Search NASA⌕ Search

SEARCH · Search NASA

Results for “Cladding design”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Multi-Functional Zirconium-Silicide Coatings on Zirconium-alloy for Improved Accident Tolerance

Recently, development of protective coatings on zirconium-alloy (Zr-alloy) fuel cladding in light water reactors has been extensively investigated to enhance safety margins in accident scenarios. One accident scenario is degradation of the reactor core under a coolant-limited condition, where a large amount of Zr-alloy claddings would react with high temperature steam, leading to oxidation, hydrogen production, and loss of intrinsic mechanical properties of the fuel cladding. Although, total replacement of Zr-alloy with SiC-SiCf composites or FeCrAl is being considered for the long-term, application of oxidation resistant coatings on zirconium-alloy is being investigated as a near-term solution to mitigate oxidation of Zr-alloy and increase the accident coping time. Coatings, such as chromium, Ti{sub 2}AlC, SiC, and Cr{sub 3}C{sub 2}- NiCr, deposited by various deposition techniques have been investigated. In addition, recent studies on modified heating surface for heat transfer enhancement provides basis on the optimal design of coated Zr-alloy claddings so that the new cladding can be achieve accident tolerance and heat transfer efficiency for both accident and normal operation conditions. Therefore, boiling heat transfer from the coated surface to water environment must be evaluated in parallel. In heat transfer performance, boiling heat transfer is characterized by nucleate boiling, critical heat flux (CHF), transition boiling, minimum film boiling temperature (T{sub min}), and film boiling. Zirconium-silicide (ZrSi{sub 2}) is a potentially attractive protective coating material for cladding applications. Although this material has not been studied extensively, it appears to be a promising candidate because of the high melting points, and the low neutron penalty, and the generally good oxidation resistance to transition metal silicides. It has been reported that oxidation of ZrSi{sub 2} powder in a flowing air environment at elevated temperatures produced ZrO{sub 2}, SiO{sub 2}, and ZrSiO{sub 4}, pointing to the possibility of achieving highly protective surface layers of SiO{sub 2} and ZrSiO{sub 4} that would protect the material from further oxidation. In this study, magnetron sputter deposition of ZrSi{sub 2} on zirconium-alloy flats and cylindrical rods was performed. The coated samples were oxidized in high temperature air (700 deg. C, 1000 deg. C and 1200 deg. C) to evaluate their oxidation resistance. The coated cylindrical rods were quenched in saturated water to observe transition boiling, T{sub min}, film boiling regime, while the coated flats were tested in pool boiling facility to investigate nucleation boiling and CHF.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Design and Characterization of High Thermal Conductivity Pipes from Roll-to- Roll Processed Hybrid Copper-Polymer Strips for Waste Heat Recovery

Metal based heat exchangers targeting low temperature (<150°C) waste heat recovery (WHR) face high maintenance and material costs [1] that result in long payback periods (>1 year). Low-cost polymer pipes suffer from poor thermal conductivity (~0.2 W/m.K), and thus a low overall heat transfer coefficient. Previous reports demonstrated thermal conductivity enhancement in polymers through systematic molecular alignment [2, 3, 4], or by metal micro-particle inclusions [5]. However, these approaches suffer from scalability issues for WHR heat exchangers in terms of cost [1, 3], and thermomechanical considerations [6, 7]. In this work, we propose hybrid metal-polymer strips that are strategically designed to enhance thermal conductivity by cladding copper around polymer strips. The polymer strips with copper-clad edges are wound helically to form the pipes for a heat exchanger. Through numerical simulations, we optimize the copper-polymer strip layouts to achieve high wall thermal conductivity (k > 1 W/m.K) and also have low volume fraction of copper (<70%) for the composite pipes. In copper-polymer composite heat exchangers, the bottlenecks for heat transfer are the thermal interfacial resistances of copper-copper and copper-polymer joints. We experimentally measured interfacial resistances to be ~1x10 -5 K.m 2 /W and ~4x10 -5 K.m 2 /W for copper-copper and copper-polymer joints, respectively. Using the measured interfacial resistances, we further optimized the design of the hybrid pipes to find high thermal conductivity (> 1 W/m.K) designs. We predict up to ~20 % enhancement in overall heat transfer relative to an all polymer (k = 0.2 W/m.K) heat exchanger system. Using a cohesive zone modelling (CZM) for predicting delamination at the weld joints, we find that our hybrid pipes can have operating internal pressures > 200 psi. The optimized designs of hybrid metal-polymer pipes provide a scalable and cost- effective means for harvesting waste heat from low temperature sources. Thekdi, A., and Nimbalkar, S. Oak Ridge National Lab (2014). Shen, Sheng, et al. Nature nanotechnology 5.4 (2010). Xu, Yanfei, et al. Science advances 4.3 (2018). Smith et al. ACS nano, 9(2), (2015). Sofian, N.M., et al., Journal of Thermoplastic Composite Materials (2001). Aymonier et al. Chemistry of Materials, 15(25), (2003). Tavman, I.H., Journal of Applied Polymer Science, 62(12), (1996).

