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A request was made by H-Canyon Process Engineering to assess the impact of blending dissolved, neutralized Spent Nuclear Fuel (SNF) with future sludge batches and their impact on downstream processing facilities. The purpose of this change is to accelerate the deinventory of SNF which is currently stored in the L-Area Disassembly Basin.
Savannah River National Laboratory (SRNL) completed a small-scale demonstration of the Defense Waste Processing Facility (DWPF) Chemical Process Cell (CPC) utilizing the nitric-glycolic acid (NGA) flowsheet to support Sludge Batch 10 (SB10) qualification. The demonstration utilized a Tank 51 slurry sample washed by SRNL (with added H-canyon material). The purpose of this document is to report the observed off-gas results from the demonstration. With the NGA flowsheet, DWPF has a CPC hydrogen generation limit of 2.4×10 -2 lb/h. The peak observed rate was nearly 90 times less than that limit during SRNL testing.
The Defense Waste Processing Facility (DWPF) is currently preparing to initiate processing of Sludge Batch 10 (SB10), which is comprised of material from Tanks 11H, 13H, 15H, and 26F, Alternate Feed Stock-2 and Sodium Reactor Experiment material from H-Canyon. Frit 473 (8B 2 O 3 -8Li 2 O-5Na 2 O-79SiO 2 , in weight percent) was recommended for sludge-only (SO) and coupled processing with the Salt Waste Processing Facility (SWPF) based on previous assessments of SB10 projections with the DWPF Product Composition Control System (PCCS) glass property models and their associated Measurement Acceptance Region (MAR) constraints. Due to the lower processing rate of Sludge Batch 9 (SB9), the heel in Tank 40 is anticipated to be greater than 40 inches at the projected start of SB10 processing to meet the upcoming Accelerated Basin Deinventory addition dates in Tank 51 for Sludge Batch 11. In December 2021 and February 2022, Savannah River Remediation System Planning provided updated SB10 Tank 40 blend projections based on heels of 74 inches, 84 inches, 94 inches and 103.4 inches. Savannah River Mission Completion (SRMC) subsequently pursued Wash Cycle Y to further reduce the total sulfur in the sludge batch and increase processing flexibility at DWPF. In August 2022, SRMC System Planning provided an updated SB10 Tank 40 blend projection with a 76 inch-heel representing Decant Y2 and 60 kilogallons of bearing water inleakage that is anticipated during the Tank 51 to Tank 40 transfer. The objectives of this task were to: • Determine the impact on the operating windows for SO and coupled processing • Determine whether any composition gaps exist between the already completed SB10 variability study and the reprojected SB10 glass composition region • Compare the SB10 reprojected glass composition region to the DWPF PCCS model development and validation ranges to ensure that compositional gaps do not exist between the data sets. This report documents the results of these evaluations. Calculation-based frit assessments were performed using the DWPF PCCS glass property models and their associated MAR constraints. Evaluated parameters for coupled processing included the following transfer volumes per Sludge Receipt and Adjustment Tank batch: 5700 gallons of Tank 40 sludge, 2400-4500 gallons of the SWPF monosodium titanate (MST) and sludge solids stream, and 15,000 gallons of strip effluent. Based on these MAR assessment results, Frit 473 remains viable for SB10 processing. A target waste loading (WL) of 36% is possible for SO operation and single strike (0.4 g/L MST) coupled processing up to 600 mg/L of SB9 insoluble sludge solids at a nominal transfer volume of ~2800 gallons. Increasing the single strike transfer volume to 4500 gallons may reduce the maximum WL below 40%. Operating windows are 12 percentage points for SO processing and 14-16 percentage points for coupled processing. Frit 625 allows for a target WL of 36% and is acceptable for use during the SB9 to SB10 transition to deplete remaining inventory as needed. Operating windows are 8 percentage points for SO processing and 12-14 percentage points for coupled processing. Predictive PCCS evaluations performed at DWPF will provide insight into batch-specific acceptability at desired WLs for compositions having expected oxide ratios during processing versus the extreme vertices (corner points) evaluated in this study. The reprojected SB10 glass composition region generally overlaps the previously evaluated SB10 variability study composition region. Thus, the minor composition shift of these updated SB10 projections indicates that no additional glasses are necessary to demonstrate acceptability relative to the chemical durability of the Environmental Assessment benchmark glass and predictability using the current PCCS models for durability. Based on a comparison of the PCCS model development and validation data to the reprojected SB10 glass composition region, the viscosity and liquidus temperature models will reliably predict SB10 compositions. No additional glasses are necessary to demonstrate predictability of these models.
