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At least 1,171 records · Page 65
All sourcesNASA NTRSDOE OSTI
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 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 OSTI2017

Preparation, Performance and Mechanism of Tc and I Getters in Cementitious Waste Forms - 17124

Cementitious waste forms offer a low temperature, cost effective technology for the solidification of liquid wastes. At the Department of Energy Hanford Site, 99-Tc and 129-I are radionuclides contained within low activity waste (LAW) which contribute significantly to environmental impacts associated with clean-up of the site. Cast Stone, a cementitious waste form comprised of 47 wt% blast furnace slag, 45 wt% fly ash and 8 wt% ordinary Portland cement, has undergone investigation as a possible candidate technology for the solidification of LAW. In screening tests using EPA Method 1315 leach testing of various mixes of Cast Stone with simulated LAW, Tc observed effective diffusivities (Dobs) had an average of 5.3 x 10-11 cm{sup 2}/s and I Dobs had an average of 5.7 x 10-9 cm{sup 2}/s. From these measurements, an opportunity exists to improve retention of Tc and I by Cast Stone. One approach to accomplish this is through the incorporation of Tc and I getters to selectively sequester Tc and I in the waste stream and eventual waste form. Initial testing conducted at PNNL identified potassium metal sulfide (KMS-2) and Sn(II)-apatite (Sn-A) as the strongest performing Tc getters and silver exchanged zeolite (Ag-Z) as the top performing I getter. This research also demonstrated that the Tc and I getters, when added to solution simultaneously, can have deleterious interactions with one another and lower the Tc and I removal levels. However, these interactions can be overcome through sequential treatment of solution with the getters. Cast Stone samples were fabricated using a 6.5 M Na average LAW simulant spiked with Tc and I with the addition of various combinations of Tc and I getters. Through EPA Method 1315 and Method 1313 testing it was found that the KMS-2 Tc getter was most successful at lowering Tc Dobs. The I getters were not effective, as apparent interference from sulfide components of the waste form may have led to dissociation of the AgI formed by the Ag-Z getter in sequestering the I from solution. (authors)

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

Simulating field conditions for the curing of salt-stone - 15501

salt-stone is a cementitious, low-level waste (LLW) material processed and stored at the Savannah River Site (SRS). It consists of fly ash (FA), blast furnace slag (BFS), and Portland cement that are blended with LLW salt solution. The dry feed materials and the salt solution are mixed at the salt-stone Production Facility (SPF), and the fresh grout subsequently pumped into a salt-stone Disposal Unit (SDU) in the salt-stone Disposal Facility (SDF) where it cures to form a rigid, cementitious monolith. In order to demonstrate the long-term performance of salt-stone, non-radioactive simulant samples are processed and cured in a laboratory environment and subsequently tested with respect to key performance properties. Of particular relevance to assessing the long-term stabilization and immobilization of radionuclides in salt-stone is the permeability, or saturated hydraulic conductivity, which determines the rate of infiltration for oxygen-bearing ground waters that can affect redox sensitive contaminants, such as {sup 99}Tc. Environmental conditions in the SDU, in particular temperature and humidity, are believed to affect the cured properties of salt-stone. With respect to humidity it has been determined that the average vapor space humidity in an SDU is at least 95%, and in addition the assumption can be made that underlying grout is subjected to a moisture-saturated environment by the grout poured over it. With respect to temperature, the SDUs contain thermo-well arrays that provide detailed vertical, and to a lesser degree radial, temperature profiles. Temperature data from the SDUs have subsequently been utilized to better simulate curing temperatures for simulant salt-stone in the laboratory. However, no two thermal profiles are the same, varying with respect to the following parameters: initial temperature, rate of temperature increase, maximum temperature attained, time at maximum temperature, and cooling rate. These temperature variables are dependent on the depth of grout (due to insulating effects of underlying and overlying grout), frequency of grout addition to the SDU (with more frequent additions causing grout temperatures to rise due to the exothermic hydration reactions), and, to a lesser degree, the ambient temperature of the external environment. Due to the variability in each thermal profile recorded in an SDU, the question arises as to which profile (or profiles) should be used in the simulated laboratory environment to best mimic the curing of field emplaced salt-stone, such that there is confidence that the properties measured for laboratory-prepared samples are representative of the grout in the SDU. Since the current SDU design contains sixty-six thermo-wells it is obviously not practical to process simulant samples according to each profile. Rather the aim would be to choose a reduced number of profiles that adequately encompass the range of recorded profiles, or to choose a single profile that is considered most conservative with respect to the resulting properties of interest. This is typically thought to be the profile exhibiting the highest peak temperature since for cement-based materials higher curing temperatures (> 60 deg. C), while resulting in enhanced reaction kinetics and strength development for short curing durations (∼28 days), have been shown to adversely affect the physical properties of the material after extended cure times. In choosing representative profiles it is important to consider not only the maximum temperature, but also the influence of rate of temperature rise to the maximum temperature, and time at or close to the maximum temperature on the post-cured properties. The intent of this paper is to consider the varied curing temperature profiles to which the salt-stone is exposed in an SDU, and to analyze the currently available data for elevated temperature curing of salt-stone to determine the potential thermal effects on the saturated hydraulic conductivity of salt-stone emplaced in an SDU. This may subsequently facilitate a more informed approach to simulating the curing of salt-stone in a laboratory environment. (authors)

