• International Journal of Contents
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• 제14권4호
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• pp.65-69
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• 2018

The objective of this study was to conduct a radiation shielding analysis for the facility equipped with radiation generator. The analysis was carried out in two aspects. First, from the aspect of the effect caused by primary and leakage radiation. Second, effect of scattered radiation was evaluated by applying a simple calculation method based on a scattering rate concept since effect of scattered radiation is significantly important at maze entrance of the radiation facility. The calculated results obtained using the simple method were compared to the results calculated using Geant4 code and the measured values. The results calculated by the suggested method indicate that slight error exists in a radiation shielding analysis done at the maze entrance comparing to other two results, while the results evaluated at the outside of the maze entrance door are relatively consistent with other values.

• Nuclear Engineering and Technology
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• 제47권3호
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• pp.380-387
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• 2015

A lead slowing-down spectrometer (LSDS) system is a promising nondestructive assay technique that enables a quantitative measurement of the isotopic contents of major fissile isotopes in spent nuclear fuel and its pyroprocessing counterparts, such as $^{235}U$, $^{239}Pu$, $^{241}Pu$, and, potentially, minor actinides. The LSDS system currently under development at the Korea Atomic Energy Research Institute (Daejeon, Korea) is planned to utilize a high-flux ($>10^{12}n/cm^2{\cdot}s$) neutron source comprised of a high-energy (30 MeV)/high-current (~2 A) electron beam and a heavy metal target, which results in a very intense and complex radiation field for the facility, thus demanding structural shielding to guarantee the safety. Optimization of the structural shielding design was conducted using MCNPX for neutron dose rate evaluation of several representative hypothetical designs. In order to satisfy the construction cost and neutron attenuation capability of the facility, while simultaneously achieving the aimed dose rate limit (< $0.06{\mu}Sv/h$), a few shielding materials [high-density polyethylene (HDPE)eBorax, $B_4C$, and $Li_2CO_3$] were considered for the main neutron absorber layer, which is encapsulated within the double-sided concrete wall. The MCNP simulation indicated that HDPE-Borax is the most efficient among the aforementioned candidate materials, and the combined thickness of the shielding layers should exceed 100 cm to satisfy the dose limit on the outside surface of the shielding wall of the facility when limiting the thickness of the HDPE-Borax intermediate layer to below 5 cm. However, the shielding wall must include the instrumentation and installation holes for the LSDS system. The radiation leakage through the holes was substantially mitigated by adopting a zigzag-shape with concrete covers on both sides. The suggested optimized design of the shielding structure satisfies the dose rate limit and can be used for the construction of a facility in the near future.

• Nuclear Engineering and Technology
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• 제52권8호
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• pp.1784-1791
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• 2020

The 14 MeV neutron generator facility is being developed by the Institute for Plasma Research India to conduct the lab scale experiments related to Indian breeding blanket system for ITER and DEMO. It will also be utilized for material testing, shielding experiments and development of fusion diagnostics. Occupational radiation exposure control is necessary for the all kind of nuclear facilities to get the operational licensing from governing authorities and nuclear regulatory bodies. In the same way, the radiation exposure for the 14 MeV neutron generator facility at the occupational worker area and accessible zones for general workers should be under the permissible limit of AERB India. The generator is designed for the yield of 10¹² n/s. The shielding assessment has been made to estimate the radiation dose during the operational time of the neutron generator. The facility has many utilities and constraints like ventilation ducts, accessible doors, accessibility of neutron generator components and to conduct the experiments which make the shielding assessment challenging to provide proper safety for occupational workers and the general public. The neutron and gamma dose rates have been estimated using the MCNP radiation transport code and ENDF -VII nuclear data libraries. The ICRP-74 fluence to dose conversion coefficients has been used for the assessment. The annual radiation exposure has been assessed by considering 500 h per year operational time. The provision of local shield near to neutron generator has been also evaluated to reduce the annual radiation doses. The comprehensive results of radiation shielding capability of neutron generator building and local shield design have been presented in the paper along with detailed maps of radiation field.

• Journal of Radiation Protection and Research
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• 제41권2호
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• pp.141-147
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• 2016

Background: A project for building a neutron time-of-flight (nTOF) facility is progressing. We expect that the construction will start in early 2016. Before that, a detailed simulation based on the current architectural drawings was performed to optimize the performance of our facility. Materials and Methods: Currently, several parts had been modified or changed from the original design to reflect requirements such as the layout of the electron beam line, shape of the vacuum chamber producing a neutron beam, and the underground layout of the nTOF facility. Detailed analysis for these modifications has been done with MCNP simulation. Results and Discussion: An overview of our photo-neutron source and KAERI nTOF facility were introduced. The numerical simulations for heat deposition, source term, and radiation shielding of KAERI nTOF facility were performed and the results are discussed. Conclusion: We are expecting that the construction of the KAERI nTOF facility will start in early 2016, and these results will be used as basic data.

• Nuclear Engineering and Technology
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• 제55권4호
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• pp.1210-1217
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• 2023

The assessment of radiation fields in the lithium loop pipes and dump tank during the operation were performed for International Fusion Materials Irradiation Facility - DEMO-Oriented NEutron Source (IFMIF-DONES) in order to obtain the radiation dose-rate maps in the component surroundings. Variance reduction techniques such as weight window mesh (produced with the ADVANTG code) were applied to bring the statistical uncertainty down to a reasonable level. The biological dose was given in the study, and potential shielding optimization is suggested and more thoroughly evaluated. The MCNP Monte Carlo was used to simulate a gamma particle transport for radiation shielding purposes for the current Li Systems' design. In addition, the shielding efficiency was identified for the Impurity Control System components and the dump tank. The analysis reported in this paper takes into account the radiation decay source from and activated corrosion products (ACPs), which is created by d-Li interaction. As a consequence, the radiation (resulting from ACPs and Be-7) shielding calculations have been carried out for safety considerations.

