Engineering topics

Stefanko, D. B.

Publications and source records attributed to Stefanko, D. B..

6 records

DOE OSTI2013

Grout Long Radius Flow Testing to Support Saltstone Disposal Unit 6 Design - 13352

The Saltstone Facility, located within the Savannah River Site (SRS) near Aiken, South Carolina, consists of two facility segments: The Saltstone Production Facility (SPF) and the Saltstone Disposal Facility (SDF). The SPF receives decontaminated legacy low level sodium salt waste solution that is a byproduct of prior nuclear material processing. The salt solution is mixed with cementitious materials to form a grout slurry known as 'Saltstone'. The grout is pumped to the SDF where it is placed in a Saltstone Disposal Unit (SDU) to solidify. SDU 6 is referred to as a 'mega vault' and is currently in the design stage. The conceptual design for SDU 6 is a single cell, cylindrical geometry approximately 114.3 meters in diameter by 13.1 meter high and is larger than previous cylindrical SDU designs, 45.7 meters in diameter by 7.01 meters high (30 million gallons versus 2.9 million gallons of capacity). Saltstone slurry will be pumped into the new waste disposal unit through roof openings at a projected flow rate of about 34.1 cubic meters per hour. Nine roof openings are included in the design to discharge material into the SDU with an estimated grout pour radius of 22.9 to 24.4 meters and initial drop height of 13.1 meters. The conceptual design for the new SDU does not include partitions to limit the pour radius of the grout slurry during placement other than introducing material from different pour points. This paper addresses two technical issues associated with the larger diameter of SDU 6; Saltstone flow distance in a tank 114.3 meters in diameter and quality of the grout. A long-radius flow test scaled to match the velocity of an advancing grout front was designed to address these technology gaps. The emphasis of the test was to quantify the flow distance and to collect samples to evaluate cured properties including compressive strength, porosity, density, and saturated hydraulic conductivity. Two clean cap surrogate mixes (Saltstone premix plus water) were designed to simulate slurry with the reference Saltstone rheology and a Saltstone with extra water from the process flushing operation. Long-radius flow tests were run using approximately 4.6 cubic meters of each of these mixes. In both tests the pump rate was 0.063 liters/second (1 gpm). A higher pump rate, 0.19 liters/second (3 gpm), was used in a third long-radius flow test. The angle of repose of the grout wedges increased as a function of time in all three tests. The final angles of repose were measured at 3.0 deg., 2.4 deg., and 0.72 deg.. The pump rate had the largest effect on the radial flow distance and slope of the grout surface. The slope on the pour placed at 0.19 liters/second (3 gpm) was most representative of the slope on the grout currently being pumped into SDU 2 which is estimated to be 0.7 deg. to 0.9 deg. The final grout heights at 1/3 of a meter from the discharge point were 115, 105, and 38 cm. Entrapped air (≥ 0.25 cm bubbles) was also observed in all of the mixes. The entrapped air appeared to be released from the flows within about 3.1 meters (10 feet) of the discharge point. The bleed water was clear but had a thin layer of floating particulates. The bleed water should be retrievable by a drain water collection system in SDU 6 assuming the system does not get clogged. Layering was observed and was attributed to intervals when the hopper was being cleaned. Heat from the hydration reactions was noticeable to the touch. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2013

