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DOE OSTI2024

AGC-4 Specimen Post-Irradiation Examination Data Interim Report

This interim report documents preliminary results of the post-irradiation examination material property testing from the fourth advanced graphite creep (AGC), AGC 4, capsule specimens. This is the fourth of a series of six irradiation test trains planned as part of the AGC experiment to fully characterize the neutron irradiation effects and radiation creep behavior of current nuclear graphite grades to moderate dose levels (=7 dpa). The AGC 4 capsule was irradiated in the Idaho National Laboratory Advanced Test Reactor at a nominal temperature of 800°C and to a peak dose of 8 dpa. Half of the AGC-4 specimens were subjected to compressive stresses to induce irradiation creep. Post-irradiation testing and measurement results are reported with the exception of thermal testing, which is still in progress, and irradiation mechanical strength testing. Additionally, some specimens initially deemed too hot to be examined in the ART Graphite laboratory may still be measured. The data reported includes specimen dimensions for both stressed and unstressed specimens to establish the irradiation creep rates, mass and dimensional data necessary to derive density, elastic constants (Young?s modulus, shear modulus, and Poisson?s ratio) from ultrasonic time of flight velocity measurements, Young?s modulus from the fundamental frequency of vibration, and electrical resistivity. A more complete evaluation of trends in the material property changes, as well as irradiation-induced creep due to the irradiation environment and applied load on the specimens, will be discussed later in AGC 4 post-irradiation examination analysis reports.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

AGC-4 Specimen Post-Irradiation Examination Data Interim Report

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Indirectly probing the structural change in ion-irradiated Zr-Based metallic glasses from small scale mechanical tests

Ion irradiation was applied to tailor the structural heterogeneities in Zr-based metallic glasses at room temperature. Experimental methods of X-ray diffraction, nanoindentation, and micropillar compression were conducted to examine the irradiation effects on their structural and mechanical property changes. It is found that the irradiated materials retained amorphous structure after room-temperature Ni ion irradiation. The reduction of elastic modulus and hardness measured by nanoindentation indicated the irradiation-induced mechanical degradation. A unified statistic model was employed to quantitatively predict the density and strength of irradiation defects, although their specific structural and physical nature is not explicitly included in this model. The transition of non-intersecting shear bands to multiple intersecting shear bands was observed on compression tests of irradiated micropillars with the increase of irradiation dose. Furthermore, the analysis of the displacement excursion of micropillar compression tests indicated a different deformation mode from the unirradiated state. These results from nanoindentation pop-in and micro-pillar compression tests suggested that irradiation eventually leads to a new state with different types and characteristics of structural heterogeneities from violent displacement cascades and non-equilibrium energy deposition/dissipation processes, which also proves ion irradiation as an effective method to tune the structure and mechanical properties of metallic glasses.

36 MATERIALS SCIENCE↗

DOE OSTI2022

A concurrent irradiation-mechanics multiscale coupling model

The vast majority of our current knowledge regarding the basic mechanisms controlling irradiation effect on mechanical properties is based almost entirely on the results of post-irradiation experiments or theoretical models. However, the concurrent effects of irradiation, mechanical stress, and thermal damage on the failure phenomena of materials and components remain largely unexplored due to its internal multiscale-multiphysics coupling nature. Here, we present here a concurrent irradiation-mechanics multiscale coupling model. The concurrent evolutions of nanoscale irradiation defect clusters, microscale dislocation configurations, and mechanical responses are well captured based on coupling cluster dynamics, discrete dislocation dynamics, and the finite element methods using an effective time marching scheme. Model predictions of defect densities and size are in general agreement with experimental observations. Irradiation hardening is shown to take place also in samples undergoing concurrent irradiation-mechanical loading, similar to samples tested post-irradiation. However, the occurrence of plastic flow localization and dislocation channel formation is not accompanied with apparent yield drop (softening) under concurrent irradiation-mechanical loading conditions, which is different from the post-irradiation case.

