The HADES (High-level rAdiowaste Disposal Evaluation Simulator) was developed by the Nuclear Fuel Cycle & Nonproliferation (NFC) laboratory at Seoul National University (SNU), based on the MOOSE Framework developed by the Idaho National Laboratory (INL). As an application of the MOOSE Framework, the HADES incorporates not only basic MOOSE functions, such as multi-physics analysis using Finite Element Method (FEM) and various solvers, but also additional functions for estimating the performance assessment of Deep Geological Repositories (DGR). However, since the MOOSE Framework does not have complex mesh generation and data analyzing capabilities, the HADES has been developed to incorporate these missing functions. In this study, although the Gmsh, finite element mesh generation software, and Paraview, finite element analysis software, were used, other applications can be utilized as well. The objectives of HADES are as follows: (i) assessment of the performance of a Spent Nuclear Fuel (SNF) disposal system concerning Thermal-Hydraulic-Mechanical-Chemical (THMC) aspects; (ii) Evaluation of the integrity of the Engineered Barrier System (EBS) of both general and high-efficiency design perspective; (iii) Collaboration with other researchers to evaluate the disposal system using an open-source approach. To achieve these objectives, performance assessments of the various disposal systems and BMTs (BenchMark Test), conducted as part of the DECOVALEX projects, were studied regarding TH behavior. Additionally, integrity assessments of various DGR systems based on thermal criteria were carried out. According to the results, HADES showed very reasonable results, such as evolutions and distributions of temperature and degree of saturation, when compared to validated code such as TOUGH-FLAC, ROCMAS, and OGS (OpenGeoSys). The calculated data are within the range of estimated results from existed code. Furthermore, the first version of the code, which can estimate the TH behavior, has been prepared to share the contents using Git software, a free and open-source distribution system.

Dry storage of nuclear fuel is compromised by threats to the cladding integrity, such as creep and hydride reorientation. To predict these phenomena, spent fuel simulation codes have been developed. In spent fuel simulation, temperature information is the most influential factor for creep and hydride formation. Traditional fuel simulation codes required a user-defined temperature history input which is given by separate thermal analysis. Moreover, geometric changes in nuclear fuel, such as creep, can alter the cask’s internal subchannels, thereby changing the thermal analysis. This necessitates the development of a coupled thermal and nuclear fuel analysis code. In this study, we integrated the 2D FDM nuclear fuel code GIFT developed at SNU with COBRA -SFS. Using this, we analyzed spent nuclear stored in TN-24P dry storage cask over several decades and identified conditions posing threats due to phenomena like creep and hydrogen reorientation, represented by the burnup and peak cladding temperature at the start of dry storage. We also investigated the safety zone of spent nuclear fuel based on burnup and wet storage duration using decay heat.

Spent nuclear fuels should be safely stored until being disposed and dry storage system is predominantly used to retain the fuels. Thermal analysis to estimate temperatures of spent nuclear fuel and the storage system should be performed to evaluate whether the temperatures exceed safety limit. Recently, thermal hydraulic analysis with CFD codes is widely used to investigate the temperature of spent nuclear fuel in dry storage. COBRA-SFS is a legacy code based on subchannel analysis code, and its fidelity is verified for evaluating the thermal analysis for licensing a dry cask system. Herein, thermal analysis result based on CFD and COBRA-SFS codes is compared and the Dry Cask Simulator (DCS) is assessed as a benchmark experiment in this study. Extended Storage Collaborating Program (ESCP) led by the Electric Power Research Institute (EPRI) is organized to address the degradation effects of spent nuclear fuel during long-term dry storage, and DCS is the first phase of the program. The dry storage system, containing a single BWR assembly in a canister, was designed to produce validation-quality data for thermal analysis model. ANSYS FLUENT was used to simulate DCS. Simulations were conducted in various decay heat and helium pressure inside the canister. In realistic conditions of decay heat and helium pressure of actual dry cask system, CFD and COBRA-SFS analysis results gave good agreement with experimental measurement. Peak temperatures of channel can, basket, canister and shell predicted by CFD simulation also showed good prediction and the discrepancies were less than 7 K while measurements uncertainty was 7 K. In high decay heat and high pressure condition, however, CFD and COBRA-SFS underestimated peak cladding temperature than experimental results.

