iync summer bulletin

IN THIS ISSUE: Safety Culture - IYNC2012 Update - ENYGF11 - Student Section -

Celebrating Achievements and Rethinking the Future of IYNC

International Youth Nuclear Congress

Youth Future Nuclear

IYNC Bulletin June 30 2011 Summer Issue N 02

Defining Safety Culture

International Youth Nuclear Congress Youth Future Nuclear

www.iync.org

1. Editorial: “Safety Culture” & “YGN Japan”

2. Alumni Section:


3. IYNC2012 update

4 Countries Reports

5. Continental Reports

6. Future Events

7. Students Section

8. Technical Articles.

Argentina, Australia, Austria, Belarus, Belgium, Bolivia,

Brazil, Canada, Croatia, Czech Republic, Finland, France,

Germany, Hungary, India, Israel, Italy, Japan, Korea,

Lithuania, Malaysia, Netherlands, Nigeria, Norway, Peru,

Romania, Russia, Slovakia, Slovenia, South Africa, Spain, Sri

Lanka, Sweden, Switzerland, Taiwan, Tanzania, United

Kingdom, Ukraine, United States.

IYNC Bulletin Committee

Chair: Miguel Millan (Spain)

Alumni: Hans Korteweg (Holland)

IYNC 2012: Craig Albers (US)

Country Reports: Misha Swason (US)

Continental Report: Rit Desai (India)

Future Events: Lavinia Rizea (Romania)

Student Section: Mahima Gupta (US)

Technical Articles: Silvia Ortega (Spain)

Frontpage: Kristine Madden (US)

Youth

Nuclear

Future


IYNC Bulletin 1 Section 1: Editorial

Editorial

“Safety Culture”

After the Chernobyl accident occurred 25 years ago, the International Nuclear Safety Advisory Group

(INSAG) defined safety culture as follows: “Safety culture is that assembly of characteristics and attitudes in

organizations and individuals which establishes that, as an overriding priority, nuclear plant safety issues

receive the attention warranted by their significance”.

In safety culture two main players exist: staff and management. The workforce has to be aware of their

responsibility and attentive to the changing conditions when a safety related activity is carried out. In these

situations, the application of human performance practices will minimize the frequency and severity of

events which are the main result of safety culture with regards of workers. The second actor, the

management team, should lead their personnel by creating a strong commitment to excellence in fostering

both personal accountability and corporate self-regulation in safety matters.

Therefore, management should demonstrate Individual and Corporate awareness of the importance of

safety; stressing the commitment, requiring demonstration at senior level management of the high priority

of safety and adopting by individuals of the common goal of safety; reinforcing the knowledge and

competence; promoting the knowledge transfer, by training and instruction of personnel and by their self-

education; motivating through leadership, recognising good performances and learning from bad

behaviours; promoting a questioning attitude through any employee; and finally, responsibility, through

formal assignment and description of duties and their understanding by individuals, in order to incorporate

prevention philosophy rather than a reaction philosophy.

As young professionals, scholars and students we should be proud of our safety culture, technology,

continuous strive for excellence, and accountability even in the worst scenarios. We salute, once again, our

Japanese colleagues, who have been recognised by the IAEA mission for their “exemplary” performance at

Fukushima.

IYNC Officers


IYNC Bulletin 2 Section 1: Editorial

Young Generation Network Japan

At first, we, YGN-Japan, would like to thank all of the people who are praying for Japan and giving us warm

messages. As you know, Japan experienced an enormous earthquake on 11 March 2011 and has since been

struggling to bring Fukushima Daiichi NPP (1F) under cold shutdown condition. Some members of YGN-

Japan have been working as an engineer for 1F or as a volunteer for Tsunami victims.

Although there have been no injury and health effect to the residents around 1F, the accident have had a

severe impact on their life. Environmental monitoring data have indicated high radiation dose rate near 1F

and residents around 20km from 1F have evacuated. Some agricultural and fishery products were under

distribution and consumption restriction. There are many critical opinions to Nuclear Power Plants (NPPs),

and many NPPs, which are on periodic inspection, are not allowed to restart.

However, we think NPPs are essential for our industry and as a countermeasure against global warming.

Therefore, under this severe situation, we are individually trying to obtain public understanding to NPPs

through personal communications, web, twitter, and etc.

Now, YGN-Japan is discussing “What can nuclear engineers and YGN do for people and future

generations?”, “What is an ideal image of future NPPs.” We will try to re-build NPPs’ confidence in future

discussions with logical manner.

For future generations, YGN-Japan would like to regain NPPs’ public trust with you, not only for Japan but

also all over the world. To do so, we would like to work together with all the YGNs in the world.

Again, we really appreciate all your support.

YGN Japan


IYNC Bulletin 3 Section2: Alumni Section

Alumni Section


In 2010, International Youth Nuclear Congress (IYNC) celebrated its 10th anniversary. Since the inaugural

Congress took place, in Bratislava, Slovakia, in 2000, the Network has brought together thousands of

students and young nuclear professionals from many countries around the globe. To mark this first

triumphant decade, IYNC organised a workshop entitled Celebrating Achievements and Rethinking the

Future of IYNC during the 2010 Congress (IYNC 2010), which took place in Cape Town, South Africa. The

workshop paid tribute to the people who conceived the IYNC idea and helped the organisation to develop

and thrive.

Summarising the IYNC’s achievements, however, brought up new questions regarding the future of the

network. The nuclear industry constantly evolves, posing new challenges to students and young

professionals. Are the IYNC’s mission and goals still relevant today, or should they be revised to reflect

changing realities? An international panel of IYNC veterans (August Fern and Alexander Tsibulya) and key

contributors (Marco Streit and Stewart Lynas), to name but a few, openly shared their thoughts on this

subject, with the active participation of the audience, in an effort to formulate IYNC’s future goals and

outline the strategies needed to achieve them.

During the workshop, it became clear that IYNC needed to continue to differentiate between the

“Network” and the “Congress”. The Congresses themselves have, in a way, become self-sustaining over the

past ten years. However, it was pointed out that the IYNC Network still required a lot of work to develop

and expand. Participants agreed that the IYNC brand was slowly becoming recognised internationally, but

that more needed to be done in terms of marketing. The website and social media were identified as key

tools to this effect. There were also calls to develop some sort of quarterly newsletter or bulletin to inform

students and young professionals about IYNC activities, as well as what is happening around the world in

national and regional young generation organisations [the IYNC Bulletin has now seen the light of day!].

It was also pointed out by a number of participants that the technical programme needed to be developed

further in order to gain international recognition as a truly high calibre programme that companies and

organisations could support by encouraging their staff to submit high-quality papers. Following the same

thought process, there were also calls for more top-level technical speakers during the plenary sessions.

There were also suggestions to streamline/simplify the statutes to prevent IYNC from becoming bogged

down by too much bureaucracy and procedure – especially when it came to choosing the locations for

future Congresses. Finally, some of the “older” participants also called for an alumni network.

Responding to the call for IYNC to grow and adapt the IYNC Board of Directors, meeting during IYNC 2010,

voted to amend the by-laws to include an additional mission point: “Provide a platform and create an

enabling environment to facilitate the building of professional networks that will open up future

opportunities.” The Board of Directors also agreed to review the statutes in the future in an effort to clarify

a number of ambiguous clauses. And as part of the process of self-renewal, a new team of young and eager

Officers was elected to lead IYNC for the next two years.


IYNC Bulletin 4 Section2: Alumni Section

One of the direct outcomes of this workshop was the conception of some sort of alumni group to help

counsel IYNC in the future. Yes, there is life after 35! As they get older, wiser and move up the ranks in our

industry, our colleagues and future alumni are now well-placed to help guide future generations of IYNC

leaders. Initial steps have already been taken to establish such a body. A small but dedicated group of IYNC

veterans (Founders, past Officers, General Co-Chairs and Technical Chairs) have been contacted in an effort

to formalise this group. The plan is to have this new advisory body up-and-running by the next biennial

Congress (IYNC 2012), in Charlotte, North Carolina (USA). The final stages of this process will mean

amending the IYNC by-laws to recognise the new advisory body.

It is clear that many challenges remain for IYNC and the nuclear community as a whole to meet. The IYNC

Network will need to consolidate its support for existing young generation networks around the world,

whilst continuing to expand and help develop young generation networks in Africa, Asia, Australia and

South America. Developing the Network will be vital to IYNC’s long-tern sustainability. However, one must

not forget the Congresses. They will need to continue to take place at various venues around the world,

spreading the “gospel” of IYNC. IYNC is now on a solid footing, with a new group of dedicated and energetic

supporters leading the way. It’s impossible not to feel extremely positive about the future of IYNC.

Hans Korteweg, Past President, IYNC

Stewart Lynas


IYNC Bulletin 5 Section 3: IYNC2012 update

IYNC2012 Update

As many already know, IYNC2012 will take place in the United States in Charlotte, North Carolina. Planning

for IYNC2012 is already underway and we are going full steam ahead!

The first meeting of the Executive Committee (Ex-Com), the planning team for IYNC2012, took place in

Brussels, Belgium in conjunction with the PiME Nuclear Communicators Conference and European Nuclear

Young Generation Forum (ENYGF) Officers meeting earlier this year. At this meeting the members from all

over the world that compose the Ex-Com team began planning the major aspects of the conference! The

budget, fundraising strategy and many other important topics were discussed. While no major decisions

were made, the foundation for an amazing conference was set up.

In May of this year, the Ex-Com held a second face-to-face in Washington DC prior to the North American

Young Generation in Nuclear’s (NA-YGN) annual professional development conference. This was the first

major young generation conference that IYNC2012 was able to advertise to and it was a major success.

Interest from the young generation has never been higher! The domestic and international interest for the

upcoming conference is so large that the Ex-Com is estimating participation at well around 600 young

members! We hope you will be one of them!

During June the Ex-com composition has been approved by the IYNC Board of Directors:

Ex-Com Member Position Contact Information

Craig Albers General Co-Chair, Local Co-Chair [email protected]

Landon Kanner Local Co-Chair [email protected]

Elizabeth McAndrew-Benavides Registration Chair [email protected]

Ryan Boyle Technical Tours Chair [email protected]

Christine Csizmadia Local Corporate Sponsorship Chair [email protected]

Erin West/Delegate Professional Development Content Chair [email protected]

Miguel Millan General Co-Chair [email protected]

Igor Vukovic Corporate Sponsorship Chair [email protected]

Melissa Crawford Finance Chair [email protected]

Amy Bird Publications Chair [email protected]

Lavinia Rizea International Relation Chair [email protected]

Wim Uyttenhove Program Chair [email protected]

The first call for papers went out in June of 2011. Registration for the conference will open in early 2012.

For more information on IYNC2012 in Charlotte and to submit an abstract visit www.IYNC.org!

mailto:[email protected]












Now Accepting Track Chair Applications for International Youth Nuclear Congress (IYNC) 2012

5-11 AugustCharlotte, USA

Each technical track will be managed by an international and US track chair. A technical track chair should at a minimum have a general understanding of the track they are leading, manage the review and acceptance of abstracts and submitted papers for the track, and organize the track session during the conference.

http://www.iync.orgAll experience levels welcome

Track 1: Strategic Planning & Professional Development Track 2: Plant Design, Construction, Operation, Maintenance & Decommissioning Track 3: Nuclear Fuel Cycle & Waste Management Track 4: Nuclear Politics, Economics & Human Resources Track 5: Nuclear Safety, Radiation Protection and Shielding Track 6: Advanced Nuclear Systems Track 7: Radiation Science, Medical Applications & N on Base-load Nuclear Applications Track 8: Reactor Physics Track 9: Thermal Hydraulics & Fluids Track 10: Materials Science & Technology Track 11: Young Generation Unique Best Practices

In addition to the comprehensive technical program, a set of interactive unique workshops will be organised in both technical and non-technical domains. Also social and cultural events are arranged. These activities will allow delegates to network and to experience the unique culture of the American Southeast.

Proposed Technical Tracks Important Due DatesAbstracts - September 30, 2011Drafts papers - January 1, 2012Final papers - February 29, 2012

Interested or Questions? Do not hesitate to contact the chairing team via the information below. Please indicate which track you are interested in and attach a small bio or resume that shows your interest in a technical track (from the list below), and join an enthusiastic conference management team!

