• Journal of Radiation Protection and Research
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• 제4권1호
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• pp.5-13
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• 1979

공기시료채칩 종료후 공기여과지에 채집된 시료중 방사능을 일정한 시간구간을 두어 계측함으로써 얻은 붕괴곡선을 이론적 방법에 의하여 분석할 수 있는 간단한 방법을 개발하였으며, 이 방법을 이용하여 라돈 붕괴생성물 각각의 공기중 방사능 농도를 결정하였다. 라돈 붕괴생성물 각 핵종의 방사능 농도는 알파붕괴, 시료채집시간, 그리고 수치계수의 함수로 표시된 방정식으로 부터 얻었다. 그리고 대기중 라돈 붕괴생성물 개개의 방사평형상태도 또한 조사하였다. TRIGA Mark-III 원자로실내에서 채집한 공기시료는 상당히 비평형상태에 있었다. 라돈 붕괴생성물들 간의 방사성 불평형의 정도는 공기와류조건과 관련된 공기시료 채집시간에 따라 상당히 달라지는 것같았다. 본 연구 결과에서 얻은 자료는 인체 내부방사선 피폭선량평가와 기체 방사성 물질 방출감시기 교정에 유용한 기초자료가될 것이 확실하다.

• 방사선산업학회지
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• 제9권3호
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• pp.111-117
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• 2015

In this paper, production of $^{166}Ho$ by double neutron capture from HANARO research reactor was evaluated. This production approach provides $^{166}Ho$ with high specific activity. $^{164}Dy$ is transmuted into $^{165g+m}Dy$ by (n,${\gamma}$) reaction, then $^{165g+m}Dy$ is transmuted into $^{166}Dy$ by (n,${\gamma}$) reaction. At the end of neutron irradiation, population of $^{166}Dy$ atoms reaches highest point. And $^{164}Dy$ exists as a mixture with $^{165m}Dy$, $^{165}Dy$, $^{166}Ho$ and $^{165}Ho$ at this point. To obtain $^{166}Ho$ with high specific activity, Ho isotopes from irradiated target is separated out. Then $^{166}Ho$ decayed from $^{166}Dy$ is eluted at radioactive equilibrium state. At each step, the number of relevant nuclide is calculated by the state equation. The neutron irradiation time for maximum $^{166}Dy$ is calculated for 283 hour. When 100 mg target of $Dy_2O_3$ (96.8% enriched $^{164}Dy$) is used, possible activity of $^{166}Ho$ is 3.54 Ci($1.31{\times}10^{11}Bq$). For separation efficiency of Dy/Ho is 99.99%, $^{166}Ho/Ho$ is 0.62.

• Nuclear Engineering and Technology
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• 제26권4호
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• pp.453-460
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• 1994

암반(Porous Rock Matrix)과 균열(fracture)에서 일차원의 이동 방정식(Migration Equation)을 3-Member Decay Chain까지 화장하고, Laplace Transform을 이용하여 초기조건이 Delta Function과 Bateman Equation인 각각에 대해 해석해를 구한다. 그 해를 이용하여 Actinide Chain 중 4n＋1과 4n＋2 Chain에서 선택된 Np$^{241}$-U$^{233}$ -Th$^{229}$ 와 U$^{234}$ -Th$^{230}$ -Ra$^{226}$ Chain의 각 핵종들의 균열에서의 농도를 상대농도로 나타낸다. 이핵종들의 지연계수(Retardation Coefficient)는 화강암에 대한 것을 사용하여 균열에서의 농도 변화를 볼 수 있다. 본 연구에 의한 결과로는 U$^{233}$ , Th$^{229}$ , Th$^{230}$ Ra$^{226}$ 같은 핵종들은 비록 초기 inventory에는 작은 양일지라도 균열과 암반에서 모핵종의 붕괴(decay)에 의해 생기므로써 처분장으로부터 먼 거리에서는 중요한 핵종이 된다는 것을 알 수 있다.

