Rod Internal Pressure Research Articles

Integral LOCA experiments to study FFRD behavior of high burnup nuclear fuels

This paper introduces the Integral Loss Of Coolant (LOCA) facility (i-LOCA) established at Seoul National University. The facility was designed to investigate the integral fuel behavior of Light Water Reactors during LOCA, encompassing aspects such as cladding oxidation, ballooning and burst, reflood quenching, secondary hydriding, and fuel pellet dispersal. Integral LOCA experiments were carried out using three types of surrogate ZrO2 pellets, representing various segment burnups: cylindrical pellets with no fuel fragmentation (<55 GWd/MTU), mixed fragments of different sizes simulating ∼68 GWd/MTU (D = 0.3, 0.5, 1.0, 2.0, 3.0, and 5.0 mm with the same mass fraction), and small single powdered fragments simulating ultra-high burnup fuel (D = 0.5 mm, ∼94 GWd/MTU). Zr-Nb-Sn, Zr-1.1Nb, and Cr-coated (15 μm, Arc Ion Plating) Zr-1.1Nb ATF cladding were employed, with rod internal pressures ranging from 1 MPa to 7 MPa. The burst size and hoop strain exhibited significant variations depending on the type of surrogate pellets used, with larger burst sizes and hoop strains observed for smaller average diameters of surrogate pellets due to the effect of azimuthal and axial temperature distribution. Fuel dispersal was influenced by rod internal pressure, burst size, and the size of pellet fragments. Only pellet fragments smaller than the burst hole width underwent dispersal upon fuel burst, while larger fragments blocked the dispersal of smaller fragments. The rapidly escalating average dispersal fraction of single powder compared to mixed powder indicated a threshold burnup for fuel dispersal between 69–94 GWd/MTU. Cladding inner oxidation length was influenced by burst hole size and remaining fuel pellets. The results of inner oxidation confirmed the validity of the U.S. NRC’s assumption regarding the length of inner wall oxidation. The tested 15 μm Cr-coated cladding tubes, produced using the arc ion plating method, exhibited no significant differences in burst geometry, fuel dispersal, inner oxidation, and secondary hydriding when compared to the uncoated reference cladding.

Modeling and design of a separate effects irradiation test targeting fission gas release from Cr-doped UO2

Fission gas release (FGR) from nuclear fuel during operation can diminish heat transfer properties across the pellet-cladding gap and increase the fuel rod internal pressure, thereby posing a concern to fuel reliability and safety during an accident. Enlarging the fuel grain size, which has been shown to improve fission gas retention, can be achieved by doping the fuel feedstock prior to sintering. In this work, the BISON fuel performance code was used to predict FGR from undoped and chromia-doped UO2 (referred to as Cr-doped UO2) fuel specimens with different grain sizes and across various temperatures. The BISON models identified the irradiation conditions for which FGR is most significant, and a separate effects irradiation experiment in the High Flux Isotope Reactor (HFIR) was then developed targeting those conditions. The experiment leveraged the MiniFuel irradiation capability at Oak Ridge National Laboratory and consisted of 12 fuel specimens of varying grain size and Cr content. A coupling scheme between BISON FGR results and the ANSYS finite element thermal model used for experiment design was formulated to predict cumulative FGR from each fuel specimen based on expected irradiation temperature histories. The fuel samples were fabricated and characterized as a part of this work, and the fuel compositions modeled in BISON were representative of the specimens used in the experiment. This combined modeling and experimental effort aims to study the effect of fuel grain size and Cr content on FGR and to provide simulated BISON FGR results that can be used for future model validation activities.

High-temperature rupture failure of high-burnup LWR-MOX fuel under a reactivity-initiated accident condition

A reactivity-initiated accident (RIA)-simulated test CN-1 on a high-burnup 64 GWd/t mixed-oxide fuel rod sheathed with M5™ cladding was conducted at the Nuclear Safety Research Reactor, resulting in fuel failure. A small opening with slight ballooning deformation characterized the post-test visual appearance of the test fuel rod. Simulation using fuel performance codes FEMAXI-8/RANNS predicted rod survival under early phase loading induced by pellet-cladding mechanical interaction and subsequent boiling transition, and the cladding-surface temperature measured online confirmed the occurrence of boiling transition. The experimental observation and simulation indicate that the failure was caused by a high-temperature rupture following increased rod-internal pressure. The RANNS sensitivity analysis revealed that a mechanical state parameter dedicated to predicting plastic instability might be an effective index for evaluating the risk of rupture failure during RIAs.

