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At least 217 records · Page 12
All sourcesNASA NTRSDOE OSTI
NASA NTRS1983

Evaluation of DSS-14 pedestal-review of top surface repair procedures

Proposed repair procedures for the top surface of the pedestal supporting the hydrostatic bearing runner for the 64m Antenna are presented. These procedures included: (1) removal of existing grout and concrete to approximately 8 in. below original concrete surface using a presplitting technique with expansive cement followed by secondary breaking; (2) preparation of exposed concrete surface including an epoxy bonding agent; and (3) replacement of material removed with 8 in. of new concrete surface including an epoxy bonding agent; and (4) replacement of material removed with 8 in. of new concrete and 4 in. of new grout.

Oesterle, R. G.↗
DOE OSTI2014

Estimating Residual Solids Volume In Underground Storage Tanks

The Savannah River Site liquid waste system consists of multiple facilities to safely receive and store legacy radioactive waste, treat, and permanently dispose waste. The large underground storage tanks and associated equipment, known as the 'tank farms', include a complex interconnected transfer system which includes underground transfer pipelines and ancillary equipment to direct the flow of waste. The waste in the tanks is present in three forms: supernatant, sludge, and salt. The supernatant is a multi-component aqueous mixture, while sludge is a gel-like substance which consists of insoluble solids and entrapped supernatant. The waste from these tanks is retrieved and treated as sludge or salt. The high level (radioactive) fraction of the waste is vitrified into a glass waste form, while the low-level waste is immobilized in a cementitious grout waste form called saltstone. Once the waste is retrieved and processed, the tanks are closed via removing the bulk of the waste, chemical cleaning, heel removal, stabilizing remaining residuals with tailored grout formulations and severing/sealing external penetrations. The comprehensive liquid waste disposition system, currently managed by Savannah River Remediation, consists of 1) safe storage and retrieval of the waste as it is prepared for permanent disposition; (2) definition of the waste processing techniques utilized to separate the high-level waste fraction/low-level waste fraction; (3) disposition of LLW in saltstone; (4) disposition of the HLW in glass; and (5) closure state of the facilities, including tanks. This paper focuses on determining the effectiveness of waste removal campaigns through monitoring the volume of residual solids in the waste tanks. Volume estimates of the residual solids are performed by creating a map of the residual solids on the waste tank bottom using video and still digital images. The map is then used to calculate the volume of solids remaining in the waste tank. The ability to accurately determine a volume is a function of the quantity and quality of the waste tank images. Currently, mapping is performed remotely with closed circuit video cameras and still photograph cameras due to the hazardous environment. There are two methods that can be used to create a solids volume map. These methods are: liquid transfer mapping / post transfer mapping and final residual solids mapping. The task is performed during a transfer because the liquid level (which is a known value determined by a level measurement device) is used as a landmark to indicate solids accumulation heights. The post transfer method is primarily utilized after the majority of waste has been removed. This method relies on video and still digital images of the waste tank after the liquid transfer is complete to obtain the relative height of solids across a waste tank in relation to known and usable landmarks within the waste tank (cooling coils, column base plates, etc.). In order to accurately monitor solids over time across various cleaning campaigns, and provide a technical basis to support final waste tank closure, a consistent methodology for volume determination has been developed and implemented at SRS.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Window to the Hot Cell: The Challenges of Repairing a Specialized Nuclear Containment

Here, at the Idaho National Laboratory (INL) nuclear R&D site we come across a myriad of challenges and health hazards. Recently, a first time activity for the INL was to replace a hot cell window put in place circa 1977. This job evolution was complex in nature and involved several health and safety hazards. A hot cell is a containment designed to protect workers from radioactivity yet continue to allow work with remote handled tools and windows for visibility. The hot cell window weighed approximately 14,000 lbs. and was held in place by 1.6 cubic yards of high density grout. Removal of the window required abrasive grout removal techniques, radiological control efforts, specialized ventilation, and personal protective equipment utilization. In order to contain the large amounts of dust from the removal of concrete and potential contamination exposure a special containment tent and ventilation system was designed to protect workers in the surrounding areas from dust and debris. Once the window was removed it had to be further taken apart for shipment. While performing this activity, several lead sheets and lead wool was required to be removed. Overall it took 38 days to accomplish this task and had several organizations involved. There were no injuries or unexpected conditions that occurred during this job task. This was the first time this activity was performed at the INL and was completed in a successful manner. Further windows are set to be replaced in the near future.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2013

