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At least 451 records · Page 25

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↗

Safety Testing and Post-Safety Test Examination of AGR-2 UCO Compact 2-3-2 and AGR-2 UO2 Compact 3-4-1

In this report, results from 1800°C safety testing and post-safety-test PIE of AGR-2 UCO Compact 2-3-2 and 1700°C safety testing of AGR-2 UO2 Compact 3-4-1 are discussed and compared to previously reported results from 1800°C safety testing of AGR-2 UCO Compact 5-4-1 and 1600°C safety testing of compact “twins”: AGR-2 UCO Compact 2-3-1 and AGR-2 UO 2 Compact 3-4-2. These compact pairs are called “twins” because their symmetric locations in the irradiation test train resulted in similar irradiation conditions. Table 1 shows the calculated burnup in percent fissions per initial metal atom (FIMA), the fast neutron fluence (neutron energies > 0.18 MeV), and the average compact temperatures during irradiation for these compacts. Capsule 2 Compacts 2-3-2 and 2-3-1 were intentionally irradiated at notably higher temperature than compacts in the other AGR-2 capsules or those previously studied in the AGR-1 irradiation test. The Capsule 2 compacts provide insight into the irradiation and post-irradiation safety test performance at higher temperatures than typical HTGR operating temperatures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Radiation Induced Segregation at Grain Boundaries in Steels: NSUF 2018 Milestone Report

This document outlines the user supported post-irradiation characterization by Oak Ridge National Laboratory under the Nuclear Scientific User Facility Consolidated Innovative Nuclear Research project, “Correlative Atom Probe and Electron Microscopy Study of Radiation Induced Segregation at Low and High Angle Grain Boundaries in Steels.” Fiscal Year 2018 was the second of a two-year project aimed at using analytical electron microscopy combined with atom probe tomography (APT) to probe the nanometer length scale segregations in austenitic stainless steels after neutron irradiation. Particularly, analytical electron microscopy using the FEI F200X Talos was successfully completed in the first year, and the second year was focused on APT study. Scanning electron microscopy (SEM) coupled with an electron backscatter diffraction (EBSD) detector was performed on selected specimens to locate high and low angle grain boundaries for targeted specimen preparation using focused ion beam (FIB). FIB prepared lift outs were used for APT examination of the grain boundaries using a Cameca Instruments LEAP 4000X HR instrument. This report briefly summarizes the specimen preparation and APT results obtained at the Center for Advanced Energy Studies.

36 MATERIALS SCIENCE↗

NSUF Tensile Testing and Characterization of Irradiated Grade 92 Ferritic-Martensitic Steels

This report summarizes the completed post-irradiation tensile testing of five ferritic-martensitic steel SS-J2 specimens at the Irradiated Materials Examination and Testing hot cell facility at Oak Ridge National Laboratory. 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 unirradiated conditions to determine how irradiation affects the steel alloy material behavior.

36 MATERIALS SCIENCE↗

AGR-5/6/7 Fuel Fabrication Feedstock Chemical Purity Specifications

This specification has been prepared to provide fuel fabricators with minimum chemical purity specifications intended to ensure that fuel components have consistent properties and that impurity levels in the as-fabricated fuel is sufficiently low to minimize in-pile TRISO particle failures and to enhance fission product retention. Process chemical purity specifications are given for the following processes: • Kernel forming and sintering • TRISO particle coating • Matrix preparation • TRISO overcoating • Compact carbonization and heat treatment

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Fission Gas Monitoring System Efficiency Calibration Summary Report

The Fission Gas Monitoring System was designed to work in conjunction with the Fuel Accident Condition Simulator Furnace and is used to measure fission gas releases during post-irradiation fuel heating tests. A high-purity helium sweep gas is swept past the heated fuel sample in the furnace and routed to a fission gas monitoring system, which cryrogenically traps the krypton (Kr) and xenon (Xe) fission gases. As the radioactive species of the trapped gas decays, high-purity germanium detectors monitor the amount of fission gas in each trap. The primary fission product present in this test is 85Kr, which has a 10.76-year half-life. The presence of 85Kr indicates particle defects and iodine release. There is also the potential for re-irradiating the fuel compacts in the Neutron Radiography Reactor in the Hot Fuel Examination Facility basement. With re-irradiation, it is possible to have xenon fission gas present, specifically 133Xe, which has a 5.243-day half-life. Re-irradiation also produces 131I, which is important for design-basis accidents and regulatory licensing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Demonstration of Integrated PostIrradiation Process Steps forMultikilogram Amounts of Low- and High-Burnup Used Nuclear Fuels

