It is necessary to prepare for cutting and storing waste materials in the reactor vessel internals (RVI) for successful decommissioning of the nuclear power plant (NPP). Since RVI contain massive components and relatively highly activated, their decommissioning process should be conducted carefully in terms of radiological and industrial safety. To achieve efficient decommissioning waste management, this study presents radiation level of RVI and cutting optimization was performed for intermediate level waste. As a result of the radiation evaluation, a part of the core side and the upper part of RVI were evaluated as intermediate-level waste, and other components were evaluated as very low-level or lowlevel waste. For intermediate-level waste cutting, the minimum cutting method that can be put into a container was reviewed in consideration of the size, thickness, and cutting method of the interior product. The final segmentation parts are expected to be loaded into two storage containers.

This paper mainly focuses on the maximum decay heat estimation generated from spent fuel assemblies in the spent fuel pool of Kori units 3&4 at the beginning decommissioning. It is assumed that the spent fuel pool is fully occupied with 2,260 spent fuel assemblies, same as its design capacity. In addition, equally 56.5 spent fuel assemblies have been generated per year. The minimum cooling time is five years considering the transition phase between the permanent shutdown and the amendment of Operating License for decommissioning. Sending and receiving of spent fuel assemblies to/from other units are neglected. Seven representative spent fuel assembly groups are established based on the burnup rate and cooling time. Conservatively high values for the burnup rates and low values for the cooling times are applied. Calculation of the decay heat of each representative group has been performed by using ORIGEN decay solver of SCALE. Then, total decay heat has been calculated based on this. Group 1, 2, and 3 contain comparatively old spent fuel assemblies with 45 GWd/tU burnup rate and 20~30 cooling years. The calculation shows 489~586 watts of decay heat per assembly. Group 4, 5, 6, and 7 contain comparatively new spent fuel assemblies with 55 GWd/tU burnup rate and 5~20 cooling years. The calculation shows 741~1,483 watts of decay heat per assembly. The total maximum decay heat therefore is estimated as 1,609,459 watts.

The spent filters stored in Kori Unit 1 are planned that compressed and disposed for volume reduction. However, shielding reinforcement is required to package high-dose spent filters in a 200 L drum. So, in this study suggests a shielding thickness that can satisfy the surface dose criteria of 10 mSv·h−1 when packaging several compressed spent filters into 200 L drums, and the number of drums required for the compressed spent filter packaging was calculated. In this study, representative gamma-emitting nuclides in spent filter are assumed that Co-60 and Cs-137, and dose reduction due to half-life is not considered, because the date of occurrence and nuclide information of the stored spent filter are not accurate. The shielding material is assumed to be concrete, and the thickness of the shielding is assumed to 18 cm considering the diameter of the spent filter and compression mold. Considering the height of the compressed spent filter and the internal height of the shielding drum, assuming the placement of the compressed spent filter in the drum in the vertical direction only, the maximum number of packaging of the compressed spent filter is 3. When applying a 18 cm thick concrete shield, the maximum dose of the spent filter can packaged in the drum is 125 mSv·h−1, so when packaging 3 spent filters of the same dose, the dose of a spent filter shall not exceed 41 mSv·h−1 and not exceed 62 mSv·h−1when packing 2 spent filters. Therefore, the dose ranges of spent filters that can be packaged in a drum are classified into three groups: 0–41 mSv·h−1, 41–62 mSv·h−1, and 62–125 mSv·h−1based on 41 mSv·h−1, 62 mSv·h−1, and 125 mSv·h−1. When 227 spent filters stored in the filter room are classified according to the above dose group, 207, 3 and 4 spent filters are distributed in each group, and the number of shielding drums required to pack the appropriate number of spent filters in each dose group is 75. Meanwhile, 8 spent filters exceeding 125 mSv·h−1 and 5 spent filters that has without dose information are excluded from compression and packaging until the treatment and disposal method are prepared. In the future, we will segmentation of waste filter dose groups through the consideration of dose reduction and horizontal placement of compressed spent filters, and derive the minimum number of drums required for compressed spent filter packaging.

