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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 1,261 records · Page 70
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Core Design Characteristics of the Fluoride Salt-Cooled High Temperature Demonstration Reactor

Fluoride salt-cooled high temperature reactors (FHRs) are a promising reactor technology option with significant knowledge gaps to implementation. One potential approach to address those technology gaps is via a small-scale demonstration reactor with the goal of increasing the technology readiness level (TRL) of the overall system for the longer term. The objective of this paper is to outline a notional concept for such a system, and to address how the proposed concept would advance the TRL of FHR concepts.Development of the proposed FHR Demonstration Reactor (DR) will enable commercial FHR deployment through disruptive and rapid technology development and demonstration. The FHR DR will close remaining gaps to commercial viability. Lower risk technologies are included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. Important capabilities that will be demonstrated by building and operating the FHR DR include core design methodologies; fabrication and operation of high temperature reactors; salt procurement, handling, maintenance, and ultimate disposal; salt chemistry control to maximize vessel life; tritium management; heat exchanger performance; pump performance; and reactivity control. The FHR DR is considered part of a broader set of FHR technology development and demonstration efforts, some of which are already underway. Nonreactor test efforts (e.g., heated salt loops or loops using simulant fluids) can demonstrate many technologies necessary for commercial deployment of FHRs. The FHR DR, however, fulfills a crucial role in FHR technology development by advancing the technical maturity and readiness level of the system as a whole.

Brown, Nicholas↗
DOE OSTI2016

Irradiation Tests Supporting LEU Conversion of Very High Power Research Reactors in the US

The US fuel development team is developing a high density uranium-molybdenum alloy monolithic fuel to enable conversion of five high-power research reactors. Previous irradiation tests have demonstrated promising behavior for this fuel design. A series of future irradiation tests will enable selection of final fuel fabrication process and provide data to qualify the fuel at moderately-high power conditions for use in three of these five reactors. The remaining two reactors, namely the Advanced Test Reactor and High Flux Isotope Reactor, require additional irradiation tests to develop and demonstrate the fuel’s performance with even higher power conditions, complex design features, and other unique conditions. This paper reviews the program’s current irradiation testing plans for these moderately-high irradiation conditions and presents conceptual testing strategies to illustrate how subsequent irradiation tests will build upon this initial data package to enable conversion of these two very-high power research reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Properties of Austenitic Stainless Steels for Nuclear Reactor Applications

Austenitic stainless steels are widely used in current commercial BWR and PWR systems as in-core and surrounding structural materials. These versatile steels have also been the main materials considered and applied for many advanced reactor technologies, including fast-breeder reactors and magnetic fusion reactors. For advanced reactors for higher-temperatures or high lifetime damage levels, modified versions of the standard commercial stainless steel grades have been developed to provide improved properties performance and radiation-resistance in specific reactor environments. An overview of the unirradiated properties and a summary of some important radiation-induced properties changes are presented. Highlights of several alloys developed for more radiation-resistance or better performance in specific reactor environments are highlighted.

Maziasz, Philip J↗
DOE OSTI2018

High aspect ratio catalytic reactor and catalyst inserts therefor

The present invention relates to high efficient tubular catalytic steam reforming reactor configured from about 0.2 inch to about 2 inch inside diameter high temperature metal alloy tube or pipe and loaded with a plurality of rolled catalyst inserts comprising metallic monoliths. The catalyst insert substrate is formed from a single metal foil without a central supporting structure in the form of a spiral monolith. The single metal foil is treated to have 3-dimensional surface features that provide mechanical support and establish open gas channels between each of the rolled layers. This unique geometry accelerates gas mixing and heat transfer and provides a high catalytic active surface area. The small diameter, high aspect ratio tubular catalytic steam reforming reactors loaded with rolled catalyst inserts can be arranged in a multi-pass non-vertical parallel configuration thermally coupled with a heat source to carry out steam reforming of hydrocarbon-containing feeds. The rolled catalyst inserts are self-supported on the reactor wall and enable efficient heat transfer from the reactor wall to the reactor interior, and lower pressure drop than known particulate catalysts. The heat source can be oxygen transport membrane reactors.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2018

