With an ultimate view to identifying abnormal releases of radioactive materials, a set of liquid and gaseous effluent data including unplanned or uncontrolled releases annually reported form the U.S. and Korean nuclear power plants were systematically analyzed. With the use of 21 years’ worth of annual discharge data for 7 radionuclide groups and 24 individual radionuclides, taken from a combined total of 1,610 reactor-years (RYs) covering 62 units of US Pressurized Water Reactors (PWRs) and 22 units of Korean PWRs, three novel formulas for estimating events were employed to calculate characteristic values. Applying these characteristic values derived from the event estimation formulas to events that transpired during 699 RYs in operational US PWRs revealed an enhanced predictive accuracy for abnormal events when considering individual radionuclides, as opposed to grouping them by radionuclide groups. This effect was particularly pronounced for specific events such as leaks caused by problems in Gas Decay Tanks, leaks in Steam Generator Power Operated Relief Valves, fuel defects, and leaks during spent nuclear fuel processing. In the case of Korean PWRs, fuel defects were identified as the primary events related to radioactive effluent releases. The methodologies and characteristic values derived from this study were applied to these events. The event estimation rate was lower in Korean compared to US PWRs, which can be attributed to the lower frequency of event occurrences in Korean PWRs (30 RYs) compared to the US. The approach proposed in this study may contribute to develop a methodology to identify implicit abnormal release data and correlate them with specific operational occurrences or events, which could improve the conventional practice of simply recording and reporting radioactive discharge data.

For safe and successful decommissioning, it is one of the most important procedures that establishing the goal and complying with regulations of which final status of decommissioned site and building. The dose criteria for cyclotron facilities should be established and applied to reuse the site and building, since building and component of a cyclotron facility have been activated by incident secondary neutrons from radioactive isotope processes (e.g. 18O(p,n)18F, etc.). Furthermore, appropriate approaches should be applied to demonstrate compliance with the dose criteria for reliability of reuse. It is of noted that U.S. NRC (Nuclear Regulatory Commission) has confirmed that the residual radioactivity which distinguishable from background radiation results in a TEDE (Total Effective Dose Equivalent) does not exceed 25 mrem (0.25 mSv) per year as radiological criteria for unrestricted use of not only nuclear power plants but also cyclotron facilities referred to 10 CFR Part 20.1402. In addition, U.S. NRC noted the two approaches (i.e. dose assessment methods and, DCGL and final status surveys) which can be applied for demonstrating compliance with the dose criteria of 10 CFR Part 20 and recommended DCGL and FSS approach based on advantages and disadvantages of the two approaches. In order to using DCGL and FSS approach, U.S. NRC suggested screening approach; using DandD Version 2 which assesses TEDE under ICRP 28 and site-specific approach; using all models or computational codes which approved by NRC staff. There are several foreign cases that release of cyclotron facilities after decommissioning (i.e. U.S. and Japan). U.S., for examples, there are two DCGL approach cases and one dose modeling case based on 25 mrem per year same as reactor facilities. The dose modeling case, however, which may not be really used in Korea because of its low applicability. On the other hand, Japan case did not establish any radiological criteria for site and building reuse such as DCGL and just confirm “no more contamination” which is all residual radioactivity is lower than MDC based on real survey. Japan case also may not be used in Korea since criteria of “no more contamination” is not clear and hard to apply for all sites. Considering regulations and criteria for site release and reuse in Korea, this study aims to suggest radiological criteria and the demonstration approach of compliance for decommissioning of cyclotron facilities based on Nuclear Safety Acts and NSSC notices.

