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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 163 records · Page 9
All sourcesNASA NTRSDOE OSTI
DOE OSTI1991

Nuclear Safety [Vol. 32, No. 4, October-December 1991]

Nuclear Safety is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 477 Report on the International Symposium on the Use of Probabilistic Safety Assessment for Operational Safety—PSA '91, S. Chakraborty and M. Khatib-Rahbar; 488 Good Relationships Are Pivotal in Nuclear Data Bases, A. S. Heger and B. V. Koen; 494 Technical Note: The Interagency Nuclear Safety Review Panel's Evaluation of the Ulysses Space Mission, J. A. Sholtis, Jr., D. A. Huff, L. B. Gray, N. P. Klug, and R. O. Winchester; ACCIDENT ANALYSIS: 502 The Severe Accident Analysis Program for the Savannah River Nuclear Production Reactors, M. L. Hyder; CONTROL AND INSTRUMENTATION: 511 A Framework for Selecting Suitable Control Technologies for Nuclear Power Plant Systems, R. A. Kisner; DESIGN FEATURES: 521 Containments for Gas-Cooled Power Reactors History and Status, P.W. Williams; ENVIRONMENTAL EFFECTS: 537 Indoor Radon: A Natural Risk, N. H. Harley and J.H. Harley; On the Importance of the Atmospheric Parameters in the Fission Products Distribution of a Severe Reactor Accident, M. C. Barla and A. R. Bayulken; Book Review: Health Effects of Exposure to Low Levels of Ionizing Radiation BEIR V, C. R. Richmond; WASTE AND SPENT FUEL MANAGEMENT: 555 Activities Related to Waste and Spent Fuel Management, M. D. Muhlheim and E. G. Silver OPERATING EXPERIENCES: 567 Effects of Component Aging on the Westhinghouse Control Rod Drive System, K. Sullivan and W. Gunther; 577 Reactor Shutdown Experience, Compiled by J. W. Cletcher, 580 Selected Safety-Related Events, Compiled by G. A. Murphy; 582 Operating U 8 Power Reactors, Compiled by M. D. Muhlheim and E. G. Silver, RECENT DEVELOPMENTS: 596 General Administrative Activities, Compiled by M. D. Muhlheim and E. G. Silver; 610 Reports, Standards, and Safety Guides, D. S. Queener; 614 Proposed Rule Changes as of June 30, 1991; ANNOUNCEMENTS: 520 Workshop on PC PRAISE: A Probabilistic Fracture Mechanics Personal Computer Code for Nuclear Power Plant Piping Reliability Assessment; 536 Fifth Workshop on Nuclear Power Plant Containment Integrity; 624 New OECD "International Information System on Occupational Exposure (ISOE)"; 625 Harvard School of Public Health Announces Short Courses; 625 Eighth Power Plant Dynamics, Control and Testing Symposium; 626 Second Training Course on Off-Site Emergency Planning and Response for Nuclear Accidents; 618 The Authors; 622 Reviewers of Nuclear Safety, Vol 32.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

FFTF Acceptance and Startup Testing for GAIN

The Fast Flux Test Facility (FFTF) is the most recent liquid metal reactor (LMR) to be designed, constructed, and operated by the U.S. Department of Energy (DOE). The 400-MWt sodium-cooled, fast-neutron flux reactor plant was designed for irradiation testing of nuclear reactor fuels and materials for liquid metal fast breeder reactors. Following the demise of the breeder reactor program in the United States, FFTF continued to play a key role in providing a test bed for demonstrating performance of advanced fuel designs and demonstrating operation, maintenance, and safety of advanced liquid metal reactors. FFTF operations ceased in April 1992 after a determination by DOE that no combination of proposed missions was financially feasible over a ten-year period. The reactor is currently deactivated and in a long-term surveillance and maintenance (S&M) mode. This report provides information on the extensive and rigorous process that was used to conduct turnover from construction followed by acceptance and startup testing of the FFTF. This paper is in support of the Gateway for Accelerated Innovation in Nuclear (GAIN), which provides the nuclear energy community with access to the technical, regulatory, and financial support necessary to move new or advanced nuclear reactor designs toward commercialization while ensuring the continued safe, reliable, and economic operation of the existing nuclear fleet. The information obtained from the design, startup, and operation of the FFTF provides valuable insight for follow-on reactor projects, such as the Versatile Test Reactor (VTR), in the areas of plant system and component design, component fabrication, fuel design and performance, prototype testing, site construction, reactor startup and operations, and reactor deactivation and decommissioning (D&D). The focus of this report is on the process used to startup the FFTF and to ensure that operations could be conducted efficiently and safely. A reference section is provided of documents detailing the successful turnover and testing process implemented for startup of the reactor and its supporting systems. The documents listed can be retrieved upon request and are believed useful for future reactor startup endeavors.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