42 ENGINEERING↗

DOE OSTI2017

Characterization of faulted dislocation loops and cavities in ion irradiated alloy 800H

Alloy 800H is a high nickel austenitic stainless steel with good high temperature mechanical properties which is considered for use in current and advanced nuclear reactor designs. The irradiation response of 800H was examined by characterizing samples that had been bulk ion irradiated at the Michigan Ion Beam Laboratory with 5 MeV Fe 2+ ions to 1, 10, and 20 dpa at 440 °C. Transmission electron microscopy was used to measure the size and density of both {111} faulted dislocation loops and cavities as functions of depth from the irradiated surface. The faulted loop density increased with dose from 1 dpa up to 10 dpa where it saturated and remained approximately the same until 20 dpa. The faulted loop average diameter decreased between 1 dpa and 10 dpa and again remained approximately constant from 10 dpa to 20 dpa. Cavities were observed after irradiation doses of 10 and 20 dpa, but not after 1 dpa. The average diameter of cavities increased with dose from 10 to 20 dpa, with a corresponding small decrease in density. Cavity denuded zones were observed near the irradiated surface and near the ion implantation peak. To further understand the microstructural evolution of this alloy, FIB lift-out samples from material irradiated in bulk to 1 and 10 dpa were re-irradiated in-situ in their thin-foil geometry with 1 MeV Kr 2+ ions at 440 °C at the Intermediate Voltage Electron Microscope. We observed that the cavities formed during bulk irradiation shrank under thin-foil irradiation in-situ while dislocation loops were observed to grow and incorporate into the dislocation network. The thin-foil geometry used for in-situ irradiation is believed to cause the cavities to shrink.

36 MATERIALS SCIENCE↗

DOE OSTI2021

Impact of neutron irradiation on the thermophysical properties of additively manufactured stainless steel and inconel

The deployment of additive manufacturing (AM) technologies to support the nuclear energy industry is limited by the current lack of data related to the irradiation performance of AM parts. The Colorado School of Mines is conducting a multi-year irradiation and post irradiation examination project to study the impacts of neutron irradiation on stainless steel 316L and Inconel 718 and 625 specimens produced by several commercially available AM techniques and irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). This paper presents initial results on the impact of neutron irradiation on the heat capacity, the thermal diffusivity, and the thermal conductivity of the specimens at temperatures between 100 oC and 400 oC. The unirradiated AM specimens have heat capacity values similar to, or lower than, the reference values, regardless of manufacturing method or material. The irradiated AM specimens have average heat capacities greater than the unirradiated specimens and reference values in every case. Additionally, the thermal diffusivities of the unirradiated AM specimens are either larger or statistically similar to the reference values. The irradiated AM specimens exhibit lower thermal diffusivities compared to the reference thermal diffusivity values. One of the stainless steel 316L laser powder bed fusion samples exhibited thermal diffusivity values with statistically significant anisotropy prior to and after irradiation. The thermal conductivities of the unirradiated AM stainless steel 316L specimens and the unirradiated Inconel 718 specimens are lower or statistically similar to the reference values. The unirradiated Inconel 625 specimens exhibit slightly higher thermal conductivity values than the reference values. The irradiated specimens show a significant increase in thermal conductivity compared to the unirradiated specimens and the reference values for all materials.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

The effect of neutron irradiation on conventional and nanocrystalline nickel: NSUF Work Package UA-22PN080501 Milestone M3UA-22PN0805012 (Final Report)

Even though nanocrystalline materials (20-100 nm) present an unprecedented potential, scientific knowledge related to the effect of neutron irradiation on the mechanical properties and microstructure is still scarce. Most of the past studies were conducted using ion irradiation which may not have the same effect as neutron irradiation because of the smaller irradiation volume and the higher dose rate. To reach a firm conclusion on the potential of nanocrystalline materials for nuclear reactor applications, extensive study of model metals with different stacking fault energy (SFE) is required to elucidate their behavior in radiation environments. Nanocrystalline copper and nickel are typically chosen because they are commonly used as model FCC metals in studies of radiation effects. Nickel is an FCC metal with a high stacking-fault energy (~125 mJ/m 2 ) compared to copper (~45 mJ/m 2 ). Hence, microcrystalline and nanocrystalline nickel samples were irradiated in the INL’s Advanced Test Reactor (ATR), as a part of FY08 North Carolina State University NSUF Irradiation Experiment# 96 to evaluate the irradiation behavior of these materials. The objective of this FY20 NSUF project# 19122 is to perform PIE (at PNNL NSUF facility) on previously ATR-neutron irradiated (1.2 and 2.6 dpa; 80-89°C) nanocrystalline and microcrystalline nickel samples to investigate the changes in mechanical properties and microstructures and evaluate whether nanocrystalline nickel is relatively more radiation resistant compared to conventional microcrystalline nickel. To perform PIE at PNNL, sixteen neutron irradiated specimens (microcrystalline and nanocrystalline) were transferred from the NSUF Nuclear Fuels and Materials Library at INL. Experimental techniques such as SEM/EBSD, XRD, TEM, Vickers microhardness and tensile testing were employed to characterize the effect of neutron irradiation on the microstructure and mechanical properties of nanocrystalline nickel and compared them with corresponding characteristics of microcrystalline nickel.

