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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 649 records · Page 36
All sourcesNASA NTRSDOE OSTI
NASA NTRS1992

Near Earth asteroid orbit perturbation and fragmentation

Collisions by near earth asteroids or the nuclei of comets pose varying levels of threat to man. A relatively small object, approximately 100 meter diameter, which might be found on an impact trajectory with a populated region of the Earth, could potentially be diverted from an Earth impacting trajectory by mass driver rocket systems. For larger bodies, such systems would appear to be beyond current technology. For any size object, nuclear explosions appear to be more efficient, using either the prompt blow-off from neutron radiation, the impulse from ejecta of near-surface explosion for deflection, or as a fragmenting charge. Practical deflections of bodies with diameters of 0.1, 1, and 10 km require interception, years to decades prior to earth encounter, with explosions a few kilotons, megatons, or gigatons, respectively, of equivalent TNT energy to achieve orbital velocity changes or destruction to a level where fragments are dispersed to harmless spatial densities.

Ahrens, Thomas J.↗
NASA NTRS1993

Next generation system modeling of NTR systems

The topics are presented in viewgraph form and include the following: nuclear thermal rocket (NTR) modeling challenges; current approaches; shortcomings of current analysis method; future needs; and present steps to these goals.

Buksa, John J.↗
NASA NTRS2005

Heat Rejection Concepts for Brayton Power Conversion Systems

This paper describes potential heat rejection design concepts for closed Brayton cycle (CBC) power conversion systems. Brayton conversion systems are currently under study by NASA for Nuclear Electric Propulsion (NEP) applications. The Heat Rejection Subsystem (HRS) must dissipate waste heat generated by the power conversion system due to inefficiencies in the thermal-to-electric conversion process. Space Brayton conversion system designs tend to optimize at efficiencies of about 20 to 25 percent with radiator temperatures in the 400 to 600 K range. A notional HRS was developed for a 100 kWe-class Brayton power system that uses a pumped sodium-potassium (NaK) heat transport loop coupled to a water heat pipe radiator. The radiator panels employ a sandwich construction consisting of regularly-spaced circular heat pipes contained within two composite facesheets. Heat transfer from the NaK fluid to the heat pipes is accomplished by inserting the evaporator sections into the NaK duct channel. The paper evaluates various design parameters including heat pipe diameter, heat pipe spacing, and facesheet thickness. Parameters were varied to compare design options on the basis of NaK pump pressure rise and required power, heat pipe unit power and radial flux, radiator panel areal mass, and overall HRS mass.

Siamidis, John↗
NASA NTRS2006

A Comparison of Coolant Options for Brayton Power Conversion Heat Rejection Systems

This paper describes potential heat rejection design concepts for Brayton power conversion systems. Brayton conversion systems are currently under study by NASA for Nuclear Electric Propulsion (NEP) and surface power applications. The Brayton Heat Rejection Subsystem (HRS) must dissipate waste heat generated by the power conversion system due to inefficiencies in the thermal-to-electric conversion process. Sodium potassium (NaK) and H2O are two coolant working fluids that have been investigated in the design of a pumped loop and heat pipe space HRS. In general NaK systems are high temperature (300 to 1000 K) low pressure systems, and H2O systems are low temperature (300 to 600 K) high pressure systems. NaK is an alkali metal with health and safety hazards that require special handling procedures. On the other hand, H2O is a common fluid, with no health hazards and no special handling procedures. This paper compares NaK and H20 for the HRS pumped loop coolant working fluid. A detailed Microsoft Excel (Microsoft Corporation, Redmond, WA) analytical model, HRS_Opt, was developed to evaluate the various HRS design parameters. It is capable of analyzing NaK or H2O coolant, parallel or series flow configurations, and numerous combinations of other key parameters (heat pipe spacing, diameter and radial flux, radiator facesheet thickness, fluid duct system pressure drop, system rejected power, etc.) of the HRS. This paper compares NaK against water for the HRS coolant working fluid with respect to the relative mass, performance, design and implementation issues between the two fluids.

