In the event of a loss of a SNF (spent nuclear fuel) transport cask during maritime transportation, it is essential to evaluate the critical depth at which the integrity of the cask can be maintained under high water pressure. SNF transport casks are classified as Type B containers and the integrity of of the containment boundary must be maintained up to a depth of 200 meters unless the containment boundary was breached under beyond-design basis accidents. However, if an intact SNF cask is lost at a depth deeper than 200-meter, release of radioactive material may occur due to breach of containment boundary with over-pressure. In this study, we developed a code for the evaluation of the pressure limit of SNF transport cask, which can be evaluated by inputting the main dimensions and loading conditions of cask. The evaluation model was coded as a computer module for ease of use. In the previous study, models with three different fidelities were developed to ensure the reliability of the calculation and maintain sufficient flexibility to deal with various input conditions. Those three models consisted of a high-fidelity model that provided the most realistic response, a low-fidelity model with parameterized simplified geometry, and a mathematical model based on the shell theory. The maximum stress evaluation of the three models confirmed that the mathematical model provides the most conservative results than the other two models. The previous results demonstrate that mathematical models can be used in the code of computer modules. In this study, additional models of transport cask were created using parametric modeling techniques to improve the accuracy of the pressure limit assessment code for different cask and situations. The same boundary conditions and loading conditions were imposed as in the previous simplified model, and the maximum stress results considering the change in the shape of the transport container were derived and compared with the mathematical model. The comparison results showed that the mathematical model had more conservative values than the simplified model even under various input conditions. Accordingly, we applied the mathematical model to develop a transportation container pressure limit evaluation code that can be simulated in various situations such as shape change and various situations.

For Korean nuclear fuel cycle project, it is necessary to design and evaluate the integrity of spent fuel storage. For the design and evaluation of spent fuel storage, it is necessary to evaluate the properties of various materials used in spent fuel storage. The materials previously considered in the design of nuclear power plants were limited to static properties and were listed in design and manufacturing code and standards. However, for the evaluation of the storage containers in scenarios such as transportation and events, dynamic material property evaluations are required. Research on the dynamic properties of materials is generally conducted in the fields of automotive and aerospace, and most of the studies are on metal materials under sheet conditions. Since the structural materials of the storage containers for used nuclear fuel are mostly composed of thick materials, consideration should be given to property evaluation methodology and quantitative comparison. In this study, the mechanical properties of stainless steel material with canister application were evaluated according to the strain rate, and the crack resistance evaluation was also performed. It was confirmed the changes in strength and crack resistance according to the increase in strain rate and observed differences in microstructural hardening behavior.

Long-term safe storage of spent nuclear fuel (SNF) determines sustainability of the current light water reactor (LWR) fleet. In the U.S., SNF is stored in stainless steel canister in dry cask storage system (DCSS) after spending several years in wet pool storage system while there is no DSCC in Republic of Korea. The SNF storage time in DSCC is expected to be multiple decades since no permanent geological repositories are identified in both countries. One limiting factor for extended storage of SNF in DSCC is chloride-induced stress corrosion cracking (CISCC) in the welded regions of the stainless steel canisters. The propensity for the occurrence of CISCC has warranted the development of the mitigation and repair technologies to ensure the safe and long-term storage for both present and new canister although no CISCC failure was reported yet. This study investigates cold spray deposition coatings of 304 L and 316 L stainless steels on prototypical stainless steel canisters such as sensitized flat and C-ring samples. The cold spray technology has been identified as the most promising approach by Extended Storage Collaboration Program (ESCP) driven by Electric Power Research Institute (EPRI). The talk includes microstructural characterization, adhesion strength measurement, residual stress evaluation, and corrosion behavior of the coated materials in boiling MgCl2 solution and electrochemical corrosion tests in NaCl solution. In addition, the capability of repair of cracks on the canister surface using the coating technology will be presented.

