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DOE OSTI2024

A Workflow to Optimize Fast Neutron Irradiation in A Thermal Neutron Spectrum Test Reactor Leveraging Open-Source Tools

The Advanced Test Reactor (ATR) located at Idaho National Laboratory (INL) is one of the key nuclear engineering research and testing facilities within the US Department of Energy (DOE). The ATR is one of few high-power research reactors in the world with different application including accelerated testing of nuclear fuel, materials irradiation in a very high neutron flux environment, and medical radioisotope production [1]. Also, the ATR offers opportunities for testing fast spectrum fission and fusion reactor materials. The key challenges in this area are in further detailing and optimizing a fast spectrum environment within a thermal test reactor. This challenge involves researching, developing, and testing novel concepts for the multiplying of neutron populations into ever higher energy spectra in high flux test reactors like ATR. The main objective of this work is to investigate candidate materials for establishing a fast neutron experiment irradiation in thermal neutron spectrum test reactors which can be accomplished by filtering thermal and epithermal neutrons and boosting fast neutrons at designated irradiation positions. However, adding these filters will render the neutron spectrum and the criticality of the system. The selection of the thickness and material layers should be accomplished by developing an optimization design algorithm that is applicable for ATR to enhance the fast neutron spectrum irradiation utilizing high-fidelity Monte Carlo methods along with advanced machine learning capabilities. This paper presents workflow for design optimization to enhance fast neutron irradiation in the ATR. The workflow leverages open-source tools to develop an algorithm that is viable to ATR and can be leveraged in other reactors. The following sections discuss the development of the experiment design optimization workflow and its application to ATR irradiation positions.

42 - ENGINEERING↗

DOE OSTI2020

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

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

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2020

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

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

Summary Report on Ion Irradiation Study of Ceramic Coating on Suppressing FCCI

Advanced cladding is critical for advanced nuclear reactors with an enhanced performance in radiation tolerance and neutron transparency. Using advanced cladding will ensure the adequate thermal conductivity and mechanical stability of the cladding base material, corrosion resistance, high-temperature coolant compatibility of the cladding surface, and chemical stability in the cladding inner wall against fuel cladding chemical interaction (FCCI). An innovative cladding with a three-layer structure (i.e., a modified surface, a clad base material, and a modified inner wall) promises to meet all these requirements. Initial research and development (R&D) regarding this innovative cladding seek to demonstrate the effectiveness of a thin ceramic coating in suppressing FCCI under ion irradiation to high dose. This United States (U.S.) Department of Energy (DOE)–Office of Nuclear Energy (NE) project report summarizes the results of recent ion irradiation studies of diffusion-couple samples with details of an ion irradiation experiment, the characterization of the interface microstructure of cerium (Ce)/titanium nitride (TiN)/oxide-dispersion strengthened (ODS), Ce/TiN/iron (Fe), Ce/ODS, and Ce/Fe samples irradiated with 80 million electron-volts (MeV) of xenon (Xe) ions to 100 displacements per atom (dpa) at 500?C using the Argonne Tandem Linac Accelerator System (ATLAS) facility at Argonne National Laboratory (ANL). The results of the ODS-substrate sample are encouraging and demonstrate the effectiveness of a thin ceramic coating on FCCI mitigation under irradiation. The results from an Fe-substrate sample revealed a complex microstructure, and the root cause is discussed.

36 MATERIALS SCIENCE↗

NASA NTRS2004

Formation and Stability of Radiation Products in Europa's Icy Shell

Spectra of Europa reveal a surface dominated by water-ice along with hydrated materials and minor amounts of SO2, CO2, and H2O2. Jovian magnetospheric ions (protons, sulfur, and oxygen) and electrons produce significant chemical modifications of the surface on time scales of a few years at micrometer depths. Our laboratory studies examine the formation and stability of radiation products in H2O-rich ices relevant to Europa. Infrared (IR) spectra of ices before and after irradiation reveal the radiation destruction of molecules and the formation of products at 86 - 132 K. In addition, spectra of ices during warming track thermal evolution due to chemical changes and sublimation processes. IR-identified radiation products in 86 - 132 K irradiated H2O + SO2 ices are the bisulfate ion, HSO4(-), sulfate ion, SO4(2-) and the hydronium ion, H3O(+). Warming results in the formation of a residual spectrum similar to liquid sulfuric acid, H2SO4, for H2O:SO2 ratios of 30:1, whereas hydrated sulfuric acid, H2SO4 4 H2O, forms for ratios of 30:1. Radiation products identified for irradiated H2O + H2S ices at 86 K are H2S2 and SO2. When irradiated at 110 and 132 K, ices with H2O:H2S ratios if either 3:1 or 30:1 show the formation of H2SO4 4 H2O on warming to 175 K. We have also examined the radiation stability of H2SO4. Addition of CO2 to H2O + SO2 ices results in the formation of CO3 at 2046 cm (sup -1) (4.89 m). This is the strongest band from a carbon-containing product in the mid-IR spectral region, and it is also seen when either pure CO2 or H2O + CO2 ice is irradiated. Experiments with CH4 added to H2O + SO2 + CO2 ices addressed the question of methane's use as a marker of methanogens in an irradiated ice environment. New results on the near-IR spectrum of pure H2O2 will be included in this presentation. Interpretations of near-IR water bands, with H2O2 present, will be discussed. Irradiations of H2O2 and H2O + H2O2 mixtures, to examine the possibility of O2 and O3 formation, are currently under investigation and new results will be discussed.