Mason, Lee S.↗
NASA NTRS2006

A Comparison of Coolant Options for Brayton Power Conversion Heat Rejection Systems

This paper describes potential heat rejection design concepts for Brayton power conversion systems. Brayton conversion systems are currently under study by NASA for Nuclear Electric Propulsion (NEP) and surface power applications. The Brayton Heat Rejection Subsystem (HRS) must dissipate waste heat generated by the power conversion system due to inefficiencies in the thermal-to-electric conversion process. Sodium potassium (NaK) and H2O are two coolant working fluids that have been investigated in the design of a pumped loop and heat pipe space HRS. In general NaK systems are high temperature (300 to 1000 K) low pressure systems, and H2O systems are low temperature (300 to 600 K) high pressure systems. NaK is an alkali metal with health and safety hazards that require special handling procedures. On the other hand, H2O is a common fluid, with no health hazards and no special handling procedures. This paper compares NaK and H2O for the HRS pumped loop coolant working fluid. A detailed Microsoft Excel (Microsoft Corporation, Redmond, WA) analytical model, HRS_Opt, was developed to evaluate the various HRS design parameters. It is capable of analyzing NaK or H2O coolant, parallel or series flow configurations, and numerous combinations of other key parameters (heat pipe spacing, diameter and radial flux, radiator facesheet thickness, fluid duct system pressure drop, system rejected power, etc.) of the HRS. This paper compares NaK against water for the HRS coolant working fluid with respect to the relative mass, performance, design and implementation issues between the two fluids.

Siamidis, John↗
NASA NTRSDate not supplied

Assessing the Needs for Space Reactor Standards

A U.S. governmental interagency space reactor standards working group (SWG) was convened to address the limited availability of voluntary consensus standards currently in place specifically for space nuclear reactor design and safety. The SWG focused on reviewing existing standards, assessing agency needs, specifying gaps that could be addressed by new standards, and prioritizing those gaps through consensus group deliberations. The SWG identified a series of findings and recommendations relative to the development of space reactor standards, including moving three specific high-priority gap items to be pursued through a consensus standards development process: Safety and Risk Analysis Methods for Space Reactors, Testing Requirements for Space Reactors, including Facility Requirements, and Safe Operating Practices for Space Reactors.

Andrew Klein↗
DOE OSTI2021

Results of the CHF-SERTTA In-Pile Transient Boiling Experiments at TREAT

To advance the current understanding of heat transfer between nuclear fuel rod cladding and the surrounding coolant under transient scenarios involving critical heat flux (CHF), the CHF-Static Environment Rodlet Transient Test Apparatus (CHFSERTTA), an in-pile experiment, was developed for and irradiated in the Transient Reactor Test (TREAT) facility at Idaho National Laboratory. The CHF-SERTTA experiment consists of a hollow, borated stainless steel rod submerged in a static water pool. The rod, heated via the (n, a) reaction in boron-10, was designed for investigating in-pile transient heat transfer between the rod surface and the coolant, avoiding the complexity of including the heat transfer between the fuel, gas gap, and cladding in typical light-water reactor (LWR) fuel rods. The instrumentation package for CHF-SERTTA focused on capturing those data that best describe the heat transfer behaviour between the rod and coolant. This paper outlines the CHFSERTTA experiment design and presents the initial results of the experiment campaign.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

The TRANSCEND Consortium - In-situ Identification of Surface Corrosion Products on Spent Nuclear Fuels - 20276

The management of spent nuclear fuel is a major ongoing concern for the UK owing to the cessation of reprocessing operations at Sellafield and the large, complex inventory arising from Magnox, AGR, PWR and prototype reactors. Retrieval and relocation operations for legacy fuels are imminent and therefore, any models that enhance our understanding of fuel evolution will help mitigate the risks associated with fuel storage and disposal. The TRANSCEND Consortium on nuclear waste management comprises four work packages, within this current paper we provide a summary overview of progress to date and illustrative results from Theme 3: Spent Nuclear Fuels. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Spent Nuclear Fuel Storage and Radioactive Waste Disposal in the United States: A Law and Policy Analysis - 20302

