• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.2
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• pp.91-100
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• 2016

The neutron activation inventories in reactor vessel and its internals, and bio-shield of a PWR nuclear power plant were calculated to evaluate the effect of impurity elements contained in the structural materials on the activation inventory. Carbon steel is, in this work, used as the reactor vessel material, stainless steel as the reactor vessel internals, and ordinary concrete as the bio-shield. For stainless steel and carbon steel, one kind of impurity concentration was employed, and for ordinary concrete five kinds were employed in this study using MCNP5 and FISPACT for the calculation of neutron flux and activation inventory, respectively. As the results, specific activities for the cases with impurity elements were calculated to be more than twice than those for the cases without impurity elements in stainless and carbon steel. Especially, the specific activity for the concrete material with impurity elements was calculated to be 30 times higher than that without impurity. Neutron induced reactions and activation inventories in each material were also investigated, and it is noted that major radioactive nuclide in steel material is Co-60 from cobalt impurity element, and, in concrete material, Co-60 and Eu-152 from cobalt and europium impurity elements, respectively. The results of this study can be used for nuclear decommissioning plan during activation inventory assessment and regulation, and it is expected to be used as a reference in the design phase of nuclear power plant, considering the decommissioning of nuclear power plants or nuclear facilities.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05b
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• pp.630-635
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• 1997

본 연구에서는 원자력 선진국들의 원자력 미래기술의 예측 사례를 조사ㆍ분석하고, 이를 우리나라의 사례와 비교하였다. 조사된 미래 원자력 기술예측 및 수준 평가는 국외의 경우 일본, 독일 그리고 프랑스의 사례가 조사되었고 국내에서는 과학기술정책관리연구소와 한국원자력연구소의 사례를 원용하였다. 이들 사례에서 공통적으로 평가하고 있는 기술의 중요성, 실현시기 그리고 제약요인을 비교하였다. 기술평가 결과에 나타난 일반적인 특징은 우선 개별기술의 중요성 평가에서는 공통적으로 방사성패기물처리, 원전내진설계 그리고 원전해체기술 등과 같이 이미 활용중인 기술로서 기존 시설의 안전성을 향상시킬 수 있는 기술의 중요성을 높이 평가하고 있다. 다음으로 실현시기 평가에서는 레이저빔 이용기술과 같이 인접과학 분야의 발전에 따른 시너지 효과가 기대되는 분야의 기술이 2010년 이전에 실현될 것으로 평가하고 있다. 마지막으로 기술개발의 저해요인의 평가는 조사사례 별로 정도의 차이는 있으나 기술적 제약요인이 가장 높고, 다음으로 경제적 제약 그리고 사회적 제약의 순으로 평가하고 있다.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.1
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• pp.91-102
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• 2020

According to the Korea Radioactive Waste Agency's (KORAD's) medium and low level radioactive waste management implementation plan, the Domestic 3rd Step Landfill Disposal Facility has planned to accept a total of 104,000 drums (2 trenches) of very low level radioactive waste (VLLW), from the decommissioning site from April 2019 - February 2026 (total budget: 224.6 billion Won). Subsequently, 260,000 drums (5 trenches) will be disposed in a 34,076 ㎡. Accordingly, KORAD is preparing a waste acceptance criteria (WAC) for this facility. Every disposal facility for VLLW in other countries such as France and Spain, operate their WAC for each VLLW facility with a reasonable application approach, This, paper focuses on analyzing the WAC conditions in VLLW sites in the USA and discusses whether these can be met in domestic VLLW WAC. It also helps in the preparation of WAC for the 3rd Step Landfill Disposal Site in Gyeongju, since the USA has prior experience on decommissioning nuclear waste.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.2_spc
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• pp.317-325
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• 2020

The Korea Radioactive Waste Agency plans to expand the storage capacity of radioactive waste by constructing a radioactive waste inspecting building to solve the problem of the lack of inspection space and drum-handling space in the radioactive waste receipt and storage building for the first-stage disposal facility. In this study, the exposure doses of radiation workers that handle new disposal containers for decommissioning waste in the storage areas of the radioactive waste inspecting building were calculated using the Monte Carlo N-particle transport code. The annual collective dose was calculated as a total of 84.8 man-mSv for 304 new disposal containers and an estimated annual 306 working hours for the radiation work. When the 304 new disposal containers (small/medium type) were stored in the storage areas, it was found that 25 radiation workers should be involved in acceptance/disposal inspection, and the estimated exposure dose per worker was calculated as an average annual value of 3.39 mSv. When the radiation workers handle the small containers in high-radiation dose areas, the small containers should be shielded further by increasing the concrete liner thickness to improve the work efficiency and radiation safety of the radiation workers. The results of this study will be useful in establishing the optimal radiation working conditions for radiation workers using the source term and characteristics of decommissioning waste based on actual measurements.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.4
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• pp.473-478
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• 2018

Radioactive aerosol generated in cutting and melting work during the NPP decommissioning process can cause internal exposure to body through workers' breath. Thus, it is necessary to assess worker internal exposure due to the radioactive aerosol during decommissioning. The actually measured value of the working environment is needed for accurate assessment of internal exposure, but if it is difficult to actually measure that value, the internal exposure dose can be estimated through recommended values such as the fraction of amount of intake and the size of particles suggested by the International Committee on Radiological Protection (ICRP). As for the selection of particle size, this study applied a value of $5{\mu}m$, which is the size of particles considering the worker recommended by the ICRP. As for the amount of generation, the amount of intake was estimated using data on the mass of aerosol generated in a melting facility at a site in Kozloduy, Bulgaria. In addition, using these data, this study calculated the level of radioactivity in the worker's body and stool and conducted an assessment of internal exposure using the BiDAS computer code. The internal exposure dose of Type M was 0.0341 mSv, that of Type S was 0.0909 mSv. The two types of absorption showed levels that were 0.17% and 0.45% of the domestic annual dose limit, respectively.