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.

In case a spent nuclear fuel transport cask is lost in the sea due to an accident during maritime transport, it is necessary to evaluate the critical depth by which the pressure resistance of the cask is maintained. A licensed type B package should maintain the integrity of containment boundary under water up to 200 m of depth. However, if the cask is damaged during accidents of severity excessing those of design basis accidents, or it is submerged in a sea deeper than 200 m, detailed analyses should be performed to evaluated the condition of the cask and possible scenarios for the release of radioactive contents contained in the cask. In this work, models to evaluate pressure resistance of an undamaged cask in the deep sea are developed and coded into a computer module. To ensure the reliability of the models and to maintain enough flexibility to account for a variety of input conditions, models in three different fidelities are utilized. A very sophisticated finite element analysis model is constructed to provide accurate response of containment boundary against external pressure. A simplified finite element model which can be easily generated with parameters derived from the dimensions and material properties of the cask. Lastly, mathematical formulas based on the shell theory are utilized to evaluate the stress and strain of cask body, lid and the bolts. The models in mathematical formula will be coded into computer model once they show good agreement with the other two model with much higher fidelity. The evaluation of the cask was largely divided into the lid, body, and bottom, bolts of the cask. It was confirmed that the internal stress of the cask was increased in accordance with the hydrostatic pressure. In particular, the lid and bottom have a circular plate shape and showed a similar deformation pattern with deflection at the center. The maximum stress occurred where the lid was in the center and the bottom was in contact with the body. Because the body was simplified and evaluated as a cylinder, only simple compression without torsion and bending was observed. The maximum stress occurred in the tangential direction from the inner side of the cylinder. The bolt connecting the lid and the body was subjected to both bending and tension at the same time, and the maximum stress was evaluated considering both tension and bending loads. In general, the results calculated by the formulas were evaluated to have higher maximum stresses than the analysis results of the simplified model. The results of the maximum stress evaluation in this study confirms that the mathematical models provide conservative results than the finite element models and can be used in the computer module.

The measurement activities to evaluate material balance of nuclear material are usually performed by operator. It is because that the IAEA does not have enough manpower to carry out nuclear measurement accountancy of all nuclear materials in the world. Therefore, the IAEA should consider scenarios which facility operator tries to divert nuclear material for misuse by distorting measurement record. It is required to verify the operator’s measurement data whether it is normal or not. IAEA measures inventory items using their own equipment which is independent of facility operator equipment for verification. Since all inventory lists cannot be verified due to limited resources, the number of items to be verified is determined through statistical method which is called as sample size calculation. They measure for the selected items using their own equipment and compares with operator’s record. The IAEA determines sample size by comprehensively considering targeted diverted nuclear material amount and targeted non-detection probability and performance of measurement equipment. In general, the targeted diverted nuclear material amount is considered significant quantity (plutonium: 8 kg, uranium-235: 75 kg). If the targeted non-detection probability or the performance of the verification equipment is low, the sample size increases, and on the contrary, in the case of high non-detection probability or good performance of verification equipment, even a small sample size is satisfied. It cannot be determined from a single sample size calculation because there are so many sample size combinations for each verification equipment and there are many diversion scenarios to be considered. So, IAEA estimates initial sample size based on statistical method to reduce calculation load. And then they calculate non-detection probability for a combination of initial sample size. Through the iteration calculation, the sample size that satisfies the closest to the target value is derived. The sample size calculation code has been developed to review IAEA’s calculation method. The main difference is that IAEA calculates sample size based on approximate equation, while in this study, sample size is calculated by exact equation. The benchmarking study was performed on reference materials. The data obtained by the code show similar results to the reference materials within an acceptable range. The calculation method developed in this study will be applied to support IAEA and domestic inspection activities in uranium fuel fabrication facility.

