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.

Based on the results of a review for various precipitation methods phosphorylation (phosphate precipitation) of metal chlorides considered as a proper treatment method for recovering of the fission products in a molten salt. In previous precipitation tests, the powder of lithium phosphate (Li3PO4) added into LiCl-KCl molten salt containing metal chlorides as a precipitation agent. The reaction of metal chlorides containing actinides and rare earths to recover with lithium phosphate in a molten salt known as solid-liquid reaction. The powder of lithium phosphate disperse in a molten salt by stirring thoroughly in order to enhance the precipitation reaction. As a result, metal phosphates as the reaction products precipitate on the bottom of the vessel and cutting at the lower part of the salt ingot considered as one of the recovery method of the precipitates. Recently, the vacuum distillation of upper part of the salt proposed as another recovering method. Cutting method of precipitate at the lower part of the salt ingot would be difficult to handle the increased size of the salt ingot produced from the practical scale equipment. In this presentation, a new method for collecting the precipitates of phosphorylation reaction into a small vessel is introduced with test results in a molten salt containing uranium and rare earths such as Nd, Ce, and La. As the first step of a series of test lithium phosphate ingot was prepared by melting the powder at a temperature 1,300°C, and the ingot put into LiCl-KCl molten salt at 500°C for more than three hours to examine the shape of ingot to be deformed or not. The phosphorylation experiments using lithium phosphate ingots carried out to collect the metal phosphate precipitates and the test result of this new method was feasible.

The Korea Atomic Energy Research Institute is developing a nuclide management process that separates high heat, high mobility, and long half-life nuclides that burden the disposal of spent fuel, and disposes of spent fuel by nuclide according to the characteristics of each nuclide. Various offgases (volatile and semi-volatile nuclides) generated in this process must be discharged to the atmosphere below the emission standard, so an off-gas trapping system is required. In this study, we introduce the analysis results of the parameters that affect the design of the off-gas trapping system. The analyzed contents are as follows. The physical quantities of the Cs, Tc/se, and I trapping filters according to the amount of spent nuclear fuel, the maximum exothermic temperature of the Cs trapping filter and the absorbed dose by distance by Cs radioactivity were analyzed according to the amount of spent nuclear fuel. In addition, a three-dimensional CFD (Computational Fluid Dynamics) analysis was performed according to operating parameters by simply modeling the off-gas trapping system, which is easy to modify mechanical design parameters. It is considered that the analysis results will greatly contribute to the development of the off-gas trapping system design requirements.

This study is to investigate fuel cladding temperature in a transport system for the purpose of developing a methodology for evaluating the thermal performance of spent fuel. Detailed temperature analysis in the transport system is important because the degradation mechanism of the fuel cladding is generally sensitive to temperature and temperature history. In such a system, the magnitude of the temperature change is determined by examining the temperature sensitivity of fuel assemblies and system components including fuel cladding temperature, considering the material properties, component specifications, component aging mechanism, and heat transfer mechanism. The sensitivity analysis is performed using heat transfer models by computational fluid dynamics for the horizontal transport system. The heat transfer within the system by convection, conduction and thermal radiation is calculated by thermal-hydraulic analysis code FLUENT. The calculation region is divided into a basket cell and a transport cask. The thermal analysis of the basket cell is for predicting the fuel cladding temperature. And the reason for analyzing the transport cask is to provide the boundary condition for the basket cell by reflecting the external environmental conditions. Here, the basket cell containing the spent fuel assembly is modeled on the homogeneous effective thermal conductivity. The purpose of this analysis is to evaluate fuel cladding temperatures for the following four main items. That is the effect of surface emissivity changes in basket due to the oxide layer of the fuel cladding, the effect of degradation of the canister backfill helium gas, the effect of fuel assembly position in basket cell on fuel cladding and basket temperatures in canister, and the effect of using the homogeneous effective thermal conductivity model instead of the fuel assembly in basket cell. As a result of the analysis, the maximum temperatures in basket cells are evaluated for the above four items. Thermal margins for each item are investigated for thermal performance requirements (e.g., peak clad temperature below 400oC).

Due to the saturation of the on-site storage capacity of spent nuclear fuel within a few years, dry storage facility should be introduced. However, it is unclear when to start operating the dry storage facility, so in case of Kori Unit 1, which is being decommissioning, the spent fuel must be stored in the spent fuel pool of another power plant. In addition, in the case of damaged fuel, it is impossible to transfer and store it with general handling methods. Therefore, a damaged fuel canister (DFC) should be able to handle damaged or failed fuel as intact fuel, and both wet and dry storage should be possible. The canister developed by Korea Hydro & Nuclear Power is designed to satisfy criticality, shielding, cooling performance, and structural integrity in accordance with NUREG-1536 and 2215. In addition, it can be handled as existing fuel handling devices rather than new handling tools. Fastening of the DFC lid and body in the spent fuel pool is possible with a hexagonal socket wrench, one of the fuel repair tools. And it is designed to facilitate visual identification of whether it is fastenedor not. The lifting method for transferring DFC to another facility is the same as the nuclear fuel lifting method. And a unique sealing and mesh structure of the lid and body is devised to completely block leakage of nuclear fuel fragments of 0.2 mm or more during vacuum drying for dry storage. The usability of DFC has been verified through test operation of the prototype, and it will be manufactured before discharging spent fuel for the decommissioning of Kori Unit 1.

