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Parker, Stephen Scott

Publications and source records attributed to Parker, Stephen Scott.

12 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

NASA: Structure and Thermophysical Properties of Delta and Epsilon Zry4H x

Small-modular and micro-reactors are part of an emerging nuclear technology aimed to provide power to decentralized grids in civilian, industrial, and defense energy sectors. Resurgence in microreactor research is partly driven by demands for a mobile, safe, and compact energy platform for remote and strategic locations. The defining characteristics of the microreactor include its high-power density on a small footprint, its highly transportable design, and the use of high-assay low-enriched uranium (HALEU) fuel. Solid-state metal hydride neutron moderator components are considered for microreactor designs where the moderator functions to thermalize neutrons into lower energies for the fuel to achieve high fission rates. Zirconium hydride (ZrH x where x denotes the atomic hydrogen-to-metal ratio) is a leading moderator candidate material due to the combination of its low neutron absorption cross section, irradiation resistance, high hydrogen atomic density (upwards of 7.5 x 10 22 H atoms/cm 3 at low temperatures), and high hydrogen desorption (dissociation) temperature around 600 °C (873 K) compared to other metal hydrides. Its high-temperature stability enables operation of fuel temperatures into several hundred degrees Celsius for overall increased efficiency.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Thermal Elimination of Pyridine from a Uranium Trichloride Precursor

Renewed interest in advanced nuclear reactors, such as Molten Salt Reactors (MSRs), has spurred studies in actinide halide chemistry and property measurements. Several proposed research-scale and commercial MSR designs incorporate uranium trichloride (UCl 3 ) fuel. There are relatively few preparations for isolated actinide halides reported in the literature, therefore novel methods for the isolation of pure material are desired. This communication describes the thermal elimination of pyridine (py) from the coordination complex UCl 3 py 2 to yield gram-scale quantities of UCl 3 . The purity of the UCl 3 product was determined through powder X-ray Diffractometry (pXRD) and Elemental Analysis (EA).

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2023

In Situ High-Temperature Raman Spectroscopy of UCl 3 : A Combined Experimental and Theoretical Study

Uranium trichloride (UCl 3 ) has received growing interest for its use in uranium-fueled molten salt reactors and in the pyrochemical processing of used fuel. In this paper, we report for the first time the experimentally determined Raman spectra of UCl 3 , at both ambient condition and in situ high temperatures up to 871 K. The frequencies of five of the Raman-active vibrational modes (v i ) of UCl 3 exhibit a negative temperature derivative ((∂ν i /∂T) P ) with increasing temperature. Here, this red-shift behavior is likely due to the elongation of U–Cl bonds. The average isobaric mode Grüneisen parameter (γ iP = 0.91 ± 0.02) of UCl 3 was determined through use of the coefficient of thermal expansion published in Vogel et al. (2021) and the (∂ νi /∂T) P values determined in this study. These results are in general agreement with those calculated here by density functional theory (DFT+U). Finally, a comparison of the ambient band positions of UCl 3 to those of isostructural lanthanide (La–Eu) and actinide chlorides (Am–Cf) has been made.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2023

Application of Particle Accelerators to Mitigate Energy and Climate Change Problems Facing America

Reliable CO 2 -free baseload power is needed to address ever-increasing demands for electricity while minimizing adverse climate change. Diversified power supply provided by solar, wind, geothermal, and nuclear reactors can displace the use of fossil fuels. The United States (U.S.) is taking a new look at nuclear power as a source of electrical energy and the nuclear industry is proposing new approaches which may minimize capital costs. However, nuclear power comes with a variety of technical problems. Chief among those is managing the used fuel from nuclear reactors. No long-term, practicable solution to this problem is available. The lack of progress on a comprehensive waste management strategy restricts growth in the nuclear power industry and minimizes the role that nuclear energy may serve as part of a zero-carbon future (Bahr, 2021). According to the Nuclear Waste Policy Act of 1982, as amended, the U.S. Government has possession of the used reactor fuel and incurs large annual storage fees paid to utilities to store and safeguard the accumulated used fuel. In effect, short term on-site storage of nuclear waste is the current waste management plan. The amount of used reactor fuel in the U.S. is approximately 80,000 metric tons. While no geologic repository exists within the U.S., the potential site at Yucca Mountain, NV would accommodate 70,000 metric tons, which fails to meet current and future needs. Given the technical and political challenges associated with establishing a single geological storage site, it is necessary for the U.S. to implement technologies that improve the suitability of geological storage by reducing the volume and radiotoxicity of stored material. No extant technology meets this need. Accelerator driven waste burners have been proposed as a scalable method to process used nuclear fuel, however considerable technical challenges remain which impede commercialization. Innovations in design require investigation of novel materials, development of accurate models and simulations, and a comprehensive assessment of safety and performance. This proposal elaborates how LANL is uniquely positioned to make key contributions to this area of research.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Density Measurement of NaCl-MgCl 2 -PuCl 3 and NaCl-UCl 3 -PuCl 3 Molten Salt Systems by Neutron Radiographic Dilatometry

A novel neutron radiographic technique was developed and implemented to measure the change in volume as a function of temperature of select plutonium-containing liquid chloride mixtures from melt to 1250 K. These measurements were performed at Los Alamos National Laboratory by members of the molten salt research group under support from Gateway for Accelerated Innovation in Nuclear (GAIN) voucher NE-21-25117: “Density Measurements of Plutonium Bearing Salts via Neutron Beam Dilatometry”. This technical report contains a description of the experimental method and a summary of the results. Initial characterization of the {NaCl, MgCl 2 , UCl 3 , PuCl 3 } salts used in this experiment includes a combination of x-ray diffraction, differential scanning calorimetry, gamma ray spectroscopy, and inductively coupled plasma mass spectrometry. Four samples of NaCl + MgCl 2 + x mol% PuCl 3 and four samples of NaCl + UCl 3 + x mol% PuCl 3 were prepared at the Plutonium Facility, shipped to the Los Alamos Neutron Science Center (LANSCE), and radiographed with the density via neutron radiography (DvNR) apparatus. Density data in the form of meniscus heights were recorded as a function of temperature. The results show a linear dependence of density with temperature and a third order polynomial dependence of density with composition as a function of PuCl 3 addition in the corresponding halide salt mixture. Additions of PuCl 3 result in a significant increase in density of the salt mixture and a slight reduction in the coefficient of volumetric expansion. These results represent a first measurement of the density of plutonium chloride salt mixtures and, in combination with additional measurements, may lead to a generalized description of the density of such salt systems for arbitrary compositions and temperature.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Crystal Structure Evolution of UCl 3 from Room Temperature to Melting

We report uranium trichloride (UCl 3 ) is actively researched to develop and improve applications ranging from molten salt reactors to actinide processing, including spent fuel reprocessing. Here, we report for the first time the crystal structure evolution between room temperature and melting point from in situ high-temperature neutron diffraction to quantify, for example, the thermal expansion of the hexagonal a and c lattice parameters. The results are compared with density functional theory calculations. The melting point of UCl 3 is determined by differential scanning calorimetry to be 1108.2 ± 0.2 K.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