Integral Test for JENDL-4 Benchmark Results with Preliminary - PowerPoint PPT Presentation

2007 Symposium on Nuclear Data, Nov. 29-30, 2007, RICOTTI Convention Center, Tokai, Ibaraki, Japan Integral Test for JENDL-4 Benchmark Results with Preliminary Version of JENDL Actinoid File 29 Nov. 2007 Keisuke OKUMURA, Go CHIBA (okumura.keisuke@jaea.go.jp, chiba.go@jaea.go.jp) Reactor Physics Group Nuclear Science and Engineering Directorate Japan Atomic Energy Agency (JAEA) 1

Background and Objective � The JENDL Actinoid File (JENDL/AC) is under developing at JAEA. � Most of the evaluations in JENDL/AC will be taken over to a part of the next general purpose file JENDL-4. Benchmark calculation for various type of reactors to confirm present performances of JENDL/AC and to polish it more and more. Goal Good performance superior to recent other nuclear data files : JENDL-3.3, JEFF-3.1, ENDF/B=VII.0, etc. 2

Framework of JENDL/AC Benchmark Test Japan Nuclear Data Committee (JNDC) Subcommittee on Reactor Constants: • Reactor Integral Test WG : Benchmark results, Work, Comments, Advices Once or twice a year Reactor constants JAEA-Tokai Nuclear Data Center Information Preliminary Benchmark results and exchange nuclear data files sensitivity data (once a month) JAEA-Oarai JAEA-Tokai data Reactor Physics Analysis Reactor Physics Group and Evaluation Group 3

Criticality Benchmark with Former Nuclear Data Files 1.015 JENDL-3.2 JENDL-3.3 1.010 JEF-2.2 ENDF/B-VI(R8) 1.005 C/E of k-eff 1.000 0.995 0.990 0.985 KRITZ2:1Hot KRITZ2:1Cold KRITZ2:13Hot KRITZ2:13Cold DIMPLE3 DIMPLE7 B&W-CoreXI TCA1.50U TCA1.83U TCA2.48U TCA3.00U MISTRAL-C1 TRACY TRX-1 TRX-2 STACY JRR4-U20 JRR4-U93 7.0wt.%10wt.% 2.6wt.% 3.0wt.% 93wt.% 1.3wt.% 2.5wt.% 1.9wt.% 20wt.% U235 3.7wt.% Proc. of the 2002 Symposium on Nuclear Data, Tokai, Japan, Nov. 21-22, JAERI-Conf 2003-006, pp15-21, (2003). 4

Benchmark Materials Handbook (Sep. 2007 Edition) of Handbook (Mar. 2007 Edition) of International Criticality Safety International Reactor Physics Benchmark Evaluation Project Experiment Evaluation Project (ICSBEP) (IRPhEP) 5

Benchmark Calculation • Continuous-Energy Monte Carlo Neutron histories : 20 ～ 60 million calculation (MVP) → 1 σ error of k eff < 0.0002 • Detailed geometrical model specified in the benchmark handbooks • Multi-group deterministic calculation for small reactivity analysis or sensitivity study with the codes: - SLAROM-UF - SN solvers in the CBG system 6

Classification of Benchmark Problems in ICSBEP Handbook Fissile Materials Fuel Forms Neutron Spectra HEU ： Highly Enriched SOL ： Solution FAST ： Fast Uranium Systems (60% ～ ) COMP ： Compound INTER ： Intermediate IEU ： Intermediate and (e.g. UO 2 ,MOX,UF 4 ) THERM ： Thermal Mixed Enrichment Uranium MET ： Metal Systems (10 ～ 60%) MIXED ： Mixed MISC ： Miscellaneous e.g. Multi-region LEU ： Low Enriched (e.g. UO 2 rods in fuel system with different Uranium Systems ( ～ 10%) solution) neutron spectra MIX ： Mixed Plutonium- Uranium Systems (e.g. MOX fuel) Example of case index for TCA-UO 2 cores U233 ： Uranium-233 Systems LEU-COMP-THERM-006 -001 PU ： Plutonium Systems (LCT6.1 ～ LCT6.18) -002 SPEC ： Special Isotope : Different critical configurations Systems -018 lattice pitch, critical water height, horizontal lattice size (NxN), etc. 7

