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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.

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At least 343 records · Page 19
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Evaluation of critical heat flux of ATF candidate coating materials in pool boiling

The critical heat flux (CHF) of potential accident tolerant fuel (ATF) cladding materials and concepts, Cr and FeCrAl coatings on zirconium-alloy substrates, was investigated experimentally. A cold spray process, which involves the propulsion of high velocity powder particles on to the substrate surface to form a coating upon impact, was used to create the coatings. Surface parameters such as contact angle, surface roughness, and morphology that are associated with CHF in pool boiling were characterized. Non-hydrodynamic effects of newly developed ATF candidate coating materials on CHF were also evaluated. The effects of thermal properties and coating thickness and expected challenges for evaluation of CHF for new ATF cladding materials development are discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Design and Evaluation of a Ring Tension Test Grip for Remote Mechanical Testing of Irradiated Tubular Specimens

Here, the ring tension test (RTT) is a mechanical testing method for determining bulk mechanical behavior in the circumferential or hoop direction for tubular materials. The test is especially useful for testing materials with anisotropic mechanical properties, such as zirconium alloys, which are commonly used as nuclear fuel cladding. Anisotropy requires direction-specific testing to determine the hoop strength. Historically, several RTT methods and grips have been used, each method has its strengths and weaknesses, and, in all cases, the measured strength is subject to uncertainty due to variations of the testing geometry and experimental tolerances. Recent analysis has shown that grips with a hemicylindrical mandrel configuration are recommended as the most robust configuration. The two strictest aspects to be controlled are the ability to determine gage region orientation and closely matching the size of the mandrels to the test specimen. This last requirement is particularly challenging when the dimensions of the specimen vary because of environmental effects such as dimensional changes due to irradiation. This paper presents a new RTT grip designed to incorporate this mandrel shape, hold the gage at the desired orientation, be suitable for remote operation in a hot-cell environment, and be adaptable for different sizes or variations in the specimen size. The general description and the unique design features of the test specimen and grips are given in detail. The performance of the grips in mechanical testing, including in a remote hot-cell environment, is also provided.

42 ENGINEERING↗
DOE OSTI2021

Fundamental Studies of Tritium Formation and Diffusivity in Pure and Defective Zircaloy-4 Getters

Zirconium (Zr) and its alloys are used as fuel cladding in nuclear reactors. These materials have their low absorption cross-section for thermal neutrons and have good mechanical and thermal properties. In addition, they are used as tritium getters in tritium-producing burnable absorber rods (TPBARs), in the form of nickel-plated Zircaloy-4 tubes. To produce 3 H, the breading blankets such as lithium aluminate are irradiated in pressurized water. 3 H produced in this process is captured by the getter and 3 H reacts chemically with Zr to form metal hydride (ZrTx). The precipitation of metal hydride results in not only the volume increase and hydrogen embrittlement of zircaloy but also changes in getter’s chemical properties. This degrades the getter’s performance. Therefore, understanding the behavior of 3 H (such as diffusion and solubility) in zircaloy is important to understand getter performance.

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

A Critical Review on the History of Fabricating Monolithic U-Mo Fuel Plates

Fabrication of uranium-molybdenum alloy fuels has been occurring since the early 2000s in support of the development of a high-uranium-density low-enrichment fuel for use in high-performance research and test reactors which operate at relatively low temperatures. The primary fuel form—a thin foil of uranium, alloyed with 10wt% molybdenum, which is coated in a layer of zirconium and then encapsulated with aluminum 6061 as a cladding material—was developed over a number of years.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1973

Feasibility of an on-line fission-gas-leak detection system

Calculations were made to determine if a cladding failure could be detected in a 100-kW zirconium hydride reactor primary system by monitoring the highly radioactive NaK coolant for the presence of I-131. The system is to be completely sealed. A leak of 0.01 percent from a single fuel pin was postulated. The 0.364-MeV gamma of I-131 could be monitored on an almost continuous basis, while its presence could be varified by using a longer counting time for the 0.638-MeV gamma. A lithium-drifted germanium detector would eliminate radioactive corrosion product interference that could occur with a sodium iodide scintillation detector.

