Evaluation of the Possible Susceptibility of Titanium Grade 7 to Hydrogen Embrittlement in a Geologic Repository Environment

2000 ◽  
Vol 663 ◽  
Author(s):  
Charles A. Greene ◽  
Alvin J. Henry ◽  
C. Sean Brossia ◽  
Tae M. Ahn

ABSTRACTTi 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.

Author(s):  
Robin Nazzaro ◽  
William Swick ◽  
Nancy Kintner-Meyer ◽  
Thomas Perry ◽  
Carole Blackwell ◽  
...  

The U.S. Department of Energy (DOE) oversees one of the largest cleanup programs in history—the treatment and disposal of 356,260 cubic meters of highly radioactive nuclear waste created as a result of the nation’s nuclear weapons program. This waste is currently stored at DOE sites in the states of Washington, Idaho, and South Carolina. In 2002, DOE began an accelerated cleanup initiative to reduce the estimated $105-billion cost and 70-year time frame required for the program. The U.S. General Accounting Office (GAO), an agency of the U.S. Congress, evaluated DOE’s high-level waste program to determine the status of the accelerated cleanup initiative, the legal and technical challenges DOE faces in implementing it, and any further opportunities to improve program management. GAO found that DOE’s initiative for reducing the cost and time required for cleaning up high-level waste is evolving. DOE’s main strategy continues to include concentrating much of the radioactivity into a smaller volume for disposal in a geologic repository. Under the accelerated initiative, DOE sites are evaluating other approaches, such as disposing of more of the waste on site or at other designated locations. DOE’s current savings estimate for these approaches is $29 billion, but the estimate is not based on a complete assessment of costs and benefits and has other computational limitations. For example, the savings estimate does not adequately reflect the timing of when savings will be realized, which distorts the actual amount of savings DOE may realize. DOE faces significant legal and technical challenges to realize these savings. A key legal challenge involves DOE’s authority to decide that some waste with relatively low concentrations of radioactivity can be disposed of on site. A recent court ruling against DOE is a major threat to DOE’s ability to meet its accelerated schedules. A key technical challenge is DOE’s approach for separating waste into high-level and low-activity portions. At the Hanford Site in Washington State, DOE is planning to implement such a method that will not be fully tested until the separations facility is constructed. This approach increases the risk and cost of schedule delays compared to fully testing an integrated pilot-scale facility. However, DOE believes the risks are manageable and that a pilot facility would unnecessarily delay waste treatment and disposal. DOE has opportunities to improve management of the high-level waste program. When it began the initiative to reduce costs and accelerate the high-level waste cleanup schedule, DOE acknowledged it had systematic problems with the way the program was managed. Although DOE has taken steps to improve program management, GAO has continuing concerns about management weaknesses in several areas. These include making key decisions without a sufficiently rigorous supporting analysis, incorporating technology before it is sufficiently tested, and pursuing a “fast-track” approach of simultaneous design and construction of complex nuclear facilities. DOE’s management actions have not fully addressed these weaknesses.


2019 ◽  
Vol 98 ◽  
pp. 10005
Author(s):  
Marek Pękala ◽  
Paul Wersin ◽  
Veerle Cloet ◽  
Nikitas Diomidis

Radioactive waste is planned to be disposed in a deep geological repository in the Opalinus Clay (OPA) rock formation in Switzerland. Cu coating of the steel disposal canister is considered as potential a measure to ensure complete waste containment of spent nuclear fuel (SF) and vitrified high-level waste (HLW) or a period of 100,000 years. Sulphide is a potential corroding agent to Cu under reducing redox conditions. Background dissolved sulphide concentrations in pristine OPA are low, likely controlled by equilibrium with pyrite. At such concentrations, sulphide-assisted corrosion of Cu would be negligible. However, the possibility exists that sulphate reducing bacteria (SRB) might thrive at discrete locations of the repository’s near-field. The activity of SRB might then lead to significantly higher dissolved sulphide concentrations. The objective of this work is to employ reactive transport calculations to evaluate sulphide fluxes in the near-field of the SF/HLW repository in the OPA. Cu canister corrosion due to sulphide fluxes is also simplistically evaluated.


