integraltestfacilitiesandthermal-hydraulicsystemcodes...

Hindawi Publishing CorporationScience and Technology of Nuclear InstallationsVolume 2012, Article ID 826732, 3 pagesdoi:10.1155/2012/826732

Editorial

Integral Test Facilities and Thermal-Hydraulic System Codesin Nuclear Safety Analysis

Klaus Umminger1 and Alessandro Del Nevo2

1 AREVA NP GmbH, Paul-Gossen Strasse 100, 91052 Erlangen, Germany2 ENEA UTIS-TCI, C.R. Brasimone, 40032 Camugnano (Bo), Italy

Correspondence should be addressed to Klaus Umminger, [email protected]

Received 1 December 2011; Accepted 1 December 2011

Copyright © 2012 K. Umminger and A. Del Nevo. This is an open access article distributed under the Creative CommonsAttribution License, which permits unrestricted use, distribution, and reproduction in any medium, provided the original work isproperly cited.

A considerable amount of resources has been devoted at theinternational level during the past three decades for estab-lishing and conducting experimental programmes in scaled-down integral test facilities (ITFs). These were aimed atsolving open issues for current nuclear power plant (NPP),demonstrating the technical feasibility of innovative designs,and generating reference databases to support code develop-ment and assessment. Since the end of the nineties, the main-tenance of the competences of experienced research teams aswell as the creation of a new generation of professionals in thearea of the nuclear safety is also a priority objective of opera-ting research facilities and promoting experimental pro-grammes.

Tens of ITFs have been built and operated so far all overthe world. Few of them, related to existing water reactor tech-nology, are currently in operation (e.g., ATLAS, PMK,PKL-III, ROSA/LSTF, INKA) or under refurbishment (e.g.,PACTEL, PSB-VVER). Some others are constructed or underdesign and are focused on innovative water reactor concepts(e.g., MASLWR, SPES-3).

The experimental data from such facilities are applicableto full-scale nuclear plant conditions; if the test facilities andthe initial and boundary conditions of experiments are pro-perly scaled, for example, the scaling will not affect the evolu-tion of physical processes important for the postulated acci-dent scenario. This evaluation determines whether the datamay be used in nuclear plant safety analyses of a postulatedaccident.

On the other side, the experimental data are fundamentalfor supporting the development and demonstrating thereliability of computer codes in simulating the behavior of an

NPP during a postulated accident scenario: in general, this isa regulatory requirement. The reliance on computer codes,among the other things, is based on the reason that, veryoften, one cannot directly apply results from test facilities toa plant, including the “reference” plant of the facility design.

Applications of computer codes to accident analyses re-quire the implicit assumptions that these codes have the cap-abilities to scale up phenomena and processes from test facili-ties to full-scale plant conditions. However, the differentscale, in terms of geometry, characterizing any facility anda nuclear plant does not ensure a priori that a code, which isable to reproduce a generic transient in a scaled facility, is alsoable to calculate with the same accuracy the same transient inNPP.

The aforementioned topics involve a number of key acti-vities, in which the research is taken up. This special issuedocuments the present scientific and technical status andrecent advances in relation to the ITF, the experimental pro-grammes, the issues connected with the code assessment, andthe scaling issue, which is related both to the representative-ness of the phenomena in the facilities as well as the appli-cability of the codes.

The special issue collects 17 papers, which are dividedinto 4 main groups, according to the following rationale: (1)Western PWR technology (8 papers), (2) Eastern PWR (orVVER) technology (4 papers), (3) BWR technology (3 pa-pers), and (4) innovative, integral-type, water reactor con-cepts (2 papers).

The paper by K. Umminger et al. presents the PKL-III integral test facility and provides a survey of test objec-tives and programs carried out to date as well as plans for

2 Science and Technology of Nuclear Installations

future investigations. The authors point out important re-sults achieved in improving the level of understanding of thePWRs system response under accident conditions as well asof the thermal-hydraulic phenomena relevant for the nuclearreactor safety and the importance for code validation.

