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All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Corrosion Testing Needs and Considerations for Additively Manufactured Materials in Nuclear Reactors

The Advanced Materials and Manufacturing Technologies (AMMT) program within the Department of Energy, Office of Nuclear Energy (DOE-NE) has developed its current recommendation for its corrosion testing strategy to deploy additively manufactured (AM) materials in advanced nuclear reactors. Additive manufacturing technologies have developed rapidly in recent years, creating new opportunities and challenges for the nuclear industry. To adopt AM technologies, the corrosion performance of AM materials needs to be adequately evaluated.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Photographic Mitigation Documentation Related to the Engine Assembly and Disassembly Facility, Area 25, Nevada National Security Site, Nye County, Nevada

This letter report provides documentation in partial fulfillment of the mitigation of adverse effects to the Engine Maintenance Assembly and Disassembly (E-MAD) facility, a major component of the Nuclear Rocket Development Station (NRDS) Historic District located in Area 25 of the Nevada National Security Site (NNSS). Construction of the E-MAD facility began in 1962 for the purpose of disassembling and maintaining nuclear-powered rocket engines. Once the building was complete in 1967, the E-MAD facility remained in continuous operation until the nuclear rocket program was terminated in 1973. The mothballed facility was repurposed in 1977 for spent nuclear fuel handling and packaging activities. By the mid-1980s, all activities at the E-MAD facility had ceased (Reno et al. 2019). This submission is intended to comply with Stipulation III.D.1 of the Memorandum of Agreement DE-GM58-20NA25534 between the U.S. Department of Energy and the Nevada State Historic Preservation Officer Regarding Corrective Action Activities and Demolition of the Engine Maintenance Assembly and Disassembly Facility and the Test Cell C Historic District, Major Components of the Nuclear Rocket Development Station Historic District Located in Area 25 at the Nevada National Security Site Nye County (hereafter referred to as the MOA). Stipulation III.D.1 states the U.S. Department of Energy (DOE) will submit the existing supplementary collection of the large-format, black-and-white photographic prints taken during the 1996 Historic American Engineering Record (HAER) documentation of the E-MAD facility to the Nuclear Testing Archive (NTA), located on the second floor of the Rogers Building at Desert Research Institute (DRI) in Las Vegas, Nevada.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2023

National Reactor Innovation Center Update

Accelerate the testing and demonstration of advanced nuclear technology by providing access to national laboratory assets and expertise: Establish & maintain four new experimental facility capabilities MSTEC – 2025; METL – Operational; HeCTF – Operational; Creep Frames – 2026 Construct and operationalize two large reactor test beds for integrated technology demonstrations and experimentation - 2028, Complete minimum of two advanced nuclear technology tests – 2030 Advanced Construction Technology – 2025 Advanced Microreactor test in DOME – 2027 & 2029

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

An Overview in the Development of a Multi-Sensor Data Science System for Monitoring a Solvent Extraction Process

To enhance nuclear nonproliferation stewardship, Idaho National Laboratory is building the Beartooth nuclear fuel cycle processing test bed. The Beartooth test bed will allow researchers the opportunity to study nuclear fuel processing operations including the use of centrifugal contactors in solvent extraction processes. The test bed is designed to support data collection and machine learning to monitor process operations. As part of this project, researchers will study data collected from a host of sensors that have not been typically used to monitor solvent extraction processes such as vibration, acoustic, colormetric, and thermal measurement data. The goal of this research is to employ machine learning and data analytics to study the confluence of signals collected from both traditionally and non-traditionally used sensors to provide operator process awareness and discover process anomalies. An overview of planned sensors and experiments that will focus on signal discovery will be presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

An Overview in the Development of a Multi-Sensor Data Science System for Monitoring a Solvent Extraction Process

To enhance nuclear nonproliferation stewardship, Idaho National Laboratory is building the Beartooth nuclear fuel cycle processing test bed. The Beartooth test bed will allow researchers the opportunity to study nuclear fuel processing operations including the use of centrifugal contactors in solvent extraction processes. The test bed is designed to support data collection and machine learning to monitor process operations. As part of this project, researchers will study data collected from a host of sensors that have not been typically used to monitor solvent extraction processes such as vibration, acoustic, colormetric, and thermal measurement data. The goal of this research is to employ machine learning and data analytics to study the confluence of signals collected from both traditionally and non-traditionally used sensors to provide operator process awareness and discover process anomalies. An overview of planned sensors and experiments that will focus on signal discovery will be presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

M3AS-25IN1002073: Analysis of data from irradiation testing of printed strain gauges in prototypic nuclear environments

