It is likely to occur internal exposure for workers in Nuclear Power Plants (NPPs) due to the intake of radionuclide. To assess the internal exposure dose the measurement of activity for remain radionuclide is necessary. The Whole Body Counters (WBCs) are commonly used for measurement of remain radionuclide activity in human body. Korea Hydro & Nuclear Power Co., Ltd. (KHNP) conduct performance test of WBCs in all NPPs for every year to confirm the performance of equipment. The performance test is conducted using unknown sources and the participants of the comparison test submit the radionuclide and activity of the unknown sources measured by WBC as a result. The performance indicator and criteria for WBC recommended in the American National Standards Institute (ANSI) N13.30 report published in 2011 are applied. The performance indicator is Root Mean Squared Error (RMSE) and criteria is 0.25 or less. The results of performance test performed in 2022 for all WBC is meet the ANSI N13.30 criteria. And the RMSE values are confirmed from 0.01 to 0.23. This means that the residual radioactivity measurement results using WBC are reliable.

Under Article 17 of the Radioactive Waste Management Act and Article 12 of the Enforcement Decree of the Radioactive Waste Management Act, KHNP shall reserve the cost for the decommissioning of NPPs as provisions. To preserve the value, an additional amount considering the discount rate is to be added annually. The initial provision is decided by estimating the decommissioning cost of NPP at the time of commercial operation, calculating the future cost by applying the inflation rate to the expected start date of decommissioning, and then discounting it at a discount rate to the present value. According to the current notice, the period for applying inflation and discount rate is defined as the period of 5 years added to the design life of NPP, which is presumed to be due to the assumption that all decommissioning costs are incurred at once 5 years after the permanent shutdown of the power plant. However, assuming that the actual decommissioning period of a domestic nuclear power plant is generally planned for 15 years, it can be expected that most of the decommissioning activities will begin after the decommissioning preparation and transition period, or 5 years after permanent shutdown of the plant. Considering this, it can be said that the current period (5 years + design life) for applying inflation and discount rate is set a little conservatively. In this paper, the initial provision is calculated by appropriately distributing the decommissioning costs of overseas NPPs categorized by International Structure for Decommissioning Costing (ISDC) during the planned decommissioning period of domestic NPPs, and then adding up the decommissioning cost each year by separately applying the inflation and discount period, which was compared with the results calculated using the current method. Through this, it was confirmed that the revised method had the effect of reducing the initial provision by 2.2% to 5.7% compared to the current method depending on the gap between inflation rate and discount rate, which can be converted to about 8 years of inflation and discount period used in the current method. It is expected that this paper will be used in the future as a basic reference for developing a more accurate method for calculating the initial provision of decommissioning cost.

In this study, a manual that can be applied to conflict management of clearance waste recycling by stakeholders was researched to recycle clearance waste that is most frequently generated when decommissioning nuclear power plants. In order to develop a manual that can be applied to conflict management, the content of the conflict should be derived first. In order to derive conflict, it is necessary to organize major issues in recycling clearance waste in consideration of domestic nuclear energy and social environment. In order to organize major issues in consideration of the domestic environment, a literature survey and a domestic current situation investigation were conducted. At this time, the subject of the major issue was selected based on the Level 1 influencing factors of the previous study. As a result of the investigation, it was confirmed that there were many major issues due to lack of reliability/understanding in nuclear energy/radiation. Through this Conflicts caused by recycling clearance waste were derived based on the organized issues. As a result of deriving conflicts, eight conflicts were derived below. 1) Reduced business availability due to lack of understanding/reliability 2) Lack of reliability in the selection and technology of nuclide analysis technology 3) Additional time and equipment required due to establishment of clearance waste regulatory requirements 4) Low economic benefits due to reduction in the effect of substituting raw materials 5) Political interference due to worsening public opinion 6) Rejection of final products due to recycling due to distrust of radiation 7) Public acceptance along the transport route from the source to the recycling plant 8) Business promotion deteriorated due to changes in energy policy As a result of the derived conflict analysis, the most conflicts related to lack of reliability/understanding in nuclear energy/radiation were derived. Accordingly, in future research, it is necessary to prepare a specific plan to enhance the understanding of stakeholders about self-disposal waste recycling. Considering that research that can solve the conflicts that will be faced when the domestic/foreign clearance waste recycling industry is activated is not activated, this study is meaningful in that it derived the conflicts that will be faced when recycling clearance waste. Also, it is expected that the conflicts derived from this study will be used meaningfully in the establishment of the clearance waste recycling management manual.

