Waste Package Container Research Articles

A study of corrosion behavior of Ni-22Cr-13Mo-3W alloy under hygroscopic salt deposits on hot surface

Alloy 22, a nickel base Ni-22Cr-13Mo-3W alloy has an excellent corrosion resistance in oxidizing and reducing environments. Most of the corrosion studies on Alloy 22 have been conducted using conventional chemical or electrochemical methods. In the present investigation, the specimen was directly heated instead of heating the electrolyte, thereby simulating the nuclear waste package container temperature profile. Corrosion behavior of Alloy 22 and evaporation conditions of water diffusing on the container were evaluated using the newly devised heated electrode corrosion test (HECT) method in simulated acidified water (SAW) and simulated concentrated water (SCW) environments. In this method, the concentration of the environment varied with test duration. The corrosion rate of Alloy 22 was not affected by the continuous increase in ionic strength of the SAW (pH 3) environment. Passivation kinetics was faster with increase in concentration of the electrolytes. The major difference between the conventional test and HECT was the aging characteristics of the passive film of Alloy 22. The heated electrode corrosion test can be used for evaluating materials for construction of heat transfer equipments such as evaporators.

Electrochemical methods to detect susceptibility of Ni-Cr-Mo-W alloy 22 to intergranular corrosion

Alloy 22 (UNS N06022), a Ni-Cr-Mo-W based alloy, is a candidate material for the outer wall of nuclear waste package (NWP) containers. Even though the alloy is highly stable at low temperatures, it could undergo microstructural changes during processing such as welding and stress relieving. Formation of topologically close-packed (TCP) phases such as μ, P, σ, etc. and Cr-rich carbides could make the material susceptible to localized corrosion. Hence, it is important to correlate the microstructural changes with the corrosion resistance of the alloy by nondestructive and rapid electrochemical tests. In this investigation, different electrochemical test solutions were used to quantify the microstructural changes associated with aging and welding of the wrought alloy 22. The results of double-loop (DL) electrochemical potentiodynamic reactivation (EPR) tests in 1 M H2SO4+0.5 M NaCl+0.01 M KSCN solution indicated Cr depletion during initial stages of aging of wrought alloy 22. Results of EPR tests in 2 M HCl+0.01 M KSCN solution at 60 °C correlated well with the Mo depletion that occurred near TCP phases formed during aging of both weld and wrought alloy 22 materials. The EPR test results were compared with standard chemical weight loss measurements specified by ASTM standard G-28 methods A and B.

Effect of Inhibiting Oxyanions on the Localized Corrosion Susceptibility of Waste Package Container Materials

AbstractThe DOE is currently preparing a license application for the permanent disposal of high level radioactive waste at Yucca Mountain, Nevada. The proposed design of waste packages for the disposal of high level radioactive waste consists of an outer container made of Alloy 22, a corrosionresistant Ni-Cr-Mo-W alloy, surrounding an inner container made of Type 316 nuclear grade stainless steel. Under conditions where passivity is maintained, the uniform corrosion rate of Alloy 22 is slow and long waste package lifetimes are projected. However, the initiation of localized corrosion such as pitting or crevice corrosion may decrease waste package lifetimes. In this study the crevice corrosion susceptibility of Alloy22 was determined in chloride solutions with additions of oxyanions that are present in the groundwater at the potential repository site. When present in sufficient concentrations relative to chloride, nitrate, carbonate, bicarbonate, and sulfate inhibited pitting and crevice corrosion of Alloy 22.

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

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.

Localized Corrosion Susceptibility of Alloy 22 as a Waste Package Container Material

ABSTRACTAlloy 22 is the material preferred by the U.S. Department of Energy for the waste package outer container for geological disposal of high-level radioactive waste at the proposed repository site in Yucca Mountain, Nevada. The susceptibility of Alloy 22 to localized corrosion is an important consideration in the evaluation of the waste package behavior and the assessment of the overall performance of the proposed repository. The effects of the environment chemical composition and temperature on localized corrosion susceptibility were examined by measuring the repassivation potential for crevice corrosion in chloride-containing solutions at temperatures ranging from 80 to 150°C. The effect of potentially inhibiting anionic species, such as nitrate, was also determined. In addition to the mill annealed material, tests were conducted on both welded and thermally aged material to evaluate microstructural effects related to container fabrication processes. The resistance of Alloy 22 to localized corrosion decreased with increasing temperature and chloride concentration. Welding and thermal aging also decreased the localized corrosion resistance of the alloy.

