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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 1,207 records · Page 67
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Proven Uranium Recovery Solutions for the Nuclear Fuel Manufacturing Market - 16060

During the Nuclear Fuel Manufacturing process a number of by-products and residues are generated from various sources. In recent years, Westinghouse has characterised and processed around 50,000 drums of residue comprising of numerous different types generated over 60 years from the manufacture of numerous types of fuel for greater than 50 reactors. These include, AGR, Magnox (uranium metal fuel), PWR, fast reactors and various test reactors. Different organisations have different definitions and names for each of these categories but essentially they fall into one of the following categories: Clean Scrap (U{sub 3}O{sub 8}, UO{sub 2}, UO{sub 3}, UF{sub 4}, powders, pellets), Dirty Scrap (U{sub 3}O{sub 8}, UO{sub 2}, UO{sub 3}, UF{sub 4}, powders, pellets), Gd Contaminated Scrap (U{sub 3}O{sub 8}, UO{sub 2}, powders, pellets), Ash, Sludge, Combustible / soft residues (e.g. ventilation filters, gloves, plastic bags), Contaminated metals (including contaminated scrap zirconium tubes), Materials arising from decommissioning. Routes have been developed across the industry for a number of the categories listed above, however in numerous examples, routes have not been established and materials have been stored creating a significant legacy. Westinghouse understands the issues regarding legacy uranic material and the financial and regulatory impact of storage. Interim storage of materials may seem like a low cost option in the short term; however ongoing costs for facilities, manpower and equipment, without a permanent and final solution can result in significant lifetime material management costs. Recovery of uranium and sending materials for safe disposal of enriched uranium residues is not easy because of the criticality risk both during processing and at a disposal site. Therefore quantities of materials than can be sent via permitted disposal routes are very limited The development of regulations often means that finding solutions is becoming more difficult. It is believed that timely resolution is ultimately the most cost effective solution. In addition, the intrinsic value of recycling this material is also high and in some cases can mitigate the cost of processing. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Development and application of modeling tools for sodium fast reactor inspection

To support the development of in-service inspection methods for the Advanced Sodium Test Reactor for Industrial Demonstration (ASTRID) project led by the French Atomic Energy Commission (CEA), several tools that allow situations specific to Sodium cooled Fast Reactors (SFR) to be modeled have been implemented in the CIVA software and exploited. This paper details specific applications and results obtained. For instance, a new specular reflection model allows the calculation of complex echoes from scattering structures inside the reactor vessel. EMAT transducer simulation models have been implemented to develop new transducers for sodium visualization and imaging. Guided wave analysis tools have been developed to permit defect detection in the vessel shell. Application examples and comparisons with experimental data are presented.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2020

Evaluations of Silicon Carbide Temperature Monitors

The effect of irradiation on materials properties is an important field of study for materials usage in both fission and fusion systems for energy production. Neutron flux and energy spectrums are well understood; however, the irradiation temperature can be more difficult to determine. Since the early 1960s, Silicon Carbide (SiC) has been used as a passive post-irradiation temperature monitor because the irradiation defects anneal out above the irradiation temperature. Irradiation temperature is determined by measuring a property change after isochronal annealing or during a continuously monitored annealing process. There are many properties that may be measured including electrical resistivity, bulk density, dimensions, thermal diffusivity, or lattice spacing. Electrical resistivity is accepted as a robust measurement technique; however, such method is time-consuming since the steps involved must be performed in a serial manner. The use of thermal expansion from continuous dilatometry is an automated process requiring minimal setup time. As part of a Nuclear Science User Facilities (NSUF) project, low dose silicon carbide monitors were irradiated in the BR2 material test reactor at SCK, Belgium, and were then evaluated at the Idaho National Laboratory (INL) High Temperature Test Laboratory (HTTL) to determine their peak temperature achieved during irradiation. The technical significance of this work was that the monitors were irradiated to a dose that was significantly less than recommended in published literature. This paper will discuss the evaluation processes available at HTTL and the performance of the low dose silicon carbide temperature monitors.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
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 OSTI2014

Three-dimensional NDE of VHTR core components via simulation-based testing. Final report

A next generation, simulation-driven-and-enabled testing platform is developed for the 3D detection and characterization of defects and damage in nuclear graphite and composite structures in Very High Temperature Reactors (VHTRs). The proposed work addresses the critical need for the development of high-fidelity Non-Destructive Examination (NDE) technologies for as-manufactured and replaceable in-service VHTR components. Centered around the novel use of elastic (sonic and ultrasonic) waves, this project deploys a robust, non-iterative inverse solution for the 3D defect reconstruction together with a non-contact, laser-based approach to the measurement of experimental waveforms in VHTR core components. In particular, this research (1) deploys three-dimensional Scanning Laser Doppler Vibrometry (3D SLDV) as a means to accurately and remotely measure 3D displacement waveforms over the accessible surface of a VHTR core component excited by mechanical vibratory source; (2) implements a powerful new inverse technique, based on the concept of Topological Sensitivity (TS), for non-iterative elastic waveform tomography of internal defects - that permits robust 3D detection, reconstruction and characterization of discrete damage (e.g. holes and fractures) in nuclear graphite from limited-aperture NDE measurements; (3) implements state-of-the art computational (finite element) model that caters for accurately simulating elastic wave propagation in 3D blocks of nuclear graphite; (4) integrates the SLDV testing methodology with the TS imaging algorithm into a non-contact, high-fidelity NDE platform for the 3D reconstruction and characterization of defects and damage in VHTR core components; and (5) applies the proposed methodology to VHTR core component samples (both two- and three-dimensional) with a priori induced, discrete damage in the form of holes and fractures. Overall, the newly established SLDV-TS testing platform represents a next-generation NDE tool that surpasses all existing techniques for the 3D ultrasonic imaging of material damage from non-contact, limited-aperture waveform measurements. Outlook. The next stage in the development of this technology includes items such as (a) non-contact generation of mechanical vibrations in VHTR components via thermal expansion created by high-intensity laser; (b) development and incorporation of Synthetic Aperture Focusing Technique (SAFT) for elevating the accuracy of 3D imaging in highly noisy environments with minimal accessible surface; (c) further analytical and computational developments to facilitate the reconstruction of diffuse damage (e.g. microcracks) in nuclear graphite as they lead to the dispersion of elastic waves, (d) concept of model updating for accurate tracking of the evolution of material damage via periodic inspections; (d) adoption of the Bayesian framework to obtain information on the certainty of obtained images; and (e) optimization of the computational scheme toward real-time, model-based imaging of damage in VHTR core components.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
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 OSTI2015

