MOX Fuel Core Physics Experiments and Analysis - Aiming for Plutonium Effective Use.

Journal of Nuclear Science and Technology

Sakai

Ando

et al. 2012

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A part of the experimental program FUBILA was dedicated to the study of core physics characteristics of full MOX BWR cores by testing five experimental cores: a core inserted with a B 4 C control blade, a core loaded with UO 2 fuel rods, the core loaded with Gd 2 O 3 -UO 2 fuel rods, a core loaded with 10 6 10 MOX assemblies and a core loaded with time-elapsed MOX fuel. The present article describes analysis results of the experimental data with deterministic analysis codes and a continuous energy Monte Carlo code coupled with major nuclear data libraries. The calculated critical k eff' s with the Monte Carlo calculations range from 0.999 to 1.007. Those of the transport calculations with sixteen energy groups are close to those of the Monte Carlo calculations while those of diffusion calculations with the same sixteen energy groups are systematically smaller by 70.3 to 70.5% Dk than those of the Monte Carlo calculations. The RMSs of differences between the calculated and measured core radial fission rates are 2 to 3%, 1 to 2%, and 1 to 2% for the diffusion, transport and Monte Carlo calculations, respectively. For the analysis of the Gd 2 O 3 -UO 2 fuel rod loaded core, the C/Es of the radial fission rates were improved by adopting a detailed lattice calculation model and a newly measured thermal and resonance cross-sections of Gadolinium.

Section: Deterministic Calculationsmentioning

confidence: 99%

mentioning

confidence: 99%

Neutronics analysis of full MOX BWR core simulation experiments – FUBILA: Part 2

Journal of Nuclear Science and Technology

Sakai

Ando

et al. 2012

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“…Resonance cross sections were obtained by a hyperfine energy group calculation module, PEACO, 12) from 961.12 eV to the thermal cutoff energy, 1.86 eV, 13) and a nuclear data library, JENDL-3.2, 14) was used. A geometrical calculation option, IGT ¼ 16, was selected in the Pij calculations.…”

Section: Calculation Methods For Rod-by-rod Fp Inventory Distribumentioning

confidence: 99%

Analysis of Rod-by-Rod FP Inventory Distributions in BWR 8×8 UO2 Assemblies Using Lattice Physics Method

Yamamoto²

2008

J. Nucl. Sci. Technol.

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Fuel rod gamma-ray spectrometry was performed for one-, two-, three-, and five-cycle irradiated BWR 8 Â 8-4 fuel assembles, and relative rod-by-rod FP inventory distributions of 137 Cs, 134 Cs, 106 Ru, and 95 Zr were obtained for the upper and lower axial height nodes of the assemblies. The measured data was corrected for the difference in gamma-ray transmission between UO 2 and Gd 2 O 3 -UO 2 rods. These distributions were compared with those calculated using the collision probability method, Pij, of the SRAC code system with an infinite lattice model of the fuel assembly. The calculated results generally well reproduce the measured distributions, and the accuracy of the analysis method of the present study was evaluated to be 1 to 2% for the inventory distributions of 137 Cs, 2 to 3% for 134 Cs, 2 to 3% for 106 Ru, and 2 to 3% for 95 Zr, which represent the distributions of the burn-up, the thermal flux multiplied by the burn-up, the buildup of 239 Pu, and the fission rate at the end of the fuel discharge cycle, respectively. One of the notable results of the analysis of this study is that the FP inventories for the Gd 2 O 3 -UO 2 rods were underestimated in most cases.

“…16) Resonance cross sections were obtained by a hyperfine energy group calculation module PEACO, 10) and a thermal cut off energy was set to be 1.8554 eV. 17) A general-purpose burnup chain model 11) that is specially prepared for analyses of isotopic inventory measurement data of irradiated fuel was used as a burnup chain library. Burnup steps were set to be less than 1 GWd/t.…”

Section: Burnup Calculations Of Br3 Mox Fuelmentioning

confidence: 99%

“…22) Resonance cross sections were obtained using the hyperfine energy group calculation module PEACO, 10) and a thermal cut off energy was set to be 1.8554 eV. 17) For core calculations, the diffusion calculation module CITATION 23) in SRAC and the transport calculation code THREEDANT 12) were used with sixteen neutron energy groups and three dimensional geometries. Macroscopic transport cross sections were created by applying B 1 and extended transport approximations to the calculations with CITATION and THREEDANT, respectively.…”

Section: D) Burnup Calculations Of Br3 Mox Fuel Rodsmentioning

confidence: 99%

Analysis of Core Physics Experiments on Fresh and Irradiated BR3 MOX Fuel in REBUS Program

Ando

et al. 2009

J. Nucl. Sci. Technol.

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As part of an international experimental program REBUS, core physics experiments have been implemented on a UO 2 core, which consists of 3.3 and 4.0 wt% UO 2 fuel rods in a square pitch of 1.26 cm, and two partial MOX cores, which replace 7 Â 7 UO 2 fuel rods in the center of the UO 2 core by fuel bundles made of fresh BR3 MOX fuel or irradiated BR3 MOX fuel with an average burnup of 20 GWd/t. Burnup calculations of the BR3 MOX fuel were performed using a general-purpose neutronic calculation code SRAC, and core calculations of the three critical cores were carried out using SRAC, a transport calculation code THREEDANT, and a continuous-energy Monte Carlo code MVP. The measured inventories of major U and Pu isotopes on a sample taken from the BR3 MOX fuel agree with the results of the burnup calculations within 3% deviation. The k eff 's of the three cores are from 0.985 to 1.002. The measured burnup reactivity of the irradiated BR3 MOX fuel was well reproduced by the three types of core calculations. The influence of the accuracy of the inventory calculations on burnup reactivity was studied by comparing between the calculated and measured inventories. The result indicates that the biases in the inventory and reactivity calculations compensate each other, and it makes the total biases of the burnup reactivity small.