The decommissioning of nuclear power plant (NPP) consists of various activities, such system decontamination, take out of activated components, segmentation of the activated components, site remediation, etc. During various activities, the generation of radioactive wastes and radiation exposure to workers is expected. The systematic waste management during the activities is important to implement the decommissioning. The inefficient waste management usually bring significant delay in decommissioning process and results in increase of decommissioning cost. The radiation exposure management is also an important issue. It is generally accepted that the hot spot, generated from operation and decommissioning of NPP, is observed in many places within containment building. Although the health physicists measure the radiation in various points, the unintended hot spots are sometimes generated and observed. The effective radiation exposure management also requires the control of personnel and space during various activities. In this study, the radiation exposure and waste management experiences of Zion NPP is reviewed. The primary nuclides and radiation exposure during various activities are systematically studied to achieve the main objectives of this paper.

The decommissioning of Kori Unit 1 is expected to generate a large amount of clearance waste. Disposing of a large amount of clearance waste is economically costly, so a recycling method has emerged. However, clearance waste recycling is expected to cause many conflicts among various stakeholders. In the previous study, possible conflicts were selected in consideration of the domestic environment and major issues. Based on this, this study classifies stakeholders involved in conflicts by group, and suggests ways to enhance understanding by stakeholder and enhance reliability. In this study, stakeholders are classified into four groups that share the same conflicts, and each of the following measures is suggested. 1) Stakeholder Engagement. 2) Common understanding of radiation risks, dialogue between the public/recycling industry/ regulatory agency. 3) Incentives to promote recycling clearance waste. 4) Reliable outlet store for recyclable clearance waste. The above understanding enhancement measures are presented so that a solution to conflict can be smoothly derived when designing a clearance waste-related consultative body composed of interested parties in the future. As a more specific solution, measures to enhance stakeholder trust can be suggested for each understanding enhancement measure. Reliability enhancement measures are also presented so that they can be applied to each stakeholder group, and these are as follows. 1) Write a stakeholder engagement plan, Measures for stakeholder participation in measuring the radioactivity concentration of clearance waste. 2) Active use of easy-to-understand radioactivity comparison data, Expansion of information on environmental radiation dose to public, nuclear/radiation education, Held a tour event at the nuclear power plant decommissioning site, New website for clearance waste information disclosure. 3) Incentives for recycling industries in which the Ministry of Environment or KHNP partially bears the losses that occur when the sales rate is low. Incentives are provided to consumers by including recyclables of clearance waste for Green Card’s green consumption points. 4) Online outlets open for recyclable clearance waste with easy-to-understand radioactivity comparison data. It is expected that if the above-mentioned reliability enhancement measures are used, it will be possible to secure the trust of stakeholders and reduce the gap between stakeholders in the future clearance-related consultative body.

