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Neutron-Photon Energy Deposition in CANDU Reactor Fuel Channels: a Comparison of Modelling Techniques Using ANISN and MCNP Computer Codes

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Neutron-Photon Energy Deposition in CANDU Reactor Fuel Channels: a Comparison of Modelling Techniques Using ANISN and MCNP Computer Codes AECL EACL CA9700305 CA9700305 AECL-11160, COG-94-387 Neutron-Photon Energy Deposition in CANDU Reactor Fuel Channels: A Comparison of Modelling Techniques Using ANISN and MCNP Computer Codes Depot d energie des photons et des neutrons dans les canaux de combustible des reacteurs CANDU: Comparaison des techniques de modelisation a l'aide des programmes de calcul MCNP et ANISN Z. Bilanovic, D.R. McCracken December 1994 decembre 8 8s 1 9 AECL NEUTRON-PHOTON ENERGY DEPOSITION IN CANDU REACTOR FUEL CHANNELS: A COMPARISON OF MODELLING TECHNIQUES USING ANISN AND MCNP COMPUTER CODES by Z. Bilanovic and D.R. McCracken System Chemistry and Corrosion Branch Chalk River Laboratories Chalk River, Ontario, CANADA KOJ 1J0 1994 December AECL-11160 COG-94-387 EACL DEP6 t D ’ENERGIE des photons et des neutrons DANS LES CANAUX DE COMBUSTIBLE DES r6a CTEURS CANDU : COMPARAISON DES TECHNIQUES DE MODELISATION A L’AIDE DES PROGRAMMES DE CALCUL MCNP ET ANISN par Z. Bilanovic et D.R. McCracken RESUME II est necessaire de determiner les profits de depot d'energie des neutrons, des photons et des electrons dans les canaux de combustible du coeur du reacteur afm d'evaluer les effets de la corrosion radio-induite, la radiolyse du caloporteur et la degradation des proprietes physiques des materiaux et des composants du reacteur. Actuellement, on doit utiliser plusieurs programmes de calcul differents pour y parvenir. La demiere version du programme MCNP en est la plus recente, la plus evoluee et la plus polyvalente, et elle pourrait bien s'averer capable de remplacer tous les autres programmes. Les divers programmes avancent des hypotheses differentes et presentent des restrictions differentes quant a la maniere de modeliser la geometric et la physique du coeur. Le present rapport donne les resultats des modeles MCNP et ANISN de depot d'energie des photons et des neutrons. Ces resultats valident 1'utilisation du programme MCNP pour la moderation geometrique simplifiee du depot d'energie des neutrons et des photons dans le cas de la geometric complexe des canaux de combustible des reacteurs CANDU. Des programmes faisant appel a la methode des ordonnees discretes, tel que le programme ANISN, constituaient les programmes de reference utilises dans les travaux anterieurs. Ce rapport presente egalement les resultats des calculs effectues a l'aide de divers modeles et ils concordent tres bien en ce qui conceme le depot d'energie des neutrons rapides. Dans le cas du depot d'energie des photons, on a du toutefois apporter quelques modifications aux methodes de modelisation. Les problemes poses par le recours a des frontieres reflectrices ont ete resolus soit par la prise en compte dans le modele des huit canaux de combustible en peripherie, soit par le recours a une source limite a la surface limite du probleme. Une fois que ces modifications ont ete incorporees, des resultats concordants ont ete obtenus entre les differents programmes de calcul. Dans le passe, des representations annulaires simples du coeur ont ete employees, en raison de la difficulty de realiser une modelisation detaillee avec les codes plus anciens. On a demontre que la modelisation a l'aide du programme MCNP, faisant appel a une geometric plus precise et plus detaillee, donne des resultats tres differents et bien meilleurs. Chimie et corrosion des systemes Laboratoires de Chalk River Chalk River (Ontario) Canada KOJ 1J0 1994 Decembre AECL-11160 COG-94-387 AECL NEUTRON-PHOTON ENERGY DEPOSITION IN CANDU REACTOR FUEL CHANNELS: A COMPARISON OF MODELLING TECHNIQUES USING ANISN AND MCNP COMPUTER CODES by Z. Bilanovic and D.R. McCracken ABSTRACT In order to assess irradiation-induced corrosion effects, coolant radiolysis and the degradation of the physical properties of reactor materials and components, it is necessary to determine the neutron, photon, and electron energy deposition profiles in the fuel channels of the reactor core. At present, several different computer codes must be used to do this. The most recent, advanced and versatile of these is the latest version of MCNP, which may be capable of replacing all the others. Different codes have different assumptions and different restrictions on the way they can model the core physics and geometry. This report presents the results of ANISN and MCNP models of neutron and photon energy deposition. The results validate the use of MCNP for simplified geometrical modelling of energy deposition by neutrons and photons in the complex geometry of the CANDU reactor fuel channel. Discrete ordinates codes such as ANISN were the benchmark codes used in previous work. The results of calculations using various