Role of hydrogen chloride in hot salt stress corrosion cracking of titanium aluminum alloys

Hot salt stress corrosion cracking of titanium aluminum alloys in presence of anhydrous hydrogen chlorid

Stress corrosion cracking of titanium alloys progress report, apr. 1 - jun. 30, 1964

Hot salt stress corrosion cracking in titanium alloys - chloride corrosion role determination using chlorine isotopes and relation between crack morphology and alloy structur

Retention and release of tritium in aluminum clad, Al-Li alloys

Tritium retention in and release from aluminum clad, aluminum-lithium alloys is modeled from experimental and operational data developed during the thirty plus years of tritium production at the Savannah River Site. The model assumes that tritium atoms, formed by the {sup 6}Li(n,{alpha}){sup 3}He reaction, are produced in solid solution in the Al-Li alloy. Because of the low solubility of hydrogen isotopes in aluminum alloys, the irradiated Al-Li rapidly becomes supersaturated in tritium. Newly produced tritium atoms are trapped by lithium atoms to form a lithium tritide. The effective tritium pressure required for trap or tritide stability is the equilibrium decomposition pressure of tritium over a lithium tritide-aluminum mixture. The temperature dependence of tritium release is determined by the permeability of the cladding to tritium and the local equilibrium at the trap sites. This model is used to calculate tritium release from aluminum clad, aluminum-lithium alloys. 9 refs., 3 figs

Stress corrosion cracking of titanium alloys Fourth quarterly progress report, 1 Jan. - 31 Mar. 1965

Stress corrosion rupturing of titanium alloy - fracture mechanic

Transitioning aluminum clad spent fuels from wet to interim dry storage

The United States Department of Energy (DOE) currently owns several hundred metric tons of aluminum clad, spent nuclear fuel and target assemblies. The vast majority of these irradiated assemblies are currently stored in water basins that were designed and operated for short term fuel cooling prior to fuel reprocessing. Recent DOE decisions to severely limit the reprocessing option have significantly lengthened the time of storage, thus increasing the tendency for corrosion induced degradation of the fuel cladding and the underlying core material. The portent of continued corrosion, coupled with the age of existing wet storage facilities and the cost of continuing basin operations, including necessary upgrades to meet current facility standards, may force the DOE to transition these wet stored, aluminum clad spent fuels to interim dry storage. The facilities for interim dry storage have not been developed, partially because fuel storage requirements and specifications for acceptable fuel forms are lacking. In spite of the lack of both facilities and specifications, current plans are to dry store fuels for approximately 40 to 60 years or until firm decisions are developed for final fuel disposition. The transition of the aluminum clad fuels from wet to interim dry storage will require a sequence of drying and canning operations which will include selected fuel preparations such as vacuum drying and conditioning of the storage atmosphere. Laboratory experiments and review of the available literature have demonstrated that successful interim dry storage may also require the use of fuel and canister cleaning or rinsing techniques that preclude, or at least minimize, the potential for the accumulation of chloride and other potentially deleterious ions in the dry storage environment. This paper summarizes an evaluation of the impact of fuel transitioning techniques on the potential for corrosion induced degradation of fuel forms during interim dry storage

Helium embrittlement model and program plan for weldability of ITER materials

This report presents a refined model of how helium embrittles irradiated stainless steel during welding. The model was developed based on experimental observations drawn from experience at the Savannah River Site and from an extensive literature search. The model shows how helium content, stress, and temperature interact to produce embrittlement. The model takes into account defect structure, time, and gradients in stress, temperature and composition. The report also proposes an experimental program based on the refined helium embrittlement model. A parametric study of the effect of initial defect density on the resulting helium bubble distribution and weldability of tritium aged material is proposed to demonstrate the roll that defects play in embrittlement. This study should include samples charged using vastly different aging times to obtain equivalent helium contents. Additionally, studies to establish the minimal sample thickness and size are needed for extrapolation to real structural materials. The results of these studies should provide a technical basis for the use of tritium aged materials to predict the weldability of irradiated structures. Use of tritium charged and aged material would provide a cost effective approach to developing weld repair techniques for ITER components

Recycle of radiologically contaminated austenitic stainless steels

The United States Department of Energy owns large quantities of radiologically contaminated austenitic stainless steel which could by recycled for reuse if appropriate release standards were in place. Unfortunately, current policy places the formulation of a release standard for USA industry years, if not decades, away. The Westinghouse Savannah River Company, Idaho National Engineering Laboratory and various university and industrial partners are participating in initiative to recycle previously contaminated austenitic stainless steels into containers for the storage and disposal of radioactive wastes. This paper describes laboratory scale experiments which demonstrated the decontamination and remelt of stainless steel which had been contaminated with radionuclides

An Evaluation of the Potential for Creep of 3013 Inner Can Lids

This report provides the technical basis to conclude that creep induced deformation of Type 304L austenitic stainless steel can lids on inner 3013 containers will be insignificant unless the temperature of storage exceeds 400 C. This conclusion is based on experimental literature data for Types 304 and 316 stainless steel and on a phenomenological evaluation of potential creep processes