The first commercial operation of Kori-1, which commenced in April 1978, was permanently shut down in June 2017, with plans for immediate dismantling. The decommissioning process of nuclear power plants generates a substantial amount of radioactive waste and poses significant radiation exposure risks to workers. Radioactivity is widely distributed throughout the primary coolant system of the reactor, including the reactor pressure vessel (RPV), steam generator (SG), and pressurizer. In particular, the SG has a considerable size and complex geometry, weighing approximately 326 tons and having a volume of 400 m3. The SG tubes are known to contain high levels of radioactivity, leading to significant radiation exposure to workers during the dismantling process. Therefore, this study aims to evaluate the workers’ radiation exposure during the cutting of SG tubes, which account for approximately 95% of the total radiation dose in the SG. Firstly, the CRUDTRAN code, developed to predict the behavior of soluble and particulate corrosion products in a PWR primary coolant system, is used to estimate the radioactive inventory in the SG tubes. Based on decontamination factors (DF) obtained in the SG tubes through overseas experience, the expected reduction in radioactivity during the Kori-1 reactor’s full-system decontamination (FSD) process is considered in the CRUDTRAN results. These results are then processed to estimate the radioactivity in both the straight and bent sections of the tubes. Subsequently, these radioactivity values are used as inputs for the MicroShield code to calculate the worker radiation exposure during the cutting of both straight and bent sections of the tubes. The cutting process assumes that each SG tube section is cut in a separate, shielded area, and the radiation exposure is assessed, taking into account factors such as cutting equipment, cutting length, working hours, and working distance. This study evaluates the worker radiation exposure during the cutting of SG tubes, which are expected to have a significantly high radioactivity due to chalk river unidentified deposit (CRUD). This assessment also considers the reduction in radioactivity within the steam generator tubes resulting from the FSD process. Consequently, it enables an examination of how worker radiation exposure varies based on the extent of FSD. This study may provide valuable insights for determining the scope and extent of the FSD process and the development of shielding methods during the dismantling of SG tubes in the future.

In South Korea, the replacement of steam generators began with Kori Unit 1 in 1995, and to date, 20 steam generators have been replaced and are currently stored in intermediate storage facilities. In the future, additional decommissioned steam generators may arise due to measures like the extension of the lifespan of nuclear power plants. In Korea, technological development for dismantling steam generators is underway, and there is no track record of actual dismantling. Although the replaced decommissioned steam generators are stored in intermediate facilities, for site recycling purposes, steam generators, which have relatively lower radiation doses compared to reactor heads and other primary equipment, should be prioritized for dismantling. While there are various specifications for steam generators, those dismantled and stored domestically are of the Recirculation Type. They can be classified into three types: the Westinghouse type WH-51 used in Kori Unit 1, the Fra-51B used in Han-ul Units 1 and 2, and the OPR-1000 used in Han-ul Units 3 and 4. The quantity of U-Tubes varies depending on the specification, but the radiation is concentrated in the primary side components, the U-Tube and Chamber. Since the parts related to the secondary side are not contaminated, they can be disposed of independently after classification. To dismantle a steam generator, it is of utmost importance to first create a scenario regarding where and how the dismantling will take place. Through the analysis of the advantages and disadvantages of each scenario, the optimal timing, location, and cutting method for dismantling should be researched. Furthermore, based on those findings, the best scenario should be derived through an analysis of worker radiation exposure and dismantling costs. To achieve this, a 3D simulation software developed by Cyclelife Digital Solutions under the French EDF was utilized to conduct simulations based on different dismantling schedules and methods. As a result, the optimal scenario for dismantling the steam generator was derived.

Kori Unit 1, pressurized water reactor, is the Korea’s first commercial nuclear power plant. It successfully generated electricity for a period of 30 years, commencing from April 19, 1978. Following its approval for continued operation in 2008, Kori Unit 1 continued to operate for an additional 9 years, resulting in a total operational period of 39 years. On June 18, 2017, Kori Unit 1 was permanently shut down. Since then, Korea is actively preparing for the decommissioning of nuclear power plant. During the decommissioning of a nuclear power plant, the heavy components such as reactor, steam generator, pressurizer, reactor coolant pump located in the containment building should be taken out of the containment building. To take out heavy components from the containment building, pipes connected to heavy component should be cut. There are numerous pipes connected to the heavy component, each with varying dimensions and material. Each pipe has a different level of contamination depending on its use. In this study, optimal cutting method of pipe connected to steam generator, one of the heavy components of nuclear power plant, is proposed during the decommissioning of Kori unit 1. In case of pipe connected to Kori unit 1 steam generator, material is stainless steel or carbon steel. These pipes have varying inner diameter, ranging from 0.6 cm to 74 cm, and thickness ranging from 0.15 cm to 7.1 cm. These pipes are classified as low and intermediate level waste (LILW) or very low level waste (VLLW). Because characteristics of pipes are different, each pipe optimal cutting methods are proposed differently considering material, dimension, contamination level, cutting cost, cutting time, and the management of secondary waste. As a result, the cutting method for pipe of reactor coolant system is selected to orbital cutting. The cutting method of main steam pipe and main feedwater pipe is selected to oxygen cutting. In case of other small pipes, cutting method is selected to circular saw.

