Advances in Development of the Fission Product Extraction Process for the Separation of Cesium and Strontium from Spent Nuclear Fuel

The Fission Product Extraction (FPEX) Process is being developed as part of the United States Department of Energy Advanced Fuel Cycle Initiative for the simultaneous separation of cesium (Cs) and strontium (Sr) from spent light water reactor (LWR) fuel. Separation of the Cs and Sr will reduce the short-term heat load in a geological repository, and when combined with the separation of americium (Am) and curium (Cm), could increase the capacity of the geological repository by a factor of approximately 100. The FPEX process 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. Results of flowsheet testing of the FPEX process with a simulated feed solution in 3.3-cm centrifugal contactors are detailed. Removal efficiencies, distribution coefficient data, coextraction of metals, and process hydrodynamic performance are discussed along with recommendations for future flowsheet testing with actual spent nuclear fuel

Liquid-Liquid Extraction Equipment

Solvent extraction processing has demonstrated the ability to achieve high decontamination factors for uranium and plutonium while operating at high throughputs. Historical application of solvent extraction contacting equipment implies that for the HA cycle (primary separation of uranium and plutonium from fission products) the equipment of choice is pulse columns. This is likely due to relatively short residence times (as compared to mixer-settlers) and the ability of the columns to tolerate solids in the feed. Savannah River successfully operated the F-Canyon with centrifugal contactors in the HA cycle (which have shorter residence times than columns). All three contactors have been successfully deployed in uranium and plutonium purification cycles. Over the past 20 years, there has been significant development of centrifugal contactor designs and they have become very common for research and development applications. New reprocessing plants are being planned in Russia and China and the United States has done preliminary design studies on future reprocessing plants. The choice of contactors for all of these facilities is yet to be determined

Design Attributes and Scale Up Testing of Annular Centrifugal Contactors

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

Summary of TRUEX Radiolysis Testing Using the INL Radiolysis Test Loop

The INL radiolysis and hydrolysis test loop has been used to evaluate the effects of hydrolytic and radiolytic degradation upon the efficacy of the TRUEX flowsheet for the recovery of trivalent actinides and lanthanides from acidic solution. Repeated irradiation and subsequent re-conditioning cycles did result in a significant decrease in the concentration of the TBP and CMPO extractants in the TRUEX solvent and a corresponding decrease in americium and europium extraction distributions. However, the build-up of solvent degradation products upon {gamma}-irradiation, had little impact upon the efficiency of the stripping section of the TRUEX flowsheet. Operation of the TRUEX flowsheet would require careful monitoring to ensure extraction distributions are maintained at acceptable levels

Characterization of radiolytically generated degradation products in the strip section of a TRUEX flowsheet

This report presents a summary of the work performed to meet the FCRD level 2 milestone M3FT-13IN0302053, “Identification of TRUEX Strip Degradation.” The INL radiolysis test loop has been used to identify radiolytically generated degradation products in the strip section of the TRUEX flowsheet. These data were used to evaluate impact of the formation of radiolytic degradation products in the strip section upon the efficacy of the TRUEX flowsheet for the recovery of trivalent actinides and lanthanides from acidic solution. The nominal composition of the TRUEX solvent used in this study is 0.2 M CMPO and 1.4 M TBP dissolved in n-dodecane and the nominal composition of the TRUEX strip solution is 1.5 M lactic acid and 0.050 M diethylenetriaminepentaacetic acid. Gamma irradiation of a mixture of TRUEX process solvent and stripping solution in the test loop does not adversely impact flowsheet performance as measured by stripping americium ratios. The observed increase in americium stripping distribution ratios with increasing absorbed dose indicates the radiolytic production of organic soluble degradation compounds

Quantifying black carbon deposition over the Greenland ice sheet from forest fires in Canada

