This study investigated the effectiveness of various chlorinating agents in partitioning light water reactor spent fuel, with the aim of optimizing the chlorination process. Through thermodynamic equilibrium calculations, the effects of using MgCl2, NH4Cl, and Cl2 as a single chlorinating agent or applying MgCl2, NH4Cl, and Cl2 sequentially for spent fuel chlorination were evaluated Furthermore, in this study, assuming the actual process operation situation, where only a part of the semi-volatile nuclides is removed during the heat treatment process, and including the process of precipitating the molten salt from the chlorination process with K3PO4 and K2CO3 precipitants, the percentage distribution of 50 nuclides in the light water reactor spent fuel into each process stream was quantitatively calculated using the simulation function of the HSC program and tabulated for intuitive viewing. Compared to a single chlorinator, sequential chlorination more effectively separated the heat and radioactivity of the spent fuel from the uranium-dominated product solids. Specifically, the sequential application of the chlorinating agents following heat treatment led to a final solid separation characterized by 93.1% mass retention, 5.1% radioactivity, and 15.4% decay heat, relative to the original spent fuel. The findings underscore that sequential chlorination can be an effective method for spent fuel partitioning, either as a standalone approach or in combination with other partitioning processes such as pyroprocessing.

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.

The types of fuel loaded and burned in domestic nuclear power plants are WH-type and OPR/ APR-type nuclear power plants, with a total of 19 types. In the case of spent nuclear fuel released in Korea, the low combustion level of 45,000 MWD/MTU or less accounts for about 75%. In terms of fuel type, WH 17×17 and CE 16×16 fuels account for about 85% of all spent nuclear fuels. The thickness of the oxide film of the fuel cladding can make the fuel rod vulnerable during reactor operation, directly affecting the integrity of the fuel rods. so, it is a very important design factor in design. Therefore, the fuel rod design code that predicts and evaluates this has also been developed to accurately predict fuel rod corrosion. And it’s being applied to the design. In this study, the ECT probe measured by inserting it between fuel rods. The thickness of the fuel cladding oxide film was measured for spent nuclear fuel. When reloading operational nuclear fuel, the IAEA recommends an oxide film thickness of up to 100 micrometers. In this study, it was confirmed that spent nuclear fuels keeping integrity burned for 2-3 cycles were sufficiently maintained within the limit. However, the difference could be confirmed according to the characteristics of the coating material, the combustion cycle, and the use of poison rods. For the reliability of the data, symmetrical to the quadrant fuels were selected, and the fuel burned at the same period was measured. The method of selecting the target fuel can produce meaningful results.

Since the first operation of the Gori No. 1 nuclear power plant in Korea was started to operate in 1978, currently 24 nuclear power plants have been being operated, out of which 21 plants are PWR types and the rest are CANDU types. About 30% of total electricity consumed in Korea is from all these nuclear power plants. The accumulated spent nuclear fuels (SNFs) generated from each site are temporarily being stored as wet or dry storage type at each plant site. These SNFs with their high radiotoxicity, heat generating, and long-lived radioactivity are currently the only type of high-level radioactive waste (HLW) in Korea, which urgently requires to be disposed of in deep geological repository. Studies on disposal of HLW in various kind of geological repositories have been carried out in such countries as Sweden, Finland, United States, and etc. with their own management policies in consideration of their situations. In Korea long-term R&D research program for safe management of SNF has also been conducted during last couple of decades since around 1997, during which several various type of disposal concepts for disposal of SNFs in deep geological formations have been investigated and developed. The first concept developed was KAERI Reference Disposal System (KRS) which is actually very much similar to Swedish KBS-3, a famous concept of direct disposal of SNF in stable crystalline rock at a depth of around 500 m which has been regarded as one of the most plausible method worldwide to direct disposal of SNF. The world first Finnish repository will be also this type. Since the characteristics of SNF discharged from domestic nuclear reactors have been changed and improved, and burnup has sometimes increased, a more advanced deep geological repository system has been needed, KRS-HB (KRS with High Burnup SNF) has been developed and in consideration of the dimensions of SNFs and the cooling period at the time point of the disposal time, KRS+, a rather improved disposal concept has also been subsequently developed which is especially focused on the efficient disposal area. Recently research has concentrated on rather advanced disposal technology focused on a safer and more economical repository system in recent view of the rapidly growing amount of accumulated SNF. Especially in Korea the rock mass and the footprint area for the repository extremely limited for disposal site. Some preliminary studies to achieve rather higher efficiency repository concept for disposal of SNF recently have already been emphasized. Among many possible ones for consideration of design for high-efficiency repository system, a double-layered system has been focused which is expected to maximize disposal capacity within the minimum footprint disposal area. Based on such disposal strategy a rather newly designed performance assessment methodology might be required to show long-term safety of the repository. Through the study some prerequisites for such methodological development will be roughly checked and investigated, which covers FEP identification and pathway and scenario analyses as well as preliminary conceptual modeling for the nuclide release and transport in near-field, far-field, and even biosphere in and around the conceptual repository system.