The potential impact of hypothetical accidents that occur during the immediate and deferred dismantling of the Kori Unit 1 steam generator has been comprehensively evaluated. The evaluation includes determining the inventory of radionuclides in the Steam Generator based on surface contamination measurements, assuming a rate of release for each accident scenario, and applying external and internal exposure dose coefficients to assess the effects of radionuclides on human health. The evaluation also includes calculating the atmospheric dispersion factor using the PAVAN code and analyzing three years of meteorological data from Kori NPP to determine the degree of diffusion of radionuclides in the atmosphere. Overall, the effective dose for residents living in the Exclusion Area Boundary (EAB) of Kori NPP is predicted, an it is found that the maximum level of the dose is 0.034% compared to the annual dose limit of 1 mSv for the general public. This implies that the potential impact of hypothetical accidents on human health discussed above is within acceptable limits.

Kori Unit 1, Korea’s first commercial nuclear power plant is preparing to dismantle after 40 years of power supply. However, unlike the public dose assessment for nuclear power plants in operation, the dose assessment for the public due to abnormal events during the decommissioning of nuclear power plants is insufficient. Therefore, in this study, the steam generator chamber is selected as hypothetical events target among metal waste, which is a major radioactive material generated during the decommissioning of nuclear power plant. In addition, the possible abnormal event scenarios and effective does to public in the Exclusion Area Boundary due to the released radioactive materials are predicted during the disassembly and transportation of the steam generator. For the source term that can be released during the dismantling of the steam generator, the inventory of each radionuclide is evaluated based on the smear test results of the steam generator replaced in Kori Unit 1 in 1998. To evaluate the diffusion of radioactive material, the atmospheric dispersion factor (χ/Q, sec/m3) is calculated through the PAVAN code of the US NRC using the meteorological data of the Kori nuclear power plant for 3 years from 2019 to 2021 according to IAEA recommendations. For the assessment of the public dose, the external dose coefficient and inhalation coefficient of the ICRP and the inhalation rate of the NRC Regulatory Guide 1.3 are referred. It is confirmed that the effective dose to the public in the Exclusion Area Boundary due to the abnormal event during the dismantling of the steam generator is much lower than the effective dose standard value of 250 mSv for 2 hours after the event in the Exclusion Area Boundary.

Severe wall thinning is found on the tube of a low-pressure evaporator(LPEVA) module that is used for a heat recovery steam generator(HRSG) of a district heating system. Since wall thinning can lead to sudden failure or accidents that lead to shutdown of the operation, it is very important to investigate the main mechanism of the wall thinning. In this study, corrosion analysis associated with a typical flow-accelerated corrosion(FAC) is performed using the corroded tube connected to an upper header of the LPEVA. To investigate factors triggering the FAC, the morphology, composition, and phase of the corroded product of the tube are examined using optical microscopy, scanning electron microscopy combined with energy dispersive spectroscopy, and x-ray diffraction. The results show that the thinnest part of the tube is in the region where gas directly contacts, revealing the typical orange peel type of morphology frequently found in the FAC. The discovery of oxide scales containing phosphate indicates that phosphate corrosion is the main mechanism that weakens the stability of the protective magnetite film and the FAC accelerates the corrosion by generating the orange peel type of morphology.

대한민국 첫 상업원전인 고리1호기는 40년간의 성공적인 운전을 끝내고 2017년 6월 18일 영구정지 되었다. 고리1호기는 본격적인 해체에 앞서 터빈건물에 폐기물처리시설 건설을 계획하고 있다. 각종 방사성폐기물은 폐기물처리시설에서 제염, 해체, 절단, 용융되어 자체처분 되거나 방사성폐기물 처분장으로 보내 진다. 해체폐기물 중 대형금속방사성폐기물은 주로 1차 계통측 기기들로 높은 방사능을 띄고 있어 해체활동 중 작업자의 피폭관리가 필요하다. 본 논문에서는 대형금속방사성폐기물 중 크기가 가장 크고 형상이 복잡한 증기발생기를 선정하여 RESRAD-RECYCLE 코드를 이용하여 작업자 피폭선량을 평가하고 저감화 방안을 수립 하고자 한다.

원자력 발전소의 증기발생기 슬러지 중에서는 이온교환수지가 발견되어서는 안 된다. 원자력 발전소의 증기발생기 슬러지 시료 중에서 발견되어 이온교환수지 입자로 의심되는 구형 입자들의 특성을 측정하였다. 미세조작기술을 이용하여 광학현미경으로 입자 크기 분포를, EPMA로 구형입자의 성분을, 그리고 IR 분광 스펙트럼 비교에 의하여 이온교환수지 여부를 조사하였다. 슬러지의 입자 크기는 1 내지 이었으나 구형 입자는 이었다. 슬러지의 주요 불순원소가 Si, Al, Mn, Cr, Ni, Zn, 그리고 Ti이었으나 구형 입자는 Si, Cu, Zn 이었다. 주성분은 두 경우 모두 철이었다. 구형 입자의 IR 분광스펙트럼은 증기발생기 취출수 정화계통에서 사용하는 이온교환수지의 스펙트럼과 비교했을 때 서로 일치하지 않음을 보여주었다. 이 결과들은 증기발생기 슬러지 시료 중에서 발견된 구형 입자가 이온교환수지는 아니며 일반적인 슬러지가 생성되는 과정에서 작은 슬러지 입자들이 크게 뭉쳐서 생성된 것임을 나타내고 있다.