This study reassess safety margin of the current Peak Cladding Temperature (PCT) limit of dry storage in terms of hydrogen migration by predicting axial hydrogen diffusion throughout dry storage with respect to wet storage time and average burnup. Applying the hydride nucleation, growth, and dissolution model, an axial finite difference method code for thermal diffusion of hydrogen in zirconium alloy was developed and validated against past experiments. The developed model has been implemented in GIFT – a nuclear fuel analysis code developed by Seoul National University. Various discharge burnups and wet storage time relevant to spent fuel characteristics of Korea were simulated. The result shows that that the amount of hydrogen migrated towards the axial end during dry storage for reference PWR spent fuel is limited to ~50 wppm. This result demonstrates that the current PCT margin is sufficient in terms of hydrogen migration.

Since SMR’s reduced reactor radius results in higher neutron leakage, SMR operates at a relatively lower discharge burnup level than traditional Light Water Reactors (LWRs). It may result in larger spent fuel amounts for SMRs. Furthermore, recent studies demonstrated that NuScale reactor will generate a significantly higher volume of low- and intermediate-level waste owing to components located near the active core including the core barrel and the neutron reflector. For spent nuclear fuel simulation, FRAPCON-4.0 was updated. Major modifications were made for fission and decay gas release, pellet swelling, cladding creep, axial temperature distribution, corrosion, and extended simulation time covering from steady-state to dry storage. In this study, typical 17×17 PWR fuel (60 MWd/kgU) and NuScale Power Module (36 MWd/kgU) was compared. NuFuel-HTP2™ fuel assembly, which has a half-length of proven LWR fuel, was employed. Owing to the lower discharge burnup and operating temperature, the maximum hydrogen pickup was 73 wppm and the maximum hoop stress was ~25 MPa. Therefore, hydride reorientation issue is irrelevant to SMR spent fuel. In this context, the current regulatory limit for dry storage (i.e. 400°C and 90 MPa) can be significantly alleviated for LWR-based SMRs. The increased safety margin for SMR spent fuel may compensate high spent fuel management cost of SMRs incurred by an increased amount. The comprehensive analysis on SMR spent fuel management implications are discussed based on simulated SMR fuel characteristics.

As the saturation rate of temporary storage facilities for spent nuclear fuel increases, regulatory demands such as interim storage and permanent disposal of spent nuclear fuel are expected to begin in earnest. Considering the domestic situation where all nuclear power plants are located on the waterfront site, the interim storage site is also likely to be located on the waterfront site, and maritime transportation is one of the essential management stages. Currently, there are no independently developed maritime transportation risk assessment code in Korea, and no research has been conducted to evaluate the release of radioactive waste due to the sinking of transport container. Therefore, it is necessary to secure technology to properly reflect the domestic maritime transportation environment and to assess the impact of the sinking accident and to carry out safety regulations. To accurately calculate the releaser rate of radionuclides contained in a cask with breached containment boundary, the flow rate through the gap generated in the containment boundary should be calculated. The fluid flow through this gap which is probably in micro scale in most situations should be evaluated combining the fluid flow inside and outside the cask. In this study, a detailed computational fluid dynamics model to evaluate the internal fluid flow in the cask and a simplified model to capture the fluid flow and the heat transfer around the cask in the sea are constructed. The results for the large scale model are compared with the analytic formula for verification of heat transfer coefficient and they showed good agreements. The heat transfer coefficient thus found can be used in the detailed model to provide more realistic data than those obtained from assumed heat transfer coefficient around the surface of the cask. In the future, fluid flow through the gap between the lid and the body of the cask will be evaluated coupling the models developed in this work.

B4C/Al composite is mainly used for neutron absorbing materials, which is one of the components of equipment that manages spent nuclear fuel. There are various processes for manufacturing neutron absorbing materials, but most of them are based on the powder metallurgy. In this study, B4C/Al composite in which the reinforcement was uniformly dispersed was manufactured by using the stir casting process. The microstructure, thermal neutron absorption rate, mechanical properties and dispersibility of the reinforcement of the prepared B4C/Al composite were analyzed.

Hydride reorientation is one of the major concerns for cladding integrity during dry storage. In this study, mechanical property of post-reorientation cladding was investigated according to the morphology and amount of the hydrides. Cladding peak temperature limit 400°C was suggested by U.S. NRC in concern of cladding creep and hydride reorientation. In line with this regulatory limit, hydride reorientation was conducted during cool-down process from the maximum temperature of 400°C, using constant internal pressurization method. The specimens were charged for hydrogen from 100 to 1,000 wppm, and various pressures range of 7.5-18.5 MPa were applied. The morphology was examined by optical microscopy. Radial hydride fraction (RHF) and radial hydride continuous path (RHCP) were calculated using image analysis software PROPHET. Finally, strain energy density (SED) was investigated via ring compress tests and the hydrogen concentration was analyzed. The result shows that when RHF is higher than 5%, SED exponentially decreases with RHF. For RHF less than 5%, SED was primarily affected by the total amount of hydrogen. Shortened length of radial hydrides with the presence of circumferential hydrides may block the radial propagation of crack. The result implies that lower burnup spent fuel with lower hydrogen concentration may be more vulnerable in terms of radial hydride compared to higher burnup fuel.

Separation of high heat generating-radioactive isotopes from spent nuclear fuel is an important issue because it can reduce the final disposal area. As one of the technologies that can selectively separate only high heat generating-radioactive isotopes without dissolving spent fuel, the methods using molten salt have recently attracted attention. Although studies on chemical changes of Sr oxides in molten salts have been reported, they have limitation in that alternative oxide reagents rather than oxide fuel were used. In this study, the separation behaviors of Sr from simulated oxide fuel using various molten salts were investigated. A powder type containing 95.7wt% of U and 0.123wt% of Sr was used as the simulated oxide fuel. LiCl, LiCl-CaCl2, MgCl2, LiCl-KCl-MgCl2 and NaCl-MgCl2 were used as molten chloride salts. The separation of Sr from the simulated oxide fuel was conducted by loading it in porous alumina basket and immersing it in a salt. The concentration of Sr in the salt was measured by ICP analysis after sampling the salt outside the basket using dip-stick technique. The separation efficiencies of Sr from simulated oxide fuel using the salts were compared. Furthermore, the causes of their separation efficiency were systematically investigated.

The guidelines for cyber security regulations at domestic and foreign nuclear facilities, such as KINAC/RS-015, NRC’s RG5.71 and NEI 13-10, require the establishment of security measures to maintain the integrity of critical digital assets (CDAs) and protect them as threats to the supply process. According to the requirements, cyber security requirements shall be reflected in purchase requirements from the time of introduction of CDAs, and it shall also be verified whether cyber security security measures were properly applied before introduction. Domestic licensees apply measures to control the supply chain in the nuclear safety sector to cyber security policies. The safety sector supply chain control policy has areas that functionally overlap with the requirements of cyber security regulations, so regulatory guidelines in the safety sector can be applied. However, since most of the emergency preparedness and physical protection functions introduce digital commercial products, there is a limit to applying the control of the supply chain in the safety field as it is. It is necessary to apply supply chain control operator policies, procedures, and purchase requirements for each SSEP function, or to establish cyber security integrated supply chain control requirements. In this paper, based on the licensee’s current supply chain control policy, the cyber security regulation plan for supply chain control according to the SSEP (Safety-Security-Emergency Preparedness) function of CDAs is considered.