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All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Status of Wrought FeCrAl-UO 2 Capsules Irradiated in the Advanced Test Reactor

Candidate cladding materials for accident tolerant fuel applications require extensive testing and validation prior to commercial deployment within the nuclear power industry. One class of cladding materials, FeCrAl alloys, is currently undergoing such effort. Within these activities is a series of irradiation programs within the Advanced Test Reactor. These programs are developed to aid in commercial maturation and understand the fundamental mechanisms controlling the cladding performance during normal operation of a typical light water reactor. Three different irradiation programs are on-going; one designed as a simple proof-of-principle concept, the other to evaluate the susceptibility of FeCrAl to fuel-cladding chemical interaction, and the last to fully simulate the conditions of a pressurized water reactor experimentally. To date, nondestructive post-irradiation examination has been completed on the rodlet deemed FCA-L3 from the simple proof-of-concept irradiation program. Initial results show possible breach of the rodlet under irradiation but further studies are needed to conclusively determine whether breach has occurred and the underlying reasons for such a possible failure. Further work includes characterizing additional rodlets following irradiation.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Analysis and Experimental Qualification of an Irradiation Capsule Design for Testing Pressurized Water Reactor Fuel Cladding in the High Flux Isotope Reactor

The Advanced Fuels Campaign within the Fuel Cycle Research and Development program of the Department of Energy Office of Nuclear Energy is currently investigating a number of advanced nuclear fuel cladding concepts to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are some of the leading candidates to replace traditional zirconium alloys due to their superior oxidation resistance, provided no prohibitive irradiation-induced embrittlement occurs. Oak Ridge National Laboratory has developed experimental designs to irradiate thin-walled cladding tubes with representative pressurized water reactor geometry in the High Flux Isotope Reactor (HFIR) under relevant temperatures. These designs allow for post-irradiation examination (PIE) of cladding that closely resembles expected commercially viable geometries and microstructures. The experiments were designed using relatively inexpensive rabbit capsules for the irradiation vehicle. The simplistic designs combined with the extremely high neutron flux in the HFIR allow for rapid testing of a large test matrix, thus reducing the time and cost needed to advanced cladding materials closer to commercialization. The designs are flexible in that they allow for testing FeCrAl alloys, stainless steels, Inconel alloys, and zirconium alloys (as a reference material) both with and without hydrides. This will allow a direct comparison of the irradiation performance of advanced cladding materials with traditional zirconium alloys. PIE will include studies of dimensional change, microstructure variation, mechanical performance, etc. This work describes the capsule design, neutronic and thermal analyses, and flow testing that were performed to support the qualification of this new irradiation vehicle.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Vanderbilt University Gamma Irradiation of Nano-modified Concrete (2017 Milestone Report)

This document outlines the irradiation of concrete specimens in the Gamma Irradiation Facility in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). Two gamma irradiation runs were performed in July of 2017 on 18 reference mortar bar specimens, 26 reference cement paste bar specimens, and 28 reference cement paste tab specimens to determine the dose and temperature response of the specimens in the gamma irradiation environment. Specimens from the first two gamma irradiations were surveyed and released to Vanderbilt University. The temperature and dose information obtained informs the test parameters of the final two gamma irradiations of nano-modified concrete planned for FY 2018.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2018

Tensile Testing of Irradiated Grade 92 Ferritic-Martensitic Steels at the IMET Hot Cell Facility

This report summarizes the completed post-irradiation tensile testing of five ferritic-martensitic (FM) steel SS-J2 specimens at the Irradiated Materials Examination and Testing (IMET) hot cell facility at Oak Ridge National Laboratory (ORNL). A brief description of the motivation for the study, the properties and irradiation conditions of the samples, and the tensile testing conditions are provided. The results will be used to compare the mechanical properties of the irradiated specimens to un-irradiated conditions to determine how irradiation affects the steel alloy material behavior.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Assembly of Rabbit Capsules for Irradiation of Pyrolytic Carbon / Silicon Carbide Diffusion Couples in the High Flux Isotope Reactor

