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All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

AGC-2 Specimen Post Irradiation Data Package Report

This report documents results of the post-irradiation examination material property testing of the creep, control, and piggyback specimens from the irradiation creep capsule Advanced Graphite Creep (AGC)-2 are reported. This is the second 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. The AGC-2 capsule was irradiated in the Idaho National Laboratory Advanced Test Reactor at a nominal temperature of 600°C and to a peak dose of 5 dpa (displacements per atom). One half of the creep specimens were subjected to mechanical stresses (an applied stress of either 13.8, 17.2, or 20.7 MPa) to induce irradiation creep. All post-irradiation testing and measurement results are reported with the exception of the irradiation mechanical strength testing, which is the last destructive testing stage of the irradiation testing program. Material property tests were conducted on specimens from 15 nuclear graphite grades using a similar loading configuration as the first AGC capsule (AGC-1) to provide easy comparison between the two capsules. However, AGC-2 contained an increased number of specimens (i.e., 487 total specimens irradiated) and replaced specimens of the minor grade 2020 with the newer grade 2114. The data reported include specimen dimensions for both stressed and unstressed specimens to establish the irradiation creep rates, mass and volume 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, electrical resistivity, and thermal diffusivity and thermal expansion data from 100–500°C. No data outliers were determined after all measurements were completed. A brief statistical analysis was performed on the irradiated data and a limited comparison between pre- and post-irradiation properties is presented. A more complete evaluation of trends in the material property changes, as well as irradiation-induced creep due to irradiation, temperature, and applied load on specimens will be discussed in later AGC-2 post-irradiation examination analysis reports.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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 OSTI2024

Status of the CERBERUS Evaluation for the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook

Modeling & Simulation (M&S) tools are used to analyze advanced reactor designs and the safety of current nuclear operations. As computers continue to improve, we are able to enhance resolution in our calculations. Therefore, the limitations of simulation capability are in the quality of data that is being used, including our ability to quantify the uncertainty and sensitivity of that data. In order to model systems of interest with increasing accuracy, the industry must improve key nuclear data measurements. The International Criticality Safety Benchmark Evaluation Project (ICSBEP) compiles and evaluates experiment data in a handbook that can be used by criticality safety engineers and others to validate computer codes and cross section libraries at nuclear facilities. Both critical and subcritical experiments are included in the handbook. These experiments, along with differential measurements, can help improve the quality of nuclear data. Concerns regarding the accuracy of Cu nuclear data have been published. The large values and trend of C-E for the Zeus intermediate energy benchmark, being one of the primary examples. Furthermore, very few experiments have been designed to be sensitive to Cu (as shown in Figure 1), so an integral, critical experiment is needed to help resolve these differences.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2025

Build-To-Replace Strategy to Reduce O&M Costs of Advanced Reactors

This paper targets the goal of reducing fixed operations and maintenance (O&M) costs to $2/MWh for advanced reactor (AR) designs, an order of magnitude reduction from current nuclear fleets. Such goal will be severely constrained if the underlying assumptions of O&M approaches and practices are not questioned and reexamined, especially today when changes can be implemented effectively and efficiently. Achieving a 90% reduction in AR O&M costs requires a completely new way of thinking – a paradigm shift – not through incremental, technology-focused approaches alone. Here, we address this challenge by evaluating the impact of moving to shorter design life for major structures, systems and components (SSCs) and shorter, more predictable refurbishment cycles as modeled by the commercial airline industry. This paper provides a brief overview of this different mindset to O&M applied to ARs: the build-to-replace approach. We provide a direct example of analysis of a build-to-replace scenario by identifying and evaluating scenarios for reduced system and component lifetimes and associated replacement and refurbishment schedules to evaluate impacts on O&M costs and other lifecycle elements such as SSC reliability.