Extended Development of a Fission Gas Release Behavior Model Inside Spherical Fuel Grains for LWR Reactors

Fission gas plays a significant role in fuel rod performance following accidents. The amount of fission gas increases dramatically under accidental conditions. This leads to a subsequent rise in the fuel rod internal pressure and temperature due to aggravated gap conductance between the fuel pellet and cladding. As a result, fuel rod performance degrades. Therefore, studying fission gas behavior is crucial for accident assessment and evaluating fuel rod performance. Minimizing the Impact of fission gas on fuel rods is essential for maintaining their integrity and safety within nuclear reactors. One important aspect of ensuring safety is predicting fission gas release (FGR). In this study, we presented an extended model to be used in light water reactors (LWRs). The FGR can be modeled using COMSOL Multiphysics with the finite element method. This modeling approach considers both normal and abnormal conditions, with the latter categorized as Class-II type incidents. The model assumes that the gas diffusion inside a spherical grain varies over time. By examining perfect sinks with gas production, perfect sinks without gas production, and imperfect sinks under steady-state conditions, different initial and boundary conditions are set. To validate the accuracy and universality of expressions used in the model, input parameters from other models and experiments are utilized. By comparing the model’s results with these inputs, the accuracy and applicability of the expressions can be confirmed. This validation process ensures that the model provides reliable predictions for fission gas behavior in fuel rods under both normal and abnormal operating conditions. Based on our findings, it is evident that the FGR fraction displays an upward trend as diffusion coefficients and temperatures rise. Conversely, larger grain sizes and higher linear heat generation rates are associated with a reduction in the FGR fraction. Notably, enhanced resolution leads to a postponed onset of FGR. Furthermore, the influence of the diffusion coefficient on the FGR fraction primarily stems from the interconnected effects of temperature and linear heat generation rate.

Radial hydride fraction with various rod internal pressures and hydrogen contents for Zr-Nb alloy cladding tube

Hydride reorientation experiments for unirradiated Zr-Nb alloy cladding tubes under multi-axial stress states induced by internal pressurization were performed with peak cladding temperature set to 400 °C. Applied internal pressure was held constant during cooling down stage. PROPHET was used to analyze Radial Hydride Fraction (RHF) of tested cladding tubes. The attained RHF behavior for a wide range of rod internal pressure (7.5-18.5 MPa) and hydrogen content (0-1100 wppm) were investigated. Empirical models for RHF with respect to rod internal pressure and hydride contents were developed. Ring compression tests (RCTs) were conducted to investigate toughness (strain energy density) change with radial hydride formation. In order to take the multi-axial stress effect into account, rod internal pressure-based regulation is necessary. Hydride reorientation threshold rod internal pressure is ∼9.32 MPa. The rod internal pressure corresponding to hoop stress of 90 MPa is 11.5 MPa, indicating that hydride reorientation in multi-axially stressed Zircaloy cladding tube occurs at hoop stresses below 90 MPa. The presented thermodynamic model for threshold rod internal pressure gives improved agreement with experimentally measured threshold with input the directly measured orientation relationship between δ-hydride and α-Zr.

Spent fuel simulation during dry storage via enhancement of FRAPCON-4.0: Comparison between PWR and SMR and discharge burnup effect

Spent fuel behavior of dry storage was simulated in a continuous state from steady-state operation by modifying FRAPCON-4.0 to incorporate spent fuel–specific fuel behavior models. Spent fuel behavior of a typical PWR was compared with that of NuScale Power Module (NPM™). Current PWR discharge burnup (60 MWd/kgU) gives a sufficient margin to the hoop stress limit of 90 MPa. Most hydrogen precipitation occurs in the first 50 years of dry storage, thereby no extra phenomenological safety factor is identified for extended dry storage up to 100 years. Regulation for spent fuel management can be significantly alleviated for LWR-based SMRs. Hydride embrittlement safety criterion is irrelevant to NuScale spent fuels; they have sufficiently lower plenum pressure and hydrogen contents compared to those of PWRs. Cladding creep out during dry storage reduces the subchannel area with burnup. The most deformed cladding outer diameter after 100 years of dry storage is found to be 9.64 mm for discharge burnup of 70 MWd/kgU. It may deteriorate heat transfer of dry storage by increasing flow resistance and decreasing the view factor of radiative heat transfer. Self-regulated by decreasing rod internal pressure with opening gap, cladding creep out closely reaches the saturated point after ∼50 years of dry storage.

