• Journal of Radiation Protection and Research
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• v.12 no.1
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• pp.34-39
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• 1987

Radiation shielding analysis for a 6MeV X-ray facility was carried out. The primary and leakage radiation for the facility can be evaluated based on the methodology in NCRP No. 49 and 51. The present study deals with radiation scattering analysis for the outside and inside door of the facility based on the albedo concept. The calculated dose rates were compared with the results of MORSE-CG code calculation and the measured data, resulting in a good agreement, even though there existed some deviation for the inside door. These results can be utilized to the radiation shielding design of the medical and industrial X and gamma ray facilities, and to the safety evaluation of these facilities.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.373-373
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• 1999

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.854-855
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• 2005

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.11a
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• pp.480-481
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• 2018

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.2
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• pp.73-80
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• 2011

The radiation shielding analysis for a Burnup-credit (BUC) cask designed under the management of Korea Radioactive Waste Management Corporation (KRMC) was performed to examine the contribution of each radiation source affecting dose rate distribution around the cask. Various radiation sources, which contain neutron and gamma-ray sources placed in active fuel region and the activation source, and imaginary nuclear fuel were all considered in the MCNP calculation model to realistically simulate the actual situations. It was found that the maximum external and surface dose rates of the spent fuel cask were satisfied with the domestic standards both in normal and accident conditions. In normal condition, the radiation dose rate distribution around the cask was mainly influenced by activation source ($^{60}Co$ radioisotope); in another case, the neutron emitted in active fuel region contributed about 90% to external dose rate at 1m distance from side surface of the cask. Besides, the contribution level of activation source was dramatically increased to the dose rates in top and bottom regions of the cask. From this study, it was recognized that the detailed investigation on the radiation sources should be performed conservatively and accurately in the process of radiation shielding analysis for a BUC cask.

• Progress in Medical Physics
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• v.15 no.4
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• pp.247-254
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• 2004

According to developing high energy linear accelerators and treatment methods, like (3 dimensional conformal radiotherapy (3D-CRT), intensity modulated radiotherapy (IMRT), many radiotherapy centers are replacing older linear accelerators with new higher technical machines. This often presents a shielding problem as the designed shield for the existing rooms is not adequate for the higher technical machines. Additional shielding in limited existing space becomes necessary. We are replacing older brachytherapy room with new higher technical linear accelerator for IMRT. This room is not adequate for the IMRT machine without additional shielding design. The logical development of optimum structural shielding designs with concrete and high density shielding blocks are presented. We obtained following results by comparison between the pre-calculating values and actual survey of completed LINAC installation. High density shielding blocks have more powerful radiation protection about 2 times.

• 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.

• Journal of Radiation Protection and Research
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• v.43 no.4
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• pp.154-159
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• 2018

Background: To prevent small leakage accidents, a real-time and direct detection system for small leaks with a detection limit below that of existing systems, e.g. $0.5gpm{\cdot}hr^{-1}$, is required. In this study, a small-size beta detector, which can be installed inside the reactor containment (CT) building and detect small leaks directly, was suggested and its feasibility was evaluated using MCNPX simulation. Materials and Methods: A target nuclide was selected through analysis of radiation from radionuclides in the reactor coolant system (RCS) and the spectrum was obtained via a silicon detector simulated in MCNPX. A window was designed to reduce the background signal caused by other nuclides. The sensitivity of the detector was also estimated, and its shielding designed for installation inside the reactor CT. Results and Discussion: The beta and gamma spectrum of the silicon detector showed a negligible gamma signal but it also contained an undesired peak at 0.22 MeV due to other nuclides, not the $^{16}N$ target nuclide. Window to remove the peak was derived as 0.4 mm for beryllium. The sensitivity of silicon beta detector with a beryllium window of 1.7 mm thickness was derived as $5.172{\times}10^{-6}{\mu}Ci{\cdot}cc^{-1}$. In addition, the specification of the shielding was evaluated through simulations, and the results showed that the integrity of the silicon detector can be maintained with lead shielding of 3 cm (<15 kg). This is a very small amount compared to the specifications of the lead shielding (600 kg) required for installation of $^{16}N$ gamma detector in inside reactor CT, it was determined that beta detector would have a distinct advantage in terms of miniaturization. Conclusion: The feasibility of the beta detector was evaluated for installation inside the reactor CT to detect small leaks below $0.5gpm{\cdot}hr^{-1}$. In future, the design will be optimized on specific data.

• Progress in Medical Physics
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• v.19 no.4
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• pp.276-284
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• 2008

A computed tomography (CT) is a powerful system for the effectively fast and accurate diagnosis. The CT system, therefore, has used substantially and developed for improving the performance over the past decade, resulting in growing concerns over the radiation dose from the CT. Advanced CT techniques, such as a multidetector row CT scanner and dual energy or dual source CT, have led to new clinical applications that could result in further increases of radiation does for both patients and workers. The objective of this study was to review the international guidelines of the shielding requirements for a CT facility required for a new installation or when modifying an existing one. We used Google Search Engine to search the following keywords: computed tomography, CT regulation or shield or protection, dual energy or dual source CT, multidetector CT, CT radiation protection, and regulatory or legislation or regulation CT. In addition, we searched some special websites, that were provided for sources of radiation protection, shielding, and regulation, RSNA, AAPM, FDA, NIH, RCR, ICRP, IRPA, ICRP, IAEA, WHO (See in Table 1 for full explanations of the abbreviations). We finally summarized results of the investigated materials for each country. The shielding requirement of the CT room design was very well documented in the countries of Canada, United States of America, and United Kingdom. The wall thickness of the CT room could be obtained by the iso-exposure contour or the point source method. Most of documents provided by international organizations were explained in importance of radiation reduction in patients and workers. However, there were no directly-related documents of shielding and patient exposure dose for the dual energy CT system. Based international guidelines, the guideline of the CT room shielding and radiation reduction in patients and workers should be specified for all kinds of CT systems, included in the dual energy CT. We proposed some possible strategies in this paper.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.4137-4141
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• 2021

As NPPs continue to operate, liquid waste continues to be generated, and containers are needed to store and transport them at low cost and high capacity. To transport and store liquid phase very low-level radioactive waste (VLLW), a container is designed by considering related regulations. The design was constructed based on the existing container design, which easily transports and stores liquid waste. The radiation shielding calculation was performed according to the composition change of barium sulfate (BaSO₄) using the Monte Carlo N-Particle (MCNP) code. High-density polyethylene (HDPE) without mixing the additional BaSO₄, represented the maximum dose of 1.03 mSv/hr (<2 mSv/hr) and 0.048 mSv/hr (<0.1 mSv/hr) at the surface of the inner container and at 2 m away from the surface, respectively, for a 10 Bq/g of ⁶⁰Co source. It was confirmed that the dose from the inner container with the VLLW content satisfied the domestic dose standard both on the surface of the container and 2 m from the surface. Although it satisfies the dose standard without adding BaSO₄, a shielding material, the inner container was designed with BaSO₄ added to increase radiation safety.