• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.197-202
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• 1997

ORIGEN2코드의 검증계산을 통해 PWR 사용 후 핵연료 조성핵종의 핵종량에 대한 핵임계측면에서 보수성을 가지는 안전인자를 산출하였고, MCNP코드의 검증계산으로 95/95 신뢰구간에서의 계산오차를 구하였다. 이를 바탕으로 직경이 1.2567cm이고 길이가 380.5cm인 196 개 금속봉을 장전한 캐니스터 ( 금속저장체 )가 x-y 방향으로 무한히 배열된 경우에 대해 캐니스터의 두께, 간격 및 외부의 공기중 수분농도에 따른 핵임계 안전해석을 수행하였다. 그 결과, 캐니스터의 두께가 7mm일 때 공기중 수분농도가 0.30 g/㎤이고 캐니스터간의 간격이 6.0cm인 경우의 최종핵 임계도값은 0.94130로서 최대허용핵임계값 (0.942)보다 적은 값을 보였다.

• Journal of the Korean Society for Nondestructive Testing
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• v.14 no.4
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• pp.228-236
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• 1995

In this paper the method to evaluate shielding deficiency by gamma scanning test was presented and verified theoretically by Monte Carlo code which is one of the best effective method for radiation shielding calculation. The cylindrical shielding model was selected to evaluate shielding deficiency by gamma scanning test. First, the reference shielding according to the design requirement of cask was fabricated specially and reference values were measured with Co-60 source and scintillation detector. As a result with which calculated the reference values, it is shown that maximum deficiency thickness for lead of true cylindrical shielding model was 12mm. To verify this, thickness of lead was calculated by MCNP code and maximum deficiency thickness was 11.6mm. The experimental result obtained by the use of reference shielding was in good agreement with the theoretical result within 4.1%. So, this method can be applied to inspect the shielding ability for great shielding or cask which the radioactive material is used. To perform measurement more exactly, the further work on the development of measuring equipment to display the results on the screen will be required.

• Journal of Radiation Protection and Research
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• v.37 no.4
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• pp.197-201
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• 2012

Annual dose at the boundary of the interim storage facility at normal condition was calculated to estimate the site area of the facility of PWR spent nuclear fuel. In this work, source term was generated by ORIGEN-ARP for 4.5 wt% initial enrichment, 45,000 MWd/MTU burnup and 10 years cooling time. Modeling of the storage facilities and radiation shielding evaluations were conducted by MCNP code depending on the storage capacity. In the case of the centralized storage system, the required site area was found to have the radius of more than 700 m.

• Journal of Radiation Protection and Research
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• v.38 no.4
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• pp.189-193
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• 2013

Annual dose on the containment building wall of the interim storage facility at normal condition was calculated to estimate the dose rate transition of the facility of PWR spent nuclear fuel. In this study, source term was generated by ORIGEN-ARP with 4.5 wt% initial enrichment, 45,000 MWd/MTU burnup and 10 years cooling time. Modeling of the storage facility and the containment building and radiation shielding evaluations were conducted by MCNP code depending on the distance between the wall and the facility in the building. In the case of the centralized storage system, the distance required for the annual dose rate limit from 10CFR72 was estimated to be 50 m.

• Journal of the Korean Society of Radiology
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• v.1 no.3
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• pp.11-16
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• 2007

In this study, the potential of a newly developed simulation toolkit, GATE for the simulation of electronic portal imaging devices (EPID) in radiation therapy was evaluated by characterizing the performance of the metal plate/phosphor screen detector for EPID. We compared the performances of the GATE simulator against MCNP4B code and experimental data obtained with the EPID system in order to validate its use for radiation therapy.

• Journal of Radiation Protection and Research
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• v.24 no.4
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• pp.205-213
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• 1999

Effective dose conversion coefficients from unit activity radionuclides contaminated on the ground surface were calculated by using MCNP4A rode and male/female anthropomorphic phantoms. The simulation calculations were made for 19 energy points in the range of 40 keV to 10 MeV. The effective doses E resulting from unit source intensity for different energy were compared to the effective dose equivalent $H_E$ of previous studies. Our E values are lower by 30% at low energy than the $H_E$ values given in the Federal Guidance Report of USEPA. The effective dose response functions derived by polynomial fitting of the energy-effective dose relationship are as follows: $f({\varepsilon})[fSv\;m^2]=\;0.0634\;+\;0.727{\varepsilon}-0.0520{\varepsilon}^2+0.00247{\varepsilon}^3,\;where\;{\varepsilon}$ is the gamma energy in MeV. Using the response function and the radionuclide decay data given in ICRP 38, the effective dose conversion coefficients for unit activity contamination on the ground surface were calculated with addition of the skin dose contribution of beta particles determined by use of the DOSEFACTOR code. The conversion coefficients for 90 important radionuclides were evaluated and tabulated. Comparison with the existing data showed that a significant underestimates could be resulted when the old conversion coefficients were used, especially for the nuclides emitting low energy photons or high energy beta particles.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.2
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• pp.91-100
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• 2016

