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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 235 records · Page 13
All sourcesNASA NTRSDOE OSTI
NASA NTRS2011

Advanced Fusion Reactors for Space Propulsion and Power Systems

In recent years the methodology proposed for conversion of light elements into energy via fusion has made steady progress. Scientific studies and engineering efforts in advanced fusion systems designs have introduced some new concepts with unique aspects including consideration of Aneutronic fuels. The plant parameters for harnessing aneutronic fusion appear more exigent than those required for the conventional fusion fuel cycle. However aneutronic fusion propulsion plants for Space deployment will ultimately offer the possibility of enhanced performance from nuclear gain as compared to existing ionic engines as well as providing a clean solution to Planetary Protection considerations and requirements. Proton triggered 11Boron fuel (p- 11B) will produce abundant ion kinetic energy for In-Space vectored thrust. Thus energetic alpha particles "exhaust" momentum can be used directly to produce high ISP thrust and also offer possibility of power conversion into electricity. p- 11B is an advanced fusion plant fuel with well understood reaction kinematics but will require some new conceptual thinking as to the most effective implementation.

Chapman, John J.↗
DOE OSTI2024

Multi-principal element alloys for fast reactor cladding applications

Given the extensive list of multi-principal element alloys (MPEAs) within literature and the overwhelming number of alloys that can be made from only a handful of elements, this article proposes a methodology for prioritizing alloys for use as cladding within advanced reactors. This paper applies neutronic, mechanical, and chemical assessments to a collection of MPEAs available within literature for use as advanced reactor cladding. The results are compared to a reference design of HT9, a ferritic/martensitic steel, for employed in sodium-cooled fast reactor. Mechanical assessments determined pressure limits for a thin-walled tube pressure boundary, thus relating the material's mechanical properties to a minimum acceptable wall thickness. Neutronic analyses reveal a maximum allowable wall thickness that a given material must meet to provide a level of neutronic economy that is equivalent to than that of HT9. Lastly, each alloying element is compared to typical fission products found in a fast reactor fuels to mitigate deleterious phenomena such as fuel-cladding chemical interactions (FCCI). These analyses indicate that Mo-Nb-Ti-V-based alloys are likely to be advantageous and that the inclusion of elements with a high neutronic penalty (e.g., Hf) could be considered with minimal consequences.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Development of a Representative Molten Chloride Fast Reactor Model to Assess the Impact of Nuclear Data

The SCALE code system was employed to conduct a preliminary investigation of nuclear data impacts for a fast spectrum molten chloride salt reactor. A computationally effective depletion model that is representative of the reactor system was successfully developed and used to conduct fuel depletion simulations. Development of this model draws from the SLICE method that was developed at Oak Ridge National Laboratory to enable generation of fuel compositions for an advanced reactor core at equilibrium operation. Eigenvalue uncertainty calculations using the ENDF/B-VII.1 nuclear data library were performed for the reactor in the fresh fuel state and an irradiated fuel state. It was determined that the primary driver of eigenvalue uncertainty was the uncertainty in the 235U (n, 𝛾) cross section. Uncertainty calculation results from this study were compared to results available for a different fast system, a sodium-cooled fast reactor, to confirm similarities and identify differences with respect to nuclear data impacts between the two fast advanced reactor systems.

Hirji, Rakim [Georgia Institute of Technology]↗
DOE OSTI2025

Idaho National Laboratory Team Visits NCERC

The DOE/NRC Criticality Safety for Commercial-Scale High-Assay Low-Enriched Uranium (HALEU) for Fuel Cycle and Transportation (DNCSH) initiative is a collaboration between the Department of Energy and the Nuclear Regulatory Commission focused on criticality safety benchmarking for processing and transportation applications related to HALEU and Advanced Reactors. The project includes design, siting and construction of a horizontal split table (HST), which has long been recognized as a need for validating advanced reactor systems.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

ARCADE Technical Pathway and Industry Impact

The Advanced Reactor Cyber Analysis and Development Environment (ARCADE) simplifies the evaluation and assessment of robustness factor and cyber resilience that support secure-by-design for advanced reactor nuclear power plants. In this manner, ARCADE supports risk-informed performance based (RIPB) evaluations of cybersecurity through its integration of plant physics with high-fidelity emulations of control systems. This cross domain approach enables comprehensive analysis of control system sensitivities, cyber-attack scenarios, and their consequences. ARCADE has been custom developed to meet the demands identified in Tier 1 of the Tiered Cyber Analysis (TCA) as outlined in NRC Draft Regulation Guide (RG) 5.96, which provides a RIPB cybersecurity approach for new reactors.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2024

