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Snead, L. L.

Publications and source records attributed to Snead, L. L..

18 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Overview of the fusion nuclear science facility, a credible break-in step on the path to fusion energy

The Fusion Nuclear Science Facility (FNSF) is examined here as part of a two step program from ITER to commercial power plants. This first step is considered mandatory to establish the materials and component database in the real fusion in-service environment before proceeding to larger electricity producing facilities. The FNSF can be shown to make tremendous advances beyond ITER, toward a power plant, particularly in plasma duration and fusion nuclear environment. A moderate FNSF is studied in detail, which does not generate net electricity, but does reach the power plant blanket operating temperatures. The full poloidal Dual Coolant Lead Lithium (DCLL) blanket is chosen, with alternates being the Helium Cooled Lead Lithium (HCLL) and Helium Cooled Ceramic Breeder/Pebble Bed (HCCB/PB). Several power plant relevant choices are made in order to follow the philosophy of targeted technologies. Any fusion core component must be qualified by fusion relevant neutron testing and highly integrated non-nuclear testing before it can be installed on the FNSF in order to avoid the high probability of constant failures in a plasma-vacuum system. A range of missions for the FNSF, or any fusion nuclear facility on the path toward fusion power plants, are established and characterized by several metrics. A conservative physics strategy is pursued to accommodate the transition to ultra-long plasma pulses, and parameters are chosen to represent the power plant regime to the extent possible. An operating space is identified, and from this, one point is chosen for further detailed analysis, with R = 4.8 m, a = 1.2 m, IP = 7.9 MA, BT = 7.5 T, βN Gr = 0.9, fBS = 0.52, q95 = 6.0, H98 ∼1.0, and Q = 4.0. The operating space is shown to be robust to parameter variations. A program is established for the FNSF to show how the missions for the facility are met, with a He/H, a DD and 5 DT phases. The facility requires ∼25 years to complete its DT operation, including 7.8 years of neutron production, and the remaining spent on inspections and maintenance. The DD phase is critical to establish the ultra-long plasma pulse lengths. The blanket testing strategy is examined, and shows that many sectors have penetrations for heating and current drive (H/CD), diagnostics, or Test Blanket Modules (TBMs). The hot cell is a critical facility element in order for the FNSF to perform its function of developing the in-service material and component database. The pre-FNSF R&D is laid out in terms of priority topics, with the FNSF phases driving the time-lines for R&D completion. A series of detailed technical assessments of the FNSF operating point are reported in this issue, showing the credibility of such a step, and more detailed emphasis on R&D items to pursue. These include nuclear analysis, thermo-mechanics and thermal-hydraulics, liquid metal thermal hydraulics, transient thermo-mechanics, tritium analysis, maintenance assessment, magnet specification and analysis, materials assessments, core and scrape-off layer (SOL)/divertor plasma examinations.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

Response of unalloyed tungsten to mixed spectrum neutrons

Neutron irradiation in mixed spectrum reactors is an effective approach to investigate the synergistic effects of solid transmutations and atomic displacement damage in tungsten. In this article the findings and interpretations from a large-scale tungsten irradiation campaign conducted using the High Flux Isotope Reactor of Oak Ridge National Laboratory are summarized. The response of originally-unalloyed tungsten to mixed spectrum neutron irradiation is characterized by enormous hardening, degradation of modulus of toughness eventually leading to embrittlement, and decreased thermal conductivity. These property changes may be attributed primarily to production, segregation, and precipitation of rhenium and osmium and assisted by displacement damage. The stages describing the evolution of the microstructural development and property changes with the neutron dose and the accompanying accumulation of transmutation products are proposed.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Recent progress in the development of SiC composites for nuclear fusion applications

Silicon carbide (SiC) fiber reinforced SiC matrix composites continue to undergo development for fusion applications worldwide because of inherent advantages of the material including low activation, high temperature capability, relatively low neutron absorption, and radiation resistance. This work presents an international overview of recent achievements in SiC-based composites for fusion applications. Key subjects include applications in fusion reactors, high-dose radiation effects, transmutation effects, material lifetime assessment, and development of joining technology (processing, test method development, irradiation resistance, and modeling capability). This paper also discusses synergy among research for fusion materials and non-fusion materials (for fission and aerospace applications). Finally, future research directions and opportunities are proposed.