The Defense Waste Processing Facility (DWPF) began processing Sludge Batch 6 (SB6), also referred to as Macrobatch 7 (MB7), in June 2010. SB6 is a blend of the heel of Tank 40 from Sludge Batch 5 (SB5), H-Canyon Np transfers and SB6 that was transferred to Tank 40 from Tank 51. SB6 was processed using Frit 418. During processing of each sludge batch, the DWPF is required to take at least one glass sample to meet the objectives of the Glass Product Control Program and to complete the necessary Production Records so that the final glass product may be disposed of at a Federal Repository. Four pour stream glass samples and two Melter Feed Tank (MFT) slurry samples were collected while processing SB6. The samples were transferred to the Savannah River National Laboratory (SRNL) where they were analyzed.
The Defense Waste Processing Facility (DWPF) is currently preparing to initiate processing of Sludge Batch 10 (SB10), which is comprised of material from Tanks 11H, 13H, 15H, 26F, 40H (heel only), and Alternate Feed Stock-2 and Sodium Reactor Experiment material from H-Canyon. Frit 473 was recommended for sludge-only (SO) and coupled processing with the Salt Waste Processing Facility (SWPF) based on previous assessments of SB10 projections with the DWPF Product Composition Control System (PCCS) glass property models and their associated Measurement Acceptance Region (MAR) constraints. Savannah River Mission Completion (SRMC) subsequently pursued Wash Cycle Y to further reduce the total sulfur in the sludge batch and increase processing flexibility at DWPF. In October 2022, SRMC System Planning provided an updated SB10 Tank 40H blend baseline projection based on the Tank 51 confirmation sample results. Due to the reduced Na concentration relative to previous SB10 projections, two additional projections based on the addition of 7,000 gallons of caustic were provided in November 2022.
The Defense Waste Processing Facility (DWPF) is currently preparing to initiate processing of Sludge Batch 10 (SB10), which is comprised of material from Tanks 11H, 13H, 15H, 26F, 40H (heel only), and Alternate Feed Stock-2 and Sodium Reactor Experiment material from H-Canyon. Frit 473 was recommended for sludge-only (SO) and coupled processing with the Salt Waste Processing Facility (SWPF) based on previous assessments of SB10 projections with the DWPF Product Composition Control System glass property models and their associated Measurement Acceptance Region constraints. Savannah River Mission Completion (SRMC) subsequently pursued Wash Cycle Y to further reduce the total sulfur in the sludge batch and increase processing flexibility at DWPF. In October 2022, SRMC System Planning provided an updated SB10 Tank 40H blend baseline projection based on the Tank 51 confirmation sample results. Due to the reduced Na concentration relative to previous SB10 projections, two additional projections based on the addition of 7,000 gallons of caustic were provided in November 2022.
The Accelerated Basin De-inventory (ABD) program increases processing of Spent Nuclear Fuel (SNF) in H-Canyon to accelerate the closure of L Basin. The dissolved SNF will be dispositioned to Sludge Batches that will be processing in the Defense Waste Processing Facility and converted to glass. The fissile glass loading directly impacts the amount of glass canisters made. This memorandum provides an estimated maximum quantity of total fissile expected to be transferred to a Sludge Batch to support increasing the fissile glass loading and thus reducing the number of canisters created across the lifetime of the ABD mission.
H=Canyon will be dissolving Fast Critical Assembly (FCA) fuel from the Japan Atomic Energy Agency (JAEA). The FCA fuel will be dissolved in the electrolytic dissolver in a solution that is a 50 wt.% nitric acid (HNO 3 ), 0.5 g/l gadolinium (Gd), and 0.05 M potassium fluoride (KF). The nitric acid concentration is expected to decrease during the batch process from 10.3 to 7 M. The temperature of the nitric acid solution may be as low as 15 °C. The fuel can will be placed in the dissolver basket insert utilizing a reusable charging device. The charging device is a coated stainless-steel rod with a clevis design. The charging device employs a linchpin to secure the FCA fuel can as it is being charged into the H-canyon dissolver. Prior to beginning the electrolytic dissolution process, the pin will be dissolved, the fuel can will remain in the dissolver basket insert and the charging device will be removed. Currently the time necessary for complete dissolution of the pin is unknown and thus the timing of the removal of the charging device cannot be planned. This process is to be performed remotely and therefore complete dissolution of the pin will not be able to be visually determined. The pin will be made of a material that dissolves in the concentrated nitric acid solution in the dissolver. The facility desired to know the time that the pin would be dissolved so that the charging device for the FCA can could be removed from the dissolver. In particular, the pin dissolution time as a function of the nitric acid concentration and the temperature was desired. The facility would like to ensure that the solution environment was such that the charging device could be removed within an operational shift (8-12 hours).
