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At least 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Effects of SRPPF ARS Effluent Variations on Magnesium Oxysulfate Solidification

A magnesium oxysulfate (MOS) grout formulation has been identified to solidify the liquid effluent from the Savannah River Plutonium Processing Facility (SRPPF) Aqueous Recovery System (ARS). The formulation uses reactive, light burnt MgO (28 wt%), anhydrous MgSO 4 (10 wt%), and dead burnt MgO (62 wt%), resulting in a grout with good mixability, acceptable density, acceptable setting times, and a leachate pH around 9.4, within the assumed Waste Isolation Pilot Plant (WIPP) brine pH range. A series of tests were then designed to observe how this formula reacted to variations in the liquid effluent. Liquid effluent temperature, the caustic pH level, concentration of NaNO 3 , and the presence of neutralization products, trace metals, and/or sodium sulfate (all potential residuals found in the effluent from upstream processing), were evaluated to see how each impacts the grout mixing time, peak temperature, setting time, presence of bleed water, leachate pH, and density. The standard mix has an average mixing time of 18 minutes, peak temperature of 100.5 °C, and set time of 60 minutes. Increasing the temperature of the liquid effluent increased the reaction rate of the mix and reduced mixing and setting times. Between 30 and 40°C the change was relatively small, raising no more than 6 °C compared to the standard mix maximum temperature. Increasing the pH of the liquid effluent had no significant impact on the mix. The addition of neutralization products and trace metals had an overall impact of decreasing the reactivity of the mix. The addition of more ions in the liquid effluent, such as an increase in the concentration of NaNO 3 and the addition of Na 2 SO 4 , generally decreased the reactivity of the mix. None of the variations to the liquid effluent hindered solidification as indicated by the lack of bleed water found on all the samples. The leachate pH values for all mixes tested did not significantly vary from the expected pH of 9.4 and none of the density values fell below 1.8 g/cm 3 . Overall, changes to the liquid effluent were found to have a minimal impact on the formulation suggesting the grout's ability to properly form despite increased temperatures and the presence of residuals typically found in the liquid effluent from the ARS.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Waste Encapsulation and Storage Facility Stabilization and Ventilation Project-18121

Department of Energy contractor CH2M HILL Plateau Remediation Company (CH2M) is responsible for the safe and compliant operation of the Waste Encapsulation and Storage Facility (WESF), a nuclear materials storage facility at Hanford containing 1,936 highly radioactive capsules containing cesium and strontium. The capsules are currently stored under approximately 4 m (13 ft) of water. The water shields workers from radiation and helps cool the capsules. WESF is a large, concrete structure nominally 48 m (160 ft) long, 30 m (98 ft) wide and 12 m (40 ft) high with the pool cells being approximately 4 m (13 ft) below grade level. It is a robust facility with walls ranging from 0.3 m to 1 m (1 ft to 3 ft) thick. While the capsules are currently in a safe configuration, WESF is an aging facility, and DOE and CH2M are modifying the facility to ensure continued safe and compliant operations and to support the eventual transfer of capsules to dry storage. Dry storage would eliminate the possibility of a release of radioactive material in the unlikely event of a major earthquake that might result in loss of pool storage water and overheating and breach of one or more capsules. The project was divided into two major deliverables: 1) Ventilation: replacement of the existing ventilation system which included a below-grade high efficiency particulate air (HEPA) filter pit with an above grade HEPA filter skid and new fans and ducting. 2) Stabilization: grouting deactivated facility hot cells and other void spaces to the maximum extent practical to reduce material at risk from legacy contamination. The aging ventilation system relied on HEPA filters that exceeded their operational life. Due to the high levels of radionuclide contamination, replacing the filters was impractical and highly hazardous. CH2M installed a new ventilation system to provide safe and compliant ventilation for future operations. The project used an existing, unused ventilation skid that was designed for another project. Repurposing the unused skid presented additional technical challenges to modify the system for the project specific requirements. The new system is above ground, easier to maintain and is built to current standards. New connections inside the facility were installed to allow the new ventilation to operate effectively. The stabilization project is unique in that core drilling and grouting was performed in an operating Hazard Category 2 nuclear facility, which brought forth safety basis and environmental regulatory challenges. The hot cells were isolated from all existing support utilities including electrical, process chemicals, manipulator ports, instrument air, etc. to seal the hot cells and prevent highly contaminated supernate or grout from leaking into adjacent galleries. A high flow grout was used to maximize the distance between injection ports. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Assessing High-energy, Laser-driven, X-ray Sources as a Technique to Analyse ILW Containers - 18242

The principle concern regarding intermediate level waste (ILW) containers is the internal volumetric expansion due to the formation of corrosion products. To tackle this (via non-destructive testing) the development of a new technique to identify problematic containers was initiated by radiographic imaging and neutron assay of small scale, simulated, nuclear waste packages. This was achieved using a high-energy laser pulse to produce a bright burst of high-energy Bremsstrahlung radiation for imaging and a pulse of thermal energy neutrons for isotopic quantification. The small scale, simulated, nuclear waste packages (typically a quarter of a 28 mm natural uranium penny encased in grout) were produced with corrosion from oxygen and hydrogen rich environments to simulate the potential effects of long term storage. These samples were characterised using electron and ion beam techniques prior to encapsulation to establish the extent of corrosion and sample morphology. The small-scale samples were analysed using high-resolution x-ray tomography (XRT). These 3D pictures identified the corrosion state of the material and the effect on the surrounding grout matrix. These data sets were used as a baseline to compare with the 2D images obtained from the high-energy radiography to ascertain the imaging quality and potential capability of a laser driven inspection system. The study established the non-destructive imaging capability of a high-energy laser driven x-ray source for identifying fine features such as corrosion products on uranium samples encased in grout and cracks, and for penetrative imaging through large thicknesses of grout. In addition, generation of laser-driven neutrons of thermal energy required for active interrogation of uranium isotopes is demonstrated experimentally and simulations of the neutron inspection method are presented, thus paving the way for criticality testing of ILW containers. This current study is single pulse exposure acquisition, but the data can be used to predict the potential capability of a diode-pumped laser-driven system operating at, or above, 10 Hz by utilising, for example, the Dipole laser technology recently delivered by the STFC Central Laser Facility. Such a system will enable 3D tomography of nuclear waste containers and also greatly reduced signal to noise ratio via multiple acquisition and image stacking. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Dynamic Leaching Method for Intact Salt-stone Samples - 18495