Safeguards Challenges for Molten Salt Reactors

There are now a wide variety of molten salt reactor (MSR) technologies being proposed or developed that are gaining international interest and momentum toward deployment. There are two sub-categories of this class of reactors - those that are liquid fueled (i.e., the molten salt is both the fuel and the coolant) and those that have solid fuel with the molten salt used as coolant only. The liquid-fueled MSRs present unique challenges to current safeguards approaches. These challenges exist because there are currently no conceptual approaches to applying safeguards to unique, tightly coupled nuclear energy systems with the reactor core, balance of plant, and fuel cycle (everything else including salt processing and material separations) combined in a single facility. The application of safeguards to liquid-fueled MSRs will have to take into consideration several technical factors including: The homogeneous mixture of fuel, coolant, fission products, and actinides (with the attendant extremely high radiation field); Continuous variation in isotopic concentrations in the fuel salt, including removal (passive or active) of fission products, rare earth elements, and noble metals; High operating temperatures of the fuel salt; Online processing where some fraction of the inventory can be removed while the reactor is operational; Unique refueling schemes, including the ability to continuously feed the core with fresh fissile or fertile material; The presence of frozen fuel, potentially requiring a different safeguards process from the liquid fuel; and The presence of fuel outside the vessel. Additionally, If the thorium fuel cycle is employed, there will be additional complications since the resulting radiation signatures will be different from those of the uranium-based fuel cycle. The existing IAEA inspection regimes are based on the uranium-plutonium fuel cycle, item counting for nuclear reactors, and bulk material accountancy for the front and back end of the nuclear fuel cycle. These techniques and associated instrumentation for bulk accountancy have been developed predominantly for enrichment, fuel fabrication, and aqueous reprocessing. However, none of these bulk accountancy measures can be directly applied to liquid-fueled MSRs. This paper will explore some of the possible challenges in applying safeguards to molten salt reactors.

Kovacic, Donald N↗
DOE OSTI2018

Advanced Reactor Regulatory Development Status Report for FY 2018

This report provides a summary status of various INL-supported activities leading to the Nuclear Regulatory Commission (NRC) publication of regulatory guidance for advanced (non-light water reactor) designs. This regulatory guidance includes new design criteria designated as advanced reactor design criteria as proposed by a U.S. Department of Energy (DOE) national laboratory team to the NRC. The advanced reactor design criteria guidance was developed in response to requests from industry. The guidance is intended to provide a basis from which future advanced reactor developers and license applicants can build principle design criteria to address regulatory requirements in their specific license applications. In addition, this report provides the current status for a number of advanced non-light water reactor (non-LWR) regulatory framework development activities, including functional containment performance criteria, advanced reactor codes and standards development, and the Regulatory Technology Development Plan.

42 ENGINEERING↗
DOE OSTI2019

Preliminary In-Vessel Shielding Analysis for the Versatile Test Reactor

Shielding design is an important part of the Versatile Test Reactor (VTR) design. Various shielding implementations are required to protect workers and reactor components from damaging radiation sources. The major radiation source in VTR during operation is the prompt neutron and photon emitted from nuclear fuel fissions as well as the activated structure and coolant. Shields are needed to protect the structures located inside the reactor vessel, such as the lower grid, the barrel wall, etc., in order to limit the degradation of the materials properties as a result of neutron damage (“displacement per atom”). Another early key challenge of shielding design for a pool-typed Sodium-cooled Fast Reactor (SFR) such as VTR is the activation of the secondary sodium as to flows through the intermediate heat exchangers (IHX). For a pool-typed SFR, the IHXs are placed in the primary sodium pool inside the reactor vessel. During operation, neutrons produced in the core can reach the IHXs due to the small absorption cross section of sodium and the large mean free path of fast neutrons. The secondary sodium gets activated inside the IHX and carries some of that activity outside of the reactor vessel. This source of radioactivity contained in the secondary sodium needs to be mitigated to meet the radiation dose limit sets for personnel working at the plant. This summary focuses on describing the work performed for the in-vessel shielding design and analysis aiming at limiting the secondary sodium activation.

