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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 361 records · Page 20
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Development of Self-Healing Zirconium-Silicide Coatings for Improved Performance Zirconium-Alloy Fuel Cladding

Zirconium-alloy fuel claddings have been used successfully in Light Water Reactors (LWR) for over four decades. However, under high temperature accident conditions, zirconium-alloys fuel claddings exhibit profuse exothermic oxidation accompanied by release of hydrogen gas due to the reaction with water/steam. Additionally, the ZrO 2 layer can undergo monoclinic to tetragonal to cubic phase transformations at high temperatures which can induce stresses and cracking. These events were unfortunately borne out in the Fukushima-Daiichi accident in in Japan in 2011. In reaction to such accident, protective oxidation-resistant coatings for zirconium-alloy fuel claddings has been extensively investigated to enhance safety margins in accidents as well as fuel performance under normal operation conditions. Such surface modification could also beneficially affect fuel rod heat transfer characteristics. Zirconium-silicide, a candidate coating material, is particularly attractive because zirconium-silicide coating is expected to bond strongly to zirconium-alloy substrate. Intermetallic compound phases of zirconium-silicide have high melting points and oxidation of zirconium silicide produces highly corrosion resistant glassy zircon (ZrSiO 4 ) and silica (SiO 2 ) which possessing self-healing qualities. Given the long-term goal of developing such coatings for use with nuclear reactor fuel cladding, this work describes results of oxidation and corrosion behavior of bulk zirconium-silicide and fabrication of zirconium-silicide coatings on zirconium-alloy test flats, tube configurations, and SiC test flats. In addition, boiling heat transfer of these modified surfaces (including ZrSi 2 coating) during clad quenching experiments is discussed in detail.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Mini-Canister Radiolysis Testing of ASNF Materials and Surrogates

An experimental irradiation campaign to investigate radiolysis behavior of ASNF was conducted using in situ gas monitoring of small, sealed stainless-steel vessels (mini-canisters) containing aluminum samples with adherent (oxy)hydroxide films under helium backfill. The samples were irradiated with gamma radiation from a Co-60 irradiator. The samples tested included aluminum plate assemblies with lab-grown (oxy)hydroxides as surrogates for fuel as well as an end cropping from an actual ASNF assembly retrieved from long-term wet storage. These experiments enabled investigation of the impacts of various fuel drying approaches on the radiolytic generation rate and measurement of the H 2 yield associated with a reactor exposed sample with reactor-formed (oxy)hydroxide. The resulting data can be incorporated into model development for ASNF in dry storage. This report presents the cumulative results from four surrogate assemblies tested after application of different preparation (drying) conditions as well as the ASNF cropping; some post-irradiation testing was included. The mini-canister results are compared to data from related experimental campaigns that also tested the impact of drying conditions using samples irradiated in glass ampoules and discusses implications of the combined data

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

Methodology of Experiment Design for Improving Transient Critical Heat Flux Model Development

One phenomenon that contributes to nuclear fuel and clad failure mechanisms is the departure from nucleate boiling (DNB). Accident scenarios such as reactivity-initiated accidents (RIAs) provide conditions where the frequency of boiling crisis is much more likely to occur and on much shorter time scales than loss of coolant (LOCA) or loss of flow (LOFA) accidents. Here, boiling crisis in the form of DNB is attributed to transient-induced heat flux values that exceed the critical heat flux (CHF). The resulting effects, such as bubble crowding and eventual dryout, can affect flow boiling heat transfer and compromise the structural integrity of nuclear reactor materials. The focus of the work proposed here lies in developing a rigorous process using out-of-pile testing, under simulated RIA conditions, to determine the transient critical heat flux of cladding materials, without contributions from aforementioned radiation-induced effects that have been observed during in-pile experiments. This study provides a road-map for developing and performing experiments that simulate the power profile and energy deposition aspects of transient events, specific to RIAs, on the laboratory scale. The goal of our project is to observe and measure DNB in an annular flow channel around a single heater rod that is operating and nominal thermal hydraulic conditions of a PWR. It is also an aim of these experiments to determine how to quantify the gap between SS and transient CHF cases, based on state-dependent properties of the cladding material type that is being interrogated. Traditionally, CHF is determined as a combination of four factors: hydraulic diameter, system pressure, mass flow, fluid quality. The authors propose that an additional measureable quantity, which is dependent on the structural properties of the cladding material itself, is partly reasonable for the difference between measured and model-developed CHF values. Using a carefully metered approach will allow for several key concepts to be realized: compensation for transient-based effects to current models for CHF; adaptation of current models for different cladding material types, including Gen-IV concept materials; and implementation of a correction factor to account for both power transient-based and radiation-induced effects on the CHF that have not yet been accounted for.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Complete Non-Radioactive Operability Tests for Cladding Hull Chlorination

