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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
NASA NTRS2011

Differential Processing of Low and High LET Radiation Induced DNA Damage: Investigation of Switch from ATM to ATR Signaling

The members of the phosphatidylinositol kinase-like kinase family of proteins namely ataxia-telangiectasia mutated (ATM) and ATM- and Rad3-related (ATR) are directly responsible for the maintenance of genomic integrity by mounting DDR through signaling and facilitating the recruitment of repair factors at the sites of DNA damage along with coordinating the deployment of cell cycle checkpoints to permit repair by phosphorylating Checkpoint kinase Chk1, Chk2 and p53. High LET radiation from GCR (Galactic Cosmic Rays) consisting mainly of protons and high energy and charged (HZE) particles from SPE (Solar Particle Event) pose a major health risk for astronauts on their space flight missions. The determination of these risks and the design of potential safeguards require sound knowledge of the biological consequences of lesion induction and the capability of the cells to counter them. We here strive to determine the coordination of ATM and ATR kinases at the break sites directly affecting checkpoint signaling and DNA repair and whether differential processing of breaks induced by low and high LET radiation leads to possible augmentation of swap of these damage sensors at the sites of DNA damage. Exposure of cells to IR triggers rapid autophosphorylation of serine-1981 that causes dimer dissociation and initiates monomer formation of ATM. ATM kinase activity depends on the disruption of the dimer, which allows access and phosphorylation of downstream ATM substrates like Chk2. Evidence suggests that ATM is activated by the alterations in higher-order chromatin structure although direct binding of ATM to DSB ends may be a crucial step in its activation. On the other hand, in case of ATR, RPA (replication protein A)-coated ssDNA (single-stranded DNA) generated as a result of stalled DNA replication or during processing of chromosomal lesions is crucial for the localization of ATR to sites of DNA damage in association with ATR-interacting protein (ATRIP). Although the majority of RPA-coated ssDNA is generally present only during DNA replication, ATR activation in G1 and G2-phase might still require formation of RPA-coated ssDNA, probably initiated by the MRN-CtIP complex and then extended by the Exo1- or BLM-dependent mechanisms at the sites of DSBs. Evidence accumulates that activation of ATM and ATR are oppositely regulated by the length of single stranded overhangs generated at the break sites by processes mentioned above and these stretches of single stranded overhangs hold the clue for ATM to ATR switch at broken DNA ends. We irradiated 82-6hTERT human fibroblast cells with low LET gamma-rays and high LET Fe and Si particles. Preliminary results with cells exposed to 1Gy gamma-rays show that the kinetics of pChk2-pT68 foci formation is comparable to that of gamma-H2AX although they appear to recede quicker. The number and intensity of observed foci reaches a maximum at 30 min and 60 min post IR for Chk2-pT68 and gamma-H2AX foci respectively and all Chk2-pT68 foci colocalize with gamma-H2AX foci. The kinetics of Chk1-pS345 and ATRIP are being determined. Results of Chk2-pT68 foci kinetics was also corroborated by western blot experiments, although phosphorylation was detected as early as 10 min and started receding 30 min post IR with 2Gy of gamma-rays. On the other hand, level of ATR-pS428 reached its maximum between 60 and 120 min and was maintained until the last measured time point of 4 hours post IR as determined by western blotting. Experiments performed with high LET Fe and Si particles will be reported.

Saha, Janapriya↗
DOE OSTI2024

Neutron resonance transmission analysis prototype system for thorium fuel cycle safeguards

Emerging thorium-based reactor designs and fuel cycles present challenges to traditional non-destructive assay techniques used in international safeguards. Specifically, assaying the masses of 233 U and 235 U when they are present together in samples with high gamma ray backgrounds is difficult because of similar passive neutron signatures and relatively weak gamma-ray emissions of 233 U. The Pacific Northwest National Laboratory (PNNL) and the Massachusetts Institute of Technology (MIT) are developing a compact neutron resonance transmission analysis (NRTA) system as one potential solution to these challenges. The NRTA technique provides isotopic information for a sample via neutron time-of-flight (TOF) measurements that exploit a sample’s epithermal neutron resonance cross-sections. A recently developed portable NRTA system uses a commercially available, pulsed deuterium-tritium neutron generator with a ~2 m flight path and a GS20 lithium glass scintillator detector. Finally, this paper describes the prototype NRTA system design, a refined radiation transport model of the system, preliminary measurements with thorium and uranium sources, and demonstration of a quantitative isotopic estimation algorithm.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Preliminary studies on diversion and misuse for TRISO-fueled heat-pipe-cooled microreactors

