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DOE OSTI2022

GIF-VHTR Materials (Graphite) Project Plan: Progress since September 2021

¦1: Data, Design Methodology and Construction 1.1: Graphite Selection and Acquisition Strategy (Original tasks completed) 1.2: Graphite Properties (Virgin Materials) 1.3: Graphite Fracture Behaviour(Original tasks completed) 1.4: Graphite Oxidation Behaviour(Original tasks completed) 1.5: Graphite Component Testing 1.6: Graphite Irradiation Effects 1.7: Graphite Irradiation Induced Creep 1.8: Graphite Codes and Standards Development (ASME & ASTM) 1.9: Graphite BehaviourModels Development 1.10: Links to Existing Graphite Irradiation BehaviourDatabases 1.11: Materials Database ¦2: Operation and Inspection 2.1: Methodologies for determining requirements for inspection of graphite cores 2.2: Methodologies for ISI of graphite cores ¦3: Decommissioning and Disposal (New task added 2021) 3.1: Technical data for decommissioning and disposal issues

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Post Irradiation Examination Results of Irradiated Yttrium Hydride

Department of Energy’s (DOE’s) Microreactor program (MRP) aims to provide the fundamental data to enable the development of microreactors. As such, material property data of critical materials for microreactor technologies are researched. Because substoichiometric yttrium dihydride (YHx, where x<2) is considered as a potential solid neutron moderator, its material property data has been combined in the Advanced Moderator Material Handbook which includes thermodynamic and thermophysical properties of YHx with the exception of irradiated material’s properties due to limited PIE. To fill the knowledge gap for the irradiated YHx, specimens and irradiation capsules were prepared at Los Alamos National Laboratory (LANL). Specimens were irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory’s (INL’s). post-irradiation examination (PIE) was performed at INL’s Materials and Fuels Complex (MFC). This report compiles the PIE results of irradiated YHx specimens through fiscal years 2022 and 2023 (FY22-23) . The PIE data will directly be incorporated into the newer version of the Advanced Moderator Material Handbook. The main takeaways include that (i) the geometrical stability and mechanical integrity of YHx was intact with couple exceptions after high-temperature irradiations (600-800°C), (ii) hydrogen content variation due to manufacturing or irradiation in YHx caused visible surface discoloration, that is also related to the microstructural changes, (iii) qualitative comparisons of PIE methods implied that H retention was significantly higher at 600°C as compared to 800°C, (iv) thermal properties included signatures correlated with the H loss or re-gain, (v) importance of manufacturing readiness and initial as-manufactured specimens history was emphasized, (vi) the needs of targeted irradiations focusing on temperature and time parameters and very targeted PIE were specified.

08 HYDROGEN↗

DOE OSTI2024

Validating Irradiance Models for High-Latitude Vertical Bifacial Photovoltaic Systems

Bifacial photovoltaic systems oriented vertically facing east-west are an emerging design, targeting production in morning and afternoon hours and providing competitive annual energy yield to traditional south-tilted modules for high latitude locations. The accuracy of existing bifacial PV models when modules are oriented vertically has yet to be examined in detail. Here, we compare four bifacial PV irradiance models in ~150 locations between 15-80 degrees N on the utility-scale, finding higher inter-model deviations for vertical PV systems than south-tilted across all latitudes less than 75 degrees N. We validate model-predicted irradiance with test-site data collected in Golden, Colorado and Fairbanks, Alaska for E-W vertical and south-tilted arrays. View factor models agree with E-W vertical test-site data in Golden with RMSE=15%. Modelling error increases for the Alaskan test-site to RMSE values between 21-30%, driven in part by high albedo measurement uncertainty during snowy months.

bifacial↗

DOE OSTI2023

Report on the Incorporation of Post-Irradiation Examination Results of Yttrium Hydride for Advanced Moderator Handbook