Tristructural isotropic (TRISO)–coated particle fuel is a promising advanced fuel concept being considered for several advanced reactor applications and for accident-tolerant fuel for light water reactors. One of the aspects studied in the development of this advanced fuel concept is the release of specific fission products (Ag, Eu, and Sr). The silicon carbide (SiC) layer of TRISO fuel serves as the primary barrier to metallic fission products and actinides not retained in the fuel kernel. The goal of this project is to evaluate the effect of irradiation on the diffusion of these fission products in the SiC layer of the fuel. For this purpose, rabbit capsules containing small slab diffusion couple specimens have been assembled to be irradiated in the High Flux Isotope Reactor (HFIR). The diffusion couple specimens have been fabricated using similar processes and equipment as those used to make TRISO particles; the desired fission products have been implanted in the specimens using an ion accelerator. Moreover, the effect of temperature on the fission products diffusion will be studied separately by performing thermal experiments in the absence of irradiation. This report describes the irradiation experiment design concept, summarizes the irradiation test matrix, and reports on the successful assembly of two rabbit capsules that will be irradiated in the HFIR.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Development of a Flexible Design for Irradiation of Miniature Tensile and Charpy Test Specimens in the High Flux Isotope Reactor

Fusion promises a sustainable, clean, safe source of energy that could play a major role in solving the great challenge of meeting the world’s growing power needs. Much of fusion technology in its current state requires further maturation to support a usable fusion reactor platform. Given that the environment in and around a sustained fusion reaction is very harsh, selecting the best materials for such a reactor is crucial. The European fusion materials community spent many years developing one such material, a low-activation martensitic structural steel named Eurofer that is intended primarily for use in the fusion blanket modules. Although this material has reached technological maturity, nuclear performance data are required to qualify Eurofer for these applications. Karlsruhe Institute of Technology (KIT), a member of the EUROfusion consortium, has sponsored an irradiation program to establish irradiation performance data for various forms of the Eurofer alloy family. This document describes the design and development of flexible irradiation capsule designs that can be used to irradiate specimens to temperatures ranging from ~200 – 400°C. This report also details as-built data for the screening and high dose phases of the KIT irradiation campaign of capsules that were submitted for irradiation in the High Flux Isotope Reactor in FY18.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

AGR 3/4 Irradiation Test Final As Run Report, Revision 1

Several fuel and material irradiation experiments have been planned for the Idaho National Laboratory Advanced Reactor Technologies Technology Development Office Advanced Gas Reactor Fuel Development and Qualification Program (referred to as the INL ART TDO/AGR fuel program hereafter), which supports the development and qualification of tristructural-isotropic (TRISO) coated particle fuel for use in HTGRs. The goals of these experiments are to provide irradiation performance data to support fuel process development, qualify fuel for normal operating conditions, support development and validation of fuel performance and fission product transport models and codes, and provide irradiated fuel and materials for post irradiation examination and safety testing (INL 05/2015). AGR-3/4 combined the third and fourth in this series of planned experiments to test TRISO coated low enriched uranium (LEU) oxycarbide fuel. This combined experiment was intended to support the refinement of fission product transport models and to assess the effects of sweep gas impurities on fuel performance and fission product transport by irradiating designed-to-fail fuel particles and by measuring subsequent fission metal transport in fuel-compact matrix material and fuel-element graphite. The AGR 3/4 fuel test was successful in irradiating the fuel compacts to the burnup and fast fluence target ranges, considering the experiment was terminated short of its initial 400 EFPD target (Collin 2015). Out of the 48 AGR-3/4 compacts, 42 achieved the specified burnup of at least 6% fissions per initial heavy-metal atom (FIMA). Three capsules had a maximum fuel compact average burnup < 10% FIMA, one more than originally specified, and the maximum fuel compact average burnup was <19% FIMA for the remaining capsules, as specified. Fast neutron fluence fell in the expected range of 1.0 to 5.5 × 10 25 n/m 2 (E >0.18 MeV) for all compacts. In addition, the AGR-3/4 experiment was globally successful in keeping the temperature in the twelve capsules relatively flat in a range of temperatures suitable for the measurement of fission product diffusion in compact matrix and structural graphite materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Development of a characterization methodology for post-irradiation examination of miniature fuel specimens

