Research Reactor Core Conversion Guidebook Vol 3 [Analytical Verification] (IAEA TECDOC-643v3)

356 Pages·1992·15.472 MB·English

IAEA-TECDOC-643
Research reactor
core conversion guidebook
Volum: 3Aen alytical verification
(Appendices G and H)
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INTERNATIONAL ATOMIC ENERGY AGENCY /A\
RESEARCH REACTOR CORE CONVERSION GUIDEBOOK
VOLUM: 3AE NALYTICAL VERIFICATION (AP)HPE DNNDAIC EGS
IAEA, VIENNA, 1992
IAEA-TECDOC-643
ISSN 1011-4289
Printed by the IAEA in Austria
April 1992
FOREWORD
n I vie het fwop roliferation concerns c fao ueshus eihet gydbh ly
enriched uranium (HEU) and in anticipation that the supply of HEU to research
d natest reactors wi elblm ore restre ihctft unetid ure, this guidebosoahk
been preparo etad ssist research reactor operaton rias ddresse ihnstga fety
and licensing issues for conversion of their reactor cores from the use of HEU
w oel nfor iescu hefheut deot l uranium (LEU) fuel.
o wTprevious guideboon kors esearch reactor core conversion have been
published by the IAEA. The first guidebook (IAEA-TECDOC-233) addressed
feasibility studies and fuel development potential for light-water-moderated
research e hsrteec aodcntdno arg suidebook (IAEA-TECDOC-324) addressed these
topics for heavy-water-moderated research reactors. This guidebook, in five
volumes, addresses the effects of changes in the safety-related parameters of
mixed cores and the converted core. It provides an information base which
should enable het appropriate approvals processer osf implementatiao fon
specific conversion proposal, wha ea thr olrhefoieaf grvr hoyt water
moderated research reactor, to be greatly facilitated.
This guidebook has been prepared and coordinated by the International
Atomic Energy Agency, with contributions volunteered by different
organizations. The IAEA is grateful for these contributions and thanks the
experts from the various organizations for preparing the detailed
investigations and for evaluating and summarizing the results.
EDITORIAL NOTE
In preparing this materia ehtp lrof ress, stae hft fofI nternational Atomic Energy Agency have
moud npntaeagd inae othertdi ginal manuscris psatus be amhutit td yhtgenboidarv se n some attention
e prhestetn otation.
The views expressed in the papers, the statements made and the general style adopted are the
re enshpat onmotnee sncdifeb osiae slvouiathity rehdiwTloy sr sr .ee fglheocvtt e rtfnhomofseoe nt s
the Member States or organizations under whose auspices the manuscripts were produced.
nthii se bsouTohke of particular designatf iocoonus nr trtoeiers ritorieyst n oidmanopelys
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 specific companies or of their products or brand names does not imply any
endorsemer nort ecome hItmA foEe nAnad.pa ehtti onon
Authorse rat hemselves responsibler of obtaininge htn ecessary permissiono t reproduce copyright
material from other sources.
This text was compiled before the unification of Germany in October 1990. Therefore the names
German Democratic Republic and Federal Republic of Germany have been retained.
PLEASE BE AWARE THAT
ALL OF THE MISSING PAGES IN THIS DOCUMENT
WERE ORIGINALLY BLANK
PREFACE
Volume3 consistsf o AppendixG which contains detailed resulta sfo
safety-related benchmark pron abi lrdoeefma lized reactd noAar ppendiH x which
contains detailed comparisons of calculated and measured data for actual cores
with MEU and LEU fuels. The results of the benchmark calculations in Appendix
G are summarized in Chapter 7 of Volume 1 (SUMMARY) and the results of the
comparisons between calculatiod nnmas easuremene trsasu mmarizn eiCd hapte8r
of Volum.1e
e haBtpo pftorh oaches describn eitd hese appendice revase ry usefnuil
ensuring that the calculational methods employed in the preparation of a
Safety Report are accurate. As a first step, it is recommended that reactor
operators/physn e ictmwsihed ousetc onith ofrstoadider ssst calecuhltat e
benchmark problem, and then compare the results of calculations with
e rr hewatohcitf corhf s moe aseurnedom eansuirn e n rmweteriahonecot it sor r
data is available in Appendix H.