08 HYDROGEN↗

DOE OSTI2024

2024 AGR 3/4 PIE Overview

An overview of work done on the AGR 3/4 PIE experiment is presented, with some modeling results and description of the R-DLBL of as-irradiated and heating test samples

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Report on partially complete post-irradiation-examination of the INL samples

Ferritic-martensitic steels (NF616 and T91), and austenitic stainless steel 800H and its Grain Boundary Engineering (GBE)-treated version 800H-TMP (ThermoMechanical Processing) were irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and their microstructures and mechanical properties were characterized. Ferritic-martensitic steels were irradiated up to 8.16 dpa in the irradiation temperatures ranging from 241 °C to 447.5 °C. Selected 800H and 800H-TMP samples were also irradiated up to 9.12 dpa at 359°C to 431°C. Vickers hardness measurement, fractography, and microstructural characterization were performed on the selected samples in the LAMDA lab. Radiation-induced hardening of NF616 and T91 was observed in the hardness measurements which show ~37% to ~65% increases, depending on irradiation dose and irradiation temperature. Within the irradiation conditions studied in this work, samples with a higher dose had a higher hardness. All the tested alloy 800H and 800H-TMP in this work showed irradiation-hardening by ~96±7% to ~152±10%. Alloy 800HTMP tended to have slightly smaller irradiation-hardening than alloy 800H. The fractography of NF616 samples D2 (2.96 dpa at 291.5°C), D4 (5.91 dpa at 359°C), and D6 (8.16 dpa at 431°C) indicated loss of ductility with negligible necking for the sample D2, yet obvious ductile failure for samples D4 and D6. Fractography of alloy 800H samples N4 (7.27 dpa at 359°C), N5 (3.9 dpa at 451.5°C), and N6 (9.01 dpa at 431°C), and alloy 800H-TMP samples P4 (7.36 dpa at 359°C), P5 (3.95 dpa at 451.5°C), and P6 (9.12 dpa at 431°C) indicated that all samples failed in a ductile mode, with obvious necking. Dimples were observed where large Ti-containing particles were observed for the samples. TEM characterization was performed on alloy 800H (N4, N5, N6) and 800H-TMP (P4, P5 and P6). Accumulation of M23C6 precipitates at grain boundaries was observed in all samples, and the presence of Ti(C,N) precipitates at grain boundaries and in the matrix was observed in alloy 800H-TMP. M23C6 precipitates were found to have the cube-on-cube orientation relationship with the matrix. In addition, nanoscale Si-rich clusters were observed in the matrix of all samples. Atom probe tomography was conducted on the same samples, and their results are being analyzed to be integrated with the TEM results for a confident description of the clustering behavior. Due to the low irradiation temperature (up to 451.5°C), dislocation loops formed in all samples, with the density and the average size of dislocation loops in the order of 1022 – 1023 m-3 and 11.7 – 15.9 nm, respectively. The loop densities in alloy 800H were higher than that in alloy 800H-TMP under the same irradiation conditions. Further systematic data analyses, together with some complementary experiments, will be pursued for these samples to foster peer-reviewed journal article publications.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2020

Radiation tolerance of the PNI RM3100 magnetometer for a Europa lander mission

The results of two radiation test campaigns on a low-cost commercial off-the-shelf magnetometer are presented. The test setup and the total ionization dose (TID) levels studied were designed to meet the requirements of a mission to land on Europa. Based on the Europa Lander Science Definition Team report, instruments inside an aluminum vault at the surface of Europa would need to withstand TID of up to 300 krad(SI). In order to evaluate the performance of the PNI RM3100 magnetometer, nine separate sensors were irradiated at two different facilities during two separate campaigns and under different configurations, including passive and active tests. Of the nine sensors, seven survived the TID of 300 krad(SI) while the other two sensors started presenting failures after reaching 150 krad(SI). Post-irradiation tests showed that eight of the nine sensors continued to work without appreciable degradation after stopping exposure, while one sensor stopped working altogether.