The miniature fuel (MiniFuel) irradiation experimental design at Oak Ridge National Laboratory (ORNL) enables benchmark testing and separate effects testing of novel fuel types in the High Flux Isotope Reactor (HFIR) in support of the Advanced Fuels Campaign (AFC). The first round of MiniFuel sub-capsules will be available for post-irradiation examination at the end of 2018. This report establishes the procedure for disassembly, fuel retrieval, post-irradiation analysis of the fission product behavior (fission gas release and fuel swelling), and microstructural characterization of the irradiated MiniFuel specimens. All methods were tested and validated out-of-cell and are ready to be implemented in-cell for use on the first set of irradiated MiniFuel specimens. The path forward regarding further characterization techniques is presented, which will involve analyzing other specimen geometries using the methods presented in this report and expanding the post-irradiation analysis of MiniFuel specimens to include additional properties.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Design of the Universal Neutron Irradiator

SHINE medical isotopes is developing a novel method to produce the medical isotope Mo-99, which is the parent of Tc-99m, the most widely used diagnostic compound in nuclear medicine. This method utilizes U-235 fission in a subcritical liquid target solution, driven by an accelerator-based neutron source that is based on a deuterium-tritium reaction in a gaseous target. In this approach, the neutron-generating target is surrounded by uranyl sulfate (low-enriched uranium); fission in the uranium salt solution efficiently produces the desired Mo-99, as well as other fission products. After irradiation, Mo-99 is separated and the uranyl sulfate solution can be irradiated again to produce Mo-99. This allows the solution to be recycled for a very long time and significantly reduces the amount of associated nuclear waste. SHINE has received U.S. Nuclear Regulatory Commission (NRC) approval to construct an Mo-99 production facility based on this method. In order to produce Mo-99 efficiently, the solution should be as close as possible to criticality. Because the solution is over moderated, any precipitation could push the solution to become critical and thereby hard to control. One of the possible precipitation scenarios is the formation of uranyl peroxide. Under irradiation, water decomposes into the products of radiolysis, some of which will form peroxide. To prevent precipitation, one has to develop a way to decompose the peroxide before it can react with uranium and form hard-to-dissolve precipitates. To study peroxide formation and develop methods to prevent precipitation, we are developing an experiment capable of mimicking radiation conditions in the SHINE system. This report describes the irradiation setup we designed to irradiate a small volume of uranyl sulfate solution for the duration of one SHINE irradiation (~6 days) at the relevant fission power (~0.5 W/cc)

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2020

Irradiation and PIE of alloys X-750 and XM-19 (EPRI Phase III)

The Nuclear Science User Facilities (NSUF) based at Idaho National Laboratory (INL), along with the Electric Power Research Institute (EPRI), formed an agreement to test representative alloys used as reactor structural materials as a pilot program to establish guidelines for future NSUF research programs. This report contains results from the portion of this program established as Phase III (of three phases), entailing irradiation and post-irradiation examination of select alloys typical of boiling water reactor (BWR) internal structural materials. Phases I and II are the subject of separate reports and represent baseline material test results and irradiation experiment design, respectively. The intent of this Phase III research program is to determine properties for the materials of interest after being irradiated at the Advanced Test Reactor (ATR) to three different target fast (E>1MeV) fluences: 5.0 x 10¹⁹ n/cm², 2.0 x 10²⁰ n/cm², and 1.0 x 10²¹ n/cm². These correspond to irradiation damage levels (displacements per atom [dpa]) of approximately 0.08, 0.30, and 1.4 dpa, which represent comparable levels to (a) a previous study which looked at X-750 irradiated to ~1 x 10¹⁹ n/cm², comparable to the lowest fluence; (b) approximately a medium level of fluence for BWR components; and (c) extended life (60 – 80 years) for BWR components. The materials chosen for this research are the nickel-based alloy X-750 and austenitic stainless steel XM-19. A spare core shroud upper support bracket of alloy X-750 was purchased by EPRI from Southern Co., and a section of XM-19 plate was purchased by EPRI from GE-Hitachi. These materials were sectioned at GE Global Research Center (GE-GRC), and parts were provided to INL for use in this pilot project.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Irradiation and PIE of alloys X-750 and XM-19 (EPRI Phase III)

