• Journal of the Korean Society of Radiology
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• v.15 no.7
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• pp.965-973
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• 2021

Since radiation therapy is irradiated with high-energy X-rays in a variety of at least 20 Gy to 80 Gy, a high dose is administered to the local area where the tumor is located, and various side effects of some normal tissues are expected. Currently, in clinical practice, lead, a representative material, is used as an effort to shield normal tissues, but lead is classified as a heavy metal harmful to the human body, and a large amount of skin contact can cause poisoning. Therefore, this study intends to manufacture a measurement sheet that can compensate for the limitations of lead using the materials Tungsten, Brass, and Copper of the 3D printer of the FDM (Fused Deposition Modeling) method and to investigate the penetration performance. Tungsten mixed filament transmission measurement sheet size was 70 × 70 mm and thickness 1, 2, 4 mm using a 3D printer, and a linear accelerator (TrueBeam STx, S/N: 1187) was measured by irradiating 100 MU at SSD 100 cm and 5 cm in water using a water phantom, an ion chamber (FC-65G), and an elcetrometer (PTW UNIDOSE), and the permeability was evaluated. As a result of increasing the measurement sheet of each material by 1 mm, in the case of Tungsten sheet at 3.8 to 3.9 cm in 6 MV, the thickness of the lead shielding body was thinner than 6.5 cm, and in case of Tungsten sheet at 4.5 to 4.6 cm in 15 MV. The sheet was thinner than the existing lead shielding body thickness of 7 cm, and equivalent performance was confirmed. Through this study, the transmittance measurement sheet produced using Tungsten alloy filaments confirmed the possibility of transmission shielding in the high energy region. It has been confirmed that the usability as a substitute is also excellent. It is thought that it can be provided as basic data for the production of shielding agents with 3D printing technology in the future.

• Nuclear Engineering and Technology
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• v.52 no.10
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• pp.2204-2220
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• 2020

An open-pool type research reactor is designed and operated considering the accessibility around the pool top area to enhance the reactor utilization. The reactor structure assembly is placed at the bottom of the pool and filled with water as a primary coolant for the core cooling and radiation shielding. Most radioactive materials are generated from the fuel assemblies in the reactor core and circulated with the primary coolant. If the primary coolant goes up to the pool surface, the radiation level increases around the working area near the top of the pool. Hence, the hot water layer is designed and formed at the upper part of the pool to suppress the rising of the primary coolant to the pool surface. The temperature gradient is established from the hot water layer to the primary coolant. As this temperature gradient suppresses the circulation of the primary coolant at the upper region of the pool, the radioactive primary coolant rising up directly to the pool surface is minimized. Water mixing between these layers is reduced because the hot water layer is formed above the primary coolant with a higher temperature. The radiation level above the pool surface area is maintained as low as reasonably achievable since the radioactive materials in the primary coolant are trapped under the hot water layer. The key to maintaining the stable hot water layer and keeping the radiation level low on the pool surface is to have a stable flow of the primary coolant. In the research reactor with a downward core flow, the primary coolant is dumped into the reactor pool and goes to the reactor core through the flow guide structure. Flow fields of the primary coolant at the lower region of the reactor pool are largely affected by the dumped primary coolant. Simple, circular, and duct type discharge headers are designed to control the flow fields and make the primary coolant flow stable in the reactor pool. In this research, flow fields of the primary coolant and hot water layer are numerically simulated in the reactor pool. The heat transfer rate, temperature, and velocity fields are taken into consideration to determine the formation of the stable hot water layer and primary coolant flow. The bulk Richardson number is used to evaluate the stability of the flow field. A duct type discharge header is finally chosen to dump the primary coolant into the reactor pool. The bulk Richardson number should be higher than 2.7 and the temperature of the hot water layer should be 1 ℃ higher than the temperature of the primary coolant to maintain the stability of the stratified thermal layer.

• Proceedings of the Materials Research Society of Korea Conference
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• 2011.05a
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• pp.15-15
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• 2011