58 GEOSCIENCES↗

NASA NTRS1992

The RBE-LET relationship for rodent intestinal crypt cell survival, testes weight loss, and multicellular spheroid cell survival after heavy-ion irradiation

This report presents data for survival of mouse intestinal crypt cells, mouse testes weight loss as an indicator of survival of spermatogonial stem cells, and survival of rat 9L spheroid cells after irradiation in the plateau region of unmodified particle beams ranging in mass from 4He to 139La. The LET values range from 1.6 to 953 keV/microns. These studies examine the RBE-LET relationship for two normal tissues and for an in vitro tissue model, multicellular spheroids. When the RBE values are plotted as a function of LET, the resulting curve is characterized by a region in which RBE increases with LET, a peak RBE at an LET value of 100 keV/microns, and a region of decreasing RBE at LETs greater than 100 keV/microns. Inactivation cross sections (sigma) for these three biological systems have been calculated from the exponential terminal slope of the dose-response relationship for each ion. For this determination the dose is expressed as particle fluence and the parameter sigma indicates effect per particle. A plot of sigma versus LET shows that the curve for testes weight loss is shifted to the left, indicating greater radiosensitivity at lower LETs than for crypt cell and spheroid cell survival. The curves for cross section versus LET for all three model systems show similar characteristics with a relatively linear portion below 100 keV/microns and a region of lessened slope in the LET range above 100 keV/microns for testes and spheroids. The data indicate that the effectiveness per particle increases as a function of LET and, to a limited extent, Z, at LET values greater than 100 keV/microns. Previously published results for spread Bragg peaks are also summarized, and they suggest that RBE is dependent on both the LET and the Z of the particle.

Non-NASA Center↗

DOE OSTI2024

Thermal Analysis of the Sirius-3 Nuclear Propulsion Fuel Experiments

To support long space missions, NASA is designing and testing different fuels for nuclear thermal propulsion. These designs include a number ceramic metallic fuels which are to be irradiated in the Tranisent Test Reactor to test the fuel’s performance at protypical conditions. The irradiation campaign of these fuels is referred to as the Sirius experiments, and includes single and stacked experiment samples. In this paper we present the thermal analysis of the Sirius-3 experiment. For this analysis, temperatures from a Bison finite element model are compared to the temperatures obtained from an experiment thermocouple where good agreement between the two are found. Because agreement of the thermocouple and Bison finite element model, insight into the heat flow is provided to experimenters as they design future experiments. As an example of this insight, the rods through the fuel specimen as shown in Sirius A and B [1] were removed and replaced with hold down rings for the Sirius-2C

33 ADVANCED PROPULSION SYSTEMS↗

DOE OSTI2021

Design Optimization for Printed Melt Wire Arrays Encapsulation

As part of the Nuclear Energy Enabling Technology (NEET) Advanced Sensor and Instrumentation (ASI) Program, Idaho National Laboratory (INL) has recently established in-house capabilities to fabricate and test new advanced-manufactured sensors for measuring peak irradiation temperature within a nuclear test reactor. Although methods of real-time temperature monitoring, such as thermocouples, may be used, the complexity of feedthroughs and attachments to collect real-time measurements greatly increases the cost of the experiment. Instead, passive monitoring techniques may be used for peak- temperature measurement that exploit the melting point of well-characterized materials (standard melt wires) to infer peak reactor temperatures. However, limited available space for instrumentation during experiments introduces an additional challenge. To accommodate this, INL has expanded its temperature- detection instrumentation capabilities to include advanced manufactured (AM) melt wires for peak irradiation temperature measurements. These melt wires can determine peak temperatures while also accommodate space limitations in irradiation experiments. In an effort to improve performance reliability of AM meltwire capabilities, a process was developed and tested to identify the significance of entrapping a high purity inert atmosphere within the packaging of printed melt wire arrays. The materials used in this study were aluminum, zinc, and tin encapsulated in high purity helium within a stainless steel (SS) 316 container. Tin, with a low melting point of approximately 230°C, Zn with a mid-melting point of approximately 420°C, and Al with a high melting point of approximately 660°C. This report describes the design, fabrication process, furnace testing and X-ray Computed Tomography (XCT) evaluation. Results show a successful outcome in creating an inert gas encapsulation and high-resolution evaluation methods.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

