IAEA-TECDOC-887
In-core fuel management benchmarks PHWRsfor
INTERNATIONAL ATOMIC ENERGY AGENCY
June 1996 The IAEA does not normally maintain stocks of reports in this series. However, microfiche copies of these reports can be obtained from
IN IS Clearinghouse International Atomic Energy Agency Wagramerstrasse 5 0 10 P.Ox Bo . A-1400 Vienna, Austria
Orders shoul accompaniee db prepaymeny db f Austriao t n Schillings 100,- for e forcheque th a mf th IAEf mo n i o n i r Aeo microfiche service coupons which may be ordered separately from the I MIS Clearinghouse. The originating Section of this publication in the IAEA was: Nuclear Power Technology Development Section International Atomic Energy Agency Wagramerstrasse 5 P.O. Box 100 A-1400 Vienna, Austria
IN-CORE FUEL MANAGEMENT BENCHMARKS FOR PHWRs IAEA, VIENNA, 1996 IAEA-TECDOC-887 ISSN 1011-4289 © IAEA, 1996 Printed by the IAEA in Austria June 1996 FOREWORD
In the framework of its reactor physics activities conducted within its nuclear power programme e IAEth , s lonAha g provide Membes dit r States wit exchang e hforua th r mfo e of technical information on in-core fuel management. This has mainly been achieved through the organizatio specialisf no technicad tan l committee meeting publicatioe th d san f technicano l documents.
Under its in-core fuel management activities, the IAEA set up two co-ordinated research programmes(CRPs) on complete in-core fuel management code packages. At a consultant meetin Novemben gi in-corn o r P 198e CR outlin fuee 8e th th l f managemeneo t benchmarks for PHWRs was prepared, three benchmarks were specified and the corresponding parameters were defined. At the first research co-ordination meeting in December 1990, seven more benchmarks were specified.
objective Th thif eo s TECDO provido t Cs i e referenc everificatioe caseth r sfo codf no e packages used for reactor physics and fuel management of PHWRs.
IAEe Th their gratefuAs i l fo participant ral P dedicateo t lCR e th sn di efforts leading thio t s report. Special thank P.Do t o sg . Krishnan . SrivenkatesaR d an i collectinr nfo e th l gal material and drafting this report. EDITORIAL NOTE
preparingIn this publication press,for IAEAthe staff of have pages madethe up from the original manuscripts submittedas authors.the viewsby The expressed necessarilynot do reflect those of the governments of the nominating Member States or of the nominating organizations. Throughout textthe names of Member States retainedare theyas were when textthe was compiled. The use of particular designations of countries or territories does not imply any judgement by publisher,the legalthe IAEA, to status the as of such countries territories,or of their authoritiesand institutions or of the delimitation of their boundaries. The mention of names of specific companies productsor (whether indicatednot or registered)as 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 materialuse or from sources already protected copyrights.by CONTENTS
1. INTRODUCTION ...... 9
. BASI2 C COMPUTER CODES USE ANALYSIR DFO S ...... 2 1 .
2.1. Lattice analysis codes ...... 12 2.1.1. POWDERPUFS-V(PPV) ...... 12 2.1.2. WIMSD-4 ...... 2 1 . 2.1.3. CLUB ...... 2 1 . 2.1.4. CLIMAX ...... 3 1 . 2.1.5. RHEA ...... 13 2.2. Computer codes for supercell calculations ...... 13 2.2.1. MULTICELL ...... 13 2.2.2. SHETAN ...... 4 1 . 2.2.3. BOXER-3 ...... 14 2.2.4. PHANTOM ...... 4 1 . 2.2.5. 3D-FAST ...... 5 1 . 2.2.6. CALC ...... 15 2.2.7. MONALI ...... 5 1 . 2.2.8. KENO-IV ...... 15 2.3. Computer codes use corr dfo e calculations ...... 6 1 . 2.3.1. PUMA-C ...... 16 2.3.2. OHRFSP ...... 16 2.3.3. DIMENTRI ...... 6 1 . 2.3.4. TRIVENI ...... 16 2.3.5. CEMESH ...... 6 1 . 2.3.6. ANAMIKA ...... 17 2.3.7. CITATION ...... 17 2.3.8. CHEBY ...... 7 1 . computee 2.4Th . r code ORIGE decar Nfo y heat calculation ...... 7 1 . 2.5. Computer code FEMXE simulatior Nfo f xenono n transient ...... 7 1 .
. DESCRIPTIO3 BENCHMARKF NO RESULTD SAN S ...... 8 1 .
3.1. Taskl: Lattice cell benchmark problems ...... 18 3.1.1. Description of the lattice cell benchmark problems ...... 18 3.1.1.1. Basic cell data ...... 18 3.1.1.2. Cases to be analysed ...... 18 3.1.1.3. Results to be provided ...... 22 3.1.2. Compariso maie th nf no result s ...... 2 2 . 3.1.2.1. Infinite multiplication factors ...... 2 2 . 3.1.2.2. Effective multiplication factors ...... 25 3.1.2.3. U-235 number densities (g/kg initial U) ...... 26 3.1.2.4. Pu-239 number densities (g/kg initia ) ...... U l 7 2 . 3.1.2.5. Relative power in the outer ring rods ...... 28 3.1.2.6. Void-induced reactivity ...... 8 2 . 3.1.2.7. Fuel temperature induced reactivity ...... 0 3 . 3.1.2.8. Moderator purity induced reactivity ...... 1 3 . 3.1.2.9. Boron moderator induced reactivity ...... 1 3 . 3.1.2.10. Weight of U/bundle induced reactivity ...... 33 3.1.2.11. Enrichment induced reactivity ...... 3 3 . 3.1.3. Access to complete information ...... 35 3.1.4. Conclusions ...... 35 3.2. Task 2: Benchmark on supercell calculations ...... 36 3.2.1. Descriptio supercelf no l benchmark ...... 6 3 . 3.2.2. Supercell modelling and computer codes used ...... 36 3.2.2.1. MULTICELL and SHETAN: Argentina ...... 36 3.2.2.2. MULTICELL: Canada ...... 9 3 . 3.2.2.3. BOXER-3, PHANTOM, 3D-FAS CALCd Tan : India ...9 .3 3.2.2.4 MULTICELL: Romania ...... 39 3.2.3. Results and discussions ...... 39 3.2.3A. Worth of reactivity devices in supercell ...... 39 3.2.3.2. Reaction rate varioun si s region supercelf so l ...... 0 4 . 3.2.3.3. Incremental cross-section variour sfo s reactivity3 4 device . .. s 3.3 core . TaseTh benchmar: k3 k ...... 3 4 . 3.3.1. Description of the benchmark ...... 43 3.3.2. Core modelling and computer codes used ...... 54 3.3.2.1. Mathematical model ...... 4 5 . 3.3.2.2. Computer codes used ...... 6 5 . 3.3.3. Results and comparison ...... 57 3.3.3.1. General ...... 57 3.3.3.2. Reference cases with the given burnup distribution ...... 58 3.3.3.3. A time step of 1 day with respect to Case 1 ...... 58 3.3.3.4. Cas wit e1 lefe core htth th hale f voideo f d ...... 0 7 . 3.3.3.5. Fresh core case ...... 0 7 . 3.3.3.6. Effect of various reactivity devices ...... 70 3.3.4. Comments and conclusions ...... 70 3.4. Task 4: Loss of regulation benchmark problem for PHWRs ...... 70 3.4.1. Specification of the problem ...... 71 3.4.1.1. Reactor model ...... 1 7 . 3.4.1.2. Incremental cross-sections ...... 2 7 . 3.4.1.3. Kinetic parameters ...... 73 3.4.1.4. Reactivity transient ...... 4 7 . 3.4.1.5. Detector representation ...... 4 7 . 3.4.1.6. Required results ...... 4 7 . 3.4.2. Descriptio calculationaf no l models ...... 4 7 . 3.4.2.1. Argentina ...... 75 3.4.2.2. India-ThPD ...... 76 3.4.2.3. Romania ...... 6 7 . 3.4.3. Result discussiond san s ...... 6 7 . 3.4.3.1. Dynamic reactivity ...... 76 3.4.3.2. Total reactor power ...... 78 3.4.3.3. Tim trif eo p ...... 8 7 . 3.4.3.4. Maximum channel and bundle powers ...... 79 3.4.4. Conclusions ...... 2 8 . 3.5. Tas : Influenck5 f isotopeo voin eo d reactivity ...... 2 8 . 3.5.1. Descriptio benchmare th f no k problem ...... 3 8 . 3.5.1.1. Basic cell data ...... 83 3.5.1.2. Cases to be analysed ...... 83 3.5.1.3. Results to be provided ...... 84 3.5.2. Computer codes used ...... 84 3.5.3. Compariso maie th nf no result s ...... 4 8 . 3.5.3.1. Multiplication factors ...... 5 8 . 3.5.3.2. Void reactivities ...... 7 8 . 3.5.3.3. WIMS reaction rates ...... 9 8 . 3.5.3.4. Fuel neutron temperature problem ...... 1 9 . 3.5.4. Conclusions and recommendations ...... 92 3.6. Task 6: Mixed loading ...... 92 3.6.1. Description ...... 92 3.6.2. Modelizatio calculatiod nan n ...... 3 9 . 3.6.3. Submissions ...... 3 9 . 3.6.4. Results ...... 4 9 . 3.6.5. Conclusions ...... 96 3.7. Tas : Calculatiok7 fissioe th f no product inventorie theid san r deca7 9 y hea. .. t 3.7.1. Benchmark description ...... 7 9 . 3.7.1.1. Cas e...... A 7 9 . 3.7.1.2. Case B ...... 98 3.7.2. Descriptio computee th f no r code used ...... 8 9 . 3.7.3 Result discussiod san n ...... 8 9 . 3.7.3.1. Comparison of the results ...... 99 3.8. Tas : Secondark8 y shut down system ...... 3 10 . 3.8.1. Description of the benchmark ...... 103 3.8.2. Modelization and calculation ...... 106 3.8.2.1. India-RED ...... 106 3.8.2.2. India-ThPD ...... 108 3.8.3. Results ...... 9 10 . 3.8.4. Conclusion recommendationd san s ...... l Il . 3.9. Task 9: Experimental validation ...... 111 3.9.1. Task 9.1: Analysi f experimentso s with 37-rod fuel clustern si ZED-2 reactor ...... Ill 3.9.1.1. Descriptio benchmare th f no k ...... 2 11 . 3.9.1.2. Computer codes used ...... 112 3.9.1.3. Main results ...... 3 11 . 3.9.1.4. Conclusions ...... 7 11 . 3.9.2. Task 9.2: Analysis of experiments with 28-rod fuel clusters in ZED-2 reactor ...... 7 11 . 3.9.2.1. Descriptio benchmare th f no k ...... 8 11 . 3.9.2.2. Computer codes used ...... 118 3.9.2.3. Main results ...... 118 3.9.2.4. Conclusions ...... 121 3.9.3. Task 9.3: Analysi f isotopiso c compositio 19-ror nfo d fuel cluster ...2 .12 3.9.3.1. Description of the benchmark ...... 122 3.9.3.2. Computer codes used ...... 122 3.9.3.3. Main results ...... 123 3.9.3.4. Conclusions ...... 127 3.9.4. Task 9.4: Analysis of the adjuster rod experiments in ZED-2 ...... 127 3.9.4.1. Benchmark problem specification ...... 128 3.9.4.2. Calculational method computed san r codes used ...... 1 13 . 3.9.4.3. Analysi f Zed-so 2 experiments ...... 2 13 . 3.9.4.4. Conclusions ...... 5 13 . 3.10. TasklO: Power distribution control ...... 5 13 . 3.10.1. Description ...... 5 13 . 3.10.1.1 Reactor model ...... 136 3.10.1.2. Problem specification ...... 137 3.10.1.3. Additional data for simulation of xenon and iodine dynamics ...... 0 14 . 3.10.2. Modelling ...... 1 14 . 3.10.2.1. Canada ...... 141 3.10.2.2. India ...... 141 3.10.3. Results ...... 142 3.10.4. Conclusions ...... 144
4. SUMMARY AND RECOMMENDATIONS ...... 147
4.1. Summary ...... 7 14 . 4.1.1. Tas Lattic: k1 e cell benchmark problems ...... 7 14 . 4.1.2. Task 2: Benchmark on supercell calculations ...... 148 4.1.3. Task 3: The core benchmark ...... 148 4.1.4. Task 4: Loss of regulation benchmark problem for PHWRs ...... 149 4.1.5. Tas : Influenck5 isotopef eo voin so d reactivity ...... 9 14 . 4.1.6. Tas : Mixek6 d loading ...... 9 14 . 4.1.7. Tas : Calculatiok7 fissioe th f no n product inventorie theid san r decay heat ...... 150 4.1.8. Tas : Secondark8 y shut down system ...... 0 15 . 4.1.9. Task 9: Experimental validation ...... 150 4.1.9.1. Task 9.1: ZED-2 experiments wit rod7 h3 s clusters .....0 15 . 4.1.9.2. Task 9.2: ZED-2 experiment clusterd s witro 8 h2 s ...... 0 15 . 4.1.9.3. Task 9.3: ZED-2 experiments with 19 rod clusters ...... 151 4.1.9.4. Task 9.4: Adjuster rod experiments in ZED-2 ...... 151 4.1.10. Tas : Powek10 r distribution control ...... 1 15 . 4.2. Recommendations ...... 1 15 .
REFERENCES ...... 153
CONTRIBUTORS TO DRAFTING AND REVIEW ...... 159 1. INTRODUCTION
A resulsa manf o t y year researcf so developmend han t activities, co-ordinated dan supported by IAEA, complete in-core fuel management packages for three types of reactors, namely PWRs, BWR PHWRsd san , became available Dat froA amNE Ban k r som[1]Fo . e reactor types, these programme package available ar s e with three level f sophisticationo s : level I is useful as a first introduction to in-core fuel management for nuclear engineers; level usefus i survey-typr n fo l e calculations, scopin sensitivitd gan y analyses computew lo t a , r degree th s accuracf eo costha t bu , y whict sufficienno s hi actuar fo t l reactor operation; level I includeH s codes sufficiently accurat provido et date eth a neede actuar dfo l reactor design and operation.
nexe Th IAEte goath f Alo supported researc developmend han t activitie thin si s field was to develop test cases appropriate to check the fuel management computer code packages and their procedures for both levels II and HI [2]. t onl no provido yt s i P epurpose CR reference Th th f eo e case verificatior sfo e th f no code packages used for reactor physics and fuel management of PHWRs, but also to provide countries tha beine tar g introduce nucleao dt r energy with sample problem resultd san s useful PHWo t R physics calculations.
PHWe th r RFo benchmark specification f realistio t se a cs reactor data (availabln ei opee th n literature) coul definee db d that provid informatioe eth n neede develoo dt inpue pth t data for running in-core fuel management codes. There are several parameters that must be calculated by the code packages, related to (1) the safety performance of the design, (2) the ability of the design to produce the rated power and (3) the economics of the design. These parameter dividee b calculate e y b PHWRe n dsma th ca intseto r firse t y odsfo tw b se Th t. scopin lever go codI lI e packages, wherea secone sth involvet dse s detailed three dimensional level III code packages.
Once the benchmark specification and parameters are available each code package can be used by any participating institute in the CRP to perform in-core fuel management calculations t wilI .e importan b l comparo t t e resultth e s calculate r thesfo d e benchmark problem differene th y sb t computer code packages code Th .e packages availabl IAEo et A Member States can also be used to calculate different types of parameters useful in physics startup tests, fuel management loading schemes, various accident analyses, etc.t , no thae ar t usually identifie in-core parth s da f o te fuel management studies.
presene Th t generatio PHWRf no naturae sus l uraniur mo fue8 2 n clusteri l, 19 f so 37 rods. Most of them employ two independent shutdown systems which comprise mechanical shut off rods, poison tubes or injection of poison in the moderator. There is a wide rang f sizeseo , viz. smale th , l size MWe0 d22 medium-sizee th , d reactors producing about 500 MWe and large sized reactors producing 750-800 MWe. The larger reactors are pron spatiao et l instabilities whic controllee har zonay db l control syste whicmn i h levef o l light water is adjusted to maintain desired power shape. The reactivity devices are mostly perpendicular to the fuel channels. Ther incentivn a s ei mako et reactore eth s more economi increasiny cb g fuel burnup. The features of a PHWR suitable for use of thorium also could be exploited. Many studies are underway because of these reasons and there is a likelihood that in future, PHWRs would use slightly enriched uraniu plutoniur mo m containing fuelr usinFo . g these advanced fuels, a suitably optimized fuel cluster may emerge.
n are A f concero a r PHWRnfo s positivi s e void coefficien f reactivityo t ; hence investigatio f voino d reactivity assumes importanc e contex th a los f n coolanf i eo so t t accident.
benchmare Th k tasks were devise asseso dt s predictive capabilitie differenf so t code package situationn si featuree s th arisin f o PHWRf t so gou s mentioned above firse Th .t three benchmarks which were discussed in the RCM at Buenos Aires were of basic nature. They dealt with the modelling of fuel cluster lattice cells, the 3-D supercell model to simulate the reactivity devices perpendicular to the fuel channels and the core simulations for various device configurations and reactor states.
Subsequently seve benchmarkw nne s were formulate proposed dan r analysidfo y sb participante th CRPf so . These benchmark relate e controe sar th safetd o dt an l y problemf so the PHWRs and new fuels ; a loss of regulation accident analysis (LORA); evaluation of void reactivity with respec differeno t t t fuel isotopes; basic characteristi hypotheticaa f co l core loaded with MOX fuel; an idealized formulation of secondary shutdown system (SDS-2); evaluatio f decano y heafunctioa s a t timf no e after shutdown flud xan ; power tiltthend san - on-line control benchmarA . directlo kt y compare calculational results with some typical experiment ZED-2n si alss owa included.
overvien a n TablI 1 benchmark1. e es providew th i l al f theid do an sr involved organization thin i s s report.
In Chapter 2, an overview is given of all computational tools used by different participants for the various benchmarks. Chapter 3 provides a description of each benchmark, the required input data, and an intercomparison of the results. More detailed informatio gives ni nationan i l reportvarioue th n so s tasks, whic collectee har IAEn a n dAi Working Document [3]. In Chapter 4, a summary of the salient conclusions and recommendation givene sar . Finally consolidatea , d lis f referenceo t includeds i s .
10 TABLE 1.1. PHWR BENCHMARK
Task Description ARG CNA ThPD1' RED2) ROK PAK ROM Rapporteur
1 Cell X X X X X X Dumitrache 2 Supercell X X X X X Lawande 3 Core X X X X X X Parvez 4 Los f Regulatioso n X X X Lawande 5 Loss of Coolant/void X X X XX Dumitrache 6 MOX X X X X Balakrishnan 7 Fission Products/ X X X Parvez decay heat 8 Secondary Shutdown X X Balakrishnan 9 Experimental Validation 9.1 37 elements X X X Lawande element8 2 2 9. s X * Lawande 9.3 19 elements X X Lawande 9.4 Adjuster rod ZED-2 X X Lawande/Fink 10 Power distribution X X Lawande/Gold Control
Theoretical Physics Division, Bhabha Atomic Research Centre, India 2) Reactor Engineering Division, Bhabha Atomic Research Centre, India
Analyze* d 19-rod cluster experiments instea 28-rof do d cluster. 2. BASIC COMPUTER CODES USED FOR ANALYSIS
brieA f descriptio basie th f cn o compute r codes use variouy db s participant comparo st e their calculations with the benchmark data is presented. These codes can be divided into following 5 categories :
(1) Computer Codes for Lattice Calculations (2) Computer Codes for Supercell Calculations of Reactivity Devices (3) Computer Codes for Core Calculation ) Compute(4 r Code r Calculatiosfo Decaf no y Heat (5) Computer Codes for Simulation of Xenon Transient
2.1. LATTICE ANALYSIS CODES
Five codes have been used for lattice analysis : WIMS (Argentina, Korea, Pakistan, Romania), POWDERPUFS-V (Argentina, Canada, Romania), CLUB (India-ThPD), CLIMAX (India-ThPD) and RHEA (India-RED). A brief description of these codes is as follows :
2.1.1. POWDERPUFS-V (PPV)
The program POWDERPUFS [4] was developed by AECL. It is the most important code e foCANDth r U lattice cell calculations availabls i t I . CANDe th onl r yfo U reactor owners. code one-dimensionala Th s ei , radial geometry, recipe-based program t calculateI . s neutron multiplication factors, two group cross sections and burnups for various fuel irradiations. The code utilizes many semiempirical relations. They have been fitte colo dt d clean experimentf so CANDU-type lattices only Westcote Th . t cross section conventio useds i ] Three-regionA n[5 . s lattice cel assumeds i l .
2.1.2. WIMSD-4
It is one of the most frequently utilized lattice cell code for thermal reactors. The program was developed at UKAEA, Winfrith, UK [6] which uses 69 group WIMS library. It usee b r plates dfo n ca , rodclustersd an s firse Th t. leve f calculationo l s involve detailea s d energy dependence calculation a simplifie n o s n celpi dl followe a transpor y b d t theory calculations focusin spatiae th n go l latticedetai e vera th s i f y lo t I versatil. e code where different e optionspatiath r lfo s condensatio e availablar n e likee DSNth , , PARSES J (collisioPI , n probability scheme) etc. There are various options available for leakage calculations too. There alse ar o additional options availabl spacr energed fo ean y condensation like homogenization over user specified regions.
2.1.3. CLUB
It is a PHWR lattice cell code developed by India-ThPD [7-9]. The multigroup integral transport equation is solved using a combination of collision probability (CP) method and interface current technique associatee Th . d nuclear data groulibrar9 6 a ps yi WIM S librarr yo its condensed version of 27 groups.
thin I s method whole th , e lattice cell (cylindricalized divides i ) d int olarga e numbef o r annular regions. Excep regionsfirsr w fo tfe t l otheal , r region homogeneouse sar . e Eacth f ho
12 non-homogeneous region contains fuel pins of a particular ring of fuel cluster completely within it. These non-homogeneous regions called rings are further subdivided into homogeneous zones. It is assumed that a ring contains the same type of fuel pins and that they are symmetrically situate thao ds t eac rina hn gi fuesee n same pi ls th e fluxinteractioe Th . n between various zones within a ring and their contribution to outgoing currents at inner and outer surfaces of the ring are calculated by the CP method. Each region is connected to neighbouring regions by interface currents t thesA . e interfaces e angulath , e incominr th flu n i xd outgoin an g g directios i n separately expanded in terms of spherical harmonics. The simplest assumption is that of isotropic angular flux or popularly known as cosine current approximation (i,e., one term expansion). This approximation has been found to be adequate for normal calculations. However, at present, up to three terms can be considered in the code CLUB. Another basic assumption is that the scalar flux is constant in each zone of the region.
The multigroup integral transport equation is first solved for the lattice cell. Then one group condensed microscopic cross sections of each element hi all the fuel regions are calculated. Thereafter, the burnup equations are solved one by one for every fuel region. These solvee ar eithey db r trapezoidal rul r Rungeo e Kutta method burnuo . Thertw e pear schemes. In the first burnup scheme called explicit burnup scheme, 33 fission products are considered explicitl pseude on d o yan fissio n produc consideres i t seconde th [10] n I d. scheme, Nephew schem uses ei treao dt t fission products thin I . s scheme fissioe th , uraniuf no plutoniud man m isotopes would yield five pseudo fission products and element Rh-105 and Xe-135. Due to short half lives of Rh-105 and Xe-135, these elements are assumed to be always at saturated concentration.
2.1.4. CLIMAX
developes wa t I India-ThPy db Dmultigroua s [11]i t I . p integral transport theory code for cylindrical geometry. The cluster is ring-wise homogenized and then interface current formalism is used. However, it applies heterogeneity correction in the fast energy range. For the resonance energy groups, the collision probability is applied for the exact cluster geometry estimato t Dancofe eth f correction factor grou7 2 A p. library derive dgrou9 6 fro e pmth WIMS library is used.
2.1.5. RHEA
developes wa t I India-REn di D [12]. Thifiva s esi group neutronics code consistinf go thre thermao etw fasd l an tgroups lattice Th . e parameter obtainee sar homogenizatioy db n over the cluster. The neutron balance equations are solved in cylindrical geometry. The heterogenous cluste thes ri n ring-wise homogenized usin hyperfine gth e flux distribution insid latticee eth e Th . burnup equation solvee sar d usin lumpega d fission product scheme.
2.2. COMPUTER CODE SUPERCELR SFO L CALCULATIONS
Various computer codes used for supercell calculations to simulate the reactivity devices by the different participants are described below:
2.2.1. MULTICELL
The incremental cross-sections for all reactivity devices were calculated using the three dimensional supercell code MULTICELL [13]. MULTICELL calculates flux distributiony sb solving the two energy group neutron diffusion equations through a finite difference iterative
13 approach. To solve the diffusion equation in the MULTICELL model, two options are available; one is the flat source option where the slowing down source is assumed to be flat, and the other grouo tw pe th optio s i n where bot epithermae hth thermad an l l fluxe calculatede ar s firse Th .t optio applicabls ni devicee th regiono d et san s over which ther littls ei e chang epithermae th n ei l flux, whil secone eth d optio uses ni d when devices greatly alte epithermae rth l flux distribution (e.g.light water zone controllers).
For materials where diffusion theory cannot be applied, such as fuel and strong absorbers, curren fluo t x ratio (CFR) boundary condition usee sar represeno dt t them. The evaluatee yar d from transport theory or analytical calculation, on the internal surface of the fuel or reactivity device for both neutron energy groups.
2.2.2. SHETAN
SHETAN [14 threa s ]i e dimensional integral transport theory cod ebloce baseth kn do method of solving the integral transport equation. This method combines the advantages of the conventional collision probability methointerface th d dan e current techniquex codmi e n Th .e ca rectangular and cylindrical coordinates which allows accurate modelling of cylindrical fuel channel reactivitd san y devices withi nrectangulaa r cell.
2.2.3. BOXER-3
In BOXER-3 [15], the two group flux distribution in the supercell is obtained by solving integrae th l transport equation. This code use combinatioa s f collisiono n probability (Pijd )an interface currents (J±) method for solving the transport equation. The supercell is divided into numbea f rectangulao r r block r majoo s r regions. Each bloc furthen kca dividee b r d intoa number of meshes which can be rectangular and/or cylindrical. Within each block, the meshes couplee ar eaco dt h othe collisioy b r n probabilities whil blocke eth couplee sar eaco dt h other through interface currents. The integral transport equation hi the "block" method are solved for flux, curren eigenvalud an t e Kusuay «b l source interaction technique.
2.2.4. PHANTOM
The computer code PHANTOM [16] is based on a diffusion iterative method which is used to preserve the net leakage into a control or a fuel region which are predominantly absorbing region obtaines sa transpora y db t theor ygrouo modeltw pe parameterTh . fuer sfo l cluster, pressure tube gapr ai , , calandria tub heavd ean y water moderato obtainee rar usiny db g computer code CLUB which is a 1-D multigroup transport theory code. The two group net leakages into fuel cluste evaluatee ar r flua r x dfo normalizatio n with total cell absorptios na unity. Equivalen diffusioD 1- t n calculatio dons ni e with computer code COMESH [17] which is corner mesh diffusion theor grouo y tw code pe leakageTh . s obtaine diffusioy db n theors yi different from those obtained by transport theory. A two group correction factor is iteratively foun thao ds t thes matcho efuetw e Th l. cluster cross-sections (except transport cross-section) are modified with these group dependent correction factors. For reactivity devices an 1-D supercell calculation is done by multigroup transport theory code PIJMURLI [18]. Thereafter an equivalent 1-D diffusion calculation is done with COMESH. The current to flux ratio (a-) boundary conditions are evaluated by diffusion iterative procedure. The 3-D supercell is then analyze COMESy db H with modified fuel cluster parameter boundare th d an s y conditionn so reactivity devices.
14 2.2.5. 3D-FAST
The computer code 3D-FAST [19] solves the 3-dimensional time dependent diffusion equation with either adiabati r improveco d quasi-static method finite Th .e difference equations are solved using successive line overrelaxation (SLOR) scheme numbee Th . energf ro y groups and numbe f spaco r e point limitee available ar sth y db e computer memory.
This code was also used for simulating the reactivity devices of PHWRs hi supercell model computee th n I . r code 3D-FAST both fuel cluste reactivitd ran y device region treatee sar d by internal boundary conditions which are determined by transport calculations using computer code DTF-LAMDA. It is assumed that a flat fast neutron source is present in the fuel. The two group diffusion equations are solved using the centre-mesh finite difference diffusion theory model 3D-FAST.
2.2.6. CALC
The code CALC [20] developed by India-RED, solves the Boltzmann linear transport equation in two dimensional geometry. The procedure of solution is based on the DSN method. formalisN DS e transpore th m th n I t equatio dimensiono tw n ni writtes i s finitn ni e difference form for angular fluxes. These equations are solved in CALC using diamond differencing (cell averaging angular flux is the arithmetic mean of the angular fluxes at the opposite boundaries space anglth d cele oi ean th fh l e phase space).
2.2.7. MONALI
The Monte Carlo code MONALI [21] developed by India-ThPD is primarily written for calculating nuclear parameters associated with thermal reactors.
crose Th s section reae sar d directly fro WIMe mth S multigroup cross sectio t (69/2nse 7 group) which have been derive NucleadK U fro e mth r Data filestreatmene Th . f anisotropo t y is possible in selective materials containing light nuclei up to first order using the PI-terms tabulate WIMe th n di S cross section set.
geometre Th y modul code handln th ef ca regioe y o ean n enclosewhice t b se n a h ca y db of quadratic surfaces of the general form:
A(X-B)2 + C(Y-D)2 + E(Z-F)2 - K2 = 0.
Provision speciax si r fo sl above formth f eo s equation including planes, cylinderd an s spheres are also included. Any number and any combination of the seven types of surfaces can enclose a region. The outermost surface can also be any combination of these surfaces. Any of the three boundary conditions, namely reflective, white or vacuum boundary condition can be applied selectively on the outermost surfaces.
2.2.8. KENO-IV
KENO-IV [22 multigroua s ]i p Monte Carlo criticality program core describes Th ei . d by box types each of which contains a different geometry configuration. The box types can be stacked togethe foro rt mthree-dimensionaa l array reflecto A buil.e b n t arounca r arrae dd th yan albedo condition can be assigned.
15 2.3. COMPUTER CODES USED FOR CORE CALCULATIONS
Various computer codes use corr dfo e simulation differene th y sb t participant briefle sar y described here:
2.3.1. PUMA-C
Develope CNEAy db , Argentina, PUMA-C [23 versioa PUMe s ]i th f no A code oriented fuee th l o management Embalse th f o t e reactor. PUM threa As i e dimensional (x,y,z r (r,<£,z)o ) reactor multigroup diffusion code, also developed by CNEA, with possibilities for burnup and fuel management calculations, fuel accounting, space dependent xenon, reactor kinetics and thermohydraulic feedback. PUMA-C solves the multigroup neutron diffusion equations in a finite difference mesh centered scheme.
2.3.2. OHRFSP
OHRFSP, Ontario Hydro Reactor Fuelling Simulation Program [24], model core sth s ea threa e dimensional arra parallelepipedsf yo . OHRFS Ontaris Pi o Hydro's versio Fuee th f lno Management Design Program FMDP [25] developed by AECL.
OHRFSP solve two-energe th s y group, three-dimensional neutron diffusion equations through an iterative finite difference approach. The calculations are performed in the SIMULATE model of OHRFSP which carries out the calculations in discrete time steps, with the bundle burnups at each step being incremented from the previous step.
2.3.3. DIMENTRI
DIMENTRI [26 threa ] e dimensional centre mesh finite difference code developey db India-ThPD employs uniform radial meshe variabld an s e axial meshes radiaA . l extrapolation distance is added to the inside notch /calandria wall radius in core calculations simulating explicitly the circular boundary. It has the 6 as well as the k-eff eigenvalue options.
2.3.4. TRIVENI
code Th e TRIVENI [27] develope India-ThPy db beinDs i g extensively use Indin di t aa all NPPs for the follow-up and fuel management of 220 MWe PHWRs. The code DIMENTRI form diffusioe th s n theory modul f TRIVENIeo . This code uses fixed axial meshe wels sa s a l radial ones.
2.3.5. CEMESH
e CEMESTh H code [28] develope India-ThPy b d o grouptw a , s Di 3-dimensiona l diffusion theory code using centre-mesh finite difference scheme. The core portion is simulated with a constant mesh size of PxPxL/2, P being the lattice pitch of the fuel cluster and L the length of the fuel bundle. Variable mesh sizes are allowed in reflector region. The constant mesh width scheme havallowo t e permanentlson y built-in informatio locatione th l typef nal o f so of reactivity devices (RDs) for a typical core in the code itself. The code determines the meshes which shall be perturbed by the presence of any given RD. A special flux related weighing schem uses ei r thi dfo s purpose consistene orden b I . o t r supercelD t 3- wit e hth l modee th l region of influence of a RD is restricted to ± L/2 distance while distributing the 'ôEs' to the three consecutive axial meshes adjoinin givega locationD nR . CEMESH allow arbitrary san y
16 combinatio configuratioD R f no cord an ne states t use I .e lattic th s e database generatey db PHANTOM code system s beeha n t I .extensivel y validated powew againslo e r th tphysic s experiments done in Indian PHWRs.
2.3.6. ANAMIKA
code Th e ANAMIKA [29] develope India-REDy db two-groupa s i , , three dimensional diffusion theory code used for core follow-up. The equations are solved using the finite difference method. This programm fixea s deha mesh spacing core Th e. follow-u dons pi e using the burnup distribution of the core.
2.3.7. CITATION
CITATION [30] developes Ridg,k wa Oa ey dNationab l Laboratoravailablw no s i d e yan Datfinite-differencA D throug3- aNE a Bank e s i hth t I . e computer code. CITATIOa s Ni comprehensive industry-size bees d ha cod nd extensivelean y use lighr dfo t water reactor core analysis and fuel management studies. The mesh sizes used in CITATION correspond to those used in OHRFSP.
2.3.8. CHEBY
CHEBY [31] is another AECL code applied to fresh core analysis only. It is a 3-D diffusion theory code. The geometrical model is prepared utilizing the MATMAP code, another AECL programme [32].
COMPUTEE 2.4TH . R CODE ORIGE DECAR NFO Y HEAT CALCULATION
e computeTh r code ORIGEN [33 bees ha ]n use r thidfo s benchmark evaluatiol al y nb three participants, i.e., Argentina, Pakistan and Romania.
The ORIGEN code developed by Oak Ridge National Laboratory, U.S.A is capable of computing isotopic compositio radioactivitd nan fuef yo l materials, fission products claddind an , g materials in both fixed and fluid fuel reactors. The code uses the matrix exponential method solvo t equatione eth radioactivf so e growt decad larga hr an yefo numbe isotopesf e o r us e Th . of the matrix exponential method permits the treatment of complex decay and transmutation schemes extensivn A . e librar nucleaf yo r dat availabls ai e wit codee hth .
2.5. COMPUTER CODE FEMXE SIMULATIOR NFO XENOF NO N TRANSIENT
FEMXEN is a three-dimensional two-group finite-difference code [34] for the simulation of xenon transients developed by India-ThPD. Delayed neutrons are assumed to be in equilibrium with prompt neutrons e longeowinth o t gr tune scale involve e xenoth n ni d redistribution. The solution of the time-dependent neutron diffusion equation is thus reduced to a series of K-eff eigenvalue calculations. The time-dependent xenon and iodine concentrations are, however, calculated at the end of each time step. The thermal neutron absorption cross- section updatee sar xeno accountinw y db ne ne spatiath r gfo l distribution. Power distributiod nan the K-eff eigenvalue are in turn updated at the beginning of the next time step with the changed cross-section direcA . t iterative criticality searc basie - th K changf carries o hn i e o th t n ei dou eff by adjusting the absorbers distributed in the core such as water levels in all the zone control compartments in PHWRs. After obtaining a converged solution in this way, the spatial control s affectedi .
17 . 3 DESCRIPTIO BENCHMARKF NO RESULTD SAN S
The benchmark problems for In-core fuel management of PHWRs initially contained three parts : (i) lattice cell calculations; (ii) supercell calculations for reactivity devices; (iii) core calculations.
A research Co-ordination meetin s hel t Argentingwa a d n 1990i a e meetin.Th g recommended seven more tasks for their inclusion in the present CRP : (iv) a loss of regulation accident analysis (LORA); (v) evaluation of void reactivity with respect to different fuel isotopes; (vi) basic characteristic of a hypothetical core loaded with MOX fuel; (vii) evaluatio f decano y a functiohea s a t f timo n e after shutdown; (viii n idealizea ) d formulation of secondary shutdown system (SDS-2); (ix) analysis of various experiments and (x) flux power tilttheid san r on-line control.
This chapter describes all the benchmarks and gives comparison of results submitted by various participants.
3.1. TASK 1 : LATTICE CELL BENCHMARK PROBLEMS
Reactor physicists from six countries solved the lattice cell benchmark problems: Argentina, Canada, India, Korea, Pakistan, Romania. Five codes have been used: PPV (Argentina, Canada, Romania), WIMS (Argentina, Korea, Pakistan, Romania), CLUB (India- ThPD), CLIMAX (India-ThPD RHEd an ) A (India-RED).
3.1.1. Description of the lattice cell benchmark problems
3.1.1.1. Basic cell data
The complete set of input data for the lattice cell reference benchmark problem is reproduce Tabln di e 3.1.1 lattice Th .e cell mode presentes i l Figurn di e 3.1.1.
3.1.1.2. Cases to be analyzed
Seven lattice cell benchmark problems have been defined. The first case is the reference cell: all the basic cell data are very similar to the real PHWR data, in normal operating conditions.
The next two problems have been defined to compute the void effect, case 2, and the fuel temperature effect perturbatioe th ,s vera casA f . yo e3 s shorni te fuetermth l l al , isotope number densitie identicae ar sreference th o t l e numbe rburnue densitieth l al p r sfo steps.
nexe Th t three problems have been define calculato dt effecte eth s associated wite hth long term perturbations moderatoe casn th I ., e4 r isotopic purit changeds yi e casn th i ; , e5 moderato poisones i r d with boron weigh e naturacase n i th ;th e6 f o t l uraniu r bundlmpe e increaseds i othee th l r initiaAl . l basic cell dat identicae aar referenc e th o lt e case onese Th . isotope number densities are independently computed for each case.
18 TABLE 3.1.1 BASIC CELL DATA
Fuel Material Compacted and sintered natura 2 pelletlUO s Total fuel mass per bundle 18.7 kg Temperature C ("effective"average° 0 70 ) Dimensions pellets Diameter 12.154 mm Stack length (nominal) m m 0 48
Fuel Bundle Assembly Numbe f elementro s (rods) 37 Length of bundle m m 5 49 Outer element ring Number of elements 18 Diameter (nominal, cold) 86.6m m 1 Intermediate element ring Numbe f elementro s 12 Diameter (nominal, cold) 57.5m m 1 Inner element ring Numbe f elementro s Diameter (nominal, cold) 29.7m 7m Centre element 1 Arrangement see Figure 3.1.1 Element Sheath Material Zirconium Density 6.55 g/cc
Dimensions Outside diameter 13.0m m 8 Thickness (average) 0.419 mm Tota masr Z lbundlr spe e 2.264 kg
19 TABLE 3.1.1. (Continued)
Pressure Trabes Material Zirconium Density 6.8775 g/cc (the density correspondo t s 6.55 normal Zr dens. * 1.05 to take into account the presence of Niobium) Temperature 290 °C
Dimensions (cold) Inside diameter 103 mm Wall thickness 4.34 mm
Calandria Tubes Material Zirconium Density 6.55 g/cc Temperature 680 °C Dimensions Inside diameter 129 mm Wall thickness m m 4 1.
Coolant
Material 0 2 D 99.7 o w/ 5 Density 0.804 g/cc Temperature 290 °C ("effective" average) Fission bundle power 620 kW Moderator Material 99.75 w/o D20 Density 1.0858 g/cc Temperature 68 °C Boron Concentration mg/kO O gD2
Miscellaneous Geometric buckling 0.76 nT2 Coolant power/fission power 0.95 Maximum burnup 8000 MWD/T Burnup step 800 MWD/T
20 13.081 mrr l
FIG. 3.1.1 The 37 Rod Cluster
21 The last problem, case 7, has been defined to calculate a slightly enriched uranium lattice cell.
All the seven cases are presented in Table 3.1.2
3.1.1.3. Results providedbe to
Table 3.1.3 indicates the results to be provided by the participants.
3.1.2. Comparison of the main results
3.1.2.1. Infinite multiplication factors
Reference Lattice cell
Durin Buenoe gth s Aires meeting, December 1990participante th , s agree o selecdt t a set of "main results" to be reported and intercompared for four buraup values: fresh clean fuel, fresh fuel with xenon, 4000 MWd/MgU and 8000 MWd/MgU. The results are reported in Refs. [35-42]. The infinite multiplication constants are presented in Table 3.1.4. frese Th h clean fuel cas missins ei PPV-Canadar gfo .
TABLE 3.1.2 LATTICE CELL ANALYZEE CASEB O ST D
Case Description
1. Burnup calculation Burnup step 800 MWd/MgU Maximum burmip 8000 MWd/MgU
2. Perturbation case Coolant density 0 g/cc
3. Sensitivity case Fuel temperature 300 °C
4. Sensitivity case Moderator purity 99.85 w/o
5. Sensitivity case Boron concentration mg/K1 g D20 6. Sensitivity case Weight U/bundle 19 kg
7. Sensitivity case Enrichment 1.2 % U235 End burnup 20 MWd/KgU
Note: For all the sensitivity cases, the isotopic densities are recalculated with the new values of the relevant parameter, wher, e ) isotopieth (3 d an c ) compositioexcep(2 r fo t keps ni s a t calculated with the nominal input data for all burnup steps.
22 TABLE 3.1.3 RESULT PROVIDEE E LATTICB TH O ST R DFO E CELL CASES
1. Cell code used 2. Energy group structure 3. Parameter e specifieb o st functioa s a d burnuf no p infinitk 1 y3. effectivk 3.2 e e fou th r eacr Fo f ho ring f fueso l rod: s 5 23 (g/K U g3. intia3 ) lU 3.4 Pu 239 (g/Kg intial U) 0 (g/K24 u P g intia3.5 ) lU 1 (g/K24 u P g intia ) 6 lU 3. 3.7 Pu 242 (g/Kg intial U) 5 (g/K13 e X g intia ) 8 lU 3. 9 (g/K14 m S g intia ) 9 lU 3. 3.10 Pin power (kW/pin)
l D 3.11 (cm) 2 D 3.12 (cm) 3.13 E al (cirr1) 2 a E 3.14(cm' 1) 2 1 E 3.15(cm' 1) 1 2 E 3.16 (cm"1) 3 .17 v E fl (cm'1) 3.18 2 vf E (cm'1) 3.19 eE fl (kW.s.crtT1) 3.20 eE f2 (kW.s.cnT1)
4. Reflector cross-section 4.1 Dl (cm) 2 D (cm 2 4. ) 4.3 E al (cm'1) 2 a E 4-.(cm"4 1) 4.5 E 12 (cm'1) 1 2 E (cnT 1)6 4.
independen0 1 f o divide e t W se e td th result s into thre efamilV groupsPP 3 y( e th : versions) WIMe th , S famil versions4 y( Indiae th d n an )famil codes)3 y( comparee W . e dth resulte familon f yo s predictionwite othere th hth d san s inside each family.
