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DOE OSTI2012

Classified Component Disposal at the Nevada National Security Site

The Nevada National Security Site (NNSS) has added the capability needed for the safe, secure disposal of non-nuclear classified components that have been declared excess to national security requirements. The NNSS has worked with U.S. Department of Energy, National Nuclear Security Administration senior leadership to gain formal approval for permanent burial of classified matter at the NNSS in the Area 5 Radioactive Waste Management Complex owned by the U.S. Department of Energy. Additionally, by working with state regulators, the NNSS added the capability to dispose non-radioactive hazardous and non-hazardous classified components. The NNSS successfully piloted the new disposal pathway with the receipt of classified materials from the Kansas City Plant in March 2012.

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DOE OSTI2013

Grout Long Radius Flow Testing to Support Saltstone Disposal Unit 6 Design - 13352

The Saltstone Facility, located within the Savannah River Site (SRS) near Aiken, South Carolina, consists of two facility segments: The Saltstone Production Facility (SPF) and the Saltstone Disposal Facility (SDF). The SPF receives decontaminated legacy low level sodium salt waste solution that is a byproduct of prior nuclear material processing. The salt solution is mixed with cementitious materials to form a grout slurry known as 'Saltstone'. The grout is pumped to the SDF where it is placed in a Saltstone Disposal Unit (SDU) to solidify. SDU 6 is referred to as a 'mega vault' and is currently in the design stage. The conceptual design for SDU 6 is a single cell, cylindrical geometry approximately 114.3 meters in diameter by 13.1 meter high and is larger than previous cylindrical SDU designs, 45.7 meters in diameter by 7.01 meters high (30 million gallons versus 2.9 million gallons of capacity). Saltstone slurry will be pumped into the new waste disposal unit through roof openings at a projected flow rate of about 34.1 cubic meters per hour. Nine roof openings are included in the design to discharge material into the SDU with an estimated grout pour radius of 22.9 to 24.4 meters and initial drop height of 13.1 meters. The conceptual design for the new SDU does not include partitions to limit the pour radius of the grout slurry during placement other than introducing material from different pour points. This paper addresses two technical issues associated with the larger diameter of SDU 6; Saltstone flow distance in a tank 114.3 meters in diameter and quality of the grout. A long-radius flow test scaled to match the velocity of an advancing grout front was designed to address these technology gaps. The emphasis of the test was to quantify the flow distance and to collect samples to evaluate cured properties including compressive strength, porosity, density, and saturated hydraulic conductivity. Two clean cap surrogate mixes (Saltstone premix plus water) were designed to simulate slurry with the reference Saltstone rheology and a Saltstone with extra water from the process flushing operation. Long-radius flow tests were run using approximately 4.6 cubic meters of each of these mixes. In both tests the pump rate was 0.063 liters/second (1 gpm). A higher pump rate, 0.19 liters/second (3 gpm), was used in a third long-radius flow test. The angle of repose of the grout wedges increased as a function of time in all three tests. The final angles of repose were measured at 3.0 deg., 2.4 deg., and 0.72 deg.. The pump rate had the largest effect on the radial flow distance and slope of the grout surface. The slope on the pour placed at 0.19 liters/second (3 gpm) was most representative of the slope on the grout currently being pumped into SDU 2 which is estimated to be 0.7 deg. to 0.9 deg. The final grout heights at 1/3 of a meter from the discharge point were 115, 105, and 38 cm. Entrapped air (≥ 0.25 cm bubbles) was also observed in all of the mixes. The entrapped air appeared to be released from the flows within about 3.1 meters (10 feet) of the discharge point. The bleed water was clear but had a thin layer of floating particulates. The bleed water should be retrievable by a drain water collection system in SDU 6 assuming the system does not get clogged. Layering was observed and was attributed to intervals when the hopper was being cleaned. Heat from the hydration reactions was noticeable to the touch. (authors)

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DOE OSTI2013

Classified Component Disposal at the Nevada National Security Site (NNSS) - 13454