Investigation of clad ballooning during NSRR RIA tests using ALCYONE fuel performance code

ALCYONE simulations of integral RIA transients performed on UO2 fuel rods prior to or within the ALPS program in the Japanese NSRR facility with stagnant liquid water coolant conditions are presented in this paper. The 11 selected tests which cover a wide range of fuel enthalpy increases (200 J/g – 800 J/g) aimed at assessing/challenging ALCYONE capabilities on four quantities of interest: clad outer surface peak temperature (293 K – 1200 K), film boiling duration (0 s – 2 s), clad residual hoop strain (0% – 25%) and transient fission gas release (2% – 25%). Despite a global consistency between measurements and ALCYONE predictions, paths are investigated to explain and reduce discrepancies. In particular, the reevaluation of fuel enthalpies by JAEA has led to revisit ALCYONE clad-to-water coolant heat exchange models which were derived from previous values of fuel enthalpies and suggested a recalibration of some of their parameters. Then, modeling the delayed gas axial flow in the free volumes of the rod is shown essential to achieve better residual hoop strain predictions in case of clad ballooning if proper timing of fission gas release, rod internal pressure increase and clad temperature elevation can be simulated. Key experimental and/or modeling research areas are shown to be at stake for future work.

Development of a 2D axisymmetric SiC cladding mechanical model and its applications for steady-state and LBLOCA analysis

A 2D axisymmetric SiC cladding mechanical model with anisotropic material properties based on finite difference method was developed. The developed model was used to analyze the behavior of SiC clad fuel (double-layer and full SiCf/SiC composites structure) in steady-state and large break loss of coolant accidents (LBLOCA) by coupling with the thermal-hydraulics simulation code MARS-KS (Multi-dimensional Analysis of Reactor Safety-Korean Standard). The multi-axial pseudo-ductile deformation model considering the characteristics of SiCf/SiC composites was employed, and the Weibull statistical failure model was applied to assess the failure probability of the cladding. Using anisotropic elastic properties, stronger tensile stress was induced in the outer monolith layer than in the isotropic case, which greatly increased the failure probability. In steady-state operation, this study shows that the failure probability of a double-layer SiC cladding is sensitive to the rod internal pressure. In-line with other studies, it was also found that the irradiation induced swelling accumulated in the refueling stage is a challenge to the structural integrity of the SiC cladding concept. The LBLOCA analysis demonstrates that the SiCf/SiC composite layer can maintain the coolable geometry of the fuel rod for both duplex and full composite cladding. Hermeticity is shown to depend on the arrestment of matrix cracks developed by pseudo-ductile deformation. Experimental support and validation on matrix crack propagation, pseudo-ductile behavior of SiCf/SiC composite under multiaxial loading, and rod-scale thermal shock with reflood quenching are necessary to further improve the model.

Analysis of Characteristics of Spent Fuels on Long-Term Dry Storage Condition

Currently, the interim storage pools of spent fuels in South Korea are expected to become saturated from 2024. It is required to prepare an operation plan of a domestic dry storage facility during a long-term period, with the researches on safety evaluation methods. This study modified the FRAPCON code to predict the spent fuel integrity evaluation such as the axial cladding temperature, the hoop stress and hydrogen distribution in dry storage. The cladding temperature in dry storage was calculated using the COBRA-SFS code with the burnup information which was calculated using the FRAPCON code. The hoop stress was calculated using the ideal gas equation with spent fuel information such as rod internal pressure. Numerical analysis method was used to calculate the degree of hydrogen diffusion according to the hydrogen concentration and temperature distribution during a dry storage period. Before 50 years of dry storage, the cladding temperature and hoop stress decreased rapidly. However, after 50 years, they decreased gradually and the cladding temperature was below 400 K. The initial temperature distribution and hydrogen concentration showed a parabolic line, but hydrogen was transferred by the hydrogen concentration and temperature gradient over time.