Moore, M. H.↗

DOE OSTI2023

TRISO Fuel Development and Qualification in the US

Presentation of TRISO Fuel Development and Qualification in the US. Including Status of HTGR and TRISO Technology Circa 2000, Advanced Gas Reactor focus on licensing and domestic TRISO fuel fabrication capabilities. Irradiation experiments discussed include AGR-1, AGR-2, AGR-5/6/7, and AGR-3/4. Results, of tests and Kernel and coating behaviors during irradiation, Fission product release from UCO Fuel Compacts, study of failed particles, safety and accident performance summaries, and Parfume modeling and capabilities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Determining the 9 $\mathrm{Be}$($n, γ$) 10 $\mathrm{Be}$ integral cross section at fission neutron energies

The 9 B neutron capture cross section has significant implications for Be materials in the nuclear industry as well as the α process in stellar nucleosynthesis. While the cross section is well constrained at thermal neutron energies, there is a lack of experimental data at higher neutron energies, and the evaluated nuclear data libraries can differ by up to two orders of magnitude. We calculate the 9 Be(n, γ) 10 Be integral cross section at fission neutron energies in an effort to resolve disagreements amongst the nuclear data libraries. Foil irradiation experiments were performed using the Flattop critical assembly at the National Criticality Experiments Research Center with either the highly enriched U or Pu cores, with target foil stacks placed at multiple locations to exploit different neutron energy profiles. Accelerator mass spectrometry was used to measure the 10 Be/ 9 Be ratio in irradiated Be foils, while all other activation products were quantified through gamma spectrometry. The experiments were simulated using the Monte Carlo N-Particle radiation transport code and combined with experimental results to determine the total neutron fluence, while the staysl-pnnl suite and fispact-ii code were used to validate the model and assess the systematic uncertainty. The new 9 Be(n, γ) 10 Be integral cross sections calculated in this work are 26.5 ± 2.2µb at 0.59 ± 0.07 MeV, 24 ± 3 µb at 0.98 ± 0.14 MeV, 21.7 ± 1.3 µb at 1.26 ± 0.11 MeV, 21.8 ± 1.4 µb at 1.32 ± 0.11 MeV, and 18.6 ± 1.1 µb at 1.46 ± 0.13 MeV. These results do not agree with integral cross sections from any of the nuclear data library evaluations. Discrepancies between the new integral cross sections reported here and the nuclear data libraries suggest a more complex cross-section structure in the MeV range which allows for more resonance contributions, and more work is needed to further constrain the evaluated cross sections.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2024

International fuel performance study of fresh fuel experiments for PCMI effects during RIA experiments

This paper presents the results of High-burnup Experiments for Reactivity-initiated Accident (HERA) Modeling & Simulation (M&S) exercise. The HERA project under the Nuclear Energy Agency (NEA) Second Framework for Irradiation Experiments (FIDES-II) program is focused on studying Light Water Reactor (LWR) fuel behavior during Reactivity-Initiated Accident (RIA) conditions. The Part I M&S cases are based on a series of tests in the Transient Reactor Test (TREAT) facility in the United States and the Nuclear Safety Research Reactor (NSRR) in Japan. The purpose of this work is to evaluate the test design to accomplish its goals in establishing clearer understanding of the effects of power pulse width during RIA conditions. Further, the blind predictions using various computational tools have been performed and compared amongst to interpret the behaviors of high burnup fuels during RIA. While many international participants evaluate the thermal–mechanical behavior of fuel rod under different conditions, a considerable scatter of outputs comes out for the cases due to the disparity between codes in predicting mechanical behaviors. In general, however, the results of thermal–mechanical analysis elaborate that nominal design conditions the shorter pulse width tests in NSRR should cause cladding failures while the TREAT tests appear to have more split prediction of failure or not. Furthermore, the sensitivity analysis varying key testing parameters reveals the considerable effect of power pulse width and total energy deposition on prediction of fuel rod failure.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRS1992

