Waste Repository In Salt Research Articles

Comparing modelling approaches for a generic nuclear waste repository in salt

This paper contains a comparison of five modelling approaches for a simplified nuclear waste repository in a domal salt formation. It is the result of a four-year collaboration between five international teams on Task F of the DECOVALEX-2023 project on performance assessment modelling. The primary objectives of Task F are to build confidence in the models, methods, and software used for performance assessment (PA) of deep geologic nuclear waste repositories, and/or to bring to the fore additional research and development needed to improve PA methodologies. This work demonstrates how these objectives are accomplished through staged development and comparison of the models and methods used by participating teams in their PA frameworks. Participating teams made a wide range of model assumptions, ranging from compartmentalized networks to full 3D models of the salt formation and repository. Despite differences in the modelling strategies, all models indicate that salt compaction and diffusion of radionuclides in brine are key processes in the repository. For the isothermal spent nuclear fuel and vitrified waste scenario with multiple early failures considered, all models indicate little of the disposed radionuclides will migrate beyond the repository seal over the 100,000-year simulations. In general, the model output quantities have the largest differences over the short term and near the waste. Disparities between the models are believed to be due to differing simplifications from the conceptual model.

Impact of the compaction behavior of crushed salt on the thermo-hydro-mechanical response of a generic salt repository for heat-generating nuclear waste

Salt formations are a promising host rock for the disposal of heat-emitting nuclear waste due to their ability to heal fractures, high thermal conductivity, and near-zero permeability. This study investigates the concept of waste emplacement in salt formations by placing waste packages in drifts and backfilling them with crushed salt. The aim is to minimize fracturing and restore the initial properties of the host rock.One of the key issues in assessing the safety and performance of nuclear waste repositories in salt is determining the reconsolidation rate of the crushed salt backfill, which has been traditionally predicted using the C-WIPP model. However, this model was calibrated using data from uniaxial oedometer lab tests, where specimens with high initial porosity values have been compacted to medium porosity values. The observed compaction behavior has then been extrapolated to the low porosity range, leading to a prediction of rapid full reconsolidation of the backfill in less than 20 years. In contrast, the new C-WIPP/TUC model used in this study was calibrated for the medium porosity range, between 17% and 8%, based on more suitable triaxial lab tests. Using coupled Thermo-Hydro-Mechanical (THM) simulations with the TOUGH-FLAC simulator, it was found that the C-WIPP/TUC model predicts a slower reconsolidation rate of the backfill compared to the C-WIPP model, with an average porosity of the backfill remaining roughly 10 times that of natural salt after 10,000 years. The effects of the relatively slow reconsolidation of the crushed salt backfill on the THM behavior of the repository are also examined, highlighting the importance of accurately capturing the behavior of crushed salt for the study of the long-term integrity of a nuclear waste repository in salt.

Deformation-assisted fluid percolation in rock salt.

Deep geological storage sites for nuclear waste are commonly located in rock salt to ensure hydrological isolation from groundwater. The low permeability of static rock salt is due to a percolation threshold. However, deformation may be able to overcome this threshold and allow fluid flow. We confirm the percolation threshold in static experiments on synthetic salt samples with x-ray microtomography. We then analyze wells penetrating salt deposits in the Gulf of Mexico. The observed hydrocarbon distributions in rock salt require that percolation occurred at porosities considerably below the static threshold due to deformation-assisted percolation. Therefore, the design of nuclear waste repositories in salt should guard against deformation-driven fluid percolation. In general, static percolation thresholds may not always limit fluid flow in deforming environments.

Disturbed Rock Zone Geomechanics at the Waste Isolation Pilot Plant

The disturbed rock zone constitutes an important geomechanical element of the Waste Isolation Pilot Plant. The science and engineering underpinning the disturbed rock zone provide the basis for evaluating ongoing operational issues and their impact on performance assessment. Contemporary treatment of the disturbed rock zone applied to the evaluation of the panel closure system and to a new mining horizon improves the level of detail and quantitative elements associated with a damaged zone surrounding the repository openings. Technical advancement has been realized by virtue of ongoing experimental investigations and international collaboration. Initial sections summarize and document theoretical and experimental results, which quantify characteristics of the disturbed rock zone as applied to nuclear waste repositories in salt. This information is then applied to operational issues pertaining to recertification of the repository.