Antonio Lafuente [email protected]

Technical Track Chair IYNC2012


IYNC Bulletin 6 Section 4: Continental Reports

Section 4: Country reports

1.1 Activities of the Austrian YG related to Fukushima Daiichi

1.2 Update from Belgium

1.3 Discussion on Fukushima Daiichi in Canada

1.4 Fukushima Daiichi is close to home for KYGN (Korea Young Generation

Network)

1.5 Safety and nuclear professionalism in the Dutch Young Generation

1.6 YG Activities in Slovakia

1.7 Education and Training Activities in Spain

1.8 The Swedish way of Young Generation

1.9 Response to Fukushima Daiichi in the UK

1.10 Successful NA-YGN response to the Fukushima events Nuclear education in

South Africa

1.11 European Nuclear Young Generation Forum2011 (enygf09).



Actually (June 2010), there are 55 members in the YG of the

Austrian Nuclear Society. Most of the YG members are

students of physics and mechanical engineering. Some of

them are young professionals in the nuclear field (which is

quite difficult in a non nuclear country)

The main activities of the Austrian YG related to the

Fukushima:

Establishing a Information Center (Call Center) at the Institute of Atomic and Subatomic Physics

The aim of information center was to provide information to the public as well as for the media (the info center could be reached by phone)

4 to 6 members of the YG have been present at the information center from Monday to Friday for about 4 weeks after Fukushima

Hundreds of calls from media (TV, Radio, Newspapers etc.) or other people to Fukushima and realted topics have been answered by the YG members

Interviews and statements for the Austrian TV Stations have been given by YG members

Factsheets and papers about NPP have been produced to be published by different media

Information about Fukushima has also been provided via the Austrian Nuclear Society Homepage and per email to the YG network

YG excursion to the Olkiluoto site in Finland 2010

This year, many activities have been organized by the

BNS-YG such as the BNS thesis contest, a technical visit

at the Westinghouse Pumps and Motor Maintenance

and Repair Center, an IYNC and ENS-YGN committees

meeting in Brussels, a workshop at Pime, the

conference on public information materials exchange

which was hold in Brussels. Also, members of BNS-YG

have participated to the European nuclear young

generation forum in Prague.

As you all know, the future of nuclear power production

seems to be jeopardized across Europe. Germany and

Switzerland took their decisions and we all respect that

democratic process (even if we are puzzled by the

justification of these decisions). For Belgium, situation is

unclear and thus it is even more important than before

for BNS-YG to provide information to as much people as

possible. The aim is not to convince people, or polarize

even more the debate, but to give tools to people to

make their own mind about what they want in the

coming years. In order to do that, we tried the last few

months to discuss with engineering students (at the

Free University of Brussels as well as at the SFEN

conference “Energie Nucleaire: le défi du temps” in

France) about the nuclear situation in Belgium and we

will focus our future Evening Lectures and events

around topics like seismic situations in Belgium.

Since March and the Fukushima incident, BNS-YG has

also taken to decision to explore new communications

channels such as a trimestrial e-newsletter and

interviews of nuclear professionals for our website.

Even more than before, we need to have a crystal clear

communication and a policy of openness to people

looking for neutral and fair facts about nuclear. We are

willing to interview key members of the Belgian nuclear

community in order to give a broader perspective on it.

We are at a crossroad for the electrical nuclear industry,

thus it is really important that we provide an

understandable and honest description about it. The

newsletter is a small, but useful, tool for that.



IYNC members in Canada, specifically Yung Hoang,

Bharath Nangia and Siddharth Das have been actively

involved with various NA-YGN chapters to organize and

oversee discussion forums and presentation with

regards to the events at the Fukushima Daiichi Nuclear

Power plant. The following presentations were

organized for NAYGN chapters and CNS

conference and industry SMEs (Subject Matter Experts)

were invited to present and participate in subsequent

discussions:

Sharing lessons learned from Fukushima Daiichi Nuclear Power plant events - presented by Victor Kreft (ph.D candidate, McMaster University)

Review actions taken in response to Fukushima events - presented by Dr. James Whitlock (Reactor Physicist, AECL)

Safety of CANDU reactors versus BWRs and PWRs - presented by Dr. James Whitlock (Reactor Physicist, AECL)

Effectively responding to public concerns with regards to Nuclear Power - presented by Dr. James Whitlock (Reactor Physicist, AECL)

Introduction to IYNC 2012 - presented by Bharath Nangia

.

Since Korea is the closest country from Japan, the

Fukushima nuclear accident is very seriously

concerned in Korea more than other countries. The

safety of nuclear power plants in Korea became

one of the top issues with public fear against

nuclear. Moreover, the media articles about the

Fukushima accident and its prediction with

uncertain information made the situation worse at

the early stage of the accident.

The KYGN (Korea Young Generation Network)

workshop has been held in spring and autumn

every year. In the workshop which is firstly held

after Fukushima, young participants from

universities, industries, and institutes shared

information they have and discussed the activities

of their institutes for post Fukushima actions. The

status of Fukushima plants, the effects of the

accident to Korea, the public concern about the

safety of Korean nuclear power plants, and future

plans for the post Fukushima actions were handled

as issues for the workshop. In order to deliver

correct information to the KYGN members, we had

discussions among young experts from various

institutes and research fields. Such activities of

KYGN have been periodically posted in Nutopia

which is a magazine published by KNS (Korean

Nuclear Society).

Nowadays, we have been performing a project for

KYGN. We are collecting and surveying the

information about the advanced nuclear power

plants in order to generate materials for education

and public information. In the relation to the

Fukushima accident, we are planning to handle the

safety features of the plants with more concern.

Through the periodic workshops and the project of KYGN, we are trying to report the activities of KYGN and to deliver the correct information about the nuclear to the KYGN members and the public



.Safety and nuclear professionalism in Netherlands

Following the Chernobyl excursion of 2010 that was organized together with the Belgian Young Generation network (See the previous IYNC-Bulletin), the Dutch Young Generation organized a workshop on 17th of February about nuclear safety.

The day was divided in two parts. In the morning session several speakers gave presentations related to Chernobyl. Andre Versteegh (Jan Runemark Award winner of 2010) discussed the details about the disaster. He was involved in the investigations in and around Chernobyl right after the disaster in 1986. He was followed by a DYG member that joined the trip in 2010, describing the situation 25 years later.

In the afternoon the focus was on safety in general. The first talk was about safety awareness, both in the nuclear and other industries. The key message was that a large number of accidents can be prevented by simply being aware that accidents can happen. Most accidents occur because they were not recognized as possible hazards, and thus no precautions were taken.

The last part of the day was a workshop in which everybody was shown in an interesting and interactive way that as a professional in the nuclear industry (or any industry for that matter), you too contribute to a safe working environment. After showing events from various fields, from Apollo 13 to Chernobyl, the participants discussed what went right and what went wrong during these accidents.

What we all learned was that accidents can happen and that we should never lose focus on safety. Less than a month after the workshop the disaster at Fukushima happened, indicating once again that we should never take safety for granted. .

.

At the end of April Young Generation of Slovak

Nuclear Society (YGN SNUS) organized altogether

with mother nuclear society annual general

assembly with more than one hundred

participants. It is annual one day meeting divided

into formal and into informal session. During

formal session 6 presentations were given. Most of

the presentations concerning Fukushima accident

and consequences of accident. Fukushima accident

was detail analyzed from the technical, seismic and

geological point of view. During general assembly

also annual meeting of the Young Generation

section was held with more than 25 participants.

Following day we organized for all the participants

of general assembly, Technical Conference of YG.

This conference is usually held every 4 years.

Twelve interesting presentations with high quality

were prepared by our members. Three best of

them were awarded. Valuable finance prizes were

prepared by our mother society.

During summer the biggest musical festival called

“Pohoda” is going to be held near Trencin town

(130 km north-east from capital Bratislava) with

more than 30 000 people. Inspired by Hungarian

colleagues we are planning to explain nuclear

energy to public during festival at the information

point. Information posters, banners and leaflets

are preparing altogether with Universities in

Slovakia. The theme of posters is “Radiation

around us”. We hope that this acclivity will be

successful and young people attending musical

festival will gain some information about nuclear

energy and radiation. And especially after

Fukushima accident as only negative information

about nuclear energy were presented in media.



Spanish YGN, Jovenes Nucleares:. In the last months,

Spanish Young Generation Network has been focused

in education and training activities.

In April, Jovenes Nucleares organized its well known

Basic Course on Nuclear Science and Technology at

Zaragoza University with more than 200 attendants.

Another course, the Course on Nuclear Safety and

Radiological Protection was organized at Valladolid

University.

One of the main activities was the second edition of

the Seminar in Advanced Nuclear Reactors

(Generation III, III+ and IV), organized with the

Politecnica University of Madrid. This is a 15 hours

course, with different expert speakers from Jóvenes

Nucleares, explaining in an easy way the differences

between the different types of advanced reactors

(AP1000, EPR; ABWR, ESBWR, Generation IV) and

comparing them with the current reactor (PWR,

BWR). This seminar was a great success, with more

than 70 attendants and some interesting discussions

and debates after the lessons.

Other educational activities were the lectures in high

schools in different cities in Spain, explaining

teenagers between the age of 14 and 18 the basics of

nuclear energy, and also, lectures in the Spanish

Nuclear Society about different topics.

Figure 1. Seminar in Advanced Nuclear Reactors.

The Swedish way of Young Generation

2011 started off with a seminar in Västerås,

arranged by Westinghouse, at which the

participants from last year and the 60 new active

members of the present year could network and

listen to lectures. The second day a study visit to

Westinghouse Electric Sweden AB was arranged.

The participants of the ongoing year formed five

workgroups with different topics; each group

containing about 12 participants:

Waste cycle

Training and knowledge transfer

Organisations within the nuclear business

Safety

Development

The next seminar will be in Nyköping in October,

jointly arranged by KSU (Nuclear Training and

Safety Center) and Studsvik Nuclear. It will contain

one day of study visits and seminars and one day

where the participants themselves will present

different topics in several parallel sessions.

Also a delegation from UK Young Generation will

visit several nuclear sites in Sweden in October as

well as parts of the seminar.



The Nuclear Industry Association, which is the representative organisation for the UK’s civil nuclear industry, led the response to the tragic events in Japan for the Nuclear Institute and the YGN body. In the weeks following the event they issued three updates a day containing press releases, analysis, and data from reputable nuclear organisations and TEPCO to the UK nuclear external stakeholders group.

As a consequence of the events in Japan, the UK Government via the Secretary of State for Energy and Climate Change, requested the HM Chief Inspector of Nuclear Installations Inspectorate (the UK regulator), produce a report on the implications for the UK nuclear industry. The purpose of the report is to identify lessons to be learnt, taking forward this work in co-operation and co-ordination with national stakeholders, utilities, vendors and international colleagues and the regulator has said it is confident that the UK’s fleet of nuclear power reactors and operators

The response to the Fukushima Daiichi tsunami, and its consequent events, has been reported extensively via the media internationally. It is not the UK YGNs purpose to provide analysis or interpretation of the events. It acknowledges the significance of the events and directs our members to appropriate, reputable media coverage, lead industry representatives and organisations.

However, we have strong links the nuclear trade association, training providers, conference organisers, and various Institutions. Via its monthly e-bulletins to all its members, the UK YGN promotes a number of websites, journals, and events to allow the reader to be fully up to date with the UK, global and Japanese nuclear industry current position. The e-bulletins also point the reader in the direction of how other countries have responded to the events and what the UK have done in assisting its colleagues overseas.

.

The North American Young Generation in Nuclear (NA-

YGN) has had a powerful closeout to the 2010-2011

year, with the post Fukushima activities and turnover

that took place during the 2011 NA-YGN Professional

Development Conference. 250 NA-YGN members went

to Capitol Hill to meet with members of Congress and

discuss their nuclear energy concerns in the United

States.

The events in Japan required NA-YGN to adapt the

initial strategic plan to match the needs of members in

this challenging period. The professional development

committee hosted two special webinars with close to

1000 participants that made North American experts

available to answer questions for NA-YGN members on

the unfolding events at Fukushima. The goal of the

webinars was to support the nuclear community for

involvement in town hall meetings and in the media.

Prior to Fukushima the NA-YGN public information

committee had instituted a network of social media

interaction. This support enabled a fast media response

to Fukushima events, including Twitter, Facebook and

the Clean Energy Insights blog. These platforms were

used to communicate between NA-YGN members,

correct inaccuracies as they were reported and tweet

facts to the general public.

The successful NA-YGN response to the Fukushima

events has led to initiatives in the strategic plan for

2011-2012 year to include: An integrated

communication Emergency Response Plan for future

events and development of a Pro-Nuclear bloggers

interface.

Click below to link to NA-YGN Vice President Interview

on NEI Youtube: The Future of the US Nuclear Industry

Post Fukushima:

.



Nuclear education in South Africa

Due to the incident at the Fukushima nuclear power

plant earlier this year, the South African Young Nuclear

Professionals Society (SAYNPS) has an aim of

intensifying public education concerning the use of

nuclear technology. This is important because the

incident resulted in many South Africans voicing their

concerns and opposition to the proposed nuclear

investment in South Africa. Our target audience is

school learners.

SAYNPS members participated in the 2011 Cell-C (a cell

communications company) “take a girl child to work”

campaign, an annual campaign aimed at encouraging

young girls to pursue tertiary education and become

future professionals. It entails young school girls

spending a day with a person/company with the aim of

giving them exposure to the work environment and

what the job responsibilities are.

