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DOE OSTI2024

The DECOVALEX international collaboration on modeling of coupled subsurface processes and its contribution to confidence building in radioactive waste disposal

Abstract The long-lived radiotoxicity of the high-level radioactive waste generated by nuclear power plants requires safe isolation from the biosphere for many hundreds of thousands of years. An international consensus has emerged that such isolation can best be provided by disposal in mined geologic repositories, a strategy that today is pursued by most countries dealing with radioactive waste. However, the need to predict the performance of such repositories over very long time periods generates large uncertainties that have to be accounted for in safety assessments. The findings from such safety assessments need to be conveyed to all stakeholders in a clear way, such that public confidence in geologic disposal solutions can be achieved. It is suggested here that close international collaboration on the technical aspects of geologic waste disposal has helped, and will continue to help, building trust and increasing confidence. This paper discusses a particular international collaboration initiative referred to as DECOVALEX, which brings together multiple teams and disciplines to collectively tackle complex experimental and modeling challenges related to geologic disposal. By describing how DECOVALEX works and by providing joint research examples, a case is made that such international collaboration contributes to knowledge transfer and confidence building in radioactive waste disposal science.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Gas migration in low- and intermediate-level waste (LILW) disposal facility in Korea

The first low- and intermediate-level waste (LILW) disposal facility is under construction in saturated granite in Korea. The safety assessment report (SAR) identified that different gases, such as hydrogen, carbon dioxide, and methane are generated at the disposal facility due to the corrosion of metal wastes and steel drum, and microbial degradation of organic matters. Reinforced concrete plays a role as an engineered barrier at the disposal facility, so its properties with regard to gas migration were evaluated in laboratory-scale experiments. Then modeling of gas migration was carried out to evaluate gas pressure build-up in the disposal facility. The gas entry pressure and relative gas permeability of the concrete was determined to be 0.97±0.15 bar, and the relative gas permeability decreased exponentially with increasing water content. The results of the modeling showed that most of hydrogen gas was dissolved in groundwater and did not significantly influence pressure build-up inside the disposal facility based on the reference case of gas generation. (authors)

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DOE OSTI2013

2013 Annual Site Inspection and Monitoring Report for Uranium Mill Tailings Radiation Control Act Title II Disposal Sites

This report, in fulfillment of a license requirement, presents the results of long-term surveillance and maintenance activities conducted by the U.S. Department of Energy (DOE) Office of Legacy Management in 2013 at six uranium mill tailings disposal sites reclaimed under Title II of the Uranium Mill Tailings Radiation Control Act (UMTRCA) of 1978. These activities verified that the UMTRCA Title II disposal sites remain in compliance with license requirements. DOE manages six UMTRCA Title II disposal sites under a general license granted by the U.S. Nuclear Regulatory Commission (NRC) established at Title 10 Code of Federal Regulations Part 40.28. Reclamation and site transition activities continue at other sites, and DOE ultimately expects to manage approximately 27 Title II disposal sites. Long-term surveillance and maintenance activities and services for these disposal sites include inspecting and maintaining the sites; monitoring environmental media and institutional controls; conducting any necessary corrective action; and performing administrative, records, stakeholder services, and other regulatory functions. Annual site inspections and monitoring are conducted in accordance with site-specific long-term surveillance plans (LTSPs) and procedures established by DOE to comply with license requirements. Each site inspection is performed to verify the integrity of visible features at the site; to identify changes or new conditions that may affect the long-term performance of the site; and to determine the need, if any, for maintenance, follow-up inspections, or corrective action. LTSPs and site compliance reports are available online at http://www.lm.doe.gov

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

International Approaches for Nuclear Waste Disposal in Geological Formations: Report on Fifth Worldwide Review

