Recently, the nuclear decommissioning and environmental restoration industries has significantly attracted as a new industry field due to the decision to decommission the KORI#1 and WOLSONG #1 nuclear power plant. In order to dispose of the decommissioning radioactive wastes generated during nuclear decommissioning, proper analysis is required, and disposal decisions are determined based on the analysis results. When dismantling a nuclear power plant, a few thousand of tons decommissioning waste are produced, so these require analysis for proper disposal. Therefore, a radionuclide facility for decommissioning waste analysis is essential for the disposal of the large quantities of decommissioning waste generated during nuclear power plant decommissioning. Korea Research Institute of Decommissioning (KRID) was established radionuclide analysis facilities to address above issues and support nuclear power plant decommissioning projects. The plan is to perform classification by type and radionuclide for all waste produced during nuclear power plant decommissioning and to support the disposal of radioactive wastes. In addition, we plan to establish validation methods for samples where verification methods are not established, in order to conduct efficient analysis and management. In this presentation, we will introduce the radionuclide facility currently under construction at KRID and present the space design, equipment layout, and utilization plans.

Nuclear power generation is expected to be enlarged for domestic electricity supply based on the 10th Basic Plan of Long-Term Electricity Supply and Demand. However, the issues on the disposal of spent nuclear fuel or high-level radioactive waste has not been solved. KBS-3 concept of the deep geological disposal and pyroprocessing has been investigated as options for disposal and treatment way of spent nuclear fuel. In other way, the radionuclide management process with 6 scenarios are devised combining chlorination treatment and alternative disposal methods for the efficient disposal of spent nuclear fuel. Various scenarios will be considered and comprehensively optimized by evaluation on many aspects, such as waste quantity, radiotoxicity, economy and so on. Level 0 to 4 were identified with the specialized nuclide groups: Level 0 (NFBC, Hull), Level 1 (Long-lived, volatile nuclides), Level 2 (High heat emitting nuclides), Level 3 (TRU/RE), Level 4 (U). The 6 options (Op.1 to 6) were proposed with the differences between scenarios, for examples, phase types of wastes, the isolated nuclide groups, chlorination process sequences. Op.1 adopts Level 0 and 1 to separate I, Tc, Se, C, Cs nuclides which are major concerns for long-term disposal through heat treatment. The rest of spent nuclear fuel will be disposed as oxide form itself. Op.2 contains Sr separation process using chlorination by MgCl2 and precipitation by K2CO3to alleviate the burden of heat after heat treatment process. U/TRU/RE will be remained and disposed in oxide form. Op.3 is set to pyroprocessing as reference method, but residual TRU/RE chlroides after electrorefining will be recovered as precipitates by K3PO4. Op.4 introduces NH4Cl to chlorinate TRU/RE from oxides after Op.2 applied and precipitates them. TRU/RE/Sr will be simultaneously chlorinated by NH4Cl without MgCl2 in Op.5. Then, chlorinated Sr and TRU/RE groups will be separated by post-chlorination process for disposal. But, chlorinated Sr and TRU/RE are designed not to be divided in disposal steps in Op.6. In this study, the mass flow analysis of radionuclide management process scenarios with updated process variables are performed. The amount and composition of wastes by types will be addressed in detail.

Support for nuclear power plant (NPP) dismantling & decommissioing (D&D) industry is necessary through development of the infrastructure and the D&D technology. Because KORI#1 and Wolsong#1 is planned to decommission until around 2030. Korea research institute of decommissioing (KRID) was established through the preliminary feasibility study. KRID has plan to support nuclear companies to join D&D industry. Normal facilities (Lv.1) of KRID infracstucture are currently being constructed and radiation management facilities (Lv.2) construction is expected to begin in October. Further, KRID is planning the construction of equipment to develop the procedure for radionuclide analysis through R&D project. A total period of the R&D project is 45 months, and the total R&D funding for this period is 19.4 billion won. The ultimate goal of the R&D project is to build the infractstucture base to analyze decommissioning radioactive wastes. Furthermore, the R&D project is important to reliably perform the NPP D&D.

