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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 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Nuclear Data Management and Analysis System Plan

The United States Department of Energy Advanced Reactor Technologies Program was formed in Fiscal Year 2015 and encompasses the Next Generation Nuclear Plant Project and Very High Temperature Reactor (VHTR) Program as they were known previously. The VHTR Program was created to support design and licensing of the first VHTR nuclear plant. Data created for and used by the program must be qualified for use, stored in a readily accessible electronic form, categorized to assure the correct data are used, and controlled to prevent data corruption or inadvertent changes. The Nuclear Data Management and Analysis System was designed to support the data needs of the VHTR Program, at the time and now the Advanced Reactor Technologies Program. Since its inception, use of the Nuclear Data Management and Analysis System has expanded to support additional projects and programs with similar requirements for control, analysis, and availability of large data sets.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2020

Modernization of Technical Requirements for Licensing of Advanced Non-Light Water Reactors: Risk-Informed and Performance-Based Evaluation of Defense-in-Depth Adequacy

This document supports the work contained in Nuclear Energy Institute (NEI) 18-04 “Risk-Informed Performance-Based Technology Inclusive Guidance for Advanced Reactor Licensing Basis Development” Revision 0. NEI 18-04 presents a modern, technology-inclusive, risk-informed, and performance-based (TI-RIPB) process for selection of Licensing Basis Events (LBEs); safety classification of structures, systems, and components (SSCs) and associated risk-informed special treatments; and determination of defense-in-depth (DID) adequacy for non-LWRs. The NEI guidance document provides one acceptable means for addressing the aforementioned topics as part of demonstrating a specific design provides reasonable assurance of adequate radiological protection. This report provides the framework and associated methodology guidelines and discussion for establishing, then evaluating, confirming, and documenting the adequacy of defense-in-depth (DID) for advanced non-light-water reactor technologies. It was developed as part of the Licensing Modernization Project led by Southern Company and cost-shared by the United States Department of Energy and has benefited from considerable NRC formal reviews and public workshops. The methodology converts the DID philosophy into a structured process that is implementable, embraces existing United States and international definitions and philosophies of DID that set the foundation for the process. It builds on the DID framework developed in the Department of Energy Next Generation Nuclear Plant Project and earlier works on this subject. The approach to establishing DID adequacy involves the incorporation of DID attributes into the plant capabilities and programmatic elements of DID. The integrated evaluation of DID adequacy includes both quantitative elements to incorporate risk-informed and performance-based (RIPB) considerations and qualitative elements that address uncertainties and limitations in the quantitative models and supporting data. Demonstration of DID adequacy ensures that there are multiple layers of defense for risk-significant challenges to the design and that the plant capabilities and programs that support each layer are provided in a manner that minimizes dependencies among these layers. The focus of this report is assurance of DID adequacy with respect to protection of the public from radiological exposures resulting from accidental releases of radioactive material. While other hazards are not specifically addressed, this methodology is expected to be beneficial for determining DID adequacy for them as well. Risk-informed evaluation of DID considers the integrated performance of all plant SSCs and associated programs to manage daily operational activities, transients, and accidents, including the evaluation of strategies for accident prevention and mitigation. The RIPB LBE scenario methodology used in this evaluation defines the challenges to the plant safety features included in the plant design basis and beyond, and the scope of all deterministic and probabilistic safety evaluations. By examining event sequences across the whole spectrum of LBEs, a systematic assessment of DID can be accomplished.

