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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 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

On the flow-stress model of Preston, Tonks, and Wallace - Observations, parameter constraints, and implementations (Rev. 1)

A detailed discussion about the PTW flow-stress model has been presented, including theoretical considerations, model calibrations to experimental data, and model implementations in LANL ASC hydro-codes. The main findings are: In order to reproduce the intended behavior of the PTW model at high strain rates, the following conditions on the model parameters need to be valid: y 1 ≥ s 0 and y 2 ≥ β ; When the conditions above are not satisfied, PTW and PTW_Mod1 have slightly different behavior in the phonon-drag regime, and should be treated as different models. PTW exhibits a discontinuity above the critical strain rate. PTW_Mod1 does not show such a discontinuity but implies a different strain-rate dependence at high strain rates. Examples of this behavior include older calibrations for stainless steel and depleted uranium; In the thermal-activation regime, some flexibility was introduced in the original PTW model to allow for different rate-dependent behaviors at intermediate strain rates. Example: copper. However, many material calibrations do not require this extra flexibility, and the complexity and ambiguity of the intended behavior can be partially blamed for the inconsistent model implementations across ASC codes; The limit for the material constant p going to zero is well defined and leads to a simplified flow-stress equation, discontinuity free; Modern calibrations of the PTW parameters are done using Impala. Constraints on the PTW parameters are enforced; Implementation of the PTW model in LANL ASC hydro-codes suffered of several shortcom ings. At the time of writing, FLAG and xRAGE contain consistent and correct implemen tations of the PTW and PTW_Mod1 models. Corresponding changes to Pagosa have been suggested.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Operation of the Fast Neutron Coincidence Collar (FNCL) with a DD-Neutron Generator

For more than 30 years, the quantitative assay of the 235 U content of light water reactor fresh fuel assemblies relied on measuring coincidence neutrons from fissions induced by an Am(Li) neutron source using 3 He based detectors. The Fast Neutron Collar (FNCL) developed by the International Atomic Energy Agency (IAEA), replaces traditional 3 He proportional counters with an array of liquid scintillator detectors arranged about the fuel assembly to provide improved measurement precision and reduced sensitivity to gadolinium poison rods. The FNCL relies on Am(Li) neutron sources that are no longer commercially available. This work examines the replacement of Am(Li) sources with a commercial off the-shelf deuterium–deuterium (DD) neutron generator. In addition to mitigating supply concerns, the neutron generator offers advantages in measurement precision and potential automation of sequential passive/active neutron measurements. This report presents the initial performance results for both the integrated DD/FNCL and Am(Li)/FNCL assays of compact depleted uranium, low-enriched uranium, and highly enriched uranium standards along with an estimate of the expected performance for fresh fuel assemblies. A discussion of the design and operation of the “FNCL Analysis and Simulation Software” is also provided.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Demonstration of In Situ Down Blending During ZIRCEX Process

The Material Recovery Pilot Plant has successfully demonstrated in situ down-blending of HEU fuel to support INL’s commitment to HALEU production technologies. Depleted uranium pellets, along with an HEU fuel piece were placed in the hydrochlorinator and subsequently declad and oxidized. UOx product was recovered, analyzed, and demonstrated to have reduced enrichment within the targeted range of 20 – 30 % U-235. This first attempt at in situ down-blending will be followed by continued process refinement over the remaining fuel operations at MRPP.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Density functional theory (DFT) study of UF 6 hydrolysis: reaction pathways, spectroscopy, and chemical kinetics

Depleted uranium hexafluoride (UF 6 ), a stockpiled byproduct of the nuclear fuel cycle, reacts readily with atmospheric humidity, but the gas-phase reaction mechanism and associated chemical kinetics are poorly understood. During the performance period we undertook development of a state-of-the-art ab initio gas-phase chemical kinetics simulation workflow to model the hydrolysis of uranium hexafluroride (UF 6 ). In doing so, we addressed several outstanding issues in the theoretical treatment of uranium-containing systems. At the outset it was unclear how to generate accurate estimates of kinetic and thermodynamic data for U-containing chemical reactions. Generation of such data has been made routine. Prior to our work, the literature associated with UF 6 hydrolysis were disparate and inaccurate. This body of work provides a modern and comprehensive theoretical assessment of the reaction mechanism, molecular clustering towards deposition, and chemical kinetics. New methodological implementations and software integrations resulting from this work are also highlighted. As much as possible, our predictions were validated against experimental data including particle morphologies, vibrational spectroscopy, atomization enthalpies, and kinetic rate constants. Nevertheless, we were unable to reconcile kinetic measurements with high-accuracy simulations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Investigations into the Ternary NaF-KF-UF4 Salt System – Phase A

