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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 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
NASA NTRS1990

Draft Environmental Impact Statement for the Ulysses Mission (Tier 2)

This Draft Environmental Impact Statement (DEIS) addresses the environmental impacts which may be caused by the preparation and operation of the Ulysses spacecraft, including its planned launch on the Space Transportation System (STS) Shuttle and the alternative of canceling further work on the mission. The launch configuration will use the STS/Inertial Upper Stage (IUS)/Payload Assist Module-Special(PAM-S) combination. The Tier 1 EIS included a delay alternative which considered the Titan 4 launch vehicle as an alternative booster stage for launch in 1991 or later. However, the U.S. Air Force, which procures the Titan 4 for NASA, could not provide a Titan 4 vehicle for the 1991 launch opportunity because of high priority Department of Defense requirements. The only expected environmental effects of the proposed action are associated with normal Shuttle launch operations. These impacts are limited largely to the near-field at the launch pad, except for temporary stratospheric ozone effects during launch and occasional sonic boom effects near the landing site. These effects have been judged insufficient to preclude Shuttle launches. In the event of (1) an accident during launch, or (2) reentry of the spacecraft from earth orbit, there are potential adverse health and environmental effects associated with the possible release of plutonium dioxide from the spacecraft's radioisotope thermoelectric generators (RTG). The potential effects considered in this EIS include risks of air and water quality impacts, local land area contamination, adverse health and safety impacts, the disturbance of biotic resources, impacts on wetland areas or areas containing historical sites, and socioeconomic impacts. Intensive analysis of the possible accidents associated with the proposed action are underway and preliminary results indicate small health or environmental risks. The results of a Final Safety Analysis Report will be available for inclusion into the final EIS.

Source record↗
NASA NTRS1988

Draft environmental impact statement for the Galileo Mission (Tier 2)

This Draft Environmental Impact Statement (DEIS) addresses the environmental impacts which may be caused by the preparation and operation of the Galileo spacecraft, including its planned launch on the Space Transportation System (STS) Shuttle and the alternative of canceling further work on the mission. The launch configuration will use the STS/Inertial Upper Stage (IUS)/Payload Assist Module-Special (PAM-S) combination. The Tier 1 EIS included a delay alternative which considered the Titan 4 launch vehicle as an alternative booster stage for launch in 1991 or later. However, the U.S. Air Force, which procures the Titan 4 for NASA, could not provide a Titan 4 vehicle for the 1991 launch opportunity because of high priority Department of Defense requirements. The only expected environmental effects of the proposed action are associated with normal Shuttle launch operations. These impacts are limited largely to the near-field at the launch pad, except for temporary stratospheric ozone effects during launch and occasional sonic boom effects near the landing site. These effects have been judged insufficient to preclude Shuttle launches. In the event of: (1) an accident during launch, or (2) reentry of the spacecraft from earth orbit, there are potential adverse health and environmental effects associated with the possible release of plutonium dioxide from the spacecraft's radioisotope thermoelectric generators (RTG).

Source record↗
NASA NTRS1997

Feasibility Investigation for Performing Fireball Temperature Tests

NASA Johnson Space Center White Sands Test Facility (WSTF) was requested by the Launch Abort Subpanel and the Power Systems Subpanel of the Interagency Nuclear Safety Review Panel to investigate the feasibility of using spectroscopic techniques to measure propellant fireball gas temperatures. This report outlines the modeling and experimental approaches and results of this investigation. Descriptions of the theoretical particle temperature and mass effusion models are presented along with the results of the survivability of small plutonium dioxide (less than or equal to 1000 microns diameter) particles entrained in various propellant fireball scenarios. The experimental test systems used to measure the hydroxide radical, water, and particle graybody spectral emissions and absorptions are discussed. Spectral results along with temperatures extracted by analyzing the spectral features are presented for the flames investigated in the laboratory environment. Methods of implementing spectroscopic measurements for future testing using the WSTF Large-scale Hydrogen/Oxygen Explosion Facility are discussed, and the accuracy expected for these measurements is estimated from laboratory measurements.