Korea Atomic Energy Research Institute is developing a radionuclide management processes as a conditioning technology to reduce the burden of spent fuel disposal. The radionuclide management process refers to a process managing radionuclides with similar properties by introducing various technology options that can separate and recover radionuclides from spent fuels. In particular, it is a process aimed at increasing disposal efficiency by managing high-heat, high-mobility, and high-toxic radionuclides that can greatly affect the performance of the disposal system. Since the radionuclide management process seeks to consider various technology options for each unit process, it may have several process flows rather than have a single process flow. Describing the various process flows as a single flow network model is called the superstructure model. In this study, we intend to develop a superstructure model for the radionuclide management process and use it as a model to select the optimal process flow. To find the optimal process flow, an objective function must be defined, and at the fuel cycle system level multiple objectives such as effectiveness (disposal area), safety (explosure dose), and economics (cost) can be considered. Before performing the system-level optimization, it is necessary to select candidates of process flow in consideration of waste properties and process efficiency at the process level. In this study, a sensitivity analysis is conducted to analyze changes in waste properties such as decay heat and radioactivity when the separation ratio varies due to the performance change for each unit process of the radionuclide management process. Through this analysis, it is possible to derive a performance range that can have waste properties suitable for following waste treatment, especially waste form manufacturing. It is also possible to analyze the effect of waste properties that vary according to the performance change on waste storage and management approaches.

n Korea, the decommissioning of nuclear power plants is being prepared, and a large amount of radioactive waste is expected to be generated. In particular, clearance level waste, which accounts for more than 90%, requires prevention of cross-contamination and prompt classification. In this study, the possible exposure route and the derivation of exposure dose for worker exposure management in a movable analysis system that can be analyzed onsite were studied. The movable radionuclide analysis system is divided into a preparatory room, a sample storage room, a radioanalysis room, a laboratory, and a waste storage room. It consists of one radioanalysis worker and one pre-treatment worker, and the main radiation exposure is expected to occur in the movement path in the sample storage room, radioanalysis room, and laboratory. The source term for the exposure evaluation, the annual usage dose presented in the radiation safety report in the movable radionuclide analysis system was used. The input data for the evaluation of the external exposure dose under normal circumstances (exposure situation, working hours, distance, etc.) is referenced at facility specifications. The internal exposure dose evaluation was assumed to be acute exposure (1 hour) assumed as internal pollution due to the drop in liquid sample during the pretreatment work. As an evaluation method, a method using a calculation formula and a method using an evaluation code was performed. For the evaluation of exposure dose using the calculation formula, a preliminary evaluation was performed using the point source method, the point kernel method, and intake and dose conversion factors. In addition, VISIPLAN and IMBA codes were used to evaluate exposure dose using the evaluation code, and the input data were supplemented for evaluation. As a result of the evaluation, the annual exposure dose limit of 20 mSv was satisfied for both normal and non-normal situations. In future research, it is planned to derive the evaluation results by particular scenarios for the detailed movement route and evaluation time according to the work process in the mobile radionuclide analysis.