VOLUME 1
VOLUM3 E SUMMARY
Topic APPENDIX Chapter
Benchmark Calculations G 7
Comparison of Calculations with Measurements H 8
CONTRIBUTING ORGANIZATIONS
ArgUoMnAn« National Laboratory United States of America
Australian Atomic Energy Commission AAEC Australia
Comlslon Chll«na de EnergTa Nuclear CChEN Chile
Comlslon Nacloa ndaEl nargfa At6mlea CNEA Argentina
Commissariat à l'Energie Atomique CEA France
Eldg. Institut fUr Reaktorforschung EIR Switzerland
Internationale Atomreaktorbau GmbH INTERATOM Federal Republic of Germany
Japan Atomic Energy Research Institute JAERI Japan
Junta de EnergTa Nuclear JEN Spain
Kyoto University Research Reactor Institute KURRI Japan
Oak Rtdge National Laboratory ORNL United States of America
Rls«f National Laboratory RIS0 Denmark
Univerf siMoti ychig -aF nord NucleaRrN FReacto r United Statesf o America
e hTIAs EIgA re hactto ernfotfurl ibutions volunteer yebtd hese organizatiod nntash anks
their experts for preparing the detailed Investigations and for evaluating and summarizing the
results presenn ttIehd is Guidebook.
CONTENTS
APPEN .DGBIEX NCHMARK CALCULATIONS
G-0. Specificationse ht rof safety-related benchmark problem .......................................11
G-l. ANL: Safety-related benchmark calculations for MTR-type reactors with HEU,
U fuEelLs .d.M.n..Ea..U. ..................................................................5..1...... .
J.E. . MMPaeEtonsn ,ington, K.E. Freese, W.L. Woodruff
G-2. INTERATOM: Benchmark calculations ......................................................5.6.....
G-3. JAERI: IAEA safety-related benchmark calculations ............................................ 91
Y. Nait. oMK, urosawa .YK, omuro. R, Oyamad. aYN, agaoka
G-4. EIR: Safety-related benchmark cR areTlcaMucltao trrisoo nf.s .......................7.0..1.....
H. Amat. oH,W inkle. rJ, Zeis
G-5. JEN: Calculations for the safety-related benchmark problem .................................. 125
G-6. AAEC: Self-limiting transients in heavy water moderated reactors .......................... 149
J.W. Connolly, E.V. Harrington, D.B. McCulloch
APPENDIX H. COMPARISON OF CALCULATIONS WITH MEASUREMENTS
H-l. CEA: Critical experiments in the ISIS reactor with the Caramel fuel element ............. 159
H-2. Kyoto University Critical Assembly (KUCA) critical experiments usinU EfgMu el
H-2.1. KURRI: KUCA critical experiments using medium enriched uranium fuel9 6..1.....
K. Kanda, S. Shiroya, M. Hayashi, K. Kobayashi, Y. Nakagome, T. Shibata
H-2.2. KURRI/ANL: Analysis of the KUCA MEU experiments using the
ANL code system ............................................................................. 183
. S Shiroya,. M Hayashi,. K Kanda, .T Shibata, W.L. Woodruff, J.E. Matos
H-2.3. JAERI: Analysis of KUCA MEU cores by the JAERI SRAC code system ...... 193
T. Mo .rKTi, suchihashi
H-2.4. KURRI: Measuremef nontse utron flux distribua tmi onenids ium
enriched uranium core ....................................................................... 201
5. Shiroya, H. Fukui, Y. Senda, M. Hayashi, K. Kobayashi
H-2.5. KURRI: Effect of reducing fuel enrichment on the void reactivity
Parti. Experimental study (Abstract) ...................................................11.2.