Tapphorn, Ralph M.↗
NASA NTRS2001

Assessment of the Free-piston Stirling Convertor as a Long Life Power Convertor for Space

There is currently a renewed interest in the use of free-piston Stirling power convertors for space power applications. More specifically, the Stirling convertor is being developed to be part of the Stirling Radioisotope Power System to supply electric power to spacecraft for NASA deep space science missions. The current development effort involves the Department of Energy, Germantown, MD, the NASA Glenn Research Center, Cleveland, OH, and the Stirling Technology Company, Kennewick, WA. The Stirling convertor will absorb heat supplied from the decay of plutonium dioxide contained in the General Purpose Heat Source modules and convert it into electricity to power the spacecraft. For many years the "potentials" of the free-piston Stirling convertor have been publicized by it's developers. Among these "potentials" were long life and high reliability. This paper will present an overview of the critical areas that enable long life of the free-piston Stirling power convertor, and present some of the techniques that have been used when long life has been achieved.

Schreiber, Jeffrey G.↗
DOE OSTI2018

Advanced Waste Management Concepts for Uranium and Plutonium Multi-Recycling - 18533

The multiple recycling of plutonium and uranium will play a major role in gradually achieving a fully closed nuclear fuel cycle in the future, once the first Generation IV fast reactors (FR) are deployed in the NPP fleet. To reach this objective, optimizations of current processes and technologies and, in some cases, the development of innovative concepts are required to meet the scientific, technical and economic challenges imposed by the treatment of spent fuels with an increasingly higher Pu content. Based on the significant past experience acquired by France in LWR and FR MOX fuel treatment, the CEA is running R and D studies to develop new efficient processes for Pu recycling and for FR MOX fuel fabrication. More compact and economically competitive processes, ensuring the safety/security of the facilities (particularly as regards criticality and proliferation resistance) have to be developed taking into account that the amounts of plutonium involved will be one order of magnitude higher than in UOX fuels. New concepts of head-end treatment, dissolution, and separation have been proposed. Consequently, effluent and waste management must be adapted to these processes and to the characteristics of fast reactor MOX fuels. New waste treatment processes are suggested: 1) Cladding waste and undissolved metallic particles could be treated simultaneously in a melting process. The molten metal could be decontaminated by a specific slag and the oxides recovered, in particular PuO{sub 2}, would be recycled at the frond-end of the process, 2) Fission Products and Actinides (FPA) containment glass will also be optimized to accommodate significant amounts of iron, nickel and chromium coming from hull corrosion, and / or decontamination slag coming from the metal melting process. 3) New peraluminous-type glass formulations are under development to accept increased amounts of FPA, as well as iron, nickel, and chromium. The first trials to evaluate the melting process have indicated that metallic compounds from the dissolution residues could be easily incorporated into a molten metal at levels of 1, 2 and 3 w%. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS1967

Magnesium-zinc reduction is effective in preparation of metals

Uranium, thorium, and plutonium are effectively prepared by magnesium-zinc reduction, using uranium oxides, thorium dioxide, and plutonium dioxide as starting materials. This technique is also useful in performing reduction of metals such as zirconium and titanium.

Knighton, J. B.↗
DOE OSTI2022

The electronic Raman scattering spectrum of PuO 2

Here the Raman spectrum of PuO 2 was measured up to 13,000 cm –1 with three different laser excitation wavelengths spanning the resonance (405 nm), near-resonance (457 nm), and preresonance (514 nm) energy range. Approximately 26 never-before-seen bands were observed between 3500 and 13,000 cm –1 . Given the very high energy of the Raman shifts of these bands and the relative insensitivity of their spectral position to the interrogating laser wavelength, they are believed to arise from an electronic origin. These bands are present in both freshly calcined and radiolytically aged PuO 2 , although a broad luminescence is observed in the aged material, which obscures many of the high frequency features. In situ laser annealing of the material attenuated this luminescence and allowed for clear observation of these never-before-seen spectral features. Discovery of these high-energy bands presents a new way of identifying PuO 2 for nuclear nonproliferation and forensics purposes.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2025

Characterization of PuO 2 With Visible and UV Raman Spectroscopy: Discrimination Between the Bulk, Surface, and an Intermediate Disordered Layer

