[en] This document represents a compilation of data and interpretive studies conducted as part of the engineered barriers program of the Basalt Waste Isolation Project. The overall objective of these studies is to provide information on barrier system designs, emplacement and isolation techniques, and chemical reactions expected in a nuclear waste repository located in the basalts underlying the Hanford Site within the state of Washington. Backfills, waste-basalt interactions, sorption, borehole plugging, etc., are among the topics discussed

[en] The purposes of the Borehold Plugging Program are: to identify issues associated with sealing boreholes and shafts; to establish a data base from which to assess the importance of these issues; and to develop sealing criteria, materials, and demonstrative test for the Waste Isolation Pilot Plant (WIPP). The Bell Canyon Test described in this report is one part of that program. Its purpose was to evaluate, in situ, the state of the art in borehole plugs and to identify and resolve problems encountered in evaluating a typical plug installation in anhydrite. The test results are summarized from the work of Peterson and Christensen and divided into two portions: system integrity and wellbore characterization tests prior to plug installation, and a series of tests to evaluate isolation characteristics of the 1.8-m-long plug. Conclusions of the Bell Canyon Test are: brine and fresh-water grouts, with acceptable physical properties in the fluid and hardened states, have been developed; the field data, taken together with laboratory data, suggest that the predominant flow into the test region occurs through the cement plug/borehold interface region, with lesser contributions occurring through the wellbore damage zone, the plug core, and the surrounding undisturbed anhydrite bed; and the 1.8-m-long by 20-cm-diameter grout plug, installed in anhydrite at a depth of 1370 m in the AEC-7 borehole, limits flow from the high pressure Bell Canyon aquifer to 0.6 liters/day

[en] A three-dimensional thermal analysis has been performed using finite difference techniques to determine the near-field response of a baseline spent fuel repository in a deep geologic salt medium. A baseline design incorporates previous thermal modeling experience and OWI recommendations for areal thermal loading in specifying the waste form properties, package details, and emplacement configuration. The base case in this thermal analysis considers one 10-year old PWR spent fuel assembly emplaced to yield a 36 kw/acre (8.9 w/m²) loading. A unit cell model in an infinite array is used to simplify the problem and provide upper-bound temperatures. Boundary conditions are imposed which allow simulations to 1000 years. Variations studied include a comparison of ventilated and unventilated storage room conditions, emplacement packages with and without air gaps surrounding the canister, and room cool-down scenarios with ventilation following an unventilated state for retrieval purposes. At this low power level ventilating the emplacement room has an immediate cooling influence on the canister and effectively maintains the emplacement room floor near the temperature of the ventilating air. The annular gap separating the canister and sleeve causes the peak temperature of the canister surface to rise by 10⁰F (5.6⁰C) over that from a no gap case assuming perfect thermal contact. It was also shown that the time required for the emplacement room to cool down to 100⁰F (38⁰C) from an unventilated state ranged from 2 weeks to 6 months; when ventilation initiated after times of 5 years to 50 years, respectively. As the work was performed for the Nuclear Regulatory Commission, these results provide a significant addition to the regulatory data base for spent fuel performance in a geologic repository

[en] The Sandia/NRC Risk Assessment Methodology consists of a procedure for assessing the post-closure, long-term risk from the disposal of radioactive waste in deep geologic formations. This procedure contains: (1) methods for selecting and screening potentially disruptive events, features and processes (i.e., scenarios); (2) models for use in simulating the physical processes and estimating the potential health effects associated with the deep geologic disposal of radioactive waste (e.g., repository evolution, ground-water flow and nuclide transport, biosphere transport and human exposure, and dose commitment and dose response); and (3) probabilistic and statistical procedures for use in risk estimates and in sensitivity and uncertainty analyses. Results of the demonstration of this methodology in the analysis of a hypothetical high-level waste repository in bedded salt are presented for the following three scenarios: (1) a hydraulic communication (boreholes or shafts) connects the middle and lower sandstone aquifers allowing water to flow through the depository; (2) a hydraulic communication allows water to flow from the middle sandstone aquifer through the depository and back to middle sandstone aquifer, referred to as a U-tube scenario; and (3) withdrawl wells completed into the middle sandstone aquifer down-gradient from the depository are coupled with the U-tube of Scenario 2

