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 rod extractor and cutter, the major requirements were considered, and the modularization design was carried out considering remote operation and maintenance. In order to design the rod extractor and cutter, these systems were analyzed and designed, also the concept on the rod extraction and cutting were considered by using the solid works tool. The main module consists of five sub-modules, and the function of each is as follows. The clamping module is an assembly fixing module using a cylinder so that the nuclear fuel assembly can be fixed after being placed. The Pusher module pushes the fuel rods by 2 inches out of the assembly to grip the fuel rods. The extraction module extracts the fuel rods of the nuclear fuel assembly and moves them to the consolidation module. The consolidation module collects and consolidates the extracted fuel rods before moving them to the cutting device. And the support module is a base platform on which the modules of the main device can be placed. The modules of level 2 can be disassembled or assembled freely without mutual interference. For the design of fuel rods cutter, the following main requirements were considered. The fuel rod cut section should not be deformed for subsequent processing, and the horizontally mounted fuel rods must be cut at regular intervals. The cutter should have the provision for aligning with the fuel rod, and the feeder and transport clamp should be designed to transfer the fuel rods to the cutting area. The main module consists of 6 sub-modules, and function of each is as follows. The cutting module is a device that cuts the fuel rods to the appropriate depth for notching. The impacting module is a device that impacts the fuel rods and moves them to the collection module. The transfer module is a device that moves the fuel rods to the cutting module when the aligned fuel rods enter the clamp module. The clamping module is a device to clamp the fuel rods before moving them to the cutting module. The collection module is a container where the rod-cuts are collected, and the support module is a base platform on which the modules of the main device can be placed. The module of level 3 can be disassembled or assembled after the cutting module of level 2 is installed, and the modules of level 2 can be disassembled or assembled freely without mutual interference.

Korea Atomic Energy Research Institute (KAERI) has been operating the Post Irradiation Examination Facility (PIEF) for spent fuel. The facility has pools and hot cells for handling and examining fuel assemblies and rods. In the first hot cell, non-destructive tests such as visual inspection, defect detection, oxide layer thickness measurement, and gamma scanning are performed on a full-length fuel rod. Then, the fuel rod is transported to the next hot cell for measuring the rod internal pressure (RIP). After the RIP measurement, the fuel rod is cut by a cutting machine to make samples for destructive tests. Currently, the existing cutting machine is broken, so a new machine needed to be designed and manufactured. The major considerations for designing the cutting machine were convenience of remote handling and decontamination. The machine was modularized and its handling parts were designed to be easily controlled by manipulators. The cover was designed to prevent radioactive contamination of the surrounding area.

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.

Currently, as the saturation capacity of wet storage pool for spent nuclear fuel (SNF) of PWR in Korea has reached approximately 75%, Dry Storage Facilities (DSF) are necessary for sustainable operation of nuclear power plants. It is necessary to develop acceptance requirements for the delivery of SNF from reactor storage site to Centralized DSF. To do this end, the mechanical integrity of SNF is directly related to its repacking, retrieving, and transporting/handling performances. And also, this integrity is a key factor associated with the criticality safety that is connected to the damaged status of SNF. According to the NUREG/CR-6835, the NRC expects that the potential for nuclear fuel failures will increase because of the increase of the fuel discharge burnup and the degradation of fuel and clad material properties. Due to such damages and/or degradation, the fuel rods in the fuel assembly may be extracted and empty for following treatments (transportation, storage, handling etc). This condition can have a detrimental effect on the criticality safety of SNF. Thus, this study investigated whether extracted and empty of damaged SNF rod affects criticality safety. In this analysis, it is assumed that up to four fuel rods are missed. As a result of the analysis, As the number of fuel rods miss up to a certain number, the value of multiplication factor value of the fuel assembly increases. In addition, since the fuel rods located at the outermost layer contained relatively less fissile material than the fuel rods located center of the lattice, and neutrons were lost by the absorption material, the effective multiplication factor value gradually decreased. Nevertheless, the criticality safety was assessed to be maintained.

A rod internal pressure increased by fission gas release is major factor that causes degradation during dry storage of spent fuel. Because rod internal pressure is greatly affected by fuel design, operation power history, it is essential to perform complex calculation using performance code to accurately predict rod internal pressure as function of burnup. However, because it is difficult to apply a complex method into dry storage design and to determine rod internal pressure based on conservative way this study presents a simple correlation that can predict an approximate rod internal pressure as function of burnup For the development of simple correlation, rod internal pressure and fuel rod void volume data measured through about 400 PIE (Post Irradiation Examination) data were used. The developed simple correlation can cover various fuel rod arrays, discharged fuel average burnup, operation history, cladding type, burnup range, and information on Westinghouse type fuel rods such as Spain ENUSA, USA EPRI/ANL/ORNL/PNNL, WEC, etc. In this paper, the data of simple correlation determination is briefly introduced, and the data analysis process and results are summarized. Two correlations that can conservatively determine rod internal pressure and free void volume in fuel rod according to fuel rod average burnup were presented, and the effect of initial He fill pressure was evaluated. In particular, the results of Post Irradiation Examination for 46 fuel rods conducted in Korea are also included, so it is expected that newly presented correlations can be used easily in various ways in the domestic research, industry, and academia.

For economic and safe management of Spent Nuclear Fuel (SNF), it is very important to maintain the structural integrity of SNF and to keep the fuel undamaged and handleable. The cladding surrounding nuclear fuel must be protected from physical and mechanical deterioration. The structural evaluation of SNF is very complicated and numerically demanding and it is essential to develop a simplified model for the fuel rod. In this study, a simplified model was developed using a new cladding failure criterion. The simplified model was developed considering only the horizontal or lateral static load utilizing the cladding material properties of irradiated Zirclaoy-4, and applicability in horizontal and vertical drop impacts was investigated. When a fuel rod is subject to bending, a very complicated 3D stress state is generated within the vicinity of the pellet–pellet interface. A very localized stress concentration is observed in the area where the edges of the pellets contact the cladding. If the failure strain criteria obtained from the uniaxial tension test or biaxial tube test is applied, failure is predicted at the beginning stage of loading with premature through-thickness stress or strain development. The localized contact stress or strain is self-limiting and is not a good candidate for the cladding failure criteria. In this work, a new cladding failure criterion is proposed, which can account for the localized stress concentration and the through-thickness stress development. The failure of the cladding is determined by the membrane plus bending stress generated through the thickness of the cladding, which can be calculated by a process called stress linearization along the stress classification line. The failure criterion for SNF was selected as the membrane plus bending stress through stress linearization in the cross-sections through the thickness of the cladding. Because the stress concentration in the cladding around the vicinity of the pellet–pellet interface cannot be simulated in a simplified beam model, a stress correction factor is derived through a comparison of the simplified model and detailed model. The applicability of the developed simplified model is checked through horizontal and vertical drop impact simulations. It is shown that the stress correction factor derived considering static bending loading can be effectively applied to the dynamic impact analyses in both horizontal and vertical orientations.

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.