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At least 1,531 records · Page 85
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Corrosion-Resistant Plug Materials for Geothermal Well Fluid Control

Conventional cements and plugs are challenged by corrosion in CO 2 -rich and extreme geothermal environments, due to the hostile chemistry and high temperatures. Thermite-based sealing and well intervention technologies are being applied in the oil and gas industry, combining the energy delivery capability of thermite materials with the sealing characteristics of low melt temperature alloys. The thermite reaction products (ceramics) and the sealing alloys used in these plugs both have very attractive corrosion properties, and their operating envelopes extend into geothermal conditions. Thermite plugs and platforms, without supplemental sealing materials, have been considered for nuclear waste isolation, carbon sequestration, and geothermal applications due to the geochemical stability of the ceramic product and its very high service temperature. This study addresses corrosion resistance of the thermite reaction products. A range of engineered thermite systems which yield thermite reaction products including pure aluminum oxide, feldspar, or aluminosilicate solid solutions (in addition to the iron produced in thermite reactions) was developed. These materials were evaluated for their strong acid resistance (pH 1), carbonate resistance (sodium carbonate) and thermal shock resistance (600 °C heating → cold water quenching repeated three times). Performance of different materials was evaluated based on the changes in mechanical properties, water-fillable porosity, phase changes under stress conditions. The aluminosilicate product exhibited very good corrosion resistance, both from material loss and strength perspectives, while the other products performed with varying degrees of stability. This paper presents the results of the thermite corrosion studies and describes the novel tools being deployed, and under development, to satisfy challenging barrier and intervention applications.

02 PETROLEUM↗
DOE OSTI2016

Phase Changes of Monosulfoaluminate in NaCl Aqueous Solution

Monosulfoaluminate (Ca 4 Al 2 (SO 4 )(OH) 12 ∙6H 2 O) plays an important role in anion binding in Portland cement by exchanging its original interlayer ions (SO 4 2- and OH - ) with chloride ions. In this study, scanning transmission X-ray microscope (STXM), X-ray absorption near edge structure (XANES) spectroscopy, and X-ray diffraction (XRD) were used to investigate the phase change of monosulfoaluminate due to its interaction with chloride ions. Pure monosulfoaluminate was synthesized and its powder samples were suspended in 0, 0.1, 1, 3, and 5 M NaCl solutions for seven days. At low chloride concentrations, a partial dissolution of monosulfoaluminate formed ettringite, while, with increasing chloride content, the dissolution process was suppressed. As the NaCl concentration increased, the dominant mechanism of the phase change became ion exchange, resulting in direct phase transformation from monosulfoaluminate to Kuzel’s salt or Friedel’s salt. The phase assemblages of the NaCl-reacted samples were explored using thermodynamic calculations and least-square linear combination (LC) fitting of measured XANES spectra. A comprehensive description of the phase change and its dominant mechanism are discussed.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2022

On the Response of a Herschel–Bulkley Fluid Due to a Moving Plate

In this paper, we study the boundary-layer flow of a Herschel–Bulkley fluid due to a moving plate; this problem has been experimentally investigated by others, where the fluid was assumed to be Carbopol, which has similar properties to cement. The computational fluid dynamics finite volume method from the open-source toolbox/library OpenFOAM is used on structured quad grids to solve the mass and the linear momentum conservation equations using the solver “overInterDyMFoam” customized with non-Newtonian viscosity libraries. The governing equations are solved numerically by using regularization methods in the context of the overset meshing technique. The results indicate that there is a good comparison between the experimental data and the simulations. The boundary layer thicknesses are predicted within the uncertainties of the measurements. The simulations indicate strong sensitivities to the rheological properties of the fluid.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Completion Design Considerations for a Horizontal Enhanced Geothermal System