The Use of Subsurface Barriers to Support Treatment of Metals and Reduce the Flux of Tritium to Fourmile Branch at the Savannah River Site in South Carolina - 13358

The Savannah River Site (SRS) produced tritium, plutonium, and special nuclear materials for national defense, medicine, and the space programs. Acidic groundwater plumes containing metals, metallic radionuclides, non-metallic radionuclides and tritium sourced from the F and H Area Seepage Basins have impacted the surface water of Fourmile Branch on SRS. Tritium releases from Fourmile Branch have impacted the water quality within areas of the Savannah River adjacent to the SRS, and this circumstance has been an ongoing regulatory concern. The F and H Area Seepage Basins operated until 1988 for the disposition of deionized acidic waste water from the F and H Separations Facilities. The waste water contained dilute nitric acid and low concentrations of non-radioactive metals, and radionuclides, with the major isotopes being Cs-137, Sr-90, U-235, U-238, Pu-239, Tc-99, I-129, and tritium. The tritium concentration in the waste water was relatively elevated because there is not a practicable removal method in water. The acid content of the waste water during the operational period of the basins was equal to 12 billion liters of nitric acid. The seepage basins were closed in 1988 and backfilled and capped by 1991. The plumes associated with the F and H basins cover an area of nearly 2.4 square kilometers (600 acres) and discharge along ∼2,600 meters of Fourmile Branch. The acidic nature of the plumes and their overall discharge extent along the branch represent a large challenge with respect to reducing contaminant flux to Fourmile Branch. The introduction of nitric acid into the groundwater over a long time effectively reduced the retardation of metal migration from the basins to the groundwater and in the groundwater to Fourmile Branch, because most negatively charged surfaces on the aquifer materials were filled with hydrogen ion. Two large pump and treat systems were constructed in 1997 and operated until 2003 in an attempt to capture and control the releases to Fourmile Branch. The operating cost, including waste disposal, for the two systems was ∼$1.3 M/month. Both systems employed reinjection of tritiated water up gradient of the extraction, and produced large quantities of waste from non-tritium isotopes and metals removal prior to reinjection. Both systems were determined to be ineffective and potentially detrimental with respect to limiting the flux of contaminants to Fourmile Branch. After it became apparent that there was very little benefit to continued operation of the systems, and the staggering cost of operations was recognized by the SRS and regulators, a new remedy was developed. The new system uses vertical subsurface barriers to redirect groundwater flow to limit the transport of contaminants to the stream. The barriers were constructed of acid resistant grout using deep soil mixing techniques. The grout mixture used low swelling clay, fly ash, and sodium hydroxide to form a pozzolana material with low permeability and low strength. The SRS and regulators agreed to a series of remedial goals, with the first goal to reduce tritium flux to the stream by 70% and bring constituents other than tritium to groundwater protection standards. (authors)

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

Development of a Performance and Processing Property Acceptance Region for Cementitious Low-Level Waste Forms at Savannah River Site - 13174

