• Nuclear Engineering and Technology
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• 제11권4호
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• pp.295-301
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• 1979

열중성자에 의한 $U^{235}$의 핵분열 시에 가벼운 쪽 핵분열 생성물의 핵전하 분포를 질량수에 따라 계산했다. 1차 핵분열 생성물의 핵전하 분포를 계산해서 실험치와의 비교를 통해, 지금까지 게시된 핵분열 조각의 핵전하 분포를 결정하는 이론의 타당성을 검토한 결과 어느 이론도 핵분열 조각의 핵전하 분포를 정확하게 기술할 수 없는 것이 판명됐다. 핵분열 조각의 최빈전하 ($Z_{P}$)와 불변전하밀도($Z_{UCD}$, UCD : Unchanged Charge Density)와의 차이는 0.45~0.5 사이로 질량수에 따라 크게 변동하지 않음을 보여준다. 핵분열 조각의 핵전하 분포에서 분산도를 나타내는 표준편차 ($\sigma$)는 ~0.5의 값을 갖는 것으로 나타났다. 핵분열 생성물에서 중성자의 Odd-Even 효과가 나타나려면, 즉발중성자 방출확률에도 Odd-Even 효과가 나타나야 한다는 결론을 얻었다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2001년도 춘계학술발표회요약집
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• pp.92.2-92
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• 2001

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2001년도 추계학술발표회요약집
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• pp.65-65
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• 2001

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 춘계공동학술발표회요약집
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• pp.53-53
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• 2002

• Nuclear Engineering and Technology
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• 제36권4호
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• pp.338-345
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• 2004

This study examines the sensitivity of several fabrication parameters for dry process fuel, using a random sampling technique. The in-pile performance of dry process fuel with irradiation was calculated by a modified ELESTRES code, which is the CANDU fuel performance code system. The performance of the fuel rod was then analyzed using a Monte Carlo simulation to obtain the uncertainty of the major outputs, such as the fuel centerline temperature, the fission gas pressure, and the plastic strain. It was proved by statistical analysis that for both the dry process fuel and the $UO_2$ fuel, pellet density is one of the most sensitive parameters, but as for the fission gas pressure, the density of the $UO_2$ fuel exhibits insensitive behavior compared to that of the dry process fuel. The grain size of the dry process fuel is insensitive to the fission gas pressure, while the grain size of the $UO_2$ fuel is correlative to the fission gas pressure. From the calculation with a typical CANDU reactor power envelop, the centerline temperature, fission gas pressure, and plastic strain of the dry process fuel are higher than those of the $UO_2$ fuel.

• Nuclear Engineering and Technology
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• 제22권4호
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• pp.410-415
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• 1990

MN(Madland-Nix) 모형에 바탕을 둔 핵분열 즉발 중성자 에너지 스펙트럼을 핵분열 유기중성자의 에너지 함수로서 구했다. 이 중성자 스펙트럼을 $^{27}$ Al(n,$\alpha$), $^{32}$S(n,p) 및 $^{115}$ In(n,n') 발단 방사화 검출기의 여기함수에 증율시켜 평균핵반응 단면적을 계산하고 상호간의 비, 즉, 중성자 스펙트럼 지수를 구했다. 그 결과 핵분열 유기중성자의 에너지에 따라 중성자 스펙트럼은 변하며 그에 따라 중성자 스펙트럼 지수도 변함을 보였다. 이것은 곧 중성자 스펙트럼 지수를 실험적으로 결정하므로써 핵분열 즉발 중성자 에너지 스펙트럼을 결정할 수 있음을 의미한다.

• Nuclear Engineering and Technology
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• 제45권7호
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• pp.921-928
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• 2013

A future nuclear energy system is being developed at Korea Atomic Energy Research Institute (KAERI), the system involves a Sodium Fast Reactor (SFR) linked with the pyro-process. The pyro-process produces a source material to fabricate a SFR fuel rod. Therefore, an isotopic fissile content assay is very important for fuel rod safety and SFR economics. A new technology for an analysis of isotopic fissile content has been proposed using a lead slowing down spectrometer (LSDS). The new technology has several features for a fissile analysis from spent fuel: direct isotopic fissile assay, no background interference, and no requirement from burnup history information. Several calculations were done on the designed spectrometer geometry: detection sensitivity, neutron energy spectrum analysis, neutron fission characteristics, self shielding analysis, and neutron production mechanism. The spectrum was well organized even at low neutron energy and the threshold fission chamber was a proper choice to get prompt fast fission neutrons. The characteristic fission signature was obtained in slowing down neutron energy from each fissile isotope. Another application of LSDS is for an optimum design of the spent fuel storage, maximization of the burnup credit and provision of the burnup code correction factor. Additionally, an isotopic fissile content assay will contribute to an increase in transparency and credibility for the utilization of spent fuel nuclear material, as internationally demanded.

• Nuclear Engineering and Technology
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• 제55권11호
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• pp.3983-3995
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• 2023

In fast reactors, restructuring of the fuel micro-structure driven by high temperature and high temperature gradient can cause the formation of columnar grains. The non-spheroidal shape and the non-uniform temperature field in such columnar grains implies that standard models for fission gas diffusion can not be applied. To tackle this issue, we present a reduced order model for the fission gas diffusion process which is applicable in different geometries and with non-uniform temperature fields, maintaining a computational requirement in line with its application in fuel performance codes. This innovative application of reduced order models as meso-scale tools within fuel performance codes represents a first-of-a-kind achievement that can be extended beyond fission gas behaviour.

• Nuclear Engineering and Technology
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• 제15권4호
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• pp.280-285
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• 1983

핵분열 중성자스펙트럼이 원자로핵계산에 미치는 영향을 고속임계로심의 임계계산을 통해 다각적으로 분석하였다. 검토된 내용은 Maxwell식과 Watt-Cranberg 형의 식 적용, 영역별 스펙트럼자료의 적용, Maxwell식에서 핵온도 선정에 따른 효과, 우라늄과 풀루토늄의 가중평균된 스펙트럼자료의 적용, 노심내 전에너지에 따른 핵온도를 고려한 스펙트럼자료의 적용 등이다.

• Nuclear Engineering and Technology
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• 제49권6호
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• pp.1211-1218
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• 2017

Monte Carlo neutron transport codes generally use the method of successive generations to converge the fission source distribution to-and then maintain it at-the fundamental mode. Recently, a phenomenon called "clustering" has been noted, which produces fission distributions that are very far from the fundamental mode. In this study, a mathematical model of clustering in Monte Carlo has been developed. The model draws on previous work for continuous-time birth-death processes, as well as methods from the field of population genetics.