A compacted mixture of Kunigeru VI bentonite and D-sand is being considered for use as an engineered barrier in low-level radioactive waste disposal facilities in Japan. An important issue is the maintenance of the retardation property of the mixture during shear that might be induced in such barriers by earthquakes and/or gradual tectonic deformations occurring over the design life of the facility. To investigate this issue, comparative tests on a bentonite-sand mixture and kaolin-sand mixture were conducted by means of a recently-developed coupled shear and permeability testing apparatus under temperature controlled condition. In addition to permeability, the specific storage of bentonite-sand specimen during shear is also systematically evaluated with the new analytical theory for the constant flow permeability test. The present study reveals that 1) temperature control is preferred for measuring the permeability of extremely-low permeability materials with the constant-flow pump method; 2) both the permeability and specific storage of the mixture of Kunigeru VI bentonite and D-sand (bentonite to sand weight ratio = 15: 85) were not significantly influenced by shear strains up to 3% whereas the permeability of the kaolin-sand mixture increased almost linearly with the increment of shear strain; 3) the swelling of bentonite in the mixture under low confining stress decreases both the permeability and specific storage of bentonite-sand mixture. This feature of bentonite helps maintain the long-term retardation property of the mixture; and 4) the constant flow permeability test method, with the newly derived theoretical analysis, promises to become a very effective means of investigating, rapidly and systematically, the permeability and specific storage of extremely-low permeability materials with relatively-low (i.e. close to in situ) hydraulic gradients. This is necessary for the design and long-term performance assessment of engineered barriers.

Compacted clay-based buffer surrounds corrosion-resistant waste containers in the Canadian concept for nuclear fuel waste disposal. Clays naturally contain small quantities of organic matter that may be resistant to bacterial degradation. The containers with highly radioactive material would subject the surrounding buffer to both heat (max. 95°C) and radiation (max. 52 Gy/h). Both could potentially break down complex organic material to smaller, more bioavailable compounds. This could stimulate microbial growth and possibly affect gas production, microbially-influenced corrosion or radionuclide migration. Experiments were carried out in which buffer (50 wt.% Na-bentonite, 50 wt.% sand) was heated at 60 and 90°C for periods of 2, 4 and 6 weeks, in some cases followed by irradiation to 25 kGy. Unheated buffer was also irradiated to 25 and 50 kGy at different moisture contents. The treated materials were subsequently suspended in distilled water, shaken for 24 h and centrifuged to remove the solids. The 0.22 μm filter-sterilized leachates were inoculated with equal volumes of fresh groundwater and incubated at room temperature for 10 d to determine the increase in total and viable bacteria compared to a groundwater control. Results indicated that leachates from buffer subjected to heat, radiation or combinations of these, had a stimulating effect on both total and viable cell counts in groundwater, compared to unamended groundwater controls. This stimulating effect was generally most pronounced for viable counts and could be larger than two orders of magnitude. Leachates from untreated buffer material also stimulated the growth of groundwater bacteria, but to a lesser extent than leachates from heat-and radiation-treated buffer material. The effects of heat and radiation on nutrient availability in clay-based sealing materials (at relevant clay/moisture ratios) should, therefore, be taken into account when attempting to quantify the effects of microbial activity on vault performance.

In this study, gas migration experiments in unsaturated and saturated states were carried out to clarify the fundamental gas migration characteristics in compacted bentonite to be used for the geological disposal of high-level radioactive waste. In unsaturated experiments, the gas permeability for Japanese bentonite (Kunigel VI) as a function of degree of saturation was measured to examine the applicability of conventional two-phase flow models to compacted bentonite. The intrinsic permeability obtained in this study was about five orders of magnitude larger than that obtained in water permeation tests with the same density. The difference seems to originate from the change of pore structure due to the swelling phenomenon of the bentonite. Since these effects have not been evaluated quantitatively yet, various relative gas permeability functions of conventional two-phase flow models were applied as a first approximation.

Saturated experiments designed to simulate the gas migration phenomenon in a repository for the waste were carried out to obtain relationship between breakthrough and swelling pressures using Kunigel VI and French Fo-Ca clay in saturation state. The reproducibility of the breakthrough pressure was also examined for Kunigel VI bentonite. The breakthrough pressure was almost the same as swelling pressure irrespective of the type of clay. As to the reproducibility of breakthrough pressure, it was observed that first and second breakthrough pressures were almost the same for Kunigel VI specimens with the dry densities of 1.7 and 1.8 g/cm3.