Raman spectroscopy is an ideal tool in the characterization of materials including PuO 2 . The wavelength-dependent absorptivity of the material defines the light penetration depth and the relative Raman scattering contribution from the bulk and the surface. Here, the surface contribution to the total Raman scattering was investigated for PuO 2 calcined at various temperatures and recorded with laser wavelengths of 355, 325, and 244 nm. These experiments provided the first glimpse of the wavelength-dependent disappearance and emergence of new phonons and electronic bands from the PuO 2 surface layers. The first indication of the wavelength transition in the Raman spectra was the loss of the 2LO2 (overtone, ~1155 cm -1 ) band and the weakening intensity of the Г 1 → Γ 5 electronic band (~2135 cm -1 ) with the 355-nm excitation laser. The Γ 5 electronic band was barely visible with the 244-nm excitation. The electronic band located at ~1050 cm -1 , corresponding to the Г 1 → Γ 4 electronic transition was observed to dramatically increase in intensity while the Г 1 → Γ 3 electronic band (2640 cm -1 ) sharpened as the UV wavelength was increased in energy from the near- to deep-UV (355–325–244 nm). The FWHM of the T 2g band was found to vary with calcination temperature (450°C and 900°C) with the 325-nm laser and the 244-nm laser. The T 2g band attributes, the strong emergence of the Г 1 → Γ 4 electronic band, and the disappearance of the 2LO2 overtone acquired with the 244-nm excitation for the different calcination temperatures suggest a shallow penetration depth.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Nuclide inventory characterization of EBR-II MOX fuel test pins

Characterization of 3 mixed oxide (MOX) fuel pins from the Experimental Breeder Reactor II (EBR-II) SPA-2/-2B experiment program was performed in support of the Advanced Reactor Experiments for Sodium Fast Reactor Fuels (ARES) joint project between Idaho National Laboratory (INL) and the Japan Atomic Energy Agency (JAEA). The characterization efforts were performed to enable transient testing in the Transient Reactor Test (TREAT) facility at INL and will include post irradiation examination (PIE) and other analyses. MOX fuel pins UW02009, 02011, and 06024 were historically irradiated from 1989 to 1994 for a total of 822.42 effective full power days (EFPDs) and peak burnup of approximately 14.3 at.% (atomic percent of heavy metal atoms, ∼134.4 GWd/t). They had since been placed in storage until a future use was identified. Characterization of the fission products and activation of each fuel pin was necessary to enable removal from storage and subsequent irradiation testing or PIE activities. The initial fuel pin geometry and composition were established using historical documentation. Irradiation history details were utilized to compute irradiated and decayed fuel pin masses, atomic densities, and activities. A pellet-wise axial neutron flux was also computed to assist in evaluation of the axial distribution of {sup 235}U and {sup 239}Pu, per pellet, within each of the test fuel pins. The results computed herein supported experiment design for advanced high-burnup MOX fuel designs under slow transient overpower (MOXTOP) conditions, which was Phase I of the ARES project. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Model development and data uncertainty integration

The effect of data uncertainties is discussed, with the epithermal neutron multiplicity counter as an illustrative example. Simulation using MCNP6, cross section perturbations and correlations are addressed, along with the effect of the 240 Pu spontaneous fission neutron spectrum, the effect of P(ν) for 240 Pu spontaneous fission, and the effect of spontaneous fission and (α,n) intensity. The effect of nuclear data is the product of the initial uncertainty and the sensitivity -- both need to be estimated. In conclusion, a multi-parameter variation method has been demonstrated, the most significant parameters are the basic emission rates of spontaneous fission and (α,n) processes, and uncertainties and important data depend on the analysis technique chosen.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

Low temperature dissolution flowsheet for plutonium metal

The H-Canyon flowsheet used to dissolve Pu metal for PuO 2 production utilizes boiling HNO 3 . SRNL was requested to develop a complementary dissolution flowsheet at two reduced temperature ranges. The dissolution and H 2 generation rates of Pu metal were investigated using a dissolving solution at ambient temperature (20-30 °C) and for an intermediate temperature of 50-60 °C. Additionally, the testing included an investigation of the dissolution rates and characterization of the off-gas generated from the ambient temperature dissolution of carbon steel cans and the nylon bags that contain the Pu metal when charged to the dissolver.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Examination of Intergranular Crack on SC0234

The microstructure of the SC0234 weld was evaluated per DA-SPEC-01067, ''General Purpose Heat Source (GPHS) Plutonium-238 Dioxide Fueled Clad Product Specification'' and it was determined that the weld microstructure conformed to the quality requirements. During the examination of weld cross section, the metallographic analysis revealed the presence of a surface crack. Prior visual examination per RPS-TA55-303 had not revealed the crack, nor did the surface photography performed per RPS-TA55-302. The presence of the crack does not violate any of the product specifications called out on DA-SPEC-01067. Although there is a well-documented history of weld hot crack susceptibility in the early development of the DOP-26 iridium alloy, a surface crack such as this had not been documented during welding at Los Alamos. In the interest of better understanding the cause; the product engineer/program recommended further examination of the crack. This report discusses the examination of the crack found in the weld cross section of simulant clad SC0234. Metallography, Scanning Electron Microscopy (SEM) and Energy Dispersive Spectroscopy techniques (EDS) where used in the analysis of the crack and the surrounding weld microstructure.

