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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 1,153 records · Page 64
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Study on neutronics calculation method of SCWR with annular fuel assembly - 26804

Many Supercritical Water Reactor designs have been proposed all over the world. Annular fuel was employed in SCWR by Xi'an JiaoTong University to achieve lower fuel center temperature and lower cladding surface temperature. Different analysis codes have been developed for these designs. In this article, a neutronics calculation code for SCWR with annular fuel is developed. This code is compared with Monte-Carlo code doing both assembly and core calculation. The results showed that the spectra difference of PWRs and SCWRs is the main reason for assembly calculation errors and conventional assembly homogenization method is not suitable for SCWRs. Thus the group-collapsing and spatial-averaging method were studied to improve the accuracy. The few-group structure which contains 14 energy groups is used. The 'super-assembly' model is employed to consider effect of reflector. The results showed that the accuracy of effective multiplication factor and radial power was improved. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Microstructure analysis for chemical interaction between cesium and SUS 316 steel in fast breeder reactor application

In this study the corrosion products on a surface after cesium corrosion examination at 650 deg. C for 100 hrs were characterized by TEM observation around the corroded area on the surface in order to understand the corrosion mechanism of cesium fission product for cladding materials in fast reactor. The experimental results suggest the main corrosion mechanism occurred in the process of the separation of cesium chromate and metal (Fe, Ni). The main reaction of corrosion process was considered to be equation, 2Cs + 7/2 O{sub 2} + 2Cr {yields} Cs{sub 2}Cr{sub 2}O{sub 7}(L). (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Investigation of tin as a fuel additive to control FCCI

Tin is being investigated as a potential additive to metallic fuel to control fuel-cladding chemical interaction (FCCI). A primary cause of FCCI is the lanthanide fission products moving to the fuel periphery and interacting with the cladding. This interaction will lead to wastage of the cladding and eventually to a cladding breach. The current study investigates the microstructure of as-cast and annealed U-10Zr-4.3Sn and U-10Zr-4.3Sn-4.7Ln (where Ln=53Nd-25Ce-16Pr-6La) using SEM, TEM, and XRD. Diffusion couple results, between U-10Zr-4.3Sn and lanthanides, and between U-10Zr-4.3Sn-4.7Ln and Fe, will also be presented. This investigation shows that tin preferentially binds the lanthanides, which will prevent lanthanide migration and interaction with the cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Mechanical properties of SiC composites neutron irradiated under light water reactor relevant temperature and dose conditions

Silicon carbide (SiC) fiber–reinforced SiC matrix (SiC/SiC) composites are being actively investigated for use in accident-tolerant core structures of light water reactors (LWRs). Owing to the limited number of irradiation studies previously conducted at LWR-coolant temperature, this paper examined SiC/SiC composites following neutron irradiation at 230–340 °C to 2.0 and 11.8 dpa in the High Flux Isotope Reactor. The investigated materials were chemical vapor infiltrated (CVI) SiC/SiC composites with three different reinforcement fibers. The fiber materials were monolayer pyrolytic carbon (PyC) -coated Hi-Nicalon™ Type-S (HNS), Tyranno™ SA3 (SA3), and SCS-Ultra™ (SCS) SiC fibers. The irradiation resistance of these composites was investigated based on flexural behavior, dynamic Young's modulus, swelling, and microstructures. There was no notable mechanical properties degradation of the irradiated HNS and SA3 SiC/SiC composites except for reduction of the Young's moduli by up to 18%. The microstructural stability of these composites supported the absence of degradation. In addition, no progressive swelling from 2.0 to 11.8 dpa was confirmed for these composites. On the other hand, the SCS composite showed significant mechanical degradation associated with cracking within the fiber. Finally, this study determined that SiC/SiC composites with HNS or SA3 SiC/SiC fibers, a PyC interphase, and a CVI SiC matrix retain their properties beyond the lifetime dose for LWR fuel cladding at the relevant temperature.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Analysis of FeCrAl cladding and UO 2 fuel including discrete and smeared cracks, and fuel relocation