While Congress plays political football, spent nuclear fuel continues to sit in de facto interim nuclear waste storage sites throughout the country. This ad hoc approach - if it can even be called an approach - to nuclear waste management is in no one's best interests, because of financial, safety, environmental, and national security concerns with the status quo. Senate leadership had asserted that an up-or-down vote on the Yucca Mountain Project would take place in September or October 2019, but that did not occur. Rather, a conference bill was agreed to, and it provided that, although the nuclear program should receive some funding, funding was not made available for centralized interim storage ('CIS') or the Yucca Mountain Project. The President's FY2021 Budget Request does not contain any proposed funding for the Yucca Mountain Project ('the Project'). This paper will discuss the implications of that process, both for the Project itself and for the U.S. nuclear waste program more generally. In order to provide context for the current situation, this paper will begin by chronicling the history of U.S. nuclear waste management, from the passage of the Nuclear Waste Policy Act of 1982 and to the present day. It will discuss the roles of the U.S. Department of Energy, the Nuclear Regulatory Commission, federal and state political delegations, and the nuclear energy industry. The paper will then identify key themes that can be expected to affect the future of nuclear waste management in the United States. For example, it will consider whether development of additional CIS facilities is a viable short- to medium-term solution. CIS facilities may relieve immediate pressure at decommissioned or soon-to-be-decommissioned nuclear power plants; but with that pressure partially alleviated, support for building a long-term repository may be reduced. It will also discuss opposition by the State of New Mexico and others to the proposed New Mexico CIS project, based not only on siting concerns but also whether it might become a de facto repository, and whether the State of Texas and others may also be opposed, for similar reasons, to the proposed CIS facility in Texas. Additionally, the paper will examine consent-based siting and its implications for Nevada and other states that are potential hosts for spent nuclear fuel storage and disposal facilities. Further, if the Project is not a politically viable solution, the paper will examine whether it is now time for serious consideration of an alternate location for a permanent repository, as well as consideration of other options entirely, such as deep geologic disposal. The paper will conclude by looking ahead to the 2020 elections. It will describe the positions of candidates (particularly Presidential candidates) on spent nuclear fuel storage and disposal options, without advocating for any particular candidate or political party. Perhaps in 2020, we will at least see a start to a solution to the seemingly intractable problem of nuclear waste management in the United States. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Measuring Muon Antineutrino Charged-Current Interactions without Mesons in the Final State, in the NOvA Near Detector

NOvA is a long-baseline neutrino experiment based at Fermilab in the US, with the primary aim of measuring neutrino and antineutrino oscillations. This will enhance our understanding of electroweak interactions by measuring the neutrino mixing angles, CP-violating phase and neutrino mass ordering. To measure these oscillations, we first need to have a deep understanding of how neutrinos and antineutrinos interact with matter. Antineutrino interaction cross sections are, at present, particularly poorly constrained, and processes such as meson exchange currents are not well understood in the antineutrino sector. This analysis will develop a cross-section measurement of muon antineutrino interactions without mesons (e.g. pions or kaons) in the final state, in the NOvA near detector. A high-statistics, high-purity sample is obtained through a cut-based selection process implementing machine learning techniques. The sample is dominated by quasi-elastic and meson exchange current interactions which are sensitive to nuclear effects such as Final-State Interactions. The cross section will be extracted as a function of the incoming neutrino energy and the kinematics of the outgoing particles. This presentation will give an overview of the analysis and discuss progress towards obtaining the cross-section measurement.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2021

