fuel behaviour and modelling under severe transient and loss of coolant accident

398 Pages·2013·11.97 MB·English

FUEL BEHAVIOUR AND MODELLING
UNDER SEVERE TRANSIENT AND LOSS OF
COOLANT ACCIDENT (LOCA) CONDITIONS
The following States are Members of the International Atomic Energy Agency:
AFGHANISTAN GUATEMALA PANAMA
ALBANIA HAITI PAPUA NEW GUINEA
ALGERIA HOLY SEE PARAGUAY
ANGOLA HONDURAS PERU
ARGENTINA HUNGARY
PHILIPPINES
ARMENIA ICELAND
POLAND
AUSTRALIA INDIA
PORTUGAL
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QATAR
AZERBAIJAN IRAN, ISLAMIC REPUBLIC OF
BAHRAIN IRAQ REPUBLIC OF MOLDOVA
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COLOMBIA LUXEMBOURG
CONGO MADAGASCAR TAJIKISTAN
COSTA RICA MALAWI THAILAND
CÔTE D’IVOIRE MALAYSIA THE FORMER YUGOSLAV
CROATIA MALI REPUBLIC OF MACEDONIA
CUBA MALTA TOGO
CYPRUS MARSHALL ISLANDS TRINIDAD AND TOBAGO
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TUNISIA
DEMOCRATIC REPUBLIC MAURITIUS
TURKEY
OF THE CONGO MEXICO
UGANDA
DENMARK MONACO
UKRAINE
DOMINICA MONGOLIA
DOMINICAN REPUBLIC MONTENEGRO UNITED ARAB EMIRATES
ECUADOR MOROCCO UNITED KINGDOM OF
EGYPT MOZAMBIQUE GREAT BRITAIN AND
EL SALVADOR MYANMAR NORTHERN IRELAND
ERITREA NAMIBIA UNITED REPUBLIC
ESTONIA NEPAL OF TANZANIA
ETHIOPIA NETHERLANDS
UNITED STATES OF AMERICA
FIJI NEW ZEALAND
URUGUAY
FINLAND NICARAGUA
UZBEKISTAN
FRANCE NIGER
VENEZUELA
GABON NIGERIA
VIETNAM
GEORGIA NORWAY
GERMANY OMAN YEMEN
GHANA PAKISTAN ZAMBIA
GREECE PALAU ZIMBABWE
The Agency’s Statute was approved on 23 October 1956 by the Conference on the Statute of the
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IAEA-tEcdoc-cd-1709
FUEL BEHAVIoUR ANd ModELLING
UNdER SEVERE tRANSIENt ANd LoSS oF
cooLANt AccIdENt (LocA) coNdItIoNS
PRocEEdINGS oF A tEcHNIcAL MEEtING HELd IN
MIto, jAPAN, 18–21 octoBER 2011
INtERNAtIoNAL AtoMIc ENERGy AGENcy
VIENNA, 2013
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IAEA Library Cataloguing in Publication Data
Fuel behaviour and modelling under severe transient and loss
of coolant accident (LOCA) conditions : proceedings of a
technical meeting held in Mito, Japan, 18-21
October 2011. – Vienna : International Atomic Energy Agency, 2013.
CD-ROM. – (IAEA-TECDOC-CD series, ISSN 1684-2073
; no. 1709)
ISBN 978-92-0-192410-0
Includes bibliographical references.
1. Nuclear reactors – Research – Congresses. 2. Nuclear fuel
elements – Mathematical models. 3. Nuclear fuels – Safety
measures – Congresses. I. International Atomic Energy Agency.
II. Series.
IAEAL 13-00812
foreword
In recent years the demands on ‘fuel duties’ have increased, including transient regimes, higher burnups and
longer fuel cycles. To satisfy these demands, fuel vendors have developed and introduced new cladding and fuel
material designs to provide sufficient margins for safe operation of the fuel components. National and international
experimental programmes have been launched, and models have been developed or adapted to take into account
the changed conditions. These developments enable water cooled reactors, which contribute about 95% of the
nuclear power in the world today, to operate safely under all operating conditions; moreover, even under severe
transient or accident conditions, such as reactivity initiated accidents (RIAs) or loss of coolant accidents (LOCAs),
the behaviour of the fuel can be adequately predicted and the consequences of such events can be safely contained.
In 2010 the IAEA Technical Working Group on Fuel Performance and Technology (TWGFPT) recommended that