Kori Unit 1 was permanently shut down in 2017 and is currently being prepared for decommissioning. Decommissioning waste generated during the decommissioning of a nuclear power plant has the characteristic of being generated in large quantities over a short period. Therefore, if proper management is not carried out, abnormal situations (i.e., unauthorized disposal, diversion, etc.) may occur. According to IAEA General Safety Report Part 6, radioactive waste shall be managed for all waste streams in decommissioning. This means ensuring that all waste streams are managed by the recorded inventory of all decommissioning waste and verifying that the recorded inventory is reasonable. The radioactive waste management has been managed in units such as mass and radioactivity. However, in the case of decommissioning waste, the amount is very large, so management by radioactivity is expected to have limitations. Therefore, in this study, a simple test was conducted to verify the decommissioning waste generated by a hypothetical scenario by mass. In this study, establish a scenario assuming various flows of decommissioning waste expected to be generated and calculate the expected inventory of decommissioning waste using Microsoft Excel. Specifically, using “Material Unaccounted For” (MUF), a material balance equation in IAEA Services Series 15, Nuclear Material Accounting Handbook, the error inventory was calculated as the difference between the physical inventory of decommissioning waste in the area and the ending inventory. We propose a simple test scenario to verify the flow of decommissioning waste by verifying that the error inventory reasonably matches the set allowable error. This study aims to verify the inventory of decommissioning waste using the material balance methodology used for nuclear material accounting. It is expected that the safety and reliability of the nuclear power plant decommissioning process can be secured by verifying that the total inventory of equipment before decommissioning and the inventory of remaining equipment and decommissioning waste after decommissioning are reasonably consistent.

Among nuclear power plants in the Republic of Korea, Kori Unit 1 and Wolsong Unit 1 have been permanently shut down, and Kori Unit 1 is preparing to be decommissioned. According to the decommissioning plan (DP) of Kori Unit 1, a radioactive waste processing complex will be built on the Kori site to reduce radioactive waste generated during decommissioning actively, and various types of decommissioning waste are expected to be treated in the complex. It is judged that matters related to the safety assessment of the complex are not included in the DP since the equipment and treatment processes have not been determined. IAEA GSR Part 5 states that radioactive waste processing complex shall be operated according to national regulations and the conditions imposed by the regulatory body. However, it has been confirmed that separate regulatory requirements for the complex have not yet been established in Korea. It is expected that the Regulation on Technical Standards for Nuclear Facilities, etc. will be applied mutatis mutandis. Liquid and gaseous radioactive materials can be expected to be released into the sea or atmosphere during the operation of the complex. Accordingly, it should be proved that standards such as discharge limits of radioactive effluents are met. Although the assessment of radioactive effluent discharged from nuclear power plants to the environment is systematically conducted, it has been confirmed that the safety assessment framework for radioactive effluents discharged from the complex has not yet been established. Currently, the SAFRAN Tool is based on SADRWMS (Safety Assessment Driving Radioactive Waste Management Solutions), an IAEA safety assessment methodology for pre-disposal management, which uses Pathway Dose Factors (PDFs) derived from generic environmental models. Therefore, in order to conduct a more detailed safety assessment of the complex on a specific site, site characteristic data should be reflected. Although safety assessment using the SAFRAN Tool was conducted at the Thailand Institute of Nuclear Technology (TINT) facility, detailed data were not provided, and PDFs reflecting site characteristic data were not applied. Also, no other studies that considered many types of waste and provided detailed data on the safety assessment were not confirmed. Therefore, this study developed K-CRAFT (Kyung Hee – Comprehensive RAdioactive waste treatment Facility safety assessment Tool), this tool that can derive PDFs by reflecting site characteristic data based on the SADRWMS methodology and conducted preliminary safety assessment for the complex which will be built in Kori site by this tool.

The solid-state chemistry of uranium is essential to the nuclear fuel cycle. Uranyl nitrate is a key compound that is produced at various stages of the nuclear fuel cycle, both in front-end and backend cycles. It is typically formed by dissolving spent nuclear fuel in nitric acid or through a wet conversion process for the preparation of UF6. Additionally, uranium oxides are a primary consideration in the nuclear fuel cycle because they are the most commonly used nuclear fuel in commercial nuclear reactors. Therefore, it is crucial to understand the oxidation and thermal behavior of uranium oxides and uranyl nitrates. Under the ‘2023 Nuclear Global Researcher Training Program for the Back-end Nuclear Fuel Cycle,’ supported by KONICOF, several experiments were conducted at IMRAM (Institute of Multidisciplinary Research for Advanced Materials) at Tohoku University. First, the recovery ratio of uranium was analyzed during the synthesis of uranyl nitrate by dissolving the actual radioisotope, U3O8, in a nitric acid solution. Second, thermogravimetric-differential thermal analysis (TG-DTA) of uranyl nitrate (UO2(NO3)2) and hyper-stoichiometric uranium dioxide (UO2+X) was performed. The enthalpy change was discussed to confirm the mechanism of thermal decomposition of uranyl nitrate under heating conditions and to determine the chemical hydrate form of uranyl nitrate. In the case of UO2+X, the value of ‘x’ was determined through the calculation of weight change data, and the initial form was verified using the phase diagram for the U-O system. Finally, the formation of a few UO2+X compounds was observed with heat treatment of uranyl nitrate and uranium dioxide at different temperature intervals (450°C-600°C). As a result of these studies, a deeper understanding of the thermal and chemical behavior of uranium compounds was achieved. This knowledge is vital for improving the efficiency and safety of nuclear fuel cycle processes and contributes to advancements in nuclear science and technology.