Analyses of mass transport in a nuclear waste repository in salt : Hwang, Y; Pigford, T H; Chambre, P L; Lee, W W L Water Resour ResV28, N7, July 1992, P1857–1868

Analyses of mass transport in a nuclear waste repository in salt : Hwang, Y; Pigford, T H; Chambre, P L; Lee, W W L Water Resour ResV28, N7, July 1992, P1857–1868

Analyses of mass transport in a nuclear waste repository in salt

Salt is the proposed host rock for geologic repositories of nuclear waste in several nations because it is “dry” and probably “impermeable”. To evaluate the safety of nuclear waste disposal in salt, it is necessary to calculate the rate of release of nuclides from solidified waste form and to predict the transport of released radionuclides. Our objective is to develop analytic methods for such prediction, based on mass transfer theory. First, we present an analysis of pressure‐driven brine migration. After consolidation, brine migration analyzed by these equations and using the particular set of parameter values is of a small magnitude, with Darcy velocities of the order of micrometers per year. Brine migration in consolidated salt is very localized, within a few meters from the waste package, and highly transient. We have analyzed diffusion of radionuclides near waste packages in a salt repository, using analytic equations for low Peclet systems. According to this analysis it is unlikely that any low‐solubility species will have difficulty meeting the U.S. Nuclear Regulatory Commission release rate requirement at the bare waste/salt interface. We have also analyzed the role that an interbed might play in being a preferential pathway for radionuclide migration. We calculate the diffusive mass fluxes from an infinitely long bare waste cylinder in salt, facing an interbed. The fractional release rates of solubility‐limited species into the interbed are low for the parameter values used in the numerical illustrations.

886463 Exploratory shaft design for a high-level waste repository in salt : Poppen, S A G; Cooley, A I; Mirza, M B Tunnlg Underground Space TechnolV3, N2, 1988, P183–192

886463 Exploratory shaft design for a high-level waste repository in salt : Poppen, S A G; Cooley, A I; Mirza, M B Tunnlg Underground Space TechnolV3, N2, 1988, P183–192

Exploratory shaft design for a high-level waste repository in salt

Abstract The authors describe the exploratory shaft design for a high-level nuclear waste repository in salt. The exploratory shaft facility in salt is one of three facilities—one each in basalt, tuff and salt—to be developed for the Office of Civilian Radioactive Waste Management Program of the U.S. Department of Energy. The authors discuss design criteria for the shaft; the influence of the geotechnical conditions on shaft lining; the shaft construction process; the two types of shaft lining to be used; and determination of the loads on the shaft lining.

Critical aspects of mining technology in excavations of a nuclear waste repository in salt : Djahanguiri, F Proc ISRM Symposium on Design and Performance of Underground Excavations, Cambridge, 3–6 Sept 1984P327–336. Publ London: British Geotechnical Society, 1984

Critical aspects of mining technology in excavations of a nuclear waste repository in salt : Djahanguiri, F Proc ISRM Symposium on Design and Performance of Underground Excavations, Cambridge, 3–6 Sept 1984P327–336. Publ London: British Geotechnical Society, 1984

