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At least 721 records · Page 40
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

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

Zirconium (Zr) alloys have excellent mechanical properties at high temperature and are resistant to corrosion in different environmental conditions. These alloys are used as tritium (3H or T) getters in nuclear reactors due to their low absorption cross section to thermal neutrons and excellent thermo-mechanical properties. Zr alloys have better mechanical and thermal properties than many other refractory alloys including stainless steel. After the irradiation of TPBAR by neutron flux, 3H produced during irradiation diffuses though the pellets and is captured by the getter. 3H chemically reacts with Zr metal to form metal hydride (ZrTx). Hydrides thus formed are brittle and adversely affect the mechanical and thermal properties of alloys. The lattice mismatch at the interface of alloy and metal hydrides also creates a stress that significantly reduces the alloy’s performance and operation life. Therefore, understanding the behavior of 3H and its species becomes significant as fuel burnup is increased, which leads to increase in hydrogen pickup and oxide formation.

Duan, Yuhua↗
DOE OSTI2021

Zeolitic catalytic conversion of alcohols to olefins

A catalyst composition for converting an alcohol to olefins, the catalyst composition comprising the following components: (a) beta zeolite; (b) at least one element selected from the group consisting of zinc, magnesium, calcium, strontium, sodium, and potassium; and (c) at least one element selected from the group consisting of hafnium, yttrium, zirconium, tantalum, niobium, and tin; wherein the components (b) and (c) are independently within or on a surface of said beta zeolite. The catalyst may also further include component (d), which is copper or silver. Also described herein is a method for converting an alcohol to one or more olefinic compounds, the method comprising contacting the alcohol with a catalyst at a temperature of at least 100° C. and up to 500° C. to result in the alcohol being converted to the one or more olefinic compounds.

Li, Zhenglong↗
DOE OSTI2021

Reinforced matrix for molten carbonate fuel cell and method for manufacturing the same

A reinforced electrolyte matrix for a molten carbonate fuel cell includes a porous ceramic matrix, a molten carbonate salt provided in the porous ceramic matrix, and at least one reinforcing structure comprised of at least one of yttrium, zirconium, cerium or oxides thereof. The reinforcing structure does not react with the molten carbonate salt. The reinforced electrolyte matrix separates a porous anode and a porous cathode in the molten carbonate fuel cell.

Surendranath, Arun↗
DOE OSTI2022

Non-noble metal based electro-catalyst compositions for proton exchange membrane based water electrolysis and methods of making

The invention provides electro-catalyst compositions for an anode electrode of an acid mediated proton exchange membrane-based water electrolysis system. The compositions include a noble metal component selected from the group consisting of iridium oxide, ruthenium oxide, rhenium oxide and mixtures thereof, and a non-noble metal component selected from the group consisting of tantalum oxide, tin oxide, niobium oxide, titanium oxide, tungsten oxide, molybdenum oxide, yttrium oxide, scandium oxide, cooper oxide, zirconium oxide, nickel oxide and mixtures thereof. Further, the non-noble metal component can include a dopant. The dopant can be at least one element selected from Groups III, V, VI and VII of the Periodic Table. The compositions can be prepared using any solution based methods involving a surfactant approach or a sol gel approach. Further, the compositions are prepared using noble metal and non-noble metal precursors. Furthermore, a thin film containing the compositions can be deposited onto a substrate to form the anode electrode.

Kumta, Prashant N.↗
DOE OSTI2022

Additive manufacturing of complex objects using refractory matrix materials

A method for the manufacture of a three-dimensional object using a refractory matrix material is provided. The method includes the additive manufacture of a green body from a powder-based refractory matrix material followed by densification via chemical vapor infiltration (CVI). The refractory matrix material can be a refractory ceramic (e.g., silicon carbide, zirconium carbide, or graphite) or a refractory metal (e.g., molybdenum or tungsten). In one embodiment, the matrix material is deposited according to a binder-jet printing process to produce a green body having a complex geometry. The CVI process increases its density, provides a hermetic seal, and yields an object with mechanical integrity. The residual binder content dissociates and is removed from the green body prior to the start of the CVI process as temperatures increase in the CVI reactor. The CVI process selective deposits a fully dense coating on all internal and external surfaces of the finished object.

