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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 145 records · Page 8
All sourcesNASA NTRSDOE OSTI
NASA NTRS2011

Transient Response to Rapid Cooling of a Stainless Steel Sodium Heat Pipe

Compact fission power systems are under consideration for use in long duration space exploration missions. Power demands on the order of 500 W, to 5 kW, will be required for up to 15 years of continuous service. One such small reactor design consists of a fast spectrum reactor cooled with an array of in-core alkali metal heat pipes coupled to thermoelectric or Stirling power conversion systems. Heat pipes advantageous attributes include a simplistic design, lack of moving parts, and well understood behavior. Concerns over reactor transients induced by heat pipe instability as a function of extreme thermal transients require experimental investigations. One particular concern is rapid cooling of the heat pipe condenser that would propagate to cool the evaporator. Rapid cooling of the reactor core beyond acceptable design limits could possibly induce unintended reactor control issues. This paper discusses a series of experimental demonstrations where a heat pipe operating at near prototypic conditions experienced rapid cooling of the condenser. The condenser section of a stainless steel sodium heat pipe was enclosed within a heat exchanger. The heat pipe - heat exchanger assembly was housed within a vacuum chamber held at a pressure of 50 Torr of helium. The heat pipe was brought to steady state operating conditions using graphite resistance heaters then cooled by a high flow of gaseous nitrogen through the heat exchanger. Subsequent thermal transient behavior was characterized by performing an energy balance using temperature, pressure and flow rate data obtained throughout the tests. Results indicate the degree of temperature change that results from a rapid cooling scenario will not significantly influence thermal stability of an operating heat pipe, even under extreme condenser cooling conditions.

Mireles, Omar R.↗
NASA NTRS2012

Large-Scale Weibull Analysis of H-451 Nuclear- Grade Graphite Specimen Rupture Data

A Weibull analysis was performed of the strength distribution and size effects for 2000 specimens of H-451 nuclear-grade graphite. The data, generated elsewhere, measured the tensile and four-point-flexure room-temperature rupture strength of specimens excised from a single extruded graphite log. Strength variation was compared with specimen location, size, and orientation relative to the parent body. In our study, data were progressively and extensively pooled into larger data sets to discriminate overall trends from local variations and to investigate the strength distribution. The CARES/Life and WeibPar codes were used to investigate issues regarding the size effect, Weibull parameter consistency, and nonlinear stress-strain response. Overall, the Weibull distribution described the behavior of the pooled data very well. However, the issue regarding the smaller-than-expected size effect remained. This exercise illustrated that a conservative approach using a two-parameter Weibull distribution is best for designing graphite components with low probability of failure for the in-core structures in the proposed Generation IV (Gen IV) high-temperature gas-cooled nuclear reactors. This exercise also demonstrated the continuing need to better understand the mechanisms driving stochastic strength response. Extensive appendixes are provided with this report to show all aspects of the rupture data and analytical results.

Nemeth, Noel N.↗
NASA NTRSDate not supplied

Overview of High Temperature Material Needs for Space Nuclear Propulsion Reactors

Space nuclear propulsion systems are capable of enabling future crewed missions to Mars. Two primary options of these systems exist: nuclear thermal propulsion (NTP) and nuclear electric propulsion (NEP). NTP systems use a reactor as a heat exchanger to heat a hydrogen propellant to produce high specific impulse (≥ 900 s) and thrust (10 - 100 klbf). A NTP materials challenge is development of nuclear fuel capable of operating for short durations (hours) under high power densities (≥ 5 MW/L) and temperatures (> 2700 K) in a corrosive hydrogen environment. In NEP systems, the reactor heats a working fluid which transfers heat to a power conversion system to power electric thrusters. NEP is capable of higher specific impulse (≥ 2000 s) but limited to low thrust (10s lbf). Light weight vessel and in-core structural materials that are radiation-resistant, creep-resistant, and dimensionally stable while operating at ~1500 K over mission times up to several years are needed for NEP. Both systems can benefit from high temperature neutron moderators. Through the Space Nuclear Propulsion project, NASA, supported by the DOE, is developing NTP and NEP systems. Technology maturation planning and risk reduction activities related to fabrication and testing of reactor material candidates is ongoing. This presentation overviews NTP and NEP materials needs and current ceramic and composite development activities.