At the Savannah River Site (SRS) low activity salt solution is stabilized via encapsulation within a grout termed salt-stone. Salt-stone is emplaced into large (multi-million liter) concrete storage containers referred to as Salt-stone Disposal Units (SDUs). Technetium-99 (Tc-99) and iodine-129 (I-129) are long-lived radionuclides contained in the low activity salt waste and subsequently incorporated into the grout waste form: both radionuclides are contributors to the potential long-term radiation exposure of the public and environment surrounding the SDUs. With respect to Tc-99, in the reducing, high pH environment within the grout, Tc-99 is expected to be relatively immobile since it exists in a reduced Tc(IV) oxidation state in the form of sparingly soluble hydrated oxides (TcO{sub 2}.xH{sub 2}O) or sulfides (TcS{sub x}). However, in the presence of O{sub 2} (associated with the future infiltration of air or oxygenated ground waters into the salt-stone monolith) it is possible for redox-sensitive Tc(IV) to transition into highly soluble (and mobile) Tc(VII) species, such as pertechnetate (TcO{sub 4}{sup -}), which are more readily transported to the surrounding environment. In contrast, I-129 is highly soluble and its leaching behavior is not anticipated to be significantly impacted by potential changes in redox environment or pH within salt-stone. Traditional approaches to quantifying the leaching behavior of Tc-99 and I-29 from cementitious matrices have involved partitioning experiments using size-reduced (crushed/ground) samples, and determination of the radionuclide fraction immobilized by the cementitious solids. Such experiments create artificially high solid-liquid contact areas that likely result in higher Tc-99 and I-129 leachate concentrations than would be expected for intact, monolithic samples. In the current study a new technique, termed the Dynamic Leaching Method (DLM), is being used to initially investigate the Tc-99 leaching behavior from monolithic salt-stone samples; the system will be used at a later date to evaluate the leaching characteristics of I-129. The data derived using this technique is intended to inform the SRS Salt-stone Disposal Facility (SDF) Performance Assessment (PA) which models the long-term transport of radionuclides from the SDUs to the environment. The DLM utilizes a flexible-wall permeameter to achieve saturated leaching under an elevated hydraulic gradient to simulate and accelerate the future transport of groundwater through salt-stone. Leachates are regularly analyzed with respect to radionuclide concentrations, key constituent concentrations (e.g., Al, Ca, Mg, Na, Si, CO{sub 3}{sup 2-}, NO{sub 2}{sup -}, NO{sub 3}{sup -}, SO{sub 4}{sup 2-}), pH, and reduction potential (E{sub h}) in order to develop an understanding of how chemical changes associated with the pore solution influence grout weathering and radionuclide leaching behavior. This paper provides a brief update with respect to data derived from preliminary Tc-99 leaching evaluations as the DLM was being developed, but the paper is predominantly focused on new sample data in which the DLM technique, pre- and post-DLM salt-stone sample characterization, and leachate analysis protocols have been optimized. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Contaminant Migration Modeling to Support the In Situ Decommissioning of Hardened Nuclear Facilities - 20215

The decommissioning of hardened nuclear facilities provides a unique challenge in balancing current worker dose with exposure to future receptors and media. In situ decommissioning (ISD) is a cost-effective and safe option for the closure of such facilities but requires a detailed understanding of the threat that radioactive and hazardous constituents remaining in the facilities pose to groundwater and surrounding surface water. The Savannah River Site (SRS) extensively employs contaminant migration models within the decommissioning process to develop site-specific removal, grouting, and monitoring strategies in support of safe and effective final end states of nuclear facilities. This work discusses the use of contaminant migration modeling in the ISD process at SRS, including specific examples from the closure of P-Reactor and R-Reactor, Building 235-F, F-Canyon and F/H Laboratory Complex. Migration models developed during the closure of P-Reactor and R-Reactor were divided into four source areas: the reactor vessel, the process area, the disassembly basin, and the purification wing. Each source area was assigned a specific inventory and migration pathway and was then subject to varying hypothetical removal, capping, and grouting schemes. Modeling indicated that groundwater Maximum Concentration Limits (MCLs) might be exceeded if no action was taken for eleven and ten constituents of concern (COCs) at P-Reactor and R-Reactor, respectively. These exceedances could occur in as few as 200 to 500 years. Alternatively, the migration modeling demonstrated that the selected ISD actions reduced contaminant mobility which allowed for significant radioactive decay and resulted in fewer predicted exceedances of groundwater MCLs (five COCs for P-Reactor; eight COCs for R-Reactor). In response to the modeling results for P-Reactor and R-Reactor, effectiveness monitoring programs were developed to target contaminants, specific to each reactor, that may migrate to groundwater. Contaminant migration modeling also revealed that roof collapse was a large factor in the release of COCs to the environment, giving rise to roof improvements, and an inspection and vegetation control program to ensure roof stability over time. Contaminant migration modeling is also aiding in the closure planning for hardened facilities in F Area. Building 235-F housed the Actinide Billet Line, which produced Np-237 billets for irradiation in SRS reactors, and the Plutonium Fuel Form (PuFF) facility that produced Pu-238 heat sources for the space program. As a result of these missions, areas within Building 235-F contain considerable residual amounts of both Pu-238 and Np-237. Contaminant migration modeling of Building 235-F was originally performed in 2012 to identify the feasibility of ISD and the amount of radioactive material removal required to prevent the exceedance of groundwater MCLs. The original model indicated that a 60% reduction in the PuFF facility Pu-238 inventory could keep groundwater concentrations below standards, while Np-237 did not pose a threat to groundwater. However, updates to the model with an emphasis on source impact pathways revealed that, due to the orientation of the source areas relative to groundwater flow, no amount of reasonable removal of Pu-238 would keep groundwater concentrations below MCLs and that Np-237 could be a large contributor to localized MCL exceedances. With this insight, the refined 2019 model is being used to assist in the development of grouting plans specific to each facility source area, where bentonite may be utilized to slow the migration of Pu-238 and its daughter products from the PuFF facility and a reducing grout may decrease Np-237 transport by ensuring the nuclide remains in the less mobile +IV oxidation state. The beginning phases of contaminant migration modeling are underway for F-Canyon and associated facilities using lessons learned from the Reactors and Building 235-F. A contaminant migration pathway, similar to the pathway used for the reactor vessels, is being developed for the hot and warm canyons. The F-Canyon inventories are being spatially refined to identify specific and localized source zones, comparable to Building 235-F, that may impact groundwater. Contaminant migration modeling provides vital input within the ISD process, from facility investigation to post closure effectiveness monitoring, making it a valuable tool in the development of safe and effective end states for hardened nuclear facilities at SRS. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Analysis of simulation methodology for calculation of the heat of transport for vacancy thermodiffusion

