High-level Nuclear Waste Containers Research Articles

Research Progress on Hydrogen Absorption and Embrittlement of Titanium and Its Alloy for High-level Nuclear Waste Container in Deep Geological Disposal Environment

Research Progress on Hydrogen Absorption and Embrittlement of Titanium and Its Alloy for High-level Nuclear Waste Container in Deep Geological Disposal Environment

On the long term estimation of hydrogen embrittlement risks of titanium for the fabrication of nuclear waste container in bentonite buffer of nuclear waste repository

Titanium is being considered as a candidate material for high-level nuclear waste container due to its superior corrosion resistance. However, problems may arise when titanium comes in contact with hydrogen-bearing environments, as it is vulnerable to hydrogen embrittlement. To assess the lifetime of titanium container, hydrogen content and hydrogen permeation efficiency were estimated in infiltrated water through bentonite prepared with simulated groundwater of Beishan, the preselected area in China, by electrochemical and metallurgical analysis. It is concluded that the absorbed hydrogen due to general corrosion will not be sufficient for hydrogen embrittlement to occur in titanium for at least 10,000 years after deep geological disposal. This represents more than 300 and 5,000 half-lives of Cs137 and Cs134 respectively; cesium being the most likely element to escape a repository due to its high solubility in groundwaters.

Effects of deep geological environments for nuclear waste disposal on the hydrogen entry into titanium

Abstract The surface of nuclear waste container will be completely wetted by the groundwater infiltrating through the bentonite or concrete buffer after the closure of its deep geological disposal. Because of its excellent corrosion resistance, titanium is one of the candidates for the container materials for high-level nuclear waste (HLNW) disposal. However, once oxygen depleted in the large time scale deep geological disposal, titanium container will suffer from hydrogen embrittlement (HE). Therefore, in order to predict the potential danger of the HLNW container caused by HE, the hydrogen entry into titanium was studied at different temperatures in simulated deep geological environment of Beishan, which is the preselected HLNW disposal area in China. Devanathan-Stachurski cell methods were firstly implemented in measuring apparent hydrogen diffusivity in titanium with hydrides being formed during hydrogen charging at low temperature ranges (298–373 K). Given the characteristic of the Devanathan-Stachurski cell method's capable of getting apparent diffusivity with combined influencing factors easily, the technique provides a simple way to measure the parameter that is necessary for the estimation of hydrogen penetration depth, facilitating the design of HLNW containers.

The Effects of Cathodic Reagent Concentration and Small Solution Volumes on the Corrosion of Copper in Dilute Nitric Acid Solutions

The exposure conditions experienced by copper-coated high-level nuclear waste containers in a deep geologic repository will evolve with time. An early exposure period involving the gamma irradiation of aerated humid vapor could lead to the formation of nitric acid condensed in limited volumes of water on the container surface. The evolution of the corrosion processes under these conditions have been studied using pH measurements in limited volumes of water containing various concentrations of nitric acid. The extent and morphology of corrosion was examined using scanning electron microscopy on surfaces and on focused ion beam cut cross sections. The composition of corrosion products was determined by energy dispersive x-ray analyses and Raman spectroscopy. In the absence of dissolved oxygen only minor corrosion was observed with the reduction of nitric acid inhibited by the formation of either chemisorbed nitrate and nitrite species or the formation of a thin cuprite (Cu2O) layer. When the solution was aerated, both oxygen and nitric acid acted as cathodic reagents. After extensive exposure periods corrosion was stifled by the formation of corrosion product deposits of Cu2O, CuO (tenorite), and Cu2NO3(OH)3 (rouaite).