Verification of Subcritical Limits in ANSI/ANS-8.1-2014

The first edition of the current American National Standards Institute (ANSI)/American Nuclear Society (ANS) standard ANSI/ANS-8.1-2014, “Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors,” was published in 1964 as ASA N6.1-1964, “Safety Standard for Operations with Fissionable Materials Outside Reactors.” In 1969, that standard was revised as ANSI N16.1-1969, “Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors.” ANSI N16.1-1969 includes a variety of subcritical limits (SCLs) for uniform aqueous solutions and metals containing fissile nuclides of 233 U, 235 U, and 239 Pu. Furthermore, SCLs are also included for uranium-water lattices. In the 1983 version of ANSI/ANS-8.1 (a revision of ANSI N16.1-1975), the suite of SCLs in the standard was expanded to include 235 U enrichment limits for homogeneous uranium-water mixtures and dry/damp oxides, uniform aqueous solutions of low-enriched uranium, and uniform aqueous mixtures of Pu(NO 3 ) 4 containing 240Pu in addition to the SCLs included in ANSI N16.1-1969. The SCLs have changed little in subsequent revisions (ANSI/ANS-8.1-1998 and ANSI/ANS-8.1-2014). The ANSI/ANS-8.1-2014 standard is currently being revised to include new SCLs (uranium metal and compounds with enrichments up to 20 wt% 235 U) and possible updates to the current SCLs already in the standard, although these new/updated SCLs will not be available to the nuclear criticality safety (NCS) community for a number of years. The original bases for these SCLs were documented in papers in journals such as Nuclear Science and Engineering and in ANS meeting transactions; however, these bases are ambiguous enough that sites and regulators in the United States have not widely endorsed them for safety purposes. Overall, the purpose of this paper is to present the results of a comparison study for the SCLs in the ANSI/ANS-8.1-2014 standard using modern codes (SCALE and MCNP) and cross sections (ENDF/B-VIII.0) to provide some assurance about their quality (bias and bias uncertainty) for use in NCS applications and for consideration by the ANS-8.1 Working Group as a reference for future revisions.

42 ENGINEERING↗
DOE OSTI2021

Nuclear diagnostics for Inertial Confinement Fusion (ICF) plasmas

The field of nuclear diagnostics for Inertial Confinement Fusion (ICF) is broadly reviewed from its beginning in the seventies to present day. During this time, the sophistication of the ICF facilities and the suite of nuclear diagnostics have substantially evolved, generally a consequence of the efforts and experience gained on previous facilities. As the fusion yields have increased several orders of magnitude during these years, the quality of the nuclear-fusion-product measurements has improved significantly, facilitating an increased level of understanding about the physics governing the nuclear phase of an ICF implosion. The field of ICF has now entered an era where the fusion yields are high enough for nuclear measurements to provide spatial, temporal and spectral information, which have proven indispensable to understanding the performance of an ICF implosion. At the same time, the requirements on the nuclear diagnostics have also become more stringent. To put these measurements into context, this review starts by providing some historical remarks about the field of ICF and the role of nuclear diagnostics, followed by a brief overview of the basic physics that characterize the nuclear phase and performance of an ICF implosion. A technical discussion is subsequently presented of the neutron, gamma-ray, charged-particle and radiochemistry diagnostics that are, or have been, routinely used in the field of ICF. This discussion is followed by an elaboration of the current view of the next-generation nuclear diagnostics. Since the seventies, the overall progress made in the areas of nuclear diagnostics and scientific understanding of an ICF implosion has been enormous, and with the implementation of new high-fusion-yield facilities world-wide, the next-generation nuclear diagnostics will play an even more important role for decades to come.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
NASA NTRS1992

Nuclear electric propulsion: An integral part of NASA's nuclear propulsion project

NASA has initiated a technology program to establish the readiness of nuclear propulsion technology for the Space Exploration Initiative (SEI). This program was initiated with a very modest effort identified with nuclear thermal propulsion (NTP); however, nuclear electric propulsion (NEP) is also an integral part of this program and builds upon NASA's Base Research and Technology Program in power and electric propulsion as well as the SP-100 space nuclear power program. Although the Synthesis Group On America's SEI has identified NEP only as an option for cargo missions, recent studies conducted by NASA-Lewis show that NEP offers the potential for early manned Mars missions as well. Lower power NEP is also of current interest for outer planetary robotic missions. Current plans are reviewed for the overall nuclear propulsion project, with emphasis on NEP and those elements of NTP program which have synergism with NEP.