Salt-stone Oxidation Study: Leaching Method - 13092

Cementitious waste forms can be designed to chemically stabilize selected contaminants, such as Tc{sup +7} and Cr{sup +6}, by chemically reduction to lower valance states, Tc{sup +4} and Cr{sup +3}, respectively, and precipitation of these species in alkaline media as low solubility solid phases. Data for oxidation of this type of cementitious waste form cured under field conditions as a function of time is required for predicting the performance of the waste form and disposal facility. The rate of oxidation (oxidation front advancement) is an important parameter for predicting performance because the solubilities of some radionuclide contaminants, e.g., technetium, are a function of the oxidation state. A non-radioactive experiment was designed for quantifying the oxidation front advancement using chromium, as an approximate redox-sensitive surrogate (Cr{sup +6} / Cr{sup +3}) for technetium (Tc{sup +7} / Tc{sup +4}). Nonradioactive cementitious waste forms were prepared in the laboratory and cured under both laboratory and 'field conditions'. Laboratory conditions were ambient temperature and sealed sample containers. Field conditions were approximated by curing samples in open containers which were placed inside a plastic container stored outdoors at SRS. The container had a lid and was instrumented with temperature and humidity probes. Sub-samples as thin as 0.2 mm were taken as a function of distance from the exposed surface of the as-cast sample. The sub-samples were leached and the leachates were analyzed for chromium, nitrate, nitrite and sodium. Nitrate, nitrite, and sodium concentrations were used to provide baseline data because these species are not chemically retained in the waste form matrix to any significant extent and are not redox sensitive. 'Effective' oxidation fronts for Cr were measured for samples containing 1000, 500 and 20 mg/kg Cr added as soluble sodium chromate, Na{sub 2}CrO{sub 4}. For a sample cured for 129 days under field conditions, leachable Cr (assumed to be the oxidized form, i.e., Cr{sup +6} was detected in the top 15 to 20 mm of the sample spiked with 1000 mg/kg Cr. Below about 20 mm, the Cr concentrations in leachates were below the detection limit (< 0.010 mg/L) which indicates that the oxidation as the result of exposure to air was limited to the top 20 mm of the sample after exposure for 129 days and that the bulk of the waste form was not affected, i.e., the Cr was stabilized and insoluble. For samples cured in the laboratory, leachable Cr was detected in the top 8 mm of the Cr 1000 sample cured in the laboratory for 37 days. Between 8 and 14 mm, the concentration Cr in the leachate dropped by a factor of about 20 to just above the detection limit. These depth of penetration results indicate that the rate of advancement of the oxidation front for samples spiked with 1000 mg/kg Cr cured under 'field conditions' for 129 days is less than that for the sample cured in the laboratory for 37 days, i.e., 0.156 and 0.216 mm/day, respectively. Additional data are presented for samples spiked with 500 and 20 mg/kg Cr. In summary, cementitious waste forms are porous solids with a network of interconnected pores ranging in diameter from 10 E-10 m to greater than a few mm. The oxidation process is assumed to occur as the result of oxygen transport through the interconnected porosity which may be filled with air, aqueous salt solution, or both. Upon oxidation, the Cr becomes soluble and can be transported in solution through aqueous pore fluid or leachate. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2013

Tanks 18 And 19-F Structural Flowable Grout Fill Material Evaluation And Recommendations

Cementitious grout will be used to close Tanks 18-F and 19-F. The functions of the grout are to: 1) physically stabilize the final landfill by filling the empty volume in the tanks with a non-compressible material; 2) provide a barrier for inadvertent intrusion into the tank; 3) reduce contaminant mobility by a) limiting the hydraulic conductivity of the closed tank and b) reducing contact between the residual waste and infiltrating water; and 4) providing an alkaline, chemically reducing environment in the closed tank to control speciation and solubility of selected radionuclides. The objective of this work was to identify a single (all-in-one) grout to stabilize and isolate the residual radionuclides in the tank, provide structural stability of the closed tank and serve as an inadvertent intruder barrier. This work was requested by V. A. Chander, High Level Waste (HLW) Tank Engineering, in HLW-TTR-2011-008. The complete task scope is provided in the Task Technical and QA Plan, SRNL-RP-2011-00587 Revision 0. The specific objectives of this task were to: 1) Identify new admixtures and dosages for formulating a zero bleed flowable tank fill material selected by HLW Tank Closure Project personnel based on earlier tank fill studies performed in 2007. The chemical admixtures used for adjusting the flow properties needed to be updated because the original admixture products are no longer available. Also, the sources of cement and fly ash have changed, and Portland cements currently available contain up to 5 wt. % limestone (calcium carbonate). 2) Prepare and evaluate the placement, compressive strength, and thermal properties of the selected formulation with new admixture dosages. 3) Identify opportunities for improving the mix selected by HLW Closure Project personnel and prepare and evaluate two potentially improved zero bleed flowable fill design concepts; one based on the reactor fill grout and the other based on a shrinkage compensating flowable fill mix design. 4) Prepare samples for hydraulic property measurements for comparison to the values in the F- and H- Tank Farm Performance Assessments (PAs). 5) Identify a grout mix for the Tanks 18-F and 19-F Grout Procurement Specification. Results for two flowable zero bleed structural fill concepts containing 3/8 inch gravel (70070 Series and LP#8 Series) and a sand only mix (SO Series) are provided in this report. Tank Farm Engineering and SRNL Project Management selected the 70070 mix as the base case for inclusion in Revision 0 of the Tanks 18-F and 19-F grout procurement specification and requested admixture recommendations and property confirmation for this formulation. Lower cementitious paste mixes were formulated because the 70070 mix is over designed with respect to strength and generates more heat from hydration reactions than is desirable for mass pour application. Work was also initiated on a modification of the recommended mix which included shrinkage compensation to mitigate fast pathways caused by shrinkage cracking and poor physical bonding to the tank and ancillary equipment. Testing of this option was postponed to FY12. Mix, LP#8-16 is recommended for inclusion in the specification for furnishing and delivering tank closure grout for Tanks 18-F and 19-F. A shrinkage compensating variation of this mix, LP#16C, has not been fully developed and characterized at this time.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2013

