Repository Environment Research Articles

Clay as sealing material in nuclear waste repositories

Nuclear waste from thermal plants poses a lasting risk to the biosphere because of its long radioactive life. The planned definitive storage place for it is in deeply buried repositories. Such repositories would need to be both impermeable to water, and plastic during deformation, in order to avoid the formation of cracks that may allow water in. One of the clay minerals, smectite, has these two properties and is an ideal candidate as a sealing material or even host rock for nuclear waste repositories. The chemical stability of smectite in the repository environment is sufficient to maintain good sealing properties during the active life of the relevant radionuclides.

Design and characterization of low-heat and low-alkalinity cements

Investigations were carried out to formulate and characterize low-alkalinity and low-heat cements which would be compatible with an underground waste repository environment. Several systems comprising Portland cement, a pozzolan (silica fume or fly ash) and blastfurnace slag were compared. All blends were characterized by high amounts of additions, the Portland cement (PC) fraction ranging only from 20 to 60%. Cement hydration was studied using several techniques: X-ray Diffraction, TGA-DTA, calorimetry, pore solution extraction and microscopy. The most important result obtained with some ternary blends was the drop in the pore solution pH by more than one unit as compared with control samples elaborated with commercial cements. The alkali content (≈ 1 to 4 mmol/L) of the interstitial solution was also strongly reduced. The blends exhibited a low-heat output as required. Leaching tests carried out in pure water indicated a very slow decalcification of the samples. Several techniques such as optical microscopy, SEM/BSE, X-ray microanalysis or X-ray diffraction were compared to estimate the degraded thickness.

Inner Material Requirements and Candidates Screening for Spent Fuel Disposal Canister

AbstractIn the context of the present Spanish ‘once-through’ nuclear fuel cycle, the need arises to complete the geological repository reference concept with a spent fuel canister final design. One of the main issues in its design is selecting the inner material to be placed inside the canister, between the steel walls and the spent fuel assemblies. The primary purpose of this material will be to avoid the possibility of a criticality event once the canister walls have been finally breached by corrosion and the spent fuel is flooded with groundwater. That is an important role because the increase in heat generation from such an event would act against spent fuel stability and compromise bentonite barrier functions, negatively affecting overall repository performance. To prevent this possibility a detailed set of requirements for a material to fulfil this role in the repository environment have been devised and presented in this paper. With these requirements in view, eight potentially interesting candidates were selected and evaluated: cast iron or steel, borosilicate glass, spinel, depleted uranium, dehydrated zeolites, haematite, phosphates, and olivine. Among these, the first four materials or material families are found promising for this application. In addition, other relevant non-performance-related aspects of candidate materials, which could help on decision making, are also considered and evaluated.

Influence of the Evolution of the Surface Area Value on the Spent Nuclear Fuel Dissolution Rate for Performance Assessment Studies

AbstractThis paper focuses on how to extrapolate current knowledge of spent fuel matrix alteration processes from laboratory to repository conditions, i.e., the influence of changes in both the initial surface oxidation level and the evolution of the specific surface area during the alteration process. Therefore, a spent fuel matrix alteration model allowing the alteration rate evolution to be predicted as a function of both the host rock considered and evaluation time scale of interest is described. At present, the model assumes that alteration of the spent fuel will start when the groundwater reaches the solid surface and that only the radiolytic species of the groundwater (oxidants generated by a-radiation of spent fuel) will produce the surface oxidation process and subsequent matrix dissolution; O2, H2O2 and OH are the species that react with UO2(s) for oxidation of the pellet surface. The dissolution process of the surface sites that are oxidized is modelled in two steps: first, a surface co-ordination of the oxidized layer with aqueous ligands and, second, detachment (dissolution) of the product species. Taking this mechanism into account, the model gives the evolution of the spent fuel matrix alteration rate over periods as long as 1,000,000 years. In this work is focussed on input the experimental results obtained of UO2 surface area behaviour (presented in previous MRS conference), on the MAM model. The matrix alteration rate results obtained, with MAM model, for repository granitic environment will be presented and compared to those performed for SFS project. Furthermore, a sensitivity analysis study has been performed on the influence of the following variables: Influence of the initial power size distribution and the initial oxidation state

Localized corrosion of alloy 22 in the potential Yucca Mountain repository environment

The U.S. Department of Energy (DOE) has indicated that it may use Alloy 22 as the waste package outer container material for the potential high-level waste repository at Yucca Mountain, Nevada. Additionally, a drip shield, made of titanium grade 7 and titanium grade 29, may extend the length of the emplacement drifts to enclose the top and sides of the emplaced waste package. Localized corrosion in the form of crevice corrosion could be one degradation process that may adversely affect the waste package performance. This paper will summarize the work conducted to evaluate the effects of environmental conditions relevant to the potential Yucca Mountain repository, metallurgical states (e.g., mill-annealed and welded plus solution annealed), and similar and dissimilar metal crevices on the crevice corrosion susceptibility of Alloy 22. This work also evaluates crevice corrosion propagation behavior resulting from contact with Alloy 22 or titanium grade 7.