The petroleum industry has had considerable success in recent decades in developing unconventional shale plays using horizontal drilling and multi-zonal isolation and stimulation techniques to fracture tight formations to enable the commercial production of oil and gas. Similar well completions could be used in Enhanced Geothermal Systems (EGS) to create multiple fractures from horizontal wells. This study assesses whether well completion techniques used in the unconventional shale industry to create multi-stage fractures can be applied to an enhanced geothermal system, with a focus on the completion of the EGS injection well. This study assumes an Enhanced Geothermal System (EGS) consisting of a central horizontal injection well flanked on each side by horizontal production wells, connected to the injection well by multiple fractures. The focus is on the design and completion of the horizontal well. For the purpose of developing design criteria, a reservoir temperature of 200 degrees C (392 degrees F) and an injection well flow rate of 87,000 barrels per day (160 kg/s), corresponding to production well flow rates of 43,500 barrels per day (80 kg/s) is assumed. The analysis found that 9-5/8 inches 53.5 pounds per foot (ppf) P110 casing string with premium connections meets all design criteria for the horizontal section of injection well. A P110 grade is fairly common and is often used in horizontal sections of shale development wells in petroleum operations. Next, several zonal isolation systems commonly used in the shale gas industry were evaluated. Three techniques were evaluated -- a 'plug and perf' design, a 'sand and perf' design, and a 'packer and port' design. A plug and perf system utilizes a cemented casing throughout the length of the injector wellbore. The sand and perf system is identical to the plug and perf system, but replaces packers with sand placed in the casing after stimulation to screen out the stimulated perforated zones and provide zonal isolation. The packer and port completion approach utilizes an open horizontal hole that zonally isolates areas through the use of external packers and a liner. A review of technologies used in these systems was performed to determine if commercially available equipment from the petroleum industry could be used at the temperatures, pressures, and sizes encountered in geothermal settings. The study found no major technical barriers to employing shale gas multi-zonal completion techniques in a horizontal well in a geothermal setting for EGS development. For all techniques considered, temperature limitations of equipment are a concern. Commercially available equipment designed to operate at the high temperatures encountered in geothermal systems are available, but is generally unproven for geothermal applications. Based on the study, further evaluation of adapting oil and gas completion techniques to EGS is warranted.

15 GEOTHERMAL ENERGY↗
DOE OSTI2012

Characterization of the liquid sodium spray generated by a pipework hole

Due to its advantageous thermodynamic characteristics at high temperature (550 deg. C), liquid sodium is the main candidate to be the cooling fluid for Generation TV nuclear reactors SFR (Sodium-cooled Fast Reactors). Now, sodium reacts very violently, both with the water and the oxygen of the air. Only few data were known about the liquid sodium behaviour when spread in the environment through micro defects. These are often present in a cooling circuit in welded or sealed joints and more rarely in the pipes. Micro defects, on the other hand, can be also generated in a cooling circuit because of the vibrations always present in a circuit into which a fluid runs. A new set-up, named LISOF, was built for testing high temperature liquid sodium when passing through micro defects and generating sprays or jets. Sprays and jets were generated by means of nozzles embedding sub milli-metric holes the diameter of which was: 0.2 mm, 0.4 mm, 0.5 mm. Tests were performed by pressurizing liquid sodium (550 deg. C) at: 3, 6 and 9 barg. Normal and high speed cinematography were used for the direct observation of the liquid sodium sprays while Phase Doppler Interferometry was used for the measurement of the droplets characteristics and velocity. Tests concerning the behaviour of the high temperature liquid sodium firing in air or in contact with the cement cover applied to a scaled down core catcher simulacrum were also performed. The paper presents the built set-up and the collected results. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Evaluation of irradiation effects on concrete structure

