• Nuclear Engineering and Technology
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• 제53권4호
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• pp.1224-1235
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• 2021

An integrated waste management approach for irradiated graphite was developed during the European Commission project 'Treatment and Disposal of Irradiated Graphite and other Carbonaceous Waste'. This included the identification of potential options for the management of irradiated graphite, taking account of storage, retrieval, treatment and disposal methods. This paper describes how these options can be assessed using multi-criteria decision analysis (MCDA) for a case study relating to a generic power reactor. Criteria have been defined to account for safety, environmental, economic and socio-political factors, including radiological impact, resource usage, economic costs and risks. The impact of each option against each criterion has been assessed using data from the project and the wider literature. A linear additive approach has been used to convert the calculated impacts to scores. To account for the relative importance of the criteria, example weightings were allocated. This application has shown that MCDA approaches can be used to support complex decisions regarding irradiated graphite management, accounting for a wide range of criteria. Use of this approach by individual countries or organisations will need to account for the specific options, scores, weightings and constraints that apply, based on their national strategies, regulatory requirements and public acceptability.

• 한국압력기기공학회 논문집
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• 제16권2호
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• pp.18-24
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• 2020

When a damage occurs in the nuclear fuel burning in the reactor, fission products that should be in the nuclear fuel rod are released into the reactor coolant. In this case, sipping test, a series of non-destructive inspection methods, are used to find leakage in nuclear fuel assemblies during the power plant overhaul period. In addition, the sipping test is also used to check the integrity of the spent fuel for moving to an intermediate dry storage, which is carried out as the first step of nuclear decommissioning, . In this paper, the principle and characteristics of the sipping test are described. The structure of the sipping inspection equipment is largely divided into a suction device that collects fissile material emitted from a damaged assembly and an analysis device that analyzes their nuclides. In order to make good use of the sipping technology, the radioactive level behavior of the primary system coolant and major damage mechanisms in the event of nuclear fuel damage are also introduced. This will be a reference for selecting an appropriate sipping method when dismantling a nuclear power plant in the future.

• 방사성폐기물학회지
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• 제15권1호
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• pp.45-63
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• 2017

한국의 가장 오래된 상업 원전인 고리 1호기가 2017년에 해체가 이루어질 예정이다. 원전 해체 폐기물의 적절한 처리는 효율적인 원전해체에 있어 중요한 역할을 할 것이다. 특히, 저준위 또는 오염되지 않은 금속폐기물의 재활용은 폐기물 발생 저감은 물론 처분장의 공간을 절약하는데 기여할 것이다. 본 논문은 재활용 시스템의 개념설계와 정의된 업무 흐름에서 발생하는 피폭 선량을 평가하는데 그 목적이 있다. 작업의 흐름과 운전 개념을 정립하기 위해 다양한 형태의 다이어그램을 설계하였다. 선량평가에 필요한 시나리오는 개념설계를 기반으로 선정되었으며, RESRAD-RECYCLE을 이용하여 선량을 평가하였다. 이를 통하여, 결정적 시나리오 선별, 핵종 특성 및 핵종 분배가 선량에 미치는 영향을 분석하였다. 더 나아가, 선량분석은 피폭 시나리오에 대한 대체 방안 수립, 필요한 제염 및 방사선방어 프로세스 그리고 허용 방사능 검토의 정보를 제공하는데 사용 될 수 있을 것이다.

• 전기학회논문지
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• 제67권2호
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• pp.337-342
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• 2018

Spent fuel is moved from the reactor into the spent fuel pool when nuclear power plant permanently shutdown. The sole function of a permanently defueled facility is to store spent fuel in a quiescent state. The function of electric system and loads are reduced. It is necessary to establish an operating strategy of electric system in the permanent shutdown nuclear plant. This paper reviews required loads and design criteria considering transition to permanent shutdown. An operating strategy of onsite electric system is proposed considering decommissioning strategy and stage of defueled condition.