A Mathematical Model for Advective Release from Perforated Waste Package Container Under Dripping Water

ABSTRACTA defense-in-depth engineered barrier system (EBS) is employed in the current design concept for the potential high-level nuclear waste repository at Yucca Mountain, Nevada, USA. Simplifying the geometry of the cylindrical waste container into the equivalent spherical configuration, and incorporating detailed analysis of the mechanics of water flow around the waste container surface, a mathematical model is developed for advective release from a “failed” (or perforated) waste container under dripping water. It is shown that the advective release rates are controlled by diffusion through the perforations in the waste container, and affected insignificantly by the dripping flow rate for the flow rate range considered. The release rates depend strongly on the number of perforations (or pit penetrations) in the waste container. The insensitivity of the release rate to the dripping flow rate is explained by the fact that radionuclide is released from the container surface to the boundary layer of the water film which contacts the container surface and is relatively stagnant. Also, since a laminar flow around the waste container surface is assumed in the model development, radionuclides transport across the water layers in the film by diffusion only. Additionally, the insignificant effect of the flow rate is contributed by the “short” penetration depth of radionuclide into the water film assumed in the model development. The analyses show that the number of perforations, the size of perforation, the container wall thickness, and the geometry (i.e., radius) of the waste container are important parameters that control the advective release rate. It is emphasized that the “failed” (or perforated) waste package container can still perform as a potentially important barrier to radionuclide release.

The Effect of Fuel Burnup and Dispersed Water Intrusion on the Criticality of Spent High-Level Nuclear Fuel in a Geologic Repository

ABSTRACTStudies of the spent fuel waste package have been conducted through the use of a Monte-Carlo neutron simulation program to determine the ability of the fuel to sustain a chain reaction. These studies have included fuel burnup and the effect of water mists on criticality. Results were compared with previous studies.In many criticality studies of spent fuel waste packages, fresh fuel with an enrichment as high as 4.5% is used as the conservative (worst) case. The actual spent fuel has a certain amount of “burnup” that decreases the concentration of fissile uranium and increases the amount of radionuclides present. The LWR Radiological Data Base from OCRWM has been used to determine the relative radionuclide ratios and KENO 5.a was used to calculate values of the effective multiplication factor, keff.1Spent fuel is not capable of sustaining a chain reaction unless a suitable moderator, such as water, is present. A completely flooded container has been treated as the worst case for criticality. Results of a previous report that demonstrated that keff actually peaked at a water-to-mixture ratio of 13% were analyzed for validity. In the present study, these results did not occur in the SCP waste package container.

Corrosion Model Validation in High Level Nuclear Waste Package Research

ABSTRACTThe strategies for waste package (WP) performance validation will be based on site specific geologic and hydrogeochemical information plus models which can be used to predict potential WP lifetimes. The development and application of such models will include the evaluation of natural analogues (NA). These analogues are needed to resolve issues related to the validation of models. Natural analogues have not had extensive use or widespread acceptance in the area of waste package failure prediction. This lack of acceptance is due to the anticipated choice of alloys for waste package containers. Few of these alloys are similar to naturally occurring metals, and the proposed HLW repositories are in general in geologic settings not very similar to those in which naturally occurring metals are generally found.Natural and archaeological analogues can be used, however, in analysis of possible waste package failures as a means of testing proposed models for failure. In fact, the analogues are the only available mechanisms for testing models of long-term waste package behavior. A strategy is outlined for incorporating natural and archaeological analogue studies into waste package research, and examples are discussed. The natural/archaeological analogues approach which appears most promising is to use archaeological and mineral samples to develop an understanding of the identities and rates of the mineral alteration reactions at or near the surface of the package, improving present capability for estmating the lifetimes of metallic waste package containers.

Modeling of release of radionuclides from an engineered disposal facility for shallow-land disposal of low-level radioactive wastes

The computer code, ENBAR-1, for the simulation of radionuclide releases from an engineered disposal facility has been developed to evaluate the source term for subsequent migration of radionuclides in and through a natural barrier. The system considered here is that a waste package (waste form and container) is placed, together with backfill materials, into a concrete pit as a disposal unit for shallow-land disposal of low-level radioactive wastes. The code developed includes the following modules: water penetration into a concrete pit, corrosion of a drum as a container, leaching of radionuclides from a waste form, migration of radionuclides in backfill materials, release of radionuclides from the pit. The code has the advantage of its simplicity of operation and presentation while still allowing comprehensive evaluation of each element of an engineered disposal facility to be treated. The performance and source term of the facility might be readily estimated with a few key parameters to define the problem.

