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

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Benchmark Results with Preliminary Version of JENDL Actinoid File

Integral Test for JENDL-4

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) 29 Nov. 2007

2007 Symposium on Nuclear Data, Nov. 29-30, 2007, RICOTTI Convention Center, Tokai, Ibaraki, Japan

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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. Good performance superior to recent other nuclear data files : JENDL-3.3, JEFF-3.1, ENDF/B=VII.0, etc.

Goal

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Framework of JENDL/AC Benchmark Test

Japan Nuclear Data Committee (JNDC) Subcommittee on Reactor Constants:

• Reactor Integral Test WG

: Work, Comments, Advices

Once or twice a year

JAEA-Tokai Nuclear Data Center JAEA-Tokai Reactor Physics Group JAEA-Oarai Reactor Physics Analysis and Evaluation Group Preliminary nuclear data files Benchmark results and sensitivity data

(once a month)

Information exchange

data

Benchmark results, Reactor constants

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• Proc. of the 2002 Symposium on Nuclear Data, Tokai, Japan, Nov. 21-22,

JAERI-Conf 2003-006, pp15-21, (2003).

Criticality Benchmark with Former Nuclear Data Files

0.985 0.990 0.995 1.000 1.005 1.010 1.015 TRX-1 TRX-2 KRITZ2:1Hot KRITZ2:1Cold KRITZ2:13Hot KRITZ2:13Cold B&W-CoreXI TCA1.50U TCA1.83U TCA2.48U TCA3.00U DIMPLE3 MISTRAL-C1 DIMPLE7 STACY TRACY JRR4-U20 JRR4-U93 C/E of k-eff JENDL-3.2 JENDL-3.3 JEF-2.2 ENDF/B-VI(R8)

1.3wt.% U235 1.9wt.% 2.6wt.% 3.0wt.% 3.7wt.% 7.0wt.%10wt.% 20wt.% 93wt.% 2.5wt.%

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Benchmark Materials

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

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Benchmark Calculation

• Continuous-Energy Monte Carlo

calculation (MVP)

• 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

Neutron histories : 20～60 million → 1σ error of keff < 0.0002

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HEU：Highly Enriched Uranium Systems (60%～) IEU：Intermediate and Mixed Enrichment Uranium Systems (10～60%) LEU：Low Enriched Uranium Systems (～10%) MIX：Mixed Plutonium- Uranium Systems (e.g. MOX fuel) U233：Uranium-233 Systems PU：Plutonium Systems SPEC：Special Isotope Systems SOL：Solution COMP：Compound (e.g. UO2,MOX,UF4) MET：Metal MISC：Miscellaneous (e.g. UO2 rods in fuel solution) FAST：Fast INTER：Intermediate THERM：Thermal MIXED：Mixed e.g. Multi-region system with different neutron spectra

Fissile Materials Fuel Forms Neutron Spectra LEU-COMP-THERM-006 Different critical configurations lattice pitch, critical water height, horizontal lattice size (NxN), etc.

• 001
• 002

:

• 018

Classification of Benchmark Problems in ICSBEP Handbook

Example of case index for TCA-UO2 cores (LCT6.1～LCT6.18)

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Selected Benchmark Problems

SOL THERM 72 9 FAST 1 INTER 3 THERM 255 63 MIXED 17 FAST 45 9 INTER 2 MIXED 1 FAST 8 THERM 56 53 MIXED 8 COMP THERM 5 INTER 29 29 THERM 192 44 MIXED 8 3 FAST 10 10 THERM 1 SOL THERM 529 208 FAST 6 INTER 1 THERM 21 MIXED 7 FAST 87 37 INTER 4 4 THERM 2 2 MIXED 1 1 SPEC MET FAST 20 20 Total 3955 930 PU COMP MET MIX COMP MET MISC U233 SOL MET

We have about 1000 results with MVP and JENDL-3.3

Fuel Form Spectra ICSBEP2006 MVP Cal. INTER 3 2 THERM 463 50 FAST 8 INTER 14 5 THERM 216 21 MIXED 45 FAST 304 41 INTER 14 9 THERM 127 3 MIXED 32 8 MISC THERM 7 SOL THERM 5 FAST 2 1 INTER 14 2 THERM 41 1 MIXED 3 MET FAST 20 11 SOL THERM 104 77 COMP THERM 1066 194 MET THERM 65 13 MISC THERM 11 LEU HEU SOL COMP MET IEU COMP

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C/E vs a specific parameter Core parameters (for trend analysis) Calculated keff and errors

Case index of benchmark problem

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5~10% 77 cases 89 ~94% 52 cases LEU-SOL HEU-SOL

JENDL-3.3 Results for LEU/HEU-SOL

0.980 0.985 0.990 0.995 1.000 1.005 1.010 1.015 1.020 400 600 800 1000 1200 1400 1600 H/U235 C/E UO2(NO3)2 UO2F2

±0.5% Reject?