Selected Benchmark Problems Fuel Form Spectra ICSBEP2006 MVP Cal. SOL THERM 72 9 INTER 3 2 FAST 1 0 SOL THERM 463 50 INTER 3 0 COMP FAST 8 0 THERM 255 63 INTER 14 5 MIXED 17 0 COMP MIX THERM 216 21 FAST 45 9 HEU MIXED 45 0 MET INTER 2 0 FAST 304 41 MIXED 1 0 INTER 14 9 FAST 8 0 MET MISC THERM 127 3 THERM 56 53 MIXED MIXED 32 8 8 0 MISC THERM 7 0 COMP THERM 5 0 SOL THERM 5 0 INTER 29 29 SOL FAST 2 1 THERM 192 44 U233 INTER 14 2 MIXED 8 3 IEU COMP THERM 41 1 FAST 10 10 MET MIXED 3 0 THERM 1 0 MET FAST 20 11 SOL THERM 529 208 SOL THERM 104 77 FAST 6 0 COMP THERM 1066 194 INTER 1 0 LEU COMP MET THERM 65 13 THERM 21 0 MISC THERM 11 0 PU MIXED 7 0 FAST 87 37 INTER 4 4 MET We have about 1000 results THERM 2 2 MIXED 1 1 with MVP and JENDL-3.3 SPEC MET FAST 20 20 Total 3955 930 8

Core parameters Calculated keff and errors (for trend analysis) Case index of benchmark problem C/E vs a specific parameter 9

JENDL-3.3 Results for LEU/HEU-SOL UO2(NO3)2 1.020 Reject? UO2F2 1.015 <C/E> = 1.0009 ± 0.0026(1 σ ) 1.010 LEU-SOL 1.005 ± 0.5% C/E C/E 1.000 5~10% 0.995 77 cases 0.990 0.985 0.980 400 600 800 1000 1200 1400 1600 H/U235 H/U235 1.020 UO2(NO3)2 1.015 UO2F2 1.010 HEU-SOL 1.005 ± 0.5% C/E C/E 1.000 89 ~94% 0.995 52 cases 0.990 <C/E> = 0.9993 ± 0.0039(1 σ ) 0.985 0.980 0 500 1000 1500 2000 H/U235 H/U235 10

Criticality of U235 Solution Fueled System (LST & HST) HSI1.1 11

Criticality of Low Enriched U235 Fueled System (LCT) ⎛ ⎞ σ 238 ⎜ ⎟ c ⎜ ⎟ σ 235 ⎝ ⎠ f Thermal 12

Criticality of Enriched U235 Fueled System (KUCA) 13

Effect of Thermal Capture Cross Section of U238 2.64 barn (Lower case), 2.68 barn (Mughabghab) 2.717 barn (JENDL-3.3) 1.005 ⎛ ⎞ σ 238 ⎜ ⎟ c ⎜ ⎟ σ 235 ⎝ ⎠ f Thermal 1.000 C/E (keff) JENDL-3.3 0.995 Mughabghab Lower case 0.990 1 2 3 4 5 6 7 U235 enrichment (wt.%) Proc. of the 2004 Symposium on Nuclear Data, Nov. 11-12, 2004, JAERI, Tokai, Japan, pp.56-63, JAERI-Conf 2005-003, (2005). 14

Light Water Moderated MOX Fueled System MISTRAL & BASALA 0.27% Δ k/kk’ B70: 0.22% Δ k/kk’ J33: 0.15% Δ k/kk’ F31: 0.17% Δ k/ JA071122: Time-change of C/E for criticality of MOX fueled TCA core due to β -decay of Pu-241 to Am-241 15