Microreactors are being developed by multiple reactor designers to mass-produce nuclear facilities for worldwide deployment. Including microreactors in a country’s energy portfolio introduces new concerns for international safeguards regarding the number of reactor locations. Microreactors have smaller quantities of nuclear material compared to current light-water reactors, which makes diverting enough material for clandestine purposes difficult for a single reactor. Diversion or misuse of multiple microreactors to obtain a significant quantity introduces a new acquisition pathway. This work examines diversion and misuse scenarios for a realistic heat-pipe-cooled microreactor to understand how these scenarios would effect reactor operations to determine if monitoring key parameters could reduce the burden of inspection on monitoring agencies. Here, we determined misuse caused too drastic of an effect on core operations to be a valid acquisition pathway for the realistic heat-pipe-cooled microreactor. Diversion was explored across varying levels of severity, where potential diversion scenarios could yield one significant quantity of material from between six and eleven microreactors. Through examining the critical control drum angle, excess reactivity, control drum worth, and power distribution changes from nominal were detected over the operational lifetime, which could indicate a divergence from normal operations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

International Nuclear Security Nuclear Material Accounting and Control Instructions for Use and Supplementary Documentation

This document describes the use of exercise kits (called NMAC Kits) developed to support training and technical engagements sponsored by DOE/NNSA Office of Nuclear Security (INS) in the use of nuclear material accounting and control (NMAC) methodologies in support of nuclear security. INS has previously used other similar kits that were originally designed to support NMAC training for international safeguards purposes. These new kits are specifically designed to be used in nuclear security training and emphasis the role NMAC plays in nuclear security as well as security against the insider threat. Planning for the development of these kits is described in LA-UR-24-30063, FY24 NMAC Kits Upgrade and served as the initial framework for the development of the new NMAC Kit material in FY25.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2025

Assessment of Self-interrogation Safeguards Signatures for Pebble Bed Reactor Fuel

New reactor designs, such as pebble-bed reactors, present challenges for the safeguarding of fissile material. Due to the multi-pass fuel circulation design and the non-uniform path a pebble may traverse through a pebble-bed reactor, there will be variations in the irradiation history of spent fuel pebbles. The standard approach of estimating actinide quantities based on fission gamma spectra or neutron emission paired with depletion modeling may yield uncertainties too large for safeguards and material accountancy. This project investigated, through modeling and experimentation, the potential of neutron self-interrogation of spent fuel pebbles as an innovative method to implement materials accountability. As an example, our feasibility studies indicate that the mass of U-235, U-238, Pu-239 and Pu-241 can be predicted to 4.1, 0.86, 13 and 13 % accuracy, respectively, when measuring 100 closely packed end-of-life spent fuel pebbles over approximately 12 days using a 4π counting geometry.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Physical Security Model Development of an Electrochemical Facility

Nuclear facilities in the U.S. and around the world face increasing challenges in meeting evolving physical security requirements while keeping costs reasonable. The addition of security features after a facility has been designed and without attention to optimization (the approach of the past) can easily lead to cost overruns. Instead, security should be considered at the beginning of the design process in order to provide robust, yet efficient physical security designs. The purpose of this work is to demonstrate how modeling and simulation can be used to optimize the design of physical protection systems. A suite of tools, including Scribe3D and Blender, were used to model a generic electrochemical reprocessing facility. Physical protection elements such as sensors, portal monitors, barriers, and guard forces were added to the model based on best practices for physical security. Two theft scenarios (an outsider attack and insider diversion) as well as a sabotage scenario were examined in order to optimize the security design. Security metrics are presented. This work fits into a larger Virtual Facility Distributed Test Bed 2020 Milestone in the Material Protection, Accounting, and Control Technologies (MPACT) program through the Department of Energy (DOE). The purpose of the milestone is to demonstrate how a series of experimental and modeling capabilities across the DOE complex provide the capabilities to demonstrate complete Safeguards and Security by Design (SSBD) for nuclear facilities.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Safeguards Material Control And Accountancy Considerations For Molten Salt Reactors