This study is for evaluation and optimization of workers’ radiation exposure for dismantling Reactor Vessel (RV) at Kori unit 1 in connection with its decommissioning process for the purpose of establishing Radiation Safety Management Plan. This is because the safety of workers in a radiation environment is an important issue. The basis of radiological conditions of this evaluation is supposed to be those of 10 years after the permanent shutdown of Kori unit 1 when dismantling work of Reactor Vessel would suppose to be started. Dose rates of work areas were evaluated on the basis of spatial dose rate derived from activation level calculated by MCNP (Monte Carlo N-Particle Transport) and ORIGEN-S code. RV are radiated by neutrons during operation, creating an environment in which it is difficult for operators to access and work. Therefore, the RV must be dismantled remotely. However, due to work such as installing devices or dismantling surrounding structures, it is not possible to completely block the access of workers. Accordingly, the exposure of workers to the RV dismantling process should be assessed and safety management carried out. The dismantling process of Kori unit 1 RV was developed based on in-situ execution in atmospheric environment using the oxigen-propane cutting technology as the following steps of Preparation, Dismantling of Peripheral Structures, Dismantling of RV and Finishing Work. For evaluation of exposure of RV dismantling work, those processes of each steps are correlated with spatial dose rates of each work areas where the jobs being done. Results of the evaluation show that workers’ collective dose for RV dismantling work would be in the range of 536–778 man- mSv. The most critical process would be dismantling of upper connecting parts of RV with 170–256 manmSv while among the working groups, the expert group performing dismantling of ICI (In-core instrumentation) nozzles and handling & packaging of cut-off pieces is evaluated as the most significantly affected group with 37.5–39.4 man- mSv. Based on the evaluation, improvement plan for better working conditions of the most critical process and the most affected workers in terms of radiation safety were suggested.

Radioactive waste generated during the decommissioning of Kori Unit 1 can be packaged in a transport container under development and transported to a disposal facility by sea transport or land transport. In this study, the cost of each transport method was evaluated by considering the methods of land transport, sea transport, and parallel transport of the radioactive waste dismantled at Kori Unit 1. In evaluating the shipping cost, the shipping cost was evaluated by assuming the construction of a new ship without considering shipping by CHEONG JEONG NURI, which is currently carrying operational waste. Since the cargo hold of CHEONG JEONG NURI was built to fit the existing operating waste transport container and is not suitable for transporting the transport container currently under development, sea transport using CHEONG JEONG NURI was excluded in this paper. In the case of on-road transportation, the final fare for each distance was calculated in accordance with the Enforcement Decree of the Freight Vehicle Transportation Business Act, and the cost of onroad transportation was evaluated by estimating the labor cost of the input manpower required for onroad transportation. The cost of on-road transportation was estimated to be approximately KRW 510 million, the product of the total number of transports 459 times the sum of the cost of transportation vehicle freight cost of about KRW 720,000 and the labor cost of input personnel of KRW 380,000. It is difficult to predict the cost of building a new ship at this point, as the cost of building new ship is determined by the cost of number of items such as ship design specifications and material prices, labor costs, and finance costs at the time of construction. Accordingly, considering the 2% annual inflation rate based on the shipbuilding cost (about KRW 26 billion) and financing cost (about KRW 12 billion) at the time of construction of the CHEONG JEONG NURI (2005 yr.), decommissioning of Kori Unit 1 (2025 yr.) construction cost finance cost was estimated and evaluated. According to the result of comparing the transport cost for each transport scenario, land transport is about 510 million won, which is advantageous in terms of economic feasibility compared to the sea transport scenario. However, when transporting by land, it is disadvantageous in terms of acceptability of residents because it is transported multiple times on general roads. The cost of building a new ship is about KRW 56.4 billion, which is disadvantageous in terms of the cost of transporting waste from the dismantling of Kori Unit 1. But, in the future, cost reduction can be expected if waste materials issued when dismantling nuclear power plants are transported.

An application of the final decommissioning plan for unit 1 of Kori NPP was submitted to NSSC on 14 May 2021. We have been implementing the project related to the radiological characterization for the plan since 2019. However, the project was not running smoothly due to the regulatory environment. The destructive sampling from the objects was not allowed, so only smear (swipe) samples are available. In this study, the sampling way and the analytical results of radionuclides are presented. In addition, we propose in-situ measurement using gamma camera and in-situ gamma spectroscopy to obtain more comprehensive radiological information on the object.

