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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 451 records · Page 25
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Signature Analysis Utilizing a Dynamic Molten Salt Reactor Model for MC&A

Moving from traditional fixed fuel nuclear reactor systems to a mobile, dynamic fuel system that has dissolved special nuclear material in a molten salt is a paradigm shift in several respects. One major consideration is how to develop effective nuclear material controls and accounting for these novel reactor systems. The Molten Salt Reactor Experiment is one example of a critical molten salt system and provided a significant reference library based on the documented effort. But it was low thermal power that was not intended to reflect a commercial scale electricity production design. Therefore, to facilitate and assist vendors with domestic licensing considerations, research is underway to identify methods that could be used for domestic safeguards approaches for these novel reactor systems. This research presents the results of a signature analysis of data generated from three simulated scenarios using the molten salt demonstration reactor model defined in the Transient Simulation Framework of Reconfigurable Modules. Each scenario provides 1 hour isotope inventories over a 180 day period. The scenarios investigated provide test cases to examine if direct gamma-ray spectroscopy of the fuel can be used to identify changes comparing a base case (no reactivity control and fixed fission contribution) to an insertion of reactivity (10 pcm no change in fission composition) and a change in fission composition. The analysis demonstrates that monitoring the total count rate in a highly collimated high-resolution photon energy spectrum is sensitive to perturbations imposed into the reactor model. The total photon count rate changes ≈2% for the fission composition change and ≈4.5% for the reactivity insertion compared to the base case. However, both scenarios show an increase in the total photon count rate. The total photon count rate can be used to identify changes due to power (number of fissions) but not due to a change in the material undergoing fission. To distinguish between the two cases of increased power, the photon spectrum would require an intensive analysis technique. A photon peak count rate ratio analysis could be used to identify changes in the fissile material fission generation in the core through identification of a static peak that shows little variation to the source of fission and a highly varying peak. The photon peak strength will ultimately be determined by the isotope’s fission yield. A preliminary analysis investigating the coupling of the isotope’s fission yield and its concentration in the fuel salt derived from the modeling has been performed. A ratio analysis of the photon peak count rates of 140 La to 99 Mo demonstrated that the reactivity insertion creates a distinct difference in the ratio compared to the fission composition change scenario.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Space Nuclear Power Autonomous Control Algorithm and Control Element Test Bed

Nuclear thermal rockets are currently NASA’s preferred option for use in a manned mission to Mars in the 2040s. The communication delay between an Earth ground station and a spacecraft heading toward Mars can be up to 20 min. Therefore, controlling the nuclear rocket engine would require either a full-time reactor operator on the mission or an autonomous control system for the reactor. The latter idea of making space nuclear reactors fully autonomous has drawn more interest from stakeholders, but such an autonomous control system must be rigorously tested and validated before it is certified for human use. The cost of a full ground test for a space nuclear reactor is tremendous, so a nonnuclear mock reactor test bed was created to test and validate control elements and control algorithms for space nuclear reactors. The test bed consists of control element hardware that inputs physical measurement data into a reactor emulator to produce the reactor’s performance under steady-state, transient, and fault conditions. The control element hardware consists of six full-sized control drums equipped with servo drives and motors and is instrumented with optical encoders, resolvers, and torque sensors for drum movement characterization. In addition to the drums, a two-phase flow loop was designed and built to mimic the valves and turbomachinery associated with the propellant flow through a nuclear thermal rocket engine; components such as pressure sensors, flow meters, thermocouples, and tachometers are instrumented throughout the loop to characterize the fluid flow, valve, and turbomachinery behavior of the system. The data from the physical hardware (e.g., drum position, propellant flow rates) are input to a nuclear reactor simulator to determine the actual nuclear reactor parameters, and the data are sent back to a control algorithm to complete the control loop. The ability to conduct numerous tests of the control systems and autonomous algorithms can help validate the instrumentation and control aspects for a space nuclear reactor for every possible fault situation.