The potential for the SCC of titanium alloys under repository-relevant environments for U.S. nuclear waste

Stress corrosion cracking (SCC) is one of the most common corrosion-related causes of premature failure of metal structural components. This paper presents an assessment of the performance of titanium alloys (titanium grade 7 for the drip shield and grade 29 for the structural support), which may be subjected to SCC under proposed U.S. nuclear waste repository environments. The threshold stresses for SCC initiation, SCC crack propagation rates, and the threshold stress intensity factors for titanium grades 7, 28, and 29 are discussed.

Formulation et caractérisation de bétons bas pH

ABSTRACT Investigations were carried out to formulate and characterize low-alkalinity and low-heat cements which would be compatible with an underground waste repository environment. Pozzolans (high-silica product such as fly ash and silica fume) and blast furnace slag were added to Portland cement to decrease the cement pore solution pH to 11. The alkali content of the interstitial solution (≈ 1 to 5 mmol/L) of blend was strongly reduced which decreased the pH by more than one unity as compared with control samples. The blends exhibited a low heat output (temperature rise below 20°C). The prepared concrete specimens had a compressive strength higher than 70 MPa after one year of curing in 100% RH, a low shrinkage (−350 to −500μm/m) and it was possible to prepare workable concrete despite the very high amounts of pozzolans in the blends.

Hydrogen Diffusion and Trapping Effects in Low and Medium Carbon Steels for Subsurface Reinforcement in the Proposed Yucca Mountain Repository

The electrochemical hydrogen permeation method was used to investigate hydrogen transport, trapping characteristics of low (0.08 pct C) and medium carbon (0.44 pct C) steels proposed for the Yucca Mountain (YM) repository environment. The presence of relatively high amounts of C, Mn, and S increased the density of trapping sites in medium carbon steel. The measured diffusivity of medium carbon steel was lower than that of the low carbon steel due to increased trapping of hydrogen at irreversible sites in the medium carbon steel. Hydrogen concentration values obtained for low carbon steels in YM ground water electrolytes indicate that increased ionic concentration decreases the uptake of hydrogen. The decrease in hydrogen permeation were due the formation of CaCO3 corrosion products on the surface of steels.

Formulating a low-alkalinity, high-resistance and low-heat concrete for radioactive waste repositories

Investigations were carried out in order to formulate and characterize low-alkalinity and low-heat cements which would be compatible with an underground waste repository environment. Several systems comprising Ordinary Portland Cement (OPC), a fast-reacting pozzolan (silica fume (SF) or metakaolin (MK)) and, in some cases, a slow-reacting product (fly ash (FA) or blastfurnace slag (BFS)) were compared. Promising results were obtained with some binary mixtures of OPC and SF, and with some ternary blends of OPC, SF and FA or BFS: pH of water in equilibrium with the fully hydrated cements dropped below 11. Dependence of the properties of standard mortars on the high contents of FA and SF in the low-pH blends was examined. Combining SF and FA seemed attractive since SF compensated for the low reactivity of FA, while FA allowed to reduce the water demand, and dimensional variations of the mortars. Finally, low-heat (Δ T < 20 °C under semi-adiabatic conditions) and high strength (≈ 70–80 MPa) concretes were prepared from two low-pH cements: a binary blend made from 60% of OPC and 40% of SF, and a ternary blend including 37.5% OPC, 32.5% SF and 30% FA.

Comparative study of cesium sorption on crushed and intact sedimentary rock

For establishing a reliable safety assessment methodology for high-level radioactive waste (HLW) geological disposal, it is essential to clarify the applicability of sorption data determined with a conventional batch sorption method to an actual repository environment. In the present study, sorption data represented as distribution coefficients (K d) were compared experimentally by applying a batch method with crushed sedimentary rock samples and a dynamic circulation method with intact sedimentary rock samples. The K d values for cesium (Cs) determined with the dynamic circulation method are around one order of magnitude lower than those obtained using the batch method, depending on Cs concentration at sorption equilibrium. This disagreement was discussed considering permeability, specific surface area, solid/liquid ratio (S/L) and solution chemistry. The K d values determined with the dynamic circulation method were corrected for different specific surface areas. As a consequence, the K d values determined with the two methods are then comparable within an order of magnitude. The K d values become even more comparable when taking into account that they decrease with increasing the S/L ratio as it is the case when going from batch to column experiments. The results basically suggest that the K d determined with the batch method are reasonably applicable to describe sorption in the actual environment in a given nuclear waste repository after appropriate corrections.