36 MATERIALS SCIENCE↗

NASA NTRS2023

Investigating the Role of Incident Ion Flux in Solar Wind Space Weathering of Carbon-Rich Asteroidal Regolith Via H+ and He+ Irradiation of the Murchison Meteorite

We present results from a set of low and high flux 1 keV/amu H + and He + irradiation experiments performed on slabs of the Murchison CM2 carbonaceous chondrite. The low flux conditions for H + and He + irradiation were ~1-1.5 orders of magnitude lower than the high flux conditions, and each experiment was irradiated to a total fluence between ~3 × 10 16 to ~6 × 10 16 ions/cm 2 . Irradiation-induced changes in the surface chemistry and optical properties of the Murchison samples were evaluated using in situ X-ray photoelectron spectroscopy (XPS) and visible and near-infrared spectroscopy (VNIR). We characterized the microstructure and composition of ion damaged rims in focused ion beam (FIB) cross-sections extracted from olivine and matrix material in each irradiated Murchison slab using transmission electron microscopy (TEM). XPS results suggest that both low flux and high flux H + and He + irradiation cause minor sputtering of surface carbon as well as a reduction in the valence state of iron, from Fe 3+ to Fe 2+ . Slope bluing is observed in VNIR spectra of the irradiated samples which may reflect carbonization and dehydrogenation of organic species and contrasts with reddening trends associated with npFe 0 formation. Although we do not observe a strong flux-dependence on the crystallinity of ion-damaged olivine, TEM analyses reveal a variety of microstructures in all olivine FIB-sections, suggesting that crystallographic orientation affects amorphization efficiency. Analyses of matrix FIB-sections indicate that phyllosilicate alteration is mainly driven by He+ irradiation, where the higher incident flux leads to greater amorphization and the formation of more distinct ion-damaged layers, similar to smooth layers in returned Ryugu particles. TEM results also provide some evidence that higher ion flux leads to greater vesiculation, with He + irradiation being more efficient at vesiculation than H + irradiation, and that higher ion flux may promote the segregation of Mg and Si into laterally extensive lenses and layers in olivine samples. We discuss the implications of these findings for constraining the role that ion flux plays in the development of space weathering characteristics in silicate phases present in carbonaceous asteroidal regoliths. These results will be important for understanding the complexity of this process and how it operates on carbon-rich airless bodies like asteroids Bennu and Ryugu.

meteorites↗

DOE OSTI2024

Promoting the regulatory acceptance of combined ion and neutron irradiation for material degradation in nuclear reactors

The Advanced Materials and Manufacturing Technologies (AMMT) program within the Department of Energy (DOE) Office of Nuclear Energy has developed its current recommendation for promoting the use of combined ion irradiation and neutron irradiation for the accelerated qualification of materials to be deployed in nuclear reactors. This plan is intended to provide a collaborative path forward that can be adopted by academia, national laboratories, and industry, and has been developed with input from the regulatory research arm of the U.S. Nuclear Regulatory Commission (NRC). To deploy new materials or materials manufactured with new technologies, such as additive manufacturing, materials must be evaluated for reactor-induced degradation from the combination of harsh temperatures, corrosive environments, and radiation fields. However, rapid deployment of materials necessitates accelerated testing methods rather than relying on years of neutron irradiation in a material test reactor. Ion irradiation has demonstrated success in reproducing material microstructure and select property evolution resulting from neutron irradiation with three to four orders of magnitude reduction in time and cost, making it an ideal candidate for accelerated irradiation testing. This presentation provides context governing both the scientific and regulatory aspects of the proposed goal. The discussion is aimed at a broad audience including researchers from industry, national laboratories, and academia. The recommended path forward is presented as a conceptual framework of specific steps. In brief, the strategy entails developing an integrated ion and neutron irradiation test plan for the material property of interest based on the fundamental tenet of the linkage of microstructure and properties in materials. Physics-based modeling interprets ion irradiation data and predicts neutron irradiation microstructure and properties with uncertainty bounds. The first round of testing is sufficient for an initial licensing application using a risk-informed approach, while a minimum required neutron irradiation test plan reduces cost and time requirements. A surveillance program with witness specimens in-reactor provides additional data over time to improve model predictions to higher damage levels and further reduce uncertainty bounds, which can be used for license extensions or longer lifetimes in new license applications.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

