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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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All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Accelerated Irradiation Testing and Post-Irradiation Characterization: U.S.-Based Capabilities for Advanced Nuclear Systems and Radioisotope Production

Irradiation experiments and post-irradiation examinations, together referred to as irradiation testing (IRT), are prerequisites for nuclear fuel and material qualification for the deployment of new and advanced reactors, as well as radioisotope production, thereby ensuring regulatory compliance. Qualified research and test reactors (RTRs) and testing facilities are essential to enable IRT to verify performance and safety under prototypical reactor conditions. In the past, qualification of new fuels or structural materials required about 20 years. Synergist strategies, advanced tools, and qualified methods are needed to greatly reduce this timeframe of IRT and radioisotope production. This study, termed accelerated-IRT, focuses on identifying gaps and leveraging U.S.-based RTRs and material testing capabilities, leveraging the preliminary evaluation and qualification of selected RTRs to provide a generic as well as specific-case solution paths forward, ensuring adherence to stringent regulatory standards. Furthermore, IRT and radioisotope production utilizing qualified RTRs necessarily includes modeling and simulation to support the design (i.e. neutronics, thermal, and structural aspects) and manufacturing of irradiation test specimens, vehicles, capsules, apparatuses, and flow loops. In addition, IRT can be improved by applying advanced manufacturing techniques and in-pile sensors and instrumentation, as discussed in this study.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Initial results from safety testing of US AGR-2 irradiation test fuel

Two cylindrical compacts containing tristructural isotropic (TRISO)-coated particles with kernels that contained a mixture of uranium carbide and uranium oxide (UCO) and two compacts with UO 2 -kernel TRISO particles have undergone 1600°C safety testing. These compacts were irradiated in the US Advanced Gas Reactor Fuel Development and Qualification Program's second irradiation test (AGR-2). The time-dependent releases of several radioisotopes ( 110m Ag, 134 Cs, 137 Cs, 154 Eu, 155 Eu, 90 Sr, and 85 Kr) were monitored while heating the fuel specimens to 1600°C in flowing helium for 300 h. The UCO compacts behaved similarly to previously reported 1600°C-safety-tested UCO compacts from the AGR-1 irradiation. No failed TRISO or failed SiC were detected (based on krypton and cesium release), and cesium release through intact SiC was very low. Release behavior of silver, europium, and strontium appeared to be dominated by inventory originally released through intact coating layers during irradiation but retained in the compact matrix until it was released during safety testing. Both UO 2 compacts exhibited cesium release from multiple particles whose SiC failed during the safety test. Europium and strontium release from these two UO 2 compacts appeared to be dominated by release from the particles with failed SiC. Silver release was characteristically like the release from the UCO compacts in that an initial release of the majority of silver trapped in the matrix occurred during ramping to 1600°C. However, additional silver release was observed later in the safety testing due to the UO 2 TRISO with failed SiC. Failure of the SiC layer in the UO 2 fuel appears to have been dominated by CO corrosion, as opposed to the palladium degradation observed in AGR-1 UCO fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Test Plan for DOE-GENIORS Dyanamic Irradiation Tests

Dynamic irradiation tests will be performed within the DOE-GENIORS collaboration, to be finished by the end of March 2021. These tests will be performed using three different setups, one at INL, one at CEA, and one at CIEMAT. The cis-mTDDGA/PTD chemical system will be used, applying both "extraction" (cis-mTDDGA - n-dodecane/HNO3) and "stripping" (cis-mTDDGA - n-dodecane/PTD - HNO3) conditions. This test plan provides the details and parameters to be used to perform irradiation at the INL using the INL Radiolysis Test Loop.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2023

Accident Tolerant Fuel Test (ATF-2c) Irradiation Test Report

A new 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 first cycle 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 fast neutron flux histories of each of the 24 test pins has been calculated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Initial examination of fuel compacts and TRISO particles from the US AGR-2 irradiation test

Post-irradiation examination was completed on two as-irradiated compacts from the US Advanced Gas Reactor Fuel Development and Qualification Program’s second irradiation test. These compacts were selected for examination because there were indications that they may have contained particles that released cesium through a failed or defective SiC layer. The coated particles were recovered from these compacts by electrolytic deconsolidation of the surrounding graphitic matrix in nitric acid. The leach-burn-leach (LBL) process was used to dissolve and analyze exposed metallic elements (actinides and fission products), and each particle was individually surveyed for relative cesium retention with the Irradiated Microsphere Gamma Analyzer (IMGA). Data from IMGA and LBL examinations provided information on fission product release during irradiation and whether any specific particles had below-average retention that could be related to coating layer defects or radiation-induced degradation. A few selected normal-retention particles and six with abnormally-low cesium inventory were analyzed using X-ray tomography to produce three-dimensional images of the internal coating structure. Four of the low-cesium particles had obviously damaged or degraded SiC, and X-ray imaging was able to guide subsequent grinding and polishing to expose the regions of interest for analysis by optical and electron microscopy. Additional particles from each compact were also sectioned and examined to study the overall radiation-induced microstructural changes in the kernel and coating layers.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Accident Tolerant Fuel Test 2B (ATF-2B) 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-2 has completed 7 cycles of prototypic steady state irradiation irradiating 41 test pins to burnups as high as 30 MWd/kgU. 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 fast neutron flux histories of each of the 41 test pins has been calculated using a combined MCNP, and ORIGIN methodology to provide accurate and reliable power histories on a per pin basis.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
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 OSTI2017