Localized corrosion of Alloy C22 nuclear waste canister material under limiting conditions

Localized corrosion behavior of Alloy C22 in simulated Yucca Mountain (YM) repository environments was studied at the highest achievable but realistic temperatures under boiling and dripping scenarios. Temperatures measured in concentrated boiling solutions of KCl and NaNO 3 were found to be stable at 140 °C, although transient boiling before dryout was observed at temperatures as high as 160 °C, as the electrolyte became progressively more concentrated. Experiments that simulated a dripping scenario with simulated J13 well water confirmed the existence of concentrated solutions stable at 142 ± 3 °C under controlled drip conditions leading to pit initiation in Alloy C22 after only a few hours. The polarization experiments conducted at 140 °C in a solution with 0.5 mol L −1 chloride concentration showed that the critical potential for localized corrosion was 250 mV (versus Ag/AgCl). Potentiostatic tests confirmed that active metal dissolution occurred only in the crevice region at this potential. The crevice corrosion of Alloy C22 required an incubation time to develop a critical crevice solution, and it was triggered by severe local chemistry (enrichment of Cl − and H +) aided by the high temperature.

The Role of Evidence in Establishing Trust in Repositories

This article arises from work by the Digital Curation Centre (DCC) Working Group examining mechanisms to roll out audit and certification services for digital repositories in the United Kingdom. Our attempt to develop a program for applying audit and certification processes and tools took as its starting point the RLG-NARA Audit Checklist for Certifying Digital Repositories [8]. Our intention was to appraise critically the checklist and conceive a means of applying its mechanics within a diverse range of repository environments. We were struck by the realization that while a great deal of effort has been invested in determining the characteristics of a 'trusted digital repository', far less effort has concentrated on the ways in which the presence of the attributes can be demonstrated and their qualities measured. With this in mind we sought to explore the role of evidence within the certification process, and to identify examples of the types of evidence (e.g., documentary, observational, and testimonial) that might be desirable during the course of a repository audit.

Computation of Ni-Cr phase diagramvia a combined first-principles quantum mechanical and CALPHAD approach

A first-principles quantum-mechanical computational code has been used to compute the energy of formation for selected ordered and topologically close-packed (TCP) phases in a Ni-base alloy. The thermodynamic data are incorporated into an existing database for Ni alloys and used in conjunction with the CALPHAD approach to compute the binary Ni-Cr phase diagram. In addition, a thermodynamic model is used to treat long-range ordering and the formation of the Ni2Cr. The phase field for ordering is compared against that predicted by the Thermo-Calc to elucidate possible implications on the long-term phase stability of Ni-base alloys in a nuclear waste repository environment.

Representing digital assets usingMPEG-21 Digital Item Declaration

Various XML-based approaches aimed at representing compound digital assets have emerged over the last several years. Approaches that are of specific relevance to the digital library community include the Metadata Encoding and Transmission Standard (METS), the IMS Content Packaging XML Binding, and the XML Formatted Data Units (XFDU) developed by CCSDS Panel 2. The MPEG-21 Digital Item Declaration (MPEG-21 DID) is another standard specifying the representation of digital assets in XML that, so far, has received little attention in the digital library community. This article gives a brief insight into the MPEG-21 standardization effort, highlights the major characteristics of the MPEG-21 DID Abstract Model, and describes the MPEG-21 Digital Item Declaration Language (MPEG-21 DIDL), an XML syntax for the representation of digital assets based on the MPEG-21 DID Abstract Model. Also, it briefly demonstrates the potential relevance of MPEG-21 DID to the digital library community by describing its use in the aDORe repository environment at the Research Library of the Los Alamos National Laboratory (LANL) for the representation of digital assets.

Modelling of the spent fuel dissolution rate evolution for repository conditions. Matrix Alteration Model results and sensitivity analysis

ABSTRACTThis paper focuses on how to extrapolate current knowledge of spent fuel matrix alteration processes from laboratory to repository conditions, i.e., the influence of changes in both the environmental conditions and the range of time scale considered. Therefore, a spent fuel matrix alteration model allowing the alteration rate evolution to be predicted as a function of both the host rock considered and evaluation time scale of interest is described.At present, the model assumes that alteration of the spent fuel will start when the groundwater reaches the solid surface and that only the radiolytic species of the groundwater (oxidants generated by α-radiation of spent fuel) will produce the surface oxidation process and subsequent matrix dissolution; O2, H2O2 and OH· are the species that react with UO2(s) for oxidation of the pellet surface. The dissolution process of the surface sites that are oxidized is modelled in two steps: first, a surface co-ordination of the oxidized layer with aqueous ligands and, second, detachment (dissolution) of the product species. Taking this mechanism into account, the model gives the evolution of the spent fuel matrix alteration rate over periods as long as 1,000,000 years.In this work the matrix alteration rate results obtained for two repository environments, granitic and argillaceous, will be presented. Furthermore, a sensitivity analysis study has been performed on the influence of the following variables: type of spent fuel considered, α-dose rate evolution, α-range in groundwater, carbonate and iron concentration in groundwater, H2 partial pressure, container time failure and specific surface area of the pellet.

