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At least 703 records · Page 39
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Microstructure and mechanical properties of FeCrAl alloys under heavy ion irradiations

FeCrAl ferritic alloys are excellent cladding candidates for accident tolerant fuel systems due to their high resistance to oxidation as a result of formation of a protective Al 2 O 3 scale at high temperatures in steam. In this study, we report the irradiation response of the 10Cr and 13Cr FeCrAl cladding tubes under Fe 2+ ion irradiation up to ~16 dpa at 300 °C. Dislocation loop size, density and characteristics were determined using both two beam bright field transmission electron microscopy and on-zone scanning transmission electron microscopy techniques. 10Cr (C06M2) tube has a lower dislocation density, larger grain size and a slightly weaker texture compared to the 13Cr (C36M3) tube before irradiation. After irradiation to 0.7 dpa and 16 dpa, the fraction of <100> type sessile dislocations decreases with increasing Cr amount in the alloys. It has been found that there is neither void formation nor α' precipitation as a result of ion irradiations in either alloy. Therefore, dislocation loops were determined to be the only irradiation induced defects contributing to the hardening. Nanoindentation testing before the irradiation revealed that the average nanohardness of the C36M3 tube is higher than that of the C06M2 tube. The average nanohardness of irradiated tube samples saturated at 1.6-2.0 GPa hardening for both tubes between ~3.4 dpa and ~16 dpa. The hardening calculated based on transmission electron microscopy was found to be consistent with nanohardness measurements.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Postirradiation examination from separate effects irradiation testing of uranium nitride kernels and coated particles

An overview of postirradiation examination results for uranium nitride kernels and uranium nitride coated particles irradiated in the High Flux Isotope Reactor are presented. This is the first postirradiation examination of the MiniFuel irradiation vehicle that was recently developed to rapidly accumulate burnup during separate effects irradiation testing. In general, the burnup and fuel temperatures measured postirradiation were consistent with the design calculations. The burnup measured by mass spectrometry ranged from 5.9 to 10 MWd/kgU and was achieved after only 68 effective full-power days of irradiation. The dilatometric evaluation of passive silicon carbide thermometry indicated that the fuel was irradiated at temperatures ranging from 410 to 460 °C. Because the irradiation temperatures and burnup were low, the UN kernels showed minimal fission gas release that was within the range of the expected recoil (athermal) release. While it is possible to measure fuel swelling using x-ray computed tomography, the observed swelling was too small to quantify in this case. Extensive microstructural characterization of the irradiated fuel was performed in this study, and no significant irradiation induced changes were observed.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Thermal properties of U-Mo alloys irradiated under high fission power density

A variety of physical and thermal property measurements made on irradiated U-Mo alloy monolithic fuel samples with a Zr diffusion barrier and clad in Al alloy 6061 are reported. These properties form the basis for additional investigations in thermal behavior of U-Mo fuel under high performance research reactor irradiation conditions. Measurements were performed on samples harvested from fuel plates irradiated under high fission power density and surface heat flux and complement previous measurements on samples subjected to moderate fission power density and surface heat flux. Specific heat capacity of irradiated monolithic U-Mo increases with increasing temperature and fission density, but carries large uncertainties associated with thickness measurements used to determine the mass fraction of the layer constituents for extraction of the U-Mo specific heat capacity, specifically the very thin and nonuniform Zr diffusion barrier. Density of irradiated monolithic U-Mo decreases with increasing fission density resulting from fission gas swelling and porosity formation. Thermal diffusivity of irradiated monolithic U-Mo increased with increasing temperature and decreased with increasing fission density. Thermal conductivity of irradiated monolithic U-Mo increased with increasing temperature, and in general, appeared to decrease and become less sensitive to temperature with increasing fission density. Thermal conductivity measurements compared to a semi-empirical model agreed within ±30%. It is suspected that the main driver in deviation between model calculations and measurements is the fission gas swelling determination, which does not account for fission power density or irradiation temperature.

