Results for “reactor experiments” — Page 37

DOE OSTI2025

Accident tolerant fuel test 2c (ATF-2C) irradiation test report

A fully prototypic testing platform for testing new accident tolerant fuel (ATF) designs for light water reactors has been established in the center flux trap of Idaho National Laboratory’s advanced test reactor (ATR). The irradiation experiment named ATF-2C has completed its four cycles of prototypic steady state irradiation. Irradiation conditions have been maintained via the loop 2a pressurized water coolant loop and have been monitored through in-situ instrumentation during each cycle. Power and burnup histories of each of the 18 fueled test pins has been calculated.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Influence of Cs Loading on Pt/m-ZrO 2 Water–Gas Shift Catalysts

Certain alkali metals (Na, K) at targeted loadings have been shown in recent decades to significantly promote the LT-WGS reaction. This occurs at alkali doping levels where a redshift in the C-H band of formate occurs, indicating electronic weakening of the bond. The C-H bond breaking of formate is the proposed rate-limiting step of the formate associative mechanism, lending support to the occurrence of this mechanism in H2-rich environments of the LT-WGS stage of fuel processors. Continuing in this vein of research, 2%Pt/m-ZrO 2 was promoted with various levels of Cs in order to explore its influence on the rate of formate intermediate decomposition, as well as that of LT-WGS in a fixed bed reactor. In situ DRIFTS experiments revealed that Cs promoter loadings of 3.87% to 7.22% resulted in significant acceleration of the forward formate decomposition in steam at 130 °C. Of all of the alkali metals tested to date, the redshift in the formate ν(CH) band with the incorporation of Cs was the greatest. XANES difference experiments at the Pt L 2 and L 3 edges indicated that the electronic effect was not likely due to an enrichment of electronic density on Pt. CO 2 TPD experiments revealed that, unlike Na and K promoters, Cs behaves more like Rb in that the decomposition of the second intermediate in LT-WGS, carbonate species, is hindered due to (1) increased basicity of Cs, (2) the tendency of Cs to cover Pt sites that facilitate CO 2 decomposition, and (3) the tendency of Cs to increase Pt particle size as shown by EXAFS results, resulting in fewer Pt sites that facilitate CO2 decomposition. As such, the LT-WGS rate was hindered overall and the rate-limiting step shifted to carbonate decomposition (CO 2 removal). Like its Rb counterpart, low levels of added Cs (e.g., 0.72%Cs) were found to improve the stability of the catalyst relative to the unpromoted catalyst; the stability comparison was made at similar CO conversion level as well as similar space velocity.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2020

AGR-5/6/7 Test Train Disassembly & Tools Phases I & II

This document provides the process for qualification of the AGR-5/6/7 Test Train Disassembly equipment for remote operation in the HFEF Main Cell. The test train is composed of five welded stainless steel capsules that are expected to be transported in two sections. The first section will be Capsules 1 and 2. The second section will include Capsules 3, 4 and 5. The capsules range in length from about 9 inches to about 12 inches. Inside the capsules is a graphite holder that contains multiple fueled AGR compacts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Initial Observations from AGR 5/6/7 Capsule 1

The fourth and final irradiation experiment in the Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) fuel development and qualification program is designated as AGR-5/6/7. Data collected from the fabrication, irradiation, and post-irradiation examination (PIE) of this tristructural isotropic (TRISO) fuel are intended to serve as the primary data set for the qualification of this fuel for use in high-temperature gas-cooled reactors (INL 2021, Collin 2018b). However, data collected from the three preceding irradiations (i.e., AGR-1, AGR-2, and AGR-3/4) may also be used to supplement data collected from AGR-5/6/7. All components of the AGR-5/6/7 fuel (i.e., UCO kernels, TRISO coatings, and fuel compacts) were produced on an engineering scale at BWXT (Lynchburg, Virginia USA) according to the fuel specification (Marshall 2017). This fuel was irradiated in the northeast flux trap (NEFT) at the Advanced Test Reactor (ATR) at Idaho national Laboratory (INL) from February 16, 2018 to July 22, 2020 (Pham et al. 2021). Measurements in the fission product monitoring system (FPMS) indicated unexpected and significant numbers of failures of TRISO particles in Capsule 1 near the end of the sixth irradiation cycle (ATR Cycle 166A). In the fourth cycle (ATR Cycle 164B) and beyond, the sweep gas flow became very low (presumably from degradation of the capsule gas outlet line via an unidentified mechanism), and the program deliberately isolated Capsule 1 from gas flow periodically. In later cycles, attempts to reestablish any kind of flow in Capsule 1 were unsuccessful. With little or no flow through Capsule 1, FPMS measurements and enumerations of failed particles in Capsule 1 were difficult or impossible as was the ability to control the helium/neon gas mixture used for temperature control. Gas flows and fission gas activity in the effluent gas from the other AGR-5/6/7 capsules were also impacted by the Capsule 1 gas flow issues and the large increase in fission gas released from the Capsule 1.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Applications of Similarity Analysis of Reactivity-initiated Accident Experiments in TREAT