Author(s):  
Robert E. Prince ◽  
Bradley W. Bowan

This paper describes actual experience applying a technology to achieve volume reduction while producing a stable waste form for low and intermediate level liquid (L/ILW) wastes, and the L/ILW fraction produced from pre-processing of high level wastes. The chief process addressed will be vitrification. The joule-heated ceramic melter vitrification process has been used successfully on a number of waste streams produced by the U.S. Department of Energy (DOE). This paper will address lessons learned in achieving dramatic improvements in process throughput, based on actual pilot and full-scale waste processing experience. Since 1991, Duratek, Inc., and its long-term research partner, the Vitreous State Laboratory of The Catholic University of America, have worked to continuously improve joule heated ceramic melter vitrification technology in support of waste stabilization and disposition in the United States. From 1993 to 1998, under contact to the DOE, the team designed, built, and operated a joule-heated melter (the DuraMelterTM) to process liquid mixed (hazardous/low activity) waste material at the Savannah River Site (SRS) in South Carolina. This melter produced 1,000,000 kilograms of vitrified waste, achieving a volume reduction of approximately 70 percent and ultimately producing a waste form that the U.S. Environmental Protection Agency (EPA) delisted for its hazardous classification. The team built upon its SRS M Area experience to produce state-of-the-art melter technology that will be used at the DOE’s Hanford site in Richland, Washington. Since 1998, the DuraMelterTM has been the reference vitrification technology for processing both the high level waste (HLW) and low activity waste (LAW) fractions of liquid HLW waste from the U.S. DOE’s Hanford site. Process innovations have doubled the throughput and enhanced the ability to handle problem constituents in LAW. This paper provides lessons learned from the operation and testing of two facilities that provide the technology for a vitrification system that will be used in the stabilization of the low level fraction of Hanford’s high level tank wastes.


1996 ◽  
Vol 465 ◽  
Author(s):  
T. H. Pigford ◽  
E. D. Zwahlen

ABSTRACTRecent proposals for a new U.S. standard for high-level waste disposal would limit the average dose to individuals in the vicinity surrounding a geologic repository. This would be a new approach to protecting the public from environmental releases of radioactivity. Heretofore, criteria adopted for geologic disposal have limited the reasonable maximum exposure to a future hypothetical individual. Here we present quantitative analyses of the relation between maximum exposure and vicinity-average exposure, resulting from future human use of ground water contaminated by radioactive releases from a repository.Estimating the vicinity-average exposure would require postulates and guesses of location and habits of future people. Exposure probabilities postulated by others show that proposed dose limit to the vicinity-average individual would be a far more lenient standard than the traditional dose limit to reasonably maximally exposed individuals. The proposed vicinity-average dose limit would allow far greater concentrations of contaminants in ground water than would be allowed by normal standards of ground water protection. A safety standard that limits vicinity-average exposure should also include limits on maximum exposure.


Author(s):  
Si Y. Lee

The engineering viability of disposal of aluminum-clad, aluminum-based spent nuclear fuel (Al-SNF) in a geologic repository requires a thermal analysis to provide the temperature history of the waste form. Calculated temperatures are used to demonstrate compliance with criteria for waste acceptance into the geologic disposal system and as input to assess the chemical and physical behavior of the waste form within the Waste Package (WP). The leading codisposal WP design proposes that a central DOE Al-SNF canister be surrounded by five Defense Waste Process Facility (DWPF) glass log canisters, that is, High-level Waste Glass Logs (HWGL’s), and placed into a WP in a geologic disposal system. A DOE SNF canister having about 0.4318m diameter is placed along the central horizontal axis of the WP. The five HWGL’s will be located around the peripheral region of the DOE SNF canister within the cylindrical WP container. The codisposal WP will be laid down horizontally in a drift repository. In this situation, two waste form options for Al-SNF disposition are considered using the codisposal WP design configurations. They are the direct Al-SNF form and the melt-dilute ingot. In the present work, the reference geologic and design conditions are assumed for the analysis even though the detailed package design is continuously evolved. This paper primarily dealt with the thermal performance internal to the codisposal WP for the qualification study of the WP containing Al-SNF. Thermal analysis methodology and decay heat source terms have been developed to calculate peak temperatures and temperature profiles of Al-SNF package in the DOE spent nuclear fuel canister within the geologic codisposal WP.