Stone, James R.↗
DOE OSTI2021

THERMAL STRESS ANALYSIS OF A SPENT NUCLEAR FUEL CANISTER

The potential for chloride induced stress corrosion cracking (CISCC) in spent nuclear fuel dry storage canisters is a current topic of research by the US Department of Energy Spent Fuel and Waste Science and Technology program. One of the important prerequisites for CISCC is a tensile residual stress state. This study utilizes computational models to provide an initial analysis of thermal stresses in a generic spent nuclear fuel canister. A STAR-CCM+ thermal fluid model provides a temperature profile analysis of the canister based on four different ambient temperature conditions. An ANSYS APDL finite element model incorporates the temperature profiles to analyze the thermal stresses in the canister. The calculated thermal stress magnitudes are in the range of 10 MPa to 80 MPa, which could be significant for crack propagation through the canister wall. The maximum thermal stress is not a concern for the structural failure of the canister, but it is high enough that it is a significant contribution to the total stress state of the canister wall, which also includes residual stress from fabrication, internal pressure from helium cover gas, and potential transient mechanical loads, such as earthquakes. This paper describes the initial finite element thermal stress analysis that was completed in 2020, reports the initial findings, and identifies areas where model refinement is still needed to complete this work in the future.

Jensen, Ben J.↗
NASA NTRS2006

Design Considerations for the Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

Nuclear Thermal Rockets or NTR's have been suggested as a propulsion system option for vehicles traveling to the moon or Mars. These engines are capable of providing high thrust at specific impulses at least twice that of today s best chemical engines. The performance constraints on these engines are mainly the result of temperature limitations on the fuel coupled with a limited ability to withstand chemical attack by the hot hydrogen propellant. To operate at maximum efficiency, fuel forms are desired which can withstand the extremely hot, hostile environment characteristic of NTR operation for at least several hours. The simulation of such an environment would require an experimental device which could simultaneously approximate the power, flow, and temperature conditions which a nuclear fuel element (or partial element) would encounter during NTR operation. Such a simulation would allow detailed studies of the fuel behavior and hydrogen flow characteristics under reactor like conditions to be performed. The goal of these simulations would be directed toward expanding the performance envelope of NTR engines over that which was demonstrated during the Rover and NERVA nuclear rocket programs of the 1970's. Currently, such a simulator is nearing completion at the Marshall Space Flight Center, and will shortly be used in the future to evaluate a wide variety of he1 element designs and the materials of which they are constructed. This present work addresses the initial experimental objectives of the Nuclear Thermal Rocket Element Environmental Simulator or NTREES and some of the design considerations which were considered prior to and during its construction.

Emrich, Bill↗
DOE OSTI2023

Initial Results of PostProcess NDE Using Advanced Techniques

The U.S. Department of Energy, Office of Nuclear Energy Advanced Materials and Manufacturing Technologies Program is developing advanced materials and manufacturing technologies that will enable the safe and economical operation of both the current fleet and the next generation of advanced nuclear reactors. Post-process nondestructive evaluation capabilities are a key enabling technology that will be required to qualify and verify the condition of components produced using advanced materials and manufacturing technologies.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Airborne Release Fractions from Surrogate Nuclear Waste Fires Containing Lanthanide Nitrates and Depleted Uranium Nitrate in 30% Tributyl Phosphate in Kerosene

