• 항공우주기술
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• 제7권1호
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• pp.68-75
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• 2008

우주방사선환경은 위성의 운용궤도와 임무 기간 및 시기에 따라 달라지고 시뮬레이션을 통 해 예측이 가능하다. 총 이온화 조사량(TID)의 경우 dose-depth 곡선으로 차폐두께에 따른 조사량을 알 수 있다. 그러나 이는 차폐두께에 따른 조사량의 정보만 보여주므로 실제 차폐 구조물의 형상에 따른 부품수준에서의 총 이온화 조사량을 예측하기 위해서는 구조물의 형태를 고려한 유효 방사선 차폐두께의 상세 분석이 필요하다. 이를 위해 다양한 구조형상을 3차원 좌표로 입력하여 모델링이 가능하게 하고 여기에 임의 지점에서 방사되는 ray를 이용하여 구조체의 전 방향에 대한 유효 차폐두께분포를 계산하는 프로그램을 개발하였다. 이 분포결과를 위성의 우주임무환경에서 예측되는 dose-depth 곡선 데이터와 결합하여 최종적으로 위성내부의 임의지점에서 예측되는 총 이온화 조사량을 계산함으로써 3차원 구조형상을 고려한 상세 분석이 가능하도록 하였다. 이를 이용하여 위성의 전자박스구조를 모델링하여 부품수준의 임의지점에서 예측되는 총 이온화 조사량을 분석하였다.

• 한국콘텐츠학회논문지
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• 제14권2호
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• pp.457-466
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• 2014

고 에너지 전자선 치료에 있어서 차폐물질은 종양조직 외 정상조직이나 주요장기를 보호하기 위해 사용된다. 하지만 이러한 물질에서 발생되어지는 산란선은 심부선량에 영향을 줄 수 있으며, 물질원자번호에 따라 다르게 나타난다. 이에 차폐물질로써 사용가능한 알루미늄, 구리, 납 등의 다양한 원자번호 물질을 전하 감약율 95% 되는 두께로 하여 측정과 MCNPX 모의계산으로 산란율을 비교분석하였다. 산란선 영향을 많이 받는 표면의 선량변화율은 최대 물질두께에서 +0.88%, 원자번호에서 +0.43%의 영향을 받으며, 전하 감약율 95% 되는 두께의 알루미늄, 구리, 납 물질은 측정에서 +19.70%, +15.20%, +12.40% 계산에서 +25.00%, +15.10%, +13.70%를 보였다. 이로 인해 산란율은 물질두께가 원자번호보다 많은 영향을 주며, 산란전자가 광자보다 많은 기여를 하고 있음을 알 수 있었다. 이에 임상에서의 적절한 차폐물질은 두께영향 산란선이 적게 방출되는 고 원자번호물질이 적당하다고 사료된다.

• Journal of Radiation Protection and Research
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• 제26권3호
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• pp.215-224
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• 2001

In order to comply with revised national regulationson radiological protection and to implement recent nuclear data and dose conversion factors, KOPEC developed a new point kernel gamma and beta ray shielding analysis computer program. This new code, named VisualShield, adopted mass attenuation coefficient and buildup factors from recent ANSI/ANS standards and flux-to-dose conversion factors from the International Commission on Radiological Protection (ICRP) Publication 74 for estimation of effective/equivalent dose recommended in ICRP 60. VisualShieid utilizes graphical user interfaces and 3-D visualization of the geometric configuration for preparing input data sets and analyzing results, which leads users to error free processing with visual effects. Code validation and data analysis were performed by comparing the results of various calculations to the data outputs of previous programs such as MCNP 4B, ISOSHLD-II, QAD-CGGP, etc.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.3073-3084
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• 2022

The propagation of radiation source uncertainties in spent nuclear fuel (SNF) cask shielding calculations is presented in this paper. The uncertainty propagation employs the depletion and source term outputs of the deterministic code STREAM as input to the transport simulation of the Monte Carlo (MC) codes MCS and MCNP6. The uncertainties of dose rate coming from two sources: nuclear data and modeling parameters, are quantified. The nuclear data uncertainties are obtained from the stochastic sampling of the cross-section covariance and perturbed fission product yields. Uncertainties induced by perturbed modeling parameters consider the design parameters and operating conditions. Uncertainties coming from the two sources result in perturbed depleted nuclide inventories and radiation source terms which are then propagated to the dose rate on the cask surface. The uncertainty analysis results show that the neutron and secondary photon dose have uncertainties which are dominated by the cross section and modeling parameters, while the fission yields have relatively insignificant effect. Besides, the primary photon dose is mostly influenced by the fission yield and modeling parameters, while the cross-section data have a relatively negligible effect. Moreover, the neutron, secondary photon, and primary photon dose can have uncertainties up to about 13%, 14%, and 6%, respectively.

• 대한방사선기술학회지:방사선기술과학
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• 제38권4호
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• pp.477-482
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• 2015

핵의학과에 근무하는 방사선작업종사자들은 방사성동위원소의 생산, 분배, 조제, 주입 등의 업무를 진행하며, 이러한 과정에서 손에 대한 방사선 피폭이 높게 발생한다. 이에 본 연구에서는 핵의학과에서 이용되는 방사성동위원소의 에너지로서 140 keV와 511 keV의 ${\gamma}$선에 대한 차폐효과를 몬테카를로 모의 모사를 통해 분석하였다. 모의실험 결과 140 keV ${\gamma}$선은 차폐체에 두께와 상관없이 모두 방사선에 대한 차폐효과가 발생되었으며, 511 keV의 ${\gamma}$선에서는 1.1 mm 이상에서 차폐효과가 발생되었다. 그러나 1.1 mm 미만에서는 2차적으로 발생된 산란선으로 인하여 차폐효과가 없었으며, 오히려 방사성동위원소의 피폭선량이 증가되었다. 따라서 효율적인 방사선 방어를 위해서는 핵종별 에너지에 따른 납 차폐체의 두께를 고려하여야 할 것이다.

