Mcc Corrosion Tests at Reference Testing Conditions for A27 Cast steel in Hanford Ground Water

AbstractThe Materials Characterization Center (MCC) at Pacific Northwest Lab- oratory is performing three kinds of corrosion tests for the Basalt Waste Isolation Project (BWIP) to establish the interlaboratory reproducibility and uncertainty of corrosion rates of container materials for high-level nuclear waste. The three types of corrosion tests were selected to address two distinct conditions that are expected in a repository constructed in basalt. An air/steam test is designed to address corrosion during the operational period and static pressure vessel and flowby tests are designed to address corrosion under conditions that bound the condi ring the post-closure period of the repository.The results of tests at reference testing conditions, which were defined to facilitate interlaboratory comparison of data, are presented. Data are reported for the BWIP/MCC-105.5 Air/Steam Test, BWIP/MCC-105.1 Static Pressure Vessel, and BWIP/MC-105.4 Flowby Test. In those cases where data are available from a second laboratory, a statistical analysis of interlaboratory results is reported and expected confidence intervals for mean corrosion rates are given. Other statistical treatment of data include analyses of the effects of vessel-to-vessel variations, test capsule variations for the flowby test, and oven-to-oven variations for air/steam tests.

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Corrosion of Lead and Lead Alloys in Simulated Repository Environments

MRS Proceedings ◽

10.1557/proc-112-783 ◽

1987 ◽

Vol 112 ◽

Author(s):

Frank E. Goodwin ◽

Richard J. Guenther

Keyword(s):

Nuclear Waste ◽

Corrosion Products ◽

Research Organization ◽

Corrosion Tests ◽

Corrosion Rates ◽

Lead Alloys ◽

Lead Zinc ◽

High Level Nuclear Waste ◽

High Level ◽

Waste Repositories

AbstractA series of corrosion tests were sponsored by the International Lead Zinc Research Organization, Inc. and conducted by Battelle Northwest to obtain information regarding the suitability of lead for applications in the containment of high-level nuclear waste. Two grades of unalloyed lead and two lead alloys were exposed to basalt, salt, and tuff simulated repository environments for times up to 1 year and temperatures from 55 to 200°C. Information was obtained on the corrosion rates, corrosion products, and concentration of lead in the test liquids under these conditions, which were intended to simulate the anticipated environments in candidate U.S. nuclear waste repositories. Leachates from the 12-month tests were analyzed in order to prepare synthetic leachates that were used to conduct scoping evaluations of the transport of lead in basalt and tuff environments.

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A Demonstration of Dose Modeling at Yucca Mountain

MRS Proceedings ◽

10.1557/proc-294-937 ◽

1992 ◽

Vol 294 ◽

Author(s):

Terri B. Miley ◽

Paul W. Eslinger

Keyword(s):

Pacific Northwest ◽

Environmental Protection Agency ◽

Yucca Mountain ◽

Department Of Energy ◽

Geologic Repository ◽

Hanford Site ◽

Regulatory Guidance ◽

Dose Limits ◽

High Level Nuclear Waste ◽

High Level

ABSTRACTThe U. S. Environmental Protection Agency is currently revising the regulatory guidance for high-level nuclear waste disposal. In its draft form, the guidelines contain dose limits. Since this is likely to be the case in the final regulations, it is essential that the U.S. Department of Energy be prepared to calculate site-specific doses for any potential repository location. This year, Pacific Northwest Laboratory (PNL) has made a first attempt to estimate doses for the potential geologic repository at Yucca Mountain, Nevada as part of a preliminary total-systems performance assessment.A set of transport scenarios was defined to assess the cumulative release of radionuclides over 10,000 years under undisturbed and disturbed conditions at Yucca Mountain. Dose estimates were provided for several of the transport scenarios modeled. The exposure scenarios used to estimate dose in this total-systems exercise should not, however, be considered a definitive set of scenarios for determining the risk of the potential repository.Exposure scenarios were defined for waterborne and surface contamination that result from both undisturbed and disturbed performance of the potential repository. The exposure scenarios used for this analysis were designed for the Hanford Site in Washington. The undisturbed performance scenarios for which exposures were modeled are gas-phase release of 14C to the surface and natural breakdown of the waste containers with waterborne release. The disturbed performance scenario for which doses were estimated is exploratory drilling. Both surface and waterborne contamination were considered for the drilling intrusion scenario.

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Corrosion of Candidate Container Materials in Air-Steam Mixtures

MRS Proceedings ◽

10.1557/proc-112-751 ◽

1987 ◽

Vol 112 ◽

Author(s):

James M. Lutton ◽

D. A. Dewees ◽

C. G. Robinson ◽

W. F. Brehm ◽

R. P. Anantatmula

Keyword(s):