Wilson, Brandon↗
DOE OSTI2015

Thick Films acoustic sensors devoted to MTR environment measurements. Thick Films acoustic sensors devoted to Material Testing Reactor environment measurements

The development of advanced instrumentation for in-pile experiments in Material Testing Reactor constitutes a main goal for the improvement of the nuclear fuel behavior knowledge. An acoustic method for fission gas release detection was tested with success during a first experiment called REMORA 3 in 2010 and 2011, and the results were used to differentiate helium and fission gas release kinetics under transient operating conditions. This experiment was lead at OSIRIS reactor (CEA Saclay, France). The maximal temperature on the sensor during the irradiation was about 150 deg. C. In this paper we present a thick film transducer produce by screen printing process. The screen printing of piezoelectric offers a wide range of possible applications for the development of acoustic sensors and piezoelectric structure for measurements in high temperature environment. We firstly produced a Lead Zirconate Titanate (PZT) based paste composed of Pz27 powder from Ferroperm, CF7575 glass, and organic solvent ESL 400. Likewise a Bismuth Titanate based paste synthesized in our laboratory was produced. With these inks we produced thick film up to 130 μm by screen printing process. Material properties characterizations of these thick-film resonators are essential for device design and applications. The piezoelectric coefficients d33 and pyro-electric P(T) coefficient are investigated. The highest P(T) and d33 are respectively 80 μC.m{sup -2}.K{sup -1} and 130 μC.N{sup -1} for the PZT transducer -which validates the fabrication process-. In view of the development of this transducer oriented for high temperature and irradiation environment, we investigated the electrical properties of the transducers for different ranges of frequencies and temperature - from 20 Hz up to 40 MHz between 30 and 400 deg. C. We highlight the evolution of the impedance response and piezoelectric parameters of screen printed piezoelectric structures on alumina. Shortly an irradiation will be realized in order to investigate their reliability towards neutron fluence. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Pretreatments for the continuous production of pyrolytic sugar from lignocellulosic biomass

Fast pyrolysis of lignocellulosic biomass yields little sugar or anhydrosugars compared to pyrolysis of pure polysaccharides because naturally abundant alkali and alkaline earth metals (AAEM) in biomass catalyze the fragmentation of pyranose and furanose rings. Sugar yields can be increased dramatically by pretreating the biomass with sulfuric acid prior to pyrolysis, which passivates the catalytic activity of the metals by converting them into thermally stable salts. However, depolymerization of lignin in biomass also depends on the catalytic activity of AAEM. Thus, passivating AAEM has the unintended consequence of slowing the rate of depolymerization and volatilization of lignin, resulting in a transient melt phase that agglomerates and can foul pyrolysis reactors. Here, to overcome this problem, various non-alkali metal sulfates were tested as replacements for sulfuric acid. Iron in the form of ferrous sulfate proved the most effective in depolymerizing lignin without fragmenting pyranose rings. Conventional nitrogen-blown pyrolysis of ferrous sulfate pretreated corn stover achieved WHSV of 4 h –1 compared to only 0.6 h –1 for acid pretreated corn stover. Autothermal (air-blown) pyrolysis of ferrous sulfate pretreated corn stover showed even more dramatic improvement, increasing WHSV from 1 h –1 to 10 h –1 compared to acid pretreated corn stover under autothermal operation. Fermentable sugar yields from the pyrolysis of corn stover increased from 0.9 wt% to 11.8 wt% on a biomass basis, a 13-fold increase as a result of the ferrous sulfate pretreatment. These advantages combine to increase volumetric sugar productivity from 62 g L –1 h –1 for conventional pyrolysis of untreated corn stover to 2041 g L –1 h –1 for autothermal pyrolysis of ferrous sulfate treated corn stover.

09 BIOMASS FUELS↗
DOE OSTI2023

BISON analysis of FeCrAl and Zircaloy cladding deformation during simulated BWR cyclic dryout conditions

The BISON fuel performance analysis code has been used to study FeCrAl and Zircaloy cladding performance to better understand cladding deformation behavior in boiling water reactors during conditions caused by anticipated operational occurrences (AOOs) or by anticipated transients without scram (ATWS). BISON models of novel separate effects tests simulating ATWS events predicted mechanical responses to thermal cycling boundary conditions. Here, we assessed FeCrAl and Zircaloy deformation and burst models to determine those best suited for cyclic dryout event analysis. In addition, we compared diametral strain profiles of the post-experiment claddings to simulation results. The capability of the BISON code for modeling BWR AOO or ATWS events is promising.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Status and Availability of OPTD, the Out-of-Pile Transient Database