<C/E> = 1.0009 ± 0.0026(1σ)

C/E

0.980 0.985 0.990 0.995 1.000 1.005 1.010 1.015 1.020 500 1000 1500 2000 H/U235 C/E UO2(NO3)2 UO2F2

H/U235 C/E

<C/E> = 0.9993 ± 0.0039(1σ)

H/U235

±0.5%

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Criticality of U235 Solution Fueled System (LST & HST)

HSI1.1

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Criticality of Low Enriched U235 Fueled System (LCT)

Thermal f c

⎟ ⎟ ⎠ ⎞ ⎜ ⎜ ⎝ ⎛

235 238

σ σ

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Criticality of Enriched U235 Fueled System (KUCA)

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0.990 0.995 1.000 1.005 1 2 3 4 5 6 7

U235 enrichment (wt.%) C/E (keff)

JENDL-3.3 Mughabghab Lower case

2.64 barn (Lower case), 2.68 barn (Mughabghab) 2.717 barn (JENDL-3.3)

Effect of Thermal Capture Cross Section of U238

• Proc. of the 2004 Symposium on Nuclear Data, Nov. 11-12, 2004,

JAERI, Tokai, Japan, pp.56-63, JAERI-Conf 2005-003, (2005).

Thermal f c

⎟ ⎟ ⎠ ⎞ ⎜ ⎜ ⎝ ⎛

235 238

σ σ

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Light Water Moderated MOX Fueled System

Time-change of C/E for criticality

• f MOX fueled TCA core

due to β-decay of Pu-241 to Am-241

MISTRAL & BASALA

B70: 0.27%Δk/kk’ J33: 0.22%Δk/kk’ F31: 0.15%Δk/kk’ JA071122: 0.17%Δk/

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<C/E> = 1.005 ± 0.012 (2σ)

PU-SOL-THERM System (J33)

Number of cases 208 Pu/HM (wt.%) 99.4∼100 Pu239/Pu (wt.%) 71.8∼99.4 Pu240/Pu (wt.%) 0.5∼23.2 Pu (g/liter) 9.5∼412

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PU-SOL-THERM System (Different Nuclear Data)

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U233 Fueled System

Thermal Intermediate Fast

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Small Reactor Benchmark

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Criticality of Uranium Fueled Fast Reactors (BFS-2)

+20%MOX

• uter region

+40%MOX central region Pure UO2 core +55%MOX central region

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Na Void Reactivity of U Fueled Fast Reactor (BFS-62-3A)

Voided Zone Name

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Criticality of MOX Fueled Fast Reactors

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SPEC-MET-FAST Benchmark (Cm244, Pu238)

Sample reactivity in Pu alloy

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SPEC-MET-FAST Benchmark (Np237)

Np237 sample reactivity in Pu or HEU fuel (SMF3)

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SPEC-MET-FAST Benchmark (Pu242, Np237)

Np HEU

(b) Center-Driven Model (c) End-Driven Model (d) Steel-Reflected Model (e) Be-Reflected Model (f) DU-Reflected Model (g) Steel-and-DU-Reflected HEU-Pu242 Model (a) Pure Pu-239 Model 3kg 239Pu Plate 1kg 239Pu Plate Low 242Pu Plate High 242Pu Plate HEU Plate Be Reflector Steel Reflector DU Reflector (b) Center-Driven Model (c) End-Driven Model (d) Steel-Reflected Model (e) Be-Reflected Model (f) DU-Reflected Model (g) Steel-and-DU-Reflected HEU-Pu242 Model (a) Pure Pu-239 Model 3kg 239Pu Plate 1kg 239Pu Plate Low 242Pu Plate High 242Pu Plate HEU Plate Be Reflector Steel Reflector DU Reflector

SMF4 SMF8

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5 10 15 20 25 30 35 40 45 50 100 200 300 400 500 600 700 800 900 1000 1100 1200 1300 1400

Time (day) Power (MW/t)

SF97-2 SF97-3 SF97-4 SF97-5 SF97-6

PIE Analysis by MVP-BURN for PWR Spent Fuel

Power History Boron History

(47GWd/t)

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PIE Analysis by MVP-BURN for PWR Spent Fuel

Pu241 Pu242 Am241 Am242m Am242g Cm242 Cm243 Am243 Cm244 Pu238

16h 141y

Cm245 Cm246 U235 U236 Np237

6.8d

U234 U238 U237 Np239 Pu240 Pu239

88y 2.4d 14y 163d (α) 29y (α) 18y (α)

Isomeric Ratio of Am241(n,γ) is based on JENDL-Act.

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Isomeric Ratio of Am241 Capture

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Capture Reaction Rate of Am241 in Typical LWR

J33 B68 B70 F31 JA070925 0.877 0.885 0.898 0.873 0.899

One-group Isomeric Ratio of Am241(n,γ)

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I.R.=0.86～0.87

a

σ

Am242g Cm242 Cm243

16h 163d (α) 29y (α)

Sensitivity on Isomeric Ratio of Am241 Capture

(about 0.88 in JA071122)

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Conclusion Good performance of JENDL/AC for various types of reactors was confirmed by comparison with the results of

• ther 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.

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Altix3700Bx2/2048CPU PC Cluster of Reactor Physics Group