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

Engineering scale demonstration of a prospective cast Stone process - 15584

A cementitious waste form is one of the alternatives being evaluated for Supplemental Immobilization of Hanford Low Activity Waste (LAW). Washington River Protection Solutions (WRPS) is collecting data on cementitious or pozzolanic waste forms such as Cast Stone. The goal of this project for WRPS is to obtain data on the performance of the Cast Stone waste form for immobilizing LAW. As part of the data package being assembled, an engineering-scale demonstration with non-radioactive simulants was performed at the Savannah River National Laboratory (SRNL) using the Scaled Continuous Processing Facility (SCPF) to fill a container with simulated Cast Stone grout. Legacy salt solution from previous Hanford salt waste testing was trimmed to resemble the average composition generated from the Hanford Tank Waste Operation Simulator (HTWOS) used in the screening tests. 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 6,056 liter (1600 gallon) container with simulated Cast Stone grout. The tank was instrumented with x-, y-, and z-axis thermocouples to monitor curing temperature and two formed core sampling vials. The target production rate was 5.7 L/min (1.5 gpm). This required a salt solution flow rate of approximately 3.78 L/min (1 gpm) and a premix feed rate of approximately 263 kh/h (580 lb/h). The final surface slope at a fill height of 100 cm was 2.5-3.8 cm across the 260 cm (8.5 ft) diameter tank. During processing, grout was collected from both the mixer discharge and the discharge into the tank. These samples were stored in a humid environment either in a closed box proximal to the tank or inside the laboratory. (authors)

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

New Formulation of Cast Stone for {sup 99}Tc Immobilization in Sulfate Rich Secondary Waste Stream - 16098

A low-temperature, cementitious waste form called Cast Stone (or grout) has been selected for solidification of the Effluent Treatment Facility (ETF)-treated wastes at the Hanford Site. Because the waste stream to be solidified may be relatively high in sulfate, the current Cast Stone dry blend formulation with ordinary Portland cement (OPC), fly ash (FA), and blast furnace slag (BFS) (47% BFS, 45% FA, and 8% OPC) may not be adequate. Thus a new formulation, called secondary waste grout, was proposed and tested. The secondary waste grout uses hydrated lime [Ca(OH){sub 2}] to tie up the sulfate initially such that ettringite [Ca{sub 6}Al{sub 2}(SO{sub 4}){sub 3}(OH){sub 12}.26(H{sub 2}O)] forms rapidly instead of forming as the cementitious waste form cures. Three different liquid waste simulants (242-A evaporator, Environmental Restoration Disposal Facility [ERDF] leachates, and Hanford Tank Waste Treatment and Immobilization Plant [WTP] off-gas condensates) were prepared and solidified using this new dry blend formulation. Baseline grout dry mixes were 20 wt% hydrated lime (HL), 35 wt% OPC, and 45 wt% BFS, and two different water-to-dry mix (w/dm) ratios (0.5 and 0.6) of free water mass to mass of dry blend were tested. After curing for a minimum of 28 days, the grout solidified waste forms were characterized and used for Tc leachability testing using the EPA 1315 Method. X-ray diffraction analysis confirmed that ∼12-15% of ettringite was formed in early stage of 28-day curing for the grouts prepared with the high sulfate WTP simulant and HL addition. In addition, the new formulation of secondary waste grout prepared with HL (20%), OPC (35%), and BFS (45%) at either 0.5 or 0.6 water to dry mix ratio yields higher average Tc interval leachability index (LI) values (12.5 to 13.9) after 28 to 90 days of leaching using deionized water, which is higher than the desired {sup 99}Tc LI value of 12 for grout destined for burial in the Integrated Disposal Facility at the Hanford Site. (authors)

