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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 199 records · Page 11
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

At-power subcritical multiplication in the Advanced Test Reactor during nuclear requalification testing

Power division information during nuclear requalification of the Advanced Test Reactor (ATR) is of considerable interest as an importance function for observed changes to core reactivity. The degree to which a given physical subdivision of a critical reactor acts as a neutron source for other lobes is not analytically characterized for general application. When ATR operates at power, individual power-producing lobes rely on each other as neutron sources in order to maintain constant power, which in general requires either exactly critical multiplication within a reactor or an external neutron source. Here, this work shows that fuel element and lobe powers in ATR can be related with subcritical multiplication theory. Subcritical multiplication factors are computed with a physically validated analytical method based on actual at-power operation, quantifying for each lobe its dependence on other lobes as an external neutron source. This explanation is significant for ATR due to the desire to irradiate a large variety of experiments simultaneously, each having its impact on the core neutron population. For any physical subdivision of any other critical reactor, it is likewise true that the subdivision undergoes only subcritical multiplication.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2026

Results of the AGR-5/6/7 UCO TRISO fuel irradiation test in the Advanced Test Reactor

AGR-5/6/7 was the last in a series of irradiation experiments sponsored by the U.S. Department of Energy in support of the development and qualification of TRISO coated particle fuel for use in a high-temperature gas-cooled reactor. This experiment was conducted to verify the performance of the reference-design TRISO-coated low-enriched UCO fuel for modular high-temperature gas-cooled reactor normal operating conditions and to explore fuel performance at temperatures substantially beyond those typical of normal operation. A total of 194 UCO fuel compacts in five capsules were irradiated in the Advanced Test Reactor for 360.9 effective full-power days, achieving final burnup ranging from 5.66% to 15.26% fissions per initial heavy metal atom and fast neutron fluence ranging from 1.62 × 10 25 to 5.55 × 10 25 n/m 2 (E > 0.18 MeV). Calculated time-averaged fuel temperatures ranged from 467 °C to 1432 °C, with a peak fuel temperature of 1536 °C. During the first five irradiation cycles (∼180 effective full-power days), 85m Kr fission gas release-rate-to-birth-rate ratios were 10 −7 –10 −6 , indicating no particle failures. Near the end of the sixth cycle, a substantial number of in-pile particle failures occurred in Capsule 1. The fission gas release from this capsule impacted the readings from the other four capsules and led to unreliable fission gas release measurements for all capsules during the last four cycles. A preliminary post-irradiation examination of Capsule 1 fuel and internal components revealed the in-pile particle failures resulted from operational issues with the capsule, not subpar performance of the fuel particles.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Phase II Development of the Surveillance Test Articles to Improve the Design, Fabrication, and Testing

Advanced reactors, such as the molten salt reactor (MSR), require materials that will withstand harsher environments than the materials used for lower temperature water reactors. Materials used in MSR construction need not only withstand elevated temperature, temperature cycling, and neutron radiation, but must be able to withstand the corrosive molten salt environment. Significant materials research has been driven by the materials needs of the MSR due to the harsh environment and the material data requirements to support licensing. Limited operation experience with MSRs has made this challenging work. Information is very limited on materials degradation due to irradiation, molten salt corrosion, elevated temperature, and the resulting fatigue, creep, and creep-fatigue loading during operation. While efforts are underway to better understand the effects of this harsh environment on construction materials, the Advanced Reactors Technologies (ART) Program has been working to develop materials surveillance test articles that could be used in a materials surveillance program and allow for the collection of information on the materials degradation during plant operation and could support timely licensing of these advanced reactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Thermal-Hydraulic Analysis of the AFIP-7 Irradiation Test in the Advanced Test Reactor - Oxide Growth Prediction and Correlation

Knowing the thickness of the oxide layer on the surface of aluminum fuel cladding is vitally important for predicting fuel temperature due to the low thermal conductivity of the oxide layer. Several correlation models for predicting oxide growth can be found in the literature. In previous research, the correlations were combined with heat transfer simulations in Abaqus, a finite element analysis code, to forecast the oxide growth. However, this approach requires heat transfer coefficients for modeling heat exchanges with the external flow field, and such coefficients were obtained through empirical equations. Since different empirical equations yield varying heat transfer coefficients, the cladding temperature and predicted oxide thickness both carry a high degree of uncertainty. Here, this research develops a new approach that integrates the fluid flow, fluid and solid heat transfer, and oxide growth correlation(s) into a single computational fluid dynamics model. We demonstrate this approach’s ability to predict oxide development on the AFIP-7 plates during two Advanced Test Reactor (ATR) irradiation cycles. The projected oxide thickness falls within the experimental measurements taken during post irradiation examination.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Testing effects of invasive fire ants and disturbance on ant communities of the longleaf pine ecosystem