Lustig, P. H.↗
DOE OSTI2014

Computer Modeling of Transport of Oxidizing Species in Grain Boundaries during Zirconium Corrosion

Zirconium (Zr) based alloys are widely used as the cladding materials in light-water reactors. The water-side corrosion of these alloys degrades their structural integrity and poses serious safety concerns. During the Zr corrosion process, a thin Zr oxide (ZrO2) layer forms on the alloy surface and serves as a barrier layer for further corrosion. The majority of the oxide has the monoclinic phase. At the transition region between the oxide and the metal, the oxide contains a thin layer of stabilized tetragonal phase. It is found that the texture of the tetragonal layer determines the protectiveness of the oxide for corrosion. The transport of oxidizing species, such as anion defects, cation defects, and electron through the tetragonal oxide layer could be the rate limiting step of the corrosion. The defect diffusion can be affected by the growing stresses and microstructures such as grain boundaries and dislocations. In this work molecular dynamics simulations are used to investigate the anion and cation diffusion in bulk and at grain boundaries in tetragonal ZrO2. The results show that defect diffusion at grain boundaries is complex and the behavior strongly depends on the grain boundary type. For most of the grain boundaries studied the defect diffusion are much slower than in the bulk, implying that grain boundaries may not be fast defect transport paths during corrosion. The connection between the modeling results and published experimental work will also be discussed. This work is funded by the Laboratory Directed Research and Development (LDRD) program at Idaho National Laboratory.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Biaxial Creep Behavior of Nb-Modified Zircaloys

Zirconium-niobium (Zr-Nb) alloys are used as cladding materials to encapsulate radioactive fuel in nuclear reactors. They possess excellent corrosion resistance at high temperatures making it possible to achieve high fuel burnup, directly increasing the thermal efficiency of the reactor. While they are commonly used in recrystallized (Rx) form in boiling water reactors, there is a need to understand the effect of cold work and stress relief (CWSR) on the biaxial creep characteristics of these materials due to their use in pressurized water reactors. In this study, the biaxial creep behaviors of as-received Zr-Nb alloys, HANA and Zirlo®, have been investigated at 500°C and 400°C, respectively, using internally pressurized tubing superimposed with axial load under varied hoop σθ to axial σz stress ratios of 0 to 2 while monitoring both the axial and hoop strains using a linear variable displacement transformer and a laser telemetric extensometer, respectively. The crystallographic textures and creep loci of these as-received Zr-Nb alloys have been evaluated to correlate with the previous studies on recrystallized HANA4 and CWSR Zircaloy-4. The creep locus of HANA4 was found to be unaffected by initial state (CWSR or Rx) and showed close correspondence to planar isotropy while the creep locus of CWSR Zirlo exhibited more resistance to axial deformation than diametrical as per CWSR Zircaloy-4 reported earlier. These differences are shown to arise from grain-shape anisotropy of the CWSR Zirlo and Zircaloy-4. The simulated creep loci using crystallite-orientation distribution functions in conjunction with prism slip models showed excellent agreement with experimental results.

Tillman, Micah↗
DOE OSTI2024

Raman spectroscopy of zirconium hydride

The characterization of zirconium hydride is important in the nuclear industry because of the hydrogen-induced embrittlement of Zircaloy cladding and its use as a neutron moderator. This paper introduces the use of Raman spectroscopy for the characterization of zirconium hydride. First-principles density functional theory (DFT) calculations were used to predict the Raman spectra of ζ-ZrH 0.5 , γ-ZrH, δ-ZrH 1.5 , δ-ZrH 1.66 , and ε-ZrH 2 with all their predicted symmetries; ζ-ZrH 0.5 (P3m1, R$\overline{3}$m, C2/m, Cm, Cmmm, and Pn$\overline{3}$m); γ-ZrH (P222, Ccce, and P4 2 /mmc); δ-ZrH 1.5 (P$\overline{4}$m2, P4 2 /mcm, Fmmm, Pn$\overline{3}$m, Ibam, P2/c, PI, and P4 2 /nnm); δ-ZrH 1.66 (Fm$\overline{3}$m); and ε-ZrH 2 (Fm$\overline{3}$m, R$\overline{3}$m, and I 4 /mmm). Two samples of Zircaloy-4 containing 133 wt ppm and 360 wt ppm hydrogen were characterized by Raman spectroscopy, showing two signal lines at 215 cm -1 and 1,187 cm -1 , which were assigned to the presence of δ-ZrH 1.66 . These signals had a good spatial correlation with visible hydride precipitates in Raman spectroscopy images. In conclusion, this work provides the basis for the characterization of all possible zirconium hydride compositions and structures using Raman spectroscopy.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Zirconium-based alloys, nuclear fuel rods and nuclear reactors including such alloys, and related methods

Zirconium-based metal alloy compositions comprise zirconium, a first additive in which the permeability of hydrogen decreases with increasing temperatures at least over a temperature range extending from 350.degree. C. to 750.degree. C., and a second additive having a solubility in zirconium over the temperature range extending from 350.degree. C. to 750.degree. C. At least one of a solubility of the first additive in the second additive over the temperature range extending from 350.degree. C. to 750.degree. C. and a solubility of the second additive in the first additive over the temperature range extending from 350.degree. C. to 750.degree. C. is higher than the solubility of the second additive in zirconium over the temperature range extending from 350.degree. C. to 750.degree. C. Nuclear fuel rods include a cladding material comprising such metal alloy compositions, and nuclear reactors include such fuel rods. Methods are used to fabricate such zirconium-based metal alloy compositions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Transmission Electron Microscopy Characterization of Fuel Cladding Chemical Interaction between Minor Actinides bearing U-Pu-Zr Fuel and AIM1 Cladding