Comparative study of spermatogonial survival after X-ray exposure, high LET (HZE) irradiation or spaceflight

Spermatogonial cell loss has been observed in rats flown on Space Lab 3, Cosmos 1887, Cosmos 2044 and in mice following irradiation with X-ray or with HZE particle beams. Spermatogonial loss is determined by cell counting in maturation stage-6 seminferous tubules. With the exception of iron, laboratory irradiation experiments (with mice) revealed a similar pattern of spermatogonial loss proportional to the radiation dose at levels less than 0.1 Gy. Helium and argon irradiation resulted in a 5-percent loss of spermatogonia after only 0.01 Gy exposure. Significant spermatogonial loss (45 percent) occurred at this radiation level with iron particle beams. The loss of spermatogonia during each spaceflight was less than 10 percent when compared to control (nonflight) animals.

Sapp, W. J.↗

NASA NTRS1988

Application of linear response theory to experimental data of simultaneous radiation and annealing response of a CMOS device

Results from the application of linear response theory are compared to experimental data from simultaneous radiation and annealing response of a CMOS device. In particular, a method is applied which was developed earlier to determine the characteristic time, t(0), as well as the parameters A and C in the 1n(t) dependence of the linear response function R(t) = -C + A1n(1-t/t(0)). The method is based on a study of the linear response for t being much less than t(0), when R(t) can be expanded in a power series of t: R(t) = R(0) + R'(0)t + 1/2R''(0)t-squared + 1/3R'''(0)t-cubed + ..., where R'(0) and R''(0) are, respectively, the first and second derivatives of R with respect to t. To find the linear response, R(t-t') is substituted in the form of this power series equation into a general equation for the shift of the threshold potential. To test the method, irradiation experiments were conducted on RCA 10(6) rad-hard CMOS IC's. A dose rate of approximately 130 rads/min was used. An IC was irradiated with Co-60 gamma rays for several hours, taking measurements of the threshold potential for one n-channel and one p-channel transistor every ten minutes. For the p-channel transistor, t(0) was found to be approximately 110 min and for the n-channel, t(0) was approximately 70 min. For the p-channel, the theoretical curve deviates from the experimental points only after 70 min; for the n-channel, the deviation takes place after 45 min. Additional findings are discussed and the application of the method to pure annealing is described.

Litovchenko, V.↗

DOE OSTI2021

Improvement of the $\mathrm{BISON U_3Si_2}$ modeling capabilities based on multiscale developments to modeling fission gas behavior

Uranium silicide (U 3 Si 2 ) is a concept explored as a potential alternative to UO 2 fuel used in light water reactors (LWRs) since it may improve accident tolerance and economics due to its higher thermal conductivity and increased uranium density. U 3 Si 2 has been previously used in research reactors in the form of dispersion fuel, but operated at lower temperatures than commercial LWRs. The research reactor data illustrated that significant gaseous swelling occurs as the fuel burnup increases. Therefore, it is imperative to understand the fission gas behavior of U 3 Si 2 under higher temperature LWR operating conditions. In this work, molecular dynamics and phase-field modeling techniques are used to reduce the uncertainty in select modeling assumptions made in developing the fission gas behavior model for U 3 Si 2 in the BISON fuel performance code. These lower length scale informed models are then utilized in the validation of BISON U 3 Si 2 modeling capabilities to simulate the ATF-1 experiments irradiated in the Advanced Test Reactor (ATR). Sensitivity analysis (SA) and uncertainty quantification (UQ) are included as part of the validation process to identify where further experiments and lower length scale modeling would be beneficial. Here, the multiscale modeling approach utilized in this work can be applied to new fuel concepts being explored for both LWRs and advanced reactors (e.g., uranium nitride, uranium carbide).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Computer vision models and advanced TEM imaging for microstructures of irradiated AM316 stainless steels