Nuclear inspection is necessary to verify nuclear activities. If North Korea takes denuclearization, North Korea’s nuclear materials should be verified through non-destructive testing and destructive testing for nuclear material production. Since destructive testing of all substances is impossible, nondestructive testing is essential. Most non-destructive tests are performed by measuring the energy of gamma rays, but the characteristics of nuclear fuel can be evaluated by measuring neutron sources when enclosed with thick shields and when shielding structures are difficult to remove. Before the neutron source evaluation of MAGNOX used by North Korea, the relative characteristics will be evaluated later by analyzing the burnup, enrichment, and cooling time of the spent nuclear fuels discharged from the domestic nuclear power plant. This study evaluated the source strength and major nuclides according to burnup for the WH17×17 nuclear fuel assembly. The depletion calculation was conducted using SCALE 6.2 ORIGEN, and 3.5wt% enrichment, 10, 20, 30, 40, 50, 60 MWd/kg burnup, and five years cooling time, the minimum requirement for transport specified in the notice of the Nuclear Safety Commission, was applied. Although the impact assessment on enrichment should be evaluated with MCNP Tally to consider the fission reaction of the generated neutrons, this study only evaluated the spontaneous fission and (a, n) reactions that occurred first because it only evaluates the burnup impact. As burnup increased, neutron generation increased, and most of the total neutron strength occurred through spontaneous fission from the 10 MWd/kg burnup step. The influence of Pu-240 nuclides was dominant in the 10 MWd/kg burnup step but most neutrons were generated in tiny amounts of Cm- 244 generated from 20 MWd/kg burnup. Since DPRK’s 5 MWe Yongbyon MAGNOX has very low burnup (about 0.7 MWd/kg), the primary neutron sources of 10 MWd/kg, Am-241 and Pu isotopes, especially Pu-240, are expected to be used as indicators for evaluating spent nuclear fuel characteristics. If only specific nuclides are evaluated as major neutron sources at lower burnup than those evaluated in this study, in that case, the accuracy of non-destructive testing can be improved. Additionally, the evaluation according to the enrichment and cooling time should be considered as well.

In the wake of the Fukushima NPP accident, research on the safety evaluation of spent fuel storage facilities for natural disasters such as earthquakes and tsunamis has been continuously conducted, but research on the protection integrity of spent fuel storage facilities is insufficient in terms of physical protection. In this study, accident scenarios that may occur structurally and thermally for spent fuel storage facilities were investigated and safety assessment cases for such scenarios were analyzed. Major domestic and international institutions and research institutes such as IAEA, NEA, and NRC provide 13 accident scenario types for Spent Fuel Pool, including loss-of-coolant accidents, aircraft collisions, fires, earthquakes. And 10 accident scenario types for Dry Storage Cask System, including transportation cask drop accidents, aircraft collisions, earthquakes. In the case of Spent Fuel Pool, the impact of the cooling function loss accident scenario was mainly evaluated through empirical experiments, and simulations were performed on the dropping of spent nuclear fuel assembly using simulation codes such as ABAQUS. For Dry Storage Cask System, accident scenarios involving structural behavior, such as degradation and fracture, and experimental and structural accident analyses were performed for storage cask drop and aircraft collision accidents. To evaluate the safety of storage container drop accidents, an empirical test on the container was conducted and the simulation was conducted using the limited element analysis software. Among the accident scenarios for spent fuel storage facilities, aircraft and missile collisions, fires, and explosions are representative accidents that can be caused by malicious external threats. In terms of physical protection, it is necessary to analyze various accident scenarios that may occur due to malicious external threats. Additionally, through the analysis of design basis threats and the protection level of nuclear facilities, it is necessary to derive the probability of aircraft and missile collision and the threat success probability of fire and explosion, and to perform protection integrity evaluation studies, such as for the walls and structures, for spent fuel storage facilities considering safety evaluation methods when a terrorist attack occurs with the derived probability.