The Saltstone Production and Disposal Facilities (SPF and SDF) at the Savannah River Site (SRS) have been treating decontaminated salt solution, a low-level aqueous waste stream (LLW) since facility commissioning in 1990. In 2012, the Saltstone Facilities implemented a new Performance Assessment (PA) that incorporates an alternate design for the disposal facility to ensure that the performance objectives of DOE Order 435.1 and the National Defense Authorization Act (NDAA) of Fiscal Year 2005 Section 3116 are met. The PA performs long term modeling of the waste form, disposal facility, and disposal site hydrogeology to determine the transport history of radionuclides disposed in the LLW. Saltstone has been successfully used to dispose of LLW in a grout waste form for 15 years. Numerous waste form property assumptions directly impact the fate and transport modeling performed in the PA. The extent of process variability and consequence on performance properties are critical to meeting the assumptions of the PA. The SPF has ensured performance property acceptability by way of implementing control strategies that ensure the process operates within the analyzed limits of variability, but efforts continue to improve the understanding of facility performance in relation to the PA analysis. A similar understanding of the impact of variability on processing parameters is important from the standpoint of the operability of the production facility. The fresh grout slurry properties (particularly slurry rheology and the rate of hydration and structure formation) of the waste form directly impact the pressure and flow rates that can be reliably processed. It is thus equally important to quantify the impact of variability on processing parameters to ensure that the design basis assumptions for the production facility are maintained. Savannah River Remediation (SRR) has been pursuing a process that will ultimately establish a property acceptance region (PAR) to incorporate elements important to both processability and long-term performance properties. This process involves characterization of both emplaced product samples from the disposal facility and laboratory-simulated samples to demonstrate the effectiveness of the lab simulation. With that basis confirmed, a comprehensive variability study using non-radioactive simulants will define the acceptable PAR, or 'operating window' for Saltstone production and disposal. This same process will be used in the future to evaluate new waste streams for disposal or changes to the existing process flowsheet. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Characterization and Remediation of the Hanford 618-10 Burial Ground Vertical Pipe Units - 17121

The 618-10 Burial Ground is approximately 148 m by 143 m (485 feet by 470 feet) and consists of 12 trenches and 94 Vertical Pipe Units (VPU's). It is located approximately 8.1 km (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). There are different variants of VPUs. Some VPUs are simply corrugated culverts or standard pipes, 36 cm (14 in.) diameter and 3.1 m to 4.6 m (10- to 15-ft) long, buried vertically. Some VPUs are carbon steel pipes 25 to 61 cm (10 to 24 in.) in diameter and up to 4.6 m (15 ft) in length. The final design of the VPUs is 56 cm (22-in) diameter, 4.4 m (14.4 ft) long waste receptacles constructed by welding five 208 L (55-gallon) bottomless drums together end-to-end and burying them vertically, separated by approximately 3.1 m (10 feet) on center in rows approximately 7.6 m (25 feet) apart. The 618-10 Burial Ground contains approximately 94 VPUs. Historical documents indicate that shielding materials (e.g., concrete or soil) may have been added to the VPUs to reduce dose rates over the openings, and that each VPU was backfilled with sand and capped with concrete at the end of its service life. Historical documents also indicate that the VPUs were covered with approximately 0.6 m (2-ft) of soil when they were closed, and an additional 0.6 m (2-ft) of topsoil was later added for surface stabilization. In general, the process for VPU remediation was as follows: - Installed four cone penetrometers (CPTs) around the exterior of each individual VPU and sent radiation detection probes down the interior of the CPTs to get initial radiation data in order to calculate the overall curie content and create maps showing the dose rate distribution throughout the VPU. - Removed the CPTs and installed a 122 cm (48-in) diameter, 1.3 cm (0.5-in) thick steel over-casing using a vibrating hammer suspended from a crane that encompassed the VPU with some surrounding soil. - Using a Bauer BG-30 drill rig with a specifically designed augur bit and a specially designed confinement system the auger process was used to stabilize and size reduce the contents. - Once the auger reached a depth of 8.5 m (28 feet) the BG-30 drill stem was separated and additional characterization was completed by sending a radiation probe down the center of the drill stem to gather radiological data. - Once auguring and characterization was completed the waste was removed using a specially designed extended clamshell retrieval bucket and placed into individual 2.7 m x 1.5 m x 1.2 m (9-ft x 5-ft x 4-ft) metal boxes. - After the waste was placed into the metal waste boxes a low strength grout was added and the grout and waste material was mixed using a large rake device attached to an excavator. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Test Methods for Cementitious Seal Materials for Borehole Disposal - 17352

