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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 1,081 records · Page 60
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Neutronic Tests of Fluoride Salt Based MSR/FHR Coolants

Research on molten salt reactor (MSR) technology based on fluoride salts as a convenient way of higher actinides incineration and/or utilization of thorium fuel has been performed at UJV Rez and later Research Centre Rez in Czech Republic since 1999. Necessary steps towards to the deployment and possible commercialization were examined in past investigations. Neutronic testing is an inseparable part of the development and was at first carried out in a series of experiments with LiF-NaF salt (60 mol% LiF, 40 mol% NaF) subjected to a neutron field shaped by different moderators within the framework of the EROS experiments. Later, neutron spectrum was measured in this salt and significant variations in calculation to experiment C/E comparison were discovered. Further tests were performed with Flibe (66 mol% LiF, 33 mol% BeF{sub 2}) reactor coolant salt containing enriched lithium (99.994 mol% Li-7), which was provided by Oak Ridge National Laboratory (ORNL). This Flibe salt originally comes from the Molten Salt Reactor Experiment (MSRE) and was supplied to the Czech Republic within the framework of a collaborative agreement on MSR and fluoride-salt-cooled high-temperature reactor (FHR) research and development (R and D) concluded between the Czech Ministry of Industry and Trade and US Department of Energy in 2012. All experiments in this work were carried out at the LR-0 reactor at Research Centre Rez. The LR-0 is a pool-type zero power light water reactor with large and flexible experimental capabilities due to its design and fuel construction.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

DESIGN CHARACTERISTICS OF THE IDAHO NATIONAL LABORATORY HIGH-TEMPERATURE GAS-COOLED TEST REACTOR

Uncertainty and sensitivity analysis is an indispensable element of any substantial attempt in reactor simulation validation. The quantification of uncertainties in nuclear engineering has grown more important and the IAEA Coordinated Research Program (CRP) on High-Temperature Gas Cooled Reactor (HTGR) initiated in 2012 aims to investigate the various uncertainty quantification methodologies for this type of reactors. The first phase of the CRP is dedicated to the estimation of cell and lattice model uncertainties due to the neutron cross sections co-variances. Phase II is oriented towards the investigation of propagated uncertainties from the lattice to the coupled neutronics/thermal hydraulics core calculations. Nominal results for the prismatic single block (Ex.I-2a) and super cell models (Ex.I-2c) have been obtained using the SCALE 6.1.3 two-dimensional lattice code NEWT coupled to the TRITON sequence for cross section generation. In this work, the TRITON/NEWT-flux-weighted cross sections obtained for Ex.I-2a and various models of Ex.I-2c is utilized to perform a sensitivity analysis of the MHTGR-350 core power densities and eigenvalues. The core solutions are obtained with the INL coupled code PHISICS/RELAP5-3D, utilizing a fixed-temperature feedback for Ex. II-1a.. It is observed that the core power density does not vary significantly in shape, but the magnitude of these variations increases as the moderator-to-fuel ratio increases in the super cell lattice models.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

High-Pressure, High-Temperature Thermal Hydraulic Test Facility for Nuclear-Renewable Hybrid Energy System Studies; Facility Design Description and Status Report

