Hydride analysis is required to assess the mechanical integrity of spent nuclear fuel cladding. Image segmentation, which is a hydride analysis method, is a technique that can analyze the orientation and distribution of hydrides in cladding images of spent nuclear fuels. However, the segmentation results varied according to the image preprocessing. Inaccurate segmentation results can make hydride difficult to analyze. This study aims to analyze the segmentation performance of the Otsu algorithm according to the morphological operations of cladding images. Morphological operations were applied to four different cladding images, and segmentation performance was quantitatively compared using a histogram, betweenclass variance, and radial hydride fraction. As a result, this study found that morphological operations can induce errors in cladding images and that appropriate combinations of morphological operations can maximize segmentation performance. This study emphasizes the importance of image preprocessing methods, suggesting that they can enhance the accuracy of hydride analysis. These findings are expected to contribute to the advancements in integrity assessment of spent nuclear fuel cladding.

In this study, the characteristics of wind pressure distribution on circular retractable dome roofs with a low rise-to-span ratio were analyzed under various approaching flow conditions by obtaining and analyzing wind pressures under three different turbulent boundary layers. Compared to the results of previous studies with a rise-to-span ratio of 0.1, it was confirmed that a lower rise-to-span ratio increases the reattachment length of the separated approaching flow, thereby increasing the influence of negative pressure. Additionally, it was found that wind pressures varied significantly according to the characteristics of the turbulence intensity. Based on these experimental results, a model for peak net pressure coefficients for cladding design was proposed, considering variations in turbulence intensity and height.

Korea Atomic Energy Research Institute (KAERI) has been operating the Post Irradiation Examination Facility (PIEF). The facility has many PIE equipment and one of them is a hydrogen analyzer for measuring hydrogen contents in Zr cladding of spent fuel. The cladding tube of fuel is oxidized in the core environment of high temperature and pressure and absorbs some of the hydrogen generated during the oxidation. The hydrogen content increases with the increase of burn-up, and causes hydriding of the material, which degrades the mechanical properties. Therefore, hydrogen content analysis of the cladding tube is required for the performance and integrity evaluation of spent fuel. In PIEF, the hydrogen analyzer extracts hydrogen gas from Zr cladding by the hot extraction method. The hydrogen gas flows with inert gas and oxidizes to H2O through a CuO reagent. Finally, an IR detector measures the hydrogen amount from the absorbed IR intensity at a specific wavelength. Because the equipment is in the glove box and has some consumable parts, the maintenance work was performed as a radiation work.

The hydride reorientation (HR) of used nuclear fuel cladding after operation affects the integrity during intermediate and disposal storage, as well as the handling processes associated with transportation and storage. In particular, during dry storage, which is an intermediate storage method, the radial hydrogen redistributes into circumferential hydrogen, increasing the embrittlement of used nuclear fuel cladding. This hydride reorientation is influenced by various key factors such as circumferential stress (hoop stress) due to internal rod pressure, maximum temperature reached, cooling rate during storage, and the concentration of precipitated hydrogen during irradiation. To simulate long-term dry storage of used nuclear fuel, hydrogenated Zircaloy-4 cladding (CWSRA) specimens were used in hydride reorientation tests under various hoop stress conditions (70, 80, 90, and 110 MPa) for extended cooling periods (3 months, 6 months, and 12 months). After the hydride reorientation tests, the cladding’s offset strain (%) was evaluated through a ring compression test, a mechanical property test encompassing both ductility and brittleness. In this study, the offset deformation of the hydride reorientation specimens was compared and evaluated through ring tensile tests. In this study, the offset deformation values were compared and evaluated through ring tensile tests of the hydride reorientation test specimens. Hydrogen in zirconium cladding reduces ductility from a physical perspective and induces rapid plastic deformation. Generally, even in hydrogenated unirradiated cladding, it maintains a tensile strength of around 800 MPa at room temperature. However, high hydrogen content accelerates plastic deformation. In contrast, samples with radial hydrogen distribution exhibit fracture behavior in the elastic region below 500 MPa. This is attributed to the directional of radial hydrogen distribution. Specimens with a hydrogen concentration of 200 ppm fracture faster than those with hydrogen concentrations exceeding 400 ppm. This is believed to be due to the ease of reorientation of radial hydrogen in cladding with relatively low hydrogen content. Although the consistency of the test results is not ideal, ongoing research is needed to identify trends in hydride reorientation from a cladding perspective.