This year it was held on 26 May where more than

100 girls were invited to spend the day at the

South African Nuclear Energy Corporation and the

Koeberg Nuclear Power Plant where the girls were

educated about safe use of nuclear technology,

they received ‘first-hand’ experience on some

aspects of nuclear technology, through spending

the day with some of our members at their work

place. Our aim is to entice learners to consider

careers that will lead them to work in the nuclear

industry in future and also to demystify nuclear

technology amongst young people in South Africa.

The day’s proceedings resulted in many young

women considering studying engineering so they

have an opportunity to contribute to the nuclear

industry in future.



ENYGF11

Czech Young Generation (CYG) was established under Czech Nuclear Society on July 1997 with main goal to

attract and educate young engineers, technicians, scientists and other skilled personnel to work in the field

of nuclear energy. CYG aims to create and maintain strong contacts among young professionals in Czech

and European nuclear environment as well as to build bonds and intensive experience sharing between

senior and junior professionals.

Almost two years ago, CYG was happy to win hosting of ENYGF, biennial forum of European nuclear youth –

since that, local Organization Committee worked hard to prepare ENYGF2011, which took place in Prague in

the middle of May 2011.

The forum was started on May 17 with goal to provide independent and unbiased ground for discussion of

topics in seven tracks, which were voted by public poll. With respect to development on Fukushima site and

requests for more extraordinary topics, tracks were slightly enhanced and expanded; final list as presented

on Forum is given as follows:

• Nuclear Safety and Sever Accidents

• Education and Training

• Service and Operation of NPP

• New Builds

• Back End Issues

• ITER and Fusion

• Nuclear Medicine

Well recognized keynote speakers from both Czech Republic and abroad opened each track with invited

presentation, which was followed by presentation of young professionals. First track, which was focused on

safety and dealing with severe accidents was introduced by Dana Drábová, Head of the State Office for

Nuclear Safety. Second part of this track was introduced by Craig Marianno from Texas A&M University, a

member of the U.S. expert mission at Fukushima site after the accident in March 2011. Track on new builds

was opened by František Hezoučký from Faculty of Mechanical Engineering, Czech Technical University in

Prague. ITER and Fusion track was covered by Jan Mlynář from Institute of Plasma Physics, Academy of

Sciences of the Czech Republic. Education and Training track was addressed by Vladimír Slugeň, President

of ENS. Keynote speech in Back End Issues track was given by Philippe Bretault, senior process expert and

deputy of expertise & innovation E&P Direction, AREVA. Daneš Burket, director of Technical Support Unit in

Power Division of CEZ Group, who is responsible for the Long-term operation project of Dukovany NPP,

started track called Service and Operation of NPP.

http://www.ipp.cas.cz/



Each oral track was also accompanied by poster section,

where rest of young nuclear professional’s

accomplishments was presented.

Discussions on various displayed and presented topics

were often overrunning to sessions, technical trips and

social events.

Social events and technical trips were prepared to relive

young nuclear experts from often highly technical

presentations: four technical trips were available – it

was possible to visit Czech Rad-waste repository “Richard”, nuclear vendor Škoda JS, research institutes at

Řež (with two research reactors) and training reactor VR-1 with side visit to tokamak COMPASS.

Social events consisted of guided tour around city of Prague with after-party at Vagon club, dinner on the

boat at Vltava river and trip to Karlštejn castle.

Organizers were happy to host ENYGF2011 and would like to express their gratitude to co-organizing

associations and sponsors: ENS-YGN, CENEN, ENS, CTU in Prague, IYNC, AREVA, CEZ Group, E.ON Kernkraft,

GNS mbH, NRI Rez, VF a.s., Tractabel Engineering, Radioactive Waste Repository Authority, Enrichment

Technology, ENVINET and Škoda JS a.s.



Continental Reports

Asia by Kenta Horio

In Asia region, national-YGNs have already established and been actively working in 5 countries

- India

- Japan

- Korea

- Malaysia

- Sri Lanka

Currently, most of YGN activities are individually and independently conducted by each national-YGN, and

we have to do much more work to tie up them and to create regional network. But in other words, every

single step toward formulation of regional network will be essential for the future progress of YGN in Asia.

One of the important efforts is to hold Asian Nuclear Young Generation Forum (ANYGF). ANYGF is still

under preparation by Japan-YGN and will need support from other Asian YGNs for realization, but it will be

the first trial to create Asian regional young generation network.

For further promotion and enhancement of young generation network in Asia, we might encourage and

support establishment of national-YGNs in various countries. We are still missing national-YGNs in several

matured nuclear countries and areas (such as China and Taiwan) and in many emerging countries (such as

Indonesia, Jordan, Thailand, U.A.E., Vietnam etc.).

No notable conferences or Participations of YGN planned in recent past.



Europe by Andrei Goicea

Active Society: European Nuclear Society - Young Generation Network

Past events

- ENEF working groups

Think-tank was formed in European Nuclear Energy Forum on Nuclear Energy issues. Three working groups

worked upon :Transparency, Risks & Opportunities.

ENS-YGN represented at transparency and Opportunities WGs

- Atomic career event

Atomic careers days were planned in Europe (Brussels). ENS-YGN present with a booth, leaflets and

promotional items

- PIME 2011 in Brussels, Belgium

SCK, GDF-SUEZ and Urenco sponsored 12 YGN participants including 4 ENS-YGN. Participation was falicited

by media training

- ENYGF in Prague, Czech Republic

On 18, May 2011 ENYGF was planned in Prague, Czech Republic. ENYGF jointly organized by Czech Nuclear

Society and European Nuclear Society

Future events

- COP17 in Durban:

South Africa Conference of the Parties of United Nations Framework Convention on Climate Change

(UNFCCC):

ENS-YGN participated with

COP14 (Poznan, Poland)

COP15 (Coppenhagen, Denmark)

ENS-YGN 2010 PIME award finalist with COP activities

17th United Nations Framework Convention on Climate Change (COP 17)

Working Group to prepare new communication material

Objective : a YGN delegation of 10-20 people (including local representatives)


IYNC Bulletin 17 Section 6: Future Events

Future Events

Happy Birthday Italy – A celebration of young Europeans

7-11 July 2011

The year 2011 marks the celebration of 150 years since Italian unification. On this occasion the Piedmont

Region and the City of Turin plans to organize a series of cultural and educational events dedicated to the

young generation, concerning all fields of activities. Discussions will be held in the framework of

conferences and workshops, and will be structured around five major topics:

Entrepreneurship and Innovation;

Youth employment

Education

Youth Participation

EU.

Facilitated by actors and experts in the fields of entrepreneurship and innovation, employment,

education, youth participation and EU integration, Happy Birthday Italy will help to ensure that the

current generation of highly qualified and socially mobile young people knows how to use all means

available to them to reach their potential.

The event is co-organized in partnertship with ThinkYoung, a youth NGO concerned with lobby activities in

favour of young people’s rights and opportunities.

What is ThinkYoung?

ThinkYoung is the first think-tank concerned with young Europeans. Created in 2007, it is a platform to

provide opportunities for young Europeans to meet, talk, share their thoughts and experiences; as a think-

tank, ThinkYoung is committed to make young voices heard in the world of EU affairs. For more details

about Think Young or if you are interested to support their activity, please visit http://www.thinkyoung.eu/.

Dates of Happy Birthday Italy: 7th – 11th July 2011

Venue: Turin, Italy

http://www.thinkyoung.eu/



ENEN Training course on Fukushima Lessons Learned

25-26 July 2011

The European Nuclear Education Network in cooperation with the Institute for Safety and Reliability are co-organizing a two day seminar on the consequences of the Fukushima Accident and the lessons learned.

Over four months after the earthquake and tsunami disaster in Japan this course will provide a comprehensive summary of the knowledge acquired on the accident in the Fukushima Daiichi NPP: the sequence of events, the role of design, accident management provisions, human and organisational factors as contributors to the accident and its progression, the consequences for public health and the environment as well as the international impact on the use and the regulation of nuclear energy. Moreover the course will explain the lessons to be learned for improving the safety of nuclear power plants.

The courses are tailored to young academic professionals in nuclear organisations as well as to university graduates in engineering and science preparing for careers at nuclear utilities, vendors, suppliers, regulators, international organisations, expert and consultancy organisations. They are also well suited for nuclear re-education of engineers and scientists working in other fields.

The seminar will take place in Garching/Munich, between July 25-26 2011. The tuition fee is 700 Euro before June 27th and 840 Euro after. The fee includes VAT and covers lectures and course material.

For more information please visit our website

www.isar.tum.de/courses Ms. Milena Dziębaj ISaR Institute for Safety and Reliability Walther-Meißner-Straße 1 85748 Garching Germany Phone: +49 89 289 13911 Fax: +49 89 289 13949

WNA 2011 – “The Future of nuclear power – Now it’s down to us” The World Nuclear Association's 36th Annual Symposium will be held 14-16 September 2011 at Central

Hall Westminster in London.

The WNA Symposium is the nuclear industry's premier event - an annual conference attended by over 700

leaders and specialists from more than 30 countries. The Symposium brings together the nuclear industry

http://www.isar.tum.de/courses



and its major stakeholders to discuss topics ranging from the nuclear fuel market to the practicalities of

building new nuclear power plants.

The key questions to be addressed at the Symposium are:

What lessons can be learned from Fukushima, and by what measures can the nuclear industry

demonstrate that it will avoid such events in the future?

How does Fukushima shape the challenges facing nuclear communicators?

Is it still possible for Europe and North America to get onto the fast track of new nuclear build?

If and as countries proceed with their plans for new reactors, what needs to be done to maximize

efficiency in the new-build process while ensuring reliability and safety in new reactors?

As regulators and industry come to terms with Fukushima, is it possible to improve the

international legal and regulatory framework so as to help the industry to commit with

confidence to long-term nuclear projects?

Registration for the conference is opened on website: http://www.wna-symposium.org/, and the

registration fees are 945 GBP for WNA members and 1795 GBP for non WNA members.

NUTECH 2011 invites you to Poland

11-14 September 2011

NUTECH is the International conference on Development and

Application of Nuclear Technologies, hosted by Poland. The

Conference has a long tradition in Poland. It has been

organized periodically as a national symposium on application

of nuclear technology in the industry, medicine, agriculture,

and environment protection. The series of national meetings

started in 1960 and continued till 2005. The last Conference,

organized in 2008 by the Institute of Nuclear Chemistry and

Technology has already attained an international status. The

main goal of the Conference is to bring together scientists

working on nuclear technologies to discuss recent developments and applications. An important role of the

Conference is to confront the opinions of scientists and academics with the key representatives of national

and international regulatory bodies dealing with nuclear technology to plan further research fulfilling the

needs of our modern society.

http://www.wna-symposium.org/



Scientific Topics

nuclear technology in medical health care and biomedical research, nuclear technology in environment protection, earth sciences, and protection of cultural heritage, radiation protection, present status and future of nuclear energy, management of radioactive wastes, industrial applications, detection and prevention of illicit trafficking of explosives and nuclear materials, radiation measurements, data processing and acquisition, quality control and quality assurance in nuclear technologies.

The fee includes conference materials with book of abstracts, coffee breaks, lunches, and tickets for the

Welcome Reception and Conference Dinner: 350 Euro until 15 May 2011/450 Euro after 15 May 2011

For students the fee is 200 Euro until 15 May 2011/300 Euro after 15 May 2011

For more information and registration to the conference please visit

http://www.ftj.agh.edu.pl/~nutech2011/index.html

BULGARIAN NUCLEAR SOCIETY - NUCLEAR POWER FOR THE PEOPLE CONFERENCE “Strengthening of the Nuclear Scientific and Educational Center” “Nuclear power for the people” is a forth coming event organized by the Bulgarian Nuclear Society in cooperation with ENS, during November 8-11, 2011 at the Bansko ski resort, Bulgaria. The main topics of the conference include:

Education and training

Advances in nuclear technology

Research reactor applications

Safety culture and public relations

Nuclear power plant operation and plant life management

Reactor physics, fuel cycle and thermal-hydraulics

Safety analyses

Nuclear in medicine

Radiation and environment

NPP Decommissioning and waste management

Regulatory issues and legislation

During the conference, the organizers have announced a competition for young authors’ for Best Presentation: “The young authors who would like to take part in Competition for Best Presentation (first author under 35) must indicate this in Participation Form. The papers for the competition will be reviewed and evaluated by special Competition Commission. Six of the presented papers will be selected for oral presentation on the conference. The best three presentations elected by both, the participants and Competition Commission will be awarded. The rest papers could participate in the competition for best poster award.”

http://www.ftj.agh.edu.pl/~nutech2011/index.html



Deadline information:

Application of Participation form and Abstracts 10.09.2011

Notification of acceptance 10.10.2011

Competition Commission notification 10.10.2011

Full Papers submission 25.10.2011 Registration Fees: Registration fee is 300 EUR. Early, before 30.09.2011, Registration fee is 250 EUR. Accompanying person fee is 70 EUR. It covers Welcome reception, Gala dinner and excursion.