Computation of the heat of transport Q{sub a}{sup *} in monatomic crystalline solids is investigated using the methodology first developed by Gillan [J. Phys. C: Solid State Phys. 11, 4469 (1978)] and further developed by Grout and coworkers [Philos. Mag. Lett. 74, 217 (1996)], referred to as the Grout-Gillan method. In the case of pair potentials, the hopping of a vacancy results in a heat wave that persists for up to 10 ps, consistent with previous studies. This leads to generally positive values for Q{sub a}{sup *} which can be quite large and are strongly dependent on the specific details of the pair potential. By contrast, when the interactions are described using the embedded atom model, there is no evidence of a heat wave, and Q{sub a}{sup *} is found to be negative. This demonstrates that the dynamics of vacancy hopping depends strongly on the details of the empirical potential. However, the results obtained here are in strong disagreement with experiment. Arguments are presented which demonstrate that there is a fundamental error made in the Grout-Gillan method due to the fact that the ensemble of states only includes successful atom hops and hence does not represent an equilibrium ensemble. This places the interpretation of the quantity computed in the Grout-Gillan method as the heat of transport in doubt. It is demonstrated that trajectories which do not yield hopping events are nevertheless relevant to computation of the heat of transport Q{sub a}{sup *}.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2012

Saltstone 4QCY11 TCLP Results

The Saltstone Production Facility (SPF) receives waste from Tank 50H for treatment. In the fourth quarter of the 2011 calendar year (4QCY11), Tank 50H accepted transfers of approximately 10 kgal from the Effluent Treatment Project (ETP), approximately 4 kgal from 211H, approximately 573 kgal from the Actinide Removal Process/Modular Caustic Side Solvent Extraction Unit (ARP/MCU) Decontaminated Salt Solution Hold Tank (DSS-HT), and approximately 5 kgal from other sources. The Saltstone Grout Sampling plan provides the South Carolina Department of Health and Environmental Control (SCDHEC) with the chemical and physical characterization strategy for the salt solution which is to be disposed of in the Z-Area Solid Waste Landfill (SWLF). During operation, samples were collected from Tank 50H and grout samples prepared to determine the non-hazardous nature of the grout to meet the requirements of the South Carolina Hazardous Waste Management Regulations (SCHWMR) R.61-79.261.24(b) and R.61-79.268.48(a). Savannah River National Laboratory (SRNL) was asked to prepare saltstone from samples of Tank 50H obtained Oct. 12, 2011 during 4QCY11 to determine the non-hazardous nature of the grout. The samples were cured and shipped to Babcock & Wilcox Technical Services Group-Radioisotope and Analytical Chemistry Laboratory (B&W TSG-RACL) to perform the Toxic Characteristic Leaching Procedure (TCLP) and subsequent extract analysis on saltstone samples for the analytes required for the quarterly analysis saltstone sample. In addition to the eight toxic metals - arsenic, barium, cadmium, chromium, mercury, lead, selenium and silver - analytes included the underlying hazardous constituents (UHC) antimony, beryllium, nickel, and thallium which could not be eliminated from analysis by process knowledge. B&W TSG-RACL provided subsamples to GEL Laboratories, LLC for analysis for the UHCs benzene, phenols and total and amenable cyanide.

36 MATERIALS SCIENCE↗
DOE OSTI2012

Impact Of Standing Water On Saltstone Placement II - Hydraulic Conductivity Data

The amount of water present during placement and subsequent curing of saltstone has the potential to impact several properties important for grout quality. An active drain water system can remove residual standing water and expose the surface of the placed saltstone to air. Oxidation of the saltstone may result in an increase in the leachability of redox sensitive elements. A dry surface can lead to cracking, causing an increase in hydraulic conductivity. An inactive drain water system can allow standing water that generates unnecessary hydrostatic head on the vault walls. Standing water that cannot be removed via the drain system will be available for potential incorporation into subsequent grout placements. The objective of this work is to study the impact of standing water on grout quality pertaining to disposal units. A series of saltstone mixes was prepared and cured at ambient temperature to evaluate the impact of standing water on saltstone placement. The samples were managed to control drying effects on leachability by either exposing or capping the samples. The water to premix ratio was varied to represent a range of processing conditions. Samples were analyzed for density, leachability, and hydraulic conductivity. Report SRNL-STI-2012-00546 was issued detailing the experimental procedure, results, and conclusions related to density and leachability. In the previous report, it was concluded that: density tends to increase toward the bottom of the samples. This effect is pronounced with excess bleed water; drying of the saltstone during curing leads to decreased Leachability Index (more leaching) for potassium, sodium, rhenium, nitrite, and nitrate; there is no noticeable effect on saltstone oxidation/leachability by changing the water to premix ratio (over the range studied), or by pouring into standing water (when tested up to 10 volume percent). The hydraulic conductivity data presented in this report show that samples cured exposed to the atmosphere had about three orders of magnitude higher hydraulic conductivity than any of the other samples. Considering these data, along with the results presented in the previous report, leads to the conclusion that small changes in water to premix ratio and the inclusion of up to 10 volume percent standing water should not be expected to have a detrimental effect on saltstone grout quality. The hydraulic conductivity results further demonstrate that curing in a moist environment is critical to maintaining saltstone quality.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Engineering scale demonstration of a prospective Cast Stone process