Invasive species and habitat disturbance are among the most important drivers of biodiversity loss and ecological change. Their individual effects, however, are difficult to disentangle because invasion and disturbance are often intimately linked. Here we test alternative hypotheses to determine if the invasive red imported fire ant, Solenopsis invicta, is a ‘driver’ or simply a ‘passenger’ of ecological change in a longleaf pine ecosystem. We randomly assigned treatments of (1) unmanipulated, (2) soil disturbance, (3) fire ant removal and (4) soil disturbance + fire ant removal to experimental blocks and measured how ant communities changed over 2 years in thirty-six 15-m 2 plots. Additionally, fire ant abundance in removal plots averaged 42% lower in pitfall traps and 95% lower on baits compared to unmanipulated, control plots. Species richness of co-occurring ants also decreased 42% in removal plots, with significant changes in community composition. Soil disturbance alone did not affect ant communities. Fire ant diet breadth—measured using carbon and nitrogen stable isotopes—increased up to 4.7-fold in soil disturbance + removal plots (i.e. 0.84‰ 2 to 3.94‰ 2 ). While non-target impacts of the fire ant removal treatment complicate interpretation, our results suggest fire ants follow an alternative ‘interacting drivers’ model in which partial recovery of some species occurs when populations of an invasive species are reduced. Further recovery of native ants may be limited by persisting, landscape-level effects of fire ants suppressing co-occurring ants below historical levels.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2021

Metal Bellows Valve Reliability Testing - Copper Stem Tip Testing

SRNL was funded in Mid-Year FY20 by NNSA NA-231 to continue evaluation of alternate valves for use in tritium service to support domestic Mo-99 production. The focus of the effort was to identify valve cycle life as a function of actuator size and stem tip material. Using the minimum size actuator to reliability open and close valves can reduce glovebox size and thus support domestic companies to “come to market” faster in supplying Mo-99 to the US market. This report serves as a continuation to the FY19 report and summarizes the task activities completed in FY20 after authorization to start work was obtained on May 5 th , 2020. Copper stem tips were tested on the Swagelok 1C and 5C actuated metal bellows valves. With ambitions to cycle each set 150,000 times, both 1C and 5C valves were met with high failure rates. The smaller 1C actuated valves required additional closing pressure to form a seal with the Cu stem tips installed however, excessive stem tip deformation may be the root cause of the majority of the valves failing before 500 cycles. The larger 5C actuated valves were cycled 150,000 times but still resulted in 80% failure rate, suspected of metal fatigue in the bellows due to high cycling frequencies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Annual Performance Testing of Tracer Gas Detectors for use in Radionuclide NESHAP Compliance Testing

Los Alamos National Laboratory’s (LANL’s) compliance with Radionuclide NESHAP0F 1 regulations is managed by the Radioactive Air Emissions Management (RAEM) team, part of LANL’s Compliance Programs group (EPC-CP). One area of the Radionuclide NESHAP addresses requirements for siting a stack sample system. Prior to commissioning a new stack sampling system, the ANSI Standard for stack sampling requires that the stack sample location must meet several criteria, including uniform mixing of tracer gas (sulfur hexafluoride, SF 6 ) and tracer aerosol (liquid oil droplets) in the air stream. This memorandum will specifically focus on the tracer gas portion of the ANSI Standard. For these mix tests, tracer gas is injected into the stack air stream and the resulting air concentrations are measured across the plane of the stack at the proposed sampling location. The coefficient of variation of the media concentration must be under 20% when evaluated over the central 2/3 area of the stack or duct.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2025

FY25 Report on Water NSTF Testing: Parametric and Accident Testing with Lower Tank Inlet

The Natural Convection Shutdown Heat Removal Test Facility (NSTF) at Argonne National Laboratory has continued to generate empirical validation data on the performance of water-based reactor cavity cooling system (RCCS) for seven years. This data is actively being used to support the development of passive decay heat removal systems for advanced reactors. Distinguishing this facility are 1) the large, ½ scale of the facility and 2) its governance under an NQA-1 qualified program for producing data of the highest pedigree for advanced reactor designers and regulators. In addition to the experimental activities discussed in this report, a computational modeling program continues to support the experimental program and further accuracy and understanding of the computational models. Together, the experimental and computational work create a mutually beneficial relationship integral to the overall program objective of advancing the understanding of the RCCS technology.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Small Scale Melter Testing of HLW Algorithm Glasses: Matrix 1 Tests

The principal objectives of the DM100 tests were to determine the effects of glass property extremes (glass viscosity, electrical conductivity, and T1%), minimum and maximum concentrations of several major and minor glass components, +/-15% batching errors in the addition of Glass Forming Chemicals (GFCs) to the feed, and variation in the sources of boron and sodium used in the GFCs on processing rates and melter conditions (off-gas characteristics, glass processing, foaming, cold cap, etc.) at constant bubbling rate.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

Advanced Test Reactor Neutron Dosimetry Report - Cycle 175C: Results of Neutron Dosimetry Measurements for the Advanced Test Reactor Cycle 175C

Results of neutron dosimetry measurements in ATR for cycle 175C are presented. A single set of cobalt and nickel wires were installed in numerous irradiation locations for the duration of the cycle. The specific activity of these wire segments can be used in conjunction with the presented cycle history information to determine the thermal- and fast-neutron fluence rates for each irradiation position during the cycle using the referenced standard test methods.

73 - NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Identifying High Burnup LOCA Test Conditions to Support LOCA Testing

Presentation summarizing work done by INL and ORNL investigating loss-of-coolant accidents involving high burnup fuel. This includes developing LOCA test conditions for TREAT and SATS experiments and highlighting potential differences between nuclear and electrical heating LOCA experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Experiment Assembly and Testing in Hot Cell in Support of Advanced Fuels Irradiation Testing

Follow-on experimentation of irradiated nuclear fuel is critical to understanding fuel behavior at different stages of the fuel lifecycle. Transient experiments using fuel with high burnup can enlarge understanding of fuel fragmentation, relocation and dispersal under various reactor accident situations. Assembling experiments with such fuel requires remote operations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