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

Roll Compaction of Powdered Radioactive Wastes - 18251

Most of low- and intermediate-level radioactive wastes must be solidified by a proper binding matrix and disposed at the centralized disposal site. So it is desirable to achieve a high loading of radwastes in a drum to reduce the disposal fee while meeting both the national regulations and the waste acceptance criteria of the disposal site. In general, the radioactive wastes (liquid, dried powder and sludge) were solidified by mixing with a proper binding matrix (cement, polymer, asphalt, etc.), but their solidified volume inevitably increases due to the limitation of incorporating of radwaste into a binding matrix. To solve this problem, some Korean NPPs operate both a CTS (Concentrate Treatment System, evaporation and granulation) and a polymer solidification system. The size distribution of granulated wastes produced by CTS is so broad (diameter; several μm ∼ several cm) that a polymer cannot impregnate the void space between particles due to their small size. However, if the powdered radwaste can be compacted into high density pellets, and the pellets can be loaded in a drum with maximum waste loading, then the binding matrix can fill the void between pellets, it might be more desirable to keep the size distribution of pellets constant to reduce the solidified waste volume. This compaction methodology can greatly reduce the disposal fee due to the reduction of the solidified waste volume, and can facilitate polymer solidification due to the constant size of the radwaste pellets. This methodology is also applicable to solidify powdered wastes such a depleted uranium used as a catalyst, ash and sludge. Loading radwaste compacted with bentonite pellets versus particles into a container serves, as an engineered or natural barrier in a deep geologic disposal system. Roll compaction is a typical agglomeration process for dry powders and depends on the device parameters and the material properties. This study, firstly, was focused on the design of roll compactor (briquetting machine), the shape and size of pellet, and secondly the loading method of pellets into a drum, their solidification in polymer, and evaluating the overall volume reduction. We used bentonite (particle size; 85 ∼ 100 μm) as a material to be compacted instead of radwaste powders (concentrates and depleted uranium wastes) without the addition of agglomeration agents. In the design of roll, we considered the diameter (200 mm) and width (41 mm) of roll, two types of pocket shapes (< 9 mm, 0.275 ml), and the surface of the pocket. Fig. 2 shows two pellet shapes, tetragonal and circular. The pressure of roll compactor, one of the main process variables, is > 320 kg f/cm{sup 2}, and the roll gap is 0.5 mm. The powders are compacted in a compaction zone between two counter-rotating rolls according to the complicated mechanisms. As a result of the state and the apparent density of the pellets obtained by changing the pressure from 0 to 286 kg f/cm{sup 2}, the pellets were unstable at pressure=0, but above 148 kg f/cm{sup 2} the pellets were more stable, and the apparent density of the pellets was 2.63 g/cm{sup 3} at 148 kg f/cm{sup 3}, and 2.68 g/cm{sup 3} at 286 kg f/cm{sup 2}. The roller speed (rpm) can determine the capacity (kg/h) of a compactor, and depends on both the compaction pressure and the feeding rate. By increasing the roll speed from 2 to 4 rpm, the apparent density of pellets was decreased gradually from 2.74 g/cm{sup 3} to 1.45 g/cm{sup 3}. We designed the special feeding system in order to feed powders uniformly to the roll gripping zone and to supply powders into the narrowest part of the roll gap. At a constant roll speed (3 rpm) and a constant compaction pressure (286 kg f/cm{sup 2}), the apparent density of the pellets was linearly proportional to the feeding rate from 2.18 to 2.74 g/cm{sup 3} by changing the feeding rate of powders from 8 to 12 rpm. In the polymer solidification of the obtained pellets, we used epoxy resin as a solidification agent, and showed their solidified waste forms in Fig. 5. The incorporation ratio (wt.%) of powders/polymer was about 71.7/28.3 in tetragonal pellets, 71.9/28.1 in circular pellets. We calculated the volume reduction in an aspect of powders itself and the solidified waste form. In powder case, through the comparison of the apparent density of the obtained pellets (2.74 g/cm{sup 3}) and raw powders (0.93 g/cm{sup 3}), the volume reduction was calculated as about 2.74/0.93 = 2.95 times. This means that powders can be compacted to 1/2.95 times. And in waste form case, the weight ratio of powder/polymer in a mixing method (workable) and a compaction method is about 0.42, 2.56, individually. Therefore the volume reduction was calculated as about 2.56/0.42 = 6.01 times. We found a compaction methodology for the volume reduction of powdered radioactive wastes, and their polymer solidification will be more effective to reduce the volume of radwastes to be disposed. After a characterization of the solidified waste form incorporated pellets (including sintered pellets), this methodology will be applied to treat depleted uranium wastes used as catalyst, and to solve the problem of NPPs' polymer solidification system, as mentioned above. (authors)