This report documents an engineering-scale demonstration with non-radioactive simulants that was performed at SRNL using the Scaled Continuous Processing Facility (SCPF) to fill an 8.5 ft container with simulated Cast Stone grout. The Cast Stone formulation was chosen from the previous screening tests. Legacy salt solution from previous Hanford salt waste testing was adjusted to correspond to the average composition generated from the Hanford Tank Waste Operation Simulator (HTWOS). The dry blend materials, ordinary portland cement (OPC), Class F fly ash, and ground granulated blast furnace slag (GGBFS or BFS), were obtained from Lafarge North America in Pasco, WA. Over three days, the SCPF was used to fill a 1600 gallon container, staged outside the facility, with simulated Cast Stone grout. The container, staged outside the building approximately 60 ft from the SCPF, was instrumented with x-, y-, and z-axis thermocouples to monitor curing temperature. The container was also fitted with two formed core sampling vials. For the operation, the targeted grout production rate was 1.5 gpm. This required a salt solution flow rate of approximately 1 gpm and a premix feed rate of approximately 580 lb/h. During the final day of operation, the dry feed rate was increased to evaluate the ability of the system to handle increased throughput. Although non-steady state operational periods created free surface liquids, no bleed water was observed either before or after operations. The final surface slope at a fill height of 39.5 inches was 1-1.5 inches across the 8.5 foot diameter container, highest at the final fill point and lowest diametrically opposed to the fill point. During processing, grout was collected in cylindrical containers from both the mixer discharge and the discharge into the container. These samples were stored in a humid environment either in a closed box proximal to the container or inside the laboratory. Additional samples collected at these sampling points were analyzed for rheological properties and density. Both the rheological properties (plastic viscosity and yield strength) and density were consistent with previous and later SCPF runs.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Hydropower Geotechnical Foundations: Current Practice and Innovation Opportunities for Low-Head Applications