The permanent shut-down of Kori unit 1, the first nuclear power plant (NPP) in Korea, generates various radioactive waste. They are dry active waste (DAW), spent resin, concentrated waste, activated metals, etc. During normal operation of NPP, activated metals are rarely generated. The decommissioning of NPP, however, generates massive amount of metallic waste including activated metals and contaminated metals, while normal operation generates small amount of metallic waste. The reactor vessel and internals are relatively highly activated components in NPP. Since they are exposed to the high concentration of neutrons during the operation, their contains relatively high radioactive nuclides. They activation analysis is usually performed to understand the radiological inventory of the activated reactor vessel and internal. The results offer various important information including, radiological inventory, waste classification, etc. The impurities in the carbon steel and stainless steel have a great impact on the radiological inventory of the activated metals. The cobalt, nickel, niobium are primary elements that affects the activation analysis and waste classification. Especially, the cobalt, which transforms to 60Co, plays an important role. The 60Co, strong gamma emitter, affects the waste classification, safety analysis of decommissioning workers, and determination of segmentation and package plans. In this paper, effects of impurity concentration on activation analysis is studied. The expected impurity from various sources, including NUREG/CR-3474, commercial NPP data, etc, and effects will be demonstrated. Also, the comparison between results and international experiences will be followed.

This study evaluates the radioactivity of concrete waste that occurs due to large amounts of decommissioned nuclear wastes and then determines the surface dose rate when the waste is packaged in a disposal container. The radiation assessment was conducted under the presumption that impurities included in the bio-shielded concrete contain the highest amount of radioactivity among all the concrete wastes. Neutron flux was applied using the simplified model approach in a sample containing the most Co and Eu impurities, and a maximum of 9.8×104 Bq·g−1 60Co and 2.63×105 Bq·g−1 152Eu was determined. Subsequently, the surface dose rate of the container was measured assuming that the bio-shield concrete waste would be packaged in a newly developed disposal container. Results showed that most of the concrete wastes with a depth of 20 cm or higher from the concrete surface was found to have less than 1.8 mSv·hr−1 in the surface dose of the new-type disposal container. Hence, when bio-shielded concrete wastes, having the highest radioactivity, is disposed in the new disposal container, it satisfies the limit of the surface dose rate (i.e., 2 mSv·hr−1) as per global standards.

A paradigm shift in the government’s energy policy was reflected in its declaration of early closure of old nuclear plants as well as cancellation of plans for the construction of new plants. To this end, unit 1 of Kori Nuclear Power Plant was permanently shut down and is set for decommission. Based on these changes, the off-site transport of spent fuels from nuclear power plants has become a critical issue. The purpose of this study is to develop an optimized method for transportation of spent fuels from Kori Nuclear Power Plant’s units 1, 2, 3, and 4 to an assumed interim storage facility by simulating the scenarios using the Flexsim software, which is widely used in logistics and manufacturing applications. The results of the simulation suggest that the optimized transport methods may contribute to the development of delivery schedule of spent fuels in the near future. Furthermore, these methods can be applied to decommissioning plan of nuclear power plants.

Chemical decontamination of primary systems in a nuclear power plant (NPP) prior to commencing the main decommissioning activities is required to reduce radiation exposure during its process. The entire process is repeated until the desired decontamination factor is obtained. To achieve improved decontamination factors over a shorter time with fewer cycles, the appropriate flow characteristics are required. In addition, to prepare an operating procedure that is adaptable to various conditions and situations, the transient analysis results would be required for operator action and system impact assessment. In this study, the flow characteristics in the steady-state and transient conditions for the chemical decontamination operations of the Kori-1 NPP were analyzed and compared via the MARS-KS code simulation. Loss of residual heat removal (RHR) and steam generator tube rupture (SGTR) simulations were conducted for the postulated abnormal events. Loss of RHR results showed the reactor coolant system (RCS) temperature increase, which can damage the reactor coolant pump (RCP)s by its cavitation. The SGTR results indicated a void formation in the RCS interior by the decrease in pressurizer (PZR) pressure, which can cause surface exposure and tripping of the RCPs unless proper actions are taken before the required pressure limit is achieved.

Currently, radioactive waste for disposal has been restricted to low and intermediate level radioactive waste generated during operation of nuclear power plants, and these radioactive wastes were managed and disposed of the 200 L and 320 L of steel drums. However, it is expected that it will be difficult to manage a large amount of decommissioning waste of the Kori unit 1 with the existing drums and transportation containers. Accordingly, the KORAD is currently developing various and largesized containers for packaging, transportation, and disposal of decommissioning waste. In this study, the radiation exposure doses of workers and the public were evaluated using RADTRAN computational analysis code in case of the domestic onroad transportation of new package and transportation containers under development. The results were compared with the domestic annual dose limit. In addition, the sensitivity of the expected exposure dose according to the change in the leakage rate of radionuclides in the waste packaging was evaluated. As a result of the evaluation, it was confirmed that the exposure dose under normal and accident condition was less than the domestic annual exposure dose limit. However, in the case of a number of loading and unloading operations, working systems should be prepared to reduce the exposure of workers.