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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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All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Transformational Challenge Reactor Autonomous Control System Framework and Key Enabling Technologies

Autonomous operation of nuclear plants has the potential for significantly improving the economics of existing plants while maintaining operation within the safety envelope of the reactor. Autonomous operation technology, if developed and deployed, may enable deployment of small reactors at remote sites while requiring minimal operator support. In addition to reducing costs, the technology can add additional layers of safety through rapid decision making and actuation in off-normal situations. Within the context of the Transformational Challenge Reactor (TCR), autonomous control along with the other unique characteristics of the reactor (advanced manufacturing) is expected to enable rapid deployment. On the other hand, technology for autonomy can add risk to the system if not properly developed. There is therefore a need for systematic development and demonstration of autonomy technology for TCR. This document describes a framework for autonomy for the TCR. It defines autonomous operation and examines whether autonomy is possible for nuclear systems. The report also describes enabling sensing technologies for autonomy and describes a framework to facilitate autonomy.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Retooling a Trade Association in a Challenged Industry: The Nuclear Energy Institute Story - 18503

The U.S. Nuclear industry is facing significant challenges to maintain economic competitiveness in an electricity supply sector where advances in natural gas, wind, and solar technologies are reshaping the marketplace. As clean and reliable baseload generation, nuclear still has a valuable role to play. But in order for the power of the atom to remain available for meeting the nation's energy, environmental, and national security needs, it will be imperative that industry leadership innovate and become more efficient at a pace that matches the competition. Industry leaders rely, to a large extent, on their trade association - the Nuclear Energy Institute (NEI) - to be the prime mover in championing initiatives that will enable a successful future. NEI realized that, in the face of an increasingly difficult economic situation, it must become both leaner and more forward looking - it would have to extend its reach and influence while at the same time reigning in costs to its membership. Therefore, on January 17, 2017 NEI completed a restructuring to improve member value and organizational effectiveness. This restructuring was designed in recognition that the nuclear industry is at a pivotal time, and NEI had a unique opportunity to change its trajectory and underline the importance of nuclear energy's role in the nation's critical infrastructure. NEI is now better positioned to support the existing reactor fleet, promote commercialization of advanced reactors, and drive American leadership in commercial nuclear exports and technology around the world. NEI began this restructuring in the fall of 2016 when it consolidated Government Affairs, Communications, Policy and Public Affairs under a new External Affairs function. The January 2017 action extended that effort across the remainder of the organization, ensuring that the organization is highly aligned, member-focused and results driven. The new NEI takes pride in the fact that it is comprised of talented employees who are passionate about nuclear energy and can make a difference for our membership. With this team in place, NEI looks forward to advancing the interests of the nuclear industry while contributing to a safer, cleaner and more productive world. This paper will examine the details of how this new organization was constructed, provide an overview of the National Nuclear Energy Strategy it developed to guide its efforts, examine the trials and tribulations that come along with disruptive change, and reflect on what has been accomplished in the retooled organizations first year. (authors)

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2020

Non-destructive Examinations of ATF-2 Baseline Rodlets

This report contains the results of non-destructive examinations of Accident Tolerant Fuels 2 (ATF-2) rodlets irradiated in the Advanced Test Reactor (ATR) Loop 2A. The experiment is part of the U.S. Department of Energy Nuclear Technology Research and Development (NTRD) program’s Advanced Fuels Campaign (AFC). The rodlets were composed of UO2 pellets and Zr-4 cladding. The data have been collected to provide baseline PIE data to support future testing in TREAT and to have performance data against which the data of the ATF concepts can be compared directly. The analyses performed included: visual examinations, axial gamma scanning, neutron radiography and profilometry. All the data collected showed a performance consistent with expectations for this fuel system at low burnup.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Specimen sizing and remaining life sample calculations for high temperature reactor material surveillance

Advanced high-temperature nuclear reactors rely on structural components that will operate for decades under combined mechanical, thermal, and environmental loading. Materials surveillance programs are a promising strategy for managing the resulting uncertainty in long-term structural integrity by monitoring degradation in service using passively actuated mechanical test articles. Previous reports have developed a simplified, spreadsheetimplementable framework for sizing these test articles and for inferring accumulated creep damage and remaining life from ex-situ test data. This report advances that work toward practical deployment by providing a sample-problem book: a collection of worked, end-toend examples in which the ASME Section III, Division 5 design analysis of a representative high-temperature reactor component is carried through to a sized surveillance article, verified with detailed finite-element analysis, and concluded with a remaining-life assessment based on an assumed ex-situ creep-rate measurement on the retrieved specimen. The report also summarizes ongoing ANL engagement with ASTM Committee E10 on Nuclear Technology and Applications toward drafting a standard covering surveillance procedures for advanced reactors.