97 - MATHEMATICS AND COMPUTING↗
DOE OSTI2025

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 OSTI2024

ANS-8 Nuclear Criticality Safety Consensus Standards - Current Initiatives

The nuclear criticality safety (NCS) consensus standards are developed as using rigorous procedures of the Standards Board of the American Nuclear Society. These procedures have been accredited by the American National Standards Institute, Inc., as meeting the criteria for American National Standards. The Nuclear Criticality Safety Consensus Committee (NCSCC) that approved all 18 NCS consensus standards is balanced to ensure that competent, concerned, and varied interests have had an opportunity to participate. The ANS-8 subcommittee (ANS-8) consists of 17 NCS experts with many years of experience as end users of ANS standards who serve on standard working groups to develop and maintain standards. ANS-8 ensures that the technical content of the standards is adequate for NCS community use. Attempts are made to ensure that ANS-8 consists of NCS professionals with a diverse range of experience such that all standards are applicable to as many sites as possible. ANS-8 is a very active subcommittee, and some active projects in progress are discussed in this paper: basis statement development for all standards, development of a glossary for consistency of definitions across all ANS-8 standards, and Considering the Criticality Safety Support Group (CSSG) Recommendation 2016-04 to the ANS Standards Board for changes in several ANS-8 standards.

96 KNOWLEDGE MANAGEMENT AND PRESERVATION↗
DOE OSTI2021

Paths Forward for Nuclear Energy: Using a Nationwide Post-Stratified Hierarchical Model to Facilitate Matching of New Nuclear Technologies to Receptive Host Communities

This project was a collaboration between the University of Oklahoma (OU) and the University of Michigan (UMich). The overall objective of the project was to address a critical problem facing the siting of nuclear facilities, where there is no mechanism for incorporating public attitudes at a state and local level into decision-making. Utilizing recent advances in data modeling over local spatial scales, the research team created a state and county level map of public attitudes towards nuclear energy. Current estimates for public acceptance of nuclear technologies primarily exists on a national level, and sub-national estimates of public opinion about reactor siting, research facilities, transport routes, or storage and disposal facilities are very scarce. Model results from this project provide a systematic basis for technology developers to evaluate public acceptance across alternative siting options as a critical component for weighing potential benefit-cost tradeoffs for alternative paths to deploy new nuclear facilities. The project utilized data from the largest database of nuclear attitudes in the US, coupled with Census and elections-related data as well as data on social vulnerability and proximity to current nuclear facilities. In addition to modeling public support, the team also developed an interactive map to showcase results, using an online tool called Public and Local Attitudes about Nuclear Energy Technologies (PLANET). The tool allows technology developers and stakeholders to interact with the model results and better understand variation in support for nuclear facility siting on a state and county level. The interactive nature of the tool also allows users to analyze results in a targeted way, by comparing levels of support across different states and/or counties as desired.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Molybdenum-99 from Molten Salt Reactor as a Source of Technetium-99m for Nuclear Medicine: Past, Current, and Future of Molybdenum-99

Technitium-99m ( 99m Tc), a widely used radioisotope, is used in tens of millions of medical diagnostic procedures annually. However, it is hard to store and must be immediately used upon production due to its short half-life (i.e., 6 h); thus, it is currently produced from 99 Mo, which itself is a result of 235 U fission. The majority of 99 Mo supplies to U.S. patients are currently provided by foreign producers and produced using highly enriched uranium (HEU). In order to minimize the proliferation risks of HEU-based medical isotope production, the U.S. Department of Energy’s National Nuclear Security Administration has funded a program to accelerate the development of technologies to produce 99 Mo without the use of HEU. Today, the global supply of 99 Mo depends on a limited number of nuclear reactors, and production has been interrupted unexpectedly since 2009 due to the fleet’s advanced age. Herein, alternative options for 99 Mo production are discussed, and one potential option is to obtain 99m Tc from molten salt reactors (MSRs). A MSR is a nuclear fission reactor that can operate at or close to atmospheric pressure with liquid fuel, which allows for producing isotopes in a timely manner. In this paper, the past and current production of 99 Mo via nuclear reactors is described, and the future of 99 Mo production by MSRs is discussed. The behavior and chemical properties of molybdenum in fluoride salts in MSRs and the possible extraction methods are also examined in addition to the limitation of current studies.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1972

Design, performance, and evaluation of a direct-current contactor for space nuclear electrical systems

A direct-current contactor for use in large space power systems was designed, built, and tested. It was developed to be operational in an environment of 540 C and at a pressure of 0.0001 N/sq m or lower. The contactor is rated to pass 10 A continuously and to interrupt a 20-A current at 10,000 V. It was tested to determine the corona threshold level and the leakage current at different temperatures. Also, it was tested for its closing and interruption ability.