Hydropower is a renewable energy resource that produces electricity from flowing water under pressure. Engineered hydropower structures, such as dams, are used to create a hydraulic head, enabling a turbinegenerator unit to convert pressurized flow into electricity. While hydropower has been a source of renewable energy since antiquity, new development in the United States has slowed in recent decades. Based on recent resource assessments, the largest opportunity to expand hydropower in the United States is from new stream-reach development (i.e., new hydropower development along stream-reaches that do not currently have hydroelectric facilities or other forms of infrastructure, such as dams). Roughly 75% of identified new stream-reach development potential is from low-head sites (less than 30 ft of head), which typically suffer from smaller power densities and higher normalized costs, given economies of scale. Hydropower developers and other stakeholders are thus interested in strategies to reduce initial capital costs while practicing sustainable development to maximize environmental compatibility with minimal disruption to natural aquatic life, sediment, and water flows. Historically, civil works have represented a significant cost driver for new hydropower development, with the foundation system representing a major cost component and source of uncertainty. The foundation system is the collection of engineered structural features (e.g. cutoff trenches, walls, grouting, anchors) constructed at or below the preconstruction ground surface that interfaces between the overlying structures (superstructures) and the bed material below (subsurface). Development of a hydropower foundation system must consider the various characteristics of the surrounding stream environment and subsurface while adhering to the engineering requirements of the superstructure that it supports. The care of water, excavation, and other construction activities are important features of foundation design and construction. The design and construction cost of the foundation system is largely dependent on the site geology and riverbed composition and is influenced by the level of geotechnical assessment required and conducted. Thus, a hydropower facility’s geotechnical foundation is often highly site-specific, with proper site selection and assessment being important to project success. The foundation system is designed to provide structural stability (of the foundation and dam), limit seepage, ensure public safety, and maintain functionality for the project life, during both construction and facility operations. Inadequate foundation or dam design can result in dam failure and the uncontrolled release of significant volumes of water, which could cause a high number of casualties and extensive property damage downstream of the failure. According to the Association of State Dam Safety Officials, approximately 30% of all historical dam failures in the United States are attributed to foundation or abutment defects, and another 20% are attributed to piping or seepage through the embankment, foundation, or abutment. To ameliorate these safety considerations, foundations often require massive amounts of construction material (e.g., grout, concrete, engineered dam fill) and long construction times. Foundation design also requires significant analysis prior to construction because the initial in-stream and abutment subsurface conditions are site-specific, and sufficient data for them often are lacking. Current practice requires on-site assessment, using expensive drilling and invasive and non-invasive investigation methods, to determine the expected cost of foundation material and treatment. Additionally, foundation construction often requires site dewatering (and other care of water activities), which involves constructing temporary diversion structures upstream and often downstream, called cofferdams, and water diversion systems that route water around the construction site. Cofferdams and water diversion systems can drastically increase construction costs and contribute to environmental disruption, including modification of flow patterns and benthic habitats. Given the technical, economic, and environmental challenges associated with hydropower foundations, opportunities exist to improve the current state of practice and to develop new and innovative solutions to Hydropower is a renewable energy resource that produces electricity from flowing water under pressure. Engineered hydropower structures, such as dams, are used to create a hydraulic head, enabling a turbinegenerator unit to convert pressurized flow into electricity. While hydropower has been a source of renewable energy since antiquity, new development in the United States has slowed in recent decades. Based on recent resource assessments, the largest opportunity to expand hydropower in the United States is from new stream-reach development (i.e., new hydropower development along stream-reaches that do not currently have hydroelectric facilities or other forms of infrastructure, such as dams). Roughly 75% of identified new stream-reach development potential is from low-head sites (less than 30 ft of head), which typically suffer from smaller power densities and higher normalized costs, given economies of scale. Hydropower developers and other stakeholders are thus interested in strategies to reduce initial capital costs while practicing sustainable development to maximize environmental compatibility with minimal disruption to natural aquatic life, sediment, and water flows. Historically, civil works have represented a significant cost driver for new hydropower development, with the foundation system representing a major cost component and source of uncertainty. The foundation system is the collection of engineered structural features (e.g. cutoff trenches, walls, grouting, anchors) constructed at or below the preconstruction ground surface that interfaces between the overlying structures (superstructures) and the bed material below (subsurface). Development of a hydropower foundation system must consider the various characteristics of the surrounding stream environment and subsurface while adhering to the engineering requirements of the superstructure that it supports. The care of water, excavation, and other construction activities are important features of foundation design and construction. The design and construction cost of the foundation system is largely dependent on the site geology and riverbed composition and is influenced by the level of geotechnical assessment required and conducted. Thus, a hydropower facility’s geotechnical foundation is often highly site-specific, with proper site selection and assessment being important to project success. The foundation system is designed to provide structural stability (of the foundation and dam), limit seepage, ensure public safety, and maintain functionality for the project life, during both construction and facility operations. Inadequate foundation or dam design can result in dam failure and the uncontrolled release of significant volumes of water, which could cause a high number of casualties and extensive property damage downstream of the failure. According to the Association of State Dam Safety Officials, approximately 30% of all historical dam failures in the United States are attributed to foundation or abutment defects, and another 20% are attributed to piping or seepage through the embankment, foundation, or abutment. To ameliorate these safety considerations, foundations often require massive amounts of construction material (e.g., grout, concrete, engineered dam fill) and long construction times. Foundation design also requires significant analysis prior to construction because the initial in-stream and abutment subsurface conditions are site-specific, and sufficient data for them often are lacking. Current practice requires on-site assessment, using expensive drilling and invasive and non-invasive investigation methods, to determine the expected cost of foundation material and treatment. Additionally, foundation construction often requires site dewatering (and other care of water activities), which involves constructing temporary diversion structures upstream and often downstream, called cofferdams, and water diversion systems that route water around the construction site. Cofferdams and water diversion systems can drastically increase construction costs and contribute to environmental disruption, including modification of flow patterns and benthic habitats. Given the technical, economic, and environmental challenges associated with hydropower foundations, opportunities exist to improve the current state of practice and to develop new and innovative solutions to challenges frequently encountered with traditional approaches. With this understanding, it is critically important to understand and document the current state of practice for hydropower geotechnical foundations, identify key challenges, and define opportunities for innovative solutions. To this end, this report documents the current state of practice across the three main phases of geotechnical foundation development: (1) geotechnical site assessment, (2) design, and (3) construction for hydropower systems. It also describes the major challenges with conventional approaches and identifies opportunities for innovation to reduce hydropower foundations costs, timelines, and risks. Key takeaways from this report include the following: Approximately 80% of available low-head sites are expected to have foundations on soil beds rather than rock beds, suggesting that rockfill and earthfill dams may be the most cost-effective conventional dam type for new projects.; Geotechnical and geologic investigation activities are time-consuming and expensive but are essential to define the parameters and criteria needed for foundation design.; Certain riverbed soil and bedrock types present significant technical challenges or require expensive foundation construction, which can prove financially prohibitive for low-head project development.; Modular hydropower design and prefabricated modular foundations represent a promising but unproven paradigm for new hydropower development. Design and construction approaches using optimized and highly repeatable, reliable components would benefit project cost, time, and risk but require additional research and development.; Temporary construction features for foundations, including cofferdams, water diversion, and water control systems, can prove costly and have inherent construction risk.; For economically viable development, hydropower geotechnical foundations should be limited to 4 to 15% of the project’s total initial capital costs. Many proposed projects have experienced cost overruns attributable to foundation difficulties or surprises during construction. These overruns may have been due to inadequate investigations, lack of adequate engineering effort to tailor the structures to site geology and topography, and/or contractual terms, among other considerations.; Challenges for hydropower foundations and opportunities for innovative technology solutions are identified in the following areas (consistent with the three main phases of foundation development): Geotechnical site assessment, Foundation design and materials, Construction methods and technology. Ultimately, this report aims to provide information about geotechnical foundations for low-head hydropower and to motivate transformative technologies to support hydropower growth.

13 HYDRO ENERGY↗
DOE OSTI2021

Corrosion of Steel During Long-Term Exposure to Evolving Cementitious Environments

SRS is proceeding with closure of the H-Area Tank Farm (HTF) and F-Area Tank Farm (FTF). Closure consists of removing the bulk waste, heel removal, and filling the tank with tailored grout formulations. The long-term performance of the concrete and steel materials of construction and the closure grout is an important consideration for Performance Assessment (PA) of closed tanks in FTF and HTF. Over PA timeframes of hundreds, thousands, to tens of thousands of years, the chemical and physical properties of the concrete and steel materials will slowly degrade due to environmental exposure and material aging. In the interim these materials will provide a barrier to the leaching of radionuclides into the soil. Savannah River Remediation (SRR) is updating relevant portions of these analysis inputs to support an imminent revision to the HTF PA and a future update to the FTF PA. This analysis provides an update to the previous PA inputs for steel corrosion. The Central Scenario for the PA analysis includes three postulated modeling cases: 1) Realistic Case, 2) Compliance Case, and 3) Pessimistic Case. This analysis also included a Fast Flow Path Case. This latter case considered the circumstance where the initial condition of the concrete was a completely degraded state and that grout shrinkage exposed the steel to the soil environment immediately. In effect, the steel was unprotected by the concrete and grout. The cases have various degrees of conservatism considered. Chemical, physical and tank configuration parameters were investigated to understand their effects on the predicted time to release of the contaminants. The assessment reviewed the initial tank and steel configuration, service life degradation of the steel, and potential corrosion mechanisms associated with degradation of the concrete materials.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS1985