The PPV - Romania, WIMS - Argentina and CLUB - India-ThPD code prediction sets are complete. During the intercomparison analyses they have been considered as representativ three th er familieefo f codeso s .
e (11-19Th ) rows presen predictioe th t n difference f reactivityo k m n i s :
= 100 j i l 0(1/k* De ;- 1/kj )
23 TABLE 3.1.4 INFINITE MULTIPLICATION FACTORS REFERENCE LATTICE CELL
Code BURNUP (MWD/MGU) Country 0 0+ 4000 8000
1 PPV - Argentina 1.125137 1.079509 1. 049259 .974061 2 PPV - Canada 1.080755 1. 048803 .971069 3 PPV Romania 1.126459 1.080696 1.048712 .971859
4 WIMS - Argentina 1.115492 1.081959 1.045623 .977738 5 WIMS - Korea 1.115919 1.092060 1.040466 .973226 6 WIMS - Pakistan 1.1068 1.0737 1.0554 .9864 7 WIMS - Romania 1.106956 1.070058 1.037644 .973647
8 CLUB - India 1.1150 1.0807 1. 0413 .9749 9 CLIMAX - India 1.1150 1.0813 1.0421 .97661 10 RHEA - India 1.11675 1. 07694 1.03719 .970568
11 De4 3 l - 8.73 1.08 - 2.82 6.19 12 8 3 l De - 9.12 0.003 - 6.79 3 .21
13 1 3 l De - 1.04 - 1,02 0.50 2.33 14 Del 32 — — — — 0.05 0.08 -0.84
15 De5 4 l 0.34 8.55 -4.74 - 4.74 16 Del 46 - 7.04 - 7.11 9.86 8.98 17 De7 4 l - 6.91 -10 .28 - 7.35 - 4.30
18 9 8 l De 0.00 0.51 0.74 1.80 19 Del 810 1.41 - 3.23 - 3.81 - 4.58
Del ij = 1000 * (1/k, - 1/k.j ) (rak)
e resultTh familsV e verinsidPP ar ye y th esimilar . However Argentineae th , n version offer largea s r multiplication facto t 800a r 0 MWd/MgU. This effect seeme b o t s related to the updating of the energy per fission values in the Argentine version of PPV.
The results insid WIMe eth S grou alse par o clos eaco et h other. Unfortunatelye th , fresh fuel case 0+ (with xenon) was not defined in an identical manner. For larger burnups, the Korean and Rumanian infinite multiplication constant estimations are similar and less reactive tha Argentineae nth n ones.
The CLUB and CLIMAX predictions are very close. However, we may note that the CLIMAX estimations become more reactive when the burnup increases. The RHEA results are less reactive, by 4.6 mk at 8000 MWd/MgU.
The PPV infinite multiplication constants are larger than the WIMS ones for the fresh clean fuel. The discrepancy is significant: 8.73 mk. As the burnup increases the
24 discrepancy decreases. At 4000 MWd/MgU the PPV predictions are still more reactive. At 8000 MWd/MgU the PPV infinite multiplication factors are less than the WIMS estimations.
CLUe Th B prediction WIMclosee e sth ar o t rS ones V one t 800PP A .s e 0thath o nt MWd/MgU the Argentinean WIMS estimation is more reactive than the PPV one by 6.19 mk, and more reactive than the CLUB one by 3.21 mk.
3.1.2.2. Effective multiplication factors
Reference Lattice Cell
The comparison analyses offer the same conclusions as for the infinite multiplication constant (Table 3.1.5). Inside each famil resulte yth rathe e sar r clos eaco et e h th other r Fo . fresh clean fuel case the WIMS and CLUB estimations are almost identical. The PPV effective multiplication constant is larger than the WIMS one by 6.42 mk. At 8000 MWd/MgU the WIMS-keff is larger than the PPV one by 9.28 mk; it is also larger than the CLUB keff by 2.72 mk.
TABLE 3.1.5 EFFECTIVE MULTIPLICATION FACTORS REFERENCE LATTICE CELL
Code BURNUP (MWD/MGU) Country 0 0* 4000 8000
Argentin- 1 V PP a 1.090689 1.046906 1.019201 .946454 Canad- 2 V PP a 1.048151 1.018798 .944345 3 PPV - Romania .1.091968 1.048056 1.018675 .944316
4 WIMS - Argentina 1.084372 1.052159 1.018482 .952661 5 WIM KoreS- a 1.086860 1.044501 1.015351 .950098 6 WIMS - Pakistan 1.0758 1.0439 1.0277 .9610 7 WIMS - Romania 1.076294 1.040906 1.010698 .948704
8 CLU IndiB- a 1.0840 1.0511 1.0145 .9502 9 CLIMA Indi- X a 1.0838 1.0515 1.0150 .95146 10 RHEA - India 1.08813 1.04981 1.01214 .947427
11 4 3 l De - 6.42 '3.72 - 0.19 9.28 12 8 3 l De - 6.73 2.76 - 4.04 6.56
13 1 3 l De - 1.07 - 1.05 0.51 2.39 14 2 3 l De 0.09 0.12 0.03
15 Del 45 2.11 - 6.97 -3.03 -2.83 16 Del 46 - 7.35 - 7.52 8.81 9.11 17 Del 47 - 6.92 - 10.27 -7.56 -4.38
18 Del 89 - 0.17 0.36 0.49 1.39 19 0 De81 l 3.50 -1.17 -2.30 -3.08
= 1000 j i . l * {1/kDe 1/k- ; 1 ,
25 3.1.2.3. U-235 number densities (g/kg initial U)
The results are presented in Table 3.1.6. Some small discrepancies may be seen even frese foth r h fuel case like 0.0. 6%
The Canadian and Rumanian versions of the PPV code are almost identical. The Argentinean PPV predicts a larger U235 density at 8000 MWd/MgU by 2.34%.
The Argentinea Koread nan n WIMS prediction vere ar sy similar Rumaniae Th . n- WIMS U235 density estimations are smaller by 5.42 % at large burnups of 8000 MWd/MgU.
CLUe Th B estimation CLIMAe close th sar o et X one smalled san r tha RHEe nth A predictions.
TABLE 3.1.6 U 235 NUMBER DENSITIES (G/KG INITIAL U )
Code BURNUP (MWD/MGU) Country 0 0+ 4000 8000
Argentin- 1 V PP a 7.110 7.110 3.822 2.044 Canad- 2 V PP a 7.1138 7.1138 3.7859 1.9957 Romani- 3 V PP a 7.1137 7.1137 3.7885 1.9972
4 WIM Argentin- S a 7.1097 7.1097 3.8368 2.0730 5 WIM Kore- S a 7.1097 7.1097 3.8001 2.0286 6 WIM Pakista- S n 7.110 7.110 3,791 2.006 7 WIM Romani- S a 7.1137 7.1137 3.7465 1.9606
8 CLUB - India 7.1138 7.1138 3.7962 2.0195 9 CLIMAX - India 7.1140 7.1140 3.8067 2.0494 10 RHEA - India 7.1138 7.1138 3.9836 2.2042
11 Rel 34 -0.06 -0.06 1.27 3.80 12 Rel 38 0.00 0.00 0.20 1.17
13 Rel 31 -0.05 -0.05 0.88 2.34 14 Re2 3 l 0.00 0.00 -0.07 -0.08
15 Rel 45 0.00 0.00 -0.96 -2.14 16 Re6 l4 0.00 0.00 -1.19 -3.23
17 Re7 l4 0.06 0.06 -2.35 -5.42
18 9 8 l Re o.oo- 0.00 0.28 1.48 19 0 Re81 l 0.00 0.00 4.94 9.15
•D=,I * inn 1 z-\
26 WIMe Th S computed densitie littla e esar large oneV ry t 3.8 sa PP b tha e 0n% th 8000 MWd/MgU CLUe Th . B estimation onesV PP . e close th sar o et
3.1.2.4. Pu-239 number densities (g/kg initial U) e codTh e prediction seee b Tabl n ni y versionV ma es PP 3.1.7 l sAl .giv e very similar results. The WIMS predicted densities are larger for the Rumanian version and smalle Koreae th r rfo n one t 800.A 0 MWd/MgU discrepanciee th , -2.21% d 7.4e an s ar 5% , the Argentinean prediction being taken as reference. The Indian results are close to each other.
TABLE 3.1.7 PU 239 NUMBER DENSITIES (G/KG INITIAL U )
Code BURNUP (MWd/MgU) Country 4000 8000
Argentin- 1 V PP a 1.928 * 2.407 2 PPV - Canada 1.9412 2.4136 Romani- 3 V PP a 1.9387 2.4125
4 WIM Argentin- S a 1.9792 2.5390 5 WIMS - Korea 1.9282 2.4828 6 WIMS - Pakistan 2.021 2.557 7 WIM Romani- S a 2.0949 2.7282
8 CLUB - India 1.9309 2.4895 9 CLIMA Indi- X a 1.9473 2.5119 10 RHEA - India 1.8613 2.4702
11 Rel 34 2.09 5.24 12 Rel 38 - 0.40 3.19
13 1 3 l Re - 0.55 - 0.23 14 2 3 l Re 0.13 0.46
15 5 4 l Re - 2.58 - 2.21 16 Rel 46 2.11 0.71 17 Rel 47 5.85 7.45
18 Rel 89 0.85 0.90 19 Rel 810 - 3.60 - 0.78
N3 - = j i l Re * 100 ('•
The WIMS and the CLUB estimations are larger than the PPV one by 5.24 % and 3.1 respectivel9% t 800ya 0 MWd/MgU.
Now, we may note that both the U235 and Pu239 number densities predictions are
27 larger for the WIMS calculations than for the PPV ones at 8000 MWd/MgU. Consequently, althougeffectivV PP e heth multiplication facto larges i r rr tha fo WIM6.4e y nb k th 2e ra Son the fresh fuel case, the WIMS excess reactivity is larger than the PPV one by 9.28 mk at 8000 MWd/MgU. 3.1.2.5 Relative powerouterthe in ring rods
The result givee ar s Tabln ni e 3.1.8.
The PPV code does not compute separate powers for the different rings of the cluster. e PakistaniTh , Rumania e Argentineath d an n n WIMS versions predict almost identical results. The Korean estimations are also very close.
CLUe Th B prediction WIMclose e sth ar o et S one fresr sfo h fuel becomt bu , e smaller burnue asth p increases e spatiaTh . l power distributio mors ni e uniform.
The CLIMAX estimations are larger than the WIMS ones for all the burnups, the relative differences being 2.66-2.7. 3%
The RHEA prediction alse sar o larger tha WIMe nth S one 1.21-1.7y sb . 8%
TABLE 3.1.8 RELATIVE POWER IN THE OUTER RING RODS
Code BURNUP (MWd/MgU) Country 0 0+ 4000 8000
1 WIM Argentin- S a 1.12682 1.12990 1.12640 1.11573 2 WIMS - Korea 1.12506 1.12762 1.11867 1.11093 3 WIM Pakista- S n 1.1257 1.1306 1.1318 1.1189 4 WIM Romani- S a 1.1256 1.1298 1.1258 1.1176
5 CLU Indi- B a 1.1222 1.1250 1.1065 1.0919 6 CLIMAX - India 1.1568 1.1606 1.1572 1.1458 7 RHE Indi- A a - — — 1.15 1.14 1.13
8 Rel 12 -0.16 -0.20 -0.69 -0.43 9 3 1 l Re 0.06 0.06 0.48 0.28 10 4 1 l Re -0.11 -0.01 -0.05 0.17 11 5 1 l Re -0.41 3 -0.4 -1.77 -2.14 12 6 1 l Re 2.66 2.72 2.73 2 .70 13 Re7 1 l — — — 1.78 1.21 1.28
= Re j i l 100
3.1.2.6 Void-induced reactivity
The Argentinean, Canadia Rumaniad nan estimationV nPP almose sar t identical (see Table 3.1.9). The void reactivity is near 15 mk for the fresh fuel, about 9 mk at 4000 MWd/MgU and close to 7 mk at 8000 MWd/MgU.
28 TABLE 3.1.9 VOID-INDUCED REACTIVITY (MK)
Code BURNUP (MWd/MgU) Country 0 0+ 4000 8000
1 PPV - Argentina 14 .49 15 .12 9.28 7.00 Canad- 2 V PP a 15 .11 9.28 6.92 3 PPV - Romania 14 .51 15 .14 9.26 6.91
4 WIMS - Argentina 16 .82 17 .37 13 .76 13 .72 5 WIMS - Korea 14 .87 15 .42 12 .12 11 .90 6 WIM Pakista- S n 16 .76 17 .15 11 .86 10 .51 7 WIM Romani- S a 13 .53 13 .98 15 .37 17 .81
8 CLUB - India 15 .97 16 .63 13 .51 13 .66 9 CLIMAX - India 16 .73 17 .16 12 .58 10 .38 10 RHE - IndiA a 19 .71 20 .27 12 .86 10 .07
11 4 3 f Di 2.31 2.23 4.50 6.81 12 8 3 f Di 1.46 1.49 4.25 6.75
13 1 3 f Di - 0.02 - 0.02 0.02 0.09 14 Dif 32 — — — — - 0.03 0.02 0.01
15 5 4 f Di - 1.95 - 1.95 - 1.64 - 1.82 16 Dif 46 - 0.06 - 0.22 - 1.90 - 3.21 17 Dif 47 _ o .29 - 3.29 1.61 4.09
18 Dif 89 0.76 0.53 - 0.93 - 3.28 19 0 81 f Di 3.7-4 3 .64 - 0.65 - 3 .59
Dif ij (mk)
Alothee lth r prediction significantle sar y different frese th hr fueFo . l cas WIMe eth S estimation e 17. (Argentina)ar sk 4m , 15. (Korea)k 4m , 17. (Pakistank 2m 14.d k an )0 m (Romania) t largeA . r burnup decrease th voie , th dinduce- f eo d reactivit mucs yi h smaller predictionsV PP e thath r nfo : 13. (Argentina)k 7m , 11. (Koreak 9m 10.d (Pakistank )an 5 m ) at 8000 MWd/MgU unknowo t e Du . n reasons, Rumanian estimation t higa s h burnupe sar similar or even larger than the fresh fuel case ones.
The Indian results are not very close to the WIMS ones, but the burnup effect is clearly differen oneV t PP tha. e nth
CLUe Th B void-induced reactivitie frese th hr e 16.t fo fue13.d a ar s k an 6k l7 m m 8000 MWd/MgU. The CLIMAX predicted reactivities are 17.2 mk for the fresh fuel and t 800a 10.k 0m 4MWd/MgU RHEe Th . A frese estimationth hr fo fue k lnea m e cass0 ar r2 e t 800a ank 0dm MWd/MgUclos0 1 o et .
29 3.1.2.7 Fuel temperature induced reactivity
Once again Argentineane th , , Canadia Rumaniad nan predictionV nPP vere sar y close (see Table 3.1.10): 6.0 mk for the fresh fuel case, 2.5 mk at 4000 MWd/MgU and 0.6 mk at 8000 MWd/MgU.
TABLE 3.1.10 FUEL TEMPERATURE INDUCED REACTIVITY (MK)
Code BURNUP (MWd/MgU) Country 0 o- 4000 8000
1 PPV - Argentina 5.78 6.04 2.50 0.64 2 PPV - Canada -___ 6.03 2.49 0.57 3 PPV - Romania 5.78 6.03 2.47 0.57
4 WIMS - Argentina 5.78 5.80 2.03 0.37 5 WIM Kore- S a 5.66 5.55 2.80 2.03 6 WIMS - Pakistan 5.8 5.9 3 .1 2.3 7 WIMS - Romania 6.18 6.27 3 .11 1.71
8 CLU - IndiB a 5 .58 5.76 2.81 1.88 9 CLIMAX - India 5 .68 5.76 2.88 1.84 10 RHEA - India 4.19 4.34 2.31 0.61
11 Dif 34 0.00 - 0.23 - 0.44 - 0.20 12 8 3 f Di - 0.20 - 0.27 0.34 1.31
13 1 3 l Re 0.00 0.01 0.03 0.07 14 Rel 32 0.00 0.02 0.00
15 Rel 45 - 0.12 - 0.25 0.77 1.66 16 Rel 46 0.02 0.10 1.07 1.93 17 Rel 47 0.40 0.47 1.08 1.34
18 Rel 89 0.10 0.00 0.07 - 0.04 19 0 81 l Re - 1.39 _ T .42 - 0.5 - 1.27
Dif ij N3 (mk)
The WIMS results indicate the same burnup evolution. The Argentinean version reactivity estimations are close to the PPV ones: 5.8 mk for the fresh fuel and 0.4 mk at 8000 MWd/MgU. The Korean, Pakistani and Rumanian WIMS versions predicted larger effects at 8000 MWd/MgU: 2.0 mk, 2.3 mk and 1.7 mk respectively.
The CLUB and CLIMAX estimations are almost identical and very close to the Korean WIMS ones. The RHEA predictions are smaller than all the other estimations, but the differences are not very significant.
30 3.1.2.8 Moderator purity induced reactivity
All the predictions are similar (see Table 3.1.11). The purity of the moderator is very importan heave th r y tfo wate r notreactivitreactorsy V ema PP tha e e W tth y . effect (4.) 0mk is a little larger than the WIMS predicted effect (3.5 mk) at 8000 MWd/MgU. The CLUB and CLIMAX estimations of 3.2 mk are near the WIMS ones.
TABLE 3.1.11 MODERATOR PURITY INDUCED REACTIVITY (MK)
Code BURNUP (MWd/MgU) Country 0 0+ 4000 8000
Argentin- 1 V PP a 3 .61 3 .68 3 .61 3 .99 2 PPV - Canada 3 .64 3 .58 3 .96 Romani- 3 V PP a 3 .61 3.67 3 .61 4.00
4 WIM Argentin- S a 3 .40 3 .43 3 .32 3 .56 5 WIM Kore- S a 3 .37 3 .42 3 .16 3 .32 6 WIMS - Pakistan 3 .04 3 .03 2.78 3 .62 7 WIM Romani- S a 3 .50 3 .52 3 .32 3 .47
8 CLUB - India 3 .22 3 .25 3 .00 3 .20 9 CLIMAX - India 3 .39 3 .42 3 .10 3 .24 10 RHEA - India 2.98 2 .87 3 .09
11 Dif 34 - 0.21 - 0.24 - 0.29 - 0.44 12 8 3 f Di - 0.39 - 0.42 - 0.61 - 0.80
13 1 3 f Di 0.00 0.01 0.00 - 0.01 14 2 3 f Di — - - - - 0.03 - 0.03 - 0.04
15 5 4 f Di - 0.03 - 0.01 - 0.16 - 0.24 16 6 4 f Di - 0.36 - 0.40 - 0.54 0.06 17 7 4 f Di 0.10 0.09 0.00 0.09
18 Dif 89 0.17 0.17 .0 .10 0.04 19 Dif 810 — _ — - 0.27 - 0.13 - 0.11
Dif ij (mk)
3.1.2.9 Boron moderator induced reactivity
The three PPV predictions are almost identical (see Table 3.1.12). The WIMS estimations are rather close to each other. The Indian codes predicted similar results. Unfortunately, ther e discrepanciear e s betwee e predictionnth e threth f eo s lattice code boroV familiesPP n e induceTh . d reactivity estimatio t variouna s burnup larges si r thae nth
31 TABLE 3.1.12 BORON MODERATOR INDUCED REACTIVITY (MK/-1 PPM) ABSOLUTE VALUES
Code BURNUP (MWd/MgU) Country 0 0+ 4000 ' 8000
Argentin- 1 V PP a 8.22 8.40 7.94 8.42 Canad- 2 V PP a 8.32 7.88 8.37 Romani- 3 V PP a 8.22 8.39 7.94 8.44
4 WIMS - Argentina 7.72 7.80 7.21 7.47 5 WIMS - Korea 8.25 8.39 7.43 7.35 6 WIMS - Pakistan 7.66 7.74 7.05 7.27 7 WIMS - Romania 7.95 8.01 7.21 7.17
8 CLU - IndiB a 7.46 7.57 6.75 6 .58 9 CLIMAX - India 7.55 7.64 6.78 6 .69 10 RHE - IndiA a — _ _ _ 7.31 6.77 1.00
11 Dif 34 - 0.50 - 0.59 - 0.73 - 0.97 12 8 3 f Di - 0.76 - 0.82 - 1.19 - 1.86
13 Dif 31 0.00 0.01 0.00 - 0.02 14 Dif 32 — — — — - 0.07 - 0.06 - 0.07
15 5 4 f Di 0.53 0.59 0.22 - 0.12 16 Dif 46 - 0.06 - 0.06 - 0.16 - 0.20 17 Dif 47 0.23 0.21 0.00 - 0.30
18 9 8 f Di 0.09 0.07 .0 .03 .0 .11 19 Dif 810 — — -*— - 0.26 0.02 0.42
Dif ij I (rappm/- k )
WIMS prediction by 0.5 ~ 1.0 mk/-lppm and also than the CLUB estimation by 0.8 -1.9 mk/-lppm frese th hr cleaFo . n fue l - Romaniacase r instanceV fo ,PP e , th ,WIM S- Argentin CLUd an a B- India-ThP D codes offer 8.22, 7.7 7.4d 2an 6 mk/-lppm boron, respectively t 400A . 0 MWd/MgU estimatione th , e 7.94ar s , 7.2 6.7d 1an 5 mk/-lppm, respectively. The discrepancies are due both to the physical treatment of the borated lattice nucleae th o t rceld dataan l .
During the Phase B Commissioning measurements the precision of the boron induced reactivity calculation is very important. Consequently there are modified PPV versions that clai moffeo t r better treatmen boratee th f o t d moderator lattice celll al . Howeverw no o t p u , the PPV versions used to solve the IAEA benchmark problems utilized the standard AECL treatment.
32 3.1.2.10 Weight U/bundleof induced reactivity
Allattice lth e code estimation similae sar r (see Table 3.1.13) effece Th negligibl.s i t e frese foth r h increasefued an l s whe burnue nth p increases. However largese th , t reactivity prediction is only 1.06 mk at 8000 MWd/MgU.
TABLE 3.1.13 WEIGHT U / BUNDLE INDUCED REACTIVITY (MK)
Code BURNUP (MWd/MgU) Country 0 0+ 4000 8000
Argentin- 1 V PP a 0.02 0.00 0.56 1.02 2 PPV - Canada 0.01 0.58 1.05 Romani- 3 V PP a 0.02 0.01 0.58 1.06
4 WIM Argentin- S a 0.16 0.15 0.46 0.87 5 WIM Kore- S a - 0.10 - 0.03 0.39 0.89 6 WIMS - Pakistan - 0.06 - 0.08 0.24 0.72 7 WIMS - Romania - 0.06 - 0.04 0.30 0.74
8 CLU - IndiB a 0 .00 0. 00 0.49 1.00 9 CLIMA - IndiX a 0.09 0.00 0.29 0.82 10 RHE Indi- A a - 0.30 - 0.33 0.09 0.53
11 4 3 f Di 0.14 0.14 - 0.12 - 0.19 12 Dif 38 - 0.02 - 0.1 - 0.09 - 0.06
13 Dif 31 0.00 0.01 - 0.02 - 0.04 14 2 3 f Di — — — — 0.00 0.00 - 0.01
15 Dif 45 - 0.26 - 0.18 - 0.07 - 0.02 16 Dif 46 - 0.22 - 0.23 - 0.22 - 0.15 17 Dif 47 - 0.22 - 0.19 - 0.16 - 0.13
18 9 8 f Di 0.0-9 0.00 - 0.20 - 0.18 19 Di0 f81 - 0.30 - 0.33 - 0.40 - 0.47
Dif ij (mk)
3.1.2.11 Enrichment induced reactivity
The PPV code is not recommended for the enriched fuel lattice cell calculations. However frese th hr fo fue, l case eved an st 400n a 0 MWd/Mg estimationV PP e Uth e ar s close to the WIMS ones. All the codes predicted similar reactivities when the burnup is not too high. (See Table 3.1.14).
33 TABLE 3.l.14 CASE 7 . ENRICHMENT INDUCED REACTIVITY (MK) ENRICHMENT = 1.2 % U235
Code BURNUP (MWd/MgU)
Country 0 0 4000 8000 20000
Argentin- 1 V PP a 147 .5 149 .1 122 .5 125 .2 -172 .7 2 PPV - Canada - 148 .6 122 .6 125 .8 -178 .0 3 PPV - Romania 147 .0 148 .7 122 .7 125 .8 -178 .0
4 WIMS - Argentina 145 .1 145 .9 123 .2 125 .6 -138 .4 5 WIMS - Korea 144 .5 148 .0 122 .3 119 .6 -139 .8 6 WIMS - Pakistan 146 .4 147 .2 120 .8 123 .8 _ _ _ _ 7 WIMS - Romania 145 .5 145 .5 121 .2 117 .7 -135 .1
8 CLUB - India 145 .2 146 .4 122 .6 119 .4 -136 .5 9 CLIMAX - India 145 .1 146 .2 121 .4 117 .1 -131 .1 10 RHE Indi- A a 142 .5 143 .5 128 .5 130 .1 -119 .2
11 4 3 f Di - 1.9 - 2.8 0.5 - 0.2 39 .6 12 Dif 38 - 1.8 - 2.3 - 0.1 - 6.4 41 .5
13 Di1 3 f 0.5 0.4 - 0.2 - 0.6 5 .3 14 Dif 32 - 0.1 - 0.1 0.0 0.0
15 Dif 45 - 0.6 2.1 - 0.9 - 6.0 -1 .4 16 Di6 4 f 1.3 1.3 - 2.4 - 1.8 17 Dif 47 0.4 - 0.4 - 2.0 - 7.9 3 .3
18 Dif 89 - 0.1 . - 0.2 - 1.2 - 2.3 5.4 19 0 Di81 f _ 2 .7 - 2.9 5.9 10 .7 17 .3
= o rh * 1 - 1/k j i f Di Nj - ^ (mk)
At 8000 MWd/MgU, an unexpected high enrichment induced reactivity of 125.6 mk, close to the PPV prediction is estimated by the Argentinean version of WIMS. All the other codes estimations RHEexcepe th r Afo t cod betweee ear n 117.Im . 119.d kan mk 6
At 20000 MWd/MgU, the enrichment induced reactivities are not available, because the natural uranium case burnup is limited to 8000 MWd/MgU. The absolute reactivities have estimationV PP bee e - 1/k nsignificantl1 e Th .calculate sar = o rh s dya largere thath l nal other ones. BotpredicteV hPP d U23 Pu23d 5an 9 number densitie lese sar s tha WIMe nth S ones. The WIMS predictions are close to each other, and close to the CLUB and CLIMAX estimations.
34 3.1.3. Access to complete information
Whe nphysicisa t solve benchmarsa k problem usee h , specifisa c code availabln a , e computer, a given nuclear data library etc. Suppose he finds some discrepancies between his marn results and the IAEA participants main results. If he wishes to have the correct explanation of these discrepancies, more information is needed.
vera r y Fo detailed investigation, even some basic physical data use preparo dt e eth code input data, are necessary: Avogadro number, atomic mass number for plutonium isotopes, isotopic compositio naturae th f no l boron etc. Such basi e founb c daty n di ama Refs. [36[42]d an ] .
The PPV and WIMS input data may be found in Ref. [35], [36], [40] and [42]. However, we may recommend the Ontario Hydro report [4] that presents the PPV input preparation details. The Korean report [40] presents not only very detailed WIMS results, but also a useful note on input data preparation. Also, we may recommend the Argentinean report [35] for the use of two different WIMS input options.
3.1.4. Conclusions
Ten codes have been used to solve the lattice cell benchmark problems: three versions - AEC V LPP codee oth f , four version WIMe th f o Ss- Winfrit h codthred an e e Indian codes: CLUB, CLIMA RHEAd Xan . Althoug resulte hth rathee sar r similar, some important discrepancies must be noted. They are related to the void effect, boron moderator reactivity and enriched Uranium cases.
The void effects predicted by the three versions of the PPV code are close to each other, but much different from the other seven estimations. The PPV void-induced reactivity significantly decreases with increasin gt 800 fres a e burnup k th 0 hr m fuefo 7 k ,o t lfro m 5 m1 MWd/MgU WIMSn i t Bu . , CLU CLIMAd Ban decrease X th voi e th dn ei induce d reactivity is much less significant.
boroe Th n moderator induced reactivity estimation t closno ee enougar s eaco ht h predictionV PP other e almose Th . ar s t identical, lik/ -Ippe mk nea m9 7. rboro t 400na 0 MWd/MgU; this value is by 10.1 % larger than the WIMS - Argentina estimation, and by 17. large6% r than CLUB Indi- a prediction WIMe Th . S estimation close sar eaco et h other, and the Indian codes predictions are also close to each other.
The PPV code is not recommended for enriched fuel lattice cell calculations. For burnups less than 4000 MWd/MgU, the PPV enrichment induced reactivity predictions are WIMclose th o et SU23e onesPu23th d s 5an A . 9 number densitie lese s predictear s V PP y db than those estimate WIMSy db discrepance th , y increases with increasing burnup t 20,00A . 0 MWd/MgU the PPV results may no longer be utilized.
Although all the predicted effective multiplication factor values are close to each noty otherema tha e burnue w ,t th p evolutio similarfrest e th no hr ns i fueFo . lV casesPP e th , effective multiplication factors are larger than the WIMS ones. As the burnup increases, the predictV PP s less Pu-239 than WIMS. Consequentlyd WIMan SV k-effectivPP e th , e estimations become closer. For large burnup case of near 8000 MWd/MgU, the WIMS effective multiplication factor predictio s largeni onee IndiaV rTh PP . tha e n nth code s estimation simila e WIMe sar th o t rS ones.
35 3.2. TAS : BENCHMARK2 SUPERCELN KO L CALCULATIONS
In PHWRs reactivite th , y devices (RDsgeneralle ar ) y located perpendiculae th o t r cele th lfuel t homogenizeorden I ge . o t r d parameter presence th n i s f suceo hreactivita y o considet s devicha a threre on e dimensional cell called 'Supercell'. This section summarize wore sth k relate supercele th o dt l calculation part (reactivity devices calculations) of the benchmark. This benchmark was analyzed by four countries, namely, Argentina, Canada, Indi Romaniad aan .
3.2.1. Description of supercell benchmark [43]
For the purpose of defining the benchmark problem, three different kinds of reactivity devices were chosen, namely adjuste) (a , r rod ) liqui(b s d zone controller ) shuf (c of td san rods basie Th .c supercell datdescriptiod aan thesf no e reactivity device givee sar Tabln i e 3.2.1 supercele Th . shows i l Fign ni . 3.2.1 reactivite Th . y device sbenchmare useth n di k problem are simplified devices which maintain the essential features of the complex actual devices. Parameters of the reactivity devices (for example rod radius etc.) are chosen so that reactivite th y worth thesf so e reactivity device nearle sar y equa thoso lt reactivitf eo y devices presen reaa n i lt reactor.
For each reactivity device, two supercell calculations namely, with and without reactivity device were required to obtain the incremental cross-sections. For zone controllers three cases were required namely, (a) light water present hi the compartment, (b) empty compartmen referenca ) (c d ean t case wit reactivito hn y device. Table 3.2.2 listfive th se supercell cases to be modelled.
Several items of the code output were to be provided hi order to mutually compare the results from different supercell codes : (a) the infinite multiplication factor of the supercell (b) the flux distribution through the cell by giving reaction rates for scattering, productio absorptiod nan functioa s na positiof no n wherever thi possibles si additionn I . e th , incremental cross sections to be used in the core calculations were to be given for comparison.
3.2.2. Supercell modellin computed gan r codes used
Since the supercell contains strong absorbers like fuel and reactivity devices either a transport theor useye b boundar r coddo o t s ei y condition reactivitfuer d sfo an l y devicee sar calculated with a transport theory code and then diffusion theory is applied for the moderator and other materials wher absorptioe eth weaks ni .
Various computer code differene s th use y db t participant describee sar d below:
3.2.2.7. MULTICELL and SHETAN: Argentina [44]
Two relatively independent chains were used. The first one is based on the Canadian secone lattic Britisth e base e d th don en an d ho cod V codePP e WIMS-D4codef o t sse o Tw . used are as follows : Chai n1 Chain2 V PP LatticWIMS-De 4 Supercell MULTICELL SHETAN/DELFIN
36 TABLE 3.2.1 IAEA BENCHMARK : SUPERCELL DATA FOR REACTIVITY DEVICES
Supercell : Cell Dimension 14.287 y 14.2875b 25.y 5b m 0c Adjuste: d rro Form Solid cylinderical rod Material Pure Fe Density 7.85 g/cc Temperature C ° 8 6 Orientation Perpendicular to fuel channel d runninan g full vertical heighf to supercell Rod radius m c 1 .6 : d ro Shuf tof Form Solid cylindricad lro Material Pure Cd (12.27 wt% cd-113) Density 8.65 g/cc Temperature C ° 8 6 Orientation Perpendicula fueo rt l channed lan running full vertical height of supercell Rod radius m c 0 5. Zone controller : Form Cylindrical tube filled with light water at 68 °C Light water density 0.9789 g/cc Temperature 68 °C Tube Material *Zr Density 6.55 g/cc Orientation Perpendicular to fuel channel and running full vertical height of supercell Rod radius 5.0 cm Dimensions : Inside diameter m c 12.2 Wall Thickness 0.14 cm
In the chain 1, the supercell calculations have been done using the computer code MULTICELL code Th e. uses boundary conditions calculated with transport theory [45n ]i strong absorbers suc soms d fueha an l e reactivity devices and, with diffusion theorn yi moderator and in materials where the absorption is weak. In chain 2, the supercell calculations are done using computer code SHETAN.
37 reactivity moderator
1/2 lattice pitch FIG. 3.2.1 The 3-D Supercellfor Reactivity Device Simulation
TABLE 3.2.2 SUPERCELL CALCULATIONS
Case Description 1 Reference Supercell 2 Reference Supercell with adjusted rro 3 Reference Supercell with control rod 4 Reference Supercell with zone controller (filled with water) Reference Supercell with zone controller (without water)
38 3.2.2.2. MULTICELL: Canada [46]
The incremental cross sections for all reactivity devices were calculated using the three dimensional supercell code MULTICELL.
3.2.2.3. BOXER-3, PHANTOM, 3D-FAST and CALC: India [47, 48]
India-ThPD used three codes namely transpora , t theory model BOXER-o tw d 3an diffusion theory models, PHANTOM and 3D-FAST.
India-RED usee computeth d r code CALC. Sinc e modeo th etw a l uses wa d dimensional one, exac te supercel parallelth l al f l o soutput s wer t generatedno e . Only incremental cross section givee sar Tablen ni s 3.2. 3.2.12- 9 .
3.2.2.4. MULTICELL: Romania [49] supercele th r Fo l calculation code sth e MULTICELL supplie AECy db L Canad useds awa . Although the code allows different user recipes, it was tried to follow the basic procedures of réf. [13]. However the "fast" group treatment is not identical to the common one. The reference case homogenized "fast" absorption cross section is identical to the PPV one. Consequently the supercell k-infand the "fast" group reaction rates are slightly modified. All e fueth l moderation ratee decreasedar s n comparisoi , e otheth o rt n participants' MULTICELL predictions, for instance. As the code cannot explicitly represent the cylindrical regions ratie th f volum, oo surfaco et e are s alteredai e surfacTh . ee areth f ao conserveds fuelwa d , ro shud adjustef .an of tConsequently d ro r volumee th , f fued so an l device rod e decreasedar s conservo T . moderatoe th e r volume annulue th , s volums ewa altered.
3.2.3. Results and discussions
All participants used the cross sections for the fuel, moderator and annulus region as calculated by their lattice cell code at 4000 MWD/T. The supercell benchmark problem was analyzed by four countries, namely, Argentina, Canada, India and Romania. These results were discusse firse th t n dIAEi A research coordinated meeting (RCM) hel Buenon di s Aires, Argentina [50]. The local worth of various reactivity devices, reaction rates in the supercell and incremental cross sections for the reactivity devices have been discussed below:
3.2.3.1. Worth reactivityof devices supercellin
Table 3.2.3 shows the infinite multiplication factors (K«,) of the supercell without reactivity devices e worthth , f adjusteo s r rods, shutoff rodzond an se controller unis a t calculate variouy db s code differenf so t countries MULTICELe th s A . L code solve sourcsa e problem for IFS = 1, the significance of "k-inf" and "local worths" are not very clear. In addition, afte flue th rx convergenc reacheds ei "faste th , " fluxe onld e "fastan syth " fluxes renormalizede ar . Consequently MULTICELe th , L reported k-inf t obtaine valueno e sar s da solution diffusioe th f so n equation mattea s factA f . o r , onl MULTICELe yth L incremental cross section utilizede ar s e averagTh . e infinite multiplication facto r supercelfo r l without reactivity devices is 1.0458 and standard deviation is 7.1 mk. The average reactivity worths of adjuster rods, shutoff rods, empty zone controllers and zone controllers filled with light water are 132.8 mk, 676.4 mk, 10.8 mk and 284.7 mk respectively. The standard deviations
39 TABLE 3.2.3 K» AND LOCAL WORTHOF REACTIVITY DEVICES
Country K* rho rho rho rho AR ZC-H2O ZC+H20 CR (nk) (ink) (ink) (HÜC) PP-V+MULTICELL Argentina 1.0481 -139.48 -12.35 -282.2 -684.6 WIMS-D4+SHETAN Argentina 1.0412 -135.109 2 9 - -287.3 -713.6 PP-V+MULTICELL Canada 1.0475 -144.0 -16.5 -293.1 -696.2 BOXER 3 India 1.0390 -120.6 ——— -282.8 -678.2 PHANTOM India 1.0417 -121.0 -11.6 -289.9 -654.2 CLUB+3D-FAST India 1.0415 -134.6 ____ -290.8 -673.05 PP-V+MULTICELL Romania 1.0614 -135.0 -4.42 -267.0 -635.0
Average 1.0458 -132.8 -10.8 -284.7 -676.4 Std. Deviation 0.0071 8.2 4.0 8.1 24.1 Percentile Difference 2.1% 16.2% 73.2% 8.9% 11.0% (max. - min.) /max.
for these four cases are 8.2 rnk, 24.1 mk, 4.0 mk and 8.1 mk. The percentage standard deviation for these four cases are 6.2, 3.6, 37.0 and 3.6 respectively. Thus it can be seen that maximum difference is in the prediction of empty zone controller tube case.
3.2.3.2. Reaction rates in various regions of supercell
Tablen I s 3.2.4-3.2. reactioe 8th n rates suc moderations ha , absorptio productiod nan n for the reference, adjuster rod, shutoff rods, empty zone controller tube and zone controllers filled with light water cases are given. Reaction rates are normalized such that total production rat unitys ei average Th . e standard deviatio percentagd nan e standard deviation for some of the important reaction rates for various cases are summarized below :
Standard Percent Case Reaction rate Average deviation standard deviation
Reference Absorptio fuen ni l 0.89670 0.0350 3.9 Absorptio Annulun ni s 0.03005 0.0005 1.6 Moderatio moderaton ni r 0.79470 0.0195 2.4
Adjuster Absorptio Adjusten ni r 0.1215 0.0090 7.5 rod rods
Shutoff Absorption in Shut off 0.6184 0.0160 2.5 rod rod
Zone ModeratioU ZC n ni 0.2331 0.0240 10.3 Controller with Absorption in ZCU 0.2678 0.0050 1.7 light water
40 TABLE 3.2.4 REACTION RATES (REFERENCE CAS) E NORMALIZE( TOTAO DT L PRODUCTIO) 0 1. N=
»Région Code Séquence Country Moderation Absorption Production PP-V + MULTICELL Argentina 0.0241 0.9059 1.0 WIMS-D4 + SHETAN Argentina 0.0272 0.9213 1.0 PP- MULTICEL+ V L Canada 0.0242 0.8211 1.0 Fuel BOXER-3 India 0.0268 0.9198 1.0 PHANTOM India 0.0284 0.9157 1.0 PP- MULTICEL+ V L Romania 0.0219 0.8963 1.0
PP- MULTICEL+ V L Argentina 0.0069 0.0304 —— WIMS-D4+SHETAN Argentina 0.0008 0.0297 Annulus PP-V+MULTICELL Canada 0.0069 0.0305 BOXER-3 India 0.0008 0,0292 —— PHANTOM India 0.0013 0.0301 —— -PP- MULTICEL+ V L Romania 0.0068 0.0304 ——
PP-V + MULTICELL Argentina 0.8138 0.0156 —— Moderator WIMS-D4+SHETAN Argentina 0.7824 0.0135 PP-V+MULTICELL Canada 0.8143 0.0156 BOXER-3 India 0.7658 0.0137 —— PHANTOM India 0.7796 0.0144 —— PP-V + MULTICELL Romania 0.8123 0.0155 ——
TABLE 3.2.5 REACTION RATES (ADJUSTE CASED RO R ) S ( NORMALIZED TO TOTAL PRODUCTION = 1.0 )
Region Code Sequence Country Moderation Absorption Production PP- MULTICEL+ V L Argentina 0.0256 0.9113 1.0 WIMS-D4 + SHETAN Argentina 0.0280 0.9242 1.0 PP- MULTICEL+ V L Canada 0.0258 0.8211 1.0 Fuel BOXER-3 India 0.0305 0.9314 1.0 PHANTOM India 0.0320 0.9272 1.0 PP-V + MULTICELL Romania 0.0233 0.9011 1.0 PP-V -I- MULTICELL Argentina 0.0073 0.0304 ——— WIMS-D4+SHETAN Argentina 0.0009 0.0298 —— Annulus PP-V+MULTICELL Canada 0.0073 0.0304 —— BOXER-3 India 0.0010 0.0297 —— PHANTOM India 0.0014 0.0306 —— PP-V + MULTICELL Romania 0.0073 0.0303 ——
PP- MULTICEL+ V L Argentina 0.8596 0.0152 ——— Moderator WIMS-D4+SHETAN Argentina 0.8160 - 0.0132 —— PP-V+MULTICELL Canada 0.8598 0.0153 ——— BOXER-3 India 0.8569 0.0133 —— PHANTOM India 0.8725 0.0137 ——— PP-V + MULTICELL Romania 0.8578 0.0152 ——
PP-V + MULTICELL Argentina 0.0007 0.1265 ——— Adjuster WIMS-D4+SHETAN Argentina 0.0012 0.1236 —— rod PP-V+MULTICELL Canada 0.0012 0.1306 ——— BOXER-3 India 0.0007 0.1086 —— PHANTOM India 0.0011 0.1095 ——— PP- MULTICEL+ V L Romania 0.0098 0.1301 ———
41 TABLE 3.2.6 REACTION RATES ( SHUTOFF ROD CASE ) ( NORMALIZE TOTAO DT L PRODUCTIO) 0 1. N=
Région Code Séquence Country Moderation Absorption Production MULTICELPP-+ V L Argentina 0.0272 0.9170 1.0 WIMS-D SHETA4+ N Argentina 0.0318 0.9366 1.0 PP-V + MULTICELL Canada 0.0274 0.8211 1.0 Fuel BOXER-3 India 0.0475 0.9856 1.0 PHANTOM India 0.0477 0.9773 1.0 PP-V + MULTICELL Romania 0.0247 0.9063 1.0 PP- MULTICELV+ L Argentina 0.0078 0.0304 —— WIMS-D4+SHETAN Argentina 0.0011 0.0302 Annulus PP-V+MULTICELL Canada 0.0078 0.0304 BOXER-3 India 0.0018 0.0323 —— PHANTOM India 0.0021 0.0328 PP-V + MULTICELL Romania 0.0077 0.030'3 ::: PP-V + MULTICELL Argentina 0.8590 0.0142 — Moderator WIMS-D4+SHETAN Argentina 0.9125 0.0122 PP-V+MULTICELL Canada 0.8595 0.0142 BOXER-3 India 1.2247 0.0120 — PHANTOM India 1.223 0.0125 — PP-V + MULTICELL Romania 0.8573 0.0141 —
PP-V + MULTICELL Argentina 0.0018 0.6143 — Shutoff WIMS-D4+SHETAN Argentina 0.0005 0.6410 rod PP-V+MULTICELL Canada 0.0022 0.6265 BOXER-3 India 0.0001 0.6108 — PHANTOM India 0.0001 0.5914 — PP-V + MULTICELL Romania 0.0027 0.6263 —~
TABLE 3.2.7 REACTION RATES (ZONE CONTROLLERS EMPTY CASE ) ( NORMALIZE TOTAO DT L PRODUCTIO) 0 1. N=
Region Code Sequence Country Moderation Absorption Production PP-V + MULTICELL Argentina 0.0376 0.9292 WIMS-D4 + SHETAN Argentina 0.0284 0.9253 PP-V + MULTICELL Canada 0.0370 0.7962 Fuel PHANTOM India 0.0299 0.9205 PP-MULTICEL+ V L Romania 0.0248 0.9066 1.0 PP-V + MULTICELL Argentina 0.0102 0.0295 WIMS-D4+SHETAN Argentina 0.0009 0.0299 Annulus PP-V+MULTICELL Canada 0.0099? 0.0295 PHANTOM India 0.0014 0.0303 PP-V + MULTICELL Romania 0.0105 0.0303 PP-V + MULTICELL Argentina 0.8034 0.0056 Moderator WIMS-D4+SHETAN Argentina 0.7751 0.0120 PP-V+MULTICELL Canada 0.7748 0.0132 PHANTOM India 0.7774 0.0125 PP-V + MULTICELL Romania 0.8140 0.0135 PP-V + MULTICELL Argentina 0.0002 0.0056 Zone WIMS-D4+SHETAN Argentina 0.0002 0.0058 Control PP-V+MULTICELL Canada 0.0002 0.0078 Tube PHANTOM India 0.0002 0.0055 PP-V + MULTICELL Romania 0.0017 0.0065
42 TABLE 3.2.8 REACTION RATES (ZONE CONTROLLERS FULL CASE ) ( NORMALIZED TO TOTAL PRODUCTION = 1.0)
Region Code Sequence Country Moderation Absorption Production
PP- MULTICEL+ V L Argentina 0.0352 0.9205 1.0 WIMS-D4 + SHETAN Argentina 0.0261 0.9175 1.0 PP-V + MULT I CELL Canada 0.0250 0.7998 1.0 Fuel BOXER-3 India 0.0321 0.9366 1.0 PHANTOM India 0.0337 0.9328 1.0 PP-V + MULTICELL Romania 0.0245 0.9054 1.0 PP- MULTICEL+ V L Argentina 0.0095 0.0295 WIMS-D4+SHETAN Argentina 0.0008 0.0292 —— Annulus PP-V+MULTICELL Canada 0.0097 0..0295 BOXER-3 India 0.0010 0.0297 —— PHANTOM India 0.0015 0.0307 —— - PP-V + MULTICELL Romania 0.0104 0.0303 —— PP-V + MULTICELL Argentina 0.7021 0.0134 ——— Moderator WIMS-D4+SHETAN Argentina 0.6321 0.0121 PP-V+MULTICELL Canada 0.7325 0.0134 BOXER-3 India 0.7387 0.0120 —— PHANTOM India 0.7752 0.0126 —— PP-V + MULTICELL Romania 0.7821 0.0138 ——— pp-V + MULTICELL Argentina 0.2020 0.2586 ——— Zone WIMS-D4+SHETAN Argentina "0.2381 0.2727 Control PP-V+MULTICELL Canada 0.2113 0.2681 Full BOXER- 3 India 0.2701 0.2668 —— PHANTOM India 0.2558 0.2686 —— PP- MULTICEL+ V L Romania 0.2213 0.2718 ———
3.2.3.3. Incremental cross-sections for various reactivity devices
Tables 3.2.9-3.2.12 giv incrementae eth l cross sections calculate variouy db s supercell code r adjustesfo r rods, shutoff rods, empty zone controller tubzond ean e controller tube filled with water respectively. These incremental cross sections were subsequen e useth n di t core calculation determino st worte eth f reactivitho y devices r emptFo . y zone controller tube, only PHANTOM incremental cross section givee sar r India nfo .