• Nuclear industry
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• v.38 no.3
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• pp.11-34
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• 2018

국민의 생명 안전 중심의 미래지향적 원자력 방사선 기술 개발을 목적으로 하는 2018년도 원자력연구개발사업의 세부사업별 시행계획이 확정되었다. 과학기술정보통신부는 이번 시행계획에 들어가 있는11개 단위 사업 추진을 위해 원자력연구개발기금 1,278억원, 일반회계 및 지역발전특별회계 813억원 등 총2,091억원을 투입할 예정이다(년도 2,335억원, 10.5% 감소) 지원 분야는 크게 원자력 기술, 방사선 기술, 기초 기반 구축으로 나누어진다. 원자력 기술 분야는 원자력 안전 연구, 원전 해체 기술, 사용후핵연료 관리 기술 및 중소형 원자로 개발 등이며, 방사선 기술 분야는 방사선 기술 고도화를 통한 핵심 기술 개발, 대형 의료 산업용 방사선 시설 장비 구축 등에 방점을 찍고 있다. 그리고 기초 기반 구축 분야는 선진기술연구센터, 국제 협력, 인력 양성 등 연구 기반 확충에 역점을 두고 있다. 2018년도 중점 추진 방향과 사업별 세부 추진 계획은 다음과 같다.

• Proceedings of the Korean Institute of Information and Commucation Sciences Conference
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• 2015.10a
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• pp.1103-1105
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• 2015

Leaked Radioactive source in nuclear power station, radiation related facilities and the aging nuclear power plant for the dismantling must need to detect and remove early to prevent major accidents. In this paper, we implemented a single sensor-based gamma-ray detectors stereo which can provide the distance to the radiation source, a direction and doserate information for fast and efficient decontamination work the radiation source. And we have carried out an algorithm development for high-speed detection of the detection equipment. Two detectors are required for stereo structure for obtaining the distance information of the radioactive source, but we designed the only sensor-based detection device for the weight reduction. We have extracted the region of interest and obtained the distance calculation result and distribution of radiation source in order to minimize a stereo image acquisition time. Detection time of the algorithm showed a shorter time of about 41%.

• Korean Journal of Mineralogy and Petrology
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• v.35 no.1
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• pp.25-39
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• 2022

This study was conducted to evaluate the manufacturing process of non-sintered cement for the safe containment of radioactive waste using low level or ultra-low level radioactive waste soil generated from nuclear-decommissioning facilities, clay minerals, and blast furnace slag (BFS) as an industrial by-product recycling and to characterize the products using mineralogical and morphological analyses. A stepwise approach was used: (1) measuring properties of source materials (reactants), such as waste soil, clay minerals, and BFS, (2) manufacturing the non-sintered cement for the containment of radioactive waste using source materials and deducing the optimal mixing ratio of solidifying and adjusting agents, and (3) conducting mineralogical and morphological analyses of products from the hydration reactions of manufactured non-sintered cement solidifier (NSCS) containing waste concrete generated from nuclear-decommissioning facilities. The analytical results of NSCS using waste soil and clay minerals confirmed none of the hydration products, but calcium silicate (CSH) and ettringite were examined as hydration products in the case of using BFS. The compressive strength of NSCS manufactured with the optimum mixing ratio and using waste soil and clay minerals was 3 MPa after the 28-day curing period, and it was not satisfied with the acceptance criteria (3.44 MPa) for being brought in disposal sites. However, the compressive strength of NSCS using BFS was estimated to be satisfied with the acceptance criteria, despite manufacturing conditions, and it was maximized to 27 MPa at the optimal mixing ratio. The results indicate that the most relevant NSCS for the safe containment of radioactive waste can be manufactured using BFS as solidifying agent and using waste soil and clay minerals as adsorbents for radioactive nuclides.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.4
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• pp.381-390
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• 2017

In Korea, commercial nuclear power plants and research reactors have on-site storage systems for the spent nuclear fuel, but it is difficult to expand the facilities used for the storage systems. If decommissioning of nuclear power plants starts, an amount of high level radioactive waste will be generated. In this study, a radiological impact assessment of the railroad transport of high level radioactive waste was carried out considering radiation workers and the public, using the developed transport container as the transport package. The dose rates for workers and the public during the transport period were estimated, considering anticipated transport scenarios, and the results compared with the regulatory limit. A sensitivity analysis was also carried out by considering the different release ratios of the radioactive materials in the high level radioactive waste, and different distances between the transport container and workers during loading and unloading phases and while attaching another freight car. For all the anticipated transport scenarios, the radiological impacts for workers and the public met the regulatory limits.

• Journal of Radiation Protection and Research
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• v.38 no.4
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• pp.228-233
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• 2013

It has been continuously generated the requirement for the replacement of the main components such as a steam generator due to the deterioration of the nuclear power plant all around the world. Also, a large amount of radioactive metal was generated during the decommissioning in a short period. It is required to make an accurate measurement of the residual radioactivity for recycling the metal waste for releasing from regulatory control. In planning the measurement procedures, the influence of geometry, self-absorption, density and other relevant factors on the representativeness of the measurements should be considered for the decommissioning metal waste. In this study, the method for measurement procedures, the source term evaluation, the ways to secure representative samples, the measurement device for wide area and the self-absorption correction factors for different density were evaluated. The metal samples for measurement were prepared for securing the simple geometry and representative by melting process. The developed correction method for measuring the radioactivity a variety density of metal waste could improve the reliability of the evaluation results for clearance.