The Nuclear and Energy Research Institute (IPEN-CNEN/SP) in Sao Paulo-Brazil, is considering borehole disposal of disused sealed radioactive sources (DSRS), which were used in smoke detectors, lightning arresters, radiotherapy, industrial gauges, irradiators, and other applications. The current IPEN concept of borehole is a modified BOSS system developed by the International Atomic Energy Agency (IAEA). It differs from the IAEA's design in that the disposal zone is below 300 m, and the DSRS are placed directly in the disposal packages constituted of stainless steel containers with 1.5 L capacity. A borehole drilled through sediment and granite, stabilized with stainless steel casing and cement grout is proposed for the conceptual model. Standard cement-material test methods and performance specific test methods, including accelerated aging methods, are needed for modeling the performance of the candidate cementitious grouts. A cross-disciplinary team from the USA Savannah River National laboratory is providing technical assistance to IPEN in this performance assessment. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Water Sealing With NOH{sub 2}O on the Fukushima Site - 17477

The Fukushima Daiichi nuclear site is still recovering from the natural disaster of March 11, 2011. Much of the recent recovery work has been focused on mitigating ground water intrusion into the radioactively contaminated basements of the reactor and turbine buildings of units 1 - 4. Other recovery efforts involve potentially sealing the vent lines in the primary containment vessels of units 1 - 3 to maintain a flooded vessel for core cooling and shielding. This paper details preliminary test results of a proprietary grout - NOH{sub 2}O{sup TM} - in sealing/plugging those vent lines to allow flooding of the primary containment vessel. Several partial and full-scale tests were conducted on primary containment vessel vent line mock-ups between March and October 2015 to ascertain the effectiveness of NOH{sub 2}O in sealing vent lines that have flowing water, which is problematic for conventional cement-based grouts that require relatively protracted set times. In contrast, NOH{sub 2}O can be used with a proprietary accelerant to provide instant set. The other key advantage of NOH{sub 2}O is its ability to penetrate even the smallest of cracks and porous soils, following and moving with the water flow and creating a continuous, impermeable barrier. It is also extremely flexible once set, conforming to its enclosure as upstream water head increases. Finally, the material's extremely high radiation resistance and longevity make it an ideal candidate for a permanent solution in high rad environments. (authors)

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

Evaluation and Impacts of Mercury in the SRS Liquid Waste System - 16121

The Savannah River Site (SRS) Liquid Waste System (LWS) stores radioactive waste in 43 underground tanks. The radionuclides in the waste are removed through a series of separation processes and the low-level fraction is immobilized in a grout waste form while the high level fraction is disposed in a glass waste form. Mercury originated from decades of canyon processing (used as a catalyst for dissolving the aluminum cladding of reactor fuel) and is present throughout the LWS (∼60 metric tons). Mercury has long been a consideration in the LWS, from both hazard and processing perspectives. Mercury is removed from the LWS in many ways including: 1) the Tank Farm evaporator condensates; 2) at the Defense Waste Processing Facility (DWPF) where it is steam-stripped and removed during the feed preparation processes; 3) it is removed at the Effluent Treatment Plant (ETP); and it is immobilized in the low-level, grout waste form. An integrated, system-wide evaluation and assessment of mercury behavior in the LWS is being performed. The recently completed Phase I evaluation focused on mercury inventory and speciation in the LWS; mercury holdup and chemical processing behavior; mercury impact Identification, including worker safety and equipment degradation; and mercury removal and disposal options. The mercury removal system at the DWPF has been unable to remove mercury successfully due to equipment and chemistry challenges since the 2008 timeframe. Phase I evaluations indicate that approximately 43 percent of the mercury in waste sent to the DWPF is now returning to the Tank Farm in the recycle stream and is concentrating to higher than expected levels (∼500 mg/l) around the 2H evaporator system. The chemical behavior of mercury is further complicated by the formation of higher than expected quantities of organomercury species, specifically methyl mercury, now present in the DWPF recycle evaporator system (2H Tank farm evaporator), which is used to volume-reduce the recycle stream. Phase II activities are building on the Phase I mercury activities, including additional sampling and characterization activities, a re-assessment of overall system knowledge, ranking and prioritizing critical gaps/information in mercury behavior across the flowsheet, assessing impacts of mercury removal and disposal options, and documenting an action plan for overall mercury management in the LWS. Phase II activities include two System Engineering Evaluations to 1) 'Reestablish Mercury removal Capability within DWPF', and 2) 'Determine the Best Alternative Mercury Removal Location with the LWS'. This paper provides the results of the Phase I and II mercury evaluations for the SRS LWS. (authors)