To address the pressing societal concern in Korea, characterized by the imminent saturation of spent nuclear fuel storage, this study was undertaken to validate the fundamental reprocessing process capable of substantially mitigating the accumulation of spent nuclear fuel. Reprocessing is divided into dry processing (pyro-processing) and wet reprocessing (PUREX). Within this context, the primary focus of this research is to elucidate the foundational principles of PUREX (Plutonium Uranium Redox Extraction). Specifically, the central objective is to elucidate the interaction between uranium (U) and plutonium (Pu) utilizing an organic phase consisting of tributyl phosphate (TBP) and dodecane. The objective was to comprehensively understand the role of HNO3 in the PUREX (Plutonium Uranium Redox Extraction) process by subjecting organic phases mixed with TBPdodecane to various HNO3 concentrations (0.1 M, 1.0 M, 5.0 M). Subsequently, the introduction of Strontium (Sr-85) and Europium (Eu-152) stock solutions was carried out to simulate the presence of fission products typically contented in the spent nuclear fuel. When the operation proceeds, the complex structure takes the following form. 􀜷􀜱􀬶 􀬶􀬾(􀜽􀝍) + 2􀜰􀜱􀬷 􀬿(􀜽􀝍) + 2􀜶􀜤􀜲(􀝋􀝎􀝃) ↔ 􀜷􀜱􀬶(􀜰􀜱􀬷)􀬶 ∙ 2􀜶􀜤􀜲(􀝋􀝎􀝃) Subsequently, separate samples were gathered from both the organic and aqueous phases for the quantification of gamma-rays and alpha particles. Alpha particle measurements were conducted utilizing the Liquid Scintillation Counter (LSC) system, while gamma-ray measurements were carried out using the High-Purity Germanium Detector (HPGe). The distribution ratio for U, Eu (Eu-152), and Sr (Sr-84) was ascertained by quantifying their activity through LSC and HPGe. Through the experiments conducted within this program, we have gained a comprehensive understanding of the selective solvent extraction of actinides. Specifically, uranium has been effectively separated from the aqueous phase into the organic phase using a combination of tributyl phosphate (TBP) and dodecane. Subsequently, samples containing U(VI), Eu(III), and Sr(II) underwent thorough analysis utilizing LSC and HPGe detectors. Our radiation measurements have firmly established that the concentration of nitric acid enhances the selective separation of uranium within the process.

Notice of the NSSC No.2021-14 defines the term ‘Neutron Absorber’ as a material with a high neutron absorption cross section, which is used to prevent criticality during nuclear fission reactions and includes neutron absorbers as target items for manufacture inspection. U.S.NRC report of the NUREG-2214 states that the subcriticality of spent nuclear fuel (SNF) in Dry Storage Systems (DSSs) may be maintained, in part, by the placement of neutron absorbers, or poison plates, around the fuel assemblies. This report mentions the need for Time-Limited Aging Analysis (TLAA) on depletion of Boron (10B) in neutron absorbers for HI-STORM 100 and HISTAR 100. Also, this report mentions that 10B depletion occurs during neutron irradiation of neutron absorbers, but only 0.02% of the available 10B is to be depleted through conservative assumptions regarding the neutron flux or accumulated fluence during irradiation, which supports the continued use of the neutron absorbers in the SNF dry storage cask even after 60 years of evaluated period. There are several types of commercially available neutron absorbers, broadly classified into Boron Carbide Cermets (e.g., Boral®), Metal Matrix Composites (MMC) (e.g., METAMIC), Borated Stainless Steel (BSS), and Borated Al alloy. While irradiation tests for neutron absorbers are primarily conducted during wet storage systems, there are also some prior studies available on irradiation tests for neutron absorbers during dry storage systems. For examples, there is an analysis of previous research on high-temperature irradiation test of metallic materials and identification of limitations in existing methodologies were conducted. Furthermore, an improvement plan for simulating the high-temperature irradiation damage of neutron absorbers was developed. In report published by corrosion society summarizes the evaluation results of the degradation mechanisms for Stainless Steel- and Al-based neutron absorbers used in SNF dry storage systems.

