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At least 145 records · Page 8
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Breakaway oxidation of ZIRLO and optimized ZIRLO cladding per 10CFR50.46c criteria

The United States Nuclear Regulatory Commission (NRC) in 10CFR50.46c proposed changes to the loss of coolant accident (LOCA) criteria for LWR licensees including a new requirement to test cladding materials for breakaway oxidation. Cladding materials must be shown to resist breakaway oxidation long enough for the LOCA transient to be terminated. The onset of breakaway oxidation is characterized by a rapidly increasing rate of oxidation and the absorption of hydrogen resulting in a loss of cladding integrity. The proposed rule changes effectively require some manner of periodic and on-going testing to monitor the population of cladding for breakaway resistance. More than 200 lots of production cladding material have been tested in accordance with the draft regulatory guidelines. Testing has been conducted in a steam environment at the critical temperature for breakaway oxidation for durations that exceed the analytical endurance limit required to comply with the breakaway oxidation test criteria. The alloys that have been tested include both ZIRLO cladding and Optimized ZIRLO cladding. Breakaway oxidation testing consisted of three basic steps: preparation of test specimens, high temperature steam oxidation and post-test specimen evaluation. The results show there is a high probability that all ZIRLO and Optimized ZIRLO lots will pass production certification testing. Variations in response to breakaway oxidation testing were observed only in testing that ran well beyond the maximum LOCA duration of 2000 seconds. No indications of early breakaway oxidation were observed in the testing of production material that has been completed thus far. This provides confidence that a sampling approach to breakaway oxidation testing and monitoring remains a sound approach.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Zircaloy Cladding as Starting Precursor for Ceramic Waste-Forms - 15562

Zircaloy cladding of spent fuel is regular waste-stream related to the process of reprocessing. It remains in the apparatus after spent fuel dissolution. The most common technique of zircaloy cladding treatment is cementation and disposal in metallic containers. In view of the high cost of the cladding material (mainly zirconium), there have been a number of proposals for this material to be recovered and re-used. However, the residual contamination of the claddings following the dissolution of the fuel and the chemical stability of the zirconium militate against any proposal for a recycling process that might be economically reasonable. At the same time there is a lot of information available about synthesis of mineral-like materials based on zirconium for immobilization of long-lived radionuclides. The object of this work was to study experimentally electrochemical dissolution of fuel rod cladding for further obtaining starting precursor for the synthesis of ceramic-waste forms. The results of experiments on electrochemical dissolution of zirconium and zirconium alloy (fuel rod cladding) in nitric acid solutions are presented. The peak dissolution rate of specimens is shown to be observed in dilute nitric acid at elevated temperatures. The rate of dissolution rises with temperature, reaching ∼40-70 mg/cm{sup 2}.h at 90 deg. C. The current yield at this point is 0.6-0.7 g/A.h. The results of analysis of the surface oxide film on specimens and residues formed during dissolution are also presented. An electrochemical dissolution of zircaloy claddings is the common technique for preparation of zirconium solutions to be used for ceramic waste form synthesis. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Advancements in FeCrAl Alloys for Enhanced Accident Tolerant Fuel Cladding for Light Water Reactors

Advanced FeCrAl alloys are being considered as an enhanced accident tolerant fuel (ATF) cladding for light water reactor (LWR) applications as they can exhibit slower reaction kinetics with high temperature steam and have a reduced production of hydrogen during accident scenarios compared both to historic 304L cladding and Zircaloy. Both of these attributes are key towards producing a fuel-form with enhanced accident tolerance over the current Zr-based-UO{sub 2} fuel form in deployment. A multi-year, multi-laboratory effort has been established to rapidly advance FeCrAl alloys as an ATF concept. This effort has focused on optimizing the microstructure and composition of these alloys in an attempt to provide performance on-par with or better than the current Zr-based cladding in both normal operating conditions and under accident scenarios. To date, a significant database supporting key performance factors in both scenarios has been developed. Based on this database, a compositional and microstructural window has been identified providing the insight for an optimized, advanced nuclear-grade FeCrAl alloy for LWRs. This work will provide a comprehensive overview of the systematic effort to advance FeCrAl alloys as an enhanced ATF cladding in commercial LWRs. Efforts undertaken by this multi-year, multi-laboratory program have shown the viability of using FeCrAl alloys as an ATF cladding in LWRs. Most importantly, it has been shown the oxidation resistance, decreased hydrogen production, and reduced heat generation of FeCrAl alloys strongly contribute to its ATF capability. Furthermore, it has been shown that the normal operational performance of a FeCrAl cladding can rival that of Zr-based alloys. Finally, it was found that optimizing the performance of FeCrAl cladding materials requires balancing the starting composition and microstructure in order to satisfy key performance indicators such as oxidation resistance and radiation tolerance. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Reactivity impact of accident tolerant claddings in an equilibrium PWR core