Generalized Approach for Determination of Thermal Conductivity of Buffer Materials

Determination of thermal conductivity of buffer materials is an important aspect in the design and characterization of engineered barrier systems (EBS) in deep geological repositories (DGRs) for safe containment of high-level nuclear waste. Several factors, viz. compaction state, particle size distribution, and mineralogical characteristics of buffer materials, influence the thermal conductivity of composite buffer materials. Therefore, it is essential to give due regards to the influence of these factors while estimating thermal conductivity of buffer materials. In view of this, the present study pertains to an extensive laboratory scale determination of thermal conductivity of a wide range of sand-bentonite based buffer materials employing a thermal needle probe. Precision and accuracy of the thermal needle probe are established in relation to a contemporary and widely endorsed thermal property analyzer. Further, the efficacy of several predictive models available in the literature is evaluated to appraise the thermal conductivity of composite buffer materials in relation to the experimental data. Realizing the need of a generalized approach considering the influence of clay mineralogy and particle size fraction present in the geomaterial, the manuscript proposes generalized thermal conductivity prediction models meant for the buffer materials. The developed models are found to deliver satisfactory performance and accuracy in appraising thermal conductivity of sand-bentonite buffers used in DGRs.

Crevice Corrosion of Grade-2 Titanium in Saline Solutions at Different Temperatures and Oxygen Concentrations

Titanium alloys are candidate materials for the fabrication of high level nuclear waste containers but may be susceptible to crevice corrosion in the initial warm oxidizing conditions anticipated in a deep geological repository. The crevice corrosion of Grade-2 titanium has been studied in 1.0 mol/L NaCl at various temperatures (120°C and 150°C) and oxygen concentrations using a galvanic coupling technique. The current due to propagation supported by oxygen reduction on the surface external to the crevice was monitored electrochemically and the total propagation determined from weight change measurements. The redistribution of impurities (Fe, V, Mo) during corrosion was determined by dynamic secondary ion mass spectrometry. Propagation was dominantly driven by the reduction of protons created inside the crevice by cation hydrolysis. Damage accumulated at the locations of TixFe intermetallic particles, which are both anodically reactive and function as catalytic cathodes for proton reduction. The compact TiO2·xH2O deposits at active sites led to stifling of corrosion when only between 9 and 56% of the available oxygen had been consumed. While propagation was more rapid at 150°C than at 120°C, it was more readily stifled at the higher temperature, leading to the accumulation of less damage.

CORROSION OF COPPER-COATED STEEL HIGH LEVEL NUCLEAR WASTE CONTAINERS UNDER PERMANENT DISPOSAL CONDITIONS

The corrosion of high level nuclear waste containers under permanent disposal conditions can occur via a number of processes which change in importance as the environment within a deep geologic repository evolves from oxic to anoxic. The container design involves a copper coating on a steel vessel. The characteristics of these coatings are being assessed using standard analytical methods, including microscopy and electron backscatter diffraction. The corrosion processes on the surface of the coatings and at through-coating defects are also being investigated electrochemically. Although minor differences exist, the surface corrosion of cold sprayed and electrodeposited coatings generally exhibited similar corrosion behaviour to standard wrought copper. Corrosion at the base of a simulated through-coating defect in a cold-sprayed coating was found to proceed via galvanic coupling to oxygen reduction on the coating surface. Besides accumulating damage at the base of the defect, corrosion propagated along the copper/steel interface, most likely a consequence of the damage inflicted during the deposition process.

Localized Corrosion of Carbon Steel in Simulated Concrete Pore Solution: Influence of Chloride Ion and Temperature

Passivity breakdown on carbon steel, which is the reference candidate material for fabricating high-level nuclear waste containers in Belgium, has been studied and the data are interpreted in terms of the point defect model (PDM). The near normal distribution in breakdown potential, measured in deaerated simulated concrete pore water solution (pH 12.5) at different chloride concentrations and temperatures are in satisfactory agreement with prediction of the PDM. An increase in temperature from 25°C to 85°C results in a decrease in localized corrosion resistance according to a linear relationship between the critical breakdown voltage, Eb , and log(chloride activity) for all chloride concentrations. The linear dependence of the passivity breakdown potential on the square root of potential scan rate (ν1/2), as predicted by the PDM, yields an estimate of the critical areal concentration of condensed cation vacancies at the metal/film interface (ξ<=5.45×1014cm-2 at 85oC and <=1.07×1014cm-2 at 55oC ) that leads to the passivity breakdown. The experimental distributions in the breakdown potential for carbon steel at 55oC and 85oC are in satisfactory agreement with the relevant calculated distributions. Acknowledgment: The authors gratefully acknowledge the support of this work by ONDRAF-NIRAS of Belgium.