Controlled flow-rate gas injection experiments have been performed on pre-compacted samples of KBS-3 specification M×801 buffer bentonite using helium as a safe replacement for hydrogen. By simultaneously applying a confining pressure and backpressure, specimens were isotropically-consolidated and fully water-saturated under pre-determined effective stress conditions, before injecting gas using a syringe pump. Ingoing and outgoing gas fluxes were monitored. All tests exhibited a conspicuous threshold pressure for breakthrough, somewhat larger than the sum of the swelling pressure and the backpressure. All tests showed a post-peak negative transient leading to steady-state gas flow. Using a stepped history of flow rate, the flow law was shown to be nonlinear. With the injection pump stationary (i.e. zero applied flow rate), gas pressure declined with time to a finite value. When gas flow was reestablished, the threshold value for gas breakthrough was found to be significantly lower than in virgin clay. There is strong evidence to suggest that the capillary pressure for the penetration of interparticle pore space of buffer bentonite is of such a magnitude that normal two-phase flow is impossible. Gas entry and breakthrough is therefore accompanied by the development of microcracks which propagate through the clay from gas source to sink. The experiments suggest that these pathways open under high gas pressure conditions and partially close if gas pressure falls, providing a possible explanation of the nonlinearity of the flow law.

Laboratory experiments were performed to study the interaction between groundwater and compacted sodium bentonite (Volclay MX-80). The solutions used were the fresh and saline groundwater simulants. The experiments were carried out in aerobic and anaerobic conditions at elevated temperature. Of main interest in the present study were the chemical changes in the reacting solution, bentonite porewater. and bentonite itself. The results for major cations display a principal difference between the interactions with fresh and saline solutions, while the differences between aerobic and anaerobic conditions within each solution case seem to be minor. The experimental results for the bentonite-water equilibria were interpreted in terms of a multi-site surface complexation model and the computer program HYDRAQL. The apparent diffusivities for sodium and sulphate in bentonite samples sandwiched between two filterplates were also determined.

Gas migration properties of bentonite/sand mixtures for a large-scale structure were estimated on the basis of laboratory test results obtained with small test specimens. Water and gas migration tests were carried out to clarify scale effects on migration properties of the mixtures.

Three-dimensional calculations that explicitly represent a realistic mixture of waste packages (WPs) are used to analyze decay-heat-driven thermal-hydrological behavior around emplacement drifts in a potential high-level waste facility at Yucca Mountain. Calculations, using the NUFT code, compare two fundamentally different ways that WPs can be arranged in the repository, with a focus on temperature, relative humidity, and liquid-phase flux on WPs. These quantities strongly affect WP integrity and the mobilization and release of radionuclides from WPs. Point-load spacing, which places the WPs roughly equidistant from each other, thermally isolates WPs from each other, causing large variability in temperature, relative humidity, and liquid-phase flux along the drifts. Line-load spacing, which places WPs nearly end to end in widely spaced drifts, results in more locally intensive and uniform heating along the drifts, causing hotter, drier, and more uniform conditions. A larger and more persistent reduction in relative humidity on WPs occurs if the drifts are backfilled with a low-thermal-conductivity granular material with hydrologie properties that minimize moisture wicking.

A coupled model concept which may be used for performance assessment of a nuclear repository is presented. The tool is developed by integration of two models, one near field and one far field model. A compartment model, NUCTRAN, is used to calculate the near field release from a damaged canister. The far field transport through fractured rock is simulated by using CHAN3D, based on a three-dimensional stochastic channel network concept. The near field release depends on the local hydraulic properties of the far field. The transport in the far field in turn depends on where the damaged canister(s) is located. The very large heterogeneities in the rock mass makes it necessary to study both the near field release properties and the location of release at the same time. In order to demonstrate the capabilities of the coupled model concept it is applied on a hypothetical repository located at the Hard Rock Laboratory in Äspö, Sweden. Two main items were studied; the location of a damaged canister in relation to fracture zones and the barrier function of the host rock. In the study of the near field rock as a transport barrier the effect of different tunnel excavation methods which may influence the damage level of the rock around the tunnel was addressed.

In a safety analysis of a repository for high level nuclear waste, it is of primary importance to identify and document all processes that are acting on the repository system. An interaction matrix methodology have been applied to the spent fuel subsystem. The purpose of this application is to identify, structure and rank the Process System and to discuss how the identified processes can be treated.

