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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 253 records · Page 14
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Assessment of the BISON Metallic Fuel Performance Models

The US Department of Energy is leading a project to design and construct a fast spectrum test reactor called the Versatile Test Reactor (VTR). The BISON nuclear fuel performance code will be used to model VTR driver fuel, including looking at the effects of differences between the VTR driver fuel element design and the legacy fuel designs and experiments on which it is based. Simulations will be conducted to help determine whether the design’s behavior and performance are properly understood and to assess the margins to cladding failure and fuel melting relative to those predicted for past metallic fuel experiments. These predictions are expected to streamline VTR design and operation by helping inform the VTR driver fuel element design and by providing supplemental information for the fuel design safety basis.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Sensitivity Analysis of Irradiated Fueled Experiments using the MOOSE Framework [Slides]

Modeling and simulation (M&S) methods are able to predict uncertainties in experimental parameters (e.g., power and fission density) during irradiation. A shortfall exists in predicting how sensitive some of the parameters will behave during the experimental process. Sensitivity and Uncertainty Quantification (SUQ) is critical in support of qualification and licensing reactor fuels. The application of a method to quantify the uncertainty in these experiments is critical to the prediction of their performance. In this work, we propose the use of a polynomial chaos expansion (PCE) method to quantify the sensitive parameters in these simulations and, in an extension, their experimental surrogates. We propose to perform M&S using PCE uncertainty quantification on a previously irradiated fueled experiment in order to provide a validation case for Griffin and expand its use as a verification and validation (V&V) tool for experiments with a neutronics component. Griffin is an advanced, deterministic neutronics analysis code built using the MOOSE (multiphysics object-oriented simulation environment) framework which can provide state-of-the-art neutronic analysis on M&S of experiments. We will use the stochastic tools module (STM) in MOOSE to provide PCE uncertainty quantification on the proposed experimental setup. Idaho National Laboratory (INL) does not yet have an in-house developed code with V&V approval for experiments performed on-site; this work would provide a necessary addition of support for experiments performed at INL. The Nuclear Regulatory Commission (NRC) has explicitly requested uncertainties in calculated values such as fuel power and burnup, and the development of this capability would benefit the relationship between INL and the NRC.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1972

Spectral absorption coefficients of argon and silicon and spectral reflectivity of aluminum

A theoretical investigation was conducted to estimate the spectral properties of argon as a function of pressure, temperature, and wave number. The spectral characteristics of the argon buffer gas exert a strong influence on radiative energy transfer in the in-reactor test configuration of the nuclear light bulb engine. An existing computer program was modified and used to calculate the spectral absorption coefficients of argon at total pressures of 50, 100, 250, 500, 750 and 1000 atm in the temperature interval between 1000 and 30,000 K. At each pressure and temperature, spectral properties were calculated for forty-seven wave numbers in the interval between 1000 and 1,000,000 cm/1. Estimates of the spectral absorption coefficients of silicon were made as part of an evaluation of silicon vapor as a possible buffer-gas seeding agent for the reference nuclear light bulb engine. Existing cross-section data were used to calculate the spectral characteristics of silicon at twenty-four temperatures in the interval between 2000 and 10,000 K.

Krascella, N. L.↗
NASA NTRS1990

Programmatic status of NASA's CSTI high capacity power Stirling space power converter program

An overview is presented of the NASA Lewis Research Center Free-Piston Stirling Space Power Converter Technology Development Program. This work is being conducted under NASA's Civil Space Technology Initiative (CSTI). The goal of the CSTI High Capacity Power element is to develop the technology base needed to meet the long duration, high capacity power requirements for future NASA space initiatives. Efforts are focused upon increasing system thermal and electric energy conversion efficiency at least fivefold over current SP-100 technology, and on achieving systems that are compatible with space nuclear reactors. The status of test activities with the Space Power Research Engine (SPRE) is discussed. Design deficiencies are gradually being corrected and the power converter is now outputting 11.5 kWe at a temperature ratio of 2 (design output is 12.5 kWe). Detail designs were completed for the 1050 K Component Test Power Converter (CTPC). The success of these and future designs is dependent upon supporting research and technology efforts including heat pipes, gas bearings, superalloy joining technologies and high efficiency alternators. An update of progress in these technologies is provided.