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

Degradation of ILW Drums Due to Internal Metallic Corrosion - 18036

Within the British nuclear inventory, metallic Intermediate Level Waste (ILW) is encapsulated in grout and stored in stainless steel drums. Based on recent inspections held in the relevant storage facilities (Sellafield, Cumbria), a proportion of these drums have exhibited considerable distortion, raising concerns about the long-term safety of the packages. Distortion may be caused by the generation of voluminous and potentially flammable substances resulting from metallic corrosion occurring within the drums' core. Potential oxygen influx in the event of steel fracture and consequently drum's failure could potentially cause release or even ignition of the encapsulated radioactive material. The mechanical degradation of the encapsulants (stainless steel and grout) induced from the progressive corrosion of the encapsulated metallic components (and the associated volume expansion) was investigated using both Finite Element Modelling (FEM) techniques as well as experimental trials. A parametric study using simplified 3D FE models was performed to determine their structural integrity under different scenarios. Synchrotron X-Ray tomography (XRT) was used to monitor the behaviour of miniaturized ILW drum-simulants, whilst the latter were subjected to controlled accelerated corrosion in situ. (authors)

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

Modelling of transport processes in concrete exposed to elevated temperatures – An alternative formulation for sorption isotherms

There is a significant need to understand, analyse and assess moisture transport in cementitious materials exposed to elevated temperatures in order to confidently predict the behaviour and ultimately the development of damage in safety critical applications such as nuclear reactor vessels, structures exposed to fire and well bore grouts. In view of this need a rigorous and robust formulation to describe water retention curves (sorption isotherms) as a function of temperature based on the evolution of physical parameters is presented. The model formulation is successfully validated against independent sets of experimental data up to temperatures of 80 °C. It is then further validated under isothermal drying conditions and then high temperature conditions through the numerical reproduction of laboratory experiments following implementation in a fully coupled hygro-thermo-mechanical finite element model. The new formulation is found to work well under a variety of conditions in a variety of cementitious material types.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Cobalt-60, Barium-133, Cesium-137, and Europium-152 migration from cementitious sources through sediment under field conditions

Safe and effective storage of radioactive waste is essential to protect human and environmental health. Due to the potential for accidental releases and the severity of the associated risks, it is imperative to further understand radionuclide transport should an accident occur. This study was the second set of measurements conducted in 2022 of an ongoing experiment that has analyzed the vadose zone migration of radionuclides from cementitious wasteforms at the Savannah River Site over the last ten years. The radionuclides introduced within the sources are prominent constituents of radioactive waste or analogs for other groups or series of radionuclides. Lysimeters were first analyzed in 2016 using a collimated high-purity germanium gamma-ray spectrometer to non-destructively measure the concentration of each radionuclide in the sediment column as a function of depth. Following these measurements, the lysimeters were redeployed for another 4 years. All radionuclides in all lysimeters were observed to transport further during the redeployment period; however, the extent of migration varied with the material used for introduction. Except for 137Cs, migration through the sediment control system increased with decreasing ionic potential (ionic charge/radius); migration order: 152Eu<137Cs<60Co<133Ba. Overall, the cementitious wasteforms were observed to decrease radionuclide migration extent relative to natural vadose zone conditions. In both cementitious wasteforms, the migration extent increased in the order 152Eu<133Ba<60Co<137Cs. However, less migration was measured when the radionuclides were incorporated into a reducing grout wasteform. The novelty of this paper is the demonstration of a technique capable of creating non-destructive measurements over decade time scales. Ultimately, this work provides insight into the long-term migration of alkali, alkali earth, divalent transition metal, and trivalent (e.g., lanthanide and actinide element) isotopes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2011

