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. The effects of direct radiation exposure from nuclear power plants as well as liquid effluents and gaseous effluents are taken into consideration in the evaluation of dose calculation for residents. In the United States, regulatory guidelines on direct radiation exposure are described in Reg. Guide 4.1, and the effects of direct radiation are evaluated through regulatory guidelines in Korea. We are going to review optimal evaluation method by reviewing the direct radiation exposure evaluation method currently being conducted in domestic nuclear power plants and the direct radiation exposure evaluation method in overseas nuclear power plants such as in the United States.

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.

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.

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 2022, new regulatory guidelines were announced in relation to the off-site dose calculation (ODC), and accordingly, measures to improve the off-site does calculation program (ODCP), kdose60, were reviewed. The main consideration is, first, that if multiple nuclear facilities are operated on the same site, the boundaries of the restricted areas shall be set as the overlapping outer boundaries of the restricted areas determined by calculation for each nuclear facility. Second, the external exposure caused by direct radiation from a number of nuclear facilities in the same site must be partially or fully applied depending on the facility and site characteristics. Third, the dose conversion coefficient should be evaluated by checking whether the effect of the daughter nuclides is properly reflected. Fourth, the soil contamination period is a factor to consider that radioactive substances deposited on the surface, such as particulate nuclides, affect residents over a long period of time. Fifth, due to the recent construction of Shin-Kori Units 5 and 6, there is a change in the site boundary of the Kori/Saeul site, so as the site boundary is expanded, it is required to add an exposure dose assessment point due to gas effluents and change the exposure dose assessment point according to crop intake. Therefore, through this study, the direction for improving the ODCP will be prepared by reviewing the recent revision of the regulatory guidelines.

Recently, Japan’s government has announced Tokyo Electric Power Company’s plan to discharge contaminated water stored from the tanks of the Fukushima Daiichi nuclear power plant site into the sea. The contaminated water is treated by advanced liquid processing system (ALPS) to remove 62 radionuclide containing cesium, strontium, iodine and etc. using co-precipitation (or precipitation) and adsorption for other nuclides (except for tritium and carbon-14). The total amount of the contaminated water generated by ALPS facility is 1,311,736 m3 (as of August 18, 2022). The amount of contaminated water is estimated same as Tokyo dome volume. Under the sea discharge plan, the contaminated water will be diluted in seawater more than 100 times, and tritium concentration lowered 1/7 of the drinking water standard set by the World Health Organization (10,000 Bq/liters). The diluted water will then move through an undersea tunnel and be discharged about 1 kilometer off the coast.

For the disposal of radioactive waste generated from nuclear power plants, characterization of radioactive waste is essential. For characterization, samples of radioactive waste are directly collected or an indirect method is used through X-ray, etc. Through indirect analysis, which is a non-destructive method, the density, filling height, homogeneity and inter structure of the waste container can be analyzed. Currently, foreign institutions are in the process of developing a technology to perform characterization of radioactive waste through indirect analysis. In particular, research on improving internal image accuracy through image analysis techniques, improving measurement methods and enhancing portability for field application is ongoing. Through the review of such technology development trends, it will be utilized in the development of domestic radioactive waste disposal technolgy.

In 2005, groundwater contamination due to unplanned releases of radioactive materials from the US. Nuclear Power Plants (NPPs) such as Braidwood and Indian Point was confirmed. The following year, in 2006, The Nuclear Regulatory Commission (NRC) established a task force team to investigate the history of unplanned release of all NPP in the US. As a results 217 events of unplanned release including leaks and spills were identified in the US NPPs. The NRC regulates the radioactivity concentration of off-site groundwater by setting a reporting levels (RLs), and if exceeds the RLs, the licensee must report within 30 days. When the off-site groundwater is used as drinking water or non-drinking water, the RLs for tritium in groundwater are 740 Bq·L−1 or 1,110 Bq·L−1, respectively. Whereas the NRC does not set the RLs for on-site groundwater. The Nuclear Energy Institute (NEI) issued the guidance document “Industry groundwater protection initiative” NEI 07-07 in 2007. And the members of the NEI promised with regulatory body and local governments to implement groundwater monitoring/protection program according to the NEI 07-07. The document states that when the on-site groundwater is used as drinking water, the RL (740 Bq·L−1) for off-site groundwater will be applied and the licensee voluntarily reported to the NRC. And also, NPPs are setting the Investigation Level (IL) below the RP and the IL is various among the NPPs. The IL is the standard by which detailed investigations are implemented when the level (radioactivity concentration) is exceeded.

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.

The decommissioning of nuclear power plant (NPP) generates large amount of waste. Since the most of the concretes are slightly surface contaminated, the accurate characterization and regionspecific surface decontamination are important for the efficient waste management. After the effective surface decontamination and separation, most of the concrete waste from decommissioning of NPP can be classified as a clearance waste. Various surface characterization and decontamination technologies are suggested. The mechanical technologies are simple and offers direct application. The laser-based technologies offer efficient separation and surface contamination. The high price, however, hesitates the application of the process. The nitro-jet technology, which is based on the evaporation of liquid nitrogen, allows the effective decontamination. However, the high price and uncertainty of large are application hinders the practical application in NPP decommissioning. In this paper, various technologies for characterization, handling, treatment, etc., will be discussed. The advantages and disadvantages of the technologies will be discussed, in terms of practical applications.

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.

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.