sponsored research and development (R&D) on the advanced, closed nuclear fuel cycle, including post-irradiation processing of used fuels, recovery of separated components, and preparation of recycle nuclear fuels. During the period 2007–2010, kilogram-scale demonstrations of the integrated processing steps were performed for actual light water reactor (LWR) used nuclear fuels (UNF). These integrated processing demonstrations were called “the Coupled-End-to-End” (CETE) campaigns and were associated with the Global Nuclear Energy Partnership program, which ran concurrently. The demonstrations were carried out in the shielded hot cells and associated gloveboxes located in the Irradiated Fuels Examination Laboratory (IFEL) and the Radiochemical Engineering Development Center (REDC) at the Oak Ridge National Laboratory (ORNL). Limited circulation program reports and a few technical papers were issued. The subject document provides a comprehensive summary of results, technical comparisons, and conclusions obtained from the demonstration campaigns and subsequent associated R&D.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Advanced Multiscale Microscopy Characterization of High Burnup LWR UO 2 Before and After LOCA Testing

Previous microscopy work within the campaign has observed increased porosity coupled with subgrain formation in the mid-radial region of the fuel. However, examinations of high burnup fuel taken from multiple reactors with different operating conditions have shown that this structure can move radially inward depending on the fuel operating conditions. It is currently theorized that inter- and intragranular fission gas bubble nucleation is precursory to the grain subdivision observed in the mid-radial and central regions of the fuel. Characterization of samples after loss-of-coolant accident (LOCA) testing has revealed that these restructured regions with a high density of bubbles and subgrains pulverize during the transient test. It appears that the increase in fission gas bubbles coupled with grain subdivision renders the fuel mechanically weaker during a LOCA transient and thus susceptible to fuel fragmentation relocation and dispersal (FFRD). Work this past fiscal year has prioritized understanding this behavior by using advanced microscopy to investigate fission gas behavior in the as-irradiated and post-LOCA state. Additional work has been performed to verify the theory described above by analyzing multiple post-LOCA optical micrographs and comparing that to as-irradiated microstructural data. This document reports progress in the post-irradiation characterization of high burnup nuclear fuel with emphasis on the restructured fuel regions, particularly how microstructural features influence FFRD and fission gas release behavior in LOCA conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Preparation and Analysis of Austenitic Steel Samples Irradiated at Very High Damage Doses

This report describes the preliminary results of the visual inspection, cutting, preparation for IASCC (irradiation-assisted stress corrosion cracking) testing, and property evaluation of austenitic stainless steel specimens irradiated at very-high-damage doses (up to 125 dpa) in the BOR-60 fast reactor. Work was performed at Oak Ridge National Laboratory using the Irradiated Materials Examination and Testing (hot cell) Facility and the Low Activation Materials Development and Analysis (LAMDA) laboratory. Report sections describe specific aspects of the research activity. Section 1 reviews the general background, material, and specimen origin, and provides detailed information on material composition and damage dose level. Section 2 describes operations performed with irradiated specimens at the hot cell facility and LAMDA laboratory. At the moment, tensile specimen preparation is complete, and the tensile bars are ready for shipment to the University of Michigan for stress corrosion cracking testing. Section 3 discusses the microstructure of archive alloys and post-irradiation material properties; because of the very limited timeframe, only selected results are given on the magnetic properties of the irradiated specimens. Section 4 summarizes the work performed, presents conclusions, and discusses future activities within the line of research.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Delayed expression of hpS2 and prolonged expression of CIP1/WAF1/SDI1 in human tumour cells irradiated with X-rays, fission neutrons or 1 GeV/nucleon Fe ions