An important issue for present and future generations is the final disposal of spent nuclear fuel. Over the past over forty years, the development of technologies to isolate both spent nuclear fuel (SNF) and other high-level nuclear waste (HLW) generated at nuclear power plants and from production of defense materials, and low- and intermediate-level nuclear waste (LILW) in underground rock and sediments has been found to be a challenging undertaking. Finding an appropriate solution for the disposal of nuclear waste is an important issue for protection of the environment and public health, and it is a prerequisite for the future of nuclear power. The purpose of a deep geological repository for nuclear waste is to provide to future generations, protection against any harmful release of radioactive material, even after the memory of the repository may have been lost, and regardless of the technical knowledge of future generations. The results of a wide variety of investigations on the development of technology for radioactive waste isolation from 19 countries were published in the First Worldwide Review in 1991 (Witherspoon, 1991). The results of investigations from 26 countries were published in the Second Worldwide Review in 1996 (Witherspoon, 1996). The results from 32 countries were summarized in the Third Worldwide Review in 2001 (Witherspoon and Bodvarsson, 2001). The last compilation had results from 24 countries assembled in the Fourth Worldwide Review (WWR) on radioactive waste isolation (Witherspoon and Bodvarsson, 2006). Since publication of the last report in 2006, radioactive waste disposal approaches have continued to evolve, and there have been major developments in a number of national geological disposal programs. Significant experience has been obtained both in preparing and reviewing cases for the operational and long-term safety of proposed and operating repositories. Disposal of radioactive waste is a complex issue, not only because of the nature of the waste, but also because of the detailed regulatory structure for dealing with radioactive waste, the variety of stakeholders involved, and (in some cases) the number of regulatory entities involved.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Assessment of Baseline Monitoring Data for the Remote-Handled Low-Level Waste Disposal Facility at Idaho National Laboratory

The purpose of this document is to summarize environmental monitoring data collected at the Remote-Handled Low-Level Waste (RHLLW) Disposal Facility during the first four years of facility operations (FY 2019 through FY 2022). This summary provides a “baseline” condition for the facility, which can be used to distinguish contaminant releases from the RHLLW Disposal Facility from pre existing contamination as well as potential future releases from other sources (i.e., upgradient aquifer sources), as specified in the facility monitoring plan. Sufficient data has been collected over the first four years of operation of the RHLLW Disposal Facility to establish baseline conditions for future compliance monitoring of aquifer wells and performance monitoring of vadose zone lysimeters. Except for seven elevated tritium measurements from lysimeter HFEF-South believed to have been impacted by waste disposals, all data was deemed appropriate for establishing baseline concentrations. Monitoring of the aquifer detected indicator analytes gross alpha and gross beta and target analytes tritium and C-14. There were a few C-14 detections prior to an increase in the required detection level (RDL) after the first round of sampling. I-129 and Tc-99 were not detected above RDLs in aquifer samples. Lysimeter samples detected indicator analytes gross alpha and gross beta and target analyte tritium. Target analytes C-14, I-129, and Tc-99 were not detected above RDLs in lysimeter samples. Except for tritium, baseline concentrations of indicator and target analytes measured during the first four years of RHLLW facility operations are expected to represent conditions through the expected 20 year operating period of the facility. Elevated tritium in the aquifer, a result of past discharges of tritium at upgradient facilities, showed a decline over the first four years of operations, and levels are expected to continue to decline with time as a result of dilution and decay. Current tritium levels in lysimeter samples may be elevated due to tritium in the water applied during construction and infiltration testing of the facility. As a result, tritium concentrations in lysimeters may decline with time as a result of dilution and decay. Statistical measures (i.e., mean, standard deviation, and 99% upper confidence level [UCL]) were calculated for all detected analytes using the four years of concentration data. These measures will be used to evaluate future monitoring results and to demonstrate the facility is performing, or not performing, as established in the facility performance assessment (PA). Future monitoring results will also be combined with the baseline data in this report to further establish temporal trends and natural variability in the measurements. Based on the data collected during the baseline period, it is recommended the gross alpha action level for performance monitoring increase from 10 to 20 pCi/L due to several baseline measurements exceeding the preliminary action level of 10. An action level of 20 pCi/L is slightly greater than the 99% UCL, is protective of the aquifer, and would reduce unnecessary sampling. It is also recommended that tritium be added to the lysimeter analyte list for post-baseline period monitoring. Tritium, while not a dose concern, is a good tracer that can provide valuable information on water flow in and around the RHLLW Disposal Facility. Recommendations for a tritium action level are provided if considered necessary.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Radionuclide Transport Simulations Supporting Proposed Borehole Waste Disposal in Israel