Non-radioactive operability tests were made to test the metal chlorination reactor and condenser and their accessories using batch chlorinations of non-radioactive cladding samples and to identify optimum operating practices and components that need further modifications prior to installation of the equipment into the hot cell for tests on actual used nuclear fuel (UNF) cladding. The operability tests included (1) modifications to provide the desired heating and reactor temperature profile; and (2) three batch chlorination tests using, respectively, 100, 250, and 500 g of cladding. During the batch chlorinations, metal corrosion of the equipment was assessed, pressurization of the gas inlet was examined and the best method for maintaining solid salt product transfer through the condenser was determined. Also, additional accessing equipment for collection of residual ash and positioning of the unit within the hot cell were identified, designed, and are being fabricated.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Small-scale mechanical testing and characterization of fuel cladding chemical interaction between HT9 cladding and advanced U-based metallic fuel alloy

Fuel cladding chemical interactions (FCCI) occurred on the interface between the nuclear metal fuel and cladding is the primary cause of cladding wastage, weakening cladding mechanical integrity, and placing fuel and cladding at risk. Although the microstructural and phase information of FCCI has been fairly understood, mechanical properties remain less studied due to limited reaction volume. Here, through a combining of advanced electron microscopy characterizations and small-scale mechanical testing techniques, including indentation and micro-tensile testing, this study investigated the microscale mechanical properties of FCCI between the ferritic/martensitic (F/M) HT9 cladding and an advanced Uranium (U)-based metallic fuel irradiated at the Advanced Test Reactor to 2.2% FIMA with peak inner cladding temperature reached to 650 °C. Mechanical testing results show significant hardening and embrittlement in the FCCI region. The brittle fracture of FCCI specimen is mainly attributed to the formation of nano-crystallized intermetallic σ-FeCr phase. Whereas mechanical softening was revealed in the unreacted HT9 matrix due to irradiation-induced microstructural and microchemical evolution, specifically, the disappearance of martensitic lath structure and the formation of Fe 2 Mo Laves phase precipitation which consumed the solid solution strengthening Mo from the F/M HT9 matrix. Due to the achieved high cladding temperature, this fuel pin is of particular significance for revealing the high-temperature irradiation effect on the mechanical properties of HT9 cladding. Therefore, the outcomes of this study are expected to contribute to the development of multi-scale mechanical behavior modeling of HT9 cladding for Generation IV reactors which requires cladding to run at higher temperature (above 600 ?).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Purification of Zirconium Tetrachloride from UNF Cladding