Accident Tolerant Fuels Series 1 (ATF-1) Irradiation Testing FY 2017 Status Report

This report contains a summary of irradiation testing of FCRD ATF-1 experiments performed at the INL in FY 2017. ATF-1 irradiation testing work performed in FY 2017 included fabrication of one ATF-1B drop in capsule ATF-1 series experiment and irradiation testing of ATF-1 capsules in the ATR. Additionally, five capsules met the sponsor-specified burnup targets, three of which were shipped from the ATR to the HFEF at INL.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Accelerated Irradiation Testing of Miniature Nuclear Fuel and Cladding Specimens

Qualification of advanced nuclear fuels and materials requires irradiation testing to provide input to predictive models for licensing. Oak Ridge National Laboratory has a long history of performing accelerated irradiation testing and post-irradiation examination (PIE) of miniature material specimens to determine the evolution of microstructure and thermophysical properties, leveraging the High Flux Isotope Reactor, dedicated hot cells, and instruments for handling small, low-level radioactive samples. This work gives a brief overview of irradiation testing of miniature samples to support qualification of advanced fuel cladding materials such as FeCrAl alloys and SiC ceramic matrix composites. More recent work has enabled separate effects irradiation testing of miniature fuel specimens. The small fuel size essentially decouples fuel temperature from fission rate, allowing for highly accelerated testing without large temperature gradients. This paper briefly highlights the new mini fuel irradiation facility and its first set of experiments.

Petrie, Christian↗
DOE OSTI2021

Destructive Examination of a FeCrAl-UO 2 Irradiation Test

Several destructive postirradiation examinations were performed on an irradiation specimen that coupled an early iron-chromium-aluminum (FeCrAl) candidate alloy cladding with UO2 pellets. This irradiation test was designed to investigate the early life performance and compatibility of the FeCrAl-UO 2 cladding-fuel system under prototypic light water reactor neutronic conditions. Additionally, these tests were expected to provide neutron irradiated samples for severe accident testing of this fuel system. The irradiation studied in this work was part of the ATF-1 series of drop-in style irradiations performed in the Idaho National Laboratory Advanced Test Reactor. The rodlet studied in this work is one of three similar rodlets irradiated in ATF-1 that had approximately 7.6 cm of cladding machined from a wrought FeCrAl alloy and fueled with a 6.1 cm stack of UO 2 pellets. The outer diameter of the cladding was ~0.94 cm and the inner diameter was ~0.83 cm. After irradiation and non-destructive examination in Idaho this rodlet was shipped to the Oak Ridge National Laboratory hot-cells for further examination. The irradiated rodlet was sectioned into several samples for microstructural, micromechanical, and severe accident testing. During sectioning, it was noted that the fuel was not firmly bonded to the cladding and could be readily removed from small cladding slices. Microstructural characterization of fuel cross sections also revealed no significant interaction between the fuel and the cladding. Samples were also prepared for microhardness testing. To prepare for high temperature oxidation testing, the fuel was dissolved from segments of the cladding. High temperature oxidation testing of cladding segments was performed at 1200°C and 1300°C in a steam environment. Comparisons between the oxidation of this FeCrAl alloy in its neutron irradiated state, as-fabricated state, and the oxidation of Zircaloy-2 are made. The oxidation testing will be followed by ring compression testing to evaluate ductility. The microstructure of the samples after oxidation and ring compression testing will also be analyzed.

Harp, Jason↗
DOE OSTI2024

Dynamic Scaling Analysis of Accelerated Irradiation Testing on Additive Manufacturing Materials by Positron Annihilation

The timely applications of Additive Manufacturing (AM) materials in nuclear environments require accelerated irradiation tests, mainly ion irradiation to enable rapid prototyping. Low dose ion irradiation would cause sub-nanostructure changes by generation of lattice defects, vacancies, vacancy clusters and voids and void swelling caused by cellular dislocations. Positron Annihilation Lifetime (PAL), a novel technology, sensitive towards sub-nanostructure morphology with high accuracy (about 10-7 vacancy per atom), supported by Transition Electron Microscope (TEM) would be applied to identify the type and total size of the defects. The subsequent PAL measurements and TEM surface studies would be followed by PAL analysis that includes sophisticated trapping model. The PAS results would become an input to dynamic scaling analysis (that predicts radiation effects from low dose studies for high dose effects), which incorporate mean-field theory model. The final effect is an in-depth understanding of the microstructure evolution of AM materials under ion irradiation which can be extrapolated to the studies of neutron irradiation, since ion-irradiation takes less time and do not cause the irradiation hazard. The working hypothesis is that PAL technology, that have excellent sensitivity to low-defect concentration would help to identify ion-induced material damage on the atomic and nano-scale level, which then could be extrapolated to understand the neutron damage better.

