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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 703 records · Page 39

Preliminary Conceptual Neutronic, Thermal, and Mechanical Design for {sup 238}Pu ATR Targets

Pu-238 is a valuable heat source material used to manufacture radioisotope power systems (RPSs), which are needed to provide electrical power for space exploration and other missions. Before 1989, the United States Department of Energy (DOE) produced Pu-238 at the Savannah River Site in South Carolina in nuclear reactors that are now decommissioned. From 1989 through 2009, existing DOE inventories of Pu- 238 were supplemented by quantities of Pu-238 purchased from Russia; however, supplies available for purchase from Russia are limited. DOE has restored domestic production to meet future demand; otherwise, NASA will be in danger of not being able to use RPSs in future missions. The Pu-238 Supply Program has the objective of reestablishing the domestic capability for producing Pu-238. In order to meet the quantities of Pu-238 needed to support NASA space mission and other federal users the ATR is being considered as an important asset to achieve the desired production of the radioisotope. In order to determine the constraints and parameters needed to maximize the production of Pu-238 at ATR a preliminary conceptual design target design was performed. The preliminary conceptual design had two principal goals: 1. Perform a trade study to determine an ATR target length that maximizes production as well as minimize the need to modify ORNL's target. 2. Evaluate the selected preliminary design from the standpoint of the ATR nuclear safety basis. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

ATR Complex and Plant Modifications 2020 2021

Allow myself to email Scott Ullsperger the training document to allow him to become more familiar with what was updated with the Core Internal Change out (CIC) for ATR. Scott is currently working on getting set up as a sub contractor. Scott Ullsperger is a retired BEA employee who has a thorough knowledge of the reactor. Scott has been selected to participate in the MSA for reactor start up.

99 GENERAL AND MISCELLANEOUS↗

Gamma-Ray Emitting Radionuclides Concentrations and Decontamination Factors of ATR Loop Liquid Samples Cycle 171B

This report contains the gamma-ray emitting radionuclides concentration and decontamination factor results from gamma-ray spectrometry measurements of ATR loop liquid samples by the Radiation Measurements Laboratory (RML) for ATR Cycle 171B Jun. 20, 2023, to Sep. 27, 2023. This report consists of five subsections, one for each loop. Each subsection contains the results for each sampling of the loop along with the date and time of sampling, spectral data identifications, sample type, and reactor power (MW). The results are reported in units of disintegrations per minute per milliliter of sample. The results are obtained with the gamma-ray spectral data analysis routines of the RML computer. Some of this report is also generated by RML data storage and handling routines.

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Evaluation of Sedimentary Structure Near the Advanced Test Reactor Complex at the Idaho National Laboratory

An analysis of relevant sediment stratigraphy near Idaho National Laboratory’s Advanced Test Reactor Complex in support of the proposed remote-handled low-level waste disposal facility is presented. The scope of analysis includes a geostatistical evaluation of sedimentary interbed top elevations and thicknesses compiled for wells near the Advanced Test Reactor Complex from land surface to the elevation of the water table. It also includes an evaluation of expected sediment texture and composition based on information contained in existing reports. Finally, based on this analysis, recommendations for characterization well spacing and appropriate transport model dimensionality are made.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Gamma-Ray Emitting Radionuclides Concentrations and Decontamination Factors of ATR Loop Liquid Samples Cycle 173A Revision 0

This report contains the gamma-ray emitting radionuclides concentration and decontamination factor results from gamma-ray spectrometry measurements of ATR loop liquid samples by the Radiation Measurements Laboratory (RML) for ATR Cycle 173A running May 3, 2024, to July 6, 2024. This report consists of five subsections, one for each loop. Each subsection contains the results for each sampling of the loop along with the date and time of sampling, spectral data identifications, sample type, and reactor power (MW). The results are reported in units of disintegrations per minute per milliliter of sample. The results are obtained with the gamma-ray spectral data analysis routines of the RML computer. Some of this report is also generated by RML data storage and handling routines.

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Gamma-Ray Emitting Radionuclides Concentrations and Decontamination Factors of ATR Loop Liquid Samples: Cycle 172A

This report contains the gamma-ray emitting radionuclides concentration and decontamination factor results from gamma-ray spectrometry measurements of ATR loop liquid samples by the Radiation Measurements Laboratory (RML) for ATR Cycle 172A running March 2, 2024, to March 16, 2024. This report consists of five subsections, one for each loop. Each subsection contains the results for each sampling of the loop along with the date and time of sampling, spectral data identifications, sample type, and reactor power (MW). The results are reported in units of disintegrations per minute per milliliter of sample. The results are obtained with the gamma-ray spectral data analysis routines of the RML computer. Some of this report is also generated by RML data storage and handling routines.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Scoping Study of Effects on the Advanced Test Reactor’s Center Flux Trap Due to Various Loadings in the H and Inner A Test Positions

This study determined the effects experienced by the In-Pile Tube, of the Advanced Test Reactor’s center flux trap, due to various loading configurations of the H and Inner A positions. The effects of concern included: the axial profile of the linear heat generation rate of each fuel rodlet, the axial flux shape, and spectrum, in the fuel, and cladding, of each rodlet, and the total fission heating rate of the entire In-Pile Tube. Additionally, the reactivity worth of each configuration, relative to an ATR configuration with water filling all concerning test positions, was determined. The test positions were filled with commonly used test specimens, backup test specimens, and homogeneous substances. The study was performed using the modeling software MCNP6.2.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Status Summary of FY16 Atom Probe Tomography Studies on UCSB ATR-2 Irradiated RPV Steels