uranium (U), molybdenum, nuclear fuel, thermal con↗
DOE OSTI2021

Irradiation stability and thermomechanical properties of 3D-printed SiC

Neutron irradiation tests were carried out on 3D-printed SiC derived from binderjet additive manufacturing and chemical vapor infiltration. Irradiation was carried to 2.3 dpa over a temperature range of 400–850 °C. Anisotropy that had been observed in the thermal conductivity of 3D-printed SiC prior to irradiation vanished after irradiation as the irradiation defect thermal resistivity accumulated in the material. No degradation in strength was observed in the material before or after irradiation, at various temperatures, or in different orientations. Electron microscopy of the microstructure after neutron irradiation showed distinct defect morphologies in the heterogenous material, but no evidence for irradiation-induced cracking or degradation in the microstructure was observed.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Comparison of unirradiated and irradiated AGR-2 TRISO fuel particle oxidation response

The silicon carbide (SiC) coating in a tristructural isotropic (TRISO) particle acts as a barrier to fission product release during reactor operation and accident scenarios. Oxidation and subsequent failure of the SiC layer during a rare air ingress event is a proposed mechanism for fission product release in a high-temperature gas-cooled reactor (HTGR). Although previous oxidation studies have analyzed unirradiated TRISO particle response, this study compared the oxidation behavior of irradiated and unirradiated TRISO particles from the second Advanced Gas Reactor Fuel Development and Qualification Program irradiation experiment (AGR-2). Particles with exposed SiC were subjected to six varying oxidizing tests in the Furnace for Irradiated TRISO Testing (FITT), examined for failure fraction with the Irradiated Microsphere Gamma Analyzer (IMGA) and characterized with focused ion beam and scanning/transmission electron microscopy techniques to analyze the oxide layer. Uncorrelated unirradiated particle failures throughout the series of exposures suggests that external factors inherent to the experiment increased particle failure sensitivity. However, irradiated particle observations indicated an increased failure response at 400 h 1400 °C in both 2% and 21% O 2 atmospheres above failure associated with external factors. Oxide thickness measurements after 400 h at 1400 °C revealed a greater oxidation rate than predicted by parabolic growth, which was attributed to the increased complexity of the oxide structure at longer exposure times. Altering the atmosphere from 21% to 2% O 2 reduced the average oxide thickness by approximately 12%–14% in both irradiated and unirradiated particles at 400 h 1400 °C. Altogether, the minor variations observed between irradiated and unirradiated particles in this study led to the conclusion that unirradiated TRISO particles can be used to approximate irradiated TRISO oxidation kinetics.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Swelling and Fission Gas Release of U-10Mo and U-17Mo Following Neutron Irradiation at 250 – 500°C

The performance of metallic fuel alloys, U-10Mo and U-17Mo, was examined using the MiniFuel test system in the High Flux Isotope Reactor. Approximately 0.8 mm thick disks were irradiated at target temperatures of 250°C, 350°C, 450°C, and 500°C up to three different fission densities, culminating at a maximum fission density of 6.8 × 10 20 cm −3 . Fission rates decayed from 3 – 6 × 10 13 cm −3 s −1 to 2 – 4 × 10 13 cm −3 s −1 over the course of the longest, eight cycle, irradiation as the 235 U was consumed and the 239 Pu concentration accumulated. Actual irradiation temperature on the last day of irradiation was measured via dilatometry using the SiC passive thermometry recovered from the MiniFuel subcapsules and compared favorably with the thermal calculations using as-built geometry and test conditions. Furthermore, average simulated temperatures were within 50°C of the target temperatures, except for the 500°C irradiation, for which the temperature variation was 62°C, 32°C, and 90°C in the in two, four, and eight cycle irradiations, respectively. Fission gas release (FGR) measurements showed no release above recoil for any U-17Mo fuels or for the U-10Mo fuel irradiated at 250°C. Finally, significant (40%–80%) FGR was found for the medium- to highest-burnup U-10Mo samples irradiated at target temperatures 350°C–500°C. Significant FGR correlated with sample thickness swelling, which was as high as 13%–35% for high-release samples and below 7% for all other (low–gas release) samples.