The US DoE's Accident Tolerant Fuels (ATF) program aims to establish a process to evaluate the feasibility of several ATF concepts. The two main aspects of such a process are to evaluate the 3-D 'whole core' neutronics and fuel (thermal-mechanical) performance. Determining the best ATF candidate usually involves a compromise between these two performance aspects. The work presented in this paper focuses solely on the 3-D neutronics, looking specifically at the end of cycle (EOC) reactivity, and upon the establishment of a realistic and modern equilibrium PWR core to use as a test-bed. Both the lattice physics code CASMO-4 and the 3-D nodal simulator NESTLE were used in this study. Eventually, operational power histories are provided to our collaborating team to carry out fuel performance evaluations to be presented in a future article. Neutronically, a goal of a prospective ATF concept is to at least match the fuel cycle length of a core with standard Zircaloy/UO{sub 2} fuel, without having to significantly change the design of the fuel bundle; meaning, without changing pin pitch or outer clad diameter. Such a 'seamless' transition to another cladding type would greatly reduce the burden of all the nuclear energy stakeholders (regulators, vendors, utilities), from both an analysis and financial perspective. For cladding materials that are less transparent to neutrons, which for this study included all the ATF concepts, the two degrees of freedom with respect to fuel fabrication are the U-235 enrichment and cladding thickness. The latter involved changing only the fuel radius and inner clad radius while keeping the gap and outer clad radius constant. The thinnest cladding value was set to a generous estimate of the minimum iron-chromium-aluminum (FeCrAl) thickness that can be manufactured. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Mechanical response and microstructure evolution from multiaxial stress relaxation in textured zircaloy nuclear cladding

Modern energy markets provide motivation for nuclear reactors to increase operational flexibility and load following capacity of existing reactor designs. The operational power changes result in fuel pellet expansion and contraction resulting in pellet-cladding mechanical interaction (PCMI) and high cladding stresses that are strongly related to stressed corrosion cracking (SCC) which can lead to fuel rod failure. This work aims to quantify the stress fields in cladding with a focus on multiaxial stress relaxation response from an imposed strain and the resulting microstructure changes in textured zircaloy claddings. Stress relaxation tests resulting in multiaxial stress fields are performed to measure the material response of Zircaloy-4 nuclear cladding. Multiple experiments using samples from the same piece of cladding are performed so samples with the same processing history may be removed at strategic points throughout mechanical testing for microstructure characterization. The microstructure evolution is analyzed with mechanical response to evaluate the static recovery impact on crystallographic texture and grain morphology. Experiments are performed at Idaho National Laboratory using an experimental device capable of applying internal pressure and independent axial load to commercial nuclear cladding at elevated temperatures with stress- or strain-controlled experimental capability. X-ray diffraction and electron backscatter diffraction are used for microstructure analysis. Rheological models are used to visualize the multiaxial stress-strain relationships and isolate elastic, plastic, and viscoplastic material properties for analysis.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Mechanical Properties of Advanced Gas-Cooled Reactor Stainless Steel Cladding After Irradiation

The production of helium bubbles in advanced gas-cooled reactor (AGR) cladding could represent a significant hazard for both the mechanical stability and long-term storage of such materials. However, the high radioactivity of AGR cladding after operation presents a significant barrier to the scientific study of the mechanical properties of helium incorporation, said cladding typically being analyzed in industrial hot cells. An alternative non-active approach is to implant He{sup 2+} into unused AGR cladding material via an accelerator. Here, a feasibility study of such a process, using sequential implantations of helium in AGR cladding steel with decreasing energy is carried out to mimic the buildup of He (e.g., 50 appm) that would occur for in-reactor AGR clad in layers of the order of 10 µm in depth, is described. The implanted sample is subsequently analyzed by scanning electron microscopy, nanoindentation, atomic force and ultrasonic force microscopies. As expected, the irradiated zones were affected by implantation damage (< 1 dpa). Nonetheless, such zones undergo only nanoscopic swelling and a small hardness increase (~ 10%), with no appreciable decrease in fracture strength. Thus, for this fluence and applied conditions, the integrity of the steel cladding is retained despite He{sup 2+} implantation.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Micro X-ray computed tomography examination of mini plate fuel with hot isostatic pressed aluminum cladding