Wipp Waste Package Testing on Simulated DHLW: Emplacement

AbstractSeveral series of simulated (nonradioactive) defense high-level waste (DHLW) package tests have recently been emplaced in the WIPP, a research and development facility authorized to demonstrate the safe disposal of defense-related wastes. The primary purpose of these 3-to-7 year duration tests is to evaluate the in situ materials performance of waste package barriers (canisters, overpacks, backfills, and nonradioactive DHLW glass waste form) for possible future application to a licensed waste repository in salt. This paper describes all test materials, instrumentation, and emplacement and testing techniques, and discusses progress of the various tests.These tests are intended to provide information on materials behavior (i.e., corrosion, metallurgical and geochemical alterations, waste form durability, surface interactions, etc.), as well as comparison between several waste package designs, fabrications details, and actual costs.These experiments involve 18 full-size simulated DHLW packages (approximately 3.0 m x 0.6 m diameter) emplaced in vertical boreholes in the salt drift floor. Six of the test packages contain internal electrical heaters (470 W/canister), and were emplaced under approximately reference DHLW repository conditions. Twelve other simulated DHLW packages were emplaced tinder accelerated-aging or overtest conditions, including the artificial introduction of brine, and a thermal loading approximately three to four times higher than reference. Eight of these 12 test packages contain 1500 W/canister electrical heaters; the other four are filled with DHLW glass.

The Salton Sea Geothermal Field as a Natural Analog for the Near-Field in a Salt High-Level Nuclear Waste Repository

AbstractThe Salton Sea Geothermal Field (SSGF), on the delta of the Colorado River in southern California, is being studied as a natural analog for the near-field environment of proposed nuclear waste repositories in salt. A combination of mineralogical and geochemical methods is being employed to develop a three dimensional picture of temperature, salinity, lithology, mineralogy, and chemistry of reactions between the reservoir rocks and the hot brines. Our aim is to obtain quantitative data on mineral stabilities and on mobilities of the naturally occurring radionuclides of concern in Commercial High-Level Waste (CHLW). These data will be used to validate the EQ3/6 geochemical code under development to model the salt near-field repository behavior.Maximum temperatures encountered in wells in the SSGF equal or exceed peak temperatures expected in a salt repository. Brines produced from these wells have major element chemistry similar to brines from candidate salt sites. Relative to the rocks, these brines are enriched in Na, Mn, Zn, Sr, Ra and Po, depleted in Ba, Si, Mg, Ti, and Al, and strongly depleted in U and Th. However the unaltered rocks contain only about 2–3 ppm of U and 4–12 ppm of Th, largely in detrital epidotes and zircons. Samples of hydrothermally altered rocks from a wide range of temperature and salinity show rather similar uniform low concentrations of these elements, even when authigenic illite, chlorite, epidote and feldspar are present. These observations suggest that U and Th are relatively immobile in these hot brines. However Ra, Po, Cs and Sr are relatively mobile. Work is continuing to document naturally occurring radionuclide partitioning between SSGF minerals and brine over a range of temperature, salinity, and lithology.

Diffusion of Colloids and Other Waste Species in Brine–Saturated Backfill Materials

ABSTRACTThe diffusion of plutonium (IV) and pertechnetate and the migration of colloidal gold in brine-saturated bentonite was measured. High ionic strength brine characterizes potentially intruding groundwater for radioactive waste repositories in salt. Plastic diffusion cells containing cylindrical diffusion columns were used for low density bentonite. Metal diffusion cells constructed entirely from Hastelloy C-276 were used for the diffusion of pertechnetate in highly compacted bentonite. Apparent distribution coefficients calculated from plutonium diffusion in two columns of low density bentonite are 2 m3/kg and 3 m3/kg. These values are in good agreement with the value of 3m3/kg that was calculated from previous batch sorption data. Pertechnetate anions migrated out of a brine-saturated high density bentonite diffusion column at rates that are too large to be rationalized with a simple diffusion theory. Additional measurements that take into account possible channeling of pertechnetate are required. Colloidal gold was excluded from low density brine-saturated bentonite, but colloidal gold may have channeled between the bentonite and the wall of the diffusion cell. These results support the effectiveness of brine-saturated bentonite as an engineered barrier to plutonium. The results also highlight the need for additional measurements of pertechnetate and gold transport in bentonite. Needs for transport measurements that take into account site-specific materials interactions are also described.

Design considerations for a nuclear waste repository in salt : Tunnlg Technol Newsl, N37, March 1982, P1–13

Design considerations for a nuclear waste repository in salt : Tunnlg Technol Newsl, N37, March 1982, P1–13