Airborne contaminants from fires containing nuclear waste represent significant health hazards and shape the design and operation of nuclear facilities. Much of the data used to formulate DOE-HDBK-3010-94, “Airborne Release Fractions/Rates and Respirable Fractions for Nonreactor Nuclear Facilities,” from the U.S. Department of Energy, were taken over 40 years ago. The objectives of this study were to reproduce experiments from Pacific Northwest Laboratories conducted in June 1973 employing current aerosol measurement methods and instrumentation, develop an enhanced understanding of particulate formation and transport from fires containing nuclear waste, and provide modeling and experimental capabilities for updating current standards and practices in nuclear facilities. A special chamber was designed to conduct small fires containing 25 mL of flammable waste containing lutetium nitrate, ytterbium nitrate, or depleted uranium nitrate. Carbon soot aerosols showed aggregates of primary particles ranging from 20 to 60 nm in diameter. In scanning electron microscopy, ~200-nm spheroidal particles were also observed dispersed among the fractal aggregates. The 200-nm spherical particles were composed of metal phosphates. Airborne release fractions (ARFs) were characterized by leaching filter deposits and quantifying metal concentrations with mass spectrometry. The average mass-based ARF for 238 U experiments was 1.0 × 10 -3 with a standard deviation of 7.5 × 10 -4 . For the original experiments, DOE-HDBK-3010-94 states, “Uranium ARFs range from 2 × 10 -4 to 3 × 10 -3 , an uncertainty of approximately an order of magnitude.” Therefore, current measurements were consistent with DOE-HDBK-3010-94 values. ARF values for lutetium and ytterbium were approximately one to two orders of magnitude lower than 238 U. Furthermore, metal nitrate solubility may have varied with elemental composition and temperature, thereby affecting ARF values for uranium surrogates (Yb and Lu). In addition to ARF data, solution boiling temperatures and evaporation rates can also be deduced from experimental data.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Hazards and Probabilistic Risk Assessments of Advanced Nuclear Reactors Coupled with Industrial Facilities

This report provides a roadmap and tool kit for site specific risk assessments across a broad range of industrial customers co-located with advanced nuclear power plants (ANPP) that are not currently built and operating in the U.S. This report builds upon the body of work sponsored by the Department of Energy (DOE) Integrated Energy Systems Pathway that has produced industrial requirements studies and techno-economic assessments on the topics of feasibility of ANPP supported industrial processes. This report also leverages the DOE Light Water Reactor Sustainability (LWRS) program that has presented hazards assessment and generic probabilistic risk assessments (PRAs) for the addition of a heat extraction system (HES) to light-water reactors (LWRs) co-located with hydrogen production facilities. Many of the hazard assessments and risk assessments performed for the LWRS report are agnostic to whether the nuclear reactor is an ANPP or were adapted to the ANPP focus. The report performs hazards assessments to include industrial facilities: an oil refinery, a methanol plant, a synthetic fuel (synfuel) plant, the production of synthetic gas (syngas) as part of the methanol and synfuel plants, wood pulp and paper mills, and hydrogen production. Hydrogen production facilities are assessed in depth through prior reports in the LWRS program and the results are leveraged in this report. All these facilities are specified through industrial process and requirements research performed by national laboratories, universities, and interaction with industry. Many of the processes used in this report are pre-conceptual designs to use for decarbonization of the current technology facilities. A process of failure modes and effects analysis (what can go wrong) and accidentology (what has historically gone wrong) was used to determine the hazards presented to the nuclear power plant by the addition of the HES and the industrial customer. Chemical properties of feedstocks and products are summarized as part of the hazards assessment. Example analysis procedures are provided for each of the hazard types identified. These deterministic analyses can be used to assess adherence to licensing criteria. They can also be used to meet other safety goals like protection of the public, workers, or industrial facility equipment. A modular high temperature gas-cooled reactor (MHTGR) PRA only existing on paper was modeled and verified in modern PRA software. This will provide a tool for representative ANPP probabilistic analyses for future research.

10 SYNTHETIC FUELS↗