Minor actinides (MA) significantly contribute to the long-term radiotoxicity of spent nuclear fuel (SNF). Separating MA from SNF and incorporating it into metallic fuels for fast reactor transmutation is a potential method to reduce this radiotoxicity. Here, this study focuses on transmission electron microscopy characterization of two samples from the fuel cladding chemical interaction (FCCI) region of an americium (Am) and neptunium (Np)-bearing (MA-bearing) uranium-plutonium-zirconium (U-Pu-Zr) fuel irradiated in the Phenix fast reactor to 9.5 % FIMA burnup at approximately 550 °C cladding temperature. The results show that despite the complex chemical interactions between MA and AIM1 cladding elements, excessive FCCI was not induced, and Am penetration depth in the cladding limited to less than 4 µm. Np remained mostly inside fuel. The Zr-rich compounds layer effectively limited the accumulation of lanthanide on the inner cladding surface. Overall, the FCCI behavior between investigated MA-bearing U-Pu-Zr fuel and AIM1 cladding is benign.

Chemical interaction↗
DOE OSTI2016

Interfacing VPSC with finite element codes. Demonstration of irradiation growth simulation in a cladding tube

This Milestone report shows good progress in interfacing VPSC with the FE codes ABAQUS and MOOSE, to perform component-level simulations of irradiation-induced deformation in Zirconium alloys. In this preliminary application, we have performed an irradiation growth simulation in the quarter geometry of a cladding tube. We have benchmarked VPSC-ABAQUS and VPSC-MOOSE predictions with VPSC-SA predictions to verify the accuracy of the VPSCFE interface. Predictions from the FE simulations are in general agreement with VPSC-SA simulations and also with experimental trends.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Advanced Low-Temperature Chlorination of Zirconium

Recovery of Zr from nuclear fuel is of high interest because the Zr cladding material comprises up to 50% of the high-level radioactive waste (HLW) that requires disposition. An ideal Zr-recovery process would not only recover the bulk Zr from the fuel or cladding but would also provide decontamination from the components that give rise to the HLW designation. A simplified decontamination pathway providing recovered Zr in compliance with low-level radioactive waste acceptance criteria could significantly reduce the burden of HLW requiring geologic disposition. A newly developed advanced low-temperature chlorination process presents an opportunity to recover a ZrCl4 product that is effectively decontaminated from the elements of concern. This advanced low-temperature chlorination process uses a mixture of the sulfur-containing chlorination compounds sulfur monochloride (S 2 Cl 2 ) and thionyl chloride (SOCl 2 ).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Annealing of (DU-10Mo)-Zr Co-Rolled Foils

Producing uranium-10wt% molybdenum (DU-10Mo) foils to clad with Al first requires initial bonding of the DU-10Mo foil to zirconium (Zr) by hot rolling, followed by cold rolling to final thickness. Rolling often produces wavy (DU-10Mo)-Zr foils that should be flattened before further processing, as any distortions could affect the final alignment and bonding of the Al cladding to the Zr co-rolled surface layer; this bonding is achieved by a hot isostatic pressing (HIP) process. Distortions in the (DU-10Mo)-Zr foil may cause the fuel foil to press against the Al cladding and thus create thinner or thicker areas in the Al cladding layer during the HIP cycle. Post machining is difficult and risky at this stage in the process since there is a chance of hitting the DU-10Mo. Therefore, it is very important to establish a process to flatten and remove any waviness. This study was conducted to determine if a simple annealing treatment could flatten wavy foils. Using the same starting material (i.e. DU-10Mo coupons of the same thickness), five different levels of hot rolling and cold rolling, combined with five different annealing treatments, were performed to determine the effect of these processing variables on flatness, bonding of layers, annealing response, microstructure, and hardness. The same final thickness was reached in all cases. Micrographs, textures, and hardness measurements were obtained for the various processing combinations. Based on these results, it was concluded that annealing at 650°C or higher is an effective treatment to appreciably reduce foil waviness.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Synchrotron micro-computed tomography analysis of neutron-irradiated U-Mo fuel