Terrani, Kurt A.↗
DOE OSTI2022

Overview of Base Model in Parametric Studies Specific to Performance of U-Mo Plates

This paper provides an overview of the base model specifically developed to perform parametric sensitivity studies on the U-10Mo monolithic fuel system. U-Mo monolithic fuels are being considered for the conversion of test reactors into high-performance research reactors that operate using proliferation-resistant, low-enriched uranium (LEU) fuels. These plate-type fuels contain a high-density, low-enrichment fuel sandwiched between zirconium diffusion barriers and encapsulated in aluminum claddings. All U.S. high-performance research reactors have released the designs of their LEU monolithic fuel reactor cores. These designs include nearly 50 distinct fuel plate geometries with different operational parameters. Consequently, a single generic plate geometry representing all the extreme points in this design matrix is unrealistic. To evaluate the performance for various parameters, a set of sensitivity studies was performed. These studies considered various input parameters (i.e., geometric, operational, and material property-related). The results revealed valuable information about plate performance and the sensitivity of this performance to various modeling inputs. To establish a reference state for comparing these result, base model featuring representative irradiation conditions was developed. To capture in-reactor behavior accurately, incorporation of representative constitutive models capable of evolving properties with respect to temperature, irradiation time, and burnup was needed. The behavioral models considered burnup-dependent properties, swelling, creep, and degradation. This paper introduces the base model created for the parametric sensitivity studies. The detailed description of the procedure includes the model geometry, model discretization, thermo-mechanical coupling, material properties and behavioral models. This paper also provides selected results and assesses the performance of the base model.

42 ENGINEERING↗
DOE OSTI2022

Design and Validation of a Gammaray Scanning System for Measuring Irradiated Nuclear Fuel

A prototype gamma-ray scanning system was designed to perform post-irradiation measurements of nuclear fuel at Idaho National Laboratory (INL). The system is composed of a coaxial high-purity germanium (HPGe) detector, a collimator, and mechanical positioning stages that translate the fuel sample across the front of the collimator. This paper documents the design process, modeling and the laboratory characterization performed to validate the system. The modeling of the system was done using the Monte Carlo N-Particle code with vendor-supplied design specifications, dimensional measurements, and x-ray radiographs of the HPGe detector. The characterization included measurements of calibrated radiation sources in different geometries, and calculations for absolute efficiency and dead time. Benchmark measurements were performed by scanning an irradiated fuel rodlet containing ten pellets of 4.9% enriched UO2 in zirconium alloy. These results were compared to simulated spectra to further help characterize the detector model. One goal of the system is to determine the number of fissions per gram of UO2 in the fuel; comparison of the results from this prototype system's assay of the irradiated fuel rodlet are in agreement with previously estimated results.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Cold spray chromium coating for nuclear fuel rods

A zirconium alloy cladding tube for use in a water cooled nuclear reactor under normal operating conditions and under high temperature oxidation conditions is described. The cladding tube has a coating uniformly deposited thereon. The coating, which may be up to 300 microns thick, is selected from the group consisting of chromium, a chromium-based alloy, and combinations thereof.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Irradiated Low-Enriched Uranium Fuel Measurements with a Gamma-Ray Scanning System

A gamma-ray scanning system was designed to perform post-irradiation measurements of nuclear fuel at Idaho National Laboratory (INL). The system is composed of a coaxial high-purity germanium (HPGe) detector, a collimator, and mechanical positioning stages that translate the fuel sample across the front of the collimator. A Monte Carlo N-Particle code simulation of the system and the fuel were created with vendor-supplied design specifications, dimensional measurements, and x-ray radiographs of the HPGe detector as well as all available fuel specifications. Benchmark measurements were performed by scanning an irradiated fuel rodlet containing eight pellets of 0.74% enriched UO2 in zirconium alloy. This fuel rodlet was part of the Static Environment Rodlet Transient Test Apparatus (SERTTA) testing campaign in the Transient Reactor Test Facility (TREAT). These results were compared to simulated spectra to help characterize the detector model and establish fidelity. A goal of the system is to determine the number of fissions per gram of UO2 in the fuel. This paper documents the modeling of the system and the calculation of the number of fissions per gram along the axial length of the SERTTA-C rodlet.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Rapid pyrolysis to form super ionic conducting lithium garnets

