Fission Gas Release, Fuel Rod Chem Related to Extended Burnup (IAEA TECDOC-697)

227 Pages·1993·15.181 MB·English

IAEA-TECDOC-697
Fission gas release and
d fcuoher el mistry
related to extended burnup
Proceeda iTn fgeosc hnical Committee Meeting
hn ePilde mbroke, Ontario, Can8 ay Ad12apa9Mr ,9il-21
INTERNATIONAL ATOMIC ENERGY AGENCY
FISSIONS AGR ELEASD ENAF UED LOCR HEMISTRY
RELATO EETDX TENDED BURNUP
IAEA, VIENNA, 1993
IAEA-TECDOC-697
ISSN 1011-4289
Prin etheItA ydbE A niA ustria
April 1993
PLEASE BE AWARE THAT
E MISHSINTGN T PHAI ISGF ED SOOCUALM LENT
WERE ORIGINALLY BLANK
FOREWORD
e ihntv tiAte aGhttio ofovne rnmef noCt ana edThate, chnical Committee
Meeting on Fission Gas Release and Fuel Rod Chemistry Related to Extended Burnup
was hen ldiP embroke, froy 8 am12AM 9 s p1jao9 wroiit2 nl tI.t ly organeihzt eybd
International Atomic Energy Agene cht Adyna tomic Energyf o Canada Limited (AECL)
and included a technical visit to Chalk River Laboratories.
Fifty-five participants f6 rcoe 1iomnnu oten rdtnrnaieatiso nal organization
attene dmhetde eting. These proceedings c8 op2na tepaheintr s presenten fidi ve
session saws es asll ummae rhsite efoss siod nnraes commendations prepaehrte ydb
participants.
The purpose of the meeting was to review the state of the art in fission gas
relead nfsd uaecoe rhl emistry relateo dte xtended burnup. Previous IAEA meetinngos
this topic were held in Erlangen and Karlsruhe (Germany) in 1979 and 1985
respectively, and in Preston (United Kingdom) in 1988.
To determine progress in the area of fuel behaviour, it was instructive to look
back to the Technical Committee Meeting on Water Reactor Fuel Element Computer
Modelln inSigt eady State, Trand Asniceacni tdent Conditions hn ePlidr esn t1oin9 88
(IWGFPTe /3tho2Tp). ics covert eatdh at meeting were:
Transient fissions ag release (FGR),
Axs imaagl ixing,
Degradation OU fot hermal conductivity.
2
Enhancemf eofnis ts sdaioigfnf usion co-efficdineant ,
Chemistry effects.
In the summary of the Preston meeting it was said: "It is clear that we are some
a m efcfwhoaayn istic mr Food Gd"fnteRh lae" , processes ine vcdoorlmvanepad lex
improperly understood."
Important recommendations were madee ht no following aspects:
The "right" level of complexity in fuel codes;
The ability to simulate in-reactor data with out-reactor tests;
The necessity for specialized microstructural examination of
irradiated U0 ;
2
"Single effects" tests to quantify chemical effects; and
Generate data from instrumental out-reactor tests to establish
degradation of U0 thermal conductivity.
2
The present meeting was held at a time when several national and international
programn wmoaets er reactor fuel irradian eteixd perimental reactors were still ongoing
d oarhre ached their conclusion (e.g. AECL, Ris0, Halde dnnwa) hen lead test
assemblied arshe ached high burnupn i power reactod rnbsa een examineeht dtA.
same time, several out-of-pile experimenn htosi gh burnup fur eow lith simulated fuel
were being carried out. As a result, significant progress has been registered since the
last meeting, particularly in the evaluation of fuel temperature, the degradation of the
global thermal conductivity wite huh tnb nduier dnrnsuatpa e nihmdti npfoagn oct
fiss sraeioglen ae sluaedr.g seo Tailtmhyw isp ortant developmen npitrs ogrammes
t nad ianntt ieotcarnuanara roliteiod nal levels.
The IAEA wishes to express its gratitude to Dr. R. Hatcher, President and Chief
Executive Officer of AECL, and to the local organizing committee for their support and
to the session chairmen and all participants for their contributions.