The deep geologic repository (DGR) concept is widely accepted as the most feasible option for the final disposal of spent nuclear fuels. In this concept, a series of engineered and natural barrier systems are combined to safely store spent nuclear fuel and to isolate it from the biosphere for a practically indefinite period of time. Due to the extremely long lifetime of the DGR, the performance of the DGR replies especially on the natural geologic barriers. Assessing the safety of the DGR is thus required to evaluate the impacts of a wide range of geological, hydrogeological, and physicochemical processes including rare geological events as well as present water cycles and deep groundwater flow systems. Due to the time scale and the complexity of the physicochemical processes and geologic media involved, the numerical models used for safety evaluation need to be comprehensive, robust, and efficient. This study describes the development of an accessible, transparent, and extensible integrated hydrologic models (IHM) which can be approved with confidence by the regulators as well as scientific community and thus suitable for current and future safety assessment of the DGR systems. The IHM under development can currently simulate overland flow, groundwater flow, near surface evapotranspiration in a modular manner. The IHM can also be considered as a framework as it can easily accommodate additional processes and requirements for the future as it is necessary. The IHM is capable of handling the atmospheric, land surface, and subsurface processes for simultaneously analyzing the regional groundwater driving force and deep subsurface flow, and repository scale safety features, providing an ultimate basis for seamless safety assessment in the DGR program. The applicability of the IHM to the DGR safety assessment is demonstrated using illustrative examples.

To obtain a license for a deep geological disposal repository for spent nuclear fuel, it is necessary to perform a safety assessment that quantifies the radiological impact on the environment and humans. One of the key steps in the safety assessment of a deep geological repository is the development of scenarios that describe how the repository evolves over the performance period and how events and processes affect performance. In the field of scenario development, demonstrating comprehensiveness is critical, which describes whether all factors that are expected to have a significant impact on the repository's performance have been considered. Mathematical proof of this is impossible. However, If the scenario development process is logical and systematic, it can support the claim that the scenario is comprehensive. Three primary approaches are being considered for scenario development: ‘Bottomup’, ‘Top-down’, and ‘Hybrid’. Hybrid approach provides a more systematic and structured process by considering both the FEPs (Features, Events, Processes) and safety functions utilized in the bottomup and top-down approaches. Many countries that develop recent scenarios prefer demonstrating scenario comprehensiveness using a hybrid approach. In this study, a systematic and structured scenario development process of a hybrid approach was formulated. Based on this, sub-scenarios were extracted that describe the phenomena occurring in the repository over the performance period, categorized by period. By integrating and screening the extracted sub-scenarios, a scenario describing the phenomena occurring over the entire period of disposal was developed.

In Korea, borated stainless steel (BSS) is used as a storage rack in spent fuel pools (SFP) to maintain the nuclear criticality of spent fuels. As the number of nuclear power plants and the corresponding amount of spent fuels increased, the density in SFP storage rack also increased. In this regard, maintaining subcriticality of spent nuclear fuels became an issue and BSS was selected as the structural material and neutron absorber for high density storage rack. Since it is difficult to replace the storage rack, corrosion resistance and neutron absorbency are required for long period. BSS is based on stainless steel 304 and 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 the low solubility of boron in austenitic stainless steel, metallic borides such as (Fe, Cr) 2B are formed as a secondary phase. Metallic borides could cause Cr depletion near it, which could decrease the corrosion resistance of the material. In this paper, the long-term corrosion behavior of BSS and its oxide microstructures are investigated through accelerated corrosion experiment in simulated SFP conditions. Because the corrosion rate of austenitic stainless steel is known to be dependent on the Arrhenius equation, a function of temperature, the corrosion experiment is conducted by increasing the experimental temperature. Detail microstructural analysis is conducted using a scanning electron microscope, transmission electron microscope and energy dispersive spectrometer. After oxidation, a hematite structure oxide film is formed, and pitting corrosion occurs on the surface of specimens. Most of the pitting corrosion is found at the substrate surface because the corrosion resistance of the substrate, which has low Cr content, is relatively low. Also, the 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 the boron content and the neutron absorption ability of the material. Using boron’s high cross-section for neutrons, the neutron absorption performance of BSS was evaluated through neutron transmission tests. The effect of the corrosion behavior of BSS on its neutron absorption performance was investigated. Samples simulated to undergo up to 60 years of degradation before corrosion through accelerated corrosion testing did not show significant changes in the neutron shielding ability before and after corrosion. This can be explained in relation to the corrosion behavior of BSS. Boron was only leached out from the secondary phase exposed on the surface, and this oxidized secondary phase corresponds to about 0.17% of the volume of the total secondary phase. This can be seen as a very small proportion compared to the total boron content and is not expected to have a significant impact on neutron absorption performance.