3.3 .COR E TASTH E: K 3 BENCHMAR K
This Section summarizes the work related to the core calculation part of the benchmark. It first describes the benchmark followed by summary of main results. Five countries participate thin di s benchmark.
3.3.1. Description of the benchmark
The benchmark is a simplified 600 MW PHWR reactor containing a limited number of well defined simple reactivity devices specificatione Th . reactoe th f so r have typically been taken from Ref givee .ar [51 Tabln nd i an ] e 3.3.1.
The reactor consists of 380 channels with each channel containing 12 bundles. It contains only three reactivity device types namely, the adjuster rods, light water zone controllers, and shutoff rods. The shut off rods normally reside out of the core. The horizontal and axial views presented in Figs.3.3.1 through 3.3.4 show the core layout and the location of these devices.
43 TABLE 3.2.9 INCREMENTAL CROSS-SECTION-ADJUSTER ROD CASE
- Cross Section Code Sequence Country Fast Transport PP-V + MULT I CELL Argentina 7.78463 E-4 ' WIMS-D4 + SHE TAN Argentina 1.62090 E-3 1 Str1 (cm" ) PP-V + MULTICELL Canada- 3.6783 E- 4 BOXER-3 India 7 . 9 E-4 PHANTOM India 8.3 E-4 CLUB + 3D FAST India 4 E- 8.2 RHEA + CALC India ~ 1.07 E-3 PP- MUL+ V TCELI L Romania 2.68603 3E- Thermal Transport PP-V +~MULTICELL Argentina 2.13823 2E- WIMS-D4 + SHETAN Argentina 2.93853 1E- 1 Str2 (cm" ) PP- MULTICEL+ V L Canada 1.5753 E-3 BOXER- 3 India 3 E- 2.58 PHANTOM India 3 E- 2.47 FASD CLU3 T+ B India 3.15 E-3 RHE CAL+ A C India 2.2 E-3 PP-V + MULTICELL Romania 9.15424 9E-
Fast Absorption PP- MULTICEL+ V L Argentina 1.41886 8E- WIMS-D SHETA4+ N Argentina 7.09950 E-6 1 Sa1 (cm" ) PP-V + MULTICELL Canada 4.04338 0E- BOXER- 3 India 5 E- 2.33 PHANTOM India 2.55 E-5 'CLUFASD 3 T+ B India 2.54 E-5 1 RHEA + CALC India 6.02 E-5
TABLE 3.2.9 ( CONTINUED )
Cross Section Code Sequence Country
Thermal Absorption PP- MULTICEL+ V L Argentina 6.62963 E-4 WIMS-D4 + SHETAN Argentina 6.48494 3E- S^ (cm"1) PP-V + MULTICELL Canada 6.861570E-4 BOXER-3 India 4 6.40E- 8 PHANTOM India 6.128 E-4 CLUB + 3D FAST India 4 6.84E- 6 RHEA + CALC India 5.846 E-4 PP-V + MULTICELL Romania 6.80524 7E- Moderation PP- MULTICEL+ V L Argentina -2.85865 7E- WIMS-D4 + SHETAN Argentina 1.94246 0E- s.,-2 (cm"1) PP-V + MULTICELL Canada -4.231196E-5 BOXER-3 India -3.78 E-5 PHANTOM India -4.06 E-5 CLUB +• 3D FAST India -3.87 E-5 RHEA + CALC India -5.47 E-5 PP-V + MULTICELL Romania -4.41538 E-5
44 TABLE 3.2.9 (CONTINUED)
Cross Section Code Séquence Country Yield PP-MULTICEL+ V L Argentina WIMS-D4 + SHETAN Argentina -1.20420 E-5 1 0 Zf1 ( cm" ) PP-V + MULTICELL Canada B OXER-3 India -5 . 7 E-7 PHANTOM India -3.5 E-7 FASD CLU3 T+ B India -5 . 2 E-7 RHE CALA+ C India 7 -5.6E- 4 PP-V + MULTICELL Romania Yield PP- MULTICEL+ V L Argentina 6.79117 E-5 WIMS-D4 + SHETAN Argentina 5.80176 E-5 1 Canada >>Sf2 (cm" ) PP-V + MULTICELL 6.851705E-5 BOXER-3 India 7.69 E-5 PHANTOM India 7 . 09 E-5 CLUB + 3D FAST India 8.14 E-5 RHE CALA+ C India5 E- 4 6 2. MULTICELPP-+ V L Romania 6.74615 7E-
TABLE 3.2.10 INCREMENTAL CROSS-SECTION-SHUTOFF ROD CASE
.Cross Section Code Sequence Country Fast Transport PP-MULTICEL+ V L Argentina -8.15682 E-4 WIMS-D4 •*- SHETAN Argentina -6.96387 E-3 1 Stp1 (cm" ) PP-MULTICEL+ V L Canada -3.202196E-3 BOXER-3 India 3 E- 2.39 PHANTOM India 3 E- 2.09 CLUB + 3D FAST India 2.57 E-3 RHEA + CALC India 1.76 E-2 PP-V + MULTICELL Romania 6.82130 E-3 Thermal Transport PP- MULTICEL+ V L Argentina -5.11145 E-3 WIMS-D4 + SHETAN Argentina 2.45140 E-3 1 Str2 (cm" ) PP-MULTICEL+ V L Canada -5.252041E-3 BOXER-3 India 2.21 E-3 PHANTOM India 1.97 E-3 CLUB + 3D FAST India 3 E- 2.68 RHE CAL+ A C India 2 E- 2.36 MULTICELf PP-- V L Romania 5.19623 E-4 Fast Absorption PP-V + MULTICELL Argentina WIMS-D4 + SHETAN Argentina 7.59494 8E- 1 PP-V + MULTICELL Canada 3.400350E-5 •Sai (cm" ) BOXER- 3 India 4.927 E-4 PHANTOM India 5.364 E-4 CLUB + 3D FAST India 4 E- 5.711 RHEA + CALC India 6 E- 5.78 nrj,,,T . vrrrj r r mTr-irrT O^m a « -î a
45 TABLE 3.2.1 CONTINUE( 0 D)
Cross Section Code Séquence Country Thermal Absorption PP- MULTICEL+ V L Argentina 3.141421E-3 WIMS-D 4SHETA+ N Argentina 3.23928 E-3 PP- MULTICEL+ V L Canada 3.506874E-3 BOXER-3 India 3.6313 E- 2 PHANTOM India 3 3.31E- 1 FASD CLU3 TB+ India 3.4032 E-3 RHEA + CALC India 2.5638 E-3 PP-V + MULTICELL Romania 3.50416 E-3 Modération PP-V + MULTICELL Argentina -4.57727 E-4 WIMS-D4 + SHETAN Argentina -2.76874 0E- 1 2 V (cm"S ) PP-V + MULTICELL Canada -1.920409E-4 BOXER-3 India -4.688 E-4 PHANTOM India -4.91 E-4 CLUB + 3D FAST India -4.981 E-4 RHE CAL+ A C India 5 E- 4.45 PP- MULTICEL+ V L Romania -4.96844 3E-
TABLE 3.2.10 (CONTINUED)
Crosg Section Code Sequence Country Yield PP- MULTICEL+ V L Argentina WIMS-D SHETA4+ N Argentina -3.57060 E-5 1 VSf1 (cm" ) PP- MULTICEL+ V L Canada BOXER-3 India 3.3 E-6 PHANTOM India 7.56 E-5 CLUB + 3D FAST India 7.68 E-6 RHEA + CALC India 2.479 E-6 PP-V + MULTICELL Romania ——— Yield PP-V + MULTICELL Argentina 3.52454 9E- WIMS-D4 + SHETAN Argentina 3.84389 E-4 1 •»Sf2 (cm" ) PP-V + MULTICELL Canada 3.620635E-4 BOXER-3 India 4.202 E-4 PHANTOM India 3.92 E-4 FASD CLU3 TB+ India 4.285 E-4 RHE CALA+ C India 2.817 E-4 PP- MULTICEL+ V L Romania 3.70121 E-4
46 TABLE 3.2.11 INCREMENTAL CROSS-SECTION-ZONE CONTROLLER TUBE EMPTY CASE
Cross Section Code Sequence Country Fast Transport PP-V + MULT I CELL Argentina -2.07752 3E- WIMS-D 4SHETA+ N Argentina -1.72766 E-2 1 Str1 (cm" ) PP-V + MULTICELL Canada -7.21433 E-4 PHANTOM India 2 -1.9E- 0 RHEA + CALC India-2 -2.0304 E-2 PP- MULTICEL+ V L Romania -2.11092 7E-
Thermal Transport PP-V + MULTICELL Argentina -3.26362 1E- WIMS-D4 + SHETAN Argentina -3.21456 E-2- 1 Str2 (cm" ) PP-V + MULTICELL Canada -2.770878E-3 PHANTOM India -3.185 E-2 RHE CAL+ A C India-2 -3.1773 E- 5 PP-V + MULTICELL Romania -3.39141 E-2 Fast Absorption PP- MULTICELV+ L Argentina -2.58475 E- 5 WIMS-D4 + SHETAN Argentina -7.27420 E-5 1 1 8 (cm"S ) PP-V + MULTICELL Canada 6.594160E-5 PHANTOM India -7.07 E-5 'RHE CAL+ A C India -2.227 E-5 PP- MULTICEL+ V L Romania -1.37620 E-5 Thermal Absorption PP-V + MULTICELL Argentina 8.59694 E-5 WIMS-D4 + SHETAN Argentina 3.57585 2E- 1 2 a (cm"S ) PP- MULTICELV+ L Canada 5.041797E-4 PHANTOM India 5 E- 8.38 RHE CAL+ A C India-2 5 4.80E- 2 PP-V + MULTICELL Romania -8.26851 E-5
TABLE 3.2.11 (CONTINUED)
Gross Section Code Sequence Country Moderation PP- MULTICEL+ V L Argentina -6.83984 9E- WIMS-D4 + SHETAN Argentina -7.78499 E-4 1 Si-2 (cm" ) PP-V + MULTICELL Canada -1.34045 E-4- PHANTOM India -8.244 E- 2 RHEA + CALC India-2 -8.753 E-3 PP- MULTICEL+ V L Romania -7.54164 7E-
Yield v **• c riUijJc i J-Uriljij Argentina WIMS-D4 + SHETAN Argentina -4.00560 E-5 1 VSf1 (cm ) r r V T PiUJj J. -LCJUjlj Canada PHANTOM India -3.874 E-4 RHE CAL+ A C India-2 -1.9378 • E-5 PP-V + MULTICELL Romania Yield PP- MULTICEL+ V L Argentina 7.72785 3E- WIMS-D4 -f SHETAN Argentina 2.22215 3E- 1 •»Sf2 (cm" ) PP- MULTICEL+ V L Canada 5.601467E-5 PHANTOM India 5 E- 7.88 RHE CALf - A C India-2 3.595 E- 4 PP-V -f MULTICELL Romania 6.94893 E-5
47 TABLE 3.2.12 INCREMENTAL GROS S-SECTION-ZONE CONTROLLER TUBE FULL CASE
Cross Section Code Séquence Country Fast Transport PP-V -H MULTICELL Argentina 2.04860 E-2 WIMS-D SHETA4+ N Argentina 2.10272 6E- 1 Str1 (cm" ) PP- MULTICEL+ V L Canada 5.77824 3E- BOXER-3 India 3 2.28E- - PHANTOM Indi3 a E- 2 0 . 2 CLU 3DFASB+ T India 2 . 65 E-3 RHEA + CALC India 2.733 E- 7 PP-V + MULTICELL Romania -3.09295 E-4 Thermal Transport PP-MULTICEL+ V L Argentina 0.184997 WIMS-D4 + SHETAN Argentina 0.171197 1 2tr2 (cm" ) PP-MULTICELf - V L Canada 0.136536 BOXER-3 India 0.13492 PHANTOM India 0.13189 CLU 3DFAS+ B T India 0.14622 RHE CALA+ C India 0.13607 PP- MULTICEL+ V L Romania 0.127362 Fast Absorption PP-V + MULTICELL Argentina 5,06230 E-5 WIMS-D SHETA4+ N Argentina 2,74854 9E- 1 Sa1 (cm" ) PP- MULTICEL+ V L Canada -3.73685 8E- BOXER-3 India 2.374 E- 5 PHANTOM India 2.167 E-4 CLUB + 3DFAST India 2.404 E-4 RHE CALA+ C India 5 7,27E- 3 PP- MULTICEL+ V L Romania 1.96134 E-5
TABLE 3.2.12 (CONTINUED)
Cross Section Code Sequence Country Thermal Absorption PP-V + MULTICELL Argentina 1.26233 E- 9 WIMS-D4 + SHETAN Argentina 1.37221 E-3 2,3 (cm'1) PP- MULTICEL+ V L Canada 9.102086E-4 BOXER-3 India 1.4763 E- 4 PHANTOM India 1.416 E-3 CLUB + 3DFAST India 3 1.35E- 5 RHEA + CALC India 1.3653 E- 4 PP- MULTICEL+ V L Romania 1.36281 E-3
Moderation PP-V + MULTICELL Argentina 1.94696 E-3 WIMS-D4 + SHETAN Argentina 3.08189 E-3 S« (cm"1) PP-V -I- MULTICELL Canada 1.492327E-3 BOXER-3 India 2.0987 E-3 PHANTOM India 2.047 E-3 CLUB + 3D FAST India 2.262 E-3 RHE CAL+ A C India 2.1813 E- 3 PP-V + MULTICELL Romania 1.43212 E-3
48 TABLE 3.2.12 (CONTINUED)
Cross Section Code Séquence Country
Yield PP- MULTICEL+ V L AX ^Cii UXIlCt WIMS-D4 + SHETAN Argentina 1.27174 3E- 1 •JSf. (cm" ) PP-V + MULTICELL Canada BOXER-3 India 1.0434 E- 9 PHANTOM India 9.41 E-5 CLUB + 3D FAST India 1.0474 E-4 RHEA + CALC India 1.2405 E-5 Yield PP- MULTICEL+ V L Argentina 1.53805 2E- WIMS-D4 + SHETAN Argentina 3.39907 E-5 -PS« (cm"1) PP-V + MULTICELL Canada -5.046475E-4 BOXER-3 India 5.54 E-5 PHANTOM India 5 E- 2.82 CLUB + 3D FAST India 5 E- -2.56 RHEA + CALC India 4.905 E- 8 PP-V + MULTICELL Romania 4.97840 E-5
TABLE 3.3.1 BASIC REACTOR DATA FOR CORE CALCULATIONS
Reactor Type Pressurized heavy water, horizontal pressure tube. Total fission power 218 W (thM 0 ) Coolant power/fission power 0.95
Moderato reflectod an r r D20 Coolant Pressurized D20 Fuel Natural UO, Refuelling method On-power, bi-directional in adjacent channels Normal direction of fuelling Directio f flono w Form Horizontal stepped cylinder Dimensions : Main shell - Inside diameter 7.6m - Inside length 4 m Sub-shell — Inside diameter 6.8 m — Inside length m m 5 96 Core data Numbe channelf ro s 380 Numbe bundles/channelf ro s 12 Cell array Square Lattice pitch 28.57m 5c Bundle length 49,5 cm Core radius 314.27 cm Core length 594 cm Extrpolàted length 606.0 cm Bundle burnup distribution (see Section III.3.1) Adjuster layout see Figures III.3.1 and'III.3.4 Zone controller layout see Figures III.3.2 and III.3.4 Shutoff rod locations see Figures III.3.3 and III.3.4 Zone controller levels 3 lattice pitches (85.72) 5cm see Figure III.3.2 49 Channel Column Designations 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22
Channel Row Designations
FIG. 3.3.1 600MWe Reactor Face View Showing Adjuster ClusterRod
50 Channel Column Designalions 1 2 3 -456781 2 0 2 9 1 8 1 9 7 1 6 1 5 1 4 1 3
Channel Row Oflsignalions A / B / C / /
FIG. 3.3.2 600MWe Reactor Face View Showing Zone Controller Locations
51 Channel Column Designations 1 2 3 456789 10 11 12 13 14 15 16 17 18 19 20 21
Channel Row Designations A / 8 / C / /
FIG. 3.3.3 ÔOOMWe Reactor Face View Showing ShutoffRod Locations
52 N Ü r •a CD U. "5 6I ( 171.45cm ^ ——— t ——— E \ ——————————————————) 1 zc J1 2 3 ZCLJ2 114.3 cm s f. ——— £ X \ 5 —————— ( 5 ——— 5 4 5 6 57 .T 5 cm 7 8 9 ) ————————————————— 0 cm — » ——— ( 5 ——————— ( zcU3 10 11 12 zc U4 57.15cm ——— i 5 —— 13 14 15 > /• m 114.c 3 ) ——————— ( 5 ————( 16 17 16 171.45cm _L3———— ( 5 ———— (. r r zc US 19 20 21 ZC U6 13<)cm 60 cm 0 cm 80 cm 13()cm r 20e 05 -<~96 5cm-*" AdjusteO • rZon Ro ShutofX ed Contro f Rol d FIG. 3.3.4 600MWE Reactor Plan View Showing Locations of Reactivity Devices 53 alln I , eight cases have been include benchmarke th n di . Thes alse ear o describen di Ref. [43]. In the first three cases, the core contains irradiated fuel bundles. To minimize the numbe f simulationo r s require r benchmardfo k calculations, onl singlya e burnup stats ei used. This state described by the instantaneous burnups of each of the 4560 fuel bundles in core th obtaines ei d fro operatinn ma g CANDU reactor (Embalse) reference Th . e casd ean two sensitivity case outlinee sar d here: ) (1 Reference case wit e giveth h n burnup distribution wit e adjusterth h d zonan s e controllers as shown in Figs. 3.3.1, 3.3.2, 3.3.4 and with the shut-off rods completely out. (2) A time step of 1 day with respect to Case 1. (3) Case 1 with the left half of the core voided. e otheTh r five case essentialle ar s y three-dimensional simulation fino t s d wortf ho reactivity devices in fresh core. All the fuel bundles are fresh and clean, and the effect of various reactivity calculatede deviceb o t s i s . Starting from cas whice4 referencee th s hi , four sensitivity cases are included. These are described below: (4) Fresh core case - A core with all fuel bundles assigned zero burnup, and with reactivity devices as in case 1. ) (5 Adjusters fully withdrawcoree th f .o t nou (6) Zone controllers empty. ) (7 Zone controllers full. ) (8 Shut-off rods fully inserte core th en di It was originally planned that the following parameters be calculated for making a comparison between various calculations. (a) the core multiplication factor and reactivity, with rho calculated as (k-l)/k; (b) the bundle and channel power distribution; and (c) the maximum bundle and channel powers, and their locations - with the power to be give termn ni f powe, coolanthae o ss th fissio i tkW o t rn i t n power times coolant powe fissioo t r n power ratio. Later, at the RCM held in Buenos Aires, it was decided that for the sake of brevity, the comparison be limited to (a) and (c) above and that for all sensitivity cases the change in reactivity with respect to the reference case be given. 3.3.2. Core modeling and computer codes used 3.3.2.1. Mathematical model s weli t I l known thae neutroth t n distribution a heav n i y water e reactob n ca r calculated with adequate precision from diffusion theory using only two energy groups. This gooo t e di du sthermalizatio large th ed siznan f theseo e reactor types. Mesh spacings a s large as the bundle itself provide satisfactory results because of the large diffusion length of neutron heavn si y water. Three dimensional analyses carrie have determino b t t o et dou e eth wort f reactivitho y device coree th n . so 54 l participantAl s have, therefore, used computer codes diffusio e baseth n do n theory and have used a three dimensional analysis. The mesh spacings have generally been comparable to the lattice pitch when away from a reactivity device and almost one-half of e pitcth hneighbourhooe th siz n i e f thesdo e devices typicaA . l mesh structure (uset da Ontario Hydro, Canada shows i ) Tabln ni e 3.3.2. TABLE 3.3.2 TYPICAL MESH SPACING CORE TH EN SI MODE L Number x-spacing (cm) y-spacing (cm) z-spacing (cm) 01 21.791667 21.791667 49.5 02 21.791667 21.791667 47.0 03 21.791667 21.791667 46.25 04 28.575 28.575 49.5 05 28.575 28.575 49.5 06 28.575 28.575 30.5 07 28.575 28.575 24.75 08 14.2875 28.575 24.75 09 14.2875 28.575 30.5 10 14.2875 28.575 49.5 11 14.2875 28.575 49.5 12 14.2875 28.575 46.25 13 14.2875 28.575 47.0 14 14.2875 28.575 49.5 15 14.2875 28.575 0.0 16 14.2875 28.575 17 14.2875 28.575 18 14.2875 28.575 19 14.2875 28.575 20 14.2875 28.575 21 14.2875 28.575 22 14.2875 28.575 23 14.2875 28.575 24 14.2875 28.575 25 14.2875 28.575 26 14.2875 28.575 27 14.2875 28.575 28 14.2875 21.791667 29 14.2875 21.791667 30 14.2875 21.791667 31 14.2875 0.0 32 - 37 14.2875 38 - 41 28.575 42 - 44 21.791667 cele lTh constant incrementae th d san l cros scasee sectionth l s al werr sfo e obtained fro supercelcele d mth lan l benchmarks calculation coveree ar d Section3.2sn dan d i an . 1 s3. These neutronic data are arranged in the form of fuel tables (cross sections as a function of burnup core interpolated th en an )i calculations e us r dfo . 55 3.3.2.2. Computer codes used Various computer codes used by the different participants are briefly described here: PUMA-C (Argentina) Develope CNEAy db , Argentina, PUMA- versioa PUMe Cs i th f no A code oriented fuee th l o management Embalse th f o t e reactor. PUMA-C solve multigroue sth p neutron diffusion equation finita n si e difference mesh centered scheme. The standard mesh spacings of one and one-half lattice pitch were employed reflectoe Th . r intervals were calculate conservo dt specifiee eth d volume th n ei notce th centran hi regionsd an l . OHRFSP (Canada) OHRFSP, Ontario Hydro Reactor Fuelling Simulation Program, [26], models the core as a three dimensional array of parallelepipeds. x-directione Inth , mesh spacing standarhale d th fan e lattice sar don e pitch size, while in the y-direction, these are defined as one lattice pitch in the core region. In both the x and y direction reflectoe th , r regio divides nwa d into three meshes z-spacinge Th . s were chosen beso t s to a outlin positione eth l reactivital f so y devices. DEMENTRI (India-ThPD) DIMENTR threa s i I e dimensional centre-mesh finite difference code which employs uniform radial meshe variabld an s e axial meshes. This cod bees eha n use estimato dt e eth reactivity worth f device o sfrese th hn i s core only. A radial extrapolatio bees ha nn adde m insiddistance c th 3 o de3. t f notch/Calandrieo a wall radiu corn i s e calculations. Radiall pitcheo ytw influencee ar s presence th y a d b f eo reactivity devicer supercellpe s A . givea , n reactivity devicaxiae th ln i e controlm c 5 +2 s direction. Becaus locatioe th f thid e o f adjuste san no r rodzond an se controllers whice har not multiples of 25 cm, 30 cm axial meshes have been used. As the above restrictions did not allow exact estimate of the bundle power (bundle lengt h49.= 5 cm), onl maximue yth m mesh power theid san r location providee sar e th y db code. The bundle locations given in the results are those covered partially by these locations. However, the channel power distributions for each case is calculated. TRIVENI (India-ThPD) The code TRIVENI has been used for the follow-up and fuel management Indian PHWRs. This code uses fixed axia s wela l s radiaa l l meshes. Becaus fixee th df eo axia l meshes having dimensions of half their bundle length (24.75 cm), the location of two rows of adjuster rods at 80 cm gets shifted by 5.75 cm and those of the zone controllers by 6.25 cm towards the central axial plane. 56 CEMESH (India-ThPD) The CEMESH code is a two-group, 3-D diffusion theory code using centre-mesh finite difference scheme with reactor dependent uniform mesh structure built int codee oth . ANAMIKA (India-RED) Another Indian code, ANAMIK Aalss i two-groupoa code D equatione 3- , Th . e sar solved usin finite gth e difference method. This programme alsfixes oha d mesh spacing. CITATION (Pakistan) CITATION is an ORNL 3-D finite difference computer code used for reactor core analysi fued an sl management studies mese Th .h sizes use i CITATIOdh N correspono dt those use OHRFSPn di . FMDP (Romania) diffusioD FMD3- a s Pi n theory AECL code fuee Th l. bundle irradiation givee sar n as input data for the first three cases. For the second case, one more day of burnup is to be fres s fuecleane d numbeth e wa lhan 8 Th . d mesf an ro 7 h , performed6 , case 5 , th e 4 r Fo . points used is 44*31*16 almost identical to the Ontario Hydro number. CHEBY (Romania) CHEBY is another AECL code applied to fresh core analysis only. It is also 3-D diffusion theory code geometricae Th . l mode prepares wa l d utilizin MATMAe gth P code, another AECL program mese Th .h spacin CHEBe th n gi Y core mode 37*30*15s i l e th r fo , X, Y, Z axes, less than the number utilized by the Ontario Hydro. 3.3.3. Result comparisod san n 3.3.3.1. General The results made available by different participants have been taken from the progress reports submitted by them, [46, 48, 52-57] and the presentations at the RCM at Buenos Aires. A summary of the results along with the sequence of codes used by various participants is given hi Table 3.3.3 which consists of 8 pages with one page for each case. Figs. 3.3.5 through 3.3.8 show the locations of 'maximum power' channels and the channels containin bundlee gth s yielding maximu referenco mtw e poweth er casesfo r . The Canadian have used outer adjusters rods of longer size (a length of 10 lattice pitche l computes Al instea . 4) f dro codes used variable mesh spacing activn si e core region except for TRIVENI, CEMESH and ANAMIKA. In Indian evaluations, reactivity device perturbations were considered in ±P distance unlike the rest where ±P/2 was considered. Additional supercell calculations wer dimensionse2 donL/ x eP witx .hP 57 Besides the differences in the cell and the incremental cross sections data used in the core calculations, which is explained in their respective reports, the only variances in the core results stem fro e e meschoicmth th f ho e spacing whic s mostli h a mattey f o r convenience, and from handling of the core -reflector interface. As different participants have used theisetn f calculateso ow r d celincrementad an l l cross sectioned, the results of the core analysis benchmark are not truly a function of the core calculational part providbu t measura e e overalth f o el capabilit o calculatt y e corth e e multiplication factors usefuls i t I . , therefore simplcarra o t t , you e statistical analysis while examining the results. This has also been done in Table 3.3.3. In general, the agreement between the multiplication factors, the reactivity changes, peae th kd bundle/channean l power distribution regardee b y sma satisfactorys da . However, the locations of the maximum power bundle/channels are not very similar, although they are in better agreement for the fresh core because burnup tables do not get involved. The variatio again n ca attributee nb differencee th o dt crosn i s handline s th section o t f d go an s geometry case-wisA . e descriptio resulte th f gives no i s n below. 5.3.5.2. Reference cases with givene th burnup distribution The core multiplication factors are less than unity by 1.2 mk to 7.9 mk. The largest difference in the maximum channel power is about 8 %. 3.3.5.5. A time step of 1 day with respect to Case 1 change Th reactivitn ei rathes yi predictione r smalth d an l thir sfo s chang fairle eli y closely. 58 Channel Column Designations 13 U 15 t6 17 -18 19 20 21 22 Channew Ro l Designations 1: Argentina (PP-V), Pakistan 2: Aigentinal (WIMS-D4), India (TRIVENI) 3: Canada 4: India-2 5: Romania (FMDP) 379.. 7cm FIG. 3.3.6 Locations of Channel Containing the Bundle Yielding Maximum Power ( Reference Equilibrium Core ) 59 Charm«! Column Designations 1 2 3 Channel Row Designations A / 8 / C / / 1: Argentina (PP-V), Romania (FMDP) 2: Argenüna (W1MS-D4), Canada 3: India (DIMENTRI) 4: India—2 Pakista: 5 n FIG. 3.3.7 Locations of " Maximum Power " Channel (Fresh Core ) 60 Chann«! Column Designations t 2 2 1 2 0 2 9 1 8 1 7 1 6 1 5 1 4 1 3 1 2 \ 1 1 0 1 9 8 7 t 6 2 5 3 4 Channew Ro l Designations A / 8 / C / / 1: Argentina (PP-V, WIMS-D/4), Canada, Pc 2: India (D1MENTRI), 3: lndia-2 : Romani4 a (FMDP) FIG. 3.3.8 Locations Channelof Containing Bundlethe Yielding Maximum Power Fresh( Core) 61 K) TABLE 3.3.3 ( 1 OF 8 ) REFERENCE CASE WITH GIVEN BURNUP DISTRIBUTION K Delta Max . Channe 1 Max. Bündle Effec- Rho Rho Power in kw, , Powekw n i r COMPUTER CODES USED Country tive mk mk Location Location 1 PP-V+MULTICELL+PUMAC Argentina 0.9940 -6 .05 --- 7016.9 ,G-06 898 .6,G-06 -06 2 WIMS -D4 +SHETAN+PUMAC Argena ti 0.994- 7-- -8 5 .2 6960.6 ,G-06 888 .0,H-05-06 3 POWERPUFS+MULTICELL+OHRFSP Canada 0.9922 -7 .86 --- 7246.0 , Fv-14 939 . 7, E-ll-06 4 PHANTOM+TRIVENI India-ThPD 0.995- 1 -- -9 4 .8 7148.0 , J-09 888 .4,H-05-06 5 PHANTOM+3D-FAST+DIMENTRI India-ThPD -_ _ _ _ _ .. _ _ _ _ — 6 PHANTOM+CEMESH India-ThPD 0.994- 9-- -2 5 .1 7037.0 , J-09 898 . 0,E-11-06 7 RHEA+CALC+ANAMIKA India-RED 7516.9 ,R-09 924 .9,3-10-07 8 WIMS-D/4+CITATION Pakistan 0.9937 -6 .296 --- 6925.2 ,F-15 876 .5,G-06 -06 9 PP-V+MULTICELL+FMDP Romania 0.998- 8-- -5 1 .2 7164.1 ,M-14 892 .3,F-14-07 10 PP-V+MULTICELL+CHEBY Romania __ ... AVERAGE 0.9948 7126.8 900 .8 DEVIATIOD ST % N 0.19% 2.51% 2.18% PERCENTILE DIFFERENCE MAX-MIN /MAX 0.66% 7.87% 6.73% TABLE 3.3.3 ( 2 OF 8 ) A TIME STEP OF 1 DAY WITH RESPECT TO CASE 1 K Delta Max. Channel Max.Bunmdle Effec- Rho Rho Power in kw, Powe kwn i r, COMPUTER CODES USED Country tive mk ink Location Location 1 PP-V+MULTICELL+PUMAC Argentina 0.9936 -6 .47 -0.42 7021.5 ,G-06 898.5 ,G-06 -06 2 WIMS-D4+SHETAN+PUMAC Argentina 0.9944 -5 .68 -0.40 6959.2 ,G-06 887.0 ,H-5- 06 3 POWERPUFS+MULTICELL+OHRFSP Canada 0.9918 -8 .27 -0.41 7248.0 ,F-14 940.0 ,E-11 4 PHANTOM+TRIVENI India-ThPD 0.9947 -5 .35 -0.45 7136.0 , J-09 887.4 ,H-05 -06 PHANTOM+3D-FAST+DIMENTR5 I India-ThPD - _ _ _ - _ 6 PHANTOM+CEMESH India-ThPD 0.9945 -5 .53 -0.41 7037.0 , J-09 898.0 ,H-05-06 RHEA+CALC+ANAMIK7 A India-RED 7542.3 ,R-09 927.6 ,S-10 -07 WIMS-D/4+CITATIO8 N Pakistan - - - 9 PP-V+MULTICELL+FMDP Romania 0.9984 -1 .64 -0.38 7160.4 ,M-14 891.6 ,F-14 -07 10 PP-V+MULTICELL+CHEBY Romania _ _ _ -______1 AVERAGE 0.9946 -0.41 7157.8 904.3 DEVIATIOD %ST N 0.20% 5.14% 2.53% 2.15% PERCENTILE DIFFERENCE MAX-MIN /MAX 0.66% 15.56% 7.73% 5.64% TABL ) CAS8 E F 1 WITE E 3.3.LEFO 3 HTH 3( TE COR HALTH EF O FCOR E VOIDED K Delt Bund. x Chann. alx Ma e Ma e1 Effec- Rho Rho , , PowePowekw kw n n i ri COMPUTER CODES USED Country tive mk mk Location Location 1 PP-V+MULTICELL+PUMAC Argentina 1.0007 0.68 6.73 9199.1,0-06 1224.1,H-05-06 2 WIMS-D4+SHETAN+PUMAC Argentina 1.0050 4.95 10 .23 10072.8 ,H-05 1340.4 ,H-05-06 3 POWERPUFX+MULTICELL+OHRFSP Canada 1.0010 1.00 8.86 8581.0 ,0-07 1053.0 ,H-04 4 PHANTOM+TRIVENI India -ThPD 1.0041 4.08 8.98 9997.0 ,0-06 1312.8 ,H-05-06 5 PHANTOM+3D-FAST+DIMENTRI India-ThPD - -- -_ _ 6 PHANTOM+CEMESH India-ThPD 1.0056 5.59 10 .70 10311.0 ,H-05 1374.0 ,H-05-06 RHEA+CALC+ANAMIK7 A India-RED 11187.3 ,P-19 1544.9 ,P-19-06 8 WIMSS-D4/+CITATION Pakistan 1.0006 0.55 9.26 9100.8 ,N-18 1153.6 ,0-18-06 9 PP-V+MULTICELL+FMDP Romania 1.0058 5.78 7.03 9387.6 ,D-11 1205.3 10 PP-V+MULTICELL+CHEBY Romania ______~ ______1 AVERAGE 1.0033 8.83 9729.6 1276.0 DEVIATIO. D ST % N 0.22% 15 .62% 7.93% 11.08% PERCENTILE DIFFERENCE MAX-MIN /MAX 0.52% 37 .10% 23.30% 31.84% TABLE 3.3.3 ( 4 OF 8 ) FRESH CORE REFERENCE CASE K Delta Max. Channel Max. Bundle Effec- Rho Rho Power in kw Power in kw COMPUTER CODES USED Country tive mk mk Location Location PP-V+MULTICELL+PUMA1 C Argentina 1.0277 26 .94 --- 7300.3 ,K-13 785 .3,E- 11- 06 2 WIMS-D4+SHETAN+PUMAC Argentina 1.0327 31 .71 --- 7245.0 ,K-10 787 .3,E- 11- 06 3 POWERPUFS+MULTICELL+OHRPSP Canada 1.0259 25 .25 --- 7456.7 ,K-10 819 .1,E- 11- 06 4 PHANTOM+TRIVENI India-ThPD - -- - _ _ _ _ _ - — _ _ 5 PHANTOM+3D-FAST+DIMENTRI India-ThPD 1.0350 33 .80 --- 7333.0 ,M-09 793 .0,E- 12- 07 6 PHANTOM+CEMESH India-ThPD 1.0358 34 .57 --- 7207.0 ,K-13 783 .5,E- 12- 07 7 RHEA+CALC+ANAMIKA India-RED 1.0304 29 .48 7462.6 ,M-10 828 .7,8- 11- 07 8 WIMS-D/4+CITATION Pakistan 1.0282 27 .44 --- 7100.2 , J-ll 799 .0,E- 11 9 PP-V+MULTICELL+FMDP Romania@ 1.0725 67 .58 --- 7410.2 ,K-13 793 .2, J-14- 07 10 PP-V+MULTICELL+CHEBY Romania@ 1.0707 66 .04 --- 7242.0 784 .9 t AVERAGE 1.0308 7306.3 797.1 % STD DEVIATION 0.34% 1.57% 1.91% PERCENTILE DIFFERENCE MAX-MIN /MAX 0. 96% 4.86% 5.45% Romaniae Th @ n value withoue sar t xenon; hence neglecte averaginn i d g k-eff. ON TABL ADJUSTER) 8 EF 3.3.O 5 3S ( CORE WITHDRAWTH E F O T NOU K Delta Max. Channel Max. Bundle Effec- Rho Rho Power in kw, Power in kw, COMPUTER CODES USED Country tive mk mk Location Location 1 PP-V+MULTICELL+PUMAC Argentina 1.0460 43 .99 17 .05 9285.1 ,L-13 1259.8,M-12- 06 2 WIMS-D4+SHETAN+PUMAC Argentina 1.0505 48 .08 16 .37 9291.7 ,L-10 1265.7,M-11- 06 3 POWERPURS+MULTICELL+OHRFSP Canada 1.0469 44 .79 19 .55 9308.0 ,L-13 1269.0,M-12 4 PHANTOM+TRIVENI India-ThPD - - - _ _ _ 5 PHANTOM+3D-FAST+DIMENTRI India -ThPD 1.0514 48 .91 15 .11 9148.0 ,L-13 1247.0,E-12- 07 6 PHANTOM+CEMESH India-ThPD 1.0517 49 .17 14 .60 9153.0 ,L-13 1237.7,L-12- 07 7 RHEA+CALC+ANAMIKA India-RED 1.0460 44 .01 14 .53 8 WIMS-D/4+CITATION Pakistan 1.0470 44 .86 18 .57 9195.2 ,K-11 1258.3,M-11 9 PP-V+MULTICELL+FMDP Romania @ 1.0916 83 .90 16 .32 9247.2 ,L-10 1233.7,M-11- 07 10 PP-V+MULTICELL+CHEBY Romania @ 1.0907 83 .17 17 .'13 9287.4 1268.8 AVERAGE 1.0485 16 .58 9239.5 1255.0 % STD DEVIATION 0.23% 9.79% 0.66% 1.04% PERCENTILE DIFFERENCE MAX-MIN /MAX 0.54% 25 .68% 1.72% 2.78% Romaniae Th @ n value withoue sar t xenon; hence neglecte averaginn i d g k~eff. TABLE 3.3.3 ( 6 OF 8 ) ZONE CONTROLLERS EMPTY K Delta Max. Channel Max. Bündle Effec- Rho Rho Power in kw, Power in kw, COMPUTER CODES USED Country tive mk mk Location Location 1 PP-V+MULTICELL+PUMAC Argentina 1.0316 30 .61 3.67 7465 .9,M-11 796.1 ,14-10-09 2 WIMS-D4+SHETAN+PUMAC Argentina 1.0368 35 .51 3.80 7416 . 8 ,M-11 792.9 ,M-10-04 3 POWERPUFS+MULTICELL+OHRFSP Canada 1.0301 29 .21 3.96 7643 . 0 , M-ll 819.0 ,M-10 4 PHANTOM+TRIVENI India-ThPD - - 5 PHANTOM+3D-FAST+DIMENTRI India-ThPD 1.0391 37 .61 3.81 7520 .0,L-11 748.0 ,L-11-06 6 PHANTOM+CEMESH India-ThPD 1.0397 38 .22 3.65 7380 . O/M-11 772.0 ,M-10-04 7 RHEA+CALC+ANAMIKA India -RED 1.0354 34 .22 4.74 8 WIMS-D/4+CITATION Pakistan 1.0323 31 .28 3.73 7319 . 8 ,M-11 808.6 ,M-11 9 PP-V+MULTICELL+FMDP Romania @ 1.0761 70 .75 3.17 7528 .8,M-13 801.3 ,M-13-09 10 PP-V+MULTICELL+CHEBY Romania@ 1.0749 69 .65 3.61 7403 .4 792.8 AVERAGE 1.0350 3.79 7459 .7 791.3 % STD DEVIATION 0.34% 10.35% 1.28% 2.62% PERCENTILE DIFFERENCX MA E -MIN /MAX 0.92% 33.10% 4.23% 8.67% Romaniae Th @ n value withoue ar s t xenon; hence neglecte averaginn i d g k-eff. TABLE 3.3.3 ( 7 OF 8 ) ZONE CONTROLLERS FULL K Delta Max. Channel Max. Bündle Effec- Rho Rho w Powek n i r Power in kw COMPUTER CODES USED Country tive mk mk Location Location 1 PP-V+MULTICELL+PUMAC Argentina 1.0231 22.54 -4.40 7322.8 ,N-09 854.9,3-11-06 2 WIMS-D4+SHETAN+PUMAC Argentina 1.0279 27.19 -4.52 7275.6 ,N-09 863.6,3-11-06 3 POWERPUFS +MUL TCELL+OHRFI SP Canada 1.0209 20.52 -4.73 7444.0 ,N-13 899.0,3-11- 4 PHANTOM+TRIVENI India-ThPD - - _ _ _ 5 PHANTOM+3D-FAST+DIMENTRI India-ThPD 1.0300 29.11 -4.69 7341.0 ,M-09 838.0,5-11-07 6 PHANTOM+CEMESH India-ThPD 1.0310 30.03 -4.54 7254.0 ,N-09 853.5,3-12-06 7 RHEA+CALC+ANAMI KA India-RED 1.0268 26.10 -3.38 8 WIMS-D/4+CITATION Pakistan 1.0233 22.80 -4.69 7133.5 ,N-09 871.0,T-11 9 PP-V+MULTICELL+FMDP Romania ® 1.0673 63.05 -4.53 7459.6 ,N-14 856.1,R-13 -07 10 PP-V+MULTICELL+CHEBY Romania @ 1.0658 61.70 -4.34 7267.9 858.7 AVERAGE 1.0261 -4.42 7312.3 861.9 % STD DEVIATION 0.34% 8.81% 1.36% 1.92% PERCENTILE DIFFERENCE MAX-MIN /MAX 0.98% 28.54% 4.37% 6.78% @ The Romanian values are without xenon; hence neglected in averaging k-eff. TABLE 3.3.3 ( 8 OF 8 ) SHUT-OFF RODS FULLY INSERTED K Delta Max. Channel Max. Bundle Effec- Rho Rho Power in kw Power in kw COMPUTER CODES USED Country tive mk mk Location Location 1 PP-V+MULTICELL+PUMAC Argentina 1.0176 17 .29 - 9.65 7294 .3, J-ll 877.6,E-11-06 2 WIMS-D4+SHETAN+PUMAC Argentina 1.0227 22 .23 - 9.48 7221 .8, J-ll 884.6,E-11-06 3 POWERPUFS+MULTICELL+OHRFSP Canada 1.0148 14 .61 -10 .64 7508 .0, J-ll 934.0,E-11 4 PHANTOM+TRIVENI India -ThPD - - _ _ _ 5 PHANTOM+3D-FAST+DIMENTRI India-ThPD 1.0242 23 .61 -10 .19 7393 .0,P-11 903.0,3-11-07 6 PHANTOM+CEMESH India-ThPD 1.0252 24 .62 -9 .95 7184 .0, J-12 891.7,E-12 -06 7 RHEA+CALC+ANAMIKA India -RED 1.0199 19 .51 -9 .97 8 WIMS-D/4+CITATION Pakistan 1.0176 17 .26 - 9.75 7088 .9, J-ll 892.1,E-11 9 PP-V+MULTICELL+FMDP Romani® a 1.0617 58 .07 - 9.51 7390 .3,1-12 876.1,E-12-07 10 PP-V+MULTICELL+CHEBY Romania@ 1.0599 56 .47 - 9.'57 7215 .8 881.4 AVERAGE 1.0203 -9 .86 7287.0 892.6 % STD DEVIATION 0.35% 3 .62% 1.75% 1.98% PERCENTILE DIFFERENCE MAX-MIN /MAX 1.01% 10 .90% 5.58% 6.20% @ The Romanian values are without xenon; hence neglected in averaging k-eff. cr\ 3.3.3.4. Case with1 left corethe the half of voided The largest variation core s th exis er voidefo t e primarile th d ar caseo t d e an s ydu difference e celth l n crosi s s section. Void reactivit s moryi r thosfo e e evaluations using WIMS multigroup library. 3.3.3.5. Fresh core case Except for the Rumanian values (calculated without considering xenon) which differ from the average by about 30 mk, the core multiplication factors are predicted within 10 mk. The largest difference in the maximum channel power is less than 5 %. 3.3.3.6. Effect variousof reactivity devices variatioe Th predictioe th n ni f deltn o lieo arh s r shu betweef casfo d of t ean % 1 n1 emptr fo ya% s zon3 hig3 s eha controller case predictione Th . maximue th r sfo m channel power lie within 6 % and those for the maximum bundle power within 10 % (The values of India-RED are not available for this case). 3.3.4. Comment conclusiond san s An exercis globan ei l core calculation bees sha n complete resulte th d differenf so d an t CRP participants have been compared. Nine codes have been use solvo t d e corth e e problem, four Indian Codes (TRIVENI, DIMENTRI, CEMESH, ANAMIKA), three codes of Canadian origin (OHRFSP, FMDP, CHEBY) Argentineae on , n code (PUMA-C)e th d an , Oak Ridge code CITATION. Considering the wide variety of code sequences used, and the differences in the cross sections, the comparisons show a reasonable agreement with the largest discrepancy being voidee inth d core case. pointes i t I that dou t onc supercelresultcele e d eth th an l r sfo l benchmark beed sha n compared, it would have been more informative if all the participants had settled on a set of cross section data and had done the core calculations using their own computer codes. As it effecte isth , s causechoice th mesy f deb o h structur handline th d ean reflectof go r regioe nar masked in the variances caused by the differences in the cross sections. 3.4. TAS LOS: K4 REGULATIOF SO N BENCHMARK PROBLE PHWRR MFO s One of the benchmarks was the simulation of loss of reactivity control event initiated by the simultaneous draining of the seven zone controllers on the one side of the reactor (one lattic esecondo pitctw i hh s linearl tunei yh ) followe insertioe th y shutofd4 b f no f rods. This transient was analyzed by three countries, namely, Argentina, India-ThPD and Romania with their code syste theid man r results were submitte IAEAo dt . This section give summare sth y of all these results. 70 3.4.1. Specificatio problee th f no m specificatione Th f thiso s problem were taken frofollowss a me ar réf d : . an [52 ] 58 , 3.4.1.1. Reactor model reactoe Th r mode similas i lcor e th e o rt stati c benchmark define Task-3r e dfo th t bu , bundle burnup distributio bees nha n replace homogeneouo tw y db s burnup zones innee Th . r zone consists of 124 channels located in a centered 12 x 12 square except five forming a right angle in each corner (see Fig.3.4.1). The average cross sections for inner zone are I - - Channel Column Designations 123456789 to It 12 13 U 15 16 17 18 19 20 21 22 Channel Row Designations FIG. 3.4.1 600MWe Reactor Face View Showing BurnupTwo Zones 71 average cross sections between 0 and 7600 MWD/T. The outer zone i$ formed by the rest of the channels with average cross sections between 0 and 6400 MWD/T. The reactivity device configuration is indicated in Figs. 3.3.2 to 3.3.4. The basic cross sections for both inner and outer zones are given in Table 3.4.1. They are calculated with PPV and are taken from réf. [35], TABLE 3.4.1 CROSS-SECTION FOR HOMOGENEOUS MODEL Exit burnup Exit burnup 6400 MWD/T 7600 MWD/t Reflector D! 1.272889 1.272889 1.313563 (cm) D2 0.938347 0.938319 0.874225 (cm) Za1 7.51545E-4 7.511370E-4 0.0 (cm"1) Ea2 4.02041E-3 4.043360E-3 8.423070E-5 (cm-1) Ef2 4.66564E-3 4.643240E-3 0.0 (cm'1) E12 7.40931E-3 7.409720E-3 1.015111E-2 (cm'1) H 0.244608 0.242116 0.0 (kw cmWlO11) f E e 6.051917E-17 5.990262E-10 0. 