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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 865 records · Page 48
All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

Potential Effect of Interfacial Bonding on Used Nuclear Fuel Vibration Reliability

This summary abstract describes the methodology used to evaluate the effect of pellet pellet and pellet clad interactions with consideration of the interfacial bonding efficiency on UNF vibration integrity. This methodology provides a solid roadmap for further protocol development with respect to effective lifetime prediction of a UNF system under normal transportation vibration. The proposed methodology that couples FEA simulations and experimental exploration efforts is also under development. The current methodology is focused on assessing the influence of interfacial bonding at the pellet pellet and the pellet clad interfaces on UNF vibration integrity. The FEA simulation results were also calibrated and benchmarked with the fatigue aging data obtained from reversible bending fatigue testing.

fuel-pellet interface↗
DOE OSTI2017

Test facility for fission chamber assemblies - 239

The Institution -Project Center ITER- with participation of the National Research Nuclear University MEPhI are developing the neutron diagnostics -Divertor Neutron Flux Monitor- (DNFM) for the International Thermonuclear Experimental Reactor (ITER). This diagnostics shall ensure measurement of the total neutron yield and ITER fusion power. Assemblies of fission chambers (FC) with different fissile materials and sensitivities are planned to be used for the neutron flux measurement. Study of fission chambers characteristics and acceptance testing of the diagnostic system will be carried out at the testing facility of the neutronic laboratory of the 'TRINITI' Institute. The facility includes: compact powerful neutron generator with the power, control and monitoring system; remote handling system for FCs positioning and movement; novel three channels data acquisition and processing system. Each acquisition channel contains some measuring circuits with signal processing. This set of circuits provides neutron flux measurement in the range of at least 10 decades for each single FC. Data acquisition and processing system features are: signal digitizing with sample rate up to 125 MHz; real time data processing with computation and data update cycle from 0.1 ms. This paper examines the structure, composition and specification of the test facility. Application software, hardware solutions and the results of fission chambers prototypes tests are discussed. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

High Temperature Test Facility Preliminary RELAP5-3D Input Model Description

A RELAP5-3D input model is being developed for the High Temperature Test Facility at Oregon State University. The current model is described in detail. Further refinements will be made to the model as final as-built drawings are released and when system characterization data are available for benchmarking the input model.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

An Adaptive Reference Scheme to Extend the Functional Range of Optical Backscatter Reflectometry in Extreme Environments

Optical frequency domain reflectometry (OFDR) measurements are performed by recording the interference pattern of light backscattered by density fluctuations or Bragg gratings along the length of an optical fiber. Changes in local temperature or strain in the fiber cause shifts in the backscattered light spectrum, which can be calibrated to the applied temperature or strain. Comparing the backscattered optical spectra from each reflection site with a reference (unstrained) spectrum allows for quantification of local spectral shifts. While mature OFDR-based technologies can provide high-precision spatially distributed measurements over kilometer lengths, the post-processing approaches used to recover spectral shifts from raw optical intensity data can limit the use of OFDR in harsh environments. This paper presents a novel approach to post-processing OFDR data which extends the usability of optical fibers exposed to harsh environments. This approach is the first to use the spectral shift quality, calculated for two OFDR measurements, to identify an appropriate reference scan from which the spectral shift can be determined. Here, three experimental data sets are used to test the algorithm: (1) an optical fiber heated to >950°C, (2) an aluminum-embedded optical fiber under strain from differential thermal expansion, and (3) an optical fiber exposed to a total neutron flux of 2×10 13 n/cm 2 /s over 40 hours in a nuclear test reactor. Results show that using a metric of quality to select an appropriate reference measurement extends the functional range of OFDR strain and temperature sensors. Furthermore, this algorithm can be applied to existing OFDR data.

47 OTHER INSTRUMENTATION↗
DOE OSTI2013

Validation of nuclear models used in space radiation shielding applications

A program of verification and validation has been undertaken to assess the applicability of models to space radiation shielding applications and to track progress as these models are developed over time. In this work, simple validation metrics applicable to testing both model accuracy and consistency with experimental data are developed. The developed metrics treat experimental measurement uncertainty as an interval and are therefore applicable to cases in which epistemic uncertainty dominates the experimental data. To demonstrate the applicability of the metrics, nuclear physics models used by NASA for space radiation shielding applications are compared to an experimental database consisting of over 3600 experimental cross sections. A cumulative uncertainty metric is applied to the question of overall model accuracy, while a metric based on the median uncertainty is used to analyze the models from the perspective of model development by examining subsets of the model parameter space.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2012

