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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 505 records · Page 28
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Open-Source Contributions to Arbiter2

Login nodes at High-Performance Computing (HPC) sites are shared resources used by a multitude of users to compile, test and submit jobs to the batch system. Because these nodes are shared by multiple users at any time, if a minority of users are using a significant proportion of the resources of the node (CPU, memory, etc.), other users' tasks may be negatively impacted. Arbiter2 is an open-source project developed by the University of Utah that aims to prevent these occurrences by dynamically limiting the resources of users depending on whether they are excessively using resources. INL has sought to adapt and develop Arbiter2 for a potential deployment at INL and plans on upstreaming the changes and improvements so that other HPC sites running Arbiter2 can benefit from their work.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2020

Magnetics Testing on Radioisotope Power Systems at the Idaho National Laboratory

National Aeronautics and Space Administration (NASA) uses Radioisotope Power Systems (RPS) to power deep-space and planetary explorations such as Cassini-Huygens, Galileo, New Horizons, Mars Science Laboratory—Curiosity, and most recently Mars 2020--Perseverance. The Space Nuclear Power and Isotope Technologies (SNPIT) division, at Idaho National Laboratory (INL), fuels, tests, and delivers the RPS to ensure it can provide the electrical power needed to complete mission objectives. Some space crafts have instrumentation specifically designed to study the subject’s magnetic field or may have instruments that can be affected by magnetic fields produced by the RPS. Analyzing and measuring the RPS’s magnetic fields ensures the required thresholds are not exceeded and accurate data can be collected for the mission. Magnetics testing provides numerical values of the RPS’s magnetic field. During Magnetics testing at INL, the RPS is placed on a resistive load to simulate actual spacecraft load characteristics (voltage and current). During this simulation, magnetometers are used to measure the magnetic field of the RPS. Raw data is collected and then analyzed to ensure the field created by the RPS is below NASA thresholds or that the magnitude of the field is accounted for to ensure mission success.

42 ENGINEERING↗
DOE OSTI2020

Mechanical Properties of Aged A709

A709 plate is in the process of being developed and qualified through a collaboration by Argonne National Laboratory (ANL), Idaho National Laboratory (INL), and Oak Ridge National Laboratory (ORNL). The goal is to qualify A709 plate in Section III, Division 5 of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC). This would permit the construction of A709 plate for fast-reactor structural applications. A tradeoff between creep and creep-fatigue properties was established in prior work. A processing route was identified as providing the optimal balance between creep-rupture and elevated-temperature cyclic properties for the solution annealed condition. This condition, however, resulted in a shorter than desired creep-rupture life. Aging is probed to evaluate its potential at further improving mechanical properties. Aging statically precipitates the solutes in solution prior to service. The processing routes identified as providing the best creep-rupture properties and best balance in mechanical properties are investigated. This report discloses the results from tensile, creep, and elevated-temperature cyclic testing performed at INL on the aged material. The aged properties are compared to solution annealed properties. A709 offers improvement in performance compared to 316H particularly at high temperatures; 316H is a material qualified in Section III, Division 5 of the ASME BPVC. For certain high-temperature fast-reactor design conditions, 316H cannot be used while A709 can.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Single-Bar SMT Testing on Alloy 617 using Software Controls

INL has performed simplified model testing using a variable in the software control program to compare with the testing performed at ORNL, which uses hardware to modify the control signal. The INL tests compared well with the ORNL tests, showing similar stress and strain values, as well as similarly shaped hysteresis loops. This demonstrates that possibility of performing simplified model testing without any hardware modifications.

36 MATERIALS SCIENCE↗
DOE OSTI2020

AGR-2 PIE at Oak Ridge National Laboratory

The Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) is currently supporting a fuel development and qualification program, which includes fuel fabrication, test irradiations, and post-irradiation examination (PIE) and safety testing to assess fuel performance during normal irradiation and under potential accident conditions. PIE work on fuel from the second test irradiation, Advanced Gas Reactor-2 (AGR-2), began at INL in July 2014. This work scope includes Oak Ridge National Laboratory (ORNL) providing technical input, performing PIE testing and analysis, and contributing expertise to this effort.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

AGR-3/4 PIE at Oak Ridge National Laboratory

The Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) is currently supporting a tristructural isotropic (TRISO) fuel development and qualification program, which includes fuel fabrication, test irradiations, and post-irradiation examination (PIE) and safety testing to assess fuel performance during normal irradiation and under accident conditions. PIE work on fuel from the third and fourth test irradiations, Advanced Gas Reactor-3/4 (AGR-3/4), began at INL in April 2015. The work scope in this memorandum purchase order (MPO) includes Oak Ridge National Laboratory (ORNL) providing technical input, preparations for PIE testing and analysis, and contributing general expertise to this effort.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Development of Mechanical Testing Capability for HFEF