Theoretical investigation of the evolution of the passive state on Alloy 22 in acidified, saturated brine under open circuit conditions

It is observed in many systems that as a metal or alloy is exposed to an aqueous solution over a long period of time, the corrosion potential drifts in the positive direction and the corrosion rate decreases. This implies the possibility that the potential may become sufficiently positive that it will exceed the transpassive potential and result in depassivation of the surface and hence to massive corrosion. On the other hand, if depassivation does not occur, the corrosion rate is predicted to continuously decrease with time, such that experiments carried out over short periods of time may seriously over-estimate the true corrosion rate over extended periods. This is an important issue in the selection of materials for the containment of high-level nuclear waste (HLNW), because of the excessively long performance horizon (up to 10 6 years), and in interpreting corrosion data from archeological artifacts. The theory developed in this work provides a physico-electrochemical explanation of the ennoblement phenomenon that is in good agreement with experiment and, in principle, allows one to optimize the model on short-term data and then predict the corrosion behavior over the very long term.

Determination of Crevice Corrosion Susceptibility of Alloy 22 Using Different Electrochemical Techniques

AbstractAlloy 22 belongs to the Ni-Cr-Mo family and it is highly resistant to general and localized corrosion. It may suffer crevice corrosion in aggressive environmental conditions. This alloy has been considered as a corrosion-resistant barrier for high-level nuclear waste containers. It is assumed that localized corrosion may occurs when the corrosion potential (ECORR) is equal or higher than the crevice corrosion repassivation potential (ER,CREV). The latter is measured by means of different electrochemical techniques using artificially creviced specimens. These techniques include cyclic potentiodynamic polarization (CPP) curves, Tsujikawa-Hisamatsu electrochemical (THE) method or other non-standard methods, such as the PD-GS-PD technique.The aim of the present work was to determine reliable critical or protection potentials for crevice corrosion of Alloy 22 in pure chloride solutions at 90°C. Conservative methodologies (which include extended potentiostatic steps) were applied for determining protection potentials below which crevice corrosion cannot initiate and propagate. Results from PD-GS-PD technique were compared with those from these methodologies in order to assess their reliability. Results from the CPP and the THE methods were also considered for comparison. The repassivation potential resulting from the PD-GS-PD technique was conservative and reproducible, and it did not depend on the amount of previous crevice corrosion propagation.