a technical meeting on “Fuel Behaviour and Modelling under Severe Transient and LOCA Conditions” be held in
Japan. The accident at the Fukushima Daiichi nuclear power plant in March 2011 highlighted the need to address
this subject, and despite the difficult situation in Japan at the time, the recommended plan was confirmed, and the
Japan Atomic Energy Agency (JAEA) hosted the technical meeting in Mito, Ibaraki Prefecture, Japan, from 18 to
21 October 2011.
This meeting was the eighth in a series of IAEA meetings, which reflects Member States’ continuing interest in
the above issues. The previous meetings were held in 1980 (jointly with OECD Nuclear Energy Agency, Helsinki,
Finland), 1983 (Risø, Denmark), 1986 (Vienna, Austria), 1988 (Preston, United Kingdom), 1992 (Pembroke,
Canada), 1995 (Dimitrovgrad, Russian Federation) and 2001 (Halden, Norway).
The purpose of the technical meeting was to provide a forum for international experts to review the current
For further information on this publication, please contact:
situation and the state of the art of the performance of nuclear fuel for water cooled reactors under severe transients
and LOCA conditions. The meeting was attended by 83 specialists representing fuel vendors, nuclear utilities,
Nuclear Fuel Cycle and Materials Section
research and development institutions, and regulatory authorities from 19 Member States. The papers submitted to
International Atomic Energy Agency
the meeting were organized into seven sessions covering analytical and experimental RIA and LOCA studies and
Vienna International Centre
international programmes, power ramp, and severe accident analysis. These proceedings contain all the papers that
PO Box 100
were presented and discussed during the meeting, and highlight key findings and recommendations based on the
1400 Vienna, Austria
summaries of the session chairpersons. While the Fukushima Daiichi accident influenced the discussions, it was
Email: [email protected]
not directly considered because of the lack of fuel behaviour data available at the time of the technical meeting.
The IAEA wishes to thank the hosts and all participants for their contributions to the technical meeting, in particular
© IAEA, 2013
K. Kamimura (JNES) and F. Nagase (JAEA). The IAEA officer responsible for this publication was V. Inozemtsev
Printed by the IAEA in Austria
of the Division of Nuclear Fuel Cycle and Waste Technology.
MAY 2013
EDITORIAL NOTE
This CD-ROM has been prepared from the original material as submitted by contributors. Neither the IAEA nor its Member States
assume any responsibility for the information contained on this CD ROM.
This publication (including the figures, tables and references) has undergone only the minimum copy editing considered necessary for
the reader’s assistance.
The use of particular designations of countries or territories does not imply any judgement by the publisher, the IAEA, as to the legal
status of such countries or territories, of their authorities and institutions or of the delimitation of their boundaries.
The mention of names of specific companies or products (whether or not indicated as registered) does not imply any intention to
infringe proprietary rights, nor should it be construed as an endorsement or recommendation on the part of the IAEA.
The authors are responsible for having obtained the necessary permission for the IAEA to reproduce, translate or use material from
sources already protected by copyrights.
.
CONTENTS