Fei, T.↗
DOE OSTI2019

Development of Molten Salt Reactor Modeling and Simulation Capabilities in VERA

In recent years, the nuclear community has seen a resurgence of interest in molten salt reactors (MSRs). Because of their molten, flowing fuel, MSRs present several new challenges not present in many other reactors, necessitating tightly coupled multiphysics modeling and simulation for accurate analysis. The Virtual Environment for Reactor Applications (VERA) has been developed to provide a high-fidelity multiphysics framework for reactor analysis. Though it was originally developed for light water reactor applications, VERA’s multiphysics capabilities make for a natural extension to MSR analysis. Recent efforts have focused on capturing the effects of moving fuel. A lumped depletion capability has been added that accounts for the moving and mixing of the fuel, as well as time spent outside the active core. A general multi-phase species transport module has also been added to track the movement and evolution of salt constituents as the salt flows through the core and around the primary loop. This paper presents a discussion of the new MSR capabilities in VERA, along with results for the lumped depletion and species transport capabilities for the Molten Salt Reactor Experiment, demonstrating VERA’s effectiveness in multiphysics simulations of MSRs.

Graham, Aaron↗
DOE OSTI2018

Experience with Chemical Looping Combustion of Coal in a 200 kW Dual Fluidized Bed Reactor

The University of Utah recently commissioned a dual fluidized bed process development unit (PDU) for investigating chemical looping combustion (CLC) and chemical looping with oxygen uncoupling (CLOU) of coal. The reactors and supporting systems were sized to support at least 200 kW of coal input based on available reaction rate data. Both the air reactor and fuel reactor are circulating fluidized beds and bed material passes between the reactors via loop seals. The reactors are refractory-lined rather than electrically heated, so the system must be preheated with natural gas before chemical looping combustion can begin. Air and steam are preheated through a combination of electric heating and natural gas burners and reactor product gas flows through heat exchangers and bag houses to cool and clean the gas of particulate. As is typical with pilot-scale systems, several operational challenges were encountered as the system was brought online and operation progressed from heatup and initial testing with gaseous fuels towards full CLOU processing of coal. The system hardware and operational procedures have undergone several modifications, but the PDU can now reliably be preheated to the operating temperature and coal can be fed at the target feed rate. This paper presents the PDU, describes the procedures for evaluating carbon conversion, CO2 capture and CO2 purity, and presents initial results from chemical looping combustion of coal with non-CLOU and CLOU carriers. General experience of operating the plant and notable challenges and solutions are also presented.

Whitty, Kevin J.↗
DOE OSTI2017

Performance of a Dual Fluidized Bed Reactor for Chemical Looping Combustion with Oxygen Uncoupling

Chemical looping with oxygen uncoupling (CLOU) is a variant of chemical looping combustion which uses as fluidized bed particles specific solid oxygen carriers which readily liberate gaseous oxygen (O2) in the fuel reactor. This enables much more efficient combustion of solid fuels such as coal than conventional chemical looping combustion, which relies on in situ gasification of char in to produce syngas that can react heterogeneously with the solid oxygen carrier. The CLOU process for coal is an especially interesting application of fluidized bed technology since several types of reactions take place within the reactors. For copper-based oxygen carrier particles, the copper cycles between cuprous and cupric oxide (Cu2O + ½ O2 ↔ 2 CuO). Volatiles released from coal during pyrolysis may react with oxidized copper or with released oxygen gas. Char reacts primarily with released oxygen. Balancing oxygen supply to the fuel reactor, heat generation and temperatures and global and internal circulation rates is a challenge. The University of Utah has been researching CLOU technology for 10 years, studying fundamental reaction kinetics for copper-based CLOU oxygen carriers, performing lab- and bench-scale experiments in single and dual fluidized beds, and developing reactor models through computational fluid dynamic modeling using Barracuda VR™. Recently the university began testing a 200 kWth dual circulating fluidized bed process development unit (PDU) for processing coal. The design process for that reactor, reactor simulations, operational experience and practical challenges of the PDU system are presented.