Black carbon (BC) concentrations observed in 22 snowpits sampled in the northwest sector of the Greenland ice sheet in April 2014 have allowed us to identify a strong and widespread BC aerosol deposition event, which was dated to have accumulated in the pits from two snow storms between 27 July and 2 August 2013. This event comprises a significant portion (57% on average across all pits) of total BC deposition over 10 months (July 2013 to April 2014). Here we link this deposition event to forest fires burning in Canada during summer 2013 using modeling and remote sensing tools. Aerosols were detected by both the Cloud‐Aerosol Lidar with Orthogonal Polarization (on board CALIPSO) and Moderate Resolution Imaging Spectroradiometer (Aqua) instruments during transport between Canada and Greenland. We use high‐resolution regional chemical transport modeling (WRF‐Chem) combined with high‐resolution fire emissions (FINNv1.5) to study aerosol emissions, transport, and deposition during this event. The model captures the timing of the BC deposition event and shows that fires in Canada were the main source of deposited BC. However, the model underpredicts BC deposition compared to measurements at all sites by a factor of 2–100. Underprediction of modeled BC deposition originates from uncertainties in fire emissions and model treatment of wet removal of aerosols. Improvements in model descriptions of precipitation scavenging and emissions from wildfires are needed to correctly predict deposition, which is critical for determining the climate impacts of aerosols that originate from fires

THE TESTING OF COMMERCIALLY AVAILABLE ENGINEERING AND PLANT SCALE ANNULAR CENTRIFUGAL CONTACTORS FOR THE PROCESSING OF SPENT NUCLEAR FUEL

Annular centrifugal contactors are being evaluated for process scale solvent extraction operations in support of United State Advanced Fuel Cycle Initiative goals. These contactors have the potential for high stage efficiency if properly employed and optimized for the application. Commercially available centrifugal contactors are being tested at the Idaho National Laboratory to support this program. Hydraulic performance and mass transfer efficiency have been measured for portions of an advanced nuclear fuel cycle using 5-cm diameter annular centrifugal contactors. Advanced features, including low mix sleeves and clean-in-place rotors, have also been evaluated in 5-cm and 12.5-cm contactors

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

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 commercially reliable ACC equipment for remote applications in the nuclear industry. Minor modifications and suggestions for improved manual remote servicing by the remote handling specialists were provided but successful removal and replacement was demonstrated in the first prototype

Temperature Profile Measurements in a Newly Constructed 30-Stage 5 cm Centrifugal Contactor pilot Plant

An annular centrifugal contactor pilot plant incorporating 30 stages of commercial 5 cm CINC V-02 units has been built and operated at INL during the past year. The pilot plant includes an automated process control and data acquisitioning system. The primary purpose of the pilot plant is to evaluate the performance of a large number of inter-connected centrifugal contactors and obtain temperature profile measurements within a 30-stage cascade. Additional solvent extraction flowsheet testing using stable surrogates is also being considered. Preliminary hydraulic testing was conducted with all 30 contactors interconnected for continuous counter-current flow. Hydraulic performance and system operational tests were conducted successfully but with higher single-stage rotor speeds found necessary to maintain steady interstage flow at flowrates of 1 L/min and higher. Initial temperature profile measurements were also completed in this configuration studying the performance during single aqueous and two-phase counter-current flow at ambient and elevated inlet solution temperatures. Temperature profile testing of two discreet sections of the cascade required additional feed and discharge connections. Lamp oil, a commercially available alkane mixture of C14 to C18 chains, and tap water adjusted to pH 2 were the solution feeds for all the testing described in this report. Numerous temperature profiles were completed using a newly constructed 30-stage centrifugal contactor pilot plant. The automated process control and data acquisition system worked very well throughout testing. Temperature data profiles for an array of total flowrates (FT) and contactor rpm values for both single-phase and two-phase systems have been collected with selected profiles and comparisons reported. Total flowrates (FT) ranged from 0.5-1.4 L/min with rotor speeds from 3500-4000 rpm. Solution inlet temperatures ranging from ambient up to 50° C were tested. Ambient temperature testing shows that a small amount of heat is added to the processed solution by the mechanical energy of the contactors. The temperature profiles match the ambient temperature of the laboratory but are nearly 10° C higher toward the middle of the cascade. Heated input solution testing provides temperature profiles with smaller temperature gradients and are more influenced by the temperature of the inlet solutions than the ambient laboratory temperature. The temperature effects of solution mixing, even at 4000 rpm, were insignificant in any of the studies conducted on lamp oil and water