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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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All sourcesNASA NTRSDOE OSTI
NASA NTRS2018

Kilopower Reactor Development and Testing

The Kilopower Reactor Using Stirling Technology (KRUSTY) was the first nuclear-powered operation of a truly new fission reactor concept in the US in over 40 years. It provided valuable experience and data, demonstrated the passive reactor operation of the Kilopower reactor class, showed that developing a small reactor is not inherently expensive, and demonstrated a space reactor concept that can be used for near-term space science and exploration. KRUSTY/Kilopower is the first step towards truly astounding space fission capabilities

Gibson, Marc↗
DOE OSTI2022

INITIAL EXPLORATION OF A NOVEL TRANSIENT ARREST SYSTEM INVOLVING FUEL HEATING

A preliminary analysis on a novel accident response system to diminish the severity of supercritical transients was conducted. The novel accident response system, called the instant shock arrest system, involves using electricity to heat the nuclear fuel at the onset of a large accidental reactivity insertion. This system is specifically designed for reactors with metallic fuel, such that the fuel is capable of conducting electricity, and being resistively heated. A reactor dynamics model of the advanced test reactor was created using the point kinetics equations and a linear reactivity feedback model to simulate how the system would effect the maximum fuel temperatures experienced during the transient. Transients with the instant shock arrest system were compare to those without it. It was found that the instant shock arrest system initially heated the fuel more than the unaffected transient but the negative reactivity inserted from such heating was enough to lower the maximum fuel temperature experienced during the transient. After simulating six different accident scenarios with reactivity insertions ranging from 0.5 \$ to 1.3 \$, it was found that the an optimal system response could reduce peak fuel temperatures during the transient by 3.5\% to 5\%. Furthermore, discussion was given on how the optimal system response could be obtained using relatively simple numerical optimization algorithms due to the smoothness of the optimization problem.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

INITIAL EXPLORATION OF A NOVEL TRANSIENT ARREST SYSTEM INVOLVING FUEL HEATING (Presentation)

A preliminary analysis on a novel accident response system to diminish the severity of supercritical transients was conducted. The novel accident response system, called the instant shock arrest system, involves using electricity to heat the nuclear fuel at the onset of a large accidental reactivity insertion. This system is specifically designed for reactors with metallic fuel, such that the fuel is capable of conducting electricity, and being resistively heated. A reactor dynamics model of the advanced test reactor was created using the point kinetics equations and a linear reactivity feedback model to simulate how the system would effect the maximum fuel temperatures experienced during the transient. Transients with the instant shock arrest system were compare to those without it. It was found that the instant shock arrest system initially heated the fuel more than the unaffected transient but the negative reactivity inserted from such heating was enough to lower the maximum fuel temperature experienced during the transient. After simulating six different accident scenarios with reactivity insertions ranging from 0.5 \$ to 1.3 \$, it was found that the an optimal system response could reduce peak fuel temperatures during the transient by 3.5% to 5%. Furthermore, discussion was given on how the optimal system response could be obtained using relatively simple numerical optimization algorithms due to the smoothness of the optimization problem.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS2010

Fission Surface Power System Initial Concept Definition

Under the NASA Exploration Technology Development Program (ETDP) and in partnership with the Department of Energy (DOE), NASA has embarked on a project to develop Fission Surface Power (FSP) technology. The primary goals of the project are to 1) develop FSP concepts that meet expected surface power requirements at reasonable cost with added benefits over other options, 2) establish a hardwarebased technical foundation for FSP design concepts and reduce overall development risk, 3) reduce the cost uncertainties for FSP and establish greater credibility for flight system cost estimates, and 4) generate the key products to allow NASA decision-makers to consider FSP as a preferred option for flight development. The FSP project was initiated in 2006 as the Prometheus Program and the Jupiter Icy Moons Orbiter (JIMO) mission were phased-out. As a first step, NASA Headquarters commissioned the Affordable Fission Surface Power System Study to evaluate the potential for an affordable FSP development approach. With a cost-effective FSP strategy identified, the FSP team evaluated design options and selected a Preliminary Reference Concept to guide technology development. Since then, the FSP Preliminary Reference Concept has served as a point-of-departure for several NASA mission architecture studies examining the use of nuclear power and has provided the foundation for a series of "Pathfinder" hardware tests. The long-term technology goal is a Technology Demonstration Unit (TDU) integrated system test using full-scale components and a non-nuclear reactor simulator. The FSP team consists of Glenn Research Center (GRC), Marshall Space Flight Center (MSFC) and the DOE National Laboratories at Los Alamos (LANL), Idaho (INL), Oak Ridge (ORNL), and Sandia (SNL). The project is organized into two main elements: Concept Definition and Risk Reduction. Under Concept Definition, the team performs trade studies, develops analytical tools, and formulates system concepts. Under Risk Reduction the team develops hardware prototypes and conducts laboratory-based testing.