2nd China International Nuclear Symposium

20-22 October 2011

China represents a growing nuclear power and the country is making clear its position in the worldwide nuclear family by organizing the 2nd CINS. The first CINS, which took place in Beijing in November 2010, attracted over 200 leaders and specialists worldwide. CINS 2 will, yet again, bring together the senior decision makers of the nuclear industry and its major stakeholders to discuss the practicalities of a new nuclear generation. The conference is dedicated to nuclear leaders in China and worldwide. The conference, organized in partnership with WNA will take place in Hong Kong and will include a cultural visit of the Lantau island. In parallel with the conference, an exhibition is going to be arranged in the lobby of the venue – Sheraton Hotel.

For more information, please visit http://www.wna-symposium.org/china/index.html

http://www.wna-symposium.org/china/index.html


IYNC Bulletin 22 Section 7: Student Section

Student Section

Work Opportunities at the IAEA Professional Posts Professional staff members of the IAEA comprise a multicultural group of experts, who carry out the programmes of the IAEA. Candidates interested in professional posts with the IAEA are expected to have a university degree and appropriate prior work experience in their chosen profession. General Service Posts General Service staff members provide support in the administrative, technical and scientific areas. Recruitment for all General Service posts is on a local basis only, and candidates must legally reside in Austria at the time of application. Technical Cooperation Programme (TCP) Experts Recruited for a limited period only and ready to travel abroad, specialists in nuclear science, engineering and related subjects possess sophisticated scientific and technical knowledge, and are an integral part of the TCP. Junior Professional Officer (JPO) Programme The JPO programme is designed for young professionals who are below the age of 32 years and hold an advanced university degree. The IAEA currently offers a limited number of JPO assignments only for applicants from Member States who hold a JPO agreement with the IAEA. Waseda University in Tokyo has had students reach out

to Canadian institutions to give an outline of TEPCO’s

operations and policies regarding nuclear plant

accident management.

Fellowship Programme for Young Professional Women This programme supports young women pursuing an advanced degree, or having recently graduated, in gaining practical international work experience in nuclear technology and applications or in technical cooperation. The programme provides fellowships of six months duration at the Agency assisting in the normal activities of an appropriate technical division.

Internship Programme The IAEA accepts a limited number of interns each year who are studying toward a university degree or have recently received a degree. Internships normally range from one month to one year in duration, and provide an excellent opportunity to gain practical experience in an international environment.

Fukushima Awareness Efforts Students at MIT started a blog: www.mitnse.com. The sole purpose of this blog is to provide up to date information on the events at the F-D plant site in simple and not very technical language. The website also provides a portal for Internet users to post questions that are answered by MIT students. Likewise, University of Wisconsin’s students organized a forum, “Understanding the Nuclear Emergency in Japan” with Dr. Michael Corradini, an expert in environmental aspects of nuclear energy, as the keynote speaker.

Outreach Efforts

Students from Texas A & M University are very

involved in outreach to high school students and

Boy and Girls Scouts merit badge programs. Their

first counseling session will be on March 26th and

April 9th, and they hope to have about 100 scouts

during the semester. During this session, the

scouts learn about the basics of nuclear power and

tour the university’s many engineering labs.

In April, students at the University of Michigan

tutored middle school and high school students

from underprivileged backgrounds of Detroit over

a period of 8 sessions about the fundamentals of

nuclear physics and took them on lab tours.

http://www.mitnse.com/


IYNC Bulletin 23 Section 7: Student Section

EUTERP

The European Training and Education in Radiation

Protection (EUTERP) had their workshop from March

28th -30th in Ayia Napa, Cyprus. The title of the

workshop was “Radiation protection training in Europe

– the next steps”. The workshop gave an overview on

the implementation of training programs in the

different member states, discussed multi-lateral

arrangements that were recently set up, reported on

the state of the art of EUTERP and the ENETRAP II

project, and learnt from EFOMP about the experiences

in MPE training.

ANS Student Conference

In April, participants of the American Nuclear Society

(ANS) student chapters from across the globe attended

the Student Conference hosted by Georgia Tech in

Atlanta from April 14th-17th. During this conference,

students presented their research at technical sessions

and were given many opportunities to network with

professionals in the field. 160 technical papers were

submissioned in 22 divisions such as Neutronics,

Materials, Thermo hydraulics, Plasmas to name a few.

Previous years’ conference attendance was close to

700 with around 450 students and remaining

professionals. This year, Georgia Tech is drew a similar

attendance. Student posters and presentations were

judged by professionals and each division had at least

one award and the Nuclear Materials division gave out

4 awards and travel scholarships.

The next ANS Student Conference will be held in

University of Las Vegas in 2012.

JAEA

Japanese Atomic Energy Agency (JAEA) is looking

to recruit a senior postdoctoral fellow (fixed term

researcher) at the Integrated Support Center for

Nuclear Nonproliferation and Nuclear Security. The

major targets for survey and research include:

·Research for strengthening the framework of

nuclear nonproliferation in Asian countries aiming

at peaceful use of nuclear energy such as Vietnam

and Indonesia, and planning and implementation

for support programs

Survey and research on trends of the U.S., Russia,

Europe and the IAEA in this field Survey and research on trends and awareness of

Asian countries for which the Center aims to

provide training and seminar in this field

AESJ

The Atomic Energy Society of Japan (AESJ) has

hosted students from numerous countries by

offering them intensive research projects at a

number of universities located around Japan.

Research topics include a variety of subjects such

as the study of free-surface flow of Liquid metal

Lithium and Effect of Irradiation on Nuclear Fuel.

For more information about these exchange

opportunities, visit the AESJ website at

www.aesj.or.jp/index-e.html. The AESJ is involved

in expansive outreach to student communities

across Japan by providing numerous professional

and funding opportunities.

Other

University of California, Berkeley and the

University of Tokyo are organizing an advance

summer school of nuclear engineering and

management with social scientific literacy from

July 31st-August 5th 2011. The school is going to

focus on Fukushima accident and related social,

ethical and regulatory problems.

http://www.aesj.or.jp/index-e.html


IYNC Bulletin 24 Section 8: Technical Papers

Technical Papers

Finalist in Capetown IYNC2010

8.1. APPLICATION OF RETRAN-3D/SIMULATE-3K COUPLING SYSTEM FOR BWR

TRANSIENT ANALYSIS WITH BEST ESTIMATE PLUS UNCERTAINTY

8.2. THE BODEX IRRADIATION EXPERIMENT TO SIMULATE TRANSMUTATION OF

AMERICIUM.

8.3. HIGH-TEMPERATURE REACTOR FUEL ELEMENT CHARACTERIZATION WITH THE

KÜFA DEVICE.



Application of RETRAN-3D/SIMULATE-3K Coupling System for BWR Transient Analysis with Best Estimate plus Uncertainty Methodology

Daisuke Fujiwara, Shoichi Suehiro, Kenichiro Nozaki, Toshiki Okamoto, Akitoshi Hotta TEPCO Systems Corporation, 2-37-28 Eitai Koto-Ku Tokyo, Japan

ABSTRACT

A topic of the best-estimate plus uncertainty (BEPU) methodology attracts attention as the advanced approaches in

the safety analysis of nuclear power plants. The RETRAN-3D/SIMULATE-3K code system (R3D/S3K) has been

developed as the candidate best-estimate codes for the Anticipated Operational Occurrences (BWR/AOO) analysis for

the boiling water reactors (BWRs). R3D/S3K is the coupled code that consists of a plant thermal-hydraulic analysis

code and a three-dimensional core kinetics analysis code. In this presentation, the on-going internal project scheme

for the application of R3D/S3K to the BWR/AOO transient analysis with the BEPU methodology is introduced.

Subsequently, primarily important two subjects are focused on practicality in applying R3D/S3K to the BWR/AOO

transient analysis. The first is the validation for the prediction capability of the time-varying axial power shape (TVAPS)

during transients caused by the change of the moderator density distribution and the control rod insertion. The

advantage of R3D/S3K is the channel-per-fuel assembly spatial core model which enables accurate consideration of

the effect of the TVAPS to the system response such as the minimum critical power ratio (MCPR). The validation was

performed based on the three turbine trip test results conducted at Peach Bottom Unit 2. The second is the

implementation of the statistical method combined with R3S/S3K for evaluating the probabilistic density distribution

of the delta-CPR/ICPR during the AOO transient. Each uncertainty for the highly ranked parameters was statistically

combined based on the R3D/S3K runs with simultaneous random sampling of input parameters. Derivation of the 95-

th percentile of the delta-CPR/ICPR with the 95% confidence limit (95%/95% value) to be compared with the

acceptance criteria for AOO transients, such as the generator load rejection with no-bypass (LRNBP) and the loss of

feed-water heating (LFWH) is demonstrated tentatively using the existing Phenomena Identification and Ranking Table

(PIRT) that can be referred in the open literature.

2. RETRAN-3D/SIMULATE-3K COUPLING SYSTEM

The major advantage of R3D/S3K is inheritance of the existent plant modelling know-how and of the

verification and validation resource qualified for R3D. Another advantage exists in the adopted detail core

models that are equivalent to that of the static core simulator, SIMULATE-3. In S3K, a channel-per-fuel

assembly mapping between thermal-hydraulic and core neutronic regions is adopted with fine axial meshes

that enables to resolve the detailed power distribution during transients induced by the change of the

moderator density distribution and insertion of the control rod. R3D is a well-known transient plant

simulator developed by Computer Simulation Analysis (CSA), and EPRI (Electric Power Research Institute in

United States). The code has been extensively assessed based on separate effect tests and integral effect

tests including actual plant transient tests. The basis of the R3D thermal-hydraulics is the one-dimensional

homogeneous-mixture two-phase compressible flow model, solving the mixture mass, momentum, and

energy conservation equations with the drift-flux correlation. A five-equation model can be optionally

adopted, where the mass continuities of vapour and liquid-phase flow are separately modelled to allow the

analysis of non-equilibrium thermodynamic conditions. R3D is capable of modelling the primary control

systems and the reactor protection systems, which are required in the various BWR/AOO transient

analyses, by combining the general control block functions implemented in the code. S3K is a three-

dimensional coupled neutronic and thermal-hydraulic core transient analysis code developed by Studsvik



Scandpower (SSP). S3K primarily consists of the three-dimensional neutron kinetics model based on the

advanced two-group nodal diffusion theory, the one-dimensional cylindrical fuel heat conduction model,

and the one-dimensional parallel-channel five-equation thermal-hydraulics model. A feature of S3K is that

its neutronic model is fully consistent with that implemented in the static core simulator, SIMULATE-3.

Therefore, S3K is capable of predicting the core criticality and power distribution as accurately as

SIMULATE-3 does. Other numerical methods such as the assembly discontinuity factors, the nonlinear

iteration and the multi-dimensional table cross section library which enhance both accuracy and

computational efficiency under the framework of the 3D coarse mesh system are also common with

SIMULATE-3. R3D/S3K employs a so-called ‘external plenum coupling’ methodology. Here, R3D analyzes the

overall plant thermal-hydraulic responses with a simplified core model represented by the single channel or

by the grouped multiple channels. The core inlet (flow rate and temperature) and outlet (pressure)

boundary conditions are provided to S3K together with the core protection signals. On the other hand, the

coupled core neutronic and thermal-hydraulic responses are analyzed by S3K with the detailed mesh

discretization, accounting the spatial and temporal variations in the fuel temperature, the coolant void

fraction, and the control rod position. The obtained three-dimensional nodal core power responses are

mapped into the coarse-mesh system employed in R3D, which are returned to R3D for the overall plant

thermal-hydraulic calculation at the next time-step. In R3D/S3K, the neutron kinetics function in R3D is

bypassed. This coupling algorithm is schematically shown in Fig. 1.

SIMULATE-3K

Thee-dimensional core kinetics calculation

RETRAN-3D

Plant thermal-hydrauliccalculation

Core Thermal PowerAxial Power Distribution

Core Inlet FlowCore Inlet temperature

Core Exit PressureCore Protection Signals

Plant ModelCore Model

Channel-per-fuel geom.With 24 or 25 axial-nodes

Plant & Core T-H calculation

Coupled Neutronic & T-H calculation

1D coarse mesh core geom.