7 (kw .cm'.s 1) Notes: The H factor is defined so that H*PHI2CELL = Fission Power of Bundle. The cross sections are averaged between 0 and the exit burnup. Fast fission and upscattering are assumed zero. •i 3.4.1.2. Incremental cross-sections Incremental cross sectione reactivitth r fo s y devices have been calculated using MULTICELL taken from e réfgivear . Tabln d ni [44 an ]e e cas f 3.4.2th zoneo r eFo . controllers filled with water, the incremental cross sections of water and zone controller empty tube given hi Table 3.4.2 should be added to basic cell cross sections without zone controllers. 72 TABLE 3.4.2 INCREMENTAL CROS SSECTION- REACTIVITR FO S Y DEVICES Case S«, St.» 2al^a2 S12,v£f2 eEf2 (cm'1) (cm'1) (cm"1) (kws/cm) Fe adjuster 7.78463E-4 1.41888E-6 -2 .85867E-5 8.70959E-19 rod 2.13822E-3 6.62963E-4 6.79117E-5 Zone Control -2.07753E-2 -2 .58475E-5 -6 .83989E-4 9.91084E-19 empty tube -3.26361E-2 8.59694E-5 7.72783E-5 Zone Control 2.04860E-2 5.06230E-5 1.94696E-3 -1.97249E-19 Water 0.184997 1.26239E-3 -1 .53802E-5 Cd Shuf tof -8.15682E-4 _ _ _ -4 .57727E-4 4.52024E-18 rod -5.11145E-3 3 .41421E-3 3 .52459E-4 - - - 3.4.1.3. Kinetic parameters kinetie Th c parameter delayed san d neutron dat givee aar Tabln i e 3.4.3. Thee yar typica CANDa f o l U reactor with equilibrium havfued an le been taken from Tablréff o . V eI [59]. TABLE 3.4.3 SIX GROUP DELAYED NEUTRON DATA I /Si X 100 Li (sec'1) 1 0.0298 0.000618 2 0.1174 0.03156 - 3 0.1043 0.122 4 0.2374 0.317 5 0.0786 1.39 6 0.0198 3.79 ß 0.5873 ___ 1 Prompt neutron life time 0.893 m see-i 6 Vfast Average fast velocity 8.19 X 10 cm/sec 5 Vtherraal Average thermal velocity 2.75 X 10 cm/sec 73 3.4.1.4. Reactivity transient lose reactivitf sTh o y control even initiates i t draininy d b sevee th f ngo compartments on one side of the reactor (in ZCU no. 1,3 and 5 shown in fig. 3.3.4). The draining starts at time t=0 sec and the water level of the seven compartments decreases by one lattice pitch (28.575 cm) in two seconds linearly in time. The level of water stays constant after two seconds. The level of water in the other seven zones remains unchanged at three lattice pitches e adjusteTh . r rods were insid t movtime no core th f drainin eo eth t o e ea d d gan durin transiente gth foue Th r. shu rodf of t s start their insertio second3 n0. s afte safete th r y system trips movemene Th . lowee indicates coordinati th d f d ren o t ro shue : f th d by of tf eo 2 4.775- 4 1 6t _ = y+ t lattice time th wher th s secondi n ei s i y pitche et d san s above midplane. 3.4.1.5. Detector representation Bundle location definee sar detectorss da , associated with trip channel followss sa : Trip channel/detector Bundle location (channel/bundle) l D L6/4 D2 L6/9 El L11-L12/4 E2 L11-L12/9 FI L17/4 F2 L17/9 wher bundle detectoe eth th f o e e power bundleon f trio ra f pso channel reaches 1.15 times beginnine th t transiente a th valu d e f gth o ha correspondin e t ei th , g trip channel ) (D,F r Eo trips. When two trip channels are tripped, the system is tripped and the shut off rods start moving 0.3 seconds later. 3.4.1.6. Required results followine Th g benchmark results wergivee b o et n ever second2 y0. s unti l shuf lal tof thed an n i afteh rod e r ar sever y second second0 upt=1 ot : s a) Total Powe unitn (i rf Ful so l power) b) Dynamic reactivity (in mk) c) Tune of Trip (seconds) ) d Maximum channe bundld an l e power theid an sr locations. 3.4.2. Descriptio calculationaf no l models Three countries namely, Argentina, India-ThPD and Romania have submitted the result r thifo ss benchmarks briee Th .f descriptio calculationae th f no l modele s th use r dfo analysis is given below : 74 3.4.2.1. Argentina [60] Code: PUMA- versioa PUMe Cs i th f no A code develope fuee th l r managemendfo t Embalse ofth e reactor r spatiaFo . l kinetics calculations PUMA-C uses improved quasi-static method. Spatial mesh spacing: The mesh intervals are given in Table 3.4.4. 37,30 and 16 meshes wer edirectionz considered an y x, s n respectivelydi . Time steps: Time interval 2 second0. f o s s were chosen upt secondso2 e tripTh . occurs at 1.98 seconds after the beginning of the transient. The time intervals were chosen TABLE 3.4.4 MESH SPACING IN CORE MODEL (37X30X16) USED IN ARGENTINEAN AND INDIAN MODELS I,J,K Dx Dy Dz I J K (cm) (cm) (cm) 2 15 .2046 15 .2046 42 .395 3 29 .0848 29 .0848 50 .0 4 22 .6606 22 .6606 49 .605 5 28 .5750 28 .5750 50 .0 6 28 .5750 28 .5750 50 .0 7 28 .5750 28 .5750 30 .0 8 28 .5750 28 .5750 25 .0 9 14 .2875 28 .5750 25 .0 10 14 .2875 28 .5750 30 .0 11 14 .2875 28 .5750 50 .0 12 14 .2875 28 .5750 50 .0 13 14 .2875 28 .5750 49 .605 14 28 .5750 28 .5750 50 .0 15 14 .2875 28 .5750 42 .395 16 14 .2875 28 .5750 17 28 .5750 28 .5750 18 28 .5750 28 .5750 19 28 .5750 28 .5750 20 28 .5750 28 .5750 21 28 .5750 28 .5750 22 14 .2875 28 .5750 23 14 .2875 28 .5750 24 28 .5750 28 .5750 25 14 .2875 28 .5750 26 14 .2875 28 .5750 27 14 .2875 22 .6606 28 14 .2875 29 .0848 29 14 .2875 15 .2046 30 28 .5750 31 28 .5750 32 28 .5750 33 28 .5750 34 22 .6606 35 29 .0848 36 15 .2046 75 so that the lower end of the shutoff rods coincides with the boundary between spatial mesh intervals used in the PUMA-C core model. In general, the time intervals were about 0.1 second and rod movement was two lattice pitches. 5.4.2.2. India-ThPD [61] Code: The computer code 3D-FAST which solves the 3-dimensional time dependent diffusion equation with either adiabatic or unproved quasi-static method was used. Spatial mesh spacing: Half corconsideres ewa - directiox symmetryo t n di e ndu . Mesh spacing used was identical to that used in Argentinean calculations. Time steps: Time steps of 0.2 s were used till 4 s and thereafter steps of l s were used. Further time steps were create t pointda s where contro insertiod ro l n start endsd san . 3.4.2.3. Romania [62] Code e MATMATh : P progra s musei o modet d e reactoth l r a threcor s a ee dimensional array positioo t , reactivite nth y devices distributiop creato t ma , e e eth th n ni mesh spacings and to calculate the coefficients of the diffusion equation. The CERBERUS cod useenergy-grous o ei solvtw o dt e eth p three dimensional neutron diffusion equations throug iterativn ha e finite difference approach. CERBERUS code uses improved quasi-static method powee Th . r distribution t eversa powee y th tim d re an puls e durin transiene gth e ar t obtained using the CERBSPOW program. These codes were supplied by Atomic Energy of Canada Limited. Spatial mesh spacing: Mesh spacinlattice on f ego pitch (28.57 ) wer5cm e user dfo x-direction in core region where no reactivity devices were present and 1/2 lattice spacings were used in the vicinity of reactivity devices (adjusters and zone controllers). Mesh spacings in y-direction were define lattic1 s a d e e cor pitcth en i hregion e z-directioTh . n mesh spacings were chosen so as to best outline the positions of all reactivity devices. Time steps: The first moment during 2 seconds contains 4 steps in time-dependent evolution, and the second during 1.9 seconds contains 25 steps and the third during 6.1 second 7 steps s ha .s Note that this choic f timo e e step doet directlno s y alloe wth enumeration of results at the required time step of 0.2 sec. 3.4.3. Result discussiond san s 3.4.3.1. Dynamic reactivity variatioe Th dynamif no c reactivit functioa s ya timf no e calculate threy db e countries is give Tabln ni e 3.4.5 maximue Th . m reactivitie drainino t e sdu f ligh go te wateon y b r lattice pitch in seven compartment calculated by PUMA-C (Argentina), 3D-FAST (India- ThPD) and CERBERUS (Romania) are 0.588 mk, 0.590 mk and 0.575 mk respectively. After 10 seconds of the initiation of transient the reactivities calculated by Argentina, India- ThP see e Romanid b Dnn an tha ca -8.5 e t -8.5 d tI , -8.5a ar thes . an 5 mk 2k 6mk em results agree well with each other. The variation of dynamic reactivity as a function of time is plotte fign di . 3.4.2. 76 TABLE 3.4.5 VARIATIO F DYNAMIO N C REACTIVITY WITH TIME Time Argentina India Romania (sec) -ThPD ** 0.0 0.000 0.000 0.000 0.2 0.052 0.050 0.057 0.4 0.104 0.100 0.114 0.6 0.157 0.160 0.170 0.8 0.213 0.210 0.227 1.0 0.270 0.270 0.284 1.2 0.329 0.330 0.342 1.4 0.391 0.400 0.400 1.6 0.454 0.460 0.459 1.8 0.520 0.530 0.517 2.0 0.587 0.590 0.575 2.2 0.587 0.590 0.575 2.4 0.588 0.590 0.576 2.6 0.588 0.590 0.557 2.8 0.585 0.540 0.418 3.0 -0.292 -0.230 -0.178 3.2 -1.500 -1.500 -1.375 3.4 -3.134 -3.080 -2.951 3.6 -5.511 -5.520 -5.248 3.8 -7.976 -8.160 -7.958 4.0 -8.584 -8.530 -8.559 5.0 -8.576 -8.560 -8.548 6.0 -8.568 -8.560 -8.539 7.0 -8.561 -8.560 -8.532 8.0 -8.556 -8.560 -8.526 9.0 -8.551 -8.560 -8.521 10.0 -8.547 -8.560 -8.518 ** Values we're obtained by interpolation ——— Argentina ——— India — - - Romania < i 45678 10 Time (sec) FIG. 3.4.2 Variation of Dynamic Reactivity with Time 77 3.4.3.2. Total reactor power The variation of reactor power relative to initial power as a function of time is given in Table 3.4.6. The maximum relative powers are found to be 1.162, 1.163 and 1.150 by Argentinean, India Rumaniad nan n simulations respectively. These values agree very well with each other. The relative reactor powers at 10 seconds after the initiation of transient are calculated 0.211, 0.210 and 0.212 by Argentinean, Indian and Rumanian simulations. The variation of relative power as a function of time for the three countries are plotted in fig. 3.4.3. 3.4.3.3. Time tripof Time at which the shutdown system trips is calculated to be 1.98, 1.98 and 1.99 sec by Argentinean, Indian and Rumanian simulations respectively. For the further calculations, Argentina and India-ThPD used 1.98 sec while Romania used 2.0 sec. TABLE 3.4.6 VARIATIO F AMPLITUDO N E FACTOR WITH TIME Time Amplitude [P(t)/P(0)3 Sec Argentina India Romania * * -ThPD 0.0 1.000 1.000 1.000 0.2 1.004 1.004 1.009 0.4 1.012 1.012 1.019 0.6 1.021 1.021 1.028 0.8 1.032 1.032 1.038 1.0 1.033 1.044 1.047 1.2 1.057 1.058 1.062 1.4 1.072 1.072 1.077 1.6 1.087 1.088 1.093 1.8 1.104 1.106 1.108 2.0 1.123 1.125 1.123 2.2 1.137 1.139 1.134 2.4 1.146 1.148 1.144 2.6 1.153 1.155 1.149 2.8 1.160 1.161 1.145 3.0 1.108 1.094 1.090 3.2 0.945 0.941 0.952 3.4 0.763 0.761 0.777 3.6 0.586 0.581 0.601 3.8 0.443 0.441 0.451 4.0 0.385 0.385 0.387 5.0 0.328 0.328 0.329 6.0 0.293 0.292 0.294 7.0 0.266 0.266 0.267 8.0 0.245 0.244 0.245 9.0 0.227 0.226 0.227 10.0 0.211 0.210 0.212 ** Values were obtaine y interpolatiodb n 78 1.2 n Argentina India - - - Romania 0.2 - 0.1 i i i i i i i i i i 456 7 8910 Time (sec) FIG. 3.4.3 Variation of Normalised Power with Time 3.4.3.4. Maximum channel bundleand powers The variation of maximum channel and bundle powers with time are given hi Tables 3.4.7 and 3.4.8 respectively. It can be seen from the Table 3.4.7, that maximum channel powers at t=0 are found to be 6.74, 6.72 and 6.66 MW by Argentinean, Indian and Rumanian simulations respectively. The value, location and time for maximum channel power durin transiene gth obtainee ar tE-1n i 0 7.8s W da channe M 6 2.8t a l 8 sec, 7.8W M 6 in E-1 E-1Argentina0n y i b 0channe c 7.6W d channese M an 7 5 t 2.7c a 2. l, se t 9Indiaa l n and Rumanian simulations respectively. It can be seen from Table 3.4.8 that maximum bundle powers at t=0 are found to be bundlt a 814.W ek 9 positio n E-ll-7 bundlt ,a 816.W k e1 position E-ll-7,and t 803.a W 4k bundle position D-ll- Argentineany 7b , India Rumaniad nan n simulations respectivelye Th . value, locatio timd f maximunan eo m bundle power durin transiene gth foune e b ar t o t d bundlt 948.a W 6ek position E-ll- 2.8t 7a 8 sec, E-ll-t 952.a W 36k bundle positio 2.7t na 9 sec, and 922.0 kW at D-ll-6 bundle position at 2.5 sec by Argentinean, Indian and Rumanian simulations respectively. Table 3.4.9 summaries the salient features of the results transiente oth f . 79 TABLE 3.4.7 VARIATIO F MAXIMUNO M CHANNEL POWER WITH TIME Time Argentina India-ThPD Romani* * a Sec MCP (MW) LOG MCP (MW) LOG P (MWMC c )Lo 0.0 6.74 E-10 6.72 E-10 6.66 E-10 0.2 6.77 E-10 6.76 E-10 6.72 E-10 0.4 6.83 E-10 6.82 E-10 6.79 E-10 0.6 6.89 E-10 6.89 E-10 6.85 E-10 0.8 6.97 E-10 6.96 E-10 6.92 E-10 1.0 7.05 E-10 7.05 E-10 6.98 E-10 1.2 7.14 E-10 7.15 E-10 7.09 E-10 1.4 7.24 E-10 7.25 E-10 7.19 E-10 1.6 7.35 E-10 7.36 E-10 7.29 E-10 1.8 7.47 •E-10 7.48 E-10 7.40 E-10 2.0 7.60 E-10 7.61 E-10 7.50 E-10 2 .2 7.70 E-10 7.71 E-10 7.58 E-10 2 .4 7.76 E-10 7.77 E-10 7.64 E-10 2.6 7.80 E-10 7.82 E-10 7.66 E-10 2 .8 7.85 E-10 7.86 E-10 7.55 E-10 3.0 7.51 S-10 7.40 S-10 7.45 -- 3 .2 6.93 S-10 6.86 S-10 6.84 S-13 3 .4 6.03 S-10 5.97 S-10 6.01 S-13 3.6 4.77 S-10 4.70 S-10 4.83 S-10 3 .8 3 .28 S-ll 3 .24 S-ll 3 .33 S-ll 4.0 2.63 E-ll 2.63 E-ll 2.60 M-04 5.0 2.26 E-ll 2.27 E-ll 2.24 E-ll 6.0 2.02 E-ll 2.03 E-ll 2.00 E-ll 7.0 1.84 E-ll 1.85 E-ll 1.82 E-ll 8.0 1.69 E-ll 1.70 E-ll 1.67 E-ll 9.0 1.57 E-ll 1.57 E-ll 1.55 E-ll 10 .0 1.46 E-ll 1.47 E-ll 1.45 E-ll ** Values were obtained by interpolation 80 TABLE 3.4.8 VARIATION OF MAXIMUM BUNDLE POWER WITH TIME Time Argentina India -ThPD Romani* * a (sec) Value LOG . Value LOG . Value LOG . (kw) (kw) (kw) 0. 0 814 .9 E-ll-7 816 .1 E-ll-7 803 .3 D-ll- 6 0.2 817 .7 E-ll-7 819 .9 E-ll-6 810 .9 D-ll- 6 0.4 824 .2 E-ll-7 826 .9 E-ll-6 818 .5 D-ll- 6 0.6 832 .2 E-ll-7 835 .2 E-ll-6 826 .1 D-ll- 6 0.8 841 .2 E-ll-7 844 .6 E-ll-6 833 .7 D-ll- 6 1.0 851 .2 E-ll-7 854 .9 E-ll-6 841 .3 D-ll- 6 1.2 862 .2 E-ll-7 866 .2 E-ll-6 853 .4 D-ll- 6 1.4 874 .3 E-ll-7 878 .5 E-ll-6 865 .5 D-ll- 6 1.6 887 .5 E-ll-7 891 .9 E-ll-6 877 .7 D-ll- 6 1.8 901 .8 E-ll-7 906 .4 E-ll-6 889 .8 D-ll- 6 2.0 917 .4 E-ll-7 922 .2 E-ll-6 901 .9 D-ll- 6 2.2 929 .0 E-ll-7 934 .2 E-ll-6 911 .4 D-ll- 6 2.4 936 .2 E-ll-7 941 .8 E-ll-6 919 .1 D-ll- 6 2.6 942 .0 E-ll-7 947 .8 E-ll-6 921 .5 2.8 947 .2 E-ll-7 952 .3 E-ll-6 908 .6 E-ll- 6 3.0 907 .8 S-ll-7 898 .1 S-ll-6 876 .3 S-ll- 6 3 .2 839 .4 S-ll-7 834 .3 S-ll-6 824 .3 S-12- 6 3 .4 732 .9 S-ll-7 728 .2 S-ll-6 726 .8 D-ll- 6 3 .6 586 .7 S-ll-7 579 .8 S-ll-6 590 .4 S-ll- 6 3.8 416 .3 S-ll-7 412 .7 S-ll-6 421 .1 S-ll- 6 4.0 337 .0 E-ll-7 338 .7 E-ll-6 334 .2 D-ll- 6 5.0 290 .2 E-ll-7 291 .7 E-ll-6 288 .0 D-ll- 6 6.0 259 .7 E-ll-7 260 .9 E-ll-6 257 .9 D-ll- 6 7.0 236 .2 E-ll-7 237 .2 E-ll-6 234 .7 D-ll- 6 8.0 217 .1 E-ll-7 217 .9 E-ll-6 215 .8 D-ll- 6 9.0 201 .2 E-ll-7 201 .8 E-ll-6 200 .0 D-ll- 6 10 .0 187 .8 E-ll-7 188 .2 E-ll-6 186 .7 D-ll- 6 ** Values were obtained b;/ interpolation 81 TABLE 3.4.9 SUMMAR RESULTF YO S Argentina India Romania -ThPD 1. Max.Channel 0 Powet= t ra (a) value (MW) 6.74 6.72 6.66 (b) location E-10 E-10 E-10 2. Max.Bundle 0 Powet= t ra (a) value (KW) 814.9 816.1 803.4 (b) location E-ll-7 E-ll-7 D-ll-6 3. Trip Time (sec) 1.98 1.98 1.99 4. Max. Positive Reactivity (a) value (mk) 0.588 0.590 0.575 5 2. 0 2. (b) 6 Tim2. e (sec) 5. Max. Relative Power (a) Value 1.162 1.163 1.150 (b) Time (sec) 2.88 2.79 2.50 6. Max. Channel Power (a) value (MW) 7.86 7.86 7.67 (b) location E-10 E-10 E-10 (c) Time (sec) 2.88 2.79 2.50 7. Max. Bundle Power (a) value (KW) 948.6 952.3 922.0 (b) location E-ll-6 E-.11-6 D-ll-6 (c) Time (sec) 2.88 2.79 2.50 8. Dynamic Reactivity after Insertio controf no l rods (a) Value (mk) -8.58 -8.50 -8.56 (b) Time (sec) 3.88 3.88 3.9 9. Dynamic Reactivity (mk) at t=10 sec -8.55 -8.56 -8.52 10. Relative Power at t=10 sec 0.211 0.210 0.212 3.4.4. Conclusions All the three contributors, namely, Argentina, India-ThPD and Romania used the codes based on improved quasi-static (IQS) method. The results are in good agreement with each othe expectes a r d sinc methode eth f solutioso n use similare dar . 3.5. TAS : INFLUENCK5 ISOTOPF EO VOIN EO D REACTIVITY s see section wa ni t I tha1 n3. t a largther s eewa differenc value th f voi o en i ed induced reactivity calculated by the multigroup codes and PPV code. In order to understand the reasons for the difference, a benchmark was formulated in which the rnfluence of various isotopes on void induced reactivity was studied. Reactor physicists from four countries solved this benchmark problem related to the influence of fuel isotopic composition on void effect: Argentina, India, Pakistan and Romania. Three codes have been used: WIMS (Argentina, India-RED, Pakistan, Romania), POWDERPUFS-V (Argentina, Romania) and CLUB (India- ThPD). Argentina, Romania, India-RED, and Pakistan used their respective WIMS libraries. India-ThPD also used a WIMS library. 82 3.5.1. Description of the benchmark problem 3.5.1.1. Basic cell data The complete set of input data for the lattice cell reference benchmark problem is reproduce Tabln di e 3.1.1 lattice Th . e cell mode presentes i l Figurn di e 3.1.1 table d Th .e an basifigure P th basee CR c ar e n reportdo , Ref. [43]. 3.5.1.2. Cases to be analyzed The problem can be described as a series of lattice cell calculations for the CANDU 37 element cluster, that form a gradual transition from a fresh fuel cell to a 4000 MWD/T changing the concentration of the most important isotopes one at a time. The objective was to study the influence of the isotopes on the void reactivity, Ref. [52]. Seven case definede ar s r eacFo .h case run o necessarye tw sar , : a) coolant case, using 0.804 g/cc coolant density; b) void case, using 0.0 g/cc coolant density. Except for the fuel composition, all the seven cases are identical. Case 1: Natural Uranium case N - U235 = 1.60309 E-4 nuclei/cm barn N - U238 = 2.21036 E-2 —- " --"— N - O16 = 4.45302 E-2 —"-—"-— fuee th l l bundlAl e rods hav same eth e fuel composition. Case 2: 4000 MWd/MeU. only U235 The fuel composition differs from one bundle ring of rods to the other. Case 3: 4000 MWd/MgU. Plus Pu239 The case 2 fuel composition is modified only by Pu239 addition. Cas : 400e4 0 MWd/MgU. Plus Pu240. Case 5: 4000 Mwd/MgU. Plus Pu241. Case 6: 4000 MWd/MgU. Plus Pu242. cass otheCase A , Pluth e6 : l er7 sal fissio n products calculate code 400r th efo y db 0 MWd/MgU. Xenon may be added, the isotopic densities being calculated by the code. 83 outeFroe th centra e mro t th ring d Case lro th , fuee6 l isotope densitie nuclen si m c i/ barn are: U23 9.92795= 8 E-5, 9.77759 E-5, 9.16703 E-5, 7.85975 7E- U238 = 2.21036 E-2 O16 = 4.45302 E-2 Pu239 = 3.92496 E-5, 3.96730 E-5, 4.12802 E-5, 4.71904 E-5 Pu240 = 7.13721 E-5, 7.37784 E-6, 8.38009 E-6, 1.13273 E-5 Pu241 = 1.00688 E-6, 1.05085 E-6, 1.24871 E-6, 1.88405 E-5 Pu242 = 1.02898 E-7, 1.09466 E-7, 1.46685 E-7, 2.75855 E-7 All the fuel bundle rods have the same U238 and O16 isotope number densities. 3.5.1.3. Results providedbe to ) a kinfcool, kinfvoid, rhoinf, keffcool, keffvoid, rhoeff, where: rho (mk) = (l/k«,,,, - l/kvoid) * 1000. totae Th l cell ) reactiob n rates, (fas thermal+ t productio: ) absorptiond nan . Thee yar normalised in order to have the U238 thermal absorption equal to unity. The relative difference alse ar so required, (voi* 0 dcool- 10 di / = )coolf . c) 11 thermal reaction rates, normalized to have U238 thermal absorption = 1.0: cell production, cell absorption, fuel absorption r eacfo hd specifiean , d isotope, taking into account the amount of each nuclide, (U235, Pu239, Pu241) production, (U235, Pu239, Pu240, Pu241, Pu242) absorption. relative Th e difference alse ar so required, dif = 100 * (void - cool) / cool. 3.5.2. Computer codes used The computer codes WIMS, POWDERPUFS- CLUd Van B have been used whice har discussed in section 2.1. 3.5.3. Comparison of the main results The participant contributions may be seen in Refs. [63-67]. 84 3.5.3.1. Multiplication factors e k-effk-ind Th s an calculatea f participante th y db s showi s Tabln ni e 3.5.1-4. Romania has calculated the WIMS values without including xenon. India-ThPD has calculated them usingroups9 6 d g orde n an i bot, 7 studo t hr2 sensitivite yth e resulte th th f o yo t s energy group structure. e WIMS-ArgentinaTh , CLUB-27 CLUB-69d Gan G prediction e similarar s e Th . difference energs9 6 betwee d yan group7 n2 s multiplication constant nearl. e sar mk y1 The India-RED and Pakistan WIMS infinite multiplication factors are close to each others. The Pakistan WIMS effective multiplication constants are much lower than the infinite values. multiplicatioV PP e Th n factor slightle ar s y different fro WIMe mth S ones. TABLE 3.5.1 K-INFINITY WIMS ESTIMATION CASE ARGENTINA INDIA-RED PAKISTAN ROMANIA NO XENON FRESH COOL 1.082149 1.077769 1.07717 1.106748 FUEL VOID 1.103189 1.097794 1.09547 1.124109 U235 COOL 0.810348 0.806705 0.80683 0.839024 VOID 0.830060 0.825590 0.82367 0.854634 ADD COOL 1.106246 1.105337 1.09313 1.121607 PU239 VOID 1,.121633 1.120422 1.11198 1.139484 ADD COOL 1.077597 1.076389 1.06453 1.091858 PU240 VOID 1.093674 1.092252 1.08411 1.110185 ADD COOL 1.087891 1.086772 1.07490 1.102072 PU241 VOID 1.104009 1.102690 1.09451 1.120425 ADD COOL 1.087813 1.086696 1.07483 1.101995 PU242 VOID 1.103928 1.102606 1.09442 1.120342 85 TABLE 3.5.2 K-INFINITY PPV AND CLUB ESTIMATION CASE ARGENTINA INDIA INDIA ROMANIA PPV -THPD -THPD PPV CLUB-27 CLUB- 69 NO XENON FRESH COOL 1.080507 1.0824 1.0829 1.126240 FUEL VOID 1.098493 1.1029 1.1015 1.144979 U235 COOL 0.799812 0.8115 0.8104 0.844718 VOID 0.815091 0.8305 0.8280 0.860945 ADD COOL 1.120674 1.1090 1.1104 1.161548 PU239 VOID 1.132333 1.1251 1.1239 1.174214 ADD COOL 1.087151 1.0797 1.0813 1.125647 PU240 VOID 1.099042 1.0964 1.0954 1.138517 ADD COOL 1.098383 1.0901 1.0917 1.136743 PU241 VOID 1.110397 1.1068 1.1058 1.149729 ADD COOL 1.098280 1.0900 1.0916 1.136633 PU242 VOID 1.110288 1.1067 1.1057 1.149614 TABLE 3.5.3 K-EFFECTIVE WIMS ESTIMATION CASE ARGENTINA INDIA-RED PAKISTAN" ROMANIA NO XENON FRESH COOL 1.052349 1.047656 0.98302 1.076088 FUEL VOID 1.071871 1.066044 0.99733 1.091826 U235 COOL 0.785515 0.78167 0.72919 0.813348 VOID 0.803957 0.79912 0.74273 0.827529 ADD COOL 1.076845 1.075658 1.00023 1.091344 PÜ239 VOID 1.090702 1.088954 1.01515 1.107640 ADD COOL 1.049131 1.047649 0.97453 1.062589 PU240 VOID 1.063716 1.061786 0.99032 1.079379 ADD COOL 1.059262 1.057876 0.98434 1.072624 PU241 VOID 1.073874 1.072038 1.00013 1.089441 ADD COOL 1.059191 1.057808 0.98427 1.072550 PU242 VOID 1.073806 1.071953 1.00008 1.089360 86 TABLE 3.5.4 K-INFINITCLUD AN B V ESTIMATIOYPP N CASE ARGENTINA INDIA INDIA ROMANIA PPV -THPD -THPD PPV CLUB-27 CLUB-69 NO XENON FRESH COOL 1.047864 1.0530 1.0532 1.091751 FUEL VOID 1.064730 1.0719 1.0704 1.109326 U235 COOL 0.772948 0.7870 0.7857 0.815672 VOID 0.787373 0.8048 0.8022 0.830999 ADD COOL 1. 087888 1.0800 1.0813 1.127138 PU239 VOID 1.098421 1.0946 1. 0834 1.138606 ADD COOL 1.055635 1.0517 1.0531 1.092622 PU240 VOID 1.066412 1.0668 1.0658 1.104310 ADD COOL 1.066664 1.0618 1.0634 1.103529 PU241 VOID 1.077551 1.0770 1.0761 1.115320 ADD COOL 1.066564 1.0617 1.0633 1.103423 PU242 VOID 1.077446 1.0770 1.0760 1.115209 3.5.3.2. Void reactivities The void reactivities and the change of void reactivity due to successive addition of isotope tabulatee sar i Tabldh e 3.5.5-6 delte Th .a value definee sar : as s delta (n) = rhoeff(n) - rhoeff(n-l) It can be seen that only U235 and Pu239 have significant influence on the void induced reactivity. The maximum change in void reactivity due to U235 is 11.9 mk and that due to Pu239 is -17.4 mk. This is due to their large number densities. Pu240 introduces about 1 mk positive reactivity. The 69 group calculations from CLUB show that there is a noticeable difference of about 2 mk consistently in each case. 87 TABLE 3.5.5 VOID REACTIVITY WIMS ESTIMATIONS (CHANGE OF VOID REACTIVITY DUE TO SUCCESSIVE ADDITION OF ISOTOPES) CASE ARGENTINA INDIA-RED PAKISTAN ROMANIA NO XENON 1. FRESH RHO 17 .307 16 .464 14 .60 13 .395 2. U235 RHO 29 .202 27.935 25 .00 21.070 DEL 11 .895 11 .471 10 .40 7.675 3. ADD RHO 11 .798 11 .351 14 .70 13 .481 PU239 DEL -17 .404 -16 .584 -10 .30 -7 .589 4. ADD RHO 13 .069 12 .709 16 .36 14 .639 PU240 DEL 1.271 1.358 1.66 1.158 5. ADD RHO 12 .846 12 .488 16 .05 14 .391 PU241 DEL -0 .223 -0 .221 -0 .31 -0 .248 6. ADD RHO 12 .850 12 .474 16 .06 14 .387 PU242 DEL 0.004 -0 .014 0.01 -0 .004 rho(tnk 1000= ) .* (1/keffcool 1/keffvoid) del(n rho(n= ) - rho(n-l) ) (n=2,3,4,5,6) TABLE 3.5.6 VOID REACTIVITY PPV AND CLUB ESTIMATIONS (CHANGE OF VOID REACTIVITY DUE TO SUCCESSIVE ADDITION OF ISOTOPES) CASE ARGENTINA INDIA INDIA ROMANIA PPV -THPD -THPD PPV CLUB -27 CLUB9 -6 NO XENON 1. FRESH RHO 15 .117 16 .76 15 .26 14 .511 2. U235 RHO 23 .702 28 .18 26 .12 22 .611 DEL 8.585 11 .42 10 .86 8.100 3. ADD RHO 8.815 12 .30 10 .2 8.936 PU239 DEL -14 .887 -15 .88 -15 .92 -13 .675 4. ADD RHO 9.573 13 .53 11 .3 9.687 PÜ240 DEL 0.758 1.23 1.1 0.751 5. ADD RHO 9.472 13 .32 11 .10 9.580 PU241 DEL -0 .101 -0 .21 -0 .2 -0 .107 6. ADD RHO 9.469 13 .31 11 .09 9.578 PU242 DEL -0 .003 -0 .01 -0 .01 -0 .002 rho(mk 1000= ) .* (I'/keffcool 1/keffvoid) del(n rho(n= ) - rho(n-l) ) (n=2,3,4,5,6) 88 3.5.3.3. WIMS reaction rates The required reaction rates for cases 4, 5, 6 and 7 are rather similar. Consequently, only cases 1, 2, 3 and 6 have been tabulated in Tables 3.5.7-10. The reaction rates of interest, namely, cell production, cell absorption, thermal cell production, thermal cell absorption, thermal fuel absorption, thermal production and absorption of U235 and Pu239, and thermal absorptio Pu24f no 0 have been tabulated. e naturaTh l normalizatio f reactiono n makrateo t s i se total cell absorption unity. However, the total cell absorption is significantly changed by the channel voiding. As the Westcott cross section of U238 is only slightly modified by the channel voiding, the normalization to unit U238 "Westcott absorption" may be suggested. As most of the participants used multigroup codes, the normalization to unit U238 thermal absorption was selected. TABLE 3.5.7 WIMS REACTION RATE R EACFO S H NUCLID RELATIVD AN E E CHANGE DU LOSO ET F COOLAN O S TCAS( : FRES 1 E H FUE- L )' ABSOLUTE VALUES RELATIVE CHANGE ARGENTINA COOLANT VOID ARG. INDIA PAK ROMANIA NO XE TOTAR LR CELL PROD. 3.8273 3.8637 0.953 1.017 0.62 0.418 CELL ABS. 3.5411 3.5067 -0.970 -0.835 -1.06 -1.133 THERMAL R R CELL PROD. 3 .5236 3.5385 0.422 0.488 -0.14 -0.118 CELL ABS. 2.9683 2.9655 -0.094 0.054 -0.26 -0.389 FUEL ABS. 2.7889 2.7919 0.108 1.040 0.04 -0.070 U235 PROD. 3.5236 3.5385 0.422 0.362 -0.14 -0.118 U235 ABS. 1.6930 1.6987 0.341 0.308 -0.13 -0.111 All reaction rates are normalized to unit U238 thermal abs. Relative chang= (void/cool-1 ) (% e 0 10 * ) 89 TABLE 3.5.8 WIMS REACTION RATE R EACFO S H NUCLID RELATIVD AN E E CHANGE DU LOSO T EF COOLAN O S TCAS( : U23 2 E 5 4000 MWD/MTU) ABSOLUTE VALUES RELATIVE CHANGE ARGENTINA COOLANT VOID ARG. INDIA PAK ROMANIA TOTAR LR CELL PROD. 2.0774 2.1068 1.416 1.468 1.08 0.746 CELL ABS. 2.5662 2.5408 -0 .989 -0 .846 -0. 99 -1.094 THERMAR LR CELL PROD. 1.8867 1.9007 0 .745 0.791 0.08 0 .055 CELL ABS. 2.1656 2.1640 -0 .074 0.079 -0.16 -0.319 FUEL ABS. 2.0022 2.0084 0.310 0.291 0.36 0 .030 U235 PROD. 1.8866 1.9007 0.745 0.671 0.08 0.055 U235 ABS. 0.9062 0.9122 0.655 0.612 0.09 0.062 All reaction rates are normalized to unit U238 thermal abs. Relative = chang(void/cool-1 ) (% e0 10 * ) TABLE 3.5.9 WIMS REACTION RATES FOR EACH NUCLIDE AND RELATIVE CHANGE DUE TO LOSS OF COOLANT ( CASE 3 : ADD PU239 ) ABSOLUTE VALUES RELATIVE CHANGE ARGENTINA COOLANT VOID ARG. INDIA PAK ROMANIA TOTAL R R CELL PROD. 4.2516 4.2100 -0 .979 -0 .944 0.70 0.496 CELL ABS. 3.8437 3 .7580 -2 .232 -2 .270 -1.01 -1 .081 THERMAR LR CELL PROD. 3 .9364 3 .8738 -1 .590 -1.507 -0.07 -0 .050 CELL ABS. 3.2352 3 .1847 -1 .561 -1 .396 -0.22 -0 .344 FUEL ABS. 3.0525 3 .0122 -1 .320 -1 .397 0.07 -0 .026 U235 PROD. 1.8791 1.8931 0.742 0.686 0.15 0.112 U235 ABS. 0.9022 0.9081 0.663 0.633 0.15 0.112 PU239 PROD. 2.0572 1.9807 -3 .719 -3 .758 -0.29 -0 .209 PU239 ABS. 1.0537 1.00-71 -4 .420 -4 .456 -0.26 -0 .184 All reaction rate normalizee sar unio dt t U238 thermal abs. Relative change (%) = (void/cool-1) * 100 90 TABLE 3.5.10 WIMS REACTION RATES FOR EACH NUCLIDE AND RELATIVE CHANGE ISOTOPEU P DUD LOSO T ECOOLANF AN SO S) U L AL TCAS( : 6 E ABSOLUTE VALUES RELATIVE CHANGE ARGENTINA COOLANT VOID ARG. INDIA PAK ROMANIA TOTAL R R CELL PROD. 4.3888 4.3434 -1.034 -0.928 0.072 0.517 CELL ABS. 4.1979 4.0936 -2.486 -2.355 -1.08 -1.129 THERMAL R R CELL PROD. 4.0470 3.9819 -1.610 -1.482 -0.04 -0.022 CELL ABS. 3.4784 3.4250 -1.534 -1.389 -0.20 -0.328 FUEL ABS. 3.2902 3 .2476 -1.296 -1.374 0.09 -0.008 U235 PROD. 1.8850 1.8991 0.747 0.682 0.15 0.114 U235 ABS. 0.9050 0.9110 0.668 0.629 0.16 0.115 PU239 PROD. 2.0700 1.9931 -3.714 -3.66 -0.21 -0.135 PU239 ABS. 1.0606 1.0138 -4.408 -4.34 -0.17 -0.095 PU240 ABS. 0.0516 0.0509 -1.235 -1.19 -0.40 -0.306 All reaction rate normalizee sar unio dt t U238 thermal abs. Relative change (%) = (void/cool-1) * 100 Argentin India-REd aan D prediction similare sar U23e Th . 5 productio absorptiod nan n rate increasee s ar voidinge th y dchange b th , e being less tha Pu23e n Th 0.7 . 9 5% productio n and absorption are significantly decreased by the voiding by -3.7 % and -4.4 % respectively. Pakistan and Romania estimations are rather similar. The changes due to channel voiding are less than the Argentina and India-RED ones, in particular for the Pu239 reaction rates. It is assumed that thi WIMd s ol relate e findine b S th y libraro dt gma y used. 3.5.3.4. Fuel neutron temperature problem e significanTh t PPV-WIMS discrepancie explainee b y sma basiy db c principlef o s model and nuclear data differences. The Romania participants believed that the PPV fuel neutron temperature prediction may play a significant role. The WIMS code was run using 69 energy groups. "Cold" lattice cells have been calculated, when all the material temperatures are equal to the moderator temperature, 68 °C [67]. The PPV fuel neutron 91 temperatur calculates ei codee th y ,d b usin empirican ga l formula that takes into accoune tth material propertie temperaturesd san WIMe Th . S equivalent temperatur computes ewa d using group fluxes, group production rate groud san p absorption rates maie Th .n findin thas gi t PPV fuel neutron temperature estimations are not very reliable, from the point of view of the accurate calculation requirements of void effect. The Rumanian report [67], could describe the noticed PPV-WIMS discrepancies as the algebraic sum of fuel neutron temperatures effect and "non-thermal" group effect. Up to now otheo n , r similar fuel neutron temperature investigatio availables ni . 3.5.4. Conclusions and recommendations e estimation codee Th th l 1sal . indicatf so e thae voith t d induced reactivity valus i e practically dictated by the U235 and Pu239 number densities. As the burnup of the fuel increases positive th ,reactivit G eLO y decreases. agreemene Th code th ef . o tpredicte2 reactivitieG dLO burnupsw bettes si lo r e rfo Th . lowes reactivitG LO t estimates yi POWDERPUFS-e th y db V code. voie Th d relate . 3 d fuel neutron spectrum perturbation play significansa t rol voin ei d effect estimation. More detailed investigation needede sar . 4. Four independent WIMS predictions have been compared. Although the input data are almost identical s closa t s expecteda e result no th , e s assumei ar s t I . d that this finding is due to the differences of the nuclear data from various WIMS libraries. An available standard PHWR-ENDF/B-6 WIMS Library will be welcome. 3.6. TAS : MIXEK6 D LOADING The intent of this benchmark is to institute an intercomparison of different calculational method reactoa r sfo r core with more tha kine nfuef on d o l presen coree th n .i t The benchmark considers two types of fuels, namely, natural UO2 and PuO2 mixed in UO2 (MOX). Three countries participated with their nine submissions in this benchmark. 3.6.1. Description kindo f fueTw so l have been considered: a) The natural uranium 37 rod cluster: standard lattice of this CRP on benchmarks. b) The same cluster, but with 0.4 % plutonium added to the 19 inner pins. core fress Th ei h with equilibrium xenon centra0 10 . l channel a 10x1 n i s 0 square constitutes the natural uranium region. The rest of the core, i.e., 280 channels, is of the above described mixed oxide variety 37 rod cluster. Power distribution calculations for the entire core have been carried out for the following cases: 1. Adjusters fully in, shut-off-rods (SRs) fully out, zonal control tubes (ZCCs) partially nominae th fille n i thes da e l yar configuration . 92 2. Adjusters fully out, shut-off-rods fully out, zonal control tubes nominal configuration. . 3 Adjusters full , shut-off-rodin y s fully out, zonal control tubes empty. . 4 Adjusters full , shut-off-rodin y s fully out, ZCCs full. 5. Adjusters fully in, SR fully in, ZCCs nominal. The following quantities have been calculated: a) k-eff b) Maximum bundle power and location. c) Maximum channel powe locationd an r . The plutonium isotopic composition has been taken as: Pu239 = 64 w/o Pu240 = 29 w/o Pu241 = 6 w/o Pu242 = 1 w/o 3.6.2. Modelizatio calculatiod nan n l participantAl s have modelle problee dth mora m n r lesi e o s similar way, wite hth fuel cell being replaced by an equivalent "lattice paste". The core thus has three major homogeneous regions, viz. e natura,th l uranium zone e mixeth , d oxide e th zone d an , reflector. reactivite Th y devices have been modelle somy db e participant modD d 3- ean e th n si modellinD mode3- D b e y2- some Th . gth treatn ei fuee sth l channereactivite th d lan y device explicitly in the rectangular cell and considers them as being oriented in directions at right angles to each other, as in fact they are. The 2-D method considers the device as inserted in the centre of a cylindrical supercell in which the fuel region has been replaced by the lattice paste. The core level calculations have been carried out by all participants in 3-D. 3.6.3. Submissions This task has received nine submissions from four participants. These nine and the methods usesummarizee b they db n mca followss da : a. Pakistan - Cell calculations by WIMS-D/4. Core calculations by CITATION, . b India-RED Cel- l calculation WIMS-D/4y sb . Core calculation ANAMIKAy sb . Romani a- Cel l calculation POWDERPUFy b s WIMSd San . Core calculationy b s CHEBXEMAX. Xenon has been treated in two different ways. Either implicitly by including 93 e latticth n ii et calculations r explicitlo , y during core calculations. Ther e thuar e s four submissions here, which have been listed as c. POWDERPUF impliciS+ t xenon d. POWDERPUFS + explicit xenon e. WIMS + implicit xenon f. WIMS + explicit xenon India-ThPD - Cell calculations were performed by CLUB and core calculations were done by DIMENTRI and CEMESH. The DIMENTRI calculation was repeated using tighter convergence criteri eigenvalueflud n ao x an . Thus ther three ear e case alln i s , listes da g. CEMESH calculations h. DIMENTRI calculations with convergence criteria of 0.001 on flux and 0.000001 on eigenvalue, i. DIMENTRI calculations with convergence criteria of 0.0001 on flux and 0.0000001 on eigenvalue. 