The DOE Advanced Reactor Technology program has supported knowledge preservation efforts to recover and preserve fuel data from the US sodium cooled fast reactor (SFR) program. Databases have been established containing experimental data generated during the Integral Fast Reactor program (from in-pile experiments at EBR-II, FFTF, and TREAT) and during related outof-pile examinations of EBR-II-irradiated fuels. The information in these databases is essential to support further development and licensing activities for advanced fast reactor designs. OPTD, the Out-of-Pile Transient Database, is an organized, searchable archive of records describing out-ofpile furnace transient tests conducted on metallic fuels. It includes records of over 150 tests on irradiated fuel pins conducted in furnace apparatuses constructed in the Alpha-Gamma Hot Cell Facility at ANL. This report presents the content, functions, and accessibility of the Out-of-Pile Transient Database (OPTD). It describes the information presently available in the database, the web interface created to provide access to this information, the activities undertaken to build OPTD, and the enabling of online access to the database for DOE users. Plans for FY2021 and beyond are also included.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

RELAP5-3D Modeling of High Temperature Test Facility (HTTF)

HTTF at Oregon State University (OSU) Reference: General Atomics? modular high-temperature gas-cooled reactor Helium cooled, electrically heated Prismatic graphite blocks in the core and reflectors Alumina ceramic blocks are used to simulate the core and top and bottom reflectors One-fourth scale in length and diameter Most of the coolant channels in the core are full scale Lower pressure compared to the prototype reactor Over 500 instruments Designed primarily to investigate depressurized conduction cooldown (DCC) transients

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2025

Step-loaded creep testing of Zircaloy-4 cladding at higher temperatures in the α-phase

A refined understanding of zirconium-based cladding thermomechanical performance during rapid transients is essential for enhancing the safety and operation of light-water reactors. Traditional models for zirconium alloys under accident conditions generally assume that creep dominates fuel cladding performance. Here, these historic models have largely remained unchanged and serve as the basis for safety criteria development. As the U.S. nuclear industry pursues higher burnup levels, the increased release of fission gases during transients raises the risk of cladding failure in the low-temperature hcp α-phase, making the fidelity of these models of greater importance. Creep testing was conducted from 550–700°C with 25–120 MPa applied hoop stresses to investigate Zircaloy-4 deformation at accident-relevant temperatures in the α-phase. Step-loading was employed to capture creep behavior across a wide stress range from a single sample. The stress-strain rate data at higher temperatures (650 and 700°C) were well-described by isotropic versions of the Erbacher and Kaddour models, while the lower temperature data (550 and 600°C) were underpredicted by both anisotropic and isotropic model variants. Greater strain rates during the initial loading step at 650 and 700°C were attributed to recrystallization and grain growth of sub-micron crystallites. Yet, texture analysis revealed the basal split texture remained after testing. These observations produced results suggesting Zircaloy-4 claddings experience higher creep rates across the α-phase than previously thought, possibly related to dynamic anisotropy due to temperature dependent activation of deformation mechanisms, effects of biaxial loading, and variation in material condition between the current testing used in previous model development.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Thermal Hydraulics Validation Activities in the ART-GCR Campaign

The Advanced Reactor Technologies - Gas Cooled Reactor Campaign (ART-GCR) is part of the United States Department of Energy - Office of Nuclear Energy's Advanced Reactor Technologies program aimed at developing a High Temperature Gas-cooled Reactor (HTGR) which will offer enhancements in safety and efficiency. The Design, Methods, and Validation arm of the ART-GCR program provides experience and advanced tools for HTGR design and analysis. The Methods and Validation arm is developing data to validate predictions of decay heat removal through an extensive experimental campaign the Natural Convection Shutdown Heat Removal Test Facility at Argonne National Laboratory. In-core thermal hydraulics validation is being facilitated through the OECD/NEA's Thermal Hydraulic Code Validation Benchmark for High Temperature Gas-Cooled Reactors using HTTF Data. That benchmark contains code-to-code and code-to-data comparisons for 3 sets of HTGR thermal hydraulics phenomena. This ongoing benchmark activity has 16 participants from 9 countries developing leading insights into HTGR thermal hydraulics code validation. The ART-GCR campaign is also working on the development of state-of-the-art MOOSE-based models of the High Temperature Engineering Test Reactor (HTTR). Through the OECD/NEA's HTTR Loss of Forced Cooling (LOFC) program, the ART-GCR campaign is developing a multiphysics transient code validation benchmark based on 3 HTTR experiments that will provide an opportunity for international collaboration on multiphysics code validation.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