Development of a new efficient method using genetic algorithm for increasing of fuel rod life time

In an effort to increase fuel burn-up and a high level of reliability for the reactor safety, this article is focused on the optimization of fuel rod design parameters such as plenum length, gap thickness, cladding thickness, and fuel rod internal pressure using the Genetic Algorithm ( GA). During the optimization process, the fuel rod design parameters are optimized according to the constraints and standard safety requirements. To increase fuel efficiency with the aim of achieving a high level of fuel burn-up and safety, a developed method was used to optimize the design parameters of the Vandellos II fuel rod. The computational results show that by optimizing the fuel rod design parameters, the fuel burn-up is increased by 12% compared to the original values while the reactor safety margin is improved by 7%. The applied method is expected to be used to review and revise fuel rod design to improve fuel performance and reactor safety.

Development of MERCURY for simulation of multidimensional fuel behavior for LOCA condition

During a loss of coolant accident (LOCA), the ballooning and rupture of fuel cladding can block coolant flow and reduce the coolability of a reactor, which can lead to violation of a safety criteria. It is crucial that fuel models consider how multidimensional thermomechanical behavior and burnup properties affect safety analysis and evaluation. In this study, a multidimensional entire fuel rod analysis module (MERCURY) based on the finite element method (FEM) was developed to simulate multidimensional fuel behavior during a LOCA. The MERCURY incorporated a transient thermal analysis model, a multidimensional gap conductance model, a nonlinear mechanical model, and a transient creep model as thermomechanical models. As fuel models, burnup-dependent material properties, an oxidation model at high temperature, a rod internal pressure model, and cladding burst criteria were developed. Each FEM-based model was verified against results using a commercial FEM package. Verifications demonstrated that the models were formulated and integrated correctly. As validation, the MERCURY simulated experiments (PUZRY) regarding cladding behavior out-of-pile at high temperature and high inner pressure, which is similar to the fuel condition during a LOCA. The simulation results show good agreement with measured hoop strain in experiment.

Initial development of an RIA envelope for dispersed nuclear fuel

A reactivity insertion accident (RIA) is a design basis accident in which reactivity is rapidly injected as a result of a control rod ejection or blade drop scenario. The resultant power increase can result in fuel rod failure and subsequent release of radioactive material into the reactor’s primary system. For light water reactors, failure under RIAs is typically a result of fuel melt, pellet-cladding interaction, or rod over-pressurization. Fuel melt occurs when the fuel system is unable to transport heat out of the fuel system. Pellet-cladding interaction occurs when an aggressive fuel pellet expansion pushes on an embrittled cladding beyond its strain limit. Rod over-pressurization occurs when a fuel rod exceeds the critical heat flux and departs from nucleate boiling, causing the rod internal pressure to rapidly increase, exceeding the systems pressure, and balloon until it bursts. To prevent failure, regulators have developed safety requirements that limit the injected enthalpy based on the state of the fuel system. However, dispersed fuel could enable the removal or relaxation of regulator-imposed safety criteria. Dispersed fuel embeds fuel particles in a highly conductive metal. Dispersed particles can have a particle radius of 1 µm to greater than 100 µm. Traditional fuel systems are plagued by poor thermal conductivity, especially at higher burnups, thus reducing the ability of the fuel system to respond to an RIA. However, dispersed fuel could improve the ability of the fuel system to transport injected energy away from the fissile material and into the coolant more efficiently. Additionally, dispersing the fuel particles mitigates hard contact that typically occurs in the traditional zirconium-uranium dioxide fuel system, thus removing pellet-cladding mechanical interaction as a potential failure mechanism. This paper evaluates fuel particles dispersed in a metal matrix using the BISON fuel performance code to develop an initial failure threshold based on melt temperature of the fuel particle and/or cladding material as a function of fission density for beginning-of-life and end-of-life conditions. BISON results show that (1) reducing the size of the particles allows fission density to increase as a function of time, (2) particle-to-particle proximity must be considered to evaluate the limiting conditions, and (3) improving the fuel particle thermal conductivity improves thermal performance.