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DOE OSTI2013

EnergySolution's Clive Disposal Facility Operational Research Model - 13475

EnergySolutions owns and operates a licensed, commercial low-level radioactive waste disposal facility located in Clive, Utah. The Clive site receives low-level radioactive waste from various locations within the United States via bulk truck, containerised truck, enclosed truck, bulk rail-cars, rail boxcars, and rail inter-modals. Waste packages are unloaded, characterized, processed, and disposed of at the Clive site. Examples of low-level radioactive waste arriving at Clive include, but are not limited to, contaminated soil/debris, spent nuclear power plant components, and medical waste. Generators of low-level radioactive waste typically include nuclear power plants, hospitals, national laboratories, and various United States government operated waste sites. Over the past few years, poor economic conditions have significantly reduced the number of shipments to Clive. With less revenue coming in from processing shipments, Clive needed to keep its expenses down if it was going to maintain past levels of profitability. The Operational Research group of EnergySolutions were asked to develop a simulation model to help identify any improvement opportunities that would increase overall operating efficiency and reduce costs at the Clive Facility. The Clive operations research model simulates the receipt, movement, and processing requirements of shipments arriving at the facility. The model includes shipment schedules, processing times of various waste types, labor requirements, shift schedules, and site equipment availability. The Clive operations research model has been developed using the WITNESS{sup TM} process simulation software, which is developed by the Lanner Group. The major goals of this project were to: - identify processing bottlenecks that could reduce the turnaround time from shipment arrival to disposal; - evaluate the use (or idle time) of labor and equipment; - project future operational requirements under different forecasted scenarios. By identifying processing bottlenecks and unused equipment and/or labor, improvements to operating efficiency could be determined and appropriate cost saving measures implemented. Model runs forecasting various scenarios helped illustrate potential impacts of certain conditions (e.g. 20% decrease in shipments arrived), variables (e.g. 20% decrease in labor), or other possible situations. (authors)

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DOE OSTI2013

Using Process Knowledge to Manage Disposal Classification of Ion-Exchange Resin - 13566

It has been previously shown by EPRI [1] that Class B and C resins represent a small portion by volume of the overall generation of radioactively contaminated resins. In fact, if all of the resins were taken together the overall classification would meet Class A disposal requirements. Lowering the classification of the ion exchange resins as they are presented for disposal provides a path for minimizing the amount of waste stored. Currently there are commercial options for blending wastes from various generators for Class A disposal in development. The NRC may have by this time introduced changes and clarifications to the Branch Technical Position (BTP) on Concentration Averaging and Encapsulation [2] that may ultimately add more flexibility to what can be done at the plant level. The BTP has always maintained that mixtures of resins that are combined for ALARA purposes or operational efficiency can be classified on the basis of the mixture. This is a point often misinterpreted and misapplied. This paper will address options that can be exercised by the generator that can limit B and C waste generation by more rigorous tracking of generation and taking advantage of the normal mix of wastes. This can be achieved through the monitoring of reactor coolant chemistry data and coupled with our knowledge of radionuclide production mechanisms. This knowledge can be used to determine the overall accumulation of activity in ion-exchange resins and provides a 'real-time' waste classification determination of the resin and thereby provide a mechanism to reduce the production of waste that exceeds class A limits. It should be noted that this alternative approach, although rarely used in a nuclear power plant setting, is acknowledged in the original BTP on classification [3] as a viable option for determining radionuclide inventories for classification of waste. Also included is a discussion of an examination performed at the Byron plant to estimate radionuclide content in the final waste stream from upstream sampling of reactor coolant and fuel pool water. (authors)

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DOE OSTI2013

Regulatory Framework for Salt Waste Disposal and Tank Closure at the Savannah River Site - 13663

The end of the Cold War has left a legacy of approximately 37 million gallons of radioactive waste in the aging waste tanks at the Department of Energy's Savannah River Site (SRS). A robust program is in place to remove waste from these tanks, treat the waste to separate into a relatively small volume of high-level waste and a large volume of low-level waste, and to actively dispose of the low-level waste on-site and close the waste tanks and associated ancillary structures. To support performance-based, risk-informed decision making and to ensure compliance with all regulatory requirements, the U.S. Department of Energy (DOE) and its current and past contractors have worked closely with the South Carolina Department of Health and Environmental Control (SCDHEC), the U.S. Environmental Protection Agency (EPA) and the Nuclear Regulatory Commission (NRC) to develop and implement a framework for on-site low-level waste disposal and closure of the SRS waste tanks. The Atomic Energy Act of 1954, as amended, provides DOE the authority to manage defense-related radioactive waste. DOE Order 435.1 and its associated manual and guidance documents detail this radioactive waste management process. The DOE also has a requirement to consult with the NRC in determining that waste that formerly was classified as high-level waste can be safely managed as either low-level waste or transuranic waste. Once DOE makes a determination, NRC then has a responsibility to monitor DOE's actions in coordination with SCDHEC to ensure compliance with the Title 10 Code of Federal Regulations Part 61 (10CFR61), Subpart C performance objectives. The management of hazardous waste substances or components at SRS is regulated by SCDHEC and the EPA. The foundation for the interactions between DOE, SCDHEC and EPA is the SRS Federal Facility Agreement (FFA). Managing this array of requirements and successfully interacting with regulators, consultants and stakeholders is a challenging task but ensures thorough and thoughtful processes for disposing of the SRS low-level waste and the closure of the tank farm facilities. (authors)