This report depicts the development of mechanical testing capability for ATF fuels at Idaho National Laboratory (INL). An aim of the study is to demonstrate the pathway of defueling an irradiated cladding segment, machining of segments into specimens with unique tensile geometries, and mechanical testing using INL’s facilities. The report consists of three parts: 1) defueling of a fuel segment, 2) mechanical test specimen machining, and 3) mechanical test demonstration.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Modeling and Analysis of the Transport and Disposal of Beryllium Moderator Blocks and Greater than Class C (GTCC) Waste

Radioactive waste in the United states is categorized based on its source, radioactivity, and security risk. The categorization method employed by the Nuclear Regulatory Committee (NRC) for low-level radioactive waste is not currently applied to any waste generated by the Department of Energy (DOE) or disposed of in DOE facilities. As a result, there is a large volume of DOE-generated waste with characteristics similar to the NRC’s “Greater than Class C” (GTCC) waste category which presently do not have a path for long-term disposal. Much of this waste cannot undergo the standard commercial disposition process due to it being categorized as Transuranic (TRU) waste under DOE guidelines, due to elevated concentrations of key fission products. Possible solutions to this dilemma are explored in the Environmental Impact Statement for the “Disposal of Greater-Than-Class-C Low Level Radioactive Wave” (EIS-0375). This paper analyzes several EIS possible paths for disposal for this “orphaned” waste. One example that highlights this categorization issue is the spent beryllium cladding that has been extracted from the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) throughout its operational lifetime. These beryllium blocks surround the reactor’s main chambers and components and serve as neutron moderators. This paper will use this particular waste form to analyze the economic and technological viability of transporting and storing this material under the storage and transportation criteria of the Waste isolation Pilot Plant (WIPP), which was deemed the most feasible disposition path according to the EIS. The example of this waste form used in this paper that highlights this categorization issue is the spent beryllium cladding extracted from Idaho National Laboratory’s (INL) Advanced Test Reactor (ATR). These beryllium blocks surround the reactor’s components and serve as neutron moderators, and thus have accumulated high concentration of high-activity transuranic isotopes. This paper analyzed the viability of transporting and storing this material under the storage and transportation criteria of the five primary disposition paths covered in EIS-0375. An array of calculations was carried out to assess the viability of the various disposition paths for the beryllium shipments as well as other waste shipments that fall within the GTCC category. Modeling with MCNP 6.2 was conducted to determine whether the beryllium blocks could be safely stored and transported within a 72-B cask; the standard shipping container for remote-handled, transuranic waste, while also meeting regulatory limits at various disposition paths. A cost analysis of the transportation and long-term disposal paths mentioned in the EIS was also carried out using available data and information from similar waste shipments. Lastly, a geochemical analysis of the various geological repository discussed in the EIS was also carried out using the Geochemists Workbench Release 14. Long-term disposal of the beryllium blocks at the Waste Isolation Pilot Plant (WIPP) proved to be the most cost-effective long-term disposition option of the ones considered in the EIS. A dose rate calculation at both contact and remote-handling distances indicate that the analyzed beryllium shipments should not exceed the exposure limits at any of the considered locations. Greater-than-class-C waste has been left in a regulatory state of limbo for years, resulting in backlogs of inventory across several research sites in the United States. Due to its high activity and presence of transuranic isotopes, it is imperative to ensure that it remains inaccessible and sequestered both in its short-term interim as well as a long-term geologic time scale. The information and assessments done in the paper could potentially serve as a reference for any future shipments of this waste form at the WIPP facility as well as other disposition paths that may be considered in the future.

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

Energy Supply Chain Strategic Plan

With energy demand rising and the desire to reduce carbon emissions becoming more prominent, government and industry are increasingly interested in developing an emissions-free, power generation source that is safe, resilient and dependable. The world is looking to expand clean energy capabilities and develop the next generation of nuclear reactors. Not only is building next-generation energy systems Idaho National Laboratory’s opportunity, it is also Idaho’s opportunity. With this in mind, INL launched Supply Chain Initiative to leverage INL’s energy projects to build an energy talent pipeline and develop a robust business ecosystem to support current and future projects.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2021

UNREVIEWED DISPOSAL QUESTION EVALUATION: Disposal of the Idaho National Laboratory Routinely Generated Remote Handled Low-Level Waste at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

This Unreviewed Disposal Question Evaluation assesses whether the Idaho National Laboratory Routinely Generated Remote Handled Low-Level Waste (NEID09MFCRLLW, Revision 7) is suitable for shallow land burial at the Area 5 Radioactive Waste Management Site on the Nevada National Security Site. Disposal of the INL Routinely Generated Remote Handled LLW meets all performance objectives of DOE Manual DOE M 435.1-1, Radioactive Waste Management Manual, Chapter IV, Section P. The INL Routinely Generated Remote Handled LLW is recommended for acceptance without conditions.