. HFukui. K, Mishima. S, Shiroya. M, Hayashi. K, Kanda. Y, Senda
P. IaAIrnt alytical study (Abstract) ................................................3..1..2....
Y. Senda, S. Shiroya, M. Hayashi, K. Kanda
H-2.6. KURRI: Study on temperature coefficients of MEU and HEU cores
e hKitn UCA .............................................................................5.1.2...
K. Kanda, S. Shiroya, M. Mon, M. Hayashi, T. Shibata
H-2.7. KURRI: Study on temperature coefficient of reactivity in KUCA light-water
moderatedd na reflected core— EffeF c/M rfto atiod na core shape
on this quantity ................................................................................ 225
K. Kanda, S. Shiroya, M. Mori, T. Shibata
H-3. JAERI: Critical experiments of the JMTRC MEU cores: Part I .............................. 237
Y. Nagaoka, K. Takeda, S. Shimakawa, S. Koike, R. Oyamada
Critical experiments of the JMTRC MEU cores: Part U ....................................... 247
S. Shimakawa, Y. Nagaoka, S. Koike, K. Takeda, B. Komukai, R. Oyamada
H-4. Comparison of calculations with measurements
H-4.1. FNR/ANL: Comparif socoan lculations withR fumNlle-Fas uerhemt ennits
core LEU demonstration reactor ........................................................... 255
H-4.2. JAERI: Analysis of critical experiments of FNR LEU cores ........................ 275
K. Arigane, K. Tsuchihashi
H-5. Comparison of calculations with measurements in the ORR whole-core LEU
demonstration reactor
H-5.1. ANL: Analytical R sRwuOhp eophtlo reorf-t cU oEUrLe Si -Al fuel
3 2
demonstration .................................................................................. 285
M.M. Bretscher
H-5.2. ANL/ORNL: Comparisof ocn alculad tmneade asured irradiated wire data
for HEU and mixed HEU/LEU cores in the ORR ..................................... 297
R.J. Cornelia, M.M. Bretscher, R.W. Hobbs
H-6. ANL/ORNL: Measuremed annnatsa lyf soics ritical assemblr iorefes search reactors
with mixed enrichments ...............................................................................503
J.R. Deen, J.L. Snelgrove, R.W. Hobbs
H-7. EIR: Comparisf oocna lculatd imnoane eSsh aAtsU ufnPEuriHeeM mlIRe fnots
reactor ..................................................................................................... 333
H. Winkler, J. Zeis
H-8. R1S0: Comparisons between calculated and measured flux and reactivity in HEU,
MEU and LEU fuel elements in DR-3 at Ris0 ................................................... 341
. KHaack
H-9. CNEA: Part I. Comparison of calculations with measurements of control rod worths
inU EH eht RA-2 reactor;
A. Gomez, A. M. Lerner, J. Testoni, R. Waldman
Part II. Uses of the method of computing control rod worths in the RA-3 reactor
U fuEelLs .d...n..a.. .U...E..w.H.i.t.h .................................................9..4...3.... .
A.M. Ler .nJTeer s,toni
H-10.CChEN: Measurementsd na analysisf o critical experimentsa L' eht ni Reina' reactor
using medium enrichment uranium fuel ...................................................1.6..3.....
. J Klein. ,R Venegas. O, Mutis
Appendix G
BENCHMARK CALCULATIONS
Abstract
Safety-related benchmark calculationn a sroif dealized, light-
water, pool-type reactor were performed (Appendices G-l through
o Gtc-5o )mpe ahcrtoe mputational methods usy ebvd arious organi-
zatie ohncTsa.l culations include contd roworol rths, power peak
ing factors, kinetics parameters, temperature and void coeffi-
cients, and postulated transients initiated by loss-of-flow and
reactivity insertions. Appendix G-6 contains analyses of self-
limiting transr ihoeenfatv sy water moderated reactors. Only
limited conclusions for actual core conversions from HEU to LEU
fuel shoe udblr dawn e fhrretos mun lAitp spen.dGi x