42 ENGINEERING↗
NASA NTRS2007

Space and Terrestrial Power System Integration Optimization Code BRMAPS for Gas Turbine Space Power Plants With Nuclear Reactor Heat Sources

In view of the difficult times the US and global economies are experiencing today, funds for the development of advanced fission reactors nuclear power systems for space propulsion and planetary surface applications are currently not available. However, according to the Energy Policy Act of 2005 the U.S. needs to invest in developing fission reactor technology for ground based terrestrial power plants. Such plants would make a significant contribution toward drastic reduction of worldwide greenhouse gas emissions and associated global warming. To accomplish this goal the Next Generation Nuclear Plant Project (NGNP) has been established by DOE under the Generation IV Nuclear Systems Initiative. Idaho National Laboratory (INL) was designated as the lead in the development of VHTR (Very High Temperature Reactor) and HTGR (High Temperature Gas Reactor) technology to be integrated with MMW (multi-megawatt) helium gas turbine driven electric power AC generators. However, the advantages of transmitting power in high voltage DC form over large distances are also explored in the seminar lecture series. As an attractive alternate heat source the Liquid Fluoride Reactor (LFR), pioneered at ORNL (Oak Ridge National Laboratory) in the mid 1960's, would offer much higher energy yields than current nuclear plants by using an inherently safe energy conversion scheme based on the Thorium --> U233 fuel cycle and a fission process with a negative temperature coefficient of reactivity. The power plants are to be sized to meet electric power demand during peak periods and also for providing thermal energy for hydrogen (H2) production during "off peak" periods. This approach will both supply electric power by using environmentally clean nuclear heat which does not generate green house gases, and also provide a clean fuel H2 for the future, when, due to increased global demand and the decline in discovering new deposits, our supply of liquid fossil fuels will have been used up. This is expected within the next 30 to 50 years, as predicted by the Hubbert model and confirmed by other global energy consumption prognoses. Having invested national resources into the development of NGNP, the technology and experience accumulated during the project needs to be documented clearly and in sufficient detail for young engineers coming on-board at both DOE and NASA to acquire it. Hands on training on reactor operation, test rigs of turbomachinery, and heat exchanger components, as well as computational tools will be needed. Senior scientist/engineers involved with the development of NGNP should also be encouraged to participate as lecturers, instructors, or adjunct professors at local universities having engineering (mechanical, electrical, nuclear/chemical, and/or materials) as one of their fields of study.

Juhasz, Albert J.↗
DOE OSTI2023

Status on Development of Graphite Analytical Tool (GAT)

The DOE-ART Graphite R&D program has been generating significant amounts of irradiated and unirradiated graphite data since 2006 when the program was part of the DOE NGNP (Next Generation Nuclear Plant) Project. This data includes critical irradiation creep and irradiated material property changes from the Advanced Graphite Creep (AGR) experiment as well as significant amounts of data on unirradiated material property values on several current nuclear graphite grades (Baseline program). Previously, Idaho National Laboratory has developed an internal analysis tool to assist with analysis of the unirradiated and irradiated data. The Graphite Analytical Tool (GAT) is intended to provide easy access to the graphite data in the form of comparing unirradiated and irradiated material property changes, comparison of material property differences between various nuclear graphite grades, and illustrate trends within the irradiated and unirradiated data generated within the DOE-ART Graphite R&D program. This report summarizes the progress to-date on the development of this analytical tool.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Thermal Predictions of the AGR-3/4 Experiment With Time-Varying Gas Gaps

In this paper, a thermal analysis was performed for the Advanced Gas Reactor test experiment (AGR-3/4) with time-varying gas gaps. The experiment was irradiated at the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). Several fuel irradiation experiments are planned for the AGR Fuel Development and Qualification Program that supports the development of the Very-High-Temperature Gas-Cooled Reactor (VHTR) under the Next-Generation Nuclear Plant (NGNP) project. AGR-3/4 combines two tests in a series of planned AGR experiments to test tristructural-isotropic (TRISO)-coated, low-enriched uranium oxy-carbide fuel. Forty-eight TRISO-fueled compacts were inserted into 12 separate capsules for the experiment (four compacts per capsule). The purpose of this analysis was to calculate the temperatures of each compact and graphite layer to obtain daily average temperatures using time (fast neutron fluence)-varying gas gaps and to compare with experimentally measured thermocouple data. A finite-element heat transfer model was created for each capsule using the commercial code ABAQUS. Finally, model results are compared to thermocouple data taken during the experiment.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Alloy 800H Weldments with Alloy 617 Filler