FGR↗
DOE OSTI2025

Microstructure, stored energy, and stability of H/He-filled nanocavities in low temperature irradiated Inconel 718

The microstructure, trapped transmutation gases, stored energy, and mechanical behavior of samples from an irradiated Inconel 718 proton beam window were characterized using transmission electron microcopy, thermal desorption spectrometry (TDS), differential scanning calorimetry (DSC), and tensile testing. In the as-irradiated condition the microstructure contained a high number density of 1–3 nm gas-filled nanocavities. Emissions of trapped gases, H and He, during TDS correlated with peaks of the energy release curves from DSC examinations, which suggest these gases were likely stored in highly stable defect traps. The stored energy from radiation damage saturated at doses of a few dpa and did not increase with increasing radiation dose, but the amount of stored H and He increased with increasing dose. Effects of post-irradiation annealing were studied as well. After exposure to 700 °C, the nanocavities grew only slightly to 2–4 nm in diameter, but after exposure to 900 °C, the cavities grew to 10–20 nm in diameter and electron energy-loss spectroscopy showed these cavities contained a core of He surrounded by a shell of H. Further, this study demonstrated that the irradiation defect structures containing H and He were remarkably stable during irradiation and after exposure up to 700 °C. The effect of the irradiation temperature, defect mobility, and interaction of H, He, and irradiation defects on mechanical behavior provides insight into the processes responsible for the unusual recovery in ductility with increasing radiation dose observed in Inconel 718 after high energy proton and spallation neutron irradiation.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Failure investigation and mitigation after experimental research reactor fuel plate deformation in an irradiation device

Experimental research reactor fuel testing is conducted in the Belgian Reactor 2 (BR2) of the Belgian Nuclear Research Centre (SCK CEN) in dedicated irradiation vehicles or rigs. One such vehicle allows flat full-size fuel plates to be irradiated by inserting them into slotted baskets that captures a narrow portion of the longitudinal edges of the plates. The motion of the fuel plates within the baskets is possible within the narrow slots and thus, the plate is considered to be unattached. The design intentionally omits fixing mechanisms of the fuel plates to the baskets to facilitate the inspection and repositioning of the plates between the irradiation cycles and the accommodation of thermal expansion of the plates in the lateral direction. However, loosely inserted fuel plates have weak structural boundary conditions allowing for larger out-of-plane deflections caused by hydrodynamic loads exerted by the flowing coolant, as compared to those of fixed plates. Unexpected large deformations of plates occurred in several irradiation cycles that further resulted in a loss of cladding integrity. These deformations could not be attributed to a single source. This triggered a series of thermal hydraulic, structural, and fluid-structure interaction analyses aiming at understanding the observed phenomenon. The analyses revealed that, for a certain combination of unfavorable manufacturing and assembly tolerances, fuel plate edges could escape out of the slots in the irradiation basket due to the hydrodynamic load. Subsequently, the plate could become wedged inside the basket coolant channel opening. This resulted in reduced coolant flow and accelerated temperature increase and thermal expansion of the plate while under irradiation. This unfavorable feedback loop could then lead to excessive plate surface temperatures, deformed plates and cladding failure, as was observed in the experiments. These analyses not only provided a probable cause of the fuel plate failures, but also resulted in a new and improved design of the irradiation basket to avoid these issues in the future. In conclusion, a series of recent successful irradiations confirm that the sources of failures were identified correctly, and the implemented mitigations were adequate.