Effect of neutron and ion irradiation on the metal matrix and oxide corrosion layer on Zr-1.0Nb cladding alloys

A detailed study has been carried out on recrystallised Zr-1.0Nb alloys corroded and irradiated under different conditions, including ex-autoclave and ex-reactor samples. After 540 days in reactor and damage around 5 dpa, the neutron irradiated sample shows no serious evidence for radiation enhanced corrosion and is still in pre-transition stage. The good corrosion resistance of the neutron irradiated Zr-1.0Nb can be related to the higher volume fraction of tetragonal phase and fewer interconnected nano porosity/cracks in the oxide. This indicates less tetragonal to monoclinic transition, leads to more protective oxide in the neutron irradiated sample, containing little evidence for short circuit paths for the penetration of oxygen or water towards the metal-oxide interface. These observations on a sample with a slow overall oxidation rate are consistent with the hypothesis that interconnected porosity can lead to early transitions and rapid oxidation. Tetragonal oxide can be either stabilised by irradiation, or stabilised by local release of impurity species from SPPs such as dissolution of Fe from Zr-Nb-Fe precipitates or radiation introduced precipitates (RIPs) which is likely to be small β-Nb clusters. The oxide consists of well-aligned columnar-equiaxed microstructure in the autoclave sample while a more complex oxide grain structure was observed in the neutron-irradiated sample. As oxide continues to grow, there are more < a > and < c > loops, dissolved Fe and RIPs in the metal matrix, however, the corrosion rate is low enough for the tetragonal oxide to stabilise and suboxide + Zr(Osat) phases exist for protectiveness, so there is no enhanced corrosion after radiation. In situ ion irradiation in the TEM revealed no visible defect clusters or voids in the oxide, suggesting that radiation damage to the metal matrix rather than oxide may have a stronger effect on corrosion mechanisms after neutron irradiation, however, cascade damages are not visible in this case. Neutron irradiation also seems to have little effect on promoting fast oxidation or dissolution of β-Nb precipitates into the surrounding oxide or metal during irradiation. Furthermore, these results are discussed in the light of the current mechanisms for corrosion of nuclear fuel cladding alloys.

(S)TEM↗

DOE OSTI2015

Indirect Tumor Cell Death After High-Dose Hypofractionated Irradiation: Implications for Stereotactic Body Radiation Therapy and Stereotactic Radiation Surgery

Purpose: The purpose of this study was to reveal the biological mechanisms underlying stereotactic body radiation therapy (SBRT) and stereotactic radiation surgery (SRS). Methods and Materials: FSaII fibrosarcomas grown subcutaneously in the hind limbs of C3H mice were irradiated with 10 to 30 Gy of X rays in a single fraction, and the clonogenic cell survival was determined with in vivo–in vitro excision assay immediately or 2 to 5 days after irradiation. The effects of radiation on the intratumor microenvironment were studied using immunohistochemical methods. Results: After cells were irradiated with 15 or 20 Gy, cell survival in FSaII tumors declined for 2 to 3 days and began to recover thereafter in some but not all tumors. After irradiation with 30 Gy, cell survival declined continuously for 5 days. Cell survival in some tumors 5 days after 20 to 30 Gy irradiation was 2 to 3 logs less than that immediately after irradiation. Irradiation with 20 Gy markedly reduced blood perfusion, upregulated HIF-1α, and increased carbonic anhydrase-9 expression, indicating that irradiation increased tumor hypoxia. In addition, expression of VEGF also increased in the tumor tissue after 20 Gy irradiation, probably due to the increase in HIF-1α activity. Conclusions: Irradiation of FSaII tumors with 15 to 30 Gy in a single dose caused dose-dependent secondary cell death, most likely by causing vascular damage accompanied by deterioration of intratumor microenvironment. Such indirect tumor cell death may play a crucial role in the control of human tumors with SBRT and SRS.