Effect Of Oxidation On Chromium Leaching And Redox Capacity Of Slag-Containing Waste Forms

The rate of oxidation is important to the long-term performance of reducing salt waste forms because the solubility of some contaminants, e.g., technetium, is a function of oxidation state. TcO 4 - in the salt solution is reduced to Tc(IV) and has been shown to react with ingredients in the waste form to precipitate low solubility sulfide and/or oxide phases [Shuh, et al., 1994, Shuh, et al., 2000, Shuh, et al., 2003]. Upon exposure to oxygen, the compounds containing Tc(IV) oxidize to the pertechnetate ion, Tc(VII)O 4 - , which is very soluble. Consequently the rate of technetium oxidation front advancement into a monolith and the technetium leaching profile as a function of depth from an exposed surface are important to waste form performance and ground water concentration predictions. An approach for measuring contaminant oxidation rate (effective contaminant specific oxidation rate) based on leaching of select contaminants of concern is described in this report. In addition, the relationship between reduction capacity and contaminant oxidation is addressed. Chromate was used as a non-radioactive surrogate for pertechnetate in simulated waste form samples. Depth discrete subsamples were cut from material exposed to Savannah River Site (SRS) field cured conditions. The subsamples were prepared and analyzed for both reduction capacity and chromium leachability. Results from field-cured samples indicate that the depth at which leachable chromium was detected advanced further into the sample exposed for 302 days compared to the sample exposed to air for 118 days (at least 50 mm compared to at least 20 mm). Data for only two exposure time intervals is currently available. Data for additional exposure times are required to develop an equation for the oxidation front progression. Reduction capacity measurements (per the Angus-Glasser method, which is a measurement of the ability of a material to chemically reduce Ce(IV) to Ce(III) in solution) performed on depth discrete samples could not be correlated with the amount of chromium leached from the depth discrete subsamples or with the oxidation front inferred from soluble chromium (i.e., effective Cr oxidation front). Exposure to oxygen (air or oxygen dissolved in water) results in the release of chromium through oxidation of Cr(III) to highly soluble chromate, Cr(VI). Residual reduction capacity in the oxidized region of the test samples indicates that the remaining reduction capacity is not effective in re-reducing Cr(VI) in the presence of oxygen. Consequently, this method for determining reduction capacity may not be a good indicator of the effective contaminant oxidation rate in a relatively porous solid (40 to 60 volume percent porosity). The chromium extracted in depth discrete samples ranged from a maximum of about 5.8 % at about 5 mm (118 day exposure) to about 4 % at about 10 mm (302 day exposure). The use of reduction capacity as an indicator of long-term performance requires further investigation. The carbonation front was also estimated to have advanced to at least 28 mm in 302 days based on visual observation of gas evolution during acid addition during the reduction capacity measurements. Depth discrete sampling of materials exposed to realistic conditions in combination with short term leaching of crushed samples has potential for advancing the understanding of factors influencing performance and will support conceptual model development.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2013

Grout long radius flow testing to support Saltstone disposal Unit 5 design

The Saltstone Facility, located within the Savannah River Site (SRS) near Aiken, South Carolina, consists of two facility segments: The Saltstone Production Facility (SPF) and the Saltstone Disposal Facility (SDF). The SPF receives decontaminated legacy low level sodium salt waste solution that is a byproduct of prior nuclear material processing. The salt solution is mixed with cementitious materials to form a grout slurry known as “Saltstone”. The grout is pumped to the SDF where it is placed in a Saltstone Disposal Unit (SDU) to solidify. SDU 6 is referred to as a “mega vault” and is currently in the design stage. The conceptual design for SDU 6 is a single cell, cylindrical geometry approximately 114.3 meters in diameter by 13.1 meter high and is larger than previous cylindrical SDU designs, 45.7 meters in diameter by 7.01 meters high (30 million gallons versus 2.9 million gallons of capacity). Saltstone slurry will be pumped into the new waste disposal unit through roof openings at a projected flow rate of about 34.1 cubic meters per hour. Nine roof openings are included in the design to discharge material into the SDU with an estimated grout pour radius of 22.9 to 24.4 meters and initial drop height of 13.1 meters. The conceptual design for the new SDU does not include partitions to limit the pour radius of the grout slurry during placement other than introducing material from different pour points. This paper addresses two technical issues associated with the larger diameter of SDU 6; saltstone flow distance in a tank 114.3 meters in diameter and quality of the grout. A long-radius flow test scaled to match the velocity of an advancing grout front was designed to address these technology gaps. The emphasis of the test was to quantify the flow distance and to collect samples to evaluate cured properties including compressive strength, porosity, density, and saturated hydraulic conductivity. Two clean cap surrogate mixes (saltstone premix plus water) were designed to simulate slurry with the reference saltstone rheology and a saltstone with extra water from the process flushing operation. Long-radius flow tests were run using approximately 4.6 cubic meters of each of these mixes. In both tests the pump rate was 0.063 liters/second (1 gpm). A higher pump rate, 0.19 liters/second (3 gpm), was used in a third long-radius flow test. The angle of repose of the grout wedges increased as a function of time in all three tests. The final angles of repose were measured at 3.0º, 2.4º, and 0.72º. The pump rate had the largest effect on the radial flow distance and slope of the grout surface. The slope on the pour placed at 0.19 liters/second (3 gpm) was most representative of the slope on the grout currently being pumped into SDU 2 which is estimated to be 0.7º to 0.9º. The final grout heights at 1/3 of a meter from the discharge point were 115, 105, and 38 cm. Entrapped air (≥ 0.25 cm bubbles) was also observed in all of the mixes. The entrapped air appeared to be released from the flows within about 3.1 meters (10 feet) of the discharge point. The bleed water was clear but had a thin layer of floating particulates. The bleed water should be retrievable by a drain water collection system in SDU 6 assuming the system does not get clogged. Layering was observed and was attributed to intervals when the hopper was being cleaned. Heat from the hydration reactions was noticeable to the touch.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2013