A scientific collaboration between the U.S. and Israel is underway to assess the suitability of a potential site for subsurface radioactive waste disposal in the Negev Desert, Israel. The Negev Desert has several favorable attributes for geologic disposal, including an arid climate, a deep vadose zone, interlayered low-permeability lithologies, and carbonate rocks with high uranium-sorption potential. These features may provide a robust natural barrier to radionuclide migration. Geologic and laboratory characterization data from the Negev Desert are incorporated into multiphase flow and transport models, solved using PFLOTRAN, to aid in site characterization and risk analysis that will support decision-making for waste disposal in an intermediate-depth borehole design. The lithology with the greatest uranium sorption potential at the site is phosphorite. We use modeling to evaluate the ability of this layer to impact uranium transport around a proposed disposal borehole. The current objective of the simulations is focused on characterizing hypothetical leakage from waste canisters and subsequent uranium migration under three infiltration scenarios. Here, we describe a hydrogeologic model based on data from a local exploratory borehole and present results for uranium flow and transport simulations under varying infiltration scenarios. We find that under the current climate conditions, it is likely that uranium will remain in the near-field of the borehole for thousands of years. However, under a hypothesized extreme climate scenario representing an increase in infiltration by a factor of 300x above present-day values, uranium may break through the phosphorite layer and exit the base of the model domain (~200 m above the water table) within 1000 years. Simulation results have direct implications for the planning of nuclear waste disposal in the Negev Desert, and specifically in intermediate-depth boreholes.

58 GEOSCIENCES↗
DOE OSTI2012

Treatment of Irradiated Graphite to meet Acceptance Criteria for Waste Disposal: A New IAEA Collaborative Research Program - 12443

World-wide, more than 250,000 tonnes of irradiated graphite have arisen through commercial nuclear-power operations and from military production reactors. Whilst most nations responsible for the generation of this material have in mind repository disposal alongside other radwaste, the lack of progress in this regard has led in some cases to difficulties where, for example, the site of an existing graphite-moderated reactor is required for re-utilisation. In any case, graphite as a radwaste stream has unique chemical and physical properties which may lend itself to more radical and innovative treatment and disposal options, including the recovery of useful isotopes and also recycling within the nuclear industry. Such aspects are important in making the case for future graphite-moderated reactor options (for example, High-Temperature Reactors planned for simultaneous power production and high-grade heat sources for such applications as hydrogen production for road fuel). A number of initiatives have taken place since the mid 1990s aimed at exploring such alternative strategies and, more recently, improving technology offers new options at all stages of the dismantling and disposal process. A new IAEA Collaborative Research Program aims to build upon the work already done and the knowledge achieved, in order to identify the risks and uncertainties associated with alternative options for graphite disposal, along with cost comparisons, thus enabling individual Member States to have the best-available information at their disposal to configure their own programs. (authors)

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DOE OSTI2012

3DD - Three Dimensional Disposal of Spent Nuclear Fuel - 12449

Three dimensional disposal is being considered as a way in which to store long-term spent nuclear fuel in underground disposal facilities in the Czech Republic. This method involves a combination of the two most common internationally recognised disposal methods in order to practically apply the advantages of both whilst, at the same time, eliminating their weaknesses; the method also allows easy removal in case of potential re-use. The proposed method for the disposal of spent nuclear fuel will reduce the areal requirements of future deep geological repositories by more than 30%. It will also simplify the container handling process by using gravitational forces in order to meet requirements concerning the controllability of processes and ensuring operational and nuclear safety. With regard to the issue of the efficient potential removal of waste containers, this project offers an ingenious solution which does not disrupt the overall stability of the original disposal complex. (authors)

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DOE OSTI2017

Safety Assessment of El Cabril, the Spanish Surface Disposal facility for LILW and VLLW - 17233

El Cabril project was launched in 1986, for Low and Intermediate Level Waste (LILW) disposal, based on concrete barriers and concrete disposal units. The operation of a Very Low Leve Waste (VLLW) complementary facility based on clay and High Density Polyethylene (HDPE) barriers and different types of disposal units started in 2008. Two main objectives are considered in the design: Immediate and post-closure protection and no radiological restrictions, after institutional control period. To fulfil these objectives several design criteria like waste isolation through a multi-barrier system, surveillance period no longer than 300 years and limitation of activity are applied. The Safety Assessment process follows a methodology which, taking into account important facts like safety and design criteria, time frame, radiological protection criteria, waste characteristics, environment, FEPs (features, events and processes), etc., develops different scenarios, models and calculations to define the safety approach. This safety approach follows the current guides and recommendations of international organizations, it covers the potential impact of the facility for the operational and long-term periods and it includes normal evolution and less likely accidental and inadvertent human intrusion events. All the studies and considerations done for the Safety Assessment have to be included in a safety Report, required for getting the Operation License for Nuclear Installations in Spain. The content of this Report is defined by the Spanish regulation and it has to be revised if there is a change in the facility that may affect the safety. It is also mandatory making a Safety Periodic Review (every 10 years in the case of El Cabril) that has to be presented to the Regulatory Council. This paper outlines the principles, objectives and criteria of El Cabril's disposal installation and it explains the safety assessment performed, for both LILW and VLLW disposal, as part of the documentation produced to obtain the authorization to operate. (authors)

