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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 721 records · Page 40
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Aerosol-Based Sealant for Air Ducts Technology

Aerosol-based sealant can be used to seal air duct leaks. This is done by pressurizing the ducts and injecting the spray sealant. This technology can be used to seal small (up to ½-inch diameter) and inaccessible duct leaks hidden in wall, ceiling, and floor cavities (Ternes and Hwang2001). The aerosol sealant itself is a nontoxic vinyl polymer (Kallett et al. 2000). The technology is a proprietary duct-sealing method developed jointly by the U.S. government and Aeroseal, the company that holds an exclusive license to use the technology (Ternes and Hwang 2001). Specifically, Lawrence Berkeley National Laboratory developed the aerosol-spray technology with funding from the U.S. Department of Energy (DOE), the U.S. Environmental Protection Agency, and others while Aeroseal developed the current version of the equipment and software to apply the technology in practice (Ternes and Hwang 2001).

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2022

Fast Modular Reactor (FMR): Overview and Conceptual Design Work

The conceptual design of the FMR including the experimental and numerical verifications will lead to: Cost estimation of the FMR based on the conceptual design results and guidelines of advanced reactor construction. Submittal of the pre-application licensing plan to the NRC for review and feedback.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Radiation Shielding Analysis of Niowave’s Uranium Target Assembly 2 (UTA-2) Facility for Molybdenum-99 Production

Niowave is seeking to develop a commercial medical isotope production process using a low-enriched uranium accelerator-driven subcritical reactor (ADSR) system to produce molybdenum-99. As part of the US Department of Energy national laboratory support to develop low enriched uranium molybdenum-99 production technologies, Oak Ridge National Laboratory is providing research and development resources to assist Niowave in the design and licensing process of its technology. The goals of this work under the radiation shielding task are to assist Niowave in developing a methodology and framework for efficient radiation shielding simulations of their Uranium Target Assembly 2 facility and to provide preliminary radiation shielding calculations and design recommendations for the facility.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2021

Cold Spray for Mitigation and Repair of Spent Nuclear Fuel Dry Storage Canisters

The purpose of this work is to investigate the use of cold spray for repair and mitigation of chloride-induced stress corrosion cracking (CISCC) in dry cask storage system (DCSS) canisters to ensure their integrity far beyond their original license period. This report provides a viability analysis of cold spray for repair and mitigation of CISCC in DCSS canisters to extend canister life.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Data Quality of Salt Property Measurements

Data quality is a key element to assessing technical uncertainties in property values of salt mixtures that are pivotal in molten salt reactor (MSR) design and safety assessments under both normal operating conditions and accident scenarios. Key properties for the salt mixtures being evaluated for use in MSRs have not been measured and the few values for relevant salt compositions available in the literature are inconsistent and not suitable for use in licensing. The applications of established techniques commonly used for property measurements to molten salts are still experimental in nature and vetted consensus standards are not yet available. Work is in progress to develop and standardize methods through DOE and commercially funded projects. The quality of measured property values limits the current ability to reliably design equipment and model normal MSR operation and the effects of operational upsets. Insights from experimental measurements pertinent to reducing technical uncertainty for regulatory development are summarized in this report. Quality controls are recommended for key properties based on the mathematical calculations used to derive property values and methods used to measure experimental values and salt produced for use in the measurements.

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 OSTI2022

Increasing Z-Strength and Testing the Capabilities of Twin Screw Extruders in Large Format Polymer Additive Manufacturing

During the first phase of this project, small and large Strangpresse extruders were assessed at Oak Ridge National Laboratory’s Manufacturing Demonstration Facility (MDF). This phase resulted in Strangpresse exclusively licensing some of ORNL’s previously developed extruder technology for high-volume thermoplastic additive manufacturing (AM). Phase II of this project will focus on further development and optimization of Strangpresse extruders for increased efficiency and increased throughput. The objective of this project is to develop a new extruder with a higher output and less energy usage than current AM thermoplastic extruders and to develop and evaluate a new tamper mechanism for increasing Z-strength resulting in an advanced and optimized extruder and tamper system.