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

OVERVIEW OF RECENT PU-238 PRODUCTION ACTIVITIES AT IDAHO NATIONAL LABORATORY

The Plutonium Fuel Services (PFS) program at Idaho National Laboratory (INL) is active in the qualification of irradiation targets containing Np-237 for irradiation in the Advanced Test Reactor (ATR) to produce Pu-238 for future NASA missions. INL qualified and loaded 7 targets in ATR’s South Flux Trap (SFT) for cycle 169A, which occurred in Spring 2021. This program was reinitiated after two baseline production targets in three positions validated significant production of Pu-238 [ref.1]. The validation model was followed by the PFS-1 experimental test in the ATRC (Critical) facility [ref.2]. This paper outlines the progress and status of the PFS program. The qualification effort, safety analysis, hardware status, and future activities for qualification of an updated target design for use in the ATR will be discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2021

AGR-5/6/7 Fuel Fabrication Report

The U.S. Department of Energy Office of Nuclear Energy (DOE NE) and the Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) Fuel Development and Qualification program (referred to as AGR Fuel program hereafter) are pursuing qualification of tristructural isotropic (TRISO) coated particle fuel for use in high temperature gas cooled reactors (HTGRs). The AGR Fuel program was established to provide a fuel qualification data set in support of the licensing and operation of an HTGR. BWX Technologies Nuclear Operations Group (BWXT-NOG) was subcontracted to fabricate the fuel for the AGR program. Several investments and innovations were realized in preparation to fabricate fuel for the AGR-5/6/7 irradiation experiments that brought fuel fabrication fully out of the laboratory and into engineering-scale operations. These included: • Increased the kernel fabrication line capacity and uniformity • Upgraded ancillary support equipment and processes for the tristructural isotropic (TRISO) coating furnace • Demonstrated efficient production of the matrix precursor powder by dry jet milling of co mingled components • Demonstrated an engineering-scale method for quick and efficient overcoating TRISO particles with the matrix precursor • Demonstrated an automated, multi cavity compacting system with a volumetric feed system • Demonstrated a combined-cycle thermal treatment furnace These changes increased production rates of some of these processes by an order of magnitude or more while eliminating the use of flammable solvents, multiple grinding and sorting operations, and the weighing out of individual die charges. Three fuel kernel lots were fabricated for production of the fuel for AGR-5/6/7. The initial lot (J52R-16-39316) was certified to fuel specifications but was not used because the kernels had a high fraction of internal fissures that caused an unacceptable fraction of the kernels to fragment when charged to the coating furnace where the TRISO coating would be deposited. Fragmented kernels increased the dispersed uranium in the particles and produced a worrisome fraction of dimpled particles with an elevated probability of in-pile failure. After some efforts to identify the cause of the fissure formation, two additional lots were produced with much lower fissure fractions, J52R-16-69317 and 69318. The latter kernel lot was a backup to the first and was not needed. Multiple kernel batches were composited to form each of the lots so as to simulate a commercial-scale operation where kernel batches would also be composited. Multiple TRISO coating runs were performed and the product characterized so that several could be composited into a TRISO lot. TRISO lot J52R-16-98005 conformed to all fuel specifications except the mean outer pyrocarbon (OPyC) layer thickness was thinner than specified. Furthermore, it was determined that the TRISO lot had a dispersed uranium fraction (DUF) that might result in the compacts not meeting the DUF specification. A review of the role of the OPyC layer and consequences of the DUF by the Technical Coordination Team and INL concluded that the fuel was acceptable for use in the AGR-5/6/7 irradiation experiment. The TRISO particles were overcoated with the matrix precursor that had been produced in a jet-milling operation. The overcoating was performed in equipment originally designed to coat pharmaceuticals. The overcoating process performed well; producing highly spherical and uniform overcoats requiring little upgrading and no recycle or rework. TRISO particles were overcoated with the matrix precursor to achieve nominal volumetric packing fractions (PFs) of TRISO particles of 25% and 40% for the irradiation experiments. The 40% PF compacts occupy the first and fifth test capsule in the test train while the inner three capsules are loaded with 25% PF compacts. The resinated graphite matrix precursor powder was a derivative of the German A3-27 matrix formulation, which differs from previous AGR irradiation campaigns that used an A3-3 formulation. Jet milling of the matrix powder precursor produced a finer mean graphite particle size than the milling operations used for the A3-3 matrix powder precursor. Changes made in the matrix formula and equipment yielded compacts with significantly higher matrix density than was attained in previous AGR irradiation campaigns. The changes in the matrix formulation and the means of milling the powders also complicated resolution of the three fuel compact defect fractions, DUF, exposed kernel fraction (EKF), and the silicon carbide defect fraction (SDF). Characterization data from BWXT-NOG had some anomalous results, so samples of the fuel compacts and overcoated TRISO particles were also analyzed by Oak Ridge National Laboratory (ORNL) to ensure that the defect fractions were accurately characterized.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Design, Manufacturing and Performance Considerations in Reactor System, Components and Materials