nuclear thermal propulsion↗
NASA NTRSDate not supplied

Heat Pipe Heat Exchanger for Nuclear Electric Propulsion Power Conversion System

Heat pipe reactors have been considered by the Space Nuclear Propulsion program for Nuclear Electric Propulsion (NEP) power conversion systems and will require the use of heat exchangers to transfer heat via heat pipes to the Brayton working fluid from the reactor. Sodium (Na) and lithium (Li) were considered as viable working fluids inside the heat pipes which were assumed to have the same geometry based on studies and information from the Los Alamos National Laboratory. The heat exchanger was assumed to be a rectangular duct with heat pipes serving as tubes from previous NEP work and recommendations. Based on this geometry, Zukauskas correlations were used to model the convective heat transfer and pressure losses. Parametric sizing of the reactor component involved operational limits-based heat pipe thermal hydraulic modeling in cohesion with required user input geometry for the in-core lattice and various subcomponents. This work considered various power conversion inlet temperatures (PCIT) of 1100 K, 1150 K, and 1200 K for Na heat pipes and 1100 K, 1150 K, 1200 K, and 1400 K for Li heat pipes based on recommendations from prior work. Using these different PCITs, the subsystem masses and pressure losses were determined and analyzed. Na showed a lower overall operating temperature and about a fifth of the maximum heat throughput capability than that of Li for the same geometry. Due to this, the entire Na-based subsystem ended up being three times more massive than the Li-based subsystem given five times the required number of heat pipes. At the low PCIT of 1100 K, the Na-based subsystem exhibited the lowest pressure losses given the large overall cross sectional flow area and relatively low frictional pressure losses. However, as the PCIT increased, the frictional pressure losses increased resulting in higher pressure losses at the 1200 K PCIT than Li-based subsystem. However, the Li-based subsystem exhibited the largest pressure losses of all analyzed cases at the 1400 K PCIT due to the low density of the Brayton working fluid at this temperature.

electric↗
DOE OSTI2022

Dynamic System Scaling Application to Accelerated Nuclear Fuel Testing

The development of nuclear fuel and materials requires a continuous effort to investigate the acute and prolonged effects of irradiation, thermal-material stress, chemical change, or other conceivable damage mechanics acquired during normal operation or accident scenarios throughout its lifetime. As in-core fuel property measurement techniques advance to support in the real-time, non-invasive, and enhanced accuracy realm, it is the future of fuel development to pursue to higher degrees of control, predictability of integral test behavior via separate effects test (SET), and shorter test time intervals. The fuel development life cycle from initial concept to commercial licensing is approximated to be 20 years and current literature suggests by optimizing fuel performance codes with SETs, the process could possibly be compressed to 5 to 10 years. Recently, a research group in the Idaho National Laboratory (INL) is testing reduced scale fuel rods and increased power density to accelerate evolution of fuel phenomena in metallic fuels. In support of nuclear fuel rod development, compressing fuel test process, and accelerating fuel phenomena, it was the purpose of this study to investigate nuclear fuel performance phenomena via literature review and effectively scale the initial conditions, boundary conditions, and geometric properties to describe the time-dependent response including to fuel burnup, thermal and mechanical stress, transmutations and inter-diffusion, and other relevant observed phenomena. The study was based on BISON simulations of historic EBR-II metallic fuel experiments and Dynamical System Scaling (DSS) method are utilized to assess effects of scaling activity including fuel phenomena acceleration and calculations of time-dependent distortions. The research successfully scaled metallic fuel phenomena, accelerated fuel testing, and assessed the distortions for each scaled case derived.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Irradiation Testing of Sensors - ASI Program Review 2022

This is a powerpoint presentation for the annual Advanced Sensors and Instrumentation (ASI) program webinar. The following is a summary of the presentation. Advanced instrumentation enables testing of nuclear fuels and materials in support of the US advanced nuclear technology industry. A number of promising sensor technologies are in the pipeline. The early part of sensor development can be done outside of the reactor environment, but full technical readiness requires experience gained from in-core performance testing. Customers usually have only one shot to conduct their irradiation experiments. Because of the high costs test sponsors are frequently reluctant to incorporate unproven technologies in reactor experiments. Therefore, it is vital to demonstrate newly-developed sensors in operational conditions, prior to incorporating them into long-term high-value experiments. Successful completion of these activities will create new monitoring capabilities in high-power test reactors (such as ATR) as well as specialty reactors such as TREAT.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