In assessing the soundness of irradiated concrete of nuclear power plants operated for more than 30 years, reference levels are employed: 1x10{sup 20} n/cm{sup 2} for fast neutrons and 2x10{sup 10} rad (2x10{sup 5} kGy) for gamma rays. Concrete structures are regarded as sound when the estimated irradiance levels after 60 years of operation are less than the reference levels. The reference levels were obtained from a paper by Hilsdorf. It was found, however, that the test conditions in which data were obtained by the researchers referred in that paper are very different from the irradiation and heat conditions usually found in a Light Water Reactor (LWR), and therefore aren't appropriate for assessing the soundness of irradiated concrete of an LWR. This paper investigates the interactions between radiation and concrete and presents the results of gamma ray irradiation tests on cement paste samples in order to provide a better understanding of the irradiation effects on concrete. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Improving the Estimates of Waste from the Recycling of Used Nuclear Fuel - 13410

Estimates are presented of wastes arising from the reprocessing of 50 GWD/tonne, 5 year and 50 year cooled used nuclear fuel (UNF) from Light Water Reactors (LWRs), using the 'NUEX' solvent extraction process. NUEX is a fourth generation aqueous based reprocessing system, comprising shearing and dissolution in nitric acid of the UNF, separation of uranium and mixed uranium-plutonium using solvent extraction in a development of the PUREX process using tri-n-butyl phosphate in a kerosene diluent, purification of the plutonium and uranium-plutonium products, and conversion of them to uranium trioxide and mixed uranium-plutonium dioxides respectively. These products are suitable for use as new LWR uranium oxide and mixed oxide fuel, respectively. Each unit process is described and the wastes that it produces are identified and quantified. Quantification of the process wastes was achieved by use of a detailed process model developed using the Aspen Custom Modeler suite of software and based on both first principles equilibrium and rate data, plus practical experience and data from the industrial scale Thermal Oxide Reprocessing Plant (THORP) at the Sellafield nuclear site in the United Kingdom. By feeding this model with the known concentrations of all species in the incoming UNF, the species and their concentrations in all product and waste streams were produced as the output. By using these data, along with a defined set of assumptions, including regulatory requirements, it was possible to calculate the waste forms, their radioactivities, volumes and quantities. Quantification of secondary wastes, such as plant maintenance, housekeeping and clean-up wastes, was achieved by reviewing actual operating experience from THORP during its hot operation from 1994 to the present time. This work was carried out under a contract from the United States Department of Energy (DOE) and, so as to enable DOE to make valid comparisons with other similar work, a number of assumptions were agreed. These include an assumed reprocessing capacity of 800 tonnes per year, the requirement to remove as waste forms the volatile fission products carbon-14, iodine-129, krypton-85, tritium and ruthenium-106, the restriction of discharge of any water from the facility unless it meets US Environmental Protection Agency drinking water standards, no intentional blending of wastes to lower their classification, and the requirement for the recovered uranium to be sufficiently free from fission products and neutron-absorbing species to allow it to be re-enriched and recycled as nuclear fuel. The results from this work showed that over 99.9% of the radioactivity in the UNF can be concentrated via reprocessing into a fission-product-containing vitrified product, bottles of compressed krypton storage and a cement grout containing the tritium, that together have a volume of only about one eighth the volume of the original UNF. The other waste forms have larger volumes than the original UNF but contain only the remaining 0.1% of the radioactivity. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Treatment of Bottled Liquid Waste During Remediation of the Hanford 618-10 Burial Ground - 13001

A problematic waste form encountered during remediation of the Hanford Site 618-10 burial ground consists of bottled aqueous waste potentially contaminated with regulated metals. The liquid waste requires stabilization prior to landfill disposal. Prior remediation activities at other Hanford burial grounds resulted in a standard process for sampling and analyzing liquid waste using manual methods. Due to the highly dispersible characteristics of alpha contamination, and the potential for shock sensitive chemicals, a different method for bottle processing was needed for the 618-10 burial ground. Discussions with the United States Department of Energy (DOE) and United States Environmental Protection Agency (EPA) led to development of a modified approach. The modified approach involves treatment of liquid waste in bottles, up to one gallon per bottle, in a tray or box within the excavation of the remediation site. Bottles are placed in the box, covered with soil and fixative, crushed, and mixed with a Portland cement grout. The potential hazards of the liquid waste preclude sampling prior to treatment. Post treatment verification sampling is performed to demonstrate compliance with land disposal restrictions and disposal facility acceptance criteria. (authors)