36 MATERIALS SCIENCE↗
NASA NTRS2017

Milli-Watt Radioisotope Power to Enable Small, Long-Term Robotic “Probe” Space Exploration

Milli-watt Radioisotope Power Systems (RPS) based on Radioisotope Heater Units (RHUs) could be an ideal power source for certain spacecraft that cannot use solar power due to large distances from the sun, or other environmental constraints, and where they enable or significantly enhance the ability of a mission to meet its scientific or operational goals. Various modular, compact RHU-based thermoelectric (TE) generator concepts developed or derived from current NASA Small Business Innovation Research (SBIR) projects satisfying this need have been investigated. These modular, compact and low mass power systems could support small, highly-mobile robotic exploration packages, and could be incorporated into different robotic package concepts, spacecraft or satellites. Current modular RPS design concepts with 40mW, 80mW and 120mW power levels use RHUs and Bi2Te3 TE converters. Skutterudites materials could be used in the future if new higher thermal energy output and higher temperature miniature heat sources were developed, for example, using technologies currently in the General Purpose Heat Source (GPHS) used in higher electric power output RTGs. Small (a.k.a., “mice-like”) robotic packages could effectively utilize these RHU-driven power levels to accommodate crawling, climbing, monitoring, taking measurements, and communicating during long-term planetary missions aimed at gathering environmental and geologic data (i.e., over multiple decades). Waste heat from the cold side of the TE converter could also be directed toward the electronics and / or energy storage (e.g. batteries) to keep them within design temperature ranges. In addition to power generation and electronics / battery heating, the RHU / TE configuration could be designed to survive an external 500°C bake out procedure for critical spacecraft sterilization, environmental certification and planetary protection. Analytical studies have been performed to optimize various design configurations for power, mass, volume and robotic mobility. Specific power (mW/kg) and volumetric specific power (mW/cm3) characteristics of various design configurations will be presented and key conceptual design tradeoffs will be discussed. Hot- and cold-side thermal interfaces required to meet power and mass goals and associated design sensitivities will also be discussed. RHU / TE systems must overcome critical design challenges to survive high-g loadings in some robotic applications and we will examine the mass impacts required to satisfy various dynamic loading environments up to 10,000 g’s. Power can be generated for a minimum of 30 years or more using plutonium-238 dioxide heat sources (given that Pu-238 has an 87.7 year half-life) with some reduction in power as the heat source naturally degrades.

Bahrami, Poyan↗
DOE OSTI2022

Unoccupied electronic structure of actinide dioxides

Ligand field density functional theory calculations of the dioxides of thorium, uranium, and plutonium have been combined with high-energy-resolution fluorescence detection (HERFD) in x-ray absorption spectroscopy and inverse photoelectron spectroscopy (IPES) measurements to provide powerful insight into the underlying composition of the unoccupied 5f electronic structure in these 5f localized systems. Fine structure in the 5f 5/2 transitions in HERFD can be directly correlated with the fine structure in the leading edge of the IPES. The shapes, intensities, and systematics in HERFD and IPES are explained in a consistent and rigorous fashion in terms of the j-specific 5f electronic structure.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Defining the Shape Density of Cerium Oxide and Titanium Dioxide Aerosol

The “shape density” of CeO 2 (cerium oxide) powder (American Elements, Los Angeles CA) is estimated to be 0.82 ± 0.02 g cm -3 and the shape density for TiO 2 (titanium dioxide) powder (American Elements, Los Angeles CA) is estimated as 0.54 ± 0.01 g cm -3 . The “shape density” is defined as the measured loose density (ASTM 2018 and ASTM 2021) of a powder divided by the ICRP (International Commission on Radiological Protection) default aerosol shape factor (Valetin 2002). This information is used for experimental measurements of the RRF (respirable release factor) and other respirable source term parameter (US DOE 1994) experiments at the Los Alamos National Laboratory (LANL). The shape density is the input parameter to the TSI Inc (Shoreview MN) AIM (Aerosol Instrument Manager) software for the TSI model 3321 APS (Aerodynamic Particle Sizer) (Figure 1).