For producing radionuclides which were mostly used in medical purposes, for instance, Positron Emission Tomography (PET), there were about 1,200 PET cyclotrons operated in 95 countries based upon IAEA database (2020). Besides, including PET cyclotrons, demands for particle accelerators are continuously increasing. In Korea, about 40 PET cyclotrons are in operating phases (2020). Considering design lifetime (about 30–40 years) of cyclotrons, there will be demands for decommissioning cyclotron facilities in the near future. PET cyclotron produces radionuclides by irradiating charged particles to the targets. During this phase, nuclear reactions (18O(p,n)18F, 14N(d,n)15O etc.) produce secondary neutrons which induce neutron activation of accelerator itself as well as surrounding infrastructures (the ancillary subsystems, peripheral equipment, concrete walls etc.). Most of the ancillary systems including peripheral equipment can be neutron activated, since, most of them were made of steels. Steels like stainless steel or carbon steel may contain some impurities, typically cobalt. Although, there were several researches evaluating activation of concrete walls and accelerator components, estimating the activation and influence on neutron interaction of the other surrounding infrastructures were insufficient. In this study, by using computational neutron transport code (MCNP 6.2), and source term calculation code (FISPACT- II), we estimated neutron distribution in cyclotron vault and activation of ancillary subsystems including some peripheral equipment. Also, using Au foil and Cd cover, we measured thermal neutron distribution at 16 points on the concrete wall, and compared it to calculated results (MCNP). Even though, the compared results matches well, there was a discrepancy of neutron distributions between presence and absence of those equipment. Additionally, in estimating activation distributions by calculating, most of the steel-based subsystems including peripheral equipment should be managed by radioactive wastes after 20 years of operation. Throughout this study, we could find that influence on neutron interaction of those equipment can affect neutron distribution in concrete walls. This results vary the activation depth as well as location of the hot contaminated spot in concrete wall. For estimating or evaluating activation distributions in cyclotron facilities, there was need to consider some equipment located in cyclotron vault.

According to the Atomic Energy Act of Korea, radioactive waste can be cleared when it meets the criteria, less than 10 uSv·y−1 for individual dose and 1 person · Sv·y−1 for collective dose. Consequently, it is necessary to evaluate radiation dose to get permission for regulatory clearance from the regulatory body of Korea. Several computational programs can be used for dose calculation depending on disposal methods such as landfill, incineration, and recycling. As for incineration, the effects of radionuclide emitted during combusting radwaste have to be considered to figure out exposure dose. In this study, GASPAR code is described to assess exposure dose from effluents released to the atmosphere during incinerating combustible radioactive wastes for regulatory clearance. GASPAR is the code programmed by Radiation Safety Information Computational Center at Oak Ridge National Laboratory for computing annual dose due to radioactive effluents released from a nuclear power plant to the atmosphere during routine operation. The calculating methods of the code are based on the mathematical model of U.S. NRC regulatory guide 1.109, about beta and gamma radiation from noble gas in semi-infinite plume, radioiodine, and particulates. GASPAR evaluates both individual dose and population dose. The considering pathways are composed of external exposure by plume and ground deposition of effluents, and internal exposure as a result of inhalation and food ingestion. Since the calculation model of GASPAR requires various variables about the radionuclide and disposal site, the accuracy of the results is decided by inputted values. The program contains the default values to parameters such as the humidity, fraction of deposition, and storage time of foods. However, to get permission, it is important to use the appropriate data representing the condition of the combustion scenario as substitutes for the default since the values are localized to the country where the code was developed. Therefore, dose assessment by GASPAR code can be applied for regulatory clearance by incineration, when reliable values depending on the disposal plan inputted.

Prior to the investigations on fuel degradation it is necessary to describe the reference characteristics of the spent fuel. It establishes the initial condition of the reference fuel bundle at the start of dry storage. In a few technology areas, CANDU fuels have not yet developed comprehensive analysis tools anywhere near the levels in the LWR industry. This requires significantly improved computer codes for CANDU fuel design. In KNF, in-house fuel performance code was developed to predict the overall behavior of a fuel rod under normal operating conditions. It includes the analysis modules to predict temperature, pellet cracking and deformation, clad stress and strain at the mid-plane of the pellet and pellet-pellet interfaces, fission gas release and internal gas pressure. The main focus of the code is to provide information on initial conditions prior to dry storage, such as fission gas inventory and its distribution within the fuel pellet, initial volumes of storage spaces and their locations, radial profile of heat generation within the pellet, etc. Potential degradation mechanisms that may affect sheath integrity of CANDU spent fuel during dry storage are: creep rupture under internal gas pressure, sheath oxidation in air environment, stress corrosion cracking, delayed hydride cracking, and sheath splitting due to UO2 oxidation for a defective fuel. To upgrade the developed code that address all the damage mechanisms, the first step was a review of the available technical information on phenomena relevant to fuel integrity. The second step was an examination of the technical bases of all modules of the in-house code, identify and extend the ranges of all modules to required operating ranges. Further improvements being considered include upgrades of the analysis module to achieve sufficient accuracy in key output parameters. The emphasis in the near future will be on validation of the in-house code according to a rigorous and formal methodology. The developed models provide a platform for research and industrial applications, including the design of fuel behavior experiments and prediction of safe operating margins for CANDU spent fuel.

