Wolsong Unit 1, a domestic heavy water reactor nuclear power plant, was permanently shut down in December 2019. Accordingly, Wolsong Unit 1 plans to prepare a Final Decommissioning Plan (FDP), submit it to the government by 2024, receive approval for decommissioning, and begin full-scale decommissioning. One of the important tasks in the decommissioning of Wolsong Unit 1 is to determine the decommissioning strategy. It is necessary to decide on a decommissioning strategy considering various factors and variables, secure the technical background, and justify it. The selection of a decommissioning strategy is best achieved through the use of formal decisionmaking assistance techniques, such as considerations related to influencing factors. It is very important to understand the basic decommissioning strategy alternatives and whether sufficient consideration has been given to situations where only a single unit is permanently shut down in a multi-unit site like Wolsong Unit 1, while the remaining units are in normal operation. As a process for selecting a decommissioning strategy, first, all considerations that could potentially affect decommissioning presented in the KINS Decommissioning Safety Review Guidelines were synthesized, influencing factors to be used in the decision-making process were determined, and the concept was defined. In order to select the most appropriate decommissioning strategy by considering various evaluation attributes of possible decommissioning alternatives (immediate dismantling and delayed dismantling), the Wolsong Unit 1 decommissioning strategy was evaluated by reflecting the AHP decision-making technique.

KHNP is conducting research to decommission Wolsong Unit 1 Calandria. Establishment of preparation and dismantlement processes, conceptual design of equipment and temporary radiation protection facilities, and waste management are being established. In particular, the ALARA plan is to be established by performing exposure dose evaluation for workers. This study aims to deal with the methodology of evaluating exposure dose based on the calandria dismantling process. The preparation process consists of bringing in and installing tooling and devices, and removing interference facilities to secure work space. The main source term for the preparation process is the calandria structure itself and crud of feeders. In the case of the dismantlement process, a structure with a shape that changes according to the process was modeled as a radiation source. It is intended to estimate the exposure dose by selecting the number of workers, time, and location required for each process in the radiation field evaluated according to the preparation and dismantlement process. In addition, it is also conducting an evaluation of the impact on dust generated by cutting operations and the human impact of C-14, H-3, which are specialized nuclides for heavy water reactors. KHNP is conducting an exposure dose evaluation based on a process based on the preparation and dismantlement process for decommissioning Calandria through computation code analysis. If additional worker protection measures are deemed necessary through dose evaluation according to this methodology, the process is improved to prepare for the dismantling of worker safety priorities.

For the release of the nuclear power plant site after the decommissioning, a reliable exposure dose assessment considering the environmental impact of residual radionuclides is essentially required. In this study, the Derived Concentration Guideline Level (DCGL) for the hypothetically contaminated surface soil at the Wolsong nuclear power plant (NPP) unit 1 site was preliminarily calculated by using the RESRAD-OFFSITE computational code and compared with the other case studies. Moreover, radiation exposure dose for local residents and relevant exposure pathways were quantitatively analyzed based on the calculation model established through this work. For the target site modeling, the source term was determined by referring to the previous case studies regarding the nuclear power plant decommissioning, quantification analysis data of pressure tubes of Wolsong NPP unit 1, and radionuclide data estimated by using the MCNP/ORIGEN-2 code. In total, 14 different radioisotopes such as Ag-108m, C-14, Co-60, Cs-134/137, Fe-55, H-3, Nb-93m/94, Ni-63, Sb-125, Sn-121m, Sr-90, and Zr-93 were considered as target radionuclides. In addition, the geological structure model of the Wolsong NPP site was established based on the final safety analysis report of Wolsong NPP unit 1. The distribution coefficients (Kd) were taken from the JAEA-SDB to estimate the migration/retardation behavior of various radionuclides under the groundwater condition of the Wolsong NPP site. In the present work, the DCGL values were calculated according to the site release criterion of 0.1 mSv/yr, which indicates the radiation protection standard for the site release. Moreover, the exposure pathway and sensitivity analyses were conducted to assess the sensitive input parameters remarkably influencing the calculation result. For the evaluation of exposure dose for local residents, a site layout centered around Wolsong NPP unit 4, located in the closest proximity to the residents’ habitation area, was alternatively established and all potential exposure pathways were considered as a comprehensive resident farmer scenario. The results obtained from this study are expected to serve as a preliminary case study for the DCGL values regarding the surface soil at the Wolsong NPP unit 1 site and for evaluating the radiation exposure dose to local residents resulting from the residual radioactivity at the site after the decommissioning.