Advancement in additively manufactured strain gauges help address critical technology gaps to accurately monitor real-time materials behavior in reactor experiments. This is critical as it provides data to inform predictive models and simulations that enhance the development of reactors and fuel cycle systems. In this report, additively manufactured strain gauges are exposed to a neutron irradiation environment at the Ohio State University Research Reactor. This report goes over a 2-week campaign for neutron irradiating printed resistive strain gauges and capacitive strain gauges. These results complement the prior separate effects (i.e., mechanical strain, temperature, etc.) testing that were performed on these additive manufactured sensors and presented in prior milestone reports. These results also help progress our understanding of their usage in harsh environment applications. The outcome of developing advanced sensing and instrumentation capabilities plays an important role in increasing the safety, reliability, and energy efficiency of both next-generation and existing nuclear reactors.

36 - MATERIALS SCIENCE↗
DOE OSTI2025

Accelerated Irradiation Testing and Post-Irradiation Characterization: U.S.-Based Capabilities for Advanced Nuclear Systems and Radioisotope Production

Irradiation experiments and post-irradiation examinations, together referred to as irradiation testing (IRT), are prerequisites for nuclear fuel and material qualification for the deployment of new and advanced reactors, as well as radioisotope production, thereby ensuring regulatory compliance. Qualified research and test reactors (RTRs) and testing facilities are essential to enable IRT to verify performance and safety under prototypical reactor conditions. In the past, qualification of new fuels or structural materials required about 20 years. Synergist strategies, advanced tools, and qualified methods are needed to greatly reduce this timeframe of IRT and radioisotope production. This study, termed accelerated-IRT, focuses on identifying gaps and leveraging U.S.-based RTRs and material testing capabilities, leveraging the preliminary evaluation and qualification of selected RTRs to provide a generic as well as specific-case solution paths forward, ensuring adherence to stringent regulatory standards. Furthermore, IRT and radioisotope production utilizing qualified RTRs necessarily includes modeling and simulation to support the design (i.e. neutronics, thermal, and structural aspects) and manufacturing of irradiation test specimens, vehicles, capsules, apparatuses, and flow loops. In addition, IRT can be improved by applying advanced manufacturing techniques and in-pile sensors and instrumentation, as discussed in this study.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Spectroscopy of states in 136 Ba using the 138 Ba(p,t) reaction

Background: The 136 Ba isotope is the daughter nucleus in 136 Xe $ββ$ decay. It also lies in a shape transitional region of the nuclear chart, making it a suitable candidate to test a variety of nuclear models. Purpose: To obtain spectroscopic information on states in 136 Ba, which will allow a better understanding of its low-lying structure. These data may prove useful to constrain future 136 Xe → 136 Ba neutrinoless $ββ$ decay matrix element calculations. Methods: Here, a 138 Ba(p, t) reaction was used to populate states in 136 Ba up to approximately 4.6 MeV in excitation energy. The tritons were detected using a high-resolution Q3D magnetic spectrograph. A distorted wave Born approximation analysis was performed for the measured triton angular distributions. Results: 102 excited states in 136 Ba were observed, out of which 52 are reported for the first time. Definite spin-parity assignments are made for 26 newly observed states, while previously ambiguous assignments for ten other states are resolved. Together with other available data, the results are used to determine level densities in 136 Ba. These were compared with theory predictions, obtained using shell model calculations with Hamiltonians previously used for 136 Xe neutrinoless $ββ$ decay matrix element evaluations. Conclusions: The shell model predicted level densities agree reasonably well for the two Hamiltonians. However the results for theory and experiment are found to agree only at lower energies, diverging from one another for the higher lying states, with the discrepancy increasing with energy. This is presumably because of lower production cross sections for a majority of the higher-lying predicted states and the experimental limitations in resolving a large number of nearly degenerate states predicted by the theory.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

NRAD Program Presentation

This is a PowerPoint presentation highlighting the Idaho National Laboratory's Neutron Radiography Reactor, NRAD, programmatic capabilities, neutron radiography quality assurance requirements, and examples of resulting training, informational, and typical radiographic capabilities for nondestructive testing of irradiated nuclear reactor fuels.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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↗
NASA NTRS1972

Nuclear light bulb propellant heating simulation using a tungsten/argon aerosol and radiation from a dc arc surrounded by a segmented mirror cavity