The International Atomic Energy Agency (IAEA) Safety Fundamentals No. SF-1 Safety Principle 7 states that people and the environment, present and future, must be protected against radiation risk. Therefore, it is important to evaluate the safety of radioactive waste repositories on a longterm time scale to ensure future safety. However, IAEA-TECDOC-767 states that the long-term time scale of interest means that the risk or dose to future individuals cannot be reliably predicted because it relies on assumptions. Therefore, evaluating the safety of long-term time scales should use safety indicators that are less dependent on assumptions. Radiotoxicity is one of the safety indicators that represent an inherent risk from radioactive waste. It has been mainly used to show the time required until the hazard presented by waste decreases to that of natural uranium ore and is easy to use in communication with the public. There are several methods for calculating Radiotoxicity. Radioactivity is multiplied by a Dose Conversion Factor (DCF) to be expressed in Sv units, or radioactivity be divided into Maximum Permissible Concentration (MPC) to be expressed in m3 units as the amount of water needed to dilute the radionuclide to the permitted level. It is also often made dimensionless through comparison with reference materials like uranium ore. Radiotoxicity varies in size several times, even if it is a waste of similar origins and components, depending on the Radiological variable (e.g., Annual Limitation Intake (ALI), Dose Conversion Factor (DCF), Maximum Permissible Concentration (MPC), Activity). Therefore, this study was conducted to determine whether there was a significant difference when different radiological variables were substituted. This study compares and analyzes their differences using various MPCs or DCFs used in each country. In addition, this study analyzes radionuclides that influence radiotoxicity with several radiological variables. This study introduces the effects of substituting different radiological variables.

International Atomic Energy Agency defines the term “Poison” as a substance used to reduce reactivity, by virtue of its high neutron absorption cross-section, in IAEA glossary. Poison material is generally used in the reactor core, but it is also used in dry storage systems to maintain the subcriticality of spent fuel. Most neutron poison materials for dry storage systems are boron-based materials such as Al-B Carbide Cermet (e.g., Boral®), Al-B Carbide MMC (e.g., METAMIC), Borated Stainless Steel, Borated Al alloy. These materials help maintain subcriticality as a part of the basket. U.S.NRC report NUREG-2214 provides a general assessment of aging mechanisms that may impair the ability of SSCs of dry storage systems to perform their safety functions during longterm storage periods. Boron depletion is an aging mechanism of neutron poison evaluated in that report. Although that report concludes that boron depletion is not considered to be a credible aging mechanism, the report says analysis of boron depletion is needed in original design bases for providing long-term safety of DSS. Therefore, this study aimed to simulate the composition change of neutron poison material in the KORAD-21 system during cooling time considering spent fuel that can be stored. The neutron source term of spent fuel was calculated by ORIGEN-ARP. Using that source term, neutron transport calculation for counting neutrons that reach neutron poison material was carried out by MCNP®-6.2. Then, the composition change of neutron poison material by neutron-induced reaction was simulated by FISPACT-II. The boron-10 concentration change of neutron poison material was analyzed at the end. This study is expected to be the preliminary study for the aging analysis of neutron poison material about boron depletion.

In nuclear facilities, a graded approach is applied to achieve safety effectively and efficiently. It means that the structures, systems, and components (SSCs) that are important to safety should be assured to be high quality. Accordingly, SSCs that consist of nuclear facilities should be classified with respect to their safety importance as several classes, so that the requirements of quality assurance relevant to the designing, manufacturing, testing, maintenance, etc. can be applied. Guidance for the safety classification of SSCs consisting of nuclear power plants and radioactive waste management facilities was developed by U.S.NRC and IAEA. Especially, in guidance for nuclear power plants, safety significance can be evaluated as following details. The single SSC that mitigates or/and prevents the radiological consequence or hazard was assumed to be failure or malfunction as the initiating event/accident occurred and the following radiological consequence was evaluated. Considering both the consequence and frequency of the occurrence of the initiating event/accident, the safety significance of each SSC can be evaluated. Based on the evaluated safety significance, a safety class can be assigned. The guidance for the safety classification of the spent nuclear fuel dry storage systems (DSS) was also developed in the United States (NUREG/CR-6407) and the U.S.NRC acknowledges the application of it to the safety classification of DSS in the United States. Also, worldwide including the KOREA, that guidance has been applied to several DSSs. However, the guidance does not include the methodology for classifying the safety or the evaluated safety significance of each SSC, and the classification criteria are not based on quantitative safety significance but are expressed somewhat qualitatively. Vendors of DSS may have difficulties to apply this guidance appropriately due to the different design characteristics of DSSs. Therefore, the purpose of this study is to evaluate the safety significance of representative SSCs in DSS. A framework was established to evaluate the safety significance of SSCs performing safety functions related to radiation shielding and confinement of radioactive materials. Furthermore, the framework was applied to the test case.

