Noble metal phase, present in used fuel, are fission products that can be found as metallic precipitates in used nuclear fuel. They exist as small particles (nm~um) in grain boundaries of the used fuels. Since they are particles deposited between the grain structures, they can be considered as defects in the pellet structure. Thermal expansion of fuels with noble metal is slightly higher than that of bare fuels. The fuels at high temperature, such as immediately after being discharged from nuclear reactors, may be subject to fuel failure if sufficient cooling is not provided. Recent research has shown that the noble metals can migrate into the rim space between the pellet and the cladding, and be deposited in the inner layer of the claddings. therefore, the mechanical integrity of the cladding can be degraded by noble metals, as well as the pellets. The concentration of the noble metal phase should be considered to evaluate the effect of the noble metals on the fuel integrity, after discharge from the reactors. SCALE/ORIGEN code was used to evaluate the noble metals in fuel assembly-scale, and the radial distribution in the fuel assembly. The radial distribution of the reactor power was derived from the SCALE/TRITON, considering Westinghouse 17×17. Square cell model was chosen for the geometry and 1/4 model was applied to reduce the computation time.

A comparison and validation between the analysis and vibration test data of a nuclear fuel assembly were conducted. During the comparison and validation process, various parameters that govern the vibration behavior of the fuel assembly were determined, including nuclear fuel rod’s stiffness, spring constants of the dimple and spring of support structures, and damping coefficients. The calibration of the vibration analysis model aimed to find analysis parameters that can accurately simulate the vibration behavior of the test data. For calibration, power spectral density (PSD) diagrams were generated for both the measured signals from the test and the calculated signals from the analysis. The correlation coefficient between these two PSD plots was calculated. To find the analysis parameters, each parameter was defined as a variable with an appropriate range. Latin hypercube sampling was used to generate multiple sample points in the variable space. Analysis was performed for the generated sample points, and PSD plot correlation coefficients were calculated. Using the generated sample points and their corresponding results, a Gaussian Process Regression model was implemented for PSD plot correlation coefficients and the maximum PSD value. Based on the constructed surrogate model, the optimal analysis parameters were easily found without additional computations. Through this method, it was confirmed that the analysis model using the optimal parametes appropriately simulates the vibration behavior of the test.

Nuclear fuel assemblies are exposed to high temperature and high pressure environments underwater for long periods of time in a reactor, leading to deterioration of the assembly structure. These assembly consists of fuel rods, grids, a top nozzle, a bottom nozzle and guide tubes. In particular, the integrity of the guide tube made of Zircaloy-4 is a very important part in handling the assembly. In the Post Irradiation Examination Facility (PIEF), there are 14×14 Westinghouse STD assemblies that have lost their handleability due to the top nozzle being removed for damaged fuel rod test. To handle these assemblies, it is reasonable to use cut guide tubes whenever possible. Therefore, it is necessary to determine the irradiation embrittlement state of the guide tube before designing or manufacturing parts that can connect the top nozzle and the guide tubes. Therefore, in this paper, the location for installing the top nozzle-guide tube connection parts was selected in the height range of 3,460 to 3,713 mm, and guide tube specimens were made within that range. Offset strain was derived from the load-displacement curve obtained through compression testing to confirm whether the ductility of guide tubes was maintained. As a result, there was no significant difference in strength and ductility of the guide tube within the above length range. In addition, it was confirmed that the ductility was maintained enough to install the top nozzle-guide tube connection parts. Therefore, it is judged that there will be no problem even if the top nozzle-guide tube connection parts are installed in the guide tube to restore the handleability of the assemblies.

Post Irradiation Examination Facility (PIEF) is a test facility for nuclear fuel research and development and performance evaluation. From the past to the present, assemblies and fuel rods have been transported from nuclear power plants (NPP) several times, and various destructive and non-destructive tests have been performed. Among these, in the case of the 14×14 Westinghouse STD assemblies that are transported as a whole assembly, the top nozzle is connected to the guide tube by welding. Therefore, the fuel rods could not be removed from the assembly at the NPP, so the assemblies were transported to PIEF as is. Then, after cutting between the top nozzle and the guide tube in the pool, and the fuel rods were extracted and tested. In order to transport the assembly in the future, it is necessary to maintain stability by inserting the dummy rod into the unit cell from which the fuel rod is extracted. However, since the length of the dummy rod is almost 4 m and the diameter is about 10 mm, the dummy rod often bends while passing through the dimple spring of the grid. Additionally, when dummy rods are inserted into unit cells that are continuously empty after the fuel rods are extracted, there may be cases where the dummy rods are not inserted into the desired unit cell but are bent and incorrectly inserted into the next unit cell. The moment the dummy rods are inserted into the dimple spring of grid, a load is applied to the dummy rod due to the tension of the spring. If it can be inserted while offsetting the load, the work can be performed more smoothly. Accordingly, an underwater handling tool was developed that can be inserted while offsetting the tension of the spring. Using this handling tool applies a load to the dummy rod and rotates the dummy rod itself, offsetting the tension of the spring and allowing the dummy rod to be inserted without bending. This handling tool is equipped with a shock absorbing device to protect the dummy rod and spring, and a module to rotate the dummy rod. As a result of inserting the dummy rod using the developed handling tool, it was possible to easily insert the dummy rod into unit cells that were previously impossible to insert.