Identity method-a new tool for studying chemical fluctuations

Event-by-event fluctuations of the chemical composition of the hadronic system produced in nuclear collisions are believed to be sensitive to properties of the transition between confined and deconfined strongly interacting matter. In this paper a new technique for the study of chemical fluctuation, the identity method, is introduced and its features are discussed. The method is tested using data on central PbPb collisions at 40 A GeV registered by the NA49 experiment at the CERN SPS.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2021

Brine Availability Test in Salt (BATS) FY21 Update

This report summarizes the 2021 fiscal year (FY21) status of ongoing borehole heater tests in salt funded by the disposal research and development (R&D) program of the Office of Spent Fuel & Waste Science and Technology (SFWST) of the US Department of Energy’s Office of Nuclear Energy’s (DOE-NE) Office of Spent Fuel and Waste Disposition (SFWD). This report satisfies SFWST milestone M2SF- 21SN010303052 by summarizing test activities and data collected during FY21. The Brine Availability Test in Salt (BATS) is fielded in a pair of similar arrays of horizontal boreholes in an experimental area at the Waste Isolation Pilot Plant (WIPP). One array is heated, the other unheated. Each array consists of 14 boreholes, including a central borehole with gas circulation to measure water production, a cement seal exposure test, thermocouples to measure temperature, electrodes to infer resistivity, a packer-isolated borehole to add tracers, fiber optics to measure temperature and strain, and piezoelectric transducers to measure acoustic emissions. The key new data collected during FY21 include a series of gas tracer tests (BATS phase 1b), a pair of liquid tracer tests (BATS phase 1c), and data collected under ambient conditions (including a period with limited access due to the ongoing pandemic) since BATS phase 1a in 2020. A comparison of heated and unheated gas tracer test results clearly shows a decrease in permeability of the salt upon heating (i.e., thermal expansion closes fractures, which reduces permeability).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Oak Ridge Response to Versatile Test Reactor Environmental Impact Statement Data Request

The Versatile Test Reactor (VTR) is a fast-spectrum test reactor being developed in the United States under the direction of the US Department of Energy Office of Nuclear Energy (DOE-NE). The VTR mission is to enable accelerated testing of advanced reactor fuels and materials required for advanced reactor technologies. The conceptual design of the 300 MWth sodium-cooled metallic-fueled pool-type fast reactor has been led by the US National Laboratories in collaboration with General Electric–Hitachi and Bechtel National, Inc. In support of the VTR project, DOE issued a Notice of Intent (NOI) in the Federal Register on August 5, 2019, announcing the intent to prepare an Environmental Impact Statement (EIS) in accordance with the National Environmental Policy Act (NEPA) and its implementing regulations. The EIS will evaluate alternatives for a versatile reactor–based fast-neutron source facility and associated facilities for the preparation, irradiation, and post-irradiation examination (PIE) of test/experimental fuels and materials. Specifically, the NOI identified two siting alternatives for the VTR reactor facility: Idaho National Laboratory (INL) or Oak Ridge National Laboratory (ORNL). In addition, the NOI also specified two siting alternatives for VTR fuel fabrication: INL and the Savannah River Site (SRS). This report provides information in response to data requests made to ORNL to fill in site-specific knowledge gaps to develop a high-quality EIS. The responses provided are not required to provide full details in every aspect; instead, they adequately bound possible environmental impacts or provide sufficient information to adequately assess likely environmental impacts. This work is being performed under a subcontract from INL to ORNL using DOE-NE funds and is directed by DOE-NE and DOE-ID. Leidos has been contracted by DOE-NE to write the VTR EIS, so most data requests have come from Leidos but were often routed through INL or DOE-ID. DOE-ID is overseeing the NEPA and EIS processes for the VTR project. Leidos will use the information provided in this report to inform the VTR EIS and will also cite this document to establish a clear, publicly available source of the information. Section 2 of this report briefly describes the proposed ORNL VTR Alternative and illustrates the location of the proposed site for the ORNL VTR Alternative. Sections 3 through 7 provide direct responses to data requests received by ORNL. These sections use a tabular format in which data requests are divided into separate items to be addressed; the items are numbered, the data requests are restated with more topical information included, and then the responses are provided. Initial data requests and follow-on requests for additional information (RAIs) are combined under the original data request fields. Finally, Section 8 presents summarized conclusions and describes future work.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Update on the Electron Source Model [Slides]