Experiments were conducted to simulate radiant heating of the propellant stream of a nuclear light bulb engine to obtain high bulk exit temperatures in the flowing simulated propellant stream by absorption of large fractions of the incident thermal radiation. A high-power, vortex-stabilized dc arc within an uncooled fused silica tube was used as the radiant energy source. It was surrounded by a mirror system to increase the radiation incident on the simulated propellant. The 12.7-cm-long by 2.3-cm-wide, diverging-duct test section had a transparent front wall and a reflecting rear wall. The central stream of seeded gas, a tungsten-particle/argon aerosol, had unseeded argon buffer layers on both sides to prevent coating of the duct walls. Arc operating times were approximately 0.5 sec with power levels up to 780 kW. Bulk exit temperatures were measured using a calorimeter downstream of the duct. The maximum simulated propellant bulk exit temperature obtained was 4515 K, compared with 3860 K in previous tests, 3300 to 3700 K expected in in-reactor tests in the Nuclear Furnace, and 6660 K in the reference nuclear light bulb engine. The maximum temperature in these tests was limited primarily by the amount of radiation incident on the test section (determined by the arc operating characteristics and the effectiveness of the mirror cavity).

Klein, J. F.↗
DOE OSTI2021

Performance-based simulants for Hanford radioactive waste treatment process testing

Radioactive wastes from nuclear fuel processing are stored in large underground storage tanks at the Hanford Nuclear Reservation. Treatment and remediation require that waste feed from the storage tanks be delivered to the Hanford Tank Waste Treatment and Immobilization Plant continually for the duration of the treatment mission. Furthermore, the complex physical and chemical processes required for this mission, the significant scale, and the hazardous nature of the waste necessitate the use of simulants in process testing. A new class of simulants, “performance-based simulants,” has been developed to match the process performance of the simulant to actual waste performance data.

Hanford tank waste treatment↗
DOE OSTI2023

Mechanical properties of Zircaloy cladding tubes and contributions to M.E.T.A. mechanical property database

To support a multi-laboratory Methodology, Evaluation, Testing, and Analysis (M.E.T.A.) cladding properties database, Oak Ridge National Laboratory’s (ORNL’s) cladding mechanical test geometries were manufactured from several nuclear-relevant cladding alloys and subsequently tested. These geometries were developed as mechanical test specimens to evaluate the properties of tube materials that may be used for irradiation testing at ORNL’s High Flux Isotope Reactor. They may also be used as test articles to be harvested—via in-cell machining—from commercially irradiated fuel rods and later tested. This report explores the differences among axial, hoop, and SSJ tensile geometries with partially recrystallized Zircaloy-2 to test ORNL correlation-based methods on a plate material that approximates, to the greatest extent possible, the characteristics of nuclear industry tubing. Furthermore, several tests were conducted with ORNL’s Zircaloy-4 tube inventory to (1) develop material properties as a standard for future tests, (2) determine the effect of the US Department of Energy’s Advanced Fuels Campaign coating processes on tube mechanical properties, and (3) evaluate the effect of specimen machining methods on the mechanical properties of tube geometries.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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 OSTI2020

Investigation of the irradiation effects in additively manufactured 316L steel resulting in decreased irradiation assisted stress corrosion cracking susceptibility

While additive manufacturing (AM) is an area of increasing interest to the nuclear industry, there are significant research and development needs prior to successful implementation of additively manufactured components into reactors. AM steel behaves differently than traditional wrought material, as has been observed in numerous mechanical testing studies. The nuclear industry must also understand the material response to irradiation. This work examines the irradiation damage and the irradiation assisted stress corrosion cracking behavior (IASCC) of both proton irradiated AM and wrought 316L steel. For IASCC testing, specimens were strained to 4% plastic strain in a simulated boiling water reactor environment (288°C, 0.2 µS/cm). Transmission and scanning electron microscopy, confocal microscopy, and X-ray computed tomography were used to examine coupons of the material as well as tensile bars used in the IASCC testing. This work found that the wrought 316L was more susceptible to IASCC than the AM steel, however, some IASCC was observed in the AM steel when tested with the tensile axis parallel to the build direction. Electron microscopy has shown significantly fewer radiation induced voids formed in the AM steel. While the pre-existing pore network in the AM steel may act as a sink to point defects, this is not expected to be the major factor in the reduced swelling observed in the AM steel. The lower density of radiation induced voids was expected to contribute to the IASCC resistance by decreasing the localization of deformation in the irradiated specimens, however, severity of localized deformation in AM steel, measured in dislocation channel spacing and height, was similar or more severe to that found in wrought. In conclusion, the presence of severe cracking in the wrought specimens did not allow for a direct comparison of localized deformation severities between the AM and wrought steel.