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

UP2 400 High Activity Oxide Legacy Waste Retrieval Project Scope and Progress-13048

The High Activity Oxide facility (HAO) reprocessed sheared and dissolved 4500 metric tons of light water reactor fuel the fuel of the emerging light water reactor spent fuel between 1976 and 1998. Over the period, approximately 2200 tons of process waste, composed primarily of sheared hulls, was produced and stored in a vast silo in the first place, and in canisters stored in pools in subsequent years. Upon shutdown of the facility, AREVA D and D Division in La Hague launched a thorough investigation and characterization of the silos and pools content, which then served as input data for the definition of a legacy waste retrieval and reconditioning program. Basic design was conducted between 2005 and 2007, and was followed by an optimization phase which lead to the definition of a final scenario and budget, 12% under the initial estimates. The scenario planned for the construction of a retrieval and reconditioning cell to be built on top of the storage silo. The retrieved waste would then be rinsed and sorted, so that hulls could subsequently be sent to La Hague high activity compacting facility, while resins and sludge would be cemented within the retrieval cell. Detailed design was conducted successfully from 2008 until 2011, while a thorough research and development program was conducted in order to qualify each stage of the retrieval and reconditioning process, and assist in the elaboration of the final waste package specification. This R and D program was defined and conducted as a response and mitigation of the major project risks identified during the basic design process. Procurement and site preparatory works were then launched in 2011. By the end of 2012, R and D is nearly completed, the retrieval and reconditioning process have been secured, the final waste package specification is being completed, the first equipment for the retrieval cell is being delivered on site, while preparation works are allowing to free up space above and around the silo, to allow for construction which is scheduled to being during the first semester of 2013. The elaboration of the final waste package is still undergoing and expected to be completed by then end of 2013, following some final elements of R and D required to demonstrate the full compatibility of the package with deep geological repository. The HAO legacy waste retrieval project is so far the largest such project entering operational phase on the site of La Hague. It is on schedule, under budget, and in conformity with the delivery requirements set by the French Safety Authority, as well as other stakeholders. This project paves the way for the successful completion of AREVA La Hague other legacy waste retrieval projects, which are currently being drafted or already in active R and D phase. (authors)

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

Project Plan, Status, and Lessons Learned for the LANL 3,706 m{sup 3} TRU Waste Campaign - 13085

The Los Alamos National Laboratory (LANL) is currently engaged in a campaign to disposition 3,706 m{sup 3} of transuranic (TRU) waste stored above grade at its Technical Area 54 (TA-54) Area G waste management facility before June 30, 2014. This campaign includes complete removal of all non-cemented above-grade TRU waste that was in storage on October 1, 2011, and is defined as 3,706 m{sup 3} of material. TRU waste containers were placed into storage up to 40 years ago, and most of the older containers must be remediated to address compliance issues before the waste can be characterized, certified as meeting the Waste Isolation Pilot Plant (WIPP) Waste Acceptance Criteria (WAC), and shipped for disposition. More than half of the remaining TRU waste volume stored above grade is contained within oversize boxes that contain waste items that must be repackaged or size reduced. Facilities and major types of equipment needed to remediate and characterize the TRU waste inventory include two additional oversize box processing lines that are being brought into service as Nuclear Hazard Category III facilities in fiscal year (FY) 2013. Multiple work shifts are scheduled for most remediation lines in FY 2013. An integrated risk-based project management schedule for all disposition activities has been developed that is based on a 'Solution Package' approach. Inventories of containers that have issues in common were compiled into about 15 waste categories and about 75 'Solution Packages' that identify all of the activities needed to disposition the inventory of TRU waste in storage. Scheduled activities include all precursor activities to begin remediation, remediation processing, characterization and certification to the WIPP WAC, and shipping of containers to WIPP. Other industrial processing practices that have been adopted to improve efficiency include staging of containers for remediation, characterization, and shipping; establishment of a transportation center; and load management practices for transportation payloads. Progress and accomplishments during FY 2012 are reviewed, and plans for FY 2013 are presented in some detail. Lessons learned on adoption of industrial processing practices are also discussed. (authors)