• Nuclear Engineering and Technology
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• 제55권8호
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• pp.3010-3016
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• 2023

In recent years, the development of radioactive waste containers for nuclear facility decommissioning and dismantling is a critical issue because the Taiwan domestic boiling water reactor nuclear power plant is going to be decommissioned. The main purpose of this research is to design a metal container that meets the structural requirements of related regulations. At first, the shielding analysis was performed by varying dimensions of radioactive waste to determine the storage efficiency of the container. Then, a series of structural analyses for operational and accidental conditions of the container with full load were conducted, such as lifting, stacking, and drop impact conditions. On the other hand, the field drop impact tests were carried out to ensure structural integrity. The present research demonstrates the structural safety of the developed container for decommissioned nuclear facilities in Taiwan.

• 방사성폐기물학회지
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• 제12권2호
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• pp.153-164
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• 2014

현재 전 세계적으로 설계단계에서 부식 생성물과 방사성 핵종의 양을 예측하는 프로그램에 대해서는 개발되거나 개발중인 프로그램이 다양하다. 그러나 원자력 발전소 해체 시 발생하는 방사화 부식생성물의 양을 평가하는 코드에 대한 개발은 이루어지지 않고 있어 정확한 산정에 어려움이 있다. 원자로 용기, 원자로 구성품 및 인접 구조물에서의 특성 원소의 중성자 조사로 인한 방사화재고량을 평가하기 위해서는 원자로의 고정된 구조물을 대표하는 모든 영역에서의 평균 중성자속과 구조물의 물질조성 및 원자로 운전이력 등을 이용하여 평가해야 한다. 본 논문에서는 설계단계에서 사용되는 1차 계통의 부식생성물과 방사성 핵종의 양을 예측하는 CORA, PACTOLE, CRUDSIM, CREAT 및 ACE 코드를 분석하였다. 향후 연구에서는 제염해체 폐기물 발생량 평가에 대한 사용가능성과 개선점을 찾아 부식생성물량 산정에 정확성을 높이고자 한다.

• 방사성폐기물학회지
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• 제16권4호
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• pp.441-454
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• 2018

2017년 고리 1 호기 영구정지를 계기로 국내 원자력발전소의 해체가 점차 가시화되고 있다. 앞으로 원전 해체가 본격적으로 추진될 경우 원전 1기 당 약 16만 t의 콘크리트 폐기물이 발생될 것으로 예측되었으며, 이들 콘크리트 폐기물은 대부분 오염준위가 매우 낮아 자체처분 대상으로 고려될 수 있다. 따라서, 국내 자체처분 폐기물(원자력안전위원회 고시 2017-65호에 따른 자체처분 허용농도 또는 자체처분 허용선량을 만족하는 폐기물)에 대한 현행 규제체계가 대량의 콘크리트 폐기물에 대한 무제한적 자체처분에 대해서도 유효성을 유지할 수 있는지를 사전에 확인할 필요가 있다. 이와 관련, 국내 자체처분 규정 개발 시 참조기준인 IAEA SRS No. 44를 심층분석하고, 국내 산업계 현황을 반영한 입력값과 계산식을 이용하여 4가지 자체처분 시나리오에 대한 예상 피폭방사선량을 평가하였다. 그 결과, 재활용 시나리오에 대한 예상선량은 대부분 정상 시나리오에 대한 자체처분 선량 기준(즉, $0.01mSv{\cdot}y^{-1}$)보다 낮은 것으로 평가되었으나, 성토 후 거주 시나리오의 경우 보수적인 가정을 적용하면 자체처분 선량 기준을 초과할 가능성도 배제할 수 없는 것으로 나타났다. 따라서, 대량의 해체 콘크리트 폐기물의 안전하고 지속가능한 자체처분을 위해서는 폐기물 처리업체 다변화, 성토 시나리오에 대한 보다 구체적인 평가, 성토를 통한 자체처분에 대한 부분적 제한조건 설정 등을 고려할 수 있다.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2007년도 학술논문요약집
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• pp.190-191
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• 2007