Experimental testing capabilities have been created to perform reactivity-initiated accident experiments in the Transient Reactor Test Facility for fuel safety testing of Accident Tolerant Fuel concepts and to extend the burnup limits for our current light water reactor fuel designs. Completely prototypic test conditions are not possible and compromises naturally have to be made. To date, the representativity of the small scale safety testing to a similar event in a commercial reactor is made by comparing a few specific phenomena of interest. The goal of this work is to apply similarity analysis as another metric to judge the representativity of in-pile safety tests to full-scaled accident scenario.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Cracking in first ramp to power of standard versus high thermal conductivity UO2 pellets via internal nuclear heating

Advanced fuel designs that incorporate thinner fuel UO2 pellets interspaced by high thermal conductivity inserts have been proposed, with the primary goals of reducing peak centerline temperatures and temperature gradients across fuel pellets and enhancing heat transfer from the fuel to the coolant. An initial series of experiments has been performed on this design, including laboratory experiments and a series of experiments using the Idaho National Laboratory (INL) Transient Reactor Test (TREAT) Facility, the latter of which compared thermal gradient driven fracture of standard pellet designs with that in the proposed advanced fuel design. Although reducing fracture is not the primary objective of the new fuel design, the lower thermal gradients are expected to reduce fracture, so it can serve as an indicator of the thermal behavior of this fuel in the reactor. The in-reactor tests were conducted at multiple linear heat generation rates and confirm the expected result that fracture in both the standard and advanced fuel pellets occurs during the first ramp to power in standard light-water reactor conditions. Post-irradiation examination of the experiment material was performed and included quantification of the extent of fracture in the fuel pellets. It was found that the advanced-design pellets reduce the extent of fracture in a statistically significant way. This confirms the expected behavior predicted by two-dimensional axisymmetric models of this experiment. This study is an important first experimental confirmation of the efficacy of the proposed inserts for achieving their desired effect on the thermal behavior of the fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

BISON Simulated and Experimental Fission Product Release Comparisons from Reradiated AGR-3/4 Compacts During High Temperature Heating Tests

The fuel performance modeling code BISON was used to predict the release of fission products iodine-131 (131I), xenon-133 (133Xe), and krypton-85 (85Kr) from four re-irradiated AGR-3/4 fuel compacts containing tristructural isotropic (TRISO) coated particles during high-temperature isothermal heating tests. The AGR-3/4 fuel compacts were irradiated in the Advanced Test Reactor (ATR) as part of the third and fourth series of planned experiments to support the Advanced Gas Reactor (AGR) Program. They were subsequently stored and re-irradiated in the Neutron Radiography (NRAD) reactor for approximately five days and then stored for another five to eight days before being subjected to isothermal heating tests in the Fuel Accident Condition Simulation (FACS furnace) for 200 to 300 hours at temperatures between 1000°C and 1600°C to evaluate fission product release at elevated temperatures. New nuclide-specific fission product source term models for the three nuclides of interest were developed using the reactor multiphysics code Griffin and implemented into BISON to support this work. The new source term models were incorporated into coupled compact- and particle-scale BISON simulations, which predict spatially- and temporally-resolved radionuclide generation, radioactive decay, transport, and release throughout the entire irradiation history, including the initial ATR irradiation, NRAD re-irradiations, FACS heating tests, and intermediate periods spent in storage. The experimentally measured fission product release from the heating tests were compared to modeling release predictions calculated by BISON to evaluate how well the code compares to experimental results. Overall, the experimental measured and BISON predicted comparative release results varied but generally agreed to within 5 particle equivalents. Comparative release results identified general observations to take into consideration to help refine future models and reduce uncertainties associated with both the measurement results and predictive results. This includes developing new uranium oxycarbide (UCO) specific kernel diffusivities for the three isotopes examined to more accurately reflect the material properties of the fuel form. Deriving new diffusivities will aid in producing a more informed BISON model