Mueller, L. A.↗
DOE OSTI2022

High-Temperature Latent Heat Storage System using Transportable Heat Pipes for Versatile Integration with Emerging Microreactors

With the growing interest in the development of micro nuclear reactors and their application to microgrid, high-temperature thermal energy storage (TES) has also been receiving great attention as a technology to reinforce the energy utilization and profitability of this class of reactors. Given the importance and research needs of the TES technology, a research project is underway at Idaho National Laboratory (INL) to develop high-temperature TES that can be easily interfaced with emerging micro nuclear reactors. Currently under development is the Heat pipe-Integrated Thermal Battery (HITB), a novel latent heat storage design that enables flexible and robust integration with any design of currently considered micro nuclear reactors. The development of detailed design based on the proposed TES concept, HITB, requires not only development and validation of modeling and simulation capabilities, but also research that spans thermal performance analysis, component- and system-scale design optimization, and experimental demonstrations including high-temperature feasibility testing. The current research efforts (and plans) are divided into the following four areas: (i) Thermal and structural modeling and analysis at both system- and component-scales for evaluating performance and conducting design optimization, (ii) Small-scale experimental demonstrations for the key components of the proposed HITB design, (iii) Modelica-based dynamic model development for developing a control strategy under dynamic operation within integrated energy systems configuration, (iv) Integrated experimental testing and demonstration. This paper introduces the overall research efforts that have been and are being conducted to develop and demonstrate the novel design of the high-temperature latent heat storage device, HITB. The key design features, ongoing research efforts including the development of modeling and simulation capabilities, small-scale demonstrations of the HITB’s key functionalities prior to scaled-up experimental demonstration, are discussed. Then, this paper concludes with a remark on the future research plans for this project.

25 ENERGY STORAGE↗
DOE OSTI2023

Feasibility Study of Advanced Manufacturing Techniques and Compositions of High Entropy Alloys

In line with the objectives of Department of Energy’s Office of Nuclear Energy Advanced Materials and Manufacturing technologies program (AMMT), this work focuses on new materials development and qualification research and development for next-generation, high-temperature nuclear reactors. While the currently qualified nuclear materials have demonstrated significant irradiation resistance at high temperatures, the community is still interested in materials that can sustain the harsh environments found in nuclear reactors for temperatures up to 1,000°C for nearly 100,000 hours. High entropy alloys (HEAs) have the potential to serve in these extreme environments of next generation nuclear reactors due to their unique phase transformation pathways and nanoscale and mesoscale microstructures. The current work focuses on understanding such nuclear energy relevant HEAs through a detailed literature survey and developing a decision matrix with criteria for identification of HEAs with may have most impact and value for further examination.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Acoustic sensing and autoencoder approach for abnormal gas detection in a spent nuclear fuel canister mock-up

Currently, spent nuclear fuel (SNF) from commercial nuclear power plants is stored in stainless-steel canisters for interim dry storage. To provide an inert environment, these canisters are backfilled with helium after vacuum drying. However, the helium environment may be contaminated during extended storage because of the material degradation. For example, the heavier fission gas xenon may be released from the fuel rods into the canister cavity should the fuel cladding be breached. Other gases such as air and water vapor may also be present as a result of leakage caused by chloride-induced stress corrosion cracking on the canister walls or by insufficient vacuum drying. Therefore, monitoring the gas composition can provide critical information about the health of SNF canisters. In this study, noninvasive testing was conducted on a 2/3-scaled SNF canister mock-up using acoustic sensing. Ultrasonic transducers were placed on the exterior surface of the canister to probe the gas composition. A dataset was collected by sealing the canister mock-up and introducing up to 1.53% argon or 1.29% air into the helium background gas. Three methods were used to detect changes in the gas composition: the time-of-flight (TOF) method, the differential method, and the autoencoder method. Results showed that the TOF method had sufficient resolution to detect abnormal gas concentrations of less than 1.0%. The differential method demonstrated a periodic in-phase and out-of-phase behavior between the benchmark (i.e., pure helium) and abnormal (i.e., with argon or air) state signals. The variational autoencoder (VAE) and the Wasserstein autoencoder (WAE) were trained on the benchmark data and were applied directly to the abnormal state data. It was found that both the unsupervised VAE and the WAE were able to distinguish the benchmark and abnormal states of the canister mock-up based on the reconstruction error.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1992