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

Hydraulic Properties of Grout Monoliths Containing Organic Ion Exchange Resin - 18369

The Waste Treatment and Immobilization Plant (WTP) at Hanford is being constructed to treat 56 million gallons of radioactive waste currently stored in underground tanks at the Hanford site. Operation of the WTP will generate several solid secondary waste (SSW) streams including spent spherical resorcinol formaldehyde cation exchange resin (IXr). The current plan for safe disposal of IXr is solidification in cementitious material (CM) typically comprised of cement, fly ash and blast furnace slag. Three CM formulations were tested to stabilize IXr for long-term land disposal. Two IXr loading rates, 0.1 and 0.3 v/v IXr:grout were used with the three CM formulations to prepare simulated waste form monoliths for hydraulic property testing. Following curing, the monoliths were removed from molds and tested to determine saturated hydraulic conductivity (K{sub sat}) and water retention properties. K{sub sat} was measured using a standard flexible permeameter method ASTM D-5084-10. A constant vapor pressure (CVP) method was used to determine van Genuchten parameters for desorption moisture retention curves for each of the SSW grout formulations. For the CVP method, vacuum saturated, intact wafers approximately 1.25 cm thick were placed above a saturated salt solution inside a sealed container for equilibration. Each saturated salt solution has an associated vapor pressure related to relative humidity in the container. Visual examination and three-dimensional x-ray micro computed tomography (μCT) indicated that the IXr was uniformly distributed in the cured waste form. Saturated hydraulic conductivity was within the range previously reported by Flach et al. in a study of the hydraulic properties of 98 pastes and grouts. The solidification of IXr in CM resulted in a waste form with water retention properties similar to cementitious pastes. (authors)

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

The Effect of Environmental Relative Humidity on Carbonation and Oxidation in a Cementitious Waste Form - 18448

Cementitious material has been developed and used for the nuclear waste disposal. Cast Stone, a type of cementitious waste form, is considered for solidification and immobilization of part of low activity radioactive waste (LAW) at DOE Hanford Site in Richland, WA. During its life cycle, the waste form, which is produced from Portland cement, fly ash, blast furnace slag and liquid waste, is subject to reaction with carbon dioxide and oxygen, both in gaseous and dissolved form. The reactions may alter the physical and chemical properties of the waste form and thus change its retention ability for the radionuclides. Therefore, understanding the rate and the impact of oxidation and carbonation in the material is important for projecting the long-term performance of Cast Stone. However, the rate of carbonation and oxidation in Cast Stone closely depends on the moisture content in the material, which varies largely with the environmental relative humidity (RH). In this study, experiments are carried out to study the effect of environmental relative humidity on the carbonation and oxidation rate as well as the leaching of the constituents from Cast Stone. Cast Stone samples doped with Cr as a reaction indicator are aged for 16 to 32 weeks under N{sub 2}, air, and 2% CO{sub 2} (in nitrogen) environments with various relative humidities (15%, 40%, 68%, 100%). After aging, samples were tested for the pH-dependent leaching (EPA Method 1313) and mass transfer rate under diffusion-controlled leaching (EPA Method 1315) of constituents. The progress of the carbonation was measured for each condition to determine the rate of carbonation front ingress. The results show that environmental relative humidity strongly influences carbonation and oxidation in Cast Stone; however, the impact is different for the rate of reaction and for the extent of reactions. The former refers to the rate at which the reaction front ingresses, while the latter refers to the fraction of the material reacted. For both carbonation and oxidation, the largest impact of the reactions at the exposure surface is observed at intermediate relative humidity (68%). However, for carbonation, the reaction front evolves more rapidly under drier conditions, with the highest rate observed at 15% RH. As a result, carbonation and oxidation have different degree of impact on the leaching behavior of Cast Stone samples aged under various relative humidities. (authors)