Report detailing comparative analysis of results from high flux isotope reactor and national institute of standards technology small-angle neutron scattering experiments

The major issues regarding irradiation effects are discussed in [1-3] and have also been discussed in previous progress and milestone reports. As noted previously, of the many significant issues discussed, the issue considered to have the most impact on the current regulatory process is that associated with effects of neutron irradiation on RPV steels at high fluence, for long irradiation times, and as affected by neutron flux. It is clear that embrittlement of RPV steels is a critical issue that may limit LWR plant life extension. The primary objective of the LWRSP RPV task is to develop robust predictions of transition temperature shifts (TTS) at high fluence ( t) to at least 1020 n/cm 2 (>1 MeV) pertinent to plant operation of some pressurized water reactors (PWR) for 80 full power years. Correlations between the high flux test reactor results and low flux surveillance specimens must be established for proper RPV embrittlement predictions of the current nuclear power fleet. Additionally, a complete understanding of defect evolution for high nickel RPV steels is needed to characterize the embrittlement potential of Mn-Ni-enriched precipitates (MNPs), particularly for the high fluence regime. While understanding of copper-enriched precipitates (CRPs) have been fully developed, the recent discovery and experimental verification [4] of late blooming MNPs with little to no copper for nucleation has stimulated research efforts to understand the evolution of these phases. New and existing databases will be combined to support developing physically based models of TTS for high fluence-low flux ( < 10 11n/cm 2 -s) conditions, beyond the existing surveillance database, to neutron fluences of at least 1 1020 n/cm2 (>1 MeV). Moreover, large number of various RPV materials have been irradiated in ATR-2 experiment and will be jointly studied by University of California Santa Barbara (UCSB) and ORNL to address majority of microstructural characteristics discussed above, see Ref. [5] and [6] for details. UCSB has performed a large number of SANS experiments in the past at the National Institute of Standards and Technology (NIST) Center for Neutron Research (NCNR). These data are taken from RPV steels irradiated in a wide range of flux-fluence space and will be very useful in comparing to the upcoming UCSB ATR-2 irradiation characterization since most of the SANS experiments with ATR-2 materials will be performed at ORNL High Flux Isotope Reactor (HFIR). However in the previous report [7], some discrepancies were observed between HFIR and NCNR generated data. One of the hypotheses was that there was some kind of extra scattering occurring off the sample holders that results in the HFIR curves falling above the NCNR curves. To test this hypothesis, UCSB provided thermally aged samples that have been previously run at NCNR to ORNL for testing at HFIR while ORNL performed some improvements to experimental set up at HFIR. This report provides the status for the Level 3 Milestone (M3LW-15OR0402013), Complete report detailing comparative analysis of results from High Flux Isotope Reactor and National Institute of Standards and Technology small-angle neutron scattering experiments. This milestone is associated with small-angle neutron scattering characterization at the High Flux Isotope Reactor of various model alloys that had been previously characterized at NCNR by UCSB.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
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 OSTI2015

Phased Array Ultrasonic Examination of Reactor Coolant System (Carbon Steel-to-CASS) Dissimilar Metal Weld Mockup Specimen

In the summer of 2009, Pacific Northwest National Laboratory (PNNL) staff traveled to the Electric Power Research Institute (EPRI) NDE Center in Charlotte, North Carolina, to conduct phased-array ultrasonic testing on a large bore, reactor coolant pump nozzle-to-safe-end mockup. This mockup was fabricated by FlawTech, Inc. and the configuration originated from the Port St. Lucie nuclear power plant. These plants are Combustion Engineering-designed reactors. This mockup consists of a carbon steel elbow with stainless steel cladding joined to a cast austenitic stainless steel (CASS) safe-end with a dissimilar metal weld and is owned by Florida Power & Light. The objective of this study, and the data acquisition exercise held at the EPRI NDE Center, were focused on evaluating the capabilities of advanced, low-frequency phased-array ultrasonic testing (PA-UT) examination techniques for detection and characterization of implanted circumferential flaws and machined reflectors in a thick-section CASS dissimilar metal weld component. This work was limited to PA-UT assessments using 500 kHz and 800 kHz probes on circumferential flaws only, and evaluated detection and characterization of these flaws and machined reflectors from the CASS safe-end side only. All data were obtained using spatially encoded, manual scanning techniques. The effects of such factors as line-scan versus raster-scan examination approaches were evaluated, and PA-UT detection and characterization performance as a function of inspection frequency/wavelength, were also assessed. A comparative assessment of the data is provided, using length-sizing root-mean-square-error and position/localization results (flaw start/stop information) as the key criteria for flaw characterization performance. In addition, flaw signal-to-noise ratio was identified as the key criterion for detection performance.

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↗