Repair of the DSS-14 Pedestal Concrete

About three years after the Goldstone Deep Space Station antenna was dedicated, grout under the hydrostatic bearing runner was found to be interacting with the runner, causing rust to form between the runner and the sole plates upon which it rests. The rust formed unevenly and the runner could not be kept flat so in 1969 the grout was removed and replaced with a Portland cement and sand dry pack grout that was less likely to produce rust. In the years that followed, oil leaking from the runner assembly caused progressive deterioration of the drypack grout. In 1982 over one thousand hours of spacecraft tracking time were lost due to this deterioration. A plan was developed to rehabilitate the bearing. The plan called for raising the rotating structure free from the concrete pedestal and placing it on three pairs of external support columns. With the weight of the structure transferred to the columns, the pads and runner could be removed and the repair started. The very successful repair included the replacement of a significant portion of the antenna pedestal.

Mcclure, D.↗
DOE OSTI2017

Contaminant Leaching From Intact Saltstone Monoliths - 17517

At the Department of Energy's (DOE's) Savannah River Site (SRS) chemically reducing materials, such as blast furnace slag (BFS), are added to grout formulations mixed with low-level radioactive salt solution in order to enhance the attenuation of redox sensitive contaminants (e.g., technetium-99 (Tc-99)). The resulting cementitious material, known as saltstone, is deposited in a series of concrete vaults for permanent disposal at the Saltstone Disposal Facility (SDF). Under oxidizing conditions, Tc persists as an anion in the +7 oxidation state, i.e., pertechnetate (TcO{sub 4}{sup -}), with very limited retention generally associated with amphoteric soil oxides. However, Tc(VII) is subject to chemical reduction to the +4 oxidation state, which is less soluble and mobile in the environment. Thus, chemically reducing grouts provide both a physical (i.e., low saturated hydraulic conductivities (K{sub sat}) that limit H{sub 2}O turnover and O{sub 2} exposure) and chemical barrier (i.e., residual reductive capacity) to contaminant release. However, many of the previous experiments evaluating the ability of saltstone to chemically reduce and immobilize Tc have been conducted using ground saltstone materials as sorbents, a practice that is likely to reduce the moisture level in the material, enhance exposure to O{sub 2} and alter sorbent properties in an unpredictable manner. Therefore, the objective of the current study was to evaluate contaminant leaching from intact monoliths that better represent the initial physical and chemical state of saltstone within the SDF disposal units. For the current study, saltstone simulants were produced utilizing SRR prescribed formulations: initial batches were spiked with rhenium (Re) to serve as nonradioactive analog for Tc-99, and subsequent batches were spiked with Tc-99 for comparison. The relative concentrations of I, Re and Tc-99 in the salt-waste simulants were consistent with the average concentrations of I-129 and Tc-99 in the low-level feed waste. In addition, SRR retrieved intact monoliths from Saltstone Disposal Unit (SDU) Cell 2A in 2015 for testing. Contaminant mass transfer rates for Tc-99 and other contaminants from spiked saltstone simulants and actual SDF saltstone samples were assessed using EPA Method 1315, Mass Transfer Rates of Constituents in Monolithic or Compacted Granular Materials Using a Semi-Dynamic Tank Leaching Procedure. This method was recently adopted for evaluating contaminant leaching from intact monolithic materials. Given the importance of redox speciation in controlling the mobility of Tc-99, the EPA mass transfer tests for the spiked simulants were initially conducted using leachates in equilibrium with three different test atmospheres: (1) oxic, (2) anoxic, and (3) anoxic reducing. Nitrate (NO{sub 3}{sup -}) leaching was evaluated as a poorly retained waste constituent. The leaching behavior of cesium-137 (Cs-137), representing about 98% of the initial radioactivity of the saltstone, was also evaluated for the SDU-extracted samples. Results from Method 1315 were also compared to a novel Dynamic Leaching Method (DLM) in which a flexible-wall permeameter was used to achieve saturated leaching under an elevated hydraulic gradient in an effort to evaluate the persistence of reductive capacity and subsequent changes in contaminant partitioning within intact saltstone monoliths. The composition of the chemical leachates from both tests was thoroughly analyzed in an effort to identify potential solid phases controlling contaminant partitioning through geochemical modeling. For the EPA 1315 tests using spiked simulants, Re leaching rates (and other poorly sorbing contaminants like NO{sub 3}{sup -} and iodine), as indicated by high effective diffusivities (D{sub e}) and low leachability indices (i.e. LI = - log[D{sub e}]), were much higher than Tc-99, which was attributed to poor initial reduction and/or rapid oxidation of Re, indicating that Re is a poor chemical analog for evaluating Tc partitioning under reducing conditions. Technetium-99 leaching rates for the spiked saltstone samples also appeared to be sensitive to the estimated reduction capacity of the dry feed materials used in making grout, a characteristic that can vary between different BFS sources. Technetium-99 leaching rates for the intact saltstone samples in the EPA 1315 test were quite similar, with less variability between the three test samples. This may be indicative of the longer curing time before the SDF samples were tested (∼3 years). Somewhat surprisingly, the leaching rates for Cs-137 from the SDF samples were generally lower than the poorly sorbing contaminants, NO{sub 3}{sup -} and iodine. For the DLM tests, Re leaching from spiked saltstone was much higher than observed for Tc-99, with rates similar to that of non-reactive grout constituents, consistent with EPA 1315 leaching rates. (authors)

