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Title: Investigation of environmentally assisted fracture of metallic nuclear waste package barrier materials in simulated basalt repository environments

Abstract

Statically loaded corrosion tests, slow strain rate (SSR) tests, and fatigue crack growth rate (FCGR) tests were conducted to evaluate the relative susceptibility of two titanium-base nuclear waste package candidate structural barrier materials Ti-grade 2 and Ti-grade 12-to environmentally enhanced cracking in a simulated repository environment. Statically loaded corrosion tests were done in oxic basalt ground water at 250/sup 0/C; SSR tests were done in oxic basalt ground water at 150, 250, and 300/sup 0/C and in air at 20 and 250/sup 0/C; and FCGR tests were done in basalt ground water, fluoride-ion-enhanced basalt ground water, high-purity water, and air at 90/sup 0/C. The following conclusions can be drawn: the general corrosion rate of statically loaded corrosion coupons was very low in a 3-mo test, and no pitting or cracking of the specimens was observed. Ti-grade 2 and Ti-grade 12 exhibited strain rate dependent ductility diminution in SSR tests. The ductility diminution was most severe in Ti-grade 2 at 300/sup 0/C and in Ti-grade 12 at 250/sup 0/C. For of Ti-grade 12 it was found to be highly orientation dependent. The ductility diminution was also found in tests conducted in air as well as in those conducted in the basaltmore » ground water environment; however, the extent of the degradation was less in air. The ductility diminution cannot be attributed to stress corrosion cracking because the fracture mode was microvoid coalescence in all tests. Evidence obtained in the current study and correlation of the present results with results obtained by other researchers indicate that dynamic strain aging is responsible for the loss of ductility. The FCGR of Ti-grade 2 and Ti-grade 12 was not affected by any of the environmental conditions used in this study, which indicates that no environmental cracking mechanism is operative under the conditions tested (90/sup 0/C, oxic ground water, and frequencies from 0.01 to 5 Hz).« less

Authors:
Publication Date:
Research Org.:
Pacific Northwest Lab., Richland, WA (USA)
OSTI Identifier:
6601741
Alternate Identifier(s):
OSTI ID: 6601741; Legacy ID: DE83005126
Report Number(s):
PNL-4379
ON: DE83005126
DOE Contract Number:
AC06-76RL01830
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
36 MATERIALS SCIENCE; 12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE WASTES FROM NUCLEAR FACILITIES; GROUND WATER; CORROSIVE EFFECTS; RADIOACTIVE WASTE DISPOSAL; GEOLOGIC DEPOSITS; HIGH-LEVEL RADIOACTIVE WASTES; TITANIUM BASE ALLOYS; CORROSION; CRACK PROPAGATION; FATIGUE; STRAIN RATE; BASALT; ENGINEERED SAFETY SYSTEMS; EXPERIMENTAL DATA; MATERIALS TESTING; PACKAGING; ALLOYS; CHEMICAL REACTIONS; DATA; HYDROGEN COMPOUNDS; IGNEOUS ROCKS; INFORMATION; MANAGEMENT; MATERIALS; MECHANICAL PROPERTIES; NUMERICAL DATA; OXYGEN COMPOUNDS; RADIOACTIVE MATERIALS; RADIOACTIVE WASTES; ROCKS; TESTING; TITANIUM ALLOYS; VOLCANIC ROCKS; WASTE DISPOSAL; WASTE MANAGEMENT; WASTES; WATER 360103* -- Metals & Alloys-- Mechanical Properties; 360105 -- Metals & Alloys-- Corrosion & Erosion; 052002 -- Nuclear Fuels-- Waste Disposal & Storage

Citation Formats

Pitman, S.G. Investigation of environmentally assisted fracture of metallic nuclear waste package barrier materials in simulated basalt repository environments. United States: N. p., 1982. Web. doi:10.2172/6601741.
Pitman, S.G. Investigation of environmentally assisted fracture of metallic nuclear waste package barrier materials in simulated basalt repository environments. United States. doi:10.2172/6601741.
Pitman, S.G. Mon . "Investigation of environmentally assisted fracture of metallic nuclear waste package barrier materials in simulated basalt repository environments". United States. doi:10.2172/6601741. https://www.osti.gov/servlets/purl/6601741.
@article{osti_6601741,
title = {Investigation of environmentally assisted fracture of metallic nuclear waste package barrier materials in simulated basalt repository environments},
author = {Pitman, S.G.},
abstractNote = {Statically loaded corrosion tests, slow strain rate (SSR) tests, and fatigue crack growth rate (FCGR) tests were conducted to evaluate the relative susceptibility of two titanium-base nuclear waste package candidate structural barrier materials Ti-grade 2 and Ti-grade 12-to environmentally enhanced cracking in a simulated repository environment. Statically loaded corrosion tests were done in oxic basalt ground water at 250/sup 0/C; SSR tests were done in oxic basalt ground water at 150, 250, and 300/sup 0/C and in air at 20 and 250/sup 0/C; and FCGR tests were done in basalt ground water, fluoride-ion-enhanced basalt ground water, high-purity water, and air at 90/sup 0/C. The following conclusions can be drawn: the general corrosion rate of statically loaded corrosion coupons was very low in a 3-mo test, and no pitting or cracking of the specimens was observed. Ti-grade 2 and Ti-grade 12 exhibited strain rate dependent ductility diminution in SSR tests. The ductility diminution was most severe in Ti-grade 2 at 300/sup 0/C and in Ti-grade 12 at 250/sup 0/C. For of Ti-grade 12 it was found to be highly orientation dependent. The ductility diminution was also found in tests conducted in air as well as in those conducted in the basalt ground water environment; however, the extent of the degradation was less in air. The ductility diminution cannot be attributed to stress corrosion cracking because the fracture mode was microvoid coalescence in all tests. Evidence obtained in the current study and correlation of the present results with results obtained by other researchers indicate that dynamic strain aging is responsible for the loss of ductility. The FCGR of Ti-grade 2 and Ti-grade 12 was not affected by any of the environmental conditions used in this study, which indicates that no environmental cracking mechanism is operative under the conditions tested (90/sup 0/C, oxic ground water, and frequencies from 0.01 to 5 Hz).},
doi = {10.2172/6601741},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Mon Nov 01 00:00:00 EST 1982},
month = {Mon Nov 01 00:00:00 EST 1982}
}