Korean MMTT project has been launched in order to clarify the vibration and shock loads under normal condition and transportation (NCT) in Korean geological and transportation conditions and to evaluate the integrity of SNF under such a transportation load. To evaluate the integrity of the SNF during normal land and sea transport tests, a representative SNF that represents the entirety of the different types of SNFs stored in the spent fuel pool of the power plant should be selected. And, it is necessary to make the test assembly to have a statically and dynamically similar behavior with the actual SNF. Therefore, in this project, we selected two types of fuel assembly that are expected to exhibit relatively conservative behavior under NCT, and these assemblies are being fabricated into surrogate test assemblies to have a similar characteristic as actual SNF based on the accumulated data from the poolside examination and the hot cell test so far. Tests were conducted for NCT conditions. In addition, a fatigue test was performed to integrity of the nuclear fuel rods under NCT conditions. Nuclear fuel assemblies are transported while being laid inside the cask under NCT, and are exposed to external shocks and vibrations. At this time, the fuel rod between the grid and grid is exposed to bending motion by this external force. For this simulation, a fixture was developed and used for static bending tests and bending fatigue tests. To simulate spent nuclear fuel rod specimens, hydrogen reorientation Zry-4 cladding was used and simulated pellets made of stainless steel were applied. And also, it was bonded using epoxy to give bonding conditions between the inside and the pellet. As a result of the test, cracks occurred due to the concentrated load between the pellets, resulting in damage to the fuel rod. The fatigue results showed a similar trend compared to the results performed by ORNL, and the lower bound fatigue curve presented by NUREG-2224 was also satisfactory.

As regulations on carbon emissions increase, the interest in renewable energy is also increasing. However, the efficiency of renewable energy generation is highly low and has limitations in replacing existing energy consumption. In terms of this view, nuclear power generation is highlighted because it has the advantage of not emitting carbon. And accordingly, the amount of spent nuclear fuel is going to increase naturally in the future. Therefore, it will be important to obtain the reliability of containers for transporting safely and storing spent nuclear fuel. In this study, a method for verifying the integrity and airtightness of a metal cask for the safe transportation and storage of spent nuclear fuel was studied. Non-destructive testing, thermal stability, leakage stability, and neutron shielding were demonstrated, and as a result, suitable quality for loading spent nuclear fuel could be obtained. Furthermore, it is meaningful in that it has secured manufacturing technology that can be directly applied to industrial field by verifying actual products.

Currently, the development of evaluation technology for vibration and shock loads transmitted to spent nuclear fuel and structural integrity of spent nuclear fuel under normal conditions of transport is progressing in Korea by the present authors. Road transportation tests using surrogate spent nuclear fuel were performed in September, 2020 using a test model of KORAD-21 transportation cask and sea transportation tests were conducted from September 30 to October 4, 2021. Finally, the shake table tests and rolling test were conducted from October 31 to November 2, 2022. The shake table test was performed with the input load produced conservatively from the data obtained from the road and sea transportation tests. The test input was produced based on the power spectral densities and shock response spectrums from the transportation tests. In addition to the test inputs from the road and sea tests, sine sweep input and half sine input were used to verify the vibration characteristics of assemblies under boundary conditions during normal conditions of transport. Because the input load of the shake table test was produced conservatively, a slightly larger strain than the strain value measured in road and sea transportation tests was measured from the shake table tests. In the case of the sea test, it is considered that the process of enveloping the data in the 20 to 80 Hz range generated by the engine propeller system was performed excessively conservatively. As a result of analyzing the test results for the difference in boundary conditions, it was confirmed that the test conditions of loading the basket generated a relatively large strain compared to the conditions of loading the disk assembly for the same input load. Therefore, it is concluded that a transportation cask having a structure in which a basket and a disk are separated, such as KORAD-21, is more advantageous in terms of vibration shock load characteristics under normal conditions of transport than a transportation cask having an integral internal structure in which a basket and a disk are a single unit. However, this effect will be insignificant because the load itself transmitted to the disk assembly is very small.