36 MATERIALS SCIENCE↗
DOE OSTI2017

The Fusion Nuclear Science Facility, the Critical Step in the Pathway to Fusion Energy

The proposed Fusion Nuclear Science Facility (FNSF) represents the first facility to enter the complex fusion nuclear regime, and its technical mission and attributes are being developed. The FNSF represents one part of the fusion energy development pathway to the first commercial power plant with other major components being the pre-FNSF research and development, research in parallel with the FNSF, pre-DEMO research and development, and the demonstration power plant (DEMO). Finally, the Fusion Energy Systems Studies group is developing the technical basis for the FNSF in order to provide a better understanding of the demands on the fusion plasma and fusion nuclear science programs.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Irradiation effects in tungsten-copper laminate composite

Tungsten-copper laminate composite has shown promise as a structural plasma-facing component as compared to tungsten rod or plate. The present study evaluated the tungsten-copper composite after irradiation in the High Flux Isotope Reactor (HFIR) at temperatures of 410-780°C and fast neutron fluences of 0.02-9.0×1025 n/m2, E>0.1 MeV, 0.0039-1.76 displacements per atom (dpa) in tungsten. Tensile tests were performed on the composites, and the fracture surfaces were analyzed with scanning electron microscopy. Before irradiation, the tungsten layers had brittle cleavage failure, but the overall composite had 15.5% elongation at 22°C. After only 0.0039 dpa this was reduced to 7.7% elongation, and no ductility was observed after 0.2 dpa at all irradiation temperatures when tensile tested at 22°C. For elevated temperature tensile tests after irradiation, the composite only had ductile failure at temperatures where the tungsten was delaminating or ductile.

Garrison, L. M.↗
DOE OSTI2016

Irradiation - High heat flux synergism in silicon carbide-based fuel claddings for light water reactors

Silicon carbide-based fuel cladding for light water reactors is anticipated to develop a unique stress state as a combined result from the high radial heat flux and irradiation-induced swelling assisted by thermal conductivity decrease. In order to verify the multi-physics finite-element analysis of this unique stress state, a novel experiment capturing the synergism of a high radiation field and a high radial heat flux in tubular silicon carbide specimens had been designed and implemented in the High Flux Isotope Reactor, Oak Ridge National Laboratory. In this experiment, small diameter tube test specimens made of monolithic or composite silicon carbide were irradiated to a dose of ∼2x10{sup 25} n/m{sup 2} (E > 0.1 MeV) under a radial heat flux of ∼0.6 MW/m{sup 2} while the outer surface temperature was maintained at a target temperature of about 300 Celsius degrees achieving a steep temperature gradient through the cladding wall thickness. The primary objective of the present paper is to discuss the technical planning and execution of the experiment. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Production of LEU fully ceramic microencapsulated fuel for irradiation testing