Dudenhoefer, James E.↗
NASA NTRS1990

Programmatic status of NASA's CSTI High Capacity Power Stirling Space Power Converter Program

An overview is presented of the NASA Lewis Research Center Free-Piston Stirling Space Power Converter Technology Department Program. This work is being conducted under NASA's Civil Space Technology Initiative (CSTI). The goal of the CSTI High Capacity Power element is to develop the technology base needed to meet the long duration, high capacity power requirements for future NASA space initiatives. Efforts are focused upon increasing system thermal and electric energy conversion efficiency at least fivefold over current SP-100 technology, and on achieving systems that are compatible with space nuclear reactors. The status of test activities with the Space Power Research Engine (SPRE) is discussed. Design deficiencies are gradually being corrected and the power converter is now outputting 11.5 kWe at a temperature ratio of 2 (design output is 12.5 kWe). Detail designs were completed for the 1050 K Component Test Power Converter (CTPC). The success of these and future designs is dependent upon supporting research and technology efforts including heat pipes, gas bearings, superalloy joining technologies and high efficiency alternators. An update of progress in these technologies is provided.

Dudenhoefer, James E.↗
DOE OSTI2024

International fuel performance study of fresh fuel experiments for PCMI effects during RIA experiments

This paper presents the results of High-burnup Experiments for Reactivity-initiated Accident (HERA) Modeling & Simulation (M&S) exercise. The HERA project under the Nuclear Energy Agency (NEA) Second Framework for Irradiation Experiments (FIDES-II) program is focused on studying Light Water Reactor (LWR) fuel behavior during Reactivity-Initiated Accident (RIA) conditions. The Part I M&S cases are based on a series of tests in the Transient Reactor Test (TREAT) facility in the United States and the Nuclear Safety Research Reactor (NSRR) in Japan. The purpose of this work is to evaluate the test design to accomplish its goals in establishing clearer understanding of the effects of power pulse width during RIA conditions. Further, the blind predictions using various computational tools have been performed and compared amongst to interpret the behaviors of high burnup fuels during RIA. While many international participants evaluate the thermal–mechanical behavior of fuel rod under different conditions, a considerable scatter of outputs comes out for the cases due to the disparity between codes in predicting mechanical behaviors. In general, however, the results of thermal–mechanical analysis elaborate that nominal design conditions the shorter pulse width tests in NSRR should cause cladding failures while the TREAT tests appear to have more split prediction of failure or not. Furthermore, the sensitivity analysis varying key testing parameters reveals the considerable effect of power pulse width and total energy deposition on prediction of fuel rod failure.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Assessment of Near-Term Fuel Screening and Qualification Needs for Nuclear Thermal Propulsion Systems

Nuclear thermal propulsion (NTP) is an in-space propulsion technology capable of both high specific impulse (850–1000 s) and thrust (44–1112 kN), which can help reduce trip times for crewed missions beyond low Earth orbit. NTP technology has been demonstrated during historic programs. Over 20 ground test reactor experiments were performed, which demonstrated the prototypic reactor operations, during the Nuclear Engine for Rocket Vehicle Application (NERVA)/Rover program (1955–1972). Although historical programs have shown that NTP is a viable in-space propulsion technology, developing NTP in modern programs is contingent on the development and qualification of ultrahigh-temperature nuclear fuel technologies that can withstand engine operating conditions. In historical NTP development programs such as NERVA/Rover, prototypic reactor/engine schemes were ground tested to assess the overall system feasibility and to qualify the reactor fuel forms for eventual flight systems. Although this approach is effective to verify fuel performance under prototypic conditions, relying solely on full-scale NTP reactor tests as the pathway for verifying or qualifying fuel is inefficient and cost prohibitive today. Additionally, modern nuclear licensing requirements state that before test reactor approval, reactor components and fuel elements should be qualified via non-nuclear (out-of-pile) and nuclear (in-pile) testing under representative operating conditions. Using this methodology, fuel matures as production scale fabrication methods are established, and as produced fuel performance is demonstrated. Here, this paper provides an overview of historical approaches to NTP fuel performance maturation, including fuel screening and qualification needs, and provides insight for establishing an efficient testing paradigm that can be implemented to rapidly and affordably develop NTP fuel forms for eventual qualification.

42 ENGINEERING↗
DOE OSTI2022

Assembling Multiphysics Nuclear Reactor Simulations Using the MOOSE Framework

The Multiphysics Object Oriented Simulation Environment (MOOSE) [1] is an open-source, parallel finite element framework which provides the foundation for many advanced modeling and simulation tools developed under the Department of Energy (DOE) Nuclear Energy Advanced Modeling and Simulation (NEAMS) Program [2] for the analysis of advanced reactors. The MOOSE framework provides the common foundational capability on which many NEAMS codes for reactor analysis are built. The MOOSE framework also includes several systems to assemble unique workflows and couplingamong MOOSE-based applications. In particular, the MultiApp and Transfer Systems are widely used to assemble different MOOSE-based or MOOSE-wrapped physics applications together to perform loosely or tightly coupled multiphysics simulations. The National Reactor Innovation Center (NRIC) Virtual Test Bed (VTB) [3] hosts publicly available nuclear reactor multiphysics simulation examples which leverage MOOSE’s MultiApp System to meet the modeling needs of different reactor types. The flexibility and robustness of coupling provided by MOOSE permits rapid development of coupled physics models for a wide range of reactor types and events

