• 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.

• 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.

• Nuclear Engineering and Technology
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• v.21 no.4
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• pp.245-258
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• 1989

A 69-group cross section library consisting of more than 130 materials was generated for thermal reactor applications using the NJOY nuclear data processing system and the recent version of evaluated nuclear data files available from IAEA Nuclear Data Section. The multigroup library was validated through the analysis of various criticality experiments and depletion results of PWR. When used with the WIMS-KAERI code, the average $K_{eff}$ obtained for 47 uranium-oxide and 41 uranium metal fueled critical configurations is 0.9997 with a standard deviation of 0.69 percent. The calculated burnup dependent isotopic inventories of uranium and plutonium generally show good agreement with measured values obtained from depleted PWR pins.s.

• Nuclear Engineering and Technology
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• v.50 no.1
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• pp.35-42
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• 2018

This article presents clean core criticality calculations and control rod worth calculations for TRIGA (Training, Research, Isotope production-General Atomics) Mark II research reactor benchmark cores using Winfrith Improved Multi-group Scheme-D/4 (WIMS-D/4) and Program for Reactor In-core Analysis using Diffusion Equation (PRIDE) codes. Cores 133 and 134 were analyzed in 2-D (r, ${\theta}$) and 3-D (r, ${\theta}$, z), using WIMS-D/4 and PRIDE codes. Moreover, the influence of cross-section data was also studied using various libraries based on Evaluated Nuclear Data File (ENDF/B-VI.8 and VII.0), Joint Evaluated Fission and Fusion File (JEFF-3.1), Japanese Evaluated Nuclear Data Library (JENDL-3.2), and Joint Evaluated File (JEF-2.2) nuclear data. The simulation results showed that the multiplication factor calculated for all these data libraries is within 1% of the experimental results. The reactivity worth of the control rods of core 134 was also calculated with different homogenization approaches. A comparison was made with experimental and reported Monte Carlo results, and it was found that, using proper homogenization of absorber regions and surrounding fuel regions, the results obtained with PRIDE code are significantly improved.

• Nuclear Engineering and Technology
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• v.47 no.7
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• pp.791-803
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• 2015

The effective cross sections (XSs) in the direct whole core calculation code nTRACER are evaluated by the equivalence theory-based resonance-integral-table method using the WIMS-based library as an alternative to the subgroup method. The background XSs, as well as the Dancoff correction factors, were evaluated by the enhanced neutron-current method. A method, with pointwise microscopic XSs on a union-lethargy grid, was used for the generation of resonance-interference factors (RIFs) for mixed resonant absorbers. This method was modified by the intermediate-resonance approximation by replacing the potential XSs for the non-absorbing moderator nuclides with the background XSs and neglecting the resonance-elastic scattering. The resonance-escape probability was implemented to incorporate the energy self-shielding effect in the spectrum. The XSs were improved using the proposed method as compared to the narrow resonance infinite massbased method. The RIFs were improved by 1% in $^{235}U$, 7% in $^{239}Pu$, and >2% in $^{240}Pu$. To account for thermal feedback, a new feature was incorporated with the interpolation of pre-generated RIFs at the multigroup level and the results compared with the conventional resonance-interference model. This method provided adequate results in terms of XSs and k-eff. The results were verified first by the comparison of RIFs with the exact RIFs, and then comparing the XSs with the McCARD calculations for the homogeneous configurations, with burned fuel containing a mixture of resonant nuclides at different burnups and temperatures. The RIFs and XSs for the mixture showed good agreement, which verified the accuracy of the RIF evaluation using the proposed method. The method was then verified by comparing the XSs for the virtual environment for reactor applicationbenchmark pin-cell problem, as well as the heterogeneous pin cell containing burned fuel with McCARD. The method works well for homogeneous, as well as heterogeneous configurations.