Source record↗
DOE OSTI2020

NRIC EBR-II Test Bed Pre-Conceptual Design Report

Authorized by the Nuclear Energy Innovation Capabilities Act (NEICA), the National Reactor Innovation Center (NRIC) provides private sector technology developers access to strategic infrastructures and assets for commercial nuclear energy research, development, demonstration, and deployment activities. The mission is to support a timely and cost-effective path to the licensing and commercialization of new nuclear energy systems. To meet these needs, NRIC is developing two test beds at Idaho National Laboratory (INL). The ZPPR Test bed (ZTB) and the EBR-II Test bed (ETB). The EBR-II test bed will support the demonstration of systems that operate at less than 10 MWt. The baseline objective is for the EBR-II Dome to act as a safety significant containment structure capable of siting reactors that utilize Safeguards Category 4 material for operations. The major areas addressed in the pre-conceptual design include: • Installation of an access door • Electrical Power • Heat Removal • Ventilation in the Dome • Module handling system Along with the design for ETB a concept of operations (COP) has also been developed. The COP is intended to facilitate a common understanding of ideas, challenges, and issues. As systems continue to evolve in complexity System Engineers and Project Directors will utilize the COP to develop and sustain a common vision of the system for stakeholders.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

AZTRAN and AZKIND verification with a BWR study case

The AZTLAN Platform project is a national initiative led by the National Institute of Nuclear Research which gather the main public institutions of higher education in Mexico: the National Polytechnic Institute, the National Autonomous University of Mexico and the Autonomous Metropolitan University. It consists in the development of a modeling platform for the analysis and design of nuclear reactors. This project aims to modernize, improve and integrate the neutronic, thermohydraulics and thermomechanical codes, developed in the national institutions, in a platform integrated and developed by Mexican experts. The development of AZTLAN Platform is divided technically into four working groups: neutronics, thermohydraulics, coupling and uncertainty analysis, and users group. The present work is the result of the activities of the Users Group focused on the follow-up, testing and use of the different codes (modules) being developed under the AZTLAN project; with the objective of providing feedback and to suggest development priorities to the team of developers, as well as to promote collaboration and synergy between the participating institutions. In particular, in this work, the AZTRAN and the AZKIND modules were verified and tested against the results of well-known codes of the nuclear industry for boiling water reactor (BWR) models. AZTRAN is a 3D transport code which solves numerically the multi-group time independent neutron transport equation in Cartesian geometry, based on the discrete ordinates (SN) method and the RTN-0 nodal scheme. AZKIND is a 3D diffusion code that solves numerically the time dependent neutron diffusion equations in Cartesian geometry, based on a Galerkin Finite Element Method using the RTN-0 polynomial nodal approximation.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

