The initial development plans for the six reactor designs, soon after the release of Generation IV International Forum (GIF) TRM in 2002, were characterized by high ambition [1]. Specifically, the sodium-cooled fast reactor (SFR) and very-high temperature reactor (VHTR) gained significant attention and were expected to reach the validation stage by the 2020s, with commercial viability projected for the 2030s. However, these projections have been unrealized because of various factors. The development of reactor designs by the GIF was supposed to be influenced by events such as the 2008 global financial crisis, 2011 Fukushima accident [2, 3], discovery of extensive shale oil reserves in the United States, and overly ambitious technological targets. Consequently, the momentum for VHTR development reduced significantly. In this context, the aims of this study were to compare and analyze the development progress of the six Gen IV reactor designs over the past 20 years, based on the GIF roadmaps published in 2002 and 2014. The primary focus was to examine the prospects for the reactor designs in relation to spent nuclear fuel burning in conjunction with small modular reactor (SMR), including molten salt reactor (MSR), which is expected to have spent nuclear fuel management potential.

To evaluate the characteristics of radioactive waste from permanently shut down nuclear power plants for decommissioning, there is a method of directly analyzing samples and, on the other hand, a computerized evaluation method based on operation history. Even if the radioactivity of the structures or radioactive wastes in the nuclear power plant is analyzed by the computerized evaluation method, a method of directly analyzing the sample must be accompanied in order to more accurately know the characteristics of the nuclear power plant’s radioactive waste material. In order to obtain such samples, we need a way to collect materials from radioactive waste. However, in the case of a permanently shut down nuclear power plant with a long operating history, human access is limited due to radiation of the material. In this study, we propose a method of remotely collecting samples that guarantees radiation protection and worker safety at the site where radioactive waste is located.

As the decommissioning of nuclear power plants progresses, interest in the inevitably generated radioactive waste is also increasing. Especially, because the containers of ILW packages are significantly more expensive than the containers of LLW packages, the special attention should be focused on minimizing the number of the containers of ILW packages. The radiation dose limit for packaging of ILW shall not exceed 2 mSv/h and 0.1 mSv/h on contact and at 2 m, respectively in South Korea. Meanwhile, The DEMplus provides various environmental geometry and all properties such as materials, absorptions, and reflections and the estimation of the radiation dose rates is based on the radiation interactions of the designed 3D geometry model. With the consideration of the radiation dose rate by using DEMplus and its strategy of packaging and cutting plan, the number of containers for ILW packages generated from decommissioning of Reactor Vessel Internal (RVI) of a nuclear power plant that has been in operation for decades was optimized in this paper. The modular shielded containers (MSC) with shielding inserted were used for radioactive wastes that require shielded packaging. In order to verify the accuracy of the estimated radiation dose rate by using DEMplus, the estimated results were compared with those obtained using MicroShield. The trends of the estimated radiation dose rates using DEMplus and the estimation of MicroShield were similar to each other. The results of this study demonstrated the feasibility of using DEMplus as a means of estimating the radiation dose limit in packaging plan of the radioactive waste.

Recently, it is being carried out the project to evaluate the properties of materials harvested from nuclear reactor after the decommissioning of Kori Unit 1. However, it is not sufficient adequate machining equipment and remote machining technique to perform the projects for evaluation of materials harvested from nuclear reactor. Thus, it is required to develop the remote machining technique in hotcell to evaluate the mechanical properties of nuclear reactor materials. The machining technique should be performed inside a hotcell to evaluate mechanical properties of materials harvested from nuclear reactor and is essential to prevent radiation exposure of workers. Also, it is essential to design the apparatus and develop the machining process so that it can be operated with a manipulator and minimize contamination in hotcell. In this research, development of remote specimen machining technique in hotcell such as machining apparatus, technique and process for compact tension specimens of material harvested from nuclear reactor are described. Remote machining technique will be useful in specimen machining to evaluate changes in mechanical properties of materials harvested in high-radioactive reactor. Also, it is expected that various types of specimens can be machining by applying the developed machining technique in the future.

Since nuclear power plant (NPP) dismantling carries the possibility of radiation exposure from a hazardous environment, it’s important to minimize that by using a remote manipulator et al. However, due to complexity of nuclear facilities, it’s necessary for operators to increase their proficiency by operating in advance in a virtual environment. In this research, we propose a virtual manipulator system using a haptic device for NPP’s reactor vessel internals (RVI) dismantling which can realistically manipulate.

강진 시 원자력발전시설의 비선형 응답이 중요하기 때문에 이 시설의 내진성능에 대한 관심이 증가하였다. 이 연구에서는 원자력 발전소 철근콘크리트 전단벽의 유한요소해석을 위한 재료모델의 적절한 변수를 제시하였다: 최대인장강도, 팽창각, 손상계수. 이를 위해 상용 유한요소 해석프로그램인 ABAQUS를 사용하여 낮은 형상비를 가진 철근콘크리트 전단벽의 비선형 거동과 전단 파괴모드 에 대한 이 주요 변수의 효과에 대한 연구를 수행하였다. 연구결과에 기반하여 비선형 시간이력해석을 통해 강진 하의 원자로건물의 비선형 응답을 평가하였다.

Less mature nuclear reactor technologies are characterized by a greater uncertainty due to insufficient detailed design information, operational data, cost information, etc., but the expected performance characteristics of less mature options are usually more attractive in comparison with more mature ones. The greater uncertainty is, the higher economic risks associated with the project realization will be. Within a comparative evaluation of less and more mature nuclear reactor technologies, it is necessary to apply economic risk measures to balance judgments regarding the economic performance of less and more mature options. Assessments of any risk metrics involve calculating different characteristics of probability distributions of associated economic performance indicators and applying the Monte-Carlo method. This paper considers the applicability of statistical risk measures for different economic performance indicators within a trial case study on a comparative evaluation of less and more mature unspecified LWRs. The presented case study demonstrates the main trends associated with the incorporation of economic risk metrics into a comparative evaluation of less and more mature nuclear reactor technologies.