A method of preparing a lithium-ion conducting garnet via low-temperature solid-state synthesis is disclosed. The lithium-ion conducting garnet comprises a substantially phase pure aluminum-doped cubic lithium lanthanum zirconate (Li 7 La 3 Zr 2 O 14 ). The method includes preparing nanoparticles comprising lanthanum zirconate (La 2 Zr 2 O 7 -np) via pyrolysis-mediated reaction of lanthanum nitrate (La(NO 3 ) 3 ) and zirconium nitrate (Zr(NO 3 ) 4 ). The method also includes pyrolyzing a solid-state mixture comprising the La 2 Zr 2 O 7 -np, lithium nitrate (LiNO 3 ), and aluminum nitrate (Al(NO 3 ) 3 ) to give the Li 7 La 3 Zr 2 O 14 and thereby prepare the lithium-ion conducting garnet. A lithium-ion conducting garnet prepared via the method is also disclosed.

Daemen, Luc L.↗
DOE OSTI2023

Cladding Ring Tension Testing (RTT) of unirradiated cladding materials using Digital Image Correlation (DIC)

Cladding Ring Tension Testing (RTT) of unirradiated cladding materials using Digital Image Correlation (DIC). RTT is an important tool for determining hoop strength of thin-walled tubing, especially in anisotropic materials such as zirconium fuel cladding. The goal is to set up a DIC system to a tensile test on these RTT specimens and study the mechanical properties.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Cast iron alloy for automotive engine applications with superior high temperature oxidation properties

A cast iron alloy is provided with a composition in weight percent (wt. %) of carbon between 2.6 to 3.4 wt. %, silicon between 2.4 to 3.2 wt. %, manganese between 0.3 to 0.6 wt. %, molybdenum between 0.4 to 1.2 wt. %, nickel between 0.6 to 1.75 wt. %, magnesium between 0.01 to 0.075 wt. %, aluminum between 1.8 to 3.5 wt. %, sulfur between 0.003 to 0.025 wt. %, zirconium between 0.001 to 0.02 wt. %, cerium between 0.001 to 0.03 wt. %, lanthanum between 0.0005 to 0.02 wt. %, and a balance of iron and unavoidable trace elements. A part formed from the cast iron alloy is also provided and the part has an Ac1 temperature equal to or greater than 895° C. and a thermo-mechanical fatigue lifetime of at least 10,000 cycles when cycled between 400° C. to 800° C. with a total cyclic strain equal to 0.001 m/m.

Godlewski, Larry Alan↗
DOE OSTI2023

Development of Surveillance Test Articles with Reduced Dimensions and Material Volumes to Support MSR Materials Degradation Management

This report details the efforts toward developing new surveillance test article designs with reduced dimensions and material volumes to support materials surveillance technology development for advanced reactors. Two fabrication methods for the surveillance test articles are described. Welded test articles were fabricated with 316H and A617 materials, and interlocking test articles were fabricated using A709 and titanium-zirconium-molybdenum (TZM). The preliminary results demonstrate the successful design and testing of the flat surveillance test articles. The report also describes ongoing efforts to use an induction heating test setup to increase the heat up and cool down rates in testing the surveillance test articles. A brief description of the planned FY-24 work is provided.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Three-way catalyst with reduced palladium loading and method of making the three-way catalyst

A three-way catalyst for reduced palladium loading is provided. The catalyst includes an inert substrate and a palladium catalyst material coating the substrate. The palladium catalyst material includes a support material formed from one of 10% CeO 2 /Al 2 O 3 , 20% CeO 2 —Al 2 O 3 (20CeAlOy), 30% CeO 2 —Al 2 O 3 (30CeAlOy), Al 2 O 3 , and MOx-Al 2 O 3 , wherein M is one of copper, iron, manganese, titanium, zirconium, magnesium, strontium, and barium. The palladium catalyst material includes a layer of CeO 2 material disposed upon the support material, wherein the layer of CeO 2 material is dispersed on a surface of the support material. The palladium catalyst material includes an active component including a layer of praseodymium oxide particles dispersed across the surface of the layer of CeO 2 material and a layer of palladium particles disposed upon and dispersed across the surface of the layer of CeO 2 material at locations each corresponding to a respective location of each of the praseodymium particles.