We conducted safety assessments for the disposal of spent resin mixed waste after the removal of beta radionuclides (3H, 14C) in a landfill facility. The spent resin tank of Wolsong nuclear power plant is generated by 8:1:1 weight ratio of spent ion exchange resin, spent activated carbon and zeolite. Waste in the spent resin tank was classified as intermediate-level radioactive waste due to 14C. Other nuclides such as 60Co and 137Cs exhibit below the low-level radioactive waste criteria. The techniques for separating mixed waste and capturing 14C have been under development, with a particular focus on microwave-based methods to remove beta radionuclides (3H, 14C) from spent activated carbon and spent resin within the mixed waste. The spent resin and activated carbon within the waste mixture exhibits microwave reactivity, heated when exposed to microwaves. This technology serves as a means to remove beta isotopes within the spent resin, particularly by eliminating 14C, allowing it to meet the low-level radioactive waste criteria. Using this method, the waste mixture can meet disposal requirements through free water and 3H removal. These assessments considered the human intrusion scenarios and were carried out using the RESRAD-ONSITE code. The institutional management period after facility closure is set at 300 years, during which accidental exposures resulting from human intrusion into the disposal site are accounted for. The assessment of radiation exposure to intruders in a landfill facility included six human intrusion scenarios, such as the drilling scenario, road construction scenario, post-drilling scenario, and post-construction scenario. Among the six human intrusion scenarios considered, the most conservative assessment about annual radiation exposure was the post-drilling scenario. In this scenario, human intrusion occurs, followed by drilling and residence on the site after the institutional management period. We assumed that some of the vegetables and fruits grown in the area may originate from contaminated regions. Importantly, we confirmed that radiation doses resulting from post-institutional management period human intrusion scenarios remain below 0.1 mSv/y, thus complying with the annual dose limits for the public. This research underscores the importance of effectively managing and securing radioactive waste, with a specific focus on the safety of beta radionuclide-removed waste during long-term disposal, even in the face of potential human intrusion scenarios beyond the institutional management period.

A seal is one of the primary means of safeguards along with surveillance. The International Atomic Energy Agency (IAEA) uses various types of seals to verify the diversion of nuclear materials and is developing new seals according to the development of technology. Independent of the IAEA, ROK uses national safeguards seals for state-level regulation. A national safeguards inspector binds the nuclear material storage by combining a seal with a metal wire and checks the serial number of the RFID chip inserted in the seal with a reader. The Wolsong spent fuel dry storage facility has 14 modules, each with 24 seals, and thus a maximum of 336 national seals will be installed. Although dependent on the sealing method, it takes about 5 minutes to verify one seal. As such, a considerable workforce is required for verification, and both the IAEA and the ROK are currently conducting random inspections. In addition, there are cases where verification is impossible because old seals are damaged due to harsh environments and long exchange periods. Therefore, in this study, we analyzed cases in areas where sealing technology has been developed to improve the problems of the existing national safeguards seals. And we proposed a method for improving national seals by finding requirements of seals considering spent fuel dry storage facility characteristics. In international logistics, sealing is essential in product transport verification, terrorism prevention, and tariff imposition. Accordingly, the field of container sealing has been extensively developed, and the International Organization for Standardization (ISO) has regulated the mechanical requirements of the seal as ISO 17712 and the electronic requirements as ISO 18185. Mechanical seals include metal and plastic seals and metal seals include bolt seals, ball seals, and cable seals. In addition, there are various electronic seals, such as radio frequency identification (RFID), near field communication (NFC), infrared (IR). Recently, there has been a trend to use active seals that have a built-in battery and can implement various additional functions. Among the various seals, the main requirements for selecting seals suitable for dry storage facilities are as follows. First, use of a sealing tube longer than 10m should be possible. Second, it should have corrosion resistance so that it can be used for more than five years in the coastal area. Third, it must be a passive seal without a power supply. Fourth, it should not be overly costly. Finally, the seal verification time should be short. As a seal that satisfies these requirements, an electronic seal with application of the passive RFID method to the mechanical form of a metal cable seal is suitable. Since it is not an active seal, it is difficult to determine the time of breakage. Therefore, designing the seal such that the RFID is also damaged when the metal seal is broken will be helpful for verification. In this study, the requirements for national safeguards seals in dry storage facilities were defined, and measures to improve the existing national seals were studied. Field applicability will be evaluated through future sealing device design and demonstration tests.