A knowledge gap exists in the data and understanding of fresh fuel salt and irradiated multicomponent fuel salt systems thermophysical properties. Quantifying these properties is necessary for the design and construction of test reactors, as well as the licensing of future commercial molten-salt reactors. To facilitate thermal property determination on a proposed fuel salt composition for Seaborg Technologies, several samples containing depleted uranium tetrafluoride (UF4), sodium fluoride (NaF), and potassium fluoride (KF) were blended, and a melt temperature analysis was performed. From the melting temperature analysis, it was determined that sample Seaborg-7, a ternary salt composition of 26.4UF4-24.7KF-48.9NaF (mol%), was very near a ternary eutectic point. Therefore, thermal properties such as melting temperature, salt stability, density, heat capacity, thermal diffusivity, and viscosity were experimentally determined on the Seaborg-7 salt. These measurements document the baseline properties of fresh fuel salt as a function of temperature, where future experiments on irradiated fuel salt will provide a holistic perspective on the change of thermophysical properties during reactor operations. Several precision instruments were used to collect property data, and instrument calibrations and data collection were performed and documented in a standardized and reproducible manner with meticulous detail. This process ensured that the measurement procedures and resulting data can readily be duplicated elsewhere. The Seaborg-7 salt was shown to be stable at temperatures up to 900°C, as no mass change was observed upon repeated heating and cooling. The peak melting temperature was determined to be 547°C (557°C endset). The enthalpy of fusion (??H?_fus^o) was determined to be 167.5 ± 2.7 J/g while the enthalpy of crystallization (??H?_c^o) was determined to be -147.8 ± 13.3 J/g. In addition to the eutectic melting peak, upon heating, several pre eutectic peaks were observed, occurring at 470°C (onset) and 499°C (peak). Specific heat capacity measurements showed a slightly increasing trend with respect to temperature in the solid phase, while the liquid-specific heat capacity showed a somewhat flat trend with an average value of 106.1 ± 1.24 J/mol·K between 600 to 800°C. Three independent trials using the Seaborg-7 salt determined the density to be ?(T) = 4.908 – 0.000363·T(°C), validated between 32 to 200°C, and ?(T) = 4.808 – 0.00113·T(°C), validated between ~575 to 850°C. Thermal diffusivity was determined for the liquid state and is represented by the linear equation y = 0.1581 + 0.000207·T(°C) between 550 to 850°C. The viscosity was determined from 600 to 800°C and is represented by the exponential fit equation, ? (mPa·s) = 736.58e^(-0.006·T(°C)). This report documents the conclusion of fuel salt thermophysical property measurements for the Seaborg SPP, Phase A project.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Rescue of Unique Liquid IP-2 Portable Tanks at Savannah River Site

Introduction • Savanah River Site procured six large IP-2 portable tanks capable of transporting liquids in 2003 to support layup of F Canyon at SRS. • Tanks are being rescued and recertified to support eventual disposition of up to 171,000 L of contaminated solvent solutions from H Canyon at SRS and to potentially transport Depleted Uranium solutions to support the ABD Program.