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DOE OSTI2017

Disposal for Canada... a Responsible Solution for 120 years of Waste Generation - 17052

Disposal is coming to Canada. After decades of interim storage of radioactive wastes, Canadian Nuclear Laboratories (CNL) is preparing to build Canada's first facility for the permanent disposal of low-level and other suitable radioactive wastes that satisfy the waste acceptance criteria. A project was launched in 2015 to establish a Near Surface Disposal Facility (NSDF) and to place it into operation in 2020 for 50 years. (authors)

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DOE OSTI2017

Overview of Features, Events, and Processes Affecting the Postclosure Performance of a Deep Borehole Disposal Facility - 17168

The United States Department of Energy's Office of Nuclear Energy (DOE-NE) is evaluating the feasibility of disposing of smaller DOE-managed waste forms in 5-km deep boreholes drilled in crystalline rock. It is thought that both the depth and the geological conditions at that depth will serve to isolate the waste from the accessible environment after the disposal facility is closed. This paper discusses the features, events, and processes (FEPs) that could affect the postclosure performance of a deep borehole disposal (DBD) facility and identifies those that would likely need to be considered in the context of conducting a performance assessment analysis. (authors)

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DOE OSTI2017

Analysis of Saltstone Cores Taken from Saltstone Disposal Unit Cell 2A - 17350

At the Savannah River Site, low-level waste is immobilized in cementitious waste form known as saltstone. The saltstone Performance Assessment models the performance of the waste form and other aspects of the Saltstone Disposal Facility over ten thousand years after closure to better understand the transport history of radionuclides and other hazardous constituents in low-level waste. As part of ongoing Performance Assessment Maintenance, Savannah River Remediation (SRR), with Savannah River National Laboratory (SRNL), designed a sampling and analyses plan to correlate the properties of field-emplaced samples and samples processed and cured in the laboratory. The primary objectives of the Sampling and Analyses Plan are; (1) to demonstrate a correlation between the measured properties of laboratory-prepared, simulant samples (termed Sample Set 3), and the field-emplaced saltstone samples (termed Sample Set 9), and (2) to validate property values assumed for the Saltstone Disposal Facility (SDF) Performance Assessment modeling. The analysis and property data for Sample Set 9 (i.e. core samples extracted from Saltstone Disposal Unit Cell 2A (SDU2A)) are documented, and where applicable, the results are compared to the results for Sample Set 3. SRNL received, stored and analyzed the samples cored from SDU2A. This paper discusses the results of the SRNL analysis of the SDU2A samples as well as the simulant saltstone samples. (authors)

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DOE OSTI2017

Using Prescribed Fire to Manage the Fernald On-Site Disposal Facility Vegetated Cap - 17401