Chemical durability of high-level waste glass in repository environment: main conclusions and remaining uncertainties from the GLASTAB and GLAMOR projects

ABSTRACTThis paper reviews the main conclusions of two European Commission funded projects, GLASTAB and GLAMOR, on the durability of HLW glass in geological disposal, and the remaining uncertainties. The progress in GLASTAB relates to the characterization of the alteration layer, the modeling of glass dissolution, the interaction of glass with near field materials, the behaviour of radionuclides, and the calculation of HLW glass dissolution as part of geological disposal. The GLAMOR programme was focused on the decrease of the glass dissolution rate upon achieving silica saturation in solution. Two conclusions from this programme are the importance of the residual rate measured beyond silica saturation and the need to reduce the uncertainties in model parameters. The overall conclusion from the two projects is that strong progress has been achieved on basic dissolution mechanisms and interaction with disposal environments - HLW glass will act as a good barrier to the release of radionuclides in geologic disposal.

Testing of a near-field biogeochemical model against data from a large-scale gas generation experiment

ABSTRACTA biogeochemical model that represents processes of metal corrosion, microbial degradation of cellulosic waste and mass transfer within a heterogeneous system has been used to represent processes of gas generation in a large-scale (20m3) experiment that has studied degradation of typical nuclear reactor operating waste. The experiment has been in operation for a period of about eight years and has established a pattern of methanogenic gas generation. A “blind testing” approach has been used to develop the model of the experiment using independently derived kinetic data for corrosion and microbial processes. The model correctly represents the anaerobic conditions leading to methane generation during the course of the experiment. The overall rate of gas generation of the experiment is well represented, as is the composition of evolved gases and geochemistry of sampled liquids. The experiment and the model together build confidence in the ability to simulate processes of gas generation and variation in chemical conditions in heterogeneous repository environments.

Erratum: “A Review of Corrosion of Titanium Grade 7 and Other Titanium Alloys in Nuclear Waste Repository Environments” (Vol. 61, No. 10, p. 987–1,003)

Erratum: “A Review of Corrosion of Titanium Grade 7 and Other Titanium Alloys in Nuclear Waste Repository Environments” (Vol. 61, No. 10, p. 987–1,003)

A Review of Corrosion of Titanium Grade 7 and Other Titanium Alloys in Nuclear Waste Repository Environments

Abstract Titanium alloy degradation modes are reviewed in relation to their performance in repository environments. General corrosion, localized corrosion, stress corrosion cracking, hydrogen-induced cracking, microbiologically influenced corrosion, and radiation-assisted corrosion of Ti alloys are considered. With respect to the Ti Grade 7 drip shields selected for emplacement in the repository at Yucca Mountain, general corrosion, hydrogen-induced cracking, and radiation-assisted corrosion either do not initiate or propagate sufficiently fast over a 10,000-year time frame in likely repository environments to cause penetration of the drip shield. Stress corrosion cracking (in the absence of disruptive events) is of no consequence to barrier performance; and localized corrosion and microbiologically influenced corrosion are not expected to occur. To facilitate the discussion, Ti Grades 2, 5, 7, 9, 11, 12, 16, 17, 18, and 24 are included in this review.

IODP's Untapped Wealth: Multi-Parameter Logging of Legacy Core

Since 1968, the Deep - Sea Drilling Project ( DSDP) and the Ocean Drilling Program (ODP) have recovered and stored approximately 300 km of core. Half of every core has been kept as an archive, normally only available for viewing. These archive half-cores are well suited for automated non-destructive geophysical measurements (core logging), including many parameters that provide essential data for reconstructing Earth’s climatic history, such as high-resolution magnetic susceptibility, natural gamma spectroscopy, and UV/VIS/IR spectrophotometry. We recently used a Geotek MSCL -X YZ core logger at the IODP West Coast Repository to log archive core halves recovered by the Glomar Challenger in 1983. The MSCL-XYZ system is specifically designed to allow multiparameter, non-destructive geophysical data to be collected easily at high spatial resolutions on up to nine split-core sections between reloading. This enables the machine to be loaded and run unattended for periods of many hours (including overnight and weekends), making it well suited for logging of archive core in a repository environment. The immediate goal was to obtain a high-resolution paleoclimate record for DSDP Site 594, east of New Zealand in the Southwest Pacific, but our underlying intention was to open up the vast reservoir of paleoclimate and other data that await extraction from well-preserved archive-half cores from the previous scientific ocean drilling programs, and now in IODP custody.