Fee, Nathaniel C. [Savannah River Nuclear Solution↗
DOE OSTI2026

HIGHER EFFICIENCY FISSION COUNTER FOR NEUTRON DETECTION

Fission chambers containing 235U have been used for detection of neutrons for many decades. In these chambers 235U is electroplated on a metal which is one plate of an ionization chamber. Since the range of fission products from induced fission in the chamber is extremely short, these deposits from electroplating are so thin that the amount of 235U results in low neutron detection efficiency and the chamber needs high internal 235U surface area to increase efficiency. A typical ~8 in. long, ~2 in. diameter fission counter only contains < 2 g of 235U total, resulting in very low detection efficiency per incident neutron impinging on the outer surface of the detector. However, because of their large electronic pulses from fission products compared to gamma rays it is very easy to discriminate against gamma rays impinging on the detector. These types of chambers are produced commercially and are used worldwide in nuclear research and nuclear power reactors. ORNL has made a higher efficiency fission ionization chamber by using many thin concentric annuli to increase the area. A higher efficiency fission counter would be useful and can be made by incorporating 235U into a glass or ceramic scintillator. For nearly 100 years the green color in antique glass was produced by incorporation of depleted uranium (up to 20 wt. %) into a glass. The concept for this scintillation-optical-based fission detector is to dope a scintillator (either ceramic or glass) with 235U. Glass scintillators are commercially available doped with as much as 8 wt. % 6Li for neutron detection. In the 235U doped scintillator the fission product pulses would be double the size of those in an ion chamber since all fission products produce light in the scintillator whereas in the ionization chamber half do not go into the ionizing gas volume. These higher efficiency fission chambers would be very useful for a wide variety of research, nuclear research, and nuclear power reactor applications. Because of their smaller size and weight, they should be useful for space application such as monitoring of space nuclear power reactors. This report reviews existing fission chamber technology and suggests the development of these small high-efficiency fission chambers. Other fission isotopes could be used for other research applications.

Mihalczo, John [ORNL] (ORCID:0000000254292681)↗
DOE OSTI2026

ZPRD Database: ZPPR-15 Monte Carlo Results

The ZPPR-15 experiments [1] were mockups of a 330 MWe Integral Fast Reactor (IFR). The ZPPR-15 assembly consisted of a clean, two zone, approximately circular core surrounded by a thin depleted uranium (DU) blanket with sodium (Na) cooling and a thick stainless steel reflector. The ZPPR-15 program was conducted in four phases: A, B, C, and D. Each phase was marked by a particular composition of the reference assembly, with the last three being representative of the three stages of the IFR fuel cycle. This report documents all MCNP [2] runs of the ZPPR-15 loadings that were included in the ZPRD database on GitLab. In the present work, eigenvalues and associated standard deviations were computed for each ZPPR-15 loading with the use of the three data libraries (ENDF/B-VII.0 [3], ENDF/B-VII.1 [4] and ENDF/B-VIII.0 [5]). Relevant results from the obtained outputs are also discussed in this report with two main objectives: a) provide and evaluate updated calculated values with respect to previous reports, notably Refs. [6] and [7] that were based on the use of the ENDF/B-VII.0 library. b) address any change in the observed reactivity effects relevant to the analysis of the experimental data when a different data library is used (previously reported results were mostly based on the use of ENDF/B-VII.0 data only).

Aliberti, Gerardo↗
NASA NTRS1972

The design of a source to simulate the gamma-ray spectrum emitted by a radioisotope thermoelectric generator

A simulated source was designed to duplicate the gamma spectrum of a uniform cylindrical 2200-watt Pu02 radioisotope thermoelectric generator containing 81% Pu-238 and 1.2 ppm Pu-236. Gamma rays from the decay of Pu-238, Am-241, Pu-239, and the 0-18(alpha,n)Ne-21 reaction were catalogued in broad energy groups. Two 46- and one 22-mc Th-228 sources provided simulation at various times in the life of the fuel capsule up to 18 years, which covers the time span of an outer planet mission. Emission from Th-228 represents the overwhelming contribution of the gamma spectrum after the first few years. The sources, in the form of 13-inch rods, were placed in a concentric hole in a cylinder of depleted uranium, which provided shielding equivalent to the self-shielding of the fuel capsule. The thickness of the U-238 cylinder (0.55cm) was determined by Monte Carlo calculations to insure that the spectrum emerging from the simulated source matched that of the fuel capsule.

Reier, M.↗
NASA NTRS1971

Shock tube technique for measuring opacities at high pressures for gaseous-core nuclear rockets.