We summarize work done in FY12 on the Los Alamos Electron Source Model (ESM), which predicts the distribution of beta-decay electrons after a high altitude nuclear explosion (HANE) as a function of L, energy, and pitch angle. In the last year we have compared model results with data taken after the Russian 2 HANE test of 1962 and presented results at the HEART conference. We discuss our future plans to continue comparison with HANE data and to develop the code to allow a more complex set of initial conditions.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗
DOE OSTI2021

Technical Documents for Gateway for Accelerated Innovation in Nuclear (GAIN)

The Gateway for Accelerated Innovation in Nuclear (GAIN) provides the nuclear energy community with access to the technical, regulatory, and financial support necessary to move new or advanced nuclear reactor designs toward commercialization. GAIN provides the nuclear community with a single point of access to the broad range of capabilities (i.e., people, facilities, materials, and data) across the U.S. Department of Energy (DOE) complex and its National Lab capabilities. The Fast Flux Test Facility (FFTF) is the most recent liquid metal reactor (LMR) to be designed, constructed, and operated by DOE. The 400-MWt sodium-cooled, fast-neutron flux reactor plant was designed for irradiation testing of nuclear reactor fuels and materials for liquid metal fast breeder reactors. Following the demise of the breeder reactor program in the United States, FFTF continued to play a key role in providing a test bed for demonstrating performance of advanced fuel designs and demonstrating operation, maintenance, and safety of advanced liquid metal reactors. The FFTF Program provides valuable information for potential follow-on reactor projects in the areas of plant system and component design, component fabrication, fuel design and performance, reactor control, prototype testing, and site construction. This report provides documents related to three important aspects of FFTF design and operation: 1) irradiation behavior of structural alloys and absorber materials, 2) thermohydraulics of rod bundles (i.e., coolant mixing), and 3) natural circulation heat transfer in the areas of modeling and validation. These technical documents are believed to be of interest to the nuclear industry and in particular to designers of new liquid metal reactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

EFFECTS OF SIZE AND GEOMETRY ON THE EQUIBIAXIAL FLEXURAL TEST OF FINE GRAINED NUCLEAR GRAPHITE

Reduction of specimen size is an essential requirement for neutron irradiation effect studies of materials since space in nuclear reactors for those studies are often very limited and costly. Reducing the specimen size can dramatically reduce the cost of these studies, the irradiation exposure to personnel involved in post irradiation examination and the volume of radioactive waste. In this study, 4 specimen sizes were systematically varied to examine the size effect on the equibiaxial flexure test of fine grain nuclear graphite in non-irradiated conditions at room temperature. A statistically significant sample population of 30 was used for each specimen geometry or dimension. The results showed that the round shaped samples followed weibull distribution with Weibull modulus of 25 to 35. The average strength followed the trend of the weakest link theory. The 2 sizes of square shaped samples had Weibull modulus of 29 and 40. Data from round and square samples agree with others reasonably and collectively follow the Weibull scaling theory. It is concluded that the equibiaxial flexural test using adequate population of small specimens provides useful and reliable information on the statistical strength properties for fine-grained nuclear graphite materials.

Shih, Chunghao↗
DOE OSTI2011

Development, calibration and experimental results obtained with an innovative calorimeter (CALMOS) for nuclear heating measurements

Nuclear heating inside an MTR reactor has to be known in order to be able to control samples temperature during irradiation experiments. An R and D program has been carried out at CEA to design a new type of in-core calorimetric system. This new development, started in 2002, has for main objective to manufacture a calorimeter suitable to monitoring nuclear heating inside the 70 MWth OSIRIS material testing reactor operated by CEA's Nuclear Energy Div. at the Saclay research center. An innovative calorimetric probe, associated to a specific handling system, has been designed to provide access to measurements both along the fissile height and on the upper part of the core, where nuclear heating still remains high. Two mock-ups of the probe were manufactured and tested in 2005 and 2009 in ex-core area of OSIRIS reactor for process validation, while a displacement system has been especially studied to move the probe along a given axial measurement range. This paper deals with the development, tests on preliminary mock-ups and the finalization of the probe. Main modeling and experimental results are presented. Moreover, alternative methods to calibration for nuclear heating rate measurements which are now possible with this new calorimeter are presented and discussed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Development, calibration, and experimental results obtained with an innovative calorimeter (CALMOS) for nuclear heating measurements