Owing to the increase in saturation rate of the spent fuel storage pond in the Kori nuclear power plant, the interim spent fuel dry storage facility is scheduled to be constructed at the Kori site. To implement safeguards in the new dry storage facility effectively, the concept of “Safeguards-by- Design” (SBD) should be applied to reflect nuclear safeguard provisions in the earliest design stages. Detailed design information pertaining to dry storage facilities has not been determined; however, the design information related to safeguards have been inferred using case studies and interviews with nuclear power plant operators worldwide. On the basis of the results of the case studies on spent fuel dry storage facilities for light water reactors, most countries apply the metal cask method in containment buildings considering safety. Furthermore, Korean operators are also considering the same method owing to tight licensing schedules and safety issues. Using the Facility Safeguardability Assessment (FSA) methodology (one of the safeguard evaluation methodologies), the difference in design between the heavy water reactor spent fuel dry storage facility, an established IAEA safeguards approach reference nuclear facility, and the light water reactor spent fuel dry storage facility (the new nuclear facility) were analyzed. Two major differences were noted as issues pertaining to potential safeguards. First, the difference in design and transport method in terms of the difference in size and weight of the spent nuclear fuel is important; light water reactor fuel is 20 times heavier than heavy water reactor that needs partial defect inspection in assemblies. Second, the difference in safeguard approach owing to the difference between the modular storage method in heavy water reactor and the container type storage method in light water reactor must be considered; movable storage cask renders the IAEA surveillance approach difficult. The results of this study can be used to identify the safeguards requirements in advance, enabling the operator to design new dry storage facilities resulting in timely and cost-effective implementation.

Currently, there are 25 nuclear power plants (NPPs) in operation in Korea, including 22 pressurized water reactors (PWRs) and three pressurized heavy water reactors (PHWRs). Two NPPs, including Kori Unit 1 and Wolsong Unit 1, are permanently shut down and awaiting decommissioning. If Kori Unit 2, which is expected to be permanently shut down soon, is included, the number of decommissioning NPPs will be increased to three. Spent fuels (SFs) are continuously generated during the NPP operation, which are stored in an SF storage pool in NPPs to cool down the decay heat emitted from SFs. For safe NPP operation, SFs must be regarded as waste, and a disposal site must be selected to isolate SFs. However, an appropriate site has yet to be selected in Korea. SFs contain long-lived nuclides with a high specific activity. For disposal, it is important to characterize the nuclides in the fuels and delay the migration of the nuclides to the environment when SFs are placed in a future disposal facility. If the disposal container is broken, the nuclides in the fuels escape from the filling material, such as bentonite. These escaped nuclides are dissolved in groundwater and migrate to the surface of the earth. Thus, it is possible to assess the radiological impact, such as the exposure dose during and after the disposal, if the types and characteristics of nuclides in SFs are known. This study investigated the nuclides in SFs and identified exposure scenarios that may occur in the disposal process of SFs and migration characteristics when the nuclides leak into groundwater to propose a dose assessment methodology for workers and the public.

Recently, the spent fuel pools withdrawn from nuclear power plants in Korea have been saturated. Therefore, specific regulations on the management of spent fuel pools, such as transportation and intermediate storage are needed. The burnup history is directly related to the management of spent nuclear fuel. This is because the decision to handle nuclear fuel may vary depending on the initial concentration of nuclear fuel, the degree to which nuclear fuel is irradiated and radioisotope nuclides are decayed, and the cooling state in the spent nuclear fuel storage tank. The purpose of this study is to determine the burnup of fuel based on the value obtained by scanning the surface of spent nuclear fuel through a neutron detector. Conversely, a database of neutron signals that scan bundles of spent nuclear fuel with an instrument with an already identified combustion history needs to be completed. First of all, the correlation between burnup history and nuclides was identified in previous studies. By setting the burnup history as the input value in the ORIGEN-ARP code, it was possible to identify the radioactive isotopes remaining in the bundle of nuclear fuel. Neutrons can finally be measured based on the amount of nuclide inventory that constitutes spent nuclear fuel. Through MCNP, the neutron detector was simulated and signals were measured to confirm how it correlates with the previously acquired burnup history database. In addition, the M (sub-critical multiplication) value, which is essential for neutron measurement, was checked to confirm the degree to which additional neutrons were generated in spent nuclear fuel in a subcritical state. The target nuclear fuel assembly was CE16×16, WH14×14, and WH17×17, which confirmed the correlation (1) between burnup, enrichment, and cooling time with the previous research topic, TNSI (Total neutron source intensity). 􀜤􀜷􁈺􀜩􀜹􀝀/􀜯􀜶􀜷􁈻 = 0.83􁈺􀜵􀯇􁈻􀬴.􀬶􀬷􀬼 ∙ 􁈺􀜫􀜧􁈻􀬴.􀬸􀬺􀬶􀬻 ∙ 􀝁􀬴.􀬴􀬴􀬼􀬷∙􀯧 􁈺1􁈻 A neutron signal will be obtained from the case according to each burnup history constituting this database. In particular, PAR=SF, a function that calculates the production amount of the fission product, was used. To confirm the computational logic of SF, it was confirmed whether a reasonable calculation was made by calculating with a nuclide spectrum.