The neutron activation inventories in reactor vessel and its internals, and bio-shield of a PWR nuclear power plant were calculated to evaluate the effect of impurity elements contained in the structural materials on the activation inventory. Carbon steel is, in this work, used as the reactor vessel material, stainless steel as the reactor vessel internals, and ordinary concrete as the bio-shield. For stainless steel and carbon steel, one kind of impurity concentration was employed, and for ordinary concrete five kinds were employed in this study using MCNP5 and FISPACT for the calculation of neutron flux and activation inventory, respectively. As the results, specific activities for the cases with impurity elements were calculated to be more than twice than those for the cases without impurity elements in stainless and carbon steel. Especially, the specific activity for the concrete material with impurity elements was calculated to be 30 times higher than that without impurity. Neutron induced reactions and activation inventories in each material were also investigated, and it is noted that major radioactive nuclide in steel material is Co-60 from cobalt impurity element, and, in concrete material, Co-60 and Eu-152 from cobalt and europium impurity elements, respectively. The results of this study can be used for nuclear decommissioning plan during activation inventory assessment and regulation, and it is expected to be used as a reference in the design phase of nuclear power plant, considering the decommissioning of nuclear power plants or nuclear facilities.

• 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.

• Journal of the Korean Society of Radiology
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• v.14 no.5
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• pp.563-575
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• 2020

The purpose of this study is to evaluate the activation characteristics that occur in a linear accelerator for container security inspection. In the computer simulation design, first, the targets consisted of a tungsten (Z=74) single material target and a tungsten (Z=74) and copper (Z=29) composite target. Second, the fan beam collimator was composed of a single material of lead (Z=82) and a composite material of tungsten (Z-74) and lead (Z=82) depending on the material. Final, the concrete in the room where the linear accelerator was located contained magnetite type and impurities. In the research method, first, the optical neutron flux was calculated using the MCNP6 code as a F4 Tally for the linear accelerator and structure. Second, the photoneutron flux calculated from the MCNP6 code was applied to FISPACT-II to evaluate the activation product. Final, the decommissioning evaluation was conducted through the specific activity of the activation product. As a result, first, it was the most common in photoneutron targets, followed by a collimator and a concrete 10 cm deep. Second, activation products were produced as by-products of W-181 in tungsten targets and collimator, and Co-60, Ni-63, Cs-134, Eu-152, Eu-154 nuclides in impurity-containing concrete. Final, it was found that the tungsten target satisfies the permissible concentration for self-disposal after 90 days upon decommissioning. These results could be confirmed that the photoneutron yield and degree of activation at 9 MeV energy were insignificant. However, it is thought that W-181 generated from the tungsten target and collimator of the linear accelerator may affect the exposure when disassembled for repair. Therefore, this study presents basic data on the management of activated parts of a linear accelerator for container security inspection. In addition, When decommissioning the linear accelerator for container security inspection, it is expected that it can be used to prove the standard that permissible concentration of self-disposal.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.746-751
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• 1998

붕소 중성자 포획 요법(BNCT, Boron Neutron Capture Therapy)용 열외중성자빔의 개발을 위하여 방사성 동위원소인 Cf-252를 중성자 선원으로 사용하였으며 상대적으로 낮은 중성자속을 높이기 위하여 미임계 증배 집합체를 이용하였다. 이전에 계시된 미임계 증배 집합체는 높은 핵연료 농축도를 필요로 하는 단점이 있어 본 연구에서는 이를 감소시키기 위한 몇 가지 설계안을 제시하였다. 중성자빔 설계를 위하여 몬테칼로 방법을 이용한 전산코드인 MCNP를 이용, 타원형두뇌 팬텀 내에서 AD, AR ADDR및 각각의 선량성분 등을 계산함으로써 설계된 중성자빔의 특성분석을 수행하였다. 새롭게 개선되어 제시된 중성자빔의 설계는 상대적으로 낮은 핵연료 농측도를 보이면서 기존의 결과와 유사한 결과를 보여주고 있으며 특히 두뇌 팬텀 내에서의 선량률은 기존에 비해 매우 높은 값을 보임으로써 짧은 시간에 효과적으로 뇌종양을 치료할 수 있는 이 점이 있다.