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

Evaluation of Resin Dissolution Using an Advanced Oxidation Process - 13241

The ion-exchange resin is widely used in nuclear reactors, in cooling water purification and removing radioactive elements. Because of the long periods of time inside the reactor system, the resin becomes radioactive. When the useful life of them is over, its re-utilization becomes inappropriate, and for this reason, the resin is considered radioactive waste. The most common method of treatment is the immobilization of spent ion exchange resin in cement in order to form a solid monolithic matrix, which reduces the radionuclides release into the environment. However, the characteristic of contraction and expansion of the resin limits its incorporation in 10%, resulting in high cost in its direct immobilization. Therefore, it is recommended the utilization of a pre-treatment, capable of reducing the volume and degrading the resin, which would increase the load capacity in the immobilization. This work aims to develop a method of degradation of ion spent resins from the nuclear research reactor of Nuclear and Energy Research Institute (IPEN/CNEN-SP), Brazil, using the Advanced Oxidative Process (AOP) with Fenton's reagent (hydrogen peroxide and ferrous sulphate as catalyst). The resin evaluated was a mixture of cationic (IR 120P) and anionic (IRA 410) resins. The reactions were conducted by varying the concentration of the catalyst (25, 50, 100 e 150 mM) and the volume of the hydrogen peroxide, at three different temperatures, 50, 60 and 70 deg. C. The time of reaction was three hours. Total organic carbon content was determined periodically in order to evaluate the degradation as a function of time. The concentration of 50 mM of catalyst was the most effective in degrading approximately 99%, using up to 330 mL of hydrogen peroxide. The most effective temperature was about 60 deg. C, because of the decomposition of hydrogen peroxide in higher temperatures. TOC content was influenced by the concentration of the catalyst, interfering in the beginning of the degradation process. It was possible to correlate it with the final amount of non-degraded resins. These results show that these conditions were favorable to destroy the resins, indicating to be the AOP an effective technique to reduce the volume of the waste. (authors)

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

Cementitious Barriers Partnership (CBP): Using the CBP Software Toolbox to Simulate Sulfate Attack and Carbonation of Concrete Structures - 13481

The Cementitious Barriers Partnership (CBP) Project is a multi-disciplinary, multi-institutional collaboration supported by the U.S. Department of Energy Office of Tank Waste Management. The CBP project has developed a set of integrated modeling tools and leaching test methods to help improve understanding and prediction of the long-term hydraulic and chemical performance of cementitious materials used in nuclear applications. State-of-the-art modeling tools, including LeachXS{sup TM}/ORCHESTRA and STADIUM{sup R}, were selected for their demonstrated abilities to simulate reactive transport and degradation in cementitious materials. The new U.S. Environmental Protection Agency leaching test methods based on the Leaching Environmental Assessment Framework (LEAF), now adopted as part of the SW-846 RCRA methods, have been used to help make the link between modeling and experiment. Although each of the CBP tools has demonstrated utility as a standalone product, coupling the models over relevant spatial and temporal solution domains can provide more accurate predictions of cementitious materials behavior over relevant periods of performance. The LeachXS{sup TM}/ORCHESTRA and STADIUM{sup R} models were first linked to the GoldSim Monte Carlo simulator to better and more easily characterize model uncertainties and as a means to coupling the models allowing linking to broader performance assessment evaluations that use CBP results for a source term. Two important degradation scenarios were selected for initial demonstration: sulfate ingress / attack and carbonation of cementitious materials. When sufficient sulfate is present in the pore solution external to a concrete barrier, sulfate can diffuse into the concrete, react with the concrete solid phases, and cause cracking that significantly changes the transport and structural properties of the concrete. The penetration of gaseous carbon dioxide within partially saturated concrete usually initiates a series of carbonation reactions with both dissolved ions and the hydrated cement paste. The carbonation process itself does not have a negative effect, per se, on the paste physical properties and can even result in reduced porosity and can help form a protective layer at the surface of concrete. However, carbonation has been shown to increase leaching of some constituents and can potentially have a detrimental effect on reinforced concrete structures by lowering pH to ca. 9 and de-passivating embedded steel (e.g. rebar) and accelerating corrosion, which are important processes related to high-level waste tank integrity and closure evaluations. The use of the CBP Software Toolbox to simulate these important degradation phenomena for both concrete vaults and high-level waste tanks are demonstrated in this paper. (authors)