The overarching objective of this project was to design and demonstrate (laboratory scale) a sublimation-based purification protocol to obtain pure zirconium chloride (ZrCl 4 ) from impure feeds derived from the direct chlorination of Zircaloy cladding materials. To achieve this objective the following goals were identified. (1) Design, construct and test an apparatus capable of subliming and purifying ZrCl 4 at room pressure and temperatures between 600 and 700K (sublimation point of ZrCl 4 606 K). (2) Investigate reactions that may be used to chemically alter the makeup of the impure ZrCl 4 feed materials to enhance their removal from ZrCl 4 in the sublimation-purification process. (3) Computationally model the reactions of metal chlorides in the gas phase under sublimation conditions (1 atm, 600 K) to determine if oligomers are formed and, if formed, how they would influence sublimation-based purification strategies. All experimental investigations performed as part of this project used stable (non-radioactive) compounds and materials as surrogates for the impure ZrCl 4 feedstocks that would be derived from the direct chlorination of Zircaloy claddings from spent nuclear fuel rods. No radiologically active materials were used at any time in this project. The impure ZrCl 4 feedstocks used in this study were derived from the direct chlorination of virgin Zircaloy claddings (Zirc-2, -4 and high Nb). The main goals of the project have been achieved during the three year period of this NEUP sponsored project. The impurities of greatest concern in the feedstocks were identified as the chloride salts of the alloying metals originally found in Zircaloy (Fe, Nb) and radiological contaminants (Cs, Sr, and Sb). A series of prototype sublimation apparatuses and heating arrangements were designed, constructed and tested. A general batch-sublimator design was identified and a 1.0 kg batch sublimator/oven system was fabricated from glass. Purification protocols developed on small scale runs were found to scale to 1.0 kg batches without problem. Purification factor ranges (PFs: wt% final / wt% initial ) for the impurity chlorides identified above have been measured: Fe (40 – 500); Nb (10 – 70); Cs (520 – 5000); Sr (250 – 1900); Sb (220 – 2300). Removal of iron is sensitive to the partial pressure of hydrogen (in nitrogen) flowing through the system. Lower PFs are obtained when the hydrogen partial pressure is lower. Lower flows are required to raise the iron PF. Removal of niobium is very sensitive to the endpoint determination of the forerun. When judged properly (color change yellow to white in sublimate) reasonable PFs (500) can be achieved. However, as discussed in the conclusions section, endpoint determination can be difficult and significantly lower PFs may be obtained when the endpoint is not correctly identified. Several methods of obtaining quantitative cesium, strontium and antimony PFs were investigated, the best being ICP-MS. The results reported here reflect very limited access to and ICP-MS instrument during the course of this project. It should be noted that, while significantly reduced amounts of antimony are observed in the product, antimony is also found in the exhaust trap after the condenser. Quantum chemical computational modeling studies of the reactions of metal chlorides in the gas phase provided important information about a chemical transport mechanism that could defeat sublimation-based purification strategies. Many metal chlorides are known to oligomerize in the gas phase to reduce their coordinative unsaturation. Dimerization of ZrCl 4 with iron or niobium chlorides would provide a mechanism by which both metal impurities could be carried over with the product during sublimation of the bulk material. Computational modeling conclusively showed that under sublimation conditions in the gas phase, dimerization reactions are not favored thermodynamically due to a significant entropic penalty for the reaction. Therefore, consistent with experimental observations, sublimation-based purification strategies remain viable. The experimental part of the report concludes with a short discussion of the issues that still remain in developing a technologically relevant Zircaloy recycling strategy. The final section of the report is a summary of the quantum chemical investigation of the reactions of metal chlorides in the gas phase under sublimation conditions.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2012

LWRS Fuels Pathway: Engineering Design and Fuels Pathway - Initial Testing of the Hot Water Corrosion System

The Advanced LWR Nuclear Fuel Development R&D pathway performs strategic research focused on cladding designs leading to improved reactor core economics and safety margins. The research performed is to demonstrate the nuclear fuel technology advancements while satisfying safety and regulatory limits. These goals are met through rigorous testing and analysis. The nuclear fuel technology developed will assist in moving existing nuclear fuel technology to an improved level that would not be practical by industry acting independently. Strategic mission goals are to improve the scientific knowledge basis for understanding and predicting fundamental nuclear fuel and cladding performance in nuclear power plants, and to apply this information in the development of high-performance, high burn-up fuels. These will result in improved safety, cladding, integrity, and nuclear fuel cycle economics. To achieve these goals various methods for non-irradiated characterization testing of advanced cladding systems are needed. One such new test system is the Hot Water Corrosion System (HWCS) designed to develop new data for cladding performance assessment and material behavior under simulated off-normal reactor conditions. The HWCS is capable of exposing prototype rodlets to heated, high velocity water at elevated pressure for long periods of time (days, weeks, months). Water chemistry (dissolved oxygen, conductivity and pH) is continuously monitored. In addition, internal rodlet heaters inserted into cladding tubes are used to evaluate repeated thermal stressing and heat transfer characteristics of the prototype rodlets. In summary, the HWCS provides rapid ex-reactor evaluation of cladding designs in normal (flowing hot water) and off-normal (induced cladding stress), enabling engineering and manufacturing improvements to cladding designs before initiation of the more expensive and time consuming in-reactor irradiation testing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Pile Fuel Cladding Silo Deflector Plate Removal Project Using Long Reach Tooling and Suspended Abrasive Water Jet Cutting System - 17541