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.

Flow-induced vibration can lead to fretting wear damage of fuel rods and spacer grids in nuclear reactors. Similarly, during the transport of spent nuclear fuel assemblies, continuous vibration and intermittent impact might also result in fretting wear due to dynamic interaction. Therefore, it is important to evaluate fuel rod damage due to fretting wear under such transport conditions. This study examines spent nuclear fuel rod specimens fabricated with hydride cladding tubes and simulated pellets, with hydrogen content ranging from 200 to 700 ppm and oxide film thickness ranging from 25 to 100 micrometers. Tests were conducted under a worst-case scenario, assuming continuous exposure to that condition during the expected transport time. Results indicate that the wear depth of all rod specimens occurred within the oxide film, suggesting a high resistance to fretting wear during transportation.

The Comprehensive Analyzer of Real Estimation for spent fuel POOL (CAREPOOL) has been developed for evaluating the thermal safety of a spent nuclear fuel pool (SFP) during the normal and accident conditions. The management of spent nuclear fuel function provides a management tool for spent nuclear fuel in the SFP. The fuel assemblies both in SFP and reactor side can be shown graphically in the screen. The loading sequence into transfer cask can be checked respectively in the CAREPOOL. A basic heat balance equation was used to estimate the SFP temperature using the heat load calculated in the previous step. The characteristics of typical SFPs and associated cooling systems at reactor sites in the Korea were applied. Accident simulation like station black out leading to loss of SFP cooling or inventory is possible. Emergency cooling water injection pipe installed subsequent to the events at Fukushima 2011 is also modeled in this system. The CAREPOOL provides four main functions- management of spent nuclear fuel, decay heat calculation by ORIGEN-S code, estimation of the time to boil/fuel uncovering by thermal-hydraulics calculations, fuel selection for periodic spent fuel transferring campaign. All of these are integrated into the GUI based CAREPOOL system. The CAREPOOL would be very beneficial to nuclear power plant operator and trainee who have responsibility for the SFP operation.

There have been a variety of issues related to spent nuclear fuel in Korea recently. Most of the issues are related to intermediate storage and disposal of spent nuclear fuel. However, recently, various studies have been started in advanced nuclear countries such as the United States to reduce spent nuclear fuel, focusing on measures to reduce spent nuclear fuel. In this study, a simple preliminary assessment of the thermal part was performed for the consolidation storage method which separates fuel rods from spent nuclear fuel and stores them. The preliminary thermal evaluation was analyzed separately for storing the spent fuel in fuel assembly state and separating the fuel rods and storing them. The consolidation storage method in separating the fuel rods was advantageous in terms of thermal conductivity. However, detailed evaluation should be performed considering heat transfer by convection and vessel shape when storing multiple fuel bundles simultaneously.

In Korea, the construction of dry storage facilities for spent nuclear fuel is being promoted through the 2nd basic plan for high-level radioactive waste management. When operating dry storage facilities, exposure dose assessment for workers should be performed, and for this, exposure scenarios based on work procedures should be derived prior. However, the dry storage method has not yet been sufficiently established in Korea, so the work procedure has not been established. Therefore, research is needed to apply it domestically based on the analysis of spent nuclear fuel management methods in major overseas leading countries. In this study, the procedure for receiving and storing spent nuclear fuel in a concrete overpack-based storage facility was analyzed. Among the various spent nuclear fuel management systems, the metal overpack-based HI-STAR 100 system and the concrete overpackbased HI-STORM 100 system are quite common methods in the United States. Therefore, in this study, work procedures were analyzed based on each final safety analysis report. First, the HI-STAR 100 overpack enters the facility and is placed in the transfer area. Remove the impact limiter of the overpack and install the alignment device on the top of the overpack. Place the HI-TRAC, an on-site transfer device, on top of the alignment unit and remove the lids of the two devices to insert the canister into the HI-TRAC. When the canister transfer is complete, reseat the lid to seal it, and disconnect the HI-TRAC from the HI-STAR 100. Raise the canister-loaded HI-TRAC over the alignment device on the top of the HI-STORM 100 overpack and remove the lids of the two devices that are in contact. Insert the canister into the HI-STORM 100 and reseat the lid. The HI-STORM 100 loaded with spent nuclear fuel is transferred to the designated storage area. In this study, the procedure for receiving and storing spent nuclear fuel in a concrete overpack-based storage facility was analyzed. The main procedure was the transfer of canisters between overpacks, and it was confirmed that HI-TRAC was used in the work procedure. The results of this study can be used as basic data for evaluating the exposure dose of operating workers for the construction of dry storage facilities in Korea.