Figure 1: R3D/S3K coupling algorithm

3. VALIDATION FOR SPATIAL CORE TRANSIENT MODEL

PB2/TT tests conducted at different power and flow conditions were analyzed using R3D/S3K for validating

of the spatial core transient model that is an essential feature of this coupled code system. The PB2 is a GE

BWR/4 plant with the jet pump coolant recirculation system. The rated core thermal power was 3293 MWt,

and the core was fueled with 764 fuel assemblies, which consisted of the GE 7X7 and 8X8 fuels. Three



turbine trip (TT) tests were conducted at the end of the cycle 2 of PB2 in 1977. R3D/S3K was applied to the

PB2/TT test analyses so as to validate its simulation ability for the TVAPS during the fast BWR pressurization

transient. The validation was performed comparing the calculated and measured LPRM responses obtained

during PB2/TT tests. The TT test was initiated by closing the turbine stop valve (TSV), which induced a rapid

increase in the reactor pressure. The core neutron fission power started to increase by the positive

reactivity insertion due to the coolant void collapse. The turbine trip scram was intentionally delayed to

observe a significant increase of the neutron flux. The core power rise was actually terminated by the

APRM (Average Power Range Monitor) high scram. The turbine bypass valve (TBV) was opened just after

the TSV closure to mitigate the rapid increase in the reactor pressure from the viewpoint of the pressure

boundary integrity, while the most-limiting pressurization transient assumes inactivation of the turbine

bypass valve in the licensing safety analysis for BWRs. Forty-three LPRM tubes were arranged horizontally

in the core, and four neutron detector channels were equipped at the different axial levels of each LPRM

tube. The CASMO-4/SIMULATE-3 code system was applied on the core burn-up follow calculations to

reproduce the core state at each PB2/TT test. S3K was directly accessible to the restart file created by

SIMULATE-3, which contained all the state parameters to reproduce the SIMULATE-3 power distribution. All

the R3D and S3K input data were generated based on the test report, and the developed control system

model in R3D referred to the analysis report prepared by EPRI. R3D plant nodalization for the PB2/TT test

analysis is shown in Fig.4. Figure 5 shows the core averaged axial power distributions at the PB2/TT test

initial steady state condition comparing between the calculated distributions by S3K and the process

computer distributions. The good agreements demonstrate that the CASMO-4/SIMULATE-3 code system is

capable of reproducing the initial core state accurately. As shown in Fig.5, the initial power shape is top-

peak for the TT1, middle-peak for the TT2, and relatively flat for the TT3 witch cover a wide range of

operating conditions. Figure 6 shows the simulated horizontally-core-averaged LPRM responses for each

four axial channels in the TT1, the TT2 and the TT3, respectively. Each LPRM response is compared with

measurement data and it is recognized that R3D/S3K well reproduced the timing of power peaks caused by

the void collapse positive reactivity insertion and the subsequent scram negative reactivity insertion. The

calculated peak powers show good agreement for the TT2, however about +/-10% prediction errors are

observed for the TT1 (LPRM-C) and the TT3 (LPRM-C, D). In this coupling system, uncertainties in the three-

dimensional power shape and its time dependent variation has the paramount importance. It should be

understood that the detailed power distribution is influenced through complex physical processes

dominated by not only thermal-hydraulics but also the neutron physics including the cross section library.

To perform the comprehensive uncertainty analysis considering this huge bunch of uncertainty parameters,

it is necessary to develop elaborate and systematic approaches as studied in the UAM project. On the other

hand, in most of the AOO transient cases, the primary dominant uncertainty parameters from the view

point of the spatial dependence can be narrowed down to TVAPS. Therefore this uncertainty can be

appropriately treated by uncertainties of those macroscopic parameters dominating the axial power shape,

such as the void coefficient, the Doppler coefficient, the scram reactivity, etc. In fast transients such as the

turbine trip, the void coefficient undoubtedly is most influential factor. The following description

demonstrates the sensitivity analysis for the LPRM response to estimate the uncertainty range of the void

coefficient. The sensitivity analyses were performed with the perturbation of the void coefficient to cover

the observed prediction errors in the PB2/TT analysis. As shown in Fig. 7, it was confirmed that +/-10%

perturbations of the void coefficient properly cover the measured LPRM response peak values. The

obtained perturbation range of void coefficient is reasonable compared to the several benchmark results,



in witch the uncertainties of void coefficient for the UO2 pin cell or fuel assembly calculations ranges from

about 10% to 25%.

Core

Upper plenum

Stand pipe

Lower plenum

Separator

Steam Dome

Dow

nco

mm

er

Jet p

um

p

Dow

nco

mm

er

Jet p

um

p

Recirculationpump

Bypass

flow

Safety/Relief

flow

Controlvalve flow(Turbine)

Turbine stopvalve

Bypass valve

Relief valve

(1st MSIV)

(2nd MSIV) Safety valve

Feedwaterflow

Core

Byp

ass

valve

0

5

10

15

20

25

0.0 0.5 1.0 1.5 2.0

Relative Power

Axi

al N

ode

TT1(PROCESS COMPUTER)



TT1(R3D/S3K)

TT2(R3D/S3K)

TT3(R3D/S3K)

Figure 1: R3D/S3K plant nordalization Figure 2: Core averaged axial power

distributions at PB2/TT test initial condition

0

1

2

3

4

5

6

7

0.2 0.4 0.6 0.8 1.0 1.2 1.4 1.6 1.8

Time[s]

LPR

M R

esp

onse

(nor

mal

ized

to initia

l le

vel)

0

1

2

3

4

5

6

7

0.2 0.4 0.6 0.8 1.0 1.2 1.4 1.6 1.8

Time[s]

LPR

M R

esp

onse

(nor

mal

ized

to initia

l le

vel)

0

1

2

3

4

5

6

7

0.2 0.4 0.6 0.8 1.0 1.2 1.4 1.6 1.8

Time[s]

LP

RM

Resp

onse

(norm

aliz

ed

to in

itia

l leve

l)

LPRM-A(MEASUREMENT)

LPRM-B(MEASUREMENT)

LPRM-C(MEASUREMENT)

LPRM-D(MEASUREMENT)

LPRM-A(R3D/S3K)

LPRM-B(R3D/S3K)

LPRM-C(R3D/S3K)

LPRM-D(R3D/S3K)

BP2/TT1 BP2/TT2

BP2/TT3

Figure 3: Simulated averaged-LPRM responses at each axial level for PB2/TT test



0.00

2.00

4.00

6.00

8.00

10.00

TT1 TT2 TT3

PB2/TT test case

LP

RM

-D r

espo

nse

peak

val

ueR3D/S3K(NOMINAL)

R3D/S3K(VOID COEFFICIENT +10%)

R3D/S3K(VOID COEFFICIENT -10%)

MEASUREMENT

Figure 4: Void coefficient sensitivity of LPRM (D-channel) response peak value

4. STATISTICAL APPROACH FOR UNCERTAINTY ANALYSIS OF AOO TRANSIENTS

Another important aspect of the BEPU methodology is the statistical uncertainty analysis for evaluating a

probabilistic distribution of relevant safety variables such as delta-CPR/ICPR. The delta-CPR/ICPR is change

in the MCPR to the initial MCPR, which is a measure of the relative severity of the transient to calculate the

operating limit of MCPR (OLMCPR). The propagation of each parameter uncertainty to the delta-CPR/ICPR

was estimated by the statistical process based on the PDFs (medians and variances) defined for the highly

ranked parameters in the PIRT. Since both PIRT and PDFs for R3D/S3K are yet to be finalized, the statistical

uncertainty quantification of the delta-CPR/ICPR was demonstrated using the existing published PIRT and

tentatively defined PDFs. The PIRTs for the fast pressurization and the core subcooling increase transients

are referred from the opend literature. A combination of uncertainties was performed by the statistical

approaches based on the simultaneous random sampling of the statistical input parameters. Benefits of the

simultaneous random sampling approach are features of easy-to-implement and flexible-to-use, however

relatively large number of code runs are required to obtain reliable statistical values. The obtained

frequency distribution was statistically analysed with two types of statistical treatments such as the

parametric approach and the non-parametric approach, to establish the bounding value that corresponds

to a priori determined confidence level like the 95%/95% value. The parametric approach is based on the

upper one-sided tolerance limit (UOSTL) method that assumes the normality of the obtained frequency

distribution. To calculate the 95%/95% value, a factor k is multiplied to the obtained standard deviation and

then added to the mean value. For example, the multiplier k is defined as 1.91 in the case of 100 samples.

This approach is applied unless the population of the obtained data is rejected by the Chi-square goodness-

of-fit testing for the normality with the significance level of 5%. On the other hand, the non-parametric

approach normally applied is the order statistics based on Wilk’s theorem , which determines the minimum

sample size for the specified upper and/or lower tolerance limits. This approach does not assume the

normality of the obtained distribution and provide the relatively conservative uncertainty bound comparing

with the UOSTL. The uncertainty analyses of the delta-CPR/ICPR were performed for limiting transients of

typical BWR plants, such as the LRNBP and LFWH. The LRNBP is the fast pressurization transient initiated by

the turbine control valve rapid closure and the LFWH is the core subcooling increase transient initiated by

the loss of feedwater heating. Sensitivity analyses were also performed for the highly ranked parameters to

decompose the uncertainty of the delta-CPR/ICPR into several important influential factors. Figure 9 shows

the sensitivity analysis of the LRNBP transient for the one-sigma perturbation of each input parameters.



The horizontal axis represents the input parameters and the positive and negative perturbations are

applied, they are described as “+” and “–“, respectively. The vertical axis represents the difference of the

delta-CPR/ICPR caused by each perturbation. The sensitivity analyses indicate the dominant parameters to

the total uncertainty of the delta-CPR/ICPR. In the case of the LRNBP, “void coefficient”, “void ratio”,

“steam line pressure drop”, and “initial core pressure” is dominant to the total uncertainty. Figure 10 shows

the result of the uncertainty analysis of the LRNBP with 100 number of R3D/S3K runs. The left figure shows

the delta-CPR/ICPR calculated for each simultaneous random sampling of input parameters, and the right

figure shows the frequency histogram of the delta-CPR/ICPR comparing with the normal distribution. The

obtained histogram was not rejected by the Chi-square goodness-of-fit testing for the normality, thus the

UOSTL method was employed to evaluate the 95%/95% value. As a result, the 95%/95% value of the delta-

CPR/ICPR was 0.172 and the nominal value was 0.148. Figure 11 shows the sensitivity analysis of the LFWH

transient for the one-sigma perturbation of each input parameters. In the case of the LFWH, “initial core

power” is dominant to the total uncertainty. An increment of initial core power make earlier to reach the

scram set point of Thermal Power Monitor (TPM), which result in the lower delta-CPR/ICPR. The

sensitivities of the initial core power to the delta-CPR/ICPR are not linear between the positive and the

negative perturbation. The non-linear sensitivity of the dominant parameter distorts the frequency

histogram of the delta-CPR/ICPR with 100 number of R3D/S3K runs as shown in Fig.12. The obtained

histogram was rejected by the Chi-square goodness-of-fit testing for normality, and then the order statistics

method was employed. The 95%/95% value of the delta-CPR/ICPR was 0.070 and the nominal value was

0.065, which conclude the smaller uncertainty of the delta-CPR/ICPR compared to the LRNBP for its smaller

sensitivities to the void and pressure related parameters.

-0.015

-0.010

-0.005

0.000

0.005

0.010

Byp

ass

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w + -

Voi

d C

oef.

+ -

Dop

pler

Coe

f. + -

Scr

am R

eact

ivity

+ -

Voi

d R

atio

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Sub

cool

ed V

oid

+ -

Pel

let

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t C

ond.

+ -

Orifice

ΔP + -

Low

er T

ie P

late

ΔP + -

Spa

cer

ΔP + -

Upp

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late

ΔP + -

Dec

ay H

eat

+ -

Upp

er P

lenu

m V

oid.

+ -

PLR

Coa

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wn

+ -

Sep

. Car

ryun

der

+ -

Sep

. L/A

+ -

Sep

. ΔP + -

Ste

am L

ine

ΔP + -

Ste

am L

ine

Geo

m. + -

Critica

l Flo

w + -

Initia

l C

ore

Pow

er + -

Initia

l C

ore

Pre

ssur

e + -

Scr

am S

peed

+ -

Jet

Pum

p Δ

P + -

Sensi

tivi

ty t

o D

elta-

CPR/IC

PR

Figure 5: R3D/S3K sensitivity analysis for LRNBP (“Δp” represents the pressure drop)

0.05

0.10

0.15

0.20

0.25

0 20 40 60 80 100

Number of R3D/S3K Runs

Delta-C

PR

/IC

PR

Nominal

95%/95% value(UOSTL)

R3D/S3K runs

0

5

10

15

20

25

30

0.107 0.116 0.124 0.132 0.140 0.149 0.157 0.165 0.173

Delta-CPR/ICPR

Fre

quency

Uncertainty Analsis

Normal Distribution

Nominal: 0.148Standard Deviation: 0.014Average: 0.14695%/95% value(UOSTL): 0.172

Figure 6: R3D/S3K delta-MCPR distribution for LRNBP



-0.015

-0.010

-0.005

0.000

0.005

0.010

Byp

ass

Flo

w + -

Voi

d C

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+ -

Dop

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f. + -

Scr

am R

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+ -

Voi

d R

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+ -

Sub

cool

ed V

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+ -

Pel

let

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t C

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+ -

Orifice

ΔP + -

Low

er T

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late

ΔP + -

Spa

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ΔP + -

Upp

er T

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late

ΔP + -

Dec

ay H

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+ -

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lenu

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oid.