The 14 MeV Neutron Irradiation Facility in MARIA Reactor

The MARIA reactor with thermal neutron flux density up to 3x10{sup 14} cm{sup -2} s{sup -1} and a number of vertical channels is well suited to material testing by thermal neutron treatment. Beside of that some fast neutron irradiation facilities are operated in MARIA reactor as well. One of them is thermal to 14 MeV neutron converter launched in 2014. It is especially devoted to fusion devices material testing irradiation. The ITER and DEMO research thermonuclear facilities are to be run using the deuterium - tritium fusion reaction. Fast neutrons (of energy approximately 14 MeV) resulting from the reaction are essential to carry away the released thermonuclear energy and to breed tritium. However, constructional materials of which thermonuclear reactors are to be built must be specially selected to survive intense fluxes of fast neutrons. Strong sources of 14 MeV neutrons are needed if research on resistance of candidate materials to such fluxes is to be carried out effectively. Nuclear reactor-based converter capable to convert thermal neutrons into 14 MeV fast neutrons may be used to that purpose. The converter based on two stage nuclear reaction on lithium-6 and deuterium compounds leading to 14 MeV neutron production. The reaction chain is begun by thermal neutron capture by lithium-6 nucleus resulted in triton release. The neutron and triton transport calculations have been therefore carried-out to estimate the thermal to 14 MeV neutron conversion efficiency and optimize converter construction. The usable irradiation space of ca. 60 cm{sup 3} has been obtained. The released energy have been calculated. Heat transport has been asses to ensure proper device cooling. A set of thermocouples has been installed in converter to monitor its temperature distribution on-line. Influence of converter on reactor operation has been studied. Safety analyses of steady states and transients have been done. Performed calculations and analyses allow designing the converter and formulate its operation limits and conditions. During first tested operation of the converter the 14 MeV neutron flux density was estimated to 10{sup 9} cm{sup -2} s{sup -1}, whereas fast fission neutrons inside converter achieved 10{sup 12} cm{sup -2} s{sup -1}, and thermal neutrons were reduced down to 109 cm-2 s-1. Taking into account the feasibility of almost incessant converter operation for a number of months, its arisen as one of the most powerful (in terms of fluence), currently available 14 MeV neutron source. Such a converter currently under operation in the MARIA reactor core will be presented. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

FY23 Status of Quality Assurance Plan for Out-of-Pile Test Data

The DOE Advanced Reactor Technology program has supported recovery and preservation of legacy metallic fuel data collected as part of the US fast reactor program, recognizing it as essential to development and licensing activities for advanced fast reactors. Databases were established as organized collections of experimental records and data generated from in-pile experiments at EBRII, FFTF, and TREAT as well as related out-of-pile examinations of irradiated fuels. The Out-of-Pile Transient Database (OPTD), includes records of over 150 out-of-pile furnace transient tests on metallic fuels conducted at Argonne’s Alpha-Gamma Hot Cell Facility to evaluate their transient performance and characterize fuel/cladding interaction. The database is accessible to registered users from US universities, laboratories, and nuclear industry. Because the data in OPTD has not been formally qualified, its applicability and ability to support licensing activities is limited. This report outlines progress and plans to quality assure data in OPTD, maximizing its impact for model validation and verification as well as qualification of fuels for safe and effective use in advanced reactor designs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Investigation of Transient Xenon Processes in VVER-1200 at the Novovoronezh NPP

The results of experimental studies of the integral and spatial transient xenon processes in the No. 6 unit of the Novovoronezh NPP with VVER-1200 are presented. An unconventional method of measurement was tested in the study of integral processes: the course of the xenon process was recorded as the change in the critical concentration of boric acid in the reactor simultaneously with a calculation of the concentration in real-time. The spatial transient processes were studied for the example of axial and diametral free xenon oscillations of energy release in the core.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Multiphysics Analysis of the MSRE Experiment Using Griffin-SAM Coupled Code System

Molten salt reactors (MSRs) with flowing fuel have the unique feature of utilizing the fuel salt for heat generation and extraction at the same time since the fuel salt is circulating through the whole primary loop of the reactor. This movement of the fuel salt results in a partial decay of the delayed neutron precursors (DNPs) outside the core and corresponding redistribution in the active core region. To capture this phenomenon accurately, neutronics and thermal hydraulics computational tools need to be modified to handle the movement of the DNPs and their decay. Idaho and Argonne National Laboratories are actively working on developing neutronics and thermal hydraulics tools to model MSRs with flowing fuel. The neutronics code Griffin and the thermal hydraulics code SAM were extended to handle flowing fuel with the drift of the DNPs under the MOOSE framework. In this work, a simplified Multiphysics model of the Molten Salt Reactor Experiment (MSRE) is developed and utilized to perform steady-state and transient analyses. The MSRE experiments were designed to show the impact of the DNPs losses on core reactivity. The following section provides a description of the MSRE experiment with the core main parameters used to develop the model. Then, the developed Multiphysics model of the MSRE is presented. Finally, steady-state verification tests and unprotected loss of flow transient tests are discussed followed by a summary and conclusions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Refabriation of HERA-HBU-2, LOC-HBU-1, and LOC-HBU-2