Wolsong unit 1, the first PHWR (Pressurized Heavy Water Reactor) in Korea, was permanent shut down in 2019. In Korea, according to the Nuclear Safety Act, the FDP (Final Decommissioning Plan) must be submitted within 5 years of permanent shutdown. According to NSSC Notice, the types, volumes, and radioactivity of solid radioactive wastes should be included in FDP chapter 9, Radioactive Waste Management, Therefore, in this study, activation assessment and waste classification of the End shield, which is a major activation component, were conducted. MCNP and ORIGEN-S computer codes were used for the activation assessment of the End shield. Radioactive waste levels were classified according to the cooling period of 0 to 20 years in consideration of the actual start of decommissioning. The End shield consists of Lattice tube, Shielding ball, Sleeve insert, Calandria tube shielding sleeve, and Embedment Ring. Among the components composed for each fuel channel, the neutron flux was calculated for the components whose level was not predicted by preliminary activation assessment, by dividing them into three channel regions: central channel, inter channel, and outer channel. In the case of the shielding ball, the neutron flux was calculated in the area up to 10 cm close to the core and other parts to check the decrease in neutron flux with the distance from the core. The neutron flux calculations showed that the highest neutron flux was calculated at the Sleeve insert, the component closest to the fuel channel. It was found that the neutron flux decreased by about 1/10 to 1/20 as the distance from the core increased by 20 cm. The outer channel was found to have about 30% of the neutron flux of the center channel. It was found that no change in radioactive waste level due to decay occurred during the 0 to 20 years cooling period. In this study, activation assessment and waste classification of End Shield in Wolsong unit 1 was conducted. The results of this study can be used as a basis for the preparation of the FDP for the Wolsong unit 1.

Wolsong Unit 1 is about 679 MW Pressurized Heavy Water Reactor (PHWR). Canada AECL was responsible for Nuclear Steam Supply System (NSSS) design and supply. Wolsong Unit 1 was operated from 1983 to 2019. Currently, Wolsong Unit 1 is under safety management after permanent shutdown. Wolsong unit 1, a heavy water reactor, has the following characteristics. • Unlike Light Water Reactor, vertical reactors, Heavy Water Reactor is installed horizontally. • The internal structure of the reactor is more complex than that of a light water reactor (380 pressure tubes in reactor as called Calandria) • The Calandria Vault, a large concrete structure filled with light water, is located outside of Calandria In the case of the decommissioning plan of PHWR in Canada, they have adopted a deferred decommissioning strategy that decommissioning begins after permanent shutdown and long-term safety management (30 to 40 years). So, Decommissioning of PHWR in Canada is expected to start in the 2050s. Nuclear Safety Act stipulates that if a commercial nuclear power plant is permanently suspended, the utility must submit a Final Decommissioning Plan (FDP) within 5 years. So, KHNP, the utility, is developing the FDP for Wolsong Unit 1 and have a plan to submit it to the government by the end of 2024. And then licensing review is expected to take at least two years. The key milestone for decommissioning project has a plan to start decommissioning in 2027 and complete it by 2034, but this is flexible depending of the government’s approval for decommissioning and the completion of prerequisites such as spent fuel transfer, etc. KHNP has prepared a strategy and system consisting of three areas such as R&D, Engineering and licensing document development to prepare the final decommissioning plan for Wolsong Unit 1. The promotion system for the preparation of the FDP for Wolsong Unit 1 is consisted of Engineering (HAS Characterization, Process/Work Package/Cost Estimation, Dismantling Safety Evaluation, Radiological Environmental Report, Radioactive Waste Treatment and Facility Construction), R&D (COG cooperation, KHNP R&D Results), Kori unit 1 lessons learned, etc. KHNP have the plan that the FDP Draft development by the end of 2023, reflecting engineering services results, R&D results, COG technical cooperation results and lessoned learned on Kori Unit 1. After collecting opinions from residents through a public hearing, the FDP will be submitted to the government by the end of 2024. It is expected that there will be many difficulties in the development process as it is the world’s first FDP development for the commercial Pressurized Heavy Water Reactors.

On-site storage facility using concrete silo dry storage systems for spent nuclear fuel at Wolsong NPP site came into operation in 1992 and was expanded four times, and a total of 300 silo dry storage systems are currently in operation. The design lifetime of silo dry storage systems has been licensed for 50 years. As the dry storage systems are subject to time constraints for a limited lifetime, countries operating the dry storage systems are working to ensure the long-term integrity of dry storage systems and IAEA also recommends that the dry storage systems be assessed for long-term storage. To demonstrate the long-term integrity due to material degradation during the licensed design lifetime, the structural integrity of silo dry storage systems was evaluated by considering the material degradation characteristics of concrete. The concrete compressive strength results measured so far by the rebound hammer method, which is an internationally standardized nondestructive test method for converting hardness into compressive strength using the correlation between rebound number and strength at the time of a Schmidt hammer strike, were analyzed in accordance with Wolsong NPP’s procedure to quantify the degradation characteristics, and the prediction of concrete strengths for 20 years and 50 years after construction of the silo dry storage systems was determined, respectively. Based on these residual compressive strengths, structural analyses of the silo dry storage systems were carried out under normal, off-normal and accident conditions of the related regulations, and the structural integrity of silo dry storage systems was reevaluated. It was confirmed the silo dry storage systems are able to maintain structural integrity up to the design lifetime of 50 years even if the concrete is deteriorated.

