scholarly journals General corrosion, irradiation-corrosion, and environmental-mechanical evaluation of nuclear waste package structural barrier materials. Progress report

1982 ◽  
Author(s):  
R.E. Westerman ◽  
S.G. Pitman ◽  
J.L. Nelson
Keyword(s):  
Nuclear Waste ◽  
Progress Report ◽  
General Corrosion ◽  
Waste Package ◽  
Barrier Materials ◽  
Structural Barrier

Corrosion of Candidate Iron-Base Waste Package Structural Barrier Materials in Moist Salt Environments

1984 ◽  
Vol 44 ◽  
Author(s):  
R. E. Westerman ◽  
S. G. Pitman
Keyword(s):  
Permian Basin ◽  
Waste Package ◽  
Barrier Materials ◽  
Iron Base ◽  
Corrosion Rates ◽  
Core Samples ◽  
Structural Barrier

AbstractMild steels are considered to be strong candidates for waste package structural barrier (e.g., overpack) applications in salt repositories. Corrosion rates of these materials determined in autoclave tests utilizing a simulated intrusion brine based on Permian Basin core samples are low, generally μm (1 mil) per year. When the steels are exposed to moist salts containing simulated inclusion brines, the corrosion rates are found to increase significantly. The magnesium in the inclusion brine component of the environment is believed to be responsible for the increased corrosion rates.

Effect of the Residual Heat Release of the Nuclear Waste Stored in an Unsaturated Zone on Radionuclide Release

2010 ◽  
Author(s):  
Lubna K. Hamdan ◽  
John C. Walton ◽  
Arturo Woocay
Keyword(s):  
Heat Release ◽  
Nuclear Waste ◽  
Unsaturated Zone ◽  
Yucca Mountain ◽  
Minimum Length ◽  
General Corrosion ◽  
Waste Package ◽  
Flow Through ◽  
Radionuclide Release ◽  
Residual Heat

Over time, nuclear waste packages disposed in geological repositories are expected to fail gradually due to localized and general corrosion. As a result, water will have access to the nuclear waste and radionuclides will be transported to the accessible environment by ground water. In this paper we consider a serious failure case in which penetrations at the top and bottom of the waste package will allow water to flow through it (flow-through model). We introduce a new conceptual model that examines the effect of the residual heat release of the nuclear waste stored in an unsaturated environment on radionuclide release. This model predicts that the evaporation of water at the hotter sheltered areas (from condensate and seepage) inside the failed waste package will create a capillary pressure gradient that drives water to wick with its dissolved and suspended contents toward these relict areas, effectively preventing radionuclides release. We drive a dimensionless group to estimate the minimum length of the sheltered areas required to sequester radionuclides and prevent their release. The implications of this model on the performance of the proposed repository at Yucca Mountain or unsaturated zone geological repositories in general are explored.

scholarly journals Review of Corrosion Modes For Alloy 22 Regarding Lifetime Expectancy of Nuclear Waste Containers

2002 ◽  
Vol 757 ◽  
Author(s):  
Raúl B. Rebak ◽  
John C. Estill
Keyword(s):  
Corrosion Rate ◽  
Nuclear Waste ◽  
Outer Layer ◽  
Localized Corrosion ◽  
General Corrosion ◽  
Corrosion Resistant ◽  
Waste Package ◽  
Environmentally Assisted Cracking ◽  
Alloy 22 ◽  
Lifetime Expectancy

ABSTRACTAlloy 22 (UNS N06022) was selected to fabricate the corrosion resistant outer barrier of a two-layer nuclear waste package container. This paper reviews the main corrosion degradation modes that are predicted for the outer layer of the container. Current results show that the containers would perform well under general corrosion, localized corrosion and environmentally assisted cracking (EAC). For example, the general corrosion rate is expected to be below 100 nm/year and the container is predicted to be outside the range of potential for localized corrosion and environmentally assisted cracking.

Retrieval and Analysis of Simulated Defense HLW Package Experiments at the WIPP

1988 ◽  
Vol 127 ◽  
Author(s):  
Martin A. Molecke ◽  
N. Rob Sorensen
Keyword(s):  
Rock Salt ◽  
Waste Form ◽  
High Level Waste ◽  
General Corrosion ◽  
Uniform Corrosion ◽  
Geochemical Environment ◽  
Waste Package ◽  
Barrier Materials ◽  
Backfill Materials

ABSTRACTIn situ waste package performance experiments involving simulated (non-radioactive) defense high-level waste (DHLW) containers have been in progress since late 1984 at the Waste Isolation Pilot Plant (WIPP) facility. These experiments involve full-size, simulated DHLW containers of several metals and designs emplaced in the WIPP bedded rock salt. These test containers are surrounded by granular backfill (packing) materials, have in many cases been intentionally injected with brines, and are heavily instrumented. A majority of the test packages also contain nonradioactive DHLW borosilicate glass waste form, either within the container and/or outside of it. The primary purpose of these WIPP simulated DHLW experiments is to evaluate the in situ durability and performance of all waste package engineered barrier materials, and to perform package concept validation testing.Twelve of the test DHLW containers, emplaced in WIPP test Room B, have been in heated operation since 1985 and had a maximum surface temperature of about 190°C. These containers were recently retrieved, after about 3 years of heated exposure, for detailed posttest laboratory analyses of: general corrosion and metallurgical degradation, waste form and backfill materials alterations, and other rock salt-brine-barrier materials near-field interactions with the “repository” geochemical environment. Test canisters and overpacks made of ASTM Grade-12 titanium showed essentially no visible degradation in either the base metal or welds; cast mild steel A216/WCA over-packs have suffered some uniform corrosion. Significant degradation of the removed instruments and associated test apparatus has been found: pieces of stainless steel (both 304L and 316) apparatus have undergone extensive stress-corrosion cracking failure and non-uniform attack; Inconel 600-sheathed instruments have undergone both extensive uniform and localized (pitting) attack. Granular backfill materials have been significantly compacted by creep closure to about a density of 2 kg/m. Laboratory analyses are still in progress. Further details on these materials results plus instrumentation data and other in situ WIPP waste package test observations are discussed.

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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