Alloy 800H is the leading candidate for the steam generator in Next Generation Nuclear Plant (NGNP) Project. It is necessary to extend the allowable stresses for Alloy 800H to 900 degrees C and 600,000 hours. This also requires data on welds in order to extend code acceptance. The focus of this presentation is on the creep-rupture properties of Alloy 800H cross welds with Alloy 617 filler. The goal is to understand the effect of an overmatched weld on creep-rupture properties.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Licensing basis event selection framework for advanced reactors - 18818

This paper describes key aspects of the risk-informed, performance-based licensing methodology as developed by the Department of Energy (DOE) for the Next Generation Nuclear Plant (NGNP) project. The discussion focuses on an event selection process using modular high temperature gas-cooled reactor (HTGR) design attributes and identifies how top-level regulatory criteria can be placed on a frequency-consequence (F-C) curve. Nuclear Regulatory Commission (NRC) staff evaluations of the proposed approach are included. In addition, important lessons learned during the NGNP prelicensing effort and their potential applications to alternative advanced reactor designs are presented. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Study on main control room and human-system interface of generation III in China - 313

This paper focuses on the study on Main Control Room (MCR) and Human-System Interface (HSI) of Generation III in China. An intelligent HSI has been developed to enhance the safety and availability of Nuclear Power Plant (NPP) by improving operational reliability. The key elements of the proposed HSI such as Advanced Alarm System, State Oriented Procedures (SOP), Large Display Panel (LDP), Computerized-base procedures..., that affects what information the operator sees or the system's response to a control input must be empirically evaluated to ensure that the new design does not compromise HSI effectiveness. Based on CPR1000's experience in China, this paper presents the approach used as well as the most relevant aspects of this kind of project for the next generation. This approach will be used in new Generation III nuclear power plant, and also used in modernizing I and C system in currently operative nuclear power plants, in addition to meeting safety requirements and the plant's operational requirements, to improve cost-effective plant and human performance and to reduce likelihood of human errors, to gain maximum benefit of the implemented technology and to increase the performance, resulting in improved plant safety, availability, reliability, and cost-effective operation. The proposed HSI has been validated and demonstrated using the CPR1000 simulator of National Key Laboratory. After sufficient validation, the characteristic design features of the proposed HSI will be reflected in the design of the main control room of the next generation reactors in China. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

IAEA Cooperative Research Project Status Report

This report provides an end-of-year summary that reflects the progress and status of United States (US) activities supporting the International Atomic Energy Agency (IAEA) Cooperative Research Project (CRP) effort to develop Modular High Temperature Reactor Safety Design Criteria. These U.S. activities are being managed by the Idaho National Laboratory (INL) on behalf of the US Department of Energy (DOE), and it is noted that this summary reflects progress during the second year of this planned three year activitiy. This status summary addresses Level 3 milestone M3AT-16IN2001104 in work package AT-16IN200110.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Advanced Reactor Technologies Advanced Reactor Technologies Program Management Plan

The Advanced Reactor Technologies (ART) program grew out of earlier Department of Energy efforts to promote next-generation, small modular, and other advanced reactor concepts. It is now sponsored by Office of Advanced Reactor Deployment, NE-52. Because of its peculiar genesis, the program resides at multiple national laboratories and brings in researchers from subcontracting entities and universities. This creates a need for Idaho National Laboratory’s (INL’s) ART to authorize and define work through memorandum purchase orders, inter-entity work orders, and subcontracts. ART manages research and development (R&D) for the Gas-Cooled Reactors (GCRs) Campaign and other advanced-reactor technologies and ensures that Nuclear Regulatory Commission (requirements and stakeholder needs are factored into the R&D activities. The GCR Campaign supports the Next Generation Nuclear Plant High Temperature Reactor as outlined in the Energy Policy Act of 20051 by