62 RADIOLOGY AND NUCLEAR MEDICINE↗

DOE OSTI2019

Temperature-dependent defect accumulation and evolution in Ni-irradiated NiFe concentrated solid-solution alloy

Temperature significantly affects defect migration and evolution in irradiated materials. However, the effects of temperature on defect evolution in concentrated solid-solution alloys (CSAs), including high entropy alloys, are not well understood, despite their potential as structural materials in advanced nuclear reactors. As an important model system of these CSAs, equiatomic Ni 50 Fe 50 (NiFe) was selected to understand the effects of temperature on defect evolution during irradiation and subsequent thermal annealing. Specifically, defect accumulation and evolution in NiFe alloy under Ni-ion irradiation at 150, 300, and 500 K were studied, and the irradiated specimens were subsequently annealed at higher temperatures. Rutherford backscattering spectrometry along the <100> channeling direction was employed to study damage accumulation and evolution before and after each irradiation and annealing experiment. Here we show that more defects survive and accumulate at 150 K, but more importantly defects can migrate to deeper depths at this low irradiation temperature. Irradiation-induced damage at 150 and 300 K does not recover substantially after post-irradiation annealing at 500 K, but dramatic recovery is observed after post-irradiation annealing at 700 K, indicating an onset temperature of defect recovery between 500 and 700 K. Here, the migration of irradiation-induced defects upon annealing is closely related to the mobility and stress state arising from the surviving defects. With the consideration of five stages of defect recovery in conventional dilute alloys, the underlying mechanisms for temperature-dependent defect accumulation and evolution in NiFe CSA are discussed.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Temperature Effects on Interdiffusion of Al and U-Mo under Irradiation

A high-energy Xe ion irradiation experiment was conducted to investigate the temperature dependence of interdiffusion in bilayer Al-UMo samples under irradiation. The amount of interdiffusion achieved at a fixed dose with the increase of temperature showed a clear transition at 175 degrees C (with an estimated error in the range of ± 10 degrees C) from temperature-independent to temperature-dependent behavior. The activation energy derived from the curve of interdiffusion quantity vs. irradiation temperature is 0.77 ± 0.16 eV. This information has been utilized to understand the temperature effect on the interdiffusion process that occurred at the interfaces of U-Mo particles and the Al matrix in U-Mo/Al dispersion fuels, whose magnitude significantly impacts the fuel's performance. Although this temperature effect was deemed important, it cannot be examined directly using in-pile irradiation data, as fuel temperatures cannot be measured in reactor irradiation and are highly correlated with fission rate and thermal conductivity evolution. To connect the knowledge accumulated from ion irradiation with in-pile irradiation data, simulation of a full-sized U-Mo/Al dispersion fuel plate irradiated in the FUTURE test in the BR2 reactor was performed with the Dispersion Analysis Research Tool (DART), a dispersion fuel performance code. DART is equipped with an interaction or interdiffusion layer (IL) growth correlation formulated to describe the temperature dependence of ion mixing results. The agreement between calculated and measured fuel meat constituent volume fractions and swelling data demonstrated that the temperature effect on in-pile Al-UMo interdiffusion is well captured with the correlation. In this case, the fitted activation energy is 0.70 eV. Considering the uncertainties associated with the ion irradiation data, the activation energy obtained from in-pile data fitting is in accord with that from ion irradiation results.

36 MATERIALS SCIENCE↗

DOE OSTI2025

The state of the art for neutron irradiation experiments from the perspective of the High Flux Isotope Reactor (HFIR)

Irradiation experiment campaigns are critical to advancing nuclear energy technologies by providing data on material performance under relevant radiation conditions. Successful irradiation experiments require integrated design efforts that balance technical goals with facility constraints. Here, this paper presents an expert-informed overview of irradiation experiment design at the High Flux Isotope Reactor. It addresses the nuclear materials research and irradiation experiment communities to guide them toward developing technically sound, facility-compatible campaigns. The High Flux Isotope Reactor is a multipurpose reactor supporting isotope production, neutron scattering, and materials testing. Its high, steady-state neutron flux is ideal for irradiation experiments, but successful execution demands coordinated thermal, structural, and reactor physics analyses. The paper outlines the complete development workflow from concept definition and design optimization to safety qualification and post-irradiation examination. Standardized capsule platforms are also discussed in terms of flexibility, specimen capacity, and thermal performance. Common failure modes such as unanticipated geometric variations, can impact temperature-dose profiles and compromise data reliability. Therefore, detailed thermal modeling and accurate as-built characterization are essential for meaningful post-irradiation data interpretation. Key recommendations include early engagement all stakeholders, clearly defined design expectations, and alignment of specimen geometries with post-irradiation examination capabilities. This approach reduces design iterations, enhances data quality, and supports more efficient use of irradiation resources. Strategic and well-planned irradiation testing not only improves individual campaign success but also accelerates the deployment of advanced nuclear technologies. By closing critical data gaps and reducing development risks, the nuclear materials community can more effectively contribute to the future of clean, resilient energy systems.