According to Article 4 and 5 of the Nuclear Safety and Security Commission (NSSC) Notice No. 2020-6, radioactive waste packages should be classified by radioactive levels, and finally permanently shipped to underground or surface disposal facilities. The level of the radioactive waste package is determined based on the concentrations of the radionuclides suggested in Article 8 of NSSC Notice No. 2021-26. Since most of the radionuclides in radioactive wastes are beta nuclides, chemical separation and quantification of the target nuclides are essential. Conventional methods to classify chemically non-volatile radionuclides such as Tc-99, Sr-90, Nb- 94, Fe-55 take a lot of time (about 5 days) and have low efficiency. An automated non-volatile nuclide analysis system based on the continuous chemical separation method of radionuclides has been developed to compensate for this disadvantages of the conventional method in this study. The features of the automated non-volatile nuclide separation system are as follows. First, the amount of secondary waste generated during the chemical separation process is very small. That is, by adopting an open-bed resin column method instead of a closed-bed resin column method, additional fittings and connector are unnecessary during the chemical separation. In addition, because the peristaltic pump is supplied for the sample and solution respectively, it is great effective to prevent cross-contamination between radioactive samples and the acid stock solution for analysis. Second, the factors that may affect results, such as solution amount, operating time and flow rate, are almost constant. By mechanically controlling the flow rate precisely, the operating time and additional factors required during the separation process can be adjusted and predicted in advance, and the uncertainty of the chemical separation process can be significantly reduced. Finally, it is highly usable not only in the continuous separation process but also in the individual separation process. It can be applied to the individual separation process because the user can set the individual sequence using the program. As a result of the performance evaluation of the automation system, recovery rates of about 80–90% and reproducibility within 5% were secured for all of the radionuclides. Furthermore, it was confirmed that the actual work time was reduced by more than 50% compared to the previous manual method. (It was confirmed that the operation time required during the separation process was reduced from 6 days to 3 days.) Based on these results, the automation system is expected to improve the safety of workers in radiation exposure, reduce human error, and improve data reliability.

Appropriateness of the minimum detectable activity in the analysis of gamma radionuclides is very important. This is reason determine the time factor among the conditions of the analysis when it is rationally determined has the advantage that radioactivity analysis can be performed accurately and quickly. In this study, 100 mL of an unknown sample was diluted in Marinelli Beaker 1L to obtain, review data on gamma radiation analysis results and minimum detectable activity for each measurement time. The measurement was used High Purity Germanium detector, target nuclides are Co-57, Co-58, Y-88 and Cs-137. Since the radioactivity analysis sample will be expected to be the waste subject to selfdisposal or less during the radioactive waste classification, the minimum detectable activity standard was set based on the detection of less than the permissible activity for self-disposal for each nuclide. The measurement methods were measured by classifying it into seven categories: 1000 seconds, 3600 seconds, 10000 seconds, 30000 seconds, 80000 seconds, 100000 seconds, and 150000 seconds. The radioactivity from this measurement are Co-57 2.89 Bq·g−1, Co-58 0.19 Bq·g−1, Y-88 0.20 Bq·g−1, Cs-137 0.15 Bq·g−1, the measurement results under all conditions were similar. On the other hand, the minimum detectable activity showed values above the allowable activity for self-disposal in not but Co-58 at 1000 and 3600 seconds. Only after taking the measurement time of 10000 seconds, the result was derived Co-57 0.0095 Bq·g−1, Co-58 0.0068 Bq·g−1, Y-88 0.0052 Bq·g−1, Cs-137 0.0062 Bq·g−1, which was confirmed to less than the allowable activity for self-disposal by nuclide. Reasonably determining the measurement time in gamma radionuclide analysis is a very important issue in terms of economy of time and accuracy of measurement. Although this study cannot be said to be able to determine a reasonable measurement time for all gamma radionuclide analysis, it is hoped that research on various samples will be made to contribute to the efficient measurement of gamma radioactivity.

