Thad M. Adams

Effects of gamma radiation on electrochemical properties of ionic liquids

Abstract The electrochemical properties of ionic liquids (ILs) make them attractive for possible replacement of inorganic salts in high temperature molten salt electrochemical processing of nuclear fuel. To be a feasible replacement solvent, ILs need to be stable in moderate and high doses of radiation without adverse chemical and physical effects. Here, we exposed seven different ILs to a 1.2 MGy dose of gamma radiation to investigate their physical and chemical properties as they related to radiological stability. The azolium-based ILs experienced the greatest change in appearance, but these ILs were chemically more stable to gamma radiation than some of the other classes of ILs tested, due to the presence of aromatic electrons in the azolium ring. All the ILs exhibited a decrease in their conductivity and electrochemical window (at least 1.1 V), both of which could affect the utility of ILs in electrochemical processing. The concentration of the irradiation decomposition products was less than 3 mol. %, with no impurities detectable using NMR techniques.

Nanostructured metal oxides for anodes of Li-ion rechargeable batteries

The aligned nanorods of Co{sub 3}O{sub 4} and nanoporous hollow spheres (NHS) of SnO{sub 2} and Mn{sub 2}O{sub 3} were investigated as the anodes for Li-ion rechargeable batteries. The Co{sub 3}O{sub 4} nanorods demonstrated 1433 mAh/g reversible capacity. The NHS of SnO{sub 2} and Mn{sub 2}O{sub 3} delivered 400 mAh/g and 250 mAh/g capacities respectively in multiple galvonastatic discharge-charge cycles. It was found that high capacity of NHS of metal oxides is sustainable attributed to their unique structure that maintains material integrity during cycling. The nanostructured metal oxides exhibit great potential as the new anode materials for Li-ion rechargeable batteries with high energy density, low cost and inherent safety.

Tensile Testing of Carbon Steel in High Pressure Hydrogen

An infrastructure of new and existing pipelines and systems will be required to carry and to deliver hydrogen as an alternative energy source under the hydrogen economy. Carbon and low alloy steels of moderate strength are currently used in hydrogen delivery systems as well as in the existing natural gas systems. It is critical to understand the material response of these standard pipeline materials when they are subjected to pressurized hydrogen environments. The methods and results from a testing program to quantify hydrogen effects on mechanical properties of carbon steel pipeline and pipeline weld materials are provided. Tensile properties of one type of steel (A106 Grade B) in base metal, welded and heat affected zone conditions were tested at room temperature in air and high pressure (10.34 MPa or 1500 psig) hydrogen. A general reduction in the materials ability to plastically deform was noted in this material when specimens were tested in hydrogen. Furthermore, the primary mode of fracture was changed from ductile rupture in air to cleavage with secondary tearing in hydrogen. The mechanical test results will be applied in future analyses to evaluate service life of the pipelines. The results are also envisioned to be part of the bases for construction codes and structural integrity demonstrations for hydrogen service pipeline and vessels.

A Framework to Develop Flaw Acceptance Criteria for Structural Integrity Assessment of Multipurpose Canisters for Extended Storage of Used Nuclear Fuel

A multipurpose canister (MPC) made of austenitic stainless steel is loaded with used nuclear fuel assemblies and is part of the transfer cask system to move the fuel from the spent fuel pool to prepare for storage, and is part of the storage cask system for on-site dry storage. This weld-sealed canister is also expected to be part of the transportation package following storage. The canister may be subject to service-induced degradation especially if exposed to aggressive environments during possible very long-term storage period if the permanent repository is yet to be identified and readied. Stress corrosion cracking may be initiated on the canister surface in the welds or in the heat affected zone because the construction of MPC does not require heat treatment for stress relief. An acceptance criteria methodology is being developed for flaw disposition should the crack-like defects be detected by periodic Inservice Inspection. The external loading cases include thermal accident scenarios and cask drop conditions with the contribution from the welding residual stresses. The determination of acceptable flaw size is based on the procedure to evaluate flaw stability provided by American Petroleum Institute (API) 579 Fitness-for-Service (Second Edition). The material mechanical and fracture properties for base and weld metals and the stress analysis results are obtained from the open literature such as NUREG-1864. Subcritical crack growth from stress corrosion cracking (SCC), and its impact on inspection intervals and acceptance criteria, is not addressed.Copyright

Fracture Property Testing of Carbon Steel in High Pressure Hydrogen

An infrastructure of new and existing pipelines and systems will be required to carry and to deliver hydrogen as an alternative energy source to meet the energy demands of the future. Carbon and low alloy steels of moderate strength are currently used in hydrogen delivery systems as well as in the existing natural gas systems. It is critical to understand the material response of these standard pipeline materials when they are subjected to pressurized hydrogen environments. The methods and results from a testing program to quantify hydrogen effects on mechanical properties of carbon steel pipeline and pipeline weld materials are provided. Fracture toughness testing has been performed for one type of steel pipe material (A106 Grade B) in base metal, welded and heat affected zone conditions. C-shaped specimens were tested at room temperature in air and high pressure (102 ATM) hydrogen. A marked reduction in JQ was documented for both the base metal and HAZ metal tested in hydrogen. The results compliment a previous study on tensile properties of A106 Grade B material in high pressure hydrogen and are envisioned to be part of the basis for construction codes and structural integrity demonstrations of piping and pipelines for hydrogen service.Copyright

