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Temperature control in a 30 stage, 5-cm Centrifugal Contactor Pilot Plant

Description: Temperature profile testing was performed using a 30 stage 5-cm centrifugal contactor pilot plant. These tests were performed to evaluate the ability to control process temperature by adjusting feed solution temperatures. This would eliminate the need for complex jacketed heat exchanger installation on the centrifugal contactors. Thermocouples were installed on the inlet and outlets of each stage, as well as directly in the mixing zone of several of the contactor stages. Lamp oil, a commercially available alkane mixture of C14 to C18 chains, and tap water adjusted to pH 2 with nitric acid were the solution feeds for the temperature profile testing. Temperature data profiles for an array of total throughputs and contactor rpm values for both single-phase and two-phase systems were collected with selected profiles. The total throughput ranged from 0.5-1.4 L/min with rotor speeds from 3500-4000 rpm. Inlet solution temperatures ranging from ambient up to 50 °C were tested. Results of the two-phase temperature profile testing are detailed
Date: September 1, 2009
Creator: Law, Jack D.; Garn, Troy G. & Meikrantz, David H.
Partner: UNT Libraries Government Documents Department

Design Attributes and Scale Up Testing of Annular Centrifugal Contactors

Description: Annular centrifugal contactors are being used for rapid yet efficient liquid- liquid processing in numerous industrial and government applications. Commercialization of this technology began eleven years ago and now units with throughputs ranging from 0.25 to 700 liters per minute are readily available. Separation, washing, and extraction processes all benefit from the use of this relatively new commercial tool. Processing advantages of this technology include: low in-process volume per stage, rapid mixing and separation in a single unit, connection-in-series for multi-stage use, and a wide operating range of input flow rates and phase ratios without adjustment. Recent design enhancements have been added to simplify maintenance, improve inspection ability, and provide increased reliability. Cartridge-style bearing and mechanical rotary seal assemblies that can include liquid-leak sensors are employed to enhance remote operations, minimize maintenance downtime, prevent equipment damage, and extend service life. Clean-in-place capability eliminates the need for disassembly, facilitates the use of contactors for feed clarification, and can be automated for continuous operation. In nuclear fuel cycle studies, aqueous based separations are being developed that efficiently partition uranium, actinides, and fission products via liquid-liquid solvent extraction. Thus, annular centrifugal contactors are destined to play a significant role in the design of such new processes. Laboratory scale studies using mini-contactors have demonstrated feasibility for many such separation processes but validation at an engineering scale is needed to support actual process design.
Date: April 1, 2005
Creator: Meikrantz, David H. & Law, Jack D.
Partner: UNT Libraries Government Documents Department

Evaluation of a New Remote Handling Design for High Throughput Annular Centrifugal Contactors

Description: Advanced designs of nuclear fuel recycling plants are expected to include more ambitious goals for aqueous based separations including; higher separations efficiency, high-level waste minimization, and a greater focus on continuous processes to minimize cost and footprint. Therefore, Annular Centrifugal Contactors (ACCs) are destined to play a more important role for such future processing schemes. Previous efforts defined and characterized the performance of commercial 5 cm and 12.5 cm single-stage ACCs in a “cold” environment. The next logical step, the design and evaluation of remote capable pilot scale ACCs in a “hot” or radioactive environment was reported earlier. This report includes the development of remote designs for ACCs that can process the large throughput rates needed in future nuclear fuel recycling plants. Novel designs were developed for the remote interconnection of contactor units, clean-in-place and drain connections, and a new solids removal collection chamber. A three stage, 12.5 cm diameter rotor module has been constructed and evaluated for operational function and remote handling in highly radioactive environments. This design is scalable to commercial CINC ACC models from V-05 to V-20 with total throughput rates ranging from 20 to 650 liters per minute. The V-05R three stage prototype was manufactured by the commercial vendor for ACCs in the U.S., CINC mfg. It employs three standard V-05 clean-in-place (CIP) units modified for remote service and replacement via new methods of connection for solution inlets, outlets, drain and CIP. Hydraulic testing and functional checks were successfully conducted and then the prototype was evaluated for remote handling and maintenance suitability. Removal and replacement of the center position V-05R ACC unit in the three stage prototype was demonstrated using an overhead rail mounted PaR manipulator. This evaluation confirmed the efficacy of this innovative design for interconnecting and cleaning individual stages while retaining the benefits of ...
Date: September 1, 2009
Creator: Meikrantz, David H.; Garn, Troy G.; Law, Jack D. & Macaluso, Lawrence L.
Partner: UNT Libraries Government Documents Department