Th diffusion in monazite

Diffusion of Th has been characterized in synthetic monazite under dry conditions. The synthetic monazites (either pure CePO 4, NdPO 4, or a mixed LREE phosphate containing Ce, Nd, and Sm) were grown via a Na 2CO 3–MoO 3 flux method. The sources of diffusant for the experiments were either synthesized ThSiO 4 or CaTh(PO 4) 2 powders mixed with CePO 4. Experiments were performed by placing source and monazite in Pt capsules and annealing capsules in 1-atm furnaces for times ranging from 3 weeks to a few hours, at temperatures from 1350 to 1550 °C. The Th distributions in the monazite were profiled by Rutherford Backscattering Spectrometry (RBS). The following Arrhenius relation was obtained for Th diffusion in monazite: D = 6.7 × 10 1 exp ( − 740 ± 63 kJ mol − 1 / R T ) m 2 s − 1 The diffusivity of Th was found to be similar for monazites containing a single REE (Ce or Nd) and the mixed LREE phosphates. Th diffusion was also similar for experiments run using the Th silicate and Ca–Th phosphate sources, suggesting that the substitutional mechanism for Th in monazite, i.e., Th + 4 + Si + 4 → REE + 3 + P + 5 for the ThSiO 4 source, and Th + 4 + Ca + 2 → 2REE + 3 for the CaTh(PO 4) 2 source, does not significantly affect Th diffusivities, and that Th is likely the rate-limiting species in diffusion. Th diffusion in monazite is about 3 orders of magnitude slower than Pb diffusion [Cherniak, D.J., Watson, E.B., Grove, M., Harrison, T.M., 2004a. Pb diffusion in monazite: a combined RBS/SIMS study. Geochimica et Cosmochimica Acta 68, 829–840.]. This contrasts with findings of Gardés et al. [Gardés, E., Jaoul, O., Montel, J.-M., Seydoux-Guillaume, A.-M., Wirth, R., 2006. Pb diffusion in monazite: An experimental study of Pb 2 + + Th 4+ → 2Nd 3+ interdiffusion. Geochimica et Cosmochimica Acta 70, 2325–2336.] who determined that diffusivities measured in (Pb, Th) ↔ REE exchange (which would under most circumstances be rate-limited by the slowest-diffusing species) are comparable to those measured by Cherniak et al. [Cherniak, D.J., Watson, E.B., Grove, M., Harrison, T.M., 2004a. Pb diffusion in monazite: a combined RBS/SIMS study. Geochimica et Cosmochimica Acta 68, 829–840.] for Pb. Th diffusion in zircon [Cherniak, D.J., Hanchar, J.M., Watson, E.B., 1997. Diffusion of tetravalent cations in zircon. Contributions to Mineralogy and Petrology 127, 383–390.] is about an order of magnitude slower than in monazite, but with similar activation energy for diffusion. The lower diffusivities in zircon may be a consequence of the larger disparity in size between Th and the Zr site in zircon as compared with Th and the REE site in monazite. Nonetheless, Th is essentially immobile in monazite with respect to exchange by volume diffusion under most geologic conditions. These findings may also have implications for containment of high-level actinide-based nuclear waste in monazite ceramic waste forms, but it should be noted that mechanisms other than diffusive exchange (e.g., recrystallization, dissolution/reprecipitation) may be important to consider along with diffusion when evaluating the retentivity of monazite for Th and other actinide elements.

The Electrochemical Response of Preoxidized Copper in Aqueous Sulfide Solutions

The conversion of a film on copper to in aqueous sulfide solutions has been followed using a combination of electrochemical techniques and in situ Raman spectroscopy. Oxide films were electrochemically grown in alkaline solutions and their composition and morphology determined using Raman spectroscopy and scanning electron microscopy. Although corrosion potential measurements indicate that the aqueous sulfide solution rapidly penetrates the porous layer, the oxide-to-sulfide reaction appears to proceed chemically at the oxide/solution interface rather than via the galvanic coupling of Cu oxidation to and reduction to Cu. In situ Raman spectroscopy confirms that the sulfide formed is , and cathodic stripping voltammetry shows that the reaction is initially rapid and then proceeds at a constant rate until the conversion is complete. Comparison of the amounts of oxide initially present and sulfide eventually formed demonstrates that the conversion is 100% efficient. These studies are part of a larger project to determine the important corrosion processes on copper high-level nuclear waste containers exposed to anoxic aqueous sulfide containing groundwaters.

Prediction of Corrosion Behavior of HLW Containers in the Framework of the French Interim Storage Concept

In the framework of a law voted in 1991, it has been decided that the management of French high-level nuclear waste (HLW) should be studied along three main lines: (a) transmutation of actinides in fast breeder reactors, (b) long-term interim storage of waste containers (whether containing vitrified waste or spent fuel elements canisters) for a maximum period of 300 yr, and (c) geological disposal of the containers after the interim storage period.The last two lines of research, among other problems, bring up the question of the long-term corrosion behavior of the containers. The Commissariat à l’Énergie Atomique (CEA) is in charge of the researches concerning the transmutation of the actinides and the long-term interim storage.As interim storage is concerned, extended studies are being carried out at the CEA in order to predict the environmental conditions of an interim storage facility and the consequent corrosion of the HLW containers over a period of 300 yr.A dry corrosion phase of a minimum duration of 100 yr was identified, followed by an atmospheric corrosion phase of a maximum duration of 200 yr. Phenomenological models were developed to obtain an estimate of the total corrosion damage during interim storage.Results led to a conservative estimate of a corroded thickness of metal comprised between 0.35 and 1 mm, hence, <2% of the wall thickness of the overpacks.Because of the extended period of time considered, this estimate, partly based on the extrapolation of experimental results, has to be reinforced. Mechanistic modeling of the elementary processes is currently in progress and has already given encouraging results for both the dry and atmospheric corrosion phases.