A Statistical Approach to Safety Margin Evaluation of Irradiation Experiments in the ATR

The current safety basis of the INL Advanced Test Reactor requires that irradiation experiments meet minimum safety margins to critical heat flux and flow instability during anticipated reactor fault conditions. Compliance to these requirements may be demonstrated by calculating safety margins using safety factors that account for uncertainties, or using a statistical approach to quantify the uncertainty in the safety margins. Safety margins obtained from the first approach are conservative since the individual events leading to the cumulative safety factor are assumed to occur simultaneously at their maximum uncertainty. The statistical approach is less conservative in the case where the uncertainties are not too large, and may reduce the conservatism resulting from an overabundance of safety factors. This report describes the procedure for safety margin evaluation using a statistical approach and identifies the important thermal-hydraulic variables to consider in an uncertainty analysis. In particular, safety margins to critical heat flux and flow instability are evaluated for three recent high power fueled experiments (MP-1, FSP-1 and AFIP-6) using RELAP5 coupled to RAVEN to perform statistical analysis and uncertainty quantification. The results demonstrate adequate safety margins to critical heat flux and flow instability during an anticipated loss of commercial power to the reactor.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Qualification of data obtained during a severe accident. Illustrative examples from TMI-2 evaluations

The accidents at the Three Mile Island Unit 2 (TMI-2) Pressurized Water Reactor (PWR) and the Daiichi Units 1, 2, and 3 Boiling Water Reactors (BWRs) provide unique opportunities to evaluate instrumentation exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during the TMI-2 accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. Post-TMI-2 instrumentation evaluation programs focused on data required by TMI-2 operators to assess the condition of the reactor and containment and the effect of mitigating actions taken by these operators. Prior efforts also focused on sensors providing data required for subsequent forensic evaluations and accident simulations. This paper provides additional details related to the formal process used to develop a qualified TMI-2 data base and presents data qualification details for three parameters: reactor coolant system (RCS) pressure; containment building temperature; and containment pressure. These selected examples illustrate the types of activities completed in the TMI-2 data qualification process and the importance of such a qualification effort. These details are described to facilitate implementation of a similar process using data and examinations at the Daiichi Units 1, 2, and 3 reactors so that BWR-specific benefits can be obtained.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Coupling of Thermal Energy Storage with Small Modular Reactors

Thermal energy storage has been proposed as a load management strategy for Small Modular Reactors (SMRs) deployed on constrained grids and/or in hybrid energy systems (1). Hybrid energy systems can include process steam applications, and the presence of intermittent energy sources such as wind and solar. Under these conditions, the reactor can be subjected to significant time varying electric loads. One strategy for accommodating these loads is to operate the reactor in a load follow mode where reactor power is modulated to match the electric demand. In addition to lost energy potential, load follow operation can result in additional stresses on the fuel and other mechanical components. A more attractive approach is to operate the reactor at or near steady state and bypass excess steam to a thermal energy storage system. The thermal energy can then be recovered later, either for electric generation during periods of peak electric demand, or for uses in process steam applications. In previous work, the design of a sensible heat Thermal Energy Storage (TES) system was proposed for SMR applications (2). In this paper we examine coupling of the TES system to a SMR, typical of Integral Pressurized Water Reactor (IPWR) designs currently under development. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Pilot-Scale Continuous Plug-Flow Hydrothermal Liquefaction of Food Waste for Biocrude Production