Experiments↗

DOE OSTI2014

Effects of ion and nanosecond-pulsed laser co-irradiation on the surface nanostructure of Au thin films on SiO{sub 2} glass substrates

Ion irradiation and short-pulsed laser irradiation can be used to form nanostructures on the surfaces of substrates. This work investigates the synergistic effects of ion and nanosecond-pulsed laser co-irradiation on surface nanostructuring of Au thin films deposited under vacuum on SiO{sub 2} glass substrates. Gold nanoparticles are randomly formed on the surface of the substrate after nanosecond-pulsed laser irradiation under vacuum at a wavelength of 532 nm with a repetition rate of 10 Hz and laser energy density of 0.124 kJ/m{sup 2}. Gold nanoparticles are also randomly formed on the substrate after 100-keV Ar{sup +} ion irradiation at doses of up to 3.8 × 10{sup 15} ions/cm{sup 2}, and nearly all of these nanoparticles are fully embedded in the substrate. With increasing ion irradiation dose (number of incident laser pulses), the mean diameter of the Au nanoparticles decreases (increases). However, Au nanoparticles are only formed in a periodic surface arrangement after co-irradiation with 6000 laser pulses and 3.8 × 10{sup 15} ions/cm{sup 2}. The periodic distance is ∼540 nm, which is close to the wavelength of the nanosecond-pulsed laser, and the mean diameter of the Au nanoparticles remains at ∼20 nm with a relatively narrow distribution. The photoabsorption peaks of the ion- or nanosecond-pulsed laser-irradiated samples clearly correspond to the mean diameter of Au nanoparticles. Conversely, the photoabsorption peaks for the co-irradiated samples do not depend on the mean nanoparticle diameter. This lack of dependence is likely caused by the periodic nanostructure formed on the surface by the synergistic effects of co-irradiation.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗

DOE OSTI2015

THE EFFECTS OF FABRICATION INDUCED RESIDUAL STRESS-STRAIN STATES ON THE IRRADIATION PERFORMANCE OF MONOLITHIC MINI-PLATES

This article evaluates the initial stress-strain state of the monolithic fuel plates and their effects on the overall irradiation performance. The monolithic fuel is a proposed fuel form for high-performance research reactors to achieve higher uranium densities within the core to allow the use of low-enriched uranium fuel. These plate-type fuels consist of a high density low enrichment Uranium-Molybdenum alloy based fuel, a diffusion barrier made of Zirconium, and a cladding made of Aluminum. Because the fabrication of monolithic plates involves multiple stages and the foil material is subjected to various thermo-mechanical processes, a presence of significant stress-strain fields is likely. In addition, the choice of the fabrication process may also cause variations in the foil’s microstructure. There are interests if the overall irradiation performance could be affected by the different initial stress-strain state and the microstructural characteristics. To investigate these concerns, a representative plate was simulated for various initial stress-strain states. First, the co-rolling process for thickness reduction stage of the foil was simulated. For this, a scaled version of the process was evaluated to identify distinct stress profiles on the selected paths. These profiles were then used to define an initial state for the foil before Hot Isostatic Pressing. Additional HIP simulations were also performed to study the cases with stress-free foils prior HIP bonding. The purpose of these simulations was to mimic the stress relief treatment after co-rolling, which aims to remove the residual stresses and minimize the material anisotropy. For the HIP process with initially stress-free foil, several bonding temperatures were considered, including the nominal bonding temperature. The final case considered a plate without initial stresses and strains, assuming a hypothetical case in which the plate would be subjected to a post-fabrication heat treatment to relieve the process-induced stresses before irradiation. Finally, the irradiation processes were simulated for all the cases with distinct pre-irradiation stress-strain profiles. For all cases, the same irradiation parameters were used. The resulting distortions, stress-strain fields and temperature profiles were gathered on selected locations. Finally, a comparative evaluation was made to determine the sensitivity to the pre-irradiation stress-strain variations. The irradiation simulations have revealed that fabrication induced stresses in fuel would be relieved quickly due to large irradiation creep strain. The results indicated that the stress-strain behavior of the fuel foil is not affected significantly by the fabrication process. It is important to note that this excludes microstructural characteristics, which may have important implications on overall behavior. In the cladding however, the stresses continue increase and additional plastic strains are generated. This is due to irradiation hardening by fast neutrons. Additional stresses and strains are formed in the cladding as a result of displacements caused by the volumetric swelling of the fuel foil.