• 대한방사선기술학회지:방사선기술과학
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• 제44권5호
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• pp.489-494
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• 2021

In this study, radioactivity quantitative analysis was performed on radon contained in cigarette, and the effective dose was calculated using the result value to determine the amount of exposure caused by smoking. A total of 5 types of cigarettes were sampled. Cigarette smoke was collected by using activated carbon, and tobacco were measured by homogenizing for quantitative analysis. For each sample, Bi-214 and Pb-214 were subjected to gamma nuclide analysis to observe the uranium-based radioactive material contained in cigarette, and a measurement time of 30,000 seconds was set for the sample based on the results of previous studies. As a result of measuring the radioactivity of tobacco, a maximum of 0.715 Bq/kg was derived, and in the case of cigarette smoke measured using activated carbon, a maximum of 3.652 Bq/kg was derived. Using this measurement, the average effective dose to the lungs is 0.938 mSv/y, and it was found that there is a possibility of receiving exposure up to 1.099 mSv/y depending on the type of tobacco. It was found that the exposure dose due to cigarette occupies a large proportion of the annual effective dose limit for the general public. Therefore, more diverse studies on radioactive substances in cigarette are needed, and measures to monitor and reduce the incidental exposure to radon should be established.

• Nuclear Engineering and Technology
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• 제53권2호
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• pp.423-429
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• 2021

The investigation of the utilization of enriched ²⁰⁸Pb as a coolant to enhance the performance of a long-life fast reactor with a Modified CANDLE (Constant Axial shape of Neutron flux, nuclide densities, and power shape During Life of Energy production) burnup scheme has performed. The analyzes were performed on a reactor with thermal power of 800 MegaWatt Thermal (MWTh) with a refueling process every 15 years. Uranium Nitride (enriched ¹⁵N), ²⁰⁸Pb, and High-Cr martensitic steel HT-9 were employed as fuel, coolant, and cladding materials, respectively. One of the Pb-nat isotopes, ²⁰⁸Pb, has the smallest neutron capture cross-section (0.23 mb) among other liquid metal coolants. Furthermore, the neutron-producing cross-section (n, 2n) of ²⁰⁸Pb is larger than sodium (Na). On the other hand, the inelastic scattering energy threshold of ²⁰⁸Pb is the highest among Na, ^natPb, and Bi. The small inelastic scattering cross-section of ²⁰⁸Pb can harden the neutron energy spectrum. Therefore, ²⁰⁸Pb is a better neutron multiplier than any other liquid metal coolant. The excess neutrons cause more production than consumption of ²³⁹Pu. Hence, it can reduce the initial fuel loading of the reactor. The selective photoreaction process was developing to obtain enriched ²⁰⁸Pb. The neutronic was calculated using SRAC and JENDL 4.0 as a nuclear data library. We obtained that the modified CANDLE reactor with enriched ²⁰⁸Pb as coolant and reflector has the highest k-eff among all reactors. Meanwhile, the ^natPb cooled reactor has the lowest k-eff. Thus, the utilization of the enriched ²⁰⁸Pb as the coolant can reduce reactor initial fuel loading. Moreover, the enriched ²⁰⁸Pb-cooled reactor has the smallest power peaking factor among all reactors. Therefore, the enriched ²⁰⁸Pb can enhance the performance of a long-life Modified CANDLE fast reactor.

• Korean Chemical Engineering Research
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• 제59권3호
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• pp.450-454
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• 2021

현재, 감마 핵종 분석은 주로 무기섬광체 또는 반도체 검출기를 활용하여 여러 분야에 사용되고 있다. 이러한 검출기는 분해능이 좋지만 크기가 제한적이며, 가공성이 낮고 경제성이 플라스틱 섬광체보다 낮다. 따라서, 나노물질인 양자점과 플라스틱섬광체의 장점을 이용하여 양자점 나노물질 기반 플라스틱 섬광체를 개발하였다. 가장 많이 활용되고 있는 Cd계열 물질인 CdS/ZnS 양자점을 플라스틱 매트릭스에 교반하여 제작하였으며, 이를 ⁶⁰Co핵종 대상 계측 실험을 하여 상용플라스틱 섬광체의 성능과 비교 분석하였다. 상용플라스틱 섬광체 대비 CdS/ZnS 양자점 기반 플라스틱 섬광체가 20~30% 높은 효율을 보였다. 이는 의료분야뿐만 아니라 원자력 해체분야에서도 방사능 분석기로 활용 가능할 것으로 판단된다.

