Evaluation of the Possible Susceptibility of Titanium Grade 7 to Hydrogen Embrittlement in a Geologic Repository Environ
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Evaluation of the Possible Susceptibility of Titanium Grade 7 to Hydrogen Embrittlement in a Geologic Repository Environment Charles A. Greene1, Alvin J. Henry2, C. Sean Brossia3 and Tae M. Ahn1 1 US Nuclear Regulatory Commission, Mail Stop T-7C6, Washington, DC 20555-0001, USA 2 Previously at US Nuclear Regulatory Commission, MPR Associates, 320 King Street, Alexandria, VA 22314, USA 3 Center for Nuclear Waste Regulatory Analyses, Southwest Research Institute, 6220 Culebra Road, San Antonio, TX 78238, USA ABSTRACT Ti grade 7 has been selected by the U.S. DOE as the current material of choice for the drip shield in the proposed high level waste (HLW) repository design. Due to the addition of Pd, Ti grade 7 exhibits enhanced resistance to hydrogen embrittlement (HE), yet there is relatively little data on HE of this material. Calculations of hydrogen absorption/recombination, solubility, and free energy of hydride formation in Ti and Pd are presented to qualitatively evaluate Keff, the stress intensity factor for crack propagation induced by hydride formation, of Ti grade 7 in relation to other Ti alloys without Pd. Calculations were performed that show concentration of hydrogen in Ti grade 7 may exceed the critical hydrogen concentration, Hc, where the material becomes embrittled, when accelerated passive dissolution of Ti grade 7 in concentrated Cl- and Cl-+F- solutions as the source of hydrogen is considered. INTRODUCTION Under the Nuclear Waste Policy Act of 1982 as amended (NWPA, Public Law 97-425) the US Department of Energy (DOE) was directed to study Yucca Mountain, Nevada as the sole site for a National high level radioactive waste (HLW) repository. The US Nuclear Regulatory Commission (NRC) is to review the DOE’s evaluation of the proposed geologic repository site to ensure protection of public health and safety through compliance with proposed draft final rule 10 Code of Federal Regulations Part 63 in accordance with NWPA. The compliance period in the draft rule is 10,000 years during which time doses to an average member of the critical group from the repository shall not exceed 25 mrem/yr. The DOE proposed repository design includes a free-standing, 15 mm thick, Ti grade 7 (0.12-0.25 wt% Pd) barrier placed over the waste packages (WP, disposal container) containing HLW [1]. The drip shield (DS) is intended to divert water that may seep into the repository disposal cavern (drift) away from the WP. The DS thus potentially prolongs the WP lifetime and delays the release of radionuclides making the DS a component of the repository safety design significant to risk [2]. The composition of Ti grade 7 (Ti 7) is presented in Table 1. In general, Ti alloys exhibit excellent corrosion resistance under many conditions as a result of a protective TiO2 passive film [3,4]. Possible degradation mechanisms of the Ti 7 DS in the repository environment are slow general corrosion, rapid or accelerated general corrosion due to the presence of chloride or chloride plus fluoride [5], hydrogen embrittlement, and mechanical deg
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