Pilot-scale hydrothermal liquefaction (HTL) of biowaste is a critical step toward commercialization of the HTL technology. Despite many HTL studies conducted with wet biomass, including food waste, few were performed with a pilot-scale continuous plug-flow reactor (PFR), with the biocrude yield and quality analysis based on dewatering (ASTM D2892 Annex X1). This paper describes the development and performance evaluation of a mobile pilot-scale HTL continuous PFR, with a processing capacity of 60 L/h of wet feedstock and 6 L/h of biocrude production. The reactor system was designed for reaction conditions of up to 325 °C and 17.25 MPa. The reactor has a volume of 28.88 L with an additional counterflow heat exchanger volume of 18.07 L. Two types of food wastes, from a food processing plant and grocery store, were processed at 280 °C for 30 min, producing biocrude oil yields of 52.19 and 47.06 wt %, energy recoveries of 68.17 and 70.77%, and carbon recoveries of 66.91 and 64.78%, respectively. Furthermore, due to its high feedstock capacity and reaction volume, large amounts of biocrude oil and post-HTL wastewater (PHW) were obtained from this pilot-scale reactor to allow downstream research on upgrading biocrude oil for transportation fuel as well as PHW treatment and nutrient recovery.

09 BIOMASS FUELS↗
DOE OSTI2016

Transmutation Fuels Development in the Advanced Fuels Campaign

Metallic fuels have a history of use that spans the entire nuclear age, especially in association with the development of the LMFBR. Initially, metallic fuels were employed due to their ease of fabrication and high heavy metal density (which led to the most favorable breeding efficiencies in fast reactors). During subsequent decades of use, metallic fuel technology has matured significantly, leading to the realization of major benefits in the areas of fuel reliability and burnup, proliferation-resistant recycling, remote fabrication, and passive reactor safety. Such characteristics continue to make metallic fuels a highly attractive technology, especially for use with future sodium fast reactors. Current research and development activities on metallic fuels in DOE's Advanced Fuels Campaign are focused on their potential use for actinide transmutation in advanced SFRs. In such a mission, metallic fuels would be employed as part of a closed fuel cycle into which plutonium and minor actinides (Am, Np) recovered from LWR spent fuel would be incorporated. All actinides in this metallic fuel would be fissioned or transmuted in a fast reactor to eliminate the extremely long-lived, heat-producing actinides in order to reduce the volume, heat load, and radio-toxicity burden on a future geological repository. Thus, the need exists to demonstrate reliable performance of metallic fuels with Pu and minor actinide constituents to very high burnup. As part of this application, there is also a need for a near zero-loss fabrication process, since losses during fuel fabrication are a potential source of long-lived actinides destined for geologic disposal. This paper presents the current state and future direction of the development of metallic fuels for actinide transmutation in the US. Metallic fuels have had a long history of use in sodium-cooled fast reactors. Continued research, development, and design innovation over a period of decades have demonstrated this fuel technology to be safe and reliable to extremely high burnups over a wide range of compositions. Their simple fabrication has historically given them a significant economic advantage over other fuel forms. Research and development on metallic fuels continues, with particular emphasis on showing that this fuel technology can incorporate significant quantities of plutonium and minor actinides for the purpose of efficiently and economically transmuting these elements in a fast-spectrum reactor. New fabrication methods to improve retention of the volatile fuel constituents during the casting process (and reduce waste generation), are being developed. Irradiation testing of the new metallic fuels has so far indicated that the new compositions, although differing significantly both in the quantities of Pu and minor actinides included and in the Zr alloying addition, perform as a function of fission density very analogously to the historic U-Zr and U-Pu-Zr fuels. In the next few years it is anticipated that a new casting furnace prototype will demonstrate that these new metallic fuel alloys can be fabricated at engineering scale, in a remote environment, with no significant fuel loss or waste generation. Irradiation tests currently underway will provide the performance data needed to assess fuel reliability/burnup capability and identify any unique behaviors that are a result of the new compositions. If these future results are consistent with the intermediate data available at present, metallic fuels are expected to continue to be a strong contender for the fuel technology choice of future fast reactors. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