In Korea, Kori Unit 1 and Wolsong Unit 1, have been permanently shut down in 2017 and 2019, and more nuclear power plants are expected to be permanently shut down after continued operation successively. Spent fuel has been generated during operation and stored in spent fuel pools. Due to the expected saturation of spent fuel pools within the next several decades, transportation of a huge amount of spent fuel is anticipated to interim storage facilities or final disposal facilities, even though the specific location is not decided. The U.S. Nuclear Regulatory Commission (NRC) states that every environmental report prepared for the licensing stage of a Pressurized Water Reactor shall contain a statement concerning risk during the transportation of fuel and radioactive wastes to and from the reactor. Thus, the licensee should ensure that the radiological effects in accidents, as well as normal conditions in transport, do not exceed certain criteria or be small if cannot be numerically quantified. Specific conditions that a full description and detailed analysis of the environmental effects of transportation of fuel and wastes to and from the reactor are exempted are specified in 10 CFR Part 51. Since there are no official requirements for radiological dose assessment for workers and public during the transportation of spent fuel in Korea, the margin when applying the U.S. regulatory criteria to the environmental impact assessment during the transport of spent fuel generated from domestic nuclear power plants is evaluated. A different approach would be needed due to the difference in the characteristics of spent fuel and geographical features.

After the Fukushima nuclear accident in Japan, concerns have increased about radioactive releases from nuclear power plants (NPPs) into the environment. Analysis of annual radioactive effluent release reports (ARERRs) shows that from 2000 to 2020, abnormal releases of radioactive effluent occurred in 703 out of 1,323 Reactor·years in the United States, accounting for 53% of the total number of reactors in 63 PWRs. Furthermore, when examining incidents and malfunctions recorded in Korea’s Operational Performance Information System of Nuclear Power Plant (OPIS) during the same period, it can be estimated that abnormal releases occurred in 9 out of the 324 Reactor·years in PWRs and PHWRs. Meanwhile, database on radioactive releases from NPPs worldwide was collected, and events of abnormal/unplanned releases were investigated. Based on the data collected from 195 NPPs in 8 countries (South Korea, the United States, Japan, France, the United Kingdom, Germany, Spain, and Canada) over a period of 21 years, totaling 4,607 Reactor·years, a program called K-IRED (KHUIntegrated Radioactive Effluent Database) was developed using MS Access. Using K-IRED, three methodologies have been developed to predict abnormal events based on the annual radioactive releases for each NPPs and radionuclide (or radionuclide group). Three newly developed methodologies were applied to the 63 NPPs (1,323 Reactor·years) in the United States, categorized by radionuclides (or radionuclide groups). Assuming an increase in radioactive effluent due to abnormal events, the annual increase rate of radioactive effluent was calculated for each methodology and the results were analyzed. The optimal methodology among the three was derived, and the applicability of predicting abnormal events in other NPPs beforehand was examined. Therefore, by predicting abnormal or unplanned releases from NPPs to the environment in advance, it is possible to prevent accidents and reduce public concerns, as suggested by results of this study.