Barua, Bipul (ORCID:0000000247184113)↗
DOE OSTI2022

Safeguards Considerations for Coated Particle Fuel Fabrication Facilities

This study aims to identify where existing fuel fabrication safeguards are limited in application to coated particle fuel fabrication facilities; perform a review of techniques that may be appropriate to address these limitations; and identify where further study is required. The purpose is to ensure the necessary technology and safeguards developments are established to facilitate safeguards-by-design for coated particle fuel fabrication facilities, analogous to the research and development (R&D) efforts in the advanced reactor space, including those requiring coated particle fuels and progress to higher technology readiness levels. Finally, this study will also help meet high priority objectives set forth by the IAEA in establishing safeguards R&D precedence and measurement/procedural standards at emerging fuel fabrication facilities. To achieve these objectives, the report is formatted to (1) distill relevant background information on fuel fabrication of oxide and coated particle fuels; (2) provide an overview of the fuel fabrication techniques for a reference oxide fuel fabrication facility and two selected TRISO fuel fabrication facilities; (3) develop facility models of safeguards approaches for the reference facility and the two TRISO facilities; and (4) identify any observed technology needs or gaps from this analysis.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Complete Optimization of LPBF Ni-Based Alloys Down-Selected from FY23 Candidate Materials Including, Thermodynamic Modeling, Sample Fabrication and Microstructure Characterization

The goal of the Advanced Materials and Manufacturing Technologies (AMMT) program is to accelerate the incorporation of new materials and manufacturing technologies into advanced nuclear-related systems. Although 316H stainless steel fabricated by laser powder bed fusion (LPBF) has already been identified as an alloy that could have a significant effect on various reactor technologies, many other materials and manufacturing techniques are being evaluated. Nickel-based alloys typically offer higher-temperature capabilities compared with advanced stainless steels, and previous reports looked at three Ni-based alloy categories: low-Co alloys with a potential use close to the reactor core; high-temperature, high-strength alloys; and molten salt–compatible alloys. In the first category, alloy 718 was studied in 2023, and creep testing at 600°C and 650°C revealed that the alloy exhibited great creep strength after the appropriate annealing but had low ductility. Advanced characterization was recently conducted to highlight the presence of strengthening γ' and γ" precipitates after creep testing and to show that brittle phases at grain boundaries might explain the low ductility of LPBF 718 compared with wrought 718. For the high-temperature, high-strength alloys, previously purchased powders of alloys 617, 230, and 625 were used to assess the printability of these three solution-strengthened alloys. Hot cracking could not be suppressed for alloy 617 and 230, and it was shown that these cracks, which were elongated along the build direction (BD), had a drastic effect on the ductility of alloy 230 at room temperature when specimens were machined perpendicular to the BD. On the contrary, LPBF printing of crack-free alloy 625 was achieved using similar printing parameters, and the alloy looked like a promising candidate for various reactor technologies. The fabrication of alloy 282 by LPBF, a γ'-strengthened alloy with great creep strength up to 800°C, was performed in 2023, and x-ray computed tomography (XCT) scans of the alloy before and after creep testing at 750°C were carried out to assess the effect of flaws on the alloy’s creep behavior. Correlation between the flaws’ volume fraction, creep ductility, and creep lifetime could be established, and future work on LPBF 625 will take full advantage of in situ printing data and ex situ XCT scans to accelerate the alloy qualification. Finally, single track experiments were performed on the two alloys previously identified as good molten salt–resistant, Ni-based candidates: Hastelloy N and 244. Various laser parameters were considered, and cracking was not observed for either of the two alloys. Wrought 244 offers better creep strength and molten salt compatibility than alloy 625, and future work will aim to establish the alloy LPBF processing window.

