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At least 451 records · Page 25
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Westinghouse Accident Tolerant Fuel Phase 2B with Higher Enriched and Higher Burnup Add-On Project Final Technical Report Deliverable Volume 2

The Westinghouse Electric Company LLC (Westinghouse) accident tolerant fuel (ATF) program funded by Westinghouse and the US Department of Energy (DOE) utilized chromium (Cr) coated zirconium alloy cladding with doped UO 2 (ADOPT(TM)) and uranium silicide (U 3 Si 2 ) high density/high thermal conductivity fuel for its lead test rod (LTR) program with irradiation beginning in 2019. ADOPT is a Cr 2 O 3 +Al 2 O 3 doped UO 2 pellet with increased oxidation resistance and density, and increased resistance to fission gas release. Due to the issues with U 3 Si 2 fuel reaction in pressurized water reactor (PWR) environments, uranium nitride (U 15 N) has been substituted for U 3 Si 2 as a long-term fuel option. Cr coated cladding with ADOPT fuel is the near-term Westinghouse EnCore ® fuel product. The long-term EnCore fuel product is SiGA ® SiC/SiC composite cladding with high density/high thermal conductivity UN fuel. In 2020, the higher burnup and higher enriched and (HBHE) program was integrated into the ATF program. The objective of this expanded program is to extend ATF burnups to at least 75 MWd/kgU to economically facilitate 24-month cycles in PWRs. The ATF program now includes ADOPT fuel with enrichments >5% 235 U. Over the past several years, Westinghouse has tested the Cr coated zirconium (Zr) and silicon carbide (SiC) claddings in Westinghouse Churchill autoclaves and the Massachusetts Institute of Technology (MIT) reactor. Additionally, the Cr coated cladding has been tested in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and in LTR programs at the Doel 4 and Bryon 2 commercial reactors. High temperature tests at the state-of-the-art Westinghouse facilities in Churchill, PA and at Karlsruhe Institute of Technology (KIT) have been carried out to determine the time and temperature limits for the Cr coated zirconium claddings. These tests indicate that Cr coated Zr cladding can take temperatures up to about 1500°C for short periods of time without becoming totally oxidized. The main issue is the formation of the Cr-Zr eutectic at 1333°C resulting in the migration of this eutectic inward with the formation of ZrCr 2 which rapidly oxidizes to ZrO 2 on the outside of the tube. The Cr coated Zr cladding also delays the bursting of the tube, reduces the burst area, and reduces hydriding and embrittlement of the Zr which is a major benefit for reducing fuel fragmentation, redistribution, and dispersal (FFRD), the major licensing issue faced by HBHE. The manufacturing parameters for the SiC have been found to have a significant effect on the corrosion rate of the SiC in light water reactor (LWR) conditions and significant improvements have been made in the yield, quality, and oxidation resistance of the SiC cladding by identifying and tightening manufacturing process parameters. Current autoclave results for SiC composite claddings indicate that a corrosion rate of fewer than 2 micrometers per year can be achieved which meets corrosion requirements under normal operating conditions. High temperature steam oxidation tests at the state-of-the-art Westinghouse facilities in Churchill, PA and at Karlsruhe Institute of Technology (KIT) have been carried and determined that SiC cladding can withstand temperatures up to about 1800°C to 1900°C before excessive corrosion reactions begin, and that the SiC will not balloon and burst. ADOPT fuel pellets were incorporated into the Westinghouse ATF program because of their increased density and oxidation resistance. A topical report supporting the use of ADOPT pellets has been submitted to the NRC and will be approved shortly. ADOPT pellets have been a product in Europe for over 15 years. The additives work to make larger grain sizes which appear to reduce fission gas release during transients as well as increase the density of the pellet. Fuel rod and assembly design in preparation for the lead test rod (LTR) and lead test assembly (LTA) programs is underway as well as licensing efforts with the Nuclear Regulatory Commission (NRC). Finally, accident analyses coupled with economic evaluations for both operating savings as well as fuel savings have been initiated. Modular Accident Analysis Program 5 (MAAP5) calculations indicate that both Cr coated cladding and SiC composite cladding can increase the time to fuel rod loss of geometry for up to two hours longer than current Zr based cladding in a station blackout scenario. This additional time is due to the much higher oxidation resistance of the SiC and Cr coated cladding. These two hours can be used to implement additional response options instituted through the Diverse and Flexible Mitigation Capability (FLEX) program by reactor operators. Both ATF cladding options also reduce hydrogen production which dramatically reduces primary system and containment pressure and the risk of fission product release beyond containment in the unlikely event of an accident. The lower pressure in the system allows more time to feed cooling water to the core, resulting in the avoidance of fuel melting. The coping time is extended indefinitely if the modest water flow provided by FLEX continues. Experimental work on methods to rapidly fabricate SiC composite structures with high density and reduce the fabrication price of SiC fibers while maintaining a high level of performance is needed. Methods for mitigating or stopping the Cr-Zr eutectic are also needed to further improve the oxidation resistance of Cr coated Zr. Minimal (<1%) swelling of U 3 Si 2 and subsequent fission gas release has been demonstrated up to 20 MWd/kgU. Irradiation experiments with U 3 Si 2 fuel in ATR to determine these properties at 40 MWd/kgU were completed but post irradiation examinations were not done since U 3 Si 2 has been replaced by UN. UN has three issues that are being addressed. The first is increasing the oxidation resistance so that there is not excessive reaction up to ~1500°C. While UN increases the pellet density and additives to the UN postponed the temperature at which rapid oxidation occurred by ~100°C to 200°C, this is not enough, and pellet coating options are now being pursued. The second is developing a method to manufacture that does not require the multi-step process of UF 6 to UO 2 to UC to UN. Efforts are underway looking at UF 6 to UN 2 to UN and UF 6 to UF 4 to UN 2 to UN reactions. Finally, an economically acceptable process for potentially enriching the 15 N content of the nitrogen used to make UN to >95% 15 N was identified though not experimentally validated. In addition to the work supported by the DOE, research and testing activities are being carried on in a world-wide effort funded by many countries such as Sweden, United Kingdom, Belgium, Netherlands, Spain, Germany, Japan, and France. This work is being facilitated through the Westinghouse led Collaboration for Advanced Research on Accident Tolerant Fuel (CARAT) program. Annual meetings were organized in 2018 and 2019 by Westinghouse as a venue for presentation of this work and to provide for the cross-fertilization of ideas among the many outstanding researchers in the ATF area. No meetings were held in 2020 and 2021 due to the COVID-19 related restrictions on traveling and gatherings. Since SiC, coated cladding, and high-density fuel options are not currently used in the nuclear industry; support from the government and industry members is needed to further the significant effort of setting new standards. The same is true for the Nuclear Regulatory Commission (NRC) which must review and approve the commercial use of these new fuels since all current regulations are oriented toward Zr/UO 2 fuel. Several meetings have been held with the NRC to generate a fast-track approach to licensing ATF using a combination of atomic modeling, in-rod sensors, and in-reactor testing. This approach was memorialized in an “Accelerated Fuel Qualification White Paper” prepared by the Accelerated Fuel Qualification Working Group lead by General Atomics.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Behavior of contaminants and simulant fission products in the advanced chlorination process