Cleanout of the F Reactor Fuel Storage Basin (FSB) is an element of the FSB decontamination and decommissioning (D&D) and is required to complete interim safe storage (ISS) of the F Reactor. Following reactor shutdown and in preparation for a deactivation layaway action in 1970, the water level in the FReactor FSB was reduced to approximately 0.6 m (2 ft) over t]to floor. Basin components and other miscellaneous items were left or placed in the FSB. The item placement was performed with a sense of finality, and no attempt was made to place the items in an orderly manner. The F Reactor FSB was then filled to grade level with 6(20of local surface material (essentially a fine sand). The reactor FSB backfill cleanout has the potential of having to remove spent nuclear fuel (SNF) that may have been left unintentionally. Based on previous cleanout of six water-filled FSBs with similar designs (i.e., the B, C, D, and DR FSBs in the 1980's), it was estimated that up to five SNF elements could be discovered in the F FSB (I). In reality about 17 full SNF elements were found in the excavation. This paper covers the technical and programmatic challenges of performing this decommissioning effort with some of the controls used for SNF management. The paper also will highlight how many various technologies were married into a complete package to address the issue at hand and show how no one tools could complete the job, but combined, good progress is being made.

• Nuclear Engineering and Technology
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• 제53권6호
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• pp.2011-2018
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• 2021

Nowadays, evaluation of amounts and distributions of radioactive waste is an important preparatory step in the process of nuclear reactor decommissioning. For tentative estimation of radioactive waste, a cell-based rigorous 2 step (R2S) method usually is used; however, a poor resolution caused by the averaged flux and spectrum in a cell is still a great challenge because of leading to underestimated or overestimated results. To overcome the poor resolution, several systems were introduced. Neither system, however, provides any function for evaluation of radioactive waste amount and distribution. Thus, it is additionally required to classify radioactive waste based on the results of activation calculation. In this study, the advanced R2S (AR2S) system was developed. To verify the performance of the system, its results for a verification problem were compared with those of the cell-based R2S method. The results showed good agreement, which is to say, within 2.0% relative error. Also, several characteristics of fine/coarse mesh were analyzed. To demonstrate the performance of the AR2S system, the radioactive waste from the Japan Power Demonstration Reactor (JPDR) was estimated, and the result indicated a high-resolution distribution. Therefore, it is expected that the AR2S system will prove useful for precise evaluation of radioactive waste.

• Journal of Radiation Protection and Research
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• 제46권1호
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• pp.24-31
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• 2021

Decommissioning efforts are underway at the reactor where the accident occurred, namely the damaged Tokyo Electric Power Company (TEPCO) Fukushima Daiichi Nuclear Power Plant (FDNPP). However, a large amount of groundwater flowing into the site has become contaminated with radioactive substances and is stored in tanks on site, which has hampered the decommissioning work. Although the inflow of groundwater has been greatly reduced through measures such as the construction of frost walls, approximately 170 ㎥ of water treated by the Advanced Liquid Processing System (ALPS) is being stored in tanks, each day. The tanks used to store this treated water are expected to become full by around the summer of 2022. It is not easy to get people to understand the efforts of all concerned parties, and providing clear information to these concerned parties is also a challenge. Questions have also been raised regarding whether other alternatives have been fully explored in the ALPS subcommittee. Some people have commented that the answers to the questions raised regarding the biological effects of tritium transmutation are inadequate. Some suspect that the answers are too detailed and incomprehensible, and that the respondents may be manipulating the public with some malicious intent. In any case, each possible plan presents both advantages and disadvantages, depending on the people who are involved. That makes it an ethical and vexing issue that can sway decisions, as perspectives change. While the environmental release plan is scientifically safe, it may represent a painful alternative. On the other hand, a more careful and imaginative approach to the idea of continued storage in tanks or other forms of storage may reveal some troublesome hidden disadvantages. Under these circumstances, experts must be prepared to answer people's questions in a comprehensive and robust manner.