Saltstone Oxidation Study: Leaching Method

Cementitious waste forms can be designed to chemically stabilize selected contaminants, such as Tc{sup +7} and Cr{sup +6}, by chemically reduction to lower valance states, Tc{sup +4} and Cr{sup +3}, respectively, and precipitation of these species in alkaline media as low solubility solid phases. Data for oxidation of this type of cementitious waste form cured under field conditions as a function of time is required for predicting the performance of the waste form and disposal facility. The rate of oxidation (oxidation front advancement) is an important parameter for predicting performance because the solubilities of some radionuclide contaminants, e.g., technetium, are a function of the oxidation state. A non-radioactive experiment was designed for quantifying the oxidation front advancement using chromium, as an approximate redox-sensitive surrogate (Cr{sup +6} / Cr{sup +3}) for technetium (Tc{sup +7} / Tc{sup +4}). Nonradioactive cementitious waste forms were prepared in the laboratory and cured under both laboratory and ?field conditions.? Laboratory conditions were ambient temperature and sealed sample containers. Field conditions were approximated by curing samples in open containers which were placed inside a plastic container stored outdoors at SRS. The container had a lid and was instrumented with temperature and humidity probes. Subsamples as thin as 0.2 mm were taken as a function of distance from the exposed surface of the as-cast sample. The subsamples were leached and the leachates were analyzed for chromium, nitrate, nitrite and sodium. Nitrate, nitrite, and sodium concentrations were used to provide baseline data because these species are not chemically retained in the waste form matrix to any significant extent and are not redox sensitive. ?Effective? oxidation fronts for Cr were measured for samples containing 1000, 500 and 20 mg/kg Cr added as soluble sodium chromate, Na{sub 2}CrO{sub 4}. For a sample cured for 129 days under field conditions, leachable Cr (assumed to be the oxidized form, i.e., Cr{sup +6} was detected in the top 15 to 20 mm of the sample spiked with 1000 mg/kg Cr. Below about 20 mm, the Cr concentrations in leachates were below the detection limit (< 0.010 mg/L) which indicates that the oxidation as the result of exposure to air was limited to the top 20 mm of the sample after exposure for 129 days and that the bulk of the waste form was not affected, i.e., the Cr was stabilized and insoluble. For samples cured in the laboratory, leachable Cr was detected in the top 8 mm of the Cr 1000 sample cured in the laboratory for 37 days. Between 8 and 14 mm, the concentration Cr in the leachate dropped by a factor of about 20 to just above the detection limit. These depth of penetration results indicate that the rate of advancement of the oxidation front for samples spiked with 1000 mg/kg Cr cured under ?field conditions? for 129 days is less than that for the sample cured in the laboratory for 37 days, i.e., 0.156 and 0.216 mm/day, respectively. Additional data are presented for samples spiked with 500 and 20 mg/kg Cr. In summary, cementitious waste forms are porous solids with a network of interconnected pores ranging in diameter from 10E-10 m to greater than a few mm. The oxidation process is assumed to occur as the result of oxygen transport through the interconnected porosity which may be filled with air, aqueous salt solution, or both. Upon oxidation, the Cr becomes soluble and can be transported in solution through aqueous pore fluid or leachate.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