The University of Tennessee, Knoxville and Oak Ridge National Laboratory have completed proof-of principle testing on an advanced sulfur-based zirconium and aluminum chlorination process for the recovery of Zr or Al from spent fuel claddings and purification of the products. This process could be used to recover hafnium-free nuclear-grade zirconium for reuse or disposal as low-level waste. Testing showed that even non-optimized reaction protocols could produce zirconium product streams from simulated irradiated fuel with purities >90%. The chlorination process was tested on representative fission product species that would be found in Advanced Test Reactor fuel (ATR). First, fission product generation was determined through modeling and simulation of the ATR. The generation of fission products was used to selected contaminants and potentially volatile fission product simulants for laboratory testing. The goal is to identify which species may react with the sulfur-chlorine reagents and be transported into the advanced chlorination off-gas.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Studies on the Vitrification Potential of High Risk Spent Nuclear Fuels

This research explored vitrification of U-Zr nuclear fuel that exhibits an explosion hazard when processed by common nitric acid dissolution. Glass frits were chosen based on properties that had been previously measured as well as being demonstrated successfully in production at the Defense Waste Process Facility (DWPF). Zirconium and Aluminum metal powders were chosen to represent the spent fuels after literature demonstrated that the potential phase that causes the issue for explosive reaction in the effluent stream is tied to the Zirconium-III phase field. DSC measurements were carried out on various simulants and waste forms to better understand thermal behavior of these materials. Zirconium (Zr) was tested with both nitric acid and the new effluent flowsheet that is under development by the SRNL Chemical Process Control (CPC) group. The nitric acid test did show the expected exothermic behavior as well as how the behavior is affected by particle size of the powders used. In contrast, the effluent samples did not show an exothermic peak. However, the concentration of metal in the samples was orders of magnitude lower than in the nitric acid testing and further testing of the effect of concentration on the exothermicity that could be measured confirmed a dampened signal. Tests were run that included frit, both DWPF Frit 510 and Iron Phosphate frit, exhibiting complex curves that included multiple endothermic and exothermic peaks during heating and cooling. The final crucible samples appear similarly to the bulk glasses pointing to these peaks being a part of the dissolution of the simulant into the base glasses.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Blending as an In-Tank Waste Processing Strategy: Year 1 Progress Report

The Hanford Site stores approximately 56 million gallons of radioactive legacy defense waste in underground tanks. Direct feed of southeast quadrant tank waste to a high-level waste treatment facility is currently being explored to accelerate tank treatment activities. This study seeks to de-risk waste transfers by favorably altering the resuspension and critical velocities of wastes through strategic waste blending. Wastes rich in aluminum- and zirconium-bearing solids pose significant risks due to their high settled strength and density. A series of tests explores how solid phase particle size and suspending phase chemistry alter waste transport properties, quantified through the just-suspended mixing speed (NJS), for pure and blended aluminum- and zirconium-bearing waste solid simulants. The study utilizes a range of gibbsites (three sizes spanning 10 to 90 µm) and zirconia solids (three sizes spanning 10 to 100 µm). As expected, the NJS of pure solids largely follows predictions based on literature correlations; blends of gibbsite and zirconium-bearing solids of differing particle sizes exhibit NJS maxima attributed to dense packing of the mixed-size settled solids. The findings of this study will assist in developing strategies to optimize the transport of waste solids for the safe and efficient delivery of tank waste.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

A Study of a Surrogate Nuclear Thermal Propulsion Fuel Element

We report a surrogate fuel, Zirconium Carbide/Titanium Nitride has been prepared using a simple method to obtain a more uniform distribution of titanium nitride in the zirconium carbide matrix. Powders were sintered using Spark Plasma Sintering at 1750oC and 1850oC using a modified sintering profile. This resulted in a fairly uniform distribution of titanium nitride in the zirconium carbide matrix with a density of 95% of theoretical at 1850oC. Scanning electron microscopy, indentation hardness and x-ray diffraction results are given.

36 MATERIALS SCIENCE↗
NASA NTRS1973

High temperature, low-cycle fatigue of copper-base alloys in argon. Part 1: Preliminary results for 12 alloys at 1000 F (538 C)

Short-term tensile evaluations at room temperature and 538 C and low-cycle fatigue evaluations at 538 C are presented for the following materials: Zirconium copper-annealed, Zirconium copper-1/4 hard, Zirconium copper-1/2 hard, Tellurium copper-1/2 hard, Chromium copper-SA and aged, OFHC copper-hard, OFHC copper-1/4 hard, OFHC copper-annealed, Silver-as drawn, Zr-Cr-Mg copper-SA, CW and aged, Electroformed copper-30-35 ksi, and Co-Be-Zr- copper-SA, aged. A total of 50 tensile tests and 76 low-cycle fatigue tests were performed using a strain rate of 0.2 percent per second.