Kori unit 1 was permanently shut down in 2007 and is currently awaiting approval for decommissioning and dismantling (D&D). The wastes generated during decommissioning is estimated to be approximately 14,500 of 200 L drums. In this study, the treatment process of decommissioning wastes will be reviewed through the case of the US Zion nuclear power station (ZNPS). Zion unit 1 and 2 received an operating license in 1973 and were permanently shut down and the spent nuclear fuel was transferred to the pool in 1998. The decommissioning was carried out according to the following five steps; (1) safe storage (SAFSTOR) dormancy, (2) preparation for decommissioning, (3) establishment of independent spent fuel storage installation (ISFSI) and transfer of the spent fuel and greater than class C radioactive materials, (4) decommissioning operations and (5) site restoration. The total volume of waste generated during decommissioning was expected to be approximately 1.7×105 m3. This is far above the Kori unit 1 waste estimation because ZNPS has a history of accidents and includes soil waste. Wastes were treated differently according to their properties and locations.

Reliable evaluation of radioactivity inventory for the nuclear power plant components and residual materials is very important for decontamination and decommissioning. This can make it possible to define optimum dismantling approaches, to determine radioactive waste management strategies, and to estimate the project costs reasonably. To calculate radioactivity of the nuclear power plant structure, various information such as interest nuclide, cross-section, decay constant, irradiation time, neutron flux, and so on is required. Especially irradiation time and neutron flux level are very changeable due to cycle specific fuel loading pattern, the plant overhaul, cycle length. However most of the radioactivity calculations have generally been performed assuming one representative or average neutron flux during the lifetime of the nuclear power plant. This assumption may include excessive conservatism because the radioactivity level has the characteristics of saturation and decay. Therefore, considering these variables as realistically as possible could prevent overestimation. In order to perform realistic radioactivity calculation, we developed monthly relative power contribution factor applying plant-specific operation history and cycle-specific neutron flux. The factors were applied to the radioactivity calculation. The calculation results ware compared with measured values of the neutron monitors that were actually installed and withdrawn from the nuclear power plant. As a result of the comparisons, there are good agreements between the calculated values and measured values. These accurate calculation results of radioactivity could contribute to the establishment of radioactive waste dismantling strategies, the classification of radioactive waste, and the deposit of disposal costs for safe and reasonable decommissioning of the nuclear power plant.

Solid radioactive waste such as rubble, trimmed trees, contaminated soil, metal, concrete, used protective clothing, secondary waste, etc. are being generated due to the Fukushima nuclear power plant accident occurred on March 11, 2011. Solid radioactive waste inside of Fukushima NPP is estimated to be about 790,000 m3. The solid radioactive waste includes combustible rubble, trimmed trees, and used protective clothing, and is about 290,000 m3. These will be incinerated, reduced to about 20,000 m3 and stored in solid waste storage. The radioactive waste incinerator was completed in 2021. About 60,000 m3 of rubble containing metal and concrete with a surface dose rate of 1 mSv/h or higher will be stored without reduction treatment. Metal with a surface dose rate of 1 mSv/h or less are molten, and concrete undergoes a crushing process. About 60,000 m3 of contaminated soil (0.005 ~1 mSv/h) will be managed in solid waste storage without reduction treatment. The amount of secondary waste generated during the treatment of contaminated water is about 6,500 huge tanks, and additional research is being conducted on future treatment methods.

In 2017, Kori unit 1 nuclear power plant was permanently shut down at the end of its life. Currently, Historical Site Assessment (HSA) for MARSSIM characteristics evaluation is being conducted according to the NUREG-1575 procedure, this is conducted through comprehensive details such as radiological characteristics preliminary investigation and on-site interview. Thus, the decommissioning of nuclear power plant must consider safety and economic feasibility of structures and sites. For this purpose, the establishment of optimal work plan is required which simulations in various fields. This study aims to establish procedure that can form a basis for a rational decommissioning plan using the virtual nuclear power plant model. The mapping procedure for 3D platform implementation consisted of three steps. First, scan the inside and outside of the nuclear power plant for decommissioning structure analysis, 3D modeling is performed based on the data. After that, a platform is designed to directly measure the radiation dose rate and mapped the derived to the program. Finally, mapping the radiation dose rate for each point in 3D using the radiation dose rate calculation factor according to the time change the measured value created on the 3D mapping platform. When the mapping is completed, it is possible to manage the exposure dose of workers according to the ALARA principle through the charge of radiation dose rate over time because of visualization of the color difference to the radiation dose rate at each point. For addition, the exposure dose evaluation considering the movement route and economic feasibility can be considered using developed program. As the interest in safety accidents for workers increases, the importance of minimum radiation dose and optimal work plan for workers is becoming increasingly important. Through this mapping procedure, it will be possible to contribute to the establishment of reasonable process for dismantling nuclear power plant in the future.