A high-pressure, high-temperature water flow loop has been designed at Idaho National Laboratory (INL) for deployment in the INL Dynamic Energy Transport and Integration Laboratory (DETAIL) within the Energy Systems Laboratory (ESL) D100 northwest high bay. This loop is the first of three thermally coupled flow loops that will comprise the Advanced Reactor Technology Integral System Test (ARTIST) facility. The ARTIST facility will ultimately include a high temperature helium loop and a liquid salt loop, in addition to the PWR loop. The high-pressure water loop will operate at Pressurized Water Reactor (PWR) conditions. It will support experimental research on nuclear-renewable hybrid energy systems as well as advanced reactor technology topics. Within DETAIL, the water flow loop will also serve as a thermal energy source, emulating a reactor system. The loop will be thermally integrated with co-located energy systems including a thermal energy transport loop, thermal energy storage system, and a 25 kW high-temperature electrolysis system for hydrogen production. DETAIL will include additional energy systems such as an integrated microgrid and renewable energy systems including solar photovoltaics. The electrically heated flow loop will be dynamically controlled to simulate nuclear fuel behavior under normal and off-normal operating conditions. DETAIL will be designed for characterization of complex system dynamics. The various components will be interfaced with Digital Real-Time Simulators for supervisory control and dynamic simulation of complete systems. The design and assembly of the high-pressure water flow loop and supporting systems has been subdivided into two Phases. Phase 1, which will be completed by the end of FY17, includes installation of cable trays and supports, electrical conduit, DI water system, drain lift station and the concrete pad for the chiller located outside of the laboratory. Phase 2 includes the detailed design and final assembly of the flow loop, supporting structures and personnel platforms, the chiller and associated piping. As of September, 2017, the final detailed design of the flow loop is complete. Funding for Phase 2 procurement, assembly, system operability checkout, shakedown testing and initial operation of the flow loop is requested for FY18.

42 ENGINEERING↗
DOE OSTI2025

Assessment and Planning of HFIR Test Articles

Burnable absorbers (BAs) are introduced into nuclear fuels to aid in controlling the activity of the irradiated fuel at the beginning of life in a light water reactor (LWR). This improves the efficiency of the reactor since it allows for higher enrichments (> 5 % 235 U) to be used in a LWR, which increases the burnup limit and therefore the cycle length. The purpose of the BA is to absorb neutrons to prevent power peaking, and the concentration of BA is consumed under irradiation. In order to understand the effect of the BAs on the microstructure and thermophysical properties of the fresh and irradiated fuel, neutron irradiation tests will be carried out at the High Flux Isotope Reactor (HFIR). These tests will include Gd 2 O 3 doped UO 2 fuels with concentrations that vary from 4 to 10 weight % Gd 2 O 3 .

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

NEA HTTR LOFC Project Test#3 Benchmark Results

In the second half of FY23, the High Temperature Engineering Test Reactor (HTTR) Loss Of Forced Cooling (LOFC)#3 data for the 9 MW test case with Vessel Cooling System (VCS) off were made available through the Nuclear Energy Agency (NEA) LOFC project framework; the neutronic model developed for the initial test (LOFC#1) achieved a satisfactory level of maturity, demonstrating its accuracy in predicting power evolution and core re-criticality, but LOFC#3 should be used primarily to investigate thermal hydraulic phenomena, as the reactor was shut down prematurely due to overheating in the upper reactor components, which prevented re-criticality; this report focuses on advancing the HTTR thermal hydraulic model to accurately simulate the LOFC#3 scenario, including simulating the LOFC#3 benchmark and generating the corresponding benchmark specifications, aiming to ensure consistency across participant models and provide essential data for future participants, including private industry stakeholders seeking to validate their computational tools.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

PELICAN Status Report for Year 2021

The Pressure Drop Experimental Loop for Investigations of Core Assemblies in Advanced Nuclear Reactors, PELICAN, is a novel experimental facility designed and constructed to recreate the prototypic full scale hydraulic flow conditions for the Versatile Test Reactor (VTR) currently under development by the U.S. Department of Energy. To support VTR design efforts, PELICAN is used to replicate VTR’s primary heat transport system (PHTS) in the core assembly region. The measurement objectives for data generated from this test facility are to mimic the prototypic VTR assembly and to meet the validation needs for code calculations and simulations of the reference design for the VTR core. These objectives focus on the need to validate pressure drop results across the various segments of the fuel assembly, including the axial reflectors, fuel, and transition regions, as they relate directly to the pumping power and safety behavior of the reactor. In the following sections, the major milestones for the experimental work performed on PELICAN over the calendar year (CY) 2021 are presented. In the next section, we discuss the assembly and shakedown testing of the facility. In Section 3, an analysis of the errors in the major measurements is performed. In Section 4 and 5 we present the results from the first set of test articles placed inside the loop. The design of the remaining test articles as part of the full assembly are shown in Section 6 and the conclusion and path forward for the following year, CY22 is given in the final section.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