The University of California Santa Barbara-2 RPV Steel Irradiation experiment was awarded in 2010 by the Nuclear Science User Facility (formerly ATR NSUF) through a competitive peer review proposal process. The experiment involved irradiation of nearly 1300 samples distributed over 13 capsules. The major objective of this experiment was to better understand embrittlement behavior of reactor pressure steels at doses beyond which available data exists yet may be achieved if reactor operating licenses are extended beyond 60 years. The experiment was instrumented during irradiation and active temperature control was used to maintain the temperature at the design temperature. Six samples were selected from a large matrix of materials to perform atom probe tomography (APT) to look at formation of high dose phases. The nature and formation behavior of these phases is discussed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

TEM Characterization of High Burn-up Microstructure of U-7Mo Alloy

As an essential part of global nuclear non-proliferation effort, the RERTR program is developing low enriched U-Mo fuels (< 20% U-235) for use in research and test reactors that currently employ highly enriched uranium fuels. One type of fuel being developed is a dispersion fuel plate comprised of U-7Mo particles dispersed in Al alloy matrix. Recent TEM characterizations of the ATR irradiated U-7Mo dispersion fuel plates include the samples with a local fission densities of 4.5, 5.2, 5.6 and 6.3 E+21 fissions/cm3 and irradiation temperatures of 101-136?C. The development of the irradiated microstructure of the U-7Mo fuel particles consists of fission gas bubble superlattice, large gas bubbles, solid fission product precipitates and their association to the large gas bubbles, grain subdivision to tens or hundreds of nanometer size, collapse of bubble superlattice, and amorphisation. This presentation will describe the observed microstructures specifically focusing on the U-7Mo fuel particles. The impact of the observed microstructure on the fuel performance and the comparison of the relevant features with that of the high burn-up UO2 fuels will be discussed.

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02-21 BWXT

7-year DOE cost-shared ARDP with $24.3M of INL scope Years 1-3: test train design Years 4-5: irradiation Years 5-7: post-irradiation examination (PIE) Irradiation of uranium nitride (UN) tri-structural isotropic (TRISO) prismatic fuel in ATR PIE will be conducted at MFC and ORNL NRIC overseeing program and ensuring continuity between disciplines/steps

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Testing in Support of Fission Surface Power System Qualification

The strategy for qualifying a FSP system could have a significant programmatic impact. The US has not qualified a space fission power system since launch of the SNAP-10A in 1965. This paper explores cost-effective options for obtaining data that would be needed for flight qualification of a fission system. Qualification data could be obtained from both nuclear and non-nuclear testing. The ability to perform highly realistic nonnuclear testing has advanced significantly throughout the past four decades. Instrumented thermal simulators were developed during the 1970s and 1980s to assist in the development, operation, and assessment of terrestrial fission systems. Instrumented thermal simulators optimized for assisting in the development, operation, and assessment of modern FSP systems have been under development (and utilized) since 1998. These thermal simulators enable heat from fission to be closely mimicked (axial power profile, radial power profile, temperature, heat flux, etc.) and extensive data to be taken from the core region. For transient testing, pin power during a transient is calculated based on the reactivity feedback that would occur given measured values of test article temperature and/or dimensional changes. The reactivity feedback coefficients needed for the test are either calculated or measured using cold/warm zero-power criticals. In this way non-nuclear testing can be used to provide very realistic information related to nuclear operation. Non-nuclear testing can be used at all levels, including component, subsystem, and integrated system testing. FSP fuels and materials are typically chosen to ensure very high confidence in operation at design burnups, fluences, and temperatures. However, facilities exist (e.g. ATR, HFIR) for affordably performing in-pile fuel and materials irradiations, if such testing is desired. Ex-core materials and components (such as alternator materials, control drum drives, etc.) could be irradiated in university or DOE reactors to ensure adequate radiation resistance. Facilities also exist for performing warm and cold zero-power criticals.

Houts, Mike↗

Hafnium Absorber Latching Mechanism Redesign

The poster entails the known problems and updated design solutions for the hafnium absorber latching mechanism. The hafnium absorber latching mechanism is a part of the safety rod system located within ATR's core. The latching mechanism's primary function was to attach hafnium absorbers to the safety rod drive system.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

ON-GOING STATUS OF KJRR FUEL (U-7MO) QUALIFICATION

In order to cope with global shortage of Mo-99 supplies and with growing demand of neutron transmutation doping, KJRR construction plan has been launched since April 2012 to provide self-sufficiency of domestic RI demand, and to extend Si doping capacity for power device market growth. Through comprehensive surveillance of the fuels in-reactor behavior, KAERI has selected the fuel meat of U-7%Mo dispersion in an aluminum matrix with 5wt%Si for the KJRR fuel. As part of the efforts for fuel licensing and qualification of the KJRR fuel, an LTA irradiation test at the ATR started from November 2015 was successfully completed by reaching at 219 EFPD in the end of February 2017. Together with the results of HAMP-1 already completed irradiation and PIE, the successful irradiation of the LTA also demonstrates the fuel integrity under more rigorous conditions than the KJRR operation conditions. This paper updates the current status of the KJRR U7Mo (8 g-U/cm3) LTA irradiation and PIE plan up to date as of February 2017.

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