Study of the Susceptibility of Oxygen-Free Phosphorous Doped Copper to Corrosion in Simulated Groundwater in the Presence of Chloride and Sulfide

AbstractCopper of the quality oxygen free and high conductivity, doped with phosphorus (Cu OFP) has been chosen as the material for the fabrication of high level nuclear waste containers in Sweden. This material will be the corrosion barrier for spent fuel in the environment of a deep geological repository. It is planned that the service life of this container will be 100,000 years. During this time water with high concentration of chloride ions might come in contact with the copper surface. If pH conditions are appropriate, this might cause pitting corrosion. This work reports a study of the susceptibility of Cu OFP to corrosion when chloride ions are present, both deionized water (DW) and in standard synthetic underground water (SUW). The techniques used were electrochemical methods such as corrosion potential evolution and Tafel curves. In addition, this system was studied with Electrochemical Impedance Spectroscopy (EIS). We also used characterization techniques such as Scanning Electronic Microscopy (SEM), Energy Disperse Spectroscopy (EDS) . The main conclusions are that copper is more susceptible to corrosion at high chloride ion concentration. Additionally, when the chloride concentration is low, it is possible to form copper chloride crystals, but at the highest concentration, copper chloride complexes are formed, leaving the copper surface without deposits. When the chloride concentration is low (&lt;0.1 M) the corrosion process is mainly controlled by diffusion, while at higher concentrations (0.1M to 1M) corrosion is controlled by transport processes. At low concentration of sulfide ( &lt;3*10-5M), copper corrosion in the presence of chloride is controlled by diffusional processes.

Influence of halide ions and alloy microstructure on the passive and localized corrosion behavior of alloy 22

Alloy 22 (N06022) is the current candidate alloy used to fabricate the external wall of the high-level nuclear waste containers for the Yucca Mountain repository. It was of interest to study and compare the general and localized corrosion susceptibility of Alloy 22 in fluoride and chloride solutions at 90 °C. Standard electrochemical tests such as cyclic potentiodynamic polarization, amperometry, and electrochemical impedance spectroscopy were used. Studied variables included the solution pH and the alloy microstructure (thermal aging). Results show that Alloy 22 is highly resistant to general corrosion in all the solutions tested. Thermal aging is not detrimental and even seems to be slightly beneficial for general corrosion at the higher solution pHs. Pitting corrosion was never observed. Crevice corrosion was found only for high chloride-containing solutions after anodic polarization. The presence of fluoride ions together with chloride ions seems to increase the susceptibility of Alloy 22 to crevice corrosion compared to pure chloride solutions.

General corrosion of alloy 22: Experimental determination of model parameters from electrochemical impedance spectroscopy data

Models used to predict general corrosion damage of Alloy 22 high level nuclear waste (HLNW) containers must be deterministic, relying upon the time-invariant natural laws, because of long time scales involved in the predictions compared with the time over which the corrosion of this alloy has been studied. The point defect model (PDM) is one such model and requires high-accuracy experimental data to determine model parameters that will accurately predict corrosion over the long times required for HLNW disposal. Electrochemical impedance spectroscopy (EIS) and steady-state polarization data were collected in deaerated 4 M NaCl, pH 6 solutions at 30 °C and 60 °C. This work, which was carried out under a quality assurance program, emphasized data reproducibility and adherence to the steady-state condition in order to ensure confluence between theory and experiment. Thus, fresh electrolyte was continuously flowed through test cells allowing test specimens to remain undisturbed for up to 6 weeks during data collection. Parameters for the PDM have been estimated by optimizing the PDM on the impedance data, and the ability of the parameter values to account for the steady-state passive current density and barrier oxide layer thickness has been evaluated using phase space analysis. The calculated impedance gives generally good agreement with experimental data showing that the PDM is a valid approach.