In order to minimize proliferation risks and improve security of nuclear material, the United States high performance research and test reactors (USHPRR) program is tasked with converting nuclear reactors that are fueled with highly enriched uranium (HEU) fuels to operate with low-enriched uranium (LEU) nuclear fuels. One favorable LEU fuel configuration is plate fuel with a metallic uranium-molybdenum foil clad within an aluminum alloy (AA 6061). In this fuel, the aluminum cladding is bonded with a hot isostatic pressing (HIP) method to seal the cladding around the fuel meat. However, the HIP process parameters influence the cladding performance, as a defective or incomplete bond can cause a pathway for corrosion. Micro X-ray computed tomography (XCT), a nondestructive technique that provides volumetric imaging, can be applied to inspect fuel plate cladding at the engineering scale. In this work, XCT methodology was developed and successfully utilized to not only observe the bond line of unirradiated mini fuel plates, but to also identify subsurface abnormalities in the plates’ cladding. Importantly, in future work, this technique could be applied to fuel plates, pre- and post-irradiation, to quantify irradiation effects on cladding defects and bond line integrity.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Fracture toughness evaluation for Zr-4 clad tubing structure with pellet inserts

Applying fracture mechanics approach to spent nuclear fuel (SNF) system reliability investigation is warranted due to the inherent flaws and hydride structures existed in a SNF system after nuclear reactor operation. However, none of the existing fracture toughness data deal with fuel cladding specific geometry or spent fuel material conditions, such as cladding structure with the pellet-inserts and the associated pellet clad mechanical interactions induced mixed-mode damage mechanisms. Thus, the development of an intrinsic fracture mechanics approach that is suitable for SNF materials is needed. Due to thin wall and small dimension of clad tubing structure, the spiral notch torsion test (SNTT) method of small specimen approach was used to estimate the clad tubing structure fracture toughness. The estimated fracture toughness for Zr-4 cladding with alumina-pellet inserts are presented in this report. For SNTT samples with a short or medium crack length, between 5.4-mm and 8-mm, the estimate JIQ upon fracture initiation for the baseline Zr-4 cladding is at 50 kJ/m2 with 2-sigma uncertainty of 3.26 kJ/m2, and the associated KIQ is at 67.46 MPa√m. For SNTT samples with a long crack length, around 13-mm, the crack initialization is deviated from that of the Mode-I tensile fracture and appears to be a mixed-mode fracture of Mode I – tensile stress and Mode III – out of plane shear stress; the estimated JMQ is at 18.9 kJ/m2, the associated KMQ is at 41.4 MPa√m.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Advanced Multiphysics Code Coupling for Cladding Surface Thermocouples During Two-Phase Heat Transfer from Nuclear Fuel

Transient testing of nuclear fuel involves the evaluation of fuel performance under off-normal and accident conditions and is essential for proving design performance. Instrumentation included in such experiments commonly includes thermocouples attached to the outer surface of the cladding to provide temperature measurements throughout the transient. However, the presence of thermocouples on the cladding surface can alter the local heat transfer characteristics with the surrounding coolant. These localized effects can influence the temperature of the nearby cladding surface and introduce uncertainties in interpreting the thermocouple data. Understanding the impact of thermocouples attached to the outer surface of the cladding is crucial for accurate data interpretation as well as its effect on the thermomechanical behavior of the cladding. This paper presents a novel methodology for simulating the impact of outer cladding thermocouples during transient testing of nuclear fuels. The simulation framework leverages the thermal-hydraulic capabilities of RELAP5-3D coupled to the BISON fuel performance code through the RELAPCouplingApp interface. The methodology is compared against Accident Tolerant Fuel Reactivity Initiated Accident-1-E experiment performed at Idaho National Laboratory. The results reveal approximately 100°C difference between thermocouple-altered temperature and virgin cladding surface. The model overpredicts surface rewet time due to conservative correlations.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Simulations of Failure via Three-Dimensional Cracking in Fuel Cladding for Advanced Nuclear Fuels