Laser cutting has been attracting attention as a next-generation tool in application for nuclear decommissioning. It enables high-speed cutting of thick metal objects, and its narrow kerf width greatly reduces the amount of secondary waste compared to other cutting methods. In addition, it only requires the relatively small cutting head without any complicated equipment, and long-distance cutting apart from a laser generator is possible using beam delivery through optical fiber. And there is almost no reaction force because it is non-contact thermal cutting. For these reasons, the laser cutting is very advantageous for remote cutting. In laser cutting, the irradiated laser power is absorbed and consumed to melt the material of the cutting target. When the applied laser power is greater than the power consumed for melting, the residual power is transmitted to the back of the cut object. This residual power may unintentionally cut or damage undesired objects located behind the cutting target. In order to prevent this, it is necessary to adjust the laser power for each thickness of the target object to be cut, or to increase the distance between the cut target and the surrounding structures so that the transmitted power density can be sufficiently lowered. In this work, safety study on residual power that penetrates laser-cut objects was conducted. Experimental studies were performed to find safe conditions for irradiation power density that does not cause surface damage to the stainless steel by adjusting the laser power and stand-off distance from the target.

Regulations on the concentration of boron discharged from industrial facilities, including nuclear power plants, are increasingly being strengthened worldwide. Since boron exists as boric acid at pH 7 or lower, it is very difficult to remove it in the existing LRS (Liquid Radwaste System) using RO and ion exchange resin. As an alternative technology for removing boron emitted from nuclear power plants, the electrochemical boron removal technology, which has been experimentally applied at the Ringhal Power Plant in Sweden, was introduced in the last presentation. In this study, the internal structure of the electrochemical module was improved to reduce the boron concentration to 5 mg/L or less in the 50 mg/L level of boron-containing waste liquid. In addition, the applicability of the electrochemical boron removal technology was evaluated by increasing the capacity of the unit module to 1 m3/hr in consideration of the actual capacity of the monitor tank of the nuclear power plant. By applying various experimental conditions such as flow rate and pressure, the optimum boron removal conditions using electrochemical technology were confirmed, and various operating conditions necessary for actual operation were established by configuring a concentrated water recirculation system to minimize secondary waste generation. The optimal arrangement method of the 1 m3/hr unit module developed in this study was reviewed by performing mathematical modeling based on the actual capacity of monitor tank and discharge characteristics of nuclear power plant.