SALTSTONE VAULT CLASSIFICATION SAMPLES MODULAR CAUSTIC SIDE SOLVENT EXTRACTION UNIT/ACTINIDE REMOVAL PROCESS WASTE STREAM APRIL 2011

Savannah River National Laboratory (SRNL) was asked to prepare saltstone from samples of Tank 50H obtained by SRNL on April 5, 2011 (Tank 50H sampling occurred on April 4, 2011) during 2QCY11 to determine the non-hazardous nature of the grout and for additional vault classification analyses. 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. Additional inorganic species determined by B&W TSG-RACL include aluminum, boron, chloride, cobalt, copper, fluoride, iron, lithium, manganese, molybdenum, nitrate/nitrite as Nitrogen, strontium, sulfate, uranium, and zinc and the following radionuclides: gross alpha, gross beta/gamma, 3H, 60Co, 90Sr, 99Tc, 106Ru, 106Rh, 125Sb, 137Cs, 137mBa, 154Eu, 238Pu, 239/240Pu, 241Pu, 241Am, 242Cm, and 243/244Cm. B&W TSG-RACL provided subsamples to GEL Laboratories, LLC for analysis for the VOCs benzene, toluene, and 1-butanol. GEL also determines phenol (total) and the following radionuclides: 147Pm, 226Ra and 228Ra. Preparation of the 2QCY11 saltstone samples for the quarterly analysis and for vault classification purposes and the subsequent TCLP analyses of these samples showed that: (1) The saltstone waste form disposed of in the Saltstone Disposal Facility in 2QCY11 was not characteristically hazardous for toxicity. (2) The concentrations of the eight RCRA metals and UHCs identified as possible in the saltstone waste form were present at levels below the UTS. (3) Most of the inorganic species measured in the leachate do not exceed the MCL, SMCL or TW limits. (4) The inorganic waste species that exceeded the MCL by more than a factor of 10 were nitrate, nitrite and the sum of nitrate and nitrite. (5) Analyses met all quality assurance specifications of US EPA SW-846. (6) The organic species (benzene, toluene, 1-butanol, phenol) were either not detected or were less than reportable for the vault classification samples. (7) The gross alpha and radium isotopes could not be determined to the MCL because of the elevated background which raised the detection limits. (8) Most of the beta/gamma activity was from 137Cs and its daughter 137mBa. (9) The concentration of 137Cs and 90Sr were present in the leachate at concentrations 1/40th and 1/8th respectively than in the 2003 vault classification samples. The saltstone waste form placed in the Saltstone Disposal Facility in 2QCY11 met the SCHWMR R.61-79.261.24(b) RCRA metals requirements for a nonhazardous waste form. The TCLP leachate concentrations for nitrate, nitrite and the sum of nitrate and nitrite were greater than 10x the MCLs in SCDHEC Regulations R.61-107.19, Part I A, which confirms the Saltstone Disposal Facility classification as a Class 3 Landfill. The saltstone waste form placed in the Saltstone Disposal Facility in 2QCY11 met the R.61-79.268.48(a) non wastewater treatment standards.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