ADVANCED TEST REACTOR (ATR) CONVERSION TO LOW ENRICHED URANIUM (LEU) FUEL

The United States Department of Energy (DOE), National Nuclear Security Administration (NNSA), Office of Material Management and Minimization (MMM) is working to convert research reactors globally from highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. MMM and its predecessor programs have converted or verified the shutdown of 88 HEU research reactors and isotope production facilities, using LEU fuel developed in the 1980s. However, there is a small set of high performance research reactors which require a new high density LEU fuel for conversion, including six U.S. high performance research reactors (USHPRRs), which include the Advanced Test Reactor (ATR) and the ATR Critical (ATRC) Facility at the Idaho National Laboratory (INL) near Idaho Falls, ID; University of Missouri Research Reactor (MURR) at Columbia, MO; Massachusetts Institute of Technology Reactor (MITR) at MIT in Cambridge, MA; High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) near Oak Ridge, TN; and National Bureau of Standards Reactor (NBSR) at the National Institute of Standards and Technology (NIST) Center for Neutron Research (NCNR) in Gaithersburg, MA. These reactors will share the same uranium-molybdenum (U-Mo) fuel plate design. The U-Mo fuel meat foil will have a zirconium (Zr) diffusion barrier with aluminum cladding. Three different application methods for the Zr diffusion barrier layer are being pursued – co-rolled, electroplating and plasma spraying. Fabrication of the higher density LEU U-Mo fuel has presented several challenges. Because the U-Mo fuel fabrication process uses the existing stockpile of 93% highly enriched uranium (HEU) to be blended with molybdenum and depleted uranium (DU) to form 19.75% LEU, cost benefits of enriching to 19.75% will not be realized. Homogeneity of the U-235 and molybdenum in the metal foil is a challenge due to U-Mo casting process results in variations between adjacent cast U-Mo plates. LEU will not require less U-235 as previously believed due to self-shielding properties of U-238 requiring additional U-235 to be loaded in each fuel element. Although the overall dimensions of the ATR fuel element remain unchanged, the weight of the LEU ATR fuel element is 72% greater than a HEU ATR fuel element. During waste recovery and recycling efforts, separation of aluminum cladding from the HEU fuel meat is a relatively simple chemical dissolution, however, separation of the Zr diffusion barrier from the U-Mo fuel meat is significantly more challenging.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Narrowing transient testing pulse widths to enhance LWR RIA experiment design in the TREAT facility

The Transient Reactor Test (TREAT) Facility provided thousands of transient irradiations and plays a crucial role in nuclear-heated safety research. The facility’s flexible design and multi-mission nature saw historic experiments for numerous reactor fuels and transient types but was never specifically adapted to address very-brief pulse transients akin to postulated Light Water Reactor (LWR) Reactivity Initiated Accidents (RIA). Since the behaviors of fuel under these conditions depends strongly on energy input duration and resulting cladding time/temperature response under pellet cladding mechanical interactions, this three-year project was conceived to investigate new pulse-narrowing capabilities. Kinetic models showed incremental improvements for minor facility enhancements including increased reactivity step insertions (to initiate the power pulse) and slightly-increased transient rod drive speed for pulse termination (“clipping”). Replacing peak fuel assemblies, repositioning non-transient control rods to hold down the “hot side” of the core, and balancing against required excess reactivity needs, the limiting fuel assembly power can be reduced by ~20 %. Furthermore, this is a remarkable discovery of latent capability in TREAT, not only in enabling the subject capabilities for reduced pulse widths, but also significantly “uprating” the core’s transient energy capacity. Incremental improvements can likely enhance TREAT’s capability into BWR-relevant missions; briefer PWR pulse shapes can only be achieved with an advanced clipping system. Helium-3 ( 3 He) was found to offer the greatest benefits for clipping design and overall performance. A unique concept showed great promise for enabling a 3 He-based system with a reasonable cost; this design concept will likely become the focal point of future work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

DCIS integration tests for Lungmen nuclear power plant - 370

The factory acceptance tests (FAT) for distributed control and information systems (DCIS) of Lungmen nuclear power plant (LMNPP) Units 1 and 2 were separately performed by individual suppliers for their supplied scopes. The FAT also included overlap interface tests between the separately supplied scopes, but did not include an overall integration test. There were concerned by the regulator and the public over the lack of an integration test, which was believed to be indispensable to verify the correctness and completeness of a digital system of such large scale ever applied by Taiwan Power Company (Taipower). To address these concerns and enhance confidence in the DCIS, the DCIS integration tests, including reactor water level integration test, reactor pressure integration test, integration test for turbine generator and reactor protection system, integration test for proper plant response to safety and non-safety network equipment failures, and DCIS dynamic integration test, were performed at site during Unit 1| site tests. It is to be noted that these tests were beyond the post construction tests, the pre-operational tests, and the start-up tests, which are the normal industry practice and formally required by the regulator. A common goal for these integration tests is to verify the capability of the DCIS to transmit signal/data properly across various suppliers' equipment to actuate the final equipment. The methods and principle, the test purpose/test items/acceptance criteria and the test results for these tests are described. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

RELAP5-3D Input Model for the High Temperature Test Facility

A RELAP5-3D model of the High Temperature Test Facility at Oregon State University has been developed. The model includes the systems, components, and controls needed to perform code assessment calculations using data from the facility. Details of the model are provided. The input was developed for RELAP5-3D Version 4.4.2.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Sirius Irradiation Experiments and Post-Irradiation Examinations for Nuclear Thermal Propulsion