PURPOSE: Differences in gene expression underlie the phenotypic differences between irradiated and unirradiated cells. The goal was to identify late-transcribed genes following irradiations differing in quality, and to determine the RBE of 1 GeV/n Fe ions. MATERIALS AND METHODS: Clonogenic assay was used to determine the RBE of Fe ions. Differential hybridization to cDNA target clones was used to detect differences in expression of corresponding genes in mRNA samples isolated from MCF7 cells irradiated with iso-survival doses of Fe ions (0 or 2.5 Gy) or fission neutrons (0 or 1.2 Gy) 7 days earlier. Northern analysis was used to confirm differential expression of cDNA-specific mRNA and to examine expression kinetics up to 2 weeks after irradiation. RESULTS: Fe ion RBE values were between 2.2 and 2.6 in the lines examined. Two of 17 differentially expressed cDNA clones were characterized. hpS2 mRNA was elevated from 1 to 14 days after irradiation, whereas CIP1/WAF1/SDI1 remained elevated from 3 h to 14 days after irradiation. Induction of hpS2 mRNA by irradiation was independent of p53, whereas induction of CIP1/WAF1/SDI1 was observed only in wild-type p53 lines. CONCLUSIONS: A set of coordinately regulated genes, some of which are independent of p53, is associated with change in gene expression during the first 2 weeks post-irradiation.

NASA Discipline Radiation Health↗

Ionizing Radiation Affects Gene Expression in Mouse Skin and Bone

Future long-duration space exploration beyond low earth orbit will increase human exposure to space radiation and microgravity conditions as well as associated risks to skeletal health. In animal studies, radiation exposure (greater than 1 Gy) is associated with pathological changes in bone structure, enhanced bone resorption, reduced bone formation and decreased bone mineral density, which can lead to skeletal fragility. Definitive measurements and detection of bone loss typically require large and specialized equipment which can make their application to long duration space missions logistically challenging. Towards the goal of developing non-invasive and less complicated monitoring methods to predict astronauts' health during spaceflight, we examined whether radiation induced gene expression changes in skin may be predictive of the responses of skeletal tissue to radiation exposure. We examined oxidative stress and growth arrest pathways in mouse skin and long bones by measuring gene expression levels via quantitative polymerase chain reaction (qPCR) after exposure to total body irradiation (IR). To investigate the effects of irradiation on gene expression, we used skin and femora (cortical shaft) from the following treatment groups: control (normally loaded, sham-irradiated), and IR (0.5 Gy 56Fe 600 MeV/n and 0.5 Gy 1H 150 MeV/n), euthanized at one and 11 days post-irradiation (IR). To determine the extent of bone loss, tibiae were harvested and cancellous microarchitecture in the proximal tibia quantified ex vivo using microcomputed tomography (microCT). Statistical analysis was performed using Student's t-test. At one day post-IR, expression of FGF18 in skin was significantly greater (3.8X) than sham-irradiated controls, but did not differ at 11 days post IR. Expression levels of other genes associated with antioxidant response (Nfe2l2, FoxO3 and Sod1) and the cell cycle (Trp53, Cdkn1a, Gadd45g) did not significantly differ between the control and IR groups at either time point. Radiation exposure resulted in a 27.0% increase in FGF18-positive hair follicles at one day post-IR and returned to basal levels at 11 days post-IR. A similar trend was observed from FGF18 gene expression analysis of skin. In bone (femora), there was an increase in the expression of the pro-osteoclastogenic cytokine, MCP-1, one day after IR compared to non-irradiated controls. FGF18 expression in skin and MCP- 1 expression in bone were found to be positively correlated (P less than 0.002, r=0.8779). Further, microcomputed tomography analysis of tibia from these animals showed reduced cancellous bone volume (-9.9%) at 11 days post- IR. These results suggest that measurements of early radiation induced changes in FGF18 gene expression in skin may have value for predicting subsequent loss of cancellous bone mass. Further research may lead to the development of a relatively simple diagnostic tool for bone loss, with the advantage that hair follicles and skin are relatively easy to acquire from human subjects.

space radiation↗

Cardiovascular Responses to Simulated Spaceflight: Molecular Signatures and Surrogate Outputs to Measure CVD Risk