Molten salt reactors are a class of nuclear reactor designs with features and operational characteristics that vary significantly. Variations in design features include the physical, chemical, and isotopic composition of fresh and irradiated fuel, neutron energy spectrums, breeding ratios, and the design’s plans for chemical processing. Specific designs within these variations affect the material control and accountancy approaches implemented by both domestic and international nuclear safeguards. This paper identifies safeguards-relevant design features for molten salt reactors and describes how each feature might influence both domestic and international nuclear safeguards implementation. Additionally, this paper describes two potential general categories of material control and accountancy approaches that may be appropriate for molten salt reactors: a black box material-balance approach with measurements on the feed and waste streams combined with robust containment and surveillance on the nuclear material within the reactor or an approach that aims to quantify the nuclear material throughout the reactor (i.e., a process monitoring approach). This paper describes the benefits and challenges of these approaches to meet technical safeguards objectives and identifies gaps in current technologies.

Hogue, Karen↗
DOE OSTI2024

Nontraditional Sensors for Aqueous Separation Research & Workforce Development

Idaho National Laboratory’s (INL) nuclear fuel cycle capabilities enable the deployment of technologies that sustain the current reactor fleet, support demonstration and deployment of new advanced reactors, and facilitate management and disposition of existing and future radiological waste materials. INL focuses on deploying nuclear energy systems with confidence by decreasing proliferation risk through research that demonstrates process transparency and supports safeguards and security by design. An example of these capabilities is the Beartooth test bed, set to begin operations toward the end of fiscal year 2026. Beartooth will include a cascade of centrifugal contactors, glove box lines, and solidification and dissolution equipment to aid in the progression of novel separation techniques and to provide hands-on experience to cultivate and maintain a robust workforce of experts. To support Beartooth’s enhanced instrumentation and monitoring equipment needs, several nontraditional sensors are being considered for future deployment. The non-traditional sensors include accelerometers, acoustic microphones, and infrared cameras. These nontraditional sensors have the potential to not only help monitor the process but also enhance nuclear safeguards.

98 - NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL↗
DOE OSTI2021

Material Control & Accountancy for Molten Salt Reactors (FY2021 Report)

There is significant domestic and international interest, investment, and research and development momentum to pursue advanced nuclear reactor technologies. Molten salt reactor (MSR) concepts display the largest variability in fuel type and design features among the current advanced concepts. MSRs have been proposed with various core designs, sizes (power), and fuel cycles. Salt-fueled molten salt systems represent the only advanced reactor type with fuel that is not in a solid form during operation. These “liquid-fueled” MSRs are unique from perspectives of fuel fabrication, spent irradiated fuel and waste components, licensing, and material control and accountability (MC&A) including the potential of fissile material holdup. The liquid fuel salt is the defining distinction in comparison to other advanced reactors that propose TRI-structural ISOtropic particle fuel pebbles, various coolant options (e.g., molten salts or metals, high temperature gas), or small modular alternatives using solid fuel variants including both light water reactors and non-light water reactors. MSRs are appealing to the nuclear energy industry because of the diverse reactor characteristics they can support including various neutron energy spectra, fueling requirements, fuel cycles, and/or fuel utilization. However, because of the significant deviation and diversity of a salt-fueled system compared to traditional solid fuel light water-cooled reactors (LWRs), the history, regulatory licensing framework, modeling capabilities, and supporting engineering technology are either lacking or, in some cases, nonexistent. Therefore, the research community is actively supporting advanced MSR development on many of these fronts in particular to assist MSR vendors with licensing requirements. ORNL is leading the research and development of respective MC&A approaches for salt-fueled MSRs. This report summarizes the research performed at Oak Ridge National Laboratory (ORNL) under the US Department of Energy, Office of Nuclear Energy, Advanced Reactor Safeguards (ARS) program to investigate safeguards and security by design concepts, licensing and regulatory considerations, and dynamic system-level modeling to understand radioisotope concentrations for salt-fueled MSRs. The report builds upon the previous research and literature, identifies the MC&A challenges inherent to a salt-fueled MSR, reviews current regulatory frameworks for LWRs and their applicability towards salt-fueled MSRs, summarizes the status and progress of an MSR dynamic modeling tool, and discusses a prospective MC&A approach based on the Molten Salt Demonstration Reactor (MSDR) model.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Inspecta Technical Report