Advancements were made in automating microscopy-based material characterization, particularly in studying irradiation effects on additively manufactured (AM) materials using machine learning (ML) and computer vision (CV). These automation efforts address the challenges of analyzing complex microstructures, accelerating the detection of irradiation-induced defects. Two CV models were developed at Argonne National Laboratory (ANL) to enhance transmission electron microscopy (TEM) analysis of irradiated AM 316 stainless steel. The first model focused on the detection of irradiation-induced dislocation loops, which contribute to material hardening and embrittlement. These loops, categorized as faulted or perfect, were automatically detected and classified using a Mask R-CNN model trained on TEM images from both in-situ and ex-situ ion irradiation experiments. The model achieved high accuracy, with precision, recall, and F1 scores of 0.839, 0.734, and 0.776, respectively, demonstrating its effectiveness in analyzing dislocation loops in irradiated AM materials. The second CV model was developed to analyze the size and wall thickness of dislocation cells in laser powder bed fusion (LPBF) 316 stainless steel. Using a U-Net++ architecture with EfficientNet as the encoder, the model was trained on TEM images to segment and measure cell size and wall thickness.

36 MATERIALS SCIENCE↗

DOE OSTI2022

Design and commissioning of an e-beam irradiation beamline at the Upgraded Injector Test Facility at Jefferson Lab

We report the Upgraded Injector Test Facility (UITF) at Jefferson Lab is a continuous-wave superconducting linear accelerator capable of providing an electron beam with energy up to 10 MeV. A beamline for electron-beam irradiation has been designed, installed and successfully commissioned at this facility, aimed at the degradation study of 1,4-dioxane and per- and polyfluoroalkyl substances (PFAS) in wastewater treatment. A solenoid with a peak axial magnetic field of up to 0.28 T and a set of raster coils were used to obtain a Gaussian beam profile with a transverse standard deviation of ~ 15.0 mm at the target location. Monte-Carlo simulations using FLUKA were carried out to calculate the total absorbed dose and the dose distribution in the sample volume inside the target cell. The simulations were benchmarked experimentally by dosimetry mapping using optichromic dosimeters. The results of the irradiation experiments showed a ~ 95% reduction of 1,4-dioxane in ultra-pure water for a dose of 1 kGy, demonstrating the potential of electron-beam irradiation towards addressing growing challenges in environmental remediation.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2022

High-burnup Experiments in Reactivity Initiated Accidents (HERA)

High-burnup Experiments in Reactivity-initiated Accidents (HERA) is a joint experimental program (JEEP) operating within the United States (U.S.) Nuclear Energy Agency’s (NEA’s) framework for irradiation experiments (FIDES). HERA is dedicated to the understanding of light water reactor (LWR) fuel performance at high burnup under reactivity-initiated accidents (RIA). In-pile RIA experiments have been performed on high-burnup fuels (e.g., above 60 gigawatt days per metric ton of uranium [GWd/MTU]) in the CABRI reactor at the Cadarache site in southern France, and the Nuclear Safety Research Reactor (NSRR) in Japan. However, most of these experiments have taken place with heavily corroded Zircaloy claddings with pulse widths that are narrower (e.g., 5 ms – 30 ms full-width-half-max [FWHM]) than what would be likely in a commercial LWR (e.g., 30 ms – 80 ms FWHM). A few tests were performed in the CABRI facility with FWHM up to 76 ms, though only one “blistered” rod experienced cladding failure. Heavy waterside corrosion and narrow pulse widths are both known to increase the vulnerability of LWR fuel to pellet cladding mechanical interaction (PCMI). The HERA proposal is designed to: (1) quantify the impact of pulse width on fuel performance, offering new insight into the applicability of existing data; (2) generate new data on high burnup fuel under pulse conditions prototypic of LWRs; (3) quantify the additional margin provided by modern cladding alloys to PCMI failure limits; and (4) offer improved data for modelers using specially designed tests that eliminate key uncertainties in high-burnup fuel tests.

(HERA)↗

DOE OSTI2026

Building a High-Throughput MiniFuel PIE Pipeline for HFIR-Based Advanced Fuel Evaluation

MiniFuel irradiation experiments have become a key capability for accelerated evaluation of advanced and high-burnup nuclear fuels in the High Flux Isotope Reactor. As the number of irradiated MiniFuel targets entering postirradiation examination (PIE) has increased, improvements to the PIE workflow were needed to support more repeatable target disassembly, subcapsule processing, specimen recovery, and downstream measurements. This report summarizes FY 2026 improvements developed and implemented at the Irradiated Fuels Examination Laboratory to improve the MiniFuel target-to-specimen (T2S) workflow and related PIE activities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Initial Observations from AGR 5/6/7 Capsule 1