The H-Canyon Facility plans on processing bundles from Non-Aluminum Spent Nuclear Fuel (NASNF) Risk Group 1 (UO 2 core fuels). Uranium-zirconium (UZr) alloys and zirconium clad UO 2 fuel containing UZr intermetallic compound have the potential to cause explosions during dissolution. UZr intermetallic typically only forms when uranium and zirconium (or zirconium alloy) cladding are in contact at high temperatures. Thermal modeling was performed to determine the maximum cladding temperature for Campaign 1 Zircoloy cladded fuel assembles during reactor operation to help determine if the UZr intermetallic compound could have formed during reactor operations. The temperature of 350°C was chosen as a conservative minimum value for the possible formation of intermetallic UZr compound. The modeling approach, inputs, and results are outlined in this document.
To eliminate future fissile mass loading constraints for the Savannah River Site H-Canyon Facility and Liquid Waste system, the Savannah River National Laboratory recommends a repository evaluation of a uranium fissile mass loading in glass at 7,144 g/m 3 in addition to the existing International Atomic Energy Agency safeguards and security limit of 2,500 g/m 3 total plutonium. This recommended increase above the authorized 2,500 g/m 3 fissile mass loading limit is based on concentrations of uranium and plutonium shown to produce an acceptable glass waste form rather than projections of the maximum fissile mass loading in future sludge batches. An authorized fissile mass loading limit greater than 2,500 g/m 3 will increase facility flexibility, reduce the number of high-level waste canisters produced, and potentially avoid an increase in the Department of Energy Environmental Management mission life without compromising safety or glass product quality.
The Savannah River Nuclear Solutions (SRNS) Fast Critical Assembly (FCA) mission is reestablishing the electrolytic dissolver for processing of Pu and Pu-U materials clad in stainless steel (SS). H-Canyon is planning to dissolve and neutralize FCA fuel without recovering the special nuclear material (i.e., Pu) prior to discarding to the Concentration, Storage, and Transfer Facilities (CSTF) operated by the Savannah River Mission Completion (SRMC) Liquid Waste (LW) Organization. The FCA discards will be combined with sludge in the CSTF after Low Temperature Aluminum Dissolution (LTAD) if needed. The combined waste will be washed, concentrated, and vitrified at the Defense Waste Processing Facility (DWPF). The high level waste canisters produced will be temporarily stored in the Glass Waste Storage Buildings prior to transfer to a future federal repository. Decants from LTAD and sludge washing will be combined with DWPF recycle and dissolved salt cake to prepare salt batches for processing in the Salt Waste Processing Facility. The resulting decontaminated salt solution will be processed in the Saltstone Production Facility.
The Savannah River Site’s H-Canyon facility has been tasked with accelerating the disposition of used research reactor fuels. • Future missions will involve dissolution of “Non-Aluminum” clad fuels – not easily dissolved – Nitric acid is not capable of dissolving stainless steel, Zirconium, and Hastelloy, clad fuels by itself • The electrolytic dissolver has been (re) selected as a disposition path for these fuels. 2
The Savannah River Site (SRS) has designed the Accelerated Basin De-inventory (ABD) program to accelerate the de-inventory of L-Basin and accelerate the Spent Nuclear Fuel (SNF) disposition mission. Similarly, the H-Canyon facility at SRS is reestablishing the 6.3D electrolytic dissolver for dissolving unirradiated stainless-steel (SS) clad Fast Critical Assembly (FCA) fuel. In both discard types (ABD and FCA), plutonium is present and its complex solubility when composited to Concentration, Storage, and Transfer Facility (CSTF) sludge is being investigated as it may have downstream impacts to the liquid waste (LW) organization. This literature review aims to highlight and compile the existing literature on plutonium solubility in waste streams relevant to ABD and FCA discards, as well as discuss some considerations in analyzing solubility data of plutonium. This review serves to help define the analysis methods for future experiments involving plutonium (and other actinides) and in designing appropriate testing conditions surrounding these studies. This review is broken up into five parts and will discuss: (i) The possible effects of testing hold time and temperature on plutonium solubility, (ii) the influence of neutralization rate and particle size of freshly precipitated discards, (iii) the coprecipitation of plutonium with iron and uranium, (iv) predictive solubility modeling and the influence of supernate anions on solubility, and (v) the speciation of plutonium in solutionas a result of supernate anions.