The Advanced Fuels Campaign of the US DOE Fuel Cycle Research/Development program is currently engaged in production of FCM fuel for irradiation testing in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). FCM fuel production is a serial process with 3 major steps namely kernel, TRISO, and compact production. The first step in the kernel process is production of air-dried UO{sub 3} gel spheres with suspended carbon particles. These spheres, roughly 2 mm in size, are then calcined to produce a mixture of UO{sub 2} and carbon. This calcination step may be performed in either argon, hydrogen, or mixtures of these gases. Once the calcination step is complete, temperature is increased (>1400 Celsius degrees) to convert this mixture to UO{sub 2} + UC{sub x}. Once dense oxide-carbide kernels are available, the inert environment is switched to a nitrogen containing one to facilitate carbothermic reduction of kernels and production of UN. Essentially, the oxygen and carbon in the oxide-carbide kernels react with one another to produce CO and are replaced with nitrogen inside the lattice. The TRISO step in fuel production is incorporation of a porous carbon buffer layer, a dense inner pyrolytic carbon layer (IPyC), a SiC layer, and finally and outer pyrolytic carbon layer (OPyC) onto the UN kernels. The coating layer deposition step is carried out inside a graphite furnace using a fluidized bed chemical vapor deposition (CVD) process. In essence, a mixture of reactant and inert gasses flow from the bottom through a bed of particles that are in turn fluidized by the gas stream. The particles undergo CVD and once a different coating layer is to be incorporated, the furnace temperature is adjusted and the reactant gases are changed. The final step in the series of fuel production activities is the consolidation of UN TRISO particles inside the SiC matrix to produce cylindrical pellets. It entails 3 critical areas: SiC powder mixing, over-coating of particles, and high-temperature densification. Optimal execution of the powder mixing step is vital to achieve a fully dense SiC matrix at the end of the process. The SiC matrix is consolidated using a special form of liquid phase sintering (LPS) involving nano-powder feedstock that is designated as the NITE process. The SiC nano-powder slurry is sprayed directly onto the particle surface using a piezoelectric nozzle. Once the over-coated particles are available, final consolidation step at high temperatures (1875 Celsius degrees) is carried out.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Thermo mechanical analysis of fully ceramic microencapsulated fuel during in-pile operation

A proposed fuel for improved performance in LWRs (Light Water Reactors) is FCM (fully encapsulated micro-ceramic) fuel concept that involves TRISO (Tri-structural-Isotropic) fuel particles embedded in a NITE (Nano Infiltrated Eutectic) silicon carbide matrix. TRISO fuel particles contain a spherical fuel kernel that will likely range from about 500 to in excess of 800 μm in diameter. The kernel and buffer layer are then typically coated with 3 isotropic layers consisting of a dense inner pyrolytic carbon (IPyC), a silicon carbide (SiC) layer, and an outer layer of dense, pyrolytic carbon (OPyC). These layers are typically around 40 μm thick, but can range from 20 to 60 μm. The TRISO particle packing fraction in the NITE-SiC matrix is expected to be 40-45 vol.%. The release of radioactivity into the coolant is dependent on the integrity of the silicon carbide layer of the TRISO particles and the NITE-SiC matrix. BISON, a code under development by Idaho National Laboratory, has been built on the Multi-physics Object Oriented Simulation Environment (MOOSE). MOOSE is a massively parallel, finite element computational system that uses a Jacobian-free, Newton-Krylov (JFNK) method to solve coupled systems and non-linear partial differential equations. A single particle comparison to fuel performance codes such as PARFUME was conducted and demonstrated that BISON is able to simulate the required physics. FCM pellets embedded with homogenous particles were simulated. It was found that the interior matrix temperature was predicted to be considerably lower than for an equivalent UO{sub 2} pellet. The swelling dependence on temperature created a compressive region around the perimeter of the pellet. However, at fluences above 6.10{sup 25} neutrons/m{sup 2}, the swelling of the pyrolytic carbon might cause significant tensile stresses and possible cracks near the surface of the pellet. This suggests that the pyrolytic carbon swelling may be an important factor in the lifetime of this fuel type.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Material selection for accident tolerant fuel cladding

Alternative cladding materials are being investigated for accident tolerance, which can be defined as >100X improvement (compared to current Zr-based alloys) in oxidation resistance in steam environments at ≥1200°C for short (≤4 h) times. After reviewing a wide range of candidates, current steam oxidation testing is being conducted on Mo, MAX phases and FeCrAl alloys. Recently reported low mass losses for Mo in steam at 800°C could not be reproduced. Both FeCrAl and MAX phase Ti 2 AlC form a protective alumina scale in steam. Therefore, commercial Ti 2 AlC that is not single phase, formed a much thicker oxide at 1200°C in steam and significant TiO 2 , and therefore may be challenging to use as a cladding or a coating. Alloy development for FeCrAl is seeking to maintain its steam oxidation resistance to 1475°C, while reducing its Cr content to minimize susceptibility to irradiation assisted Cr-rich α’ formation. The composition effects and critical limits to retaining protective scale formation at >1400°C are still being evaluated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