Field Programmable Gate Array-Based Reactor Protection Systems and Potential for Inclusion of Secure Elements to Improve Cybersecurity

For acceptable implementations of technologies like wireless communications, remote monitoring, etc., strong mitigations must be developed and evaluated to ensure that new attack pathways do not increase risk for Advanced Reactors. Secure Elements can be adopted and adapted for this purpose based on tamper resistance and cryptographic abilities, but research must be done to properly integrate into critical components such as FPGA-based Important to Safety systems in conjunction with current and future regulations on cyber security features in Advanced Reactors. Typically, the integration of a Secure Element happens during the POST and UEFI boot of a computing platform, performed by the Operating System, which is not possible with FPGAs because they do not include these firmware components. Work must be done to identify a reliable and secure method for integration in FPGA-based systems which lack Operating Systems and therefore complex boot procedures, system calls, etc.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2024

Field Programmable Gate Array-Based Reactor Protection Systems and Potential for Inclusion of Secure Elements to Improve Cybersecurity

For acceptable implementations of technologies like wireless communications, remote monitoring, etc., strong mitigations must be developed and evaluated to ensure that new attack pathways do not increase risk for Advanced Reactors. Secure Elements can be adopted and adapted for this purpose based on tamper resistance and cryptographic abilities, but research must be done to properly integrate into critical components such as FPGA-based Important to Safety systems in conjunction with current and future regulations on cyber security features in Advanced Reactors. Typically, the integration of a Secure Element happens during the POST and UEFI boot of a computing platform, performed by the Operating System, which is not possible with FPGAs because they do not include these firmware components. Work must be done to identify a reliable and secure method for integration in FPGA-based systems which lack Operating Systems and therefore complex boot procedures, system calls, etc.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

The Advanced Scale Up Reactor Experiment (ASURE) Facility: A Testbed for Advancing the Art of Biomass and Waste Co-Gasification Systems

Powerpoint slides for an oral presentation at the 2024 Clearwater Clean Energy conference. It includes a brief discussion of the plans to construct a small pilot-scale co-gasification test facility known as the Advanced Scale Up Reactor Experiment (ASURE) Facility at NETL’s Morgantown, WV site in support of FECM’s goals for advancing biomass and waste co-gasification systems for carbon-neutral hydrogen production.

Rowan, Steven↗
DOE OSTI2024

The Advanced Scale Up Reactor Experiment (ASURE) Facility: A Testbed for Advancing the Art of Biomass and Waste Co-Gasification Systems.

This paper was an executive summary for an oral presentation to be included in the conference proceedings of the 2024 Clearwater Clean Energy conference. It includes a brief discussion of the plans to construct a small pilot-scale co-gasification test facility known as the Advanced Scale Up Reactor Experiment (ASURE) Facility at NETL’s Morgantown, WV site in support of FECM’s goals for advancing biomass and waste co-gasification systems for carbon-neutral hydrogen production.

Rowan, Steven↗
DOE OSTI2025

Supporting New Advanced Nuclear Technologies for Commercial-Maritime Applications

The ANS summary doesn't require an abstract, but I will produce one for the purpose of LRS: The large demand for maritime nuclear power underscores the need for experimental campaigns and modeling and simulation of new advanced reactors, which offer numerous advantages in terms of safety, efficiency, and compactness. INL, through the work conducted by NRIC and ABS, has addressed some of the technical, regulatory, and economic aspects of potential nuclear commercial maritime applications. However, on the technical side, there remain important physical phenomena, particularly for advanced reactors, that are not yet fully understood. Addressing these knowledge gaps requires a combination of experiments and advanced modeling and simulation techniques. INL possesses significant expertise in Multiphysics modeling and simulation. By collaborating with INL, the maritime nuclear sector can leverage this expertise to advance the development and deployment of innovative nuclear technologies.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1973

An evolutionary approach for a compact-split-core reactor

An economical approach for advanced reactor power development is presented, and systems that result from the several stages of this plan are described. The development starts with a highly modularized heat pipe, radioisotopic design and evolves into a low specific weight high performance reactor system.

Breitwieser, R.↗
NASA NTRS1973

An evolutionary approach for a compact-split-core reactor.

An economical approach for advanced reactor power development is presented, and systems that result from the several stages of this plan are described. The development starts with a highly modularized heat-pipe, radioisotopic design and evolves into a low-specific-weight, high-performance reactor system.