BR2↗
DOE OSTI2018

Understanding of a Novel Irradiation-induced Nanostructuring Process in SiC layer of TRISO fuel particles via Transmission Electron Microscopy

Metallic fission products are found to transport through the SiC layer of the tristructural isotropic (TRISO) nuclear fuel particle and various studies have been undertaken to understand the transport behavior [1, 2]. While the grain boundaries in SiC acts as a primary pathway to dissipate fission products in the form of multi-phased complex compounds, intragranular fission product precipitation has always been observed. Research on neutron irradiation damage tolerance of silicon carbide (SiC) and its composites, used in TRISO fuel, light water reactors (LWR) fuel cladding etc., has traditionally ignored the irradiation-induced “polymorphism” of SiC. In the post-irradiation advanced microscopy effort at Idaho National Laboratory, it has been recently discovered that the intragranular precipitation of fission products in SiC layer of TRISO coated fuels takes place via a dual step nucleation mechanism that involves cubic (?) ?hexagonal (?) SiC polymorphic transition, and subsequent metamorphosis of ?-SiC into fission product precipitates, as shown in Figure 1 [3]. This precipitation behavior cannot be explained by any reported nucleation mechanism. The extent of this precipitation has been observed to be influenced by the nature of the neutron irradiation induced damage structures. These damage structures (e.g., defect morphology and density) are also indicative of the irradiation temperature and neutron fluence levels experimented by the individual TRISO particles. The distinction of neutron irradiation damage structures as a function of the fuel kernel type, burn up level and fabrication histories of TRISO coated particles from the Advanced Gas reactor (AGR-1 and AGR-2) experiments has been examined and compared. This discovery not only adds new fundamental understanding of physical behavior of polymorphic ceramics under irradiation, but also provides the following beneficial traits for high temperature applications. 1) Since the extent of precipitation of the desired phase is directly dependent on the amount of precursor phase, the volume fraction of desired phase can be easily controlled by irradiation parameters. This eliminates the conventional limitation of volume fraction of a desired phase in the system as outlined by the phase diagram of the specific system. 2) Most of the precipitate-strengthened materials for high temperature applications suffer from degradation of size and shape of precipitate upon prolonged thermal exposures. But, precipitates are confined to the shape and size of the morphological template of surrogate phase (e.g. a Pd silicide precipitate is confined by the ?-SiC as shown in Figure 1(c)) in the present nucleation pathway. Hence, there is no possibility of desired phase to undergo any morphological degradation at high temperature that eventually will extend the lifetime of these materials at high temperature.

42 ENGINEERING↗
DOE OSTI2022

Ultra-high dose rate FLASH irradiator at the radiological research accelerator facility

The Radiological Research Accelerator Facility has modified a decommissioned Varian Clinac to deliver ultra-high dose rates: operating in 9 MeV electron mode ( FLASH mode), samples can be irradiated at a Source-Surface Distance (SSD) of 20 cm at average dose rates of up to 600 Gy/s (3.3 Gy per 0.13 µs pulse, 180 pulses per second). In this mode multiple pulses are required for most irradiations. By modulating pulse repetition rate and irradiating at SSD = 171 cm, dose rates below 1 Gy/min can be achieved, allowing comparison of FLASH and conventional irradiations with the same beam. Operating in 6 MV photon mode, with the conversion target removed ( SuperFLASH mode), samples are irradiated at higher dose rates (0.2–150 Gy per 5 µs pulse, 360 pulses per second) and most irradiations can be performed with a single very high dose rate pulse. In both modes we have seen the expected inverse relation between dose rate and irradiated area, with the highest dose rates obtained for beams with a FWHM of about 2 cm and ± 10% uniformity over 1 cm diameter. As an example of operation of the ultra-high dose rate FLASH irradiator, we present dose rate dependence of dicentric chromosome yields.