The past as prologue - A look at historical flight qualifications for space nuclear systems

Currently the U.S. is sponsoring production of radioisotope thermoelectric generators (RTGs) for the Cassini mission to Saturn; the SP-100 space nuclear reactor power system for NASA applications; a thermionic space reactor program for DoD applications as well as early work on nuclear propulsion. In an era of heightened public concern about having successful space ventures it is important that a full understanding be developed of what it means to 'flight qualify' a space nuclear system. As a contribution to the ongoing work this paper reviews several qualification programs, including the general-purpose heat source radioisotope thermoelectric generators (GPHS-RTGs) as developed for the Galileo and Ulysses missions, the SNAP-10A space reactor, the Nuclear Engine for Rocket Vehicle Applications (NERVA), the F-1 chemical engine used on the Saturn-V, and the Space Shuttle Main Engines (SSMEs). Similarities and contrasts are noted.

Gary L Bennett↗
DOE OSTI2021

Global Market Analysis of Microreactors

Microreactors are smaller and simplified versions of nuclear power plants using current generation and advanced nuclear technology designs. The purpose of this report is to assesses the unique capabilities of microreactors and assess potential deployment in specific global markets in the 2030-2050 timeframe, with consideration for regulatory limits. The methods include developing unique microreactor deployment indicators and matching use cases to define future profile markets. Top-down and bottom-up analysis techniques were used to evaluate emerging market trends, derive a range of possible demands, and rank potential markets in 63 countries including current nuclear users and newcomer countries. Deployment is considered from regulatory and planning perspectives. Results indicate significant potential for global deployment of microreactors, but also significant challenges in achieving the technical capacities, meeting regulatory requirements and international accords, achieving competitive costs, and for gaining public acceptance. Profile microreactor markets were identified that include isolated operations, distributed energy, resilient urban applications, marine propulsion, and disaster relief. Market demands could be strong in Western Europe and possibly even the U.S., but particularly strong across Asia and Eastern Europe in isolated operations and distributed energy applications. Build rates in the hundreds of units by 2040 and thousands by 2050 would be needed to attain market penetration at scale.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRSDate not supplied

Ideas For Infusing In-Space Servicing, Assembly and Manufacturing Concepts into Nuclear Electric Propulsion Architectures

NASA is currently investigating nuclear electric propulsion (NEP) for human Mars transport within the space nuclear propulsion portfolio. NEP spacecraft have the following characteristics, they: 1) include very large structures (~100-meter length); 2) are comprised of many components/modules; and 3) have very long lifetimes (e.g., 50 years for fuel rods). Thus, NEP spacecraft can be classified as a “persistent asset,” which is any zero-g or planetary surface system that benefits from in-space assembly (ISA) or multiple visits for servicing, repairs, and upgrades. NEP spacecraft will benefit from taking advantage of, and incorporating, In-space Servicing, Assembly, and Manufacturing (ISAM) capabilities in the spacecraft architecture from the onset, enabling system maintenance, repair, and evolution. ISA has a long history of being proposed for, and studied as, a means for achieving large systems in space. More recently, the benefits of ISA have been recognized by NASA, the Department of Defense (DOD), other government agencies, and commercial space companies, and thus, ISAM is being actively pursued at a national level. Past and current strategies for achieving large structures in space have relied largely on two strategies; the first is to launch monolithic structures (designed to meet launch vehicle requirements for payload size and mass) that are docked or berthed to other monolithic structures on-orbit to form a larger structure (e.g., the International Space Station [ISS]); the second is folding and packaging large structures to fit inside a payload fairing and deploying the full-sized structure (unaided) once on-orbit (e.g., the James Webb Space Telescope [JWST]). To date, conceptual architecture studies performed for NEP spacecraft capable of human-rated Mars transport have only included a combination of the two previously mentioned strategies. This paper will propose ideas for infusing ISAM strategies into NEP vehicle architectures that leverage existing and near future technologies and enable the resulting NEP systems to be realized in a more time- and cost-efficient manner.