Data Package for Secondary Waste Form Down-Selection—Cast Stone

Available literature on Cast Stone and Saltstone was reviewed with an emphasis on determining how Cast Stone and related grout waste forms performed in relationship to various criteria that will be used to decide whether a specific type of waste form meets acceptance criteria for disposal in the Integrated Disposal Facility (IDF) at Hanford. After the critical review of the Cast Stone/Saltstone literature, we conclude that Cast Stone is a good candidate waste form for further consideration. Cast stone meets the target IDF acceptance criteria for compressive strength, no free liquids, TCLP leachate are below the UTS permissible concentrations and leach rates for Na and Tc-99 are suiteably low. The cost of starting ingredients and equipment necessary to generate Cast Stone waste forms with secondary waste streams are low and the Cast Stone dry blend formulation can be tailored to accommodate variations in liquid waste stream compositions. The database for Cast Stone short-term performance is quite extensive compared to the other three candidate waste solidification processes. The solidification of liquid wastes in Cast Stone is a mature process in comparison to the other three candidates. Successful production of Cast Stone or Saltstone has been demonstrated from lab-scale monoliths with volumes of cm3 through m3 sized blocks to 210-liter sized drums all the way to the large pours into vaults at Savannah River. To date over 9 million gallons of low activity liquid waste has been solidified and disposed in concrete vaults at Savannah River.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

Radionuclide Retention in Concrete Wasteforms

Assessing long-term performance of Category 3 waste cement grouts for radionuclide encasement requires knowledge of the radionuclide-cement interactions and mechanisms of retention (i.e., sorption or precipitation); the mechanism of contaminant release; the significance of contaminant release pathways; how wasteform performance is affected by the full range of environmental conditions within the disposal facility; the process of wasteform aging under conditions that are representative of processes occurring in response to changing environmental conditions within the disposal facility; the effect of wasteform aging on chemical, physical, and radiological properties; and the associated impact on contaminant release. This knowledge will enable accurate prediction of radionuclide fate when the wasteforms come in contact with groundwater. The information present in the report provides data that (1) measures the effect of concrete wasteform properties likely to influence radionuclide migration; and (2) quantifies the rate of carbonation of concrete materials in a simulated vadose zone repository.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

Diffusion and Leaching Behavior of Radionuclides in Category 3 Waste Encasement Concrete and Soil Fill Material – Summary Report

One of the methods being considered for safely disposing of Category 3 low-level radioactive wastes is to encase the waste in concrete. Such concrete encasement would contain and isolate the waste packages from the hydrologic environment and would act as an intrusion barrier. The current plan for waste isolation consists of stacking low-level waste packages on a trench floor, surrounding the stacks with reinforced steel, and encasing these packages in concrete. These concrete-encased waste stacks are expected to vary in size with maximum dimensions of 6.4 m long, 2.7 m wide, and 4 m high. The waste stacks are expected to have a surrounding minimum thickness of 15 cm of concrete encasement. These concrete-encased waste packages are expected to withstand environmental exposure (solar radiation, temperature variations, and precipitation) until an interim soil cover or permanent closure cover is installed, and to remain largely intact thereafter. Any failure of concrete encasement may result in water intrusion and consequent mobilization of radionuclides from the waste packages. The mobilized radionuclides may escape from the encased concrete by mass flow and/or diffusion and move into the surrounding subsurface environment. Therefore, it is necessary to assess the performance of the concrete encasement structure and the ability of the surrounding soil to retard radionuclide migration. The retardation factors for radionuclides contained in the waste packages can be determined from measurements of diffusion coefficients for these contaminants through concrete and fill material. Some of the mobilization scenarios include (1) potential leaching of waste form before permanent closure cover is installed; (2) after the cover installation, long-term diffusion of radionuclides from concrete waste form into surrounding fill material; (3) diffusion of radionuclides from contaminated soils into adjoining concrete encasement and clean fill material. Additionally, the rate of diffusion of radionuclides may be affected by the formation of structural cracks in concrete, the carbonation of the buried waste form, and any potential effect of metallic iron (in the form of rebars) on the mobility of radionuclides. The radionuclides iodine-129 ({sup 129}I), technetium-99 ({sup 99}Tc), and uranium-238 ({sup 238}U) are identified as long-term dose contributors in Category 3 waste (Mann et al. 2001; Wood et al. 1995). Because of their anionic nature in aqueous solutions, {sup 129}I, {sup 99}Tc, and carbonate-complexed {sup 238}U may readily leach into the subsurface environment (Serne et al. 1989, 1992a, b, 1993, and 1995). The leachability and/or diffusion of radionuclide species must be measured to assess the long-term performance of waste grouts when contacted with vadose-zone pore water or groundwater. Although significant research has been conducted on the design and performance of cementitious waste forms, the current protocol conducted to assess radionuclide stability within these waste forms has been limited to the Toxicity Characteristic Leaching Procedure, Method 1311 Federal Registry (EPA 1992) and ANSI/ANS-16.1 leach test (ANSI 1986). These tests evaluate the performance under water-saturated conditions and do not evaluate the performance of cementitious waste forms within the context of waste repositories which are located within water-deficient vadose zones. Moreover, these tests assess only the diffusion of radionuclides from concrete waste forms and neglect evaluating the mechanisms of retention, stability of the waste form, and formation of secondary phases during weathering, which may serve as long-term secondary hosts for immobilization of radionuclides. The results of recent investigations conducted under arid and semi-arid conditions (Al-Khayat et al. 2002; Garrabrants et al. 2002; Garrabrants and Kosson 2003; Garrabrants et al. 2004; Gervais et al. 2004; Sanchez et al. 2002; Sanchez et al. 2003) provide valuable information suggesting structural and chemical changes to concrete waste forms which may affect contaminant containment and waste form performance. However, continued research is necessitated by the need to understand: the mechanism of contaminant release; the significance of contaminant release pathways; how waste form performance is affected by the full range of environmental conditions within the disposal facility; the process of waste form aging under conditions that are representative of processes occurring in response to changing environmental conditions within the disposal facility; the effect of waste form aging on chemical, physical, and radiological properties, and the associated impact on contaminant release. Recent reviews conducted by the National Academies of Science recognized the efficacy of cementitious materials for waste isolation, but further noted the significant shortcomings in our current understanding and testing protocol for evaluating the performance of various formulations.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Analysis of the Tank 5f Final Characterization Samples-2011