A new method of opacity determination is described that uses an improved shock tube design and is applicable to measurements of Planck and Rosseland mean opacities in hydrogen or seeded hydrogen at hydrogen partial pressures of 500 atm, at temperatures from 2300 to 8000 K. The selection of seed materials to be added to the gases to make them semiopaque is discussed. Currently favored seeds are silicon and carbon or graphite for reactive seeds, and depleted uranium and tungsten 184 for nonreactive seeds.

Patch, R. W.↗
NASA NTRS1973

Induction simulation of gas core nuclear engine

The design, construction and operation of an induction heated plasma device known as a combined principles simulator is discussed. This device incorporates the major design features of the gas core nuclear rocket engine such as solid feed, propellant seeding, propellant injection through the walls, and a transpiration cooled, choked flow nozzle. Both argon and nitrogen were used as propellant simulating material, and sodium was used for fuel simulating material. In addition, a number of experiments were conducted utilizing depleted uranium as the fuel. The test program revealed that satisfactory operation of this device can be accomplished over a range of operating conditions and provided additional data to confirm the validity of the gas core concept.

Poole, J. W.↗
NASA NTRS1973

Shield materials recommended for space power nuclear reactors

Lithium hydride is recommended for neutron attenuation and depleted uranium is recommended for gamma ray attenuation. For minimum shield weights these materials must be arranged in alternate layers to attenuate the secondary gamma rays efficiently. In the regions of the shield near the reactor, where excessive fissioning occurs in the uranium, a tungsten alloy is used instead. Alloys of uranium such as either the U-0.5Ti or U-8Mo are available to accommodate structural requirements. The zone-cooled casting process is recommended for lithium hydride fabrication. Internal honeycomb reinforcement to control cracks in the lithium hydride is recommended.

Kaszubinski, L. J.↗
NASA NTRS1986

Pb isotopic evidence for early Archaean crust in South Greenland

The results of an isotopic remote sensing study focussed on delineating the extent of Early Archean crust north and south of the Nuuk area and in south Greenland is presented. Contamination of the Late Archean Nuk gneisses and equivalents by unradiogenic Pb uniquely characteristic of Amitsoq gneiss was detected as far south as Sermilik about 70 km south of Nuuk and only as far north as the mouth of Godthabsfjord. This study was extended to the southern part of the Archean craton and the adjoining Early Proterozoic Ketilidian orogenic belt where the Pb isotopes suggest several episodes of reworking of older uranium depleted continental crust. The technique of using the Pb isotope character of younger felsic rocks, in this case Late Archean and Early Proterozoic gneisses and granites to sense the age and isotopic character of older components, is a particularly powerful tool for reconstructing the evolutionary growth and development of continental crust.

Taylor, P. N.↗
NASA NTRS2017

Nuclear Rocket Ceramic Metal Fuel Fabrication Using Tungsten Powder Coating and Spark Plasma Sintering

Nuclear thermal propulsion is an enabling technology for crewed Mars missions. An investigation was conducted to evaluate spark plasma sintering (SPS) as a method to produce tungsten-depleted uranium dioxide (W-dUO2) fuel material when employing fuel particles that were tungsten powder coated. Ceramic metal fuel wafers were produced from a blend of W-60vol% dUO2 powder that was sintered via SPS. The maximum sintering temperatures were varied from 1,600 to 1,850 C while applying a 50-MPa axial load. Wafers exhibited high density (>95% of theoretical) and a uniform microstructure (fuel particles uniformly dispersed throughout tungsten matrix).