42 ENGINEERING↗

DOE OSTI2014

SU-E-T-222: Investigation of Pre and Post Irradiation Fading of the TLD100 Thermoluminescence Dosimetry for Photon Beams

Purpose: The pre-irradiation and post-irradiation fading of the Thermoluminescense dosimeter signals were investigated in this study. Methods: Two groups of TLD chips with pre-determined ECC values were used in this study. The two groups were divided into 6 series, each composing of 5 TLD chips.The first group was used for pre-irradiation fading. 5 TLDs were exposed to a known amount of radiation from Cs-137 source, and were read out the next day. After seven days, the other 5 TLDs were exposed to the same amount of radiation and were read out after a day. The other series of 5 TLDs were also exposed after 7,19,28, 59, and 90 days, and were read out a day after irradiation. The loss in TLD signal were obtained for all the above cases. The second group, was used for postirradiation fading. All the TLDs of this group were exposed to a known amount of radiation from Cs-137 source. The 6 series composed of 5 TLDs were read out after 1,7,19,28,59, and 90 days. The above-mentioned procedures for obtaining pre-irradiation, and post-irradiation fading were performed for three storage temperatures (25°C, 4°C, and −18°C). Results: According to the results obtained in this study, in case of pre-irradiation fading study, the signal losses after 90 days are 12%, 24%, and 17% for 25°C, 4°C, and −18°C respectively. In case of post-irradiation fading study, the sensitivity losses after 90 days are 25%, 216%, and 20% for 25°C, 4°C, and −18°C respectively. Conclusion: The results indicate that the optimized time between exposing and reading out, and also the optimized time between annealing and exposing is 1 day.The reduction of Storage temperature will reduce the post-irradiation fading, While temperature reduction does not have any effect on pre-irradiation fading.

60 APPLIED LIFE SCIENCES↗

DOE OSTI2015

SU-C-BRB-01: Development of Dynamic Gimbaled X-Ray Head Swing Irradiation Technique

Purpose: The Vero4DRT has a unique gimbaled x-ray head with rotating around orthogonal two axes. The purpose of this study was to develop a new irradiation technique using the dynamic gimbaled x-ray head swing function. Methods: The Vero4DRT has maximum field size of 150Χ150 mm2. The expanded irradiation field (expanded-field) for the longitudinal direction which is vertical to the MLC sliding direction, was created by the MLC motion and the gimbaled x-ray head rotation. The gimbaled x-ray head was rotated ± 35 mm, and the expanded-field size was set as 150Χ220 mm2. To irradiate uniform dose distribution, the diamond-shaped radiation field was created and continuously moved for the longitudinal direction. It was achieved by combination of opening and closing of the MLC and gimbal swing rotation. To evaluate dosimetric characteristic of the expanded-field, films inserted in water-equivalent phantoms at 100 mm depth were irradiated and the field size, penumbra, flatness and symmetry were analyzed.In addition, the expanded-field irradiation technique was applied to virtual wedge irradiation. Wedged beam was acquired with the delta–shaped radiation field. 150Χ 220 mm2 fields with 15, 30, 45, and 60 degree wedge were examined. The wedge angles were measured with irradiated film and compared with assumed wedge angles. Results: The field size, penumbra, flatness and symmetry of the expanded-field were 150.0 mm, 8.1–8.4 mm, 2.8% and −0.8% for the lateral direction and 220.1 mm, 6.3–6.4 mm, 3.2% and −0.4% for the longitudinal direction at 100 mm depth. The measured wedge angles were 15.1, 30.2, 45.2 and 60.2 degrees. The differences between assumed and measured angles were within 0.2 degrees. Conclusion: A new technique of the gimbal swing irradiation was developed. To extend applied targets, especially for whole breast irradiation, the expanded-field and virtual wedge irradiations would be effective.

60 APPLIED LIFE SCIENCES↗