Irradiation-Corrosion Evaluation of Metals For Nuclear Waste Package Applications in Grande Ronde Basalt Groundwater

1983 ◽  
Vol 26 ◽  
Author(s):  
J.L. Nelson ◽  
R.E. Westerman ◽  
F.S. Gerber
Keyword(s):  
Nuclear Waste ◽  
Irradiation Dose ◽  
Nuclear Waste Repository ◽  
Waste Package ◽  
Iron Base ◽  
Dose Rates ◽  
Corrosion Rates ◽  
Corrosion Evaluation ◽  
Similar Materials ◽  
Waste Repository

ABSTRACTThe corrosion behavior of several iron-base and titanium-base alloys was studied in synthetic Grande Ronde Basalt groundwater at temperatures of 150°C to 2500°C and under irradiation dose rates to 2 × 106 rad/hr. The objective of these ongoing studies is to help select one or more materials for waste-package canisters that will maintain their integrity for time periods up to 1,000 yr in a nuclear waste repository constructed in basalt. The corrosion rates of iron-base alloys under irradiated conditions were generally 2 to 3 times as high as those obtained on similar materials under nonirradiated conditions. The titanium alloys exhibited low corrosion rates but absorbed significant amounts of hydrogen under irradiated conditions.

scholarly journals Preclosure analysis of conceptual waste package designs for a nuclear waste repository in tuff

10.2172/59344 ◽  
1984 ◽  
Author(s):  
W.C. O`Neal ◽  
D.W. Gregg ◽  
J.N. Hockman ◽  
E.W. Russell ◽  
W. Stein
Keyword(s):  
Nuclear Waste ◽  
Nuclear Waste Repository ◽  
Waste Package ◽  
Waste Repository

Experimental Interaction of Basalt, Bentonite, and Water: Implications for Groundwater Chemistry in Waste Package Packing Material in a Nuclear Waste Repository

1987 ◽  
Vol 112 ◽  
Author(s):  
Gail L. McKeon ◽  
E. C. Thornton ◽  
D. J. Halko ◽  
M. I. Wood
Keyword(s):  
Nuclear Waste ◽  
Near Field ◽  
Sodium Concentration ◽  
Packing Material ◽  
Deionized Water ◽  
Solution Chemistry ◽  
Nuclear Waste Repository ◽  
Waste Package ◽  
Waste Repository ◽  
C 30

AbstractExperiments have been conducted by the Basalt Waste Isolation Project (BWIP) to assess changes in solution chemistry in the near-field environment of a nuclear waste repository in basalt. These Dickson autoclave experiments were carried out using Grande Ronde basalt ± bentonite and synthetic groundwater or deionized water at 300°C, 30 MPa, and solution-to-solid mass ratio of 10 for up to two years. Groundwater solution changes during reaction of the basalt and basalt/bentonite included initial decreases in pH and sodium concentration presumably due to smectite formation. This initial trend subsequently reversed in the basalt system with pH rising to ca. 7.5 and sodium increasing to the starting value. Steady state pH values for the basalt/bentonite system were ca. 6.4. The basalt + deionized water test exhibited a constant rise in pH to ca. 7.9 and release of sodium to solution in response to basalt dissolution. Slightly oxidizing conditions characterized the early part of all of the experiments followed by a decrease in fO2 to 10−31 to 10−32 These results are consistent with other work at similar and lower temperatures, suggesting that the packing material will react in the waste package environment to produce slightly alkaline, reducing conditions.

Corrosion Behavior of Alloy 22 and Ti Grade 7 in a Nuclear Waste Repository Environment

CORROSION ◽  
2004 ◽  
Vol 60 (8) ◽  
pp. 764-777 ◽  
Author(s):  
F. Hua ◽  
G. Gordon
Keyword(s):  
Corrosion Rate ◽  
Corrosion Behavior ◽  
Nuclear Waste ◽  
Crevice Corrosion ◽  
Nuclear Waste Repository ◽  
General Corrosion ◽  
Corrosion Rates ◽  
Significant Difference ◽  
Alloy 22 ◽  
Waste Repository