The Savannah River National Laboratory (SRNL) was requested by SRR to provide sample preparation and analysis of the Tank 5F final characterization samples to determine the residual tank inventory prior to grouting. Two types of samples were collected and delivered to SRNL: floor samples across the tank and subsurface samples from mounds near risers 1 and 5 of Tank 5F. These samples were taken from Tank 5F between January and March 2011. These samples from individual locations in the tank (nine floor samples and six mound Tank 5F samples) were each homogenized and combined in a given proportion into 3 distinct composite samples to mimic the average composition in the entire tank. These Tank 5F composite samples were analyzed for radiological, chemical and elemental components. Additional measurements performed on the Tank 5F composite samples include bulk density and water leaching of the solids to account for water soluble species. With analyses for certain challenging radionuclides as the exception, all composite Tank 5F samples were analyzed and reported in triplicate. The target detection limits for isotopes analyzed were based on customer desired detection limits as specified in the technical task request documents. SRNL developed new methodologies to meet these target detection limits and provide data for the extensive suite of components. While many of the target detection limits were met for the species characterized for Tank 5F, as specified in the technical task request, some were not met. In a few cases, the relatively high levels of radioactive species of the same element or a chemically similar element precluded the ability to measure some isotopes to low levels. The Technical Task Request allows that while the analyses of these isotopes is needed, meeting the detection limits for these isotopes is a lower priority than meeting detection limits for the other specified isotopes. The isotopes whose detection limits were not met in all cases included the following: Al-26, Sn-126, Sb-126, Sb-126m, Eu-152 and Cf-249. SRNL, in conjunction with the plant customer, reviewed all these cases and determined that the impacts were negligible.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Analysis of the tank 6f final characterization samples-2012

The Savannah River National Laboratory (SRNL) was requested by Savannah River Remediation (SRR) to provide sample preparation and analysis of the Tank 6F final characterization samples to determine the residual tank inventory prior to grouting. Fourteen residual Tank 6F solid samples from three areas on the floor of the tank were collected and delivered to SRNL between May and August 2011.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