Buckling Design Criteria for Waste Disposal Containers

This report documents analytical and experimental results from a survey of the technical literature on buckling of thick-walled cylinders under external pressure. Based upon these results, a load factor is suggested for the design of waste package containers for disposal of high level radioactive waste in repositories mined in salt formations. The load factor is defined as a ratio of buckling pressure to allowable pressure. Specifically, a load factor which ranges from 1.5 for plastic buckling to 3.0 for elastic buckling is included in a set of proposed buckling design criteria for waste disposal containers.

Corrosion Studies of A216 Grade WCA Steel in Hydrothermal Magnesium-Containing Brines

AbstractThe U.S. Department of Energy's Salt Repository Project (SRP) is investigating the general corrosion resistance of cast mild steel as a candidate material for waste package containers. Evaluation of this material is being performed at the Pacific Northwest Laboratory in environments simulating expected repository conditions.General corrosion studies of mild steel (ASTM A216 grade WCA) in the as-cast and normalized conditions were conducted in hydrothermal halitesaturated (saturated at ambient temperature) brine environments simulating a “dissolution” and an “inclusion” brine. Corrosion tests were also performed in brines similar to the inclusion brine but containing magnesium concentrations ranging from 1000 to 30,000 ppm to investigate the effect of magnesium on the corrosion behavior.Corrosion rates of the cast mild steel were found to increase with increasing temperature and with increasing magnesium concentration. Some possible mechanisms that explain the observed behavior are presented.

Modeling The Dissolution Behavior of Defense Waste Glass in A Salt Repository Environment

AbstractA mechanistic model describing a dynamic mass balance between the production and consumption of dissolved silica was found to describe the dissolution behavior of SRL-165 defense waste glass in a high-magnesium brine (PBB3) at a temperature of 90°C. The synergistic effect of the waste package container on the glass dissolution rate was found to depend on a precipitation reaction for a ferrous silicate mineral.The model predicted that the ferrous silicate precipitate should be variable in composition where the iron/silica stoichiometry depended on the metal/glass surface area ratio used in the experiment. This prediction was confirmed experimentally by the variable iron/silica ratios observed in filtered leachates. However, the interaction between dissolved silica and iron corrosion products needs to be much better understood before the model can be used with confidence in predicting radionuclide release rates for a salt repository.

Corrosion Behavior of Container Materials in Grande Ronde Basalt Groundwater

AbstractThree candidate waste package container materials were tested for 5 mo at 200 °C in Hanford Site Grande Ronde Basalt groundwater (9.75 pH) under anoxic conditions (<0.1 mg/L dissolved oxygen in water). The materials were cast carbon steel (American Society for Testing and Materials (ASTM) A27), wrought carbon steel (American Iron and Steel Institute (AISI) 1020), cupro- nickel 90-10, and Fe9CrlMo steel. Testing was performed in 1-L titanium autoclaves at a pressure of 6.9 MPa and a flow rate of ∼O.02 mL/min. Anoxic conditions were achieved by constantly sparging the synthetic groundwater in the reservoir with argon. In addition, the groundwater was conditioned by placing a 2.54-cm layer of crushed basalt (∼O.635 cm average size) at the bottom of each autoclave. The average corrosion rates at 200 °C were 0.9 μm/yr for cupronickel 90-10 and Fe9CrlMo steel, 1.1 μm/yr for cast carbon steel, and 1.4 μm/yr for wrought carbon steel. Pitting was not detected in any of the specimens. Posttest analysis of the corrosion speci- mens indicated the formation of a thin film of iron smectite clay on the surface of all specimens. In addition, magnetite formation immediately adjacent to the specimen surface was observed in the iron-base alloys, consistent with previous investigations. Based on the present investiga- tions, all the materials exhibited significantly lower corrosion rates than were used in the Environmental Assessment (∼5 μm/yr) [1] for calculating acceptable container lifetimes. However, longer term general corrosion tests and tests to study the container materials susceptibility to other degradation modes are necessary prior to making a final evaluation of the suitability of these materials for use in fabricating high-level nuclear waste containers.