Conway, J. B.↗
NASA NTRS1974

Compatibility of refractory materials for nuclear reactor poison control systems

Metal-clad poison rods have been considered for the control system of an advanced space power reactor concept studied at the NASA Lewis Research Center. Such control rods may be required to operate at temperatures of about 140O C. Selected poison materials (including boron carbide and the diborides of zirconium, hafnium, and tantalum) were subjected to 1000-hour screening tests in contact with candidate refractory metal cladding materials (including tungsten and alloys of tantalum, niobium, and molybdenum) to assess the compatibility of these materials combinations at the temperatures of interest. Zirconium and hafnium diborides were compatible with refractory metals at 1400 C, but boron carbide and tantalum diboride reacted with the refractory metals at this temperature. Zirconium diboride also showed promise as a reaction barrier between boron carbide and tungsten.

Sinclair, J. H.↗
NASA NTRS1989

Study of metal corrosion using ac impedance techniques in the STS launch environment

AC impedance measurements were performed to investigate the corrosion resistance of 19 alloys under conditions similar to the STS launch environment. The alloys were: Zirconium 702, Hastelloy C-22, Inconel 625, Hastelloy C-276, Hastelloy C-4, Inconel 600, 7Mo + N, Ferralium 255, Inco Alloy G-3, 20Cb-3, SS 904L, Inconel 825, SS 304LN, SS 316L, SS 317L, ES 2205, SS 304L, Hastelloy B-2, and Monel 400. AC impedance data were gathered for each alloy after one hour immersion time in each of the following three electrolyte solutions: 3.55 percent NaCl, 3.55 percent NaCl-0.1N HCl, and 3.55 percent NaCl-1.0N HCl. The data were analyzed qualitatively using the Nyquist plot and quantitatively using the Bode plot. Polarization resistance, Rp, values were obtained using the Bode plot. Zirconium 702 was the most corrosion resistant alloy in the three electrolytes. The ordering of the other alloys according the their resistance to corrosion varied as the concentration of hydrochloric acid in the electrolyte increased. The corrosion resistance of Zirconium 702 and Ferralium 255 increased as the concentration of hydrochloric acid in the electrolyte increased. The corrosion resistance of the other 17 alloys decreased as the concentration of the hyrdochloric acid in the electrolyte increased.

Calle, Luz M.↗
NASA NTRS1996

Cryogenic Brush Seal Test Results

Brush seals are compliant, contact seals that have long-life, low-leakage characteristics desirable for use in rocket engine turbopumps. 50.8-mm (2.0 inch) diameter brush seals with a nominal initial radial interference of 0.127-mm (0.005 inch) were tested in liquid nitrogen at shaft speeds up to 35,000 rpm and differential pressure loads up to 1.21 MPa (175 psi) per brush. The measured leakage rate of a single brush was 2-3 times less than that measured for a 12-tooth, 0.127-mm (0.005 inch) radial clearance labyrinth seal used as a baseline. Stage effects were studied and it was found that two brush seals with a large separation distance leaked less than two brushes tightly packed together. The maximum measured groove depth on the Inconel 718 rotor was 25.4 (mu)m (0.001 inch) after 4.31 hours of shaft rotation. The Haynes-25 bristles wore approximately 25.4-76.2 (mu)m (0.001-0.003 inch) under the same conditions. Three seal runner coatings, chromium carbide, Teflon impregnated chromium, and zirconium oxide, were tested in liquid hydrogen at 35,000 and 65,000 rpm with separate 50.8 mm diameter brush seals made of Haynes-25 bristles and having a nominal initial radial interference of 129 rpm. Two bare Inconel-718 rotors were also tested as a baseline. The test results revealed significant differences between the wear characteristics of the uncoated and coated seal runners. At both speeds the brush seal with the bare Inconel-718 seal runner exhibited significant bristle wear with excessive material transferring to the runner surface. In contrast, the coated seal runners inhibited the transfer and deposit of bristle material. The chromium carbide coating showed only small quantities of bristle material transferring to its surface. The Teflon impregnated chromium coating also inhibited material transfer and provided some lubrication. This coating, however, is self-sacrificing. The Teflon remained present on the low speed runner, but it was completely removed from the high speed brush seal, which was tested considerably longer. The tests of the Teflon coating revealed the importance of using a lubricating and low friction coating for brush seals to reduce bristle and seal runner wear. The zirconium oxide coating exhibited the greatest amount of coating wear, while the brushes incurred only slight wear. Further testing of ceramics is recommended before making a final judgement on the viability of ceramic coatings for brush seals because of the contrast between the results reported by Carlile and the results presented herein. Strictly based on the results presented hereinabove, the chromium carbide and Teflon impregnated chromium coatings were considered preferable to the uncoated Inconel-718 and zirconium oxide coatings because of their good wear resistance and characteristics to inhibit bristle material wear and transfer to the seal runner.