Given the domestic situation, all nuclear power plants are located at the seaside, where interim storage sites are also likely to be located and maritime transportation is considered inevitable. Currently, Korea does not have an independently developed maritime transportation risk assessment code, and no research has been conducted to evaluate the release rate of radioactive waste from a submerged transportation cask in the sea. Therefore, secure technology is necessary to assess the impact of immersion accidents and establish a regulatory framework to assess, mitigate, and prevent maritime transportation accidents causing serious radiological consequences. The flow rate through a gap in a containment boundary should be calculated to determine the accurate release rate of radionuclides. The fluid flow through the micro-scale gap can be evaluated by combining the flow inside and outside the transportation cask. In this study, detailed computational fluid dynamic and simplified models are constructed to evaluate the internal flow in a transportation cask and to capture the flow and heat transfer around the transportation cask in the sea, respectively. In the future, fluid flow through the gap will be evaluated by coupling the models developed in this study.

For deep geological repository of the spent nuclear fuel, the fuel assemblies loaded in the storage cask are transferred to the disposal cask and the operation is performed in the fuel handling hot cell at the fuel re-packaging facility. As the fuel handling hot cell shielding is accomplished by the concrete wall and the viewing glass window, the required shielding thickness was evaluated for both materials. The ordinary concrete is applied to hot cell wall and two kinds of glasses, i.e., single layer of lead glass and double layer of lead glass and borosilicate glass, are considered for the viewing glass window. A bare spent PWR fuel assembly exposed to the environment in the hot cell was considered as the neutron and gamma radiation sources. The neutron and gamma transport calculations were performed using the MAVRIC program of the SCALE code system for the dose rate evaluation. The dose limit of 10 μSv/h is applied as the target dose to establish the required shielding thickness. The concrete wall of 94 cm thickness reduces the total dose rate to 6.9 μSv/h, which is the sum of neutron dose and gamma dose. Penetrating the concrete wall, both of the neutron dose and the gamma dose decrease constantly with shield thickness and the gamma dose is always dominant through whole penetrating distance. Single layer lead glass of 74 cm thickness reduces total dose rate to 6.2 μSv/h. Applying double layer shield glass combined of lead glass and borosilicate glass, the total dose rate reduces to 3.6 μSv/h at same shield thickness of 74 cm. Through the shield glass, gamma dose decreases rapidly and neutron dose decreases slowly compared with those for concrete wall. In result, neuron dose becomes dominant on the window glass shielding. The more efficient dose reduction of double layer glass is achieved by the borosilicate glass’s superior neutron shielding power. Thus, the use of double layer glass of lead glass and borosilicate glass is recommended for the viewing glass of the fuel handling hot cell. Finally, it is concluded that about 1 m thick concrete wall and 75 cm thick viewing glass window are sufficient for the radiation shielding of the hot cell at the spent fuel repackaging facility.

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.

In order to dispose of spent nuclear fuel (SNF) in deep geological repository, source term evaluation considering its specification, enrichment, burnup, cooling time should be performed. In this study, the measured values of Takahama-3 pressurized water reactor SNF (WH 17×17) samples were analyzed with SCALE 6.1/ORIGEN-S and TRITON code calculation results for validation. Unlike the ORIGENS code, TRITON code calculations differed from two-dimensional neutron flux distribution by using the multi-group cross-section library. Both calculation results from ORIGEN-S and TRITON code showed higher errors in 234U, 239Pu, and 241Pu compared to other actinide nuclides. In the case of axial locations of fuel rods in fuel assembly, fuel rods located at the edge of the fuel assembly presented increased errors due to nuclear reaction cross-section. Overall, the ORIGEN-S predictions informed more accurate agreement with the measured results compared with TRITON results. Especially to 235U, 239Pu, and 240Pu radionuclides, ORIGEN-S errors were denoted more than twice as low as the TRITON results. Comparing the calculation results with experimental results implied that the ORIGENS code was more accurate code than the TRITON code for source term evaluation.