Enhancing performance of fuel cladding and duct alloys is a key means of increasing fuel burnup. This project will address the failure of fuel cladding via three-dimensional cracking models. Researchers will develop a simulation code for the failure of the fuel cladding and validate the code through experiments. The objective is to develop an algorithm to determine the failure of fuel cladding in the form of three-dimensional cracking due to prolonged exposure under varying conditions of pressure, temperature, chemical environment, and irradiation. This project encompasses the following tasks: 1. Simulate 3D crack initiation and growth under instantaneous and/or fatigue loads using a new variant of the material point method (MPM); 2. Simulate debonding of the materials in the crack path using cohesive elements, considering normal and shear traction separation laws; 3. Determine the crack propagation path, considering damage of the materials incorporated in the cohesive elements to allow the energy release rate to be minimized; 4. Simulate the three-dimensional fatigue crack growth as a function of loading histories; 5. Verify the simulation code by comparing results to theoretical and numerical studies available in the literature; 6. Conduct experiments to observe the crack path and surface profile in unused fuel cladding and validate against simulation results; and 7. Expand the adaptive mesh refinement infrastructure parallel processing environment to allow adaptive mesh refinement at the 3D crack fronts and adaptive mesh merging in the wake of cracks. Fuel cladding is made of materials such as stainless steels and ferritic steels with added alloying elements, which increase stability and durability under irradiation. As fuel cladding is subjected to water, chemicals, fission gas, pressure, high temperatures, and irradiation while in service, understanding performance is essential. In the fast fuel used in advanced burner reactors, simulations of the nuclear fuels are critical to understand the burnup, and thus the fuel efficiency.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Effects of Multiple Drying Cycles on High-Burnup PWR Cladding Alloys

The purpose of this research effort is to determine the effects of canister/cask vacuum drying and storage on radial hydride precipitation in high-burnup (HBU) pressurized water reactor (PWR) cladding alloys during cooling for a range of peak drying-storage temperatures, internal gas pressures, and hoop stresses. The HBU PWR cladding alloys have a wide range of hydrogen contents and varying hydride morphology after in-reactor service. Radial hydrides are a potential embrittlement mechanism for HBU cladding subjected to hoop-stress loading, which may be significant during normal cask transport. Ring compression tests (RCTs), which simulate pinch-type loading at grid spacers, are used to determine cladding ductility as a function of RCT temperature and the ductile-to-brittle transition temperature (DBTT). Previous tests were conducted with pressurized and sealed cladding rodlets heated to 400°C (the NRC ISG-11, Rev. 3, limit for all fuel burnups under normal conditions of storage and short-term loading operations) and cooled slowly at 5°C/h. Following this drying-storage simulation, the DBTT for HBU M5 ® decreased from 80°C to 70°C to <20°C as the peak hoop stress decreased from 140 MPa to 110 MPa to 90 MPa. Under similar drying-storage conditions, the DBTT for HBU ZIRLO™ decreased from 185°C to 125°C to 20°C to <20°C as the peak cladding hoop stress decreased from 140 MP to 110 MPa to 90 MPa to 80 MPa. These tests were conducted with a single heating-cooling cycle as simulation of the vacuum drying process with one drying cycle. (ISG-11, Rev. 3 allows <10 drying cycles, with cladding temperature variations that are less than 65°C each.)