SERPENT 2 and TRIPOLI-4 neutronics benchmarking of the Jules Horowitz reactor

Jules Horowitz Reactor (JHR) is a future center for nuclear material testing in Europe, currently under construction at CEA Cadarache in southern France. It has been modelled extensively with CEA's HORUS3D/N neutron calculation package, that includes a Monte-Carlo neutron transport code TRIPOLI-4. In order to support the future computational analyses on the reactor core physics, VTT's Monte Carlo neutronics code SERPENT 2 has been used to create an alternative model of the JHR, taking advantage of the pre-existing Constructive Solid Geometry (CSG) structures on which the TRIPOLI-4 model was built. This work describes a benchmark, in which two JHR core configurations were analyzed using both of these models. The first configuration is a fresh fuel core at the beginning of life, in which aluminum based dummy-devices are installed in the reactor test positions. The second configuration represents the end-of-cycle equilibrium core with a heterogeneous burnup distribution, in which CALIPSO devices containing cladding material samples are installed at the in-core experimental positions, while the reflector positions are occupied by DEN-REP devices containing UO{sub 2} fuel. Neutron flux spectra and heat deposition are the parameters of interest in this study, due to their high importance in creating the desired experimental irradiation environment. The impact of delayed gamma energy deposition will be studied in detail, since the two models account for it with different methods. This work aims to establish a SERPENT 2 model of JHR, understanding its differences and agreements with TRIPOLI-4, and discussing its potential future applications. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Simulation of Peach Bottom turbine trip transient with RETRAN-3D using BXFGEN

The Peach Bottom 2 Turbine Trip transient is used to evaluate the capabilities and accuracy of coupled 3-D neutronics/thermal hydraulics best estimate calculations for support of safety analysis. Selected results, obtained with RETRAN-3D, are compared to available measurements and benchmark results. The comparisons are analyzed and the application of RETRAN-3D for a best estimate analysis approach within the safety analysis methodology is discussed. Overall, both the thermal-hydraulic and neutronic results show a good comparison to the experimental data, and the model is considered to accurately predict test data. In particular, the initial acoustic response of the main steam line and the subsequent main pressurization stages are well-predicted. During the main pressurization stage, the timing and magnitude of the in-core local pressurization and associated void compression are properly predicted based upon the agreement between local power range monitors (LPRM) predictions upwards, but the shift is slightly underpredicted. This indicates the methodology is well-posed for simulating this class of transient response.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

FFTF HT9 Cladding Microstructure Characterization

The sodium-cooled fast reactor (SFR) is a promising candidate for next generation nuclear reactors, operating at extreme conditions which include high temperatures (>500?C core outlet temperature) and significant neutron damage. High-Cr martensitic HT9 steel is an excellent candidate for SFR cladding and duct material due to its compatibility with liquid sodium, good thermal conductivity, resistance to void swelling, and strong creep rupture strength [1-4].However, the harsh in-core environment of SFRs can cause complex microstructural changes and mechanical property degradation in HT-9. Ensuring the safe use of HT9 cladding for metallic fuel requires both a thorough understanding of its mechanical response to microstructure evolution as well as reliable microstructure-sensitive modeling predictions. Microstructure-sensitive modeling of high temperature creep behavior in HT9 cladding for SFR applications currently lack experimental data to model the phenomena accurately. To fill this need, methods to perform microstructural characterization have been developed and performed on HT9.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Advanced Characterization Capabilities for Nuclear Materials via Nuclear Science User Facilities (NSUF)

Advanced post-irradiation examination (PIE) techniques are required to design new or improved nuclear materials, characterize, and understand in-core behavior of fuel and materials, and support the qualification of new reactor materials. The Nuclear Science User Facilities (NSUF) is the U.S. Department of Energy Office of Nuclear Energy's only designated nuclear science user facility. NSUF provides researchers access to PIE capabilities at Idaho National Laboratory and at a diverse mix of university, national laboratory and industrial partner institutions. The PIE capabilities include novel destructive and non-destructive techniques for radiation damage characterization, such as advanced diffraction techniques (X-ray, electron, or neutron) coupled to extreme environments; in-situ observation of microstructural evolution under irradiation; in-situ irradiation to monitor corrosive attack in coolant environments; in-situ irradiation and mechanical testing; and test methods for synergistic effects of superimposed extreme environments (temperature, irradiation, stress, corrosion) on materials behaviors. This talk will provide an overview of NSUF PIE capabilities.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Innovative Nuclear Materials Outboard-A Project Specimen Preparation Guide