• 시스템엔지니어링학술지
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• 제17권1호
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• pp.76-84
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• 2021

Radioactive waste generated from nuclear power plant decommissioning shall be strictly managed so that radioactive materials above the allowable limit are not leaked into the environment. Radioactive wastes shall be classified and treated for management based on characteristics such as the type of waste, physicochemical properties, nuclide concentration and radioactivity. Waste characterization report shall be prepared and submitted to the disposal facility operator to ensure that the treated waste is suitable for disposal. The disposal facility operator shall review the waste Characterization report and visit the nuclear power plant decommissioning site to ensure that the wastes are processed step by step according to the plan. The waste Characterization report may be used as input data to evaluate disposal facility safety. Domestic and foreign data are collected and reviewed to confirm the entire processes from waste generation to delivery. This paper proposes the method to prepare the waste Characterization report which contains data and information on waste characteristics, treatment facilities & method and packaging method & container.

• Nuclear Engineering and Technology
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• 제54권2호
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• pp.507-513
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• 2022

When X-ray energy above 8 MV is used, photoneutrons are generated by the photonuclear reaction, which activates the components of linear accelerator (linac). Safely managing the radioactive material, when disposing linac or replacing components, is difficult, as the standards for the radioactive material management are not clear in Korea. We surveyed the management status of radioactive components occurred from medical linacs in Korea. And we also measured the activation of each part of the discarded Elekta linac using a survey meter and portable High Purity Germanium (HPGe) detector. We found that most medical institutions did not perform radiation measurements when disposing of radioactive components. The radioactive material was either stored within the institution or collected by the manufacturer. The surface dose rate measurements showed that the parts with high surface dose rates were target, primary collimator, and multileaf collimator (MLC). ⁶⁰Co nuclide was detected in most parts, whereas for the target, ⁶⁰Co and ¹⁸⁴Re nuclides were detected. Results suggest that most institutions in Korea did not have the regulations for disposing radioactive waste from linac or the management procedures and standards were unclear. Further studies are underway to evaluate short-lived radionuclides and to lay the foundation for radioactive waste management from medical linacs.

• Journal of Radiation Protection and Research
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• 제46권2호
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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.

• Nuclear Engineering and Technology
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• 제54권5호
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• pp.1754-1759
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• 2022

The activity of gamma-ray emitting nuclides is calculated assuming that each gamma-ray is detected individually; thus, the magnitude of the coincidence summing signal must be considered during activity calculations. Here, the correction factor for the coincidence summing effect was calculated, and the detection efficiencies of two HPGe detectors were compared. The CANBERRA Inc. GC4018 high-purity Ge detector provided an estimate for the peak-to-total ratio using a point source to determine the coincidence summing correction factor. The ORTEC Inc. GEM60 high-purity Ge detector uses EFFTRAN in LVis to obtain the parameters of the detector and source model and the gamma-gamma and gamma-X match estimates, in order to determine the coincidence summing correction factor. Nuclide analyses, radioactivity comparisons, and analyses of reference material samples were performed utilizing certified reference materials to accurately determine the detection efficiencies. For both Co-60 and Y-88, the detection efficiency for a point source increased by an average of at least 12-13%, whereas the detection efficiency determined using LVis increased by an average of at least 13-15%. The calculated radioactivity values of the certified reference material and reference material samples were accurate to within 3% and 6% of the measured values, respectively.