3.6.4. Results The results from the nine submissions have been summarized in four tables (Tables 3.6.1 through 3.6.4). These tables give k-eff, the maximum bundle power and location, the maximum channel power and location, and the reactivity worths of the reactivity devices. A look at the tables shows a fair degree of broad agreement among the various submissions. TABLE 3.6.1 K-eff VALUES ADJ SR zee a b c d e f g h i IN OUT REF 1.0877 1.0886 1.0972 1.1106 1.0882 1.0850 1.0964 1.0962 1.0962 OUT OUT REF 1.0971 1.0981 1.1087 1.1216 1.0998 1.0962 1.1043 1.1043 1.1043 IN OUT EMPT 1.0914 1.0919 1.1012 1.1145 1.0919 1.0885 1.0998 1.0997 1.0997 IN OUT FULL 1.0832 1.0843 1.0924 1.1058 1.0837 1.0807 1.0922 1.0918 1.0919 IN IN REF 1.0801 1.0834 1.0901 1.1035 1.0811 1.0780 1.0889 1.0886 1.0886 a PAKISTAN b INDIA - 2 c ROMANIA (PPV-IM) d ROMANIA (PPV-EX) e ROMANIA (WIMS-IM) f ROMANIA (WIMS-EX) g INDIA (CEMESH} h INDIA (DIMENTRI-Flux Convergence Criteria 0.001} i INDIA (DIMENTRI-Flux Convergence Criteria 0.0001) 94 TABLE 3.6.2 BUNDLE POWERS ( KWth ) , Ptotal = 2071 MWth ADJ SR zee a b c d e f g h i IN OUT REF 1082 996 1097 1085 1059 1044 1060 1015 1090 D-11,7 D-11,7 D-12,6 D-12,6 D-12,7 D-12,7 D-12,6 D-11,7 D-12,7 OUT OUT REF 1019 962 1088 1084 1053 1041 1020 998 1037 D-ll. 6 , E-11,7 D-12,6 D-12,6 D-12,7 D-12,7 D-ll, 6 D-11,7 D-12,7 IN OUT EMPT 937 917 947 946 922 917 936 956 962 D-ll, 6 D-11,7 D-11,7 D-11,7 D-11,7 D-11,7 T-12, 6 T-11,7 T-ll, 7 IN OUT FULL 1149 1018 1183 1166 1147 1122 1162 1095 1195 S-11,6 T-11,6 T-11,7 T-11,7 T-11,7 T-11,7 T-12, 7 T-11,7 T-12, 7 IN IN REF 1101 999 1104 1095 1072 1050 1107 1123 1147 K-03,6 L-04,7 D-12,7 D-12,7 D-12,7 P-12,7 L-20,7 M-04,7 L-04,7 a PAKISTAN b INDI2 - A c ROMANIA (PPV-IM) d ROMANIA (PPV-SX) e ROMANIA (WIMS-IM) f ROMANIA (WIMS-EX) g INDIA {CEMESH) h INDIA {DIMSNTRI-Flux Convergence Criteria 0.001) i INDIA (DIMENTRI-Flux Convergence Criteria 0.0001) TABLE 3.6.3 CHANNEL POWERS ( KWth ) , Ptotal = 2071 MWth ADJ SR zee a b C d e f g h i IN OUT REF 8403 7743 8590 8513 8284 8163 8284 7750 8310 D-ll D-10 D-12 D-12 D-12 D-12 D-13 D-10 D-ll OUT OUT REF 7526 7107 7592 7616 7338 7306 7589 7320 7541 D-ll E-10 D-12 D-12 D-12 D-12 D-ll M-05 D-10 IN OUT EMPT 7516 7309 7729 7726 7508 7478 7582 7590 7627 D-ll T-10 D-ll D-ll D-ll D-12 T-12 T-ll T-ll IN OUT FULL 8494 7736 8769 8679 8542 8399 8720 8010 8702 S-10 T-10 T-10 T-10 T-10 T-10 T-13 T-10 T-10 IN IN REF 8349 7832 8424 8419 8224 8178 8515 8470 8526 L-04 L-04 M-19 M-19 M-19 . M-19 M-19 M-04 M-04 a PAKISTAN b INDI2 - A c ROMANIA (PPV-IM) d ROMANIA (PPV-EX) e ROMANIA (WIMS-IM) f ROMANIA (WIMS-EX) g INDIA (CEMESH) h INDIA (DIMENTRI-Flux Convergence Criteria 0.001) i INDIA (DIMENTRI-Flux Convergence Criteria 0.0001) 95 TABLE 3.6.4 WORTHS OF REACTIVITY DEVICES DEVICE a b c d e f g h i ADJ 9.41 7.95 9.45 8 .87 9.70 9.35 6.55 6.76 6.76 zee 6.89 5.47 7.27 6 .98 6.88 5.70 6.29 6.54 6.50 SR 6.48 4.34 5.98 5.76 6.09 6.00 6.27 6.37 6 .37 PAKISTAN INDIA - 2 ROMANIA (PPV-IM) ROMANIA (PPV-EX) ROMANIA (WIMS-IM) ROMANIA (WIMS-EX) INDIA (CEMESH) INDIA (DIMENTRI-Flux Convergence Criteria 0.001) INDIA (DIMENTRI-Flux Convergence Criteria 0.0001) 3.6.5. Conclusions All methods appea e locatio agreo th maximue t r th n o ef no m bundl channed ean l powers. The differences only exist in either a neighbouring channel or a channel symmetrically placed on the other side of the core. It is generally seen that the maximum bundle and channel powers predicted by India-RE lowee Dar r tha otherse nth . This coul explainee db face th t thay db t they have taken a bundle as one mesh point, while most of the others have considered a bundle as two mesh points. A comparison of the two cases from India-ThPD, cases i and h leads to the conclusion that tighter convergence criteria will result in the prediction of higher maximum powers. This appears to reinforce the finding of the last paragraph concerning smaller mesh lengths. Explicit xenon predicts lower power peaking than implicit xenon, which is only to be expected. Convergence criteri appeat no o havao d rt e much impac k-efe th fn tvalueo judges sa d from the submissions h and i. K-eff values from WIMS seem to be generally lower than those from POWDERPUFS. Explicit expectexenoe b n nca reduco dt k-effe eth , sinc thin ei s case, xenon density is higher at higher flux points. This effect is seen in the two submissions e and f with WIMS. However, the effect appears to be reversed in the POWDERPUFS results, as seen from submissions c and d. This would, of course, depend upon the effective flux used for xenon calculation at the lattice level. wortZCCe faie n i th Th re f hagreemeno sar casesl al wort e r Th fo t.adjuste f ho r rods show rathea s r wide spread l threAl . e submissions from India-ThPD show lower values, whil foue eth r submissions fro frome on mRomani e Pakistath d aan n show higher values, with the India-RED value falling somewhere in between. A comparison with the worth calculations made for the fully natural uranium core shows a much closer agreement in the adjuster worth value f Romaniaso , India-ThP India-REDd Dan . 96 The SR worths agree fairly well over the spectrum of submissions given here, except India-REfoe rth D value whic significantls hi y lower referencA . naturae th o et l uranium core showed that there too, the value given by India-RED is lower. 3.7. TASK 7: CALCULATION OF THE FISSION PRODUCT INVENTORIES AND THEIR DECAY HEAT During the first Research Co-ordination Meeting (RCM) held in 1990 in Buenos Aires, Argentina benchmara , k relate fissioo dt n products inventorie theid san r decay heat rates definewa s d [52]. Three countries, namely, Argentina, Pakista d Romanian n a participated in this benchmark. The summary of the results obtained by the various participant presentes i s thin di s section. 3.7.1. Benchmark description core Th e considere definee originae on th e n di th s ldi IAEA PHWR benchmarke Th . bundle burnup distribution has been taken from the Embalse Nuclear Power Plant. The following two cases are defined for the fission product inventories and the decay heat calculation benchmark. 3.7.1.1. Case A Core inventories of radioactive isotopes (and their decay heat rates) are calculated for reference th e burnup distribution Table Th . e 3.7.1 shows group-wise fuel bundle burnup distribution. TABLE 3.7.1 GROUP-WISE DISTRIBUTION OF FUEL BUNDLE BURNUPS Group MWD/Mg Numbef ro Weighted No. Bundles Av eBurnu. p 1. 0 - 1600 1110 778.29 2. 1601 - 3200 759 2392.13 3. 3201 - 4800 799 3981 3.2 4 . 4801 - 6400 851 5627.10 5. 6401 - 8000 777 7127.38 6. 8001 - Onwards 264 8627.42 All bundles range lyingroua th f assignee n ego i p ar weightee dth d average burnup groupe oth f . Core average power densit bees yha n use fluencr dfo e calculation. resulte th nuclidee n I sth s have been ordee listeth f ascendinn o rdi g mass number. The isobars (nuclides having the same mass numbers) are listed in the order of ascending charge numbers. The radioactivity (in curies) is given for each of these isotopes. The total thermal power and the gamma power is tabulated for each of the fission product elements. The aggregate (total core) valu alss ei o given. 97 The results are provided for the following time steps after the reactor shut down: 5 seconds 10 " 102 " 103 " 104 " 10s " 106 " 107 " 108 " 109 " 1010 " 3.7.1.2. CaseB above th Al f eo l dat generatee aar d cas e agaith er whernfo assumes i t ei d that after reference th e burnu bees pha n achieved reactoe th , r remain shutdowa n si nday0 stat3 r s efo and is then operated again at full power for 30 days with all bundles exposed to the same(average) power density. 3.7.2. Description of the computer code used The computer code ORIGEN has been used for this benchmark evaluation by all three participants, i.e., Argentina, Pakistan and Romania. The ORIGEN code develope Ridgk Oa ey db Nationa l Laboratory ,capabls i U.S.A e of computing isotopic compositio radioactivitd nan fuef yo l materials, fission productsd an , cladding materials in both fixed and fluid fuel reactors. The code uses the matrix exponential method to solve the equations of radioactive growth and decay for a large number of matrie th f o x isotopes e exponentiaus e Th . l method permit treatmene sth complef o t x decay transmutatiod an n schemes extensivn A . e librar f nucleayo r dat availabls ai e wit codee hth . 3.7.3. Result discussiod san n resulte Th s have been taken fro progrese mth s reports submitte e IAEe th th y Ao dt b various participants, References [68-70]. There are some differences in input data used by the participants, which are described as under: The valuespectrae th f so l indices, THERMFASd an TS (requireRE , ORIGENy db ) use Argentiny db Pakistad aan differene givee nar ar Tabln d i an t e 3.7.2. Romanit no s aha mentione valuee dth f thesso e spectral indices. Argentina has used a constant specific power of 0.62 MW/bundle, while Pakistan and Romania have used total fission power of 2180 MW for 4560 fuel bundle i.e. a constant specific powe 0.4780f o r 7 MW/bundle. Argentin reactoR LW uses re aha cros dth s section librar f ORIGEyo whicNn i e hth cross section isotopee th f so s U235, U238, Pu239, Pu240, Pu241, Pu24 Np23d 1an e 9ar modified. Pakistan and Romania have used the original LWR reactor cross section library of ORIGEN. 98 TABLE 3.7.2 SPECTRAL INDICES USE BENCHMARDN I K CALCULATIONS Country Code Used Burnup Spectral Indices MWD/T THERM RES FAST ARGENTINA WIMS 4000 0.675453 0,.0843526 0,.5532061 * PAKISTAN WIMS-D/4 4000 0.815754 0,.027155 0,.062991 PAKISTAN WIMS-D/4 0 0.815754 0,.030012 0..070800 ROMANIA ...... ______ * Used in Benchmark calculations 3.7.3.1. Comparison of the results A summary of the results is presented for comparison in Table 3.7.3 through Table 3.7.8. For Case A, Argentina has presented the results for the case in which fuel is taken out of the reactor after irradiated upto a burnup of 7500 MWD/T, while Pakistan and Romania have used bundle burnup distributio benchmare s givena th n ni k (Table 3.7.1). Pakistan obtained the results for all six groups using weighted average burnup (MWD/T) of each group. The results of all the six groups are then added to get core aggregates. Romania has not mentioned the method of obtaining core aggregates. For Case B, all three participants have considered the case in which after they have achieved the burnup of Case A, the reactor has been shutdown for 30 days and is then operated again at full power for 30 days with all bundles exposed to the power density. cleas i t I r from Table 3.7.3 through Table 3.7.8 that ther variancee ear resulte th n si s TABLE 3.7.3 TOTAL CORE RADIOACTIVITY(CURIES F FISSIOO ) N PRODUCTS CAS( BUNDL: A E E BURNUP DISTRIBUTIO E BENCHMARKS GIVENA TH N NI ) Country Time after Shutdown (Seconds) OE+000 5 OE+001 OE+011 OE+O'OE+031 Z 1 OE+041 OE+0 OE+061 5 1 OE+07 fOE+11 OE+08 ARGENTINA 1 279E+10 1 185E+10 1 143E+10 9 209E+09 6 336E+C9 4 050E+09 2 439E+09 1 177E+09 2 967E+08 2 029E+07 6 933E+03 PAKISTA 559E+09 N 757E+08 9 373E+08 9 694E+06 9 530E+04 9 750E+02 9 407E+01 9 333E+05 9 399E+09 8 145E+04 7 814E+0l 5 3 ROMANIA 9 593E+09 8 800E+09 8 444E+09 6 820E+09 4 653E+09 2 920E+09 1 684E+09 8 443E+08 1 757E+08 l 120E+07 4 278E+03 VARIANCE % 1 3 7 -2 2 7 6 -2 2 1 5 -2 3 8 3 -7 7 2 5 (P-A)/ -2 8 5 A9 -7 3 3 8 -6 8 6 4 -5 1 3 2 -4 1 1 2 -3 -2 0 85 5 9 5 -2 0 1 6 -2 6 7 5 -2 0 0 5 (R-A)/-2 A -26 56 -27 91 -30 95 -28 25 , -40 79 -44 81 -38 29 3 8 5 13 1 (R-P)/2 0 17 P 3 9 6 08 36 2 3 8 05 49 9 6 9 01 85 8 1 16 88 2 7 2 A - Argentina P - Pakistan R - Romania 99 TABLE 3.7.4 TOTAL CORE THERMAL POWER (WATTS F FISSIOO ) N PRODUCTS ( CAS : BUNDL A E E BURNUP DISTRIBUTIO E BENCHMARKS GIVEA NTH N NI ) Country Time after Shutdown (Seconds) 0 OE+00 5 OE+00 1 OE+01 1 OE+02 1 OE+03 1 OE+04 1 OE-05 1 OE+06 1 OE+07 1 OE+8 1 OE+10 ARGENTINA 1 433E+Û8 1 271E+08 1 194E+0S 9 039E+07 5 185E+07 2 40SE+07 1 194E+07 5 338E+06 1 228E+06 7 811E+04 l 680E+01 PAKISTAN 1 136E+08 9 931E+07 9 320E+07 6 957E+07 3 902E+07 1 726E+07 7 190E+05 2 574E+06 3 745E+05 1 369E+04 4 136E+00 ROMANI lOlE-f-01 A 9 678E+08 099E+09 7 878E+05 7 908E+03 7 775E+D1 7 383E+08 7 534E+03 6 7 166E-06 993E<-03 5 OÛ6E+01 4 1 VARIANCE Ï CP-A5/A -20 70 -21 84 -21 93 -23 03 -24 74 -28 22 -39 77 -51 78 -59 50 -82 47 -75 38 (R-A)/A -23 15 -23 83 -23 78 -23 91 -24 62 -26 19 -29 78 -33 80 '-41 64 -48 88 -40 11 7 3 2 - 5 5 2 - 8 0 3 (R-PV- P -1 14 0 15 2 84 16 59 37 30 91 35 191 67 143 23 A - Argentina Pakista- P n R - Romania TABLE 3.7.5 TOTAL CORE GAMMA POWER (WATTS F FISSIOO ) N PRODUCTS ( CAS : BUNDL A E E BURNUP DISTRIBUTIO E BENCHMARKS GIVENA TH N NI ) Country Time after Shutdown (Seconds) 0 OE+00 5 OE+00 1 OE+01 1 OE+02 1 OE+03 1 OE+04 1 OE+05 1 OE+06 1 OE+07 1 OE+8 1 OE+10 ARGENTIN 5 057E+0A 4 912E+07 4 818E+07 4 170E+07 2 771E+07 313E+01 7 6 839E+07 2 993E+05 5 628E+06 2 089E+05 5 551E+04 0 PAKISTA 975E+03 N 849E+03 7 765E+03 7 188E+03 7 041E+02 7 210E+09 7 4 131E+06 459E+01 6 808E+01 6 476E+02 5 : 038E+03 0 ROMANIA 3 870E+07 3 759E+07 3 687E+07 3 167E+07 2 064E+07 9 6Q2E+06 4 800E+06 1 981E+06 3 291E+05 9 218E+03 2 879E+OC VARIANCE % (P-A)/A -21 0 33 9 1 -8 -2 15 1 6 48 -8 -2 17 8 8 57 -6 -2 35 2 5 51 -5 -2 69 5 3 59 -3 -27 98 (R-A)/A -23 47 -23 47 -23 47 -24 05 -25 52 -26 88 -29 81 -33 81 , -41 52 -55 88 -48 14 (R-P)/P -2 64 -2 34 -2 07 -0 66 1 13 4 26 16 19 35 78 82 02 272 29 177 36 A - Argentina Pakista• P n R - Romania 100 TABLE 3.7.6 TOTAL CORE RADIOACTIVITY(CURIES) OF FISSION PRODUCTS ( 0 DAYCAS3 : B ES SHUTDOWN AFTER BENCHMARK BURNUP FOLLOWEY DB 30 DAYS AT FULL POWER) Country Time after Shutdown (Seconds) 0 OE+0 5 OE+0 1 OE+0 1 OE+0 1 OE+0 1 OE+0 0OE+01 1 OE+00 OE+01 2 3 45 67 l.OE+ I OE+180 ARGENTINA l 194E+10 PAKISTAN 9 622E+09 8 827E+09 8 451E+09 6 775E+09 4 601E+09 2 828E+09 1 491E+09 5 931E+08 l 166E+08 6 207E+06 2 817E+03 ROMANIA 9 455E+09 8 783E+09 8 442E+09 6 805E+09 4 656E+09 2 942E+09 1 740E*09 8 237E+08 2 405E-08 2 319E*C7 8 783E+D3 VARIANCE Ï 0 4 (P-AV9 -1 A - -- , ------2 7 CR-A)/0 -2 A 9 7 1 21 1 1 6 1 3 0 - 27 6 0 2 5 6 0 - 10 8 6 3 5 8 (R-P)/1 3 - P 0 7 6 1 3 0 4 0 2 1 4 4 0 Argentin- A a Pakista- P n Romani• R a TABLE 3.7.7 TOTAL CORE THERMAL POWER(WATTS F FISSIOO ) N PRODUCTS ( CASE B : 30 DAYS SHUTDOWN AFTER BENCHMARK BURNUP FOLLOWED BY 30 DAYS AT FULL POWER) Country Time after Shutdown (Seconds) 0 OE+0 OE+05 OE+011 OE+01 OE+03 1 OE+01 ÛE+01 0 0OE+01 OE+01 2 4 5 67 l.OE+OE+11 08 ARGENTIN l 381E+0A 8 PAKISTA 1 126E+0N 9 872E+08 7 9.268E+0 6 942E+07 3 912E+07 746E+01 7 7 522E+07 2 809E+06 4 693E+06 2 054E+05 6 395E+04 0 ROMANIA 1 066E+08 9 380E+07 8.868E+07 6 72SE+07 3 880E+07 1 748E+07 8 527E+05 3 760E+06 1 015E+06 8 527E+04 2 098E+01 VARIANC% E (P-A)/A -18 49 (R-AJ/A -22 83 . ------. CR-PJ/P -5 33 -4 98 -4 32 -3 08 -0 82 0 11 13 36 33 86 116 28 315 14 228 07 - AArgentin a - PPakista n R - Romania 101 TABLE 3.7.8 TOTAL CORE GAMMA POWER (WATTS) OF FISSION PRODUCTS ( CASE B : 30 DAYS SHUTDOWN AFTER BENCHMARK BURNUP FOLLOWED BY 0 DAY3 FULT A S L POWER) Country Time after Shutdown (Seconds) 0 OE+0 OE+05 OE+0 OE+01 1 OE+01 OE+0 1 0OE+0 1 0 OE+0OE+11 1 OE+02 1 OE+ 1 3 4 5 60 S7 ARGENTIN 852E+04 A 7 PAKISTAN 3 949E+07 3 824E+07 3 742E+07 3 184E+07 2 054E+07 9 355E+06 4 320E+06 1 586E+06 2.208E+05 3 819E+03 l 604E+OC ROMANIA 3 735E+07 3 633E+07 3 564E+07 3 087E+07 2 038E+07 9 4SSE+06 4 826E+06 2 038E+06 4 076E+05 2 277E+04 6 933E+00 VARIANCE % 1 6 (P-AJ/8 -1 A (R-A3/A -23 02 ------' ------, --- 3 2 2 33 3 2 6 49 0 5 4 8 0 5 8 2 1 7 1 1 8 1 1 8 7 0 - 5 0 3 - 6 7 4 - 9 9 4 - 2 4 (R-PJ/5 - P ArgentinA- e 0 - Pakistan Romani- R a thest attributee bu ear face th t o dtha t participant e tth s have used different burnup valued san adopted different approaches to get core aggregates. The difference hi results could also be caused by the difference in using specific power (MW per unit weight of fuel) and the spectral indices, THERM, RES and FAST and by the difference in cross sections. To eliminate the effect of these various factors, a comparison for both the cases A and B has also been made in Table 3.7.9 through Table 3.7.12, between Pakistani and Argentinean results, for different time steps after a shutdown at 7500 MWD/T. In this comparison the values of spectral indices and specific power are kept the same as those used by Argentina. It can be seen from these tables that Argentinean and Pakistani results are in very close agreement with each other smale Th . l variance attributee b y difference th sma o dt e in neutronic data library. TABLE 3.7.9 RADIOACTIVIT F FISSIOO Y N PRODUCT CURIES/TON I S N ( CASE A : BURNUP = 7500 MWD/T ) Country Time after Shutdown (Seconds) 0.OE+00 5 OE+00 1 OE+01 1 OE+02 1 OE+03 1 OE+04 1 OE+05 1 OE+06 1 OE+07 1 OE+8 1 OE+10 ARGENTIN 500E+01 A 390E+Û1 8 340E+01 8 080E+01 8 430E+07 8 750E+04 7 860E+02 7 380E+01 7 480E+03 7 380E+02 6 130E+08 5 1 PAKISTA 510E+01 N 390E+01 8 340E+01 8 090E+01 8 540E+07 8 850E+04 7 840E+02 7 370E+01 7 490E+03 7 470E+02 6 200E+0l 5 2 0 6 7 4 8 7 3 9 2 0 2 7 0 - 0 7 0 - 1 1 2 8 4 1 3 9 0 0 0 0 0 0 0 7 6 0 VARIANC % E 102 TABLE 3.7.10 THERMAL POWER OF FISSION PRODUCTS IN WATTS/TON Country Time after Shutdown (Seconds) 0 OE+00 5 OE+00 1 OE-01 1 OE+02 1 OE+03 1 OE+04 1 OE+05 1 OE+06 1 OE+07 1 OE+8 1 OE+10 ARGENTIN 68E+01 A 49E+01 640E+01 605E+01 608E+06 682E+02 540E+01 526E+06 544E+01 416E-G9 497E-0l 2 1 PAKISTA 69E+01 N49E+01 640E+01 606E+01 610E+06 683E+02 537E+01 516E+06 542E+01 446E+08 4 63E-0 2 2 1 0 5 3 3 4 6 7 - 9 3 1 - 0 6 1 - 4 1 2 - 5 3 0 3 3 0 0 0 0 0 0 0 0 0 0 0 5 VARIANC0 % E TABLE 3.7.11 GAMMA POWER OF FISSION PRODUCTS IN WATTS/TON Country Time after Shutdown (Seconds) 0 OE+0 OE+05 OE+01 OE+01 OE+01 OE+001 OE+01 0 OE+0OE+11 1 OE+021 OE+ 1 3 4 5 60 S7 ARGEffTIN 93E+05 A 76E+05 565E+05 589E+04 525E+03 554E+01 502E+08 551E+03 4 50E+0 6 445E+02 351E-06 2 2 PAKISTAN 5 95E+05 5 77E+05 5 66E+05 4 88E+05 3 23E-05 1 53E+05 7 84E+04 3 42E+04 6 36E+03 1 70E+02 6 50E-02 VARIANCE % 0 34 0 17 0 18 -0 20 -0 62 0 65 -2 24 -2 56 -3 64 30 61 -0 15 TABLE 3.7.12 RADIOACTIVITY AND DECAY POWER OF FISSION PRODUCTS AT THE TIME OF DISCHARGE ( CASE B : 30 DAYS SHUTDOWN AFTER BENCHMARK BURNUP FOLLOWED BY 30 DAYS AT FULL POWER) Country Radioactivity Thermal Power Gamma Power (Curies/Ton) (Watts/Ton) (Watts/Ton) ARGENTINA 1.40E+08 1.62E+06 5.69E+05 PAKISTAN 1.44E+08 1.62E+06 5.69E+05 VARIANCE % 2.86 0.00 0.00 3.8. TAS : KSECONDAR8 Y SHUT DOWN SYSTEM 3.8,1. Description of the benchmark The secondary shutdown system (SDS2) consists of liquid poison injection (LPI) into the moderator through nozzles along the horizontal tubes hi modern PHWRs. The actual propagatiot je shape variatioe th f th e o d n an poisof no n concentration profil complexs ei r Fo . problem specification some idealized assumptions are made as described below. The problem was to calculate the reactivities and snapshot flux distributions for four different poison distributions in the core at four different times during poison propagation as specified in [43, 52]. 103 Initial core conditions: 600 MW(e) benchmark core (cf: Task 3) Fresh fuel with saturated fission products N I s AR l Al T OU s AlSR l l ZCUAl t nominaa s l fills Poison geometry: Six poison tubes at locations ±5 lattice pitches from core mid-plane along y-direction fro) an mcm t (±15 da 0 cor , e0cm mid-plane along z-direction s showa , Fign ni . 3.8.1. Poison tubm ec radiu0 3. = s Poison continuou a tub s e ha hole f o t ssse betwee lattico ntw e tubes, that make loot si k like almost a long slit. Four such slits are oriented at 0°, 90°, 180° and 270° around the poison tube. (Fig. 3.8.2). The poiso t divergencnje ° (Fig15 s .ei 3.8.2) . The cross section of the poison would then be wedge shaped. Longitudinal view of the poison propagation is shown in Fig. 3.8.3 at four different tunes. The initial gadolinium concentration is C0 = 8000 ppm. variatioe Th concentration ni describes ni belows da : Lattice Pitch Gd Concentration 0-2 C0 2-4 0.25 C0 4-6 0.10 C0 Y ~~)7 +5 P 1150 cm FIG. 3.8.1 Locations of Liquid Poison Injection Nozzles 104 FIG. 3.8.2 Cross Sectional Poison Viewthe of Jet t=t 1—— 1—— 1—— —— oc0 0 0 'L c=o. 25( o 0 o o t o o 0 o 0 o o o 0 o o o o o 0 o o 0 o o ] ooooooooo O0.10C 'C ooooooooooo ooooooooooooo 0=0.250•o oooooooooooooo C=Cg oooooooooooooo oooooooooooooo 'I FIG. 3.8.3 Longitudinal View of Poison Sheath for Four Snapshots 105 This benchmark problem involved the modelling of these poison jets, calculation of negative reactivity insertio functioa s powee na th f timd o nr an edistributio n (artificially normalized to full power) for each snapshot geometry. The required results were, a. Supercell calculations corresponding to cells of various gadolinium concentrations, b. Static core calculations using the above supercell results to determine reactivity and power distribution snapshoe r eacth sfo f ho t poiso t geometriesnje . 3.8.2. Moderation and calculation Thi Indiao s tascarrietw s ne k wa t onlth participants dy ou y b resulte Th . s presented in the CRP meeting by these two were essentially for two different poison profiles [71,72]. s suggeste wa participante t th I f do thae on ts should recalculate b probleo e t eth s a o ms consistent with the other calculation. India-RED has since recalculated it according to the way the problem had been defined by India-ThPD in their submission. 3.8.2.1. India-RED thin I s formalismbees ha n t treateje e th ,conicals da flue xTh . distribution ovee th r poison and the corresponding lattice was calculated using a 2-D transport theory model [73]. flue xTh distributio associates it d nan d lattic calculates ewa plana n di e perpendiculae th o rt axis of the jet. The poison was assumed to be confined to a cross sectional area formed by the angle of the cone. A series of such calculations was done along the axis of the jet over the extent of the poison taking into account the variation of the concentration of the poison (See Fig. 3.8.5). FIG. 3.8.4 Propagation of Two Half Jets along O° and 9CP Angles (India -ThPD) 106 cnn POISONTUBE FIG. 3.8.5 Modelling of Poison Jet by India - RED 107 contributioe Th absorptioe th o nt n cross sectio lattice th f thes neo i n calculatea s da functioe distancth f o n e fro e poisomth n tub y integratinb e g ove cele th rl wit e fluhth x distribution over the poison-lattice domain weighing. The extra absorption was then determine functioa distance s da th f no e fro poisoe mth n tube poisonee Th . d lattices were then identified with respec theio t t r positio e poisoth f no n coree tubee extrth Th n .ai s absorption for these cells was calculated depending on the amount of poison in that particular s distancit celd an l e fro poisoe mth n tube. The whole core calculations were then done using diffusion theory code ANAMIKA where this extra absorptions of the different perturbed cells were used explicitly. The reactivity e poisowortth f o hn jets correspondin e fou th o eact g r f o hsnapsho t poison distribution was calculated as 1/K1 - 1/K2 where Kl is the K-effective without the poison K-effective th s i 2 K ejetd wit san poisoe hth n jets. 5.5.2.2. India-ThPD Supercell Modelling of the Poison Jets Due to inherent complexity of the problem the calculation is broken into two parts viz, conceivin manf go ysupercell D type3- f so usind informatioe an s gth n from supercell analyse subsequenn i s t core simulationpoisoe th r npe geometrs A . y description poisoe th , n jet propagates to a maximum distance of six lattice pitches in both y and z directions. Eleven supercell type conceivee sar designated dan (thV eH s originatinda g cell with double poison regions), VI to V5 (vertical cells) and HI to H5 (horizontal cells). This is schematically indicate Fign di halo . 3.8.ftw jetr 90°4fo sd alonan , , anglesassumes g0° wa t I . d thae th t poison slab engulf fuee th sl channel whenever they cros e fueth sl channel boundarien i s respective supercells considered. e supercelTh l proble s beemha n analyze y Montb d e Carlo codes KENd Oan MONALI collisioD 3- , n probability code BOXE PHANTOd Ran M code system basen do diffusion iterative technique in two energy groups. The 3-D cell consists mainly of four regions, the fuel cluster, the pressure tube + air ga coolanp+ t tube, D20 moderato witO 2 hgrouD o poisond tw ran pe crosTh . s sectionr sfo the first three types of regions were obtained by the code CLUB [9] by treating the normal fuel cluster without poison in 27 energy groups using WIMS library. To obtain the few group cross sectio poisoe th f no n slaregioD b1- superceln na dons i l e wit poisoe hth n sla givef bo n concentratio thicknesd n an centr e th t sa e surrounde fuea . y ldb cm past 0 e±5 regio o t p nu This calculation is done by PIJMURLI in 27 groups. The wedge shape of the poison is simulated as such in MONALI code. s approximatei t I equivalenn a y db t slab regio othen ni r models. Diffusion method required proper judgement on treating the poison region. Internal boundary condition coul appliee db d wheneve poisoe rth n regio fuee th lt n doecu t sno cluster. Additional CLUB/PIJMURLI calculations were needed to obtain the cross section of fuel cluster engulfed by poison. Core Simulations o corTw e simulation models wer e centre-meseth employeds wa e h On .finit e difference diffusion code CEMESH which consider constansa t mesh siz PxPxL/f eo e th n 2i 108 core region. The other was DIMENTRI which uses-a constant radial mesh spacing PxP and variable mesh spacing in the axial direction as described for Task-3. By processing the information on the fuel zone description and the locations of reactivity devices (RDs) the codes determine the meshes influenced by any RD and ascribe appropriate deltaE/E to them. The CEMESH simulations employed the deltaS/E 's of all the four models of supercells. The DIMENTRI simulations considered the equivalent deltaE/E for thermal absorption cross section which conserves the reactivity for the supercell with the particular choice of poison curtain, BOXEe baseth n do R model. 3.8.3. Results resulte Th f supercelso l simulation India-ThPy sb givee Dar Tabln i e 3.8.1. The Table 3.8.2 give e e corresultth th s ef o sanalysis . Initially small negative reactivities of 2.7 mk and 6.7 ink are inserted at times t2 and t3, but large negative reactivity of nearly 30 mk is inserted at the final snapshot of t4. The results the CEMESH analysis were almos foue t identicath rl typeal r f supercelfo so l l modellin f SDS2go resulte Th . f so DIMENTRI are also quite close to those by CEMESH simulations. The results of the initial calculation by India-RED [71] are tabulated in Table 3.8.3. In this case the problem analyzed was a different one where it was assumed that there was onl single yon betweet eje pitcheso ntw . Althoug problee hth interpretes mwa d differently yb the two participants , the reactivity worths at time t4 was different only by 25 %. The problem was recalculated by India-RED and the results were in good agreement with results of India-ThPD as seen from Table 3.8.2. This shows that the poison jets are so black to neutrons that numbee eveth f ni f holeo r increaseds si increase th , negativn ei e reactivits yi t altogetheno r proportionat increase th numbee o th et n ei holesf o r . TABLE 3.8.1 COMPARISON OF K-INFINITY VALUES FOR VARIOUS SUPERCELLS ( INDIA -ThPD ) Cell Poison Slainfinit- K b y values from codes Type Thickness KENO BOXER PHANTOM MONALI (cm) ± 0 Normal 1.0833 1.081 1.0824 1.0802 + 0.0011 cluster HV (1.88,1.62) 0.4196 0.396 0.4043 0.3993 + 0.0013 VI ( 5.61) 0.4516 0.448 0.4294 0.4464 ± 0.0014 V2 ( 9.35) 0.3771 0.376 0.3712 0.3737 ± 0.0011 V3 (13.09) 0.2886 0.285 0.2824 0.2842 ± 0.0011 V4 (16.83) 0.2477 0.248 0.2497 0.2487 ± 0.0009 V5 (20.57) 0.1975 0.198 0.1984 0.2005 ± 0.0006 HI (4.860) 0.5251 0.510 0.5415 0.5200 ± 0.0012 H2 (8.099) 0.4597 0.445 0.4601 0.4536 ± 0.0014 H3 (11.34) 0.4016 0.369 0.4019 0.3829 ± 0.0012 H4 (14.58) 0.3701 0.352 0.3795 0.3600 ± 0.0010 H5 (17.82) 0.2480 0.297 0.3192 0.2978 ± 0.0009 In Monte Carlo simulation by KENO code, the standard e ordedeviatio0.001± th f f o ro n eigenvalue i 5s nwa . 109 TABLE 3.8.2 RESULTS OF THE CORE SIMULATION FOR FOUR SNAPSHOTS OF SDS2 TRANSIENT Time K-eff p l de MBP* loc. . c MCPlo * Step (mk) (KWt) (MWt) 1.03124 0.0 (1) tl 1.03585 0.0 783 .4 E-12, 7 7.200 K-13 (2) 1.03500 0.0 800 .9 E-12, 7 7.331 M-09 (3) 1.02824 2.99 t2 1.03291 2.75 998 .4 0-17,6 8.971 M-16 1.03223 2.58 985 .1 0-17,7 8.733 L-15 1.02411 6.75 t3 1.02874 6.67 1425 .7 0-18,7 12.285 N-17 1.02754 7.00 1445 .9 0-18,7 12.264 N-17 1.00011 30 .18 t4 1.00459 30 .04 3898 .0 M-20, 7 31.196 M-19 1.00292 30 .88 3980 .1 M-19, 7 31.073 M-19 (1) - India - RED (2) - India - ThPD - CEMESH - Indi ) ThP- a(3 DIMENTR- D I * The MBP and MCP values are obtained by artificially normalising the reactor power to 2180 MWth at each instant. TABLE 3.8.3 CORE SIMULATION RESULTS WIT BETWEET H JE ONL E YON N PITCHES (INDIA-RED) time k-eff Deltp a mk tl 1.07883 2 t 1.07846 0.318 t3 1.07453 3.709 t4 1.05119 24.373 110 3.8.4. Conclusions and recommendations The complex problem of modelling the SDS2 has been tackled by methods with varying degree of sophistication at the supercell level. The diffusion iterative method is found to give results comparable to transport methods. The differences between the calculational models seen at supercell level is not observable from core results. The representation of a series of jets as poison curtains emerging from a long slit is an idealized model specially chose r thinfo s benchmark realitn I . poisoe yth n emerges a s cones from closely spaced holes betwee fueo ntw l channels e individuaTh . l jets quickly coalesce to form a finite wedge. This wedge shape is rectangularised for supercell calculations which is felt adequate. However if individual conical jets are far apart or in different directions mora , e involved treatmen necessarye b y stema te pTh . functioe th r nfo concentratio poisoe th diffuset i f nphysicas e o n a th n si int e moderatoe lb osenst th no ey rma a true picture of what happens hi reality. propea f I r safety analysidonee b o t , s thessi e factor consideredhave y b o sema t e Th . intent of this task was to see how best one could model the poison jets as they diffuse throug moderatoe hth rfels sinct wa tha conventionat e i tth l practice modellinr sfo solie gth d reactivit adequate b y t device no r thi y efo s mora cases ma r eFo . detailed safety analysis, some experimental result requirede b s alsy oma . 3.9. TAS : EXPERIMENTAK9 L VALIDATION One of the task (Task 9) was regarding the validation of codes by analyzing four set of experiments given below: 1) First experimen analyzee b o t regardin s dwa lattice gth e measurements done wit- h37 rod fuel cluster at ZED-2 reactor hi Canada [74]. The experiment was performed at only one lattice pitch of 28.58 cm (square) using heavy water and air as coolants (Task 9.1). 2) The second set of experiments to be analyzed was regarding the lattice measurements done with 28-rod fuel cluster at ZED-2 reactor in Canada [75-76]. The experiments were performed at number of lattice pitches using heavy water and air as coolants. (Task 9.2) 3) The third experiment to be analyzed was regarding the isotopic composition of fuel t varioua s burnups [77] (Task 9.3). 4) The fourth experiment to be analyzed was the measurements done with stainless steel adjusters hi the form of rods and tubes of various sizes [78] (Task 9.4). This section summarize resulte sth s submitte variouy db s participant above th r esfo mentioned tasks. 3.9.1. Task 9.1: Analysis of experiments with 37-rod fuel clusters in ZED-2 reactor The summary of the results are submitted by three countries, namely, Argentina [79], India-ThPD [80 Romanid an ] e computeTh a [81]. r code se analysi th use n i df thi o s s experiment are CLUB (India-ThPD) and WIMS (Argentina, Romania). Both the codes used 111 the WIMS cross section library. However, différera version f WIMso S library might have been used. This experimen alss owa t analyze Canady db a using WIMS cod theied [74an r] results have also been included. 3.9.1.1. Description of the benchmark Description of experiment In Canada, an experiment was performed to determine cell parameters in lattices of 37-rod fuel cluste t ZED-a r 2 reactor. Measurements were mad singla t ea e pitch, namely, 28.5 squarem 8c , using heav coolantss a yr wateai d .an r The 37-rod fuel assembly contained eighteen, twelve, six and one fuel rods on circles of respectivelydiameterm c 0 d , 2.97 an , s5.75 cm bundl8.66A m .cm 71 c forme s 1 ewa d by welding thes zircaloy-o rod7 tw e3 o st 4 plates. Each fuecontained ro l dstaca naturaf ko l uranium oxide pellet diameters m (12.m 1 , density 10.5 g/cc) withi zircaloy-na 4 sheatf ho wall thickness 0.4 mm and I.D. 12.2 mm. Fuel stack length was about 48 cm and overall bundle length 49.5 cm. The pressure and calandria tubes were of 1050 aluminium alloy. The I.D. and O.D. of pressure tuber were 10.39 cm and 11.02 cm respectively and corresponding dimension f calandriso a tube respectivelywer13.3m d c an 4e m 12.7c 0 . The fuel assembly loaded at the centre of the reactor called demountable fuel assembly was slightly different from the above mentioned fuel assembly. All detailed activation measurements were made within or about this assembly. Each fuel sheath was nominall same th f eyo material , diamete wald ran l thicknes thas sa tabove useth n dei fuel assembly. Thirtthirte th f yyo seven fuel rods contain UO2 pelletsame th ef s o materia d an l dimension s abovea s . These thirty elements were fitted with welde d capen ds which incorporated a screw threaded projection. The remaining seven rods could be completely disassembled to enable foils to be positioned between fuel pellets. Although the fuel used in these seven rods was UO2, all pellets were flat ended and of reduced diameter (11.73 mm). -These reduced diameter pellets could be packaged, together with foils, within very thin walled (0.13 mm) 2S aluminium cans of I.D. 11.8 mm and length about 10 cm. The resulting package obviated foi peller o l t jamming problems allowed an , d accurate alignment foile th sf o with adjacent fuel pellets. Parameters to be Compared Following parameters measured with 37-rod fuel clusters have to be calculated: fast fission ratio throughout a fuel cluster initial conversion ratio throughou fuea t l cluster U-235 fission rate throughou fuea t l cluster detailed relative copper activity distribution throughout a lattice cell lutetium-manganese activity ratios relative to a thermal reference location. 