42 ENGINEERING↗
DOE OSTI2022

Development and Demonstration of a Risk-Informed Approach to the Regulatory Required Fuel Reload Safety Analysis

The United States (U.S.) nuclear industry is facing a strong challenge to maintain regulatory-required levels of safety while ensuring economic competitiveness to stay in business. Safety remains a key parameter for all aspects related to the operation of light water reactor (LWR) nuclear power plants (NPPs) and can be achieved more economically by using a risk-informed ecosystem such as that being developed by the Risk-Informed Systems Analysis (RISA) Pathway under the U.S. Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) Program. The LWRS Program is promoting a wide range of research and development (R&D) activities with the goal to maximize both the safety and economically efficient performance of NPPs through improved scientific understanding, especially given that many plants are considering second license renewal. The RISA Pathway has two main goals: (1) the deployment of methodologies and technologies that enable better representation of safety margins and the factors that contribute to cost and safety; and (2) the development of advanced applications that enable cost-effective plant operation. The plant reload optimization framework development project aims to build an artificial intelligence, i.e., Genetic Algorithm (GA), based reactor core designing tool taking into account reactor safety and fuel performance analyses.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Effect of thermal aging on microstructure and crack growth response of Alloy 152 weld

Nickel-based Alloy 690 and the associated weld Alloys 52 and 152 are typically used for nozzle penetrations in replacement heads for pressurized water reactor (PWR) vessels, because of their excellent overall resistance to general corrosion and environmental degradation, primarily stress corrosion cracking (SCC). However, many of the existing PWRs are expected to operate for 40-80 years. Likewise, advanced water-cooled small modular reactors (SMRs) will use Ni-Cr alloys and are expected to receive initial operating licenses for 60 years. Hence, the thermal stability of Ni-Cr alloys is a potential concern for the long-term performance of both existing and advanced nuclear power plants, and possibly spent fuel storage containers. The objective of this research is to understand the microstructural changes occurring in high-Cr, Ni-based Alloy 152 weldments during long time exposure to the reactor operating temperatures, and the effect of these changes on the service performance. One area of particular concern is the potential for long range ordering (LRO), i.e. formation of the intermetallic Ni2Cr phase under prolonged exposure to reactor temperatures and/or irradiation, which can increase strength, decrease ductility, and cause dimensional changes or lead to in-service embrittlement of components made with these alloys. Hence, this research focused on the microstructural evolution and the SCC response of Alloy 152 under accelerated thermal aging. The materials studied involved three heats of Alloy 152 used to produce a dissimilar metal weld (DMW) joining an Alloy 690 plate to an Alloy 533 low alloy steel (LAS) plate, thermally aged at three different temperatures (370°C, 400°C and 450°C) for up to 75,000h (equivalent to 60 years of service). The microstructural characterization by means of synchrotron X-ray did not find evidence of LRO in any of the three heats aged to an equivalent of 60 years of service. Testing in a primary water environment of a heat of Alloy 152 aged at 370°C to a 60-year service equivalent revealed a fatigue and corrosion fatigue crack growth responses similar to those measured on the un-aged alloy. However, the SCC CGR response of the aged sample appears to show a deterioration in performance.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Molten Salt Reactor Technical and Safety Considerations Outside of Guidance Documents

This document provides information on distinctive characteristics of liquid salt–fueled molten salt reactors (MSRs) for US Nuclear Regulatory Commission (NRC) consideration as it seeks to achieve effective and efficient advanced reactor mission readiness. The NRC has requested that Oak Ridge National Laboratory provide advice regarding technical and safety considerations that might be advantageous to address via policy, rulemaking, or license conditions (i.e., beyond the scope of guidance documents) for accommodating MSRs. It is recognized that MSRs could be regulated based on existing rules, with exceptions, to reflect their distinctive characteristics and technologies. However, directly applying the existing regulatory processes that evolved with the light water reactor fleet to MSRs could be unduly burdensome to the point that their application would inhibit deployment of MSRs in the United States.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