Nuclear Thermal Propulsion (NTP) systems hold promise in reducing transit times for exploration of the solar system by both crewed and uncrewed missions. NTP systems currently under investigation include a once-through high temperature gas-cooled fission reactor to provide thermal energy to heat the coolant which also serves as the propellant. The fuel systems of angular UN fuel particles dispersed in a matrix of W/Re, creating a ceramic and metallic composite or cermet, has been irradiated in Idaho National Laboratory’s (INL) Transient Reactor Test Facility (TREAT) enabling evaluation of these materials under representative nuclear heating rates (~95 K/s) and peak temperatures (~2527 K). These tests named Sirius-1 (UN-W/Re), have been irradiated and this paper will present post-irradiation examination results. The Sirius-1 test produced cracks in the fuel specimen and spalling of surface material. Uranium soot was found on the inner wall of the irradiation capsule indicating loss of some fissile material from the fuel specimen. Spalling from the surfaces was also noted upon visual inspection. Uranium diffusion from the fuel particles resulted in the formation of U/Re phases and edge features producing a laminar microstructure.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Gear Test Assembly – Experimental Testing and Gear Analysis (FY2021 Year End Report)

The Gear Test Assembly (GTA) is an experimental apparatus designed to test mechanical components used in fuel handling systems of liquid-sodium cooled fast-spectrum nuclear reactors. The performance and lifetime of gears, bearings, and dynamic seals are the primary focus of study. Three experimental campaigns have been completed since the start of operation in 2019, and the GTA is currently assembled and ready to begin the fourth experimental campaign. Sodium operations at the Mechanisms Engineering Test Loop (METL) have been paused while repairs are made to the Building 308 Alkali Metal Scrubber. This work is nearly complete, and once METL is brought back into operation the GTA will begin sodium testing immediately.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2012

Small Fast Spectrum Reactor Designs Suitable for Direct Nuclear Thermal Propulsion

Advancement of U.S. scientific, security, and economic interests through a robust space exploration program requires high performance propulsion systems to support a variety of robotic and crewed missions beyond low Earth orbit. Past studies, in particular those in support of the Space Exploration Initiative (SEI), have shown nuclear thermal propulsion systems provide superior performance for high mass high propulsive delta-V missions. The recent NASA Design Reference Architecture (DRA) 5.0 Study re-examined mission, payload, and transportation system requirements for a human Mars landing mission in the post-2030 timeframe. Nuclear thermal propulsion was again identified as the preferred in-space transportation system. A common nuclear thermal propulsion stage with three 25,000-lbf thrust engines was used for all primary mission maneuvers. Moderately lower thrust engines may also have important roles. In particular, lower thrust engine designs demonstrating the critical technologies that are directly extensible to other thrust levels are attractive from a ground testing perspective. An extensive nuclear thermal rocket technology development effort was conducted from 1955-1973 under the Rover/NERVA Program. Both graphite and refractory metal alloy fuel types were pursued. Reactors and engines employing graphite based fuels were designed, built and ground tested. A number of fast spectrum reactor and engine designs employing refractory metal alloy fuel types were proposed and designed, but none were built. The Small Nuclear Rocket Engine (SNRE) was the last engine design studied by the Los Alamos National Laboratory during the program. At the time, this engine was a state-of-the-art graphite based fuel design incorporating lessons learned from the very successful technology development program. The SNRE was a nominal 16,000-lbf thrust engine originally intended for unmanned applications with relatively short engine operations and the engine and stage design were constrained to fit within the payload volume of the then planned space shuttle. The SNRE core design utilized hexagonal fuel elements and hexagonal structural support elements. The total number of elements can be varied to achieve engine designs of higher or lower thrust levels. Some variation in the ratio of fuel elements to structural elements is also possible. Options for SNRE-based engine designs in the 25,000-lbf thrust range were described in a recent (2010) Joint Propulsion Conference paper. The reported designs met or exceeded the performance characteristics baselined in the DRA 5.0 Study. Lower thrust SNRE-based designs were also described in a recent (2011) Joint Propulsion Conference paper. Recent activities have included parallel evaluation and design efforts on fast spectrum engines employing refractory metal alloy fuels. These efforts include evaluation of both heritage designs from the Argonne National Laboratory (ANL) and General Electric Company GE-710 Programs as well as more recent designs. Results are presented for a number of not-yet optimized fast spectrum engine options.