C Rajagopal, Manjunath↗
DOE OSTI2020

Augmented Monitoring and Condition Assessment Program for SNF Wet Storage Life Extension - 20489

Approximately 27 MTHM of spent nuclear fuel (SNF) owned and managed by the U.S. Department of Energy, Office of Environmental Management is stored in the L Basin at the Savannah River Site (SRS). This 'DOE SNF' is comprised of approximately 12,000 aluminum-clad, aluminum-based fuel assemblies (∼7 MTHM), and approximately 2000 non-aluminum fuel assemblies (∼20 MTHM). A program is in progress to perform Non-Destructive Examination (NDE) of the fuel and their storage containers to characterize the materials' condition, and to evaluate the effects of service to enable continued safe wet storage of the SNF. The Augmented Monitoring and Condition Assessment Program (AMCAP), a two-part program to develop and implement remote underwater Non-Destructive Examination of the aluminum SNF, and of the containers of the non-aluminum SNF, respectively, is aimed at the characterization and evaluation of corrosion degradation of aluminum fuel and container materials. The predominant design of the aluminum SNF (ASNF) stored in L Basin is the Materials Test Reactor (MTR) equivalent design, a plate fuel design. These SNF are stored in 5'' diameter tubes or 5'' x 5'' squares (called bundles) that are nominally 12' or up to 14' long. The SNF is de-bundled and inspected using a custom-designed MTR Fuel Inspection Table. The inspection table provides for indexed fuel positioning for a video camera examination with controlled lighting. A total of 10 of the SNF originating from foreign research and test reactors were selected for examination based on burnup, enrichment, and prior damage caused by service/storage history. A special inspection campaign of these 10 assemblies is in progress. The observed corrosion damage included minor to moderate attack from general corrosion, pitting, crevice, end grain and galvanic corrosion. Example results from the completed inspections are shown and discussed. The focus is a comparison of the as-received condition versus the as-found current storage condition that will serve to validate Water Chemistry and Corrosion Monitoring Programs. The non-aluminum SNF (NASNF) stored in L Basin are of diverse design that includes various geometries with claddings of stainless steel, Zircaloy, and Hastelloy. The fuel core materials include uranium alloys, oxides/mixed oxides, and carbides. These SNF materials, originating from early experimental and test reactors, are in various physical forms including single fuel elements and cut pieces. The fuel is stored in L Basin in various configurations including in direct bundled storage in aluminum tubes and in isolation cans that are in the bundles or in a larger over-size storage container (OSC). Concern with inside-out corrosion and the potential loss of configuration control and ability to handle the storage containers prompted the development of remote NDE methods that include visual and UT technologies to assess the condition of the containers. Candidate stored materials were selected for a special inspection campaign; the development of the NDE methods for the inspection for galvanic, crevice and sediment-induced corrosion (inside-out) are discussed. A summary of the SNF storage in L Basin at the SRS, and an overview of the AMCAP to enable continued safe storage of DOE SNF are described. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Design, Test and Demonstration of Fault Current Limiting HTS Transformer