Flowsheet Testing of the Fission Product Extraction Process as Part of Advanced Aqueous Reprocessing

Description: As part of the Advanced Fuel Cycle Initiative (AFCI), the reduction in volume and heat generation of spent nuclear fuel requiring geologic disposal is currently being addressed. The goal is to optimize utilization of the nation’s first repository and potentially reduce or eliminate the need for additional repositories. This will be achieved through separating long-lived, highly toxic elements, reducing high-level waste volumes and the toxicity of spent nuclear fuel, and reducing the heat generation of spent nuclear fuel. The Idaho National Laboratory (INL) is working closely with a team of national laboratories and other organizations to support development of these separations processes. Key to the reduction of short-term heat load in a geological repository is the separation of 137Cs and 90Sr. Removal of these highly radioactive fission products reduces the short-term (~100 yr) heat generation source of the spent nuclear fuel. Once separated, the Cs and Sr would be placed in storage until the activity has decayed to LLW levels, at which time it could potentially be disposed of as a non-transuranic (TRU) low-level waste (LLW).
Date: June 1, 2007
Creator: Law, Jack; Peterman, Dean R.; Riddle, catherine; Meikrantz, David H. & Todd, Terry
Partner: UNT Libraries Government Documents Department

ADVANCED TECHNOLOGIES FOR THE SIMULTANEOUS SEPARATION OF CESIUM AND STRONTIUM FROM SPENT NUCLEAR FUEL

Description: Two new solvent extraction technologies have been recently developed to simultaneously separate cesium and strontium from spent nuclear fuel, following dissolution in nitric acid. The first process utilizes a solvent consisting of chlorinated cobalt dicarbollide and polyethylene glycol extractants in a phenyltrifluoromethyl sulfone diluent. Recent improvements to the process include development of a new, non-nitroaromatic diluent and development of new stripping reagents, including a regenerable strip reagent that can be recovered and recycled. This new strip reagent reduces product volume by a factor of 20, over the baseline process. Countercurrent flowsheet tests on simulated spent nuclear fuel feed streams have been performed with both cesium and strontium removal efficiencies of greater than 99 %. The second process developed to simultaneously separate cesium and strontium from spent nuclear fuel is based on two highly-specific extractants: 4',4',(5')-Di-(t-butyldicyclo-hexano)-18-crown-6 (DtBuCH18C6) and Calix[4]arene-bis-(tert-octylbenzo-crown-6) (BOBCalixC6). The DtBuCH18C6 extractant is selective for strontium and the BOBCalixC6 extractant is selective for cesium. A solvent composition has been developed that enables both elements to be removed together and, in fact, a synergistic effect was observed with strontium distributions in the combined solvent that are much higher that in the strontium extraction (SREX) process. Initial laboratory test results of the new combined cesium and strontium extraction process indicate good extraction and stripping performance.
Date: February 1, 2005
Creator: Law, Jack D.; Todd, Terry A.; Herbst, R. Scott; Meikrantz, David H.; Peterman, Dean R.; Riddle, Catherine L. et al.
Partner: UNT Libraries Government Documents Department

Hydraulic and Clean-in-Place Evaluations for a 12.5-cm Annular Centrifugal Contactor at INL