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DOE OSTI2017

Integrated Disposal Facility Performance Assessment: Modeling Consistency to Support Regulatory Decision Making at Hanford - 17398

Under DOE Order 435.1, 'Radioactive Waste Management', Washington River Protection Solutions, LLC and INTERA, Inc are currently preparing a performance assessment for the Hanford Integrated Disposal Facility, a near-surface low-activity waste disposal facility located on the Hanford Site in southeast Washington State. The disposal facility is expected to receive vitrified low-activity waste produced at the Hanford Waste Treatment and Immobilization Plant, as well as other secondary solid wastes generated during the treatment process. The performance assessment performs the analyses that will be relied upon to receive the Disposal Authorization Statement, establish waste acceptance criteria, modify the RCRA permit for the facility and make a waste-incidental-to-reprocessing determination for the vitrified tank waste. In January 2006 in the case State of Washington v. Bodman (Civil No. 2:03-cv-05018-AAM), DOE, the Washington State Department of Ecology (Ecology) and the State of Washington Attorney General's office reached a settlement agreement in Ecology's claim regarding off-site waste shipments to Hanford and the adequacy of modeling in the environmental impact statement. In response to the agreement, the DOE Office of Environmental Management set a policy to establish a consistent use of commercially available and public domain software for a vadose zone and groundwater model platform to support regulatory decisions at Hanford. In October 2012, the Associate Principal Deputy Assistant Secretary for Environmental Management released a memorandum to the manager of the Richland Operation Office directing how future environmental modeling activities to support regulatory decision making on the Hanford Site would be undertaken to be consistent with the 2006 settlement agreement. The preparation of the performance assessment must meet DOE Order 435.1 requirements and the additional direction provided by DOE. This paper will identify the approach taken to comply with the requirements in the 2012 memorandum. The paper will describe how environmental modeling performed for the 2017 Integrated Disposal Facility Performance Assessment was planned in a phased approach and how the current modeling is tied to the analyses that supported the decisions made in the environmental impact statement published in 2012. Additionally, the selection and qualification of the software selected to perform modeling activities for the performance assessment will be described. (authors)

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DOE OSTI2017

Achieving Leak Tightness in New Savannah River Site Saltstone Disposal Units - 17454

Saltstone Disposal Unit (SDU) 6 is a new disposal unit at the Department of Energy's Savannah River Site Saltstone Disposal Facility. The Saltstone Facility is part of the site's overarching waste management and disposal system. It receives the low-level waste stream and will immobilize in grout more than 90 percent of the liquid volume generated during treatment of the site's liquid high level waste. SDU-6 is the first in a series of large disposal units to be constructed on site. Its internal diameter and height are 114 m (375 ft) and 13.1 m (43 ft), respectively. It uses a cylindrical wire-wrapped concrete water tank design (AWWA D110, Type I), and at 121 ML (32 Mgal), it is one of the largest structures of its kind in the country. Because of its size, SDU-6's base slab was placed in ten sections. Earlier placed sections restrained newer adjacent sections, resulting in shrinkage cracking. Visible cracks were repaired with epoxy. However, the SDU basemat leaked at less than an estimated 4 L/min rate during its first full-height hydrostatic test (2015). The project baseline was to install a coating system to chemically protect the concrete from the high-pH, high-sulfate Saltstone solutions. The coating was not installed during the first hydrostatic test, and subsequently, an adhered elastomeric liner system was chosen to provide higher assurance of bridging floor cracks and preventing leakage. A robust process was used to select and install a liner that will not only provide chemical protection but also achieve leak tightness, which was successfully demonstrated in a second hydrostatic test late in 2016. (authors)

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DOE OSTI2015

Approaches to consider covers and liners in a low-level waste disposal facility performance assessment - 15300

