• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.10 no.1
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• pp.70-74
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• 2014

An Operating Experience Report(OER) has written about events and accidents happened at a Nuclear Power Plant(NPP). The purpose of publishing the OER is to prevent the similar event or accident repeatedly by spreading the experience of a single plant to other plants personnel. In this paper, it is analyses that the foreign NPPs' OERs on JIT published by the International Nuclear Agency(WANO, INPO, COG, BE). The analysis introduced in this paper is performed along with the various factors such as type of work, root-cause, and equipment. The root-cause analysis about the OERs shows that the Human-error is the major factor in foreign NPPs, but on the other hand equipment problem is the main part of the Domestic NPPs. The ratio of the foreign NPP's OERs on JIT according to the type of work was applied to KHNP-JIT developed nowadays for the first time in KOREA.

• Journal of Soil and Groundwater Environment
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• v.23 no.4
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• pp.1-15
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• 2018

Despite of safety issues related to radiological hazards, 31 countries around the world are operating more than 450 nuclear power plants (NPPs). To operate NPPs safely, safety regulations from radiation protection organizations were developed and adopted in many countries. However, many cases of radionuclide releases at foreign NPPs have been reported. Almost all commercial NPPs routinely release radioactive materials to the surrounding environments as liquid and gas phases under control. These releases are called 'planned releases' which are planned, regularly monitored, and well documented. Meanwhile, the releases focused in this review, called 'unplanned releases', are neither planned nor monitored by regulatory and/or protection organizations. NPPs are generally composed of various structures, systems and components (SSCs) for safety. Among them, the SSCs near reactors are closely related to safety of NPPs, and typically fabricated to comply with stringent requirements. However, some non-safety related SSCs such as underground pipes may be constructed only according to commercial standards, causing the leakage of radioactive fluids usually containing tritium ($^3H$). This paper discusses SSCs of NPPs and introduces several cases of unplanned releases at foreign NPPs. The current regulation on the environmental radiological surveillance and assessment around the NPPs in South Korea are also examined.

• Journal of the Korean Society of Safety
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• v.14 no.1
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• pp.177-184
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• 1999

PSA(Probabilistic Safety Assessment) methodology is widely used on assessing the safety of Nuclear Power Plants(NPPs) quantitatively in the domestic nuclear field. Initiating event frequencies are absolutely needed to conduct PSA, and they considerably affect PSA results. There is no domestic database where domestic trip event cases are reflected, so they are used to assess the safety of NPPs that are from the foreign database. In this paper, operating experience data from the Korean NPPs was collected and analyzed for the trip event cases, which are necessary to determine the initiating events and their frequencies. Korean NPPs have experienced five of 16 initiating events, which we LOFW. LOCV, LOCCW, LOOP and GTRN as a result of analyzing the trip event cases. Initiating frequencies based on the domestic trip event cases are analyzed, and they are similar to that from the foreign database.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.9 no.1
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• pp.35-39
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• 2013

The Operating Experience Report(OER) has written about the event and accident happened at a Nuclear Power Plant(NPP). The purpose of publishing the OER is to prevent the similar event or accident repeatedly by spreading the experience of a single plant to other plants personnel. Before initiating the analysis mentioned in this paper, 2,298 review reports for the same number of OER published from 2007 to June 2012 have been written to achieve the correct and objective statistics. The analysis introduced in this paper is performed with the various factors such as year, plant type, equipment, type of work, root-cause. The root-cause analysis is showed that the equipment problem is the major factor in domestic NPPs, but on the other hand human-error is the main part of the foreign NPPs. Moreover, while the number of the man-made event is decreasing, the equipment-made event is rapidly increasing in domestic NPPs.

• Journal of Welding and Joining
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• v.31 no.6
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• pp.1-7
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• 2013

In nuclear power plants, lined carbon steel pipes or PCCPs (pre-stressed concrete cylinder pipes) have been widely used for sea water transport systems. However, de-bonding of linings and oxidation of PCCP could make problems in aged NPPs (nuclear power plants). Recently at several NPPs in the United States, the PCCPs or lined carbon steel pipes of the sea water or raw water system have been replaced with HDPE (high density polyethylene) pipes, which have outstanding resistance to oxidation and seismic loading. ASME B&PV Code committee developed Code Case N-755, which describes rules for the construction of buried Safety Class 3 polyethylene pressure piping systems. Although US NRC permitted HDPE materials for Class 3 buried piping, their permission was limited to only 10-year operation because of several concerns including the quality of fusion zone of HDPE. In this study, various requirements for fusion qualification test of HDPE and some regulatory issues raised during HDPE application review in foreign NPPs are introduced.