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

Westinghouse Modular Grinding Process - Enhancement of Volume Reduction for Hot Resin Supercompaction - 13491

In nuclear power plants (NPP) ion exchange (IX) resins are used in several systems for water treatment. Spent resins can contain a significant amount of contaminates which makes treatment for disposal of spent resins mandatory. Several treatment processes are available such as direct immobilization with technologies like cementation, bitumisation, polymer solidification or usage of a high integrity container (HIC). These technologies usually come with a significant increase in final waste volume. The Hot Resin Supercompaction (HRSC) is a thermal treatment process which reduces the resin waste volume significantly. For a mixture of powdered and bead resins the HRSC process has demonstrated a volume reduction of up to 75 % [1]. For bead resins only the HRSC process is challenging because the bead resins compaction properties are unfavorable. The bead resin material does not form a solid block after compaction and shows a high spring back effect. The volume reduction of bead resins is not as good as for the mixture described in [1]. The compaction properties of bead resin waste can be significantly improved by grinding the beads to powder. The grinding also eliminates the need for a powder additive.Westinghouse has developed a modular grinding process to grind the bead resin to powder. The developed process requires no circulation of resins and enables a selective adjustment of particle size and distribution to achieve optimal results in the HRSC or in any other following process. A special grinding tool setup is use to minimize maintenance and radiation exposure to personnel. (authors)

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

The Polymers for Liquid Radioactive Waste Solidification: a Lost Chapter in the History of Engineering or a Step Forward? - 13529

Ideas on the application of polymers for the liquid radioactive waste immobilization go a way back, and the first studies in the area were published 30-40 years ago. One should admit that regardless of the fairly large number of publications appeared in the past years currently the interest in this work came down greatly. It was the successful assimilation and worldwide implementation of the LRW cementation technology caused a slump in the interest in polymers. But today it's safe to say that the situation slowly changes, particularly due to the market appearance of the high-tech polymers manufactured by Nochar Company, and unique properties of these polymers gradually raise the demand in various countries. The results of multiple experiments performed with the simulated solutions have passed the comprehensive tests with actual waste. The economic effect from the implementation of the new technology is defined by the volume reduction of waste coming onto the repository, by the decline in the cost of transportation and of the repository construction on account of cutting down the construction volume. Interesting results have been obtained during the search for the technical decisions that would allow using the polymer materials in the processing technology of the industrial toxic waste. One more promising area of the possible application of polymers should be pointed out. It is the application of polymer materials as the assets for the emergency damage control when the advantages of the polymers become obvious. (authors)

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

Mitigation of Hexavalent Chromium in Storm Water Resulting from Demolition of Large Concrete Structure at the East Tennessee Technology Park - 12286

