In accordance with the notification of the Nuclear Safety and Security Commission (NSSC), environmental impact assessments around nuclear power plants are conducted annually and the results are disclosed to the public. KHNP evaluates the dose of residents around nuclear power plants using the K-DOSE60 program that reflects ICRP-60. K-DOSE60 calculates the expected exposure dose for residents by modifying the atmospheric dispersion and deposition factors evaluation module (XOQDOQ), gaseous effluent evaluation module (GASDOS) and liquid effluent evaluation module (LIQDOS) developed by the US NRC. The current evaluation program is the Bounding Assessments method, which evaluates under the assumption that residents reside at the exclusion area boundary (EAB), and has a disadvantage in that the estimated exposure dose is evaluated too conservatively. In the EPRI, instead of the conservative method that is conventionally performed for the residents’ dose evaluation method, a plan to improve the accuracy of the dose evaluation reflecting the site characteristics was reviewed. In addition, improvements were derived through the review of NPPs operation status, experience cases and the latest technology.

There are many Systems, Structures, and Components (SSCs) in Nuclear Power Plants (NPPs). The systems include radiological waste treatment system, spent fuel pool cooling, emergency core cooling systems, etc. The structures include reactor building, piping vaults, radioactive waste storage facilities, etc. The components include valves, pumps, piping segments, etc. Radionuclides exist in some of these SSCs and unplanned release may occur when leaks or spills from them. And also Work Practice (WP) is another reason of unplanned release in NPPs. The WP is defined as an action taken by individuals during maintenance, operational or support activities, which could result in or prevent a spill or leak of a radioactive solid, liquid or gas that has a credible mechanism for contamination of groundwater. According to the results of the Electric Power Research Institute (EPRI) survey, a total of 323 unplanned release event occurred at US NPPs from 1970 to 2014. Among them, 219 events were counted to have occurred at pressurized water reactors (PWRs). In addition, it was confirmed that 41 of the 44 PWR sites (about 93%) in the US, operated at the time of the survey period, had experienced at least one unplanned release events of licensed material which impacted groundwater. This means that the US PWR sites have experienced an average of approximately 5 unplanned release event per site. The source with the most unplanned releases, including SSCs and WP, was miscellaneous systems with a percentage of about 33% (72 events). Miscellaneous systems include pipes, and it was confirmed that unplanned releases mainly occurred in pipes such as the main steam system, condensate and feedwater system, and emergency core cooling system. And the percentage was high in the order of WPs (21%, 45 events), radioactive effluents (20%, 43 events), refueling water storage (8%, 17 events), radioactive waste/material operations (7%, 16 events), spent fuel storage (5%, 12 events), unknown (4%, 9 events), and structures (2%, 5 events). The history of the unplanned release of the US NPPs will be considered when revising major SSCs in the domestic NPP groundwater monitoring program.

In order to monitor the contamination of groundwater due to unplanned release of radioactive materials and the spread to off-site environments, the nuclear power plants (NPPs) conduct groundwater monitoring program (GWMP) in Korea. The GWMP should be established based on the groundwater flow model reflecting the conceptual site model (CSM) of the NPP’s site. In this study, in order to optimize the GWMP, the existing CSM and the groundwater flow model of the domestic NPPs site was updated by reflecting the latest groundwater level. As part of the CSM improvement, the hydrogeological units were subdivided more detailed from three to six through the review of hydrogeological characteristics of the NPPs site. In addition, major variables that affect groundwater flow, such as water conductivity, have been updated. The groundwater flow model was revised overall as the CSM was improved. In particular, the excavation depth of the structure and backfill area generated during the construction stage of the NPP structures was accurately reflected, and the drainage boundary conditions were realistically reflected. To verify the revised groundwater flow model, steady-state correction was performed using the groundwater level measured in April, 2021. As a results of the steady-state correction, the standard error of estimate, root mean square (RMS), normalized RMS, and the correlation coefficient were 0.32 m, 1.692 m, 5.608%, and 0.964, respectively. This means that the groundwater flow model is reasonably constructed. The CSM and groundwater flow model improved in this study will be used to optimize the monitoring location of groundwater in NPPs.

In Korea, Starting with the permanent shutdown of Kori Unit1, decommissioning of commercial nuclear power plant is underway. Although various technologies are required to decommissioning a nuclear power plant, the most important technology is the characterization of radioactive waste. In particular, it is possible to establish an accurate decommissioning plan and estimate cost for radioactive waste through accurate characterization of reactor vessel (RV), reactor vessel internal (RVI) and bioshield, which are highly activated waste. In Slovakia’s V1 nuclear power plant, two units were shutdown in 2006 and 2008, respectively, and decommissioning license was approved in July 2011. Before approving the decommissioning license, the decommissioning database project was carried out from 2008 to 2011. At this time, radioactive evaluation was performed through sampling and radiological analysis of radioactive structures.

