Earticle

Background: Iridium-192 (192Ir) is one of the radioisotopes most commonly utilized for industrial radiography worldwide. In accordance with the Nuclear Safety Act of the Republic of Korea, the use of projectors containing 192Ir sources requires conducting a shielding evaluation for a given workplace. This evaluation usually uses gamma constants and half-value layers (HVLs). However, significant differences in these values among references can result in over- or underestimation of values. Materials and Methods: The radiological characteristics of the gamma constant and HVL of 192Ir sources employed in industrial radiography were studied with a focus on the effects of source configurations and usage conditions. The source configurations were categorized into point source, volume source (lateral/upward direction), and encapsulated source. The usage conditions, defined by variations in the lead thickness and the distance among the source, lead shield, and tally, enabled derivation of the HVL. Monte Carlo N-Particle version 6.2 (MCNP6.2) was used in these calculations. Results and Discussion: The gamma constant substantially varied depending on the source configuration, exhibiting up to 43%-lower dose rates for the volume source oriented in an upward direction in comparison to the point source. In contrast, usage conditions substantially affected the HVL, which increased from 3.27 mm to 5.51 mm depending on the lead thickness under the broad-beam geometry. Furthermore, the HVLs under narrow-beam conditions were approximately 16.49% lower than those under the broad-beam geometry. Conclusion: A method with new data on the gamma constants and HVLs for shielding evaluation of industrial radiography with 192Ir was presented. The relevant findings resolve inconsistencies in the data presented in the literature and enhance shielding evaluation accuracy. The proposed method not only provides practical standards for shielding evaluations but also ensures regulatory consistency.

Background: FCTC10 container is designed to transport 60Co radioactive sources used in irradiation industry. It belongs to Type B(U) Category III (yellow) package when being loaded with a 60Co source of 1.8 × 105 Ci. Materials and Methods: The container is constituted of shielding container, basket, protective cover and bracket. Shielding ability is provided mainly by stainless steel shells, tungsten alloy and lead among steel shells. Radiation level around the container has been calculated with both Monte Carlo simulations and measurements. Results and Discussion: It is proven that the shielding performance of the container fulfills the requirements in GB11806-2004 (Regulations for the safe transport of radioactive material, China Standard Press). Exposure doses to workers and to critical groups of public were calculated based on hypothetical exposure scene according to transport practice experience. Conclusion: The results show that doses to workers and public are less than the constraint dose considered in design, and the radiation level would be increased less than a factor of 2 under design basis accidents.

유방 X선 촬영장치는 유방암의 발생률이 급증하고 건강한 삶에 대한 관심이 높아지면서 검사건수 또한 매년 급격히 증가하고 있다. 이는 의료용 방사선을 사용하는 것으로 유방암의 조기발견에 중요한 역할을 하는 동시에 의료피폭의 원인이 되고 있다. 이에 자체 제작한 차폐복을 이용하여 검사부위를 제외한 타 부위에 대한 차폐 전 · 후의 방사선량율을 측정하고 감소율을 계산하였다. 그 결과, 차폐 전·후의 감소율은 최저 85.52%에서 최고 92.67%로 나타났다. 이는 유방촬영 시 효과적인 차폐복의 사용으로 촬영목적과 관계가 없는 부위의 피폭을 최소화 할 수 있고, 의료영상의 질적인 향상을 유지할 수 있을 것이다.

As the incidence of breast cancer increases rapidly and interest in a healthy life increases, the number of tests for mammography is also increasing rapidly every year. This is the use of medical radiation, which plays an important role in early detection of breast cancer and is a cause of medical exposure. Accordingly, using self-made shielding apron, the radiation dose rate before and after shielding against other areas except the inspection area was measured and the reduction rate was calculated. As a result, the reduction rate before and after shielding was found from the lowest 85.52% to the highest 92.67%. This can minimize exposure to areas unrelated to the purpose of imaging by using effective shielding apron during mammography, and also maintain the quality improvement of medical images.

Background: Heat energy produced in nuclear reactors and nuclear fuel cycle facilities interactions modifies the physical properties of the shielding materials containing water content. Therefore, in the present paper, effect of the heat on shielding effectiveness of the concretes is investigated for gamma and neutron. The mass attenuation coefficients, effective atomic numbers, fast neutron removal cross-section and exposure buildup factors. Materials and Methods: The mass attenuation coefficients, effective atomic numbers, fast neutron removal cross-section and exposure buildup factors of ordinary and heavy concretes were investigated using NIST data of XCOM program and Geometric Progression method. Results and Discussion: The improvement in shielding effectiveness for photon and reduction in fast neutron for ordinary concrete was observed. The change in the neutron shielding effectiveness was insignificant. Conclusion: The present investigation on interaction of gamma and neutron radiation would be very useful for assessment of shielding efficiency of the concrete used in high temperature applications such as reactors.

