Neutronic Analysis Code for Fuel Assembly Using a Vectorized Monte Carlo Method

Morimoto, Yuuichi; Maruyama, Hiromi; Ishii, Kazuya; Aoyama, Motoo

doi:10.13182/nse89-a23688

Cited by 45 publications

(14 citation statements)

References 4 publications

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“…Group constants of 12 energy groups for the core neutronic calculation were evaluated for the horizontal cross section of the fuel bundle lattice by the Monte Carlo calculation code with 190 energy groups [5]. In the burn-up calculation, 45 actinides from 253 Es and 84 fission products (83 nuclides treated explicitly and 1 lumped fission product) were treated.…”

Section: Core Calculation Methodsmentioning

confidence: 99%

Application of the Resource-Renewable Boiling Water Reactor for TRU Management and Long-Term Energy Supply

Hino

Ohtsuka

Takeda

et al. 2014

Nuclear Back-End and Transmutation Technology for Waste Disposal

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The RBWR (resource-renewable boiling water reactor) is an innovative BWR that has a capability to breed and burn trans-uranium elements (TRUs) using a multi-recycling process. The RBWR can be used as a long-term energy supply, and it reduces the negative environmental impact that TRUs cause as they are otherwise long-lived radioactive wastes. Various design concepts of the RBWR core have been proposed. The RBWR-AC is a break-even reactor and the RBWR-TB and RBWR-TB2 are TRU burners. The RBWR-TB is designed to burn TRUs from the RBWR-TB itself and to burn almost all the TRUs by repeating their recycling. The RBWR-TB is assumed to be applied for a nuclear power phase-out scenario. The RBWR-TB2 is intended to burn TRUs from LWR spent fuels. The RBWR-TB2 is assumed to be applied for reducing the amount of TRUs to be managed in storage facilities. The RBWR cores achieve their TRU multi-recycling capability under the constraint that the void reactivity coefficient must be negative by introducing the parfait core concept. This chapter reviews details of the specific design and core characteristics of the RBWR.

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Section: Core Calculation Methodsmentioning

confidence: 99%

Application of the Resource-Renewable Boiling Water Reactor for TRU Management and Long-Term Energy Supply

Hino

Ohtsuka

Takeda

et al. 2014

Nuclear Back-End and Transmutation Technology for Waste Disposal

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“…The ACS calculations used 200 batches of 2,000 neutrons and were compared to the reference MC calculation, which used 10,000 batches of 2,000 neutrons. The VMONT code 6,7) was used for the reference MC calculation and the number of energy groups was 190 for both the ACS and reference calculations. For the ACS calculations, initial neutron weights were set based on the neutron source distribution updated batch-wise.…”

Section: Applicability Of Acs Methods For Single Assembly Problemmentioning

confidence: 99%

“…Some of the most popular approaches for the pinwise whole core calculation are the Monte Carlo (MC) methods 1,[5][6][7] because of their flexible handling of targeted geometries. On the other hand, the methods have a shortcoming in that their calculation time would be enormous, particularly for cases with thermal-hydraulic iterations and/or burnup calculations.…”

Section: Introductionmentioning

confidence: 99%

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Multi-Assembly Analysis with an Advanced Correlated Sampling Method

Moriwaki

2003

Journal of Nuclear Science and Technology

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The Advanced Correlated Sampling (ACS) method, which can greatly enhance the calculation efficiency of Monte Carlo (MC) methods using a perturbation technique and a pseudo-scattering technique, is improved so as to be able to handle multi-assembly analysis. Two boundary connection methods-the neutron path connection and neutron current connection methods-are developed, and their accuracies for a 2×2 multi-assembly system are confirmed. The speedup factor against the corresponding direct MC calculation is a maximum of about 100. These boundary connection methods are also applied in pin-wise 2-D quarter core calculations as an example of a large-scale system. A newly developed perturbed neutron source iteration technique allows for very smooth source convergence. Furthermore, by incorporating the Direct Response Matrix (DRM) method as the acceleration means for the source convergence, the number of source iteration cycles is significantly shortened. The results confirm that this improved ACS method has the ability to perform pin-wise whole core calculations very accurately and efficiently.

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“…In the 3D-DRM method, the subresponse matrices were calculated using the fuel assembly analysis code VMONT, 7) which is based on a Monte Carlo neutron transport method. The VMONT code uses a multigroup model for the neutron spectrum calculation, and the total number of energy groups is 190.…”

Section: Verification Of Spectral History and Pin Power Correctimentioning

confidence: 99%

Development of Spectral History and Pin Power Correction Methods for Pin-by-Pin Core Analysis Method Using Three-Dimensional Direct Response Matrix

Mitsuyasu

Ishii

Hino

et al. 2010

J. Nucl. Sci. Technol.

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Spectral history and pin power correction methods have been developed for the pin-by-pin core analysis method using the three-dimensional direct response matrix (3D-DRM). The direct response matrix is formalized using four subresponse matrices in order to respond to a core eigenvalue k and thus it can be recomposed at each outer iteration in the core analysis. For core analysis, it is necessary to take into account the historical effect, which is related to spectral heterogeneity. The spectral history method is used to evaluate the nodal burn-up spectrum obtained by using the outgoing neutron current instead of the nodal flux because the 3D-DRM method does not use the nodal flux. The pin power correction method corrects the fuel rod neutron production rates obtained in the pin-by-pin calculation. These two methods were tested in a heterogeneous system. The test results show that the neutron multiplication factor error and nodal neutron production rate errors can be reduced by half during burn-up. The root-mean-square differences between the relative fuel rod neutron production rate distributions and the maximum error of relative fuel rod production rate can also be reduced by half. This means that the developed methods can reflect the effects of intra-and interassembly heterogeneities during burn-up and can be used for core analysis.

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Neutronic Analysis Code for Fuel Assembly Using a Vectorized Monte Carlo Method

Cited by 45 publications

References 4 publications

Application of the Resource-Renewable Boiling Water Reactor for TRU Management and Long-Term Energy Supply

Application of the Resource-Renewable Boiling Water Reactor for TRU Management and Long-Term Energy Supply

Multi-Assembly Analysis with an Advanced Correlated Sampling Method

Development of Spectral History and Pin Power Correction Methods for Pin-by-Pin Core Analysis Method Using Three-Dimensional Direct Response Matrix

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