Whitty, Kevin J.↗
DOE OSTI2019

Gravity-based, non-invasive reactor system and method for coolant inventory monitoring

A gravity-based, non-invasive method of measuring a level of fluid in a container comprises use of at least one gravity meter located as proximate a center of mass of the fluid as possible. In a nuclear reactor system a method for monitoring the level of fluid in a nuclear reactor module, a report of a loss or gain of fluid within a cylindrical module may be generated from capturing a time series of gravity data from a first gravity meter mounted as an upper gravity meter and a second gravity meter mounted as a lower gravity meter, for example, proximate a cylindrical nuclear reactor module so as not to require any invasive conduit through, for example, a containment pressure vessel (CPV) or a reactor pressure vessel (RPV). In one embodiment, the upper and lower gravity meters are mounted on stable mounts as close to the fluid in the module as possible within a coolant pool or a structure containing cooled air. If a coolant pool of water surrounds a nuclear reactor module, the meters may be housed within a dry housing in the coolant pool such that the meters may be accessed from above the coolant pool and are located as close as possible to the reactor module and its contained mass of fluid.

Bloom, Scott Harris↗
DOE OSTI2020

Nuclear reactor support and seismic restraint with core retention cooling features

A nuclear island includes a nuclear reactor, a lateral seismic restraint, and a reactor core retention cooling system. The lateral seismic restraint includes a vertically oriented pin attached to one of the bottom of the lower vessel head and the floor underneath the nuclear reactor, and a mating pin socket is attached to the other of the bottom of the lower vessel head and the floor. The reactor core retention cooling system includes one or more baffles, optionally thermally insulating material, disposed alongside the exterior surface of a lower portion of the reactor pressure vessel including at least the lower vessel head. A plenum is defined between the one or more baffles and the exterior surface of a lower portion of the reactor pressure vessel. The one or more baffles may define a lower plenum inlet surrounding the lateral seismic restraint.

Edwards, Tyler A.↗
DOE OSTI2021

Cyber Threat Assessment Methodology for Autonomous and Remote Operations for Advanced Reactors (Conference Presentation)

The next generation of Advanced Reactors include planned capabilities for both Autonomous (operating without human interaction for a set period-of-time) and Remote (operating with human interaction from a separate physical location) Operations. Existing Nuclear Reactor architectures include a set of safety and security constraints tightly coupled with personnel policies and procedures. As Advanced Reactors are fielded with these new Autonomous and Remote operational capabilities, the architectures and associated infrastructure services and components will perceivably expand the overall attack surface and risk calculations with regards to safe and secure operations. This paper is part of an FY21 work program focused on ensuring Advanced Reactor designs are informed with threat-based guidance on design and operation of Secure Architectures with a specific focus on the deployment of Autonomous Systems in support of Advanced Reactor Operations. The next phase of this research program is to complement produce a methodology for assessment of the cyber threat against these architectures as well as a catalogue of Use Cases to support the Advanced Reactor community in their implementation of Autonomous and Remote Operations.

97 MATHEMATICS AND COMPUTING↗
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 OSTI2021

Heat pipe reactor core and heat exchangers formation and deployment

A heat pipe reactor may include a reactor core and one or more heat exchangers positioned on one or both sides of the reactor core. The heat pipe reactor may also include a plurality of heat pipes extending from the reactor core and out through the one or more heat exchangers. The reactor core may be composed of a plurality of monolithic blocks.