American Recovery and Reinvestment Act (ARRA) funding was provided to supplement the environmental management program at several DOE sites, including the East Tennessee Technology Park (ETTP) in Oak Ridge, Tennessee. Demolition of the ETTP K-33 Building, the largest building to be demolished to date in Oak Ridge, was awarded to LSRS in FY-2010 under the ARRA program. The K-33 building was an 82 foot tall 2-story structure covering approximately 32 acres. Once this massive building was brought down to the ground, the debris was segregated and consolidated into piles of concrete rubble and steel across the remaining pad. The process of demolishing the building, tracking across concrete debris with heavy equipment, and stockpiling the concrete rubble caused it to become pulverized. During and after storm events, hexavalent chromium leached from the residual cement present in the large quantities of concrete. Storm water control measures were present to preclude migration of contaminants off-site, but these control measures were not designed to control hexavalent chromium dissolved in storm water from reaching nearby receiving water. The following was implemented to mitigate hexavalent chromium in storm water: - Steel wool was distributed around K-33 site catch basins and in water pools as an initial step in addressing hexavalent chromium. - Since the piles of concrete were too massive and unsafe to tarp, they were placed into windrows in an effort to reduce total surface area. - A Hach colorimetric field meter was acquired by the K-33 project to provide realtime results of hexavalent chromium in site surface water. - Three hexavalent chromium treatment systems were installed at three separate catch basins that receive integrated storm water flow from the K-33 site. Sodium bisulfite is being used as a reducing agent for the immobilization of hexavalent chromium while also assisting in lowering pH. Concentrations initially were 310 - 474 ppb of hexavalent chromium in surface water at the out-falls that discharge to nearby receiving water. After implementation of the actions described above, concentrations of hexavalent chromium have been effectively reduced to less than 25 ppb at the out-falls. The LSRS team completed demolition of K-33 five months ahead of schedule, and debris removal was completed three months ahead of schedule. A total of 164,000 tons of steel and concrete from the building demolition, accounting for 13,000 shipments, were disposed to the EMWMF. Because of the high toxicity of hexavalent chromium at low concentrations, hexavalent chromium had to be controlled at ppb levels. Hexavalent chromium contaminant concentrations were successfully reduced by over 90% in surface water discharged from the K-33 demolition site into nearby receiving water. Initial efforts of wind-rowing debris piles and obtaining real-time hexavalent chromium measurements to focus initiatives coupled with placement of steel wool in pools or catch basins had some effectiveness. More significant reductions were obtained as the debris piles were removed/disposed in EMWMF, and treatment of surface water with sodium bisulfite in integrated manholes occurred. (authors)

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

Grout Isolation and Stabilization of Structures and Materials within Nuclear Facilities at the U.S. Department of Energy, Hanford Site, Summary - 12309

Per regulatory agreement and facility closure design, U.S. Department of Energy Hanford Site nuclear fuel cycle structures and materials require in situ isolation in perpetuity and/or interim physicochemical stabilization as a part of final disposal or interim waste removal, respectively. To this end, grout materials are being used to encase facilities structures or are being incorporated within structures containing hazardous and radioactive contaminants. Facilities where grout materials have been recently used for isolation and stabilization include: (1) spent fuel separations, (2) uranium trioxide calcining, (3) reactor fuel storage basin, (4) reactor fuel cooling basin transport rail tanker cars and casks, (5) cold vacuum drying and reactor fuel load-out, and (6) plutonium fuel metal finishing. Grout components primarily include: (1) portland cement, (2) fly ash, (3) aggregate, and (4) chemical admixtures. Mix designs for these typically include aggregate and non aggregate slurries and bulk powders. Placement equipment includes: (1) concrete piston line pump or boom pump truck for grout slurry, (2) progressive cavity and shearing vortex pump systems, and (3) extendable boom fork lift for bulk powder dry grout mix. Grout slurries placed within the interior of facilities were typically conveyed utilizing large diameter slick line and the equivalent diameter flexible high pressure concrete conveyance hose. Other facilities requirements dictated use of much smaller diameter flexible grout conveyance hose. Placement required direct operator location within facilities structures in most cases, whereas due to radiological dose concerns, placement has also been completed remotely with significant standoff distances. Grout performance during placement and subsequent to placement often required unique design. For example, grout placed in fuel basin structures to serve as interim stabilization materials required sufficient bearing i.e., unconfined compressive strength, to sustain heavy equipment yet, low breakout force to permit efficient removal by track hoe bucket or equivalent construction equipment. Further, flow of slurries through small orifice geometries of moderate head pressures was another typical design requirement. Phase separation of less than 1 percent was a typical design requirement for slurries. On the order of 30,000 cubic meters of cementitious grout have recently been placed in the above noted U.S. Department of Energy Hanford Site facilities or structures. Each has presented a unique challenge in mix design, equipment, grout injection or placement, and ultimate facility or structure performance. Unconfined compressive and shear strength, flow, density, mass attenuation coefficient, phase separation, air content, wash-out, parameters and others, unique to each facility or structure, dictate the grout mix design for each. Each mix design was tested under laboratory and scaled field conditions as a precursor to field deployment. Further, after injection or placement of each grout formulation, the material was field inspected either by standard laboratory testing protocols, direct physical evaluation, or both. (authors)