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

Solid Secondary Waste Immobilization in Cementitious Waste Forms at the Hanford Site - 19081

During operations of the Hanford Tank Waste Treatment and Immobilization Plant, solid secondary wastes (SSW) will be generated. These materials include HEPA filters, granular activated carbon (GAC), and silver mordenite (AgM). The planned disposal pathway for SSW is through solidification (blending) or encapsulation in a cementitious waste form and then disposal on site at the Hanford Site's Integrated Disposal Facility. Performance assessment evaluations of SSW forms have been performed; however, limited experimental data on simulated SSW is available as input. This paper will cover efforts to fill this gap in experimental data through testing of simulated SSW materials. Two grout formulations were used in the tests: 1) Hanford grout mix 5, a mix of 25 wt% ordinary Portland cement (OPC) and 75 wt% fly ash (FA); and 2) the Cast Stone dry blend of 47 wt% blast furnace slag, 45 wt% FA, and 8 wt% OPC. The sorption (Kd) of potential contaminants (iodide, iodate, technetium) to the grouts and SSW materials was investigated in simulated grout pore water. Waste forms containing GAC and AgM were prepared using the two grout formulations and characterized using scanning electron microscopy energy dispersive X-ray spectroscopy and X-ray diffraction. Acquiring this information will further support the candidacy of cementitious waste forms to be used for disposal of SSW at Hanford. (authors)

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

Mercury Speciation and Stabilization in Saltstone - 19103

Mercury is a contaminant of concern in Savannah River Site (SRS) high level waste (HLW). For five decades elemental mercury was used in the catalytic dissolution of aluminum cladding from the enriched-alloy uranium recovery process. Over 60,000 kg of mercury are now distributed in the SRS HLW tanks. The mercury speciation in the liquid waste streams resulting from conditioning HLW sludge for vitrification and in the decontaminated low activity sodium salt solution removed from the tanks was initially thought to be primarily ionic with a minor amount of elemental mercury. Recent sampling detected organic mercury, primarily methylmercury, HgCH{sub 3}{sup +} (MeHg{sup +}), and a trace amount of ethylmercury (EtHg) in these streams in addition to the ionic and elemental forms of mercury. The objective of this work was to identify the fate of the organic mercury in saltstone, a waste form containing portland cement, slag cement, and Class F fly ash. This information is needed to evaluate total mercury and organic mercury limits for the saltstone Waste Acceptance Criteria to assure regulatory compliance with mixed waste disposal requirements and to evaluate opportunities to enhance the waste form performance. Results to date indicate that the MeHg{sup +} reacts with sulfide in the cementitious reagents, slag in particular, to form nanoparticles of β-HgS, meta cinnabar, which is very insoluble and has a solubility product in water, pK{sub sp}, at 25 deg. C of 53. (authors)

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

Overview of Kinectrics' Experience in Characterization of Low and Intermediate Level Waste (L and ILW) - 19144

Kinectrics has over two decades of experience in physico-chemical and radiological characterization of L and ILW generated from the operation and refurbishment of nuclear reactors owned by Ontario Power Generation (OPG). This characterization is required to develop an inventory and also to support the safety case, as well as licensing, of OPG's waste management facilities including its planned Deep Geological Repository (DGR) for Low and Intermediate Level Waste (L and ILW). The typical waste matrices characterized include reactor artefacts, non-processable, compactable, incinerable, ash, assorted resin, filters, and sludge. The objective of this paper is to present an overview of the characterization work undertaken over the years and further elaborate on some recent activities pertaining to characterization of a select subset of L and ILW streams such as pressure tubes, feeder pipes, bulk non-processable and active liquid waste. (authors)

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

How do Early Age Geopolymers Respond to Gamma Irradiation Exposure? - 19515

Geopolymers have been proposed by the UK nuclear industry as a potential replacement for the currently accepted blend of Portland cement and supplementary cementitious materials, such as fly ash and blast furnace slag. This preferred grouting system is not ideal for the encapsulation of some problematic waste streams, such as oils and reactive metals. Metakaolin based geopolymers have been selected as the alternative grouting system, as this mitigates the supply risk which is currently associated with the supplementary cementitious materials described above. To allow these systems to be used as an alternative nuclear grouting system, the effect of irradiation on the 'neat' grout needs to be understood. The majority of the literature appears to present studies of the irradiation of aged grouts (3 - 7 days). This study will assess samples that have been irradiated at early age curing (20 hours). The effects of clay type and total dose will also be studied, and a comparison between the effects of rotary and flash calcined clays, irradiated to doses up to 1 MGy will be presented. The thermogravimetry data indicates that irradiation promotes the liberation of water and also the formation of carbonate species. The rotary calcined metakaolin liberates more water than the flash calcined metakaolin. Therefore, it is expected that the flash calcined system is more stable to irradiation. (authors)