The purpose of this study is to provide lessons learned in the experience of improvement work of fuel handling equipment at operating nuclear power plants. The upgrade of fuel handling equipment for safety enhancement and performance improvement has been going on for 15 years since the early 2000’s. The main goal is to increase fuel loading/unloading capability of the equipment from about 2.5 fuel assemblies per hour to more than six (6). It is achieved with sequential operations of three (3) fuel handling equipment, which consists of the refueling machine, the fuel transfer system and the spent fuel handling machine. The scope of the upgrade for fuel handling equipment is summarized as follows. The PC data control system based on PLC for interlocks and high speed motor drive system is introduced for better operating efficiency. The motors and drives for bridge, trolley, and hoist are replaced with AC servomotors and drivers, respectively. The fuel transfer system has an auto-initiation feature operating from the refueling machine or the spent fuel handling machine. The winch motor and drive for the carriage of fuel transfer system is also replaced with AC servomotors and drivers. And some of HPU (hydraulic power units) equipment for each building (reactor containment building and fuel handling building) are replaced to improve their function. The considerations for improvement of fuel handling equipment are as belows. 1) Fuel handling should be consistent with the instructions provided by the fuel designer and/or manufacturer, which are for Standard type fuel and Westinghouse type fuel, used in domestic nuclear power plants. Each fuel has unique design characteristics, which are PLC setpoints for overload and underload, slow speed zones for a bridge, trolley and hoist, allowable acceleration/deceleration value in handling, hoist elevation and manual speed in off-index operation at reactor. 2) The interlock system should be designed in accordance with design criteria specified by the utilities of nuclear power plant. 3) Performance should be improved according to the operating characteristics of the fuel handling equipment. High-speed and optimization of FTS upender and carriage are essential for operating performance so that its modification should be considered first. And the low speed and range in the operation mechanism of the hoist should be designed to comply with guidelines. 4) The accident analysis through self-diagnosis function and operation history in modification at domestic operating nuclear power plants would be good lessons learned. It is advisable to utilize such various information as it can help to improve reliability of nuclear fuel handling operation and power plant operation rate.

Nuclear power plants (NPPs) are designed in consideration of redundancy, diversity, and independence to prevent leakage of radioactive materials from safety of view, and a contingency plan is established in case of DBA (Design Basis Accident) occurrence. In addition, NPPs have established contingency plans for physical attacks, including terrorist intrusions and bomb attacks. However, the level of contingency plan caused by cyberattacks is quite insufficient compared to the contingency plan in terms of safety and physical protection. The purpose of this paper is to present the problems of cyberattack contingency plan and methods to supplement it. The first problem with cyberattack contingency plan is that the initiating event for implementing the contingency plan is undecided. In terms of safety, the DBA is identified as an initial event, and each contingency plan is based on the initial events specified in the DBA such as Loss of Coolant Accident and Loss of Offsite Power. In terms of physical protection, each has a contingency plan by identifying bomb attacks and terrorist intrusions in Protected Area and Vital Area as initial events. On the other hand, in the contingency plan related to a cyberattack, an initial event caused by a cyberattack is not identified. For this, it is necessary to classify the attack results that may occur when the CDA is compromised based on the attack technique described in Design Basis Threat. Based on this, an initiating event should be selected and a contingency plan according to each initiating event should be established. The second problem is that there is no responsibility matrix according to the occurrence of the initiating event. From a safety point of view, when a DBA occurs, the organization’s mission according to each initial event is described in the contingency plan, and related countermeasures are defined in case of an accident through Emergency Operation Procedure. In the case of physical protection, referring to IAEA’s Regulatory Guide 5.54, the organization’s responsibility is defined in matrix form when an initial event such as a bomb attack occurs. In this way, the responsibility matrix to be carried out in case of initiating events based on the defined initial event should be described in the contingency plan. In this paper, the problems of the cyberattack contingency plan are presented, and for this purpose, the definition of the initial event and the need for a responsibility matrix when the initial event occurs are presented.

