profile comprises of very modest uranium but abundant thorium resources. A unique three stage programme, based on optimum utilization of indigenous resourc-es, offers a solution for the country’s long term energy security and sustainability. In-dia’s nuclear power programme is based on a closed fuel cycle. The philosophy, apart from increasing the energy potential of the resource manifold, reduces the amount waste considerably. There is also the benefit of nuclear power being clean free. While the indigenous nuclear power programme is robust and on course, a much faster nuclear capacity addition in the near term, to meet the rising demand and mitigating existing energy shortages, is contemplated through additionalities based on interna-tional cooperation.


13A - Nuclear Power and Sustainable DevelopmentSession Chairs: Kevin W. Hesketh (UK National Nuclear Lab), Sudhinder THAKUR (Nuclear Power Corporation of India

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Sunday May 9, 2010, 8:00 am – 12:00 pm Workshop # 1 The AP1000 Reactor: Passive Safety Design and Core Design Characteristics Grand Station Ballroom 5 Chairs: Jim Winters and Bob Fetterman (Westinghouse)

Sunday May 9, 2010, 8:00 am – 12:00 pm Workshop # 2 A Short Course on Analytical Benchmarking in Neutron Transport Theory Grand Station Ballroom 3 Chair: Barry D. Ganapol (University of Arizona)

Sunday May 9, 2010, 8:00 am – 12:00 pm Workshop # 3 SCALE 6 TRITON Lattice Physics Grand Station Ballroom 4 Chairs: Mark DeHart, Matt Jessee, Jess Gehin, and Steve Bowman (Oak Ridge National Laboratory)

Sunday May 9, 2010, 8:00 am – 12:00 pm Workshop # 4 AEGIS/SCOPE2: A Next Generation Core Analysis System Grand Station Ballroom 2 Chairs: Masahiro Tatsumi (NFI) and Akio Yamamoto (Nagoya University)

Sunday May 9, 2010, 1:30 pm – 5:30 pm Workshop # 5 The Advanced Boiling Water Reactor (ABWR) Grand Station Ballroom 5 Chair: Juan Casal (Westinghouse)

Sunday May 9, 2010, 1:30 pm – 5:30 pm Workshop # 6 Transport Methods for Reactor Core Calculations Grand Station Ballroom 3 Chair: Richard Sanchez (CEA, Saclay)

Sunday May 9, 2010, 1:30 pm – 5:30 pm Workshop # 7 PARCS Reactor Core Simulator Grand Station Ballroom 4 Chair: Tom Downar (University of Michigan)

Sunday May 9, 2010, 1:30 pm – 5:30 pm Workshop # 8 Reactor Physics Analysis with Monte Carlo Grand Station Ballroom 2 Chair: Forrest Brown (LANL)

Friday May 14, 2010, 8:00 am - 4:30 pm Workshop # 9 The Pebble Bed Modular Reactor: From V.S.O.P. (Very Superior Old Product) to Generation IV candidate Grand Station Ballroom 3 Chair: Frederik Reitsma (PBMR)

PHYSOR 2010 - Advances in Reactor to Power the Nuclear RenaissanceMay 9 and 14, 2010

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