Billone, M. C.↗
DOE OSTI2017

Al x In 1-x P LEDs with II-VI Cladding Layers for Efficient Red and Amber Emission. Final Scientific & Technical Report

Realizing solid state lighting systems with high efficiency and enhanced functionality will require the adoption of color-mixing (cm-LEDs) architectures combining highly efficient red, green, blue and amber light emitting diodes. Already, the hybrid architectures (hy-LEDs) pair phosphor-converted wide bandgap nitride LEDs with (Al x Ga 1-x ) 0.51 In 0.49 P red LEDs to improve the emission spectrum and reduce losses in the infrared. However, the efficiencies of (Al x Ga 1-x ) 0.51 In 0.49 P red and amber LEDs are still too low to truly enable these approaches. The DOE Solid-State Lighting R&D Plan highlights red and amber LED improvement as a long-term research goal. Excessive electron leakage out of the cladding layers, in particular, leads to strong thermal droop and red LED efficiencies of ~25% at normal operating temperatures. The typical device architecture for red and amber LEDs is based on indirect bandgap (Al y Ga 1-y ) 0.51 In 0.49 P cladding layers used to confine injected electrons and holes within a lower, direct bandgap (Al x Ga 1-x ) 0.51 In 0.49 P light-emitting active layer (x < y). The challenge with this approach is that the electron confinement is limited and decreases as the emission wavelength is shortened. Hence, thermal droop and overall device efficiency are very poor for amber LEDs. Innovative solutions to this problem must be sought through the use of new semiconductor materials and structures that will increase electron confinement. We plan to develop Al x In 1-x P LEDs with a II-VI semiconductor electron cladding layer to greatly increase the carrier confinement. Such an approach has the potential to improve red and amber LED efficiency, temperature stability, and light extraction. Al x In 1-x P active layers provide high direct bandgap energies needed to reach amber wavelengths without detrimental internal losses. ZnSe-based alloys are attractive candidates for the electron cladding layer because they can be engineered to provide high conduction band energies relative to the Al x In 1-x P active layer while maintaining roughly the same lattice constant. The lower index of refraction of the II-VI cladding layer will also aid light extraction. While extended defects have largely precluded the use of II-VI semiconductors as the light emitting layer in LEDs and lasers, the effects of these defects on device performance can be mitigated by limiting the use of II-VI materials to the thin, heavily doped cladding layer. Proposed research will focus on optimizing the quality of the Al x In 1-x P active layer and the heteropolar growth of the III-V/II-VI interface, tuning the II-VI cladding layer composition, and device design to maximize electron confinement and LED efficiency. Significant improvement in the performance of red and amber LEDs through this new approach is expected to advance the capabilities of next generation solid-state lighting technologies.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Purification of Zirconium Cladding Using a Chloride Volatility Process

The goal of this work is to investigate the thermochemistry of used fuel Zr cladding component chlorides as well as their activation products and residual radionuclide metal chlorides, in order to provide the DOE-NE community with fundamental data that will support the development of a chlorination process for the separation and purification of ZrCl 4 . Our studies will focus on the following elements: Zr, Sn, Fe, Cr, Sb, Nb and Cs; in the following sections, the term “chloride species” is applied to the following elements: Sn, Fe, Cr, Sb, Nb and Cs. Experimental and theoretical research will focus on the characterization, reactivity and volatility of the chloride species present in the ZrCl 4 obtained after chlorination of Zr cladding. The methods previously used for the purification of ZrCl 4 will be reexamined and the thermochemistry of the chloride species evaluated for each of the methods. Claddings of interest include Zircaloy-4, Zircaloy-2 and Zr-Nb alloys (Zirlo TM and Zr-2.5Nb). In this context, we propose five Tasks: Task 1: Computational study of the chemical form and thermochemical behavior of the chloride species produced after chlorination of Zr, Task 2: Characterization the chloride species obtained after chlorination of Zr cladding and Zr cladding components, Task 3 and Task 4: Study the thermochemical behavior of Zr cladding and Zr cladding complements in the presence of Cl 2 and HCl gas, and Task 5: Laboratory scale demonstrations of the removal of the impurity chloride species from ZrCl 4 .

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Fracture Toughness Evaluation for Spent Nuclear Fuel Clad Systems Using Spiral Notch Torsion Fracture Toughness Test