3.9.1.2. Computer codes used The experiment with 37-rod cluster has been analyzed by India-ThPD using the computer code CLUB, Argentin Romanid aan a usin WIMe gth S code. 112 3.9.1.3. Main results In order to study the effect of various options on different parameters, Argentina analyze experimene dth usiny tb g three option solutioe transporth e r th sfo f no t equatione Th . three options are 1-D collision probability, PERSIUS, 2-D collision probability, PIJ, upto the calandria tube, callecollisioD 2- dd partianan probabilitJ PI l whole th r eyfo cell, called full PU. We have included the results of first (PERSIUS Model) and third option (full PIJ Method) in this report. The report [74] also gives the results of calculations performed by AECL using WIM J optionsSPI cod havd e an eW . e wit N includehDS N resulte dth DS f so option in this report. The results of WIMS codes are labeled by CNEA (Argentina), INPR (Romania) and AECL (Canada) results in the following tables. These tables give measured values followed by deviations (%) from the measured values obtained as follows: D = 100 * (calculated/measured - 1) Fast Fission Ratio The PHWRs use natural uranium as fuel. In this a few percent of all fissions take place in U-238. The calculable and measurable parameter related to these fast fissions in U-23fase th t s 8fissioi n ratio (8), poinwhicy defines i an t x th a s da ô[U-23= 8 fissions/U-235 fissions]x Table 3.9.1.1 give experimentae sth deviatione th d l valuean ô s f calculateso d with different codes. It can be seen from the table that all the codes underestimate the ô for D2O TABLE 3.9.1.1 FAST FISSION RATIO (S) FOR 37-ROD FUEL CLUSTER ô Description Ring A Rin gB RingC Ring D Bundle Average D20 Coolant Expereimental 0.0768 0.0719 0.0592 0.0411 0.0511 Deviation(%) CLUB - 1.8 - 1.5 0.2 - 0.7 - 0.8 Déviât ion (%) INPR -10 .46 -10 .1 8.36 -11 .0 - 9.8 Deviation (%) CNEA-PER - 9.5 - 3.7 1.8 0.1 - 0.4 Deviation (%) CNEA-PIJ - 4.0 - 3.7 2.0 - 1.3 - 2.1 Deviation (%) AECL 7.7 4.5 1.7 - 3.2 - 0.1 Air Coolant Experimental 0.0810 0.0770 0.0664 0.0481 0.0583 Deviation(%) CLUB - 2 .6 - 1.4 2.0 - 2.5 - 2.2 Deviation (%) CNEA-PER -11 .0 - 4.4 0.4 - 1.6 - 2.0 Deviatio) n{% CNEA-PIJ - 6 .4 - 4 7 - 4.8 - 4.8 - 4.8 Deviatio) n(% AECL 8.9 6.0 1.7 - 3.0 0.2 refeD d o centralt ran C , RingB ,, innerA s outermosd an , t rinf go the fuel cluster. INP Institut= R r Nucleafo e r Power Reactors, Pitesti, Romania CNE GerenciA= Aree D a a Centrales Nucleares, Argentina AECL = Atomic Energy of Canada Limited, Canada (WIMS-CRNL computations, AECL-5307) ô = [U-238 fissions/U-235 fissions] 113 coolant. In INPR results, there is relatively large error of approximately -10 %. The results of WIMS are sensitive to the particular option (PERSIUS/PIJ/DSN) used hi the code. There is mor ee spatia erroth n i rl distributio f fasno t fission ratio withi e fuenth l cluster with PERSIUS model. The results of WIMS given by different institutes differ substantially which ma differeno t partle yb e ydu t versioWIMe th f no S librar e optioyth used n dan use r dfo solving neutron transport equatio WIMe th n i S code CLUe Th . B underestimates fast fission ratio by 0.8 % compared to 0.1 % by AECL results. However, when one considers the spatial distribution of ô within the fuel cluster, the error in the results of CLUB is relatively r coolantai lessr CLUe Fo .th , B underestimateAECLy b compare% % -2 2 0. y b o dt t i s WIMS. Relative Conversion Ratio The long term reactivity behaviour of a PHWR is very significantly affected by the productio fissiof no n Pu-239 formed from neutron captur U-238n ei therefors i t I . e important to be able to calculate reliably some parameters related to U-238 captures. A suitable parameter is the relative conversion ratio, C which is defined as [U-238 captures/U-235 fissions]x [U-238 captures/U-235 fissions]p where x and p refer to a fuel location and a thermal reference location respectively. Cp is calculated by assuming the flux distribution to be pure Maxwellian at the physical temperature of the experimental pit, that is, (NO238 235 (N The N and crm are U-235/U-238 number density and Maxwellian averaged fission/capture 5 23 cross sections respectively e calculateTh . y Canadiandb valu/<7 f 238 ,a o eusiny b s g f c m WIM grou9 S6 p librar 4.7625xlQ-s [75]K ywa ° !5 29 t a 3 m Table 3.9.1.2 lists the experimental values of C and their deviation. There is a good agreement between calculated and experimental values and bundle averaged value of C are codese th Dr l 2OFo al . y coolantwithib % n1 , ther slighs ei t overestimatio CLUy b % B 3 n0. compare underestimatioo dt INPy b AECd R% an 7 L0. resultsf no largese Th . t discrepancy is only -2.5 % for the fuel rods of third ring for CNEA results. Relative U-235 Fission Rate Distribution Table 3.9.1.3 gives the experimental values of normalized U-235 fission rate (i.e., average U-235 fissio nfuee 1.0s ratth deviationi d l n e)i an valuee th f so s calculatee th y db CLU WIMd Ban S codes agreemene Th . t between CLU experimend Ban bees ha t n founo dt excellene b calculatee th d an t d valuef experimentao founwithie e % li sar o 5 dt n 0. l values. In the CNEA and AECL results, there is slightly more error for central rod for both the coolants. 114 TABLE 3.9.1.2 RELATIVE CONVERSION RATI 37-ROR OFO D FUEL CLUSTER Relative Conversion Ratio Description Ring A Ring B Ring C Ring D Bundle Average O 2 D Coolant Experimental 1.4936 1.4681 1.4254 1.3925 1.4138 Deviatio) n(% CLUB 0.0 0.0 0.5 0.3 0.3 Deviatio) n(% INPR - 0.18 - 0.15 - 1.12 - 0.70 - 0.72 Deviatio) n(% CNEA-PER - 0.50 0.4 - 0.4 - 0.1 - 0.1 Déviât ion (%) CNEA-PIJ 0.20 0.2 - 0.6 0.1 - 0.4 Deviatio) n{% AECL 0.20 0.10 - 1.13 - 0.50 - 0.70 Air Coolant Experimental 1.3909 1.3590 1.3620 1.3880 1.3765 Déviât ion (%) CLUB - 1.1 0.2 - 1.5 - 0.9 - 0.9 Deviation (%) CNEA-PER - 2.3 - 0.1 - 2.1 0.0 - 0.6 Deviatio) n(% CNEA-PIJ - 2.0 - 0.2 - 2.5 - 0.3 - 1.0 Deviation (%) AECL 0.0 1.0 - 2.0 - 0.6 - 0.8 refeD d centralo ran t C Ring, B ,, sA inner outermosd ,an t rinf go the fuel cluster. INP Institut= R Nuclear fo e r Power Reactors, Pitesti, Romania CNEA = Gerencia De Area Centrales Nucleares AECL = Atomic Energy of Canada Limited, Canada (WIMS-CRNL computations, AECL-5307) [U-238 captures/U-235 fisssions]x ______= C [U-238 captures/U-235 fissions]p where x and p refer to a fuel location and a thermal reference location respectively. TABLE 3.9.1.3 U-235 FISSION RATES U-235 Fission Rates Description Rin gA Rin gB Rin C g Rin gD Bundl e Average D2O Coolant Experimental 0.746 0.787 0.907 1.147 1.000 Deviation (%) CLUB 0.7 0.5 0.0 - 0.2 0.0 Deviation (%) CNEA-PER - 1.7 1.4 0.6 0.1 0.0 Deviatio) n(% CNEA-PIJ _ 2 3 2.5 1.0 - 1.0 0.0 Deviatio) n(% AECL - 2.6 - 1.5 0.4 0.6 0.0 r CoolanAi t Experimental 0.814 . 0.840 2 927 1.112 1.000 Deviation (%) CLUB - 0.1 - 0.5 0.0 0.0 0.0 Deviatio) n(% CNEA-PER - 2.0 0.8 0.4 0.1 0.0 Deviation (%) CNEA-PIJ - 3.4 0.7 0.5 - 0.4 0.0 Déviât ion (%) AECL - 4.7 - 2.1 0.5 1.1 0.0 115 — Relative Copper Activity Distribution Table 3.9.1.4 gives the calculated and experimental values of normalized Cu-63 absorption rates at various locations. The calculated values by CLUB are in good agreement with the experiment and the maximum deviation is 0.5 % for the second ring. The maximum deviatio r INPnfo R result secone th s 1.5i sn i d 5 rin% tha d r AECgan fo t L results -4.8% r coolantai r PERSIUe centrae fo th Th .d n i ro l S optio f WIMno S give relativela s y large moderatoe erroth n i r r (5.. 7%) TABLE 3.9.1.4 NORMALISED COPPER ACTIVITIES Copper Activités Description Rin gA Rin gB Rin gC Rin gD Moderator Average D20 Coolant Experimental 0.757 0.794 0.913 1.140 2.096 Deviation (%) CLUB 0.4 0.5 - 0.3 0.0 0.2 Déviât ion (%} INPR 1.40 1.55 - 0.16 - 0.31 - Déviât ion (%) CNEA-PER 1.8 0.3 - 1.3 0.6 5.7 Déviât ion (%) CNEA-PIJ 1.5 1.4 0.2 - 0.5 0.6 Deviatio) n(% AECL - 2.7 - 1.4 - 0.6 0.7 4.0 Air Coolant Experimental 0.824 0.844 0.928 1.110 2.024 Déviât ion (%) CLUB - 0.5 0.0 0.2 - 0.2 - 0.2 Deviation (%) CNEA-PER 2.3 0.6 - 0.6 0.1 - 0.5 Deviatio) n(% CNEA-PIJ 0.5 0.3 0.3 - 0.4 0.6 Deviatio) n(% AECL - 4.8 - 2.2 - 0.2 0.8 4.0 Copper activitie e normalizear s d such thae bundlth t e average copper activit unitys yi . Lutetium-Manganese Ratio Integral information on neutron spectrum shape in the thermal energy range can be obtained from Lu-Mn relative activities. The lutetium-manganese ratio is calculated in the form: [ALu/AMn]p e wheractivatioar A e n rates. Table 3.9.1.5 give e experimentath s d an ln M valueR f o s deviation e valueth f o s calculate CLUy b d WIM d Ban S code t varioua s s locationse Th . agreement between the experimental and calculated values by both the codes is satisfactory for bot coolantse hth . 116 TABLE 3.9.1.5 LUTETIUM-MANGANESE RATIO Lutetium-Manganese Ratio Description RingA Ring B Ring C RingD Bundle Average D2O Coolant Expérimental 1.290 1.273 1.235 1.171 1.206 Déviâ) (% n tio CLUB - 2.8 - 2.3 - 1.5 0.1 - 0.8 Déviation (%) CNEA-PER2 0. - - 1.7 2 1. - 0.9 0.2 Déviatio) n(% CNEA-PIJ - 1.8 - 1.4 - 0.6 0.9 0.1 Déviation (%) AECL 0 0. - 0.3 3 0. - 1.1 0.5 r CoolanAi t Expérimental 1.289 1.272 1.2421.197 1.223 Déviatio) n(% CLUB - 1.0 - 0.3 - 0.1 - 0.3 - 0.2 Déviation (%) CNEA-PER -2 0.1. 3 8 0. 0.7 0.9 Déviatio) n(% CNEA-PIJ - 0.7 0.2 0.5 0.4 0.4 Déviation (%) AECL 1.1 1.3 1.2 0.6 0.9 the fuel cluster. 3.9.1.4. Conclusions Fast fission ratio is well predicted by CLUB and WIMS (AECL) /WIMS (CNEA) codes spatiae Th . l distributio fasf no t fission ratio withi fuee nth l cluste bettes i r r predicted by the CLUB and there is relatively more error in the WIMS results which are very sensitive to the choice of option used for solving the transport equation in the WIMS code. There is relatively large error of -10 % in the INPR-WIMS results. relative Th e conversion rati wels oi e erroe l th codeth e predicte n i d th r l san al y db predictio less ni s . Similarlytha% n1 U-23e th , 5 fission rat predictes ei y db withi% 7 n0. computee th r code CLUB considerewhice b n hca d quite satisfactory. Ther slightls ei y more error for central rod in the AECL results. neutroe Th n flux spatial distribution, tha , coppeis t r activities calculatee th l al y db codes are also found to be in good agreement with the experimental values. However, the PERSIUS optio f WIMno S gives relatively larg emoderatore erroth n i r . Similarlye th , lutetium-manganese rati wels oi l predicte CLUy db WIMd Ban S codes. 3.9.2. Task 9.2 : Analysis of experiments with 28-rod fuel clusters in ZED-2 reactor The results of analysis of the experiments were submitted by India-ThPD [82, 83] only. The experiments were analyzed using the computer codes CLUB and CLIMAX. Romania presented the results of analysis of experiments with 19-rod cluster [81] instead of 28-rod cluster using WIMS code. Even thoug presene part th h f no theo t te benchmarkyar , they have been included. 117 3.9.2.1. Description benchmarkthe of Descriptio f Experimenno t In Canada, experiments were performed to determine cell parameters in lattices of 28- fued ro l cluste t ZED-2ra reactor usincoolants a (Her g)ai D d 2sO an [75,76] buckline Th . g measurements were performed at 8 triangular lattice pitches of 24, 26, 28, 30, 32, 34, 36 and 40 cm for both the coolants and detailed reaction rate measurements were carried out at 4 lattice pitches of 24, 28, 32 and 40 cm only. The 28-rod fuel assembly contained sixteen, eight and four rods on circles of diameters 8.412 cm, 5.308 cm and 2.326 cm respectively. Each fuel rod contained a stack of natural uranium oxide pellets (14.2 mm diameter, density 10.45 g/cc) within a zircaloy-2 sheath of I.D. 14.3 mm and O.D. 15.2 mm. Fuel stack length was 47.73 cm and overall bundle length 49.67 cm. Five of these fuel clusters were stacked hi a 65s Al pressure tube which was surrounded by a 50s Al calandria tube to form a fuel assembly. The I.D. and O.D. of pressure tube were 10.19 cm and 10.78 cm respectively and corresponding dimensions of calandria tube were 12.46 cm and 12.74 cm respectively. Fifty-five fuel assemblie triangulan si r arrays were use measurementr dfo pitchet sa s 24 to 40 cm inclusive. When these 55 assemblies were not sufficient for criticality they were augmented with driver assemblies positioned in an annulus outside the 55 test assemblies. Parameter Comparee b o st d Following parameters measured with 28-rod fuel clusters have to be calculated: calculation of K-eff using measured material buckling; fast fission ratio initial conversion ratio neutron density measurements 3.9.2.2. Computer codes used The experiments with 28-rod fuel clusters have been analyzed by the computer code CLUB (India-ThPD) [82], CLIMAX (India-ThPD) [83 thosd an ]f 19-ro eo d cluster with WIMS code (Romania) [81]. 3.9.2.3. Main results The various tables in this report give measured values followed by deviation (%) from measuree th d values obtaine followss da : D = 100* (calculated/measured -1) K-eff K-efe calculates Th wa f d fro measuree mth d material buddings. Tables 3.9.2.d 1an 3.9.2.2 give K-eff calculate 28-ror dfo d cluster experiments usin computee gth r codes CLUB and CLIMAX respectively seee b nn froca t I m. these tables thaK-efe r Dth t fo 2f0 coolant is underpredicte boty codesdb e haveragth e Th . e valu f K-efeo CLUy b f B0.997s i e th r 9fo 118 TABLE 3.9.2.1 ANALYSI F 28-ROD-FUEL-CLUSTEO S R EXPERIMENTS WITH CLUB Hexagonal ô C Coolant pitch (cm) K-eff Expt. D (%) Expt. D (%) 24 0.9987 0.0580 2.59 0.9258 - 2.25 26 0.9963 28 0.9968 0.0582 -1 .89 0.8459 - 2.45 D20 30 0.9983 32 0.9978 0.0554 1.26 0.8054 - 2.89 34 0.9986 36 0.9982 40 0.9985 0.0547 1.164 0.7663 - 2.88 24 1.0021 0.0691 -1 .01 0.9144 - 2.67 26 0.9980 28 0.9994 0.0632 2.69 0.8117 - 0.87 Air 30 1.0015 32 1.0022 0.0619 2.42 0.7663 - 0.98 34 1.0020 36 1.0010 40 1.0009 0.0624 -0 .16 0.7312 - 1.79 TABLE 3.9.2.2 ANALYSI F 28-ROD-FUEL-CLUSTEO S R EXPERIMENTS WITH CLIMAX Hexagonal ô C Coolant pitch (cm) K-eff Expt. D (%) Expt. D (%) 24 0.9937 0.0580 -11 .21 0.9258 - 2.84 28 0.9930 0.0582 -18 .04 0.8459 - 3.78 D2O 32 0..9939 0.0554 -16 .61 0.8054 - 4.36 36 0.9935 40 0.9937 0.0547 -16 .82 0.7663 - 4.02 24 1.0012 0.0691 -11 .58 0.9144 - 4.21 28 0.9987 0.0632 -11 .55 0.8117 - 3 .08 Air 32 1.0010 0.0619 -13 .08 0.7663 - 3 .25 36 0.9992 40 0.9986 0.0624 -15 .22 0.7312 - 3.68 eigh correspondine t th case d san g valu CLIMAr efo 0.9936Xs i r coolan K-efe ai r Th . fo f t is overpredicte averag e CLUy th db d B an eeigh e valuth r t efo case s 1.0009si case th en I . of CLIMAX, the average value of K-eff is 0.9997. Thus, the K-eff can be predicted within k witm h2 CLU withid witk Ban m hn6 CLIMA r 28-roXfo d clusterr ai r Dd fo 2sO an coolants. Table 3.9.2.4 gives the K-eff for 19-rod cluster experiments calculated with WIMS seee b ncoden froca t mI . this table tha K-efe tth underpredictes i f botr coolantde fo hth d san average th e valur coolant ai K-eff eo d 0.997s an i f s O 0.996respectivelyd 2 5D an r 6fo . 119 Fast Fission Ratio PHWRs use natural uranium as fuel, and a few percent of all fissions take place in U-238. A parameter related to these fast fissions in U-238 is the fast fission ratio (o), which is defined as [U-23= 6 8 fissions/U-235 fission] Tables 3.9.2.1 and 3.9.2.2 give the experimental values of ô and deviations of the values calculate e codeth y sb dCLU d CLIMABan r 28-roXfo d cluster experiments respectively. It can be seen from these tables that the fast fission ratio can be predicted within 2 % by CLUB and the error is as large as - 18 % in the CLIMAX. Table 3.9.2.4 gives the deviations of the calculated values of fast fission ratio for 19-rod cluster experiments using WIMS code systematicalls i t I . y underpredicte casee th r bot l sfo hal n di . % coolante 6 erroe largs 1 th a th s s i erd a san Initial Conversion Ratio The long term reactivity behaviour of a PHWR is very significantly affected by the productio fissilf no e Pu-239 formed from neutron captur U-238n ei therefors i t I . e important e abl b o calculat t eo t e reliably some parameters relate o U-23t d 8 capture suitablA . e paramete initiae th s i rl conversion ratio whicC , defines hi s da C[U-23= 8 captures/U-235 absorptions] TABLE 3.9.2.3 COMPARISON OF NEUTRON DENSITY EXPERIMENTS FOR 28-ROD-FUEL-CLUSTER EXPERIMENTS WITH CLUB Hex. nPT/nf nCT/nf r^/nf pitch (cm) Expt D {%) Expt D (%) Expt D (%) D20 Coolant 24 1.478 - 3.11 1.553 - 2.64 1.947 - 2.41 28 1.504 - 3.66 1.585 - 3.41 2 .115 - 3.12 32 1.489 - 2.01 1.575 - 2.09 2 .215 - 1.58 40 1.1503 - 2.26 1.559 - 0.45 2 .452 - 2.36 Air Coolant 24 1.390 - 3.45 1.449 - 1.52 1.895 - 2.74 28 1.389 - 2.73 1.466 - 2.05 2 .050 - 3.17 32 1.398 - 3.00 1.470 - 1.97 2 .175 - 2.94 40 1.4114 - 3.75 1.490 - 2.95 2 .422 - 4.17 120 TABLE 3.9.2.4 ANALYSI F 19-ROD-FUEL-CLtISTEO S R EXPERIMENTS WITH WIMS Hexagonal pitch (cm) K-eff Coolant Expt D (%) Expt D (%) 24 0.9937 0.0580 - 10 .22 0.9258 - 1.86 28 0.9958 0.0582 - 15 .71 0.8459 - 2.51 D2O 32 0.9982 0.0554 - 14 .30 0.8054 - 3 .22 40 1.0023 0.0547 - 16 .13 0.7663 - 3 .56 24 0.9959 0.0691 - 7.54 0.9144 - 2.38 28 0.9947 0.0632 - 6.01 0.8117 - 0.88 AIR 32 0.9983 0.0619 - 7.88 0.7663 - 1.12 40 0.9973 0.0624 - 12 .57 0.7312 - 2.07 Tables 3.9.2.1 and 3.9.2.2 give the experimental values of C and deviations of the values calculated by the codes CLUB and CLIMAX for 28-rod cluster experiments. It can seee b n from these tables tha initiae tth l conversion rati systematicalls oi y underpredicted dan the error lies within 1-3 % in CLUB and within 2-4 % in CLIMAX results. Table 3.9.2.4 give deviatione sth calculatee th f o s d value f initiao s l conversion rati r 19-roofo d cluster experiments using WIMS code. The WIMS also underpredicts it and the maximum error is -3.56 % for D2O coolant at lattice pitch of 40 cm. Neutron Density Distribution Table 3.9.2.3 give experimentae th s l value f variouo s s neutron density ratiod an s deviatio valuee th f no s calculate CLUy db 28-ror Bfo d cluster experiments experimentae Th . l error in these measurements is of the order of 2 %. From this table, it is clear that neutron density ratio underpredictee ar s casese mose th l Th .al t n importandi t ratis i ns oi md /nan f underpredicted by 2-3 %. 3.9.2.4. Conclusions The K-eff is predicted within 2 mk by the computer code CLUB for 28-rod cluster experiments with D2O and air coolants. In the case of CLIMAX code, the K-eff is predicted errocoolante O 2 th mor s D i rr d 19-ro r an Fo e.withifo k dm clusten6 r experiments, WIMS code underpredicts the K-eff for both the coolants and the average value of K-eff is 0.9975 r coolantai d anan d s O 0.996respectively2 D r 6fo . Fast fission rati predictes oi CLUy db withi 28-ror % B fo 4 n3- d cluster experiments. The CLIMAX code underpredicts and it lies between 11-18 %. For 19-rod cluster experiments WIMe th , S code underpredict. % erroe 6 largs 1 a th s s d i rea an t si The initial conversion ratio is systematically underpredicted by CLUB and CLIMAX codes for 28-rod cluster experiments and the error lies within 1-3 % for CLUB and within r CLIMAXfo % r 4 19-ro2- Fo . d cluster experiments WIMe th , S also underpredictd an t i s the maximum error is -3.56 %. 121 The relative neutron density distributio underpredictens i computey b % 3 r2- cody db e CLU 28-ror Bfo d cluster experiments. 3.9.3. Task 9.3: Analysi f isotopiso c compositio 19-ror nfo d fuel cluster The third set of experiment to be analyzed was regarding the analysis of isotopic compositio f fue no t variou a l s burnups. This report present resulte sth s submitteo tw y db countries, namely, India-ThPD [84] and Argentina [85]. The computer codes used in the analysis of this experiment are CLUB (India-ThPD) and WIMS (Argentina). Both the codes use WIMe dth S cross section library. However, different version f WIMo s S library might have been used. 3.9.3.1. Description of the benchmark Descriptio Irradiatiof no n Experiment The irradiatioe MW n 5 experimen2 a s reactori D conductes D NP wa t e NP . th n di heavy water coole moderated dan d reactor principae forTh .e th fuef mo n i l l s fuebundlewa l s of 19 zircaloy-clad, natural UO2 fuel elements in a 1-6-12 circular array with an O.D. of 6.4 cm. The fuel element radius was 0.714 cm and outside radius of the cladding was 0.763 cm. reactoe Th r also contained some periphera elemen7 l t bundle similaf so r uranium content, a few experimental bundles and a highly enriched uranium booster rod for startup and xenon transient override. The eight bundles were selected to have an irradiation history typical of an unperturbed CANDU-type reactor with bi-directional on-power fuelling. In particular, none of the bundles spent significant time in "non-standard" positions like near boozer rods experimentar o l fuel bundles, which might pertur neutroe bth n spectrum burnue Th . thesf po e bundles ranged from 1000 to 10000 MWD/T. r eacFo h bundle pinx si f twelv,oute se o th rn ei ring innee , th thre n ri ex pinsi f so centrae th ring wern d pi lan ,e dissolve analyzedd dan . Uranium analyses were made witha UF6 mass spectrometer and thermoionization device was used for plutonium analyses. These measurements provided relative concentration f U-235so , U-236, U-238, Pu-239, Pu-240, Pu-241 and Pu-242 hi each ring of each bundle. Burnup values were determined from the U-235/U-238 ratios. CaseAnalyzee b o st d The benchmark consists of analyzing the following isotopic concentration ratios for the fuel pins of each ring and bundle average values for the above eight cases: N235 N236 N239 N240 N241 N242 N238 N238 N238 N239 N239 N239 3.9.3.2. Computer codes used The experiments with 19-rod fuel clusters have been analyzed by the computer code CLUB (India-ThPD) [84] and by Argentina with the WIMS code [85]. 122 •~ 3.9.3.5. Main results Tables 3.9.3.1-3.9.3.6 give the results of calculations for various isotopic ratios obtained by the computer codes CLUB (India-ThPD) and WIMS (Argentina). These tables give measured values followe deviationy db fro) measuree sm(% th d values obtainee th y db two codes as follows: D100= * (calculated/measure) 1 d- f eaco hd en table e e averagth th , n I e deviatio standard nan d case e deviatioth l s al r nfo corresponding to different burnups are given. In the case of U-236, the correct measured values are not given for the first case, so the first case has been omitted in the calculation of average deviation and standard deviation. Similarly, the concentration of Pu-240, Pu-241 and Pu-242 is very small at very low burnups and there can be large error in the measured values so the first two cases have been omitted for the calculation of average deviation and standard deviation for these isotopes. TABLE 3.9.3.1 COMPARISON OF U-235 ISOTOPIC RATIO Burnup Central Middle Outermost Bundle Description (MWD/T) Ring Ring Ring Average Experimental 980 0.636 0.626 0.603 0.612 Deviatio) (% n CLUB 7.14 -0.47 0.00 0.00 0.00 Deviation (%) WIMS 8.09 -0 .72 0.10 0.04 0.02 Experimental 1500 0.596 0.586 0.552 0.565 Deviation (%) CLUB 3.53 -0.34 -0.34 0.18 0.00 Deviatio) (% n WIMS 4.77 -0.67 -0.28 0.19 0.00 Experimental 3250 -0.484 0.468 0.420 0.438 Deviation (%) CLUB -2.55 0.21 0.00 -0.24 0.00 Deviation (%) WIMS -1.27 -0.45 0.24 -0.27 0.01 Experimental 5000 0.384 0.363 0.301 0.325 Deviatio) (% n CLUB 0.26 -0.52 -0.55 0.33 0.00 Deviatio) (% n WIMS 2.01 -1.45 -0.17 0.24 -0.03 Experimental 6500 0.309 0.285 0.223 0.247 Deviatio) (% n CLUB 2.55 -0.65 -0.35 0.00 0.00 Deviation (%) WIMS 4.50 -2.08 0.10 -0.21 -0.17 Experimental 7900 0.262 0.238 0.174 0.199 Deviation (%) CLUB 0.35 -1.53 -1.26 1.15 0.00 Deviation {%) WIMS 2.25 -3.10 -0.65 0.70 -0.14 Experimental 9100 0.226 0.203 0.143 0.166 Deviatio) (% n CLUB -1.50 -1.33 -1.48 0.70 0.00 Deviatio) (% n WIMS 0.24 -2.96 1 -0.4 0.40 0.05 Experimental 10800 0.174 0.156 0.101 0.122 Deviation (%) CLUB -0.94 0.00 -1.92 0.99 0.00 Deviation (%) WIMS 0.79 -2.20 -1.22 0.64 -0.13 Average D (%} CLUB 1.10 -0.58 -0.74 0.39 0.00 STD (%) CLUB 2.96 0.56 0.68 0.47 0.00 Average D (%) WIMS 2.67 -1.71 -0.29 0.22 -0.05 STD (%) WIMS 2.80 0.98 0.45 0.33 0.08 123 TABLE 3.9.3.2 COMPARISO F U-236-TO N O U-238 ISOTOPIC RATIO Central Middle Outermost Bundle Description Ring Ring Ring Average Experimantal < 0.020 < 0.020 < 0.020 < 0.020 Deviation {%) CLUB Deviation (%) WIMS Experimental 0.021 0.023 0.026 0.025 Deviation {%) CLUB 0.00 -4 .35 3.85 0.00 O Deviation (%} WIMS -5 .02 O .60 _2 .14 -4 .97 Experimental 0.037 0.039 0.047 0„ 044 Deviation (%} CLUB 0.00 2.56 0.00 0.00 Deviatio} (% n WIMS -1 .79 — 2 .28 -3 .99 -3 .53 Experimental 0.052 0.059 0.067 0.062 Deviation {%) CLUB 1.92 -5 .08 -2 .99 0.00 Deviatio) (% n WIMS -0 .75 -8 .80 -7.30 -4.95 Experimental 0.065 0.066 0.078 0.074 Deviatio) (% n CLUB 0.00 3 .03 -1 .28 -1.35 Deviation (%) WIMS -3 .54 -1 .47 -5 .90 -5.15 Experimental 0.070 0.074 0.085 0,081 Deviatio) (% n CLUB 2.86 1.35 -1 .18 0.00 Deviatio) (% n WIMS -0 .36 -2 .63 -5 .85 -4 .98 Experimental 0.076 0.079 0.088 0,085 Deviation (%) CLUB 1.32 2.53 0.0 0.00 Deviation (%) WIMS -1 .71 -2 .62 -4 .11 -4 .10 Experimental 0.082 0.087 0.091 0 089 Deviatio} (% n CLUB 3 .66 1.15 3.30 3.37 Deviatio) {% n WIMS -0 .42 -3 .83 -1 .11 -1.62 Average D (%} CLUB 1.39 0.17 0.24 0.29 STD {%) CLUB 1.38 3 .16 2.30 1.34 Averag) {% eD WIMS — 1 .94 -4 .32 -4 .34 -4 .19 STD {%} WIMS 1.62 2.84 2.03 1.18 TABLE 3.9.3.3 COMPARISON OF PU-239 TO U-238 ISOTOPIC RATIO Central - Middle Outermost Bundle Description Ring Ring Ring Average Experimentl a 0.066 0.069 0.083 0.078 Deviation (%) CLUB 4.55 5.80 1.20 1.28 Deviation (%) WIMS 0.61 -0 .32 0.13 -0 .40 Experimental 0.092 0.095 0.115 0.107 Deviation (%) CLUB 3.26 5.26 -0 .87 0.93 Deviatio) {% n WIMS 0.80 1.47 -0 .12 0.56 Experimental 0.160 0.164 0.186 0.178 Deviatio) (% n CLUB -1 ,25 Q .00 -2 .69 -2 .25 Deviatio} (% n WIMS -2 .34 -2 .00 -0 .26 -1 .04 Exp eimenr 1 ta 0.204 0.211 0.229 0.222 Deviation (%) CLUB 0.98 0.00 -1.75 -0 .90 Deviation {%) WIMS -0 .17 -1 .58 1.19 0.29 Expérimenta 0.234 0.233 0.248 0.242 Deviation (%) CLUB 0.00 2.15 0.00 0.83 Deviatio) (% n WIMS -1 .41 0.31 2.60 1.92 Experimental 0.249 0.243 0.260 0.254 Deviatio) (% n CLUB 0.00 3.70 -0 .77 0.79 Deviation (%) WIMS 1.76 1.56 1.78 1.54 Experimental 0.256 0.252 0.265 0.261 Deviatio) (% n CLUB 0.78 3.17 -0 .38 0.77 Deviatio) (% n WIMS -1 .24 0.92 1.86 1.18 Experimental 0.264 0.261 0.271 0.267 Deviatio} (% n CLUB 1.89 3.45 -0 .37 1.12 Deviation (%} WIMS -0 .72 0.62 1.50 1.29. Averag) (% eD CLUB 1.28 2.94 -0 .70 0.32 STD (%) CLUB 1.77 2.01 1.08 1.16 Average D (%) WIMS -0 .78 0.12 1.08 0.67 STD {%) WIMS ~ 1.05 1.25 0.98 0.95 124 TABLE 3.9.3.4 COMPARISON OF PU-240-TO PU-239 ISOTOPIC RATIO Central Middle Outermost Bundle Description Ring Ring Ring Average Experimental 5.060 5.570 6.500 6.170 Deviatio) (% n CLUB -12.08 -15.3 -15.20 -15.15 Deviatio) (% n WIMS - 4.55 - 9.57 - 9.49 - 9.25 Experimental 7.230 7.840 9.370 8.850 Deviation (%) CLUB - .7.75 - 9.63 -11.27 -10.95 Deviation (%) WIMS - 2.06 - 5.69 - 7.41 - 6.87 Experimental 13.600 14.700 17.500 16.500 Deviatio} (% n CLUB 0.30 - 1.05 - 1.77 - 1.60 Deviation (%} WIMS 3.31 0.11 - 0.47 - 0.11 Experimental 21.500 23.100 27.500 25.900 Deviatio) {% n CLUB 0.07 - 0.48 - 1.32 - 1.16 Deviatio) (% n WIMS 2.99 0.57 0.13 0.41 Experimental 28.600 30.700 36.400 34.300 Deviation (%) CLUB 0.62 0.17 - 1.10 - 0.82 Deviation {%) WIMS 3 .44 1.12 0.62 0.89 Experimental 34.200 36.600 42.700 40.400 Deviatio) (% n CLUB 0.53 0.28 - 0.26 - 0.16 Deviation (%) WIMS 3.30 1.23 1.74 1.74 Experimental^ 38.600 41.500 47.800 45.400 Deviation (%) CLUB 1.03 0.16 0.05 0.02 Deviation (%) WIMS 3.67 1.03 2.19 1.97 Experimental 47.300 49.300 56.100 53.600 Deviatio) (% n CLUB -1.43 0.45 - 0.12 - 0.22 Deviation (%) WIMS 1.14 1.44 2.52 2.08 Averag) (% eD CLUB 0.18 - 0.08 - 0.75 - 0.66 ) (% D ST CLUB 0.78 0.52 0.68 0.59 Average D (%) WIMS 1.41 - 1.22 - 1.27 - 1.14 SRTD (%) WIMS 2.88 3.85 4.28 4.10 TABLE 3.9.3.5 COMPARISOF NO PU-24O 1T PU-239 ISOTOPIC RATIO Central Middle Outermost Bundle Description Ring Ring Ring Average Experimental 0.330 0.410 0.490 0.460 Deviatio) (% n CLUB -36.06 -43.41 -38.98 -40.00 Deviatio) (% n WIMS -25.94 -36.19 -31.01 -32.07 Expérimenta 0.590 0.690 0.900 0.830 Deviatio) (% n CLUB -23.05 -27.68 -28.11 -28.31 Deviation (%) WIMS -13.54 -20.84 -21.38 -21.30 Experimental 1.660 1.830 2.470 2.240 Deviation (%) CLUB -5.24 - 5.03 - 7.21 - 6.56 Deviatio} (% n WIMS 1.49 - 1.28 - 3.93 - 2.79 Expe riment a 1 3.110 3.450 4.690 4.240 Deviation (%) CLUB 0.10 0.00 - 1.58 - 1.23 Deviatio) (% n WIMS 7.06 3.61 0.81 1.90 Experimental 4.570 5.050 6.980 6.290 Deviatio) (% n CLUB 1.93 2.06 - 1.32 - 0.78 Deviatio) {% n WIMS 8.18 5.01 - 0.02 1.38 Experimental 5.840 6.450 8.740 7.900 Deviation (%} CLUB 0.74 0.74 - 1.35 - 1.01 Deviatio) (% n WIMS 6.79 3.54 - 0.42 0.85 Experimental 6.850 7.710 10.230 9.260 Deviation (%) CLUB 0.67 - 1.32 - 2.21 - 1.80 Deviatio) (% n WIMS 6.82 1.42 - 1.44 - 0.06 Experimental 8.840 9.420 12.120 11.140 Deviatio) (% n CLUB -2.18 0.73 0.91 0.32 Deviatio} (% n WIMS 3.69 3.55 1.56 2.01 Average D (%) CLUB -0.67 - 0.47 - 2.13 - 1.84 STD (%) CLUB 2.39 2.27 2.47 2.20 Average D (%) WIMS 5.67 2.64 - 0.57 0.55 STD (%) WIMS 2.32 2.04 1.77 1.65 125 TABLE 3.9.3.6 COMPARISO PU-24F O N PU-23O 2T 9 ISOTOPIC RATIO Central Middle Outermost Bundle Description Ring Ring Ring Average Experimental 0.019 0.031 0.027 0.028 Deviatio) (% n CLUB -73.68 -80.65 -66.67 -71.43 Deviatio) (% n WIMS -69.84 -79.13 -62.32 -68.07 Experimental 0.032 0.036 0.061 0.050 Deviatio) (% n CLUB -50.00 -47.22 -50.82 -48.00 Deviatio) (% n WIMS -42.74 -42.44 -46.16 -42.46 Experimental 0.130 0.160 0.270 0.220 Deviatio) (% n CLUB - 3.08 - 6.25 - 9.63 - 4.09 Deviatio) (% n WIMS 3.73 - 4.17 - 7.71 - 1.72 Experimenl ta 0.420 0.540 0.890 0.750 Deviation {%) CLUB 0.48 - 5.74 - 4.94 - 3.73 Deviation (%) WIMS 9.77 - 2.47 - 1.86 0.18 Experimental 0 .870 1.050 1.880 1.580 Deviatio) (% n CLUB 2.99 3.14 - 3.30 - 2.22 Deviation (%} WIMS 11.84 6.10 - 0.99 0.87 Experimental 1.360 1.660 2.890 2.440 Deviatio) (% n CLUB 3 .38 2.47 - 1.07 - 0.66 Deviation (%) WIMS9 .2 2 1 5.37 1.08 2.34 Experimental 1.760 2.330 3.900 3.310 Deviation (%) CLUB 9.55 0.09 0.31 0.12 Deviation (%) WIMS 19.02 2.94 2.35 3.05 Experimental 3.060 3.510 6.050 5.110 Deviatio) (% n CLUB - 0.59 4.64 . 0.83 1.23 Deviation (%) WIMS 8.13 7.75 2.89 . 4.22. Average D (%) CLUB 2.12 - 0.27 - 2.97 ^ - 1.56 STD (%) CLUB 3.97 4.26 3 .58 1.96 Average D {%) WIMS 10.80 2.59 - 0.71 1.49 STD (%) WIMS 4.64 4.44 • 3.56 1.96 Table 3.9.3.1 gives the values of deviation (%) from the measured values for U-235/U-238 isotopi seee b c n n ratio froca t mI . this table tha deviatione pine th tth s l al n si burnupe th l usuallal e r sar bundle ofo th f yd smallean concentratioe Th . f U-23no e th n 5i fue average l th bundl n o es ei underpredicte r inne ringo dfo overpredicted tw r san e th r dfo outer ring by both the codes. Table 3.9.3.2 give e valueth s f deviatioo s ) fro e measure(% nmth d valuer fo s U-236/U-238 isotopic ratio e bundlTh . e average deviatio r CLU s 0.2nfo i 9% B whics hi quite small compared to large underprediction of 4.20 % using WIMS code. Table 3.9.3.3 gives the values of deviation (%) from the measured values for PU-239/U-238 isotopic ratio e concentratioTh . f Pu-23no overpredictes 9i pine f th o s n di inner ring underpredicted san pine outermosf th so n di t rin CLUr gfo B results bundle Th . e average deviatio e concentratio Th s 0.3n i . 2% f Pu-23no underpredictes 9i centrae th n di l overpredicted an n pi pine f remaininth so n di g rings bundle Th . e average deviatio 0.6s ni 7 % and can be considered satisfactory for both the codes. 126 Table 3.9.3.4 give e valueth s f deviatioo s ) froe measure(% nmth d valuer fo s Pu-240/Pu-239 isotopic ratio. The concentration of Pu-240 is on the average overpredicted underpredicted e centrafoan th r n pi le pin f th otheo sr fo dr ring r bot e fo codess hth . However, there is slightly less error in the results of CLUB. The bundle average deviation in the prediction of Pu-240/Pu-239 ratio is -0.66 %.in the CLUB and -1.14% in the WIMS. Table 3.9.3.5 gives the values of deviation (%) from the measured values for Pu-241/Pu-239 isotopic ratio. It is underpredicted for all the rings in the results of CLUB and overpredicted for the central and middle rings and underpredicted for the outermost ring in the WIMS results erroe , howeverTh . is r , smal bot n casesi le hth . Table 3.9.3.6 give e valueth s f deviatioo s ) froe measure(% nmth d valuer fo s Pu-242/Pu-239 isotopic ratio overpredictes i t I . underpredicte d centrae an th r n dpi fo l r dfo the remaining pins in the CLUB results whereas it is overpredicted for the central pin and the pin middlf so e rinunderpredicted gan outermose th r dfo WIMte rinth n gi S resultse Th . bundle average deviation is only -1.56 % for CLUB results and 1.49 % for WIMS results. furthee b n ca r t seeI n from Table 3.9.3.1 that bot codee hth s overpredic burnue e th t th n po average average Th . e deviatiocodeo e 2.61.1d e tw burnu Th e s ar e an . th 70 th % % n i y pb predictioe erroth n ri burnuf highesne o th r pfo t burned ordee fue boty th codee b lf hrth o s si of 100 MWD/T which can be considered very small. 3.9.3.4. Conclusions The relative concentration of U-235 is predicted satisfactorily by both the codes. It is underpredicted for the inner two rings and overpredicted for the outermost ring by both the codes. relative Th e concentratio f U-23no predictes 6i d wel CLUBy b l . However, thers ei large WIM e erroth n i rS results. e relativTh e concentratio f variouno e bundlth sr e predicte isotopefo ear u P f do s within reasonable error by both the codes. However, there are small differences in the sign of deviations within the fuel cluster. Further, the average concentration of Pu-241 and Pu-242 withi clustee nth underpredictes i r CLUy db overpredicted Ban WIMSy db . predictio e erroe th Th n MWD/i 0 r ordee 10 f burnuth no f f r o boto Te fo s hpth i codes consideree whicb n hca d very small. 3.9.4. Task 9.4: Analysi adjustee th f experimentso d rro ZED-2n si Measurement mock-un so p adjusters were mad ZED-n ei 2 reactor core- usin28 , g52 element natural uranium oxide, D2O cooled, fuel clusters [77]. Reactivity worths were obtained from critical heigh d levean t l coefficien f reactivito t y measurements. Flux distributions on and around the absorbers and throughout the reactor core were obtained from the activity of Cu and In foils. These set of experiments have been recommended for analysis by the participants of the IAEA CRP on In-core Fuel Management Code Package Validation forPHWRs [52]. experimente Th s have beeparticipatine th nf o analyze o tw gy db countries , namely Argentina (one submission) [86] and India-ThPD (two submissions) [87,88]. 