M Poirier

Review: Waste-Pretreatment Technologies for Remediation of Legacy Defense Nuclear Wastes

Abstract The U.S. Department of Energy (DOE) is responsible for retrieving, immobilizing, and disposing of radioactive waste generated during the production of nuclear weapons in the United States. The general strategy for treating the radioactive tank waste consists of pretreating the wastes by separating them into high-level and low-activity fractions. The high-level fraction will be immobilized in a glass form suitable for disposal in a geologic repository. The low-activity waste will be immobilized in a waste form suitable for on-site. This paper reviews recent developments in the application of pretreatment technologies to the processing of the DOE radioactive tank wastes.

Full-Scale Testing of a Caustic Side Solvent Extraction System to Remove Cesium from Savannah River Site Radioactive Waste

Abstract Savannah River Site (SRS) personnel have completed construction and assembly of the Modular Caustic Side Solvent Extraction Unit (MCU) facility. Following assembly, they conducted testing to evaluate the ability of the process to remove non-radioactive cesium and to separate the aqueous and organic phases. They conducted tests at salt solution flow rates of 3.5, 6.0, and 8.5 gpm. During testing, the MCU Facility collected samples and submitted them to Savannah River National Laboratory (SRNL) personnel for analysis of cesium, Isopar® L, and modifier [1-(2,2,3,3-tetrafluoropropoxy)-3-(4-sec-butylphenoxy)-2-propanol]. SRNL personnel analyzed the aqueous samples for cesium by Inductively-Coupled Plasma Mass Spectroscopy (ICP-MS) and the solvent samples for cesium using a Parr Bomb digestion followed by ICP-MS. They analyzed aqueous samples for Isopar® L and modifier by gas chromatography (GC). The conclusions from the cesium analyses follow. The cesium in the feed samples measured 15.8 mg/L, in agreement with expectations. The decontamination factor measured 181–1580 at a salt solution flow rate of 3.5 gpm, 211–252 at a salt solution flow rate of 6.0 gpm, and 275–878 at a salt solution flow rate of 8.5 gpm. The concentration factor measured 11.0–11.1 at 3.5 gpm salt solution flow rate, 12.8–13.2 at 6.0 gpm salt solution flow rate, and 12.0–13.2 at 8.5 gpm salt solution flow rate. The organic carryover from the final extraction contactor (#7) varied between 22 and 710 mg/L Isopar® L. The organic carryover was less at the lowest flow rate. The organic carryover from the final strip contactor (#7) varied between 80 and 180 mg/L Isopar® L. The organic carryover in the Decontaminated Salt Solution Hold Tank (DSSHT) and the Strip Effluent Hold Tank (SEHT) was less than 10 mg/L Isopar® L, indicating good recovery of the solvent by the coalescers and decanters.

Removal of Sludge Heels in Savannah River Site Waste Tanks with Oxalic Acid

The Savannah River Site (SRS) is preparing two tanks for closure. The first step in preparing the tank for closure is mechanical sludge removal. In mechanical sludge removal, a liquid such as inhibited water or salt solution is added to the tank, the liquid is mixed with the sludge to form a slurry, and the slurry is transported from the tank. Mechanical cleaning removes a large fraction of the sludge in the tank, but it leaves a sludge heel of several thousand gallons. SRS employs chemical cleaning to remove this sludge heel. In chemical cleaning, oxalic acid is added to the tank to dissolve the sludge, and the liquid, containing the dissolved sludge, is transported from the tank. The authors conducted demonstrations of the chemical cleaning process with simulated SRS waste and actual SRS waste to assess the effectiveness of oxalic acid in dissolving SRS sludge. Following these demonstrations, SRS conducted chemical cleaning in two waste tanks (referred to as Tank A and Tank B). During chemical cleaning, the authors analyzed samples to assess the effectiveness of the chemical cleaning in removing the sludge heel. The conclusions from this work follow. With the exception of iron, the dissolution of sludge components from Tank A agreed with results from the actual waste demonstration performed in 2007. The fraction of iron removed from Tank A by chemical cleaning was significantly less than the fraction removed in the SRNL demonstrations. The likely cause of this difference is the high pH following the first oxalic acid strike. The dissolution of sludge components from Tank B agreed with results from the actual waste demonstration performed in 2007. The fraction of plutonium removed from Tank B by chemical cleaning was slightly higher than the fraction removed in the SRNL demonstrations. Most of the sludge mass remaining in the tank is iron and nickel. The remaining sludge contains significant amounts of barium, chromium, and mercury. Most of the radioactivity remaining in the residual material is beta emitters and 90Sr. The chemical cleaning removed a large fraction of the uranium, aluminum, calcium, sodium, strontium, and cesium. The chemical cleaning was not effective at removing nickel, mercury, plutonium, americium, and curium.