As UO 2 fuel heats during the rise to operating power in an LWR, the fuel pellets begin to fracture due to thermal stresses arising due to the temperature profile radially across the fuel. As the fuel fractures, it expands diametrically, reducing the fuel cladding gap and improving the heat transfer characteristics of the fuel. By increasing the effective fuel diameter, gap closure is expected to occur much sooner and greater stresses are expected to form in cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Complete Status Report Documenting the Processing, Microstructure and High-Temperature Mechanical Properties of the Nanostructured Ferritic Alloy OFRAC

This report provides a summary of the new nanostructured ferritic alloy, OFRAC (Oak Ridge Fast Reactor Advanced Fuel Cladding), that was developed for the advanced sodium fast reactor fuel cladding and core internals. The composition of OFRAC, which is Fe-12Cr-1Mo-03.Ti-0.3Nb + 0.3Y2O3 (wt. %), was chosen to achieve high temperature strength as well as high temperature creep performance and swelling resistance requirements that past structural materials such as austenitic stainless steels and ferritic/martensitic steels, i.e. HT-9, are not suitable as SFR cladding materials. The mechanical alloying procedures were described for producing OFRAC and the detailed characterization studies revealed uniformly distributed, high concentration of Y-Ti-O enriched oxide particles, or nanoclusters. The Nb addition that was intended to sequester interstitial C and N that are commonly introduced as contamination during ball milling of powder was found to be very effective in forming coarse size Nb carbonitrides along with excess Ti that removes these interstitial elements from the bcc-Fe matrix of OFRAC. The tensile tests on the OFRAC in the extruded condition showed excellent balance in strength and ductility from room temperature to 800°C. Both the microstructure and tensile properties were not significantly degraded after annealing the extruded OFRAC samples for 8 h at 1050°C and 1150°C. The latter temperature is ~76% of the melting point of OFRAC. These results demonstrated the stability of the microstructure and tensile properties to extremely high temperatures. The thermal creep resistance at elevated temperatures and stresses was investigated using strain rate jump testing. The results showed that OFRAC possessed high stress exponents at high stresses and temperatures of 550°C, 600°C, 700ºC and 800°C. This thermal dependence of minimum creep rate indicates that OFRAC will show quite low creep rates at higher temperatures and stresses compared to HT-9 and stainless steels. These significant improvements indicate that the NFA OFRAC alloy is a compelling and viable choice for sodium fast reactor fuel cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

GPHS Module Pre-Impact test Temperature Prediction with ANSYS [PowerPoint]

The purpose of the project was to: Develop a finite element thermal model of the General Purpose Heat Source (GPHS) Module loaded with four fueled clads; Perform steady state analysis to validate the thermal model via energy balance; Perform transient thermal analysis to predict clad temperatures at various times; and, Transient model results will be used to inform module soak time needed to reach a desired clad temperature prior to impact.

42 ENGINEERING↗
DOE OSTI2023

ECAR-7210 Steady-state Thermo-mechanical Analysis of the MARVEL Fuel Bundle

This document reports the steady-state thermal-hydraulic and static mechanical analyses of the MARVEL micro-reactor core. The objective of this work is to investigate whether fuel-to-fuel contact could occur due to the thermal deformation of the fuel cladding during the normal operation and to assess the resulting impact on the peak cladding temperatures (PCT) if contact occurs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Modular In-Chamber Electron Beam Welding – Process Planning: Welding, Cladding, Inspection, and Manufacturing