• The SRS H-Canyon Building is a critical facility under the responsibility of DOE-EM. • It includes an Air Exhaust Tunnel (HCAEX) that allows for ventilation of the process airflow. • Inspections are performed remotely because of hazards, e.g. radioactivity, debris, high airflow, and nitric acid vapors.
Stainless-steel clad Pu from Japan’s Fast Critical Assembly (FCA) reactor is currently being dispositioned at the Savannah River Site. The electrolytic dissolver, operated by Savannah River Nuclear Solutions in H-Canyon, is being utilized to dissolve the material. The resulting solutions are transferred to the Concentration, Storage, and Transfer Facilities, operated by Savannah River Mission Completion (SRMC), for subsequent vitrification at the Defense Waste Processing Facility (DWPF). In support of the FCA mission startup, a preliminary evaluation was conducted by the Savannah River National Laboratory to assess the impact of the FCA discards on the liquid waste system.
Mission Statement: Concrete wall characterization for structural integrity evaluation of H-Canyon exhaust tunnel using a remote controlled robotic arm with a NDT Concrete Instrument. Challenges: Rough and Curved Surfaces, Remote location. The UR5 is a collaborative robotic arm capable of: Payload: 11 lbs (5 kg), Reach: 33.5 in (850 mm), Footprint: 5.87 in (149 mm) diameter, Weight: 45.4 Ibs (20.6 kg). The force torque sensor reacts to a set value inputted by the user that can be utilized for sensitive products. This allows the UR5 to react to surfaces using the built-in function to search for a wall and orient itself normal to that plane. The Proceq Pundit 250 array is a nondestructive device that the user applies against a concrete wall and scans using ultrasonic transducers. The back wall and other defects can be show through the touchscreen. A fixture to hold the Proceq Pundit 250 array onto the UR5 robotic arm was designed and 3D printed. It includes openings for easy access to the buttons from the Pundit array as well as a reinforced structural design. The scan from the Pundit array can show the back-wall, rebars, and other defects to assist in determining the structural integrity. In order to scan, a program was made to search for the wall first. 1. The UR5 robotic arm will go to the desired location away from the wall. 2. It will then search for the wall by moving towards it slowly. 3. After contact, the force applied will slowly but steadily increase to a preset value inputted by the user. 4. The force torque sensor will use those values and communicate with the UR5 by adjusting the arm to become normal to the wall. 5. The arm will then apply force to push back the transducers on the Pundit array to avoid gaps. 6. It will then prompt the user to collect data and wait until finished. The UR5 teach pendant allows the user to program the robot to perform automated and repeatable tasks. It includes a free drive mode which allows the user to move the arm to a desired position by hand. The user can also apply restricted planes which include either stopped or slower motion to ensure safety. The concrete sample being tested includes rebars and different grades of surface roughness to understand the readings from the Proceq Pundit 250 array. Desired concrete samples are currently being fabricated that will simulate the terrain being tested.
The Savannah River Site’s (SRS) H-Canyon recently re-activated its 6.3D electrolytic dissolver to dissolve stainless steel clad research reactor fuels from Japan. A potential new mission was identified to dissolve other long-cooled fuel stored at SRS that is not aluminum clad. In the basic process of dissolution, the spent fuel dissolves into nitric acid (bulk solution). The eight ft. high and four ft. radius 6.3D dissolver is unique in that it has a platinum-coated niobium insert to resist corrosion, and the niobium basket forms an anode and cathode arrangement within the charge chute (Fig 1,2). DC current flows from anode to cathode, at up to 10,000 amps at 45 volts, through the potential gradient in the nitric acid solution, which allows the stainless steel (or other non-aluminum cladding) of the fuel to dissolve.