+ -

PLR

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wn

+ -

Sep

. Car

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+ -

Sep

. L/A

+ -

Sep

. ΔP + -

Ste

am L

ine

ΔP + -

Ste

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Geo

m. + -

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l Flo

w + -

Initia

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Pow

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Initia

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Pre

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e + -

Scr

am S

peed

+ -

Jet

Pum

p Δ

P + -

Sen

sitiv

ity t

o D

elta

-CPR/I

CPR

Figure 7: R3D/S3K sensitivity analysis for LFWH (“Δp” represents the pressure drop)

0.04

0.06

0.08

0.10

0 20 40 60 80 100

Number of R3D/S3K Runs

Delta-C

PR

/IC

PR

Nominal

95%/95% value(ORDER)

R3D/S3K runs

0

5

10

15

20

25

30

35

0.059 0.060 0.062 0.063 0.065 0.067 0.068 0.070

Delta-CPR/ICPR

Fre

quency

Uncertainty Analsis

Normal Distribution

Nominal: 0.065Standard Deviation: 0.0026Average: 0.06595%/95% value(ORDER): 0.070

Figure 8: R3D/S3K delta-MCPR distribution for LFWH

5. CONCLUSIONS

The project is introduced for the application of R3D/S3K to the BWR/AOO analysis with BEPU. The

establishment of PIRT and the determination of input parameter’s PDF for R3D/S3K are still underway and

they requires the further development of the systematic process to exclude the engineering judgment in

the ranking and/or quantification process. Subsequently, the primary important two issues are focused on

relating to employ R3D/S3K to the BEPU methodology. The first is the validation of the spatial core

transient model to predict the three dimensional power and flow variation during BWR/AOO transients.

The prediction capability of TVAPS was assessed with PB2/TT tests and it was concluded that R3D/S3K is

almost able to reproduce the LPRM response peak values occurring under the fast pressurization transient

with the prediction errors covered by reasonable void coefficient uncertainty. The second is the

implementation of statistical approaches for the uncertainty analysis of the delta-CPR/ICPR during the

BWR/AOO transients. The uncertainties of individual input parameters were combined to the probability

density distributions of the delta-CPR/ICPR with the simultaneous random input perturbations. The

95%/95% values of the delta-CPR/ICPR were evaluated with parametric and non-parametric statistics for

the LRNBP and LFWH transients. The shape of the obtained total uncertainty distributions were

decomposed with sensitivity analyses. The presented validation and implementation of statistical

approaches enhance the applicability of R3D/S3K to the advanced BWR/AOO licensing practice.



The BODEX Irradiation Experiment to Simulate Transmutation of Americium

S. Knol*, P.R. Hania, J.D. Bruin, A.V. Fedorov, E. de Visser - Týnová, K.O. Broekhaus, F.C. Klaassen

Nuclear Research and Consultancy Group, PO Box 25, NL-1755 ZG Petten, Netherlands;

*[email protected]

ABSTRACT

The burning of minor actinides in inert matrices in Accelerator Driven Systems is a promising way of reducing the long

term radiotoxicity of radioactive waste. Irradiation of these actinides in inert matrices results however in significant

production of helium which has its effect on the matrix in which for example the americium is embedded. Internal

pressures cause the matrix to swell which has its influence on fuel temperature and fuel-cladding interaction.

To study this swelling behavior without the complex influence of fission products, an irradiation experiment has been

designed where the americium has been replaced by boron. Boron produces a large amount of helium in a short

period of time due to the large cross-section for the 10B(n, α)7Li reaction. Molybdenum, magnesium oxide and yttria

stabilized zirconia are the three matrices studied as candidate matrix materials for americium transmutation in ADS

systems. Gas release and swelling behavior of the different matrices are studied at two different temperatures, 800°C

and 1200°C, resulting in 6 capsules.

The experiment is irradiated for 57 full power days in the High Flux Reactor in Petten. Post Irradiation Examinations

demonstrated that the molybdenum matrix has the lowest release which resulted in large cracks in the matrix. The

high porosity of the MgO matrix caused the largest gas release but large deformations were also measured.

1. INTRODUCTION

In high level nuclear waste the so-called minor actinides are the largest contributors of the long-term radiotoxicity of

the waste. These minor actinides are created during irradiation of uranium and plutonium and although relatively

small in amount, they require the most attention when handling the waste. Where most lighter fission products decay

in several decades to a maximum of several hundreds of years, some isotopes of the minor actinides have a decay

time up to several thousands of years. The general consensus is that there are two ways of disposing of these waste

products, long term storage and transmutation.

Storage focuses on isolating the waste from the human environment, usually in deep geological repositories, to let the

waste decay for a long period of time until it is of no harm to the biological environment anymore. Transmutation

focuses on transmuting the waste products in nuclear reactors. With the latter, the waste is treated by separating the

minor actinides and developing ways of putting them next or into the fuel of nuclear reactors. The current research on

transmutation is mainly focusing on two minor actinides, americium and curium.

The work described in this paper addresses one of the challenges related to the transmutation of americium in an

Accelerator Driven System (ADS), as part of the EUROTRANS FP6 project [1]. One way of realizing this is by

implementing the americium in an inert matrix material. This material is called inert for its transparency for neutron

radiation. The objective is to perform transmutation of the americium, while keeping the fission products in the matrix

which stays intact. For americium this matrix is subject to an additional problem of internal gas pressures, caused by a

large production of helium.



Looking at the entire decay chain, a total of 75% of the initial 241Am forms a helium atom. This occurs when 241Am

captures a neutron and turns into to 242

Am which decays via beta decay to 242

Cm which in turn is a strong α-emitter.

Due to the low solubility of helium this results in a significant internal gas pressure which the matrix material has to

withstand.

The response of the matrix to this helium production is and important factor in the design of fuel pins dedicated to

transmutation. Too much swelling may cause fuel-cladding interaction and possibly even cladding failure. To reduce

swelling it might be beneficial to stimulate the release of helium from the matrix. This however requires larger plenum

volumes and thus longer fuel pins.

To study He production in the matrix materials an alternative has been developed by substituting the americium by

boron. This element has a large cross section for a 10B(n, α)7Li reaction. By using 10B it is possible to study the changes

in structural integrity in a very short amount of time, and without the complex chemical interaction of a large number

of fission products. Additionally the absence of fissile material greatly simplifies the handling and increases the

possibilities in the Post Irradiation Examinations. Based on previous experiments [2] three different matrix materials

were selected; molybdenum, magnesium oxide (MgO) and yttria stabilized zirconia (YSZ), all irradiated at both 800°C

and 1200°C.

2. FABRICATION

The BODEX irradiation experiment consists of two legs with in total six capsules. Both legs contained all three different

matrix materials but operated at different temperatures, e.g. 800 and 1200°C. Each capsule contained 5 pellets with

matrix material: one without boron as a reference for irradiation damage to the matrix, one with 11

B as a reference to

the chemical interaction between boron and the matrix material without helium production, and three pellets with 10B. All pellets with a boron compound were mixed in such a way that they contain approximately 1.5 mmol 10B per

cm3 pellet material. Important requirements for the B-doped pellet are:

The boron compound does not chemically interact with the inert matrix material, and can be sintered

together with the inert matrix. During irradiation the boron is gradually converted into lithium therefore the inert

matrix should be compatible with the lithium compound as well.

As the produced alpha particles (1,47 or 1,78 MeV) have a typical mean free path of 1-5 µm, a micro-

dispersed boron compound is used to study specifically the properties of the inert matrix under the production of

He gas. By successive sieving, milling and acetone sedimentation steps the boron-containing particles are sieved

to a particle size < 5-10 m.

In order to fabricate pellets of the matrix materials with sufficient strength, integrity and required porosity, high

calcination temperatures are required in the order of 1500-1600°C. Additionally, sintering during irradiation must be

avoided for this would change pellet dimensions and as a consequence temperatures during irradiation. This means

that during fabrication, the pellets must have been fully sintered.

Pellets of inert materials containing a specific B-containing compound were pressed, calcined at 1600°C for 6 hours

and reground for analysis. The amount of B-containing compound amounted to 5wt%, sufficient for detection by

means of X-ray diffraction (typically 1-2 wt%). The obtained X-ray diffractograms were compared with the

diffractograms of the non-calcined mixtures to check if the ratio of inert matrix and B-compound had changed and if

other reaction products were formed. Pellets were visually inspected by means of light-microscopy to check on

possible melting behavior.



3. EXPERIMENT DESIGN & IRRADIATION

The BODEX experiment was designed and manufactured by NRG and irradiated in the High Flux Reactor (HFR) in

Petten, the Netherlands for 57 full power days. The goal of the BODEX experiment is to study the effect of an intensive

helium gas production on the properties of various matrix materials at different temperatures. For this purpose, a

sample holder has been designed with two legs, which are placed in the two channels of a rig that is suitable for

positioning the experiment in the Pool Side Facilities (PSF) of the HFR. The PSF is a facility next to the core, which has

the possibility of moving the experiment from and to the core, controlling the neutron flux and temperature during

irradiation. Aditionally, temperatures are controlled by the design of appropriate gas gaps in the experiment between

the capsule, shroud and containment tube. These gas gaps are filled with a mixture of helium and neon.

Each leg (second containment) contains three capsules (first containment). These capsules contain pellets of the

different matrix materials. For monitoring the required parameters, two pressure transducers and twelve

thermocouples (two per capsule) are used. Additionally, 16 neutron fluence-monitoring sets are embedded in the

Shroud for neutron dosimetry. A schematic view of both sample holder legs with capsules is presented in Figure 9.

Figure 9: Schematic view of BODEX, left is the bottom of the experiment, right the top. The online instrumentation consisted of Thermocouples (TC) and Pressure Transducers (PT)

Figure 10: Schematic view of the capsules. The location of the five pellets with the different boron compounds is indicated.

Each capsule contains 5 pellets (Ø 5 mm x 5 mm) of one matrix material, kept apart by 6 spacers for positioning the

pellets and increased heating purposes. From bottom to top, three pellets are doped with a 10B-compound, one pellet

is doped with an 11

B-compound and one pellet consists of the matrix material only. These last two types of pellets in

each capsule are for reference purposes. In Figure 10, the position of the pellets in each capsule, including the spacer

discs between the pellets is shown. Each capsule is filled in the same order. The volume surrounding the pellets and

spacers is filled with neon gas to keep a clear distinction between filling gas and the helium produced by irradiation

The pellets in the warm leg of the sample holder reached an irradiation temperature of approximately 1200°C and in

the cold leg approximately 800°C. In this way the influence of the temperature on the helium gas production and

release is addressed. This difference in temperature design has some consequences for the fabrication of the legs as

can be seen in Table 3-1. The most important difference is the different materials for the spacers and the shroud,

which is a large tube around the capsules designed to realize the right size of the gas gaps and to contain the

instrumentation. The influence of the different materials is described in the Post Irradiation Examinations.



Table ‎3-1: Experimental specifications of BODEX.

Warm leg Cold leg

Pellet temperatures 1200°C 800°C

Gas filling at BOI Ne at 1 atm Ne at 1 atm

Structural materials molybdenum and tungsten Mostly molybdenum

Targeted He production at EOI 0.29 mmol 0.27 mmol

Maximum He production at EOI 0.44 mmol 0.44 mmol

Burn up target of 10

B at EOI 65 % 62 %

ONLINE PRESSURE MEASUREMENTS

In Figure 11 the amount of helium released in the gas line of the middle and top capsules are compared to the

calculated amounts of helium produced in the pellets due to the nuclear reaction. These calculations are performed

with the MCNP code [3]. The MCNP calculations are based on the actual fluence rates determined by the fluence

detectors located in the experiment. The relative difference between calculated and measured helium gives the

fractions of the helium released from the pellets. Relative increase of the released fractions in the beginning of

irradiation is explained by little trapping inside the pellets. As the build up of irradiation induced defects, e.g. helium

bubbles, proceeds the released fractions decrease in both matrices. After about 30 days of irradiation the trapping

inside the pellets saturates and the released fractions reach the steady state values: 35 ± 3% for MgO (middle capsule)

and 9 ± 1% for Mo (top capsule).

Figure 11: Amount of measured (red line) helium released in the gas line of the middle (left) and top (right)

capsules are compared to the helium produced in the pellets calculated with the MCNP code (blue line).