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 OSTI2020

The application of synchrotron micro-computed tomography to characterize the three-dimensional microstructure in irradiated nuclear fuel

This research demonstrated the first time that synchrotron X-ray micro-computed tomography (μ-CT) was utilized to analyze the three-dimensional (3-D) morphology of the microstructure of irradiated nuclear fuel. Both absorption contrast and propagation-based phase contrast enhanced μ-CT techniques were evaluated. Porosity and phase region volumes were determined in neutron irradiated uranium 10 wt.% zirconium (U-10Zr) using propagation-based phase contrast enhanced μ-CT. Volumes were determined by image processing software and algorithms, with the porosity volume subsequently compared to the ASTM E562 method. The 3-D porosity of a small region (~ 8 x 10 5 µm 3 ) located in the intermediate fuel redistribution zone that was removed via a focused ion beam was determined to be 7.2% via image processing. A detailed 3-D analysis of the porosity with respect to volume and morphology was provided. Five growth stages of pore evolution were observed including: nucleation, growth, coalescence, interconnected porosity, and extended/interconnected porosity. Four distinct phase regions were identified, including the pores, via both absorption contrast and propagation-based phase contrast enhanced μ-CT. A quantitative comparison of fuel swelling between this localized microscopic U-10Zr fuel region, previously irradiated in the Fast Flux Test Facility, and historic U-10Zr fuels irradiated in the Experimental Breeder Reactor-II (EBR-II) was conducted. It was determined that for similar burnups, the localized swelling in this microscopic fuel region was 7.7%, which was approximately one-half of the value of the macroscopic swelling reported in EBR-II irradiated fuel. With the advent of new specimen preparation tools for irradiated materials, as well as advances in characterization techniques, researchers can now obtain previously inaccessible microstructural insight into irradiated fuels. This possibility naturally lends itself to improved fuel performance models for the nuclear community.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Photon Irradiation Effects on the Enduray Vision System for Nuclear Inspections

Video cameras with greater radiation hardening are needed for the operation and remote handling of equipment in nuclear reactor inspection and refueling applications. Vega Wave Systems has developed a radiation-hardened vision system for nuclear energy applications. The program described in this report provided radiation-hardness testing of this vision system using the Argonne Low-Energy Accelerator Facility (LEAF). This report presents the results of irradiation tests performed on several of Vega Wave System’s cameras at dose rates ranging from 1 kGy/hr up to 10 kGy/hr and up to total doses of >500 kGy. The results were excellent, providing proof of the camera system’s immunity to radiation effects at all levels tested.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Initial Temperature Testing of Advanced Manufactured Melt Wire Package

As a part of the In-Pile Instrumentation Program, The Idaho National Laboratory (INL) has recently established in-house capabilities to fabricate and test new advanced manufactured sensors for measuring the peak irradiation temperature inside a nuclear test reactor. Although nuclear test reactors can use standard methods of real-time temperature monitoring, such as thermocouples, the complexity of feedthroughs and attachment complications to collect real-time measurements greatly increases the cost of the experiment. Passive monitoring techniques may be utilized that exploit the melting point of well characterized material (standard melt wires), however, the limited space available for instrumentation during experiments is also a challenge. To accommodate the challenge, INL has expanded its temperature detection instrumentation capabilities paired with the advancement of in-house advanced manufacturing options to include printed melt wires for peak irradiation temperature measurement. The advanced manufactured melt wire package can determine peak temperatures in experiments that can fit the challenging space available during irradiation. Recent results and knowledge from this ongoing work have made it possible to develop unique temperature sensors that can accomplish user specific temperature ranges inside the harsh environments experienced during radiation testing. This report summarizes the initial temperature testing using fabricated melt wires with two different melting temperatures, silver (melting temperature of ~960 °C) and solder (melting temperature of ~160 °C), that were initially tested for their relative melting points using non-standard method of furnace and hot plate testing under normal air conditions. Future work will include more controlled way of testing printed melt wires under inert gas with multiple temperature measurements followed by comparison of standard melt wire package performance for baseline characterization.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2025