Nuclear heating inside an MTR reactor has to be known in order to be able to control samples temperature during irradiation experiments. An R and D program has been carried out at CEA to design a new type of in-core calorimetric system. This new development, started in 2002, has for main objective to manufacture a calorimeter suitable to monitoring nuclear heating inside the 70 MWth OSIRIS material testing reactor operated by CEA's Nuclear Energy Division at the Saclay research center. An innovative calorimetric probe, associated to a specific handling system, has been designed to provide access to measurements both along the fissile height and on the upper part of the core, where nuclear heating still remains high. Two mock-ups of the probe were manufactured and tested in 2005 and 2009 in ex-core area of OSIRIS reactor for process validation, while a displacement system has been especially studied to move the probe along a given axial measurement range. This paper deals with the development, tests on preliminary mock-ups and the finalization of the probe. Main modeling and experimental results are presented. Moreover, alternative methods to calibration for nuclear heating rate measurements which are now possible with this new calorimeter are presented and discussed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Monolithic U-10Mo Fuel Development for U.S. High Performance Research Reactors

To support the U.S Department of Energy (DOE), National Nuclear Security Administration (NNSA) mission to reduce the threat posed by the civilian use of highly enriched uranium (HEU) worldwide, the US High Performance Research Reactor (US HPRR) fuel development program is pursuing development and qualification of a new high-density monolithic U-Mo fuel to facilitate conversion of five higher power research reactors located in the U.S. (ATR, HFIR, NBSR, MIT and MURR) and one critical facility (ATR-C) from HEU to low enriched uranium (LEU). These reactors require higher uranium fuel density than is achievable with dispersion fuel systems used to convert lower power research reactors. The down selected fuel system consists of U-10Mo alloy fuel foils having a thin Zr diffusion barrier interlayer, clad in 6061 Al alloy by hot isostatic pressing. In order to support fabrication development, fuel performance evaluations, generic fuel qualification through the Nuclear Regulatory Commission (NRC), and eventual reactor licensing, a series of irradiation experiments and testing campaigns are being planned. The tests are designed to provide data needed for fabrication process down selection, to establish acceptable fuel performance under normal operating conditions and anticipated transients, and to demonstrate successful scale-up to prototypic reactor-specific fuel element designs. A comprehensive testing program is being executed as part of the US HPRR fuel development program to develop and qualify a new monolithic UMo fuel system to allow conversion of five high power research reactors and one critical facility within the US in support of the NNSA mission to convert research reactors worldwide from HEU to LEU fuels. A significant outcome of this program will be the qualification and eventual licensing of a new plate-type fuel system for use in research reactors through the NRC for the first time in nearly 3 decades. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Heat transfer from glass melt to cold cap: Computational fluid dynamics study of cavities beneath cold cap

Efficient glass production depends on the continuous supply of heat from the glass melt to the floating layer of batch, or cold cap. Computational fluid dynamics (CFD) are employed to investigate the formation and behavior of gas cavities that form beneath the batch by gases released from the collapsing primary foam bubbles, ascending secondary bubbles, and in the case of forced bubbling, from the rising bubbling gas. The gas phase fraction, temperature, and velocity distributions below the cold cap are used to calculate local and average heat transfer rates as a function of the bubbling rate. It is shown that the thickness of the cavities is nearly independent of the cold cap shape and the amount of foam evolved during batch conversion. Furthermore, it is ~7 mm and up to ~15 mm for the cases without and with forced bubbling used to promote circulation within the melt, respectively. Using computed velocity and temperature profiles, the melting rate of the simulated high-level nuclear waste glass batch was estimated to increase with the bubbling rate to the power of ~0.3 to 0.9, depending on the flow pattern. The simulation results are in good agreement with experimental data from laboratory- and pilot-scale melter tests.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

Corrective Action Investigation Plan for Corrective Action Unit 104: Area 7 Yucca Flat Atmospheric Test Sites, Nevada National Security Site, Nevada, Revision 0