Idaho National Laboratory (INL) is the national and global leader of the Department of Energy (DOE) ’s nuclear energy research, development, and demonstration programs. Developing critical design criteria of new advanced reactor systems, components and materials requires an understanding of fabrication and the irradiation environment during normal operating and accident conditions. Next generation researchers and designers, are, therefore challenged not only by demands for ever improved performance but must also work to shorten the development and commercialization lifecycle for new nuclear reactors and systems to remain competitive. This provides unique and exciting opportunities for all contributors to this field of study. This presentation will offer a strategic overview of the competitively funded R&D awards by the Advanced Methods for Manufacturing (AMM) program of the DOE’s office of Nuclear Energy (NE). This program conducts research and development (R&D) to accelerate innovations by reducing the cost and schedule of constructing new nuclear plants and to make fabrication of nuclear power plant components faster, cheaper, and more reliable. By evaluating state-of-the-art practices found in other large manufacturing industries, the nuclear community has identified six major areas of innovation to reach the objectives of this DOE-NE program. Additionally, typical research case studies performed at INL, will be presented, showing the vast variety of opportunities for which, the next generation of researchers can seek solutions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Energy Arbitrage: Comparison of Options for use with LWR Nuclear Power Plants

Arbitrage is the opportunistic buying and selling of a commodity during local pricing valleys and peaks respectively to maximize economic value. This report evaluates options for energy arbitrage integrated with existing light water reactor (LWR) nuclear power plants (NPPs) where nuclear energy could be stored in a variety of forms and later recovered to generate electrical power during periods when grid electricity demand and pricing are high. The forms of energy storage examined in this report include the potential value of batteries, hydrogen, and thermal energy storage for coupling with nuclear power. Various large demand response options are also analyzed, including the production of liquid nitrogen via air separation and liquefaction, liquefaction of hydrogen, compressed hydrogen, and the cryogenic capture of CO2. Demand response refers to dispatchable loads that can cycle up or down depending on-grid electricity demand to aid in balancing the grid. Large demand response options could dispatch to aid nuclear power stations in avoiding power turndowns by providing an alternate disposition for electrical energy by producing marketable products (e.g., liquid nitrogen, hydrogen, or captured CO2). Static conditions were chosen and analyzed in this report for each option. Dynamic operation or optimization of energy arbitrage or demand response are out of scope for this report. The analysis is based on storage systems with discharge capacities of 500 MW for which various durations of storage and costs of charging (electricity cost) are examined. While the value of thermal energy to an industrial user for flexible plant operations has been previously proven as a business case, this report evaluates costs of hydrogen energy storage and leading thermal energy storage options, and large demand response loads that could be integrated with LWRs in comparison to utility-scale battery storage for use of off-peak nuclear energy. Compilation of this information will be used by the Idaho National Laboratory (INL) RAVEN/HERON systems integration and economics tool to evaluate thermal energy dispatch to industrial users. Relative ranking of energy storage options was done using a levelized cost of storage (LCOS) metric which calculates a rough breakeven cost for the system, taking into account the capital and operating costs as well as the revenue from arbitrage. Table ES1 below shows the LCOS for each of the energy storage options considered. First, in the table, lithium iron (Fe) phosphate batteries are listed as the base case for comparison against the other options. Next is hydrogen storage where most of the hydrogen analyses assumed the hydrogen to be produced using solid oxide electrolytic cell (SOEC) high temperature steam electrolysis (HTSE). The others used existing models of polymer electrolyte membrane (PEM) low temperature electrolysis to produce hydrogen. HTSE performance parameters and costs were taken from existing INL models. Various means were assumed to convert the hydrogen to electricity, including PEM fuel cells (FCs) and a gas turbine mixed in a 30 vol% mixture with natural gas. Physical storage (pressure vessels) and geological storage (natural underground features) were used to store the hydrogen as noted. Geological storage is more economical, but the locations are limited because of the requirement for pre-existing geological formations that will support storage. Thermal energy storage (TES) options were also analyzed including electro-thermal energy storage (ETES) and four different liquid sensible heat TES storage media as noted (Hitec, Hitec XL, Therminol-66, and Dowtherm A). The ETES process considered was modified using existing public documentation on an Echogen process and uses a separate supercritical CO2 charge and discharge cycle with sand as the heat storage media.