Many countries have been developing their own FEP (Feature, Event, Process) lists to formulate radionuclide release scenarios in deep disposal repository of spent nuclear fuels and to assess the safety. The main issue in developing a FEP list is to ensure its completeness and comprehensiveness in examining all plausible scenarios of radionuclide release in a repository of interest. To this end, the NEA International FEP (IFEP) list as a generic reference have been developed and updated through long-term international collaborations. Leading countries advanced in the research field of deep geologic disposal of spent nuclear fuels have comparatively mapped their project-specific FEP (PFEP) lists with the IFEP list. Recently in 2019, NEA has published an updated version of IFEP list (ver. 3.0) which has a different classification system: the IFEP version 3.0 has the five main categories including the waste package, repository, geosphere, biosphere and external factors while the previous IFEP versions were mainly classified into the external, environmental, and contaminant factors. Most leading countries in this field, Finland and Sweden, recently succeeded to obtain the design and/or construction licenses for deep geologic disposal of spent nuclear fuel. Therefore, their PFEP lists should be good benchmark cases to the following countries. However, their PFEP lists have not comparatively mapped with the most recent version of IFEP and thus some gaps may exist in showing completeness and comprehensiveness in comparison to the IFEP version 3.0. In this study, we comparatively map the PFEP lists of Finland and Sweden to the IFEP version 3.0. The comparatively mapped PFEP list could be used as the basis for verifying the comprehensiveness and completeness of the domestic PFEP list currently under development in Korea.

CYPRUS is a web-based waste disposal research comprehensive information management program developed by the Korea Atomic Energy Research Institute over three years from 2004. This program is stored as existing quality assurance documents and data, and the research results can be viewed at any time. In addition, it helps to perform all series of tasks related to the safety evaluation study of the repository in accordance with the quality assurance system. In the future, it is necessary to improve the user convenience by clarifying the relationship between FEP and scenarios and upgrading output functions such as visualization and automatic report generation. This purpose of this study is to research and develop the advanced program of CYPRUS. This study is based on building FEP, DIM and scenario databases. It is necessary to develop an algorithm to analyze and visualize the FEP, DIM and scenario relationship. This project is an integrated information processing platform for DB management and visualization considering user convenience. The first development goal is to build long-term evolutionary FEP, DIM, and scenarios as a database. The linkage by FEP item was designed in consideration of convenience by using a mixed delimiter of letters and numbers. This design provides information on detailed interactions and impacts between FEP items. Scenario data lists a series of events and characteristic change information for performance evaluation in chronological order. In addition, it includes information on FEP occurrence and mutual nutrition by period, and information on whether or not the repository performance is satisfied by item. The second development goal is to realize the relationship analysis and visualization function of FEP and scenario based on network analysis technique. Based on DIM, this function analyzes and visualizes interactions between FEPs in the same way as PID, RES, etc. In addition, this function analyzes FEP and DIM using network analysis technique and visualizes it as a diagram. The developed platform will be used to construct and visualize the FEP DB covering research results in various disposal research fields, to analyze and visualize the relationship between core FEP and scenarios, and finally to construct scenarios and calculation cases that are the evaluation target of the comprehensive performance evaluation model. In addition, it is expected to support the knowledge exchange of experts based on the FEP and scenario integrated information processing platform, and to utilize the platform itself as a part of the knowledge transfer system for knowledge preservation.