Description: Hydraulic and Clean-in-Place Evaluations for a 12.5 cm Annular Centrifugal Contactor at the INL Troy G. Garn, Dave H. Meikrantz, Nick R. Mann, Jack D. Law, Terry A. Todd Idaho National Laboratory Commercially available, Annular Centrifugal Contactors (ACC) are currently being evaluated for processing dissolved nuclear fuel solutions to selectively partition integrated elements using solvent extraction technologies. These evaluations include hydraulic and clean-in-place (CIP) testing of a commercially available 12.5 cm unit. Data from these evaluations is used to support design of future nuclear fuel reprocessing facilities. Hydraulic testing provides contactor throughput performance data on two-phase systems for a wide range of operating conditions. Hydraulic testing results on a simple two-phase oil and water system followed by a 30 % Tributyl phosphate in N-dodecane / nitric acid pair are reported. Maximum total throughputs for this size contactor ranged from 20 to 32 liters per minute without significant other phase carryover. A relatively new contactor design enhancement providing Clean-in-Place capability for ACCs was also investigated. Spray nozzles installed into the central rotor shaft allow the rotor internals to be cleaned, offline. Testing of the solids capture of a diatomaceous earth/water slurry feed followed by CIP testing was performed. Solids capture efficiencies of >95% were observed for all tests and short cold water cleaning pulses proved successful at removing solids from the rotor.
Date: September 1, 2008
Creator: Garn, Troy G.; Meikrantz, David H.; Mann, Nick R.; Law, Jack D. & Todd, Terry A.
Partner: UNT Libraries Government Documents Department

Development of Technologies for the Simultaneous Separation of Cesium and Strontium from Spent Nuclear Fuel as Part of an Advanced Fuel Cycle

Description: As part of the Advanced Fuel Cycle Initiative, two solvent extraction technologies are being developed to simultaneously separate cesium and strontium from dissolved spent nuclear fuel. The first process utilizes a solvent consisting of chlorinated cobalt dicarbollide and polyethylene glycol extractants in a phenyltrifluoromethyl sulfone diluent. Recent improvements to the process include development of a new, non-nitroaromatic diluent and development of new stripping reagents, including a regenerable strip reagent that can be recovered and recycled. Countercurrent flowsheets have been designed and tested on simulated and actual spent nuclear fuel feed streams with both cesium and strontium removal efficiencies of greater than 99 %. The second process developed to simultaneously separate cesium and strontium from spent nuclear fuel is based on two highly-specific extractants: 4,4',(5')-Di-(t-butyldicyclo-hexano)-18-crown-6 (DtBuCH18C6) and Calix[4]arene-bis-(tert-octylbenzo-crown-6) (BOBCalixC6). The DtBuCH18C6 extractant is selective for strontium and the BOBCalixC6 extractant is selective for cesium. A solvent composition has been developed that enables both elements to be removed together and, in fact, a synergistic effect was observed with strontium distributions in the combined solvent that are much higher that in the strontium extraction (SREX) process. Initial laboratory test results of the new combined cesium and strontium extraction process indicate good extraction and stripping performance. A flowsheet for treatment of spent nuclear fuel is currently being developed.
Date: April 1, 2005
Creator: Law, Jack D.; HErbst, R. Scott; Meikrantz, David H.; Peterman, Dean R.; Riddle, Catherine L.; Tillotson, Richard D. et al.
Partner: UNT Libraries Government Documents Department

Development of a novel solvent for the simultaneous separation of strontium and cesium from dissolved Spent Nuclear Fuel solutions