• Radiation Oncology Journal
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• 제13권1호
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• pp.101-104
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• 1995

High Dose Rate Remote Afterloading system was installed at Wonkwang University Hospital in January 1994. In this report, the calibration of a Gammamed 12-i High Dose Rate Remote Afterloading system and the radiation survey around the facility after design and construct a shieding room are discussed. The radiation survey of the facility indicates that the use of ordinary concrete shielding of existing room will provide adequate shielding. Also, the methodologies for performing source calibration are presented.

• 방사선산업학회지
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• 제17권4호
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• pp.443-449
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• 2023

KAERI(Korea Atomic Energy Research Institute) is conducting research and development of large-scale radiation generators and the latest radiation measuring instruments. In particular, research and development of security screening equipment using an electron beam accelerator and a neutron generator is in progress recently. Globally, 20 ft containers are used to transport imports and exports, and electron beam accelerators are radiation sources to measure the shape of the material inside the container during customs inspections in each country. KAERI is developing a device that can use an electron beam accelerator and a neutron generator sequentially to grasp the shape of various materials as well as the location of the internal target material. In this study, when using the neutron generator, the radiation dose and the degree of activation by neutron for the facility and surrounding environment, facility equipment were simulated using MCNP and FISPACT code. As a result, the shielding structures inside and outside the radiation control area were satisfactory to the reference level established conservatively based on the Korean Nuclear Act.

• Journal of Radiation Protection and Research
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• 제32권4호
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• pp.190-193
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• 2007

The objective of this study is to recommend the radiation protection design parameters from the shielding point of view for concrete wall between the decay tank room and the primary pump room in TRIGA Mark-II Research Reactor Facility. The shield design for this concrete wall has been performed with the help of Point-kernel Shielding Code Micro-Shield 5.05 and this design was also validated based on the measured dose rate values with Radiation Survey Meter (G-M Counter) considering the ICRP-60 (1990) recommendations for occupational dose rate limit ($10{\mu}Sv/hr$). The recommended shield design parameters are: (i) thickness of 114.3 cm Ilmenite-Magnetite Concrete (IMC) or 129.54 cm Ordinary Reinforced Concrete (ORC) for concrete wall A (ii) thickness of 66.04 cm Ilmenite-Magnetite Concrete (IMC) or 78.74 cm Ordinary Reinforced Concrete (ORC) for concrete wall B and (iii) door thickness of 3.175 cm Mild Steel (MS) on the entrance of decay tank room. In shielding efficiency analysis, the use of I-M concrete in the design of this concrete wall shows that it reduced the dose rate by a factor of at least 3.52 times approximately compared to ordinary reinforced concrete.

• Journal of Radiation Protection and Research
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• 제12권1호
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• pp.34-39
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• 1987

X-선 발생장치 시설에 대한 방사선 차폐해석은 두가지 측면에서 고려된다. 먼저 1차 방사선과 누설방사선에 의한 영향을 고려하여야 하며 본 연구에서는 NCRP-49 및 51에 제시된 방법을 사용하였다. 둘째, 시설의 미로 출입구에서는 산란방사선에 의한 영향이 보다 중요한데, 본연구에서는 이 영향을 평가하기 위하여 Albedo 개념을 근거로 한 단순한 방법이 제시되었고, 이를 토대로 계산된 결과를 컴퓨터 코-드(MORSE-CG)의 계산치 및 실측된 결과와 비교하였다. 본 연구에서 제시된 방법에 의한 결과는 미로출입구 안쪽의 경우에 대한 해석에서는 약간의 오차를 보이고 있으나 미로출입구 문바깥에서는 비교적 잘 일치한다. 여기서 출입문바깥의 경우에 대한 해석이 방사선방어 측면에서 보다 중요하다는 관점에서 볼 때 이 방법은 의료용 또는 산업용의 X-선 및 감마선시설의 차폐설계 및 해석에 응용될 수 있을 것이라고 판단된다.

• Nuclear Engineering and Technology
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• 제45권5호
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• pp.695-700
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• 2013

Thirty-eight radiation safety reports for brachytherapy equipment were evaluated to determine the current status of brachytherapy units in Korea and to assess how radiation oncology departments in Korea complete radiation safety reports. The following data was collected: radiation safety report publication year, brachytherapy unit manufacturer, type and activity of the source that was used, affiliation of the drafter, exposure rate constant, the treatment time used to calculate workload and the HVL values used to calculate shielding design goal values. A significant number of the reports (47.4%) included the personal information of the drafter. The treatment time estimates varied widely from 12 to 2,400 min/week. There was acceptable variation in the exposure rate constant values (ranging between 0.469 and 0.592 ($R{\cdot}m^2/Ci{\cdot}hr$), as well as in the HVLs of concrete, steel and lead for Iridium-192 sources that were used to calculate shielding design goal values. There is a need for standard guidelines for completing radiation safety reports that realistically reflect the current clinical situation of radiation oncology departments in Korea. The present study may be useful for formulating these guidelines.