Uncertainty analysis of a loss of cooling and loss of coolant accident in a spent fuel pool using TRACE

In this work the fundamentals and application results of an uncertainty analysis of a loss of coolant and loss of cooling accident in a spent fuel pool are presented. The spent fuel pool was modeled with TRACE V5.0 Patch 5 using a VESSEL 3D component in Cartesian coordinates. A more detailed nodalization was necessary as compared with a previous study of the same accident scenario to better develop the uncertainty analysis in the framework of the Best Estimate Plus Uncertainty (BEPU) approach. First, the model was calibrated using Main Yankee spent fuel pool experimental data of temperatures and flows measured in steady state conditions. Using the calibrated model, the analysis of the accident was performed considering uncertainty in boundary conditions and in all the coefficients included in the TRACE uncertainty quantification data module. The figure of merit selected in this work is the time at which the cladding oxidation reaches 0.17 times the cladding thickness before oxidation (CFR 50.46 b2). The uncertainty analysis was performed using Wilks’ method and surrogate models. In particular, the Elastic Net surrogate model has demonstrated its potential not only to predict the evolution of the figure of merit in an appropriate way with lower computational effort than the Wilks’ method but also its capability to identify the most relevant uncertain parameters, such as the fuel and cladding specific heats, rod internal pressure coefficient, burst temperature coefficient, a group of fluid regime heat transfer coefficients, the cladding metal-water reaction rate coefficient and the cladding tolerance for the average cores. In addition, the surrogate model provides more information of the distribution of the evolution of the figure of merit. The estimations of the 5/95 tolerance limits, both with the Wilks’ method and with the surrogate models, are very similar, ranging from 30.432 to 30.461 h.

Parametric study on the structural response of a high burnup spent nuclear fuel rod under drop impact considering post-irradiated fuel conditions

Abstract A parametric study of several parameters relevant to design safety on the spent nuclear fuel (SNF) rod response under a drop accident is presented. In the view of the complexity of interactions between the independent safety-related parameters, a factorial design of experiment is employed as an efficient method to investigate the main effects and the interactions between them. A detailed single full-length fuel rod is used with consideration of post-irradiated fuel conditions under horizontal and vertical free-drops onto an unyielding surface using finite-element analysis. Critical drop heights and critical g-loads that yield the threshold plastic strain in the cladding are numerically estimated to evaluate the fuel rod structural resistance to impact load. The combinatory effects of four uncertain parameters (pellet-cladding interfacial bonding, material properties, spacer grid stiffness, rod internal pressure) and the interactions between them on the fuel rod response are investigated. The principal finding of this research showed that the effects of above-mentioned parameters on the load-carrying capacity of fuel rod are significantly different. This study could help to prioritize the importance of data in managing and studying the structural integrity of the SNF.

Methodology for a realistically conservative characterization of spent fuel in dry storage

The characterization of spent fuel in dry storage is fundamental to assess and assure the cladding integrity as one of the main safety criteria. To do that, reliable methodologies based on applicable analytical tools are pursued. The present paper shows an easy to apply methodology to characterize dry stored fuel from a realistically conservative approach. It has been validated through EOL rod internal pressure data from PWR fuel rods. The application to old and current designs of discharged fuel rods shows cladding stresses below the limit usually applied for embrittlement due to hydride radial reorientation (90 MPa). An analysis of the stress reduction with respect to enveloping conditions (licensing limits concerning irradiation and stored heat load) points out the importance of approaching realistic power histories (stress reduction higher than 30%), whereas the realism sought in the stored heat load is shown to be less important (stress reduction around 12%).

Investigations on a typical small modular PWR using coupled neutronic-thermal-mechanical evaluations to achieve long-life cycle-length

The main purpose of this study was to design small modular PWR with power of 150 MWth for long life operation, based on the coupled neutronic and thermal-mechanical evaluations. In this regard, pitch and fuel rod diameter were changed from 0.9 cm to 2.5 cm and 0.6 cm–0.96 cm, respectively; and desirable pairs (pitch - fuel rod diameter) were obtained. Neutronic parameters such as excess reactivity, maximum to average ratio of the radial neutron flux and fuel burnup were calculated using MCNP code. Also, thermal-mechanical parameters of the fuel rod including fuel centerline temperature, cladding hoop stress and fuel rod internal gas pressure were computed using FRAPCON code. Independency of results from number of axial nodes in FRAPCON code was analyzed. The number of axial nodes was considered 40 to ensure the accuracy of the FRAPCON results. Neutronic and thermal-mechanical cost values for each pair were calculated and Pareto front of the desirable pairs were obtained. Based on the results, eight desirable pairs were selected. By assuming equal importance for neutronic and thermal-mechanical parameters, total cost value was defined and pair of (1.1, 0.64) was selected as the most desirable one. Finally, fuel cycle lifetime was assessed for desirable pair considering neutronic (criticality and burnup) and thermal-mechanical (cladding strain, oxidation and hydriding as well as fuel rod internal gas pressure) limits. It was observed that fuel cycle lifetime was enhanced to 1750 days (40 GWD/ton).