Proctor, Margaret P.↗
NASA NTRS1995

A Method Based on Radiative Cooling for Detecting Structural Changes in Undercooled Metallic Liquids

We introduce a structure-sensitive parameter for undercooled melts which can be measured in containerless processing experiments. We have established that the ratio, R(T), of hemispherical total emissivity epsilon(sub T)(T) to constant-pressure specific heat c(sub p)(T) can serve as an indicator which is sensitive to any changes in short range atomic order in undercooled metallic melts. R(T) (triple bonds) epsilon(sub T)(T)/c(sub p)(T) values for nickel, zirconium, and silicon have been obtained using the high temperature electrostatic levitator while the levitated melts were undergoing purely radiative cooling into the deeply undercooled region. R(T) plots for undercooled liquid nickel and zirconium indicate no significant change in short-range structure from their melting temperatures to 15% undercooling. In contrast, liquid silicon shows marked short-range structural changes beginning above its melting temperature and extending throughout the undercooled region. The short-range structure of liquid silicon is related to the highly-directional covalent bonding which characterizes its solid form. The nickel and zirconium data show that epsilon(sub T) varies linearly with T, in support of metal emissivity theories.

Rulison, Aaron J.↗
NASA NTRS1998

Experiments on Nucleation in Different Flow Regimes

The vast majority of metallic engineering materials are solidified from the liquid phase. Understanding the solidification process is essential to control microstructure, which in turn, determines the properties of materials. The genesis of solidification is nucleation, where the first stable solid forms from the liquid phase. Nucleation kinetics determine the degree of undercooling and phase selection. As such, it is important to understand nucleation phenomena in order to control solidification or glass formation in metals and alloys. Early experiments in nucleation kinetics were accomplished by droplet dispersion methods. Dilitometry was used by Turnbull and others, and more recently differential thermal analysis and differential scanning calorimetry have been used for kinetic studies. These techniques have enjoyed success; however, there are difficulties with these experiments. Since materials are dispersed in a medium, the character of the emulsion/metal interface affects the nucleation behavior. Statistics are derived from the large number of particles observed in a single experiment, but dispersions have a finite size distribution which adds to the uncertainty of the kinetic determinations. Even though temperature can be controlled quite well before the onset of nucleation, the release of the latent heat of fusion during nucleation of particles complicates the assumption of isothermality during these experiments. Containerless processing has enabled another approach to the study of nucleation kinetics. With levitation techniques it is possible to undercool one sample to nucleation repeatedly in a controlled manner, such that the statistics of the nucleation process can be derived from multiple experiments on a single sample. The authors have fully developed the analysis of nucleation experiments on single samples following the suggestions of Skripov. The advantage of these experiments is that the samples are directly observable. The nucleation temperature can be measured by noncontact optical pyrometry, the mass of the sample is known, and post-processing analysis can be conducted on the sample. The disadvantages are that temperature measurement must have exceptionally high precision, and it is not possible to isolate specific heterogeneous sites as in droplet dispersions. Levitation processing of refractory materials in ultra high vacuum provides an avenue to conduct these kinetic studies on single samples. Two experimental methods have been identified where ultra high vacuum experiments are possible; electrostatic levitation in ground-based experiments and electromagnetic processing in low earth orbit on TEMPUS. Such experiments, reported here, were conducted on zirconium. Liquid zirconium is an excellent solvent and has a high solubility for contaminants contained in the bulk material as well as those contaminants found in the vacuum environment. Oxides, nitrides, and carbides do not exist in the melt, and do not form on the surface of molten zirconium, for the materials and vacuum levels used in this study. Ground-based experiments with electrostatic levitation have shown that the statistical nucleation kinetic experiments are viable and yield results which are consistent with classical nucleation theory. The advantage of low earth orbit experiments is the ability to vary the flow conditions in the liquid prior to nucleation. The put-pose of nucleation experiments in TEMPUS was to examine.