4. POST IRRADIATION EXAMINATION

NEUTRON FLUENCE ANALYSIS

Ten flux monitor sets were imbedded in the shroud, which are used to determine the received fluence. With the

received fluence, the boron burnup (or helium production) is determined by MCNP calculations and the FISPACT code

(Table ‎4-1). Additional to these calculations the amount of released helium is measured for each capsule, two by

online measurements presented before and two by puncturing of the capsules. Puncturing of the last two capsules

0 10 20 30 40 50 6 10 17

10 18

10 19

10 20

10 21

Time / days

BODEX-top

0 10 20 30 40 50 60 10 17

10 18

10 19

10 20

10 21

Time / days

He / atoms BODEX-mid He / atoms



failed due to mechanical reasons. The difference between the two numbers is the amount of helium trapped in the

samples.

Table ‎4-1: Calculated 10

B burnup and release based on the measured fluence and helium for each capsule.

Capsule 10B burnup He Produced He measured release Measuring method

height [mm] [%] mmol mmol %

Mo-800 -134 65.8 % 0.277 - -

Mo-1200 37 62.1 % 0.261 0.023 9.0 ± 1 Online measurement

MgO-800 -54 70.0 % 0.323 0.104 32.2 ± 3 Puncturing

MgO-1200 -54 63.0 % 0.296 0.104 35.0 ± 3 Online measurement

YSZ-800 37 68.4 % 0.300 - -

YSZ-1200 -134 58.6 % 0.256 0.068 26.7 ± 3 Puncturing

SWELLING BEHAVIOR

Dimensions of each pellet have been measured before and after irradiation to assess the swelling behavior. Volumes

are also determined by helium pycnometry. Comparing the geometric and pycnometric volumes give insight in the

development of the open porosity of the samples.

The molybdenum samples show a clear difference in geometric swelling between the three 10

B doped samples and the

two reference samples (Figure 13). Small but consistent differences are visible between the samples irradiated at

800°C. The geometric volume change of the 10B doped samples is limited to 2%, as well as the volume measured with

pycnometry, meaning almost no change in open porosity. The difference between radial and axial swelling was very

only very slightly anisotropic, most likely due to a slight pressure caused by the spring to position the samples.

Much larger swelling can be seen in the samples from the warm leg. The reference samples show identical behavior to the samples irradiated at 800°C but the

10B doped samples have increased significantly. In the samples irradiated at

1200°C, the geometric volume shows changes up to 9%.

800 °C 1200 °C

Figure 12: Two molybdenum 10B samples.

Before (top row) and after (bottom row) irradiation.

Figure 13: Volumes of molybdenum samples. No pycnometry measurements

have been done on the last 10B samples

Volume change Molybdenum

-2,0

0,0

2,0

4,0

6,0

8,0

10,0

No Boron B-11 B-10 B-10 B-10 No Boron B-11 B-10 B-10 B-10

Relative change (%)

Geom. Volume Pycn. Volume

800 °C 1200 °C



Please note that the pictures have been taken under different circumstances, therefore no conclusions on dimensions

can be drawn from the pictures. The swelling for all Mo samples is slightly anisotropic since the increase in height is

always lower than the increase in diameter (results not presented here). For these samples the porosity has increased

significantly as can be seen in the pycnometric measured volumes. Most likely this is explained by the cracks that are

visible in the irradiated warm samples (Figure 12) and contribute to the porosity measurements.

For the MgO samples, the swelling of the cold samples is much larger compared to that of the molybdenum samples.

Swelling over 10% is measured, although a large spread is visible in the volumetric swelling between the samples

(Figure 15). Also interesting is that the swelling is almost completely due to radial swelling, most samples even showed

a decrease in axial direction. This large anisotropy is visible in all samples, although more pronounced in the 10B

samples. The cold samples show a large increase in geometric volume, but very limited increase in pycnometric

volume. This indicates a noticeable increase in porosity.

800°C 1200°C

Figure 14: Two MgO 10B samples.


Figure 15: Volumes of MgO samples. No

pycnometry measurements have been done on the last 10B samples

Some problems occurred while handling the MgO samples since some of them appeared to be very fragile. During the

PIE examinations several samples were found with broken-off discs or grit in the containers used to keep the samples

separated. The warm samples appear to show quite stable behavior in volume, but this is for a large part caused by

the anisotropic deformations. Where the radius has increased significantly, the height has decreased just as much,

cancelling each other in the volume calculation. Most likely the high porosity of the samples (~25%) causes the

structural weakness, which additionally causes a number of samples to be very fragile.

Volume change MgO

-15,0

-10,0

-5,0

0,0

5,0

10,0

15,0


Relative change (%) Geom. Volume Pycn. Volume

800 °C 1200 °C



800 °C 1200 °C

Figure 16: Two YSZ 10B samples.


Figure 17: Volumes of YSZ samples. No

pycnometry measurements have been done on the last 10B samples

A clear difference can be seen in the volume change of the YSZ samples due to the irradiation temperature. The

samples in the warm leg had increased in size more then than the cold samples, 4% versus 2%, respectively. The

reference samples show almost no difference between the two temperatures, indicating the irradiation temperature

does not influence the matrix itself. Similar to the MgO samples, but in much lesser extend, the YSZ samples showed

signs of brittleness. This can be observed in Figure 16 where the irradiated pellets show some rough edges.

The YSZ samples (Figure 16) all show limited swelling. Especially the cold samples have not changed a large amount in

height or diameter, with the 11B doped-reference samples even showing the largest change. All cold samples have

swollen less than 1% in one direction. The increase of the warm samples is slightly larger for the 10B doped samples,

whereas the reference samples show similar behavior for both irradiation temperatures.

5. CONCLUSIONS

The irradiation of the BODEX experiment was successful and the burnup reached the desired value within 10%,

resulting in a helium production of approximately 1 mmol/cm3.

The post irradiation examinations showed that molybdenum has the advantage that a large amount of the helium gas

is retained in the matrix, with only a release of 9% at high temperatures. This has the advantage that a smaller plenum

is required when designing fuel pins. Consequently however a large volumetric swelling is observed at high

temperature. Although cracks are also visible at the 800°C irradiated samples the swelling is still limited. The matrix

however is not able to deal with the larger internal pressures present at 1200°C, based on the significantly larger

amount of cracks and the swelling of 9%. The samples irradiated at 800°C have swollen only up to 2%, but due to loss

of the gas pressure the helium release could not be studied.

The MgO matrix was fabricated with a high open porosity, which stimulated the helium release from the matrix and

was expected to reduce the swelling. Up to 35% of the helium was released from the matrix but still large

deformations have been measured. Although the swelling seems limited looking at the volume only, it is very

anisotropic. Practically all samples have decreased in height and increased in diameter significantly, which could cause

Volume change YSZ

-5,0 -4,0 -3,0 -2,0 -1,0 0,0 1,0 2,0 3,0 4,0 5,0


Relative change (%) Geom. Volume Pycn. Volume

800 °C 1200 °C



problems for cladding design. The high initial porosity apparently did have an effect on the strength of the matrix,

making the material very soft and the swelling unpredictable. Additionally some difficulties in the PIE occurred by

broken of pieces.

From a swelling point of view YSZ shows the most stable behavior during irradiation. The swelling is limited to 3-4% for

the warm samples and slightly less for the cold samples, in the order of ~2%. Although the porosity is comparable to

the porosity of the molybdenum 3-5%, the release is much larger, in the order of 27%. Disadvantages of the YSZ matrix

are the brittleness, which is comparable to the MgO matrix, complicating reliable measurements, and the high

activity. This activity made PIE in a glovebox complicated with the extra safety precautions.

REFERENCES

[1] EUROTRANS: EUROpean Research Programme for the TRANSmutation of High Level Nuclear Waste in an

Accelerator Driven System, 6th Framework Programme EURATOM, Management of Radioactive Waste,

Contract No. FI6W-CT-2004-516520.

[2] R.J.M.‎ Konings,‎ et‎ al.,‎ “The‎ EFTTRA-T4‎ experiment‎ on‎ americium‎ transmutation.”,‎ Journal‎ of‎ Nuclear‎

Materials, 282.2-3(2000): 159-70

[3] S.C. van der Marck and C.M. Sciolla,‎“BODEX‎(368-01)‎Nuclear‎Analysis”,‎NRG‎Report‎2.1668/07.86021/I,‎

Revision B, Petten, December 2007.



High-Temperature Reactor Fuel Element Characterization with the Küfa Device

A. I. Kellerbauer, P. D. W. Bottomley, D. Freis1, V. V. Rondinella, P. Van Uffelen, European Commission, Joint Research Centre, Institute for Transuranium Elements, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen, Germany.

ABSTRACT

Modern fuel elements for high-temperature reactors (HTRs) contain a large number of spherical fuel

particles embedded in three layers of coating materials. These are designed to ensure mechanical stability

and retention of fission products under normal and transient conditions, regardless of the radiation damage

sustained in-pile. In hypothetical depressurization and loss-of-forced-circulation (D-LOFC) accidents, fuel

elements of modular HTRs are exposed to temperatures several hundred degrees higher than during

normal operation, causing increased thermo-mechanical stress on the coating layers. At ITU, a vigorous

experimental program is being pursued with the aim of characterizing the performance of irradiated HTR

fuel under such accident conditions. A cold finger device (Küfa), operational in ITU’s hot cells since 2006,

has been used to perform heating experiments on eight irradiated HTR fuel pebbles from the AVR

experimental reactor and from dedicated irradiation campaigns at the High-Flux Reactor in Petten, The

Netherlands. Gaseous fission products are collected in a cryogenic charcoal trap, while volatiles are plated

out on a water-cooled condensate plate. A quantitative measurement of the release is obtained by gamma

spectroscopy. In this paper, we present an overview of experimental results from the Küfa testing as well as

the on-going development of new experimental facilities.

1. INTRODUCTION

Gas-cooled high-temperature reactors (HTRs) feature the remarkable property of being inherently safe

against power excursions even under adverse conditions. With a combination of low power density, high

thermal mass and high surface-to-power ratio, the core of an HTR can withstand certain accident scenarios

without significant mechanical failure and release of fission products. As a result of a hypothetical

depressurization and loss of forced circulation (D-LOFC) accident, all active heat sinks are removed. Despite

the fact that decay heat is carried away from the core only by means of natural heat transfer mechanisms,

the core temperature rises very slowly over the course of many hours before reaching a plateau several

hundred degrees above the operating temperature (see, for instance, Reference [1]).

1 Present address: Westinghouse Electric GmbH, Dudenstr. 44, 68167 Mannheim, Germany



The amount of fission products released under such accident conditions depends to a large extent on the

quality of the fuel elements (FEs). Modern HTR fuel contains TRISO (tri-structural isotropic) coated particles

(CPs) consisting of fuel kernels covered with a porous carbon buffer, an inner layer of pyrolytic carbon, a

layer of ceramic silicon carbide and finally an outer pyrolytic carbon layer. Together, the coatings provide a

barrier against the release of fission products both in normal operation and under off-normal conditions. In

order to provide well-qualified data for reactor design and licensing, the performance of HTR FEs is tested

experimentally by subjecting them to conditions that simulate a D-LOFC accident. The release of fission

products as a function of time (and hence, of temperature) provides insight into the underlying physical

and chemical mechanisms for CP failure and fission product transport. Both single CPs and entire FEs can be

studied. In the latter case, a large number of particles is effectively investigated at the same time, providing

a result with high statistical relevance.

Some 25 years ago, a method to expose a spherical HTR FE to a simulated D-LOFC accident and to

simultaneously collect all released fission products was developed at Kernforschungsanlage Jülich (KFA),

today Forschungszentrum Jülich (FZJ). The Küfa (an abbreviation of the German “Kühlfingerapparatur”) is

an experimental device in which spherical HTR FEs can be heated to up to 1800°C at ambient pressure in a

helium atmosphere while the release of fission gases and volatile fission products is measured. The original

apparatus was taken into operation at KFA in 1984 [2]. When the German HTR program was ended, the

Küfa at KFA was decommissioned. In the framework of the European HTR-F project, parts of the Küfa as

Figure 18: Schematic overview of the Küfa device at ITU.



well as drawings and technical documents were transferred to ITU in 2001 [3]. The new Küfa was taken into

operation at ITU’s Hot Cells unit in 2005 [4]. Since then, it has been used to perform simulated accident

testing on eight spherical FEs. Two similar instruments have been built at Idaho National Laboratory (INL)

and Oak Ridge National Laboratory (ORNL) [5, 6], but the ITU Küfa is currently the only device worldwide

for the accident testing of irradiated HTR fuel.

2. KÜFA EXPERIMENTAL FACILITY

An overview sketch of the Küfa is shown in Figure 1. Its main components are a bell-shaped outer stainless-

steel vessel, a gas conduction cylinder encompassed by a heating element, and a mobile cold finger. The

heater is mounted in a water-cooled copper block and surrounded by heat shields. The Küfa was designed

as a metallic resistive furnace; a power transformer installed behind the hot cell delivers the necessary high

current at low voltage. All high-temperature parts of the Küfa consist of tantalum because this material is

the least prone to carbidization from contact with the graphite FE. The pebble is placed in a three-point

bearing at the lower end of the gas conduction cylinder. A thermocouple measures the temperature about

3 mm below the FE and delivers the control variable for the regulation of the furnace power. The furnace

temperature as well as all operating processes and parameters are controlled centrally by a programmable

logic controller and an associated computer program. All operating parameters are collected and recorded.

