Evaluation of the Susceptibility of Astm A216 Grade Wca Mild Steel to Stress Corrosion Cracking in Simulated Salt Repositopy Environments

1986 ◽  
Vol 84 ◽  
Author(s):  
S. G. Pitman
Keyword(s):  
Mild Steel ◽  
Stress Corrosion ◽  
Nuclear Waste ◽  
Dynamic Tests ◽  
Static Tests ◽  
Waste Package ◽  
Test Conditions ◽  
High Level Nuclear Waste ◽  
High Level ◽  
Container Material

AbstractIn current conceptual designs, a mild steel (ASTM A?16 Grade WCA) is the relerence container material for use in high level nuclear waste packages intended for emplacement in a salt repository. The resistance of the steel to stress corrosion crackinq (SCC) is being investigated as part of the effort underway to verify the suitability of the material for waste package applications. Static tests (U-bend and bolt-loaded fracture toughness specimens) and dynamic tests (slow strain rate and corrosion fatigue) were conducted on both as-cast and weldment specimens of the material, in both low-Mg and high-Mg halite-saturated brines, in the temperature range of 90 to 200°C. The investigations indicate that the steel is not susceptible to SCC under the test conditions employed.

A Corrosion Localization Assessment of the Mild Steel Used for Nuclear Waste Package

1989 ◽  
Vol 176 ◽  
Author(s):  
Masatsune Akashi ◽  
Takanori Fukuda ◽  
Hiroshi Yoneyama
Keyword(s):  
Mild Steel ◽  
Penetration Depth ◽  
Nuclear Waste ◽  
Ultrasonic Inspection ◽  
Gumbel Distribution ◽  
Distribution Model ◽  
Waste Package ◽  
Maximum Penetration Depth ◽  
Maximum Penetration ◽  
High Level

ABSTRACTThis paper describes a study of corrosion behavior of a mild steel as a candidate of the high-level nuclear waste package in the geological disposal situations, conducted to establish a model for estimating the corrosion allowance requirement to achieve the 1,000 year lifetime for the package. In several series of galvanostatic tests, the maximum penetration depth and the depth distribution were measured for each specimen with a sophisticated ultrasonic inspection technique. The Gumbel distribution model was successfully used in analyzing each set of data for the maximum penetration depth. Relations among the average penetration depth, the maximum penetration depth, and the corrosion allowance requirement were discussed.

scholarly journals Functions of the Materials Review Board

1984 ◽  
Vol 44 ◽  
Author(s):  
M. J. Steindler ◽  
W. B. Seefeldt
Keyword(s):  
Nuclear Waste ◽  
Environmental Protection Agency ◽  
Nuclear Regulatory Commission ◽  
Department Of Energy ◽  
Waste Package ◽  
Performance Predictions ◽  
High Level Nuclear Waste ◽  
Regulatory Commission ◽  
High Level ◽  
Nuclear Waste Policy

Some nuclear waste is destined for disposal in deep geological formations. The disposal system for wastes from commercial nuclear activities, and perhaps also for high-level wastes from defense-related activities, is to be designed and operated by the Department of Energy (DOE) and licensed by the Nuclear Regulatory Commission (NRC). The Nuclear Waste Policy Act [1] outlines some of the procedures and schedules that are to be followed by DOE in carrying out its assignment in the disposal of high-level nuclear waste (HLW). The regulations of the NRC that deal with HLW [2] are only partly in place, and amendments (e.g., related to the unsaturated zone) are yet to be approved and issued. The Environmental Protection Agency (EPA) has issued only draft versions of the regulations pertaining to HLW disposal [3], but key features of these drafts are at present in adequate agreement with NRC documents. On the basis of the trends that have become evident in the last few years, the DOE will be required to substantiate performance predictions for all pertinent aspects of a repository, especially the performance of the engineered waste package. The basis for demonstrating that the waste package performance in the repository will be in concert with the requirements is data on the waste package materials. These key materials data must clearly be highly reliable, and DOE will be required to assure this reliability. This paper addresses the organization and functions that have been assembled to aid in establishing the quality of materials data that are important in the licensing of a waste repository.

Development of a Backfill for Containment of High-Level Nuclear Waste

1981 ◽  
Vol 11 ◽  
Author(s):  
Floyd N. Hodges ◽  
Joseph H. Westsik ◽  
Lane A. Bray
Keyword(s):  
Thermal Conductivity ◽  
Thermal Diffusivity ◽  
Nuclear Waste ◽  
Release Rate ◽  
Sodium Bentonite ◽  
Waste Package ◽  
Diffusivity Measurements ◽  
High Level Nuclear Waste ◽  
High Level ◽  
Host Rocks

ABSTRACTSodium and calcium bentonites, pressed to densities between 1.9 and 2.2 g/cm3, have hydraulic conductivities in the range of 10−11 to 10−13 cm/s. Batch sorption distribution ratios (Rd) indicate that Sr, Cs, and Am are strongly sorbed on bentonites and zeolites, that Np and U are moderately sorbed on bentonites and zeolites, and that Am, Np, U, I, and Tc are strongly sorbed on charcoal. Sorption results with basalt and tuff ground waters are similar; however, iodine in tuff ground water sorbs more strongly on bentonites Thermal diffusivity measurements for dry, compacted (p ∼ 2.1 g/cm3) sodium bentonite indicate that the thermal conductivity of a high density bentonite backfill should be roughly similar to that of silicate host rocks (basalt, granite, tuff). These results indicate that a bentonite backfill can significantly delay the first release of many radionuclides into the host rock and that by forming a diffusion barrier a bentonite backfill can significantly decrease the longterm release rate of radionuclides from the waste package.