EDITORIAL NOTE
In preparing this materiae htlp rof ress, stae hft ffoI nternational Atomic Energy Agency have
d givenne a e soud oaprtmhihahgogeinn ri nastamtatlt ote ameu nds yantsintuoe ubandbsmcrititpe tds
e hpt otresentation.
The views expressed in the papers, the statements made and the general style adopted are the
responsibility of the named authors. The views do not necessarily reflect those of the governments of
the Member States or organizations under whose auspices the manuscripts were produced.
n tih iessT uhbeo ok of particular designatiof oncos untrr iotees rritories dt oyoimnnepsa ly
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 mentio fno specific companf ioe trohs eir product rosb rand names doet onsi mpynlay
endorsement or recommendation on the part of the IAEA.
Authors are themselves responsible for obtaining the necessary permission to reproduce copyright
material from other sources.
CONTENTS
Summary of the Technical Committee Meeting ......................................................... 7
Opening address: The energy dilemma ................................................................... 19
S.R. Hatcher
EXPERIMENTAL I (Session 1)
Fission gas release and fuel temperature during power transients in water reactor fuel at
extended burnup ............................................................................................ 25
. PKnud. sCeBna ,g. gMMero ,gens, HeTno ,fiegaard
Fission gas release below 20 kWirr1 in transient tested water reactor fuel at
extended burnup ............................................................................................ 32
M. Mogensen .C,B agger .H, Tofiegaard. P,K nudsen, C.T. Walker
Experimental assessmen a ftote mperature threshol rdoft hermally induced fissiosang
relean tsirea nsient tested water reactor fuel with extended burnup .....................8..3......
C. Bagger, M. Mogensen
Fission gas release behavior of high burnup fuels during power ramp tests ...................... 44
T. Aoki, S. Koizumi, H. Umehara, K. Ogata
Fission-gas releasn eif uel performino gte xtended burnupn isO ntario Hydro
nuclear generating stations ...............................................................................35
M.R. Floyd, J. Novak, P.T. Truant
EXPERIMENTAL II (Session 2)
Fission gas release of high burnup fuel ................................................................... 63
. RManzel, R.P. Bod. mGBearr ,t
Fission gas release enhancement at extended burnup: Experimental evidence from
French PWR irradiation .................................................................................. 68
C. Forai, B. Blanpain, B. Kapusta, P. Guedeney, P. Permezel
The effect of fuel pellet variants on fission gas release following power ramps ................. 75
D.A. Howl, I.R. Topliss
s ag PFrelease behaviourf o high burn-uX pOMf uelsr of thermal reactors .....................28.
K. Kamimura
FUEL ROD CHEMISTRY AND RELATED PROPERTIES (Session 3)
Fission gas release from high burnup PWR fuels under transient conditions ..................... 91
H. Kanazawa, H. Sasajima, K. Homma, K. Ichise, T. Fujishiro, T. Yamahara
Fission product be dhcad ohvfnrueioema lui rt seatrxy tended burnup: Results, models
and programmes of analytical experiments carried out at the CEA, Grenoble ................ 98
M. Charles, L. Caillot, P. Dehaudt, C. Lemaignan
Impact of systematic stoichiometry differences among BWR rods on fission gas release ...... 103
B. Grapengiesser, D. Schrire
Rim effect observations from the third Ris0 fission gas project ..................................... Ill
N. Kjaer-Pedersen
Experimental techniqud neraes sults relateo dth igh burn-up investigateihot ntas
OECD Halden Reactor Project ........................................................................81.1.
W. Wiesenack
Review of Studsvik's international fuel R&D projects ................................................. 124
M. Grounes, S. Djiirle, G. Lysell, H. Mogard
Methodologf yoi n-pile experiments including experimental studief osW ER-1000
refabricated R fIrMue eehstl tsea arch reactor ....................................................0.31..
Yu.K. Bibilashvili, A.V. Grachev, V.V. Novikov, V.P. Smirnov, A.V. Smirnov
Some experimental results of investigations on the nuclide composition and
burnup fracf tioWon ER-1000 standard fuel ...............................................3..3..1.... .