As you know, boron compounds, borax ($Na_2B_4O_5(OH)_4{\cdot}8H_2O$) etc. were known thousands of years ago. As for natural boron, it has two naturally occurring and stable isotopes, boron 11 ($^{11}B$) and boron 10 ($^{10}B$). The neutron absorption $^{10}B$ is included about 19~20% with 80~81% $^{11}B$. Boron is similar to carbon in its capability to form stable covalently bonded molecular networks. The mass difference results in a wide range of ${\beta}$ values between the $^{11}B$ and $^{10}B$. The $^{10}B$ isotope, stable with 5 neutrons is excellent at capturing thermal neutrons. For example, it is possible to decrease a thermal neutron required for the nuclear reaction of uranium 235 ($^{235}U$). If $^{10}B$ absorbs a neutron ($^1n$), it will change to $^7Li+^1{\alpha}$ (${\alpha}$ ray, like $^4He$) with prompt ${\gamma}$ ray from $^{11}B$ $^{11}B$ (equation 1). $$^{10}B+^1n\;{\rightarrow}\;^{11}B\;{\rightarrow}\; prompt \;{\gamma}\;ray (478 keV), \;^7Li+4{\alpha}\;(4He)\;\;\;\;{\cdots}\; (1)$$ If about 1% boron is added to stainless steel, it is known that a neutron shielding effect will be 3 times the boron free steel. Enriched boron or $^{10}B$ is used in both radiation shielding and in boron neutron capture therapy. Then, $^{10}B$ is used for reactivity control and in emergency shutdown systems in nuclear reactors. Furthermore, boron carbide, $B_4C$, is used as the charge of a nuclear fission reaction control rod material and neutron cover material for nuclear reactors. The $B_4C$ powder of natural B composition is used as a charge of a control material of a boiling water reactor (BWR) which occupies commercial power reactors in nuclear power generation. The $B_4C$ sintered body which adjusted $^{10}B$ concentration is used as a charge of a control material of the fast breeder reactor (FBR) currently developed aiming at establishment of a nuclear fuel cycle. In this study for new boron compound, silicon boride ceramics for capturing thermal neutrons, preparation and characterization of both silicon tetraboride ($SiB_4$) and silicon hexaboride ($SiB_6$) and ceramics produced by sintering were investigated in order to determine the suitability of this material for nuclear power generation. The relative density increased with increasing sintering temperature. With a sintering temperature of 1,923 K, a sintered body having a relative density of more than 99% was obtained. The Vickers hardness increased with increasing sintering temperature. The best result was a Vickers hardness of 28 GPa for the $SiB_6$ sintered at 1,923K for 1 h. The high temperature Vickers hardness of the $SiB_6$ sintered body changed from 28 to 12 GPa in the temperature range of room temperature to 1,273 K. The thermal conductivity of the SiB6 sintered body changed from 9.1 to 2.4 W/mK in the range of room temperature to 1,273 K.

• The Journal of Korean Society for Radiation Therapy
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• v.25 no.1
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• pp.87-90
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• 2013

Purpose: Introducing CR (Computed Radiography) system created a process of printing therapy irradiation images and converting the degree of enlargement. This is to increase job efficiency and contribute to work improvement using a computerized method with home grown software to simplify this process, work efficiency. Materials and Methods: Microsoft EXCEL (ver. 2007) and VISUAL BASIC (ver. 6.0) have been used to make the software. A window for each shield block was designed to enter patients' treatment information. Distances on the digital images were measured, the measured data were entered to the Excel program to calculate the degree of enlargement, and printouts were produced to manufacture shield blocks. Results: By computerizing the existing method with this program, the degree of enlargement can easily be calculated and patients' treatment information can be entered into the printouts by using macro function. As a result, errors in calculation which may occur during the process of production or errors that the treatment information may be delivered wrongly can be reduced. In addition, with the simplification of the conversion process of the degree of enlargement, no copy machine was needed, which resulted in the reduction of use of paper. Conclusion: Works have been improved by computerizing the process of block production and applying it to practice which would simplify the existing method. This software can apply to and improve the actual conditions of each hospital in various ways using various features of EXCEL and VISUAL BASIC which has already been proven and used widely.

• Journal of the Korean Society for Aeronautical & Space Sciences
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• v.38 no.12
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• pp.1209-1216
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• 2010

This paper dealt with an alternative for maximizing mass savings in spacecraft design by replacing conventional aluminum alloy housing used for various spacecraft avionics by composite materials. Key requirements were defined for the purpose of composite housing design with sufficient durability and various functionalities as well as more lightweight characteristics as compared with aluminum alloy widely-used for conventional electronics housing. Conceptual design was also carried out for manufacturing modular, lightweight composite electronic housing equipped with high thermal and electrical conductivities, EMI protection, and radiation shielding characteristics as well as excellent structural performance; feasibility of enhancing mass savings in spacecraft design was presented.

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.1036-1044
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• 2023

The combination of MCNP6 and the FISPACT codes was used to predict inventories of radioisotopes produced by neutron exposure of a sample in a research reactor. The detailed MCNP6 model of the Budapest Research Reactor and the specific irradiation geometry of the NAA channel was established, while realistic material cards were specified based on concentrations measured by PGAA and NAA, considering the precursor elements of all significant radioisotopes. The energy- and spatial distributions of the neutron field calculated by MCNP6 were transferred to FISPACT, and the resulting activities were validated against those measured using neutron-irradiated small and bulky targets. This approach is general enough to handle different target materials, shapes, and irradiation conditions. A general agreement within 10% has been achieved. Moreover, the method can also be made applicable to predict the activation properties of the near-vessel concrete of existing nuclear installations or assist in the optimal construction of new nuclear power plant units.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.1
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• pp.1-11
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• 2004