The Nuclear Science User Facilities (NSUF) based at Idaho National Laboratory (INL), along with the Electric Power Research Institute (EPRI), formed an agreement to test representative alloys used as reactor structural materials as a pilot program to establish guidelines for future NSUF research programs. This report contains results from the portion of this program established as Phase III (of three phases), entailing irradiation and post-irradiation examination of select alloys typical of boiling water reactor (BWR) internal structural materials. Phases I and II are the subject of separate reports and represent baseline material test results and irradiation experiment design, respectively. The intent of this Phase III research program is to determine properties for the materials of interest after being irradiated at the Advanced Test Reactor (ATR) to three different target fast (E>1MeV) fluences: 5.0 x 10¹⁹ n/cm², 2.0 x 10²⁰ n/cm², and 1.0 x 10²¹ n/cm². These correspond to irradiation damage levels (displacements per atom [dpa]) of approximately 0.08, 0.30, and 1.4 dpa, which represent comparable levels to (a) a previous study which looked at X-750 irradiated to ~1 x 10¹⁹ n/cm², comparable to the lowest fluence; (b) approximately a medium level of fluence for BWR components; and (c) extended life (60 – 80 years) for BWR components. The materials chosen for this research are the nickel-based alloy X-750 and austenitic stainless steel XM-19. A spare core shroud upper support bracket of alloy X-750 was purchased by EPRI from Southern Co., and a section of XM-19 plate was purchased by EPRI from GE-Hitachi. These materials were sectioned at GE Global Research Center (GE-GRC), and parts were provided to INL for use in this pilot project.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Assembly and Delivery of Rabbit Capsules for Irradiation of Prototype Metal and Nanocomposite Specimens in the High Flux Isotope Reactor

The Massachusetts Institute of Technology has developed a manufacturing process for nanodispersion-strengthened materials that could be used as fuel cladding or reactor core materials. This report describes the design concept of irradiation capsules intended to accommodate specimens from such materials. The capsules will be irradiated in the flux trap of Oak Ridge National Laboratory’s High Flux Isotope Reactor at three different fluence levels. The goal of this neutron irradiation is to investigate the effects of irradiation on those materials at the nanoscale. In addition, the report describes the irradiation test matrix and the successful assembly of the irradiation capsules.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

High Temperature Steam Oxidation of Irradiated FeCrAl in the Severe Accident Test Station

FeCrAl-UO2 test capsules were fabricated at Oak Ridge National Laboratory (ORNL) and irradiated at the Advanced Test Reactor. Following irradiation, samples were sectioned from the irradiated rod and oxidation kinetics were evaluated to access the candidate cladding high-temperature oxidation performance following irradiation. The irradiation was performed at approximately 400°C to a burnup of 10 GWd/MT. The high-temperature oxidation tests were conducted in the ORNL Severe Accident Test Station at 1200 and 1300°C. Weight measurements were taken before and after oxidation testing. Cross-sections of the cladding were metallographically mounted and optical microscopy was performed. Measurements of the oxidation layer before and after high-temperature testing were collected. The results indicate the irradiated FeCrAl C35M alloy provided good thermal stability up to 1200°C.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Effects of Irradiation on Thermal Properties of Polyurethane Foam

Polyurethane foam is used as a thermal insulator, fire protective material, and impact absorber in the 9977 shipping package. The 9977 shipping package is used to transport special nuclear materials (SNM), and store SNM in the K-area complex at the Savannah River Site. The primary fire-retardant mechanism in the foam is that it produces intumescent char when exposed to high temperatures. Intumescence occurs in the direction of heat application and can allow the foam to “repair” cracks in the event of a mechanical impact, in addition to serving as a secondary thermal barrier with decreased heat conductivity. Because the foam is being used for long-term storage applications, it is vital that it retain its physical, thermal, and mechanical properties over the course of time and with the events of exposure of heat and irradiation. The goal of this project was to determine the effects of gamma irradiation on the thermal properties of polyurethane foam, both in terms of cumulative irradiation and rate of irradiation. At the dose rates and cumulative doses of irradiation measured, there was no correlation between higher cumulative irradiation and thermal stability, specific heat capacity, or chemical structure. Based on the results of the experiments, the fire protective properties in the 9977 shipping containers should be retained with the expected amount of radiation exposure.