국내 가동원전 중 2-루프 가압경수로인 고리1호기는 약 40년 운전한 후, 2017년 6월 18일 영구정지되었다. 영구정지된 고리 1호기는 주요 해체작업을 수행하기전에 계통내 선량률을 저감시켜 작업자피폭을 최소화하기 위한 계통제염을 수행할 예정이다. 일반적으로, 계통제염 범위는 원자로압력용기, 가압기, 증기발생기, 화학 및 체적제어계통, 잔열제거계통 및 원자로 냉각재계통 주요배관을 포함한다. 이러한 계통 및 기기 등을 효율적으로 제염하기 위해서는 제염과정에서 원자로냉각재계 통내 유동특성을 평가할 필요가 있다. 계통제염을 위해 순환유량을 제공하는 방법은 다양하나, 본 논문에서는 잔열제거펌프 운전에 따른 고리1호기 원자로냉각재계통내 유동특성을 평가하였다. 잔열제거펌프를 이용한 계통제염은 원자로냉각재 내 유량의 불균형을 초래하여 계통내 기기 및 배관 등에 불순물을 침적시켜 제염이 효율적이지 않다는 것으로 평가되었다.

An air cleaning unit(ACU) in a nuclear power plant plays a role on cooling the safety-related components whose function is involved in the reactor shutdown, and maintains the suitable temperature and humidity for work in the rooms where an operator is working on. To guarantee the performance of the unit the design of ACU should be subject to credible codes and standards, such as ASME, ANSI and ASHRAE, etc. On top of them, the desorption of the carbon adsorber is addressed in ASME N509 which causes the adsorber no longer to capture the iodine isotopes produced by the severe accident. In this study, the equations of the heat source were derived from the radioactive decay heat of each iodine isotopes. From the resultant equation, the maximum temperature below 300oF for the desorption was calculated under a proposed condition and analysed with the reference results.

가압경수로(PWR)에서 배출되는 고준위폐기물을 지하 500m의 화강암 암반의 처분장에 장기간(약 10,000년 동안) 처분하기 위하여 여러 구조적 안전성 평가 수행을 통하여 처분용기모델이 개발되었다. 기존에 설계된 가압경수로용 처분용기 모델은 구조적 안전성은 문제가 없으나 너무 무거운 단점이 지적되었다. 따라서 구조적 안전성을 유지하면서 좀 더 경량화 된 처분용기모델을 개발하는 것이 요구된다. 기존의 처분용기모델이 무거워진 한가지 이유는 처분용기 개발 시 적용된 외력조건 및 안전계수 등에 대한 조건들을 너무 엄격하게 적용했기 때문이라고 사료되기 때문에 이런 조건들을 완화하여 처분용기의 재원들을 조정하여 구조해석을 다시 수행하는 것이 요구된다. 따라서 본 논문에서는 설계 완성된 기존의 처분용기에 대하여 외력 조건 및 용기의 재원(두께 등) 들을 변화시키면서 구조해석을 재 수행하여 구조적 안전성 평가를 보완하였다. 이를 바탕으로 외력 조건에 따른 처분용기의 재원 등을 재 산출한다. 보완 해석 결과 기존의 122cm의 처분용기의 직경을 102cm까지 줄여 경량화 시킬 수 있음이 확인되었다.

연구용원자로 해체비용은 해체대상물에 대한 특성 및 제원에 맞게 해체작업을 분류하고 구성요소를 설정하여 단위비용인자를 바탕으로 한 공학적 비용 산정 방법으로 해체비용을 산정한다. 연구용원자로에 대한 해체비용은 크게 인건비, 장비 및 재료비로 구성이 되는데 해체작업에 소요되는 인건비는 해체대상물에 소요되는 작업시간을 바탕으로 계산을 한다. 본 논문에서는 연구용원자로 해체비용 산정 시 인건비 계산에 필요한 단위비용인자 및 작업 난이도 인자를 산출하였다.

본 논문에서는 깊은 지하 500m에 처분된 가압경수로(PWR) 고준위폐기물 처분용기에 지하수압과 벤토나이트 팽윤압이 가해지는 동안 처분용기에 발생하는 크립변형을 예측하기 위하여 처분용기에 대한 구조해석을 수행하였다. 보통 이러한 크립변형은 처분용기에 추가적인 외력이 작용하지 않더라도 처분용기에 작용하는 압력과 내부의 높은 열에 의하여 발생될 수 있다. 처분용 기내부의 열분포의 복잡성 덴 시간의존성으로 인하여 일단 외부 지하수압 및 팽윤압만 고려하여 크립해석을 수행하였다. 이를 위하여 적당한 크립함수를 사용하였으며, 해석은 1억초 동안 수행하였다. 해석결과 1억초 동안 발생하는 크립 변형률은 매우 작으며 주희 처분용기의 위아래 덮개에 발생함을 알 수 있었다. 그러나 처분용기의 구조강도에 중요한 내부 주철삽입물에는 훨씬 더 작은 미소한 변형률만 발생하여 처분용기에 발생하는 크립변형은 처분용기의 구조적인 안전성에는 큰 영향을 미치지 않음을 알 수 있었다. 해석 초기에 처분용기 내에 급격히 응력이 증가하여 최고치에 도달한 후 잠깐동안 이 응력 값을 유지하다가 그 이 후에는 급격히 응력 값이 감소하는 응력이완현상을 보이고 있기 때문에 발생 응력 측면에서도 전혀 처분용기의 구조적인 안전성에 문제가 없음이 확인되었다.