36 MATERIALS SCIENCE↗
DOE OSTI2026

MARVEL Reactor Digital Engineering Developments

The MARVEL reactor project has served to introduce a new generation of engineers to the processes required to transform a reactor design from simply an idea on paper into what will be an approved, constructed, and operational nuclear power system. Much as there have been advances in materials, analysis, and evaluation methodologies over the 50 years since the last reactor was built at INL, so too has the technology for managing the engineering process itself advanced. Digital Engineering tools and methods provide improved coordination between previously siloed engineering disciplines, reduced burdens of non-value-added data transcription processes and bring forward insights and improvements that might otherwise fall later in the design stage, where changes are much more costly. While the tools and techniques to support the full digital engineering vision are not yet complete, the MARVEL design processes provide valuable demonstrations and validations of key aspects and illuminate further areas for implementation by subsequent projects.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2016

MOCUM Verification with a Heterogeneous MOX Whole Core C5G7 Benchmark Problem

Recent advancements in computer technology has allowed for the possibility to solve complex reactor problems without homogenization, but the present benchmark problems are all limited to small scale with few fuel assemblies and homogeneity at the core level. Therefore, to verify the capabilities of the advanced neutronics codes, heterogeneous full core model shall be developed. Recently several whole-core benchmark problems such as BEAVRS and VERA are developed with measured data to which computational codes can be compared. With a particular research interest on the MOX fuel, this study develops a two-dimensional (2-D) whole core MOX neutronics benchmark based on the C5G7 benchmark concept. The developed MOX core benchmark will then be used to verify the accuracy of the MOCUM code, which is developed using the method of characteristic (MOC) as the flux solver with an advanced geometry processor designed for arbitrary core geometries. MOCUM results will be compared against MCNP5 results as the reference solutions. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Assessing Government and Industry Roles to Accelerate Commercialization of Advanced Reactors and Supporting Fuel Cycles

Recent growth in private sector interest and investment in advanced reactor designs signals a new period in the greater than six decade pursuit of commercial nuclear technology. This nuclear entrepreneurship is primarily centered in the North America; meanwhile, national programs in China, France, India, Russia, and South Korea also feature active efforts to design, construct, operate and export advanced reactors .The business case for most private ventures calls for an aggressive commercialization timeline-tied to expectations for returns on investment. Finally, strategic energy portfolio projections and utility resource planning point to the 2030 - 2050 window as the period of greatest need for dispatchable, energy intensive, low-carbon generation options to address retirement of baseload capacity, increased penetration of renewables and economy-scale decarbonization. However, lead times for commercialization range from 30 to 50 years based on more traditional government-centric timelines. Fuel development and fuel cycle infrastructure deployment represent parallel timelines that directly influence the timeline for commercialization of many advanced reactor concepts. While multi-decade development periods appear to be commonplace for new nuclear projects, such timelines were not always the norm-three separate countries set out to develop civilian commercial power from fundamental nuclear principles, in just over a decade, on average. The US led light-water reactor (LWR) development in the 1950's, the UK pioneered a gas-cooled reactor (GCR) concept in the same time period, and Canada forged the path to commercializing the heavy-water reactor (HWR) starting in the mid-1950's and into the mid-1960's. The necessary fuel cycle facilities to support the commercial reactor design also did not appear suddenly, but were a collaborative effort between government and private partners. In this paper, the efforts of these three governments and their private industry partnerships in reactor and fuel cycle development are explored, and a select sample of representative milestones for each is presented, with the roles that each group played. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

High power ring methods and accelerator driven subcritical reactor application

High power proton accelerators allow providing, by spallation reaction, the neutron fluxes necessary in the synthesis of fissile material, starting from Uranium 238 or Thorium 232. This is the basis of the concept of sub-critical operation of a reactor, for energy production or nuclear waste transmutation, with the objective of achieving cleaner, safer and more efficient process than today’s technologies allow. Designing, building and operating a proton accelerator in the 500-1000 MeV energy range, CW regime, MW power class still remains a challenge nowadays. There is a limited number of installations at present achieving beam characteristics in that class, e.g., PSI in Villigen, 590 MeV CW beam from a cyclotron, SNS in Oakland, 1 GeV pulsed beam from a linear accelerator, in addition to projects as the ESS in Europe, a 5 MW beam from a linear accelerator. Furthermore, coupling an accelerator to a sub-critical nuclear reactor is a challenging proposition: some of the key issues/requirements are the design of a spallation target to withstand high power densities as well as ensure the safety of the installation. These two domains are the grounds of the PhD work: the focus is on the high power ring methods in the frame of the KURRI FFAG collaboration in Japan: upgrade of the installation towards high intensity is crucial to demonstrate the high beam power capability of FFAG. Thus, modeling of the beam dynamics and benchmarking of different codes was undertaken to validate the simulation results. Experimental results revealed some major losses that need to be understood and eventually overcome. By developing analytical models that account for the field defects, one identified major sources of imperfection in the design of scaling FFAG that explain the important tune variations resulting in the crossing of several betatron resonances. A new formula is derived to compute the tunes and properties established that characterize the effect of the field imperfections on the transverse beam dynamics. The results obtained allow to develop a correction scheme to minimize the tune variations of the FFAG. This is the cornerstone of a new fixed tune non-scaling FFAG that represents a potential candidate for high power applications. As part of the developments towards high power at the KURRI FFAG, beam dynamics studies have to account for space charge effects. In that framework, models have been installed in the tracking code ZGOUBI to account for the self-interaction of the particles in the accelerator. Application to the FFAG studies is shown. Finally, one focused on the ADSR concept as a candidate to solve the problem of nuclear waste. In order to establish the accelerator requirements, one compared the performance of ADSR with other conventional critical reactors by means of the levelized cost of energy. A general comparison between the different accelerator technologies that can satisfy these requirements is finally presented. In summary, the main drawback of the ADSR technology is the high Levelized Cost Of Energy compared to other advanced reactor concepts that do not employ an accelerator. Nowadays, this is a show-stopper for any industrial application aiming at producing energy (without dealing with the waste problem). Besides, the reactor is not intrinsically safer than critical reactor concepts, given the complexity of managing the target interface between the accelerator and the reactor core.