The amount of waste that contains or is contaminated with radionuclides is increasing gradually due to the use of radioactive material in various fields including the operation and decommissioning of nuclear facilities. Such radioactive waste should be safely managed until its disposal to protect public health and the environment. Predisposal management of radioactive waste covers all the steps in the management of radioactive waste from its generation up to disposal, including processing (pretreatment, treatment, and conditioning), storage, and transport. There could be a lot of strategies for the predisposal management of radioactive waste. In order to comply with safety requirements including Waste Acceptance Criteria (WAC) at the radioactive waste repository however, the optimal scenario must be derived. The type and form of waste, the radiation dose of workers and the public, the technical options, and the costs would be taken into account to determine the optimal one. The time required for each process affects the radiation dose and respective cost as well as those for the following procedures. In particular, the time of storing radioactive waste would have the highest impact because of the longest period which decreases the concentrations of radionuclides but increases the cost. There have been little studies reported on optimization reflecting variations of radiation dose and cost in predisposal management scenarios for radioactive waste. In this study, the optimal storage time of radioactive waste was estimated for several scenarios. In terms of the radiation dose, the cumulative collective dose was used as the parameter for each process. The cost was calculated considering the inflation rate and interest rate. Since the radiation dose and the cost should be interconvertible for optimization, the collective dose was converted into monetary value using the value so-called “alpha value” or “monetary value of Person-Sv”.

Kori Unit 1 was permanently shut down in 2017 and is preparing to be dismantled. Decommissioning nuclear power plants is expected to generate a lot of decommissioning waste. Therefore, a radioactive waste treatment complex will be built on the site to safely and effectively the process of decommissioning waste generated from the Kori Unit 1, and the details are specified in the decommissioning plan. Therefore, a safety assessment should be conducted according to the facility’s normal and abnormal operations to construct a radioactive waste treatment complex. Currently, a safety assessment for a radioactive waste treatment complex can be conducted by the Safety Assessment Framework (SAFRAN) Tool based on the Safety Assessment Driving Radioactive Waste Management Solutions (SADRWMS) methodology developed by the International Atomic Energy Agency (IAEA). The SAFRAN Tool can be calculated radiation dose and hazard quotient (HQ) for workers and the public under normal and abnormal conditions of the radioactive waste treatment complex. When evaluating the radiation dose for the public due to releasing radioactive materials into the air or discharging radioactive materials into liquids, the radiation dose is calculated using the amount discharged or released from the treatment complex, and the Pathway Dose Factors (PDFs) derived from the generic environmental model given in the IAEA Safety Reports Series No.19. PDFs, which reflect the specific site data rather than the generic environmental model data, should be calculated and evaluated when performing the safety evaluation of the radioactive waste treatment complex to be built on the Kori site. In addition, in the SAFRAN tool, there is an inconvenience in that it must be calculated separately by radionuclides to calculate the contribution of dose or HQ for each radionuclide. Therefore, in this study, a safety assessment tool for a radioactive waste treatment complex was developed using Visual Basic by supplementing the limitations of the SAFRAN tool. This tool was developed to allow users to choose whether to apply PDFs based on the IAEA SRS-19 based on the generic environmental model or PDFs calculated to reflect the specific site data. Furthermore, the tool considered all types of decommissioning wastes that may occur during the decommissioning of the Kori Unit 1 and the treatment process scheduled to be introduced. Therefore, this study is expected to be used as basic data when conducting the safety assessment of radioactive waste treatment complex scheduled to be introduced in Korea.

Spent nuclear fuel (SNF) characterization is important in terms of nuclear safety and safeguards. Regardless of whether SNF is waste or energy resource, the International Atomic Energy Agency (IAEA) Specific Safety Guide-15 states that the storage requirements of SNF comply with IAEA General Safety Requirement Part 5 (GSR Part 5) for predisposal management of radioactive waste. GSR Part 5 requires a classifying and characterizing of radioactive waste at various steps of predisposal management. Accordingly, SNF fuel should be stored/handled as accurately characterized in the storage stage before permanent disposal. Appropriate characterization methods must exist to meet the above requirements. The characterization of SNF is basically performed through destructive analysis/non-destructive analysis in addition to the calculation based on the reactor operation history. Burnup, Initial enrichment, and Cooling time (BIC) are the primary identification targets for SNF fuel characterization, and the analysis mainly uses the correlation identified between the BIC set and the other SNF characteristics (e.g., Burnup - neutron emission rate) for characterizing. So further identification of the correlation among SNF characteristics will be the basis for proposing a new analysis method. Therefore, we aimed to simulate a SNF assembly with varying burnup, initial enrichment, and cooling time, then correlate other SNF properties with BIC sets, and identify correlations available for SNF characterization. In this study, the ‘CE 16×16’ type assembly was simulated using the SCALEORIGAMI code by changing the BIC set, and decay heat, radiation emission characteristics, and nuclide inventory of the assembly were calculated. After that, it was analyzed how these characteristics change according to the change in the BIC set. This study is expected to be the basic data for proposing new method for characterizing the SNF assembly of PWR.