The Fernald Preserve is a former uranium processing plant that underwent extensive remediation pursuant to the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA). It is situated on a 425 ha (1,050 acre) tract of land, approximately 29 km (18 miles) northwest of Cincinnati, Ohio. Remedial activities and subsequent ecological restoration have converted the site from an industrial production facility to an undeveloped park, encompassing a series of wetlands, prairies, and forested communities. The remediation of buildings and soil contamination was completed in 2006, but aquifer remediation is ongoing. Since 2007, the site has been managed by the US Department of Energy (DOE) Office of Legacy Management. Portions of the site are open to the public, with approximately 11 km (7 miles) of walking trails that are open year-round and a visitors center. Site remediation included the construction of an On-Site Disposal Facility (OSDF) for the long-term disposal of contaminated soil and building materials. The OSDF consists of eight disposal cells that encompass approximately 30 hectares (75 acres) on the eastern portion of the Fernald site. Cells are lined and capped, and each has a separate leak detection and leachate collection system. The integrated cap is vegetated with native grasses and forbs. Management of the Fernald Preserve includes the monitoring and maintenance of ecologically restored communities and the vegetated OSDF cap, as required by the site's institutional controls plan. Prescribed burning is recognized as the preferred method for prairie management. Prescribed fire has been used at the Fernald site since 2009 for established prairies in restored areas, but not on the OSDF cap. Stakeholders had reservations regarding the safety of using prescribed fire on the OSDF cap. Because of that, vegetation on the cap was managed as a grassland community. It was mowed, raked, and baled to remove thatch, and promote the continued establishment of native grasses and forbs. Since agricultural use is restricted at the site, the resulting hay bales were kept on property and reused as organic matter and mulch. Over time, more hay bales were being generated in a season than were reused. Therefore in 2015, DOE decided to revisit stakeholder acceptance of prescribed fire on the OSDF cap. Ongoing relationships with local stakeholders allowed for extensive discussions regarding the current management and proposed use of prescribed fire on the OSDF cap. The public was provided information about the planning and implementation of prescribed burns at the site and invited to observe the execution of a successful site burn. This approach led to stakeholder acceptance of DOE's proposal, and vegetation on the caps of cells 4, 5, and 6 was burned in late March 2016. The burn was successful and resulted in several benefits from a land stewardship and cost perspective. This paper details the stakeholder engagement approach and the resulting benefits from the use of prescribed fire on the OSDF cap. (authors)

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DOE OSTI2015

Disposing of waste package with respect to physico-chemical characteristics - 15041

The waste inventory to be taken into account for the design of the Cigeo project is defined in the industrial program for waste management established in cooperation between nuclear operators (EDF, AREVA, CEA) and Andra. This waste inventory has been constituted with a list of different families of radioactive waste packages from High Level Waste (HLW) to Intermediate Level Long-lived Waste (IL-LLW). IL-LLW include a large variety of items such as structural elements from fuel assemblies (cladding from the fuel rods called 'hulls', end pieces called 'end caps' and assembly spacer grids, etc.), sludge from effluent treatment or miscellaneous equipment (filters, pumps, etc.). Most ILLW (except solidified sludges) are basically metallic but organic and inorganic components such as plastics (cellulose, PVC...) and salts may also be included. The rules used for designing Cigeo are based on the separation of waste having different physico-chemical characteristics in different cells. However it could be possible to dispose of waste packages with different physico-chemical characteristics in the same cell if we can demonstrate that a type of waste does not generate significant physico-chemical perturbations on other co-stored families. The first step of our analysis has been the identification of categories of materials present in waste likely to interact with each other. The second step consisted in analysing qualitatively the physico-chemical interactions between these materials. In the third step, we defined criteria allowing evaluation of the consequences of these interactions on (i) gas production (mainly H{sub 2}), (ii) radionuclide release and transfer and (iii) degradation of concrete components of the disposal cell. We built then a table allowing qualitative comparison of the consequences of the interactions between the materials types defined in the first step. The conclusion of this study shows that it is necessary to pay a very careful attention before disposing of waste packages containing organic components nearby waste packages containing actinides radionuclides. The geochemical phenomena coupling must also be quantitatively evaluated so as to analyse possible interaction processes between the released organic compounds and the cementitious components of the cells, the near-field clay-stone and the radionuclides migration. (authors)

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DOE OSTI2015

Status of HLW disposal in the USA and rational, progressive, paths forward based upon lessons learned in the USA and abroad since 1973 - 15103