A general, integrated performance assessment code, AREST-CT, was used to analyze the influence of various factors on the release rates of radionuclides from a proposed facility for disposal of low-activity tank wastes. The code couples various process models together based on the framework of reaction-transport theory. The disposal facility was modeled as a 1-D column surrounded by soil. A borosilicate waste glass, LD6–5412 was the waste form considered in the analysis. Included in the simulations were 38 aqueous species, 14 minerals, 21 equilibrium reactions, and 16 kinetic reactions. Dissolution rate of the glass and the release rates of Te, Pu, U, Np, I, Se under different conditions were calculated for 50,000 years. The simulations revealed that 1) open exchange between the atmosphere and pore-water within the vault significantly improves the performance; 2) an ion-exchange reaction between the glass and aqueous phase increases the release rates significantly; and 3) at the hydrogeologie conditions under consideration, variation of the pore-water velocity has little effect on the release rate of radionuclides. These results provide a scientific basis for formulation of waste forms and engineering design of the disposal facility. Reaction-transport modeling can provide information on the long-term performance of disposal systems that are not obtainable from laboratory experiments alone or by conventional decoupled process models.

A new mechanistic low-level waste source term model was developed. Key features of this effort are: use of cumulative probability functions to describe the failures of waste containers; capability to describe diffusion controlled release which is dependent on the conditions of the waste package surroundings; consideration of the effects of gas generation on the source term, and; use of a source inventory characterization routine to provide a built-in capability for defining the distributions of radionuclides in various waste forms and streams. The model is capable of describing the diffusion of radionuclides in waste forms with the use of concentration and flux continuity boundary conditions. Release of radionuclides from various waste streams is modeled by the combination of diffusion, dissolution, and surface release. Failures of waste containers are portrayed by the use of probabilistic failure functions based on Weibull and lognormal distributions. The model for radionuclide release from waste packages is coupled with the near-field transport model to describe the effects of migration and dispersion within the disposal unit. Characteristics of the new model were evaluated through sensitivity analysis and compared with existing source term codes.

A 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.

A two-layer waste package (carbon steel outer barrier and Alloy 825 inner barrier) is specified to dispose of high-level nuclear waste at the potential repository at Yucca Mountain. A set of improvements and more realism have been added to a stochastic waste-package degradation model which was developed for a recent total system performance assessment of the potential repository [1]. The waste-package surface is divided into “patches” to better represent the general corrosion of the carbon-steel outer barrier. The “corrosion-time” concept is developed to represent the corrosion of the carbon-steel outer barrier in changing exposure conditions with time such as those expected in the potential repository. With the patches approach and the corrosion-time concept implemented into the waste-package degradation model, sensitivity of the waste package degradation (failure and pitting degradation) to different threshold spalling thicknesses of the corrosion products from the carbon-steel outer barrier is analyzed. The results show that the waste-package pitting degradation is sensitive to the corrosion-products spalling thickness of the carbon-steel outer barrier. A greater pitting degradation of the waste packages is predicted with a smaller spalling thickness. Further understanding of the corrosion-products spalling in different exposure conditions (i.e., water chemistry, water contact mode, etc.) and its effects on carbon steel corrosion is needed to enhance the confidence in the waste-package performance modeling in the potential repository.

The Phase 3 Total System Performance Assessment (TSPA) sponsored by the Electric Power Research Institute (EPRI) has led to new insights into critical models and parameters affecting estimated doses to humans from a potential repository of high-level radioactive wastes at Yucca Mountain, Nevada. The Phase 3 model has been extended to encompass time-varying climate and infiltration, detailed modeling of the source term and hydrology, and detailed specification of possible interaction between percolating ground water and waste containers. The model estimates doses to a time of one million years.

The three key radionuclides contributing to estimated total doses are Tc-99,1–129, and Np-237. Five other nuclides contributing to dose in lesser (but significant) amounts are U-233, Th-229, Pa-231, Ac-227, and Se-79. These results are consistent with other TSPAs.

From sensitivity studies, the most critical models and parameters are as follows. Infiltration and percolation assumptions, including the amount of lateral diversion of infiltration water, are important and need verification with site data and/or more detailed modeling. Parameters of the unsaturated zone (UZ) and saturated zone (SZ) determine dilution and delay of concentrations and peak doses downstream. The fraction of containers that become wet are not critical in our model, but this lack of sensitivity reflects our coupling of the fraction with a model of focused flow past the containers; an different model might indicate higher sensitivity. Also, the degree of coupling between fracture and matrix flow is important in affecting the times of peak doses but not their magnitudes.