Sandia National Laboratories (SNL) has developed Inspecta (International Nuclear Safeguards Personal Examination and Containment Tracking Assistant), an AI-powered smart digital assistant (SDA) equipped with robotic capabilities. This innovative tool aims to enhance the effectiveness, efficiency, and safety of international nuclear safeguards inspections. Inspecta is designed to assist inspectors on-site by supporting or automating tasks that are often mundane, hazardous, or prone to errors. In 2021, the development team established the specifications for Inspecta by thoroughly analyzing International Atomic Energy Agency (IAEA) documents and consulting with former IAEA inspectors and subject matter experts. This process involved aligning in-field inspection tasks with existing commercial and open-source technologies, thereby creating a roadmap for the initial prototype of Inspecta and identifying areas requiring further research and development. From 2022 to 2025, the focus shifted to integrating a critical inspection activity—the examination of seals—into an early version of Inspecta. This phase has involved the development of both software and hardware capabilities essential for this task. This report outlines the advancements in Inspecta’s functionalities, particularly those that support the seal examination process.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2021

Misuse Detection for a Generalized SFR Test Reactor

Sodium-cooled Fast Reactors (SFRs) present unique challenges for international safeguards. SFRs possess neutron physics characteristics that if configured appropriately could produce more fissile material than consumed. An adversary state may choose to build an SFR, justified by a lack of domestic natural uranium and limited access or interest to procuring uranium from international markets. Once constructed, the state may choose to misuse the SFR for the purpose of diverting fissile plutonium from declared operation. This work shows that a demonstration SFR does not need to be configured as a plutonium breeder to create one Significant Quantity (SQ) of plutonium in a short amount of time (e.g., one to few years). However, such an extreme case of misuse would change the core reactivity in such a way as to be easily indicated by deviations of control rod position compared to declared operation. In this work a contrived SFR demonstration reactor was modeled for the purpose of exploring proliferation scenarios and how such misuse could be detected using the SFR's Reactor Data Acquisition System (RDAS). Typically, the International Atomic Energy Agency (IAEA) does not have access to the control rod position, power, thermal, pressure sensing and indicating systems of nuclear power plants. However, this work shows that such data streams can be compared against a parallel detailed simulation model (a Digital Twin) to detect possible misuse.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2023

Nuclear Material Control & Accounting for Pebble Bed Reactors (FY 2023 Summary Report)

This report discusses the work done under the US Department of Energy NE-5 Advanced Reactor Safeguards and Security Program during FY 2023. It provides a summary of material control and accounting (MC&A) for pebble bed reactors (PBRs) and addresses some of the main challenges with current PBR MC&A approaches that will inform safeguards and security by design efforts. The efforts to date have focused on tristructural isotropic (TRISO) pebble fuel material accounting and control including working with partners in industry, loss and production of nuclear material as part of reactor operations, burnup modeling and measurements, uncertainty quantifications for such modeling and measurements, statistical approaches needed, and measurement methods. The unique fuel management and utilization in a PBR, where the fuel in spherical form is introduced and circulates through the reactor, poses special challenges for MC&A. This contrasts with traditional water-cooled reactors in which the fuel is contained in large assemblies and can be easily identified and counted. Even online fueled reactors, such as the CANDU reactors (none of which operate in the United States), are significantly different because the fuel is still contained in relatively large assemblies, is uniquely identified, and the number of assemblies that pass through the core on an annual basis is much fewer than the hundreds of thousands that circulate in a PBR, none of which are uniquely identified. Additionally, the nature of the TRISO fuel results in very low heavy metal loading with each pebble containing less than 10 g of uranium and on the order of less than 1 g of fissile material. This low fuel density and the robustness of the TRISO particles are major features of the TRISO fuel from a safety basis as each TRISO particle and pebble acts as a containment for the nuclear material and fission products during normal and accident conditions. This also results in very low plutonium loading per pebble during normal operations, which is on the order of 0.1 g at full burnup. A major feature of PBRs is that they will allow for significantly higher burnup, on the order of 160 GWd/THM compared to the burnup of traditional LWRs, which is on the order of 45 GWd/THM. This is achieved by monitoring the pebbles as they circulate through the reactor and allowing them to be reintroduced into the core until the desired burnup is achieved and they are removed from the reactor and enter the spent fuel storage areas.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1973

Effects of weightlessness in man.