models are presented, and they show very good agreement for fast-neutron energy deposition. In the case of photon energy deposition, however, some modifications to the modelling procedures had to be incorporated. Problems with the use of reflective boundaries were solved by either including the eight surrounding fuel channels in the model, or using a boundary source at the bounding surface of the problem. Once these modifications were incorporated, consistent results between the computer codes were achieved. Historically, simple annular representations of the core were used, because of the difficulty of doing detailed modelling with older codes. It is demonstrated that modelling by MCNP, using more accurate and more detailed geometry, gives significantly different and improved results. System Chemistry and Corrosion Branch Chalk River Laboratories Chalk River, Ontario, CANADA, KOJ 1 JO 1994 December AECL-11160 COG-94-387 TABLE OF CONTENTS Page 1. INTRODUCTION ........................................................................................................ 1 2. THE ANISN COMPUTER CODE............................................................................... 2 2.1 General Introduction ....................................................................................... 2 2.2 CANDU Lattice Cell Calculations.................................................................. 3 3. THE MONTE CARLO NEUTRON AND PHOTON TRANSPORT CODE. MCNP........................................................................................................................... 5 3.1 General Introduction ........................................................................................ 5 3.2 MCNP - ANISN: Comparative Study .............................................................. 5 3.3 MCNP 18- Element Annular Geometry ......................................................... 7 3.4 Study of MCNP 37-Element Geometry........................................................... 8 4. CONCLUSIONS ........................................................................................................... 8 5. DIRECTION OF FUTURE WORK............................................................................ 9 6. ACKNOWLEDGEMENTS........................................................................................ 10 7. REFERENCES ........................................................................................................... 10 LIST OF TABLES Table 1: Energy structure of the 39-group coupled neutron-gamma cross-section library ..................................................................................................................... 12 Table 2: Materials found in the 39-group cross section library.......................................... 13 Table 3: Dimensions and regions used in the ANISN 37-element lattice site................... 14 Table 4A: Compositions of materials for the fuel regions, used in ANISN calculations. The data are based on WIMS calculations for a 37-element lattice at a bumup of approximately 3300 MW.d/MgU .................. 15 Table 4B: Composition of materials used in ANISN calculations for the fuel channel region. The data are based on WIMS calculations for a 37 -element lattice at a bumup of approximately 3300 MW.d/MgU .................. 15 Table 5: Neutron fission spectmm for the centre pin based on a fission rate of 7.03xl012 fissions/cm3.s........................................................................................ 16 Table 6 A: Energy deposition in the outer coolant region for ANISN calculations using annular geometry. Neutrons were generated by SOURCE. Photons were generated internally by ANISN................................................... 17 Table 6B: Energy deposition in the outer coolant region for ANISN calculations using annular geometry. Neutrons were generated by SOURCE. Photons were generated internally by ANISN................................................... 17 - 11 - Table 7A: Photon source strength (photons/s) and spatial distribution as calculated by GAMSRC....................................................................................... 18 Table 7B: Photon source strength (photons/s) and spatial distribution as calculated by GAMSRC............................................................................................ 19 Table 8A: Energy deposition in the outer coolant region for ANISN calculations using annular geometiy. The photon source was generated by GAMSRC.............................................................................................................. 20 Table 8B: Energy deposition in the outer coolant region for ANISN calculations using annular geometiy. The source was generated by GAMSRC.................... 20 Table 9A: Ratio of ANISN to MCNP results in percent, for neutron energy deposition.............................................................................................................
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