Electron beam welding (EBW) is commonly employed by industry today to join relatively thin section components. Industrial sectors which utilize the technology include; automotive, aircraft, aerospace, defense, medical, oil & gas, etc. The ability to produce single pass, thick section (up to 6 inches, 150mm) welds has also been available for more than 40 years; however, the technology has not been deployed for civil nuclear manufacturing and fabrication such as reactor pressure vessels, steam generators, or pressurizers. Instead, thick section joining is performed much in the same manner it has been for decades using submerged arc or gas tungsten arc welding processes. Recent research performed under U.S. Department of Energy (DOE) project DE-NE0008629 has demonstrated that the overall welding time to produce one major reactor pressure vessel (RPV) girth weld can be reduced from weeks to less than 90 minutes, while maintaining exceptional quality and inspection characteristics. The subject project (DE-NE0008846) investigates a modular in-chamber (MIC) EBW approach which allows various RPV sections to be joined within a “modular” chamber. The MIC-EBW approach utilizes individual modules that can be stacked and de-stacked to various heights to enable production girth welds to be performed at a given elevation. This report describes the various processing methodology necessary to complete major girth weldments for the NuScale Power small modular reactor including welding development, manufacturing, production sequence, inspection, and cladding.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Modeling of gap conductance for LWR fuel rods applied in the BISON code

A new gap conductance model is proposed in this study as a combination of Toptan’s model and the Ross-Stoute model. A variance-based sensitivity analysis is performed to understand how simulation results depend on all input parameters of the proposed model. Additionally, new modeling options (e.g. fill gas thermal conductivity, temperature jump distance, thermal accommodation coefficient, etc.) are added into the nuclear fuel performance code, BISON. The need for further investigation of the gap heat transfer between fuel and cladding in BISON motivated this study to evaluate its impact on the code’s predictions. New gap conductance modeling is proposed. A series of integral-effects validation tests is performed: (1) to demonstrate the impact of the proposed model on the code’s fuel temperature predictions at the beginning of life and through the reactor’s life; (2) to ensure that the proposed model is capable of accurately modeling gap heat transfer characteristics in real-world problems; and (3) to investigate the impact of the estimation of fission gas release on the fuel temperature predictions with the proposed model. Overall, the results indicate that the proposed gap conductance model improves BISON’s predictions.

42 ENGINEERING↗
DOE OSTI2020

An Investigation of Liquefaction in Irradiated TRIGA Fuel Exposed to Relatively High Temperatures

This report presents the findings of an investigation into high temperature fuel cladding chemical interactions (FCCI) in Training, Research, Isotopes, General Atomics (TRIGA) fuel rods. A TRIGA fuel-rod core or meat is principally composed of uranium (U) particles dispersed in a zirconium hydride (Zr H) matrix. The fuel is clad in sealed 304SS or Incoloy 800 tubes. At high temperatures, the fuel will interact with the cladding, resulting in FCCI. To investigate the FCCI (in this case, liquefaction), irradiated TRIGA fuel rods were exposed to relatively high temperatures in furnace tests. The tests were performed at different temperatures using segments from an irradiated TRIGA fuel rod at the Hot Fuel Examination Facility located at Idaho National Laboratory (INL). The data from this study is important for developing a better understanding of the performance of U-ZrH fuel type during transient events (e.g., a Loss of Coolant Accident (LOCA) event) that may result in high temperatures while operating a TRIGA reactor. Furnace tests were run for 6 hours at 730, 800, 900, 950, and 1,000°C. One test was run for 12 hours at 950°C. Post-test microstructural characterization was performed on the heat-treated samples using optical metallography to look for evidence of liquefaction. Additionally, microstructural characterization was performed using scanning electron microscopy to investigate the microstructure of the as-irradiated fuel before heat treatment and the microstructural changes that occurred after heat treating for the 730 and 950°C heat treatments for 6 hours. Results of the analysis were compared to those produced during diffusion studies performed using unirradiated TRIGA fuel meat and type 304 stainless steel. A gap is typically present in as-irradiated fuel that limits any possibility for chemical interaction between the fuel and cladding. Optical metallography (OM) was used to identify areas in the specimens heat treated at 950 and 1000°C where fuel and cladding contact was satisfactory for the FCCI to transpire. The bulk of the interaction zones that form due to FCCI developed in the fuel meat and not as much in the cladding. Even in the sample tested at 1,000°C, no evidence was found that gross fuel meat or cladding melting had occurred. Only limited porosity was found in the fuel meat that possibly could have been due to melting of a particular precipitate phase or phases that had formed during FCCI. These results agreed with those reported for the unirradiated diffusion couple studies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Radiation response of Ti 2 AlC MAX phase coated Zircaloy-4 for accident tolerant fuel cladding