The project developed new technology that enables the creation of a high temperature superconductor-based FCL power transformer. SuperPower’s research and development created new methods to bond HTS conductor to a supporting substrate, test, and insulate the resulting bonded conductor, reduce winding ac losses, ensure FCL functionality during a transformer fault and build firm superconducting joints in the transformer harnesses and cabling. The bonded conductor in this program was shown to meet the critical operating parameters of providing the superconducting transformer operation while being able to meet the target normal state resistance required for FCL operation. The bonded conductor was also shown to be able to handle the fabrication stresses associated with the manufacture of the FCL transformer while also being able to handle the high hoop stresses and axial forces during a fault transient. Much of the technology developed here is applicable to the broader applied superconductivity community. The ability to tailor the clad conductors performance characteristics gives the designer of devices utilizing HTS a broader capability to address the particular needs of an given application. SuperPower worked with its sub-recipients Waukesha Electric Systems, ORNL, Southern California Edison and University of Houston to develop the design, fabrication, installation and operational aspects of a fault current limiting transformer on the electrical grid.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2022

FY2022 Progress Report for Advanced Re-fabrication/Re-instrumentation Capability Development

In support of performing follow-on irradiation experiments with previously irradiated materials, the Halden Reactor Project developed unique and state of the art capabilities to refabricate and re-instrument previously irradiated materials. Such materials were used in in-pile tests at the Halden reactor, and out-of-pile tests for example using furnaces as a heat source. The decision to close the Halden Reactor Project results in the loss of this refabrication and re-instrumentation capability. As a result, the United States (U.S.) Department of Energy (DOE) has determined to develop refabrication and re-instrumentation capability at the unique shielded facilities at Idaho National Laboratory (INL). The development of refabrication capability has been completed and demonstrated. This report focuses on the complementary aspects of reinstrumentation and the progress to-date. Halden spent nearly 30 years developing both refabrication and re-instrumentation. Collaboration with Halden is allowing Idaho National Laboratory (INL) to develop this capability much more rapidly. In FY-21 the results include development of the capability to drill annular center holes in ceramic UO2 fuel pellets, development of fuel rod end caps with feedthroughs for centerline instrumentation inside the rodlet, The procurement of both fuel drilling and welding demonstration equipment from Halden, evaluation of surface thermocouple attachments to support better understanding of temperature measurement uncertainties, and finally, the conceptual design of a new shielded enclosure where advanced refabrication and re-instrumentation equipment can be housed. In FY-22, the results include; Completing set up of the drilling and welding modules procured from Halden, and early experimental trials using that equipment.; Completing set up of an out-of-cell circumferential weld system to allow for further weld development to take place and support fabrication of fuel for fresh fuel experiments.; Evaluation of the Hot Fuel Examination Facility (HFEF) infrastructure to support future installation of advanced re-fabrication/re-instrumentation equipment, specifically related to necessary infrastructure for cryo-drilling.; Developments in dry-drilling alternative. Including experimental studies showing cordierite is the most suitable surrogate for UO 2 for performing drilling studies. That drilling performance is enhanced when a fuel-clad bonding condition is simulated.; Conceptual design completed for attaching surface thermocouples to irradiated fuel in support of TWIST capsule experiments. The authors would like to thank the numerous colleagues at INL and Halden who provided support in accomplishing this work. Their support both material and intellectual is invaluable in advancing the state of the art and establishing the capabilities for refabrication and re-instrumentation at INL.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Qualitative and quantitative analysis of neutron irradiation effects in SiC/SiC composites using X-ray computed tomography