43 PARTICLE ACCELERATORS↗
NASA NTRS2004

Improved Nuclear Reactor and Shield Mass Model for Space Applications

New technologies are being developed to explore the distant reaches of the solar system. Beyond Mars, solar energy is inadequate to power advanced scientific instruments. One technology that can meet the energy requirements is the space nuclear reactor. The nuclear reactor is used as a heat source for which a heat-to-electricity conversion system is needed. Examples of such conversion systems are the Brayton, Rankine, and Stirling cycles. Since launch cost is proportional to the amount of mass to lift, mass is always a concern in designing spacecraft. Estimations of system masses are an important part in determining the feasibility of a design. I worked under Michael Barrett in the Thermal Energy Conversion Branch of the Power & Electric Propulsion Division. An in-house Closed Cycle Engine Program (CCEP) is used for the design and performance analysis of closed-Brayton-cycle energy conversion systems for space applications. This program also calculates the system mass including the heat source. CCEP uses the subroutine RSMASS, which has been updated to RSMASS-D, to estimate the mass of the reactor. RSMASS was developed in 1986 at Sandia National Laboratories to quickly estimate the mass of multi-megawatt nuclear reactors for space applications. In response to an emphasis for lower power reactors, RSMASS-D was developed in 1997 and is based off of the SP-100 liquid metal cooled reactor. The subroutine calculates the mass of reactor components such as the safety systems, instrumentation and control, radiation shield, structure, reflector, and core. The major improvements in RSMASS-D are that it uses higher fidelity calculations, is easier to use, and automatically optimizes the systems mass. RSMASS-D is accurate within 15% of actual data while RSMASS is only accurate within 50%. My goal this summer was to learn FORTRAN 77 programming language and update the CCEP program with the RSMASS-D model.

Robb, Kevin↗
DOE OSTI2025

Reactor System Safety: A case for New and Advanced Reactors

This presentation covers a generic overview of reactor system safety testing, analysis, and related research and development (R&D) focusing on new and advanced reactor systems. This presentation is prepared for graduate-level student seminar talks. No specific reactor design information is provided. Only publicly available information and related published articles and book contents are utilized.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2022

Advanced Reactor Control and Operations (ARCO): A University Research Facility for Developing Optimized Digital Control Rooms

The Advanced Reactor Control and Operations (ARCO) facility was constructed in January 2018 to serve as a test bed for advanced reactor control rooms and operator support systems. Since then, it has supported human-machine interface user experience research, fault detection and mitigation technology development, control room concept of operations development, and remote operations research. ARCO serves as the control room for the Compact Integral Effects Test (CIET) facility, which replicates the primary-side flow paths and thermal-hydraulic behavior of a fluoride-salt-cooled high-temperature reactor (FHR) using simulant fluids and scaling principles. New reactor designs feature different operating conditions and scenarios than those in existing reactors. ARCO supports the research and development of digital tools for operator communications, intuitive real-time data analysis, online health monitoring and prognostics, and control room cybersecurity. By integrating these different technologies, ARCO acts as a prototypical control system to iteratively develop methods and tools of operation in advanced small modular nuclear reactors. This paper describes the features of and challenges to operating advanced small modular reactors underlying the design basis for ARCO and its operator support systems.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Idaho national laboratory to demonstrate collaboration first versus competition to accelerate achieving a secure clean energy future by 2031