This study provides technical information about the nuclear fuel handling process, which consists of various subprocesses starting from new fuel receipt to spent fuel shipment at a nuclear power plant and the design requirements of fuel handling equipment. The fuel handling system is an integrated system of equipment, tools, and procedures that allow refueling, handling and storage of fuel assemblies, which comprise the fuel handling process. The understanding and reaffirming of detailed code requirements are requested for application to the design of the fuel handling and storage facility. We reviewed the design requirements of the fuel handling equipment for its adequate cooling, prevention of criticality, its operability and maintainability, and for the prevention of fuel damage and radiological release. Furthermore, we discussed additional technical issues related to upgrading the current code requirements based on the modification of the fuel handling equipment. The suggested information provided in this paper would be beneficial to enhance the safety and the reliability of the fuel handling equipment during the handling of new and spent fuel.

Radiological impact analyses were carried out for a near-surface radioactive waste repository at Gyeongju in South Korea. The RESRAD-ONSITE code was applied for the estimation of maximum exposure doses by considering various exposure pathways based on a land area of 2,500 m2 with a 0.15 m thick contamination zone. Typical influencing input parameters such as shield depth, shield materials’ density, and shield erosion rate were examined for a sensitivity analysis. Then both residential farmer and industrial worker scenarios were used for the estimation of maximum exposure doses depending on exposure duration. The radiation dose evaluation results showed that 60Co, 137Cs, and 63Ni were major contributors to the total exposure dose compared with other radionuclides. Furthermore, the total exposure dose from ingestion (plant, meat, and milk) of the contaminated plants was more significant than those assessed for inhalation, with maximum values of 5.5×10−4 mSv‧yr−1 for the plant ingestion. Thus the results of this study can be applied for determining near-surface radioactive waste repository conditions and providing quantitative analysis methods using RESRAD-ONSITE code for the safety assessment of disposing radioactive materials including decommissioning wastes to protect human health and the environment.

희토류(Rare Earth) 함량이 높은 TRU 생성물 중의 RE 원소를 감소시키기 위하여 RE 원소와 UCl3의 산화반응을 이용한 RE 제거 공정의 타당성을 HSC Chemistry 코드를 이용하여 검토하였다. 사용후핵연료에 포함된 TRU 원소 및 RE 원소의 조성 및 열역학적 자료를 검토하였으며, UCl3와의 산화 반응에 따른 평형 자료를 계산하여 공정 가능성을 검토하였다. 실제 파이로 프로세싱 처리를 가정한 물질수지로부터 TRU 생성물의 RE 함량이 다른 두 가지 경우에 대하여 RE 원소 제거율과 TRU 회수율을 평가하였다. TRU 생성물을 산화제인 UCl3와 반응시켰을 때 각 원소의 Gibbs free energy의 차이에 의한 선택적 산화 반응이 일어났다. 투입된 UCl3 양을 조절하여 TRU 회수율을 최대로 유지하면서도 RE 원소를 제거하여 최종생성물의 TRU/ RE 비를 증가시킬 수 있는 가능성을 계산 결과로 확인하였다. 본 연구의 결과들은 열역학적 평형 자료에 기반한 결과이므로 실제 공정에 적용하기에는 많은 차이점이 존재한다. 그러나 TRU 물질을 취급하기 어려운 환경에서 파이로프로세싱의 TRU 생성물을 고속로의 핵연료로 공급하기 위한 공정 설계에 중요한 자료로 활용될 것으로 기대된다.