During the decommissioning of nuclear power plant (NPP), massive amount of concrete wastes is generated, which are non-radioactive and radioactive. The concrete is a representative construction material which affords reliable structural stability, good formability, and trustful integrity. Also, its reasonable neutron absorbing property allows the various application for many components, including building construction material, bio-shield concrete, etc. Due to the noted aspects of concrete, the radiological concrete characterization is classified as an important process for development of effective strategy for concrete management, in terms of process management and financial control during the decommissioning. The characterization of bio-shield concrete is important in waste management. The understanding and characterization of activation depth is essential for the determination of waste management strategy, classification of bio-shield concrete, and process development of decommissioning. On the other hand, concrete for construction application requires the evaluation of surface contamination of them. The concrete for containment building and its structure is rarely activated, but surface contaminated. In this paper, the reactor data from representative PWR reactors in the US is studied. The experience on Yankee Rowe, Maine Yankee, and Connecticut Yankee NPPs are systematically studied. The Yankee Rowe was a 4-loop PWR of Westinghouse design with 185 MWe. The Main Yankee was a 3- loop PWR of Combustion Engineering design with 864 MWe. The Connecticut Yankee was a 4-loop Westinghouse type with 560 MWe. The characterization studies on bio-shield concrete will be discussed in this paper, including activation depth, primary nuclides, etc.

The Fukushima nuclear power plant accident, which was caused by the Great East Japan Earthquake on March 11, 2011, is of great concern to the Korean people. The scope of interest is wide and diverse, from the nuclear accident itself and the damage situation, to the current situation in Fukushima Prefecture and Japan, and to the safety of Japanese agricultural and fishery products. Concerns about nuclear safety following the Fukushima nuclear accident have a significant impact on neighboring nation’s energy policy. It has been 11 years since the Fukushima nuclear accident. In neighboring nation society, the nature and extent of damage caused by the Fukushima nuclear accident, the feasibility of follow-up measures at home and abroad, the impact on neighboring nations, and the direction of nuclear policy reflecting the lessons of the accident are hotly debated topics. Recently, the controversy has grown further as it is intertwined with Japan’s concerns about the safety and discharge of the contaminated water into the sea, and conflicts over domestic nuclear power policies. About 1.29 million tons, as of March 24, 2022, of the contaminated water are generated, which is close to the 1.37 million tons of water storage capacity. In response, the Japanese government announced on April 13, 2021, that it plans to discharge the contaminated water into the sea from 2023. This study evaluates the amount of the contaminated water that has passed through the ALPS and reviews the preparations and related facilities for ocean discharge after diluting the contaminated water. In addition, it is intended to forecast the various impacts of ocean discharge.

Numerous nuclear power plants that had been built in the late 20th century have entered the aging phase and are scheduled to be decommissioned. The decommissioning project of a commercial nuclear power plant is an array of complex processes involving the activities of site characterization, decontamination, dismantling, and site restoration. Hence, a number of essential factors, such as scheduling, work progress, and staffing, should be taken into account while the decommissioning plan is drafted and modified. Guidances on managerial and social aspects of decommissioning have been rare as compared to those of technical viewpoints. Nonetheless, the nuclear industry in the US has presented no little amount of experience on their decommissioning projects dealing with those perspectives. Thus, three sets of the case study were conducted to obtain useful lessons learned. The Maine Yankee nuclear power plant initially acquired 40 years of the operating license, it was in operation for only 25 years from 1972 until 1996. The owner group decided to shut down because of the deterioration of the profitability in 1997. The case of the Maine Yankee project enlightened the importance of the contract management and stakeholder relations. The Rancho Seco nuclear power plant is a single-unit nuclear power reactor site with 913 MWe output that commenced commercial operation in 1975. The Rancho Seco that had become the first-ever reactor shut down by a public voting introduced several innovative approaches for the decommissioning, some of which turned out to be very successful. The SONGS 1 commenced the commercial operation in 1968 and had been decided to cease its operation permanently due to a steep decline in profitability in 1992. The SONGS 1 presented worthwhile lessons in terms of project management. In this study, several lessons learned related on managerial, engineering, and regulatory/social aspects considered during the NPP decommissioning will be reviewed and discussed.

HIC refers to a radwaste packaging container that can maintain integrity for more than 300 years in the general underground environment and disposal conditions in Korea. For HIC, the integrity of containers is verified according to the HIC regulation guideline for LLW and ILW disposal. Existing material tests include mechanical stability, permeability resistance, corrosion properties, chemical durability and biological resistance. In this study, a chemical durability test was conducted to prove the suitability of the HIC material by measuring the degree of chemical influence other than corrosion from the disposal environment. The chemical resistance evaluation method was used to simulate the disposal environment in the underground repository, and the amount of change in the physical properties of the degraded polymer concrete specimens according to the test time was confirmed. The technical standards considered leaching of material components, sulfation attack, acid attack, alkali, carbonate, and salt crystallization. The compressive strength and weight change of the specimens with time were checked by immersing them in a chemical solution that could leak major hazardous substances and wastes in the groundwater of the repository for several months. In addition, in order to evaluate the integrity in condition severe than the disposal environment, a flow was applied to a chemical solution having a concentration twice that of the basic chemical resistance test conditions, and the test period was extended twice to accelerate the deterioration of the specimen.