[en] Radioactive aqueous wastes generated by the solvent extraction of uranium from expended fuels at ICPP will be calcined in the New Waste Calcining Facility (NWCF). The calcined solids are pneumatically transferred to stainless steel bins enclosed in concrete vaults for interim storage of up to 500 years. The Fifth Calcined Solids Storage Facility (CSSF) provides 1000 m³ of storage and consists of seven annular stainless steel bins inside a reinforced concrete vault set on bedrock. Storage of calcined solids is essentially a passive operation with very little opportunity for release of radionuclides and with no potential for criticality. There will be no potential for fire or explosion. Shielding has been designed to assure that the radiation levels at the vault exterior surfaces will be limited to less than 0.5 mRem/h. A sump in the vault floor will collect any in-leakage that may occur. Any water that collects in the sump will be sampled then removed with the sump jet. There will be an extremely small chance of release of radioactive particulates into the atmosphere as a result of a bin leak. The Design Basis Accident (DBA) postulates the spill of solids from an eroded fill line into the vault coupled with a failure of the vault cooling air radiation monitor. For the DBA, the maximum calculated radiation dose to an exposed individual near the site boundary is less than 1.2 μRem to the bone and lung

[en] The report reviews the geologic and hydrologic data base for penetration seal designs referenced to the Los Medanos bedded salt site in New Mexico and to four candidate salt domes in the Gulf Interior. Experience with existing shafts highlights the importance, for shaft decommissioning as well as operation, of achieving an adequate seal at and immediately below the top of salt. Possible construction procedures for repository shafts are reviewed, noting advantages and disadvantages with respect to repository sealing. At this stage, there does not appear to be a clear preference for excavation by drill and blast or by drilling. If conventional drill and blast methods are used, it may be necessary to grout in permeable zones above the salt. An important consideration with respect to sealing is that grouting operations (or freezing should it be used) should not establish connections between the top of salt and water-bearing zones higher in the stratigraphic section. Generally, it is concluded that Los Medanos and the dome salt sites are favorable candidate repository sites from the point of view of sealing

[en] The Canadian research program to investigate the disposal of nuclear fuel waste is well established with participation from a wide cross section of the scientific and engineering community. Some phases of the program have not proceeded at the pace originally envisaged. Nevertheless, significant progress has been made. The resources available and difficulties in obtaining approval for additional field research areas have restricted geotechnical research to granite formations. With the prospect of increased levels of funding and new initiatives for approvals for field research, it is expected that the program will proceed on a broader front

[en] To approach the subject of high level nuclear waste disposal in deep ocean sediments it is convenient to differentiate between processes occurring in a near field environment, that region arbitrarily defined as lying between the canister surface and the maximum extent of the 100⁰C isotherm, and those which occur at lower temperatures and beyond the influence of intense radiation. A variety of considerations related to the chemistry of seawater-sediment mixtures suggests that about 200⁰C is the maximum temperature advisable in the near field environment. Results of a coupled fluid flow - thermal transport computer model show the maximum convection rate adjacent to a canister having surface temperature of 200⁰C is 0.3 m/100 years, and that this velocity is halved with the passage of each thermal half life of the assumed waste form (30 years). Based on this convective model, it follows that compounds formed in the near field environment during the first thousand years following emplacement would be restricted to a region lying within two meters of the canister surface

[en] Safe deposition of radioactive canisters in boreholes or tunnels requires a buffer mass for which the following criteria are specified: sufficient bearing capacity, suitable plastic mechnical properties, low permeability, ion-adsorption ability, sufficiently good heat conduction. Its required physical and mechanical properties must also be stable for thousands of years at the elevated temperature in the radiation field close to the canisters. So for all evidence given in the current investigation shows that a combination of Na-bentonite and silt/sand-sized quartz components gives a buffer mass with optimum properties. The bentonite should form at least 10 weight percent of the solid mass where the bearing capacity is important. In other parts of the bore-holes, tunnels and shafts the bentonite content can be increased to 20 or even higher percentages

[en] A thermoelastic/plastic finite element analysis of the influence of an air-gap on hole closure about a waste-container/sleeve assembly emplaced in a typical repository room (SALT/4T Model) indicates that hole closure would be of the order of hundredths of an inch. Acceptable air-gap width is thus governed by the hole size required for emplacement efficiency. A refined mesh analysis and laboratory testing is suggested in order to further explore the possibility of eliminating the engineering necessity of the sleeve