Wolsong unit 1 (W1), which is a CANDU-6 type PHWRs that had been operated for 30 years since 1983, was shutdown in 2019. In this study, the radioactive waste levels of calandria and concrete structures were calculated to establish a decommissioning plan for W1. The specific systems within the scope of this study were grouped into 6 major categories as follows: Calandria, End Shield, Fuel Channel Assembly, Reactivity Control Device, End Shield Support, Vault. The main operating history of W1 is that the re-tubing project was performed. These characteristics and operation history were reflected in the evaluation. The neutron flux and energy spectrum of each structure were calculated by using MCNP code, and ORIGEN code is implemented to the calculation of radioactivity for each nuclide using the results from MCNP and the material information of the structure. As for the impurity information, ASTM B350, B351, B353 standard was used for zircaloy alloy. For other alloy, impurity information provided by NUREG/CR-3474 was applied. Since W1 is expected to be decommissioned immediately, the waste level was evaluated under cooling conditions for 5 years after permanent shutdown. Through the level evaluation of each component obtained as a result of the study, it can be used as basic data for the radioactive waste management of the decommissioning plan.

To obtain confidence in the safety of disposal facilities for radioactive waste, it is essential to quantitatively evaluate the performance of the waste disposal facilities by using safety assessment models. Thus, safety assessment models require uncertainty management as a key part of the confidencebuilding process. In application to the numerical modelling, the global sensitivity analysis is widely employed for dealing with parametric and conceptual uncertainties. In particular, the parametric uncertainty can be effectively reduced by minimizing the uncertainty of critical parameters in the safety assessment model. In this paper, the numerical model of each step disposal facility (Silo, Near surface, and Trench type) at Wolsong Low and Immediate Level Waste (LILW) Disposal Center is designed by using a two-dimensional finite element code (COMSOL Multiphysics). In order to determine the critical parameters for non-adsorbed nuclides such as H-3, C-14, Tc-99, we introduced the variance-based sensitivity analysis methodology of the global sensitivity analysis. In the case of Silo type, the density of waste is highly sensitive to the total leakage quantity of all nuclides. Additionally, the initial nuclide concentration of H-3 was identified as another important parameter of H-3. On the other hands, the mass transport coefficient showed a high contribution in C-14 and Tc-99. In other types of disposal facilities, the leaking properties of H-3 are significantly affected by the amount of infiltration water. However, C-14 and Tc-99 were found to be more sensitive to the density of waste.

Numerical model was developed that simulates radionuclide (3 H and 14C) transport modeling at the 2nd phase facility at the Wolsong LILW Disposal Center. Four scenarios were simulated with different assumptions about the integrity of the components of the barrier system. For the design case, the multi-barrier system was shown to be effective in diverting infiltration water around the vaults containing radioactive waste. Nevertheless, the volatile radionuclide 14C migrates outside the containment system and through the unsaturated zone, driven by gas diffusion. 3 H is largely contained within the vaults where it decays, with small amounts being flushed out in the liquid state. Various scenarios were examined in which the integrity of the cover barrier system or that of the concrete were compromised. In the absence of any engineered barriers, 3 H is washed out to the water table within the first 20 years. The release of 14C by gas diffusion is suppressed if percolation fluxes through the facility are high after a cover failure. However, the high fluxes lead to advective transport of 14C dissolved in the liquid state. The concrete container is an effective barrier, with approximately the same effectiveness as the cover.

원자력 사고 후 대기로 누출된 방사성물질이 지표 토양내 침적된 후 강우에 의하여 주변 환경으로 이동하여 지표수계를 오염시킨다. 지표 토양내 침적된 방사성핵종의 거동 평가를 위하여 수립된 지표 수계 및 토양 유실 모델의 주요 입력자료를 수 집하여 분석하였다. 월성 원전이 위치한 낙동강권역의 하천과 호수에서의 물리적 특성과 주요 생물상의 변화를 파악하기 위해서 원전 주변 수생 환경의 조사 및 분석을 수행하였다. 이를 위해 국내 여러 기관에서 제공하는 수치지도, 수문자료, 수질 및 생태환경자료 등을 수집 분석하여 자료간 상호 연계성을 갖도록 체계적인 DB를 구축하였다. 구축된 수생환경 자료는 지표수계에 흡착된 방사성물질의 중장기 거동 평가를 위하여 수립된 지표수계 유동, 토사유실 및 생태계 모델의 기본 입력자 료로 제공되어 종합적인 방사선영향평가에 활용될 예정이다.