SCALE Calculations Replicating ARH-600 Data

This report documents Standardized Computer Analyses for Licensing Evaluation (SCALE) calculations replicating data from the ARH-600 criticality handbook. These calculations complement those reproduced with MCNP in PRC-NS-00009. Calculations were performed with SCALE version 6.1, using the v7-238 cross section library. The original ARH-600 data were calculated by the Atlantic Richfield Hanford (ARH) Company starting in the 1960s, and Monte Carlo N-Particle (MCNP) calculations were performed by the Hanford Plateau Remediation Company in 2010. This report is not intended to replace final analysis of fissile systems by qualified criticality personnel.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

SAS4A/SASSYS-1 Validation with EBR-II Tests Performed During the SHRT Testing Program

The EBR-II sodium-cooled fast reactor supported a wide range of fast reactor research activities. The final fifteen years of operations at EBR-II were used for experiments and tests to demonstrate the importance of passive safety in liquid metal reactors and the capability of such systems to provide strong passive responses during off-normal conditions. The Shutdown Heat Removal Test program at EBR-II provided test data supporting the validation of computer codes for design, licensing, and operation of LMRs, among other objectives. The program included nearly sixty tests, most of which were protected and unprotected loss of flow and loss of heat sink tests run at various initial powers and flow rates. This report documents the results of modeling and simulation of tests from the Shutdown Heat Removal Test program to support validation of Argonne’s SAS4A/SASSYS-1 fast reactor safety analysis code. SHRT-17 was a protected loss of flow test where a loss of electrical power to all sodium coolant pumps was simulated to demonstrate the effectiveness of natural circulation cooling characteristics. SHRT-45R was an unprotected loss of flow test where the control rod scram function of the plant protection system was disabled to demonstrate the effectiveness of EBR-II’s passive reactivity feedbacks. BOP-301 and BOP-302R were unprotected loss of heat sink tests that demonstrated how reactivity feedback effects driven by changes to the core inlet temperature can shut down the fission process. This validation activity leverages modeling and simulation efforts that originated under an International Atomic Energy Agency Coordinated Research Project for benchmark analysis of the EBR-II Shutdown Heat Removal Tests. The SAS4A/SASSYS-1 models originally developed for the CRP have been modernized to utilize new code features and modeling practices. Differences between the predicted and measured test data were explored and the causes of these differences were identified as being due to modeling approximations, uncertainty in transient component behavior, and instrumentation uncertainties. Overall, it was concluded that for SHRT-17, SHRT-45R, BOP-301, and BOP-302R there was good agreement between the flow, power, and temperature predictions and the measured data.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Impact of Coating Defects on Performance of Coated Zirconium Cladding

Research on accident tolerant fuels (ATF) has started after the Fukushima accident [1–3]. While efforts have been expended on both fuel and cladding ATF concepts, the bulk of work has been devoted to improved cladding. The overarching goal of these approaches is to extend the coping time available during a severe accident before any event would result in release of radioactivity to the public. The most basic ATF cladding concept is obtained by applying a thin coating of highly corrosion-resistant material on the surface of a licensed zirconium cladding alloy. This thin coating is intended to not interfere with the neutronic or mechanical performance of the base cladding under normal operating conditions. Different coating materials, thicknesses, coating processes, process parameters, and testing methods have an impact on the microstructure and mechanical properties and therefore on the results of the applied investigation methods. These challenges have motivated an initial focus on demonstrating that the presence of coatings do not perturb the critical performance benchmarks of uncoated material. Ongoing lead test assembly irradiations of coated zirconium concepts in commercial reactions is intended to establish baseline performance in this regard in the coming years.