Other critical design assumptions that could lead to reduced and/or delayed doses are a more robust container design, a capillary barrier around each container, the dilution during hydrologie transport from the repository to a potential agricultural community downstream, and the characteristics of an “average” individual in that community who might receive a dose.

In performance assessment of geological disposal systems for High Level Radioactive Waste (HLW), the change of environment over the long-term must be considered. Therefore, it is necessary to consider a wide range of parameters concerned with radionuclides migration, especially the dependence of solubility on geochemical environment. In this study, assuming that the release rate of the nuclides from buffer material is limited by inventory ultimately, the relationship between the initial inventory and the solubility that produces a solubility-invariant maximum release rate from the buffer is examined by using a simple steady-state analytical solution without decay. As the result, the threshold of “effective” solubility in the performance assessment of the geological disposal systems for HLW is obtained as a function of initial inventory, distribution coefficient (Kd), diffusion coefficient, and thickness, porosity and density of the buffer. Also, the threshold of “effective” steady dissolution rate corresponding to the threshold of “effective” solubility is obtained.

Recent proposals for a new U.S. standard for high-level waste disposal would limit the average dose to individuals in the vicinity surrounding a geologic repository. This would be a new approach to protecting the public from environmental releases of radioactivity. Heretofore, criteria adopted for geologic disposal have limited the reasonable maximum exposure to a future hypothetical individual. Here we present quantitative analyses of the relation between maximum exposure and vicinity-average exposure, resulting from future human use of ground water contaminated by radioactive releases from a repository.

Estimating the vicinity-average exposure would require postulates and guesses of location and habits of future people. Exposure probabilities postulated by others show that proposed dose limit to the vicinity-average individual would be a far more lenient standard than the traditional dose limit to reasonably maximally exposed individuals. The proposed vicinity-average dose limit would allow far greater concentrations of contaminants in ground water than would be allowed by normal standards of ground water protection. A safety standard that limits vicinity-average exposure should also include limits on maximum exposure.

Thermal/hydrological modeling of the Radioactive Scrap and Waste Facility (RSWF) has been completed with the TOUGH2 (Transport of Unsaturated Groundwater and Heat) Code.[1] The RSWF will be utilized as an interim storage facility for ceramic and metallic waste forms developed from the electrometallurgical treatment of spent nuclear fuel. The RSWF is an array of 1,350 carbon steel liners located at grade near Argonne National Laboratory-West on the Idaho National Engineering Laboratory (INEL).

The primary driving force for this modeling research was to assess thermal capacity limits for RSWF liners so that heat generating materials can be safely stored. Maximum wasteform temperatures will be governed by both the amount of heat the system can dissipate and the orientation and characteristics of the wasteform. The focus of this report is on the amount of heat the interim storage system can safely dissipate. The effect of the temporal variation of soil moisture on the performance of the RSWF is assessed. The facility was analyzed to determine the maximum allowable thermal loading of the RSWF liners.

A relatively simple model is suggested for radionuclide dispersion and assessment of additional population dose and radiation risk for radionuclide release from the nuclear submanne (NS) #601 with Pb-Bi coolant that was dumped in the Kara Sea.

Bias is a difference between model and reality. Bias can be introduced at any stage of the modelling process during a site characterisation or performance assessment programme. It is desirable to understand such bias so as to be able to optimally design and interpret a site characterisation programme. The objective of this study was to examine the source and effect of bias due to the assumptions modellers have to make because reality cannot be fully characterised in the prediction of ground-water fluxes. A well-defined synthetic “reality” was therefore constructed for this study. A limited subset of these data were independently interpreted and used to compute groundwater fluxes across specified boundaries in a cross section. The modelling results were compared to the “true” solutions derived using the full dataset. This study clarified and identified the large number of assumptions and judgements which have to be made when modelling a limited site characterisation dataset. It is concluded that bias is introduced at each modelling stage, and that it is not necessarily detectable by the modellers even if multiple runs with varied parameter values are undertaken.

In parallel to studies of disposal of vitrified high-level waste from reprocessing, projects have been initiated to examine options for direct disposal of spent fuel in Switzerland. The basic concept involves in-tunnel emplacement of encapsulated spent fuel in a deep repository which is backfilled with compacted bentonite. Two possible host rocks are considered - crystalline basement and Opalinus Clay. This paper reports the results of a thermal analysis which was carried out to evaluate constraints on repository layout set by the desire to limit temperatures experienced by the bentonite backfill.