The development of the USA's only candidate (since 1987) HLW-repository at the Yucca Mountain (YM) site, Nevada, was halted by the Obama administration in 2009. Four years later, the Obama administration proposed a new strategy for the back end of the nuclear fuel cycle that would open a consent-based HLW repository 'by 2048'. At the end of 2014, both the YM HLW repository and the new HLW-disposition (storage and disposal) strategy remain on hold due to lack of enabling legislation. With this as background, and based upon our active involvement in, monitoring of, and reporting on HLW management programs in the USA and abroad since 1973, we submit that: 1. Politics, political will, and legal actions and outcomes will continue to govern progress. 2. Public acceptance and sustained support (PASS) may influence politics and political will. 3. PASS is time-dependent and governed by: a. Ideologies; b. Perceived benefits and risks; c. Level and equity of radiation protection provided by applicable regulations; d. Perceived robustness/safety of the proposed disposition concept; and e. Perceived credibility and competency vested in the involved organizations and their respective leaders (= TRUST). We therefore recommend that the USA concurrently pursues at least two, 'centralized' HLW-storage and -disposal solutions based upon best-available relevant knowledge and majority-PASS at facility-host locations. Furthermore, safe(r) and more secure disposal of defense/government-generated HLW could probably commence at least 15 years earlier than 'by 2048' in/at: 1. The YM repository/site. 2. The WIPP repository/site. 3. 'Local' deep boreholes. (authors)

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DOE OSTI2015

Impact of Dutch nuclear fuel cycle scenarios on radionuclide inventory intended for disposal - 15201

One of the main subjects of Dutch five-year research program into geological disposal of radioactive waste, OPERA, is the post-closure radiological safety assessment. Reliable estimates of the radionuclide inventory and matrix composition are an important input for the safety assessment. The amounts and categories of waste foreseen to be disposed in 2130 the Netherlands have been compiled by the Dutch waste management organisation COVRA. Additionally, NRG has made a characterisation of the waste (radionuclides and matrix). This reference inventory is based on the Dutch base nuclear energy scenario of operating the Dutch single reactor, the Borssele NPP, until its presently scheduled shut-down in 2033. However, changes in the presently adopted nuclear fuel cycle strategy in the Netherlands may impact both the quantities of generated radioactive waste and its composition, and therefore the source term of the radiological safety assessment. To get a grip on this uncertainty a set of alternative future fuel cycle scenarios in the Netherlands, which are in compliance with scenarios formulated in the Dutch vision documents, has been considered and evaluated. The scenarios have been analysed with the computer code DANESS, 'Dynamic Analysis of Nuclear Energy System Strategies'. The result of the DANESS analyses concerns the amount and type of the radioactive waste to be finally disposed in a deep geological repository. This also includes the inventories of the individual radionuclides that are relevant for the post-closure safety assessment to be performed in a subsequent stage of the OPERA program. (authors)

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DOE OSTI2015

Transport of French PWR steam generators to very low level waste disposal - 15203

When considering the complexity of IAEA nuclear transport regulations, the transport of a Steam Generator (SG) in one un-packaged block from a dismantling plant to a nuclear waste disposal is a challenge. The choices made, the difficulties encountered and the feedback that has been learnt from it are described. The disposal of decommissioning waste requires taking into account not only the constraints of acceptance of the repository but also the specific constraints of transportation from the dismantling plant to the final disposal site. In France, these criteria and regulations are respectively under the supervision of the French National Radioactive Waste Management Agency (ANDRA) and the French Nuclear Safety Authority (ASN). In the course of the dismantling of the EDF PWR plant Chooz A, the four decontaminated SGs were shipped to the ANDRA Very Low Level Waste (VLLW) repository. After removal, the SGs were decontaminated, allowing them to attain the criteria of VLLW and thus be accepted by the ANDRA. However, to transport these decontaminated SGs as single un-packaged blocks, it was necessary to ensure compliance with the Surface Contaminated Object SCO-I regulation criteria. For this, transport safety files were written and surface contamination and radiation criteria checked. One of the four SGs had to be decontaminated again and another one had to be transported under special arrangement. Due to the exceptional nature of the transportation, specialized means of transport and stowage were also studied. (authors)

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DOE OSTI2015

SORTIE DB - Source term integrated environmental management disposal facility database - 15249

The Environmental Management Disposal Facility (EMDF) in Oak Ridge, Tennessee, is a proposed U.S. Department of Energy low-level Comprehensive Environmental Response Compensation and Liability Act (CERCLA) waste disposal facility with an operational start in the early 2020's. As a DOE low-level waste disposal facility, it is subject to DOE Order 435.1, requiring a composite analysis and possibly a performance assessment. The Source Term Integrated EMDF Data Base - SORTIE DB - provides source term estimates used in these assessments. An Oak Ridge Office of Environmental Management (OREM) Federal and Contractor team developed this state-of-the-art descriptive analytics and visualization schema. DOE OREM senior management has recognized complementary beneficial uses of SORTIE DB in decisions for accelerating clean-up progress, dealing with conflicting sequencing and budget priorities, and enhancing stakeholder and regulatory interfaces. SORTIE DB is designated as an OREM planning tool. SORTIE DB contains estimated forecast waste volumes and masses for over 800 Decontamination and Decommissioning (D and D) and Remedial Action (RA) sequenced baseline packages over the next 25+ years. Available characterization data, radionuclide identification and concentrations, and data from safety basis and emergency management documents are incorporated into SORTIE DB. (authors)