in-space assembly↗
NASA NTRSDate not supplied

Ideas for Infusing In-Space Servicing, Assembly, and Manufacturing Concepts into Nuclear Electric Propulsion Architectures

NASA is currently investigating nuclear electric propulsion (NEP) for human Mars transport within the space nuclear propulsion portfolio. NEP spacecraft have the following characteristics, they: 1) include very large structures (~100-meter length); 2) are comprised of many components/modules; and 3) have very long lifetimes (e.g., 50 years for fuel rods). Thus, NEP spacecraft can be classified as a “persistent asset,” which is any zero-g or planetary surface system that benefits from in-space assembly (ISA) or multiple visits for servicing, repairs, and upgrades. NEP spacecraft will benefit from taking advantage of, and incorporating, In-space Servicing, Assembly, and Manufacturing (ISAM) capabilities in the spacecraft architecture from the onset, enabling system maintenance, repair, and evolution. ISA has a long history of being proposed for, and studied as, a means for achieving large systems in space. More recently, the benefits of ISA have been recognized by NASA, the Department of Defense (DOD), other government agencies, and commercial space companies, and thus, ISAM is being actively pursued at a national level. Past and current strategies for achieving large structures in space have relied largely on two strategies; the first is to launch monolithic structures (designed to meet launch vehicle requirements for payload size and mass) that are docked or berthed to other monolithic structures on-orbit to form a larger structure (e.g., the International Space Station [ISS]); the second is folding and packaging large structures to fit inside a payload fairing and deploying the full-sized structure (unaided) once on-orbit (e.g., the James Webb Space Telescope [JWST]). To date, conceptual architecture studies performed for NEP spacecraft capable of human-rated Mars transport have only included a combination of the two previously mentioned strategies. This paper will propose ideas for infusing ISAM strategies into NEP vehicle architectures that leverage existing and near future technologies and enable the resulting NEP systems to be realized in a more time- and cost-efficient manner.

in-space assembly↗
DOE OSTI2021

A High-Assay Low-Enriched Uranium Fuel Transportation Concept

The uranium 235U enrichment commonly used in fuel production for U.S. light water nuclear reactors typically does not exceed 5 wt%. In contrast, many of the currently investigated advanced reactor concepts demand fuel with higher enrichments. This includes high-assay low-enriched uranium (HALEU), characterized by a 235U enrichment of 5 to 20 wt%. The necessity of HALEU transportation in the fuel production cycle leads to new challenges caused by various technical and regulatory hurdles. Current U.S. Nuclear Regulatory Commission–approved transportation package designs for UF6 with enrichments above 5 wt% provide relatively small payloads [=116 kg (250 lb)]. Furthermore, in accordance with 10 CFR 71.55, package design activities for fissile material enriched above 5 wt% need to consider water infiltration in the containment as part of the criticality safety evaluations. This study presents a transportation package concept for HALEU advanced nuclear reactor fuel with a significantly higher payload of up to 376 kg (830 lb) of fissile material per package and up to 1881 kg (4149 lb) of HALEU per legal weight truck. The anticipated chemical form of the transported material is UO2 downblended from available highly enriched uranium. The concept utilizes a combination of existing transportation packaging, 18 inner canisters, and a novel basket design that includes a borated aluminum flux trap. Criticality and shielding evaluations; fundamental structural, confinement, and thermal assessments; and studies on package operations are presented. The results of this study build significant confidence in the technical feasibility of a high-capacity HALEU transportation package concept while demonstrating the concept’s potential to meet U.S. regulatory requirements.

Eidelpes, Elmar↗