Schnitzler, Bruce G.↗
DOE OSTI2017

Test based reliability assessment method for a safety critical software in reactor protection system - 284

Digitalization of the instrumentation and control systems in nuclear power plant entails some new features which do not exist in analog systems. When the new features are applied in safety critical systems, the risk stemming from them should be quantified properly to assure the reliability of the entire power plant. Among the new features, software is considered as the most important factor because it can cause common cause failure to many redundant systems. Therefore, in this work, a test based reliability assessment method for a safety-critical software is developed. An output of a software is determined by not only inputs but also the internal state of the software at the time. In the case of safety critical software developed for usage in nuclear power plant, it is possible to directly investigate probable internal states thanks to detailed design specification and programing features, and the probable inputs also can be identified based on the physical linearity of each process parameter and hardware characteristics related to data acquisition. In this work, a development method for practical exhaustive test case consisting of investigation of internal state and probable input sets to a specific internal state are suggested. In addition to this, software logic simulator for execution of software logic test and the reliability quantification method based on the test result are developed. The feasibility of the suggested method is demonstrated via a case study. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Laboratory Instrument Software Controlled Spread Spectrum Time Domain Reflectometry for Electrical Cable Testing

This research discusses development of a software-controlled laboratory instrument based spread spectrum time domain reflectometry system (SSTDR). This constitutes one task within PNNL’s Light Water Sustainability Program (LWRS) whose mission includes advancing nondestructive examination (NDE) techniques for off-line and on-line in-situ cable condition monitoring. In 2022, PNNL evaluated SSTDR for detection and characterization of a number of cable anomalies (Glass et al. 2022). The review included comparison of SSTDR to Frequency Domain Reflectometry (FDR) techniques which have enjoyed encouraging feedback and are starting to be used in nuclear power plants for periodic cable condition monitoring of cable systems as part of the plant’s overall cable aging management program. The FDR test introduces a broad-band chirp onto the cable at the cable end then listens for any reflection from a change of impedance along the cable caused by a damaged conductor or insulation, splices, contact with moisture, or other cable anomalies. The signal is captured in the frequency domain then transformed back to the time domain using an inverse Fourier transform (IFT). Based on the velocity of propagation, the impedance response signal is plotted against distance along the cable. Peak locations along the X-axis indicate the distance along the cable where a portion of the signal has been reflected back to the instrument as a result of a cable anomaly. The FDR test is considered the gold standard of reflectometry however it does require the cable to be de-energized to perform the test. The LIVEWIRE commercial SSTDR produces a similar plot to the FDR however all processing is in the time domain. A pseudo-random noise code (PN code) is input onto the cable conductor and the instrument listens for any reflected response from cable anomalies. The SSTDR processes the signal as an autocorrelation comparing the input PN code to any reflected signal detected. The autocorrelation analysis for thermal aging, water and water ingress detection, ground fault and phase-to-phase fault detection at various locations along the cable and with the cable attached and detached from a motor load, and on both energized and un-energized conditions were performed. These results were contrasted to Frequency Domain Reflectometry (FDR) measurements of the un-energized cable. Results were encouraging but indicated more work was warranted – particularly with the SSTDR, it seemed that the insulation damage would likely be better evaluated with multiple bandwidth cable tests particularly including larger bandwidths than were possible with the current commercial instrument. The commercial instrument’s bandwidth was set at 6, 12, 24, and 48MHz but note that SSTDR and FDR definitions of bandwidth trend similarly but are not the same. The FDR response could be more broadly adjusted, and the bandwidth of 100 to 500 MHz produced the best responses. FDR responses to anomalies were clearer than SSTDR responses and indications were that a broader bandwidth SSTDR may lead to improved SSTDR detection capability. This project used a laboratory instrument based SSTDR (primarily using an Arbitrary Waveform Generator (AWG) and a digital oscilloscope plus Python in-house software) that allowed software adjustment of the SSTDR bandwidth, window functions applied to the exciting Pseudo-random Noise (PN) code plus and other aspects of the SSTDR signal processing. Hereafter, this will be referred to as the PNNL SSTDR. Evaluating specific performance of the PNNL SSTDR is left to a separate report. This report documents hardware and software development to produce the SSTDR cable test system.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