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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 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Results of Tests of VVER-1000 Fuel Elements with High Fuel Burnup in the MIR Reactor for LOC Incident Conditions

The main results of tests performed on VVER-1000 fuel elements with high burnup of fuel in a channel of the MIR channel reactor for LOC accident conditions are examined. The experimental setup contains one fuel element, which is arranged along the central axis. The aim of the tests is to determine the parameters of cladding depressurization. In the experiment, questions concerning fragmentation, axial displacement, and possible escape of fuel to outside the cladding were studied and the properties of the cladding material and fuel, which were subjected to the extreme parameters, were determined. The main parameters and the results of two experiments are presented. The conditions under which fuel-element depressurization occurred are determined.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Thermionic Fuel Element with Cermet Fuel

A method of calculating and the computational results obtained for the time variation of the radial deformation of the fuel-element cladding of a single-element EGC based on cermet with a polycrystalline molybdenum matrix in which uranium dioxide particles are distributed are presented. It is shown that the maximum admissible deformation of cladding comprised of 1-mm thick, hardened, single-crystalline, tungsten-tantalum alloy with a 10 mm in diameter central channel in the kernel is reached in 10 years.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Nuclear fuel elements including protective structures, and related method of forming a nuclear fuel element

A nuclear fuel element includes a core comprising a fissile element and an additional element. A protective structure surrounds the core and comprises at least a first material surrounding the nuclear fuel. The first material comprises the fissile element and the additional element and comprises a greater than stoichiometric amount of the additional element. An outer portion of the nuclear fuel element comprises a metal. Related nuclear fuel elements, and related methods are also disclosed.

Griffith, George W.↗
DOE OSTI2024

Elongate SiC fuel elements

An elongate fuel element is described that has a silicon carbide cladding enclosing a fuel, such as UO 2 , wherein the fuel is dimensioned relative to the cladding to define gaps at each lateral end of the enclosure sufficiently large such that upon swelling in use, the fuel does not increase the strain on the cladding beyond the limits of the claddings strain tolerance. The lateral gaps at the ends of the fuel allow lateral expansion during swelling that reduces the strain on the cladding.

Long, Yun-Bo↗
DOE OSTI2014

Methods for making a porous nuclear fuel element

Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Swelling of Uranium Dioxide–Silumin Dispersion Fuel Composition in Experimental Fuel Elements of the SM Reactor

A variant of experimental dispersion fuel elements with a displacer and uranium dioxide-silumin fuel composition, which were proposed for modernizing the core of the SM reactor, was investigated. The fuel elements were irradiated in a wide spectrum of neutron- and thermophysical parameters of the reactor’s reflector channel. The fuel elements showed satisfactory radiation resistance at thermal flux density 5.8–8 MW/m{sup 2} and burnup to 45%. The swelling of the fuel composition was analyzed. It is shown that of three possible components of the swelling the one due to gaseous fission products makes the main contribution, approximately 3/4 of the total amount. The interaction of the fuel components results in compaction of the kernel and promotes some compensation of the swelling, increasing as the interaction increases. These investigations show that the depth of burnup at the achieved thermal flux density for fuel elements with a displacer is not the limit.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

University of Missouri Research Reactor LEU Fuel Element Flow Test Conceptual Design—Hydraulic Reactor Design Parameters

The University of Missouri-Columbia Research Reactor (MURR®) is one of five U.S. high performance research reactors (USHPRR), plus one critical facility, that actively collaborates with the National Nuclear Security Administration (NNSA) Material Management and Minimization(M 3 ) Reactor Conversion Program to convert to the use of low-enriched uranium (LEU, < 20 wt.% U-235) fuel. A new type of LEU fuel with very high density, based on an alloy of uranium and 10 weight percent molybdenum (U-10Mo), is expected to allow the conversion to LEU of USHPRR that have been found unable to be converted with previously qualified uranium silicide-aluminum (U 3 Si 2 -Al) dispersion fuel. MURR has been working with the USHPRR Reactor Conversion (RC) Pillar at Argonne National Laboratory to perform fuel element design and fuel cycle performance analyses, steady-state thermal hydraulics safety analyses, and accident safety analyses in preparation for the conversion of MURR and to support a preliminary Safety Analysis Report (SAR) for conversion to LEU fuel. This work is performed in preparation for the flow test campaign that will be conducted by the USHPRR RC Pillar. The purpose of the hydraulic performance evaluation of the MURR LEU fuel element designed by the RC Pillar is to test a prototypic commercially fabricated LEU fuel element to determine whether any failure modes are observed or predicted in the fuel element, including significant deformations such as plate bending, twisting, or plate detachment from the side plate under selected safety-basis limits for reactor hydraulic conditions. To support the design of the flow test for MURR LEU fuel element hydraulic performance evaluation, design parameters for hydraulic testing of the LEU fuel element are laid out in this report.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Nuclear thermal propulsion fuel element packing device

