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.

Modeling Alloy 22 general corrosion in the Yucca Mountain nuclear waste repository

JOM ◽  
2008 ◽  
Vol 60 (1) ◽  
pp. 52-57 ◽  
Author(s):  
Kevin G. Mon ◽  
Brvan E. Bullard ◽  
Fred Hua ◽  
Gopal C. De
Keyword(s):  
Nuclear Waste ◽  
Yucca Mountain ◽  
Nuclear Waste Repository ◽  
General Corrosion ◽  
Alloy 22 ◽  
Waste Repository

The Corrosion Behavior of Niobium in Hydrochloric Acid Solutions

CORROSION ◽  
1980 ◽  
Vol 36 (10) ◽  
pp. 554-558 ◽  
Author(s):  
B. S. COVINO ◽  
J. P. CARTER ◽  
S. D. CRAMER
Keyword(s):  
Hydrochloric Acid ◽  
Corrosion Rate ◽  
Corrosion Behavior ◽  
Mineral Resources ◽  
The United States ◽  
General Corrosion ◽  
Ferric Ion ◽  
Acid Solutions ◽  
Corrosion Rates ◽  
Hydrochloric Acid Solutions

Abstract Studies to better understand the corrosion behavior of niobium in hydrochloric acid solutions have been conducted as part of the Bureau of Mines’ effort to conserve the United States’ mineral resources. A study of the general corrosion behavior of niobium as a function of acid concentration, time, oxygen, and added ferric ion is reported. Weight loss (corrosion) tests were performed in air saturated (agitated and static) and helium saturated (agitated) hydrochloric acid solutions ranging from 1N to 10N at temperatures from 35 to 100 C and exposures up to 60 days. The 30 day corrosion rates ranged from 0.0 μm/y in 35 C air saturated (agitated) 1N HCl to 230 μm/y in air saturated (static) 9.8N HCl at 85 C. Although the corrosion rates decreased rapidly for times up to 30 days, the rates for 30 and 60 days were essentially identical. The presence of air did not affect the corrosion rate of niobium, while the presence of only 70 ppm of ferric ion reduced the corrosion rate in most air saturated 5N and 10N HCl solutions.

scholarly journals Long-Term Corrosion Behavior of Alloy 22 in 5 M CaCl2 at 120°C

2006 ◽  
Author(s):  
John C. Estill ◽  
Gary A. Hust ◽  
Kenneth J. Evans ◽  
Marshall L. Stuart ◽  
Rau´l B. Rebak
Keyword(s):  
Corrosion Rate ◽  
Salt Concentration ◽  
Crevice Corrosion ◽  
Immersion Time ◽  
Corrosion Potential ◽  
Open Circuit ◽  
General Corrosion ◽  
Critical Potential ◽  
Alloy 22

In conditions where tight crevices exist in hot chloride containing solutions Alloy 22 may suffer crevice corrosion. The occurrence (or not) of crevice corrosion in a given environment (e.g. salt concentration and temperature), is governed by the values of the critical potential (Ecrit) for crevice corrosion and the corrosion potential (Ecorr). This paper discusses the evolution of Ecorr and corrosion rate (CR) of creviced Alloy 22 specimens in 5 M calcium chloride (CaCl2) at 120°C. Tested specimens included non-creviced rods and multiple creviced assemblies (MCA) both non-welded (wrought) and welded. Results show that Alloy 22 suffers crevice corrosion under the open circuit conditions in the aerated hot CaCl2 brine. However, after more than a year immersion the propagation of crevice corrosion was not significant. The general corrosion rate decreased or remained unchanged as the immersion time increased. For rods and MCA specimens, the corrosion rate was lower than 100 nm/year after more than a year immersion time.

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.

Effects of Fluoride and other Anions on the Corrosion of Alloy 22

2002 ◽  
Vol 757 ◽  
Author(s):  
A. L. Pulvirenti ◽  
K. M. Needham ◽  
M. A. Adel-Hadadi ◽  
A. Barkatt ◽  
C. R. Marks ◽  
...  
Keyword(s):  
Corrosion Rate ◽  
Phosphoric Acid ◽  
Corrosion Inhibitor ◽  
General Corrosion ◽  
Corrosion Resistant ◽  
Electrochemical Tests ◽  
Corrosion Rates ◽  
Reducing Conditions ◽  
Constant Ph ◽  
Alloy 22