Gu, Yuntao↗
DOE OSTI2022

DESIGN AND VALIDATION OF A GAMMA-RAY SCANNING SYSTEM FOR MEASURING IRRADIATED NUCLEAR FUEL (ABSTRACT)

A prototype gamma-ray scanning system was designed to perform post-irradiation measurements of nuclear fuel at Idaho National Laboratory (INL). The system is composed of a coaxial high-purity germanium (HPGe) detector, a collimator, and mechanical positioning stages that translate the fuel sample across the front of the collimator. This paper documents the design process, modeling and the laboratory characterization performed to validate the system. The modeling of the system was done using the Monte Carlo N-Particle code with vendor-supplied design specifications, dimensional measurements, and x-ray radiographs of the HPGe detector. The characterization included measurements of calibrated radiation sources in different geometries, and calculations for absolute efficiency and dead time. Benchmark measurements were performed by scanning an irradiated fuel rodlet containing ten pellets of 4.9% enriched UO2 in zirconium alloy. These results were compared to simulated spectra to further help characterize the detector model. One goal of the system is to determine the number of fissions per gram of UO2 in the fuel; comparison of the results from this prototype system's assay of the irradiated fuel rodlet are in agreement with previously estimated results.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

Status Report on Fast Flux Test Facility Mechanistic Fuel Failure Experiment Analysis with BISON for Post Irradiation Examination Support

The renewed interest in metallic U-Zr nuclear fuel alloy has led to a drive for deeper understanding of the mechanisms driving the phenomena observed under irradiation conditions. The Department of Energy Advanced Fuel Campaign has developed infrastructure to support metallic fuel development, including Post Irradiation Examination (PIE) of legacy Fast Flux Test Facility (FFTF) Mechanistic Fuel Failure (MFF) experiments. The PIE performed on legacy FFTF MFF experiments gives insight on metallic fuel performance and can address the lack of knowledge and scarcity of reliable data identified in several studies over recent years. Unfortunately, PIE efforts can cost significant time and resources which can impede the progress of metallic U-Zr fuel development. Metallic U-Zr fuel performance modeling can be used to inform PIE efforts on regions of interest for relevant investigations and can help understand phenomena observed in PIE. This report demonstrates the current progress of FFTF MFF fuel performance simulations using the BISON fuel performance code and discusses the support provided by simulation to PIE efforts. Progress in temperature, profilometry, fission gas release, plenum pressure, and zirconium redistribution simulation results have been demonstrated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Advances in Metallic Fuel Database Development and Data Qualification

The Fuels Irradiation and Physics Database (FIPD [1]) is a comprehensive repository of data and documents related to Uranium-Zirconium based metallic fuel test pins. This database stores operational conditions of these pins, calculated using a suite of Argonne National Laboratory analysis codes developed during the Integral Fast Reactor (IFR) program. Key calculated data include axial distributions of power, temperature, fluence, burnup, and isotopic densities. Additionally, the FIPD holds post-irradiation examination (PIE) data such as fission gas release, gas chemistry measurements, and axial distributions derived from profilometry, gamma scanning, and neutron radiography. Complementing these data is an extensive archive of documents related to various pins and experiments. These include raw PIE records, design details, safety analyses, and operational reports. More detail about FIPD can be found in ref. [2]. The database development is an ongoing effort covering metallic fuel experiments from the Experimental Breeder Reactor II (EBR-II) and the Fast Flux Test Facility (FFTF). The recent improvements to the database and the data QA status are summarized in this paper.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