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DOE OSTI2015

Siting considerations for a deep borehole disposal facility - 15403

This paper describes considerations for siting and designing a facility for deep borehole disposal of spent nuclear fuel (SNF) and/or high-level radioactive waste (HLW). Technical considerations include geological, hydrogeochemical, and geophysical characteristics that are related to the suitability of the site for deep drilling and borehole construction, waste emplacement activities, waste isolation (seals and surrounding geology), and long-term safety of the deep borehole disposal system. Logistical factors to be considered include: local or regional availability of drilling contractors (equipment, services, and materials) capable of drilling a large-diameter hole to approximately 5,000 m depth; construction of surface facilities and waste handling and emplacement equipment; legal and regulatory requirements associated with drilling, operation, and post-closure; and access to transportation systems. Social and political considerations include the distance from population centers and the support or opposition of local and state entities and other stakeholders to the facility and its operations. These technical, logistical, and socio-political factors are examined in the context of a consent-based siting process for a deep disposal borehole, which would be the first phase in developing an operating facility. (authors)

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DOE OSTI2016

Final Closure of the Maxey Flats Low-Level Radioactive Waste Disposal Site - 16088

This paper describes the final closure remedy for the former Maxey Flats Low-Level Radioactive Waste Disposal Site (MFDS), the challenges presented at the site due to the condition of the disposal facility and waste, and their resolution through comprehensive geo-synthetic design. The MFDS is an inactive landfill located in Kentucky on a ridge surrounded by a buffer zone. Historically, waste disposal occurred without reducing voids or without compacting the placed waste. Settlement and water management issues led to releases of radioactive contamination into surrounding waters, resulting in the closure of the facility in 1977. The final closure remedy consists of 1) development and implementation oversight of a sump abandonment design to ensure no direct pathways for infiltrating water to contact the waste and 2) development and implementation oversight of the Final Closure Period (FCP) cap design. Challenges successfully addressed by the FCP scope included: Abandon sumps to minimize storm water migrating into the waste; Address potential existing groundwater waste interaction; Maximum use of on-site cap construction materials; Design cap to fit very restricted site footprint; Assess and address potential isolated differential settlement; Maximize passive storm water management; Provide alternative site access during construction. The FCP remedy and resultant design satisfies these challenges through use of layers of geo-grid-reinforced soil and geo-synthetic materials, contours and slopes that minimize the quantities of construction materials while maximizing the use of available material, probabilistic subsidence modeling, an integrated storm water management system ensuring the slow release of storm water into receiving streams, and a dedicated construction haul road system. (authors)

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DOE OSTI2016

Hydrogeologic Lessons Learned from Deep Drilling Projects Applied to Deep Borehole Disposal of Radioactive Wastes - 16133

Hydrogeological results from the Kola SG-3, Gravberg-1, Cajon Pass, and KTB-VB and KTB-HB boreholes are applied to the US deep borehole disposal concept to assess whether old, saline, and reducing deep groundwater, in low bulk permeability crystalline basement rocks, could be found in a DOE proposed deep borehole field test. Each of the drilling projects faced characterization challenges, including obtaining sufficient information to characterize key parameters and their uncertainty, equipment limitations and failures, and obtaining samples that are representative of in situ conditions while maintaining sample integrity. Despite these characterization challenges, each project obtained sufficient data to directly measure or infer, for redox state, conditions at the 3 to 5 km (1.9 - 3.1 miles) target disposal depth of the US deep borehole disposal concept. The hydrogeologic data suggest that old, saline, and reducing deep groundwater could be found at a test site located in geologically stable crystalline basement. Although the DOE deep borehole testing program has not defined the magnitude of an acceptably low bulk permeability, bulk permeability results indicate that values used in current US deep borehole modeling efforts may be too low by an order of magnitude or more, as compared with in situ conditions. (authors)

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