When exporting nuclear power plants to a third country, the U.S. conditions import countries to join the International Atomic Energy Agency (IAEA) Additional Protocol. At the Korea-U.S. summit, Korea also agreed to maintain equal non-proliferation standards. This paper first analyzes how the U.S. applies the conditions for joining additional protocols to export control policies. The U.S. Atomic Energy Act is a general law in the field of nuclear power that governs both civilian and military use of nuclear power. Article 123 stipulates matters related to “cooperation with other countries.” According to Article 123, the United States must conclude a peaceful nuclear cooperation agreement with another country that stipulates nuclear non-proliferation obligations for nuclear cooperation to a “significant” extent. Article 123 of the Nuclear Energy Act presents nine conditions for signing the Nuclear Cooperation Agreement, and matters related to safeguards are stipulated in Nos. 1 and 2, and only IAEA’s Comprehensive Safeguards Agreement (CSA) is specified as requirements under the current law. As a result of analyzing the countries of the nuclear cooperation agreements currently signed by the United States, the United States is evaluating the AP in terms of the policy as an essential item. Among the nuclear agreements with the United States, three countries, Egypt, Brazil, and Argentina do not have AP in effect. Among them, Brazil and Argentina are recognized by the IAEA as replacing the ABACC with the AP, so only Egypt is not a member of the AP. The nuclear agreement between the U.S. and Egypt was signed in 1981 before the AP existed, and all recently signed agreements were identified as AP-effective countries. As a result of reviewing the U.S. export control laws, the U.S. did not legislate the AP as a condition for peaceful nuclear exports. Reflecting the NSG export control guidelines, AP was legislated as an export license requirement only in exporting sensitive nuclear technology (enrichment, reprocessing). However, it is confirmed that the U.S. policy applies AP entry into force as one of the main requirements for determining whether it is harmful to nuclear exports, along with the conclusion of the Nuclear Cooperation Agreement, the application of the Comprehensive Safeguards Agreement, and military alliance. The appropriate scope of application of the Additional Protocol in Korea and its application plan will be suggested through future research.

UAVs (Unmanned Aerial Vehicle) are a rising threat to national facilities due to their cheap price and accessibility. Incidents such as the terrorism attack in Saudi Arabia’s oil facilities and the paralysis of the airport system in England’s Gatwick airport shows the need for integrating CUAS (Counter- Unmanned Aerial Systems) in important national facilities. Recently efforts have been made to evaluate the technical performance of the CUAS. Especially SNL (Sandia National Laboratory) modified the methodology used for PPS (Physical Protection Systems) to develop a performance metrics for CUAS. The performance metrics can be used to effectively analyze the facilities capability of countering drone attacks in a probabilistic way. In this study, we managed to derive the safety boundary of a reference nuclear power plant model based on its current CUAS and protection capabilities with a simplified methodology. Based on the outermost boundary of the model, the time table of the UAS consist of 4 variables which are the assessment time, transmission time, neutralization time and the maximum vehicle velocity. Dividing the maximum velocity to the net time derived, we estimated the minimum sensing point of the CUAS which is the minimum safety boundary of the facility to safely manage the UAV attack. Two practice cases were evaluated with the methodology which is based on the UAV groups classified by the United States DOD (Department Of Defense) that matches the classification of the UAV in Korea. Each variable was assumed to fit the process of a realistic nuclear power plant. Using the variables, we calculated the minimum safety boundary of the facility. With the methodology introduced in this study, regulators and stakeholders can easily evaluate the capability of the facilities CUAS for a design basis UAV attack. Also it can be used as a simple tool to analyze the facilities vulnerability for specific UAV specifications and a guideline to check the protective procedures of the facility.

Nuclear power plants, which are important national facilities, require special attention against the threat of terrorism using various methods. Among the terrorist threats, as structural damage and human casualties due to explosions continue to occur, interest in the blast load is increasing. However, domestic nuclear power plants do not have sufficient design requirements for protection against the threat of explosives. To prepare for the threat of terrorism using explosives, it is necessary to evaluate the physical protection performance of nuclear power plants against blast load, and to use this to improve protection performance and establish regulatory standards. Most of the explosion-proof designs used abroad use the empirical chart presented by UFC 3-340- 02 (DoD 2008), which does not take into account the effect of near-field explosions. When explosions occur inside nuclear power plants, near-field explosions occur in most cases. In this study, it was assumed that explosives were installed in the corridor inside nuclear power plants. A spherical TNT was placed in the middle of the corridor floor to simulate near-field explosions, and the structure response according to the weight of the TNT was evaluated. The corridor was modeled with a reinforced concrete material and the LS-DYNA program was used for analysis. For the explosion model, the Arbitrary-Lagrangian-Eulerian (ALE) analysis technique applying the advantages of the Lagrangian and Eulerian methods were used. By analyzing the pressure history and the degree of deformation of the structure according to the explosion, the degree of threat caused by the explosion was analyzed. Based on the analysis of this study, physical barriers performance database (DB) using Modeling & Simulation (M&S) will be constructed by performing sensitive analysis such as representative structure shape setting, boundary conditions, material of structures, etc. The constructed DB is expected to be used to establish regulatory standards for the physical barriers of nuclear power plants related to explosives.