Spent nuclear fuels are temporarily stored in nuclear power plant site. When a problem such as cracking of spent nuclear fuel assembly or cladding occurs or uranium that has not been separated during the reprocessing remains, it is necessary to treat it. The borosilicate glasses have been considered to vitrify whole spent nuclear fuel assembly. However, a large amount of Pb addition was necessary to oxidize metals in assembly to make them suitable for oxide glass vitrifcation. Furthermore, these borosilicate glasses need to be melted at high temperatures (> 1,400°C) when UO2 content is more than 20wt%. Iron phosphate glasses can be melted at a relatively low temperature (< 1,300°C) even with a similar UO2 addition. A composition of iron phosphate glass for immobilization of uranium oxide has been developed. The glasses have glass transition temperatures of ~555°C that are high enough to maintain its phase stability in geological repositories. The waste loading of UO2 in the glass is ~33.73wt%. Normalized elemental releases from the product consistency test were well below the US regulation of 2 g/m2. Nuclear criticality safety and heat generation in deep geological repositories were calculated using MCNP and computational fluid dynamics simulation, respectively. The glass had effective neutron multiplication factor (keff) of 0.755, which is smaller than the nuclear- criticality safety regulation of 0.95. Surface temperature of the disposal canister is expected to lower than the limit temperature (< 100°C). Most of the U in the glass is in the 4+state, which is more chemically durable than the 6+state. As a result of long-term dissolution experiment, chemically-durable uranium pyrophosphate (UP2O7) crystals were formed.

Dry head end process is developing for pyro-processing at KAERI (Korea Atomic Energy Research Institute). Dry processes, which include disassembling, mechanical decladding, vol-oxidation, blending, compaction, and sintering shall be performed in advance as the head-end process of pyro-processing. Also, for the operation of the head-end process, the design of the connecting systems between the down ender and the dismantling process is required. The disassembling process includes apparatus for down ender, dismantling of the SF (Spent Fuel) assembly (16×16 PWR), rod extraction, and cutting of extracted spent fuel rods. The disassembling process has four-unit apparatus, which comprises of a down ender that brings the assembly from a vertical position to a horizontal position, a dismantler to remove the upper and bottom nozzles of the spent fuel assembly, an extractor to extract the spent fuel rods from the assembly, and a cutter to cut the extracted spent fuel rods as a final step to transfer the rod-cuts to the mechanical decladding process. An important goal of dismantling process is the disassembling of a spent nuclear fuel assembly for the subsequent extraction process. In order to design the down ender and dismantler, these systems were analyzed and designed, also concept on the interference tools between down ender and dismantler were considered by using the solid works tool.

In Korea, research on the development of safety case, including the safety assessment of disposal facility for the spent nuclear fuel, is being conducted for long-term management planning. The safety assessment procedure on disposal facility for the spent nuclear fuel heavily involves creating scenarios in which radioactive materials from the repository reach the human biosphere by combining Features, Events and Processes (FEP) that describe processes or events occurring around the disposal area. Meanwhile, the general guidelines provided by the IAEA or top-tier regulatory requirements addressed by each country do not mention detailed methods of ‘how to develop scenarios by combining individual FEPs’. For this reason, the overall frameworks of developing scenarios are almost similar, but their details are quite different depending on situation. Therefore, in order to follow up and clearly analyze the methods of how to develop scenarios, it is necessary to understand and compare case studies performed by each institution. In the previous companion paper entitled ‘Research Status and Trends’, the characteristics and advantages/disadvantages of representative scenario development methods were described. In this paper, which is a next series of the companion papers, we investigate and review with a focus on details of scenario development methods officially documented. In particular, we summarize some cases for the most commonly utilized methods, which are categorized as the ‘systematic method’, and this method is addressed by Process Influence Diagram (PID) and Rock Engineering System (RES). The lessons-learned and insight of these approaches can be used to develop the scenarios for enhanced Korean disposal facility for the spent nuclear fuel in the future.

