• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.62-62
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• 2003

• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.63-63
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• 2003

• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.62.2-62
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.94-99
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• 1996

최근 열중성자 산란법칙 라이브러리 ENDF/B-VI Release-2가 제공된 바 있다. 여기에는 경수내 수소와 흑연내 탄소에 대한 산란법칙이 포함되어 있어, 이를 경수격자인 TRX와 BAPL로 WIMS 계산을 통하여 검증하였다. 온도에 따른 변화를 검증하기 위해 가압경수로와 흑연감속 기체냉각로의 단위격자에 대한 WIMS계산을 수행하였다. WIMS 라이브러리 생산에 Release-1, Release-2 및 자유기체모델을 이용하여 상대적 차이를 검증한 결과 Release-2는 대체적으로 Release-1보다 개선되었으나, 그 개선의 정도는 현저하지 않음을 보이 주고 있다.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.163-168
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• 1996

Simulations of Wolsong-1 Phase-B commissioning measurements have been performed, as part of the program to validate WIMS-AECL lattice cell calculations for application to CANDU reactor simulations in RFSP. A required component of these simulations is the calculation of incremental cross sections representing reactivity control devices in the reactor. The incremental cross section properties of the Wolsong-1 adjusters, Mechanical Control Absorbers (MCA) and liquid Zone Control Units (ZCU) are based on the WIMS-AECL/MULTICELL modelling methods and the results are compared with those of WIMS-AECL/DRAGON-2 modelling methods.

• Proceedings of the Korean Nuclear Society Conference
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• 2003.10a
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• pp.69.1-69.1
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• 2003

• Nuclear Engineering and Technology
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• v.45 no.7
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• pp.951-960
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• 2013

Energy group structure has a significant effect on the results of multigroup transport calculations. It is known that $UO_2-PuO_2$ (MOX) is a recently developed fuel which consumes recycled plutonium. For such fuel which contains various resonant nuclides, the selection of energy group structure is more crucial comparing to the $UO_2$ fuels. In this paper, in order to improve the accuracy of the integral results in MOX thermal lattices calculated by WIMSD-5B code, a swarm intelligence method is employed to optimize the energy group structure of WIMS library. In this process, the NJOY code system is used to generate the 69 group cross sections of WIMS code for the specified energy structure. In addition, the multiplication factor and spectral indices are compared against the results of continuous energy MCNP-4C code for evaluating the energy group structure. Calculations performed in four different types of $H_2O$ moderated $UO_2-PuO_2$ (MOX) lattices show that the optimized energy structure obtains more accurate results in comparison with the WIMS original structure.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.65-70
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• 1997

Computational benchmark calculations have been performed for CANDU DUPIC fuel lattice and core using a Monte Carlo code MCNP-4B with ENDF/B-V library. The eigenvalues of the DUPIC fuel lattice have been predicted by an integral transport code WIMS-AECL using ENDF/B-V library for different burnup steps and lattice conditions. The comparison has shown that the eigenvalues match those of MCNP-4B within 0.20% $\Delta$k difference between WIMS-AECL and MCNP-4B results. The calculation of a 2-dimensional CANDU core loaded with DUPIC fuel has shown that the eigenvalue predicted by a diffusion code RFSP using lattice parameters generated by WIMS-AECL matches that of MCNP-4B within 0.12%Δk and the largest bundle power prediction error is around 7.2%.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.214-219
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• 1997

CANDU원자로의 노심계산을 위한 WIMS-AECL에 적합한 최적의 cross section library를 선정하기 위하여, 연소도에 대한 무한증배계수(k$\infty$)의 변화와 핵연료온도반응도계수, 냉각재온도반응도계수, 감속재온도반응도계수 및 기포반응도들을 계산하여 그 결과를 비교하였다. 그 결과, WIMS-AECL에 WINFRITH와 ENDF/B-V를 사용한 경우 핵연료 온도계수를 제외하고는 무한증배계수와 반응도계수들의 계산차이는 유효연소도 영역에서 크지 않았다. 그러나 연소가 진행됨에 따라 차이가 커짐을 보여주고 있으며, 기존의 POWDERPUFS-V(PPV)결과와는 초기 연소도에서는 차이가 적으나, 연소도가 커짐에 따라 많은 차이를 보여주고 있다. 따라서, 연소된 핵연료 또는 Pu이 함유된 핵연료에 대한 격자실험의 자료에 의하여 평가 및 검증될 수 있을 것이다.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.64-69
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• 1997

The code WIMS-AECL has been used for the lattice analysis of DUPIC fuel. The lattice parameters calculated by the code is sensitive to the choice of number of parameters, such as the number of tracking lines, number of condensed groups, mesh spacing in the moderator region, other parameters vital to the calculation of probabilities and burnup analysis. We have studied this sensitivity with respect to these parameters and recommend their proper values which are necessary for carrying out the lattice analysis of DUPIC fuel.