The fourth and final irradiation experiment in the Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) fuel development and qualification program is designated as AGR-5/6/7. Data collected from the fabrication, irradiation, and post-irradiation examination (PIE) of this tristructural isotropic (TRISO) fuel are intended to serve as the primary data set for the qualification of this fuel for use in high-temperature gas-cooled reactors (INL 2021, Collin 2018b). However, data collected from the three preceding irradiations (i.e., AGR-1, AGR-2, and AGR-3/4) may also be used to supplement data collected from AGR-5/6/7. All components of the AGR-5/6/7 fuel (i.e., UCO kernels, TRISO coatings, and fuel compacts) were produced on an engineering scale at BWXT (Lynchburg, Virginia USA) according to the fuel specification (Marshall 2017). This fuel was irradiated in the northeast flux trap (NEFT) at the Advanced Test Reactor (ATR) at Idaho national Laboratory (INL) from February 16, 2018 to July 22, 2020 (Pham et al. 2021). Measurements in the fission product monitoring system (FPMS) indicated unexpected and significant numbers of failures of TRISO particles in Capsule 1 near the end of the sixth irradiation cycle (ATR Cycle 166A). In the fourth cycle (ATR Cycle 164B) and beyond, the sweep gas flow became very low (presumably from degradation of the capsule gas outlet line via an unidentified mechanism), and the program deliberately isolated Capsule 1 from gas flow periodically. In later cycles, attempts to reestablish any kind of flow in Capsule 1 were unsuccessful. With little or no flow through Capsule 1, FPMS measurements and enumerations of failed particles in Capsule 1 were difficult or impossible as was the ability to control the helium/neon gas mixture used for temperature control. Gas flows and fission gas activity in the effluent gas from the other AGR-5/6/7 capsules were also impacted by the Capsule 1 gas flow issues and the large increase in fission gas released from the Capsule 1.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Microstructural Simulation of Ion-Irradiated Natural Rocks and Minerals

This report details the contribution of Oak Ridge National Laboratory to the Nuclear Energy University Program (NEUP) project, "Rapid Characterization of Concrete Mineralogy using Multi-Scale Tools" led by the University of Illinois at Urbana-Champaign. To support the subsequent license renewal (SLR) of the US nuclear power plants (NPPs) fleet, the performance of concrete-forming aggregates against neutron irradiation needs to be assessed. Ion irradiation is proposed as a rapid, cost-effective surrogate method to full neutron irradiation testing. To complement the characterization of ion-irradiated rock specimens, ORNL has run finite-elements and fast-Fourier transform (FFT-based) simulations using the codes MARS and Microstructure-Oriented Scientific Analysis of Irradiated Concrete (MOSAIC). The main conclusion of this analysis is that the apparent post–ion-irradiation step-height underestimates the accumulated free radiation-induced volumetric expansion (RIVE) in the ion-implanted depth by about 15% at full amorphization and about 25% toward the beginning of the ion irradiation experiment. The discrepancy is explained by the fact that the step height is proportional to the sum of the RIVE (i.e., one third of the RIVE) and the irradiation-assisted plastic strains in the vertical direction (lower than two thirds of the RIVE). Because of the large lateral compressive stresses caused by the restraining effect of the unirradiated substrate, the stress field in the mineral grains and at the grain boundary (GB) in the ion-implanted layer is much different than the comparable stress field occurring during neutron irradiation. Hence, the mismatch RIVE causing cracks in the rock-forming minerals leads to different cracking patterns. Ion irradiation continues to be used as a rapid technique to assess the RIVE potential of rocks.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Novel Materials R&D for Next-Generation Accelerator Target Facilities

High-Entropy Alloys and Electrospun Nanofiber materials are two novel classes of materials that can offer improved resistance to beam-induced radiation damage and thermal shock. Research to develop these new materials specifically for multi-megawatt accelerator target applications, such as beam windows and particle-production targets, has recently begun. The research program will combine in-beam experiments with complementary simulations to tailor the microstructures of these novel materials for use in next-generation accelerator target facilities. Iterative simulations to optimize the material composition, physics performance and beam-induced thermomechanical response will guide the material design and fabrication processes based on established figures of merit. Ensuing material irradiation experiments using low-energy ions and prototypic high-energy protons, followed by extensive post-irradiation material characterization, will then assess and qualify the selected novel materials. This talk will provide an overview of the novel materials development research program initiated at Fermilab through my DOE Early Career Research Program award.

43 PARTICLE ACCELERATORS↗