Once systems, structures and components (SSCs) of dry storage systems are classified with respect to safety function or safety significance (i.e., safety classification), appropriate engineering rules can be applied to ensure that they are designed, manufactured, maintained, managed (e.g. aging management) etc. In Unites States, the systems, structures and components (SSCs) consisting DSSs are classified into two or several grades (i.e., class A, B and C or not important to safety, and important to safety (ITS) or not important to safety (NITS)) with respect to intended safety function and safety significance. This classification methods were based on Regulatory Guide 7.10 (i.e., guidance for use in developing quality assurance programs for packaging). Also, in Korea, SSCs of DSSs should be classified into ITS and NITS in much the same as method based on Regulatory Guide 7.10. In that guidance, for providing graded approach to manage the SSCs of packaging, they were trying to classifying SSCs in accordance with radiological consequences. But there was limitations that the provided classification criteria was still qualitative, so that it was not enough for managing the SSCs according to graded approach. On the other hand, in some other nuclear facilities (i.e., nuclear power plant, radioactive waste management facility and disposal facility etc.), quantitative criteria relevant to radiological consequence (i.e., radiation doses to workers or to the public) or inventory of radioactivity are existed so that it can be applied for classifying safety classes. In summary, the study on the application safety classification that applied quantitative criteria to perform safety classification of SSCs in DSS is inadequate or insufficient. The purpose of this study is proposing the preliminary framework for estimating safety significance of SSCs in DSS which can be utilized in our further advanced studies. In this study, a framework was established to estimate the safety significance of SSCs related to radiation shielding and confinement using MCNP® 6.2 and Microsoft Excel. Referring to the methodology of IAEA Specific Safety Guide 30, we assumed severity for failures of components that could lead to degradation of the SSC’s performance. The safety class of SSC was decided based on the impact of SSC’s failure on consequences.

Discharge limits for nuclear power plant gaseous effluents are presented as dose constraints or on the basis of radioactivity or radioactivity concentration. Accordingly, the operator evaluates the amount of radioactive material discharged from a specific nuclear power plant to the environment and periodically reports them to regulators. Multi-step sampling and analysis and calculation are performed during the radioactivity evaluation process of radioactive effluent, and the uncertainty generated in each step causes the uncertainty of the final radioactivity. Considering that the purpose of evaluating radioactivity discharged from nuclear power plants to the environment is to verify the satisfaction of discharge limits and safety margins, it is necessary to accurately evaluate the discharged radioactivity as much as possible, understanding of the uncertainty contained in the reported value of radioactivity and efforts to reduce it. In this study, modelling of the radioactivity evaluation procedure in gaseous effluent discharged as batch mode from nuclear power plant has performed, a generalized framework was established to evaluate the uncertainty based on ISO/IEC Guide 98-3 (GUM: 1995) involved in the whole process, and the uncertainty contained in the calculated radioactivity of each radionuclide (group) was evaluated and its characteristics. In addition, through probabilistic evaluation, the actual probabilistic distribution and statistical characteristics of radioactive effluent releases reported as a single value were confirmed. As a result, the range of values expected to be included in the confidence level of approximately 95% of the distribution of values for radioactivity in a gaseous effluent discharged as a batch mode from nuclear power plant was calculated. And, the priority of each input parameter turned out to be (1) gaseous waste volume, (2) sample bottle volume, and (3) measured radioactivity of the sample. In addition, the probability distribution of the radioactivity was simulated by Monte Carlo method. As such, the mean, minimum, and maximum values in confidence level of 95% were obtained, and they were reasonably matched the calculated value within 5% deviation. It was shown that radioactivity to the environment, which has been reported as a single value, has a specific probabilistic distribution form.