ABSTRACTSamples of Alloy 22 were tested in solutions containing various anions in order to determine their effect on the corrosion of the alloy. It was found that Alloy 22 is relatively corrosion resistant in HCl and HNO3 at pH 1 and 160°C (general corrosion rates on the order of 10 μm/year), but more susceptible to phosphoric acid, especially under reducing conditions. The presence of fluoride raised the corrosion rate of Alloy 22 to the order 1 mm/year at pH 1, and fluoride is still active towards Alloy 22 at pH levels as high as 3.5. Samples tested in solutions of 1000xJ13 in which the pH was altered during testing showed an increase in corrosion rate over solutions of constant pH. Preliminary electrochemical tests suggest that nitrate may be an effective corrosion inhibitor in fluoride containing solutions, while sulfate is not.

scholarly journals Methods to Calculate Corrosion Rates for Alloy 22 From Polarization Resistance Experiments

2006 ◽  
Author(s):  
Lana L. Wong ◽  
Sue I. Martin ◽  
Rau´l B. Rebak
Keyword(s):  
Corrosion Rate ◽  
Rapid Screening ◽  
General Corrosion ◽  
Absolute Value ◽  
Electrochemical Tests ◽  
Corrosion Rates ◽  
Alloy 22 ◽  
Fixed Set ◽  
Faraday Law ◽  
Current Potential

The general corrosion rate may be measured using immersion tests or electrochemical tests. The electrochemical tests are fast and can be used for a rapid screening of environmental effects such as temperature and electrolyte composition. The electrochemical tests are described in ASTM standards G 59 and G 102. The basis of these tests is to calculate the resistance to polarization (Rp) in a voltage vs. current plot and to convert these values to corrosion rates using the Faraday law. Commercial software can calculate the corrosion rate based on inputs from the operator. This paper discusses three ways of calculating the corrosion rate (Methods 1, 2 and 3) based on a fixed set of acquired data of voltage vs. current. The conclusions are that the way the corrosion rate is calculated does not impact greatly on the absolute value of the corrosion rate. Variations in the acquired data (current, potential) from one experiment to another seem more important that the manner the data is fitted with the Rp slope.

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.

Modelling of the Degradation of Cement in a Nuclear Waste Repository

1988 ◽  
Vol 127 ◽  
Author(s):  
A. Haworth ◽  
S. M. Sharland ◽  
C. J. Tweed
Keyword(s):  
Nuclear Waste ◽  
Transport Model ◽  
Thermodynamic Description ◽  
Nuclear Waste Repository ◽  
General Corrosion ◽  
Chemistry Transport Model ◽  
High Ph ◽  
Preliminary Model ◽  
Waste Repository ◽  
Aqueous Species

ABSTRACTThe current UK concept for a low- or intermediate-level nuclear waste repository includes a largely cementitious backfill. The cement provides a high pH environment in which the general corrosion rate of the metal canisters is reduced and the solubilities of many nuclides low. It has previously been assumed that this high pH will exist for a period of 107 years, however cement will degrade due to leaching of the solid components and attack from aqueous species in groundwater. In this paper we describe the preliminary stages of a model of the degradation of cement in a repository. The modelling involves the incorporation of a thermodynamic description of cement[2] into the static code PHREEQE[5]. This is then used in a coupled chemistry-transport model of simple leaching of cement using the code CHEQMATE[4]. This preliminary modelling also provides a useful verification of CHEQMATE as a direct comparison with a THCCDM (a coupled code based on CHEMTRN) model is possible. Results from this preliminary model suggest that the fall in pH due to leaching is slow. The model is sufficiently flexible to form the basis of more detailed investigations of the effect of groundwater interactions on the degradation of cement.

Effects of Heavy Metal Ions on Corrosion of Titanium Grade 7 In Concentrated Groundwaters

2002 ◽  
Vol 713 ◽  
Author(s):  
April L. Pulvirenti ◽  
Karen M. Needham ◽  
Mohammad A. Adel-Hadadi ◽  
Charles R. Marks ◽  
Jeffery A. Gormana ◽  
...  
Keyword(s):  
Nuclear Waste ◽  
Heavy Metal Ions ◽  
Yucca Mountain ◽  
Nuclear Waste Repository ◽  
General Corrosion ◽  
Reasonable Approximation ◽  
Waste Repository ◽  
Concentrated Solution ◽  
Titanium Grade ◽  
Service Environments

ABSTRACTA Titanium Grade 7 (Ti-7) drip shield is being considered for use in the potential nuclear waste repository at Yucca Mountain, Nevada. The purpose of the drip shield is to keep water from reaching the surfaces of the canisters containing nuclear waste. Among the species that may be expected to be present in the water are heavy metals. This paper explores the effects of five species, lead, mercury, cadmium, tin and sodium, on the dissolution rate of Ti-7. The results of these tests indicated that lead, mercury, cadmium and tin ions do not significantly accelerate the general corrosion of Ti-7 under the conditions studied. A highly concentrated solution of NaOH was found to cause significant dissolution of Ti-7. Such a solution is a reasonable approximation of possible service environments.

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