Then, K. Y. Choi et al. describe ATLAS integral testfacility, the main findings, and lessons learned obtained usingthe data of the past integral effect tests. The paper discussesalso the future prospects of the application of ATLAS facilityin the framework of national nuclear R&D program in Korea.

T. Takeda et al. discuss the analysis of a PWR loss offeedwater (LOFW) by means of a ROSA/LSTF experimentconducted for the OECD/NEA ROSA project and the appli-cation of RELAP5 code. The description of relevant phenom-ena of the transient and the results of assessment of the codeare reported in the paper.

C. Addabbo and A. Annunziato illustrate the LOBI faci-lity project. The paper provides a historical perspective andsummarizes major achievements of the research programmewhich has represented an effective approach to internationalcollaboration in the field of reactor safety research and devel-opment. Focus is also given to the issue of the managementof research data.

The paper of J. Freixa and A. Manera addresses the issueof the validation process of system codes for the transientanalyses of PWR. Five posttest analyses by TRACE code arepresented. They are related to small and intermediate LOCAexperiments executed in ROSA/LSTF integral test facility.The paper underlines the relevance of the validation activityin relation to the modeling of an NPP for safety analysis pur-poses.

A. Bousbia Salah and J. Vlassenbroeck present the anal-ysis of three experimental tests performed in PKL-III integraltest facility, executed in the framework of the OECD/NEAPKL-2 project. The tests are devoted to the study of the cool-down procedures operated after the reactor trip in order tobring the primary side temperature and pressure to the resi-dual heat removal system (RHRS) operating conditions. Theobjective is to assess the impact of a chosen cool-downstrategy upon the occurrence of natural circulation inter-ruption (NCI) and the capabilities of CATHARE2 code inpredicting such process. Relevance is given to the interactionbetween the key parameters governing the transient and thephenomena involved.

F. Reventos et al. analyze the results of three posttest cal-culations of the LOBI integral test facility experiments. Theselected experiments are LOCA scenarios of different breaksizes and with different availability of safety injection com-ponents. The objective of the analysis is to improve the know-ledge of the phenomena reproduced in the facility in order touse them for nodalization qualification purposes of nuclearpower plants or for establishing accuracy databases for un-certainty methodologies.

The last paper of the Western PWR technology group byA. Prosek discusses the assessment of how the accuracy ofa simulation depends on the code version used. The codeselected is RELAP5, widely used all over the world for safetyanalysis of nuclear power plants. The study is based on an ex-perimental test executed in the BETHSY integral test facility,

identified as 9.1b, which is the OECD International StandardProblem no. 27.

A. Del Nevo et al. discuss a benchmark activity amongdifferent thermal-hydraulic system codes (i.e., ATHLET,RELAP5-3D, KORSAR, and TECH-M) and institutions per-formed on the basis of the PSB-VVER integral test facilitydata. The activity is preformed in the framework of theOECD/NEA PSB-VVER project. The objective is to collect,analyze, and document the numerical activity (pre- andposttest) performed by the participants, describing the per-formances of the code simulations and their capabilities toreproduce the relevant thermal-hydraulic phenomena obser-ved in the experiment.

G. Ezsol et al. illustrate the PMK-2 facility, which was thefirst integral-type facility for VVERs. The paper gives com-prehensive information on the design features of PMK-2facility with a special respect to the representativeness ofphenomena, the experiments performed, and the results ofthe validation of ATHLET, CATHARE2, and RELAP5 codes.It discusses also the safety significance of the PMK-2 projects.

The paper by V. Kouhia et al. describes the constructionand the experimental research activities of two integral testfacilities: the former, PACTEL, devoted to the simulation ofVVER-440 NPP and the latter, PWR-PACTEL, for researchactivities associated with the EPRTM reactor. The OECDInternational Standard Problem no. 33 and the PWR-PACTEL benchmark launched in 2010 are presented assample activities.

V. Blinkov et al. present three LB LOCA tests, two MSLBtests, and one SB LOCA test carried out in the integral testfacility BC V-213, which represents the bubble condenser sys-tem of Kola NPP (Unit 3). The paper provides informationabout the application of ATHLET code for supporting thedesign of the experiments and, then, describes in detail thevalidation activity related to the COCOSYS code.