A micro-jet array for economic intensification of gas transfer in bioreactors

Bioreactors are of interest for gas-to-liquid conversion of stranded or waste industrial gases, such as CO, CH 4 , or syngas. Process economics requires reduction of bioreactor cost and size while maintaining intense production via rapid delivery of gases to the liquid phase (i.e., high k L a). Here, we show a novel bioreactor design that outperforms all known technology in terms of gas transfer energy efficiency (k L a per power density) while operating at high k L a (i.e., near 0.8 s -1 ). The reactor design uses a micro-jet array to break feedstock gas into a downward microbubble flow. Hydrodynamic and surfactant measurements show the reactor's advanced performance arises from its bubble breakage mechanism, which limits fluid shear to a thin plane located at an optimal location for bubble breakage. Power dissipation and k L are shown to scale with micro-jet diameter rather than reactor diameter, and the micro-jet array achieves improved performance compared to classical impinging-jets, ejector, or U-loop reactors. Additionally, the hydrodynamic mechanism by which the micro-jets break bubbles apart is shown to be shearing the bubbles into filaments then fragmentation by surface tension rather than “cutting in half” of bubbles. Guided by these hydrodynamic insights, strategies for industrial design are given.

09 BIOMASS FUELS↗
DOE OSTI2021

Survey of Cyber Risk Analysis Techniques for Use in the Nuclear Industry

Using traditional probabilistic risk analysis methods for severe accident safety risk management on non-digital systems, structures, and components at nuclear power plants is well-established. In contrast, cyber risk analysis of digital assets is still an immature field with unproven techniques due, in part, to the continuously changing threat environment and the challenge of digital assets failing in unexpected ways. As the nuclear fleet continues to adopt digital instrumentation and control systems, it is increasingly important to have effective and efficient cyber risk analysis techniques to support risk management decisions, such as risk elimination by system redesign or risk mitigation by implementation of prioritized security controls. To understand the state of the art in cyber risk analysis for future research, we surveyed 36 publications across ten application domains. We describe our survey methodology and rate each technique based upon scope, adoptability, and repeatability. In this work, we examine the unique constraints of the nuclear industry and outline the strengths and weaknesses of using the cyber risk analysis techniques in the industry, highlighting gaps with current techniques. We also discuss challenges and potential research directions for advancing the science for both existing and new advanced reactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

MELCOR Modeling of Non-LWR Systems Draft Report for the Nuclear Regulatory Commission

This report provides an overview of technical issues and design features relevant to advanced reactors and reviews MELCOR's current readiness for modeling accidents in such reactor types. This report describes advanced reactor physics models currently available or under development, and gauges the level of effort required to develop new models and capabilities applicable to assessing advanced reactor safety issues. Finally, this report reviews the available database that can be used in verification and validation of new models. Four general advanced reactor types are considered in this report: 1) High Temperature Gas-Cooled Reactor (HTGR); 2) Sodium Fast Reactor (SFR); 3) Molten Salt Reactor (MSR); 4) Fluoride Salt-Cooled High Temperature Reactor (FHR)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

AGR-5/6/7 Fuel Fabrication, Process Documentation, and Disposition Planning

The need for U. S. energy self-sufficiency dictates the development of new low-cost, environmentally-safe forms of energy generation, hydrogen production and industrial heat systems using high temperature gas-cooled reactor (HTGR) technology. Efficient fabrication of ultra-high quality, coated particle fuel is a central requirement for implementing gas reactor technology. Babcock & Wilcox Nuclear Operations Group-Lynchburg (B&W NOG-L) is uniquely qualified in the U.S. with the institutional knowledge, access to equipment, and a Nuclear Regulatory Commission (NRC) license to produce tristructural isotropic (TRISO) coated particle nuclear fuel above 5 wt% 235U. This Statement of Work (SOW) defines the Subcontractor, B&W NOG-L and identifies the work scope to perform Advanced Gas Reactor (AGR)-5/6/7 fuel fabrication, process documentation, and material disposition planning.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

AGR-5/6/7 Fuel Fabrication, Process Documentation, and Disposition

The need for U. S. energy self-sufficiency dictates the development of new low-cost, environmentally-safe forms of energy generation, hydrogen production and industrial heat systems using high temperature gas-cooled reactor (HTGR) technology. Efficient fabrication of ultra-high quality, coated particle fuel is a central requirement for implementing gas reactor technology. Babcock & Wilcox Nuclear Operations Group-Lynchburg (B&W NOG-L) is uniquely qualified in the U.S. with the institutional knowledge, access to equipment, and a Nuclear Regulatory Commission (NRC) license to produce tristructural isotropic (TRISO) coated particle nuclear fuel above 5 wt% 235U. This Statement of Work (SOW) defines the Subcontractor, B&W NOG-L and identifies the work scope to perform Advanced Gas Reactor (AGR)-5/6/7 fuel fabrication, process documentation, and material disposition planning.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