In nuclear fuel development research, consideration of the back-end cycle is essential. In particular, a review of an in-reactor performance of nuclear fuel related to the various degradation phenomena that can occur during spent fuel dry storage is an important area. The important factors affecting the degradation of zirconium-based cladding during dry storage are the cladding’s hydrogen concentration and rod internal pressure after irradiation. In this study, a preliminary analysis of the in-reactor behavior of the HANA cladding, which has been developed and is currently undergoing licensing review, was performed, and based on this result, a comparative analysis between nuclear fuel with HANA cladding and current commercial fuel under storage conditions was performed. The results show that the rod internal pressure of nuclear fuel with HANA cladding is not significantly different from that of commercial cladding, and the hydrogen concentration in the cladding tends to reduce due to the increased corrosion resistance, so fuel integrity in a dry storage conditions is not expected to be a major problem. Although the lack of cladding creep data under dry storage conditions, the results from the Halden research reactor test comparing in-reactor creep behavior with Zircaloy-4 showed that there is sufficient margin for degradation due to creep during storage.

It has been known that as oxide layer (ZrO2) forms on the nuclear fuel cladding during irradiation in nuclear power plants, the corrosion kinetics are influenced by various parameters such as chemical environments. One of those environments, crud deposition driven by coolant chemistry has an adverse effect on the formation of oxide (ZrO2) and leads to increase thickness of the layer. In this study, crud formation was performed through loop experiment equipment on the surface of intentionally-made oxide layer (ZrO2) on cladding tubes and then the composition and characteristics of cruds were examined for the investigation of nuclear power plant environment. As a result, various cruds in composition and microstructure were formed depending on the exquisite methods and conditions such as metal ion concentration.

Zircaloy-4 is utillzed in nuclear fuel rod cladding due to its strength and corrosion resistance. However, it can undergo deformation over time, known as creep, which poses a safety risk in reactors. Furthermore, hydrogen absorption during reactor operation can alter its properties and affect creep rates. Previous research suggests a trend in which hydrogen concentration corelates unidirectionally with creep rates, either increasing or decreasing as the concentration rises. This trend can also be observed in EPRI’s creep model, EDF-CEA Model-3. However, recent literature has suggested that creep behavior may vary depending on the state of hydrogen presence. Therefore, it has become evident that creep behavior can be influenced not only by hydrogen concentration but also by the state of hydrogen presence, whether it is in a solid solution state or precipitated as hydrides. Our study aimed to compare creep behavior in specimens with hydrogen concentrations below and above solubility limits. We fabricated Zircaloy-4 plate specimens with varying hydrogen concentrations and conducted creep tests. The results revealed that specimens below the solubility limit exhibited decreasing creep rates as hydrogen concentration increased, while those above the limit displayed increasing creep rates. This investigation confirms that the state of hydrogen presence significantly impacts creep behavior within Zircaloy-4 cladding. As part of our additional research plans, we intend to conduct creep tests on the material based on its orientation, whether it is in the rolling direction (RD) or the transverse direction (TD). We also plan to perform creep tests on ring specimens. Additionally, for the ring specimens, we aim to evaluate how creep behavior differs between the cold-worked stress-relieved (CWSR) condition and the recrystallized annealed (RXA) condition achieved through high-temperature heat treatment.