Silicon carbide fiber-reinforced silicon carbide matrix (SiC/SiC) composites are candidate materials for cladding of light water reactor (LWR) fuels. Loss of fission product gas retention due to the formation of microcrack networks is considered a potential failure mechanism for SiC/SiC-cladded fuels. In this work, a variety of SiC/SiC composite tubes were irradiated with and without an LWR-relevant radial heat flux in the High Flux Isotope Reactor, followed by detailed characterization with X-ray computed tomography (XCT). This first set of XCT data for neutron-irradiated samples confirmed that the internal stresses arising from a combination of temperature gradients and irradiation-induced swelling act as the primary driver for cracking. Consequently, while the observed cracking patterns varied depending on the tube architectures, the sharp edges of relatively large pores were found to be the common stress concentrator. These findings are useful to help improve the design and manufacturing of SiC/SiC fuel claddings for reduced failure probability.

42 ENGINEERING↗
DOE OSTI2021

Development and Implementation of Digital Image Correlation for Out-of-Cell Burst Testing

This report details the progress made towards utilizing in-situ methods to observe and quantify unirradiated Zircaloy-4 (Zr4) fuel cladding deformation during out-of-cell design basis accident conditions, specifically during loss-of-coolant accident (LOCA) burst testing in the Severe Accident Test Station at Oak Ridge National Laboratory. Digital image correlation (DIC) was implemented to calculate cladding strain during burst and infrared (IR) thermography was used to attempt a correlation of strain with temperature and to investigate thermal gradients. For both techniques, early experimentation revealed experimental modifications were necessary for implementation. For DIC, this manifested as shortened cladding lengths which could be centered in front of the viewing chamber bored through the IR furnace side, and for IR thermography this meant conducting burst tests in air without a reaction tube and the incorporation of a SiC shell to eliminate direct IR reflections from the tungsten lamps. In the following, the technical issues and subsequent origin of these modifications are discussed, laying the framework for a proof-of-concept DIC/IR characterization of a burst test in the final section.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Neutron Imaging With Li-Glass Based Multicore SCIntillating FIber (SCIFI)

The improvement of neutron imaging towards and beyond the microscale is a well-documented need for the iterative characterization and modeling of numerous microstructured X-ray opaque materials. This work presents the recent progress in evaluating a SCIntillating FIber (SCIFI) proof-of-concept towards micron-level thermal neutron radiography. These SCIFIs are composed of 6 Li-enriched silicate glass cores doped with a Ce activator. The cores possess ~8.5 μm diameters and ~10 μm pitch following fiber drawing with a cladding glass into an all-solid multicore fiber. A polished 5 mm × 5 mm array of 100 microstructured multicore SCIFI pixels was fabricated into a 1 mm thick faceplate. The neutron efficiency and light yield of the faceplate are characterized as functions of the 7.38 weight percent of Li 2 O, thickness, and the 70% active volume. It was determined that approximately 39% of a thermal neutron (2 Å) beam can be absorbed by the faceplate. The 6 L i ( n , &#x03B1;<!-- α --> ) t reaction is estimated to produce 7,700 ± 1,000 scintillation photons per event, referencing light collection from 241 Am irradiation of the faceplate. Simulations suggest that on average 17.5 ± 1.4% of these photons will be transported to an end of the fiber array for a thermal beam, with at least 7.2% of that total scintillation light being confined into the fiber cores in which it originated. The SCIFI faceplate was integrated into the Neutron Microscope (NM) at the Pulse OverLap DIffractometer (POLDI) beamline located at the Paul Scherrer Institut to image a Siemens star test object. Processed neutron radiographs acquired with the proof-of-concept faceplate resolved features at a state-of-the-art resolution of 16.1 ± 0.5 μm. The potential for even high resolution designs having smaller pitch or different cladding material is discussed.