Additive manufacturing↗
DOE OSTI2020

Report Summarizing Boiling Transition (Dryout) Testing of FeCrAl Cladding

Nuclear fuel and cladding in a light water reactor must maintain integrity under all normal operational and anticipated operational occurrences. In a Boiling Water Reactor (BWR) some Anticipated Operational Occurrences (AOOs) and Anticipated Transients Without SCRAM (ATWS) events have been known to result in cladding damage due to dryout with conventional Zircaloy-2 claddings. This work presents an assessment of commercially-fabricated, accident tolerant iron-chromium-aluminum (FeCrAl) alloy C26M cladding in such events with regard to the ballooning failure mode for unirradiated cladding In the severe accident test station (SATS), an experiment has been developed to gauge accident tolerance of fuel cladding materials under simulated cyclic dryout conditions. Thermal cycling from 300°C to 650°C was implemented on C26M and Zircaloy-2 tubes. The C26M tube did not fail after 54 cycles. The Zircaloy- 2 tube run under the same parameters as the C26M tube failed during the fourth cycle, consistent with expectations from the literature. Due to thermal gradients present in the furnace, both tubes experienced a maximum temperature at ~780°C. Due to differences in tube geometry, the hoop stress of the C26M tube was significantly higher than that of the Zircaloy-2 tube, 76 MPa at the maximum peak temperature compared to 55 MPa at burst, respectively. The cyclic dryout experiments show an increased performance of C26M over Zircaloy-2 under simulated conditions, an indication of C26M possessing superior high temperature oxidation resistance, mechanical properties and potentially resistance to fatigue.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Hydrogen Recombiner Catalyst Evaluations for Waste Storage

Radiolysis of water in nuclear waste storage generates hydrogen gas that can accumulate within sludge style waste and be rapidly released during agitation events, creating a significant flammability hazard. Engineering controls are therefore required to limit hydrogen concentrations during both quiescent storage and transient disturbances. Catalytic recombination of hydrogen in waste storage offgas is a proven mitigation strategy, maintaining hydrogen levels below flammability limits and managing sudden concentration spikes. Conventional recombiners rely on platinum and/or palladium catalysts, with development efforts focused on extending service life, increasing active surface area, and ensuring safe deployment in radioactive environments. Savannah River National Laboratory (SRNL) is evaluating a newly developed hydrogen recombiner catalyst from Canadian Nuclear Laboratories as a cost-effective and durable alternative for nuclear waste applications. Testing was conducted in SRNL’s Shielded Cells facility, which enables reduced-scale experimental modeling under radiation fields and near-use-case conditions relevant to radioactive waste storage. Catalyst performance was evaluated using a custom offgas characterization system designed for near-zero flow conditions. The experimental apparatus consisted of a gas-tight 2.7 L PTFE vessel equipped with temperature monitoring, gas flow controls, and a variable-speed mixer to simulate sludge agitation. Offgas composition was monitored using a dedicated gas chromatograph with argon carrier gas and a krypton internal standard. Measurements were obtained for an empty vessel, the vessel containing a well characterized radioactive tank waste sample, and the same configuration with the candidate catalyst installed. Results demonstrate that the new catalyst effectively reduced hydrogen concentrations in the offgas within the constraints of the experimental design. In addition to confirming catalytic activity, the testing provided valuable insights into experimental optimization and considerations for future performance evaluations. These findings support the potential scalability of the technology and highlight its applicability to broader nuclear waste management operations, offering improved safety and reduced operational costs through enhanced catalyst durability and lower replacement frequency.

Tener, Zachary P. [Savannah River National Laborat↗
DOE OSTI2023

National Reactor Innovation Center Annual Report FY 2023

It is my privilege to present this year’s annual report for the National Reactor Innovation Center (NRIC). Now rounding out our fourth year of operation, NRIC is poised for successfully delivering our most foundational promise: to accelerate the testing and demonstration of advanced nuclear technology by providing access to national laboratory assets and expertise. Some specific goals include: • Establish and maintain four new experimental facility capabilities: o MSTEC – 2025 o METL – Operational o HeCTF – Operational o Creep Frames – 2026 • Construct and operationalize two advanced reactor test beds (DOME and LOTUS) for integrated technology demonstrations and experimentation – 2028 • Complete minimum of two advanced nuclear technology tests – 2030 o Advanced Construction Technology – 2025 o Advanced Microreactor test in DOME or LOTUS – 2027 & 2029

22 GENERAL STUDIES OF NUCLEAR REACTORS↗