Various types of spent fuel assembly in nuclear power plants have been transported to a post irradiation examination facility (PIEF) in KAERI to examine the mechanical and chemical properties of fuel and cladding. Once the fuel assembly arrive at PIEF, it is dismantled in a pool area to extract the fuel rods. Dismantling of the fuel assembly is performed by cutting the top nozzle. Currently, couple of dismantled assemblies have been stored in a storage pool without the top nozzle in PIEF. These assemblies cannot be handled directly using a gantry crane in the pool, and thus are contained in a special basket to handle. In this research, we developed a restoration method for a dismantled spent fuel assembly, especially for 16×16 Korea Optimized Fuel Assembly (KOFA). After reviewing the original design document and reports of KOFA, two tools are devised; an assembly tool and a tightening tool for a bolt. Since the top nozzle and dismantled KOFA can be re-assembled using a bolt, we follow the original design, size, and materials of the previously used bolt. The bolt to restore the top nozzle of KOFA is made of 321 stainless steel and has a design that fits the guideline of DIN 13-21 international standard. Our procedure can potentially be used to restore and repair the dismantled spent fuel assembly.

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.

This study investigates the behavior of the thermal conductivity among material properties in order to develop a thermal evaluation methodology of spent fuel assembles in a transport cask. It is inefficient to model each element of the spent fuel assembly in detail, and it is generally calculated by modeling the effective thermal conductivity (ETC). The ETC model was developed to allow a much simpler representation of a spent fuel assembly within a fuel compartment by treating the entire spent fuel rod array and the surrounding fill gas within the confines of the compartment as a homogenous solid material. The fuel rod assembly and surrounding gas are modeled with an effective conductivity that is designed to yield an overall conduction heat transfer rate that is equivalent to the combined effect of local conduction and radiation heat transfer in a plane through the assembly. When this model is applied to the transport cask, it tends to predict the cladding peak temperature lower than the results of detailed model in which the fuel rod arrangement and shape of the fuel assembly are simulated. As for the tendency of the error, the model tended to under-predict when basket temperature was lower than a certain temperature, and over-predict when it was higher. The purpose of this study is to investigate the attenuation effect of the cladding peak temperature on the related variables when the ETC model is applied to the transport cask. In addition, based on the thermal characteristics of this model, a correction factor that can compensate for this attenuation effect is presented. This correction factor is obtained by finding the difference between a separate ETC homogeneous model and a separate detailed fuel model, rather than directly applying the ETC calculated from the detailed fuel model to the transport cask.

Spent nuclear fuels released from the reactor are stored in cooling pools and then stored in dry storage casks. During the transition from the wet storage to dry storage cask, a vacuum drying process is used to remove residual water in the cask. During the vacuum drying process, gas pressure is reduced to below 400 Pa to promote evaporation and water removal. KAERI is developing a PWR single assembly (PLUS7) test equipment to simulate the thermal flow in spent fuel assembly. In this study, the thermal conductivity of air at low pressure was derived to perform the thermal analysis of the canister in vacuum. In addition, thermal analyses were performed for the canister with backfill gases of helium, air, and a vacuum in the vertical orientation using the COBRA-SFS code. At low pressure, the thermal conductivity of air depends on pressure and temperature. The reduced thermal conductivity, kr (W/m-K) was calculated using the curve fit for air at reduced pressure in thin gaps presented in the General Electric Fluid Flow Handbook. 􀝇􀯥/􀝇􀬴 = 􀬵 􀬵􀬾 􀮼􀯍/􀯉􀰋 Where, k0 is the thermal conductivity at atmospheric pressure (W/m-K), P is the reduced (vacuum) pressure (Pa), δ is the gap size (m), T is the temperature (K), and C is the Lasance constant (7.657E-5 N/m-K). The thermal conductivity of air decreases as the pressure decreases. The reduced thermal conductivity of air at pressures of 400 Pa and 40 Pa was calculated to be 0.97 and 0.77, respectively. For the analysis in vacuum, no enhancement of the convective heat transfer was assumed (Nu=1.0). For the helium backfill, the peak cladding temperature was the lowest and the axial temperature profile was the flattest due to the higher thermal conductivity and lower density of the helium. For the vacuum backfill, the peak cladding temperature was the highest and temperature gradient was the sharpest due to the only radiative heat transfer effect in the fuel assembly.

Dry head end process is developing for pyro-processing at KAERI (Korea Atomic Energy Research Institute). Dry processes, which include disassembling, mechanical decladding, vol-oxidation, blending, compaction, and sintering shall be performed in advance as the head-end process of pyro-processing. Also, for the operation of the head-end process, the design of the connecting systems between the down ender and the dismantling process is required. The disassembling process includes apparatus for down ender, dismantling of the SF (Spent Fuel) assembly (16×16 PWR), rod extraction, and cutting of extracted spent fuel rods. The disassembling process has four-unit apparatus, which comprises of a down ender that brings the assembly from a vertical position to a horizontal position, a dismantler to remove the upper and bottom nozzles of the spent fuel assembly, an extractor to extract the spent fuel rods from the assembly, and a cutter to cut the extracted spent fuel rods as a final step to transfer the rod-cuts to the mechanical decladding process. 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 down ender and dismantler, these systems were analyzed and designed, also concept on the interference tools between down ender and dismantler were considered by using the solid works tool.

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).

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.

This paper mainly focuses on the maximum decay heat estimation generated from spent fuel assemblies in the spent fuel pool of Kori units 3&4 at the beginning decommissioning. It is assumed that the spent fuel pool is fully occupied with 2,260 spent fuel assemblies, same as its design capacity. In addition, equally 56.5 spent fuel assemblies have been generated per year. The minimum cooling time is five years considering the transition phase between the permanent shutdown and the amendment of Operating License for decommissioning. Sending and receiving of spent fuel assemblies to/from other units are neglected. Seven representative spent fuel assembly groups are established based on the burnup rate and cooling time. Conservatively high values for the burnup rates and low values for the cooling times are applied. Calculation of the decay heat of each representative group has been performed by using ORIGEN decay solver of SCALE. Then, total decay heat has been calculated based on this. Group 1, 2, and 3 contain comparatively old spent fuel assemblies with 45 GWd/tU burnup rate and 20~30 cooling years. The calculation shows 489~586 watts of decay heat per assembly. Group 4, 5, 6, and 7 contain comparatively new spent fuel assemblies with 55 GWd/tU burnup rate and 5~20 cooling years. The calculation shows 741~1,483 watts of decay heat per assembly. The total maximum decay heat therefore is estimated as 1,609,459 watts.