• Corrosion Science and Technology
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• v.19 no.3
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• pp.131-137
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• 2020

A number of piping components in the secondary system of nuclear power plants (NPPs) are exposed to aging mechanisms, such as flow-accelerated corrosion (FAC), cavitation, flashing, solid particle erosion, and liquid droplet impingement erosion. Those mechanisms may lead to thinning, leaking, or rupture of the components. Due to the pipe ruptures caused by wall thinning in Surry unit 2 in the USA in 1986 and Mihama unit 3 in Japan in 1994, pipe wall thinning management has emerged as one of the most important issues in the nuclear industry. To manage pipe wall thinning, a foreign program has been utilized for NPPs in Korea since 1996. As our experience and knowledge of pipe wall thinning management have accumulated, our program needs to reflect our experience, requests from users, and the result of recent experiments using Flow Accelerated Corrosion Testing System (FACTS). FACTS is the empirical experimental facility developed by Korea Atomic Energy Research Institute (KAERI) for tests. Accordingly, KEPCO-E&C developed a 3D-based pipe wall thinning management program called ToSPACE in 2016. This paper describes a comparison between the FAC analysis results using ToSPACE and the experimental results using FACTS to verify their applicability to pipe wall thinning management in NPPs.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.4
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• pp.19-25
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• 2011

Steam Generator(SG) tube is an important component of Nuclear Power Plant(NPP), which is the pressure boundary between the primary and secondary systems. The integrity of SG tube has been confirmed by the eddy current test every outage. The eddy current technique adopting bobbin probe is currently the primary technique for the steam generator tubing integrity assesment. The bobbin probe is one of the essential components which consist of the whole ECT examination system and provides us a decisive data for the evaluation of tube integrity. Until now, all of the ECT bobbin probes in Korea which is necessary to carry out inspection are imported from overseas. However, KHNP has recently developed the bobbin probe design technology and transferred it to domestic manufacturers to fabricate the probes. This study has been conducted to establish technical requirements applicable to the steam generator tube inspection using the bobbin probes fabricated by the domestic manufactures. The results have been compared with the results obtained by using foreign probe to identify the availability to the steam generator tube inspection. As a result, it is confirmed that the domestic bobbin probe is generally applicable to SG tube inspection in the NPPs.

• Journal of Radiation Protection and Research
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• v.14 no.2
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• pp.1-9
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• 1989

An analytic method for quantitative comparisions between the alternatives for radiation protection optimization is required to aid the decision making process. This paper introduces the dynamic Markov model to evaluate the effect of inservice inspection, testing, and repair activities of the plant on radiation protection. In the example to put the Markov model into practice, the steam generator inspection intervals which minimize expected cost and total exposure dose were determined using the data for Kori-2 unit and foreign plants. The results show that the effect of the radiation exposure on the steam generator inspection interval is determined by the cost rather than the radiation exposure. The Markov model used in the example can be applied easily to the domestic NPPs by replenishing the data and also can be used in evaluating the comparative priority between various alternatives for radiation protection optimization.

• Economic and Environmental Geology
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• v.54 no.2
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• pp.285-297
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• 2021

Globally, nuclear-decommissioning facilities have been increased in number, and thereby hundreds of thousands of wastes, such as concrete, soil, and metal, have been generated. For this reason, there have been numerous efforts and researches on the development of technology for volume reduction and recycling of solid radioactive wastes, and this study reviewed and examined thoroughly such previous studies. The waste concrete powder is rehydrated by other processes such as grinding and sintering, and the processes rendered aluminate (C3A), C4AF, C3S, and ��-C2S, which are the significant compounds controlling the hydration reaction of concrete and the compressive strength of the solidified matrix. The review of the previous studies confirmed that waste concretes could be used as recycling cement, but there remain problems with the decreasing strength of solidified matrix due to mingling with aggregates. There have been further efforts to improve the performance of recycling concrete via mixing with reactive agents using industrial by-products, such as blast furnace slag and fly ash. As a result, the compressive strength of the solidified matrix was proved to be enhanced. On the contrary, there have been few kinds of researches on manufacturing recycled concretes using soil wastes. Illite and zeolite in soil waste show the high adsorption capacity on radioactive nuclides, and they can be recycled as solidification agents. If the soil wastes are recycled as much as possible, the volume of wastes generated from the decommissioning of nuclear power plants (NPPs) is not only significantly reduced, but collateral benefits also are received because radioactive wastes are safely disposed of by solidification agents made from such soil wastes. Thus, it is required to study the production of non-sintered cement using clay minerals in soil wastes. This paper reviewed related domestic and foreign researches to consider the sustainable recycling of concrete waste from NPPs as recycling cement and utilizing clay minerals in soil waste to produce unsintered cement.