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2023

Dose reconstruction for plutonium-239 intakes at the Rocky Flats Plant

Purpose: The Rocky Flats (RF) Plant was a weapons manufacturing facility that operated from the early 1950s to 1989. Its primary missions were the production of plutonium (Pu) pits for thermonuclear weapons and the processing of retired weapons for Pu recovery. The purpose of this study was to estimate radiation doses to a cohort of 4499 RF workers from an intake of 239 Pu, the primary plutonium isotope handled at the site. Materials and methods: The latest biokinetic models of the International Commission on Radiological Protection, or site-specific variations of those models, were used to estimate 239 Pu intakes for each worker based on model fits to bioassay data often coupled with lung measurements. Results: Urinary excretion and lung retention data for most 239 Pu intakes could be fit reasonably well by a mixture of Pu dioxide and moderately soluble material. For some workers, better fits were obtained by application of other absorption types including Type S, 239 Pu nitrate, or pure 239 Pu dioxide, or by assuming intake via a wound. The lungs typically received the highest tissue doses, with fifty-year committed equivalent doses in the range of 0.5–1 Sv for 275 workers, 1–5 Sv for 115 workers, and greater than 5 Sv for 12 workers. Conclusions: RF was a unique site regarding a large number of lung measurements available for determining the appropriate absorption types for inhaled material. Further, this provided higher confidence in reconstructed 239 Pu doses than is generally gained from urinary data alone.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2018

Cocrystallization of Uranyl and Plutonyl Nitrate Hexahydrates

In the system UO{sub 2}(NO{sub 3}){sub 2}–PuO{sub 2}(NO{sub 3}){sub 2}–HNO{sub 3}–H{sub 2}O with the ratio Pu/(U + Pu) = 0.3–0.5 and stabilization of the pair U(VI)–Pu(VI), the crystallization of (U, Pu)O{sub 2}(NO{sub 3}){sub 2}·6H{sub 2}O occurs in accordance with the metric of the phase diagram of the system UO{sub 2}(NO{sub 3}){sub 2}–HNO{sub 3}–H{sub 2}O with coefficient of cocrystallization of plutonium 0.804 ± 0.028. This condition holds in the sector of the working lines of the crystallization process that correspond to the minimum content of uranium in the mother solution. The degree of enrichment of the mother solution is determined by the crystallization temperature, i.e., the mass fraction of the mother solution in the total mass of the system. The constancy of the coefficient of cocrystallization with the technologically important parameters of the process makes it possible to use the metric of the solubility diagram of the system UO{sub 2}(NO{sub 3}){sub 2}–HNO{sub 3}–H{sub 2}O to calculate the crystallization of (U, Pu)O{sub 2}(NO{sub 3}){sub 2}·6H{sub 2}O with the crystalline phase with the required composition being obtained.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Quantification of calculation accuracy for code systems in burn-up credit applications by recalculations of experimental data - 14480

For application in criticality safety, numerical code systems require profound validation against experimental data. In terms of burn-up credit, this refers to the inventory determination as well as the subsequent criticality calculation. Within this framework, a dedicated, conservative estimation of uncertainties, bias and bias uncertainties for the resulting calculated multiplication factor k{sub eff} is indispensable. However, measures to reduce excess conservatism may become desirable, depending on the application case. For this reason GRS implemented a methodology to validate a calculation code system for criticality safety analyses for systems with irradiated nuclear fuel. The work at hand demonstrates this methodology by means of the GRS home-built inventory determination system KENOREST and ORNL's well-known SCALE 6 CSASS criticality safety calculation sequence. A number of 40 radiochemical analysis samples taken from SFCOMPO and elsewhere have been calculated to derive isotopic correction factors in different ways, depending on the required level of conservatism and the calculational efforts to be carried out. This includes bounding methods as well as different Monte Carlo sampling approaches. Furthermore, from the ICSBEP a number of 515 critical benchmark experiment configurations comprising water moderated low enriched UO{sub 2} fuel rods and water moderated mixed oxide (MOX) fuel rods have been selected and evaluated. Important absorber and structure materials are also covered. Various trending and sensitivity and uncertainty analysis methods have been applied, again to derive conservative estimates on the various contributions to k{sub eff} uncertainties. The overall methodology is applicable to low enriched, water moderated PWR uranium dioxide fuels in a burn-up range of about 10 to 60 GWd/tHM, and validates the reactivity contributions of the major uranium and plutonium isotopes. The methodology itself has no limitation which requires the exclusion of further isotopes as e.g. fission products. However, its application is constrained by the experimental database being publicly available. After the potential provision of an extended database in the future, the range of applicability can easily be extended to more nuclides. In addition, it is not limited to the code systems being used here but can also be applied using different numerical tools. For a generic PWR spent fuel pool rack and different enrichment and burn-up combinations, the impact of the current level of validation on k{sub eff} is demonstrated and quantified. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