A packing device for filling fuel elements with a powder through a fill aperture in an outer shell of the fuel element, including a stationary base, a clamp assembly including a body defining a bore therein, the bore being configured to slidably receive a fuel element therein, wherein the clamp assembly is movable along a vertical axis with respect to the stationary base, a cam assembly including a cam and a drive motor configured to rotate the cam, wherein rotation of the cam alternatingly raises the clamp assembly up along the vertical axis and subsequently drops the clamping assembly, and a powder reservoir assembly including a powder reservoir and a fill needle in fluid communication with the powder reservoir.

Lake, Russell L.↗
DOE OSTI2022

Ordered Particle Packing in Dense TRISO/SiC Fuel Elements and Preliminary Assessment of Neutronic and Thermomechanical Characteristics

Detailed analysis of the particle distribution in Transformational Challenge Reactor fuel elements indicates that particle packing is not random; instead, it follows a relatively ordered structure near fuel element surfaces. Discrete particle neutronic simulations indicate that the core reactivity is not impacted when assuming homogenization of particles with the silicon carbide matrix. However, the neutronic power distribution resulting from the ordered packing structure indicates that the highest-power particles reside at the top and bottom of the fuel elements and nearest the YH 1.85 moderator rods. The power distribution results were applied to thermo-mechanical simulations using mesh-based power distributions. Previous results indicated high stress at the bottom of the fuel element, where packing is most ordered. Additionally, to reduce this stress concentration, additively manufactured protrusions were added to the bottom of a test fuel element to disrupt dense particle packing. These protrusions reduced the overall power peaking, but the thermomechanical simulations did not indicate a significant change in the fuel element’s maximum stress or failure probability.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Methods of preserving a nuclear fuel element

A method of preserving a nuclear fuel includes exposing a surface of a fuel element comprising aluminum to a phosphorus-containing acid and reacting the phosphorus-containing acid with the aluminum to form aluminum phosphate (AlPO4). A nuclear fuel element includes a nuclear fuel and a shell surrounding the nuclear fuel. The shell comprises aluminum phosphate.

Rezvoi, Aleksey↗
DOE OSTI2024

Evaluation of flow-induced plate deflection for University of Missouri research reactor low-enriched uranium fuel element

The University of Missouri Research Reactor (MURR), located on the campus of the University of Missouri in Columbia, Missouri, is one of the six United States (U.S.) High Performance Research Reactors (USHPRR), including one critical facility, that are actively collaborating with the U.S. Department of Energy (DOE) National Nuclear Security Administration (NNSA) Office of Material Management and Minimization (M3) Reactor Conversion Program to convert from highly enriched uranium (HEU, ≥20 wt% U-235) fuel to low-enriched uranium (LEU, <20 wt% U-235) fuel. A new type of very high-density LEU fuel based on a monolithic alloy of uranium and 10 wt% molybdenum (U-10Mo) is expected to allow conversion of some USHPRR, including MURR. In the design of its fuel elements, MURR is using thin parallel curved fuel plates separated by coolant channels. In this work, fluid-structure interaction (FSI) analysis of the MURR LEU fuel element is performed at the element level (as compared to the plate level analysis), which models all components of the LEU fuel element, including fuel plates and the supporting structures. Therefore, the effect of supporting structures on the flow distribution within the element and the fuel plate deflection are evaluated. In addition to the element nominal flow rate and dimensions, the tolerances in the geometry of the coolant channel and plate thickness, the effect of a comb on plate deflection, and the uncertainty of the flow rate per element are evaluated. For the LEU fuel plates, which are thinner than the current HEU plates, the predicted plate deflection is found to be small compared to the fabrication and assembly tolerances. Thus, the FSI-induced deflections are not expected to noticeably reduce the coolant flow rate or predicted safety margins in the limiting channels for the MURR LEU fuel element. In addition to the simulation work, a hydraulic performance test of the MURR LEU fuel element is currently being planned to support conversion to the use of LEU fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Study on the mechanism of diametral cladding strain and mixed-oxide fuel element breaching in slow-ramp extended overpower transients