127 — 3.9.4.1. Benchmark problem specification Core Geometry The calandria housin reactoe gth r cor cylindricaa s ei l aluminium tank calandrie Th . a is surrounde graphita y db mean i e m nreflectoc radia0 6 f lo r thicknesf o sidee d th san n so bottome thicth m c n ko 0 .9 presen s Radiall2.8f i o m 6p c betwee n tga i ya calandrie nth d aan the graphite reflector. The core layout is given in Fig. 3.9.4.1. There are 52, 28 element natural U02 cooleO 2 D , d fuel bundle double-wallea n si d aluminium tube. Five fuel bundles are stacked vertically with the bottom of fuel located 15 cm from the calandria floor. Table 3.9.4.1 gives more details about the fuel cluster. Adjusters The adjusters consis eithef o t S.Sra . concentrituba r eo botd ro hca madtubd ean e of S.S. Six types of adjusters considered for measurement are serially numbered 1 to 6 and describee ar Tabln di e 3.9.4.2 adjustee ende th f Th .so r tubes were ope alloo nt ingrese wth s of D2O. REVOLVING :" i •":•'.'•'.'. —REMOVABLE TOP SHIELD REFLECTOR REACTOR TANK 0,0-MOOERAKIR Alt? OUCI DUMP VALVE AM dimtncitn* in FIG. 3.9.4.1 Vertical Section of the ZED-2 Reactor ( Taken from the Directory of Nuclear Reactors ) 128 above adjusterTh x esi s were hel botn di h horizonta verticad an l l orientations during measurements. Horizontally the rod was hung symmetrically in the reactor along the K direction (Fig. 3.9.4.1) with its axis 130 cm from the calandria floor. Vertically the rod was at the centre of the reactor (lattice site KO) with the lowest point of the tube 2 cm from the calandria floor. TABLE 3.9.4.1 DESCRIPTION OF 28 ROD CLUSTER LATTICE CELL OF ZED-2 Fuel pellet radius 0.710m 5c Sheath inner radius 0.7155 cm Sheath outer radius 0.7609 cm Effective fuel density (10.45*47.73/49.67) 10.042 g/cc Radii of the centres of the pins in the fuel cluster Inner ring (4 pins) 1.163 cm Intermediate ring (8 pins) 2.652 cm Outer ring (12 pins) 4.206 cm Pressure tube inner radius 5.0965 cm Pressure tube outer radius 5.3925 cm Pressure tube material Al 65S Calandria tube inner radius 6.2m 3c Calandria tube outer radius 6.36m c 9 Calandria tube material Al 5OS Lattice pitch 28.575 cm Temperature 23.3 °C Heavy water purity 99.80% 4wt Areas of the bundle end region aluminum 38.474 cm2 zircaloy 10.222 cm2 air 35.411 cm2 coolant 43.329 cm2 TABLE 3.9.4.2 DESCRIPTIO ABSORBERF NO S MATERIALS Outer Diameter of all Adjusters = 7.62 cm Adjuster Tube Wall Rod Length Stainless no. Thickness Diameter Steel (cm) (cm) (cm) Type 1 0.1713 _ _ _ 287 304 L 2 0.1713 1.273 286 316 3 0.1713 1.905 286 316 4 0.087 ___ 287 304 L 5 0.087 0.634 286 316 6 0.087 1.273 286 316 129 Locatio f Detectorno s Flux distribution core th e n si wer e measured usin , In-AgCu l foil activationse Th . foils were hun t variouga s radial location axiad san l elevation core th en si includin gsouta h reference thimble (Fig. 3.9.4.1). In the vicinity of the adjusters the locations of foils, wires or strips around vertically and horizontally suspended absorbers are shown in Figs. 3.9.4.2 and 3.9.4.3 measuremene Th . tvertica e directioth r fo l absorbernN alons i s e th glina n ei moderator symmetricall i betweeyh fueo s directei ntw l E clusterN dd towardan s fuea s l cluster. For horizontal orientation the measurement direction Al is similar to N while the directio narroe th lie 2 betweep n A si w ga absorbe e fuee nth th l d clusteran r . N Cu WIRES ON AI STRIP PLAN Cu AND In FOILS -1.3m 5 Cu WIRES ON AI STRIP - 1.30m Cu BAND -1.2m 5 0 O n I FOIL D SCAN u ELEVATION FIG. 3.9.4.2 Locations of Foils, Stripsabout Wiresand and on Vertical Adjusters 130 N (A!- • ) Cu WIRES ON AI ' (A2) STRIP Cu BAND u STRIC P Cu AND In FOILS ON BOTTOM SURFACE FIG. 3.9.4.3 Locations of Foils, Strips and Wires on and about Horizontal Adjusters 3.9.4.2. Calculational methods and computer codes used The problem has been analyzed by three models up to supercell level. Two evaluations wer coreo t carrieep u level t varioue dou Th . s corceld an le modelss a use e dar below: Model-A: WIMS-D4-SHETAN with and without leakage correction for cell and PUMA- C for core analysis (Argentina) [86]. Model-B: PHANTOM with and without leakage correction for cell and CEMESH for core analysis (India-ThPD) [87]. Model-C: CLUB-DTF-IV-BOXER celr 3fo l analysis (India-ThPD) [88]. Model-A and Model-C were used to analyze the experiments with adjusters 1 and 3 in horizontal or vertical orientations while model-B was used to analyze all the experiments adjusterx o si forientationso tw n i s . 131 3.9.4.3. Analysis ZED-2of experiments 28 Rod Fuel Cluster Cell Table 3.9.4.3 gives cell average parameters for uncontrolled fuel cluster cell with and without leakage correction havC s modeld eA . an use B sd same lattice code CLUr Bfo analyzin fuee gth l cluster f resultcello t ,gives se si onl e nyon here. Measured bucklinf go uses 3.7obtai2wa o d t 7m" n parameters with leakage correction sees i t nI . that predictions of integral parameters, k-inf and k-eff are quite close. The maximum difference in homogenized cross section fasr observes fo s ti production% 5 s da . TABLE 3.9.4.3 CELL AVERAGE PARAMETERS FOR 28 ROD CLUSTER OF ZED-2 Before leakage correction After leakage correction Argentina India Percent Argentina India Percent Parameter WIMS-D4 CLUB Diff. WIMS-D4 PIJMURLI Diff. K-inf inity 1.12988 1.12968 -0 .02 1.12988 1.12968 2 -0.0 K-eff ective 1.12988 1.12968 -0 .02 0.99936 0.99956 0.02 D-l 1.3348 1.3616 1.99 1.3370 1.3791 3 .10 D-2 0.8907 0.8920 0.15 0.8910 0.8917 0.08 Sigma a-1 1.6284E-3 1.5938E-3 -2 .15 1.6250E-3 1.5912E-3 -2.10 Sigma a- 2 3. 9293E-3 3 .9831E-3 1.36 3 .9214E-3 3. 9746E-3 1.35 Sigm2 1- a 9.0908E-3 9.2236E-3 1.45 8.9198E-3 9.0478E-3 1.42 Sigma 2-1 3 .9328E-5 4.0246E-5 2.31 4.2417E-5 4.3437E-5 2.38 Nusigma f-1 9.1730E-4 9.6853E-4 5.43 9.2567E-4 9.7807E-4 5.51 Nusigma f-2 4.8424E-3 4.8625E-3 0.41 4.8319E-3 4.8514E-3 0.40 Phi-l/Phi-2 0.4365 0.4362 -0 .07 0.4821 0.4813 -0.17 Percent Diff = (India-Argentina. / Averag 0 10 X )e 3-D Supercell Analyses Table 3.9.4.4 give comparisoe sth supercelD 3- f no l homogenized parameters without any adjusters. Adjuste , tub1 r e typadjusted typd ean ro , e tub 3 d r(se ean e Table 3.9.4.2) have been chosen for comparison of results of various models. Table 3.9.4.5 gives the TABLE 3.9.4.4 HOMOGENISED PARAMETER CELD 3- LR WITHOUFO S T ADJUSTERS Argentina Argentina India India Parameter SHETAN WIMS-D4 COMESH BOXER3 K-inf inity 1.12352 1.12929 1.12971 1.13090 D-l 1.3370 1.3370 1.3791 1.3825 D-2 0.8913 0.8910 0.8918 0.8931 Sigma -1 1.7340E-3 1.6250E-3 1.6431E-3 1.6369E-3 Sigma -2 4.0360E-3 3 .9210E-3 4.0666E-3 4.1928E-3 Sigm1 a -2 8.7903E-3 8.9200E-3 9.5109E-3 9.1922E-3 Sigma 2-1 4.2566E-5 4.2450E-5 3 .9804E-5 4. 0487E-5 Nusigma f-1 9.8805E-4 9.2570E-4 9.9898E-4 9.4327E-4 Nusigma f-2 4.9801E-3 4.8320E-3 4.9642E-3 5.1599E-3 132 comparison of two group parameters for various materials of the adjuster 3 i.e., 1.905 cm rod in 0.1713 cm tube. Table 3.9.4.6 gives the 3-D supercell homogenized parameters with adjuster 1 and 3. It must be mentioned that the supercell dimensions for Indian evaluations are PxPxL/2 while for Argentina it is P/2xPxL/2. Thus the two group parameters are not comparable Indiao tw e n resultTh . reasonable ar s y well comparable. TABLE 3.9.4.5 TWO GROUP CROSS SECTIONS FOR VARIOUS ADJUST ROD MATERIALS ( TUBE THICKNESS : 0.1713 CM , ADJUSTER ROD DIA. : 1.905 CM ) Tube of Thickness 0.1713 cm Inside Moderator Adjuster Rod of Dia 1.905 cm Argentina India India Argentina India India Argentina India" India Parameter WIMS-D4 PIJMURLI DTP-IV WIMS-D4 PIJMURLI DTF-IV WIMS-D4 PIJMURLI DTP-IV D-l 0.5333 0.6973 0.5651 1.2728 1.3154 1.2861 0.5414 0.6729 0.5708 D-2 0.3074 0.3029 0.3016 0.8685 0.8713 0.8712 0.3101 0.3047 0.3073 Sigma a-1 1.0220E-2 8.0093E-3 1.0344E-2 2.0310E-6 7.8293E-6 6.3476E-6 1.7122E-2 1.4575E-2 1.0222E-2 Sigma a-2 2.0226E-1 2.0209E-1 2.0667E-1 6.9490E-5 7.0938E-5 7.1928E-5 2.0179E-1 2.0176E-1 1.9517E-1 Sigm 2 1- a3.4982E- 3 2.5706E-3 3.6154E-3 1.5338E-2 1.1259E-2 1.5411E-2 3.3980E-3 2.5285E-3 3.4753E-3 Sigma 2-1 2.9900E-4 2.7624E-4 1.7606E-4 5.7759E-5 5.4253E-5 3.4041E-5 3.5138E-4 3.3715E-4 2.1123E-4 TABLE 3.9.4.6 CELL AVERAGE PARAMETER SUPERCELR SFO L WITH VARIOUS ADJUSTERS Horizontal Tube and Rod Adj. Horizontal Tube Adjuster Vertical Tube and Rod Adj. Argentina India India Argentina India India Argentina India India Parameter SHETAN COMESH BOXERS SHETAN COMESH BOXERS SHETAN COMESH BOXERS k infinity 0.95568 1.0311 1.0400 1.00340 1.0594 1.0643 0.85427 1.0457 1.0516 D-l 1.2953 1.3606 1.3803 1.3326 1.3622 1.3818 1.2720 1.3615 1.4445 D-2 0.8515 0.8901 0.8902 0.8872 0.8912 0.8911 0.8192 0.8908 0.8907 Sigma a-1 1.7546E-3 1.6152E-3 1..6958E-3 1.7401E-3 1.5991E-3 1.6877E-3 1.7636E-3 1.6107E-3 1.6556E-3 Sigm2 a- a 4.8785E-3 4.4160E-3 4..5786E-3 4.5978E-3 4.2782E-3 4,.4549E-3 5.8069E-3 4.3639E-3 4.5831E-3 Sigma 1-2 8.8707E-3 9.0291E-3 9..1593E-3 8.8797E-3 9.0376E-3 9.1654E-3 8.9252E-3 9.0320E-3 9.2070E-3 Sigm1 2- a 4.4059E-5 4.4174E-5 4..0705E-5 4.3466E-5 4.3833E-5 4,.0440E-5 4.5554E-5 4.4061E-5 4.0704E-5 nusigma f-1 9.7388E-4 9.7563E-4 9..6645E-4 9.7916E-4 9.7519E-4 9..6625E-4 9.6262E-4 9.7674E-4 9.4531E-4 nusigm2 f- a 5.0524E-3 4.8939E-3 5..1638E-3 5.0143E-3 4.8761E-3 5,.1486E-3 5.2553E-3 4.9087E-3 5.2193E-3 Table 3.9.4.7 gives Cu activation cross sections around a fuel bundle and adjusters. e valueTh s have been linearly interpolate r respectivdfo fro0 9 m d ereferencean 9 8 , 88 s models. Table 3.9.4.8 and 3.9.4.9 give the variation of calculated Cu activity in the vicinity tube th ef adjusterso adjusterd ro tube valuee d th , ean Th . s have been linearly interpolated from references 88, 89 and 90 for respective models at the points of measurement and renormalized to measured value at r=3.81 cm. It is observed that activities of all models are compared fairly well with measured ones. The PHANTOM model which uses diffusion theory shows slightly higher deviation moderaton si r regions sinc normalizatioe eth bees nha n done on the surface of absorber. Root mean square percent deviation from measurement for all the models are within 2.7 %. 133 TABLE 3.9.4.7 CU ACTIVATION CROSS SECTIONS AROUND A BUNDLE AND ADJUSTERS ( INTERPOLATED FROM REFERENCES 3, 4 AND 5 FOR RESPECTIVE MODELS) Aroun a dfue l bundle For adjuster-1 (tube) For adjuster-3 (tube & rod) r (cm) Cross section in barns r '(cm) Cross secti on in barns Cross secti on in barns SHETAN COMESH BOXERS SHETAN COMESH BOXERS SHETAN COMESH BOXERS 6.5 3-0 .17 3.048 3.137 0.5 3.012 3.212 3.005 3.192 2.975 3.207 7.0 3.178 3.081 3.157 1.0 3.070 3.211 3.097 3.191 3.077 3.206 7.5 3.187 3.106 3.173 2.0 3.158 3.206 3.163 3.187 3.152 3.202 8.0 3.197 3.123 3.184 3.0 3.153 3.197 3.167 3.176 3.168 3.190 8.5 3.207 3.136 3.191 4.0 3.139 3.178 3.155 3.152 3.166 3.169 9.0 3.212 3.146 3.197 5.0 3.184 3.208 3.203 3.194 3.195 3.212 9.5 3.217 3.153 3.202 6.0 3.200 3.219 3.218 3.208 3.210 3.224 10.0 3.220 3.159 3.205 8.0 3.201 3.228 3.221 3.207 3.222 3.226 11.0 3.224 3.167 3.209 10.0 3.203 3.229 3.224 3.208 3.224 3.228 12.0 3.227 3.172 3.213 12.0 3.200 3.228 3.222 3.205 3.223 3.225 13.0 3.229 3.176 3.214 14.0 3.194 3.224 3.215 3.198 3.220 3.217 14.0 3.230 3.178 3.215 16.0 3.170 3.219 3.202 3.174 3.216 3.204 15.0 3.230 3.180 3.217 18.0 "3.126 3.213 3.183 3.129 3.210 3.186 16.0 3.231 3.180 3.217 20.0 3.060 3.202 3.160 3.064 3.199 3.162 r - Distance from centre of fuel cluster : r' - Distance from centre of adjuster TABLE 3.9.4.8 VARIATIO CALCULATEF NO ACTIVITU D C VICINIT E TUBE TH TH EYN I F YO ADJUSTE RINTERPOLATE( D FRO RESPECTIVMR REFERENCEFO 5 D AN E 4 MODELS , S3 ) r'(cm) Horizontal direction - Al Horizontal directio2 A - n Vertical direction - N Meas. SHETAN COMESH BOXERS Meas. SHETAN COMESH BOXERS Meas. SHETAN COMESH BOXERS 3.81 20.50 20.50 20.50 20.50 16.93 16.93 16.93 16.93 19.71 --- 19.7119.71 4.81 21.57 22.39 21.37 22.12 17.29 17.26 17.40 17.36 21.00 --- 20.55 21.35 5.81 22.06 22.39 21.70 22.78 16.78 16.71 17.19 17.08 21.31 - -- 20.86 22.05 6.81 22.40 22.54 21.96 22.64 15.96 15.81 16.67 16.10 21.66 - -- 21.12 22.04 7.81 22.82 22.74 22.13 22.73 21.91 - -- 21.29 22.22 8.81 22.59 22.75 22.26 22.80 22.12 --- 21.4122.31 9.81 22.55 22.71 22.34 22.86 22.02 --- 21.5022.38 10.81 22.69 22.69 22.42 22.90 22.20 --- 21.5722.44 11.81 22.61 22.69 22.49 22.93 22.12 - -- 21.64 22.51 12.81 22.90 22.71 22.56 22.94 22.32 - -- 21.71 22.58 13.81 23.01 22.75 22.64 22.94 22.43 — 21.79 22.64 14.31 23.04 22.78 22.68 22.93 22.45 — 21.84 22.68 r.m.s. dev.(S) 1.3 1.5 1.4 0.3 1.5 0.6 2.5 1.6 - Distanc ' r e from centr f adjusteeo r 134 TABLE 3.9.4.9 VARIATION OF CALCULATED CU ACTIVITY IN THE VICINITY OF THE TUBE AND ROD ADJUSTER ( INTERPOLATED FROM REFERENCES 3, 4 AND 5 FOR RESPECTIVE MODELS ) r'(cm) Horizontal direction - Al Horizontal direction - A2 Vertical direction - N Meas. SHETAN COMESH BOXERS Meas. SHETAN COMESH BOXERS Meas. SHETAN COMESH BOXERS 1.01 15.05 15.81 14.74 15.58 13.25 13.45 12.27 13.13 15.87 16.65 15.52 16.59 1.81 16.71 16.73 16.85 17.31 14.73 14.23 14.08 14.59 17.66 17.62 17.73 18.44 2.61 17.32 17.42 17.44 17.90 15.02 14.83 14.56 15.02 18.33 18.30 18.35 19.04 3.61 17.58 17.66 17.45 17.84 14.68 14.77 14.53 14.81 18.30 18.54 18.36 18.83 3.81 17.79 17.79 17.79 17.79 14.82 14.82 14.82 14.82 18.72 18.72 18.72 18.72 4.81 19.16 19.79 19.05 19.72 15.57 15.53 15.64 15.59 20.27 21.30 20.04 20.74 5.81 20.02 20.20 19.55 20.60 15.56 15.28 15.64 15.59 21.18 21.40 20.57 21.71 6.81 20.50 20.60 20.01 20.58 14.83 14.61 15.38 14.80 21.72 21.72 21.06 21.87 7.81 20.94 20.91 20.34 20.78 22.00 22.12 21.41 22.21 8.81 21.19 21.01 20.61 20.96 22.50 22.34 21.69 22.44 9.81 21.37 21.06 20.82 21.11 22.80 22.52 21.91 22.63 10.81 21.42 21.13 21.00 21.25 22.74 22.70 22.10 22.81 11.81 20.70 21.19 21.15 21.36 22.95 22.87 22.26 22.98 12.81 21.69 21.26 21.29 21.45 23.29 23.05 22.42 23.14 13.81 22.06 21.33 21.44 21.51 23.43 23.22 22.59 23.30 14.31 22.20 21.37 21.51 21.54 23.47 23.31 22.67 23.37 r.rn.s. dev.OO 2.2 2.1 2.3 1.2 2.5 0.4 1.9 2.7 2.2 r' - Distance from centre of adjuster Core Analyses The core calculations have been performed using models A and B. Table 3.9.4.10 gives analysis of critical configurations of bare core and cores with adjusters. It is observed that the calculated k-effective values are very close to unity for both the models. To find the wort adjustersf ho same th , e cases have been analyzed without adjusters worte bees Th . hha n assessed using two recipes, one by subtracting the reactivities of the corresponding adjuster OUT and adjuster IN cases and the other by subtracting reactivity of bare core from that of the adjuste casesT founs i t OU rI . d thacalculatee th t d worth f Argentinso India-ThPd aan D are fairly close and are both less than measured ones by about 10-15 %. 3.9.4.4. Conclusions A comparison of analysis of ZED-2 experiments performed by Argentina and India- ThPD has been presented. Models with varying degree of sophistication for 1-D as well as celD fol3- r analysis have been used. Predictio f integrano l parameters k-in k-efd e an far f cells differenceD e 3- Th .d quitan homogenizeen D i sclos 1- r efo d cross section observes si d below 5 % in general. Comparison of Cu activity profiles around a fuel bundle and adjusters are fairly measuree closth o et d activities worthe Th . f adjustero s s calculate Argentiny db a and India-ThPD are reasonably close to each other but are both less than measured ones by about 10-1. 5% 3.10. TASK 10: POWER DISTRIBUTION CONTROL 3.10.1. Description This tas s intendeki teso capabilite dt th t f modellinyo effece gth f xenon-iodino t e dynamics on power distribution and the reactor regulating system response (i.e. zone controller level change and adjuster rod movement) to the changes in power distribution. 135 TABLE 3.9.4.10 RESULTS OF CORE SIMULATIONS - CRITICAL CONFIGURATIONS WITH ADJUSTE CORE TH ERN I Mod K-eff Rhol K-eff Rhol Worth of Adjusters Wort Adjusterf o h s Height (mk)(mk) Rho2-Rhol Rho2-RhoO Meas (cm) PUMA-CCArgen ) CEMESH (India) PUMA-C CEMESH PUMA-C CEMESH Worth No adj IN 215 500 1 00006 0 06 0 99913 -0 87 RhoO Horizontal adjusterN I s Adjuster -1 233 290 1 8 00044 0 8 0 99916 0 84 Adjuste2 - r 239 684 0 99877 1 23 Adjuster0 00 - 30 24 1 2 00000 0 2 0 99870 1 30 Adjuster -4 225 479 0 99882 1 18 Adjuster -5 227 000 0 99885 1 15 Adjuster3 05 - 60 23 0 99881 -1 19 Vertical adjusters IN Adjuster -1 236 057 0 99944 -0 56 Adjuster -2 238 347 0 99907 -0 93 Adjuste3 - r 236 078 1 00011 0 11 0 99898 -1 02 Adjuster5 09 - 44 22 0 99917 -0 83 Adjuste5 - r 225 208 0 99913 -0 87 Adjuster -6 227 729 0 99907 -0 93 Horizontal adjusters OUT Adjuster -1 233 290 1.00817 8 11 00734 7 29 7 62 8 13 8 04 8 16 9 33 Adjuster -2 236 849 00879 8 71 9 94 9 58 11 00 Adjuste3 - r 240 000 1 01084 10 73 01003 9 93 10 71 11 23 10 67 10 80 12 50 Adjuste4 - r 229 547 00395 3 93 5 11 4 80 5 41 Adjuster -5 227 000 00463 4 61 5 76 5 48 6 16 Adjuste6 - r 230 053 00597 5 93 7 12 6 80 7 60 Vertical adjusters OUT Adjuster -1 230 576 1 00620 6 16 6 72 7 03 8 03 Adjuste2 r- 238 347 1 00742 7 36 8 29 8 23 9 44 Adjuster -3 236 078 1 00931 9 23 1 00848 8 41 9 12 9 43 9 17 9 28 10 70 Adjuste4 - r 224 095 1 00332 3 30 4 13 4 17 4 74 Adjuster8 20 - 55 22 1 00382 3 81 4 68 4 68 5 33 Adjuster9 72 - 67 22 1 00495 4 93 5 86 5 80 6 62 3.10.1.1. Reactor model Bur zonp nu e specification lose regulatiof th As o n si n benchmark reactor cor composes ei homogeneouo tw f do s burnup zones. Inner Zone: Average cross section between 0 and 7600 MWD/T in the inner 124 channels. They are located in a centered 12x12 square except five forming a right angle in each corner (see Fig. 3.4.1) Outer Zone: Average cross section between 0 and 6400 MWD/T, formed by the rest of channels. Two group cross sections for the two zones are listed in Table 3.4.1. 136 Reactivity Devices Zone Control Compartments: There are fourteen zone control compartments as shown in Fig. 3.10.1. The level..of compartmente th l wateal n i r s changes simultaneousl controo t y globae th l l powee th f o r reactor, while a differential change in the water level of the compartments is effected to control average zonal powers proportionaA . l contro uses i thir dw fo s la lpurpose . Tha, is t the change in level in each compartment is proportional to the deviation from the average of the average zone power relative to its initial (t=0) value. There is no neutron kinetics simulation in this benchmark problem. So, the global power control is reduced to a criticality search calculation and the multiplication factor, Keff , is kept constant throughout the calculation. Xenon and iodine dynamics are, however, simulated. The average level of the zone control compartments is kept between 20 and 70 percent, while individual levels are allowed between 10 and 90 percent. Adjuster Rods: adjusteTher1 2 e ear r rods (see Fig. 3.4.1). The dividee yar d int onumbea groupf ro s called banks for convenient operation and for providing appropriate supplementary reactivity to the zone control system. An adjuster bank is completely and sequentially withdrawn/inserted whenever average zone level drifts beyon above dth e mentioned limits. Partial withdrawal/insertio adjusten a f no rt permitted banno s ki . Also e dynamicth , f o s adjuster bank movemen t simulateno s ti thin di s benchmark. Adjuste bankind ro r specifiegs i d in Table 3.10.1. Contro t usel no thin rod di e s ar sproblem theio S . r specification t givenno e .ar s orden I retaio t r n simplicity assumes i t i , d that only thermal group absorption cross section changes consequent to the change in zone water level and movement of adjuster rods. Incremental cross sections are limited to incremental thermal absorption cross sections only and equal E& = 6.5 x 10 cm"1 for zone controller compartments and 2.5 x 10 cm"1 for adjuster rods with both values defined ove crosra s sectional are 57.1f a49.o x (sem m 5c c e Fig.3.10.2). n appropriatA e mesh size coul e finite chosedb th r e nfo differenc e calculations. However, the generation of individual bundle power limits the maximum mesh dimension to one fuel bundle length in the axial direction and a mesh dimension of one-pitch length in radiae eitheth f o rl direction respectively calculatioe th f I . performes ni d wit hfinea r mesh size, e.g. meshe4 2 ,axia e th n lsi direction (1/2 bundle length mesh) powee ,th fuli rh l bundle f powere calculateo b o i t m sf adjacen o ssu s da meshes2 t e centre reactoth Th . f eo s i r bundleassumeh 7t mesbetweee e d b Th an .o hd t h identification6t standara f o s ni d one. 3.10.1.2. Problem specification Reacto assumes i r fuloperatine % b l 0 o powedt 10 t ga r with equilibrium xenod nan t tuniodin a ste0 d et= preductio a ean ful % l 0 powe7 o nt effectes i r d (Fig. 3.10.3). After continuous operatio t thina s power leve threr fo l e hours powee th , raises i r % d0 bac10 o kt full power in four successive time intervals of thirty minutes each. The power raise in the first and fourth step is 5 % full power while 10 % full power is increased in the second and 137 Channel Column Designations 1 2 3 4156781 2 0 2 9 1 8 91 7 1 6 t 5 1 4 11 3 Channel Row Designations A / B / C / / FIG. 3.10.1 600MWe Reactor Face View Showing Zone Controller Zone Specifications 138 TABLE 3.10.1 ADJUSTER ROD BANKING Bank No . S. No. of Adjuster Rods 1 I/ , , 117 15 , 21 2 2, 6, 18 3 4, 160 2 , 4 8, 9, 13, 14 5 3, 19 6 5, 17 7 10, 12 Bank Withdrawa throug1 : l sequentiallh7 y Bank Insertio throug7 : n sequentiall1 h y 1 23 A 5678910 11 12 13 R 15 151718 19 2021 2223 242526 27 28 1 1 0 1 3 £5 9 2 1 78 6 12 13 U 15 1B 17 18 19 20 21 22 24 23 22 21 20 19 18 21-2 C -1 23-5 Cr**2> Z6 -7 17 16 I A 1 A 2 A3 A A5 A6 A T CD 15 2 U r h 13 ^ r— ^ - A8 A9 A' 0 A' 1 A1 2 A13 A14 rU>J 12 u. 11 10 A15 A16 AI 7 A 8 A' S A:>0 A: 1 9 1 h- 8 ? Z8-9 C ZU H2 — C4 Î13-K 7 G - 5 4 3. 2 1 ' F/G. 5. Viewp 10.To 2 of Reactivity Devices after Homogenisation ( Note : A, C and Z Refer to Adjuster Rods , Control Rods and Zone Control Compartments Respectively) 139 third step, respectively. In this way, the reactor power is 100 % full power at time t=4.5 hr and it is assumed to remain constant thereafter. The transient is followed for 6 hours. Step reductio powen ni s assumei r instantaneoue b o t d s dynamicit t simulated no san s i ss A d mentioned earlier, reactivity is kept constant by operating zone control compartments. Adjuster operatee whesar d average an ns th d a e zone level drifts beyon above dth e mentioned limits. Various indicators of power distribution such as maximum channel power, maximum bundle power, spatial tilt differenn si t directions etc generatee .ar t eacda h time step. Tilts are normalized to the total reactor power. A top to bottom spatial tilt is defined as the difference bottoe betwee thahalth d p n to man ti f reacto powee e e nhalth th th f n fo i r r divided by total reactor power. Result r comparisofo s e compilenar tima t a de interva f fifteeo l n minutes each [89]. 100% 90% - 70% ^ 60% 50%————'————————————————————————————————————————'———— 0 6. 5 5. 0 5. 5 4. 0 4. 5 3. 0 3. 5 2. 0 2. 5 1. 0 1. 5 0. 0.0 Time (hr) FIG. 3.10.3 Power Trajectory 3.10.1.3. Additional data for simulation of xenon and iodine dynamics Microscopic cell averaged cross-section for thermal neutron absorption in xenon aXe = 1.2x 10-18cm2 Fission yield of xenon 7Xe = 0.006 Fissio 0.06= n2 yiel iodinf do , e7 1 Decay constant of xenon XXe = 0.0756 hr" 2.1= 010-x 5 s'1 Decay constant of iodine \ = 0.1058 hr"1 2.9= 4IQ'x 1 s' 5 140 3.10.2. Modelling 3.10.2.1. Canada benchmare Th k proble solves mwa d usin POWDERPUFe gth XEMAd San X modules OHRFSe ofth P code t solveI . three-dimensionae sth grouo ltw p finite-difference e forth f mo neutron diffusion equation over a core model. The POWDERPUFS module provides burnup- dependent fuel cell cross sections and xenon parameters. The XEMAX module includes in core th e calculation :(1explicin a ) t modellin xenof go iodind nan e distribution variationn si each fuel bundle (with each bundle thermal absorption cross sections adjusted as a result of xenon variations) modellina ) (2 d bul f ;an gspatiad o kan l control (i.e., differential variations zonn i e compartment levels proportiona bulo t l variatiokK relativd nan zono et e power variation are effected iteratively during flux calculations: 20 iterations are carried out zone level adjustede sar anothed an , iteration0 2 r carriee sar d out, etc). Core mesh spacings and notch dimensions are given in [90]. The parameters used in Canadiae th n solution deviated fro benchmare mth k specificatio respectso tw n ni . First, rather than inputin cele gth l cross section Tabln si e 3.4.1, cross sections were generated directly from the POWDERPUFS module for the reference 37-element fuel bundle defined in Task-1, at exit burnups of 7600 and 6400 MWd/T for the inner and outer burnup zones. Secondly, the xenon-related parameters are also generated from the POWDERPUFS module which are as follows: Thermal neutron absorption in xenon crxe Westcotx 10'2 x 15 8cm 3. t= fuel flux = 1.3 x 10'18 cm2 x Thermal cell flux. Fission yield of xenon yXe = 0.006 Fission yield of iodine yj = 0.062 1 Decay 0.07= constan6hr xenof o te X nX 2.10= x lO'5^1 Decay constant of iodine \ =0.105 hr"1 = 2.92 x IG'5 s'1 Calculational time-steps varie d5 minutes 1 betwee d e variabilitan Th . 5 n s wa y required in order to manually search for the time at which adjusters would be withdrawn or inserted. The initial (t=0) level of all 14 zone controllers was assumed to be 45 %. 3.10.2.2. India problee Th solves mwa d usin FEMXEe gth N code. FEMXE three-dimensionaa Ns i l two-group finite-differenc esimulatioe codth r efo xenof no n transients t carrieI . t direcsou t iterative criticality searc averagn ho e zone level, while differential change individuan si l zone level made sar correco et powee th t r distribution thir sFo . problem unifora , m mesh sizf eo radiae th n axiae i l28. planth m 24.7l6n d c i direction ean m 5c , respectively useds o wa , N . iterations were done to restore the transient power distribution to the reference power distribution. Instead a relativel, y short time ste f 3.7o p 5 minute e s choseth wa sr fo n simulation of transient [91]. 141 3.10.3. Results A brief summary of results is presented in Table 3.10.2 through Table 3.10.5. As per benchmark specifications results were require uniforn a t da m time interva minutes5 1 f o l . However, in this summary report results are given in a comparative format only after a tune period of 30 minutes each. Details regarding method of calculation and results at short time intervals coul e foundb Refn i d . [90, 91] Tabl.n I e 3.10.2 have w , e maximum channel power, maximum bundle powe spatiad an r l tilt t eaca s h time intervals. Table 3.10.d 3an 3.10.4 give variation individuae th n si l zone powe leveld an r , respectively transiene th s a , t progresses. Average zone leve alss i l o listed. Finally Tabln i , e 3.10. have time 5w eth e elapsed since the start of the transient at which adjuster banks are operated. It can be seen from these results that generally ther s gooi e d agreement between Canadia Indiad nan n results. Some of the differences particularly in the adjuster bank withdrawal /insertion tunes coul e attributedb e manneth o dt whicn i r e xenohth n parameter burnud an s p date ar a generate Canadiae th i dh n model t resultI . differenn si t xenon bul spatiad kan l reactivitien si their simulation. TABLE 3.10.2 CHANNEL POWERS, BUNDLE POWER SPATIAD SAN L TILTS Time Reactor Maximum Channel Maximum Bundle Top/Bottom Left/Right Front/Back (hr) Power ( %FP ) Power (kW) Power (kW) Tilt (%) Til) (% tTil ) (% t 0.0 100 .0 6653.0 E-10 801.2 E-ll- 6 2.4 0.1 0 .0 . India 6815. J-10 822.8 E-ll- 6 1.64 0.00 0. 0 Canada 0.50 70 .0 4793.9 E-ll 563.1 E-ll- 7 2.3 0.0 0. 1 4850. E-ll 578.6 E-ll- 6 1.71 0.04 -0.01 1.00 70 .0 4708.4 E-12 564.0 E-12- 7 2.0 0.1 -0.3 4841. P-ll 576.8 E-ll- 6 1.16 0.00 -0.08 1.50 70 .0 4830.4 M- 8 576.2 G-ll- 7 2.0 0.1 -0.2 5059. 0-11 604.9 N-ll- 6 1.27 0.00 -0.01 2.00 70 .0 4856.2 P-ll 581.8 N-ll- 7 1.6 0.1 0.0 5078. P-ll 610.5 N-ll- 6 1.02 0.00 -0.01 2.50 70 .0 4878.0 P-ll 587.2 N-ll- 7 1.6 0.1 0.0 5105. 0-11 616.8 N-ll- 6 0.87 0.00 0.00 3.00 75 .0 5245.4 M- 8 634.4 N-ll- 7 1.6 0.1 0.0 5497. 0-11 666.9 N-ll- 6 0.96 0.00 0.00 3.50 85 .0 5986.4 M- 8 733.4 M-ll- 6 2.5 0.1 0.1 6313. 0-11 774.0 M-ll- 6 1.42 0.00 0.00 4.00 95 .0 6799.9 M- 8 853.9 M-ll- 7 2.1 0.1 0.0 7139. 0-11 902.9 M-ll- 7 1.43 0.00 0.00 4.50 100 .0 6907.0 L-15 861.4 E-12- 7 1.9 -0.1 0.0 7212. 0-11 908.3 M-ll- 6 1.45 0.00 0. 10 5.00 100 .0 6676.6 E-13 852.5 E-12- 6 2.0 -0.1 0.0 6920. M-14 876.8 E-ll- 6 1.47 0.00 0.00 5.50 100..0 6805.0 E-10 837.7 E-ll- 7 2.9 0.1 0.0 6826. L-14 843.8 E-ll- 6 1.57 0.00 0.00 6.00 100..0 6652.5 E-13 819.8 E-ll- 7 2.4 0.1 0.0 6840. K- 9 841.7 E-ll- 6 1.67 0.00 0.00 FP Full Power 142 TABLE 3.10.3 INDIVIDUAL-ZONE POWERS Time Power(%FP) Zone Powe( r Zone 1 through 14) in MWth (hr) 1 234 5 6 7 8 9 10 11 12 13 14 0.0 100. 155.6 150.3 159.9 115.3 147.5 155.4 149.8 155.7 150.4 160.0 115.4 147.4 155.5 149.8 India 153.3 149.7 158.7 120.9 149.9 153.3 149.7 153.3 149.7 158.7 120.9 149.9 153.3 149.7 Canada 0.50 70. 108.8 105.4 111.6 81.4 103.2 108.7 105.3 108.8 105.1 111.5 81.4 102.9 108.4105.1 107.4 104.8 111.1 84.7 104.9 107.3 104.7 107.4 104.8 111.1 84.7 104.9 107.3104.7 1.00 70. 109.0 105.3 109.5 80.9 103.1 109.0 105.0 109.5 105.6 110.4 81.9 103.7 109.5105.3 107.2 105.2 109.1 85.0 105.3 107.2 105.2 107.2 105.2 109.1 85.0 105.3 107.2105.2 1.50 70. 108.4 104.4 110.1 83.9 103.0 108.3 104.1 109.0 105.3 110.3 84.2 102.7 108.9105.0 107.1 104.4 109.5 87.8 104.6 107.0 104.4 107.1 104.4 109.5 87.8 104.6 107.0104.4 2.00 70. 108.8 105.3 108.5 84.2 103.3 108.7 104.9 108.9 105.3 108.4 84.3 103.2 108.8104.9 107.2 104.7 108.1 88.0 104.8 107.2 104.7 107.2 104.7 108.1 88.0 104.8 107.2104.7 2.50 70. 109.0 105.3 107.8 84.5 103.3 108.9 105.0 109.1 105.4 107.8 84.6 103.2 108.9 105.0 107.2 104.8 107.5 88.4 104.9 107.2 104.8 107.2 104.8 107.5 88.4 104.9 107.104.2 8 3.00 75. 116.7 112.7 115.5 91.0 110.6 116.6 112.3 116.8 112.8 115.5 91.0 110.5 116.7112.3 114.8 112.1 115.3 95.1 112.2 114.9 112.1 114.8 112.1 115.3 95.1 112.2 114.112.9 1 3.50 85. 131.8 126.4 134.9 104.7 124.2 131.6 126.0 131.8 126.3 134.9 104.5 123.3 131.6125.8 129.3 126.0 133.5 109.5 127.4 129.3 126.0 129.3 126.0 133.5 109.5 127.4 129.3126.0 4.00 95. 145.8 141.0 150.4 119.5 139.4 145.7 140.6 145.8 141.1 150.4 119.4 139.2 145.6140.6 143.8 140.4 148.8 125.4 140.7 143.8 140.4 143.8 140.4 148.8 125.4 140.7 143.8140.4 4.50 100. 154.1 149.9 158.1 120.8 147.3 154.5 149.9 153.4 148.8 158.7 122.3 147.4 154.2 148.8 152.2 148.6 157.4 128.6 148.8 152.2 148.6 152.2 148.6 157.4 128.6 148.8 152.2 148.6 5.00 100. 154.8 149.3 160.0 116.2 148.5 155.5 150.3 155.2 150.5 158.6 116.1 148.0 155.2150.0 153.1 149.5 158.5 122.2 149.6 153.1 149.5 153.1 149.5 158.5 122.2 149.6 153.1 149.5 5.50 100. 156.0 149.4 162.9 113.4 147.5 155.7 148.8 156.1 149.5 163.0 113.5 147.3 155.8 148.9 153.3 149.7 158.7 120.9 149.9 153.3 149.7 153.3 149.7 158.7 120.9 149.9 153.3 149.7 6.00 100. 155.6 150.3 160.0 114.9 147.9 155.4 149.9 155.6 150.4 160.1 115.0 147.8 155.149.4 8 153.4 149.6 158.8 120.8 149.8 153.4 149.6 153.4 149.6 158.8 120.8 149.8 153.4149.6 TABLE 3.10.4 INDIVIDUA AVERAGD LAN E ZONE LEVELS Time Power(%FP) Individual Zone LevelCZon Averagd througe1 an ) eh14 Zone Level (hr) 1 2 3 4 5 6 7 8 9 10 11 12 13 14 Ave. 0.0 100. 0.45 0.45 0.45 0.45 0.45 0.45 0.45 0.45 0.45 0.45 0.45 0.45 0.45 0.450.450 India 0.45 0.45 0.45 0.45 0.45 0.45 0.45 0.45 0.45 0.45 0.45 0.45 0.45 0.45 0.450 Canada 0.50 70. 0.23 0.25 0.17 0.32 0.22 0.23 0.26 0.22 0.24 0.16 0.31 0.22 0.22 0.260.240 0.26 0.28 0.12 0.39 0.16 0.27 0.28 0.26 0.28 0.12 0.39 0.16 0.27 0.20.258 0 1.00 70. 0.30 0.35 0.10 0.62 0.19 0.30 0.36 0.14 0.18 0.11 0.71 0.24 0.14 0.10.279 6 0.29 0.37 0.10 0.59 0.10 0.29 0.37 0.10 0.15 0.10 0.73 0.18 0.10 0.10.255 9 1.50 70. 0.26 0.37 0.10 0.90 0.28 0.26 0.38 0.25 0.33 0.10 0.90 0.36 0.25 0.340.351 0.31 0.39 0.10 0.90 0.28 0.31 0.39 0.31 0.39 0.10 0.90 0.28 0.31 0.390.380 2.00 70. 0.21 0.34 0.10 0.90 0.27 0.21 0.35 0.21 0.32 0.10 0.90 0.33 0.21 0.330.327 0.24 0.34 0.10 0.90 0.23 0.24 0.34 0.24 0.34 0.10 0.90 0.23 0.24 0.340.341 2.50 70. 0.20 0.34 0.10 0.90 0.26 0.20 0.35 0.20 0.32 0.10 0.90 0.32 0.20 0.30.323 3 0.23 0.34 0.10 0.90 0.22 0.23 0.34 0.22 0.34 0.10 0.90 0.22 0.22 0.340.336 3.00 75. 0.21 0.35 0.10 0.90 0.27 0.21 0.35 0.21 0.33 0.10 0.90 0.33 0.21 0.340.330 0.24 0.36 0.10 0.90 0.24 0.24 0.36 0.24 0.36 0.10 0.90 0.24 0.24 0.360.350 3.50 85. 0.33 0.41 0.11 0.90 0.35 0.33 0.42 0.33 0.41 0.10 0.90 0.38 0.33 0.420.401 0.40 0.47 0.12 0.90 0.37 0.40 0.47 0.40 0.47 0.12 0.90 0.37 0.40 0.470.449 4.00 95. 0.57 0.65 0.33 0.90 0.64 0.56 0.66 0.57 0.65 0.33 0.90 0.65 0.57 0.60.616 6 0.63 0.71 0.38 0.90 0.68 0.63 0.71 0.63 0.71 0.38 0.90 0.68 0.63 0.70.661 6 4.50 100. 0.70 0.77 0.38 0.90 0.67 0.69 0.78 0.51 0.56 0.41 0.90 0.68 0.51 0.50.647 2 0.73 0.88 0.40 0.90 0.69 0.73 0.88 0.57 0.62 0.49 0.90 0.74 0.57 0.620.693 5.00 100. 0.57 0.61 0.39 0.88 0.55 0.57 0.63 0.57 0.60 0.40 0.88 0.59 0.56 0.610.600 0.62 0.68 0.42 0.90 0.61 0.62 0.68 0.63 0.69 0.42 0.90 0.61 0.63 0.690.649 5.50 100. 0.50 0.50 0.47 0.63 0.54 0.50 0.51 0.51 0.50 0.47 0.62 0.54 0.50 0.50.521 2 0.55 0.56 0.53 0.60 0.57 0.55 0.56 0.55 0.56 0.53 0.60 0.57 0.55 0.50.556 8 6.00 100. 0.54 0.54 0.58 0.56 0.63 0.54 0.54 0.55 0.54 0.57 0.55 0.63 0.54 0.50.565 4 0.59 0.60 0.57 0.58 0.64 0.59 0.60 0.59 0.60 0.57 0.58 0.63 0.59 0.60.590 4 143 TABLE 3.10.5 OPERATIO ADJUSTEF NO BANKD RRO S R BanA # k Time(hr) Withdrawn Inserted 1 0.31 5.37 India 0.32 5.25 Canada 2 0.56 4.56 0.60 4.52 3 1.25 4.12 1.25 4.05 banAR nA withdraws i k sequencn i n throug1 e whe7 h n Average Zone Level drift Totaf so belo% l 0 Capacit2 w y An AR bank is inserted in sequence 7 through 1 when Average Zone Level drifts above 70 % of Total Capacity A simple point-kinetics calculation was carried for the power trajectory in Fig. 3.10.3, usin inpus ga t bot Indiae h th xeno d nan n related parameters founs wa dt I . thae th t maximum difference in predicted xenon reactivity between the two input sets was 0.34 mk, and the difference at the end of the trajectory (6 hours) was 0.13 mk. These correspond to maximua m differenc averagn i e e differenczond % en e3 e levea th t d roughlf ea o l an % y6 ohours6 f , which agree closely wit differencee hth Fign i s . 3.10.4. Fig. 3.10.5 plot maximue sth m channel power simulationso s tw (MCPs e th r r )fo Fo . tune steps corresponding to moments of core power variation, the results have been normalized in the figure so as to correspond to the same power. It is found that the MCP for the two simulations differ by 160 kw initially, and the difference remains in the range of 160- througw k 0 h20 transient. 3.10.4. Conclusions Both the computer code packages agree reasonably well in their ability to simulate xenon transient consequend san actioS thir RR tn fo s simpl slod ean w transient. 144 80°/c 70% 60% - 20% h 10% 0 6. 5 5. 0 5. 5 4. 0 4. 5 3. 0 5 3. 0. 5 0 2. 0. 0 2. 5 1. 1.0 Time (hr) Canada India FIG. 3.10.4 Comparison Averageof Zone Levels ON 4.5 0 6. 5 5. 0 5. 5 4. 0 4. 5 3. 0 3. 5 2. 