Barnes, M. W.↗
NASA NTRS2017

Kilopower: Small and Affordable Fission Power Systems for Space

The Nuclear Systems Kilopower Project was initiated by NASA's Space Technology Mission Directorate Game Changing Development Program in fiscal year 2015 to demonstrate subsystem-level technology readiness of small space fission power in a relevant environment (Technology Readiness Level 5) for space science and human exploration power needs. The Nuclear Systems Kilopower Project centerpiece is the Kilopower Reactor Using Stirling Technology (KRUSTY) test, which consists of the development and testing of a fission ground technology demonstrator of a 1 kWe-class fission power system. The technologies to be developed and validated by KRUSTY are extensible to space fission power systems from 1 to 10 kWe, which can enable higher power future potential deep space science missions, as well as modular surface fission power systems for exploration. The Kilopower Project is cofounded by NASA and the Department of Energy National Nuclear Security Administration (NNSA).KRUSTY include the reactor core, heat pipes to transfer the heat from the core to the power conversion system, and the power conversion system. Los Alamos National Laboratory leads the design of the reactor, and the Y-12 National Security Complex is fabricating it. NASA Glenn Research Center (GRC) has designed, built, and demonstrated the balance of plant heat transfer and power conversion portions of the KRUSTY experiment. NASA MSFC developed an electrical reactor simulator for non-nuclear testing, and the design of the reflector and shielding for nuclear testing. In 2016, an electrically heated non-fissionable Depleted Uranium (DU) core was tested at GRC in a configuration identical to the planned nuclear test. Once the reactor core has been fabricated and shipped to the Device Assembly Facility at the NNSAs Nevada National Security Site, the KRUSTY nuclear experiment will be assembled and tested. Completion of the KRUSTY experiment will validate the readiness of 1 to 10 kWe space fission technology for NASAs future requirements for sunlight-independent space power. An early opportunity for demonstration of In-Situ Resource Utilization (ISRU) capability on the surface of Mars is currently being considered for 2026 launch. Since a space fission system is the leading option for power generation for the first Mars human outpost, a smaller version of a planetary surface fission power system could be built to power the ISRU demonstration and ensure its end-to-end validity. Planning is underway to start the hardware development of this subscale flight demonstrator in 2018.

Space nuclear power↗
NASA NTRS2021

Electrolytic Co-deposition Neutron Production Evaluation

Co-deposition electrochemical cells are a simple means to examine novel nuclear reactions. In this study, palladium and deuterium atoms were co-deposited on a cathode at stoichiometric densities, forming dendritic morphologies. Bubble detector neutron dosimeters were used to measure equivalent dose levels during electrolytic deposition. Standard configuration cells expected to produce excess neutrons were denoted as experimental cells and contained an electrolyte consisting of palladium(II) chloride, lithium chloride, and heavy water. The control cells used copper(II) chloride, lithium chloride, and heavy water electrolyte. Experimental and control cells were supplied current, increasing from 0.1 to 100.0 mA over a period of 20 days. For test days 9 through 20, the standard configuration experimental cells exhibited significantly higher average neutron radiation than the controls at a 99% confidence level. Alternative configurations involving more increasing current over a shortened time scale, substitutions of water in place of heavy water or potassium chloride for lithium chloride, as well as inclusion of depleted uranium near the cell cathodes all resulted in inconclusive evidence for promoting neutron production.

neutron↗
DOE OSTI2023

Statistical Approaches for Pebble Bed Reactor Operations and Safeguards

The design of pebble bed reactors (PBRs) and their method of operation align more closely with statistical approaches used in manufacturing and process control than traditional safeguards statistical approaches. The reason is PBRs will employ a nondestructive assay (NDA) measurement (burnup measurement system [BUMS]) that is part of the fuel handling system supporting discharge decisions in addition to the reactor code which monitor performance. The integration of these two approaches provides the opportunity to monitor reactor performance statistically for both operations and safeguards in ways not achievable using other reactor designs. For light water reactors (LWRs), knowledge about reactor code performance in predicting irradiated special nuclear material (SNM) content historically was only achieved from special measurement campaigns or from fuel reprocessing. Conversely, through statistical comparison of the BUMS with the reactor code-predicted values, PBRs can achieve this in real time. The resulting SNM distribution is an indicator of reactor performance because factors such as transit time and path of the pebble fuel through the reactor determine the plutonium production and uranium depletion. By analyzing the predicted and measured values, opportunities exist to adjust operating parameters, fuel design, and other characteristics to optimize performance and fuel utilization. From a safeguards perspective, this approach also provides the information necessary to validate declared values and evaluate whether the reactor is being operated as expected. This paper outlines statistical approaches for PBRs that can be used to support both operations and safeguards.

Ball, Cory↗