Cladding strain caused by fuel/cladding mechanical interaction (FCMI) was evaluated for mixed-oxide fuel elements subjected to 70–90% slow-ramp extended overpower transient tests in the experimental breeder reactor II. Calculated transient-induced cladding strains were correlated with cumulative damage fractions (CDFs) using cladding strength correlations. In a breached high-smeared density solid fuel element with low strength cladding, cladding thermal creep strain was significantly increased to approximately half the transient-induced cladding strain that was considered to be caused by the tertiary creep when the CDF was close to the breach criterion (=1.0), with the remaining strain due to instantaneous plastic deformation. In low-smeared density annular fuel elements, FCMI load was significantly mitigated and resulted in little cladding strain. The CDFs of the annular fuel elements were lower than 0.01 at the end of the overpower transient, indicating a substantial margin to breach. A substantial margin to breach was also maintained in a high-smeared density fuel element with high strength cladding.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2016

The electric current effect on electrochemical deconsolidation of spherical fuel elements - 18447

For High-Temperature Gas-cooled Reactor in China, fuel particles are bonded into spherical fuel elements by a carbonaceous matrix. In conjunction with the development of a system to determine fuel failure fractions through {gamma}-analysis of individual and undamaged fuel particles, an automated electrochemical deconsolidation apparatus was developed in this study to separate the particles from the carbonaceous matrix by disintegrating the matrix into fine graphite powder, based on the anodic oxidation of graphite in an electrolyte containing nitric acid. A series of tests were conducted under a range of deconsolidation conditions to develop procedures and refine the equipment. The easy-handle un-irradiated spherical fuel elements were used as samples for initial tests. It is found that the morphology, size, and element content of deconsolidated graphite matrix and free particles were notably affected by electric current intensity. During the electrolysis, the nitric acid electrolyte gradually infiltrated into the micropores of graphite matrix, and the electrochemical reaction would take place around micropore structure as well as matrix surface. With much higher surface reaction capability due to the large specific area of micropore structure, the expansion stress originated from inner surface of micropores widened the c-spacing of the layer lattice of carbonaceous matrix, which was thought to be one of the main driving force of deconsolidation. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Fuel – clad chemical interaction evaluation of the TREAT reactor conceptual low-enriched-uranium fuel element

The Transient Reactor Test (TREAT) facility resides at the Materials and Fuels Complex (MFC) at Idaho National Laboratory (INL). The TREAT reactor is currently undergoing design and engineering studies for its conversion from a high enriched uranium (HEU) to a low-enriched uranium (LEU) core. The conceptual design of the LEU fuel element identified two main design differences compared with the HEU fuel element; namely, it will contain four times more fissionable material in its graphite matrix and distinct nuclear-grade Zirconium alloy, as Zircaloy-3 was used in the HEU fuel assembly and is not commercially available currently. These design changes may impact the magnitude of chemical interaction between fuel and cladding materials during physical contact under expected TREAT operation conditions and, therefore, was evaluated through a combination of experimental testing and thermodynamic modeling in order to determine implications for the fuel assembly. In this study, two potential cladding material types, Zircaloy-4 or Zr-1Nb alloys, were evaluated, and it was found for both material types that the extent of interaction and specific chemical reactions are minimal and no detrimental effect on the overall cladding properties is observed. Here, the resulting interaction layer of 3-6 µm was measured after a 2-week exposure at 820°C. The thermodynamic analysis was extended to temperatures beyond the TREAT reactor operation and accident conditions in order to give some insight that may be of interest for other reactor systems as the High Temperature Gas Reactors (operation above 1,000°C) and for Nuclear Reactor Severe Accident phenomenology study where the UO 2 fuel could reach temperatures over 2,800°C and melt.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Post irradiation examination of a uranium-zirconium hydride TRIGA fuel element

Low-enriched (LEU) U-ZrH fuel, with a 235 U content less than 20% of the total uranium, is being evaluated for possible use in different types of reactors, including space nuclear systems, light water reactors (LWRs) and micro-reactors. As a result, it is beneficial to better understand the macrostructural and microstructural changes that occur in this fuel during irradiation. This paper reports the results of the post irradiation examination of an LEU U-ZrH fuel element (30 wt.% U, <20% 235 U) using neutron radiography, precision gamma scanning, chemical analysis, optical metallography and scanning electron microscopy combined with energy dispersive spectroscopy and wavelength dispersive spectroscopy, where the fuel element was irradiated in a Training, Research, Isotope, General Atomics (TRIGA) reactor. Results of microstructural characterization indicated some dehydriding and cracking of the U-ZrH fuel occurred during irradiation; an axial and radial burnup gradient existed in the fuel during irradiation, as measured by gamma scanning and chemical analysis; negligible microstructural changes transpired during irradiation, based on comparison of irradiated and as-fabricated U-ZrH fuel microstructures; and, negligible, fission product-rich, phases could be resolved in a U-ZrH fuel that was irradiated to a calculated 20% depletion of 235 U.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Engineering Considerations for Remote Refabrication of EBR-II Fuel Elements