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

Low and intermediate level radioactive waste processing in plasma reactor

Methods of low and intermediate level radioactive waste processing comprise: cementation, bituminization, curing in polymer matrices, combustion and pyrolysis. All these methods are limited in their application in the field of chemical, morphological, and aggregate composition of material to be processed. The thermal plasma method is one of the universal methods of RAW processing. The use of electric-arc plasma with mean temperatures 2000 - 8000 K can effectively carry out the destruction of organic compounds into atoms and ions with very high speeds and high degree of conversion. Destruction of complex substances without oxygen leads to a decrease of the volume of exhaust gases and dimension of gas cleaning system. This paper presents the plasma reactor for thermal processing of low and intermediate level radioactive waste of mixed morphology. The equipment realizes plasma-pyrolytic conversion of wastes and results in a conditioned product in a single stage. As a result, the volume of conditioned waste is significantly reduced (more than 10 times). Waste is converted into an environmentally friendly form that suits long-term storage. The leaching rate of macro-components from the vitrified compound is less than 1.10{sup -7} g/(cm{sup 2}.day). (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Setting an optimal procedure for transferring intermediate radioactive effluents: contribution of laboratory experiments - 15234

The INB35 nuclear facility at the CEA Saclay center is in charge of collecting and treating radioactive effluents generated from site's research activities. The first step of the process consists of an evaporation. The concentrate is an IL radioactive suspension with high salt concentration and suspended matter content. Until the 1990's, the resulting effluents were bitumen coated and since 2012 they have been cemented. During the 1990's, concentrates were stored in 7 tanks, named MA50X with X from 1 to 7. With today's changes to safety regulations, the effluents must now be transferred before tank dismantling. According to the site records, the compositions of the liquid wastes MA501, MA502, MA503, MA504, MA506 and MA507 are similar. From 2013 to 2014, effluents contained in MA502, MA503, MA504 tanks have been transferred. Several difficulties have been encountered: - The tanks were historically different, especially manhole tank access - Since the 1990's, effluent composition has changed due to water evaporation. The salt concentration has increased and insoluble elements have formed a deposit in the bottom of the tanks resulting in high composition heterogeneity - In some tanks, a wall breeze-blade exists dividing the tank into two zones, which complicates the interventions. Major challenges have been met and will be described in the presentation. - To obtain an effluent which is as homogeneous as possible, the duration of mixing has been optimized. - To pump the liquid, it has been necessary to treat the effluent with nitric acid. This step has been optimized thanks to experiments in laboratory and hydrodynamic modelling. - A mobile workshop has been designed in order to perform the effluent transfer. It is set up directly above the tanks and can be moved from one tank to another on rails. Laboratory experiments have been carried out to characterize the suspension and sludge in order to understand its chemical behavior. - The treatment of the effluent has been studied in the laboratory and is now mastered. - The analytical strategy has been optimized. The close collaboration between the analytical laboratory and the dismantling project was a real advantage. Thanks to this, the duration of the last transfer on MA 504 was reduced by half, compared to the first operations. Transfer operations will be described and the differences and similarities between laboratory and in-situ observations will be pointed out. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