New Designed Containers for Activated ILW from NPP Decommissioning - 20208

The paper focuses on packaging of dismantled and fragmented reactor components into containers and their transport to their storage and/or disposal facility within the immediate dismantling strategy applied on decommissioning of nuclear power plant (NPP) finally shutdown after standard operation. The main purpose of the paper is to present the process of design, manufacture, testing and use of new transport and storage containers for activated intermediate level waste (ILW) from V1 NPP decommissioning at Jaslovske Bohunice site. Legislation conditions, container selection process, requirements on their design and manufacture, scope of associated licensing documentation and technical procedures applied up to their use on site are described in detail. (authors)

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

Materials and Fuels Complex (FY22 – FY26 Five-Year Investment Strategy)

The Department of Energy Office of Nuclear Energy (DOE-NE) vision is to “Advance nuclear energy science and technology to meet U.S. energy, environmental, and economic needs.” The Materials and Fuels Complex (MFC) serves as the foundation of a nuclear RD&D enabling test bed at Idaho National Laboratory (INL) and is an integral part of a National Reactor Innovation Center (NRIC) strategy. MFC facilities focus on developing and maintaining RD&D capabilities that can increase research throughput, reduce barriers to deployment, and facilitate commercialization of new ideas and technologies for clean and secure sources of energy.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2016

A comparison of the irradiation creep behavior of several graphites - 18583

Graphite creep strain data from the irradiation creep capsule Advanced Graphite Creep-1 (AGC-1) are reported. This capsule was the first (prototype) of a series of five or six capsules planned as part of the AGC experiment, which was designed to fully characterize the effects of neutron irradiation and the radiation creep behavior of current nuclear graphite. The creep strain data and analysis are reported for the six graphite grades incorporated in the capsule. The AGC-1 capsule was irradiated in the Advanced Test Reactor at Idaho National Laboratory (INL) at approximately 700 deg. C and to a peak dose of 7 dpa (displacements per atom). The specimen s final dose, temperature, and stress conditions have been reported by INL and were used during this analysis. The derived creep coefficients (K) were calculated for each grade and were found to compare well to literature data for the creep coefficient, even under the wide range of AGC-1 specimen temperatures. Comparisons were made between AGC-1 data and historical grade data for creep coefficients. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Development of programmable logic in a rapid, automated, and repeatable fashion for safety-related applications - 385

Digital safety platform considerations for power plant modernization or new build efforts involve a burdensome task of implementing a large number of software-based functions onto a designated platform to match the plant's architecture. This problem is not unique to any particular safety-related platform solution, but is prevalent across the landscape of current Qualified and Licensed Safety-related Instrumentation and Control (I and C) Solutions. As an example when implementing a RG 1.97 compliant Post Accident Monitoring System (PAMS) solution with the Lockheed Martin Nuclear Protection and Control (NuPAC) safety platform, 72 unique software loads are needed to configure the Qualified Display System (QDS). Large scale development and customization efforts to upgrade existing plant functions or systems require the licensee to assume the cost and technical risk necessary to undergo a digital based upgrade. Often the cost, schedule, and technical risk outweighed the benefit of modernization efforts due to the unknown hurdles and past projects that overran cost and schedule projections. Lockheed Martin's Engineering Production Suite (EPS) minimizes the licensee risk and expense associated with modernization efforts through a developer's toolkit for rapid, automated, and repeatable development of Programmable Logic used to implement plant functions and architectures for use on the Lockheed Martin NuPAC platform. This paper describes a use case where by EPS can create system level representations, simulate system functions, partition logic designs to accommodate hardware constraints through logic partitioning, automate the assembly of Application Specific Programmable Logic (ASPL), generate system test cases, program the NuPAC hardware, and perform testing throughout the design workflow. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Flavor-dependent radiative corrections in coherent elastic neutrino-nucleus scattering