The program for the Apollo 16 flight was designed to include both safeguards against and investigations of the physiological problems arising from increase in the period of manned space flight. Precautions included the provision of a controlled diet with high potassium content, carefully controlled work loads and work-rest cycles, and an emergency cardiology consultation service, and investigations were made to enable preflight vs postflight comparisons of metabolic, cardiovascular, and central nervous system data. Results of these investigations indicate that adjustment to weightlessness can be satisfactorily assisted by appropriate countermeasures, including attention to diet.

Berry, C. A.↗
DOE OSTI2020

Cooperative Research and Development Agreement (CRADA) NFE-18-07194 with TerraPower LLC (Final Report)

Because of the potential economic and safety benefits of the molten salt reactor (MSR) concept, development of several designs has been initiated around the world over the past decade. New international nuclear safeguards needs and verification challenges are likely to arise because of the commercial interests in MSRs and the number of MSR design variants. As a result, work was undertaken to explore the cross-cutting issues specific to the application of international nuclear safeguards (i.e., safeguards) to liquid-fueled molten salt reactors (LFMSRs). Through a public-private partnership, a report was developed that focuses on the TerraPower Molten Chloride Fast Reactor (MCFR) design that has received a funding award from the US Department of Energy (DOE), Office of Nuclear Energy. The report is intended to provide a preliminary analysis for a safeguards-by-design (SBD) effort to inform designers about how safeguards could be applied to LFMSRs by the International Atomic Energy Agency (IAEA).Although the report specifically focuses on the TerraPower design, the conclusions are applicable to the main design features of LFMSRs and can be used to extrapolate how existing IAEA safeguards measures for other fuel cycle facilities can be appropriately applied or modified. The report evaluates the appropriate safeguards approaches for LFMSRs, presents existing safeguards inspection technologies are still valid for LFMSRs, and identifies new challenges that will require novel measurement instruments to meet verification standards of the IAEA and the international safeguards regime. This document summarizes the results of the work performed under the full report developed as part of the Cooperative Research and Development Agreement (CRADA). This summary does not contain any protected CRADA information and is intended for public release.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Technical report Letter: RAFM, ODS steels and MMLC for Nuclear energy application

The lifetime, thermodynamic efficiency, safety and economic viability of new generation fission and fusion reactor concepts can largely be tied to the mechanical performance and stability of structural alloys under extreme environments. In this context, engineered nano materials could have broad-reaching impact on the future of advanced nuclear fuel-cycle and reactors. These systems are characterized by a large number density of interfaces which are efficient sinks for point defects and moderately biased; therefore limiting the deleterious effects of irradiation. Broadly, nuclear nano-technology deals with the use of the latest engineered-nanomaterials for improving the nuclear power performances and safety in all areas of nuclear energy production to bring new generations of nuclear power units. New advanced fuel assembly designs also have implications for securities and safeguards. To support the readiness for potential future license applications, an understanding of the technologies that would enable new reactor designs in the areas of component performance and domestic safeguards is necessary. This technical report letter work explores the technical issues and potential regulatory considerations associated with developing and adopting fuel claddings made of advanced nano- materials. Specifically three classes of nanomaterials are considered: (i) reduced activation ferritic/martensitic (RAFM) steels, (ii)oxide dispersed steels (ODS) and (iii) multi-metallic layered composites (MMLC).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Evaluation of uncertainty in antineutrino spectra normalization calculations for advanced nuclear reactor monitoring

Antineutrino detection systems have been envisioned as an important aspect of safeguarding the next generation of nuclear reactors, especially considering designs utilizing exotic fuel cycles. Deployment of antineutrino detection systems for safeguarding applications is hindered by the uncertainties associated with the calculations required for antineutrino spectra measurements and predictions. The focus of this work is to assess the impact of system components on antineutrino spectra normalization uncertainties and their significance in reactor power monitoring sensitivity. The dominant limitation in antineutrino detection calculations is typically the uncertainty associated with a cosmogenic background. This limitation becomes more pronounced when signals are weak, although the issue is mitigated in larger reactors due to their stronger source strength. Additionally, antineutrino emission uncertainties vary with the isotopic composition of the reactor fuel. Unconventional fuel cycles, featuring less common fissioning isotopes, such as Pu-240, introduce larger antineutrino yield uncertainties. The findings from this study suggest that future research on safeguard-targeted antineutrino detection should prioritize background mitigation, particularly when background simulation is necessary. Advanced nuclear reactor designs have a major influence on background understanding and successful system implementation.

Physics↗