Many radioactive materials within the nuclear fuel cycle present a significant hazard. Such materials include spent nuclear fuel, high-level waste, legacy waste and other nuclear materials. These materials are often held in long-term storage as an interim stage within their lifecycle, including reuse or disposal. Safety and security of spent nuclear fuel (SNF) interim storage installations and follow-on SNF transportations to its final repository sites are very important, due to a great concentration of fission products, actinides and activation products. Fracture mechanics approach in applying to SNF system reliability investigation, especially for the high burn up (HBU) SNF, during SNF long term dry storage or SNF transportation is necessary due to the inherited flaws and inhomogeneity structures existed in a SNF system from the reactor operation. For instance, such as hydride and oxide formation, surface flaw induced by SNF assembly contact interactions, or internal flaws induced by pellet-clad mechanical interaction (PCMI). However, none of the existing fracture toughness data deal with fuel cladding specific geometry or SNF material conditions, such as cladding structure with the segment fuel pellets configuration and the PCMI mechanism. Consequently, the application of existing fracture toughness data to SNF system reliability investigation can be error prone; thus, the development of an industry-wide consensus fracture mechanics approach is needed. The objective of this research is to develop an in-situ fracture mechanics testing protocol, including the associated analytical procedure, that is suitable for evaluating the SNF fracture toughness. Fracture toughness data was obtained under quasi-static fracture loading using Oak Ridge National Laboratory (ORNL) developed Spiral Notch Torsion Test (SNTT) technology carried out on a biaxial tension/torsion tester. These data will be used to support SNF reliability investigation during long-tern SNF dry storage stage or in the follow-on SNF system transport to the final repository sites. SNTT has been a recent breakthrough in measuring fracture toughness for different materials, including metals, ceramics, concrete and polymers composites. Due to its high geometry constraint and unique loading condition, SNTT can be used to measure the fracture toughness with smaller specimens without concern of size effects. The application of SNTT to brittle materials has been proved to be successful. The micro-cracks induced by original notches in brittle materials could ensure crack growth in SNTT samples. Therefore, no fatigue pre cracks are needed for brittle materials specimens. The application of SNTT to the ductile material to generate valid toughness data will require a test sample with sufficient pre-crack length to increase the sample localized constraint at crack front. Fatigue pre-crack growth techniques with the associated compliance function estimated was developed for estimating the crack penetration depth to monitor the fatigue crack growth evolution. In order to extend SNTT approach to a thin shell cladding structure material, a new testing protocol and the associated analytical procedure for estimating the SNF fracture toughness was developed. The detailed SNTT approach and its estimated fracture toughness for the Zr-4 cladding structure with segment alumina inserts are presented in this report. The SNTT test results indicate that SNTT method is a reliable test approach with good repeatability in applying to Zr-4 cladding material. For the medium and the short crack length, the estimate J Q upon fracture for the baseline Zr-4 cladding is at 285.7 lb/in with 2-sigma uncertainty of 18.59 lb/in., and the associated K Q is at 61.4 Ksi√in. For a long crack length, the crack initialization orientation is apparently deviated from the principle tensile stress contour – 45° spiral crack front (i.e., Mode I + Mode III failure mechanism); and the estimated J Q is at 108 lb/in., the associated K Q is at 37.7 Ksi√in. The estimate Mode I J Q (along the 45° pitch principle tensile stress contour) is at 200 lb/in.; which indicate a significant reduction in the fracture toughness in a tubing structure under mixed mode loading condition, Mode I + Mode III, compared to that of Mode I (tensile stress dominated) alone.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Report on DIC Cyclic Dryout Testing of ATF Cladding Materials

Boiling water reactors can experience anticipated operational occurrences, where an expected deviation from normal operating conditions is witnessed for short time scales. One example of such transients arises from cladding dryout, where stable boiling becomes unstable, resulting in cladding heat up before re-wetting occurs. To understand the impact of such transients on fuel cladding lifetime and performance, cyclic dryout conditions were imposed upon fresh fuel cladding in the absence of irradiation. Pressurized cladding tube segments were exposed to rapid temperature increases to intermediate temperatures below a cladding rupture threshold and deformation was monitored in situ via 3-D digital image correlation. The impact of cycle frequency on cladding deformation was assessed. A clear understanding of elastic and plastic strain responses were developed for Zr alloy deformation under temperature transients, and these results can be used to support material model development and validate fuel performance codes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