Medical cyclotrons have been used for dedicated medical of commercial applications such as positron emission tomography (PET) for the past tens of years. These cyclotron facilities have produced positron-emitting radionuclides (i.e. 11C, 13N, 15O, 18F, etc.). Among them, 18F, produced by 18O(p,n)18F reaction is the most widely used which has longer half-life (around 110 m) and lower energy of emitted positrons (around 0.63 MeV). Secondary neutrons produced during 18O(p,n)18F reaction could cause neutron activation of structures, systems, and components of cyclotron facilities. Therefore, International Atomic Energy Agency (IAEA) had addressed that during the operation of cyclotrons, concrete walls become radioactive over time and this radioactivity needs to be characterized for planning of the facility decommissioning. Moreover, several prior studies had estimated the neutron activation and levels of radioactivity of concrete wall of cyclotron facilities. Although those studies assessed the neutron activation of actual cyclotron facilities, however, the purpose of assessment was only for decommissioning each individual facility. Also, the assumptions, conditions or insights of conclusion may be limited to each individual case. For these reasons, this study focused on analysis of effects of major factors (e.g. concrete type, impurity contents of structural materials, etc.) about neutron activation of cyclotron facilities. In this study, the well-known methodology of neutron activation estimation was established and neutron activation products of concrete wall of cyclotron vault was calculated. Also, sensitivity analyses were conducted to figure out the effects of major factors of neutron activation and production of radioactive wastes during decommissioning of the facility. The methodology and results were validated by two steps: comparing with prior studies and comparing with another computer code. Concrete type did not affect that the decision of level of radioactivity waste criteria. Because of relatively longer half-lives, impurity contents of structural materials especially Co and Eu were turned out one of the most important factors for planning the facility decommissioning. It is hard to simply figure out the radioactivity levels of cyclotron facilities, however, rough predictions of minimum period for decay-in-storage as radioactive waste management can be possible with using information of thermal neutron spectra and major impurity nuclides (e.g. 59Co, 151Eu and 153Eu) for minimization of radioactive waste production and relief of charge of radioactive waste management.

Organic scintillator is easy to manufacture a large size and the fluorescence decay time is short. However, it is not suitable for gamma measurement because it is composed of a low atomic number material. Organic scintillation detectors are widely used to check the presence or absence of radiation. The fluorescence of organic scintillators is produced by transitions between the energy levels of single molecules. In this study, an organic scintillator development study was conducted for use in gamma measurement, alternative materials for secondary solute used in basic organic scintillators were investigated, and the availability of alternative materials, detection characteristics, and neutron/gamma identification tests were performed. In other words, a secondary solute showing an improved energy transfer rate than the existing material was reported, and the performance was evaluated. 7-Diethylamino -4-methylcoumarin (DMC), selected as an alternative material, is a benzopyrone derivative in the form of colorless crystals, has high fluorescence and high quantum yield in the visible region, and has excellent light stability. In addition, it has a large Stokes shift characteristic, and solubility in solvent is good. Through this study, it was analyzed that the absorption wavelength range of DMC coincided with the emission wavelength range of PPO, which is the primary solute. Through this study, it was confirmed that the optimal concentration of DMC was 0.04wt%. As a result of performing gamma and neutron measurement tests using a DMC-based liquid scintillator, it showed good performance (FOM=1.42) compared to a commercial liquid scintillator. Therefore, the possibility of use as a secondary solute was demonstrated. Based on this, if studies on changes in the composition of secondary solute or the use of nanoparticles are conducted, it will be possible to manufacture and utilize a scintillator with improved efficiency compared to the existing scintillator.

The U.S. Nuclear Regulatory Commission (NRC) states that every environmental report prepared for the licensing stage of a Pressurized Water Reactor shall contain a statement concerning risk during the transportation of fuel and radioactive wastes to and from the reactor. Thus, the licensee should ensure that the radiological effect in accidents, as well as normal conditions in transport, do not exceed certain criteria or be small if cannot be numerically quantified. These are specified in 10 CFR Part 51 and applied in NUREG-1555 Supplement 1 Revision 1, which deals with Environmental Standard Review Plan. Corresponding regulations in Korea would be the Nuclear Safety and Security Commission Notice No. 2020-7. In Appendix 2 of the Notice, guides on the radiological environmental report for production and utilization facilities, spent nuclear fuel interim storage facilities, and radioactive waste disposal facilities. In this guide, unlike the regulations in the U.S., there are no obligations for radiological dose assessment for workers and public during the transportation. Therefore, overall regulations and their legal basis related to risk assessment during transportation conducted for the environmental report in the U.S. were analyzed in this study. On top of that, through the comparison with regulations in Korea, differences between the two systems were figured out. Finally, this study aims to find the points in terms of assessing transport risk to be revised in the current regulatory system in Korea.