At the lower end of the cold finger, facing the FE, an exchangeable stainless-steel plate is positioned.

Thanks to efficient water cooling, the surface temperature of the plate itself is kept at less than 100°C even

at the maximum furnace temperature of 1800°C. In this way, volatile fission products such as 137Cs and 110Agm plate out on the condensate plate with high efficiency. The cold finger can be pulled up through a

system of locks containing a water-cooled gate valve, and the condensate plate can be exchanged semi-

automatically using a telemanipulator. The used plates are then transported to a laboratory with low

radiation background, where a gamma spectrum is recorded with a high-purity germanium detector to

quantify the gamma-emitting isotopes [4]. During the experiment the device is flushed with helium. The

helium enters the gas conduction cylinder below the FE, flows past it, over the condensate plate and into

the bell. From there it is evacuated by a membrane pump to an activated-carbon cold trap cooled with

liquid nitrogen. Fission gases condense in the activated carbon, and their activity can be measured by a

sodium iodide gamma detector mounted below the cold trap [7].

3. RECENT EXPERIMENTAL RESULTS

As part of the European Integrated Project RAPHAEL (Reactor for process heat, hydrogen and electricity

generation), five HTR FEs of type Nukem GLE-4.2 were irradiated in the High-Flux Reactor (HFR) in Petten,

The Netherlands. The irradiation campaign HFR-EU1bis lasted for 10 cycles and a total of 249 equivalent full

power days. The aim of the experiment was to test the FE behavior at increased temperature as well as

increased burn-up and fast-neutron flux, resulting in increased radiation-induced damage to the coating

layers [8]. The target burn-up was 15.4% fissions per initial metal atom (FIMA); however, only slightly more

than 11% FIMA was actually achieved in the FE with the highest burn-up. The central temperature of the

FEs was maintained at 1250°C during the entire irradiation campaign, except for a brief temperature

excursion due to an operator error involving the gas mixing system in the irradiation rig [9]. The irradiation

ended on 18 October 2005. After irradiation, four of the five FEs were transferred to ITU in order to be

heated in Küfa accident simulation tests, while extensive post-irradiation examinations were carried out on

the fifth at the Nuclear Research and Consultancy Group (NRG) in Petten [10].



Between March 2008 and November 2009, Küfa accident tests were performed on all four HFR-EU1bis

pebbles transported to ITU. In the following, the result of the Küfa heating experiment of pebble

HFR-EU1bis/1 will be presented as a representative example. It was placed in one of the upper irradiation

positions and therefore exhibited a somewhat lower burn-up of 9.3% FIMA and a lower fast fluence of

2.4×1021 cm−2 as compared with the FEs irradiated in the central positions. The Küfa heating experiment

began on 28 March 2008. Due to the short cooling time since the end of irradiation (less than three years),

the relatively short-lived fission products 110Agm and 106Ru were still present. Therefore, it was possible to

also measure their release in addition to the comparatively long-lived cesium isotopes. While 110Agm is not

particularly relevant to accident scenarios, it is of the utmost interest for reactor operation at increased

temperatures due to its high volatility. This is particularly important for future HTR variants with direct

cycles where gas turbine contamination can be substantial.

The heating test began with a long phase (200 h) of simulated operation at 1250°C to determine the

equilibrium release of 110Agm during operation and to give an indication of the possible contamination of a

future HTR’s primary circuit. The cesium contamination of the graphite matrix and thus potential particle

failure during irradiation was also assessed in this way. After the simulated operation, an accident

simulation up to 1600°C (corresponding to the nominal peak fuel temperature obtained from thermal-

hydraulic calculations [1]) was conducted during 200 h. This phase was followed by an additional heating to

1700°C during 150 h. In-between each of the heating phases the FE was cooled to room temperature. A

graph of the temperature profile and the cumulative fractional releases of all measured fission products is

shown in Figure 2. The cumulative fractional releases of gaseous and volatile fission products are

summarized in Table 1.

Table 2: Summary of the results of the Küfa heating test with pebble HFR-EU1bis/1.

Heating

temperature

(°C)

Heating

duration

(h)

Fractional release

85Kr 137Cs 134Cs 110Agm 106Ru

1250 210 1.4×10−6 7.2×10−5 7.0×10−5 2.3×10−3 6.4×10−7

1600 200 1.0×10−5 1.2×10−3 1.2×10−3 7.2×10−3 —

1700 150 2.4×10−5 4.3×10−3 4.3×10−3 — —

The 85Kr equilibrium fractional release during the simulated operation at 1250°C amounted to 1.4×10−6.

During the second phase at 1600°C a fractional release value of 1.0×10−5 was measured, whereas during the

third phase at 1700°C the release increased to 2.4×10−5. This corresponds to the inventory of about 0.2

particles. Already during the first 200 h of simulated operation at 1250°C, a marked fractional release of 137Cs and 134Cs on the order of about 7×10−5 was found. During the second phase, the accident simulation at

1600°C, the fractional release of Cs increased to 1.2×10−3. During the third phase at 1700°C the release of

Cs increased less strongly than expected, so that finally a cumulative fractional release of 4.3×10−3 was

obtained.



110Agm was already strongly released during the first heating phase, with a fractional release up to 2.3×10−3.

During the 1600°C phase this value increased to 7.2×10−3. It was, however, not possible to reliably measure

the release during the third phase at 1700°C. Due to the strong release of 137Cs and 134Cs it was not possible

to bag out the condensate plates and they instead had to be measured through the window of a shielded

glove box near the hot cells. Because of the high background in that location the gamma activity of 110Agm

could not be detected. In the graph the cumulative detection limits and the corresponding uncertainties are

plotted as conservative limits. It was only possible to measure the release of 106Ru during the first phase, up

to a fractional release of 6.4×10−7. For the second and third phases, the detection limits and corresponding

error bars were again used in the graph, as for 110Agm.

4. DISCUSSION

The release curve of 85Kr doesn’t show any obvious jumps either at 1250°C or during the accident

simulation. Furthermore, the overall low release of 85Kr, well below the inventory of a single CP, indicates

that no particles failed during the experiment. It merely represents the expected release from free uranium

contamination in the graphite matrix. The most striking observation is the relatively high release of cesium,

about two to three orders of magnitude higher than observed in the HFR-K6 pebbles [4]. This can be

explained either by increased diffusion through intact particle layers, by an external contamination of the

graphite matrix, or both. We carried out a detailed analysis based on a mechanical-failure model as well as

a diffusion simulation to investigate the origin of the observed metallic fission product release. This yielded

a fractional release of 110Agm of the order of 2×10−4 during the irradiation phase, about an order of

magnitude below the measured value. Apparently the leading causes for this large release fraction are the

high irradiation temperature of 1250°C and the known high volatility of silver. The calculated release of 137Cs and 134Cs by diffusion from intact particles is around 3×10−6, at the level of the yield from matrix

uranium contamination. The higher release during irradiation would therefore have to originate from

defective particles. Both for silver and for cesium the calculation actually overestimates the actual release

measured during the accident simulation phases.

Core

Upper plenum

Stand pipe

Lower plenum

Separator

Steam Dome

Dow

ncom

mer

Jet pum

p

Dow

ncom

mer

Jet pum

p

Recirculationpump

Bypass

flow

Safety/Relief

flow

Controlvalve flow(Turbine)

Turbine stopvalve

Bypass valve

Relief valve

(1st MSIV)

(2nd MSIV) Safety valve

Feedwaterflow

Core Bypass

valve

0

5

10

15

20

25

0.0 0.5 1.0 1.5 2.0

Relative Power

Axia

l N

ode




TT1(R3D/S3K)

TT2(R3D/S3K)

TT3(R3D/S3K)

FIGURE 19: Temperature profile and fractional fission product release for fuel element HFR EU1bis/1.



The model is able to reproduce the experimental data reasonably well under the assumption that the

graphite matrix of the FE was contaminated with the equivalent inventory of about 20 CPs, or 2.1×10−3 of

the 9560 CPs contained in the pebble. This may have been caused by an increased irradiation temperature

resulting in an enhanced diffusion through the intact coatings of the particle. It is also possible that an

external contamination of the graphite matrix occurred during irradiation stemming from failed particles in

this pebble or one of the other FEs irradiated as part of experiment HFR-EU1bis. During one of the

irradiation cycles, the FE capsule was subjected to an unexpectedly high temperature excursion. This may

have led to the failure of a number of CPs, whose inventory was then released into the graphite matrix. This

hypothesis is also supported by the leveling-off of the release curve for cesium during the third heating

phase. Possibly the fission product “reservoirs” in the outer particle layers and the graphite matrix were

depleted, and diffusion through the SiC layer could not replenish them quickly enough.

5. CONCLUSION

Diffusion calculations suggest that the high release of Ag and Cs from HFR-EU1bis fuel elements at relatively

low temperatures was due to a contamination of their graphite matrix with substantial amounts of fission

products prior to the Küfa accident simulations. We therefore conclude that the irradiation temperature of

TRISO fuel should remain well below 1250°C to prevent fission product release at a level which is too high

for inherently safe HTR concepts. Assuming this prerequisite is fulfilled, earlier accident simulations have

shown that HTRs may withstand accident temperatures ≥ 1600°C without losing their integrity [4]. Clearly,

additional Küfa accident simulation experiments with FEs irradiated under carefully controlled conditions

are required to fully characterize their response to off-normal conditions. The irradiation of five more HTR

fuel pebbles of German and Chinese production has recently been completed within the HFR-EU1

irradiation campaign. It is planned that they will be subjected to Küfa heating tests at ITU as part of a

follow-up project to RAPHAEL.

6. OUTLOOK

In addition to further Küfa measurements, a direct method for the quantification of the fraction of failed

particles in a heated FE would be desirable. For this purpose, the graphite matrix of the FE can be dissolved

and the CPs extracted (deconsolidation). Subsequently the condition of individual particles can be

determined by measuring the inventories of fission products with different volatility. Besides indicating the

number of defective particles, this can provide information on the release behavior of different fission

products inside the CP and in the matrix. Such a device was first developed and applied under the name

IMGA (Irradiated-microsphere gamma analyser) at KFA [11], followed by a similar facility at ORNL [12].

Recently, the deconsolidation of a FE and the separation of its CPs have been successfully performed at ITU.

A newly developed IMGA device is currently being tested with activated dummy particles. Another

important aspect that has yet to be addressed is the hypothetical ingress of air or water during a D-LOFC

accident. Earlier studies have shown that the ability of FEs to resist such an accident is very limited [13].

Therefore further experimental work on this problem is required. For this purpose, a new corrosion device

Kora (from German “Korrosionsapparatur”), which was initially developed at KFA Jülich together with the

Küfa, is currently being constructed at ITU and is being prepared for use in a hot cell.



7. ACKNOWLEDGMENTS

This work was supported by the European Commission within the 6th Euratom Framework Program under

Contract No. 516508 (RAPHAEL Integrated Project).

8. REFERENCES

[1] H. Haque, W. Feltes and G. Brinkmann, Nucl. Eng. Design 236 (2006) 475

[2] W. Schenk, D. Pitzer and H. Nabielek, Jülich Rep. 2234, Jülich, Germany, 1988

[3] E. H. Toscano et al., 2nd International Topical Meeting on High Temperature Reactor Technology,

Beijing, China, 2004, p. B13

[4] D. Freis et al., J. Eng. Gas Turbines Power 132 (2010) 042901

[5] D. Petti et al., JOM Journal of the Minerals, Metals and Materials Society 62 (2010) 62

[6] K. Verfondern (ed.), IAEA-Tecdoc-978, 1997, p. 174

[7] K. A. Stradal, Jülich Rep. 707-RW, Jülich, Germany, 1970

[8] M. A. Fütterer et al., Nucl. Eng. Design 238 (2008) 2877

[9] S. de Groot et al., Nucl. Eng. Design 238 (2008) 3114

[10] S. de Groot et al., 4th International Topical Meeting on High Temperature Reactor Technology,

Washington (DC), USA, 2008, vol. 1, p. 307

[11] M. J. Kania and C. A. Baldwin, Experts Meeting on MHTGR/HTR Fuel Performance under Accident

Conditions, Oak Ridge National Laboratory, 1989

[12] C. A. Baldwin and M. J. Kania, Oak Ridge National Laboratory Technical Report ORNL/TM-11455,

1990

[13] W. Schenk, R. Gontard and H. Nabielek, Jülich Rep. 3373, Jülich, Germany, 1997

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