The Spent Nuclear Fuel (SNF) cladding serves as the first barrier that prevents the release of radioactive materials. It is very important to maintain cladding integrity in SNF management. It is known that the pinch load applied to the cladding can lead to Mode-3 failure and the cladding becomes more vulnerable to this failure mode with the existence of radial hydrides and other forms of mechanical defects. In this study, a numerical analysis process was proposed to scientifically and systematically evaluate the fracture resistance of cladding with reoriented hydrides under pinch load. The mechanical behavior and fracture of the irradiated cladding under pinch load can be evaluated by Ring Compression Test (RCT). Under the stress field generated by RCT, the cracks propagate more easily through radial hydrides than circumferential hydrides. The δ-hydride which form within the α-zirconium matrix causes a large expansion strain due to the volume difference and voids form at the interface between the hydride and the zirconium matrix. Chan demonstrated that the load needed to form voids and separate the hard hydride precipitates from the Zr matrix is considerably lower than that which initiates brittle fracture of hydrides using a micro-cantilever test. Therefore, we propose a microstructure crack propagation analysis method based on Continuum Damage Mechanics (CDM) that can simulate fracture of hydride, zirconium matrix, and Zr/hydride interface. CDM is possible to simulate the hydride, zirconium matrix, and interface cracking in a continuum model based on cladding deformation. The RCT simulation model was constructed from the microscopic images of irradiated cladding. A pixel-based finite element model was created by separating the hydride, zirconium matrix, and interface using the image segmentation method on a morphology operation basis. The appropriate element size was selected for the efficiency of the analysis and crack propagation using CDM. The force-displacement curves and strain energy from RCT were compared and analyzed with the simulation results of different element sizes. The finalized RCT simulation model can be used to evaluate the fracture resistance of the irradiated cladding under the quantified pinch load and to establish the failure criterion of fuel rods under pinch load. The advantages and limitations of the proposed process are discussed.

Hydride reorientation is widely known as one of the major degradation mechanisms in Zirconium cladding during dry storage. Some previous theoretical models for hydride reorientation used assumption of an ideal radial basal pole orientation for HCP structure of Zirconium cladding. Under this assumption, circumferential hydride was considered to precipitate in the basal plane while radial hydride was considered to precipitate in the prismatic plane, thereby giving energetical penalty on thermodynamical precipitation of radial hydrides. However, in reality, reactor-grade Zirconium cladding exhibits average 30° tilted texture, adding complexity to the hydride precipitation mechanism. In this study, reactor-grade Zirconium cladding was charged with hydrogen and hydride reorientation -treated specimens were fabricated. Microstructural characterization of hydrides was conducted via following three methods in terms of interface and stored energy. And this study aimed to compare these characteristics between circumferential and radial hydrides. Using Electron Back Scattered Diffraction (EBSD), the interface was investigated assuming that interface lies parallel to the axial axis of the tube. These were further validated with Transmission Electron Microscope (TEM). In addition, Differential Scanning Calorimetry (DSC) analysis was conducted to calculate the stored energy. This investigation is expected to establish fundamental understanding of how hydrides precipitate in Zirconium cladding with different orientations. And it will also increase the predictability of radial hydride formation and help understanding the mechanical behavior of Zirconium cladding with radial hydrides.

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.

In the process of spent fuel dry storage, which is an intermediate management method, it was found that hydrides in the circumferential direction rearranged into radial hydrides. Various factors, such as hoop stress, peak temperature, cooling rate during the storage period, and hydrogen concentration accumulated during the burnup process, significantly affect the susceptibility of spent fuel cladding. In recent studies based on the hydrogen solubility value of about 210 ppm corresponding to the peak temperature of 400°C, if the threshold stress decreases as the hydrogen concentration increases in the low hydrogen range under 210 ppm, the threshold stress increases as the hydrogen concentration increases in the low hydrogen range under 210 ppm. The fundamental cause of this trend is the diffusion of hydrogen into the high-stress region due to the stress gradient formed in the specimen, and hydrogen compounds which remain undissolved in the circumferential direction, even at the peak temperature, play a crucial role to determine the magnitude of the threshold stress. This study evaluated the behavior of hydride reorientation under various hoop stress conditions (70, 80, 90, and 110 MPa) using unirradiated Zircaloy-4(CWSRA) cladding tubes under long-term cooling conditions (3, 6, and 12 months). The results of analyzing the offset strain by hydrogen concentration for long-term cooling showed that specimens with low hydrogen concentration exhibited higher integrity than specimens with high hydrogen concentration at hoop stresses of 90 and 110 MPa. The HR test using irradiated fuel cladding showed that specimens with low hydrogen concentrations exhibited relatively higher susceptibility. To quantify these results, it is necessary to research further in detail by repeated tests.