Operating and decommissioning nuclear power plants generates radioactive waste. This radioactive waste can be categorized into several different levels, for example, low, intermediate, and high, according to the regulations. Currently, low and intermediate-level waste are stored in conventional 200-liter drums to be disposed. However, in Korea, the disposal of intermediate-level radioactive waste is virtually impossible as there are no available facilities. Furthermore, large-sized intermediate- level radioactive waste, such as reactor internals from decommissioning, need to be segmented into smaller sizes so they can be adequately stored in the conventional drums. This segmentation process requires additional costs and also produces secondary waste. Therefore, this paper suggests repurposing the no-longer-used spent nuclear fuel casks. The casks are larger in size than the conventional drums, thus requiring less segmentation of waste. Furthermore, the safety requirements of the spent nuclear fuel casks are severer than those of the drums. Hence, repurposed spent nuclear fuel casks could better address potential risks such as dropping, submerging, or a fire. In addition, the spent nuclear fuel casks need to be disposed in compliance with the regulations for low level radioactive waste. This cost may be avoided by repurposing the casks.

Nuclear power plant decommissioning generates significant concrete waste, which is slightly contaminated, and expected to be classified as clearance concrete waste. Clearance concrete waste is generally crushed into rubble at the site or a satellite treatment facility for practical disposal purposes. During the process, workers are exposed to radiation from the nuclides in concrete waste. The treatment processes consist of concrete cutting/crushing, transportation, and loading/unloading. Workers’ radiation exposure during the process was systematically studied. A shielding package comprising a cylindrical and hexahedron structure was considered to reduce workers’ radiation exposure, and improved the treatment process’s efficiency. The shielding package’s effect on workers’ radiation exposure during the cutting and crushing process was also studied. The calculated annual radiation exposure of concrete treatment workers was below 1 mSv, which is the annual radiation exposure limit for members of the public. It was also found that workers involved in cutting and crushing were exposed the most.

Major accidents at nuclear power plants generate huge amounts of radioactive waste in a short period of time over a wide area outside the plant boundary. Therefore, extraordinary efforts are required for safe management of the waste. A well-established remediation plan including radioactive waste management that is prepared in advance will minimize the impact on the public and environment. In Korea, however, only limited plans exist to systematically manage this type of off-site radioactive waste generating event. In this study, we developed basic strategies for off-site radioactive waste management based on recommendations from the IAEA (International Atomic Energy Agency) and NCRP (National Council on Radiation Protection and Measurements), experiences from the Fukushima Daiichi accident in Japan, and a review of the national radioactive waste management system in Korea. These strategies included the assignment of roles and responsibilities, development of management methodologies, securement of storage capacities, preparation for the use of existing infrastructure, assurance of information transparency, and establishment of cooperative measures with international organizations.

In assessing the seismic safety of nuclear power plants, it is essential to analyze the structures using the observed ground motion. In particular, spatial variation in which the characteristics of the ground motion record differ may occur if the location is different within the site and even if the same earthquake is experienced. This study analyzed the spatial variation characteristics of the ground motion observed at the structure and site using the earthquake records measured at the Hamaoka nuclear power plant. Even if they were located on the same floor within the same unit, there was a difference in response depending on the location. In addition, amplification was observed in Unit 5 compared to other units, which was due to the rock layer having a slower shear wave velocity than the surrounding bedrock. Significant differences were also found in the records of the structure’s foundation and the free-field surface. Based on these results, the necessity of considering spatial variation in the observed records was suggested.