• Proceedings of the Korean Nuclear Society Conference
• /
• 1996.05c
• /
• pp.547-552
• /
• 1996

확률론적 방법을 이용한 MCNP4A 전산코드를 이용하여 두꺼운 차폐체내에서 효과적인 분산감소기법에 대하여 가장 단순화된 모델을 이용하여 고찰하여 보았다. 등방점선원과 이를 둘러싼 반경 50cm의 납차폐체를 계산을 위한 모빌로 사용하여 차폐체 내부 각 영역과 외부에서의 평균선속을 계산하였다. 분산감소기법으로는 구역분할법과 Exponential transform을 적용하여 각 구역에서의 분산의 변화를 비교하였다. 계산결과 두꺼운 차폐체문제에서는 exponential transform이 가장 효과적인 분산감소기법으로 나타났고 이때 구역분할법을 통하여서는 상대오차의 크기를 더욱 줄일 수 있었다.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1996.11b
• /
• pp.639-646
• /
• 1996

직접적 또는 실험적인 방법들에 의한 밀도측정계기의 설계는 많은 시간의 소비와 인적, 물적인 비용의 소모가 요구되기 때문에 계산된 속(flux) 분포에 근거한 비실험적인 방법들이 일반적으로 사용된다. 이전에는 2그룹 확산방정식으로 밀도측정계기를 설계해 왔으나 복잡한 기하학적 모사에서의 한계로 인하여 계산시간의 문제로 도외시되었던 몬테카를로(Monte Carlo) 방법이 컴퓨터 기술의 발전으로 유용하게 되었다. 본 연구에서는 3차원 모델링이 가능하고 검증용 프로그램으로 알려져 있으며 몬테카를로방법을 사용하는 MCNP 코드를 이용하여 밀도측정계기의 기하학적 배치를 제시하고자한다.

• Journal of the Korean Society for Nondestructive Testing
• /
• v.14 no.4
• /
• pp.228-236
• /
• 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
• /
• v.37 no.4
• /
• pp.197-201
• /
• 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
• /
• v.38 no.4
• /
• pp.189-193
• /
• 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.

• Proceedings of the Korean Institute of Information and Commucation Sciences Conference
• /
• 2014.05a
• /
• pp.707-709
• /
• 2014

The radiation sources leaked from large-scale radiation leak accident like the Fukushima nuclear power plant accident or nuclear explosions can cause to the very large damage for us. So that the damage can be minimized, we have being developed a detector that can providing information about the location of the source to remove dangerous substances quickly than the conventional single detector. In this paper, we designed and implemented the radiation shield and the collimator for the development of the stereo radiation detector to detect contamination things using MCNP Simulation. And we analysed the test results of the radiation shield and collimator using the radiation source. The results of this paper will be used as the basis for improving the efficiency of the stereo radiation detector being studied currently.

• Journal of the Korean Society of Radiology
• /
• v.1 no.3
• /
• pp.11-16
• /
• 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
• /
• v.24 no.4
• /
• pp.205-213
• /
• 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)
• /
• v.14 no.2
• /
• pp.91-100
• /
• 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 the Korean Society of Radiology
• /
• v.14 no.5
• /
• pp.563-575
• /
• 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.