The three-dimensional (3D) microstructure of neutron-irradiated uranium-10 wt.% molybdenum (U-10Mo) fuel with a burn-up of 9.8 × 10 21 fissions/cm 3 was characterized using a novel, multi-modal synchrotron micro-computed tomography approach combining propagation-based phase-contrast enhanced and absorption contrast techniques. The porosity development, porosity interconnectedness, swelling, composition, local thickness of the zirconium (Zr) diffusion barrier, and the influence of the fuel–cladding interaction on the local composition and pore morphology, were uniquely determined in 3D. Two cuboids were produced using a focused ion beam-scanning electron microscope at the Zr diffusion barrier–fuel interface and in the bulk fuel. The bulk fuel sample swelled by 53.3 [+9.7/−3.1]%, while the fuel near the Zr–fuel interface swelled by 63.3 [+14.7/−7.2]%. The average local thickness of the Zr diffusion barrier decreased by 53 %, compared to the expected pre-irradiated thickness. Four pore morphology regions were identified initiating parallel to the fuel–Zr interaction region: (1) an interaction layer of suppressed porosity, (2) a layer of elongated and interconnected porosity, (3) a transition zone of low porosity, and (4) a layer of unoriented porosity representative of the bulk fuel behavior. The increase in porosity near the diffusion barrier corresponded to a higher U concentration compared to that in the bulk fuel. The interconnected porosity in the fuel near the diffusion barrier was extensive and oriented parallel to the diffusion barrier, while the bulk fuel had more compact and isolated pore networks. The interaction layer, despite having suppressed porosity, was nearly 100 wt.% U. Porosity suppression at the diffusion barrier corresponds to the expected reduction in radiation-driven diffusion of Xe at the interface despite the anticipated increase in fission product nucleation originating from a higher U concentration. In conclusion, the novel 3D insights of the porosity, swelling, and compositional variations characterized herein can improve the fidelity of fuel performance codes for proliferation-resistant fuels for research and test reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Application of Phase Field Simulations to Fuel Behavior

The application of the phase filed method to simulate the formation and the stress induced re-orientation of zirconium hydride during dry storage of the spent fuel and clad assembly is discussed. The phase field technique is able to capture qualitatively the effect of external stress on the hydride orientation in Zr-H system. However, the modeling effort to-date is far from adequate and several issues remain to be addressed before the simulations can be used as a predictive tool for the behavior of the clad during long time dry storage.

elastic energy↗
DOE OSTI2024

UNCERTAINTIES FOR IN-REACTOR TESTING OF TRITIUM LOADED GETTERS WITH TEMPERATURE CONTROL

The fourth experiment in TPBAR Materials Irradiation Separate-Effects Test (TMIST) series (TMIST-4) is planned to experimentally determine the effect of neutron irradiation on the equilibrium vapor pressure of tritium over a tritium loaded Nickel-Plated Zirconium (NPZ) getter in a simulated TPBAR service environment. This pressure will be used to estimate the tritium permeation rate. In the thermal design in the TMIST-4, two evaluations are required. 1) The cladding temperature is calculated to estimate the feasibility of detecting potential irradiation enhancement of the partial pressure for an NPZ getter. This work is required to obtain a precise cladding temperature for measurement of a low tritium partial pressure. This work has been completed and indicated that the required/target cladding temperatures based on the TMIST-4 test plan can be satisfied under the test condition range by modifying the capsule component shapes and inserting a thermally conductive metal sleeve. 2) The evaluation of uncertainties associated with the test design and operation is performed. This includes fabrication tolerances, and power and nuclear operating conditions. These factors will impact the component temperatures inside capsules and resulting the tritium partial pressure. The tritium permeation rate is a function of temperature and tritium partial pressure. The prediction of tritium permeation rates affects the feasibility of obtaining measurement resolutions required to reach reasonable conclusion for potential irradiation enhanced performance. This work will investigate the uncertainties.

TTP, TPBAR, TMIST-4, TPBAR Materials Irradiation S↗
DOE OSTI2021

Microstructural characterization of as-fabricated monolithic plates with boron carbide, aluminum boride, and zirconium boride burnable absorbers

The use of burnable absorbers can be beneficial for nuclear reactors by extending the fuel’s operational cycle, providing additional criticality control, and flattening the power profile. In this work, three burnable absorber materials (boron carbide, aluminum boride, and zirconium boride) embedded in aluminum have been fabricated into foils and clad in AA-6061 for potential use in high performance research reactors. The as-fabricated boron-containing phases were determined using transmission electron microscopy to be AlB2, B4C, and ZrB2. TEM also revealed weak bonding at the B4C-matrix interface. SEM showed a relatively uniform spatial distribution of boron-containing phases for all the candidate materials. Higher porosity was observed in the foil containing ZrB2 in its as-rolled condition. The porosity in the ZrB2 foil was reduced by hot isostatic pressing. The size and shape distributions of the boron-containing phases were analyzed on the criteria of cross-sectional area, perimeter, roundness, circularity, and aspect ratio. A method of converting the 2D burnable absorber dispersoids seen in cross-sectional microscopy images into 3D volumes was derived using both spherical and ellipsoidal geometry models. The difference in calculated burnable absorber dispersoid average volume between the two models ranges from 20% to 100%, which could impact burnable absorber burnout rates due to differences in neutron self-shielding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