SiC cladding R&D execution plan

To advance the development of accident-tolerant fuel cladding for light-water reactors, the Advanced Fuels Campaign actively addresses the R&D gap between the specialized technological progress in SiC-based composite materials, led by industrial partners, and the fundamental scientific understanding of material behavior under reactor operating conditions. The laboratory activities are systematically organized based on the Development Strategy document, which outlines critical barriers to the use of SiC composites in nuclear energy applications for the coming decade and beyond. To complement the strategy, this execution plan specifies a detailed year-by-year work package and establishes key milestones for assessing progress in technological advancement. This plan has been formulated in collaboration with principal investigators at national laboratories, stakeholders, and industry partners with the goal of closing technology gaps and readying SiC-based cladding technologies for full-scale commercial testing. The research conducted at the national laboratories is categorized into three primary subjects: (1) addressing challenges associated with the performance of SiC cladding under normal operating environments, (2) identifying the failure limits of SiC cladding under accident conditions, and (3) modeling fuel performance for SiC cladding systems. Comprehensive laboratory activities integrate the development of advanced modeling tools for SiC-based cladding systems with experimental validation via integrated testing of multiphysics phenomena. This document lays out the laboratory plan to support the development of SiC-based technologies, which have the potential for significant operational and safety benefits for the U.S. nuclear industry. This plan aligns with industry developer needs, relevant Nuclear Energy University Programs, and Nuclear Energy Advanced Modeling and Simulation Programs to ensure SiC cladding technologies are matured as efficiently as possible and will be maintained and updated to accommodate the dynamically evolving progress of development, capability needs, and resource availability.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Effect of Clad Fast Neutron Flux Distribution on Quarter-Core Fuel Performance Calculations with BISON

As part of the Consortium for Advanced Simulation of Light Water Reactors (CASL), the Virtual Environment for Reactor Applications (VERA) is being developed to provide high-fidelity multiphysics simulations of nuclear reactor cores. Over the past year, a one-way coupling capability has been developed, taking power and temperature data from MPACT/CTF calculations and running standalone BISON fuel rod cases. This has allowed CASL contributors to gain insight into the fuel performance characteristics while a more tightly coupled methodology between MPACT, CTF, and BISON has been under development. Many of the initial analyses have focused on the Watts Bar Unit 1 core using a fast flux factor—the default approach—which defines the clad fast flux as a linear function of the local linear heat rate. While this approach is more valid when considering single rod cases in isolation, in larger problems such as full-sized reactors, the local linear heat rate does not directly correlate to the clad fast neutron flux. For example, variations in assembly enrichment can lead to larger neutron fluxes in neighboring assemblies that may be at lower enrichment and power. Similarly, considering axial effects, spacer grids suppress power locally, but the fast neutron flux is significantly less affected, and the approach can lead to an underestimation of the fast neutron flux at that time. A poor estimation of the clad fast flux values can affect the clad creep rate, fuel-clad gap, and clad stresses. Additionally, the fast flux factor has been considered constant throughout the life of the rod. As the cycle progresses, the spectrum will harden as burnable absorbers and 235U deplete, so the fast flux would be expected to increase as the spectrum hardens. The effect on quantities of interest are highlighted for the first cycle of the Watts Bar Unit 1 core, comparing the results from an explicit clad fast flux representation from MPACT to the constant fast flux factor approach used in previous analyses. Future work will consider multicycle effects to assess the impact at higher burnups.

Stimpson, Shane G.↗
DOE OSTI2019

OVERVIEW OF ALUMINUM CLADDING OXIDE PREDICTION MODELS FOR EUROPEAN HIGH POWER RESEARCH REACTORS

UMo/Al dispersion fuel clad with aluminum alloy is a primary candidate fuel form that is being developed for the European high power research reactors (EUHPRR). In BR2 tests, water oxidation of cladding grew thick due to high powers. Excessive cladding oxidation is a cause for serious fuel performance degradation in a test while fuel performance itself may be sound. It is necessary for a fuel designer to have a reliable model to predict cladding oxidation for fuel licensing activities. Using oxide data from BR2 tests, this study reviews the prominent oxide prediction models and cladding-to-coolant heat transfer correlations. Because the oxide growth models were based upon different coolant flow conditions from those of the EUHPRR, the effect of the heat transfer correlations at the cladding-to-coolant interface on the oxide thickness prediction were also examined. The best combination of the oxide prediction model and the heat transfer correlation applicable for the EUHPRR is recommended. Since the existing oxide models were all developed for aluminum alloy types different from those currently considered for the EUHPRR, the effect of cladding alloy type is also discussed.

Research reactor fuel, aluminum oxidation, in-pile↗