36 MATERIALS SCIENCE↗
DOE OSTI2022

MSR Salt Spill Accident Testing Using Eutectic NaCl-UCl 3

Assessing the potential consequences of identified accident scenarios is an essential part of the licensing process for a new nuclear reactor and is achieved by using accident progression models. A likely accident scenario that will be evaluated by developers of molten salt reactors (MSRs) is a spill of radionuclide-bearing fuel salt onto the reactor containment floor (i.e., a salt spill accident). Models to determine the consequences of a molten salt spill accident need to be parameterized and validated using experimental data, but little experimental effort has been dedicated to fill these data gaps to date. Specifically, data is needed to enable model development for individual processes (e.g., spreading, heat transfer, corrosion, radionuclide vaporization, and aerosol generation) that quantifies the sensitivities towards the initial conditions of the spill, the ambient environment, and the features of the containment. In addition, integrated experiments that simulate salt spill accidents will need to be conducted to provide insight into coupled processes and data for model validation, but these experiments will require the use of proven methods to quantify the processes under evaluation.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Mechanical Testing of Additively Manufactured Materials

This report summarizes research activities conducted at Argonne National Laboratory in support of the development, qualification and certification of additively manufactured (AM) metallic components to allow for innovative reactor design and licensing. In this program, AM 316L stainless steel (SS) was evaluated in both light water reactor (LWR) environments as well as at higher temperatures expected in advanced reactors.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Comparative Structural Rigidity Analysis of the NBSR DDE and the NBSR LEU Fuel Element

The National Bureau of Standards Reactor (NBSR) is a 20-MW heavy-water moderated and cooled, enriched-fuel, tank-type reactor located in the NIST Center for Neutron Research laboratory complex. NBSR is one of six U.S. high performance research reactors (USHPRR), including one critical facility, that actively collaborates with the National Nuclear Security Administration (NNSA) Material Management and Minimization (M3) Reactor Conversion Program to convert from the use of highly enriched uranium fuel to the use of low-enriched uranium (LEU, < 20 wt% U-235) fuel. On the LEU U-10Mo fuel qualification side, mini-plate and large-plate irradiations have been successfully performed, and more experiments are either ongoing or planned. As an additional experimental campaign to support licensing of the LEU fuel for the use in the NBSR, the NBSR Design Demonstration Element (DDE) will be a full-sized test element similar to the NBSR LEU fuel element and irradiated in a test reactor under conditions as prototypic of the NBSR LEU fuel element operating conditions as possible.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Multiphysics Analysis of Load Following and Safety Transients for MicroReactors

The tools developed within the U.S. Department of Energy (DOE) Office of Nuclear Energy’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program aim at providing high fidelity multiphysics modeling capabilities to support design and licensing of various types of advanced nuclear reactors, including the technologies being developed by U.S. microreactor vendors relying on heat pipe and gas-cooled technologies. In FY-2022, the NEAMS Multiphysics Applications team made significant progress both in demonstrating capabilities applied to microreactor problems, and in supporting NEAMS developers.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Ageing Management for Extended Long-Term Dry Storage of Spent Nuclear Fuel and Transportation

The principal objectives of the Coordinated Research Project (CRP) and CRADA were to (1) investigate how ageing effects leading to degradation of materials used in the spent fuel dry storage systems could be managed by ageing management programs (AMPs) and (2) using existing AMPs as a basis, establish guidance on how to develop, generate, and maintain AMPs for dry storage systems of spent nuclear fuel (SNF) that can be accomplished in various ways. The PI (Dr. Liu) is an internationally recognized expert on ageing management for license renewal of nuclear power plants and independent spent fuel dry storage installations. Dr. Liu was invited by IAEA to participate in this CRP and served as Chair of the Working Group, supporting the IAEA CRP Lead, with other members from Argentina, the Czech Republic, France, Germany, Hungary, Japan, Pakistan, Spain, Switzerland, United Kingdom, and the United States of America. Argonne’s scope of work in the CRADA for the CRP included (1) ageing management guidance documents developed by Argonne for DOE and used by the Nuclear Regulatory Commission (NRC) and industry and (2) research leveraged from the DOE Office of Nuclear Energy on the mechanical properties of high-burnup fuel cladding and the ARG-US remote monitoring systems technology developed for the DOE Packaging Certification Program, Office of Packaging and Transportation, Office of Environmental Management.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