A Ti 2 AlC MAX phase material was deposited on Zircaloy-4 substrate using the cold spray process, with the goal of enhancing the accident tolerance of zirconium-alloy fuel cladding in light water reactors. The samples were annealed up to 923 K to study the intermetallic phase formation between the coating and Zircaloy-4 substrate. Three interface compounds, ZrAl 2 , ZrAl, and Zr 3 Al, were identified and the activation energy of the interfacial intermetallic compound growth was determined. The radiation response of the coating material and interfacial intermetallic compounds after 3.5 MeV Zr 2+ ion irradiation was investigated by transmission electron microscopy and nanoindentation. Ion irradiation disrupted the nanolamellar structure of the Ti 2 AlC coating while amorphization was observed some of the intermetallic compounds. Furthermore, nanoindentation revealed overall hardening of the Ti 2 AlC coating and Zr 3 Al phase, with no significant change observed in the ZrAl 2 and ZrAl phases.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

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

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

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Analysis of a Loop for Fission Gas Release and Monitoring at the PULSTAR Reactor

Modern nuclear fuel performance analysis tools aim at establishing a predictive approach for examining macroscopic fuel behavior due to fuel gaseous swelling and the associated pellet cladding mechanical interaction (PCMI). In general, the macroscopic behavior is highly sensitive to understanding microscopically driven phenomena, which in this case is the release of fission gas (i.e., Kr and Xe) from the fuel. Previous studies identified the uncertainties associated with the intra-granular fission gas diffusion coefficient as most limiting for the accuracy of fission gas release calculations. In particular, it was concluded that, in order to improve fission gas release as well as gaseous swelling predictions using fuel performance codes, a better characterization of the diffusion coefficient is needed. This work explores the implementation of an on-line fission gas release and measurement loop at the North Carolina State University (NCSU) PULSTAR reactor. A unique characteristic of this loop is the ability to control the temperature of the fuel samples during irradiation to establish a more accurate correlation between the irradiation conditions and an observable such as the intragranular gas diffusion coefficient. Analysis was performed of an 'in-pile' fission gas release experiment that is currently under implementation at the NCSU PULSTAR reactor. Using MCNP6 calculations and COMSOL simulations the fission rate density and temperature distributions in UO{sub 2} microspheres were estimated. To explore the feasibility of detecting fission gas isotopes, estimates of R/B for {sup 138}Xe were obtained using the equivalent sphere model and combined with MCNP6 predicted B estimates to obtain R. Assuming detection in a typical gamma-ray spectrometry HPGe system, it was conclude that reasonable count rates will be achieved to allow extraction of release information. At this stage, accurate assessment is underway of the performance of various loop components to set the final design criteria in support of loop fabrication. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

FY 2014 Status Report: of Vibration Testing of Clad Fuel (M4FT-14OR0805033)

The DOE Used Fuel Disposition Campaign (UFDC) tasked Oak Ridge National Laboratory (ORNL) to investigate the behavior of light-water-reactor (LWR) fuel cladding material performance related to extended storage and transportation of UNF. ORNL has been tasked to perform a systematic study on UNF integrity under simulated normal conditions of transportation (NCT) by using the recently developed hot-cell testing equipment, Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT). To support the testing on actual high-burnup UNF, fast-neutron irradiation of pre-hydrided zirconium-alloy cladding in the High Flux Isotope Reactor (HFIR) at elevated temperatures will be used to simulate the effects of high-burnup on fuel cladding for help in understanding the cladding materials properties relevant to extended storage and subsequent transportation. The irradiated pre-hydrided metallic materials testing will generate baseline data to benchmark hot-cell testing of the actual high-burnup UNF cladding. More importantly, the HFIR-irradiated samples will be free of alpha contamination and can be provided to researchers who do not have hot cell facilities to handle highly contaminated high-burnup UNF cladding to support their research projects for the UFDC.t