Long-Term Thermal Stability of Alloy 825 as a High-Level Nuclear Waste Container Material

1995 ◽  
Vol 412 ◽  
Author(s):  
D. S. Dunn ◽  
Y.-M. Pan ◽  
G. A. Cragnolino ◽  
N. Sridhar
Keyword(s):  
Grain Boundary ◽  
Nuclear Waste ◽  
Elevated Temperatures ◽  
Thermal Exposure ◽  
Boundary Region ◽  
Waste Container ◽  
High Level Nuclear Waste ◽  
High Level ◽  
Welding Processes ◽  
Container Material

AbstractThe thermal exposure of Fe-Cr-Ni-Mo materials to certain temperature regimes often results in the formation of grain boundary carbides and the associated depletion of alloying elements. This phenomenon, termed sensitization, is frequently the result of welding processes or in service exposure to elevated temperatures. In this investigation, alloy 825, a candidate high-level nuclear waste (HLW) container material, was thermally exposed to temperatures in the range of 550 to 800 °C for periods of up to 1,000 hr. Sensitization of the material was evaluated by corrosion tests and grain boundary analyses using an analytical electron microscope. The sensitized microstructure was found to contain M23C6-type carbides as well as a Cr-depleted region in the vicinity of the grain boundaries. The degree of sensitization was correlated to the extent of Cr depletion in the grain boundary region.

Critical Conditions for Initiation of Localized Corrosion of Mild Steels in Contact with Bentonite Used in Geological Disposal Packages of Nuclear Waste

1992 ◽  
Vol 294 ◽  
Author(s):  
Guen Nakayama ◽  
Masatsune Akashi
Keyword(s):  
Mild Steel ◽  
Nuclear Waste ◽  
Localized Corrosion ◽  
Critical Conditions ◽  
Geological Disposal ◽  
Alkaline Environments ◽  
Current Design ◽  
Long Time ◽  
High Level Nuclear Waste ◽  
High Level

ABSTRACTIn the current design of geological disposal of high-level nuclear waste, the use of bentonite to stand as an artificial barrier-cum-buffer between the host rock and the packages made of mild steel is being investigated. Although mild steels commomly have been considered to be passivity in alkaline environments, under certain circumstances, they become liable to localized corrosion, e.g., pitting corrosion and crevice corrosion. Since bentonite can turn the environment alkaline to a pH of approximately 10 when it is mixed with groundwater, critical conditions for the initiation of localized corrosion of mild steel must be known to evaluate the extremely long time integrity of disposal packages serving in such an environment. This paper presents and discusses the observations and results acquired in a series of critical conditions for the initiation of localized corrosion of mild steels in various groundwater-bentonite environments at 20C, with a deaerated aqueous solution of 1 mMol/L [HCO3−] +10 ppm [CI−], simulating the natural groundwater and varying the bentonite content.

Integrated Analysis for Long-Term Degradation of Waste Package at the Potential Yucca Mountain Repository for High-Level Nuclear Waste Disposal

2002 ◽  
Vol 713 ◽  
Author(s):  
Joon H. Lee ◽  
Kevin G. Mon ◽  
Dennis E. Longsine ◽  
Bryan E. Bullard ◽  
Ahmed M. Moniba
Keyword(s):  
Corrosion Rate ◽  
Nuclear Waste ◽  
Yucca Mountain ◽  
Integrated Analysis ◽  
Total System ◽  
General Corrosion ◽  
Waste Package ◽  
High Level Nuclear Waste ◽  
High Level

ABSTRACTThe technical basis for Site Recommendation (SR) of the potential repository for high-level nuclear waste at Yucca Mountain, Nevada has been completed. Long-term containment of the waste and subsequent slow release of radionuclides from the engineered barrier system (EBS) into the geosphere will rely on a robust waste package (WP) design, among other EBS components as well as the natural barrier system. The WP and drip shield (DS) degradation analyses for the total system performance assessment (TSPA) baseline model for the SR have shown that, based on the current corrosion models and assumptions, both the DSs and WPs do not fail within the regulatory compliance time period (10,000 years). From the perspective of initial WP failure time, the analysis results are encouraging because the upper bounds of the baseline case are likely to represent the worst case combination of key corrosion model parameters that significantly affect long-term performance of WPs in the potential repository. The estimated long life-time of the WPs in the current analysis is attributed mostly to the following two factors that delay the onset of stress corrosion cracking (SCC): (1) the stress mitigation to substantial depths from the outer surface in the dual closure-lid weld regions; and (2) the very low general-corrosion rate applied to the closure-lid weld regions to corrode the compressive stress zones. Uncertainties are associated with the current WP SCC analysis. These are stress mitigation on the closure-lid welds, characterization of manufacturing flaws applied to SCC, and general corrosion rate applied to the closurelid weld regions. These uncertainties are expected to be reduced as additional data and analyses are developed.

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