accelerated irradiation testing↗
DOE OSTI2022

Irradiation Testing Summary for 2020

This is a one page summary of irradiation testing worked performed under the Advanced Sensor Initiative in FY2021.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2019

Characteristics of Fuel Articles for Irradiation Testing in INL-TREAT and MIT-NRL

To obtain valuable data that will help in the design and commissioning of the Transformational Challenge Reactor, a series of graded, increasingly complex irradiation tests are planned that will demonstrate the feasibility of advanced manufactured reactor core components. In the first irradiation test, conventionally- fabricated uranium nitride tristructural isotropic (UN TRISO) fuel kernels will be embedded into compacts that are fabricated using advanced manufacturing methods. These compacts are to be irradiated at Massachusetts Institute of Technology’s Nuclear Reactor Laboratory and Idaho National Laboratory’s Transient Reactor Test Facility. This report presents the characteristics of the proposed samples and experiments.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

AFC FY 2023 HFIR Irradiation Test Matrix – Supported by the Design of a Ring Specimen Irradiation Vehicle

The development and deployment of accident-tolerant cladding materials in light water nuclear reactors requires the collection of neutron-irradiated properties of these materials with specific consideration to the material properties orientation (i.e., tube geometry). To that end, a test plan to irradiate several cladding tube and ring specimens in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) was proposed, along with the development of a new capsule design that features ring specimens with a preexisting gauge section. This ring specimen geometry removes the need to machine a gauge section on the irradiated specimen and thus allows for direct application of a ring-pull test post-irradiation. The test matrix includes uncoated and chromium-coated Zircaloy-4 and advanced zirconium alloy cladding specimens, pre-characterized at ORNL. Plans are to irradiate the specimens at about 4 and 10 dpa—corresponding to two and five cycles at the HFIR centerline—with a target specimen gauge section temperature of 330°C.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Separate Effects Miniature Fuel Irradiation Testing to Support Fuel Safety

The recent interest in advanced fuel forms that offer increased thermal conductivity and uranium density, as well as modifications (e.g., use of dopants or additives) to traditional UO2 fuel, necessitates a vehicle for performing high-throughput irradiation testing of these fuels before they can be safely deployed. This interest, combined with the recent decommissioning of the Halden Reactor, places a strong burden on the remaining test reactors to provide the scientific data needed to advance these fuels toward commercialization. Oak Ridge National Laboratory is actively working to alleviate some of this burden by commissioning a new “MiniFuel” test facility that will provide a vehicle for performing accelerated separate-effects irradiation testing of sub-sized nuclear fuel specimens. The small size of the fuel allows for a flexible irradiation vehicle capable of exploring a wide range of fuel performance variables within a reasonable time and cost. This paper describes the experiment facility, and the initial irradiation experiments being performed including uranium nitride (UN) kernels, coated tristructural-isotropic (TRISO) UN particles, doped and un-doped (reference) UO2 disks, U3Si2 disks, and fully ceramic microencapsulated (FCM) pellets with UN kernels and a SiC matrix. A focus is placed on the performance of miniature FCM pellets, which are being considered as the ultimate accident tolerant fuel for light water reactors. The experiments described in this paper will provide basic scientific data (e.g., swelling, fission gas release) regarding irradiation performance under well-characterized conditions.

Petrie, Christian↗
DOE OSTI2023

In-Rod Sensor System Overview, Benefits and Recent Irradiation Test Results

Westinghouse is developing an In-Rod Sensor (IRS) System capable of measuring critical fuel rod parameters such as center-line fuel temperature, rod internal pressure and axial fuel pellet stack elongation without having penetrations to the fuel rod. The technology is similar to the Halden Reactor instrumentation, but the IRS system focuses on commercial fuel rod instrumentation. Both systems circumvent fuel rod penetrations, however the IRS system overcomes unique geometric challenges found in commercial fuel rod assemblies. The IRS could also play a key role in accelerated fuel qualification by allowing real-time measurements during separate effects tests, integral fuel testing or during irradiation of lead test rods or lead test assemblies. Combining these measurements with data acquired from targeted separate effects testing and multi-scale modelling could significantly reduce the time required to qualify new fuel and reduce the safety margin uncertainty of operating plants. The IRS is actively in development for Light Water Reactors (LWR) based fuel assemblies but can also be applied to Gen IV reactor designs with minimal modifications.

Carvajal, Jorge↗