43 PARTICLE ACCELERATORS↗
DOE OSTI2018

Total dose effect of Al 2 O 3 -based metal–oxide–semiconductor structures and its mechanism under gamma-ray irradiation

In our work, insights into the total dose response and native point defect behavior in the Al2O3 gate dielectric during irradiation were gained by gamma-ray irradiation experiments and first-principles calculations. It is found that the O vacancy (VO) can act as a hole trap in the Al2O3 gate dielectric during irradiation, leading to the negative shift of the capacitance–voltage (C–V) curves of the Al2O3-based metal–oxide–semiconductor (MOS) structure. Our calculations show that the neutral defect VO becomes a +2 charged center after irradiation, and the positively charged VO is a kind of conductive path for electrons, which contributes to an increase of the leakage current in the irradiated MOS capacitors. Additionally, the trapped holes are accumulated with irradiation doses, which can lower the barrier height of the Al2O3 gate oxide and further cause the increase of the leakage current. The other native point defects in the Al2O3 layer, such as aluminum vacancy (VAl), aluminum interstitial (Ali) and oxygen interstitial (Oi), only act as fixed charge centers during irradiation. Net negative charges existing in the Al2O3 layers before irradiation are mainly induced by the negatively charged defects of VAl and Oi.

Engineering↗
DOE OSTI2020

Irradiation of the kidneys causes pathologic remodeling in the nontargeted heart: A role for the immune system

Cardiac disease is a frequent and significant adverse event associated with radiotherapy for cancer. Identifying the underlying mechanism responsible for radiation injury to the heart will allow interventions to be developed. In the present study, we tested if local kidney irradiation results in remodeling of the shielded, nontargeted heart. One kidney, two kidneys, or the total body of male WAG and Dahl SS rats were irradiated with 10 Gy of X-rays. Local kidney irradiation resulted in systemic hypertension, increased BUN, infiltration of T lymphocytes, natural killer cells, and macrophages into the renal cortex and medulla, and renal fibrosis. Local irradiation of kidneys in WAG rats resulted in remodeling in the nontargeted heart after 120 days, manifested by perivascular fibrosis and increased interventricular septal thickness, but was not seen in Dahl SS rats due to a high baseline level of fibrosis in the sham-irradiated animals. Genetic depletion of T cells mitigated the nephropathy after local kidney irradiation, indicating a role for the immune system in mediating this outcome. Local kidney irradiation resulted in a cascade of pro-inflammatory cytokines and low-molecular weight metabolites into the circulation associated with transmission of signals resulting in pathologic remodeling in the nontargeted heart. A new model is proposed whereby radiation-induced cardiac remodeling in susceptible animals is indirect, with lower hemi body organs such as the kidney exporting factors into the circulation that cause remodeling outside of the irradiated field in the shielded, nontargeted heart. This nontargeted effect appears to be mediated, in part, by the immune system.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2023

Impacts of irradiation-induced nanostructure on phonon linewidths and thermal conductivity in U-Zr alloy

Ion irradiation can significantly change material properties. Thermal conductivity degradation of reactor fuel is detrimental to safety and efficiency—driving research on how irradiation impacts heat-carrying phonons. Here, we report the measurements of phonon dispersion and lifetimes (inverse linewidths) in freestanding irradiated crystals and reveal the relationship between thermal conductivity and proton irradiation induced nanostructures in UZr 2.6 using inelastic x-ray scattering, electron microscopy, modulated thermoreflectance, and first-principles calculations. Irradiation results in more significant increases in phonon linewidths and decreased thermal conductivity in the basal plane than that along the c axis, decreasing thermal conductivity anisotropy. Furthermore, microscopy reveals that this results from irradiation induced nanoprecipitates and metastable phase separation. Surprisingly, irradiation leads to smaller phonon linewidths near some zone boundaries, which we attribute to irradiation induced short-range order.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Spectral Effects in Albedo and Rearside Irradiance Measurement for Bifacial Performance Estimation