On-site disposal cells are in use and being considered at several US DOE sites as the final disposition for large amounts of waste associated with cleanup of contaminated areas and facilities. These disposal cells are typically regulated by States and/or the USEPA in addition to having to comply with requirements in DOE Order 435.1, Radioactive Waste Management. The US DOE-EM Office of Site Restoration formed a working group to foster improved communication and sharing of information for personnel associated with these Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) disposal cells and work towards more consistent assumptions, as appropriate, for technical and policy considerations related to performance and risk assessments in support of a Record of Decision and Disposal Authorization Statement. One task completed by the working group addressed approaches for considering the performance of covers and liners/leachate collections systems in the context of a performance assessment (PA). A document has been prepared which provides recommendations for a general approach to address covers and liners/leachate collection systems in a PA and how to integrate assessments with defense-in-depth considerations such as design, operations and waste acceptance criteria to address uncertainties. Specific information and references are provided for details needed to address the evolution of individual components of cover and liner/leachate collection systems. This information is then synthesized into recommendations for best practices for cover and liner system design and examples of approaches to address the performance of covers and liners as part of a performance assessment of the disposal system. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Simplifying the Tracking of Waste Disposal Inventories: Process Improvement for Rapid Data Interpretation and Reduced Errors - 16143

Tracking the inventory of low-level radioactive waste at an operating disposal facility offers unique challenges. At the Savannah River Site's Saltstone Disposal Facility (SDF) the inventory used in Performance Assessment modeling is determined based on radionuclide concentrations from the aqueous waste in a feed tank and the volume of the waste that is transferred from the feed tank to the Saltstone Production Facility (SPF). These inputs are reported quarterly. As such, the disposal inventories are also updated quarterly; however, due to decay and ingrowth, the inventory values for some radionuclides can be misrepresented if the data is not tracked appropriately. Additionally, it is not feasible to analyze the feed tank samples for every possible radionuclide; therefore, a series of adjustments are necessary to estimate the concentrations of radionuclides that are not explicitly measured but still pose a potential dose risk. Prior to 2015, the SDF disposal inventories were updated using a series of spreadsheets that required users to perform a significant number of steps (including manually transferring data back and forth between a separate decay and ingrowth calculation software). Despite this labor intensive approach, the process only provided single point-in-time results. With multiple steps, minimal automation, and limited result data, this approach to inventory tracking was not ideal and minor mistakes were often identified during the equally cumbersome review process. As a process improvement, the simplified SDF Waste Inventory Disposed Estimator (SDF-WIDE) Model was developed using GoldSim software. The software allows users to identify specific dates to define the duration of the simulation, from the beginning of operations until the time of future facility closure, and provides graphical interfaces for displaying data over time. Further, GoldSim's Contaminant Transport Module is used to perform the decay and ingrowth calculations, making the model completely self-contained (i.e., there is no need to transfer intermediate data back and forth between different programs). GoldSim also includes model 'versioning' features that significantly reduce the time needed to perform the appropriate quality assurance reviews as required for each quarterly inventory update. By transitioning from the complex spreadsheets to the simplified SDFWIDE model, the time needed to perform updates to the SDF inventory has been reduced from days to hours, the risk of making human errors has been greatly mitigated, and users are able to quickly observe trends in the data over time. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Evaluating Contaminant Transport in the Vadose Zone to Support Remedy Decisions at Aqueous-Waste Disposal Sites - 16195

Inorganic contaminants in aqueous waste disposed at land surface have a potential to contaminate groundwater. However, contaminant concentrations are attenuated during transport through the vadose zone. Quantifying contaminant attenuation and the resulting temporal concentration profile of in underlying groundwater is important for assessing the need for, and type of, remediation in the vadose zone and groundwater. An approach was developed for evaluating vadose zone transport and attenuation of aqueous wastes containing inorganic (non-volatile) contaminants that were disposed at land surface (i.e., directly to the ground in cribs, trenches, tile fields, etc.) and their effect on underlying groundwater. The approach provides a structured method to estimate transport of contaminants through the vadose zone and the resulting temporal profile of groundwater contaminant concentrations. A series of numerical model simulations was applied to investigate how different factors affect contaminant transport in the vadose zone and the conditions for which individual factors are important. This analysis demonstrated that, while water flow and contaminant transport can be complex during aqueous-waste disposal, conditions will change over time after disposal ceases and fewer factors will control the contaminant transport rate and resulting temporal profile of groundwater contaminant concentrations. Data verifying that these conditions have been reached can be provided by a few key metrics such as the vertical profile of moisture content, the temporal profile of recharge at the site, and basic information about waste disposal and subsurface properties. Under these conditions, and with information about specific contaminant transport parameters (e.g., for partitioning), estimates of contaminant transport can be made and used to support remedy decisions. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Macro-encapsulation as Manufactured Components for Disposal at the Nevada National Security Site (NNSS) - 16213