Integrating HPC simulations and physical experiments to characterize the effects of gamma radiation on seismic protective devices

Seismic protective systems, composed of seismic isolators and dampers, can substantially reduce the effects of earthquake shaking on nuclear power plants and components therein. To enable the use of these devices to protect equipment inside a plant and close to a source of radiation, the U.S. Department of Energy (DOE) funded a project at the Idaho National Laboratory (INL) and the University at Buffalo to characterize the effects of absorbed gamma dose on their mechanical properties. An early task in the project was to determine the exposure time required in the INL Foss Therapy Services (FTS) 60 Co gamma irradiator to achieve a target absorbed dose in the materials used to construct isolators and dampers, including fluids, polymers, composites, and metals. This task required the novel integration of high-performance computing (HPC), Monte Carlo N-Particle (MCNP) simulations, and irradiation experiments using Fricke dosimetry. An MCNP model of the FTS irradiator at INL was developed and validated using Fricke dosimetry. Simulations of three experiments in the irradiator, two with Fricke vials only and one with Fricke vials and a large-size isolator, predicted the Fricke-measured absorbed dose rate to within 15% in all three cases, providing high confidence in the calculation of the gamma dose absorbed in the materials comprising the seismic protective devices. The simulations demonstrated that the effects of photon scattering on absorbed dose rate in the FTS irradiator are negligible for test articles installed close to the cobalt sources and near the rear of the irradiator. The validated MCNP model of the FTS irradiator is being used to support ongoing DOE-funded experiments on seismic protective devices and could be applied to future, non-seismic-related experiments. In conclusion, the novel validation process successfully deployed for the FTS irradiator at INL could be applied to other irradiators, requiring new MCNP models and simulations, and irradiation experiments using dosimeters.

42 - ENGINEERING↗

NASA NTRS1985

Spectroscopic analysis of radiation-generated changes in tensile properties of a polyetherimide film

The effects of electron radiation on Ultem, a polyetherimide were studied for doses from 2 x 10 to the 9th power to 6 x 10 to the 9th power rad. Specimens were studied for tensile property testing and for electron paramagnetic resonance and infrared spectroscopic measurements of molecular structure. A Faraday cup design and a method for remote temperature measurement were developed. The spectroscopic data show that radiation caused dehydrogenation of methyl groups, rupture of main-chain ether linkage, and opening of imide rings, all to form radicals and indicate that the so-formed atomic hydrogen attached to phenyl radicals, but not to phenoxyl radicals, which would have formed hydroxyls. The observed decays of the radiation-generated phenoxyl, gem-dimethyl, and carbonyl radicals were interpreted as a combining of the radicals to form crosslinking. This crosslinking is the probable cause of the major reduction in the elongation of the tensile specimens after irradiation. Subsequent classical solubility tests indicate that the irradiation caused massive crosslinking.