HYDROGEN EMBRITTLEMENT TESTING OF A ZIRCONIUM BASED ALLOY

Nuclear fuel rods in power reactors are typically clad with zirconium based alloys. These materials undergo corrosion and consequent hydriding during reactor operation. Due to limited storage space in the utilities’ spent fuel pool for used and discharged fuel, many of the utilities are implementing dry cask storage. There is a concern that the relatively high temperature drying process for dry storage coupled with the nascent hydrides and internal gas pressures, may promote radial hydride reorientation that is favorably oriented to promote clad cracking clad breach during storage, transport and handling. Samples were charged with 200 and 800 wppm gaseous hydrogen and subjected to a radial hydride growth treatment at three stresses. These samples were characterized and tested using a simple ring compression test. Ductile to brittle transition temperatures were determined for these samples.

The Development/Demonstration of Friction Stir Welding (FSW) for Closure of Radioactive Materials Containers in Support of Fully-Remote Operations and Extended Storage

Over the past 10 years at the Hanford Site, Fluor has successfully completed several radioactive materials packaging campaigns, all of which have included container closure welding. These campaigns utilized current, fusion welding processes and were performed on a semi-remote basis; that is, the Welding Operator had access to the weld joint to make repairs and equipment adjustments as needed, but performed the welding via cameras and remote video. Upcoming packaging activities will be performed under fully-remote operations (no human access to the weld joint), making weld repair difficult. Fluor believes a more robust joining process, one that can make high-quality, defect free-welds on a consistent basis, will be needed to successfully complete this work. In addition, improved long-term degradation properties, associated with container welds, may be required.Current United States (US) Used Nuclear Fuel (UNF) containers are made of austenitic stainless steel (S/S) and fabricated using fusion welding processes. Fusion welds in this material can be sensitive to environmental degradation, and in particular, Stress Corrosion Cracking (SCC), depending on conditions and length of service. With uncertainty surrounding the status of a US national repository and its impact on the disposal of UNF (significant delay), the nuclear industry is preparing for extended, on-site storage. Because of sensitivity to SCC and the need to consider extending container storage terms, there is concern regarding the performance of UNF container welds.In an effort to address these two issues, container weld quality (process robustness) and long-term corrosion performance, Fluor, along with the Pacific Northwest and Savannah River National Laboratories, are evaluating the use of Friction Stir Welding technology for the fabrication of UNF containers.Copyright

Production and Characterization of ZrC-UC Inert Matrix Composite Fuel for Gas Fast Reactors

Abstract Dispersed fuel composites consisting of uranium carbide particles (microspheres) in a zirconium carbide (inert) matrix were fabricated and characterized. Advanced fuels including refractory inert matrix fuels are being considered for gas fast reactors, which can accommodate a variety of feed materials including recycled transuranics that include minor actinides for incineration and high-level waste reduction. The particles for this effort were fabricated by employing a custom built rotating electrode machine. This process employed a uranium carbide electrode manufactured by combustion synthesis of uranium hydride and graphite powders. Two process parameters, namely, arc intensity and rotational speed, were varied to assess their effects on the size of the particles produced. The particles were characterized for microstructure, density, and composition (homogeneity). These particles were mixed with pure zirconium and graphite powders in different matrix to particle volumetric ratios of 90/10, 80/20, and 70/30 and inductively heated to 1850°C to initiate combustion synthesis to produce composites of zirconium carbide with the embedded uranium carbide particles. The aim was to limit process temperature and in particular process time, bearing in mind the possible future extensions of these processes to minor actinide-bearing fuels and also to avoid any changes in the structural integrity of the particles and large-scale diffusion of uranium into the matrix. The composites were characterized for microstructure, phase composition, density, and porosity distribution. The results are presented.

Tritium Sequestration in Gen IV NGNP Gas Stream via Proton Conducting Ceramic Pumps

This project is aimed at addressing issues related to tritium sequestration for effective utilization of heat in the Next-Generation Nuclear Plant (NGNP). Trace levels of tritium are present in the exhaust gas streams, posing a technical hurdle to using heat from the high-temperature exhaust, with an outlet temperature expected to be 850°C 900°C. This presents a significant challenge, since the removal of tritium from the high-temperature gas stream, usually accomplished at low temperatures through a tritium-permeable membrane or hydrogen isotope getters, must be accomplished at elevated temperatures in order to make use of this heat in downstream processing. The objective of this project is to experimentally evaluate alternative tritium separation technology based on ceramic separation membrane materials with high-temperature proton conductivity. The focus will be demonstrating the feasibility of tritium sequestration via high-temperature proton pumps, fabrication and performance evaluation of thin-film high-temperature nanocrystalline proton conductors for tritium separation, and development of mixed protonic and electronic conducting thin films for high-temperature tritium permeation systems.