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.

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.

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.

This study reassess safety margin of the current Peak Cladding Temperature (PCT) limit of dry storage in terms of hydrogen migration by predicting axial hydrogen diffusion throughout dry storage with respect to wet storage time and average burnup. Applying the hydride nucleation, growth, and dissolution model, an axial finite difference method code for thermal diffusion of hydrogen in zirconium alloy was developed and validated against past experiments. The developed model has been implemented in GIFT – a nuclear fuel analysis code developed by Seoul National University. Various discharge burnups and wet storage time relevant to spent fuel characteristics of Korea were simulated. The result shows that that the amount of hydrogen migrated towards the axial end during dry storage for reference PWR spent fuel is limited to ~50 wppm. This result demonstrates that the current PCT margin is sufficient in terms of hydrogen migration.

Hydride reorientation is one of the major concerns for cladding integrity during dry storage. In this study, mechanical property of post-reorientation cladding was investigated according to the morphology and amount of the hydrides. Cladding peak temperature limit 400°C was suggested by U.S. NRC in concern of cladding creep and hydride reorientation. In line with this regulatory limit, hydride reorientation was conducted during cool-down process from the maximum temperature of 400°C, using constant internal pressurization method. The specimens were charged for hydrogen from 100 to 1,000 wppm, and various pressures range of 7.5-18.5 MPa were applied. The morphology was examined by optical microscopy. Radial hydride fraction (RHF) and radial hydride continuous path (RHCP) were calculated using image analysis software PROPHET. Finally, strain energy density (SED) was investigated via ring compress tests and the hydrogen concentration was analyzed. The result shows that when RHF is higher than 5%, SED exponentially decreases with RHF. For RHF less than 5%, SED was primarily affected by the total amount of hydrogen. Shortened length of radial hydrides with the presence of circumferential hydrides may block the radial propagation of crack. The result implies that lower burnup spent fuel with lower hydrogen concentration may be more vulnerable in terms of radial hydride compared to higher burnup fuel.

The effect of oxidation time on the characteristics and mechanical properties of spent nuclear fuel cladding was investigated using Raman spectroscopy, tube rupture test, and tensile test. As oxidation time increased, the Raman peak associated with the tetragonal zirconium oxide phase diminished and merged with the Raman peak associated with the monoclinic zirconium oxide phase near 333 cm−1. Additionally, the other tetragonal zirconium oxide phase peak at 380 cm−1 decreased after 100 d of oxidation, whereas the zirconium monoclinic oxide peak became the dominant peak. The oxidation time had no effect on the tube rupture pressure of the oxidized zirconium alloy tube. However, the yield and tensile stresses of the oxidized nuclear fuel cladding tube decreased after 100 d of oxidation. The results of the scanning electron microscopy and transmission electron microscopy were represented with the in-situ Raman analysis result for the oxide characteristics generated on the cladding of spent nuclear fuel.

A long-term cooling effect on hydride reorientation of a cladding tube can affect the integrity of spent nuclear fuel transportation and long-term storage. In this study, experimental setup for investigating the degree of radial reorientation of hydrides in the circumferential direction during the long-term cooling was established. The experimental setup was designed to be simplified since the long-term evaluation requires a long term period such as 12, 18 and 24 months when the cladding tube specimen is gradually cooled down from 400°C to 100°C. For the test, hydrogen-charged specimens of 100 ppm, 200 ppm, and 500 ppm were prepared. The specimen was sealed with fixtures and check valve, and was pressurized up to 90 Mpa. To heat the specimen, a box-type furnace was used while the temperature of the specimen was measured from thermocouples attached to the specimen. After the heat treatment, the long-term cooling was performed by developing temperature control program to investigate several cooling rate conditions of the specimen. As a reference case, microstructure and brittle property of the hydrogen-charged specimens of 100 ppm, 200 ppm, and 500 ppm without the long-term cooling was observed. In the case of the hydrogen content, it was uniformly distributed in circumferential direction although it was non-uniform in the axial direction. In the case of the brittle property, a compression test was performed. For the future work, the microstructure and brittle property of the hydrogencharged specimens after the several long-cooling conditions were investigated. Then, the degree of radial reorientation of hydrides in the circumferential direction during the long-term cooling was studied.

The evaluation of the damage ratio of spent nuclear fuel is a very important intermediate variable for dry storage risk assessment, which requires an interdisciplinary and comprehensive investigation. It is known that the pinch load applied to the cladding can leaded 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, the failure resistance of Zircaloy-4 cladding against the pinch load is investigated using numerical simulations assuming the existence of radial hydrides. The simulation model is based on the microscopic images of cladding. A pixel-based finite element model was created by separating the Zircaloy-4 and hydride using the image segmentation method. The image segmentation method uses a morphology operation basis, which is a preprocessing method through erosion operation after image expansion to enable normal segmentation by emphasizing pixels corresponding to hydrides. The segmented images are converted into a finite element model by assigning node and element numbers together with corresponding material properties. Using the generated hydride cladding finite element model, several numerical methods are investigated to simulate crack propagation and cladding failure under pinch load. Using extended finite element (XFEM) models the initiation and propagation of a discrete crack along an arbitrary, solution-dependent path can be simulated without the requirement of remeshing. The applicability of fracture mechanical parameters such as stress intensity, J-integral was also investigated.