AGR-5/6/7 Fuel Fabrication, Process Documentation, and Disposition

The need for U. S. energy self-sufficiency dictates the development of new low-cost, environmentally-safe forms of energy generation, hydrogen production and industrial heat systems using high temperature gas-cooled reactor (HTGR) technology. Efficient fabrication of ultra-high quality, coated particle fuel is a central requirement for implementing gas reactor technology. Babcock & Wilcox Nuclear Operations Group-Lynchburg (B&W NOG-L) is uniquely qualified in the U.S. with the institutional knowledge, access to equipment, and a Nuclear Regulatory Commission (NRC) license to produce tristructural isotropic (TRISO) coated particle nuclear fuel above 5 wt% 235U. This Statement of Work (SOW) defines the Subcontractor, B&W NOG-L and identifies the work scope to perform Advanced Gas Reactor (AGR)-5/6/7 fuel fabrication.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

RELAP5-3D modelling of the HTTR-GT/H 2 secondary system and turbomachinery

High Temperature Gas-cooled Reactors (HTGRs) can be used for the cogeneration of electricity and hydrogen. The JAEA-operated High-Temperature engineering Test Reactor (HTTR-GT/H 2 ) is exploring using the reactor for electricity and hydrogen production. Here, a RELAP5-3D model of the HTTR-GT/H 2 secondary system has been developed using design information. The various components and heat exchangers in the secondary system were investigated, modified, and improved where applicable. With these changes made, the model was run, and the results were compared to the design conditions. The results for the sole-power generation mode were shown to fit the design conditions very well. The largest temperature difference was on the order of 7 K, and the largest pressure difference was on the order of 0.05 MPa. The results for the hydrogen cogeneration mode did not fit the design conditions as well as the other case. The largest temperature difference was about 39 K and the largest pressure difference was about 0.27 MPa at the compressor outlet. The larger differences for the hydrogen cogeneration mode are attributed to the various complex components and the flow being split in the secondary loop. A transient reduction in heat removal capability of the secondary system was investigated. Reactor temperatures were predicted to rise as a result. The core reactivity response due to this increase in temperature is investigated and is expected to add negative reactivity to the reactor.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
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 OSTI2012

Linear variable differential transformer (LVDT)-based elongation measurements in Advanced Test Reactor high temperature irradiation testing

New materials are being considered for fuel, cladding and structures in next generation and existing nuclear reactors. These materials can undergo significant dimensional and physical changes during high temperature irradiations. Currently, such changes are determined by repeatedly irradiating a specimen for a specified period of time in the Advanced Test Reactor (ATR) and then removing it from the reactor for evaluation. The labor and time to remove, examine and return irradiated samples for each measurement make this approach very expensive. In addition, such techniques provide limited data and may disturb the phenomena of interest. To resolve these issues, an instrumented creep testing capability is being developed for specimens irradiated under pressurized water reactor coolant conditions in the ATR at the Idaho National Laboratory (INL). This paper reports the status of INL efforts to develop this testing capability. In addition to providing an overview of in-pile creep test capabilities available at other test reactors, this paper focuses on efforts to design and evaluate a prototype test rig in an autoclave at INL’s High Temperature Test Laboratory.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2014

Reactor pressure vessel head vents and methods of using the same

Internal head vents are usable in nuclear reactors and include piping inside of the reactor pressure vessel with a vent in the reactor upper head. Piping extends downward from the upper head and passes outside of the reactor to permit the gas to escape or be forcibly vented outside of the reactor without external piping on the upper head. The piping may include upper and lowers section that removably mate where the upper head joins to the reactor pressure vessel. The removable mating may include a compressible bellows and corresponding funnel. The piping is fabricated of nuclear-reactor-safe materials, including carbon steel, stainless steel, and/or a Ni--Cr--Fe alloy. Methods install an internal head vent in a nuclear reactor by securing piping to an internal surface of an upper head of the nuclear reactor and/or securing piping to an internal surface of a reactor pressure vessel.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