We investigate the impact of spectral dependence of ground surface reflectivity on albedo and rearside irradiance measurements necessary for bifacial photovoltaic (PV) module performance estimation and monitoring. Because PV modules are spectrally selective, albedo and irradiance measurements performed with common irradiance sensors may require spectral mismatch corrections when used for performance prediction. We investigate via simulation the differences in spectrally responsive albedo measured with thermopile pyranometers and crystalline silicon PV reference cells in comparison to a typical crystalline-silicon bifacial PV module. Simulations are performed for nine different representative ground surface materials using simulated solar spectra together with spectral reflectivity data distributed with the SMARTS simulation software. For the materials considered, the results show that albedo spectral mismatch relative to the bifacial module is distributed over a range of ±9.2% for thermopile pyranometers versus only ±3.7% for a typical PV reference cell. We consider the impact of this spectrally-responsive albedo mismatch on bifacial PV module rearside irradiance measurements. Using synthesized rearside spectral irradiance distributions, we find that for the nine different ground surface materials the predicted rearside irradiance measurement deviates from the effective irradiance observed by the PV module by on the order of 16.5 W/m 2 for the pyranometer and 3.6 W/m 2 for the PV reference cell. We discuss the implications for bifacial albedo and irradiance measurement.

41 EE - Solar Energy Technologies Office (EE-4S)↗
DOE OSTI2016

Modeling Changes in Measured Conductance of Thin Boron Carbide Semiconducting Films Under Irradiation

Semiconducting, p-type, amorphous partially dehydrogenated boron carbide films (a-B 10 C 2+x :H y ) were deposited utilizing plasma enhanced chemical vapor deposition (PECVD) onto n-type silicon thus creating a heterojunction diode. A model was developed for the conductance of the device as a function of perturbation frequency (f) that incorporates changes of the electrical properties for both the a-B 10 C 2+x :H y film and the silicon substrate when irradiated. The virgin model has 3 independent variables (R1, C1, R3), and 1 dependent variable (f). These samples were then irradiated with 200 keV He + ions, and the conductance model was matched to the measured data. It was found that initial irradiation (0.1 displacements per atom (dpa) equivalent) resulted in a decrease in the parallel junction resistance parameter from 6032 Ω to 2705 Ω. Further irradiation drastically increased the parallel junction resistance parameter to 39000 Ω (0.2 dpa equivalent), 77440 Ω (0.3 dpa equivalent), and 190000 Ω (0.5 dpa equivalent). It is believed that the initial irradiation causes type inversion of the silicon substrate changing the original junction from a p-n to a p-p+ with a much lower barrier height leading to a lower junction resistance component between the a-B 10 C 2+x :H y and irradiated silicon. In addition, it was found that after irradiation, a second parallel resistor and capacitor component is required for the model, introducing 2 additional independent variables (R2, C2). This is interpreted as the junction between the irradiated and virgin silicon near ion end of range.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Mechanical response of HFIR-irradiated M5FRAMATOME cladding under simple and complex loading conditions

This study investigates the mechanical behavior of High Flux Isotope Reactor (HFIR) irradiated M5FRAMATOME cladding under simple and complex loading conditions through axial tensile and reversible cyclic bending. Axial tension specimens were pre-machined prior to HFIR irradiation while cyclic bend specimens were inserted as intact tubes. Tests articles were neutron-irradiated to 4 and 16 dpa, and specimens were tested at ORNL's hotcell facilities. The axial tension tests were conducted under constant displacement control, and the reversible cyclic bend tests were performed using ORNL's Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) apparatus. Results showed that mechanical response of Cr-coated and uncoated M5FRAMATOME cladding were similar and independent of irradiation dose for axial tension tests, while Cr-coated specimens’ reversible cyclic bend behavior differed from uncoated counterparts. For all tests, irradiation temperature showed a significant impact on the mechanical behavior. Below 280°C, all axial tensile specimens whether coated or not behaved similarly. Above 280°C, YS and UTS showed decrease with increasing irradiation temperature. A similar behavior was also observed in cyclic bend tests as well. The mechanical damage during cyclic bend tests was linked to damage accumulation in unirradiated Cr-coated zircaloy-4 specimens, and the effect of irradiation temperature was related to changing characteristics of defect mobility during high temperature irradiation.