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

In-Can Incineration and Vitrification Process for the Treatment of Transuranic Wastes - 20035

Technological alpha contaminated wastes are produced by the nuclear fuel cycle management industry, essentially from MOX fuel production units. These Plutonium Contaminated Materials (PCM) contain a mixture of organic, metal and mineral materials, contaminated with uranium, plutonium and americium. One treatment scenario under study is to incinerate the organic materials, melt the metallic fraction and incorporate the mineral fraction (including ashes from organics incineration) into a glass matrix without sorting or pretreatment (grinding for instance). A scale 1 pilot of this PIVIC process (Process for Incineration and Vitrification In Can) has been commissioned in 2018 at the CEA R and D vitrification facility in Marcoule site. A R and D program aiming to demonstrate the feasibility of this process is currently on going. This paper starts with a description of PIVIC process with explanation of the technologies involved. Then, main results of the first year of operation with the technological demonstration goal are presented. (authors)

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

Pyrolysis of Radioactive Spent Resins in the PRIME Installation - 20190

Montair and Belgoprocess developed the pilot PRIME (Pyrolysis Resins In Mobile Electric) Installation for treatment of spent resins in one single reactor without pre-treatment and without intermixing of other waste, allowing for easy traceability and isotopic characterization of the end product. Both low level waste (LLW, < 2 mSv/h) and medium level waste (MLW, > 2 mSv/h) can be treated in the installation. The PRIME pilot installation is designed to treat 2 batches of 100 liter of saturated ion exchange resins over a period of 24 hours. A dosing unit with screw feeder is used for the dosing of the saturated ion exchange resins to the pyrolysis reactor. This completely new developed innovative pyrolysis reactor contains a special design mechanical stirrer and can operate at a temperature of up to 600 deg. C under inert conditions. As a result, the organic molecules will decompose at an accelerated rate until only the mineralized material remains. The lower temperature and the absence of oxygen also prevent the breakdown of ion exchange resins into (powdery) ash, which is a considerable advantage during further handling. Gases that are released during drying and pyrolysis are treated in an also newly developed thermal oxidizer, which uses an innovative, electrically heated design instead of a traditional natural gas- or fuel oil secondary combustion chamber design. After passing through the thermal oxidizer, the gases are cooled down to saturation temperature and treated in the wet scrubbing system. An extraction fan keeps the system under negative pressure to prevent the possibility of any off-gas leakage out of the system. A series of tests were performed successfully with wet cationic resins, anionic resins and a mixture of both. The results of these tests are described in the presentation. For intellectual property protection, the system and process of pyrolysis of organic waste and the system of thermal oxidation of pyrolysis gases described in this paper are patent pending. (authors)

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

Investigation of Radiological Contamination at Parks in the Gateway National Recreation Area - 20350

In 2005, the National Park Service was notified that elevated radiation levels were detected at Great Kills Park during a baseline radiological mapping effort conducted by the Counter Terrorism Bureau of the New York City Police Department and the United States Department of Energy. Great Kills Park is in the Staten Island Unit of the Gateway National Recreation Area. A records search determined that approximately 280 acres of the park was established over an area in which the New York City Department of Sanitation placed approximately 15.1 million cubic yards of waste fill from 1944 to 1948. Since 2005, a series of response actions and investigations were conducted on behalf of the National Park Service to mitigate risk to human health and the environment and to determine the nature of radiological contamination at Great Kills Park. The findings support that radiological contamination is related to certain components of waste fill, including incinerator residue from both on and offsite incinerators, coal ash from the historical use of coal-fired boilers for heating, and incidental radiological artifacts in the waste fill due to their historical use and subsequent disposal. A majority of the radiological artifacts were determined to contain radium-226 and included radioluminescent markers, radium impregnated switch board buttons, radium paint containers, and brachytherapy devices. The National Park Service has several other Parks in the Gateway National Recreation Area that were established in areas with a similar history to Great Kills Park, and overlie waste fill historically placed during landfilling operations, including Spring Creek Park and Dead Horse Bay. Over the last year, the National Park Service has performed additional radiological investigations at Great Kills Park as part of the ongoing Remedial Investigation under the Comprehensive Environmental Response, Compensation, and Liability Act. In addition, preliminary radiological investigations were performed at Spring Creek Park and Dead Horse Bay Park in the Jamaica Bay Unit of Gateway National Recreation Area. These investigation activities identified gamma radiation levels above the ambient level, several of which were attributable to radiological artifacts similar in nature to those recovered at Great Kills Park. Ongoing investigations at Parks in the Gateway National Recreation Area support an association between radiological contamination and historical landfilling practices that were common prior to the regulation and control of radioactive materials. (authors)