The hot cell facility for research activities related to the lithium reduction of spent fuel, which is designed to permit safe handling of source materials with radioactivity levels up to 1,385 TBq, is planned to be built. To meet this goal, the facility is designed to keep gamma and neutron radiation lower than the recommended dose-rate in normally occupied areas. The calculations peformed with QAD-CGGP and MCNP-4C are used to evaluate the proposed engineering design concepts that would provide acceptable dose-rates during a normal operation in hot cell facility. The maximum effective gamma dose-rates on the surfaces of the facility at operation area and at service area calculated by QAD-CGGP are estimated to be $2.10{\times}10^{-3}, 2.97{\times}10^{-3} and 1.01{\times}10{-1}$ mSv/h, respectively. And those calculated by MCNP-4C are $1.60{\times}10^{-3}, 2.99{\times}10^{-3} and 7.88{\times}10^{-2}$ mSv/h, respectively, The dose-rates contributed by neutrons are one order of magnitude less than that of gamma sources. Therefore, it is confirmed that the radiological design for hot cell facility satisfies the Korean criterion of 0.01 mSv/h for the operation area and 0.15 mSv/h for the service (maintenance) area.

• Journal of Radiation Protection and Research
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• v.46 no.2
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• pp.58-65
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• 2021

Background: The evaluation of skyshine distribution, release of airborne radioactive nuclides, and soil activation and groundwater migration were required for radiological assessment of the impact on the environment surrounding In-Flight (IF)-system facility of the RAON (Rare isotope Accelerator complex for ON-line experiment) accelerator complex. Materials and Methods: Monte Carlo simulation by MCNPX code was used for evaluation of skyshine and activation analysis for air and soil. The concentration model was applied in the estimation of the groundwater migration of radionuclides in soil. Results and Discussion: The skyshine dose rates at 1 km from the facility were evaluated as 1.62 × 10^-3 μSv·hr^-1. The annual releases of ³H and ¹⁴C were calculated as 9.62 × 10^-5 mg and 1.19 × 10^-1 mg, respectively. The concentrations of ³H and ²²Na in drinking water were estimated as 1.22 × 10^-1 Bq·cm^-3 and 8.25 × 10^-3 Bq·cm^-3, respectively. Conclusion: Radiological assessment of environmental impact on the IF-facility of RAON was performed through evaluation of skyshine dose distribution, evaluation of annual emission of long-lived radionuclides in the air and estimation of soil activation and groundwater migration of radionuclides. As a result, much lower exposure than the limit value for the public, 1 mSv·yr^-1, is expected during operation of the IF-facility.

• Journal of the Korean Society of Radiology
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• v.11 no.5
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• pp.335-341
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• 2017

Cyclotron is a device that accelerates positrons or neutrons, and is used as a facility for making radioactive drugs having short half-lives. Such radioactive drugs are used for positron emission tomography (PET), which is a medical apparatus. In order to make radioactive drugs from a cyclotron, a nuclear reaction must occur between accelerated positrons and a target. After the reaction, unncessary neutrons are produced. In the present study, radioactivation generated from the collisions between the concrete shielding wall and the positrons and neutrons produced from the cyclotron is investigated. We tracked radioactivated radioactive isotopes by conducting experiments using FLUKA, a type of Monte Carlo simulation. The properties of the concrete shielding wall were comparatively analyzed using materials containing impurities at ppm level and materials that do not contain impurities. The generated radioactivated nuclear species were comparatively analyzed based on the exposure dose affecting human body as a criterion, through RESRAD-Build. The results of experiments showed that the material containing impurities produced a total of 14 radioactive isotopes, and $^{60}Co$(72.50%), $^{134}Cs$(16.75%), $^{54}Mn$(5.60%), $^{152}Eu$(4.08%), $^{154}Eu$(1.07%) accounted for 99.9% of the total dose according to the analysis having the exposure dose affecting human body as criterion. The $^{60}Co$ nuclear species showed the greatest risk of radiation exposure. The material that did not contain impurities produced a total of five nuclear species. Among the five nuclear species, 54Mn accounted for 99.9% of the exposure dose. There is a possibility that Cobalt can be generated by inducive nuclear reaction of positrons through the radioactivation process of $^{56}Fe$ instead of impurities. However, there was no radioactivation because only few positrons reached the concrete wall. The results of comparative analysis on exposure dose with respect to the presence of impurities indicated that the presence of impurities caused approximately 98% higher exposure dose. From this result, the main cause of radioactivation was identified as the small ppm-level amount of impurities.

• Journal of Radiation Protection and Research
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• v.9 no.2
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• pp.90-96
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• 1984

Estimated dose rates in spent fuel pool storage with the extended fuel cycle core management were reviewed and compared with design limit after calculation with the aid of DLC-23/CASK(22 n, 18 g) nuclear data and ANISN code. Radioactivity and gamma spectrum within spent fuel assemblies were calculated with ORIGEN code by extended fuel cycle model. In the calculation of dose rate, the fuel pool geometry was assumed to be infinite slab. Also, composition materials and radiation source within assemblies which are being stored in pool storage were assumed to be uniformly distributed throughout all the assemblies. As a result of culculation of dose rate from stored assemblies and waterborne radionuclides in pool water, the calculated dose rates appear to be lower than design basis limit under normal condition as well as abnormal condition.