36 MATERIALS SCIENCE↗
DOE OSTI2022

A Procedure to Correct the Historical Atmospheric Longwave Irradiance Data When the World Reference Is Established with Respect to the International System of Units

Historical atmospheric longwave irradiance data sets with traceability to the International System of Units (SI) are essential for renewable energy and atmospheric science research and applications. To date, all pyrgeometers used to measure the irradiance are traceable to the interim World Infrared Standard Group (WISG), not to SI units. In 2013, the Absolute Cavity Pyrgeometer (ACP) (Reda et al. 2012) was developed at the National Renewable Energy Laboratory (NREL) to measure the atmospheric longwave irradiance. The ACP has been compared against the InfraRed Integrating Sphere (IRIS), developed by the Physikalisch-Meteorologisches Observatorium Davos/World Radiation Center (PMOD/WRC) (Gröbner 2012). The ACP and the IRIS are absolute instruments traceable to SI units through the International Temperature Scale of 1990. Results of six comparisons between the ACP and the IRIS at different locations have shown that the irradiance measured by WISG pyrgeometers underestimates clear-sky atmospheric longwave irradiance by 2 W/m 2 to 6 W/m 2 (Gröbner et al. 2014); therefore, once the world reference is established with traceability to SI units, the WISG would be corrected, then used to calibrate field pyrgeometers with traceability to SI units. The following described method is used to correct the historical atmospheric longwave irradiance data sets in anticipation of the WISG scale change.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2021

Effects of Electron Irradiation on Candidate Materials for Target Window in Accelerator-Driven Mo-99 Production

Room-temperature (RT) axial tensile tests were conducted in 2018 and 2021 with flat, sub-sized, dog-boned samples of non-irradiated and irradiated Inconel 718 and irradiated beryllium (Be). The sample gauge length was 7.62 mm. Tensile properties of interest are the engineering yield stress (YS), ultimate tensile stress (UTS), uniform elongation (UE) and total elongation (TE). The electron dose of the 2021 samples was ≈2 times the dose of samples tested in 2018. Relative dose rates of 0, 1 (2018), and 2 (2021) are assigned to these samples. The 2021 electron-irradiated Be samples were brittle with a low failure stress of 338±64 MPa based on two samples. The failures occurred in the gauge section. The 2018 results for electron-irradiated Be indicated a slight decrease in ductility and strength for the one electron-irradiated sample that failed in the gauge section.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2022

Irradiation of MiniFuel Targets Bearing TRISO Fuel Compacts (Status Report)

Irradiation testing of MiniFuel compacts bearing tristructural isotropic (TRISO) fuel particles was performed at Oak Ridge National Laboratory (ORNL) to support the development of Kairos Power’s (KP’s) fluoride salt–cooled high-temperature reactor concept. The fuel compacts were fabricated with TRISO fuel particles of different types—including low-enriched uranium oxide, uranium carbide (LEUCO), natural uranium oxide, uranium carbide (NUCO), and low-enriched uranium dioxide (LEUO 2 )—and inserted into MiniFuel irradiation targets. Five targets were assembled and inserted in the High Flux Isotope Reactor (HFIR) for four cycles. The data collected post-irradiation will provide experimental input to validate TRISO fuel performance models for high particle power operations. This report summarizes the completion of the HFIR irradiation, the as-irradiated numerical analysis, and the post-irradiation work performed to date. This work was performed under the Nuclear Science User Facility program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Uncertainty in Annual Energy Resulting from Uncertain Irradiance Measurements

We report an analysis quantifying the contribution to uncertainty in annual energy projections from uncertainty in ground-measured irradiance. Uncertainty in measured irradiance is quantified for eight instruments by the difference from a well maintained, secondary standard pyranometer which is regarded as truthful. We construct a statistical model of irradiance uncertainty and apply the model to generate a sample of 100 annual time series of irradiance for each instrument. The sample is propagated through a common performance model for a reference photovoltaic system to quantify variation in annual energy. Although the measured irradiance varies from the reference by a few percent (standard deviation of 1-2%) the uncertainty in annual energy is on the order of a fraction of one percent. We propose a model for a factor that represents uncertainty in modeled annual energy that arises from uncertainty in ground-measured irradiance.

14 SOLAR ENERGY↗