Cinbiz, Nedim [ORNL] (ORCID:0000000346268515)↗
DOE OSTI2016

Irradiation at Different Fetal Stages Results in Different Translocation Frequencies in Adult Mouse Thyroid Cells

While it is generally believed that fetuses are at high risk of developing cancers, including leukemia, after low doses of radiation, it has been reported that atomic bomb survivors exposed in utero did not show a dose response for translocations in blood T lymphocytes when they were examined at approximately 40 years of age. Subsequent mouse studies confirmed that animals irradiated during the fetal stage did not show evidence of radiation effects in lymphocytes and bone marrow cells when they were examined after reaching adulthood. However, in a study of rat mammary epithelial cells, radiation effects were clearly observed after fetal irradiation. These results indicate that the fate of chromosome aberrations induced in a fetus could vary among different tissues. Here we report on translocation frequencies in mouse thyroid cells, which were irradiated at different stages of fetal development. Cytogenetic examination was then conducted using fluorescence in situ hybridization (FISH) painting of chromosomes 1 and 3. Adult mice, 2 Gy X-ray irradiated at 15.5-day-old fetuses (E15.5), showed a higher translocation frequency (30/1,155 or 25.3 x 10 -3 ) than nonirradiated adult controls (0/1,007 or 0.1 x 10 -3 ), and was near that experienced by irradiated mothers and non-pregnant adult females (43/1,244 or 33.7 x 10 -3 ). These results are consistent with those seen in rat mammary cells. However, when fetuses were irradiated at an earlier stage of development (E6.5) before thyroid organogenesis, the resulting observed translocation frequency was much lower (3/502 or 5.8 x 10 -3 ) than that in E15.5 mice. These results suggest that after fetal irradiation, tissue stem cells record radiation effects primarily when the exposure occurs in cells that have been integrated into tissue. Embryonic stem cells that have been damaged prior to integration into the niche may undergo negative selection due to apoptosis, mitotic death or stem cell-niche cell interactions. The implications of these results in interpreting cancer risks after fetal irradiation are also discussed.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2016

AGC-2 Irradiation Report

The Next Generation Nuclear Plant (NGNP) will be a helium-cooled, very high temperature reactor (VHTR) with a large graphite core. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor (HTGR) designs.[ , ] Nuclear graphite H 451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphites have been developed and are considered suitable candidates for the new NGNP reactor design. To support the design and licensing of NGNP core components within a commercial reactor, a complete properties database must be developed for these current grades of graphite. Quantitative data on in service material performance are required for the physical, mechanical, and thermal properties of each graphite grade with a specific emphasis on data related to the life limiting effects of irradiation creep on key physical properties of the NGNP candidate graphites. Based on experience with previous graphite core components, the phenomenon of irradiation induced creep within the graphite has been shown to be critical to the total useful lifetime of graphite components. Irradiation induced creep occurs under the simultaneous application of high temperatures, neutron irradiation, and applied stresses within the graphite components. Significant internal stresses within the graphite components can result from a second phenomenon—irradiation induced dimensional change. In this case, the graphite physically changes i.e., first shrinking and then expanding with increasing neutron dose. This disparity in material volume change can induce significant internal stresses within graphite components. Irradiation induced creep relaxes these large internal stresses, thus reducing the risk of crack formation and component failure. Obviously, higher irradiation creep levels tend to relieve more internal stress, thus allowing the components longer useful lifetimes within the core. Determining the irradiation creep rates of nuclear grade graphites is critical for determining the useful lifetime of graphite components and is a major component of the Advanced Graphite Creep (AGC) experiment.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