Carbon Steel ◽

Nuclear Waste Repository ◽

Corrosion Mechanism ◽

Ferritic Alloy ◽

Corrosion Rates ◽

Air Atmosphere ◽

Base Materials ◽

High Level Nuclear Waste ◽

High Level ◽

Corrosion Of Iron

AbstractThe environment during the operating period of a high-level nuclear waste repository in basalt is expected to be air saturated with steam. Liquid groundwater is not expected to be in contact with the container surface during that time.The report presents corrosion findings from tests conducted for one to twenty-five months in an air-steam environment. Tests were carried out with bare metal specimens exposed to an air atmosphere containing 12% moisture in chambers maintained at temperatures between 150°C and 300°C. Cast carbon steel exhibited total penetrations less than 0.002 mm for exposures up to 25 months. A ferritic alloy steel, Fe9CrlMo, showed corrosion results very similar to cast carbon steel. Unalloyed copper materials showed essentially linear corrosion rates, with total penetrations between 0.002 mm at 150°C and 0.14 mm at 300°C in 25 months. Cupronickel 90−10 exhibited total penetrations between 0.001 mm at 150°C and 0.05 mm at 300°C in 25 months. There was a tendency for the corrosion rate to increase with time for cupronickel at 250°C and 300°C possibly because of a mid-test change in the corrosion mechanism. Limited testing of specimens surrounded with bentonite/basalt packing material indicated that the presence of packing has no strong effect on the corrosion of iron-base materials; however, copper-base and cupronickel materials corroded at higher rates in the presence of packing, with a possible shift towards the lower bare specimen corrosion rates with increasing time.

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Release and speciation of carbon from Zircaloy-4 in anaerobic and highly alkaline conditions: Comparison of simple immersion and potentiostatic corrosion tests

Radiocarbon ◽

10.1017/rdc.2018.105 ◽

2018 ◽

Vol 60 (6) ◽

pp. 1787-1796

Author(s):

Sebastien Caes ◽

Frank Druyts ◽

Peter Thomas

Keyword(s):

Corrosion Rate ◽

Nuclear Power ◽

Power Plants ◽

Corrosion Potential ◽

Fuel Cladding ◽

Corrosion Tests ◽

Potentiodynamic Polarization Curves ◽

Alkaline Conditions ◽

High Level Nuclear Waste ◽

High Level

ABSTRACTThe gas release and speciation of carbon species from irradiated and unirradiated Zircaloy-4 samples, representative for the fuel cladding as used in Belgian nuclear power plants, were studied in a saturated Ca(OH)2 solution in anaerobic conditions. This environment is relevant for the Belgian Supercontainer design, as perceived for the geological disposal of high-level nuclear waste. To achieve this, we performed simple immersion and potentiostatic corrosion tests. Potentiodynamic polarization curves, recorded prior to the potentiostatic tests, revealed that irradiation seems to induce changes on the Zircaloy-4 corrosion behavior, such as a shift of the corrosion potential. Potentiostatic corrosion tests on unirradiated Zircaloy-4 provided a corrosion rate of ~54 nm/yr over a 7 day-experiment, whilst a corrosion rate of only ~4 nm/yr was calculated for the irradiated sample. Gas chromatography revealed that during simple immersion tests, which lasted 195 days, hydrogen, methane, ethane, and CO2 were produced, with methane being the major compound. Assuming that all carbon released from the metal was transformed into gaseous compounds, this yields to a corrosion rate ranging from 57 to 84 nm/yr for the irradiated sample. However, caution has to be taken on these corrosion rate and more tests should be performed to confirm these results.

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Ruthenium solubility and its impact on the crystallization behavior and electrical conductivity of MoO3-containing borosilicate-based model high-level nuclear waste glasses

Journal of Non-Crystalline Solids ◽

10.1016/j.jnoncrysol.2020.120356 ◽

2020 ◽

Vol 549 ◽

pp. 120356

Author(s):

Hrishikesh Kamat ◽

Blanca I. Arias-Serrano ◽

Aleksey Yaremchenko ◽

Ashutosh Goel

Keyword(s):

Electrical Conductivity ◽

Nuclear Waste ◽

Crystallization Behavior ◽

High Level Nuclear Waste ◽

High Level

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Conductivity Behavior of Salt Deposits on the Surface of Engineered Barrier Materials for the Potential High-Level Nuclear Waste Repository at Yucca Mountain, Nevada

MRS Proceedings ◽

10.1557/proc-824-cc1.4 ◽

2004 ◽

Vol 824 ◽

Cited By ~ 3

Author(s):

Lietai Yang ◽

Miriam R. Juckett ◽

Roberto T. Pabalan

Keyword(s):

Electrical Conductance ◽

Yucca Mountain ◽

Nuclear Waste Repository ◽

Conductivity Measurements ◽

Barrier Materials ◽

Salt Mixtures ◽

High Level Nuclear Waste ◽

Waste Repository ◽

High Level ◽

Conductivity Behavior

AbstractThe electrical conductance or conductivity of three salt mixtures, Na-K-Cl-NO3, Ca-K-Cl and Ca-Na-Cl, were measured at 25, 50 and 70°C [77, 122, and 158 °F] as a function of relative humidity (RH). Mutual deliquescence and efflorescence RH (MDRH and MERH) values were determined based on the conductivity measurements. It was found that the conductivity of the three salt mixtures started to increase at RH values that are approximately 40 % of their MDRH and increased by 1to 2 orders of magnitude just before reaching the MDRH. At the MDRH, a significant increase in conductivity was observed. The MDRH and MERH for the Ca-K-Cl and Ca-Na-Cl mixtures were found to be approximately 15 % in the temperature range of 50 to 70 °C [122 to 158 °F]. The MDRH and MERH for the Na-K-Cl-NO3system were found to be approximately 54 % at 50 °C [122 °F] and decreased significantly with an increase in temperature.

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