0 2. 5 1. 0 1. 5 0. 0 0. Time (hr) Canada India FIG. 3J0.5 Comparison of Maximum Channel Powers (Mw) 4. SUMMARY AND RECOMMENDATIONS develomaie o t s Th npresene wa goa pth P f benchmarklo CR t s whic appropriate har e to chec improvd kan fuee eth l management computer code package theid san r procedured san also to provide countries that are being introduced to nuclear energy with sample problems d resultan s usefu o start l t with reactor physics calculation n PHWi s R type reactors. Therefore, benchmark specifications were established which includef realistio t se cda data for in-core fuel management codes. The results of measurements were also provided to verify compard an e with these parameters calculate in-core th y db e fuel management code codr so e packages. The benchmark problems for In-core fuel management of PHWRs initially contained three task lattic) (i s: e cell calculations; (ii) supercell calculation reactivitr sfo y devices; (iii) core calculations. The research co-ordination meeting held at Argentina in 1990, recommended seven more tasks for their inclusion in the present CRP : (iv) a loss of regulation accident analysis (LORA); (v) evaluation of void reactivity with respect to different fuel isotopes; (vi) basic characteristics of a hypothetical core loaded with MOX fuel; (vii) evaluation of decay heat functioa s a timf no e after shutdown; (viii idealizen a ) d formulatio f secondarno y shutdown system (SDS-2); (ix) analysis of various experiments and (x) flux power tilts and their on-line control. countriesx Si , namely, Argentina, Canada, India, Republi f Koreaco , Pakistad nan Romania participated in these benchmarks for PHWRs. From India, two groups belonging to Reactor Engineering Division (India-RED) and Theoretical Physics Division (India-ThPD) participated. Table 1.1 gives an indication of the participation of various countries in the different benchmarke taskth f so . 4.1. SUMMARY 4.1.1. Task 1: Lattice cell benchmark problems Task 1 consisted of solving seven lattice cell problems for 37-rod fuel cluster. The first was the reference case under normal operating conditions. The next two problems were defined to compute the void effect and the fuel temperature effect. The next three problems were defined to calculate the effects associated with the long term perturbations: changes in moderator isotopic purity, addition of poison in the moderator, increase in the weight of e naturath l uraniu r bundlempe . Finally lattice cell calculation slightla r sfo y enriched uranium lattice defineds celwa l . countriex si Ale th l s participate lattice th n dei cell benchmark problem sArgentina: , Canada, India, Korea, Pakistan, Romania. Five codes were used: PPV (Argentina, Canada, Romania), WIMS (Argentina, Korea, Pakistan, Romania), CLUB (India-ThPD), CLIMAX (India-ThPD) and RHEA (India-RED). The PPV is a recipe based four-factor formula code specifically adjusted for natural uranium clusters for CANDU reactors. RHEA is a five group code with a diffusion theory formalism othere th multigroue l sar Al . p transport theory codes. 147 Although the results of the five codes were rather similar, some discrepancies relatevoie th do t deffect , boron moderator reactivit enriched an y d uranium cases were observed. The void effects predicted by simple codes were different from those calculated by multigroup transport theory codes e voiTh d. induced reactivit y formeb y r methods significantly decreased with increasing burnup unlike transport theory estimations. This led to a more detailed analysis of components of void effect as a separate Task 5. The boron moderator induced reactivity estimations by PPV were nearly 10 % higher than WIMS predictions while they highe wer% 8 e1 r tha CLUe nth B results. The enriched fuel lattice cell calculations by all codes other than PPV which is not meant for such lattices agreed well. 4.1.2. Tas Benchmar: k2 superceln ko l calculations In PHWRs, the reactivity devices (RDs) are generally located perpendicular to the cele th lfuel t homogenizeorden I ge . o t r d parameter presence th n si f suceo hreactivita y consideo t s devicha three ra eon e dimensional cell called 'Supercell'. Three different kinds of reactivity devices were considere thir dfo s benchmark, namely adjuste) (a , r rod liqui) (b s d zone controllers and (c) shut off rods. The supercell benchmark problem was analyzed by four countries namely Argentina, Canada, IndiRomaniad an a varietA . f computeo y r codes were use : dMULTICEL L (Argentina, Canad d Romania)an a , SHETAN (Argentina), BOXER-3 (India-ThPD), PHANTOM-CEMESH (India-ThPD), 3D-FAST (India-ThPD CALd )an C (India-RED)e Th . computer codes SHETAN and BOXER-3 are based on transport theory and remaining codes basee ar diffusion do n theory with boundary condition calculated from transport theory. It was seen that the maximum difference in reactivity was in the prediction of empty zone controller tube case. The reaction rates predictions for moderation, absorption and production for all the cases were within aboustandar% 10 t d deviation. 4.1.3. Task 3: The core benchmark The Task- relates core th 3wa e o dt calculation PHWsimplifiea W r M fo s R0 60 d reactor containing a limited number of well defined simple reactivity devices. In all, eight cases were analyzed. In the first three cases, the core contained irradiated fuel bundles typica operatinn a f o l g CANDU reactor (Embalse, Argentina) othee Th . r five cases were essentially three-dimensional simulations to find worth of reactivity devices in fresh core. Five countries participate thin di s benchmark. Various computer codese useth y db different participants wer ePUMA-: C (Argentina), OHRFSP (Canada), DIMENTRI (India- ThPD), TRIVENI (India-ThPD), CEMESH (India-ThPD), CITATION (Pakistan), FMDP (Romania), CHEBY (Romania). All the codes are based on diffusion theory. 148 basiAe sth c metho calculatiof do same-fos codesne wa th differencee l th al ,r e th n i s results were basically due to the lattice methods used for calculating the two group cross section largese th s- t differenc observes caswa e voidef th eo n di d core. 4.1.4. Task 4: Loss of regulation benchmark problem for PHWRs This benchmark was the simulation of loss of reactivity control event initiated by the simultaneous drainin sevee th f ngo zon ereactoe controllersidth e f eon o re (onth n eso lattice pitch in two seconds linearly in time) followed by the insertion of 4 shutoff rods. This transient was analyzed by three countries, namely, Argentina, India-ThPD and Romania with their codes PUMA-C, 3D-FAS CERBERUd Tan S respectively codee th l s Al .wer e based on improved quasi-static method resulte Th . s wer goon ei d agreement with each othes a r expected sinc methode eth f solutioso n used were similar. 4.1.5. Task 5: Influence of isotopes on void reactivity As was noticed during the analysis of Task-1 there was a large difference in the value f voio s d induced reactivity calculate multigroue th y db p transport theory coded an s PPV code. In order to understand the reasons for the difference, a benchmark was formulated in which the influence of various isotopes on void induced reactivity was studied. Participants from four countries took par thin i t s benchmark problem Argentina: , India, Pakistad nan Romania. Three codes were used: WIMS (Argentina, India-RED, Pakistan, Romania), POWDERPUFS-V (Argentina, Romania CLUd an ) B (India-ThPD). estimatione codee Th th l sal indicatef so d thavoie th t d induced reactivity valus ewa practically dictate U23e Pu23th d y 5an db 9 number densities agreemente Th . s were better at lower burnups than at higher values. It was further observed that the void related fuel neutron spectrum perturbation playe dsignificana t rol voin ei d effect estimation. 4.1.6. Task 6: Mixed loading The aim was to analyze a core with MOX fuel. The benchmark considered two types of fuels, namely, naturainnee th fue2 fuen X ri lUO lregionl clusterMO d outee th san i rsh regions. This task received rune submissions from four participants froe on m, Pakistad nan India-RED respectively, four from Romani thred aan e from India-ThPD. l methodAl s appeare locatioe agreo dt th maximue n th e o f no m bundl channed ean l powers. It was further seen that axial mesh structure and convergence criteria influenced in predictio f poweno r distribution. 4.1.7. Task 7: Calculation of the fission product inventories and their decay heat This benchmark is related to fission products inventories and their decay heat rates standare foth r d core studied (cf. Tas wit) k3 bundle hth e power distribution taken froe mth Embalse NPP. Three countries, namely, Argentina, Pakistan and Romania participated in this benchmark, all of them employing the code ORIGEN. There were certain differences in the results which were attributed to the fact that the participants had used different burnup values and adopted different approaches to get core aggregates. Keepin valuee gth spectraf so l indice specifid san c power same Argentineae th , n 149 and Pakistani results were in very close agreement "with each other. The minor variations ma attributee yb difference th o dt neutronin i e c data library. 4.1.8. Task 8: Secondary shut down system This benchmar r evaluatiofo s secondare kwa th f no y shutdown system (SDS2r fo ) idealizen a d mode f liquio l d poison injection intmoderatoe oth r through nozzles alone gth horizontal tubes. This task was carried out only by the two Indian participants with different models for poison jets at supercell levels. The complex problem of modelling the SDS2 has been tackle methody db s with varying degre sophisticatiof eo supercele th t na l level, like diffusion theory, transport theory as well as Monte Carlo methods. While the Monte Carlo method simulated the wedge shape of the poison jets, others used equivalent rectangular shapes. The agreements were still quite good. When the poison propagated sufficiently, regardles e simulationth f o s s employin a gsingl e poiso ne abov th con r eo e wedg/ e rectangular shapes difference th , t reactivitne n i e y values narrowe t corda e level. bescoulw e ho intene on t de Th modethif se o t o st idealizee tass th l kwa d poison jets thes a y diffuse throug e moderatorhth . Howeve actuae th r l geometr poisoe th f yo n jetn si a practical core could be quite complex which requires detailed off-pile experimental results for modelling. The participants opined that from the safety angle the initial addition rate of negative reactivit dynamie th d yan c characteristic vere ar sy crucial. 4.1.9. Task 9: Experimental validation s regardin e Taswa Th 9 ke validatio th g f codeo n y analyzinb s f go fout se r experiments. 4.1.9.1. Task 9.1: ZED-2 experiments clusters withrod 37 This experiment was regarding the lattice measurements done with 37-rod fuel cluster at ZED-2 reactor in Canada [75]. The experiment was performed at only one lattice pitch of 28.5 (squarem 8c ) using heav coolantss a yr wateai d . an rThi s experimen analyzes wa t y db Argentina, India-ThP Romanid Dan a usin codee gth s WIMS (Argentina, Romaniad an ) CLUB (India-ThPD). s observewa t I d thae fasth tt fission ratis welowa l predicte boty db codese hth . However, there was relatively more error in the WIMS results in predicting its spatial distribution. Other reaction rates like relative conversion ratio, U-235 fission rate, copper activities and lutetium-manganese ratio were predicted well by both the codes. 4.1.9.2. Task 9.2: ZED-2 experiments clusters withrod 28 This experiments was regarding the lattice measurements done with 28-rod fuel cluster at ZED-2 reactor in Canada [76,77]. The experiments were performed at a number of lattice pitches using heavy water and air (He) as coolants. The results of analysis of these experiments were submitted by India-ThPD only using the computer codes CLUB and CLIMAX. Romania presented the results of analysis of experiments with 19-rod cluster instead of 28-rod cluster using WIMS code. observes wa t I d thaK-ef e predicte s computee th t th wa f l al dy rb withi codesk m n6 . Fast fission rati predictes owa CLUy db withiunderpredicted % B an 4 n3- 11-1y db y b 8% 150 CLIMA r 28-roXfo d cluster experiments r 19-roFo . d cluster experiments WIMe th , S code . underpredicte% 6 largs 1 a erroe s s th ea wa d r an t di initiae Th l conversion ratiunderpredictes owa codese th relative l Th al . y db e neutron density distribution was underpredicted by 2-3 % by the computer code CLUB. 4.1.9.3 Task 9.3: ZED-2 experiments on 19 rod clusters This experimen regardins twa analysie gth f isotopiso c compositio f fuen o variout a l s burnups [78]. The experiment was analyzed by two countries, namely, India-ThPD and Argentina using the computer codes CLUB and WIMS respectively. The relative concentratio e bundlth r f efo variouo n wer u P esd predictean isotope U f o sd within reasonable error by both the codes. The error in the prediction of burnup by both the codes was of the order of 100 MWD/T which can be considered very small. 4.1.9.4 Task 9.4: Adjuster rod experiments in ZED-2 This experimen measuremente th s twa s done with stainless steel adjuster fore th m n i s of rods and tubes of various sizes in ZED-2 reactor [79]. Reactivity worths were obtained from critical height and level coefficient of reactivity measurements. Flux distributions on and around the absorbers and throughout the reactor core were obtained from the activity of Cu and In foils. The experiments were analyzed by Argentina and India-ThPD. Comparison of Cu activity profiles around a fuel bundle and adjusters are fairly close e measureth o t d activities worthe Th . adjusterf so s calculate Argentiny db India-ThPd aan D are reasonably close to each other but both are less than measured ones by about 10-15%. 4.1.10. Task 10: Power distribution control This task was intended to test the capability of modelling the effect of xenon-iodine dynamics on power distribution and the reactor regulating system response (i.e. zone controller level chang adjusted ean movementd ro r changee th o t ) powe n si r distribution. o countrieTw s participate n thii d s benchmark, India-ThP d CanadaDan . There were reasonable agreements in their evaluations. Differences if any were due to the manner in whic xenoe hth n parameter burnud san p data were generated. 4.2. RECOMMENDATIONS Base resultn do f abovso e benchmark desird san f participanteo furtheo st r improve techniquese th followine th , g recommendation madee b n .sca lattice th t eA leve mose th l . t1 important effect which needs further detailed stude th s yi spectrum effec voin o t d reactivity. superceln I . l2 calculations mose th , t challenging tas relates ki simulatioo dt shapef no s and variatio f poisono n concentratio e poisoth n i nn jet f secondaro s y shutdown system. A more realistic benchmark based on experiments should be proposed for this case. 3. The benchmark for LORA was rather a mild transient with no drastic changes in spatial power distributions. Henc eLORa A considering stronger spatial effecte du s 151 to the asymmetric insertion of reactivities "with temperature feedbacks could be considered. The "shutoff rods" considered in the standard benchmark were really the solid control absorber typicaa n i s l CAND reactor0 U60 realitn I . y8 2 ther e ear shutoff rods constituting the Shut Down System -1 (SDS-1) in these reactors. For acciden strongese t analysith f o o t stw rod availablt assumee sno ar e b o det introducing appreciable flux tilts. Such models should als e considereob n futuri d e analysis. Incidentl partiaya l LOG A (half core voiding) could als modellede ob . simultaneouA . 4 s ruptur pressurf eo e tub calandrid ean a tube leadin LOCAgo t , addition of coolant water into moderator and consequent dilution of moderator (originally containing some boron) with unavailability of a few shutoff rods is another interesting case that could be taken up for analysis. 5. The power distribution control transient considered was also a relatively simple one and the transient was considerably slow. In order to bring out important features of control algorithm one can choose an off-normal initial core configuration such as one with significantly different zone levels, unavailability of one or two zone compartment core th tilea n i powetd san r distribution etc relativel.A y quick power cycling may be introduced to involve frequent movement of different reactivity devices resultin changina n gi g spatial distributio f xenonno naturaA . l extensiof no this problem is modelling of flux mapping system and its integration in the detailed simulatio f Reactoo n r Regulating System particulan (I . e fluth rx tilt incidenn i t Pickering NGS, Canada, coul modellee db d [97]). 6. In PHWRs with on-line flux mapping systems based on distributed in-core SPN detectors e fluth ,x distributio s evaluateni d basemodaa n do l synthesis technique. These modal methods are also employed in schemes for real time follow-up of the cores, or as an operator aid for channel selection during refuelling etc. The modes employed are the fundamental and a number of higher harmonics. There are many techniques employe r evaluatindfo g them benchmarA . calculato kt compard ean e them will be quite useful. . 7 Analysi variouf so s fuel cycle s, slightl sucTh s ha y enriched uranium- (SEUu P d )an recycling concept anothes si r area requiring multi-national evaluations. 8. Inclusion of more experiments for analysis could be taken up, such as Th, Th-MOX and clusters with burnable poison etc. 9. Employing higher order nodal codes for core optimization is another interesting area for detailed studies. 152 REFERENCES [I] In-core Fuel Management Programs for Nuclear Power Reactors, Final Report of CRP Coden o s Adaptabl Smaleo t Mediur lo m Size Computers, IAEA-TECDOC-314, Vienna, 1984. [2] Consultants Report on In-Core Fuel Management Aspects of Nuclear Power Reactors and Related Core Physical Matters, Doc. IFM-6, Vienna, December 1987. [3] Working Document of the CRP on In-Core Cuel Management Benchmarks of PHWRs, Vienna, 1993. [4] P.Cd E.S.Yan . n LokenTi . , "POWDERPUFS-V Physics Manual", TDAI-31, AECL- Engineering Company, July 1979. [5] C.H. Westcott, "Effective Cross Section Values for Well Moderated Reactor Spectra" d Editio3r n Corrected CRR P- 9600 , EANDC-4, TNNC-30, Chalk River, Ontario, Nov.l, 1960. ] [6 J.R .Genera A Aske" , al wt le Descriptio e Latticth f o ne Code WIMS", JBNES, Vol. 5, p. 564, 1966. [7] P.D.Krishnani, "CLU MultigrouA B - p Integral Transport Theory Cod Analysir efo f so Cluster Lattices", Ann. nucl. Energy, Vol. 255p , .9 , 1982. [8] P.D. Krishnani, "Interface Current Method for PHWR Cluster Geometry with Anistopy Angulae inth r Flu t Interfaces"xa , Ann. nucl. Energy, Vol. 287p , .9 , 1982. ] [9 P.D. Krishnani, "CLU MultigrouA B- p Integral Transport Theory Cod r Latticefo e Calculation f PHWso R Cells", Repor . BARC/1992/E/017No t , 1992. [10] F.J. Payers et al., AEEW - 785 Part l, 1972. [II] H.C. Huria, IAEA-TECDOC-314, Vienna, 1984. [12] B.S.Singh, Kamala Balakrishnan and M.R. Balakrishnan, "RHEA - A Five Group Neutronics Code for Burnup Analysis of Cluster Geometry Lattices", Report No. B.A.R.C/I-608, 1980. [13] A.R. Dastur and D.B. Buss, "MULTICELL-A 3-D Program for the simulation of Reactivity Devices in CANDU Reacotrs", AECL-7544, Feb. 1983. [14] H.C. Chow and M.H.M. Roshd, "SHETAN-A Three Dimensional Integral Transport cod Reactor efo r Analysis", AECL-6878, September 1980. [15] S.B. Degwekar, "BOXER-3 : A Three Dimensional Integral Transport Code for PHWR supercell", Repor . BARC-1295no t , 1985. . Jagannatha[16V Diffusio] A " , al t nne Iterative Mode Simulatior fo l Reactivitf no y Devices PHWRs"n i , Nucl. Sei. Eng., Vol. 104 . 222p , , 1990. 153 . JagannathaV [17] , "COMESal t ne CorneA H- r Mesh Finite Difference Cod Solvo eT e Multigroup Diffusion Equations in Multidimensions", Report BARC.-1367, 1987. [18] H.C.Huria, "PIJMURLI Code", unpublished, 1986. [19] H.P. Gupta, V.K. Jain and S.D. Paranjape, "A simplified scheme to simulate the Reactivity Devices in PHWRs", A paper presented in Vlllth National Symposium on Radiation Physics held in Bombay during 17-19 January, 1990. [20] S.K .. BalaramanV Kapi d f CALan o l Two-Dimensionae ManuaA A " C, ,Us r fo l l multigroup Transport Theory Code", Report No. TRPS-221, 1975. [21] H.C.Gupta, Report BARC-1543, 1991. [22] L.M.Petrie and N.F.Cross, "KENO-IV - An Improved Monte Carlo Criticality Program", ORNL-4938, 1975. . GrantC [23] , "UsPUMA-e th f eo C SysteFuee th lr mManagemenfo Embalsf o t e Nuclear Station", CNEA, Engineering Branch, Internal Repor 1166/88. No t , July 1988. [24] A.L. Wight, "Ontario Hydro Reactor Fuelling Simulation Progra mOHRFSP"- , Ontario Hydro Design and Development Division Report 86000, January 1987. [25] A.L. . SibleyWighR d an ,t "Fuel Management Design Progra m- FMD P Par- 1 t Program Description", Atomic Energ f Canadyo a Limited, Internal Report, TDAI-105, August 1977. [26] Kamala Balakrishnan and K.R. Srinivasan, "DIMENTRI - A Two Group Three Dimensional Diffusion Theory Programme", Report RED/TPS/136, 1966. [27] K.R. Srinivasan, "TAQUIL and TRIVENI Computer Codes for Fuel Management of PHWRs", Note PHWR/500/PHY/18, 1986. . Jagannatha Analysi[28V n ]Reactoe "A , f PHWal o sMW t nre 5 witR23 h PHANTOM- CEMESH", BARC/1991/I/016, 1991. [29] Kamala Balakrishnan, "ANAMIKA - A Two group Diffusion Theory Code for Core Follow-up Studies", Repor . RED/RP/FC/45No t , 1990. [30] T.B. Fowle D.Rd ran . Vondy, "Nuclear Reactor Analysis Code CITATION" Ridgk Oa ,e National Laboratory, Tennessee, 1969. [31] AECL Internal Repor CHEBn o t Y Code. [32] AECL Internal Report relate MATMAn do P Code. [33] M.J.Bell, "ORIGEN-The ORNL Isotope Generatio Depletiod nan n Code" ORNL-4628, May 1973 (Updated 1979). [34] Arvind Kumar Compute" , r Code Developmen Physicr tfo PHWe s MW RDesig 0 ", 50 f no BARC Report No. 1313, 1986. 154 . BassarskyF [35] , C.R. Calabres J.Md ean . Fink, "Calculatio f IAEno A Reactor Physics Benchmark Problems for PHWR in CNEA (Argentina)", Progress Report No. 1 : Cell calculations, 21-6-90, Buenos Aires. [36] M.R. Lauzon and M.B. Gold, "Description and Solution of the IAEA Fuel Management Benchmark for PHWRs", Ontario Hydro, Report No. 90156, June 1990, Canada. [37] P.D. Krishnani, "Lattice Analysis for IAEA PHWR Benchmark by the Computer Code CLUB", BARC, Internal Note No. RPS27, 1990. [38] H.P. Gupta and A.N. Nakra, "Analysis of IAEA Lattice Benchmark for PHWRs", BARC, Internal Note No. ThPD/327, May 1990 [39] A. Kakodkar, K.Balakrishnan and P. Yaso, "In-Core Fuel Management Code Package Validation for PHWR", Yearly Progress Report, IAEA Research Contract Project No. 5204/RB, April 1992, BARC, Bombay, India. [40] Hark Rho Kirn et al, "Lattice Calculation of CANDU PHWR in KAERI as Part of the IAEA In - Core Fuel Management Benchmarks for PHWR", Oct.30, 1990, Design Analysis Department, Korea Atomic Energy Research Institute. . Parvez[41A ] , "Progress Repor Researcn o t h Contrac o 5706/RB"N t , 1990, Karachi, Pakistan. [42] I. Dumitrache et al, "Solution of the PHWR Lattice cell Benchmark Problems Defined by the IAEA Vienna", Nov. 1990, Contract 5707 RB, RENEL Institute for Nuclear Research, Pitesti, Romania. [43] Consultants Report on In-Core Fuel Management Benchmarks for PHWRs, Vienna, 15-18 November 1988. . BassarskyF [44] , C.R. Calabrese, J.M. Fink, "Calculation f IAEo s A Reactor Physics Benchmark Problems for PHWR in CNEA (Argentina) Progress Report No.2 : Reactivity Devices Calculations" I.T. 1042/90, 1990. [45] S.A. Kushneriuk, "Neutron Captur y Lonb e g Cylindrical Bodies surroundey b d Predominantly Scattering Media", AECL-462, June 1957. [46] M.R. Lauzo M.Bd nan . Gold, "Descriptio Solutiod nIAEan e th f Ano Fuel Management Benchmar Pressurizer kfo d Heavy Water Reactors", Repor . 90156no t , June 1990. [47] H.P. Gupt R.D.Sd an a . Yadav, "Descriptio Analysid nan f IAEA-Benchmarko s r fo s PHWRs", Not . Th.P.D./Phy/358,eNo 1992y Ma ' . [48] Anil Kakodkar, Kamala Balakrishnan and P.Yaso, "Progress Report for the year 1991, CRP on In-core Fuel Management code Package for PHWRs". [49] I. Dumitrache, V. Toma, C. Margeanu, I.Prodea and C.Rizoil, "Solution of the PHWR Reactivity Device Benchmark Problems defined by the IAEA Vienna", Progress Report for contract 5707/RB, November 1992. 155 [50] "Summary Report" First Research Coordination Meetin In-corn go e Fuel Management Benchmar PHWRsr kfo , Buenos Airs, Argentina, 14-21 December 1990. [51] "Operational Physics of Power Reactors", Proceedings of the Training Course held at Trieste, IAEA-SMR-68/II, 1982. A.A.Pasanen, "Fundamental f CANDso U Reactor Nulcear Design" 131-214p p , , 1982. G.M. Frescur A.Ld an a . Wight, "CAND FueW l PH Management"U - . 451-526pp , , 1982. [52] Summary e FirsReporth f t o tResearc h Co-ordination Meetin n In-Coro g e Fuel Management Benchmarks for PHWRs, Buenos Aires, Argentina, 14-21 December 1990. . Bassarsky[53F ] , C.R. Calabrese, J.M. Fink, "Calculation f IAEo s A Reactor Physics Benchmark-Core Calculations", Progress Report Submitte IAEAo dt , 1990. [54] R. Srivenkatesan, H.P. Gupta and R.D.S. Yadav, "Analysis of IAEA PHWR Benchmark-Core Calculations", Progress Report Submitte IAEAo dt , 1990. [55] Ansar Parvez, Mohammad Saleem, Tasneem Hakeem, "Core Global Calculatioe th r nfo IAEA PHWR Benchmark and KANUPP using the Computer Code CITATION", Progress Report Submitte IAEAo dt , 1990. . Toma[56V . ]Dumitrache T , . PatrulescuI , . RizoiuC , . TomaG , . MargeanuC , . ProdeaI , , "Benchmark for In-core Fuel Management Programs of PHWRs", Progress Reoport Submitte IAEAo dt , 1990. [57] V.Jagannatha d R.P.Jainan n , "Analysi e IAEth Af o s PHWR (600 MWe) Core Benchmark with PHANTOM-CEMESH Code System", Th.P.D./Note No./371, November , 19926 , . . FinkJ [58] , "Specification f Inpuo s t Dat r Los afo f Regulatio so n Benchmark Probler mfo PHWR", June 1991. . Fresema[59M ] , "Reactor Physics Aspect Candf so u Accident Analysis", Winter Collegn ei ICTP on Nuclear Physics and Reactors, Trieste, Italy, Jan.25-Mar.19, 1982. [60] A.M. Lerner, C.R. Calabrese, J.M. Fink, "Solutio lose f Regulatioth o sf no n IAEA Benchmark Proble PHWr mfo C.N.E.A."Rn i , I.T.Argentina 1021/92 Jun9 2 , e 1992. [61] V.K. Jain and H.P. Gupta, "IAEA-Loss of Regulation Benchmark Problem for PHWR (Task-4)", ThPD. Note No./369/92 Sept5 2 , . 1992. [62] Toma Victor, Dumitrache Ion, Toma Ginamihaela, "The Los Regulatiof so n Benchmark Problem for PHWR", Romania S-l-NT, 193, 20 June 1992 (Rev. -1). [63] M. Pomerantz and J. Fink, "Calculation of the Benchmark Problem on LOG A Reactivity Effect from Different Nuclides in CNEA", CNEA, Buenos Aires, Argentina, November 1992. [64] BARC - RED Task 5 Results, November 1992, Bombay, Trombay, India. 156 [65] P.D. Krishnan H.Pd an i . Gupta Influence"o" , f Isotopic Compositio Voin no d Effect", Not . ThPD/365eNo , September 1992, BARC, Bombay, India. [66] M. Saleem, M. Arshad, A. Parvez, "Influence of Isotopic Composition on Void Effect", IAEA PHWR Benchmark Tas , Revk5 , Novembe1 . r 1992, Karachi Nuclear Power Plant, PO Box 3183, Paradize Point, Karachi, Pakistan. . DumitracheI [67] . MargeanuC , . ProdeaI , Isotopie Th " , c Composition Influenc Voin eo d Effect, PHWR Benchmark Problem", Sl-NT-197, September 30, 1992, RENEL, Institute for Nuclear Research, Reactor Physics and Nuclear Safety Division, Pitesti, Romania. [68] Zamonsky, "Solutio Fissioe th f no n Product Inventorie Decad an s y Heat Probler mfo PHWR in CNEA", Informe Tecnico CNEA-RE-CA.93-06, Argentina, September 1993. [69] Farrukh Iqbal, Ansar Parvez, "Calculation of the Fission Product Inventories and their Decay Heat", Progress Report submitted to IAEA, Pakistan, November 1992. [70] Didita Liana Petree Hi , , "Fission Product Inventorie Decad an s y Heat Calculations", Progress Report submitte IAEAo dt , 1992. [71] Kamala Balakrishnan et al., Submissions at the TCM, Bombay 1992. [72] V. Jagannathan, Vinod Kumar, S. B. Degwekar, H.C. Gupta and R.P. Jain, Analysis of SDS-2: Tas , Park8 Lattic - tI e Calculations, ThPD Not . 373eNo , Par Nov.d an t ; ,92 I -CorI e Simulations, ThPD Not . 374eNo , Nov.. ,92 [73] Kamala Balakrishna , "Reactivital t ne y Worth Liquif so d Poiso Moderator"ne Jetth n si , Ann nuclf o . . Energy ,. 391Volp , , .20 1993. [74] R.E. Kay, "Lattice Measurements with 37-Element Bruce Reactor Fue Heavn i l y Water Moderator: Detailed Lattice Cell Parameters", AECL-5307, 1976. [75] K.J. Serdula, AECL-2606, 1966. [76] K.J. Serdula and R.E. Green, AECL-2772, 1967. [77] M.F. Dure t al.e t , "Plutonium ComparisoProductioA - D NP n i n between Experiment and Calculation", AECL-3995, 1971. [78] R.T. Jones, AECL-5853, 1977. [79] A.M. Lerner, "Lattice Calculation 37-Elemene th r fo s t Bruce Reactor Fue Heavn i l y Water Moderator: Detailed Lattice Cell Comparisons Between WIMS Calculations and Measurements Performed in the Experimental Facility ZED-2", Report No. CNEA-RE- CA 93-08, CNEA, Argentina, 1993. [80] P.D. Krishnani, "Analysi f Experimento s s with 37-Rod Fuel Cluster n ZED-i s 2 Reactor", Note No. PHWR 500/PHY/16, 1986. 157 Dumitrachen [81Io ] , "Benchmar In-Corr kfo e Fuel Managements Progra PHWRs"mf o , Final Report, National Electric Power Authority, Institut r Nucleaefo r Research, Romania, 1992. [82] P.D. Krishnani, "Analysi f Experimento s s with 28-Rod Fuel Clusters using Computer Code CLUB", Note No. ThPD/367, 1992. [83] A.N. Nakra, "Analysis of 28 Rod Uranium Oxide Cluster Experiments using Computer Code CLIMAX", Not . ThPD/438eNo , 1994. [84] P.D. Krishnani, "Analysis of Isotopic Composition for 19-Rod Fuel Cluster Using Computer Code CLUB", Note No. ThPD/366, 1992. [85] A.M. Lerner and J.M. Fink, "Plutonium Production in NPD Irradiated Fuel - Comparison of Measurements With WIMS", Informe Tecnico CNEA-RE-CA 93-10, Argentina, October 1993. [86] J.M.Fink, "Stainless Steel Adjuste Experimentd Ro r ZED-e Th n 2si Critical Facility- Comparison Between Measurement d Calculationan s s wite Codeth h s WIMS- D4/SHETAN/PUMA-C", I.T. 1033/92, Sep. 17, 1992. [87] V. Jagannathan, R.P. Jain, A.N.Nakra and R.D.S. Yadav, "Analysis of The Adjuster Rod Experiments in ZED-2 by PHANTON-CEMESH Code System", Report BARC/1993/I/007, 1993. [88] S.B.Degweker, P.D.Krishnani, R.S.Moda d R.Srivenkatesanan k , "Analysie th f o s Adjuste Experimentd Ro r ZED-e th n si 2 ReactoSupee Th ry b rCel l Code BOXER3", Not . ThPD/427eNo June0 2 , , 1994. [89] Arvind Kumar, "Problem Description and Input Data for the Power Distribution Control and Flux Mapping Benchmark Problem for PHWRs for Task-10", Internal Note Th.PD, BARC, 1992. [90] P. de Buda and M. Gold, Paper Presented in Second RCM on In-Core Fuel Management Benchmarks for PHWRs, Bombay, India, Nov. 1992. [91] Arvind Kumar, Paper Presented in Second RCM on In-Core Fuel Management Benchmark PHWRsr sfo , Bombay, India, Nov. 1992. 158 CONTRIBUTOR DRAFTINO ST REVIED GAN W Arshad, M. Karachi Nuclear Power Plant, KANUPP Pakistan Arvind Kumar Bhabha Atomic Research Centre, BARC ThPD, Bombay, India Balakrishnan, Kamala Bhabha Atomic Research Centre, BARC RED, Bombay, India Bashir, Farhat Karachi Nuclear Power Plant, KANUPP Pakistan Bassarsky. F , Comision Nacional de Energia Atomica, CNEA Argentina Buda, P. de Ontario Hydro, Toranto, Canada Calabrese, C.R. Comision Nacional de Energia Atomica, CNEA Argentina Crijns. MJ , International Atomic Energy Agency, IAEA VIC, Vienna Scientific Secretary Division of Nuclear Power Nuclear Power Technology Development Section Degwekar, S.B. Bhabha Atomic Research Centre, BARC ThPD, Bombay, India Didita, Liana Institute for Nuclear Research Romania Dumitrache, Ion Institut Nuclear efo r Research Romania Fink, J.M. Comision Naciona Energie d l a Atomica, CNEA Argentina Giil, Choong Sup Korea Atomic Energy Research Institute, KAERI Republic of Korea Gold, M. Ontario Hydro, Toranto, Canada Grant, Carlos Comision Naciona Energie d l a Atomica, CNEA Argentina Gupta, H.C. Bhabha Atomic Research Centre, BARC ThPD, Bombay, India Gupta, H.P. Bhabha Atomic Research Centre, BARC ThPD, Bombay, India 159 Hakim, Tasneem Karachi Nuclear Power- Plant, KANUPP Pakistan Hie, Petre Institut r Nucleaefo r Research Romania Iqbal, Farrukh Karachi Nuclear Power Plant, KANUPP Pakistan Jagannathan. V , Bhabha Atomic Research Centre ARB , C ThPD, Bombay, India Jain, R.P. Bhabha Atomic Research Centre ARB , C ThPD, Bombay, India Jain, V.K. Bhabha Atomic Research Centre ARB , C ThPD, Bombay, India Kannan, Umasankari Bhabha Atomic Research Centre, B ARC RED, Bombay, India Kim, Bong Ghi Korea Atomic Energy Research Institute, KAERI Republic of Korea Kim, Haro kRh Korea Atomic Energy Research Institute, KAERI Republi f Koreco a Kim, Jung-Do Korea Atomic Energy Research Institute, KAERI Republic of Korea Krishnani, P.D. Bhabha Atomic Research Centre, B ARC ThPD, Bombay, India Lanzon, M.R. Ontario Hydro, Toranto, Canada Lawande, S.V. Bhabha Atomic Research Centre, BARC ThPD, Bombay, India Lerner, Ana Maria Comision Nacional de Energia Atomica, CNEA Argentina Margeanu, Crisina Institut r Nucleaefo r Research Romania Modak, R.S. Bhabha Atomic Research Centre, BARC ThPD, Bombay, India Nakra, A.N. Bhabha Atomic Research Centre, BARC ThPD, Bombay, India 160 Parvez, Ansar Karachi Nuclear Power Plant, KANUPP Pakistan Patrulescu, Ilie Institute for Nuclear Research Romania Pomerantz, M. Comision Naciona Energie d l a Atomica, CNEA Argentina Prodea, losif Institute for Nuclear Research Romania Pushpam, Neelima Bhabha Atomic Research Centre, BARC RED, Bombay, India Rastogi, B.P. Bhabha Atomic Research Centre, BARC ThPD, Bombay, India Razak, Abdul Karachi Nuclear Power Plant, KANUPP Pakistan Rizoiu, Andrei Institute for Nuclear Research Romania Saleem, Mohammad Karachi Nuclear Power Plant, KANUPP Pakistan Srinivasan, K.R. Bhabha Atomic Research Centre, BARC ThPD, Bombay, India Srivenkatesan, R. Bhabha Atomic Research Centre, BARC ThPD, Bombay, India Suh, Doo-Suhk International Atomic Energy Agency, IAEA VIC, Vienna Scientific Secretary Division of Nuclear Power Nuclear Power Technology Development Section Tazeen, Ahmad Karachi Nuclear Power Plant, KANUPP Pakistan Toma, Gina Mihaela Institute for Nuclear Research Romania Toma, Victor Institute for Nuclear Research Romania Vinod Kumar Bhabha Atomic Research Centre, BARC ThPD, Bombay, India Wight, A.L. Ontario Hydro, Toranto, Canada 161 Yadav, R.D.S. Bhabha Atomic Research Centre, BARC ThPD, Bombay, India Yaso, P. Bhabha Atomic Research Centre, BARC RED, Bombay, India Zamonsky, G. Comision Nacional de Energia Atomica, CNEA Argentina Research Coordination Meetings Buenos Aires, Argentina: 14-21 December, 1990 Bombay, India: 23-26 November, 1992 Consultants Meetings Vienna, Austria: 15-18 November, 1988 Vienna, Austria: 8-12 August, 1994 Vienna, Austria: 28 November - 2 December, 1994 0> CD CD 162