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Mapping the Microstructure During the Historic U-Mo Monolithic Fuel Foil Fabrication Process

Since 2004, there has been extensive effort towards the development of a uranium molybdenum monolithic fuel system to convert high performance research and test reactors. The RERTR-6 experiment was the first to attempt a monolithic fuel instead of a dispersed fuel form. The fabrication methods evolved overtime and provided the basis for current fabrication methods. The various steps in the process inevitably tailor the fuel alloy microstructure which is known to influence irradiation behavior. This document aims to present and discuss the fabrication evolution that transpired through a recounting of historical data from the various fabrication campaigns. By overlaying this data with basic science studies on the U-Mo that explored transformation kinetics, it is possible to estimate a measure of the impact heat treatments have on the final as-fabricated microstructure.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Post-irradiation 1600°C Heating Test of AGR-1 UCO Fuel Kernels

Five bare kernels were selected from the U.S. Advanced Gas Reactor (AGR) 1 irradiation experiment Compact 5-3-1 to perform post-irradiation safety tests. The safety test involved heating the kernels in the Fuel Accident Condition Simulator (FACS) furnace in the inert atmosphere to a peak temperature of 1600°C and isothermally holding at this temperature for about 47 hours while collecting fission products released. This test was to assess the retention of fission products in bare kernels without the effects of the other TRISO layers (buffer, IPyC, SiC, and OPyC) or the graphitic matrix material. The bare kernels released nearly 100% of cesium and antimony, while they were able to maintain about 30% of europium, 50% of strontium, and the majority of cerium and ruthenium. In addition, about 48% of the calculated inventory of Kr-85 released during the test, indicating that kernels were capable of retaining a considerable fraction of fission gas K-85 during irradiation.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

MARVEL Utilization Plan

This report provides a high-level overview of the operational plan for the MARVEL microreactor. The main focus is on discussing testing and application demonstration opportunities that leverage the reactor. With the reactor rapidly progressing towards demonstration, along with the short operational window (2 years), it was deemed critical to start establishing a basis for how stakeholders can engage with the program and leverage the reactor as a testbed. The report discusses potential opportunities for using MARVEL leveraging from data/design access, novel controls testing, novel nuclear-electric applications, to novel nuclear-heat applications. It also provides guidance for interested stakeholders on potential funding opportunities that can be pursued to support tests during the operational lifetime of the reactor. The report also discusses the recommended strategy for outreach and the organization structure to review and down select project. It is important to emphasize that this is intended to be a living document with revisions issued as the MARVEL demonstration timeline evolves. Similarly, the framework for engagement is subject to change as the project nears construction. As such, this first revision to the original test plan highlights the new Microreactor Program-sponsored pathway for engaging with stakeholders via an Expression of Interest and Notice of Opportunity.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

Road to PROSPECT-II

The Precision Reactor Oscillation and SPECTrum (PROSPECT) experiment is based in a segmented liquid scintillator antineutrino detector situated approximately 7 meters from the highly enriched High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. Its main goal is to investigate short-baseline antineutrino oscillations. The first phase of data collection, known as PROSPECT-I, was held from 2018 to 2019 and was used for several high-precision analyses, including multiple measurements of the $^{235}U$ antineutrino spectrum and searches for $eV$-scale sterile antineutrino oscillations. The collaboration is now preparing for the second phase, PROSPECT-II, which features an upgraded detector design. This advancement will enhance sensitivity and statistical power, allowing for a broader range of analyses beyond those achieved in PROSPECT-I. As we transition into this new phase, new questions have arisen concerning background simulation and its potential differences from those conducted during the initial phase of the experiment. Moreover, it is essential to ascertain, through simulations, the positive effects that an improved detector could have on the study of oscillations. This information is crucial for justifying the proposed enhancements, and I will cover all of this in the talk.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗

DOE OSTI2024

Road to PROSPECT-II

The Precision Reactor Oscillation and SPECTrum (PROSPECT) experiment is based in a segmented liquid scintillator antineutrino detector situated approximately 7 meters from the highly enriched High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. Its main goal is to investigate short-baseline antineutrino oscillations. The first phase of data collection, known as PROSPECT-I, was held from 2018 to 2019 and was used for several high-precision analyses, including multiple measurements of the $^{235}U$ antineutrino spectrum and searches for $eV$-scale sterile antineutrino oscillations. The collaboration is now preparing for the second phase, PROSPECT-II, which features an upgraded detector design. This advancement will enhance sensitivity and statistical power, allowing for a broader range of analyses beyond those achieved in PROSPECT-I. As we transition into this new phase, new questions have arisen concerning background simulation and its potential differences from those conducted during the initial phase of the experiment. Moreover, it is essential to ascertain, through simulations, the positive effects that an improved detector could have on the study of oscillations. This information is crucial for justifying the proposed enhancements, and I will cover all of this in the talk.

43 PARTICLE ACCELERATORS↗

DOE OSTI2020

Characterization of a minor actinides bearing metallic fuel pin irradiated in EBR-II

The X501 transmutation experiment irradiated in Experimental Breeder Reactor II (EBR-II) was intended to evaluate the safety and performance of the addition of minor actinides (Am, Np) into the metallic nuclear fuel system. The irradiation of this experiment is also of paramount importance for comparisons between true fast spectrum irradiations (e.g. EBR-II) and shrouded irradiation capsules performed at the thermal neutron spectrum Idaho National Laboratory Advanced Test Reactor. The X501 experimental fuel pins were irradiated to a peak measured burnup of 6.2% fission per initial heavy metal atom. This paper reports and discusses Postirradiation Examination results for one of two minor actinide bearing pins from the EBR-II X-501 experiment (X501-G591) with a focus on fuel microstructure evolution. Additionally, fuel swelling, fission product distribution, cladding strain, fission gas release, fuel microstructure and fuel cladding chemical interaction were all evaluated. The fission gas release was 79.9% and the microstructure along the fuel pin presented several characteristics very similar to other EBR-II U-Pu-Zr ternary metallic fuel experience. Minor actinides seem to not dramatically affect the overall performance of this candidate transmutation fuel. Scanning Electron Microscopy was also performed and an in-depth characterization of the different regions and phases is presented. Performance data from these irradiations can be used to inform the feasibility of minor actinide transmutation in future reactor systems.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Thermal-Epithermal eXperiments – Hafnium Experiment Performed at the DOE’s National Criticality Experiments Research Center (NCERC)

The Thermal-Epithermal eXperiments (TEX) – Hafnium (TEX-Hf) experiments were completed in October 2022 at the DOE National Criticality Experiments Research Center (NCERC) at the Nevada National Security Site (NNSS). This measurement campaign provides seven unique configurations that can be used to validate the neutron absorption and scattering cross sections of hafnium in a highly enriched uranium (HEU) system. Hafnium is a strong neutron absorber important for marine propulsion systems. Mike Zerkle of the Naval Nuclear Laboratory (NNL) states, “The experiments were requested by Naval Reactors to provide a suite of unclassified, clean, and well-characterized benchmarks to drive improvements to hafnium nuclear data.” The experiments were a collaboration between Los Alamos National Laboratory (LANL), Lawrence Livermore National Laboratory (LLNL), and NNL, funded by the Nuclear Criticality Safety Program (NCSP). The Hf was procured directly by NNL and provided for the experiment.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2022

Utilizing Sensitivity and Correlation Coefficients from MCNP and Whisper to Guide Microreactor Experiment Design

When designing experiments for full-scale reactor systems, MCNP®* and Whisper can be used to create neutronic models and compare the similarity of two nuclear systems via correlation coefficients for κ eff , effective multiplication factor. This thesis applies this framework to a conceptual heat-pipe, yttrium-hydride moderated microreactor system and experiments. The framework is intended as a supplement to other neutronics/thermal/multiphysics analyses and provides a concrete method to measure the neutronic similarity of two systems. By analyzing the shared nuclear data uncertainty, as well as sensitivity to nuclear data over all neutron energies, highly informative experiments can be designed to aid in the development of microreactor and other advanced reactor technologies and systems.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2022

Advanced Moderator Material Handbook (FY22 Rev. 2)