SUMMARY ............................................................................................................................... 1
R&D PROGRAMME ON FUEL BEHAVIOUR UNDER TRANSIENTS (Session 1)
Fuel safety research at JAEA ................................................................................................................ 17
F. Nagase
IRSN R&D studies on fuel behaviour under RIA conditions ............................................................... 25
M. Petit, O. Marchand, F. Barré, P. Giordano
EXPERIMENTAL RIA STUDIES (Session 2)
Status of RIA related fuel safety research at JAEA .............................................................................. 45
T. Sugiyama, H. Sasajima, M. Amaya, F. Nagase
New techniques for the testing of cladding material under RIA conditions.......................................... 55
H.K. Yueh, R.J. Comstock, V. Grigoriev, J. Karlsson, W. Lees, Y.P. Lin, D. Lutz,
D. Mitchell, D. Schrire
Current status of RIA study in the Republic of Korea .......................................................................... 67
S.-K. Kim, J.-G. Bang, D.-H. Kim, Y.-S. Yang, Y.-H. Koo
The cladding fracture behaviour under biaxial stress conditions .......................................................... 77
T. Fukuda, T. Mihara, M. Amaya, T. Sugiyama, F. Nagase
Fracture behaviour of hydrided cladding tubes with radial incipient crack in periphery ...................... 85
T. Mihara, T. Fukuda, Y. Udagawa, T. Sugiyama, F. Nagase
ANALYTICAL RIA STUDIES (Session 3)
Recent RIA and LOCA analyses performed at Technical Research Centre Finland (VTT)
using fuel performance codes SCANAIR and FRAPTRAN-GENFLO ............................................ 97
A. Arffman
Assessment of steady state uncertainties impact on RIA modelling ................................................... 117
A. Huguet, L.E. Herranz
RCCA ejection fault studies: Application of a burnup threshold for the evaluation of the
radiological releases ........................................................................................................................ 131
C. Bernaudat, J. Guion, N. Waeckela
Present status of the verifications and model development of FEMAXI and RANNS
codes in JAEA ................................................................................................................................. 141
M. Suzuki, Y. Udagawa, T. Sugiyama, F. Nagase
RCCA ejection fault studies: Application of a burnup threshold for the evaluation of the
radiological releases ........................................................................................................................ 153
Y. Udagawa, T. Sugiyama, M. Suzuki, F. Nagase
LOCA R&D (Session 4)
LOCA: An operator’s view on German licensing practice ................................................................. 163
A. Wensauer
IRSN R&D studies on fuel behaviour under LOCA conditions ......................................................... 179
M. Petit, C. Grandjean, N. Trégourès, F. Barré, P. Giordano
Investigation of the RU-43LV fuel behaviour under LOCA conditions in CANDU reactor .............. 191
G. Horhoianu
EXPERIMENTAL LOCA STUDIES (Session 5)
High burnup fuel behaviour under LOCA conditions as observed in Halden reactor
experiments ..................................................................................................................................... 207
E. Kolstad, W. Wiesenack, B. Oberländer, T. Tverberg
Results of integral, high burnup, fuelled LOCA tests and companion testing with
as-fabricated and pre-hydrided cladding ......................................................................................... 219
P. Askeljung, M. Billone, M. Flanagan
Behaviour of high burnup fuel during LOCA; key observations and test plan at JAEA .................... 229
F. Nagase, T. Chuto
UJP LOCA oxidation criteria “K” and “O ” ....................................................................................... 241
Β
M. Négyesi, L. Novotný, J. Kabátová, S. Linhart, V. Klouček, J. Lorinčík,
V. Vrtílkováb
Quantification of the chemical elements partitioning within pre-hydrided Zircaloy-4
after high temperature steam oxidation as a function of the final cooling scenario
(LOCA conditions) and consequences on the (local) materials hardening ..................................... 253
J.C. Brachet, D. Hamon, J.L. Béchade, P. Forget, C. Toffolon-Masclet, C. Raepsaet,
J.P. Mardon, B. Sebbari
ANALYTICAL LOCA STUDIES (Session 6)
Evaluation and analysis of high burnup fuel behaviour under LOCA conditions ............................... 269
Y. Yun, M.A. Zimmermann
The RAPTA-5.2: Code for modelling of VVER type fuel rod behavior under design
basis accidents conditions ............................................................................................................... 279
P.V. Fedotov, A.A. Goncharov, A.V. Kumachev, O.A. Nechaeva, V.V. Novikov,
A.V. Salatov, Y.V. Pimenov, Y.N. Dolgov
Assessment of fuel behaviour under large break LOCA condition for Indian pressurized
heavy water reactor ......................................................................................................................... 289
A. Dixit, S.K. Yadav, N. Kumar, T.A. Khan, M. Singhal, N. Mohan, P.K. Malhotra,
S.G. Ghadge, U. Chandra
Matare coupled codes package validation and assessment of potential clad ballooning behaviour
for limiting ATWS and LBLOCA fault sequences in support of development of the
SIZEWELL ‘B’ PWR safety case with an increased fuel internal pressure limit and
M5TM clad fuel ................................................................................................................................. 301
R. Page, S. Jenkins
Radiological consequences for CHASHMA NPP Unit 2 due to large break LOCA without
ECCS and CSS ................................................................................................................................ 327
S.A.A. Rizvi, M.A. Ahmed, Q. Hasnain, A. Minhaj
POWER RAMP AND SEVERE ACCIDENT (Session 7)
Status of power transient test program on LWR fuels using JMTR .................................................... 349
J. Ogiyanagi, S. Hanawa, F. Nagase
Fission gas release from high burnup fuel during normal and power ramp conditions ...................... 357
M. Amaya, J. Nakamura, F. Nagase
Investigation of VVER 1000 core degradation during SBO accident scenario in case of
pressurizer SV stuck in open position ............................................................................................. 367
P. Groudev, A. Stefanova, R.Gencheva
List of Abbreviations ........................................................................................................................... 379
List of Participants .............................................................................................................................. 383
SUMMARY