42 ENGINEERING↗
DOE OSTI2022

MELCOR Accident Progression and Source Term Demonstration Calculations for a FHR

MELCOR is an integrated thermal hydraulics, accident progression, and source term code for reactor safety analysis that has been developed at Sandia National Laboratories for the United States Nuclear Regulatory Commission (NRC) since the early 1980s. Though MELCOR originated as a light water reactor (LWR) code, development and modernization efforts have expanded its application scope to include non-LWR reactor concepts. Current MELCOR development efforts include providing the NRC with the analytical capabilities to support regulatory readiness for licensing non-LWR technologies under Strategy 2 of the NRC's near- term Implementation Action Plans. Beginning with the Next Generation Nuclear Project (NGNP), MELCOR has undergone a range of enhancements to provide analytical capabilities for modeling the spectrum of advanced non-LWR concepts. This report describes the generic plant model developed to demonstrate MELCOR capabilities to perform fluoride-salt-cooled high-temperature reactor (FHR) safety evaluations. The generic plant model is based on publicly-available FHR design information. For plant aspects (e.g., reactor building leak rate and details of the cover-gas system) that are not described in the FHR references, the analysts made assumptions needed to construct a MELCOR full-plant model. The FHR model uses a TRi-structural ISOtropic (TRISO)-particle fuel pebble-bed reactor with a primary system rejecting heat to two coiled tube air heat ex changers. Three passive direct reactor auxiliary cooling systems provide heat removal to supplement or replace the emergency secondary system heat removal during accident conditions. Surrounding the reactor vessel is a low volume reactor cavity that insulates the reactor with fire bricks and thick concrete walls. A refractory reactor liner system provides water cooling to reduce the concrete wall temperatures. Example calculations are performed to show the plant response and MELCOR capabilities to characterize a range of accident conditions. The accidents selected for evaluation consider a range of degraded and failed modes of operation for key safety functions providing reactivity control, the primary system decay heat removal and also a piping leak of the line to the coolant drain tank.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

MELCOR Accident Progression and Source Term Demonstration Calculations for a Heat Pipe Reactor

MELCOR is an integrated thermal hydraulics, accident progression, and source term code for reactor safety analysis that has been developed at Sandia National Laboratories for the United States Nuclear Regulatory Commission (NRC) since the early 1980s. Though MELCOR originated as a light water reactor (LWR) code, development and modernization efforts have expanded its application scope to includ e non-LWR reactor concepts. Current MELCOR development efforts include providing the NRC with the analytical capabilities to support regulatory readiness for licensing non-LWR techno logies under Strategy 2 of the NRC?s near- term Implementation Action Plans. Beginning with the Next Generation Nuclear Project (NGNP), MELCOR has undergone a range of enha ncements to provide analytical capabilities for modeling the spectrum of advanced non-LWR concepts. This report describes the generic plant model developed to demonstrate MELCOR capabilities to perform heat pipe reactor (HPR) safety evaluations. The generic plant mode l is based on a publicly-available Los Alamos National Laboratory (LANL) Megapower design as modified in the Idaho National Laboratory (INL) Design A description. For plant aspects (e.g., reactor building size and leak rate) that are not described in the LANL and INL references , the analysts made assumptions needed to construct a MELCOR full-plant model. The HP R uses high assay, low-enrichment uranium (HALEU) fuel with steel cladding that uses heat pipes to transfer heat to a secondary Brayton air cycle. The core region is surrounded by a stainless-steel shroud, alumina reflector, core barrel and boron carbide neutron shield. The reactor is secured inside a below-grade cavity, with the operating floor located above the cavity. Example calculations are performed to show the plant response and MELCOR capabilities to characterize a range of accident conditions. The accidents selected for evaluation consider a range of degraded and failed modes of operation for key safety functions providing re activity control, the primary and secondary system heat removal, and the effectiveness of th e confinement natural circulation flow into the reactor cavity (i.e., a flow blockage).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