The Innovative Nuclear Materials (INM) Program was recently established by the U.S. Department of Energy (DOE) to develop advanced material technologies for use in nuclear reactors. The INM program is presently focused on researching in-core non-fueled materials for application in fast spectrum nuclear reactors. The widespread deployment of fast reactors continues to be a prominent aspiration for advanced nuclear technology developers. However, companies working to license these reactors have no choice but to rely on historic material technologies since further optimization and advancement of these materials is impeded by the lack of fast neutron irradiation test facilities. INM-OA is a non-fueled drop-in experiment which will irradiate material specimens of interest to fast reactor applications. This experiment will be irradiated at Idaho National Laboratory (INL) in the Advanced Test Reactor (ATR) outboard-A (OA) position during normal and high temperature steady state (HTSS) cycles. It will include material specimens supplied by members of the INM program and will utilize a cadmium-lined basket to filter out incident thermal neutrons, thus simulating a faster neutron energy spectrum. Material specimens will undergo post-irradiation examination including microscopy and mechanical testing. In addition to absorption reactions, fast neutrons cause microstructural damage in materials by atom displacement, which can cause exacerbated changes in physical properties and behavior. Thus, the data obtained from the INM-OA experiment will be crucial for understanding the engineering-scale behavior of reactor materials. This document is intended for the Principal Investigators providing samples for this project. Topics included are a general description of the experiment, the irradiation experiment/capsule design, sample geometries, number of samples to be provided, documentation to be provided, a brief list potentially useful characterization methods that can be leveraged at INL, and other miscellaneous requirements specific to this project. This document is intended for informational use only.

innovative nuclear materials↗
DOE OSTI2022

Multi-Physics Investigation of a Natural Circulation Molten Salt Micro-Reactor that Utilizes an Experimental In-Pile Device to Improve Core Physics and System Thermal-Hydraulic Performance.

INL employee PhD Dissertation - The Molten Salt Reactor (MSR) concept is a rapidly evolving Generation IV design that has recently attracted favorable attention due to the potential for reducing waste generation, realizing passive safety features, and seizing on the opportunity for cost effective economics. This thesis investigates the performance benefit of a new device invented by the doctoral candidate. The device is referred to as a Wrapped Helix around an Inclined Plane or WHIP. The WHIP is protected under a provisional patent filed with the USA Patent Office on 28 September 2021 under application number 63/261,776, BEA docket number BA-1254. The WHIP device can be located in-core or near-core to promoted enhanced thermal-hydraulics and neutronics performance. While the WHIP can be employed in a variety of solid or liquid fuel designs, this thesis investigates the device’s benefit in the application of a natural-circulation, micro-molten salt nuclear battery concept (MsNB). This thesis will specifically investigate the temperature coefficient of the MsNB fuel (FLiNaK) using novel temperature sensitivity techniques unique to Serpent particle transport code, evaluate the WHIP’s thermal-hydraulic and neutronic performance effects using both established (STAR-CCM+) and novel (Python code developed to estimate the circulation effective delayed neutron fraction, ßef f ) analytical and numerical methods, evaluate the neutron noise behavior of the MsNB and how the WHIP may alter the character of the MsNB’s transfer function, and how the WHIP affects the autonomous, load-following performance under transient power conditions using Python code developed by the candidate. The sum body of this work, in part, has been published in three journal articles as the timing of the provisional patent process has allowed. Results show that creative utilization of WHIP engineering design and function reduces reactor system volume, fuel loading, control/stability in the buoyant regime

buoyant flow↗
DOE OSTI2025

Modeling Continuous Online Refueling with SCALE 6.3.1 and Serpent-2 in the EIRENE Novel Molten Salt Reactor Design

The Molten Salt Reactor (MSR) is a Generation IV advanced fission reactor design in which the coolant, and in some cases the fuel itself, is in the form of liquid molten alkali-halide salts with a fluoride or chloride ionic base. In liquid-fueled MSRs, reactor refueling may be performed online, where fresh fuel salt is added to the core during operation without the need for shutdown periods. Refueling for these reactor designs is typically modeled using either a batch or continuous refueling approach, both of which can be simulated with the SCALE 6.3.1 and Serpent-2 code systems. However, the specific manner in which they are implemented can vary depending upon the desired rate of refueling, whether the refueling rate is constant or variable, and consideration of salt drainage for systems where the in-core salt volume is kept constant. In a novel MSR fuel cycle concept termed the “Sourdough” fuel cycle, fuel salt is allowed to “grow” within the core, with excess salt either being transferred to an external holding tank to maintain a constant core volume or diverted to an upper plenum within the core to allow for volume growth. In this work, continuous refueling was modeled in a thermal-spectrum, LEU-fueled, small MSR design operating with the Sourdough fuel cycle using the SCALE 6.3.1 and Serpent-2 codes. Two different continuous refueling approaches were simulated, with the performance of each being compared to determine which is most suitable for use with the Sourdough fuel cycle concept.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Development of an In Situ Fission Gas Release Instrument for Fuel Sample Irradiations in the High Flux Isotope Reactor