Bayuzick, R. J.↗
NASA NTRS2010

Coating for components requiring hydrogen peroxide compatibility

The present invention provides a heretofore-unknown use for zirconium nitride as a hydrogen peroxide compatible protective coating that was discovered to be useful to protect components that catalyze the decomposition of hydrogen peroxide or corrode when exposed to hydrogen peroxide. A zirconium nitride coating of the invention may be applied to a variety of substrates (e.g., metals) using art-recognized techniques, such as plasma vapor deposition. The present invention further provides components and articles of manufacture having hydrogen peroxide compatibility, particularly components for use in aerospace and industrial manufacturing applications. The zirconium nitride barrier coating of the invention provides protection from corrosion by reaction with hydrogen peroxide, as well as prevention of hydrogen peroxide decomposition.

Yousefiani, Ali↗
NASA NTRS2006

Alpha-Voltaic Sources Using Liquid Ga as Conversion Medium

A family of proposed miniature sources of power would exploit the direct conversion of the kinetic energy of alpha particles into electricity. In addition to having long operational lives, these sources are expected to operate with energy-conversion efficiencies from 70 to 90 percent. A power source as proposed (see figure) would be an electrolytic cell in which liquid gallium would serve as both an electrolyte and an energy-conversion medium. The cell would contain an iridium cathode and a zirconium anode. The alpha particles, each with a kinetic energy approx.5.8 MeV, would be emitted by radioactive decay of Cm-244, which has a half-life of 18 years. The Cm-244 source would be positioned so that the a particles would enter the liquid gallium, where their kinetic energy would be dissipated mostly through ionization of Ga atoms, creating Ga(+) ions and free electrons. The electrons would be collected by iridium cathode, and the Ga(+) ions would be neutralized at the zirconium cathode by electrons returning after flowing through an external circuit. Gallium is a candidate for use as the electrolyte and the energy-conversion medium because in the liquid state it is a semimetal: its electrical conductivity is greater than that of a typical semiconductor but small in comparison with the conductivities of metals. Consequently, in liquid gallium, electrons and Ga(+) can exist without immediate recombination and can be moved by electric fields. It is expected that electric fields, resulting at least partly from the difference between the work functions of the electrode metals, would move the electrons and ions to their respective electrodes. The open-circuit potential of the cell is expected to be 1.62 V - equal to the difference between the work functions of iridium and zirconium. Unlike in a solid-state energy conversion medium, the impingement of energetic a particles would not give rise to displacement damage in the liquid gallium. Hence, the cell should have a long life, limited only by the half-life of Cm-244. A cell having a volume less than 25 cu mm, containing 1 curie of Cm-244 (the curie is a unit of radioactivity equal to 3.7 10(exp 10) disintegrations per second) is expected to deliver a current between 7 and 12 mA, which, at the expected open-circuit potential, would provide a power in the approximate range of 11 to 20 mW.

Patel, Jagdish U.↗
DOE OSTI2023

Carbide-based fuel assembly for thermal propulsion applications

Carbide-based fuel assembly includes outer structural member of ceramic matrix composite material (e.g., SiC—SiC composite), insulation layer of porous refractory ceramic material (e.g., zirconium carbide with open-cell foam structure or fibrous zirconium carbide), and interior structural member of refractory ceramic-graphite composite material (e.g., zirconium carbide-graphite or niobium carbide-graphite). Spacer structures between various layers provide a defined and controlled spacing relationship. A fuel element bundle positioned between support meshes includes a plurality of distributively arranged fuel elements or a solid, unitary fuel element with coolant channels, each having a fuel composition including high assay, low enriched uranium (HALEU). Fuel assemblies are distributively arranged in a moderator block and the upper end of the outer structural member is attached to a metallic inlet tube for hydrogen propellant and the lower end of the outer structural member is interfaced with a support plate, forming a nuclear thermal propulsion reactor.