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DOE OSTI2016

Managing Low-Level Waste at Dounreay - 16269

Dounreay Site Restoration Limited (DSRL) is the site licence company responsible for the clean-up and demolition of Britain's former centre of fast reactor research and development. The Dounreay site requires facilities to dispose of up to 175,000 cubic metres of low level waste (LLW) that has been and will continue to be generated during the decommissioning and restoration of the site. In May 2014, construction of two new vaults for the disposal of LLW at Dounreay was completed, including the necessary roads, infrastructures and services, followed by an encapsulation plant in September 2014. The vaults are located immediately to the northeast of the Dounreay licensed site, on NDA-owned land. The facilities are for disposal of the solid LLW generated at Dounreay and the adjacent Vulcan site only. In 2004, following consultation with stakeholders and members of the public, Dounreay issued the 'Best Practicable Environmental Option (BPEO) Report' on the management of Dounreay LLW. This report assessed potential management options for LLW against a range of criteria, including technical, environmental, cost, health and safety issues. The study was conducted in line with best practice and involved consultation with a range of people, including the local community and organisations with an interest in radioactive waste management. The recommendations and conclusions from the study provided the basis for the 'Dounreay Solid LLW Overall Strategy', which was published in March 2005. A fundamental component of this strategy was the development of new below-ground disposal vaults for LLW at Dounreay. Siting the vaults on land at Dounreay avoids any need to transport the LLW away from Dounreay on public roads. Disposal at Dounreay therefore satisfies the proximity principle of managing the waste at source. Disposal at Dounreay is also consistent with UK Government Policy on LLW management, the Scottish Government's Radioactive Waste Management Policy and the Nuclear Decommissioning Authority's (NDA's) strategy on management of LLW from the UK nuclear industry. A planning application was initially lodged with the Highland Council in June 2006. The period for the determination of this application was extended to allow the Scottish Environment Protection Agency (SEPA) sufficient time to assess fully the supporting Environmental Safety Case (ESC). The planning application was re-activated in May 2008, and planning permission was granted by Highland Council in April 2009. The planning permission included 26 conditions on DSRL, covering a wide range of environmental issues such as construction noise/dust, traffic, visual impact and also a condition to develop a community fund aimed at socio-economic development of the area around the Dounreay site. After undertaking the necessary detailed design work, construction of the phase one vaults commenced in November 2011 by principal contractor, Graham Construction. Work was undertaken on both vaults simultaneously, excavating around 243,000 m{sup 3} of rock. When excavation was complete, steel reinforcement was placed and concrete poured to construct the floor slabs. This was followed by the construction of the reinforced walls and then the installation of the steel support structure and materials that form the roof. Mechanical and electrical installation then followed, in parallel with the construction of the vault apron, ramps and access roads. A pumping system and control building was also constructed to remove groundwater from around the vaults to allow the vaults to be kept dry while they are being filled with waste. A layer of excavated material, approximately 3 m thick, was placed across the existing ground surface between the vaults and the coast, to ensure the water table lies well below the new ground surface after closure of the vaults, and that any future trace contaminated groundwater flows out to the sea rather than to the soil zone. This layer is referred to in project documents as the 'enhanced geosphere'. After disposals are completed and the vaults are closed, this layer will be joined to the capping over the vaults so that the ensemble merges visually with the surrounding landscape. The original vegetation and soil mix will be returned as far as is practicable. An encapsulation plant has been constructed to fill the LLW containers with grout prior to disposal in the vault, similar to the process used at the national Low-Level Waste Repository facility (LLWR), near Drigg in Cumbria. The plant has the capacity to grout up to four containers per day. The grout fills voids in the container and helps provide chemical conditions in the vault that limit the release of radionuclides in the far future, after the vault has been penetrated by groundwater. On April 24 2015, active commissioning began at the Encapsulation Plant and LLW Vaults. Half-height ISO containers packed with low level waste are currently being filled with grout and disposed of in the LLW Vault. Also packaged Demolition Low Level Waste (DLLW) is being processed and disposed of in the Demolition Low Level Waste Vault. To date over 2500 packages of DLLW and 70 containers LLW have been disposed of. Completion of construction and the start of operating the Phase One vaults and encapsulation plant are fundamental to the successful and efficient decommissioning of the Dounreay site, upon which many millions of pounds of investment and hundreds of jobs rely.

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DOE OSTI2018

Characterization and Remediation of the Hanford 618-10 Burial Ground Steel-pipe Vertical Pipe Units - 18238

The 618-10 Burial Ground is approximately 485 feet by 470 feet and consisted of 12 trenches and 94 VPUs. It is located approximately 5 miles north of Richland, Washington in the 600 Area of the Hanford Site. The 618-10 Burial Ground received low- to high-activity waste (e.g., fission products and some plutonium-contaminated waste). Most of the waste originated at the 300 Area laboratory facilities. Lower activity radioactive material and shielded drums were disposed in trenches, while laboratory wastes with higher levels of radiation were disposed in Vertical Pipe Units (VPUs). The 618-10 Burial Ground contained approximately 94 VPUs (uncertain mix of first and second generation). There are different variants of VPUs. The first generation of VPUs was simply corrugated culverts or standard pipes, 14-in diameter and 10- to 15-ft long, buried vertically. The second generation of VPUs was 22-in diameter, 15-ft long waste receptacles constructed by welding five 55-gallon bottomless drums together end-to-end and burying them vertically, separated by approximately 10 feet on center in rows approximately 25 feet apart. A subset of the first generation VPUs consisted of the steel pipe VPUs represented by carbon-steel pipes, 10 to 24 in. in diameter with various wall thicknesses. This paper will present the remediation of the steel-pipe VPUs, which were the last of the VPUs remediated in this project. The remediation process developed for steel pipe VPUs included exposing a section of the VPU within an excavated trench, placing a mix box around the VPU, filling the mix box with grout, shearing the VPU and contents, and then removing the grout waste mixture into a waste box for disposal at the Hanford Environmental Restoration Disposal Facility (ERDF). Prior the start of remedial activities extensive mock-ups were performed on surrogate steel-pipe VPUs to ensure the effectiveness of the process before introducing the hazards of radioactive VPU waste. The critical steps in the process of size reducing steel-pipe VPUs and mixing the contents with grout was successfully demonstrated. This process was completed for each steel-pipe VPU in successive lifts until the entire set of steel-pipe VPUs were mixed in grout and removed from the excavation. The remediation of the steel-pipe VPUs proved be successful in that all the VPUs were removed and transported to ERDF without a release to the local environment. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Performance Assessment for Closure of the Calcined Solids Storage Facility at the INL Site - 19289

