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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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Application of research reactor environments for validation of the IRDFF-II dosimetry cross section library
Activation data from seven different reactor-based reference neutron fields are examined to provide enhanced validation evidence for the newly released IRDFF-II library. A least-squares based spectrum adjustment methodology is used and rigorous statistical metrics demonstrate the consistency of the set of IRDFF-II dosimetry cross sections. The use of these reactor-based neutron fields provides validation evidence for nine more reactions than have been addressed in the 252 Cf(sf) and 235 U(th) benchmark validation testing. The use of covers to shift their energy response increases the power of the response validation but makes it challenging to properly capture response correlations.
Miniature Sensor Irradiation Tests Under Steady-State and Transient Conditions at MIT Research Reactor (MITR) and Transient Reactor Test Facility (TREAT)
Not provided.
Parametric Analysis of an Optical Fiber–Based Gamma Thermometer for University Research Reactors Using an Analytic Thermal Model
Not Available
University of Missouri Research Reactor (MURR) Design Demonstration Element End Fitting Structural Rigidity Analysis
The primary objective of this work is to assess the extent to which the stiffness (measured by means of maximum displacement) of the end fittings in the DDE contributes to the stiffness of the entire element, and how it compares to the equivalent stiffness of the end fittings in the LEU element. Structural analysis of both the LEU element and the DDE were performed using COMSOL 5.3a finite element software. Supporting combs are used on the leading and trailing edges of fuel plates for both the LEU element and the DDE. Therefore, simulations with and without combs are performed as two bounding boundary conditions on the leading edge of the fuel plates. Three types of loads are analyzed in this work: the hydraulic load due to the channel flow disparity-induced pressure differential, the thermal load due to the thermal expansion of the fuel plates, and a point load equal in magnitude to the LEU element’s weight.
Final Design for Thermal/Epithermal eXperiments using High 240 Pu Content Plutonium/Aluminum Zero Power Research Reactor Plates with Polyethylene Moderator (IER 520 Final Design CED-2 Report)
The US Department of Energy Nuclear Criticality Safety Program (NCSP) convened a multinational Thermal Epithermal eXperiments (TEX) meeting in July of 2011 to discuss the data and experimental needs of criticality safety practitioners. The number one and two priority integral experiment data needs were for 239 Pu and 240 Pu, with special emphasis on cross section performance in the intermediate energy range (from 0.625 eV to 100 keV). LLNL measured five critical configurations with LANL for the plutonium test bed (IER-184) and published the experiments as International Criticality Safety Benchmark Evaluation Project evaluation PUMET-MIXED-002. Modeling of the benchmark configurations using ENDF/B-VIII.0 nuclear data showed significant overprediction of reactivity for configurations that had a large percentage of fissions in the intermediate energy regime. This report documents a variation on the TEX plutonium test bed to provide a test of 240 Pu cross sections, with sensitivity of the configuration to 240 Pu radiative capture and fission cross sections a priority for the design.
Examining Sources of Uncertainty in Coincidence Neutron Measurements of Spent Research Reactor Fuel [Slides]
The nature of spent fuel introduces many sources of uncertainty into measurement results. Quantification of those uncertainties is essential for understanding the relationship between NDA measurements and fissile mass. Modeling is a useful tool for identifying those characteristics that contribute significantly to measurement uncertainties and understanding how to account for those uncertainties when performing coincident neutron measurements.
FY21 Progress of the Oak Ridge Health Physics Research Reactor CAAS Benchmark Evaluation [Slides]
The FY21 version of the evaluation, focusing on element 57 dose at 3 meters, was presented at the ICSBEP Technical Review Group meeting in October 2021 and was not accepted for 2022 publication in the handbook. The main issue was identified and replacement of the element 57 neutron dose benchmark metric by neutron fluence (closer to what was measured) is required. A subgroup was formed, confident for the updated neutron fluence evaluation to be accepted in the ICSBEP handbook. The updated evaluation will be presented again at the 2022 ICSBEP TRG for publication in the 2023 handbook.
Application of Research Reactor Environments for Validation of the IRDFF-II Dosimetry Cross Section Library
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Natura Resources Molten Salt Research Reactor (MSRR) Fuel Workshop: Summary Report
Idaho National Laboratory (INL) hosted the Natura Resources team in Idaho Falls for a 3-day fuel workshop, July 16-18, 2024. The purpose of this workshop was to understand Natura’s fuel requirements, determine the feasibility of producing that fuel at INL, and define the scope of work that would enable fueled operation of the Natura MSR-1 being deployed at Abilene Christian University (ACU).
Neutron Radiography of Element 300 at Sandia National Laboratory Annular Core Research Reactor, April 2024
This is a presentation on the work done in collaboration between NTESS and Oregon State University in April 2024 at the ACRR in TA-V. It will be presented by the OSU grad student, but contains Sandia information shared on an External Collaboration Network.
Neutronic Effects of Missing Elements in the Annular Core Research Reactor (ACRR)
Neutronic effects of ACRR operating with a missing element are analyzed using MCNP. Results are focused on assessing the peaking factors associated with a single missing element. Intra-element power gradients are also assessed. A missing tri-element is evaluated as unique case. Auxiliary studies are provided that analyze the effects of cavity buckets, rod positions, and reflector elements. Neutronic effects of a nearby FREC-II (coupled and decoupled/tilted) are presented for select scenarios.
Advanced Reactor Technologies: Very High Temperature Reactor Research and Development Quarterly Report
Major accomplishments for the months of January, February and March 2020 in the fields of Fuels Development Qualification, High Temperature Materials Development, Graphite and GRC Modeling and Validation.
Advanced Reactor Technologies: Very High Temperature Reactor Research and Development Quarterly Report April, May, June 2020
Report of activity for the Fiscal Year 2020 third quarter
LOW ENRICHED U-10MO FUEL PLATE CHARACTERIZATION AND REVIEW OF DE-CLADDING FOR RESEARCH REACTOR APPLICATIONS
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MONTE CARLO PERTURBATION ANALYSIS OF FUEL TEMPERATURE VARIATIONS IN THE MCNP MODEL OF THE ANNULAR CORE RESEARCH REACTOR.
Abstract not provided.
MONTE CARLO PERTURBATION ANALYSIS OF DIMENSION AND DENSITY VARIATIONS OF THE ANNULAR CORE RESEARCH REACTOR MODEL FUEL.
Abstract not provided.