MELCOR Accident Progression and Source Term Demonstration Calculations for a HTGR

MELCOR is an integrated thermal hydraulics, accident progression, and source term code for reactor safety analysis that has been developed at Sandia National Laboratories for the United States Nuclear Regulatory Commission (NRC) since the early 1980s. Though MELCOR originated as a light water reactor (LWR) code, development and modernization efforts over the past decades have expanded its application scope to include non-LWR reactor concepts. Current MELCOR development efforts include providing the NRC with the analytical capabilities to support regulatory readiness for licensing non-LWR technologies under Strategy 2 of the NRC's near-term Implementation Action Plans. Beginning with the Next Generation Nuclear Project (NGNP), MELCOR ha s undergone a range of enhancements to provide analytical capabilities for modeling the spectrum of advanced non-LWR concepts. This report describes the generic plant model developed to demonstrate MELCOR capabilities to perform high-temperature gas reactor (HTGR) safety evaluations. The generic plant model is based on publicly available PMBR-400 design information. For plant aspects (e.g., reactor building size and leak rate) that are not described in the PBMR-400 references, the analysts made assumptions needed to construct a MELCOR full-plant model. The HTGR model uses a TRi-structural ISOtropic (TRISO)-particle fuel pebble-bed reactor with a primary system rejecting heat to a recuperative heat exchange r. Surrounding the reactor vessel is a reactor cavity contained within a confinement room cooled by the Reactor Cavity Cooling System (RCCS). Example calculations are performed to show the plant response and MELCOR capabilities to characterize a range of accident conditions. The accidents selected for evaluation consider a range of degraded and failed modes of operation for key safety functions providing reactivity control, primary system heat removal and reactor vessel decay heat removal, and confinement cooling.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Preliminary Framework for Human-Automation Collaboration

The Department of Energy’s Advanced Reactor Technologies Program sponsors research, development and deployment activities through its Next Generation Nuclear Plant, Advanced Reactor Concepts, and Advanced Small Modular Reactor (aSMR) Programs to promote safety, technical, economical, and environmental advancements of innovative Generation IV nuclear energy technologies. The Human Automation Collaboration (HAC) Research Project is located under the aSMR Program, which identifies developing advanced instrumentation and controls and human-machine interfaces as one of four key research areas. It is expected that the new nuclear power plant designs will employ technology significantly more advanced than the analog systems in the existing reactor fleet as well as utilizing automation to a greater extent. Moving towards more advanced technology and more automation does not necessary imply more efficient and safer operation of the plant. Instead, a number of concerns about how these technologies will affect human performance and the overall safety of the plant need to be addressed. More specifically, it is important to investigate how the operator and the automation work as a team to ensure effective and safe plant operation, also known as the human-automation collaboration (HAC). The focus of the HAC research is to understand how various characteristics of automation (such as its reliability, processes, and modes) effect an operator’s use and awareness of plant conditions. In other words, the research team investigates how to best design the collaboration between the operators and the automated systems in a manner that has the greatest positive impact on overall plant performance and reliability. This report addresses the Department of Energy milestone M4AT-15IN2302054, Complete Preliminary Framework for Human-Automation Collaboration, by discussing the two phased development of a preliminary HAC framework. The framework developed in the first phase was used as the basis for selecting topics to be investigated in more detail. The results and insights gained from the in-depth studies conducted during the second phase were used to revise the framework. This report describes the basis for the framework developed in phase 1, the changes made to the framework in phase 2, and the basis for the changes. Additional research needs are identified and presented in the last section of the report.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Factors Impacting Nuclear Energy Share in U.S. Energy Markets