The Korea Nuclear Safety Act defines a high integrity container (HIC) as “a radioactive waste packaging container that can maintain its integrity for more than 300 years under the general underground environment and disposal conditions in Korea”, and detailed technical standards are not described. The US Nuclear Safety Commission’s “Low-Level Waste Licensing Branch Technical Position on Waste Form” describes the detailed requirements for solidification and HIC. The main contents of the US NRC technical position include limiting the free standing water, minimum design life, demonstrating mechanical, thermal and radiation stability, etc. In this study, the stability evaluation was performed to understand the mechanical strength with respect to horizontal and vertical loads. The basic property of polymer concrete was carefully evaluated, including compressive strength, structural fatigue resistance, etc. The long term creep test, loading of 40% of compressive strength, indicates that the polymer concrete exhibits good long term mechanical integrity.

The off-site dose calculation is regularly carried out at the nuclear power plants in order to evaluate off-site dose from gaseous and liquid effluent during normal operation. In 2009, the off-site calculation program (K-DOSE60) was developed in accordance with ICRP-60 by KHNP. This software needs meteorological data, gaseous and liquid effluent data, and various other input parameters to evaluate off-site dose. As a result, it takes a certain amount of time for the user to enter accurate input data and verify calculated results, and it is difficult to intuitively determine them because of providing textbased calculated results. Therefore, in this study, the improvement of the calculation program was considered so that a more reliable and effective evaluation could be performed when calculating the off-site dose. The main improvements of the off-site dose calculation program (ODCP) are as follows. First, it is developed as the network-based program to link with meteorological data, and gaseous and liquid effluent data to remove input errors and simplify data transfer. Second, through validation process of input data, input errors are eliminated. Third, the input data and calculated results are visually provided so that the user can easily determine the evaluation results. Fourth, database of input and calculated results is constructed to facilitate evaluation result history management.

After the Fukushima accident in 2011, a huge amount of radioactively contaminated water is being generated by cooling the melted fuel of units 1, 2 and 3. Most of contaminated water is seawater and underwater containing not only salt elements but also nuclear fission products with radioactivity. To treat the contaminated water, Cs/Sr removal facilities such as KURION and SARRY are being operated by TEPCO. Additionally, three ALPS facilities are on operation to meet the regularity standards for discharge to the sea. However, massive secondary wastes such as Zeolite, sludge and adsorbent is being generated by these facilities for liquid water treatment. The secondary wastes containing various radionuclide with Cs and Sr is difficult to store due to highly radioactive concentration and corrosive properties. In Japan, a variety of technologies such as GeoMelt vitrification, In-Can vitrification and CCIM vitrification is considered as a promising solution. In this study, they were reviewed, and the advantage and disadvantage of each technology were evaluated as the candidate technologies for thermal treatment of sludge radwaste.

In KHNP CRI, the 100 kW PTM (plasma torch melting) system was designed for the treatment and disposal technology of various radioactive wastes including the metal, concrete, liquid waste and insulator. The facility consists of melting chamber, thermal decomposition chamber, waste feeding system and off-gas treatment system. In this study, to evaluate the applicability of the PTM system, demonstration test was conducted using the radiation hazmat suit as combustible waste. The plasma melting chamber is pre-heated by 2nd combustion device and plasma torch for 5 hours. The temperature inside the plasma melting chamber is approximately 1,600°C. The combustible waste was put into the melting chamber by the pusher feeding device with the throughput of maximum 50 kg/hour. During the test, the power of plasma torch is 60–96 kW on the transferred mod. It was evaluated in terms of long-term integrity of PTM system on operation according to the waste throughput ratio.

Plasma melting technology has been considered as promising technology for treatment of radioactive wastes. According to the IAEA TECDOC-1527 report (2006), the technology has an advantage that it can treat regardless of waste types which is both combustible and non-combustible wastes. In particular, it is expected that a large amount of concrete, a representative non-combustible wastes, will be generated during the operation and dismantling of nuclear power plants. In order to treat the concrete waste in plasma torch melting system, various factors could be considered like the slag of electric conductivity, viscosity and melting temperature. Above all, as a critical factor, the viscosity of the melt is very important to easily discharge the melt. The viscosity of slag (SiO2-CaO-Al2O3 system) can be lowered by adding a basic oxide such as CaO, Na2O, MgO and MnO. The basic oxides are donors of oxygen ions. These oxides are called notwork breakers, because they destroy the network of SiO2 by reacting with it. In this study, the slag composition of the concrete waste was developed to apply the plasma torch melting. Also, demonstration test was performed with the developed slag composition and 100 kW plasma torch melting system.