Gap conductance modeling I: Theoretical considerations for single- and multi-component gases in curvilinear coordinates

Accurate estimation of heat transfer across the gap is important in nuclear fuel performance because heat transfer across the fuel-to-cladding gap heavily impacts fuel temperatures and the thermo-mechanical performance of nuclear fuel rods. Better understood physics will allow a better prediction of the gap behavior. This paper focuses on providing an overview of the gap conductance model including theoretical considerations and underlying assumptions. The gap conductance is calculated considering three summed heat paths: fill gas conductance, direct thermal radiation, and solid contact conductance. Each heat transfer mechanism is described in detail. First, the models are generalized to curvilinear coordinates for diatomic/polyatomic molecules. Traditional models use one-dimensional Cartesian equations for a monatomic gas. Second, expressions for temperature jump distance and thermal accommodation coefficients are made consistent with the kinetic theory for both single- and multi-component gases. Lastly, fill gas thermal conductivity is updated to include its dependence on rod internal pressure.

Numerical simulation of the effects of localized cladding oxidation on LWR fuel rod design limits using a SLICE-DO model of the FALCON code

Abstract A methodology for evaluation of mechanical and thermal effects of localized non-axisymmetric oxidation in zircaloy claddings on LWR fuel reliability is proposed. To this end, the basic capabilities of the FALCON fuel behaviour code are used. Examples of methodology application to adjustment of selected operational limits for modern BWR fuel rods, to capture effects of the excess local oxidation, are presented. Specifically, the limiting rod internal pressure for the onset of cladding lift-off is reduced, depending on initial excess oxidation spot sizes. Also, the power limits for Anticipated Operational Occurrences are adjusted, to preclude fuel melting and cladding failure due to PCMI and PCI-SCC in the affected fuel rods.

Modeling of flow blockage due to swell and rupture for analysis of LBLOCA using a multiple fuel rods scheme under high burnup condition

The present study discusses a model of Large Break Loss-Of-Coolant Accident (LBLOCA) calculation using the multiple fuel rods modeling scheme which was developed to consider the fuel rods with different burnup in preparation for the recently proposed revision of Emergency Core Cooling System (ECCS) rule in the United States of America and in Korea. The thermal conductivity of the fuel pellet, the initial oxide layer thickness of the cladding, the size of gap and the rod internal pressure, and the effective thermal conductivity of the cladding considering the oxide layer are reflected by an input as functions of the burnup. Further, through the analysis on the division of hydraulic channels in which the multiple fuel rods are located, the influences of division of the hydraulic channel and the crossflow between channels on cladding thermal response are evaluated. Also a modeling scheme to calculate the flow blockage due to swell and rupture of cladding for the current system thermal-hydraulic code such as MARS-KS code is discussed. To consider the uncertainty of the swell and rupture model of the code and the uncertainty in the present flow blockage modeling scheme, a multiplication factor is introduced and used to evaluate the influence of the flow blockage on ECCS performance.

Probabilistic Assessment of Peak Cladding Hoop Stress and Hydrogen Content of PWR SNF Rod Cladding

This study presents two statistical models that were developed to estimate the expected peak cladding hoop stress (CHS) and the amount of hydrogen in pressurized water reactor (PWR) spent nuclear fuel (SNF) rod cladding. Peak CHS is caused by high rod internal pressure during vacuum drying performed when transferring SNF to dry storage. During in-reactor operation of PWR fuel, the rod cladding tends to corrode and uptake hydrogen. The hydrogen content and CHS control hydride-related cladding embrittlement at low material temperatures. The two methodologies developed in this study were used to create a generic rod database with information on PWR SNF conditions. This database provides information on 100 000 randomly selected rods that form part of the current U.S. SNF inventory. According to the statistical results, the expected hydrogen content of PWR rod cladding is in a sensitive interval that may facilitate hydride reorientation. However, the modeling results suggest that the expected peak CHS of the selected rods is significantly below 90 MPa, which is the estimated lower bound stress necessary to trigger significant radial hydride embrittlement in cladding after being cooled to room temperature. Further, the results indicate that hydride embrittlement due to excessive hydrogen in cladding is unlikely. Therefore, a low probability of hydride-related embrittlement of PWR SNF cladding currently stored in the U.S. inventory is anticipated, even under consideration of low cladding temperatures after long-term SNF dry storage.