A.V. Smirnov, A.P. Chetverikov, V.V. Novikov, V.N. Proselkov
MODELLING AND MODELLING SUPPORT I (Session 4)
Hot cell fission gas release studies on UO .............................................................. 145
2
M. Coquerelle, D. Bottomley
Fission-product release kinetics from CANDU and LWR fuel during high-temperature
steam oxidation experiments ............................................................................. 153
D.S. Cox, Z. Liu, P.M. Elder, C.E.L. Hunt, V.l. Arimescu
Thermal conds uarcgetl idevnaitasye from SIMFUEL .........................................5..6.1....
P.G. Lucuta, R.A. Verrait, H. Matzke, IJ. Hastings
Intergranular fission gas bubbles and solid precipitates in UO irradiated at high burnup
2
in various conditions ...............................................................................2..7.1....
M. Charles,. G Eminet,. C Lemaignan
R OF model improvement rof high burnup fuel analysis ..............................................671
.KMor. iH,I keda. N,F uruya
Modeling CANDU-type fuel behaviour during extended burnup irradiations usjnga
revised version of the ELESIM code .................................................................. 183
V.I. Arimescu, W.R. Richmond
Improvemf eoEnLt ESIM CANDU fuel performance analysis code:
Fission prs oradeguleca tse, fuel densifd icnnaetauiotrn on flux depression ........3..9...1..... .
H.C.Su .WHk, wang, B.C. Kim, K.S. Sim, Y.oHeH.
MODELLING AND MODELLING SUPPORT II (Session 5)
Evaluati foomn easured high burnup fuel temperatut raRe is0 project phas3 e. ...........5..0.2....
S. Kitajima, M. Kinoshita
A simple s friaeslsgeioans e/gaseous swelling model .......................................1..1...2..... .
T. Kogai
An approach to modelling fuel behaviour using data from some international
high burnup fuel programmes ........................................................................... 219
L-À. NordströttmO .C,
Fuel performance modelinf ogh igh burnup fuel ....................................................5.2.2..
S.H. Shann, L.F. van Swam
Discussion between S.H. Shann. M dna Mogensen ....................................................132
Lif soPt articipants .......................................................................................3.3.2...
SUMMARYE HT FO TECHNICAL COMMITTEE MEETING
EXPERIMENTAL I
1 Sess ion
Chairmen: M. Coquerelle (CEC)
H.C. Suk (Republic of Korea)
Summary
Improvements in fuel utilization have led to a growing interest in extending the
burnupf ot hermal power reactors. Fissios anrge lease (FGR) cou al ledibf e limiting
factor under either steady irradiation conditions or moderate transient conditions.
Inca ebn etritovteefrs under esrhtealten adfsioneg mechanismsd venaxraiiso tus
internatior nonaat lional programe tamrragee e eisd nhhtiiffnltftueg erfneocnet
parameters which can play a role in FGR.
Three papers from Denmark (M. Mogensen et al.) reported on the Third Ris0
Fissions aG Project. This project aimedt a continuous monitoringR GFf fo romOU
2
during power transients «400 W/cm) conducted in the Riso reactor by irradiating
prefabricatedd sfoueerg lments instrumented with pressure transdudcnears
thermocouples. Furthermore, a detailed study of the fuels before and after the power
transients allow acs orrelatiod nfnu beael RtrweGesteFrnu cturing.
s sa0fuuwUpep ly liAebmd erican Nuclear Fuel (AN2 M4F(W) .d, /Ugkrg ain
2
siz =e6 //m). General Electric (GE) (15-44 W.d/, kU1g 6-m 2g/1/r ain sd iRnzaeis) 0 (44-
48 MW.d/kg U, 10 //m grain size) and used for the transient tests. The major results
can be summarized as follows:
These transient tests allow a determination of FGR as a function of
centreline fuel temperature dna displaye ht existencea fo threshold
temperature of about 1200°C at which fission gas release occurs. This
onset temperature is independent of burnup;
No grain sizR ceGo ee uFdbflfd en etoect rminy ecbdo mparing tesnots
fuels with increasing grain size and supplied by a same vendor (GE);
An unexpected release was determined at low power transients
(125-170 W/cmF fNuA, el); this resus acwlto nfirmey dbr adiochemical
analysis and could be explained by a release of gas contained in large
pores (diameter 20-50 //m);
The results from the ANF fuel were attributed to a typical fuel structure.