The development of advanced fuels for Light Water Reactors (LWRs) is essential for improved fuel performance and safety, including performance margins, uprates, and higher burnup. Because available commercial LWR fuels often cannot be used in research reactors, the refabrication of preirradiated fuel is necessary to enable follow-on testing of these fuels for characterizing fuel behavior during steady and transient irradiations.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Forced flow transient safety analysis of irradiation device with adjustable orifice for research reactor fuel assemblies

The Belgium Reactor 2 (BR2) of the Belgian Nuclear Research Centre (SCK CEN) has several irradiation devices or rigs that are dedicated to the fuel performance and qualification demonstration testing of research reactor fuels. In support of the U.S. High Performance Research Reactor (USHPRR) LEU conversion project, a new flexible irradiation apparatus, MUSTANG-R, has been constructed. SCK CEN has completed the design and safety study, in cooperation with Idaho National Laboratory (INL) and Argonne National Laboratory (ANL), to allow for the irradiation testing of a full-size fuel assembly in a 200 mm diameter channel in the BR2 reactor. The moveable valve is a key design feature of the device and acts like an adjustable orifice enhancing or restricting the flow through a coolant channel inlet located in the BR2 upper plenum. This moveable valve allows the flow through the device to be adjusted prior to each BR2 cycle to obtain the necessary conditions for the fuel qualification test. This ensures accurate and representative thermal-hydraulic conditions of the fuel design are achieved. The device was designed and qualified as passively safe, implying verification by a combination of mechanical and thermal-hydraulic analysis and testing. This includes characterization of the safety margin required for a scenario where the moveable valve is assumed to be erroneously closed during irradiation. A simplified and conservative method is proposed for analyzing the corresponding forced flow transient using a critical heat flux criterion. In conclusion, this allows the required minimum valve opening to be determined for the experiments' design and safety studies.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Improvements to PHISICS/RELAP5-3D{sup C} Capabilities for Simulating HTGRs

Accurate, high-resolution full core simulations of prismatic high temperature reactors for burnup and transient analysis remain a challenge. The cores are geometrically large, but some features are quite small (tri-structural isotropic (TRISO) particles, burnable poison pellets). The treatment of both neutron scattering in graphite and resonance capture are complex and not adequately captured using the methods traditionally used in High Temperature Gas-cooled Reactors (HTGR) and commonly used in Light Water Reactors (LWR). For transient analysis, temperature feedback is a function of the TRISO fuel form, but full core models cannot resolve phenomena at this scale without averaging over space and energy. For burnup calculations, a second level of heterogeneity must be resolved to capture the local effects of fuel and burnable poisons, while accurately propagating their effects through and between blocks. The HTGR Methods and Simulation group has been utilizing the INL-developed PHISICS/RELAP5-3D{sup C} (P/R), code suite for participation in the Organization for Economic Cooperation and Development (OECD) Modular High Temperature Gas Cooled Reactor (MHTGR) Benchmark, funded by the US Department of Energy (DOE) under the Advanced Reactor Technologies (ART) program. The MHTGR-350 benchmark consists of several lattice, steady-state and transient exercises, and the P/R code has been successfully applied to most of the cases defined. In addition to this code-to-code verification activity, it is planned that P/R will be validated using data from several Loss of Forced Cooling (LOFC) tests performed at the High Temperature Test Reactor (HTTR) in a joint effort with Japan Atomic Energy Agency (JAEA) under the Civil Nuclear Energy Working Group (CNWG) project. These verification and validation simulations will identify code and model discrepancies, sensitivities, and establish the current state-of-the-art in HTGR simulation. The Department of Astronautical, Electrical and Energy Engineering (DIAEE), 'Sapienza' University has been involved in the development of P/R since 2014 through various graduate and PhD student projects. In the framework of this collaboration the P/R optimization for application on the OECD MHTGR-350 and HTTR models will be carried out.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