• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 추계학술발표회요약집
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• pp.49.1-49
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• 2002

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1999년도 춘계학술발표회요약집
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• pp.31-31
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• 1999

• Nuclear Engineering and Technology
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• 제53권8호
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• pp.2445-2453
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• 2021

Since the nuclear data forms a vital component in reactor physics computations, the nuclear community needs processing codes as tools for translating the Evaluated Nuclear Data Files (ENDF) to simulate nuclear-related problems such as an ACE format that is used for MCNP. Errors, inaccuracies or discrepancies in library processing may lead to a calculation that disagrees with the experimentally measured benchmark. This paper provides an overview of the processing and preparation of ENDF/B-VIII.0 incident neutron data with NECP-Atlas and NJOY codes for implementation in the MCNP code. The resulting libraries are statistically inter-compared and tested by conducting benchmark calculations, as the mutualcomparison is a source of strong feedback for further improvements in processing procedures. The database of the benchmark experiments is based on a selection taken from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP handbook) and those proposed by Russell D. Mosteller. In general, there is quite good agreement between the NECP-Atlas1.2 and NJOY21^(1.0.0.json) results with no substantial differences, if the correct input parameters are used.

• Nuclear Engineering and Technology
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• 제55권9호
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• pp.3229-3240
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• 2023

Due to neutron radiation, atomic displacement has a significant effect on material in nuclear reactors. A range of secondary displacement models, including the Kinchin-Pease (K-P), Lindhard, Norgett-Robinson-Torrens (NRT), and athermal recombination-corrected displacement per atom (arc-dpa) have been suggested to calculate the number of displacement per atom (dpa). As neutron elastic interaction is the main cause of displacement damage, the focus of the current study is to calculate the atomic displacement caused by the neutron elastic interaction in order to estimate the exact amount of yield strength in a WWER-1000 reactor pressure vessel. To achieve this purpose, the reactor core is simulated by MCNPX code. In addition, a program is developed to calculate the elastic radiation damage induced by the incident neutron flux (RADIX) based on different models using Fortran programming language. Also, due to non-elastic interaction, the displacement damage is calculated by the HEATR module of the NJOY code. ASME E-693-01 standard, SPECTER, NJOY codes, and other pervious findings have been used to validate RADIX results. The results showed that the RADIX(arc-dpa)/HEATR outputs have appropriate accuracy. The relative error of the calculated dpa resulting from RADIX(arc-dpa)/HEATR is about 8% and 46% less than NJOY code, respectively in the ¼ and ¾ vessel wall.

• Nuclear Engineering and Technology
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• 제31권6호
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• pp.631-637
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• 1999

The module LEAPR of NJOY data processing system has been improved to have the capability of computing the thermal elastic scattering cross sections for silicon, which has a diamond-like structure. Silicon lattice was assumed as an fcc lattice with two atoms at each lattice point. The calculation formulas for thermal neutron elastic scattering by silicon were introduced and incorporated into LEAPR, and then the scattering cross sections for silicon were computed. The results were compared with experimental data, and they were found to give a good agreement with experimental data.

• Nuclear Engineering and Technology
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• 제46권4호
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• pp.475-480
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• 2014

We developed a simple method to prepare the displacement damage cross section of SiC using NJOY and SRIM/TRIM. The number of displacements per atom (DPA) dependent on primary knock-on atom (PKA) energy was computed using SRIM/TRIM and it is directly used by NJOY/HEATR to compute the neutron energy dependent DPA cross sections which are required to estimate the accumulated DPA of nuclear material. SiC DPA cross section is published as a table in DeCART 47 energy group structure. Proposed methodology can be easily extended to other materials.

• Nuclear Engineering and Technology
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• 제25권3호
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• pp.457-469
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• 1993

NJOY 코드 씨스템으로 JEF-1 평가 핵자료를 처리하여 고속로용 50군 군정수 SET을 생산하였다. 이를 이용하여 스물일곱가지 고속 임계로심 실험에서 얻어진 임계도 및 노심 중앙에서의 반응 율비를 계산하고 측정치와 비교·분석하였다. 아울러 ENDF/B-IV와-V 자료로 해석한 결과와도 비교·검토하였다. 일반적으로, 임계실험의 적분량 추정에서 JEF-1의 결과는 지금까지 사용해온 ENDF/B-IV의 결과보다 개선되었고, ENDF/B-V의 결과에 근접하고 있다.

• Nuclear Engineering and Technology
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• 제49권8호
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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.

• Nuclear Engineering and Technology
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• 제24권1호
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• pp.1-13
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• 1992

ORIGEN2 전산코드를 위한 연구로용 1군 단면적 데이타베이스를 개발하였다. 여기에는 EN-DF/B-lV와 -V가 기 본 핵 자료로 사용되었고 이들은 NJOY 코드시스템을 사용하여 69군으로 처리되었다. 1군 축약을 위한 가중함수는 핵연료의 연소에 따른 KMRR의 중성자 스펙트럼을 WIMS-KAERI코드로 계산하여 사용하였다. 새로 개발된 데이타베이스는 KMRR핵연료의 연소에 따른 악티나이드 생성량 평가를 통해 상세 다군 수송계산 결과와 잘 일치함이 입증되었다. 그리고 새로운 데이타베이스를 이용하여 KMRR의 사용후 핵연료 특성을 분석하였다.

• Nuclear Engineering and Technology
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• 제21권4호
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• pp.245-258
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• 1989

열중성자로의 핵계산을 위한 69군 단면적 라이브러리를 생산하였다. 기본 평가핵자료로는 IAEA Nuclear Data Section에서 수집된 자료가, 그리고 이를 처리하여 군정수화 하는데는 NJOY코드가 이용되었다. 새로이 마련된 라이브러리의 유용성을 검증하기 위해 각기 산화우라늄과 금속 우라늄 연료로 구성된 임계실험치를 WIMS-KAERI 코드로 계산된 결과와 비교, 검토하였다. 총 88임계결과에 대해 평균 $K_{eff}$ 값 0.9997, 그리고 표준 편차 0.69%를 보였다. PWR 연료의 연소결과로 얻어진 우라늄과 플루토늄 생성량에 대한 평가에서도 전반적으로 좋은 결과를 얻었다.다.