We calculate coherent elastic neutrino-nucleus scattering cross sections on spin-0 nuclei (e.g. 40 Ar and 28 Si) at energies below 100 MeV within the Standard Model and account for all effects of permille size. We provide a complete error budget including uncertainties at nuclear, nucleon, hadronic, and quark levels separately as well as perturbative error. Our calculation starts from the four-fermion effective field theory to explicitly separate heavy-particle mediated corrections (which are absorbed by Wilson coefficients) from light-particle contributions. Electrons and muons running in loops introduce a non- trivial dependence on the momentum transfer due to their relatively light masses. These same loops, and those mediated by tau leptons, break the flavor universality because of mass-dependent electromagnetic radiative corrections. Nuclear physics uncertainties significantly cancel in flavor asymmetries resulting in subpercent relative errors. We find that for low neutrino energies, the cross section can be predicted with a relative precision that is competitive with neutrino-electron scattering. We highlight potentially useful applications of such a precise cross section prediction ranging from precision tests of the Standard Model, to searches for new physics and to the monitoring of nuclear reactors.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2023

Investigation of thermal hydraulic behavior of the High Temperature Test Facility's lower plenum via large eddy simulation

A high-fidelity computational fluid dynamics (CFD) analysis was performed using the Large Eddy Simulation (LES) model for the lower plenum of the High–Temperature Test Facility (HTTF), a ¼ scale test facility of the modular high temperature gas-cooled reactor (MHTGR) managed by Oregon State University. In most next–generation nuclear reactors, thermal stress due to thermal striping is one of the risks to be curiously considered. This is also true for HTGRs, especially since the exhaust helium gas temperature is high. In order to evaluate these risks and performance, organizations in the United States led by the OECD NEA are conducting a thermal hydraulic code benchmark for HTGR, and the test facility used for this benchmark is HTTF. HTTF can perform experiments in both normal and accident situations and provide high-quality experimental data. However, it is difficult to provide sufficient data for benchmarking through experiments, and there is a problem with the reliability of CFD analysis results based on Reynolds–averaged Navier–Stokes to analyze thermal hydraulic behavior without verification. To solve this problem, high-fidelity 3-D CFD analysis was performed using the LES model for HTTF. It was also verified that the LES model can properly simulate this jet mixing phenomenon via a unit cell test that provides experimental information. As a result of CFD analysis, the lower the dependency of the sub-grid scale model, the closer to the actual analysis result. In the case of unit cell test CFD analysis and HTTF CFD analysis, the volume-averaged sub-grid scale model dependency was calculated to be 13.0% and 9.16%, respectively. As a result of HTTF analysis, quantitative data of the fluid inside the HTTF lower plenum was provided in this paper. As a result of qualitative analysis, the temperature was highest at the center of the lower plenum, while the temperature fluctuation was highest near the edge of the lower plenum wall. The power spectral density of temperature was analyzed via fast Fourier transform (FFT) for specific points on the center and side of the lower plenum. FFT results did not reveal specific frequency-dominant temperature fluctuations in the center part. It was confirmed that the temperature power spectral density (PSD) at the top increased from the center to the wake. The vortex was visualized using the well-known scalar Q-criterion, and as a result, the closer to the outlet duct, the greater the influence of the mainstream, so that the inflow jet vortex was dissipated and mixed at the top of the lower plenum. Additionally, FFT analysis was performed on the support structure near the corner of the lower plenum with large temperature fluctuations, and as a result, it was confirmed that the temperature fluctuation of the flow did not have a significant effect near the corner wall. In addition, the vortices generated from the lower plenum to the outlet duct were identified in this paper. It is considered that the quantitative and qualitative results presented in this paper will serve as reference data for the benchmark.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2020