47 OTHER INSTRUMENTATION↗
DOE OSTI2016

BISON Fuel Performance Analysis of FeCrAl cladding with updated properties

In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromium-aluminum (FeCrAl) alloys due to much slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high temperature steam, making accident mitigation more likely. As a continuation of the development for these alloys, suitability for normal operation must also be demonstrated. This research is focused on modeling the integral thermo-mechanical performance of FeCrAl-cladded fuel during normal reactor operation. Preliminary analysis has been performed to assess FeCrAl alloys (namely Alkrothal 720 and APMT) as a suitable fuel cladding replacement for Zr-alloys, using the MOOSE-based, finite-element fuel performance code BISON and the best available thermal-mechanical and irradiation-induced constitutive properties. These simulations identify the effects of the mechanical-stress and irradiation response of FeCrAl, and provide a comparison with Zr-alloys. In comparing these clad materials, fuel rods have been simulated for normal reactor operation and simple steady-state operation. Normal reactor operating conditions target the cladding performance over the rod lifetime (~4 cycles) for the highest-power rod in the highest-power fuel assembly under reactor power maneuvering. The power histories and axial temperature profiles input into BISON were generated from a neutronics study on full-core reactivity equivalence for FeCrAl using the 3D full core simulator NESTLE. Evolution of the FeCrAl cladding behavior over time is evaluated by using steady-state operating conditions such as a simple axial power profile, a constant cladding surface temperature, and a constant fuel power history. The fuel rod designs and operating conditions used are based off the Peach Bottom BWR and design consideration was given to minimize the neutronic penalty of the FeCrAl cladding by changing fuel enrichment and cladding thickness. As this study progressed, systematic parametric analysis of the fuel and cladding creep responses were also performed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

In-situ determination of strain during transient burst testing and the temperature dependence of Zircaloy-4 claddings

Understanding fuel system behavior during postulated loss-of-coolant accidents is pertinent for continued safe and efficient operation of light water reactors, particularly as higher burnups are being pursued and safety margins re-evaluated. Conventional mechanical models for the incumbent Zr alloys typically rely on the assumption that steady-state creep is the dominant fuel cladding response during transient accident conditions. To investigate this assumption, simulated accident burst testing was performed on Zircaloy-4 claddings with balloon behavior measured in-situ. Here, two distinct loading conditions were utilized during burst testing: (1) constant-gas-inventory where pressure was allowed to increase with temperature and (2) constant pressure. In-situ strains and strain rates were measured via 2-dimensional digital image correlation techniques and synchronized with temperature to determine deformation dependencies. The temperature dependence of strain rate was characterized by a two segment Arrhenius relationship, with a distinct transition between the high and low temperature/strain regimes. The average activation energy of the lower temperature/strain regime was 328 ± 25 kJ/mol, in agreement with the ~320 kJ/mol used for conventional LOCA models. However, the higher temperature/strain segment, which encompassed most of ballooning, showed increased activation energies as well as a dependence on whether the burst region was in view. For tests that burst away from the camera view, the average high temperature/strain segment activation energy was 635 ± 150 kJ/mol. For samples where the rupture opening formed in view, the average activation energy was 1015 ± 179 kJ/mol. This observed shift in temperature dependence indicates a transition in deformation mechanism at the end of life, possibly to time independent failure mechanisms, which has not yet been visualized in the literature for Zr alloys. Parameters at the transition points were analyzed to determine thresholds for this change in behavior, which occurred at an average hoop strain of 6.9 ± 2.1 %.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Mechanisms of stored energy release in silicon carbide materials neutron-irradiated at elevated temperatures

Understanding the stored energy release behavior of SiC is crucial for the design of SiC-based fuel cladding for light water reactors because stored energy release affects response during accident conditions. Differential scanning calorimetry was used to evaluate the stored energy of monolithic SiC and SiC fiber–reinforced SiC matrix composite variants following neutron irradiation under light water reactor–relevant dose and temperature conditions. The tests were performed at heating rates of 0.5, 3, and 20 K/min up to 1,273 K. The onset temperature of the energy release roughly corresponded to the irradiation temperature, and energy release was delayed as the heating rate increased. The monolithic and composite specimens exhibited similar stored energy release behavior up to ~1,000 K, and above that temperature, the energy release depended more on material type. The energy release measured in this study is below the material-specific heat capacity at temperatures below ~1,000 K but exceeds it under certain material and annealing conditions at temperatures above ~1,000 K. The amount of irradiation-induced volumetric swelling of the specimens was an indication of the total stored energy. Based on this observation, the energy release behavior was described by a swelling recovery model.