PU-SOL-THERM System (J33) Number of cases 208 <C/E> = 1.005 ± 0.012 (2 σ ) 99.4 ∼ 100 Pu/HM (wt.%) 71.8 ∼ 99.4 Pu239/Pu (wt.%) 0.5 ∼ 23.2 Pu240/Pu (wt.%) 9.5 ∼ 412 Pu (g/liter) 16

PU-SOL-THERM System (Different Nuclear Data) 17

U233 Fueled System Thermal Intermediate Fast 18

Small Reactor Benchmark 19

Criticality of Uranium Fueled Fast Reactors (BFS-2) +20%MOX +40%MOX +55%MOX Pure UO 2 core outer region central region central region 20

Na Void Reactivity of U Fueled Fast Reactor (BFS-62-3A) Voided Zone Name 21

Criticality of MOX Fueled Fast Reactors 22

SPEC-MET-FAST Benchmark (Cm244, Pu238) Sample reactivity in Pu alloy 23

SPEC-MET-FAST Benchmark (Np237) Np237 sample reactivity in Pu or HEU fuel (SMF3) 24

SPEC-MET-FAST Benchmark (Pu242, Np237) 3kg 239 Pu Plate 3kg 239 Pu Plate 1kg 239 Pu Plate 1kg 239 Pu Plate (a) Pure Pu-239 Model (a) Pure Pu-239 Model Low 242 Pu Plate Low 242 Pu Plate High 242 Pu Plate High 242 Pu Plate (c) End-Driven Model (c) End-Driven Model Be Reflector Be Reflector (b) Center-Driven Model (b) Center-Driven Model Steel Reflector Steel Reflector (e) Be-Reflected Model (e) Be-Reflected Model (d) Steel-Reflected Model (d) Steel-Reflected Model DU Reflector DU Reflector (f) DU-Reflected Model (f) DU-Reflected Model HEU Plate HEU Plate (g) Steel-and-DU-Reflected HEU-Pu242 Model (g) Steel-and-DU-Reflected HEU-Pu242 Model SMF4 Np HEU SMF8 25

PIE Analysis by MVP-BURN for PWR Spent Fuel Boron History 50 SF97-2 SF97-3 SF97-4 SF97-5 45 SF97-6 40 35 Power (MW/t) (47GWd/t) 30 25 20 15 Power History 10 5 0 0 100 200 300 400 500 600 700 800 900 1000 1100 1200 1300 1400 Time (day) 26

PIE Analysis by MVP-BURN for PWR Spent Fuel Isomeric Ratio of Am241(n, γ ) is based on JENDL-Act. Pu241 Pu242 14y U234 U235 U236 U237 U238 6.8d Am242m 141y Am241 Am243 Np239 Np237 2.4d Am242g 88y 16h Pu238 Pu239 Pu240 Cm242 Cm243 Cm245 Cm244 Cm246 18y ( α ) 29y ( α ) 163d ( α ) 27

28 Isomeric Ratio of Am241 Capture

Capture Reaction Rate of Am241 in Typical LWR One-group Isomeric Ratio of Am241(n, γ ) J33 B68 B70 F31 JA070925 0.877 0.885 0.898 0.873 0.899 29

30 Sensitivity on Isomeric Ratio of Am241 Capture Am242g Cm243 29y ( α ) 16h Cm242 163d ( α ) σ a (about 0.88 in JA071122) I.R.=0.86 ～ 0.87

Conclusion Good performance of JENDL/AC for various types of reactors was confirmed by comparison with the results of other recent nuclear data files, JENDL-3.3, JEFF-3.1, and ENDF/B-VII.0. However, further investigation is recommended for: � Criticality of PU-SOL-THERM system, � Criticality of U233-SOL-INTER system, � Generation of Cm-242 and Cm-243 in the LWR spent fuel. 31

32 Reactor Physics Group PC Cluster of Altix3700Bx2/2048CPU