Reid, Robert Stowers↗
DOE OSTI2021

Systems and methods for enhancing isolation of high-temperature reactor containments

A high-temperature containment-isolation system for transferring heat from a nuclear reactor containment to a high-pressure heat exchanger is presented. The system uses a high-temperature, low-volatility liquid coolant such as a molten salt or a liquid metal, where the coolant flow path provides liquid free surfaces a short distance from the containment penetrations for the reactor hot-leg and the cold-leg, where these liquid free surfaces have a cover gas maintained at a nearly constant pressure and thus prevent high-pressures from being transmitted into the reactor containment, and where the reactor vessel is suspended within a reactor cavity with a plurality of refractory insulator blocks disposed between an actively cooled inner cavity liner and the reactor vessel.

Peterson, Per F.↗
DOE OSTI2022

High Energy Physics Opportunities Using Reactor Antineutrinos

Nuclear reactors are uniquely powerful, abundant, and flavor-pure sources of antineutrinos that continue to play a vital role in the US neutrino physics program. The US reactor antineutrino physics community is a diverse interest group encompassing many detection technologies and many particle physics topics, including Standard Model and short-baseline oscillations, BSM physics searches, and reactor flux and spectrum modeling. The community's aims offer strong complimentary with numerous aspects of the wider US neutrino program and have direct relevance to most of the topical sub-groups composing the Snowmass 2021 Neutrino Frontier. Reactor neutrino experiments also have a direct societal impact and have become a strong workforce and technology development pipeline for DOE National Laboratories and universities. This white paper, prepared as a submission to the Snowmass 2021 community organizing exercise, will survey the state of the reactor antineutrino physics field and summarize the ways in which current and future reactor antineutrino experiments can play a critical role in advancing the field of particle physics in the next decade.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2018

Means of Meeting Chapter 15 Requirements for a Small Modular Reactor Cooled by Lead-Bismuth

The mission of the Department of Energy Office of Nuclear Energy (DOE-NE) is to advance nuclear power as a resource capable of meeting the nation’s energy, environmental, and national security needs by resolving technical, cost, safety, proliferation resistance, and security barriers through Research, Development, and Demonstration (RD&D). To assist in accomplishing this, DOE-NE has established the Gateway for Accelerated Innovation in Nuclear (GAIN) to provide the nuclear community with access to the technical, regulatory, and financial support necessary to move innovative nuclear energy technologies toward commercialization. As part of this program, a GAIN Voucher was awarded to the Columbia Basin Consulting Group (CBCG) to develop for the Lead-Bismuth Reactor (LBR): (1) a means of meeting Chapter 15 requirements and (2) to perform an assessment against a gap analysis developed by the U.S. Nuclear Regulatory Commission (NRC) for Chapter 15. These two assessments will determine to what extent the LBR can adapt to or meet the Chapter 15 Accident Analysis requirements, and determine gaps that may still exist and what additional information is needed in order to close them. The CBCG LBR design is being developed with a high level of passive safety. It has the advantages of the classical liquid metal reactor, i.e. small size and passive safety, while having the additional advantage of a chemically inert coolant. The CBCG and the Pacific Northwest National Laboratory (PNNL) collaborated to develop an initial licensing pathway for the CBCG Lead-Bismuth conceptual reactor design. It was found that the reactor is well along the licensing pathway and is licensable although some additional design information is needed to establish a complete licensing basis. The missing information is identified in this report and it is noted that much of the missing information can be obtained from the Fast Flux Test Facility (FFTF), a DOE liquid metal test reactor. It was also found that the traditional Chapter 15 requirements for assessing the response of a reactor to possible accident sequences can be adapted to the LBR concept. The Chapter 15 review examined several classes of initiating events and accidents that are normally covered in Chapter 15: reactivity initiated events, reduction in heat removal events, local fuel failure events, and events initiated on the steam supply side. Although some design information is lacking at present, the LBR can be expected to respond as required by Chapter 15 requirements to almost all of the first three events. For events initiated on the steam side, design information is lacking and so conclusions cannot be reached at present. On the other hand, it should be quite possible to develop the design information for the steam supply side and the impact of events initiated on this side can be assessed in a subsequent study.

Wootan, David W.↗