During extended space missions beyond low Earth orbit, astronauts will encounter prolonged periods of weightlessness and low dose space radiation. Previous studies have shown that exposure to small doses of high LET radiation (< 50 cGy) can lead to both short-term and long-term alterations in heart function, structure and underlying molecular mechanisms. In this study, we aim to identify the molecular signature associated with the cardiovascular response to simulated galactic cosmic radiation (5-ion GCR) alone or in combination with simulated weightlessness at time intervals relevant to mission length and recovery. Additionally, we aim to determine whether sex impacts cardiovascular responses to these spaceflight factors. Our overarching goal is to enhance our understanding of the cardiovascular risks associated with extended space missions and the clinical endpoints they suggest. We hypothesize that exposure to simulated space radiation leads to enduring alterations in the transcriptome, redox signaling and cytokine environment of cardiovascular tissue, some which have known links with reduced cardiovascular performance, aging, and increased risk of cardiovascular disease (CVD). Furthermore, we posit that simulated space radiation exposure in combination with simulated microgravity exacerbates cardiovascular deficits compared to single factor exposure. Female and male C57BL/6J mice, aged 23-24 weeks, were exposed to a single dose of 5, 15, or 50 cGy of 5-ion GCR, or sham-treated (0 cGy). Euthanasia was performed at 14 days and ~4 months post-irradiation. Hearts, aorta and blood plasma were collected shortly thereafter. RNA-sequencing of left ventricles at ~4 months post-GCR exposure revealed sex differences in the heart transcriptome with a few genes showing radiation-dependent changes in expression levels. Notably, some of the differentially expressed genes in 15 and 50 cGy GCR groups are known to play roles in the development of CVD. Analysis of protein levels of a subset of inflammatory cytokines in the heart indicated sex differences but no differences between sham and 50 cGy groups. Results also showed correlations among differentially expressed genes and a subset of inflammatory cytokines, with some correlations altered by GCR exposure. These findings suggest that GCR exposure can modify protein and gene networks linked to inflammation and CVD progression. In the aorta, telomere lengths were comparable across treatment groups sexes. Mitochondrial copy number is a biomarker for mitochondrial function with decreased copy numbers associated with cardiometabolic disease traits. Mitochondrial copy numbers of aorta also showed no sex nor dose differences. In a second study, mice underwent one week of simulated microgravity by hindlimb unloading (HU) and then exposed to a single dose of 15 cGy of 5-ion GCR. HU was conducted for an additional two weeks following GCR exposure. Single factor exposure groups (HU or GCR only) also were included in the study. Euthanasia was then performed and the same tissues were collected. Protein levels of select inflammatory cytokines in the heart showed sex-dependent differences in expression. In the aorta, telomere lengths and mitochondrial copy number also showed sex differences. In summary, our results indicate differences between sexes in biomarkers related to cardiovascular health. Exposure to 5-ion GCR or HU, alone or in combination, did not result in changes in most of the cardiovascular biomarkers that were examined. However, in the heart, simulated space radiation at doses of 15 and 50 cGy led to long-term alterations in the expression levels of a small group of genes known to be associated with the progression of CVD. The long-term transcriptomic changes resulting from exposure to simulated space radiation should be carefully investigated to mitigate adverse cardiovascular events during and after deep space missions. Our results also highlight the importance of sex-specific strategies in monitoring and maintaining cardiovascular health during and after deep space missions.

cardiovascular↗

Support for Fuel Fabrication and Characterization

Oak Ridge National Laboratory (ORNL) has participated in the Advanced Gas Reactor (AGR) program for the Next Generation Nuclear Plant (NGNP) and the transition to the Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) Technology Development Office (TDO) since the inception of the program. In this effort, ORNL has fabricated tristructural isotropic (TRISO) fuel and fuel compacts for AGR-1, AGR-2, and AGR-3/4 experiments, provided matrix only components for AGC-2, AGC-4, and AGR 3/4 experiments and for university studies, and provided characterization, coating, and compaction support for INL and BWX Technologies Nuclear Operations Group (BWXT). In fiscal year (FY) 2017, the activities for the INL ART TDO fuel development are focused on preparations for the AGR-5/6/7 experiments and technical support to the Generation-IV International Forum (GIF) pre-irradiation leach-burn-leach (LBL) round robin effort with China and South Korea. With the successful transfer of fuel fabrication technology to BWXT, ORNL’s involvement has shifted toward transferring fuel characterization methods, providing technical support, verification of BWXT laboratory results, and fabricating surrogate fuel materials for university collaborations. ORNL will also provide analytical support to BWXT and INL for uranium dispersion analysis and confirmatory fuel characterization analyses as needed.