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.

The thermal integrity of spent nuclear fuels has to be maintained during their long-term dry storage. The detailed temperature distributions of spent fuel assemblies are essential for evaluating the integrity of their dry storage systems. In this study, a subchannel analysis model was developed for a canister of a single fuel assembly using the COBRA-SFS code. The thermal parameters affecting the peak cladding temperature (PCT) of the spent fuel assembly were identified, and sensitivity analyses were performed based on these parameters. The subchannel analysis results indicated the presence of a recirculation flow, based on natural convection, between the fuel assembly and downcomer region. The sensitivity analysis of the thermal parameters indicated that the PCT was affected by the emissivity of the fuel cladding and basket, convective heat transfer coefficient, and thermal conductivity of the fluid. However, the effects of the wall friction factor of the canister, form loss coefficient of the grid spacers, and thermal conductivities of the solid materials, on the PCT were predominantly ignored.

To ensure radiological safety margin in the transport and storage of spent nuclear fuel, it is crucial to perform source term and shielding analyses in advance from the perspective of conservation. When performing source term analysis on UO2 fuel, which is mostly used in commercial nuclear power plants, uranium and oxygen are basically considered to be the initial materials of the new fuel. However, the presence of impurities in the fuel and structural materials of the fuel assembly may influence the source term and shielding analyses. The impurities could be radioactive materials or the stable materials that are activated by irradiation during reactor power operation. As measuring the impurity concentration levels in the fuel and structural materials can be challenging, publicly available information on impurity concentration levels is used as a reference in this evaluation. To assess the effect of impurities, the results of the source term and shielding analyses were compared depending on whether the assumed impurity concentration is considered. For the shielding analysis, generic cask design data developed by KEPCO-E&C was utilized.

Most of the spent nuclear fuel generated by domestic nuclear power plants (NPPs) is temporarily stored in wet storage which is spent fuel pool (SFP) at each site. Currently, in case of Kori Unit 2, about 93.6% of spent nuclear fuel is stored in SFP. Without clear disposal policy determined for spent nuclear fuel, the storage capacity in each nuclear power plant is expected to reach saturation within 2030. Currently, the SFP stores not only spent fuel but also various non-fuel assembly (NFA). NFA apply to all device and structures except for fuel rods inserted in nuclear fuel assembly. The representative NFA is control element driving mechanism (CEDM), in-core instrument (ICI), burnable poison, and neutral resources. Although these components are irradiated in the reactor, they do not emit high-temperature heat and high radiation like nuclear fuel, so if they are classified as intermediate level waste (ILW) and low level waste (LLW) and moved outside the SFP, positive effects such as securing spent fuel storage space and delaying saturation points can be obtained. Therefore, this study analyzes the status of spent fuel and Non Fuel Assembly (NFA) storage in SFP of domestic nuclear power plants. In addition, this study predict the amount of spent fuel and NFA that occur in the future. For example, this study predicts the percentage of current and future ICIs and control rods in the SFP when stored in the spent fuel storage rack. In addition, the positive effects of moving NFA outside the SFP is analyzed. In addition, NFA withdrawn from SFP is classified as ILW & LLW according to the classification criteria, and the treatment, storage, and disposal methods of NFA will be considered. The study on the treatment, storage, and disposal methods of NFA is planned to be conducted by applying the existing KN-12 & KN-18 containers and ILW & LLW containers being developed for decommissioning waste.

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).

