• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.71-76
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• 1997

CANDU 원자로용 핵연료 다발의 양 끝에 있는 endcap과 endplate가 원자로의 노물리 특성에 미치는 영향이 MCNP와 WIMS-AECL 계산코드로 계산되었다. 이 계산에 의하면 end region을 고려한 경우의 차이가 0.15% 이내로 거의 무시할 수 있다. 그러므로 end region을 고려할 수 없는 격자코드로 계산을 수행해도 노물리 특성에 미치는 영향이 거의 무시될 수 있으므로 CANDU 원자로의 격자 특성 계산에 사용될 수 있음이 증명되었다.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.103-108
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• 1997

압력용기로의 속중성자 조사량 평가를 4군 노달 노심해석코드로 수행하였다. 이 코드는 MCNP에 비해 정확성은 떨어지나, 핵연료 연소의 효과나 핵연료 장전 모형의 영향을 쉽게 고려할 수 있었다. 속중성자 조사량 감소 방안으로서 반사체 차폐 구조물을 설치하는 방안과 노심외곽에 대체 핵연료 집합체를 장전하는 방안을 비교하였다. 신형원전의 경우 가장 효과적인 방안은 물 반사체 영역에 금속 차폐 구조물을 설치하는 것이나 운전중인 원자로의 경우 비록 주기길이의 감소와 핵연료 비용의 증가는 있으나 속중성자 감소 효과에 있어서는 대체 핵연료 집합체의 장전이 대안일 수 있다.

• Journal of Radiation Protection and Research
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• v.21 no.1
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• pp.17-25
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• 1996

In this study, to set-up the calculation method of radiation shielding of the KSC-4 shipping cask which is being used for spent fuel transportation, the pre-existing two calculation methods, deterministic and probabilistic methods were tested. For the first, the DOT4.2 computer code adopting the deterministic theory was applied for the calculation of effective neutron shielding under assumption of continuous wall thickness of the cask. To verify the first results, the probabilistic theory was used as an alternate calculation. In this case MCNP4A computer code adopting the probabilitic theory was used. And same approximation was obtained from the two different shielding calculations. From the results, it could be confirmed that the design and calculation method used for the radiation shielding of the KSC-4 was adequate and sufficiently safe to meet the design and QA requirements of 10CFR71 Appendix H.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.931-936
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• 1995

서울시 중심부 300m 상공에서 약 22kT의 플루토늄 원폭이 폭발했을 때를 가정하고 폭발시 나오는 초기 방사선에 의한 선량을 계산하였다. 계산을 위하여 몬테칼로 코드인 MCNP4A를 이용하였으며 방사선의 위해도를 알아보기 위하여 선량당량으로 환산 하였다. 계산 결과 가까운 거리에서는 평균자유행로가 짧은 중성자에 의한 선량이 높게 나왔으나 거리가 멀어질수록 감마선에 의한 영향이 더 큰 것으로 나타났다.

• Journal of Radiation Protection and Research
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• v.22 no.1
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• pp.47-58
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• 1997

The conversion coefficients of effective dose per unit air kerma and equivalent dose per unit fluence were calculated by MCNP4A code for antero-posterior(AP) and postero- anterior(PA) incidence of broad, unidirectional beams of photons into anthropomorphic MIRD phantom. Calculations have been performed for 20 monoenergetic photons of energy ranging from 0.03 to 10 MeV. The conversion coefficients showed a good agreement with the corresponding values given in the draft publication of joint task group of ICRP and ICRU within 10%. The deviations may arise from the differences of geometry in the MIRD phantom and the ADAM/EVE phantoms, and the differences in the codes and cross-section data used. Inclusion of a specific oesophagus model results in effective dose slightly different(5% at most) from the effective doses obtained by adopting the equivalent doses for the thymus or pancreas. Deletion of the ULI from the remainder organ appeared not to be significant for the cases of photon dosimetry covered in this study.

• Journal of the Korea Computer Industry Society
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• v.8 no.1
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• pp.9-16
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• 2007

In this study, the change of neutron detection can be use the basic data of asphalt content detector under the influence of the jurisdiction and usage of radioisotopes are limited of $100[{\mu}Ci]$ or less. To obtain neutron detector's properties using design materials in first step phase, the change of neutron detection is to be calculated how can be increase or decrease due to the change of asphalt content, also it look over the change results which is installed absorber(cadmium plate) around moderator(polyethylene) using MCNP Code.