36 MATERIALS SCIENCE↗
DOE OSTI2022

High-burnup Experiments in Reactivity Initiated Accidents (HERA)

High-burnup Experiments in Reactivity-initiated Accidents (HERA) is a joint experimental program (JEEP) operating within the United States (U.S.) Nuclear Energy Agency’s (NEA’s) framework for irradiation experiments (FIDES). HERA is dedicated to the understanding of light water reactor (LWR) fuel performance at high burnup under reactivity-initiated accidents (RIA). In-pile RIA experiments have been performed on high-burnup fuels (e.g., above 60 gigawatt days per metric ton of uranium [GWd/MTU]) in the CABRI reactor at the Cadarache site in southern France, and the Nuclear Safety Research Reactor (NSRR) in Japan. However, most of these experiments have taken place with heavily corroded Zircaloy claddings with pulse widths that are narrower (e.g., 5 ms – 30 ms full-width-half-max [FWHM]) than what would be likely in a commercial LWR (e.g., 30 ms – 80 ms FWHM). A few tests were performed in the CABRI facility with FWHM up to 76 ms, though only one “blistered” rod experienced cladding failure. Heavy waterside corrosion and narrow pulse widths are both known to increase the vulnerability of LWR fuel to pellet cladding mechanical interaction (PCMI). The HERA proposal is designed to: (1) quantify the impact of pulse width on fuel performance, offering new insight into the applicability of existing data; (2) generate new data on high burnup fuel under pulse conditions prototypic of LWRs; (3) quantify the additional margin provided by modern cladding alloys to PCMI failure limits; and (4) offer improved data for modelers using specially designed tests that eliminate key uncertainties in high-burnup fuel tests.

(HERA)↗
DOE OSTI2022

Hydriding, Oxidation, and Ductility Evaluation of Cr-Coated Zircaloy-4 Tubing

Accident-tolerant fuel concepts have been developed recently in diverse research programs. Recent research has shown clear advantages of Cr-coated Zr cladding over bare cladding tubes regarding oxidation behavior under the design basis loss-of-coolant accident condition. However, limited data are available about the hydriding behavior of the Cr coating. For that purpose, Cr-coated Zricaloy-4 tubes were tested to investigate the effects of hydriding, oxidation, and postquench ductility behavior on coated Zr cladding. A high-power impulse magnetron sputtering (HiPIMS) process was used to produce a high-density coating on the Zircaloy-4 tube surface. Coated and uncoated Zircaloy-4 tube specimens underwent one-sided hydriding in a tube furnace filled with pure hydrogen gas at 425 °C. The tubing specimen ends were sealed with Swagelok plugs before the hydriding runs. For uncoated specimens, H analysis of the hydrided specimens indicated that the H content increased as the test time and initial pressure increased. However, almost no change was observed for the coated specimens that were hydrided under the same test conditions. After one-sided hydriding, the hydrided coated and uncoated specimens were exposed to steam at high temperatures for two-sided oxidation studies to simulate accident conditions. The coated specimens showed a slower oxidation: oxygen pickup was 50% lower than the uncoated specimens tested under the same conditions. Ring compression testing was performed to evaluate the embrittlement behavior of the Cr-coated specimens after hydriding and oxidation. The results indicated that the HiPIMS coating provides excellent protection from hydriding and oxidation at high temperatures.