In the design of a spent-fuel (SF) storage, the consideration of burnup credit brings the benefits in safety and economic views. According to it, various SF burnup measurement systems have been developed to estimate high fidelity burnup credit, such as FORK and SMOPY. Recently, there are a few attempts to localize the SF burnup measurement system in South Korea. For the localization of SF burnup measurement systems, it is very important to build the isotope inventory data base (DB) of various kinds of SFs. In this study, we performed DeCART2D/MASTER core follow calculations and McCARD single fuel assembly (FA) burnup analyses for Hanbit unit 3 and confirmed the characteristic of the isotope inventory over burnup. Firstly, the core follow calculations for Cycles 1~7 were performed using DeCART2D/MASTER code system. The core follow calculation is very realistic and practical because it considers the design conditions from its nuclear design report (NDR). Secondly, the Monte Carlo burnup analyses for single FAs were conducted by the McCARD Monte Carlo (MC) transport code. The McCARD code can utilize continuous energy cross section library and treat complex geometric information for particle transport simulation. Accordingly, the McCARD code can provide accurate solutions for burnup analyses without approximations, but it needs huge computing resources and time burden to perform whole-core follow calculations. Therefore, we will confirm the effectiveness of the single McCARD FA burnup analyses by comparing the DeCART2D/MASTER core follow results with the McCARD solution. From the results, the use of single FA burnup analyses for the establishment of the DBs will be justified. Various FAs, that have different 235U enrichments and loading pattern of fuel rods and burnable absorbers, were considered for the burnup analyses. In addition, the results of the sensitivity analyses for power density, initial enrichment, and cooling time will be presented.

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.

Detailed temperature distributions of the spent fuel are required to evaluate the long-term integrity of the dry storage system. In this study, a subchannel analysis method was established to obtain the detailed temperatures of a spent fuel using the COBRA-SFS code. The SAHTT (Single Assembly Heat Transfer Test) model was selected as the subchannel analysis. It was developed at the PNL to investigate heat transfer characteristics of spent PWR fuel under dry storage conditions. The SAHTT has a 15×15 rod array with simulated rods 0.42 in. (10.7 mm) in diameter. Control rod thimbles were modeled with unheated rods. The COBRA-SFS input consists a detailed subchannel model with 256 subchannels, 225 rods, and 8 slab nodes. The heat generation rate was axially uniform with total power of 1.0 kW. Subchannel analyses were performed for the vertical orientation under three different backfills of air, helium, and vacuum. For the vacuum backfill, the peak temperature was the highest and temperature gradients the sharpest only due to the radiation heat transfer effect. For the helium backfill, peak temperature was lowest and the axial profiles flattest due to the higher conductivity and lower density of helium. Subchannel analyses were also performed to evaluate the effect of thermal parameters such as surface emissivity, convective heat transfer coefficients, and flow resistance coefficients on the PCT (Peak Cladding Temperature). The PCT was affected by the emissivity of the fuel rod and the basket, and in particular, the basket emissivity had a greater effect. The PCT was affected by the Nusselt number, but the range of the Nusselt number is around 3.66. Therefore, the effect of the Nusselt number on the PCT will not be significant. As a result of the analysis according to the flow resistance coefficients, the PCT was affected by the wall friction factor, but the loss coefficients from the space grid had little effect. Subchannel technique obtained from this work can be used to predict the detailed temperature distributions of spent fuel assembly.

The temperature of the spent fuel cladding is the basis for the evaluation of integrity. It is almost impossible to directly measure the temperature of spent nuclear fuel. Because spent nuclear fuel is dangerous. We are preparing a test to measure the cladding temperature with an equivalent fuel assembly by simulating the characteristics of spent nuclear fuel. PLUS7 was selected as the test target in consideration of the amount of generation, thermal water retention, residual moisture content, and manufacturability of domestic spent nuclear fuel. The nuclear fuel assembly is planned to be manufactured in two main ways. Except for the cladding tube that simulates decay heat, the structure will be manufactured by KEPCO Nuclear Fuel, and fuel rods and canisters will be manufactured by SUKEGAWA Electric in Japan. The same nuclear fuel assemblies as commercial skeleton will be applied. The temperature of the fuel cladding will be measured by attaching a thermocouple directly to the surface of the cladding tube. The canister is composed of a basket, a basket supporter, a heater and drain tube, a lead, and an observation window. The working fluid is water and helium, and the maximum pressure inside the canister is 1.1 MPa and the minimum pressure is 0.05 kPa. The maximum temperature of the surface of the cladding was designed to be 500°C, and the maximum temperature of the sealing to keep airtightness was designed to be 250°C. To satisfy this condition, we plan to evaluate the leak rate below 10−5 std.cm3·s−1, which is equivalent to helium tightness. The maximum heat of decay per fuel rod is 13 W, and one assembly is up to 3 kW. Production of the test equipment is expected to be completed in the first half of next year, and testing is scheduled to begin in the second half of next year. The test will evaluate all environments that the spent nuclear fuel may experience, such as dry normal conditions, abnormal conditions, wet conditions, and dry conditions. All data will be used for interpretation verification purposes.