Technical Report:

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  • Simulated spent fuel, simulated defense high-level waste, and simulated and {sup 99}Tc-doped commercial high-level waste (PNL 76-68) were reacted with groundwater, both in the presence and absence of basalt, simulating expected conditions for a nuclear waste repository located in basalt at the Hanford Site, Washington. Experiments were performed in gold bag sampling autoclaves at temperatures between 90{degree}C and 300{degree}C, at 30 megapascals pressure. During the course of the experiments, samples of the fluid phase were periodically withdrawn from the autoclaves and were analyzed for pH as well as major, minor, and trace cations and anions. At 200{degree}C and 300{degree}C, allmore » dissolved species displayed either steady-state concentrations or decreasing concentrations after the first 1,000 hrs. At 100{degree}C, some dissolved components had not reached steady-state concentrations after 6000 hr. Solids characterization suggests that the formation of secondary alteration phases such as alkali feldspar, smectite clays, scapolite, and a variety of uranium-bearing silicate phases, imposes solubility limits on the release of many analog elements of potential radionuclides. These steady-state (or solubility) concentration limits can be coupled with measured hydrologic flow rates to calculate radionuclide release rates from the waste form for a nuclear waste repository located in basalt.« less
  • This progress report summarizes work on the interaction of waste package materials with simulated salt repository environments. Research focused on the mechanistic understanding of corrosion processes involving carbon steels in acid chloride solutions, a literature review of the solubility of selected compounds of several elements in simulated salt repository brines, and the development of a plan to investigate the mechanisms for mobilization and movement of water and pathways for the flow of brine in a bedded salt formation. A detailed literature search was performed to compile relevant corrosion data for carbon steels in anoxic acid chloride solutions and simulated saltmore » repository brines at temperatures between /approximately/20 and 400/degree/C. An analytical model for general corrosion was developed to calculate the amount of penetration (i.e., thinning) as a function of time, temperature, and the pressure of corrosion product hydrogen that can build up during long-term exposure. The model was used to calculate the amount of corrosive penetration of a carbon steel container for three waste package configurations for emplacement time periods up to 1000 years. A literature review of solublity data and other related thermodynamic information was conducted for compounds of Sn, Pb, Sr, and Se from a list of 17 elements that form an important component of the waste. A methodology for investigating the principal transport processes and pathways for water movement in a bedded salt formation was developed, and laboratory studies to obtain more quantitative estimates of water movement in the repository were planned. 85 refs., 27 figs., 18 tabs.« less
  • Simulated spent fuel, simulated defense high-level waste, and simulated and /sup 99/Tc-doped commercial high-level waste (PNL 76-68) were reacted with groundwater, both in the presence and absence of basalt, simulating expected conditions for a nuclear waste repository located in basalt in gold bag sampling autoclaves at temperatures between 90/sup 0/ and 300/sup 0/C, at 30 MPa pressure. During the course of the experiments, samples of the fluid phase were periodically withdrawn from the autoclaves and were analyzed for pH as well as major, minor, and trace cations and anions. At 200 and 300/sup 0/C, all dissolved species displayed either steady-statemore » concentrations or decreasing concentrations after the first 1000 h. At 100/sup 0/C, some dissolved components had not reached steady-state concentrations after 6000 h. Solids characterization suggests that the formation of secondary alteration phases such as alkali feldspar, smectite clays, scapolite, and a variety of uranium-bearing silicate phases, imposes solubility limits on the release of many analog elements of potential radionuclides. These steady-state (or solubility) concentration limits can be coupled with measured hydrologic flow rates to calculate radionuclide release rates from the waste form for a nuclear waste repository in basalt. The experimentally determined steady-state concentrations of analog elements are compared to calculated solubilities for individual elements. In many cases, the experimentally determined concentrations are several orders of magnitude higher than the calculated solubility concentrations. Possible reasons for these discrepancies include invalid assumptions on stable alteration solids, temperature differences, and kinetic effects. The steady-state concentrations reported in this document provide realistic and defensibly conservative data that can be used on a preliminary basis for evaluating waste form performance both alone and in the presence of basalt.« less
  • This paper describes the reference waste package design for the Basalt Waste Isolation Project. The waste form, functional requirements, and configuration for the reference design are described. A short horizontal borehole emplacement configuration containing a single waste package can meet the functional requirements during the operating, containment, and post-containment periods. The waste form will be placed in a thick-walled, carbon steel container. Surrounding the container is packing material made from a mixture of crushed basalt and sodium bentonite clay. 3 refs.
  • A method of developing waste package performance requirements for specific nuclides is described, and based on federal regulations concerning permissible concentrations in solution at the point of discharge to the accessible environment, a simple and conservative transport model, and baseline and potential worst-case release scenarios.