The development of remote controlled methods for manufacturing EBR-II fuel elements was influenced by many interacting factors. Radiation levels within the process cell have been predicted to range from 10 3 to 10 7 rad per hour. Radiation damage to organic lubricant, electrical insulations, elastic seals, and protective coatings precludes the use of many standard machine components. Heat generated in the fuel by absorbed radiation makes forced cooling necessary in many operations. Oxygen must be exluded from all operations where the fuel is exposed. Equipment must be designed for remote maintenance and component replacement within the limitation of available manipulators. The EBR-II fuel consisted of fissium alloy pins sodium bonded in stainless steel tubes. Precision casting was chosen as the basis for refabricating the fuel pins. Remote controlled equipment was developed to cast, assemble, and inspect the EBR-II fuel elements. Radiation resistant, plug-in machine components were developed to give reasonable life expectancy and to allow remote maintenance and replacement.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Evaluation of Erosion of the Dummy “EE” Plate 19 in YA Type ATR Fuel Element During Reactor PALM Cycles

Advanced Test Reactor (ATR) Cycle 153B-1 was a 14-day, high-power, powered axial locator mechanism (PALM) operating cycle that completed on April 12, 2013. Cycle 153B-1 was a typical operating cycle for the ATR, and did not result in any unusual plant transients. ATR was started up and shut down as scheduled. The PALM drive physically moves the selected experiments into and out of the core to simulate reactor startup and heat up, and shutdown and cooldown transients, while the reactor remains in steady-state conditions. However, after the cycle was over, when the fuel elements were removed from the core and inspected, several thousand flow-assisted erosion pits and “horseshoeing” defects were readily observed on the surface of the several YA-type fuel elements (these are aluminum “dummy” plates that contain no fuel). In order to understand these erosion phenomena, a thermal-hydraulic model of coolant channel 20 on a YA-M fuel element was generated. The boundaries of the model were the aluminum EE plate of a YA-M fuel element and a beryllium reflector block with 13 horizontal saw cuts which represented regions of zero flow. The heat generated in fuel plates 1 through 18 was modeled to be passing through the aluminum EE plate. The coolant channel 20 width was set at 0.058 in. (58 mils). It was established that the horizontal saw cuts had a significant effect on the temperature of the coolant. The flow, which was expected to vary linearly with gradual heating of the coolant as it passed through the channel, was extremely turbulent. The temperature rise, which was expected to be a smooth “S” curve, was represented by a series temperature rise “humps,” which occurred at each horizontal saw cut in the beryllium reflector block. Each of the 13 saw cuts had a chamfered edge which resulted in the coolant flow being re-directed as a jet across the coolant channel into the surface of the EE plate, which explained the temperature rise and the observed scalloping and pitting degradation on the YA-M fuel elements. In the case of scalloping (horseshoeing) a surprising similarity of that defect to those appearing on aluminum plate rolled in over-lubrication conditions, were established. In turn, this made us think that the principal feature responsible for the appearance of these defects, was horizontal cuts in the beryllium reflector block created to arrest the propagation of large vertical crack(s) in Be in PALM cycles with higher overall fluence. This assumption was fully confirmed by the results of thermo-hydraulic simulations. The neutronics data for these modeling experiments were provided using advanced irradiation simulations (MCNP, HELIOS). In the case of pitting erosion the following corrective measures were proposed based upon the results of JMatPro v.8.2 modeling (TTT- and CCT-diagrams): change the fabrication process by adding blister anneal before program anneal, immediately after cold rolling of AA6061plate. This step will allow achieving complete recrystallization, eliminating of strengthening due to metastable precipitates, and reduce the possibility of forming sharp microstructural features upon the surface.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Behavior of Modified Zirconium-Alloy Fuel-Element Cladding Under Irradiation

The first reactor tests and post-reactor studies of fuel elements with cladding made from the zirconium alloys E-636M, E-635opt, and E-635M1, developed at VNIINM, have been conducted. The purpose of the studies was to search for alloys of the type E-635 that surpass the base alloy in terms of resistance to corrosion and hydration. The studies make it possible to recommend the alloys E-635M and E-635opt for use as cladding of the fuel elements in extended-life cores.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