25 ENERGY STORAGE↗
DOE OSTI2021

FY 2021 High Performance Computing Overall Systems Availability

INL HPC capabilities were utilized by individuals at universities, industry, and government laboratories for a wide range of research and engineering activities between October 1, 2020 and September 30, 2021. The HPC systems themselves experienced minimal downtime over the course of the fiscal year. Most downtime was planned to apply critical security patches to system software. The most serious of the unplanned downtime events occurred due to maintenance on the Kohler backup generator in July. Users were kept notified of all events, both planned and unplanned, via 36 total email notifications in FY-21. The INL HPC systems continue to provide high availability for collaboration and innovation in nuclear energy systems research.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2022

Initial Observations from AGR 5/6/7 Capsule 1

The fourth and final irradiation experiment in the Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) fuel development and qualification program is designated as AGR-5/6/7. Data collected from the fabrication, irradiation, and post-irradiation examination (PIE) of this tristructural isotropic (TRISO) fuel are intended to serve as the primary data set for the qualification of this fuel for use in high-temperature gas-cooled reactors (INL 2021, Collin 2018b). However, data collected from the three preceding irradiations (i.e., AGR-1, AGR-2, and AGR-3/4) may also be used to supplement data collected from AGR-5/6/7. All components of the AGR-5/6/7 fuel (i.e., UCO kernels, TRISO coatings, and fuel compacts) were produced on an engineering scale at BWXT (Lynchburg, Virginia USA) according to the fuel specification (Marshall 2017). This fuel was irradiated in the northeast flux trap (NEFT) at the Advanced Test Reactor (ATR) at Idaho national Laboratory (INL) from February 16, 2018 to July 22, 2020 (Pham et al. 2021). Measurements in the fission product monitoring system (FPMS) indicated unexpected and significant numbers of failures of TRISO particles in Capsule 1 near the end of the sixth irradiation cycle (ATR Cycle 166A). In the fourth cycle (ATR Cycle 164B) and beyond, the sweep gas flow became very low (presumably from degradation of the capsule gas outlet line via an unidentified mechanism), and the program deliberately isolated Capsule 1 from gas flow periodically. In later cycles, attempts to reestablish any kind of flow in Capsule 1 were unsuccessful. With little or no flow through Capsule 1, FPMS measurements and enumerations of failed particles in Capsule 1 were difficult or impossible as was the ability to control the helium/neon gas mixture used for temperature control. Gas flows and fission gas activity in the effluent gas from the other AGR-5/6/7 capsules were also impacted by the Capsule 1 gas flow issues and the large increase in fission gas released from the Capsule 1.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Case Study - Intentional Inclusion: How embedding inclusion into the fabric of Idaho National Laboratory positioned the organization for a sustainable future

Idaho National Laboratory, a Department of Energy national laboratory headquartered in Idaho Falls, Idaho, set out to embed inclusion into every aspect of its organization. The goal was to ensure the laboratory is a place where everyone feels they are valued, belong and can bring their best selves to work each day. The laboratory viewed inclusion as a journey in which a consistent, collective commitment to intentional decision-making and incremental progress ultimately achieves amazing results. INL’s intentional inclusion strategy has allowed the organization to: • Retain and build a next-generation workforce to transform the world’s energy future and secure the nation’s most critical infrastructure. • Lead inclusively through individual, team-based, and labwide actions that elevate organizational effectiveness. • Grow laboratory cultural acumen by establishing a clear connection between inclusion and mission success. This case study shares how INL brought employees along on the inclusion journey and inspired everyone to keep moving forward together.

99 GENERAL AND MISCELLANEOUS↗