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

MIT reactor (MITR) driven subcritical facility for the fluoride-salt-cooled high-temperature reactor (FHR) - 18623

The Fluoride-salt-cooled High temperature Reactor (FHR) uses clean salt coolants, carbon-matrix coated-particle fuel similar to that used in High-temperature Gas-cooled Reactors (HTGRs) and can be coupled to a Nuclear Air-Brayton Combined Cycle (NACC). Recent assessments indicate favorable economics and safety characteristics, but no FHR has been built. The question is what experimental facilities should be constructed to reduce technical uncertainties before a decision to build a test/demonstration reactor? The MIT Reactor (MITR) design and license would allow the construction and operation of a subcritical facility with 700 deg. C salt circulating through multiple full-width partial-height fuel assemblies operating with a power density up to 30% of a commercial FHR. This option would allow hot systems testing as a major step toward building the test/demonstration reactor. Preliminary system design, power control options, testing capabilities, and key nuclear characteristics of such a reactor-driven subcritical facility are described. A method of deriving subcritical multiplicity using surface source has been proposed and verified in this study. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

SCALE Code System 6.2.1

The SCALE Code System is a widely-used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor and lattice physics, radiation shielding, spent fuel and radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules including three deterministic and three Monte Carlo radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

SCALE Code System 6.2.2

The SCALE Code System is a widely used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor physics, radiation shielding, radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules including 3 deterministic and 3 Monte Carlo radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results. SCALE 6.2 represents one of the most comprehensive revisions in the history of SCALE, providing several new capabilities and significant improvements in many existing features.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

SCALE Code System

The SCALE Code System is a widely-used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor and lattice physics, radiation shielding, spent fuel and radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules including three deterministic and three Monte Carlo radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2020

SCALE Code System

The SCALE Code System is a widely used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor and lattice physics, radiation shielding, spent fuel and radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules, including three deterministic and three Monte Carlo radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Enhancing Post-Irradiation Examination (PIE) Efficiency with High-Throughput Experiments

Nuclear materials testing is essential for understanding changes in reactor environments, but traditional methods are costly and time-consuming. High-Throughput Experimentation (HTE) enables rapid screening and optimization of material properties by significantly increasing the number of specimens in the experiment. In this work, HTE methods were employed to irradiate novel materials at the Advanced Test Reactor (ATR) at Idaho National Laboratory. A total of 367 sub-sized tensile specimens were irradiated using the standard capsule design. To facilitate HTE, automated systems such as the Automated Mechanical Testing System were implemented for mechanical testing within hot cell environments, thereby reducing the need for manual interaction with hot cell manipulators and significantly lowering both the cost and duration of experiments. Digital image correlation (DIC) was used to measure displacement during mechanical testing, providing precise, non-contact measurement capabilities. The methodologies developed in this research will support the advancement of HTE techniques for future nuclear materials testing experiments at ATR. This approach allows for more efficient and cost-effective evaluation of new materials, ultimately leading to improved reactor performance, enhanced safety, and extended longevity within the nuclear industry.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

KRUSTY Reactor Design

The Kilowatt Reactor Using Stirling TechnologY (KRUSTY) was a reactor design, development, and test program to demonstrate the nuclear operation of a Kilopower reactor. Kilopower systems are intended to provide between 1 and 10 kW(electric) in space, or on the surface of planets or moons, with a clear evolution to substantially higher power systems. KRUSTY was a prototype of a 1-kW(electric) highly enriched uranium–fueled Kilopower system. In March of 2018, KRUSTY successfully operated as a fission power system and was the first nuclear-powered operation of any truly new reactor concept in the United States in over 40 years. This paper discusses the design of the KRUSTY reactor along with the philosophy, goals, and engineering work that ultimately led to KRUSTY’s success.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS2020

KRUSTY Reactor Design

The Kilowatt Reactor Using Stirling TechnologY (KRUSTY) was a reactor design, development, and test program to demonstrate the nuclear operation of a Kilopower reactor. Kilopower systems are intended to provide between 1 and 10 kW(electric) in space, or on the surface of planets or moons, with a clear evolution to substantially higher power systems. KRUSTY was a prototype of a 1-kW(electric) highly enriched uranium–fueled Kilopower system. In March of 2018, KRUSTY successfully operated as a fission power system and was the first nuclear-powered operation of any truly new reactor concept in the United States in over 40 years. This paper discusses the design of the KRUSTY reactor along with the philosophy, goals, and engineering work that ultimately led to KRUSTY’s success.

Kilopower↗