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

N-BAR PROBLEMS AS APPROXIMATIONS TO THE BREE PROBLEM

The Bree solution to the problem of a ratcheting cylinder under constant pressure and cyclic thermal load is a fundamental result in nuclear engineering and widely used as the technical basis for the ASME Boiler and Pressure Vessel Code and other design methods. However, because the loading conditions in the Bree problem are difficult to achieve experimentally there have been relatively few works experimentally examining the problem and extending it to other relevant design situations, for example cladded components. In contrast, 2-bar problems are widely studied experimentally and are relatively easy to setup. These 2-bar problems are thought to be representative of Bree-type geometries, but a formal connection has not been demonstrated. This work formally establishes the connection between the Bree cylinder and an n-bar problem – a coupled bar experiment with, in general, more than two bars linked in parallel. The connection suggests that n-bar experiments using a fairly limited number of bars might be an experimentally-accessible setup that better represents ratcheting phenomenon in actual nuclear pressurized components. Such experiments could test surrogate cladded or multi-material components by using bars of different materials. Finally, this work suggests control schemes that yield optimally efficient n-bar experiments – experiments that best replicates a Bree cylinder with a limited number of bars.

Messner, Mark C.↗
DOE OSTI2019

{ital n}-bar problems as approximations to the Bree problem

The Bree solution to the problem of a ratcheting cylinder under constant pressure and cyclic thermal load is a fundamental result in nuclear engineering and widely used as the technical basis for the ASME Boiler and Pressure Vessel Code and other design methods. However, because the loading conditions in the Bree problem are difficult to achieve experimentally there have been relatively few works experimentally examining the problem and extending it to other relevant design situations, for example cladded components. In contrast, 2-bar problems are widely studied experimentally and are relatively easy to setup. These 2-bar problems are thought to be representative of Bree-type geometries, but a formal connection has not been demonstrated. This work formally establishes the connection between the Bree cylinder and an n-bar problem a coupled bar experiment with, in general, more than two bars linked in parallel. The connection suggests that n-bar experiments using a fairly limited number of bars might be an experimentally-accessible setup that better represents ratcheting phenomenon in actual nuclear pressurized components. Such experiments could test surrogate cladded or multi-material components by using bars of different materials. Finally, this work suggests control schemes that yield optimally efficient n-bar experiments experiments that best replicates a Bree cylinder with a limited number of bars.

Messner, M. C.↗
DOE OSTI2020

Evaluation of the effects of neutron irradiation on first-generation corrosion mitigation coatings on SiC for accident-tolerant fuel cladding

In this work, high purity SiC and SiC/SiC composites coated with commercial TiN, Cr, CrN, or CrN/Cr multilayer coatings were irradiated in Ar or flowing PWR water in the Massachusetts Institute of Technology Nuclear Reactor Laboratory (MITR). Irradiation in Ar was performed in the core. In the water environment, identical samples were placed in one of three different locations: in-core, providing exposure to neutron damage and radiolysis-affected water; above-core, where samples were exposed to radiolysis-affected water but not neutron damage, or outside of the core, where samples were exposed to the coolant water without the effects of radiation. Radiation in Ar revealed significant cracking of all but the TiN coatings, attributed to differential swelling between the coating and substrate. Lattice swelling was not observed in any of the coatings, but 0.2% void swelling was observed in the Cr coating. All of the coatings failed during water exposures in the core. Likewise, CrN/Cr spalled in each condition. Cr was protective, except under radiation damage as a result of cracking, and TiN severely degraded in the core with no coating was found following exposure. A SiC/coating ATF cladding system is anticipated to perform adequately following improvements in coating ductility and purity.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