Barringer, Eric A.↗
DOE OSTI2022

Post-Transient Examination Results of RIA Commissioning Teats at the Transient Reactor Test Facility

Six reactivity-initiated-accident (RIA) commissioning tests have been performed at Idaho National Laboratory’s (INL) Transient Reactor Test Facility. Five of these tests were performed using fresh fuel rodlets, and the sixth test was performed using a previously irradiated rodlet from the ATF-2 irradiation experiment. These experiments demonstrate the ability to perform RIA testing including the ability to perform experiments using previously irradiated materials. Post-transient examinations revealed pellet cladding interactions and cladding ballooning depending on the initial boundary conditions of the test. Increased thickness of both zirconium oxide and alpha-zirconium were found in the cladding with tests performed at higher levels of total energy deposition. No oxide or alpha-zirconium was found in the test with the greatest amount of ballooning, which was also the test with the highest initial rod internal pressure. The highest energy deposition test (1110 J/g) resulted in failure of the rodlet. Fragmentation of the fuel pellets occurred in this test with the size of the fragments decreasing inversely to the pellet radius.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Formation and characterization of hydride rim structures in Zircaloy-4 nuclear fuel cladding tubes

Zirconium alloy tubes used as nuclear fuel cladding are subject to oxidation and subsequent hydrogen pickup during their long service in commercial light water reactors. The hydrogen picked up in the cladding can precipitate as a brittle hydride rim feature on the cladding outer surface. Here, to better understand the effect of hydride rims on the fracture behavior of stress-relieved zirconium alloy cladding tubes, a procedure to produce these rim-like structures has been developed and is described herein. Extensive characterization of the ‘as hydrided’ tubes is performed. The hydrogen charging apparatus consists of a tube furnace with a quartz chamber that is connected to a vacuum pump as well as a bottle of pure hydrogen gas. The charging station uses a static charge of hydrogen as opposed to a flowing gas. The chief advantage of this approach is the ability to monitor the pressure drop in the hydriding chamber and correlate this pressure drop to a known rate of hydrogen pickup in the cladding tube. It was found that the hydrogen partial pressure, metal temperature, and surface treatment all clearly played a role in whether a hydride rim was formed. Extensive characterization of the hydride rims shows they consist of needle like platelets of δ phase hydrides (ZrH1.66) oriented in the circumferential direction with a radial spacing of several microns in a sandwich-like structure.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Step-loaded creep testing of Zircaloy-4 cladding at higher temperatures in the α-phase

A refined understanding of zirconium-based cladding thermomechanical performance during rapid transients is essential for enhancing the safety and operation of light-water reactors. Traditional models for zirconium alloys under accident conditions generally assume that creep dominates fuel cladding performance. Here, these historic models have largely remained unchanged and serve as the basis for safety criteria development. As the U.S. nuclear industry pursues higher burnup levels, the increased release of fission gases during transients raises the risk of cladding failure in the low-temperature hcp α-phase, making the fidelity of these models of greater importance. Creep testing was conducted from 550–700°C with 25–120 MPa applied hoop stresses to investigate Zircaloy-4 deformation at accident-relevant temperatures in the α-phase. Step-loading was employed to capture creep behavior across a wide stress range from a single sample. The stress-strain rate data at higher temperatures (650 and 700°C) were well-described by isotropic versions of the Erbacher and Kaddour models, while the lower temperature data (550 and 600°C) were underpredicted by both anisotropic and isotropic model variants. Greater strain rates during the initial loading step at 650 and 700°C were attributed to recrystallization and grain growth of sub-micron crystallites. Yet, texture analysis revealed the basal split texture remained after testing. These observations produced results suggesting Zircaloy-4 claddings experience higher creep rates across the α-phase than previously thought, possibly related to dynamic anisotropy due to temperature dependent activation of deformation mechanisms, effects of biaxial loading, and variation in material condition between the current testing used in previous model development.

36 MATERIALS SCIENCE↗