SESSION 1: R&D PROGRAMME ON FUEL BEHAVIOUR UNDER TRANSIENTS

Chairpersons: T. Fuketa (JAEA) and M. Flanagan (NRC)

1. BACKGROUND

The presentations of this session highlight research activities at two of the major nuclear test reactor
facilities, CABRI reactor and the Nuclear Safety Research Reactor (NSRR), operating in the world
today.

2. SUMMARIES AND COMMENTS

F. Nagase of the Japan Atomic Energy Agency (JAEA) in Japan presented an overview of the fuel
safety research activity at the JAEA. F. Nagase explained that JAEA’s research activities span
reactivity initiated accidents (RIAs), loss of coolant accidents (LOCAs), fuel behaviour code
development and verification, as well as fundamental studies of specific phenomena. The JAEA
studies in the Nuclear Safety Research Reactor (NSRR), simulating RIA events, have produced a
fundamental understanding of pellet cladding mechanical interaction (PCMI) failure for RIA events
and semi-integral quench tests have confirmed the adequacy of the Japanese LOCA regulatory limits
for the high burnup of fuel. Fundamental studies on oxidation behaviour and crack propagation in
hydrided cladding have been critical for evaluating and interpreting the results of both the LOCA and
RIA studies. The knowledge from the transient testing and the fundamental studies has been
successfully integrated into fuel behaviour codes to produce improved modelling capabilities. In the
future, JAEA research programs will continue to use these demonstrated experimental approaches to
study advanced cladding and fuel materials. Recently, a new emphasis has emerged on fuel and
cladding response in severe accidents and beyond design basis accidents.

M. Petit of the Institut de Radioprotection et de Sûreté Nucléaire (IRSN) in France presented an
overview of research being conducted at IRSN to study fuel behaviour under RIA conditions. M. Petit
presented that the experiments in the sodium loop of the CABRI reactor led to a fundamental
understanding of the PCMI failure mode of RIA transients and the findings indicated a need to revise
RIA criteria. Now, a new program will be conducted in a water loop in the CABRI reactor which will
illuminate both the PCMI and departure from nucleate boiling (DNB) phase of a postulated RIA
transient. A test matrix was presented for the new program with a total of 10 tests. Two tests in sodium
have already been completed. Many of these planned tests are specifically designed to complement the
experimental databases of the CABRI sodium loop tests and the NSRR RIA tests.

3. PROBLEMS, CHALLENGES AND PERSPECTIVES

Even with these 12 planned tests, it is clear some aspects of fuel behaviour under RIA conditions will
still remain largely unaddressed.

Despite many years of research and operating experience of light water reactors, research questions
still remain in order to characterize the fuel and cladding behaviour in response to accident conditions
in a detailed way. At the same time, there are fewer and fewer facilities capable of simulating RIA,
LOCA and transient conditions. Combining test reactor studies with fundamental studies further
increases the contribution of test reactor findings to the understanding of fuel and cladding response to
1
transient conditions. With many countries facing decreased financial resources, international
collaboration is becoming increasingly necessary to run large scale test programs and fundamental
understanding of the complex phenomena of nuclear fuel and cladding is necessary to direct the
limited resources to generate the maximum benefit. The JAEA and IRSN research programs provide
critical coordination of large scale test programs and fundamental research which has resolved, and
will continue to resolve critical questions for safe operation of nuclear power plants.

4. RECOMMENDATIONS FOR FUTURE WORK

An extension in the test program is therefore proposed to address aspects such as the performance of
fuel as intermediate burnups, where large uncertainty and disagreement between expert interpretation
remains.
1. Encouragement of partnerships and synthesis of research program findings from various
international programs.
2. Combining test reactor studies with fundamental studies further increases the contribution of
test reactor findings to the understanding of fuel and cladding response to transient
conditions and should be supported.

SESSION 2: EXPERIMENTAL RIA STUDIES

Chairpersons: M. Petit (IRSN) and T. Sugiyama (JAEA)

1. BACKGROUND

The behaviour of nuclear reactor fuels during off-normal and postulated accident conditions such as
RIA are a matter of investigations. In order to define adequate criteria for fuel design effects and for
burnup related effects, the fuel response to RIA transients is being studied experimentally and
analytically. Considerable experimental evidence has been gathered in recent years covering a variety
of materials and conditions.

2. SUMMARIES AND COMMENTS

In the session 2, entitled “Experimental RIA Studies”, 5 papers were presented.

The first paper by T. Sugiyama, JAEA, Japan, showed an overview of the 14 RIA experiments
performed in the NSRR reactor during the Advanced LWR Fuel Performance and Safety (ALPS)
programme on high burnup light water reactor (LWR) fuel. Both low and high initial temperature tests
were conducted. The paper discussed the outcome of the experiments in terms of PCMI failure
criterion. Also addressed was the influence of the initial temperature on PCMI failure. In conclusion
the fuels to be tested in the phase 2 of ALPS (that were received in January 2011) were presented.

The second paper by K. Yueh, Electric Power Research Institute (EPRI), United States of America
(USA), discussed the results of 2 new mechanical tests performed by Studsvik, Sweden on behalf of
EPRI. The first type of tests on Zr-2 with radial hydrides — called rapid heating and load test —
showed that even brief exposure at high temperature enhances the ductility of the hydrided cladding
samples. There appears to be an abrupt recovery between 70–100°C, comparable to a brittle to ductile
transition, followed by a more gradual ductility improvement when temperature increases. The second
type of test, called modified burst test, consists of a loading of a cladding sample by the use of a tube
2