Recently, the deep geological disposal system isolating a spent nuclear fuel (SNF) is considered a disposal method of high-level radioactive waste for the safety of humans or the natural environment. The one of important requirements for maintaining the thermal stability of these systems is that the temperature of the buffer does not exceed 100°C even though the decay heat is emitted from highlevel radioactive wastes loaded in the disposal container. In 2007, a deep geological disposal system based on the Swedish disposal concept was developed for the SNF in Korea. To respond to the development process, the thermal stability of the deep geological disposal system developed for the disposal of domestic pressurized light water reactor (PWR) SNFs with discharged burn-up of 55 GWD/MTU was evaluated in 2019. The thing is that the recent fuel activity is pursuing to operate further high burn-up fuel conditions, and it leads to emergency core cooling system (ECCS) revision for extending the license for up to 60 or more than 60 GWD/MTU in the world. In this regard, this study evaluates numerically the thermal stability of the deep geological disposal system for the high burn-up PWR SNF having large decay heat compared to previous conditions for two different length disposal containers classified according to the length of PWR SNFs discharged from domestic nuclear power plants. A finite element analysis using a computational program was used to evaluate the thermal design requirements. Results show that both types of disposal containers would increase the temperature which reduces or fails to meet the safety margin of the disposal system. This study suggests that the design of the previous disposal system is needed to be further developed for the high burn-up PWR SNF which would be used in future nuclear power plant systems.

Radiation dose rates for spent fuel storage casks and storage facilities of them are typically calculated using Monte Carlo calculation codes. In particular, Monte Carlo computer code has the advantage of being able to analyze radiation transport very similar to the actual situation and accurately simulate complex structures. However, to evaluate the radiation dose rate for models such as ISFSI (Independent Spent Fuel Storage Installation) with a lot of spent fuel storage casks using Monte Carlo computational techniques has a disadvantage that it takes considerable computational time. This is because the radiation dose rate from the cask located at the outermost part of the storage facility to hundreds of meters must be calculated. In addition, if a building is considered in addition to many storage casks, more analysis time is required. Therefore, it is necessary to improve the efficiency of the computational techniques in order to evaluate the radiation dose rate for the ISFSI using Monte Carlo computational codes. The radiation dose rate evaluation of storage facilities using evaluation techniques for improving calculation efficiency is performed in the following steps. (1) simplified change in detailed analysis model for single storage cask, (2) create source term for the outermost side and top surface of the storage cask, (3) full modeling for storage facilities using casks with surface sources, (4) evaluation of radiation dose rate by distance corresponding to the dose rate limit. Using this calculation method, the dose rate according to the distance was evaluated by assuming that the concrete storage cask (KORAD21C) and the horizontal storage module (NUHOMS-HSM) were stored in the storage facility. As a result of calculation, the distance to boundary of the radiation control area and restricted area of the storage facility is respectively 75 m / 530 m (KORAD21C case), and 20 m / 350 m (NUHOMS-HSM case).

In Korea, borated stainless steel (BSS) is used as spent fuel pool (SFP) storage rack to maintain nuclear criticality of spent fuels. As number of nuclear power plants and corresponding number of spent fuels increased, density in SFP storage rack also increased. In this regard, maintain subcriticality of spent nuclear fuels was raised as an issue and BSS was selected as structural material and neutron absorber for high density storage rack. Because it is difficult to replace storage rack, corrosion resistance and neutron absorbency are required for long period. BSS is based on stainless steel 304 and it is specified in the ASTM A887-89 standard depending on the boron concentration from 304B (0.20-0.29% B) to 304B7 (1.75-2.25% B). Due to low solubility of boron in austenitic stainless steel, metallic borides such as (Fe, Cr)2B are formed as secondary phase metallic borides could make Cr depletion near it which could decrease the corrosion resistance of material. In this paper, long-term corrosion behavior of BSS and its oxide microstructures are investigated through accelerated corrosion experiment in simulated SFP condition. Because corrosion rate of austenitic stainless steel is known to be dependent on the Arrhenius equation, a function of temperature, corrosion experiment is conducted by increasing the experimental temperature. Detail microstructural analysis was conducted with scanning electron microscope, transmission electron microscope and energy dispersive spectrometer. After oxidation, hematite structure oxide film is formed and pitting corrosions occur on the surface of specimens. Most of pitting corrosions are found at the substrate surface because corrosion resistance of substrate, which has low Cr content, is relatively low. Also, oxidation reaction of B in the secondary phase has the lowest Gibbs free energy compared to other elements. Furthermore, oxidation of Cr has low Gibbs free energy which means that oxidation of B and Cr could be faster than other elements. Thus, the long-term corrosion might affect to boron content and the neutron absorption ability of the material.