The paper of D. Paladino and J. Dreier is focused on themultipurpose facility PANDA. The applications cover inte-gral containment response tests, component tests, primarysystem tests, and separate effect tests. The paper provides anoverview of the research programs performed in relation toBWR containment systems and those planned for PWR con-tainment systems.

S. Leyer and M. Wich describe the INKA (integral testfacility Karlstein) test facility, which was designed and con-structed to test and demonstrate the performance of the pas-sive safety systems of KERENA, the new AREVA boiling waterreactor (BWR) design. The authors discuss the facility feat-ures and capabilities, the instrumentation systems, and theexperimental program. INKA is within the KERENA devel-opment program devoted to single component/system testsof the emergency condenser, the containment cooling con-denser, and the passive core flooding system. The paperoutlines the integral system tests, which will be performedto simulate transients and LOCA (loss of coolant accident)scenarios. These experiments will involve the testing of theKERENA Passive Pressure Pulse Transmitter System.

S. Racca and T. Kozlowski present a validation activitycarried out with TRACE code related to the simulation ofspray cooling injection in a BWR reactor. The data, acquired

Science and Technology of Nuclear Installations 3

by experiments of the Swedish GOTA test facility, are com-pared with the results derived from the code modeling. Focusis given to the countercurrent flow limiting (CCFL) pheno-menon. The activity includes the application of the propaga-tion of input errors (PIEs) method, which is used to performthe uncertainty analysis and to identify the input parametershaving more influence on the figure of merit selected (in thiscase the peak cladding temperature).

The first paper of the innovative (integral-type) water re-actor concepts group by F. Mascari et al. describes the integ-ral test facility MASLWR. The multiapplication small light-water reactor (MASLWR) is the scaled-down model of asmall modular PWR, relying on natural circulation duringboth normal and accident conditions. The paper includes areview of the main characteristics of the facility and sum-marizes the tests already executed and the related code vali-dation activity. Finally, the reader can find information onthe IAEA International Coordinated Research Program, cur-rently ongoing, based on the MASLWR experimental tests.

The paper of A. Achilli et al. presents the SPES-3 integraltest facility. It is designed by SIET to simulate the W-IRISreactor concept under normal and accident conditions. Theauthors give an overview of the design of the test facilityunderlining the uses of RELAP5 and RELAP5/GOTHICcoupled codes for the achievement of the final scaling con-cepts and layout of the facility.

In conclusion, the scientific and technical contributionsfrom the authors provide the readers with useful informationrelated to 14 experimental facilities and cover a broad spec-trum of the recent past and current activities dedicated to thisspecial issue.

Acknowledgments

The guest editors acknowledge all authors who have submit-ted papers to this special issue. Special thanks are due to ourcolleagues for the kind collaboration in reviewing thesepapers.

Klaus UmmingerAlessandro Del Nevo

TribologyAdvances in

Hindawi Publishing Corporationhttp://www.hindawi.com Volume 2014

International Journal of

AerospaceEngineeringHindawi Publishing Corporationhttp://www.hindawi.com Volume 2010

FuelsJournal of


Journal ofPetroleum Engineering


Industrial EngineeringJournal of


Power ElectronicsHindawi Publishing Corporationhttp://www.hindawi.com Volume 2014

Advances in

CombustionJournal of


Journal of


Renewable Energy

Submit your manuscripts athttp://www.hindawi.com


StructuresJournal of

International Journal of

RotatingMachinery


EnergyJournal of


Hindawi Publishing Corporation http://www.hindawi.com

Journal ofEngineeringVolume 2014

Hindawi Publishing Corporation http://www.hindawi.com Volume 2014

International Journal ofPhotoenergy


Nuclear InstallationsScience and Technology of


Solar EnergyJournal of


Wind EnergyJournal of


Nuclear EnergyInternational Journal of


High Energy PhysicsAdvances in

The Scientific World JournalHindawi Publishing Corporation http://www.hindawi.com Volume 2014

integraltestfacilitiesandthermal-hydraulicsystemcodes...

Documents