Interests in molten salt reactor (MSR) using a fast spectrum (FS) have been increased not only for having a high power density but for burning the high-level waste generated from nuclear power plants. For developing the FS-MSR technologies, chloride-based fuels are considered due to the advantage of higher solubility of actinides and lanthanides over fluoride-based salts. Despite significant progress in development of MSR technology, the manufacturing technology for production of the fuel is still insufficiently understood. One of the option to prepare the MSR fuel is to use products from pyroprocessing where oxide form of spent nuclear fuel is reduced into metal form and useful elements can be collected via electrochemical methods in molten salt system at high temperature. In order to chlorinate the products into chloride form, previous study used NH4Cl to chlorinate U metal into UCl3 in an airtight reactor. It was found that the U metal was completely chlorinated into chloride forms; however, impurities generated by the reaction of NH4Cl and reactor wall were found in the product. Therefore, in this work, the air tight reactor was re-deigned to avoid the reaction of reactor wall by insertion of Al2O3 crucible inside of the reactor. In addition, the reactor size was increased to produce UCl3 over 100 g. Using the newly designed reactor, U metal chlorination experiments using NH4Cl chlorinating agent were performed to confirm the optimal experimental conditions. The detailed results will be further discussed.

The effect of Li2O addition on precipitation behavior of uranium in LiCl-KCl-UCl3 has been investigated in this study. 99.99% LiCl-KCl eutectic salt is mixed with 10wt% UCl3 chips at 550°C in the Pyrex tube in argon atmosphere glove box, with 10 ppm O2 and 1 ppm H2O. Then, Li2O chunks are added in mixed LiCl-KCl-UCl3 and the system has been cooled down to room temperature for 10 hours to form enough UO2 particles in the salt. The solid salt has been taken out from the glove box, and cut into three sections (top, middle and bottom) by low-speed saw for further microscopic analysis. Three pieces of solid salt are dissolved in deionized water at room temperature and the solution is filtered by a filter paper to collect non-dissolved particles. The filter paper with particles is baked in vacuum oven at 120°C for 6 hours to evaporate remaining moisture from the filter paper. Further analysis was performed for the powder remaining on the filter paper, and periphery of the powder (cake) on the filter paper. Scanning electron microscopy (SEM), electron diffraction spectroscopy (EDS), and X-ray powder diffraction (XRD) are adopted to analysis the characteristic of the particles. From SEM analysis, the powders are consisted of small particles which have 5 to 10 m diameter, and EDS analysis shows they are likely UO2 with 23 at. % of uranium and 77 at. % oxygen. Cake is also analyzed by SEM and EDS, and needle like structures are widely observed on the particle. The length of needle is distributed from 5 to 20 m, and it has 6 to 10 at. % of chlorine, which are not fully dissolved into deionized water at room temperature. From XRD analysis, the particles show the peak position of UO2, and the result is well matched with the SEM-EDS results. We are planning to add more Li2O in the system for fully reacting uranium in UCl3, and compare the results to find the effect of Li2O concentration on UO2 precipitation.

The skeleton of fuel assembly is composed of top nozzle, bottom nozzle, grids, and guide tubes. In the reactor core, all the parts of the fuel assembly suffer degradations due to the condition of high temperature, pressure and water environment. Therefore, many material properties of high temperature mechanical strength, corrosion and irradiation resistance have been considered to choose the material for fuel assembly parts in the fuel development stage. The guide tubes have important roles to connect each parts and support the load of fuel assembly while the fuel is lifted. In Westinghouse 14×14 standard fuel assembly, Zircaloy-4 was used for the material of the guide tubes. Zircaloy-4 has a resistance to water corrosion and maintain good mechanical properties after the discharge from the core, so this alloy is also utilized for a fuel rod cladding material although the microstructure is slightly different due to the heat treatment difference. Thus, it is expected that there is no issue regarding the guide tube integrity after the discharge and during the storage in the pool, especially in case of low burn-up. However, the surface oxidation and resultant hydrogen pick-up can affect to the embrittlement to the Zr alloy. So, it is needed to know the actual status of spent fuel assembly by performing post-irradiation examination. In this study, the degradation level of the guide Tubes in low burn-up spent fuel assembly was investigated using the KAERI PIE facility in order to make some data which can be utilized to the baseline for evaluating the integrity of the spent fuel skeleton.

As the zircaloy cladding absorbs an excessive amount of hydrogen and cooled down under hoop stress, radial hydride may be precipitated by hydride reorientation phenomenon. There have been many previous studies about the threshold stress of the reorientation, but it is known that the quantitative degree of hydride reorientation rather than the threshold is important for the prediction of mechanical properties. A thermodynamic model for Radial Hydride Fraction (RHF) prediction has been developed in this study. The model calculates RHF with respect to temperature, cooling rate, hydrogen content, and applied stresses. Once the cooling rate is given, the solid solution concentration at each temperature is determined by Hydrogen-Nucleation-Growth-Dissolution model. Subsequently, the increment of radial hydride is derived by nucleation and growth theory. The code based on the thermodynamic theory can provide the prediction of RHF under hoop stress, as well as a change in precipitation behavior over time. RHF of the zircaloy cladding in long-term dry storage can be obtained by the implementation of the code and the degradation of the cladding is directly estimated according to the correlation between RHF and mechanical properties. Ongoing experimental validation of the developed model is discussed.