. T Aoki (Japan) reportede ht no ramp behaviour fo Zr-liner fuels, with emphasis
on the correlation between fission gas release and pellet microstructural changes.
e hfuTels tested wo wdet irffoef erent types, which were base-irradiaott epud
20 and 40 GW.d/t U, respectively.
e phowTer ramp tests were pe eJhrafpotar mtna eMd aterial Test Reactor
(JMT 2Rr eRhet ta dn)aa ctorn i Sweden: ni JMTR, powes arwh elda romf aximumfo
four a hrao mtuarps termin 0aW6l /l cefmvoe lmaximun mSiw ; eden, spoawwe r
cycled between 220 and 440 W/cm, where the number of cycles was 100 and 1000.
e hTaccumulated hold timet a ramp terminal le0 0v2 dnhae 52 soalw uowrt esht rof
types of test.
Experimental investigations show thae tht fissions agr eleases i proportionaolt
the ramp terminal level and also to the square root of the accumulated hold time at
ramp terminae lf ohlellTvoewl .ing conclusions were drawn:
(1 ) The power ramp fission gas release mechanism is essentially the same as the
fission gas release from steady state fuel.
o mwajto r ecohnTtrol me(c2 h)ae :cneoirsnbamsis d eoretd
e (htauT) nnel formation that depes dangid fnfsou s ehgiot ronat in boundaries;
growth and connection of the grain boundary bubbles;
(bs adG) iffusion frome htg rain after tunnel formation.
(3e )hT tunnel formation process mighte b determineds ag ehat yb ccumulationno
grain boundaries. Therefore, the burnup dependency might appear in the fission
gas release of power ramp tests.
(4) Power cycling effect on FGR is expected to be small because of extended time
operation.
(5) A short time power increase, as in an abnormal transient, is not expected to
induce additional FGR.
. rMFloyd (Canada) reportn eiond vestigation nCos ANDU 37-element bundles
e iBrhrrautdc iean tNeidG S-A reactor. Stress corrosion cracking (SCO related defects
have been observed in ramped bundles having a burnup <450 MW.h/kg U, and under
steady-state conditions at higher burnup. Only a few bundles are involved. It was
noted that graphite-based CANLUB coating decreases SCC susceptibility, but also
seems to inhibit the inner surface of the sheath from acting as a getter for liberated
oxygen. The retention of graphite CANLUB coating is greatly reduced above a burnup
0 0M4ofW .d/ .kUgH ence, guidelin eseraus ggest reopdfo wer ramping.
Extended burnup fissions ag releaset on saw predicted wol yb burnup fuel
modelling code extrapolations. Ths isbi elie epba vrr ei eomodtdt auerucildtyi onni
U0 thermal conductivity, not accounted for in the models. The presence of solid
2
fission products is at least in part responsible for this. More work needs to be
concentrated on the possible onset of hyperstoichiometry and any possible link with
CANLUBd nas heath oxidation behaviour.
Recommendations
Incentives for a better understanding of FGR mechanisms still exist and the
interest in more detailed, sophisticated studies is growing. Future studies should focus
oe nhfto llowing points:
FGR determinationn o fuel rods irradiatedt a very high burnup
(> 60 000 MW.d/t) should be envisaged;
Detee rmthhientra mftiooanl conducr tif eovoixriftraya mda0ipaiUvltee ,d
thermal diffusivity, heat capacity, etc. muse tbc arrier odf thuo igh burnup;
2
The chemical variations of irradiated U0 must be analysed by means of hot-cell
techniques (Knudsen cell, solid state che2 mistry, annealing techniques);
e hTexperimental determinatioe hnt ffoi ssions arg etainm eirm ehdt ni usetb
y cmbar etruaief onodms ethods other thay fanlur oeXrme rpsoac ence.
Additional effort nee cebod onst centre ahptto ensods ible onsfeot
hyperstoichiometry (O/Uy rnpaao tdsionsa)ib, le link with graphite coatd isnnahg eath
oxidation.
Developments and studies similar to those detailed above (particularly
instrumented ramp testsX f)u OeshlM.e olae ub fulhifden otlcd hneid
e ihnTfluene f RotscbhhheGu troe e rnFnmusbuolhod pfor oeld precisely
determined.