1998 ◽  
Vol 123 (1) ◽  
pp. 67-81 ◽  
Author(s):  
Patrick Jollivet ◽  
Michèle Nicolas ◽  
Etienne Vernaz

1983 ◽  
Vol 26 ◽  
Author(s):  
Jan Prij ◽  
Leo H. Vons

ABSTRACTResults are presented of in-situ measurements, performed in a 300 m deep dry-drilled borehole, in the ASSEmine. Convergence measurements at ambient as well as elevated temperatures and pressure measurements at elevated temperatures are discussed. Creep equations derived from these experiments are used for the numerical analysis of the time dependent behaviour of a salt dome with a HLW repository. The analyses show that the total stresses in the salt remain compressive with deviatoric components smaller than 3 MPa.


Author(s):  
R. Senger ◽  
J. Ewing

This study is part of a generic investigation for the assessment of the required minimum distance between a Spent Fuel/High-Level Waste/ Intermediate-Level Waste (SF/HWL/ILW) repository and a Low/ Intermediate-Level Waste (L/ILW) repository. For this, a large-scale numerical model was constructed to investigate the two-phase flow behavior for such a repository configuration in a low-permeability claystone formation. The modeling focused on the pressurization mechanisms associated with (a) resaturation of backfilled underground facilities, (b) thermal effects caused by heat generation from the SF/HLW canisters, and (c) gas generation from corrosion and degradation of different wastes in the L/ILW and ILW caverns and in the SF/HLW emplacement tunnels. The model accounts for gas generation from corrosion and degradation of both L/ILW and ILW wastes indicating decreasing rates with time, and from corrosion of the SF/HLW canisters characterized by a constant rate. Heat generation from radioactive decay of radionuclides of MOX/UO2 wastes is described by an exponential decay with time. The preceding operational phases of the different repository components were simulated representing the transient initial conditions for the post-closure phase. The simulated pressure buildup in the L/ILW repository shows a near linear increase between 10 and 4,000 years when the peak pressure of 6.5 MPa is reached for a repository at about 370 m bg. This is followed by a similar decline, recovering to near hydrostatic pressures after 1 million years. The SF/HLW repository (repository level 600 m bg) indicates a pressure rise between 100 and 1,000 years affected by the early thermal effects, followed by a steep increase between 3,000 and 100,000 years when the pressures level off to a maximum of 6.5 MPa after 160,000 years (corresponding to a steel corrosion rate of 1 μm/year). This is the time when all the metal is corroded and the gas generation stops resulting in a sudden decline, and the pressures level off to about 4.5 MPa in the SF/HLW emplacement tunnel after 1 million years. The numerical modeling demonstrates that the main pressurization mechanism is from gas generation in the different repository components. The pressure histories show a distinct separation of the pressure peaks between the L/ILW repository and the SF/HLW/ILW repository. Moreover, the thermal phenomena affect the pressures in the SF/HLW repository at early time only (prior to about 2,000 years). The thermal expansion of the pore water in the nearfield around the SF/HLW tunnels does produce a relatively steep pressure buildup after 100 years, but it dissipates rapidly prior to the main pressure buildup caused by the gas generation and gas accumulation in the SF/HLW repository. The thermally induced pressure buildup is restricted to the vicinity of the SF/HLW emplacement tunnels (decameter range) and thus, significant interference of the thermally induced pressure perturbation around the SF/HLW/ILW repository with the early gas pressure buildup in the L/ILW repository can be excluded.


1989 ◽  
Vol 176 ◽  
Author(s):  
B. P. McGrail ◽  
M. J. Apted ◽  
D. W. Engel ◽  
A. M. Liebetrau

ABSTRACTA mechanistic model describing a dynamic mass balance between the production and consumption of silicic acid was coupled to a near-field mass transport model to predict the dissolution kinetics of a high-level waste glass in a deep geologic repository. The effects of interactions between an iron overpack and the glass are described by a time-dependent precipitation reaction for a ferrous silicate mineral. The kinetic model is used to transform radionuclide concentration-versus-reaction progress values, predicted from a geochemical reaction path computer code, to concentration-versus-time values that are used to calculate the rate of radionuclide release by diffusive mass transfer to the surrounding host rock. The model provides for both solubility-limited and kinetically limited release; the rate-controlling mechanism is dependent on the predicted glass/groundwater chemistry.


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