Breitwieser, R.↗
DOE OSTI2026

Report on the Update and Codification of Expert Knowledge into ASTM Standard Practice E521

The overall purpose of the Grand Challenge Integrated Research Project (IRP), Accelerating the Qualification of Materials to Enable Rapid Deployment of Advanced Reactors, is to fulfill several objectives. The first objective is to complete and demonstrate/establish the process for predicting the microstructure and properties of structural materials in reactor and at high doses using ion irradiation as an accurate predictive tool for assessing behavior under reactor irradiation. The second objective is for this process to be adopted as part of an ASTM standard, and to work with the US Nuclear Regulatory Commission to utilize this standard in licensing decisions on advanced reactor designs for which such data are nonexistent or impractical to achieve. The need for standardized procedures is apparent based on the outcomes of a recent round-robin experiment from laboratories using ion beams for radiation damage studies and the roadmap for ion beam technologies to address challenges for the advancement of nuclear energy technologies.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Bayesian calibration of irradiated graphite property models under high temperatures

Graphite under high temperatures and irradiation is central to advanced reactors. We develop a Bayesian calibration framework for graphite property models that explicitly represents model-data mismatch via a Gaussian-process discrepancy. The approach propagates uncertainty from parameters, experimental noise, and model form, with a hierarchical variance structure to capture group and cross-group noise. Using two predictive models across five grades (IG-110, NBG-18, PCEA, NBG-17, 2114) and four properties-irradiation-induced dimension change, creep, Young’s modulus change ratio, and coefficient of thermal expansion change ratio-we obtain average predictive-error reductions of 54%, 65%, 17%, and 17% when discrepancy is included. We illustrate engineering impact with a multiphysics model of a very-high-temperature reactor prismatic reflector brick, analyzing stresses under high fluence and temperature. Accounting for model discrepancy markedly improves predictive accuracy and provides a robust basis for reliable graphite component design in advanced reactors.

36 - MATERIALS SCIENCE↗
DOE OSTI2024

Design of a first-of-A-kind instrumented advanced test reactor irradiation Capsule experiment for In situ thermal conductivity measurements of metallic fuel

Metallic fuel undergoes dramatic microstructural changes early in life due to fission gas swelling until ~2–3 at% burnup which affects the conductivity of the material, however the evolution of metallic fuel thermal conductivity during this early phase burnup has never been successfully measured in situ. The Irradiated Material Properties Accelerated Characterization Test (IMPACT) experiment will be the first in a series of experiments to irradiate advanced nuclear metallic fuel specimens with novel embedded thermal conductivity probes in ATR. In the current work the IMPACT experiment final design and supporting analysis is reported in detail. Results are evaluated for various reactor operational conditions to meet the functional requirements of the experiment. Finally, the first iteration of this IMPACT experiment will provide data regarding thermal properties evolution in uranium-zirconium (U10Zr) fuel, but this experiment vehicle is envisioned for future advanced fuels and structural materials irradiations in ATR.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Generation IV International Forum – Knowledge Management and Knowledge Preservation: An initiative from the Education and Training Working Group

Dr. Patricia Paviet is the Chair of the Generation IV International Forum – Education and Training Working Group. As such, she is overseeing and directing several initiative: The GIF webinar series and the Pitch your Gen IV Research competition. This group has launched a new initiative on knowledge management and knowledge preservation of advanced reactor systems and how lessons learned are passed on in an international context. The paper is a summary of the various GIF member countries’ strategies to preserve and transfer the knowledge of advanced reactors.

96 KNOWLEDGE MANAGEMENT AND PRESERVATION↗
NASA NTRS2020

Modeling Electrolytic Conversion of Metabolic CO2 and Optimizing a Macrofluidic Electrochemical Reactor for Advanced Closed Loop Life Support Systems

The International Space Station (ISS) is currently equipped with a complex, heavy, and power consuming system that recovers approximately 50% of O2 from metabolic CO2. Future long duration missions will require a sustainable and highly efficient system capable of yielding a minimum of 75% O2 recovery. A Macrofluidic Electrochemical Reactor (MFECR) technology development effort is currently underway at NASA Marshall Space Flight Center (MSFC) to significantly increase current O2 recovery efficiency and reduce complexity of the system. This paper presents a comprehensive multi-physic 3D model developed at MSFC on CO2 conversion to O2 and C2H4 at standard conditions via MFECR. The 3D spatial domain of the model is a replica of the actual MFECR’s 3D drawing generated for the MFECR fabrication and operated to recover O2 from CO2 yielding C2H4 as byproduct. Electrochemical (EC) physics that includes EC multicomponent reaction mechanisms, mass transport, and current density distributions is coupled in the model with all the other physics phenomena involved in the process, such as free and porous fluid flow, multicomponent mass transfer, heat transfer, and DC electrical current generation along with Joule heating effect. The authors plan to use experimental results to validate this comprehensive and rigorous model and build a reliable simulator that will not only assist the authors on the MFECR design but also optimize its operation.

Dominguez, Jesus A.↗