• Aerospace Engineering and Technology
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• v.7 no.1
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• pp.68-75
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• 2008

Space radiation environments depend on satellite mission orbit, period, and date, and it can be predicted by simulation. Total Ionizing Dose(TID) can be predicted by Dose-depth Curve which only inform the dose level depending on the shielding thickness. So detail effective shielding analysis considering real structure is necessary to predict part level TID. For this purpose, program is developed to calculate shielding thickness distribution by structure modeling and ray trace from certain point in the structure. Finally TID at certain point in the 3-dimensional structure can be calculated by integration of shielding distribution result and dose-depth curve data. Using this program, TID is analyzed at part level certain point by modeling of equipment box structure in the satellite.

• Journal of Radiation Protection and Research
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• v.26 no.3
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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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• v.54 no.8
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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.

• Journal of radiological science and technology
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• v.38 no.4
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• pp.477-482
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• 2015

Workers in nuclear medicine have performed various tasks such as production, distribution, preparation and injection of radioisotope. This process could cause high radiation exposure to wokers' hand. The purpose of this study was to investigate shielding effect for r-rays of 140 and 511 keV by using Monte-carlo simulation. As a result, it was effective, regardless of lead thickness for radiation shielding in 140 keV r-ray. However, it was effective in shielding material with thickness of more than only 1.1 mm in 511keV r-ray. And also it doesn't effective in less than 1.1 mm due to secondary scatter ray and exposure dose was rather increased. Consequently, energy of radionuclide and thickness of shielding materials should be considered to reduce radiation exposure.

• The Journal of the Korea Contents Association
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• v.14 no.2
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• pp.457-466
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• 2014

During irradiation of lesions in cancer treatment with high energy electrons, normal tissue and critical organs are protected by the shielding material. Scattered radiation that generated the shielding materials affect the depth dose and atomic number. Therefore, we want to examine secondary particles and the scattered photons through calculation and its associated analysis, and compare the measurement for the aluminum, copper, and lead shielding substance of which thickness has 95% charge reduction. Dose change rate which effected scattering radiation was found to be +0.88% for material thickness, +0.43% for atomic number, and +19.70%, +15.20%, +12.40% for measurement, +25.00%, +15.10%, +13.70% for calculation on the aluminum, copper, and lead materials of which thickness has 95% charge reduction, respectively, As a result, we found that scattering rate was dependent on thickness than atomic number. In the dose increasing rate, scattered electrons are more important than scattered photon. For the above mentioned reasons, I think that high atomic number materials should be applied to reduce scattered radiation that generated with thickness effect.

• Proceedings of the Korea Concrete Institute Conference
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• 2003.05a
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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.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.2
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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.

• 대한방사선방어학회:학술대회논문집
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• 2011.11a
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• pp.178-179
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• 2011

• Nuclear Engineering and Technology
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• v.46 no.4
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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.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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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.