Risk-Informed Safety Analysis for Accident Tolerant Fuels

Accident Tolerant Fuels (ATF) are being tested by different nuclear vendors and research organizations and their introduction in the US light water reactors fleet is planned for the second half of the 2020`s. In the framework of the US-DOE Light-Water Reactor Sustainability program, Risk-Informed Systems Analysis pathway, researches are being conducted at the Idaho National Laboratory for developing tools and methods that can help the industry quantifying the benefits from the adoption of ATF technology. In this paper we describe the developed risk-informed methodology including the safety analysis codes improvements, and we present some results for selected accident scenarios. The developed methodology combines INL state-of-the-art deterministic Best Estimate code RELAP5-3D, and Probabilistic Risk Analysis tools RAVEN and SAPHIRE. The analyses are performed on a three-loops pressurized water reactor, simulating Station Blackout and LB-LOCA accidents and considering near-term ATFs, or FeCrAl and Chromium-coated clads. Finally, we show how, applying our methodology, the new Core Damage Frequency can be assessed. The results indicated that the main benefits in introducing near-term ATFs concern a significant reduction in the hydrogen production during accident conditions. No significant Core Damage Frequency reduction was found.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Technical Requirements for Application of Operator Support Technology across Multiple Plant Systems

Argonne National Laboratory (ANL) and Idaho National Laboratory (INL) are jointly developing and evaluating operator aids to: (i) facilitate more timely response to plant faults and grid disturbances; (ii) achieve better management of plant upsets and improved operator performance; and (iii) ultimately improve nuclear plant safety, production, and cost management. The overall objective of the collaborative effort in this third and final year of the project was to conduct human performance tests using nuclear plant operators in the Human Systems Simulation Laboratory (HSSL) control room at INL to assess the value of a computerized operator support system (COSS) as an aid to the operator. This report describes the work performed by ANL. A companion report describes the work performed by INL and the results of the human performance tests.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1975

Helium heater design for the helium direct cycle component test facility

The paper describes a helium heater to be used to conduct non-nuclear demonstration tests of the complete power conversion loop for a direct-cycle gas-cooled nuclear reactor power plant. Requirements for the heater include: heating the helium to a 1500 F temperature, operating at a 1000 psia helium pressure, providing a thermal response capability and helium volume similar to that of the nuclear reactor, and a total heater system helium pressure drop of not more than 15 psi. The unique compact heater system design proposed consists of 18 heater modules; air preheaters, compressors, and compressor drive systems; an integral control system; piping; and auxiliary equipment. The heater modules incorporate the dual-concentric-tube 'Variflux' heat exchanger design which provides a controlled heat flux along the entire length of the tube element. The heater design as proposed will meet all system requirements. The heater uses pressurized combustion (50 psia) to provide intensive heat transfer, and to minimize furnace volume and heat storage mass.

Larson, V. R.↗
NASA NTRS2015

Nuclear Thermal Propulsion

Development efforts in the United States for nuclear thermal propulsion (NTP) systems began with Project Rover (1955-1973) which completed 22 high-power rocket reactor tests. Results indicated that an NTP system with a high thrust-to-weight ratio and a specific impulse greater than 900 s would be feasible. John F. Kennedy, in his historic special address to Congress on the importance of Space on May 25, 1961, said, "First, I believe that this nation should commit itself to achieving the goal, before this decade is out, of landing a man on the Moon and returning him safely to the Earth..." This was accomplished. He also said, "Secondly ... accelerate development of the Rover nuclear rocket. This gives promise of someday providing a means for even more exciting and ambitious exploration of space... to the very end of the solar system itself." The current NTP project focuses on demonstrating the affordability and viability of a fully integrated NTP system with emphasis on fuel fabrication and testing and an affordable development and qualification strategy. The goal is to enable NTP to be considered a mainstream option for supporting human Mars and other missions beyond Earth orbit.

Mitchell, Sonny↗