Passive Dissolution and Localized Corrosion of Alloy 22 High-Level Waste Container Weldments

Localized corrosion of high level nuclear waste containers is considered an important factor that will have a strong influence on the overall performance of the proposed repository at Yucca Mountain, NV. The present candidate container material, Alloy 22 [UNS N06022 (57Ni-22Cr-13.5Mo-3W-3Fe)], is highly resistant to localized corrosion. Assessing the performance of the HLW containers also requires an evaluation of localized corrosion resistance and determination of the passive corrosion rate of Alloy 22 weldments. The localized corrosion resistance of welded Alloy 22 specimens was evaluated by measuring the repassivation potential in chloride containing solutions whereas potentiostatic tests were used to determine the passive corrosion rate of the welded material. The results for the welded material are compared to those for the base metal.

Methodologies for Predicting the Performance of Ni-Cr-Mo Alloys Proposed for High Level Nuclear Waste Containers

AbstractFor the geologic disposal of the high level nuclear waste (HLW), aqueous corrosion is considered to be the most important factor in the long-term performance of containers, which are the main components of the engineered barrier subsystem. Container life, in turn, is important to the overall performance of the repository system. The proposed container designs and materials have evolved to include multiple barriers and highly corrosion resistant Ni-Cr-Mo alloys, such as Alloys 625 and C-22. Calculations of container life require knowledge of the initiation time and growth rate of localized corrosion. In the absence of localized corrosion, the rate of general or uniform dissolution, given by the passive current density of these materials, is needed. The onset of localized corrosion may be predicted by using the repassivation and corrosion potentials of the candidate container materials in the range of expected repository environments. In initial corrosion tests, chloride was identified as the most detrimental anionic species to the performance of Ni-Cr-Mo alloys. Repassivation potential measurements for Alloys 825, 625, and C-22, conducted over a wide range of chloride concentrations and temperatures, are reported. In addition, steady state passive current density, which will determine the container lifetime in the absence of localized corrosion, was measured for Alloy C-22 under various environmental conditions.

Long-Term Prediction of Localized Corrosion of Alloy 825 in High-Level Nuclear Waste Repository Environments

Abstract Long-term prediction of localized corrosion of high-level nuclear waste container materials is a necessary step in the performance assessment of the engineered barrier system. Localized corrosion of corrosion-resistant materials may occur if the containers are exposed to chloride (Cl−)-containing ground water at elevated temperatures. Potentiostatic tests conducted on alloy 825 (UNS N08825), a candidate container material, have shown that the potential at which localized corrosion can be initiated is a function of surface conditions and exposure time. The initiation potentials for localized corrosion measured in short-term tests with polished specimens overestimated the long-term initiation potentials by as much as 300 mV. On the other hand, the repassivation potential obtained from short-term tests on deep pits can be used to predict the initiation of localized corrosion in long-term tests.

Effects of Surface Chromium Depletion on Localized Corrosion of Alloy 825 as a High-Level Nuclear Waste Container Material

Abstract Effects of the chromium-depleted, mill-finished surface on the localized corrosion resistance of alloy 825 (UNS N08825) were investigated. Tests were conducted in solutions based on the ground water at Yucca Mountain, Nevada, but with a higher concentration of chloride (Cl−). Results indicated that breakdown (Ep) and repassivation (Erp) potentials for mill-finished surfaces were more active than those for polished surfaces. Potentiodynamic polarization tests indicated pits could be initiated on the chromium-depleted surface at potentials of 220 mVSCE in a solution containing 1,000 ppm Cl− at 95°C. Potentiostatic tests identified a similar pit initiation potential for the mill-finished surface. However, under long-term potentiostatic tests, a higher potential of 300 mVSCE was needed to sustain stable pit growth beyond the chromium-depleted layer. An increase in surface roughness also was observed to decrease localized corrosion resistance of the material.