36 MATERIALS SCIENCE↗
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↗
DOE OSTI2019

Progress Report on Modified Burst Testing and Alternative Test Methodologies

Performance of accident-tolerant fuel candidates must be assessed during the pellet-cladding mechanical interaction (PCMI) phase of a design-basis reactivity-initiated accident (RIA). Therefore, in this report, we describe Fiscal Year (FY) 2019 research activities on the development of separate-effects testing for ATF cladding candidate iron-chromium-aluminum (FeCrAl) tube specimens under loading conditions comparable to the PCMI. The first section of this report provides modified-burst test (MBT) results of unirradiated FeCrAl tubes at various pressurization-rates and an RIA-relevant temperature (275°C). The failure strain of FeCrAl was observed and calculated using a high-speed digital-image correlation (DIC) technique developed recently under the Department of Energy (DOE)’s Advanced Fuel Campaign (AFC). To overcome the complexities of performing MBT in a hot-cell environment on irradiated samples, a “plane-strain tension” (PST) test was proposed as an alternative methodology for PCMI testing. The relevance of the PST test as an alternative to MBT for irradiated specimens was explicated using the data available in the literature. The design of hot-cell compatible PST fixtures and a specific holder for PST hot-cell specimen machining were developed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

M5 Alloy Specification for TREAT Conceptual Fuel Cladding Selection

The Idaho National Laboratory (INL) Transient Reactor Test Facility (TREAT) is a nuclear test facility designed to provide safety data for fuel designs by simulating conditions during various types of nuclear excursions. The M5 alloy cladding material for the conceptual TREAT Low Enriched Uranium fuel core shall be supplied according to this specification.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

In-pile thermal conductivity of uranium dioxide at low burnup

In-pile thermal conductivity of uranium dioxide (UO 2 ) was investigated at low burnup levels (<0.2 kWd/kg-UO 2 ) to elucidate the concurrent effects of the fission damage and the thermal recovery. In-pile experiments were performed in the Halden reactor. The fuel centerline and the cladding temperatures were measured during the experiments, and the power level of specially designed test rodlets were also monitored. The uncertainty of the pellet-cladding gap's thermal resistivity was minimized with a liquid-metal bond, while comparable measurements were made in reference test rodlets without liquid metal bonding. The experimental data were analyzed using the inverse heat transfer approach. The heat conduction equation was solved by applying the measured temperature of the cladding as a boundary condition, and the calculated fuel centerline temperature was compared with the measured fuel centerline temperature to determine the relative matrix resistivity. Results showed that the in-pile matrix resistivity, the phonon-lattice interaction terms of the thermal conductivity, was ~1.5–2.5 times higher than its unirradiated value due to fission-induced damage. Finally, the in-pile annealing caused a significant recovery of the matrix resistivity, and the amount of recovery increased from ~20 to ~33%, while the annealing temperature increased from 700 to 1,000 °C.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Development of Innovative Non-Concentrated Alloy (NCA) Fuel Cladding for Advanced Nuclear Reactors

As the nuclear energy sector advances toward next-generation reactors, the need for high-performance fuel cladding materials has become increasingly urgent. Traditional alloys like zirconium and stainless steel are reaching their performance limits under higher temperatures, more corrosive coolants, and extended irradiation. This report presents the development of a new class of fuel cladding materials based on Non-Concentrated Alloys (NCAs) ? multi-element systems designed to deliver enhanced mechanical strength, corrosion resistance, and radiation tolerance. Through a combination of computational modeling (Computer Coupling of Phase Diagrams and Thermochemistry: CALPHAD), simulation-guided alloy selection, and experimental fabrication via arc melting and spark plasma sintering (SPS), three strategic alloy design paths were explored: (1) FeCrAl-based NCAs, (2) refractory-lean neutron-efficient alloys, and (3) equimolar high-entropy compositions. Microstructural analysis confirmed the formation of stable body-centered cubic BCC_A2 phases, while mechanical testing demonstrated hardness values significantly exceeding those of conventional cladding materials. The results highlight the tunability of NCA systems and their potential for balancing strength and ductility ? a critical consideration for in-reactor performance. Looking forward, future work will focus on thermomechanical optimization, CALPHAD refinement, and benchmarking against industry standards to enable scalable deployment. This work not only advances the science of nuclear materials but also supports broader goals in nuclear safety, performance, and nuclear energy innovation.

36 - MATERIALS SCIENCE↗