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

Overview of Upcoming Process Improvement Efforts at the Saltstone Facilities at the Savannah River Site - 20474

The Saltstone Facilities at the Savannah River Site (SRS) process and dispose of low activity salt solution waste generated by other liquid waste facilities. Today in the Liquid Waste System (LWS), approximately 125 million liters of salt waste remain to be processed. The Saltstone Production Facility (SPF) receives the decontaminated salt waste, which is mixed with dry feeds consisting of cement, fly ash and slag to produce the saltstone. The mixed slurry is pumped to the Saltstone Disposal Units (SDUs) for safe disposal in the Saltstone Disposal Facility (SDF). To support continuous Salt Waste Processing Facility (SWPF) operations at annual processing volumes many times higher than present, significant modifications are required to ensure that the SPF has the capacity to support the throughput of salt waste. These changes include several infrastructure upgrades that are being worked through the Enhanced Low Activity Waste Disposal (ELAWD) project. In addition to infrastructure upgrades, Savannah River Remediation (SRR), the Savannah River Site's Liquid Waste Contractor for the U.S. Department of Energy, has identified a flowsheet improvement opportunity of implementing a cement-free saltstone formulation to eliminate cement from the current three component formulation. This two-component formulation would simplify the procurement, transport, off-loading, and storage of the premix materials, and increase the storage capacity for the individual components. This flowsheet, when implemented, will reduce operational risk and improve throughput necessary to achieve disposal rates necessary for SWPF operations. Research has been completed to confirm the feasibility of this formulation, with the new cement-free flowsheet expected to be implemented in the next few years. Improvements have also been identified downstream of the SPF to increase SDU utilization. Specifically, a concentrated effort has taken place to raise the fill height in SDU 6 to the full disposal height, 13 meters, to maximize the usage of each SDU in support of the liquid waste disposal mission. This improvement has been made through additional thermal modeling and improved process understanding to create a SDU flammability model and control strategy more representative of facility operations. This work will allow for the completion of construction and operational readiness for SDU 7 while filling SDU 6 to ensure there is no lapse in available space once SWPF starts up. A summary of these improvements as they relate to increased throughput will be provided along with associated status. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2010

Integration of Feedstock Assembly System and Cellulosic Ethanol Conversion Models to Analyze Bioenergy System Performance

Research barriers continue to exist in all phases of the emerging cellulosic ethanol biorefining industry. These barriers include the identification and development of a sustainable and abundant biomass feedstock, the assembly of viable assembly systems formatting the feedstock and moving it from the field (e.g., the forest) to the biorefinery, and improving conversion technologies. Each of these phases of cellulosic ethanol production are fundamentally connected, but computational tools used to support and inform analysis within each phase remain largely disparate. This paper discusses the integration of a feedstock assembly system modeling toolkit and an Aspen Plus® conversion process model. Many important biomass feedstock characteristics, such as composition, moisture, particle size and distribution, ash content, etc. are impacted and most effectively managed within the assembly system, but generally come at an economic cost. This integration of the assembly system and the conversion process modeling tools will facilitate a seamless investigation of the assembly system conversion process interface. Through the integrated framework, the user can design the assembly system for a particular biorefinery by specifying location, feedstock, equipment, and unit operation specifications. The assembly system modeling toolkit then provides economic valuation, and detailed biomass feedstock composition and formatting information. This data is seamlessly and dynamically used to run the Aspen Plus® conversion process model. The model can then be used to investigate the design of systems for cellulosic ethanol production from field to final product.

09 BIOMASS FUELS↗