As the plan for the nuclear dismantlement due to the permanent shutdown of Kori-1 and Wolseong- 1 nuclear power plants has been concretized, a “movable radionuclide analysis system” is being developed that can quickly and accurately analyze large amounts of radioactive waste generated on the sites during dismantling. This system has various advantages from the perspective of strict regulations on the radioactive waste movement and social acceptability, such as preventing unexpected accidents while moving on the national highway or expressway, reducing various documents and immediate response to dismantling plans. Currently the system is being developed to be equipped with previously developed sample pretreatment and radioactivity measuring equipment and automated volatile and nonvolatile nuclide separation equipments, but to ensure mobile stability, it needs to analyze factors and establish stability standards. In the KS Q ISO/IEC 17025:2017 standard, the requirements for “facilities and environmental conditions” are a very important factor in building reliability for consumers as part of the quality guarantee for this facility. In order to meet the requirements, the technical standards of various test equipment to be installed in this facility were investigated. The physical, chemical, and radiological hazards that could affect the safety of the equipment and workers in the process of moving the equipment between nuclear power plants or between nuclear dismantling sites were derived from vibrations, rapid changes in temperature and humidity, and the spread of contamination from radioactive waste samples. Therefore, the scope of application of the law, which is the basis for securing stability during movement, was classified into two situations: movement from facility manufacturer to installation site (non-contaminated) and movement from primary to secondary use (contaminated). And in order to investigate the Nuclear Safety Act, enforcement ordinances, and radiation safety management, and to establish standards for packaging and transportation of radioactive materials, the results of transportation tests and transport details were compared and analyzed. Finally, the air suspension systems and the automatic temperature and humidity control devices were analyzed to establish standards for securing stability against the vibration and the sharp changes in the temperature and humidity, and countermeasures such as accident measures in accordance with the Enforcement Decree of the Nuclear Safety Act were also investigated.

Site characterization for decommissioning Kori Unit 1 is ongoing in South Korea after 40 years of successful operation. Kori Unit 1’s containment building is assumed to be mostly radioactively contaminated, and therefore radiation exposure management and detailed contamination investigation are required for decommissioning and dismantling it safely. In this study, site-specific Derived Concentration Guideline Levels (DCGLs) were derived using the residual radioactivity risk evaluation tool, RESRAD-BUILD code. A conceptual model of containment building for Kori Unit 1 was set up and limited occupational worker building inspection scenario was applied. Depending on the source location, the maximum contribution source and exposure pathway of each radionuclide were analyzed. The contribution of radionuclides to dose and exposure pathways, by source location, is expected to serve as basic data in the assessment criteria of survey areas and classification of impact areas during further decommissioning and decontamination of sites.

본 연구에서는 인체 내에서 방사성핵종의 거동에 관하여 국제방사선방호위원회에서 권고한 최근 호흡기 모델, 소화기 모델 및 생체동역학 모델을 사용하여 생물학적분석 자료로부터 섭취량과 예탁유효선량을 평가하기 위한 BIDAS 프로그램을 개발하였다. 프로그램은 생물학적분석 자료를 관리하는 데이터베이스 모듈, 각 방사성 핵종에 대한 예측 생물학적분석 양을 내장하고 있는 모듈, 측정된 생물학적분석 양에 근거하여 급성 및 만성 피폭으로부터 섭취량과 선량을 평가하는 계산모듈 등으로 구성되어 있다. 본 논문은 프로그램의 특성과 검증결과에 대해 기술한다.

방사선과의 상호작용에 의하여 섬광이 발생하는 무기형광체(inorganic fluor)인 cerium activated yttrium silicate(CAYS)를 폴리설폰 고분자막에 함침시킴으로써, 형광 용액의 도움 없이 방사능 오염도를 측정할 수 있는 새로운 측정막을 제조하였다. 막의 제조는 두 가지 공정으로 나누어진다. 우선 고분자와 용매의 균일한 1차 제막용액을 유리판 위에 제막 후 용매증발을 통해 폴리설폰 고분자막이 생성되도록 하였다. 고형화된 폴리설폰 필름 표면에 CAYS가 분산된 고분자 제막용액을 2차로 도포시킨 후, 비용매 욕조에 침지시키는 상전환 공정을 이용하거나 용매의 증발을 통한 유리화에 의해 2차 용액의 고형화를 유도함으로써 함침막을 제조하였다. 이렇게 제조된 막의 형상은 치밀한 구조를 지니는 고분자 지지체와 이에 완전히 고착된 CAYS함침막의 이중구조를 지니게되며, 지지체 부분은 막의 안정성을 2차 제막에서 생성된 부분은 기능성의 향상을 이룰 수 있는 구조적 특성을 지닌다. 제조된 함침막에 방사성핵종을 직접 도포하여 방사성핵종의 탐지 특성을 측정하였을 때 효율적인 탐지 특성을 지니는 것으로 확인되었다.