The Pile Fuel Cladding Silo (PFCS) at the Sellafield decommissioning, reprocessing and nuclear waste management facility in the UK was constructed in the late 1940-s and used for the storage of Intermediate Level Waste (ILW) arising from the Windscale Piles, Calderhall and Chapelcross Reactors until the 1960's. The PFCS is 21 metres high and comprises 6 identical tall compartments with the waste added to the silo via a charge-hole at the top of each compartment; each compartment is split by a central dividing wall that runs from the base of the silo to a few metres below the underside of the silo roof. A deflector plate assembly (DPA) supported on two horizontal beams is located between the top of the dividing wall and the roof and allowed the waste to spread as it fell into the silo and prevent excessive mounding. Sellafield Ltd is currently engaged in an overall programme of works to remove the waste for storage within modern ILW containers. Prior to the commencement of these works the Deflector Plate Assemblies (DPA) and beam structures within the silos are required to be removed to allow access for the waste retrievals equipment through the side of the silo at the point where the DPAs are situated. James Fisher Nuclear (JFN) and Sellafield Ltd have developed a system using suspended water jet cutting equipment combined with custom long reach tooling, installed through penetrations made in the side of the silo, to enable the remote size reduction of the DPA and beams within each compartment. The use of a low pressure suspended water jet cutting system has enabled a single water/abrasive hose to be routed down a tool deployment pole and allowed long hose lengths up to 80 m from pump to nozzle enabling siting of equipment on a congested and difficult access site. The use of lightweight, yet high strength, carbon fibre poles within the tools has enabled the deployment of tool end effectors at distances of up to 6.5 m from the compartment walls using a relatively small 60 mm diameter pole. This maximised articulation of the pole within the 300 mm wall penetrations to reach all areas of the DPA. Capability of the operators to install the tooling and size reduce DPA and beam assemblies have been demonstrated through the training on a Full Scale Test Rig (FSTR). The FSTR was integral to the development of the equipment and techniques and to providing assurance to the client that the project was technically achievable given the tight building restraints and safety issues. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

High-resolution characterization of ceramic-metal interface of TiN coating on ferritic-steels for nuclear application

Advanced fuel cladding is critical for fast reactors, offering sufficient thermal conductivity, mechanical and dimensional stability and radiation tolerance of the cladding base material. Additionally, it must provide corrosion resistance and high temperature coolant compatibility on the cladding outer surface, as well as chemical stability on the cladding inner wall against fuel cladding chemical interaction (FCCI). TiN ceramic coating has been considered an effective diffusion barrier for inner and outer cladding-walls for enhanced performance. The TiN-metal interface microstructure and chemistry play a critical role in coating bond strength and integrity under harsh conditions. High-resolution transmission electron microscopy characterization of ceramic-metal interface at atomic resolution in unirradiated, irradiated and thermal cycled conditions were performed. The interface remained intact after irradiation up to 200 dpa or thermal cycling five times up to 550 °C. In conclusion, this work discusses the potential impact of these results on coating performance and design for advanced claddings.

36 - MATERIALS SCIENCE↗
DOE OSTI2016

Nanoscale Mechanical Behavior of Nuclear Materials

One of the primary performance objectives targeted for improved performance of nuclear fuels is fracture toughness. New materials or composite fuel architectures that offer greater resistance to cracking under the extreme environments encountered during nuclear reactor service would provide significant improvements to steady state (e.g. heat transfer, fuel redistribution) and transient (e.g. radionuclide release at elevated temperatures) conditions. Options currently under investigation include 'nontraditional' nuclear fuels designated around high uranium density. These include uranium silicides, uranium borides, and composite fuel materials constructed of these and uranium nitride or uranium dioxide. The increase uranium density compared with uranium dioxide (UO{sub 2}) has made them interesting to a new generation of nuclear fuels research driven by the renewed push for accident-tolerant light water reactor (LWR) fuels. Measurement of the thermophysical and thermodynamic properties of these constituent materials has progressed in recent years, but evaluation of mechanical properties at relevant temperatures is necessary to support further evaluations. While thermophysical properties such as thermal conductivity, heat capacity, and thermal expansion are generally the primary properties of interest, mechanical properties will play an important role in governing the deformation and fracture of the fuel during both normal operation and potential transients. It is therefore necessary to develop techniques capable of measuring these properties to predict the response of a candidate fuel form to anticipated thermal gradients. Nanoindentation techniques provide a demonstrated alternative to traditional mechanical test methods. Much work has also been carried out over the past few years to obtain stress-strain curves using a spherical nanoindentation approach. The techniques are expected to be useful for measurement of the mechanical properties of local structure of various materials. Nanomechanical test techniques have particular value in the area of nuclear materials, where numerous factors motivate limiting the sample volume to the extent possible. However, few attempts at application of nanoindentation to evaluate the mechanical properties of the nuclear fuels and cladding materials are reported in the literature. The effect of a limited set of variables (e.g. porosity, temperature dependence) on the mechanical properties of UO{sub 2} has been explored using early nanomechanical testing techniques. However, nanomechanical evaluation has not been extended to other nuclear fuels. In this work, we demonstrate the effectiveness of using instrumented spherical nanoindentation test techniques to capture the mechanical response of regions within individual grains in ion-irradiated W and Zr over a range of ion-irradiation conditions. As well as the nanomechanical response of U-Si and UO{sub 2} fuel materials Nanoindentation results obtained at room temperature will be used to express nano- and bulk- mechanical properties as a function of material stoichiometry. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Designed porosity materials in nuclear reactor components