방사선 차폐 물질인 납을 대신하여 싸고 친환경적인 산화철, 알루미나로 차폐체를 제작 후 차폐체의 성능 을 평가하고자 한다. 산화철과 알루미나 각각 석고와 혼합하여 차폐체를 제작 이후 X-ray Tube를 이용하여 석고 보드, 납복과 비교하여 성능을 평가한다. 관전압에 변화를 주며 실험한 결과 알루미나의 차폐 성능은 석고보드보 다 차폐 성능이 떨어졌으며 50%의 알루미나가 함유되었을 때 석고보드와 차폐 성능이 비슷하였다. 산화철은 약 75%가 함유되었을 때 납복의 차폐 성능과 비슷해졌다. 알루미나를 이용한 차폐체는 석고와 비슷한 차폐 성능을 보여줘 납 대체물질로 적합하지는 않다. 하지만 산화철은 납복과 비슷한 차폐 성능을 나타내어 향후 납을 대체하 는 X선 차폐 물질로 활용이 가능하기에 추가 연구가 필요하다.

It is intended to evaluate the performance of the shield after manufacturing a shield with cheap and eco-friendly iron oxide and alumina instead of lead, which is a radiation shielding material. After manufacturing the shield by mixing iron oxide and alumina with gypsum, the performance is evaluated by comparing it with gypsum board and lead apron using an X-ray tube. As a result of the experiment, the shielding performance of alumina was lower than that of the gypsum board, and when 50% of alumina was contained, the shielding performance was similar to that of the gypsum board. Iron oxide became similar to the shielding performance of lead apron when it contained about 75%. A shielding material using alumina shows shielding performance similar to that of gypsum, so it is not suitable as a substitute for lead. However, since iron oxide exhibits similar shielding performance to lead, it can be used as an X-ray shielding material to replace lead in the future, so further research is needed.

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.

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 μSv/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.

본 의료기관에서 사용되는 방사선 차폐체는 주로 납을 활용하여 제품과 부속품을 제작한다. 납은 가공성과 경제성이 우수하지만 폐기 시 환경 문제로 인해 사용량을 줄이고 있으며, 오랫동안 사용했을 시 크랙 현상과 중력 에 의한 처짐 현상으로 인해 차폐막, 차폐벽, 의료기기 부픔 등으로 장기간 사용하기에는 한계가 있다. 이러한 문 제점을 해결하기 위해 구리, 주석 등을 사용하지만, 아직 차폐성능을 제어하기에는 제작 공정에 어려움이 있어 대부분 텅스텐을 많이 사용하고 있다. 그러나 아직 텅스텐의 종류에 따른 특성이 잘 제시되지 못해 다른 차폐재와 의 비교가 어렵다. 따라서 본 연구에서는 순수 텅스텐, 탄화텅스텐, 산화 텅스텐을 이용하여 동일한 공정으로 의료 방사선 차폐시트를 제작하여 시트 단면의 입자 구성과 차폐성능을 비교하였다. 비교 결과 순수 텅스텐, 탄화텅스 텐, 산화 텅스텐 순으로 차폐성능이 우수한 것으로 나타났다.

Radiation shields used in medical institutions mainly use lead to manufacture products and fitments. Although lead has excellent processability and economic efficiency, its use is being reduced due to environmental issues when it is disposed of. In addition, when used for a long time, there is a limit to using it as a shielding film, shielding wall, medical device parts, etc. due to cracking and sagging due to gravity. To solve this problem, copper, tin, etc. are used, but tungsten is mostly used because there is a difficulty in the manufacturing process to control the shielding performance. However, it is difficult to compare with other shielding materials because the characteristics according to the type of tungsten are not well presented. Therefore, in this study, a medical radiation shielding sheet was manufactured in the same process using pure tungsten, tungsten carbide, and tungsten oxide, and the particle composition and shielding performance of the sheet cross-section were compared.As a result of comparison, it was found that the shielding performance was excellent in the order of pure tungsten, tungsten carbide, and tungsten oxide.