The dismantling of the reactor pressure vessel has been carried out at a number of commercial nuclear power plants, including the Zion nuclear power plant in the United States and the Stade nuclear power plant in Germany. The dismantling method for the reactor pressure vessel is either in the air or in the water, depending on the utility. In general, a mechanical cutting method is used when dismantling the reactor pressure vessel in the water. And when dismantling a nuclear reactor pressure vessel in the air, the thermal cutting method is applied. However, there is no case of dismantling commercial nuclear reactor pressure vessel by applying a mechanical method in the air. In this study, when a nuclear reactor pressure vessel is dismantled by applying a mechanical method in the air, the applicability was evaluated by testing it using a demonstration mockup of Kori Unit 1. For the evaluation, the mockup was made in the actual size of Kori Unit 1. Mechanical cutting devices used the band saw and the circular saw. In the test, the cutting of the reactor pressure vessel was performed remotely by reflecting the working conditions of the decommissioning site. The band saw cutting method was applied to vertical cutting, and the circular saw cutting method was applied to horizontal cutting. In order to dismantle one cut-off piece, mockup test was performed according to a series of dismantling processes, it consists of preparatory work, vertical cutting process, horizontal cutting process, packaging process and finishing work. The cutting speed of the band saw is 3–10 mm·min−1, and the cutting speed of the circular saw is 2–4 mm·min−1. As a result of the test, when the mechanical cutting method was applied, as is known, the kerf width was smaller than when the thermal cutting method was applied. The cut surface showed a clean state without drag lines generated during thermal cutting. However, the working time was much slower than when the thermal cutting method was applied.

Cutting reactor pressure vessels (RPV) into acceptable sizes for waste disposal is a key process in dismantling nuclear power plants. In the case of Kori-1, a remote oxyfuel cutting method has been developed by Doosan Heavy Industry & Construction to dismantle RPVs. Cutting radioactive material, such as RPV, generates a large number of fine and ultrafine particles incorporating radioactive isotopes. To minimize radiological exposure of dismantling workers and workplace surface contamination, understanding the characteristics of radioactive aerosols from the cutting process is crucial. However, there is a paucity of knowledge of the by-products of the cutting process. To overcome the limitations, a mock-up RPV cutting experiment was designed and established to investigate the characteristics of fine and ultrafine particles from the remote cutting process of the RPV at the Nuclear Decommissioning Center of Doosan Heavy Industry & Construction. The aerosol measurement system was composed of a cutting system, purification system, sampling system, and measurement device. The cutting system has a shielding tent and oxyfuel cutting torch and remote cutting robot arm. It was designed to prevent fine particle leakage. The shielding tent acts as a cutting chamber and is connected to the purification system. The purification system operates a pressure difference by generating an airflow which delivers aerosols from the cutting system to the purification system. The sampling system was installed at the center of the pipe which connects the shielding tent and purification system and was carefully designed to achieve isokinetic sampling for unbiased sampling. Sampled aerosols were delivered to the measurement device. A high-resolution electrical low-pressure impactor (HR-ELPI+, Dekati) is used to measure the size distribution of inhalable aerosols (Aerodynamic diameter: 6 nm to 10 μm) and to collect size classified aerosols. In this work, the mock-up reactor vessel was cut 3 times to measure the number distribution of fine and ultrafine particles and mass distribution of iron, chromium, nickel, and manganese. The number distribution of aerosols showed the bi-modal distribution; two peaks were positioned at 0.01−0.02 μm and 0.04–0.07 μm respectively. The mass distribution of metal elements showed bi-modal and trimodal distribution. Such results could be criteria for filter selection to be used in the filtration system for the cutting process and fundamental data for internal dose assessment for accidents. Future work includes the investigations relationships between the characteristics of the generated aerosols and physicochemical properties of metal elements.

상부 탑재형 노내계측기(TM-ICI) 개발은 원자로하부헤드 대신 원자로상부헤드로 계측기를 삽입함으로써 중대사고 위험을 줄이기 위해 진행 중이다. 이 개발 과제의 일환으로, NUREG/CR-6909와 Code Case N-761의 두 방법에 따라 TM-ICI 노즐에 대한 환경피로평가가 수행되었다. TM-ICI 노즐은 level A, level B 및 시험 조건에서의 과도조건에 따른 하중을 받는데 이에 대해 피로평가를 해야 한다. 원자로냉각재환경이 고려된 TM-ICI 노즐의 누적사용계수는 1이하로 평가되었고, 이는 ASME Code 허용기준을 만족한다.

본 논문에서는 원통형 쉘에 부착된 노즐의 구조 건전성평가를 수행하고 그 결과를 비교하기 위해 2차원(2D)과 3차원(3D)해석이 수행되었다. 현재 원자력 발전소에서 사용되는 3개의 노즐을 구조 건전성평가를 위해 선정하였고, 각각 노즐은 내부압력, 온도변화 및 외부하중을 받는다. 내부압력에 대한 2D 해석은 1.5이상의 계수 값을 이용하거나 응력집중 계수를 적용하여야 하고, 온도변화에 대한 2D와 3D 해석결과는 피복재의 유무와 상관없이 서로 거의 비슷하며, 외부하중에 대한 WRC Bulletin 297에 의한 해석결과는 3D 해석결과보다 더 보수적임을 확인할 수 있었다.

본 논문에서는 노심용융사고 시 관통노즐이 제거된 원자로용기 하부헤드의 구조 건전성 평가를 수행하였다. 열응력, 노심용융물의 질량 그리고 내압조건의 해석결과를 고려할 때, 하부헤드의 열응력에 의한 영향이 가장 크게 나타났다. 손상 가능성은 파손기준에 따라 평가하였으며, 등가소성변형률이 임계변형률 파손기준보다 낮은 수준으로 평가되었다. 열-구조물 연성해석 결과 하부헤드의 두께 중간층에서 항복강도보다 낮은 응력이 발생한 탄성영역 구간을 확인하였다. 내압이 커지면서 탄성영역 범위가 점차 좁아지면서 탄성영역이 내벽으로 이동하는 결과를 확인하였고, 노심용융사고 시 구조적 건전성을 만족하는 것으로 평가되었다.

원자로 용기의 온도-압력 한계곡선을 위하여 국내공동비교연구를 수행하였다. 국내 원전의 데이터를 이용하여 국내 각 기관에서 온도-압력 한계곡선 작성에 사용하고 있는 방법 및 기법을 비교하기 위하여 round robin 해석을 제안하였고 주어진 문제에 대하여 각 기관이 문제를 해석한 후 결과를 제출하여 이들을 분석함으로써 온도-압력 한계곡선 작성에 대한 표준 해석 자료를 만들어 추후 평가에 이용할 수 있도록 하였다.

본 연구에서는 국내에서 가장 취약할 것으로 예상되는 원자력 발전소에 가압열충격 사고를 유발할 수 있는 주증기관 파단사고를 가정하여 열수력 해석과 파괴역학 해석을 수행하였다. 원전수명관리연구의 일환으로 계통열수력 해석 및 혼합열유동 해석에 의하여 구한 냉각제의 온도와 압력의 이력 및 용기의 재질성분으로부터 용기의 응력확대계수와 파괴인성치를 계산하고 이들을 비교하여 균열의 진전여부를 판단하여 형상계수가 1/6인 표면균열이 견딜 수 있는 최대 기준무연성천이온도를 결정하였다.