• Nuclear Engineering and Technology
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• v.27 no.1
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• pp.74-83
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• 1995

The accurate determination of the fast neutron flux/fluence onto the pressure vessel is an essential part of the reactor pressure vessel surveillance program. It has been reported recently that the iron cross section data in ENDF/B versions III through V might underestimate the flux/fluence of fast neutrons in steel structures such as reactor pressure vessel. In this study, for the comparison of iron data of ENDF/B-IV and VI we produced two 47-group cross section sets, CXFe-IV and CXFe-Ⅵ, which are based on Yonggwang nuclear unit-3/4 model and the iron data of ENDF/B-IV and VI, respectively. A comparison was made of the results obtained from DOT4.3 calculation using CXFe-IV and CXFe-VI. From the results, it was found that the fast flux(E 〉 1.0 MeV), which is important for the pressure vessel embrittlement analysis, increases by about 7.6％ at the inner wall and 20％ at the outer wall of the vessel, if the iron data are used from ENDF/B-VI instead of ENDF/B-IV.

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.56.1-56
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• 2002

• Nuclear Engineering and Technology
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• v.49 no.8
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• pp.1610-1616
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• 2017

To validate the new Evaluated Nuclear Data File $(ENDF)/B-VIII.0{\beta}4$ library, 31 different critical cores were selected and used for a benchmark test of the important parameter keff. The four utilized libraries are processed using Nuclear Data Processing Code (NJOY2016). The results obtained with the $ENDF/B-VIII.0{\beta}4$ library were compared against those calculated with ENDF/B-VI.8, ENDF/B-VII.0, and ENDF/B-VII.1 libraries using the Monte Carlo N-Particle (MCNP(X)) code. All the MCNP(X) calculations of keff values with these four libraries were compared with the experimentally measured results, which are available in the International Critically Safety Benchmark Evaluation Project. The obtained results are discussed and analyzed in this paper.

• 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.251-256
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• 1996

일반화된 다군의 material 단면적 라이브러리 형식인 MATXS와 이를 각종 수송계산 코드에 적용할 수 있도록 하는 TRANSX 코드 체제를 소개하고 그 유용성을 검토하였다. 이를 위해 ENDF/B-VI.2를 이용하여 열 및 고속 임계노심 해석을 위한 각각의 라이브러리를 생산하고, 수송계산 코드인 ONEDANT를 이용하여 검증계산을 수행하였다. 열중성자 임계노심 해석결과 유효증배계수에서 약 0.3% 내외로 실험치에 근사한 결과를 얻었으며, 고속임계노심에서도 임계도 및 중심반응율비 결과가 실험치에 접근하고 있다.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.67-72
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• 1995

최근 선진 원자력 개발국들이 새로운 평가 핵자료집들을 계속하여 공개하고 있다. 이 핵자료들을 노심 채석에 적용하기 위한 연구의 일환으로 최근에 발표된 ENDF/B-V1.2 평가 핵자료집을 이용하여 CASMO-3의 단면적 Data Library를 만들어 검증해 보았다. 평가 핵자료집의 개선 정도와 계산상으로 만들어진 Library와 실제 사용되는 Library의 차이를 알아보기 위하여 BNDF/B-IV도 같이 처리하여 검증하였다. 유효증배계수와 integral parameter들을 비교한 결과 ENDF/B-VI의 유용성과 일관성이 입증되었고, 단면적 Library의 수정 작업에 관한 연구의 필요성도 제기되었다.

• 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.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.82-82
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• 2001

• Nuclear Engineering and Technology
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• v.37 no.1
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• pp.101-108
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• 2005

The neutron induced nuclear cross-section data for Pd-105, Ag-109, Xe-131, and Cs-133 were calculated and evaluated from an unresolved energy to 20 MeV. The energy dependent optical model potential parameters were extracted based on recent experimental data and applied up to 20 MeV. A spherical optical model and a statistical model for the equilibrium energy, and a multistep direct and a multistep compound model for the pre-equilibrium energy were used in the calculation. The direct capture model was recently introduced for fast neutron capture. The theoretically calculated cross-sections were compared with the experimental data and the evaluated files. The total and capture cross-sections calculated using the model were in good agreement with the reference experimental data. The evaluated cross-section results were compiled in ENDF-6 format and merged with the resonance component, already adopted in the ENDF/B-VI release 8. New data library files covering from thermal to 20 MeV were created. They are at the preliminary stage of an ENDF/B- VII release.

• The Journal of the Korea Contents Association
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• v.4 no.2
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• pp.76-82
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• 2004

The thermal neutron capture cross-section (at 2,200 m/s value) of the $^{141}$Pr(n,$\gamma$)$^{142}$Pr reaction was measured by an activation method by using the heavy water ($D_2$O) thermal neutron facility at the KUR(Kyoto University Reactor). The thermal neutron fiux used in this experiment was monitored with the$^{197}$Au(n,$\gamma$)$^{198}$Au standard cross-section. The previous results and the evaluated data of JENDL-3.2, ENDF/B-VI, and JEF-2.2 were in good agreement with the current result.