Processing Macrobatch 2 at the Savannah River Site Integrated Salt Disposition Process (ISDP)

The Savannah River Site (SRS) is currently removing liquid radioactive waste from the tanks in its Tank Farm. To treat waste streams that are high in 137Cs, 90Sr, and/or actinides, SRS developed the Actinide Removal Process (ARP) and the Modular Caustic Side Solvent Extraction (CSSX) Unit. Collectively, these two processes make up the Integrated Salt Disposition Process (ISDP). The ARP part is responsible for the removal of strontium and actinides, while the MCU part is responsible for removing cesium. This paper discusses the qualification testing of the second batch of caustic waste that is being processed through ISDP currently. This paper also describes the tests conducted and compares results with current facility performance. The ARP contacts the salt solution with monosodium titanate (MST) to sorb strontium and select actinides. After MST contact, the resulting slurry is filtered to remove the MST (and sorbed strontium and actinides) and entrained sludge. The filtrate is transferred to the MCU for further treatment to remove cesium. The solid particulates removed by the filter are concentrated to ∼5 wt%, washed to reduce the sodium concentration, and transferred to the Defense Waste Processing Facility (DWPF) for vitrification. The CSSX process extracts the cesium from the radioactive waste using a customized solvent to produce a Decontaminated Salt Solution (DSS), then strips and concentrates the cesium from the solvent with dilute nitric acid. The DSS is incorporated in grout while the strip acid solution is transferred to DWPF for vitrification. In order to predict waste behavior, the Savannah River National Laboratory (SRNL) personnel performed tests using actual radioactive samples of the second waste batch – Macrobatch 2 – for processing prior to the start of the operation. Testing included MST sorption to remove strontium and actinides followed by CSSX batch contact tests to verify expected cesium mass removal and concentration. This paper describes the tests conducted and compares results from MCU facility operations. The results include strontium, plutonium, and cesium removal, cesium concentration, and organic entrainment and recovery data. Our work indicates that the bench scale tests are a conservative predictor of actual waste performance.

BLENDING TIME AND VELOCITY VARIATIONS DURING BLENDING IN A TANK USING DUAL OPPOSING JETS

Blending times are required for many process industries, and statistical analysis of the measured blending times was used to determine a relationship between CFD (computational fluid dynamics) predictions and experiments. A 95% blending time occurs when tank contents are sufficiently blended to ensure that concentration throughout the tank is within ±5% of the total change in concentration. To determine 95% blending times, acid and base tracers were added to an eight foot diameter tank, and the pH data were recorded to monitor blending. The data for six pH probes located throughout the tank were normalized to a range of 0 to 1. Then the blending time was established when the pH converged between 0.95 and 1.05 on the normalized graphs. Evaluation of results from 79 different tests concluded that the maximum blending time occurred randomly at any one of the six pH probes. The research then considered the calculated 95% blending times, which had uncertainties up to more than 100% at a 95% confidence level. However, this uncertainty is considered to be an actual variation in blending time, rather than an experimental error. Not only were there significant variations in the blending times, but there were significant variations in the velocities measured at different points in the blending tank.Copyright

Evaluation of Alternative Filter Media for the Rotary Microfilter

The Savannah River Site is currently developing and testing several processes to treat high level radioactive liquid waste. Each of these processes has a solid-liquid separation process that limits its throughput. Savannah River National Laboratory researchers identified and tested the rotary microfilter as a technology to increase solid-liquid separation throughput. The authors believe the rotary microfilter throughput can be improved by using a better filter membrane. Previous testing showed that asymmetric filters composed of a ceramic membrane on top of a stainless steel support produced higher filter flux than 100% stainless steel symmetric filters in crossflow filter tests. Savannah River National Laboratory and Oak Ridge National Laboratory are working together to develop asymmetric ceramic–stainless steel composite filters and asymmetric 100% stainless steel filters to improve the throughput of the rotary microfilter.

Comparison of Experimental Results to CFD Models for Blending in a Tank Using Dual Opposing Jets

Research has been completed in a pilot scale, eight foot diameter tank to investigate blending, using a pump with dual opposing jets. The jets re-circulate fluids in the tank to promote blending when fluids are added to the tank. Different jet diameters and different horizontal and vertical orientations of the jets were investigated. In all, eighty five tests were performed both in a tank without internal obstructions and a tank with vertical obstructions similar to a tube bank in a heat exchanger. These obstructions provided scale models of several miles of two inch diameter, serpentine, vertical cooling coils below the liquid surface for a full scale, 1.3 million gallon, liquid radioactive waste storage tank. Two types of tests were performed. One type of test used a tracer fluid, which was homogeneously blended into solution. Data were statistically evaluated to determine blending times for solutions of different density and viscosity, and the blending times were successfully compared to computational fluid dynamics (CFD) models. The other type of test blended solutions of different viscosity. For example, in one test a half tank of water was added to a half tank of a more viscous, concentrated salt solution. In this case, the fluid mechanics of the blending process was noted to significantly change due to stratification of fluids. CFD models for stratification were not investigated. This paper is the fourth in a series of papers resulting from this research (Leishear, et.al. [1- 4]), and this paper documents final test results, statistical analysis of the data, a comparison of experimental results to CFD models, and scale-up of the results to a full scale tank.