Experimental measurement of gaseous fission product release with respect to temperature and burnup is a critical aspect of understanding nuclear fuel performance, validating predictive models, and qualifying new fuels. To measure this phenomenon in real-time, Oak Ridge National Laboratory has developed an instrument for measuring in situ fission gas release from small-scale fuel samples irradiated in the High Flux Isotope Reactor (HFIR). The instrument uses a continuous flow of Heover the fuel samples to sweep gaseous fission products from a sealed capsule in the HFIR Be reflector to an instrument station adjacent to the reactor. The instrument station houses two high-purity germanium (HPGe) detectors that measure decay gamma rays from fission products passing through a room temperature dwell chamber placed over the detector crystal. The sealed capsules in the reactor are designed to modulate fuel sample temperatures between 700 and 1,100°C by changing the Ar/He gas mixture surrounding the capsules during irradiation. N-type thermocouples are incorporated into the capsule housing to record real-time fuel temperatures. The capsules are heated primarily by prompt gamma rays emitted from the HFIR core with minimal heat contributions from fission in the fuel samples to minimize temperature gradients in the specimens for separate-effects characterization of the material. This paper describes modeling of time-dependent nuclear heating and fission product formation in fuel samples, thermal characteristics of the in-core capsules, and expected gaseous fission product gamma spectra at the HPGe instrument station.

Mulligan, Padhraic L [ORNL] (ORCID:000000025826540↗
DOE OSTI2024

Influence of Joule Heating on the Stability of High Temperature Irradiation Resistant Thermocouples

Development of in-core instrumentation is driven by the pursuit of safer and more economic energy production from both existing nuclear reactors and Generation IV reactor designs. Idaho National Laboratory (INL) has developed high temperature irradiation resistant thermocouples (HTIR-TCs) for temperature sensing inside Generation IV nuclear reactors. These thermocouples are composed of phosphorus-doped niobium (Nb-P) and lanthana-doped molybdenum (Mo-LaO) thermoelements, an alumina (Al2O3) insulation, and a niobium sheath. HTIR-TCs require an initial heat treatment exceeding the maximum service temperature to stabilize the generated electromotive force (EMF) signal; however, the mechanism behind this stabilization is not well understood. This work evaluates the impact of Joule heating on the thermoelements' microstructures, chemical stability, and mechanical properties to determine the mechanisms by which the EMF signal stabilization occurs. Accordingly, during the Joule heat treatment, a secondary Nb3P phase coarsened along the length of the Nb-P thermoelement, along with the formation of an interaction region at the Al2O3/niobium interface. The interaction between the alumina insulation and the Nb-P thermoelement was also observed within the Mo-LaO thermoelement. Joule heating induced stability within the generated EMF signal of HTIR-TCs through the formation of secondary phases within the Nb-P and the interaction of the alumina insulation with the thermoelements.

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗
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 OSTI2024

Influence of Joule Heating on the Stability of High Temperature Irradiation-resistant Thermocouples

Development of in-core instrumentation is driven by the pursuit of safer, more economic energy production from the perspective of both existing nuclear reactors and Generation IV reactor designs. Idaho National Laboratory has developed high-temperature irradiation-resistant thermocouples (HTIR-TCs) for temperature sensing inside Generation IV nuclear reactors. These thermocouples are composed of phosphorus-doped niobium (Nb-P) and lanthana-doped molybdenum (Mo-LaO) thermoelements, an alumina (Al2O3) insulation, and a niobium sheath. HTIR-TCs require an initial heat treatment exceeding the maximum service temperature to stabilize the generated electromotive force (EMF) signal. The mechanism behind the stabilization of the HTIR-TCs through traditional heat treatment methods is understood; however, the traditional heat treatment method is expensive, time consuming, and results in a heterogeneous microstructure. Therefore, we investigated a rapid method for stabilization and microstructure homogeneity, through Joule heating. This work evaluates the impact of Joule heating on the thermoelements’ microstructures, chemical stability, and mechanical properties so as to determine the mechanisms by which stabilization of the EMF signal occurs. Accordingly, during the Joule heat treatment, a secondary Nb3P phase coarsened along the length of the Nb-P thermoelement, along with the formation of an interaction region at the Al2O3/niobium interface. The interaction between the alumina insulation and the Nb-P thermoelement was also observed within the Mo-LaO thermoelement. Joule heating induced stability within the generated HTIR-TC EMF signal via the formation of secondary phases within the Nb-P and the interaction between the alumina insulation and the thermoelements.

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