The hydride reorientation (HR) of the post-irradiated nuclear fuel cladding after use affects the integrity of the spent nuclear fuel. During the dry storage process, which is an intermediate storage method, it was found that the hydride in the circumferential direction is rearranged into radial hydride, and this is believed to be due to factors such as hoop stress, peak temperature, accumulated hydrogen concentration, and cooling rate during the storage period. f(HR) = f(Tmax) + f(σH) + f(CH) + f(△T) + f(10Cy) + f(cooling rate) + ...... To simulate long-term dry storage of spent nuclear fuel, the hydride reorientation behavior was evaluated using unirradiated Zircaloy-4 (CWSRA) cladding with hydrogen charged under various hoop stresses (70, 80, 90, and 110 MPa) at long-term cooling periods (3, 6, and 12 months). Test results showed that as the cooling time increased, the sample with 90 MPa hoop stress at a maximum temperature of 400°C approached the ductility recommendation limit of 2%. In a 90 MPa hoop stress specimen with 3 months cooling period at peak temperature of 400°C, the offset strain was 4.24% at room temperature RCT, while it showed the result of 2.86% for the cooling period of 12 months. On the other hand, the specimen with hoop stress of 110 MPa and cooling period of 12 months showed result of 1.4%. The test results need to take into account errors in hydrogen charging and hydrogen analysis, and it is necessary to consider reproducibility through repeated tests. These results indicate the need for continued attention to the evaluation of the effects of hydride reorientation due to long-term cooling in the context of the integrity of spent fuel.

The damage ratio of Spent Nuclear Fuel (SNF) is a very important intermediate variable for dry storage risk assessment which require an interdisciplinary and comprehensive investigation. It is known that the pinch load applied to the cladding can lead to Mode-3 failure and the cladding becomes more vulnerable to this failure mode with the existence of radial hydrides and other forms of mechanical defects. In this study, a sensitivity analysis was performed to evaluate the importance of the damage parameters that need to be calibrated for the simulation of zircaloy-4 cladding failure using computational mechanics. The simulation model was generated from a microscopic image of the cladding with hydride. The image segmentation method was used to separate the Zircaloy-4, hydride, and hydride- Zircaloy matrix interfaces to create a pixel-based finite element model. The ring compression test (RCT) was simulated because the resistance of the cladding under pinch load can be evaluated by this test. It was assumed that the damage starts with the formation and growth of voids or small cracks in the material, which grow and combine to form larger cracks, eventually leading to the complete fracture of the material. Therefore, the ductile damage criterion was applied to all materials to simulate crack formation and propagation. The sensitivity analysis was performed based on the design of experiments using L8 orthogonal array. The effects of five factors on the fracture resistance of hydrided cladding were quantified, and they are the fracture strains describing the damage initiation in zircaloy-4 matrix, hydride, and hydride-zirconium matrix, and yield stress and Young’s modulus for hydride-zirconium matrix. Information on those parameters are hardly available in literature and experimental data which enable the estimation of those are also very rare. It is planned to build a computational model which can accurately simulate the fracture behavior of hydrided cladding by calibrating significant fracture parameters using reverse engineering. The results of this study will help to figure out those significant parameters.