The Calcined Solids Storage Facility is composed of six discrete stainless steel 'bin sets' at the Idaho Nuclear Technology and Engineering Center at the INL Site. Approximately 4,440 m{sup 3} of the dry, granular calcine waste is stored in the facility. Closure of the facility will entail removal of the calcine waste followed by grouting. A 5-cm thick residual amount of calcine was assumed to remain at the bottom of the bins. Important radionuclides in the residual layer for the groundwater pathway were identified as the fission and activation products Tc-99, Se-79, I-129, Cs-135, H-3, and actinides Pu-239, and Pu-240. The stainless-steel bins prevent near-term release of grouted residual calcine waste to the environment. Future releases were conservatively based on the corrosion rate of the weld seams on the stainless-steel bins, which site-specific studies measured mean tunnel corrosion rates of 0.0132 mm/yr with a coefficient of variation of 0.333. Based on tunnel corrosion rates, mean bin hydrologic containment times were estimated to be between 238 years and 1,191 years, depending on the thickness of the bin. Any breach in the weld seam was assumed to result in water flow and radionuclide leaching through the bin. The structural integrity of the bin was assumed to be controlled by the general stainless-steel corrosion rate, which was much slower (3.5 x 10{sup -5} mm/yr), resulting in structural integrity times between 24,000 years to over 100,000 years. A mixing-cell model was used to compute water flow and transport through the hydrologically failed bins and the vadose zone. Credit was taken for the geochemical properties of the grout used to fill the bins at closure, which significantly slows the releases from the bins for key radionuclides like Tc-99. A semi-analytical model for three-dimensional dissolved-phase transport in an aquifer of constant unidirectional flow, infinite length and width, and finite thickness was used to compute concentrations in the aquifer across the model domain for a time varying source. Cumulative impacts from all bin sets were calculated by summation across the bin set sources that were superimposed on a uniform Cartesian grid. Effective doses were calculated for direct groundwater ingestion from wells located 100-m downgradient from each of the bin sets and ingestion of meat, milk, and vegetables irrigated with water derived from the well. Intruder doses were based on a well drilled through the grouted waste and cuttings spread around a home site after a 500-year institutional control period. Peak doses during the 1,000-year compliance period were less than 5% of the 250 μSv/yr (25 mrem/yr) performance objective and intruder doses were generally less 5% of the of the 5000 μSv/yr (500 mrem) and 1000 μSv/yr (100 mrem) performance objective for acute and chronic intruder exposure, respectively. Groundwater concentrations were all less than the maximum contaminant limits. In all cases, doses were highest for the oldest bin set because, compared to the other bin sets, it had the highest radionuclide concentrations in the calcine waste. The performance assessment results indicate that performance objectives can be met leaving considerably more than 5-cm of residual calcine waste in the bin sets. (authors)

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DOE OSTI2025

Foam and Unreacted Material Reduction in Magnesium Oxysulfate for SRPPF Waste Solidification

A magnesium oxysulfate (MOS) grout formulation has been proposed to solidify the liquid effluent from the Savannah River Plutonium Processing Facility (SRPPF) Aqueous Recovery System (ARS). This formulation uses a combination of light-burnt MgO, anhydrous MgSO 4 , and dead-burnt MgO resulting in a grout with good mixability and acceptable density. The setting time for the grout is within the operational limits of the SRPPF facility and the leachate pH is ~9.4, which is within the assumed pH range of the brine from the Waste Isolation Pilot Plant (WIPP). When first tested at the 1-gallon small-scale and 55-gallon full-scale, there was a significant foam layer that raised concerns of a nonhomogeneous final form. A nonhomogeneous mixture could have variable density throughout the waste form and therefore allow for unequal shielding. To reduce the potential for unequal shielding in the final form, a series of foam reduction tests were carried out and a final formulation of 28 wt% light burnt MgO, 10 wt% MgSO 4 , and 62 wt% dead burnt MgO was used to remove the foam in under two hours while still maintaining a mix that has an acceptable density, mix time, peak temperature, and set time.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Reduction of Porosity in Cementitious Waste Forms to Reduce Effective Diffusivity

Although current practices meet regulations for grout waste forms, development of sequestration materials and techniques for contaminants of concern could expand the use of grout waste forms and reduce risk to established processes from future compositional changes. There are contaminants in waste streams that are not permanently and completely sequestered in current grout waste forms under certain conditions. These contaminants include methyl mercury, 129 I, 99 Tc, and nitrate. Each of these species have minor fractions that are not permanently and completely isolated from the environment and therefore potentially contribute to long-term environmental impacts. An approach of interest includes the technique of reducing porosity to result in reduced leachability of all contaminants simultaneously. The current Savannah River Site (SRS) saltstone formulation was used to evaluate the relationship between leachability and porosity reduced waste forms by altering the water-to-binder ratio. Results indicated decreasing the water-to-binder directly reduced porosity and resulted in the waste form’s ability to more effectively immobilize nitrate. This work was funded by the Department of Energy Office of Environmental Management (DOE-EM) Technology Operations Office (TOO) Project # HQ221818.

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