Idaho National Laboratory (INL) announced at COP27 it would reach net zero greenhouse gas (GHG) emissions by 2031. As a Nuclear, Energy and Environment, and National Homeland Security laboratory, the predominant solution to closing the clean energy gap will include nuclear as a safe, clean, reliable and affordable electricity source with the additional benefit of producing heat and hydrogen to fuel INL’s large transportation fleet. INL’s collaboration first vs. competition is essential to the program’s success. The focused actions in INL’s Nuclear Roadmap include: Infrastructure, Licensing/Regulatory, Financial, Time to Market, Fuel Cycle and Public Confidence/Communications. The roadmap also includes nuclear technology innovations and creative partnerships with utility providers, regulators, businesses, community members, and Indigenous Peoples to accelerate deployment of advanced reactors. Through development of the Net-Zero Nuclear Roadmap, INL will offer a model to provide safe and secure energy for the nation and the world by: (1) establishing the necessary infrastructure on its 890-square mile site to support demonstration, (2) showing proven pathways through the licensing and regulation process, (3) partnering with utilities to ensure commercial application, and (4) collaborating with industry to site new technologies.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Levels of Digitization, Digitalization, and Automation for Advanced Reactors

Much has been written about levels of automation (LOA), but comparatively little has been written about levels of digitization (LODi) and levels of digitalization (LODa). Digitization is a digital representation of analog information and is typical of migration from analog to digital control systems, digitalization involves enhancing the functionality of digital information, and automation changes control from humans to machines. Each of these technology implementations has its own scales, and each forms a viable type of functionality that should be considered not as a continuum toward automation but rather as separate categories of solutions that meet the needs of advanced reactors. In this paper we develop separate LODi, LODa, and LOA scales and demonstrate how conflation of these technologies, using the example of computer-based procedures, can lead to confusion in the design process. With the race to develop advanced reactors, the surest metric of success and safety is proper consideration of the right technology requirements for different control systems.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2012

Assessment of NDE Technologies for Detection and Characterization of Stress Corrosion Cracking in LWRs

Stress corrosion cracking (SCC) in light water reactors (LWRs) has been a persistent form of degradation in the nuclear industry. Examples of SCC can be found for a range of materials in boiling and pressurized water reactor environments, including carbon steels, stainless steels, and nickel-base stainless alloys. The evolution of SCC is often characterized by a long initiation stage followed by a phase of more rapid crack growth to failure. This provides a relatively short window of opportunity to detect the start of observable SCC, and it is conceivable that SCC could progress from initiation to failure between subsequent examinations when managed by applying periodic in-service inspection techniques. Implementation of advanced aging management paradigms in the current fleet of LWRs will require adaptation of existing measurement technologies and development of new technologies to perform on-line measurements during reactor operation to ensure timely detection of material degradation and to support the implementation of advanced diagnostics and prognostics. This paper considers several non-destructive examination (NDE) technologies with known sensitivity to detection of indicators for SCC initiation and/or propagation, and assesses these technologies with respect to their ability to detect and accurately characterize the significance of an SCC flaw. Potential strategies to improve SCC inspection or monitoring performance are offered to benefit management of SCC degradation in LWRs.

NDE↗
DOE OSTI2023

Industrial Requirements Status Report and Down-Select of Candidate Technologies

The report discusses the potential for implementing Integrated Energy Systems (IES) using advanced nuclear reactors to provide thermal and electrical energy to carbon-emitting industries while reducing their carbon footprint. Nine different industries were reviewed to determine their decarbonization potential, how nuclear power can fit in, and the carbon footprint of each. The report also identifies specific integration points and streams for most processes, including temperature and pressure requirements of integrating nuclear power with industrial processes. The report also provides insights on advanced nuclear reactors and their potential for us in IES for decarbonization. The target audience for the report is broad, but mainly addressed to the industry leads and nuclear reactor developers. The report recommends that in the short term, a few processes should be identified to continue research and investigation so that front end engineering design work can provide technical analysis for integrating nuclear heat and replacing existing carbon-emitting thermal sources. The work emphasized the significant decarbonization opportunity across each of the technologies looked at as well as how novel and useful nuclear integration is to reduce the environmental impact of these significant industries.

02 PETROLEUM↗