Description: The recovery of Cs and Sr from acidic solutions by solvent extraction has been investigated. The goal of this project was to develop an extraction process to remove Cs and Sr from high-level waste in an effort to reduce the heat loading in storage. Solvents for the extraction of Cs and Sr separately have been used on both caustic and acidic spent nuclear fuel waste in the past. The objective of this research was to find a suitable solvent for the extraction of both Cs and Sr simultaneously from acidic nitrate media. The solvents selected for this research possess good stability and extraction behavior when mixed together. The extraction experiments were performed with 4 ,4,(5 )-Di-(tbutyldicyclohexano)- 18-crown-6 {DtBuCH18C6}, Calix[4]arene-bis-(tert-octylbenzocrown-6) {BOBCalixC6} and 1-(2,2,3,3-tetrafluoropropoxy)-3-(4-sec-butylphenoxy)-2-propanol {Cs-7SB modifier} in a branched aliphatic kerosene {Isopar® L}. The BOBCalixC6 and Cs-7SB modifier were developed at Oak Ridge National Laboratory (ORNL) by Bonnesen et al. [1]. The values obtained from the SREX solvent for DSr in 1 M nitric acid ranged from 0.7 to 2.2 at 25oC and 10oC respectively. The values for DCs in 1 M nitric acid with the CSSX solvent ranged from 8.0 to 46.0 at 25oC and 10oC respectively. A new mixed solvent, developed at the Idaho National Engineering and Environmental Laboratory (INEEL) by Riddle et al. [2], showed distributions for Sr ranging from 8.8 to 17.4 in 1 M nitric acid at 25oC and 10oC respectively. The DCs for the mixed solvent ranged from 7.7 to 20.2 in 1 M nitric acid at 25oC to 10oC respectively. The unexpectedly high distributions for Sr at both 25oC and 10oC show a synergy in the mixed solvent. The DCs, although lower than with CSSX solvent, still showed good extraction behavior.
Date: October 1, 2004
Creator: Riddle, Catherine L.; Baker, John D.; Law, Jack D.; McGrath, Christopher A.; Meikrantz, David H.; Mincher, Bruce J. et al.
Partner: UNT Libraries Government Documents Department

Development of Cesium and Strontium Separation and Immobilization Technologies in Support of an Advanced Nuclear Fuel Cycle

Description: As part of the Advanced Fuel Cycle Initiative, two solvent extraction technologies are being developed at the Idaho National Laboratory to simultaneously separate cesium and strontium from dissolved spent nuclear fuel. The chlorinated cobalt dicarbollide/polyethylene glycol (CCD/PEG) process utilizes a solvent consisting of chlorinated cobalt dicarbollide for the extraction of Cs and polyethylene glycol for the synergistic extraction of Sr in a phenyltrifluoromethyl sulfone diluent. Countercurrent flowsheets have been designed and tested on simulated and actual spent nuclear fuel feed streams with both cesium and strontium removal efficiencies of greater than 99%. The Fission Product Extraction (FPEX) process is based on two highly-specific extractants: 4,4',(5')-Di-(t-butyldicyclo-hexano)-18-crown-6 (DtBuCH18C6) for the extraction of Sr and Calix[4]arene-bis-(tert-octylbenzo-crown-6) (BOBCalixC6) for the extraction of Cs. Laboratory test results of the FPEX process, using simulated feed solution spiked with radiotracers, indicate good Cs and Sr extraction and stripping performance. A preliminary solvent extraction flowsheet for the treatment of spent nuclear fuel with the FPEX process has been developed, and testing of the flowsheet with simulated spent nuclear fuel solutions is planned in the near future. Steam reforming is currently being developed for stabilization of the Cs/Sr product stream because it can produce a solid waste form while retaining the Cs and Sr in the solid, destroy the nitrates and organics present in these aqueous solutions, and convert the Cs and Sr into leach resistant aluminosilicate minerals. A bench-scale steam reforming pilot plant has been operated with several potential feed compositions and steam reformed product has been generated and analyzed.
Date: February 1, 2006
Creator: Law, Jack D.; Garn, Troy G.; Herbst, R. Scott; Meikrantz, David H.; Peterman, Dean R.; Riddle, Catherine L. et al.
Partner: UNT Libraries Government Documents Department