Macro-encapsulation as Manufactured is a relatively new terminology that supports the disposal of classified legacy weapon components at the Area 5 Disposal Cell at the Nevada National Security Site (NNSS). These components are composed of hundreds and sometimes thousands of parts. Examples of this type of item include firing sets, trainers, radar assemblies, neutron generators and other similar components. Sometimes there are hazardous Resource Conservation and Recovery Act-regulated (RCRA) metals involved in the parts that make up the components. The typical regulated metals in the components can be lead and silver from circuit boards and wiring, cadmium or chromium from plating operations of the parts. To perform the Toxicity Characteristic Leaching Procedure (TCLP) components have to be completely disassembled and sanitized prior to being sent to laboratories for analysis. The results vary but the majority of times there are regulated metals present in these samples. The levels are relatively low and since analysis has been performed then a disposal process needs to be followed. A process needed to be developed and approved by governing boards to allow disposal of these items as long as this material meets the Waste Acceptance Criteria (WAC) (No liquids, compressed gases, explosives, SNM, etc.) before being sent to the Nevada desert for disposal. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Documentation of Radioactive Waste Packages for Disposal in the Konrad Repository - 16253

For the Konrad repository, the plan approval procedure has been completed in 2002. In 2007 all complaints filed against the repository were dismissed, so that there is a legal and un-appealable license decision available. The waste acceptance requirements were revised taking into account additional requirements from the plan approval decision. Today, there are already several hundred waste packages produced that meet all requirements for final disposal in Konrad. The acquisition, verification and quality assurance of the data relevant for disposal is performed in parallel to the waste treatment and packaging. This constitutes the development of database applications, on the one hand to capture and handle specific data for the waste packages to another to manage the disposal procedures. The development of a database application for recording disposal-relevant data of waste packages has already been completed with regard to the radiological requirements. Additional aspects based on the requirements of the plan approval decision, e.g. water law relevant aspects, have to be taken into account. Thus, an upgrade relating to the recording of groundwater debilitating substances will be carried out in a second step, which is expected to be completed in 2016. In parallel, the development of a database application is carried out to coordinate the waste package deliveries to the repository and to document their disposal. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Advances in Thermal-Hydrologic Modeling of Nuclear Waste Disposal in Deep Boreholes - 16303

Deep borehole disposal is one of the potential options for safely isolating high-level radioactive waste. It is expected that existing drilling technology can provide reliable and cost-effective construction of suitable deep boreholes. In addition, favorable disposal conditions such as low permeability host rock, high salinity, and geochemically reducing conditions, exist at depth in many locations. Coupled thermal-hydrologic processes induced by heat from the radioactive waste may impact fluid flow and the associated migration of radionuclides. This work looks at those processes as was also done with recent studies. Simulations of thermal-hydrology for the emplacement of cesium and strontium capsules in a deep borehole are presented. The simulations looked at disposal options such as different disposal configurations and aging of waste to reduce thermal effects. The simulations studied temperature and fluid flux in the vicinity of the borehole. Simulation results include temperature and vertical flux profiles around the borehole at selected depths. Of particular importance are peak temperatures, and fluxes above the disposal zone. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Secondary Waste Form Development and Data Package for Hanford's Integrated Disposal Facility Performance Assessment - 16479

The Hanford Low-Activity Waste (LAW) Vitrification facility is planned to immobilize Hanford tank waste for on-site, near-surface disposal at Hanford's Integrated Disposal Facility (IDF). During this process, liquid and solid secondary wastes will be generated which are also slated for disposal at the IDF. In order to receive authorization to dispose of the waste in the IDF, a Performance Assessment (PA) must be completed in accordance with DOE Order 435.1, Radioactive Waste Management, and its accompanying Manual. The PA must include calculations of potential doses to a member of the public for a 1,000-year period to provide a reasonable expectation that performance objectives will not be exceeded and that the waste will be managed in a manner that is protective of worker and public health and safety, and the environment. Liquid and solid secondary wastes are planned to be immobilized within cementitious waste forms similar to those used elsewhere in the United States and around the world for permanent disposal of low-level radioactive waste (LLW). Data packages and supporting documents are being developed to provide the technical underpinning for the analyses of contaminant releases to the environment from these waste forms for the IDF PA. Traditionally, releases from cementitious waste forms have utilized a diffusion-limited release model using contaminant-specific diffusivities measured in standard leach tests as inputs. More mechanistic modeling techniques are under development to increase the rigor of the PA analyses. In addition, recent testing has shown diffusivities of key contaminants can be reduced by several orders-of-magnitude through optimized formulation of the waste form and use of additives to bind specific contaminants. (authors)

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