High hydrogen density moderators such as metal hydrides are an important research topic within the DOE NE Microreactor Research, Development, and Deployment (RD&D) Program due to their ability to retain hydrogen to much higher temperatures than other hydrogenous media. This class of moderators, which includes yttrium dihydride (YH 2 ), thermalizes neutrons in the system such that the overall fuel mass or the required uranium enrichment in the system can be significantly reduced. Knowledge of material properties, both in the as-fabricated and irradiated state, are important to understanding moderator performance during steady-state and transient reactor operation. This document provides a detailed summary of the literature data on yttrium dihydride, thermomechanical and other property data, and a critical evaluation of that data. This handbook also provides a description of ongoing experiments to understand in-reactor performance, such as irradiations in ATR, as well as nuclear data from an integral critical experiment at NCERC. This report focuses on measured quantities but also includes some modeling results for comparison where applicable.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2024

FUEL PERFORMANCE SIMULATION OF HIGH BURNUP FUELS IN PLANNED INTEGRAL DESIGN BASIS ACCIDENT EXPERIMENTS

High burnup (HBu) fuel rods from the Byron Nuclear Generating Station (BNGS) were recently received at Idaho National Laboratory (INL) to support a variety of planned Nuclear Energy fuel cycle R&D objectives ranging from fuel performance, fuel recycle, and spent fuel research topics. Among these R&D activities, these fuel rods will be the subjects of multiple in-pile experiment programs at the Transient Reactor Test (TREAT) facility as well as detailed characterization and testing in the hot cells at INL and Oak Ridge National Laboratory (ORNL). TREAT RIA experiments are planned for the Nuclear Energy Agency Framework for Irradiation Experiments (FIDES) Joint Experimental Program called High burnup Experiments in Reactivity Initiated Accident (HERA) program. TREAT and ORNL-furnace LOCA experiments are part of the Department of Energy (DOE) Advanced Fuels Campaign (AFC) program U.S. consensus LOCA test plan, and the in-pile experiments have also been proposed in a FIDES project called Loss of Coolant-High Burnup (LOC-HBu). The results of these test programs will provide crucial data about safety performance enabling extended licensable burnup limits for these fuels. The purpose of this paper is to document fuel performance computational simulations of the BNGS fuel using the Bison code. The detailed assessments include (1) the irradiation history of the fuel to provide prediction of as-run fuel conditions and (2) extending the irradiated fuel conditions into the as-designed experiment conditions for the HERA-HBu RIA experiments and for the LOC-HBu LOCA experiments. The results of these assessments will inform post-irradiation examinations (PIE) of the BNGS parent rods and detailed final design of the planned experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

FUEL PERFORMANCE SIMULATION OF HIGH BURNUP FUELS IN PLANNED INTEGRAL DESIGN BASIS ACCIDENT EXPERIMENTS

High burnup (HBu) fuel rods from the Byron Nuclear Generating Station (BNGS) were recently received at Idaho National Laboratory (INL) to support a variety of planned Nuclear Energy fuel cycle R&D objectives ranging from fuel performance, fuel recycle, and spent fuel research topics. Among these R&D activities, these fuel rods will be the subjects of multiple in-pile experiment programs at the Transient Reactor Test (TREAT) facility as well as detailed characterization and testing in the hot cells at INL and Oak Ridge National Laboratory (ORNL). TREAT RIA experiments are planned for the Nuclear Energy Agency Framework for Irradiation Experiments (FIDES) Joint Experimental Program called High burnup Experiments in Reactivity Initiated Accident (HERA) program. TREAT and ORNL-furnace LOCA experiments are part of the Department of Energy (DOE) Advanced Fuels Campaign (AFC) program U.S. consensus LOCA test plan, and the in-pile experiments have also been proposed in a FIDES project called Loss of Coolant-High Burnup (LOC-HBu). The results of these test programs will provide crucial data about safety performance enabling extended licensable burnup limits for these fuels. The purpose of this paper is to document fuel performance computational simulations of the BNGS fuel using the Bison code. The detailed assessments include (1) the irradiation history of the fuel to provide prediction of as-run fuel conditions and (2) extending the irradiated fuel conditions into the as-designed experiment conditions for the HERA-HBu RIA experiments and for the LOC-HBu LOCA experiments. The results of these assessments will inform post-irradiation examinations (PIE) of the BNGS parent rods and detailed final design of the planned experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