Long, E. R., Jr.↗

DOE OSTI2020

NDE Technology Engineering Program for Hanford DST Non-Visual Volumetric Inspection Technology: Phase II RAVIS Radiation Tolerance Test Report

This test report provides the results of radiation tolerance robustness testing that was performed on samples of robotic components and an ultrasonic guided wave air-slot sensor that represent components/sub-systems of the Robotic Air-slot Volumetric Inspection System (RAVIS) that has been engineered for volumetric inspection of Hanford tank bottom plates via under-tank refractory pad air-slots. The specific components tested for 1) functionality during active irradiation and 2) tolerance to cumulative radiation dose (until failure or upon reaching a cumulative dose test limit) were: • four samples each of a printed circuit board (PCB) and direct current (DC) motor, which are robotic components, and • 26 ultrasonic piezoelectric elements (samples) inside an air-slot sensor. The robotic components are part of the RAVIS air-slot inspection crawler drive control system that is responsible for remote communication with and actuation of the air-slot inspection crawler. The failure of either of these components during under-tank deployment would require manual retrieval via the crawler’s tether, which risks damage to the robot/refractory/tank. Preemptive replacement of the components at appropriately conservative dose/time intervals informed by failure dose would reduce the likelihood of under-tank failure. The components were included in radiation tolerance testing to quantify their failure doses to inform replacement intervals. The air-slot sensor is responsible for collecting ultrasonic inspection data (scan images) for the tank bottom plates during under-tank deployment. Compromised signal quality due to elevated noise levels caused by gamma radiation would compromise inspection performance. The air-slot sensor was included in radiation tolerance testing to quantify the impact of active irradiation on sensor signal quality. The irradiation and in-situ functional tests of the PCBs, DC motors and air-slot sensor took place in June and July 2020 at the Pacific Northwest National Laboratory. Testing was performed at a gamma dose rate near 300 rad/hr., which, in the absence of under-tank dose rate data, has been conservatively estimated to be the upper-bound dose rate beneath the primary tanks at Hanford. Irradiation took place at elevated temperatures of 150-200°F to determine failure doses that reflect the compounding effects of gamma radiation and heat. The test results revealed: • The DC motors can tolerate being actively irradiated at the high dose rate at 200°F and can tolerate a cumulative dose of 300,000 rad, that which would be incurred after 5 years of service at the 300 rad/hr dose rate. The component therefore meets minimum and preferred radiation tolerance and lifecycle requirements for robotic components. • The air-slot sensor can tolerate being actively irradiated at the high dose rate at 150°F and can tolerate a cumulative dose of 60,000 rad, that which would be incurred after 1 year of service at the 300 rad/hr dose rate. The sensor therefore meets minimum radiation tolerance and lifecycle requirements. • The PCB can tolerate being actively irradiated at the high dose rate, but can only tolerate a cumulative dose of 19,000 rad at 150-200°F. The PCB does not meet minimum radiation tolerance and lifecycle requirements; however, because the component is considered replaceable, it can be replaced before a cumulative dose of 19,000 rad is reached, determined through either monitoring with a dosimeter or scheduled time intervals that are calculated based on conservative estimates of under-tank dose rates.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2024

Contribution of Lobe Power to Experiment Heating in the Advanced Test Reactor

In order to ease the computational burden associated with designing irradiation experiments in the Advanced Test Reactor (ATR), scaling factors are often used to estimate design parameters at different lobe powers. Here, this paper examines the validity of long-standing assumptions about the contribution of lobe power to total experiment heating in the ATR. For each of the ATR’s 77 different experiment positions, the fractional contribution of each of the ATR’s five lobes to the total heating in that position is calculated and compared to traditional assumptions. The updated fractional contributions are then used to scale heating rates in a sample problem, and the results are compared to traditional scaling methods as well as explicit MC21 heating calculations. It is concluded that for experiment locations in close proximity to the ATR driver fuel (i.e. flux traps and the A, H, and B positions), heating rates scaled with the updated fractional contributions generally agree better with explicit MC21 calculations than do heating rates scaled using the traditionally assumed contributions. For the I positions, which are located on the very periphery of the ATR core, both scaling methods led to poor results when compared against explicit calculations due to the effect that movement of the outer shim control cylinders has on the experiment heating in those positions.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