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.

As temporary storage facilities for spent nuclear fuel (SNF) are becoming saturated, there is a growing interest in finding solutions for treating SNF, which is recognized as an urgent task. Although direct disposal is a common method for handling SNF, it results in the entire fuel assembly being classified as high-level waste, which increases the burden of disposal. Therefore, it is necessary to develop SNF treatment technologies that can minimize the disposal burden while improving long-term storage safety, and this requires continuous efforts from a national policy perspective. In this context, this study focused on reducing the volume of high-level waste from light water reactor fuel by separating uranium, which represents the majority of SNF. We confirmed the chlorination characteristics of uranium (U), rare earth (RE), and strontium (Sr) oxides with ammonium chloride (NH4Cl) in previous study. Therefore, we prepared U-RE-SrOx simulated fuel by pelletizing each elements which was sintered at high temperature. The sintered fuel was again powdered by heating under air environment. The powdered fuel was reacted with NH4Cl to selectively chlorinate the RE and Sr elements for the separation. We will share and discuss the detailed results of our study.

A radiation shielding resin with thermal stability and high radiation shielding effect has been developed for the neutron shielding resin filled in the shielding shell of dry storage/transport cask for spent nuclear fuel. Among the most commercially available neutron shielding resins, epoxy and aluminum hydroxide boron carbide are used. But in case of the resin, hydrogen content enhances the neutron shielding effect through optimization of aluminum hydroxide, zinc borate, boron carbide, and flame retardant. We developed a radiation shielding material that can increase the boron content and have thermal stability. Flame retardancy was evaluated for thermal stability, and neutron shielding evaluation was conducted in a research reactor to prove the shielding effect. As a result of the UL94 vertical burning test, a grade of V-0 was received. Therefore, it was confirmed that it had flame retardancy. According to an experiment to measure the shielding rate of the resin against neutron rays using NRF (Neutron Radiography Facility), a shielding rate of 91.54% was confirmed for the existing resin composition and a shielding rate of 96.30% for the developed resin composition. A 40 M SANS (40 M Small Angle Neutron Scattering Instrument) neutron shielding rate test was performed. Assuming aging conditions (6 hours, 180 degrees), the shielding rate was analyzed after heating. As a result of the experiment, the developed products with 99.8740% and 99.9644% showed the same or higher performance.

One of the most important factors in the delivery and acceptance requirements for dry storage of spent fuel is the burnup of spent fuel. Here, burnup has a unit of MWD/MTU and is used as a measure of how much nuclear fuel is depleted in a nuclear reactor. In addition, since it is one of the most basic characteristic information for the soundness evaluation of spent nuclear fuel, it is a required item not only by regulatory agencies but also by KORAD, the acquiring agency. The burnup of spent nuclear fuel is the burnup calculated through flux mapping using signals measured from in-reactor instruments during nuclear power plant operation (hereinafter: actual burnup) and the burnup calculated using the core design code (hereinafter: design burnup). In this paper, the design burnup of spent nuclear fuel discharged from OPR100 NPPs (Nuclear Power Plants) in Korea was recalculated to confirm the reliability of the actual burnup currently managed at the nuclear power plant. Basically, since spent nuclear fuel must maintain subcriticality under wet storage or dry storage, a burnup error of about 5% is considered as a conservative approach when evaluating the criticality safety of wet storage tanks and dry storage systems. Therefore, in this paper, we tried to verify whether the difference between actual burnup and design burnup for all spent nuclear fuel released from domestic OPR100 type light water reactor nuclear power plants is within 5%. As a result of the evaluation, the largest deviation between actual burnup and design burnup was about 1,457 MWD/MTU, and when converted into a percentage, it was about 3.3%. Therefore, it was confirmed that the actual burnup managed by OPR1000 NPPs in Korea has sufficient reliability. In the future, we plan to check the reliability of the performance burnup managed in WH NPPs, and some of them will be verified through measurement.