CAU 104 comprises the 15 CASs listed below: (1) 07-23-03, Atmospheric Test Site T-7C; (2) 07-23-04, Atmospheric Test Site T7-1; (3) 07-23-05, Atmospheric Test Site; (4) 07-23-06, Atmospheric Test Site T7-5a; (5) 07-23-07, Atmospheric Test Site - Dog (T-S); (6) 07-23-08, Atmospheric Test Site - Baker (T-S); (7) 07-23-09, Atmospheric Test Site - Charlie (T-S); (8) 07-23-10, Atmospheric Test Site - Dixie; (9) 07-23-11, Atmospheric Test Site - Dixie; (10) 07-23-12, Atmospheric Test Site - Charlie (Bus); (11) 07-23-13, Atmospheric Test Site - Baker (Buster); (12) 07-23-14, Atmospheric Test Site - Ruth; (13) 07-23-15, Atmospheric Test Site T7-4; (14) 07-23-16, Atmospheric Test Site B7-b; (15) 07-23-17, Atmospheric Test Site - Climax These sites are being investigated because existing information on the nature and extent of potential contamination is insufficient to evaluate and recommend corrective action alternatives (CAAs). Additional information will be obtained by conducting a corrective action investigation before evaluating CAAs and selecting the appropriate corrective action for each CAS. The results of the field investigation will support a defensible evaluation of viable CAAs that will be presented in the Corrective Action Decision Document. The sites will be investigated based on the data quality objectives (DQOs) developed on April 28, 2011, by representatives of the Nevada Division of Environmental Protection and the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Site Office. The DQO process was used to identify and define the type, amount, and quality of data needed to develop and evaluate appropriate corrective actions for CAU 104. The releases at CAU 104 consist of surface-deposited radionuclides from 30 atmospheric nuclear tests. The presence and nature of contamination at CAU 104 will be evaluated based on information collected from a field investigation. Radiological contamination will be evaluated based on a comparison of the total effective dose (TED) to the dose-based final action level (FAL). The presence of TED exceeding the FAL is considered a radiological contaminant of concern (COC). Anything identified as a COC will require corrective action. The TED will be calculated as the total of separate estimates of internal and external dose. Results from the analysis of soil samples will be used to calculate internal radiological dose. Thermoluminescent dosimeters will be used to measure external radiological dose. Based on process knowledge of the releases associated with the nuclear tests and radiological survey information about the location and shape of the resulting contamination plume, it was determined that the releases from the nuclear tests are co-located and will be investigated concurrently. A field investigation will be performed to define areas where TED exceeds the FAL and to determine whether other COCs are present at the site. The investigation will also collect information to determine the presence and nature of contamination associated with migration and excavation, as well as any potential releases discovered during the investigation. Appendix A provides a detailed discussion of the DQO methodology and the DQOs specific to each CAS.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2012

Rainier Mesa/Shoshone Mountain CAU Technical Basis Agreement Document: Hydrologic Source Term

The Hydrologic Source Term (HST) represents the transient release of residual radioactivity into groundwater associated with an underground nuclear test. An HST model seeks to quantitatively calculate these releases from a mathematical representation of release processes and physical and chemical data related to the test and the altered test environment. The development of an HST model requires a phenomenological understanding of underground detonation effects and their role in altering the physical and chemical environment surrounding a test, an estimate of the abundance, physical form, and spatial distribution of radioactivity in the post-test environment – the radiologic source term (RST), and consideration of mechanisms and processes that serve to transfer radioactive compounds into groundwater following a detonation. Simplified representations of the HST model are required for use in populating individual source regions in larger scale models of groundwater flow and contaminant transport within the sub CAU and CAU scale models. By necessity, such models must abstract and generalize many of the complicated release mechanisms over the large number of tests encountered in the CAU.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2024

Recent High Burnup LOCA Testing at Oak Ridge National Laboratory

To improve fuel cycle economics, increasing the fuel burnup limit in light-water reactors requires a solid technical foundation. Observations from experiments at Halden and Studsvik highlighted severe fuel fragmentation during loss-of-coolant-accident (LOCA) conditions, indicating a need for further technical considerations. These experiments suggest that the fragmentation threshold for high-burnup fuel might be influenced by pre-transient power levels. Consequently, additional LOCA test data are essential to complement existing findings and to deepen our understanding. Oak Ridge National Laboratory’s Severe Accident Test Station has been instrumental in advancing knowledge about high-burnup fuel fragmentation, relocation, and dispersal. This milestone report details two high-burnup LOCA tests designed to evaluate the effects of terminal temperature and grid spacers (or cladding restraints) on fuel fragmentation and relocation susceptibility. In addition, BISON fuel performance modeling and out-of-cell benchmark testing were conducted to better interpret the in-cell test results. These tests were developed in collaboration with fuel vendors to ensure that the results provide valuable data for topical reports and support the Nuclear Regulatory Commission’s review.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