• Journal of the Korea Computer Industry Society
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• v.7 no.1
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• pp.7-14
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• 2006

In Korea, under the influence of the jurisdiction, usage of radioisotopes are limited. The limitation is $100[{\mu}Ci]$ or less. Therefore, in this study, basic data were designed, and the following data are needed in order to improve content measuring instrument which is suitable for radioisotopes limitation. Owing to the source and detector's properties, measuring instrument was designed geometrically, neutron and photon's particle transportation was analysed by using the MCNP code which is in Monte Carlo Method, also the location of source and detectors, geometrical structure of neutron absorber and moderator was designed.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.55-60
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• 1995

CANDU형 원자로의 노심해석을 위해 핵연료 격자 및 반응도 설비(reactivity devices)에 대한 2군 군정수가 필요하다. 특히 CANDU형 원자로의 노심해석에 있어서 반응도 설비나 구조물은 증분격자 상수(Incremental Cross Section)에 의해 묘사된다. 현재 CANDU형 원자로의 반응도 설비의 증분격자 상수를 계산하기 위해 MULTICELL 코드를 사용하여 계산하고 있다. 그러나 weak absorber에 대해 기존의 증분격자 상수를 이용하여 계산한 반응도가는 시운전(Phase-B)조건에서의 노물리 시험치보다 다소 과소평가하고 있다. 본 연구에서는 증분격자 상수 계산 방법의 개선 방향을 모색하기 위해 SHETAN 및 MCNP 코드로 단일 격자에서의 반응도가를 계산하여 비교, 평가하였다. HCNP 계산의 결과는 조정봉(Adjuster rods)과 흡수봉/정지봉 (Mechanical Control Absorber/Shutoff rod)은 MULTICELL의 계산 결과보다 적으며, 경수영역 조절기(Liquid Zone Controller)는 크게 나타났다. 또한 SHETAN 코드를 이용한 결과는 MULTICELL의 결과보다 약간 크게 나타났다.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.2
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• pp.117-123
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• 2012

In order to dispose of the LILW(low and intermediate level radioactive waste) stored at KAERI, the radwaste drum assay system will be introduced to evaluate the radioisotopes inventory of stored drums. At present, the construction project of the dedicated assay facility to operate it and carry out routine maintenance of that equipment has been conducting at the radwaste treatment facility. Since that facility will be constructed in front of a 1st radwaste storage facility as well as the radwaste drums to be assayed and the transmission source in the radwaste drum assay system are in that facility, they could act as the radioactive sources and then, would affect the dose rate at the inside and the outside of the facility. Therefore, the radiation shielding should be evaluated through the concrete wall near to the radioactive sources whether the wall thickness is sufficient against the regulations. In this study, the radiation safety for the concrete wall around the radiation controlled area in the radwaste drum assay facility was evaluated by the MCNP code. From the evaluation results, the thickness of those concrete walls which are under consideration of about 30 cm was enough to shield the radiation from the radioactive sources.

• Journal of Radiation Protection and Research
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• v.30 no.2
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• pp.55-60
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• 2005

The evaluation of neutron flux distribution was performed for the ex-core detector design of SMART-P. DORT and MCNP code were used for the calculation of energy-dependent neutron flux distribution at 100% full power condition. Two code results show that maximum thermal flux appears at the $1^{st}$ water region in IST region and agree within 10% difference. In addition, another evaluation was performed code with assumptions that cote was composed of fission source and control rod without fuel assemblies. These assumptions make neutron count rate to be minimized. As a results, maximum thermal flux showed $6.99{\times}10^{-2}(n/cm^2-sec)$, when the strength of initial fission source was assumed as $1.0{\times}10^8(n/sec)$. The main reason of these results is due to the thermalization of fast neutrons in the water region and thermal flux is proportional to 80% of total neutron flux. Therefore, optimization of filler material of detector guide tube, position of installation and axial length of detector segments is necessary for the design of ex-core detector to enhance the neutron count rate and above results could be used in ex-core detector design as a fluence requirement.