Abstract Alloy 22 (UNS N06022) and Ti Grade 7 (UNS R52400) are the current corrosion-resistant materials of choice for fabricating the waste package outer barrier and the drip shield, respectively, for the proposed high-level nuclear waste repository at Yucca Mountain. In this work, the general and crevice corrosion behavior of annealed and welded Alloy 22 and Ti Grade 7 exposed in basic saturated water (BSW-12) for four and eight weeks at 60°C to 105°C were evaluated using the ASTM G78 method combined with surface analysis and statistical analysis of corrosion rate. The general corrosion rates for Alloy 22 and Ti Grade 7 were found to increase linearly with temperature but decrease with the exposure time. The mean corrosion rate was found to be 0.003 mpy (0.075 μm/y) at 60°C and 0.010 mpy (0.25μm/y) at 105°C for Alloy 22 and 0.008 mpy (0.20 μm/y) at 60°C and 0.022 mpy (0.56 μm/y) at 105°C for Ti Grade 7. No significant difference in corrosion behavior between the annealed and welded materials was observed. For both materials the surface imperfections inherited from materials processing did not seem to deteriorate the excellent corrosion resistance of the materials but might serve as the “traps” for corrosion products. The apparent activation energies for the temperature dependence of corrosion rates of Ti Grade 7 and Alloy 22 in BSW-12 environment were obtained as 25.3 (±5.5) KJ/mol and 23.7 (±4.5) KJ/mol, respectively. Although none of the materials was found susceptible to crevice corrosion under the test conditions, to conclude that these materials are immune to crevice corrosion in BSW-12 would require longer-term testing.

Estimated Effects of Temperature-Relative Humidity Variations on the Composition of In-Drift Water in the Potential Nuclear Waste Repository at Yucca Mountain, Nevada

2004 ◽  
Vol 824 ◽  
Author(s):  
Lauren Browning ◽  
Randall Fedors ◽  
Lietai Yang ◽  
Osvaldo Pensado ◽  
Roberto Pabalan ◽  
...  
Keyword(s):  
Relative Humidity ◽  
Nuclear Waste ◽  
Yucca Mountain ◽  
Nuclear Waste Repository ◽  
Aqueous Corrosion ◽  
Waste Package ◽  
Waste Repository ◽  
Effects Of Temperature ◽  
Drift Conditions ◽  
Simulated Time

AbstractWe define four distinct thermohydrochemical environments for drip shield and waste package corrosion in the potential nuclear waste repository, referred to here as the Dry, Seepage + Evaporation, Seepage + Condensation + Evaporation, and the Seepage + Condensation environments. These environments are bounded by temperature and relative humidity conditions at drift wall and drip shield/waste package surfaces judged most likely to initiate fundamental changes in the quantity and/or chemistry of in-drift waters. The duration in which different environments might exist is evaluated by comparing simulated, time-dependent temperature and relative humidity curves for three different locations within repository drift 25. In-drift conditions and processes postulated to cause drip shield/waste package corrosion are evaluated within the context of the thermohydrochemical environments by various means, including analytical calculations and geochemical simulations. Of the four environments considered here, the Seepage + Evaporation environment presents the most significant potential for aqueous corrosion of drip shield and waste package materials, and may persist for approximately 500 years in center drift locations. The likelihood for corrosion in other thermohydrochemical environments is significantly lower, but may increase with the acquisition of new data or the demonstration of extenuating circumstances.

Chemical, Physical and Engineering Characterization Of Candidate Backfill Clays and Clay Admixtures for a Nuclear Waste Respository-Part I

1981 ◽  
Vol 6 ◽  
Author(s):  
Sudesh K. Singh
Keyword(s):  
Nuclear Waste ◽  
Morphological Changes ◽  
Deionized Water ◽  
Engineering Properties ◽  
Nuclear Waste Repository ◽  
Exchange Cation ◽  
Mineral Components ◽  
Waste Repository ◽  
Mineralogical Compositions

ABSTRACTFourteen Canadian clays and clay admixtures were subjected to simulated nuclear waste repository environments. The present work is concerned with the montmorillonite-dominant materials only. The montmorillonite-dominant samples showed significant leaching on interaction with deionized water. On heating the samples at 200°C for 500 hours, montmorillomites lost intermicellar water completely and acquired cusp-like to cylindrical morphologies. The loss of water and the morphological changes in montmorillonites significantly altered the engineering characteristics. Permeability, shrinkage limits, compactability and shear strength varied in response to the dominant exchange cation in the structure of montmorillonites and the presence of other mineral components in the materials. The synthetic granite water reacted with montmorillonites and led to changes in chemical and mineralogical compositions, crystalline state and engineering properties.

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