The purpose of this report is to collate information and findings from recent studies conducted by national and international bodies/institutes to identify approaches for maintaining/enhancing the role of nuclear energy in the current and future energy mix of the United States. This report shows how nuclear generation grew quickly to provide 20% of U.S. electricity, sustaining that level for three decades without the benefit of new construction, but is now projected to decline going forward. Nuclear construction costs in the U.S. spiraled out of control, ending construction for two decades and the recent resumption of construction has continued the pattern of schedule delays and cost overruns. However, nuclear operations exhibited strong learning, achieving and sustaining the highest capacity factor of any electricity generation technology and license extensions and uprates have sustained nuclear market share. The share of nuclear energy in the U.S. electricity market is projected to decline by ~1/3rd over the next 30 years The report provides an overview of how the markets work in theory and in practice. It indicates how market deregulation and clean energy policies have created conditions where nuclear plants are being retired for economic rather than technical reasons. The report also shows how many of the markets do not in practice have free competition but instead have outcomes that are being determined more and more by policy instead of market forces. While electricity costs from existing nuclear plants are low, electricity from new builds is projected to be too expensive to be competitive head-to-head with natural gas, even for nth-of-a-kind costs. Wind and solar energy have enjoyed an extended period of sustained subsidy. This protected environment has resulted in a sustained reduction in plant level costs to the point that some of these Variable Renewable Energy (VRE) technologies are becoming competitive even if their direct subsidies are removed. But plant level costs underestimate total VRE costs which include a number of system-level externalities. The incremental system value of additional VRE capacity was shown to decline as market share increases, with solar value declining more quickly than wind. The report closes with examination of a possible future for nuclear generation as part of deep decarbonization of the electricity sector. This approach avoids direct competition with natural gas. The two options for achieving 100% decarbonization are to use only renewables or to use all zero emissions technologies, and the report shows the second approach is much less expensive than the first.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2018

Fundamental study of key issues related to advanced sCO 2 Brayton cycle: Prototypic HX development and cavitation

Diffusion bonded heat exchangers are the leading candidates for the sCO 2 Brayton cycles in next generation nuclear power plants. Commercially available diffusion bonded heat exchangers utilize set of continuous semi-circular zigzag micro channels to increase the heat transfer area and enhance heat transfer through increased turbulence production. Such heat exchangers can lead to excessive pressure drop as well as flow maldistribution in the case of poorly designed flow distribution headers. The goal of the current project is to fabricate and test potential discontinuous fin patterns for diffusion bonded heat exchangers; which can achieve desired thermal performance at lower pressure drops. Prototypic discontinuous offset rectangular and Airfoil fin surface geometries were chemically etched on to 316 stainless steel plate and sealed against an un-etched flat pate using O-ring seal emulating diffusion bonded heat exchangers. Thermal-hydraulic performance of these prototypic discontinuous fin geometries was experimentally evaluated and compared to the existing data for the continuous zigzag channels. The data generated from this project will serve as the database for future testing and validation of numerical models.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

PMDT: AI-Enabled Predictive Maintenance Digital Twins for Advanced Nuclear Reactors

Our team made substantial technical progress on various fronts during the course of the program. Multiple milestones were geared towards demonstrating the feasibility of machine learning based predictive maintenance digital twins towards reducing O&M costs, whereas some other milestones actually focused on identifying technical gaps and developing technologies such as humble AI to provide necessary robustness to the ML-based models. We were able to demonstrate in many cases that Machine learning-based methods can be successfully adapted for Nuclear plant environments especially for remote monitoring applications. Detailed analyses were carried out with plant and full scope simulation data along with capabilities of enhanced analytics to assess and set realistic expectations on cost reductions in O&M. These assessments are paving the way for investments towards reactor design improvements as well project planning for SMR projects as they develop and mature in the next few years. Technology developed under this program got direct visibility to GE Hitachi and their utility customers and resulted in positive intents to deploy some of the elements from design phase. The project additionally resulted in several reports, publications, software and data generation that will be useful in deployment and O&M services for BWRX300 fleets.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