AGR-5/6/7 Fuel Specification

The Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) Technology Development Office (TDO) program is developing and qualifying fuel for a first of a kind high temperature gas-cooled reactor (HTGR) and associated systems to generate electrical power and produce hydrogen or process heat. One of the critical technology development areas of the program is nuclear fuel development and qualification. Fuel qualification is needed to support licensing, operation, and near term deployment of an HTGR reactor.5 Toward this end, the Advanced Gas Reactor (AGR) program includes manufacturing fuels for a series of irradiation experiments in the Advanced Test Reactor (ATR) at INL.5 The fifth, sixth, and seventh irradiation experiments, called “AGR 5/6/7,” will use fuel manufactured entirely in engineering scale equipment to demonstrate the performance of fuels fabricated in a production environment. This fuel will be composited from multiple low enriched uranium carbide/oxide (LEUCO) kernel forming runs, multiple tristructural isotropic (TRISO) coating runs, and multiple overcoating runs. The fuel will be composited into lots as kernels, TRISO particles, and compacts. Previous AGR fuel specifications1,2,8,9 provide the basis for most of the following fuel specifications, which are specific to the AGR 5/6/7 experiment, and may not be immediately applicable to fuel produced for an HTGR reactor. Fuel specifications will be influenced by the final reactor design.

Support for Fuel Fabrication and Characterization

Oak Ridge National Laboratory (ORNL) has participated in the Advanced Gas Reactor (AGR) program for the Next Generation Nuclear Plant (NGNP) and the transition to the Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) Technology Development Office (TDO) since the inception of the program. In this effort, ORNL has fabricated tristructural isotropic (TRISO) fuel and fuel compacts for AGR-1, AGR-2, and AGR-3/4 experiments, provided matrix only components for AGC-2, AGC-4, and AGR 3/4 experiments and for university studies, and provided characterization, coating, and compaction support for INL and BWX Technologies Nuclear Operations Group (BWXT). In fiscal year (FY) 2017, the activities for the INL ART TDO fuel development are focused on preparations for the AGR-5/6/7 experiments and technical support to the Generation IV (GenIV) pre-irradiation leach-burn-leach round robin effort with China and South Korea. With the successful transfer of fuel fabrication technology to BWXT, ORNL’s involvement has shifted toward transferring fuel characterization methods, providing technical support, verification of BWXT laboratory results, and fabricating surrogate fuel materials for university collaborations. ORNL will also provide analytical support to BWXT and INL for uranium dispersion analysis and confirmatory fuel characterization analyses as needed.

Support for Fuel Fabrication and Characterization

Oak Ridge National Laboratory (ORNL) has participated in the Advanced Gas Reactor (AGR) program for the Next Generation Nuclear Plant (NGNP) and the transition to the Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) Technology Development Office (TDO) since the inception of the program. In this effort, ORNL has fabricated tristructural isotropic (TRISO) fuel and fuel compacts for AGR-1, AGR-2, and AGR-3/4 experiments, provided matrix only components for AGC-2, AGC-4, and AGR 3/4 experiments and for university studies, and provided characterization, coating, and compaction support for INL and BWX Technologies Nuclear Operations Group (BWXT). In fiscal year (FY) 2018, the activities for the INL ART TDO fuel development are focused on final characterization to support the AGR-5/6/7 experiments and technical support to the Generation-IV International Forum (GIF) pre-irradiation leach-burn-leach (LBL) round robin effort with China and South Korea.

AGR-5/6/7 Fuel Fabrication Feedstock Chemical Purity Specifications

This specification has been prepared to provide fuel fabricators with minimum chemical purity specifications intended to ensure that fuel components have consistent properties and that impurity levels in the as-fabricated fuel is sufficiently low to minimize in-pile TRISO particle failures and to enhance fission product retention. Process chemical purity specifications are given for the following processes: • Kernel forming and sintering • TRISO particle coating • Matrix preparation • TRISO overcoating • Compact carbonization and heat treatment

Support for INL ART TDO Advanced Gas Reactor Fuel Fabrication and Characterization

Oak Ridge National Laboratory (ORNL) has participated in the Advanced Gas Reactor (AGR) program for the Next Generation Nuclear Plant (NGNP) since the inception of the program. In this effort, ORNL has fabricated tri-structural isotropic (TRISO) fuel and fuel compacts for AGR-1, AGR-2, and AGR-3/4 experiments, provided matrix only components for AGC-2 and AGR-3/4 experiments and for university studies, and provided characterization, coating, and compaction support for Babcock and Wilcox (B&W). In FY 2017, the activities for the Very-High Temperature Reactor (VHTR) Technology Development Office (TDO) fuel development are focused on preparations for the AGR-5/6/7 experiments. With the successful transfer of fuel fabrication technology to B&W, ORNL’s involvement has shifted toward transferring fuel characterization methods, providing support for fabricating surrogate fuel materials international collaborations. ORNL will also provide analytical support to B&W for uranium dispersion analysis and confirmatory fuel characterization analyses as needed.