A nuclear fuel pellet with a porous substrate, such as a carbon or tungsten aerogel, on which at least one layer of a fuel containing material is deposited via atomic layer deposition, and wherein the layer deposition is controlled to prevent agglomeration of defects. Further, a method of fabricating a nuclear fuel pellet, wherein the method features the steps of selecting a porous substrate, depositing at least one layer of a fuel containing material, and terminating the deposition when the desired porosity is achieved. Also provided is a nuclear reactor fuel cladding made of a porous substrate, such as silicon carbide aerogel or silicon carbide cloth, upon which layers of silicon carbide are deposited.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Ion irradiation testing and characterization of FeCrAl candidate alloys

The Fuel Cycle Research and Development program’s Advanced Fuels Campaign has initiated a multifold effort aimed at facilitating development of accident tolerant fuels. This effort involves development of fuel cladding materials that will be resistant to oxidizing environments for extended period of time such as loss of coolant accident. Ferritic FeCrAl alloys are among the promising candidates due to formation of a stable Al₂O₃ oxide scale. In addition to being oxidation resistant, these promising alloys need to be radiation tolerant under LWR conditions (maximum dose of 10-15 dpa at 250 – 350°C). Thus, in addition to a number of commercially available alloys, nuclear grade FeCrAl alloys developed at ORNL were tested using high energy proton irradiations and subsequent characterization of irradiation hardening and damage microstructure. This report summarizes ion irradiation testing and characterization of three nuclear grade FeCrAl cladding materials developed at ORNL and four commercially available Kanthal series FeCrAl alloys in FY14 toward satisfying FCRD campaign goals.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Evaluation Metrics Applied to Accident Tolerant Fuels

The safe, reliable, and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the United States’ nuclear industry. Continual improvement of technology, including advanced materials and nuclear fuels, remains central to the industry’s success. Decades of research combined with continual operation have produced steady advancements in technology and have yielded an extensive base of data, experience, and knowledge on light water reactor (LWR) fuel performance under both normal and accident conditions. One of the current missions of the U.S. Department of Energy’s (DOE) Office of Nuclear Energy (NE) is to develop nuclear fuels and claddings with enhanced accident tolerance for use in the current fleet of commercial LWRs or in reactor concepts with design certifications (GEN-III+). Accident tolerance became a focus within advanced LWR research upon direction from Congress following the 2011 Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex. The overall goal of ATF development is to identify alternative fuel system technologies to further enhance the safety, competitiveness and economics of commercial nuclear power. Enhanced accident tolerant fuels would endure loss of active cooling in the reactor core for a considerably longer period of time than the current fuel system while maintaining or improving performance during normal operations. The U.S. DOE is supporting multiple teams to investigate a number of technologies that may improve fuel system response and behavior in accident conditions, with team leadership provided by DOE national laboratories, universities, and the nuclear industry. Concepts under consideration offer both evolutionary and revolutionary changes to the current nuclear fuel system. Mature concepts will be tested in the Advanced Test Reactor at Idaho National Laboratory beginning in Summer 2014 with additional concepts being readied for insertion in fiscal year 2015. This paper provides a brief summary of the proposed evaluation process that would be used to evaluate and prioritize the candidate accident tolerant fuel concepts currently under development.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