• 대한방사선방어학회:학술대회논문집
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• 대한방사선방어학회 2011년도 추계 학술발표회 및 심포지엄
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• pp.178-179
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• 2011

• 한국콘크리트학회:학술대회논문집
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• 한국콘크리트학회 2003년도 봄 학술발표회 논문집
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• pp.985-990
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• 2003

In order to demonstrate Advanced Spent Fuel Conditioning Process (ACP), shielding facilities such as hot cell suitable to handling radionuclides and process property will be necessary. But the construction of new facilities needs much money, man-power and time, it is now scheduled to remodel the hot cell, which has already been installed and maintained at Irradiated Material Experiment Facility (IMEF) in the Korea Atomic Energy Research Institute (KAERI). The basic structure and concrete shield wall of hot cell partly have been constructed on the base floor in IMEF building in current status. And hot cell after remodeling will be used for carrying out the lab-scale experiment of ACP. The hot cell was built in accordance with 35 curies of fe-59(1.2 MeV) as design criteria of radiation dose limit. But the radioactive source of ACP is expected to be much higher than design criteria of IMEF, shielding ability of the hot cell in the current status is unsatisfactory to the hot test of ACP. Therefore shield wall shall be reinforced with heavy concrete, steel or lead. In this paper, dose rates are calculated according to ACP source, shielding materials, etc., and reinforcement structures are determined considering the current situation of hot cells, installation of shield windows and the easiness of work.

• 방사성폐기물학회지
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• 제21권2호
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• pp.193-204
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• 2023

To ensure radiological safety margin in the transport and storage of spent nuclear fuel, it is crucial to perform source term and shielding analyses in advance from the perspective of conservation. When performing source term analysis on UO₂ fuel, which is mostly used in commercial nuclear power plants, uranium and oxygen are basically considered to be the initial materials of the new fuel. However, the presence of impurities in the fuel and structural materials of the fuel assembly may influence the source term and shielding analyses. The impurities could be radioactive materials or the stable materials that are activated by irradiation during reactor power operation. As measuring the impurity concentration levels in the fuel and structural materials can be challenging, publicly available information on impurity concentration levels is used as a reference in this evaluation. To assess the effect of impurities, the results of the source term and shielding analyses were compared depending on whether the assumed impurity concentration is considered. For the shielding analysis, generic cask design data developed by KEPCO-E&C was utilized.

• Nuclear Engineering and Technology
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• 제46권4호
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• pp.547-556
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• 2014

Korea expects a shortage in storage capacity for spent fuels at reactor sites. Therefore, a need for more metal and/or concrete casks for storage systems is anticipated for either the reactor site or away from the reactor for interim storage. For the purpose of interim storage and transportation, a dual purpose metal cask that can load 21 spent fuel assemblies is being developed by Korea Radioactive Waste Management Corporation (KRMC) in Korea. At first the gamma and neutron flux for the design basis fuel were determined assuming in-core environment (the temperature, pressure, etc. of the moderator, boron, cladding, $UO_2$ pellets) in which the design basis fuel is loaded, as input data. The evaluation simulated burnup up to 45,000 MWD/MTU and decay during ten years of cooling using the SAS2H/OGIGEN-S module of the SCALE5.1 system. The results from the source term evaluation were used as input data for the final shielding evaluation utilizing the MCNP Code, which yielded the effective dose rate. The design of the cask is based on the safety requirements for normal storage conditions under 10 CFR Part 72. A radiation shielding analysis of the metal storage cask optimized for loading 21 design basis fuels was performed for two cases; one for a single cask and the other for a $2{\times}10$ cask array. For the single cask, dose rates at the external surface of the metal cask, 1m and 2m away from the cask surface, were evaluated. For the $2{\times}10$ cask array, dose rates at the center point of the array and at the center of the casks' height were evaluated. The results of the shielding analysis for the single cask show that dose rates were considerably higher at the lower side (from the bottom of the cask to the bottom of the neutron shielding) of the cask, at over 2mSv/hr at the external surface of the cask. However, this is not considered to be a significant issue since additional shielding will be installed at the storage facility. The shielding analysis results for the $2{\times}10$ cask array showed exponential decrease with distance off the sources. The controlled area boundary was calculated to be approximately 280m from the array, with a dose rate of 25mrem/yr. Actual dose rates within the controlled area boundary will be lower than 25mrem/yr, due to the decay of radioactivity of spent fuel in storage.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(2)
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• pp.847-852
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• 1995

A shielding analysis for KSC-7, the shipping cask for transporting the 7 PWR spent fuel assemblies, has been carried out. Radiation source term has been calculated on spent fuel with burnup of 50,000 MWD/MTU and 1.5 years cooling time by ORIGEN2 code. The shielding calculation for the cask has been made by using MCNP4A code with continuous cross section data library from ENDF/B-V. As a result of neutron dose rate analysis, another shielding calculational model on spent fuel shipping cask was provided which is using the Monte Carlo method.