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At least 19 records

Enhancement of a Process Flowsheet for Recovering and Concentrating Critical Materials from Bituminous Coal Sources

Recovery of rare earth elements (REE) from coal-related resources has recently received significant interest due to supply concerns and their key roles in critical industries and defense-related technologies. In this study, an integrated flowsheet consisting of sorting, crushing/grinding, physical separation, acid leaching, solvent extraction, and selective precipitation was designed to achieve continuous production of rare earth products from various coal-related resources such as coarse refuse and acid mine drainage. A pilot plant was constructed that enabled continuous testing of the circuitry on a range of different coal-based feedstocks. Despite achieving significant success in producing high-grade rare earth oxide products, low REE recovery values and high operational costs necessitated a re-evaluation of the process flowsheet. This article describes the circuitry improvements that were based on both laboratory and pilot plant test results. Roasting as a pre-leach treatment was found to significantly improve the leaching rate and overall REE recovery values. After roasting West Kentucky No. 13 and Fire Clay coal refuse materials at 600 °C, REE recovery increased by 60 and 40 absolute percentage points, respectively. Rare earth concentration in the pregnant leachate solution (PLS) was increased to around 700 ppm prior to solvent extraction using a staged precipitation method. REEs in the precipitate were re-dissolved, purified, and extracted from the PLS using solvent extraction and selective precipitation to values of around 80%. As a result of the improvements, chemical consumption in the acid leaching and solvent extraction processes was significantly reduced. A techno-economical assessment of the improved flowsheet indicated significant commercialization potential.

01 COAL, LIGNITE, AND PEAT↗

Process evaluation and flowsheet development for the recovery of rare earth elements from coal and associated byproducts

Coal and coal byproducts produced annually contain enough rare earth elements (REEs) to meet current U.S. demand. Here, the REEs exist in the form of minerals, ion-exchanged elements associated with clays and chemically bonded elements within the organic coal matrix. Researchers at three universities have collaborated to evaluate the effectiveness of existing physical and chemical concentration processes for the recovery of the REEs. The efforts have found that physical separation processes can increase the total REE content in the thickener underflow material collected from a Central Appalachian coal cleaning facility from around 300 ppm to values greater than 17,500 ppm on an ash basis. Leaching followed by solvent extraction was found to be the most effective to recover the REEs from coarse middlings materials collected from multiple coal basins as indicated by recovery values exceeding 80%. Based on material characterization data and process evaluations conducted for several potential feed coal sources, economic recovery of the REEs will require byproduct production of both clean coal and REE concentrate using a process flowsheet that is unique for each source.

01 COAL, LIGNITE, AND PEAT↗

Evaluation of Changes to the MTR Fuel Dissolution Flowsheet to Increase the Number of L-Bundles Charged to a Dissolver

Material Test Reactor (MTR) fuels are dissolved in the H-Canyon facility using a Hg-catalyzed, HNO₃ dissolution flowsheet. The flowsheet is based on experimental work in which laboratory-scale experiments were used to provide bounding values for both offgas and H₂ generation rates. The generation of H₂ gas, primarily from the dissolution of Al, is controlled by limiting the surface area exposed to the solution. The number of L-Bundles charged and their submergence level defines the exposed surface area which is directly proportional to the H₂ generation rate. For the MTR fuel dissolution flowsheet, the maximum number of L-Bundles which can be charged to a dissolver was specified as a function of the Al concentration in the solution and the submergence level of the bundle. The MTR fuels are currently dissolved in the 6.1D dissolver. To reach the desired terminal Al concentration, a total of 12 L-Bundles of fuel are dissolved per batch. Depending on the submergence level being used in the dissolver, the 12 bundles are dissolved in three charges of four bundles per charge or in a sequence of three, four, and five bundles. However, if six L-Bundles could be dissolved in the initial charge, a dissolver batch could be completed in two dissolutions instead of three. To assess the feasibility of dissolving six bundles in the first charge, the Savannah River National Laboratory evaluated changes to the MTR flowsheet which increase the number of bundles which can be charged to a dissolver. Since the development of the existing MTR fuel dissolution flowsheet, improved offgas measurement methods have been developed to more accurately calculate the offgas composition and generation rate during laboratory-scale dissolutions. In the first part of this study, improvements in the calibration model for the Raman spectrometer (used for offgas characterization) and a more effective data smoothing routine were applied to existing data. The uncertainty in the offgas composition data used to define the MTR fuel dissolution flowsheet were subsequently re-evaluated which resulted in an approximate 50% reduction in the uncertainty of the H₂ gas concentration measurements. The reduced uncertainty was used in new calculations to define the maximum number of L-Bundles which can be charged to either the 6.1D or 6.4D dissolver as a function of the Al concentration. With the reduced uncertainty in the H₂ concentration measurements, a flowsheet was defined which supported initially charging 6 bundles to a dissolver. This flowsheet is based on the use of 0.002 M Hg, an L-Bundle immersion depth of 54 in., an air sparge/purge flowrate of 40 scfm, an iodine reactor operating temperature of 200 °C, and not exceeding 60% of the H₂ LFL in the dissolver offgas. In the second phase of this study, calculations were performed to assess the impact of changing various operating parameters to increase the number of L-Bundles charged to an H-Canyon dissolver. Applying the improved offgas measurement techniques including the reduced uncertainty in the H₂ concentration measurements to the existing data, the number of bundles of MTR fuel which can be charged to either the 6.1D or 6.4D dissolver were calculated based on varying the bundle immersion depth (44-60 in.), air sparge/purge flowrate (40-60 scfm), iodine reactor operating temperature (150-200 °C), and the percentage of the H₂ LFL which must not be exceeded (60-76%). Qualitatively, the percentage of the LFL had the largest influence on the maximum number of bundles which can be charged to a dissolver followed by the air sparge/purge flowrate, L-Bundle immersion depth, and the iodine reactor operating temperature. During the evaluation of each variable, the other variables were held at the baseline flowsheet conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Recommendation of ruthenium source for sludge batch flowsheet studies

Included herein is a preliminary analysis of previously-generated data from sludge batches 7a, 7b, 8, and 9 sludge simulant and real-waste testing, performed to recommend a form of ruthenium for future sludge batch simulant testing under the nitric-formic flowsheet. Focus is given to reactions present in the Sludge Receipt and Adjustment Tank cycle, given that this cycle historically produces the most changes in chemical composition during Chemical Process Cell processing. Data is presented and analyzed for several runs performed under the nitric-formic flowsheet, with consideration given to effects on the production of hydrogen gas, nitrous oxide gas, consumption of formate, conversion of nitrite to nitrate, and the removal and recovery of mercury during processing. Additionally, a brief discussion is given to the effect of ruthenium source selection under the nitric-glycolic flowsheet. An analysis of data generated from scaled demonstration testing, sludge batch 9 qualification testing, and antifoam degradation testing under the nitric-glycolic flowsheet is presented. Experimental parameters of interest under the nitric-glycolic flowsheet include N2O production, glycolate destruction, conversion of glycolate to formate and oxalate, and the conversion of nitrite to nitrate. To date, the number of real-waste experiments that have been performed under the nitric-glycolic flowsheet is insufficient to provide a complete understanding of the effects of ruthenium source selection in simulant experiments with regard to fidelity to real-waste testing. Therefore, a determination of comparability between the two ruthenium sources as employed under the nitric-glycolic flowsheet is made based on available data in order to inform ruthenium source selection for future testing under the nitric-glycolic flowsheet.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Prioritized List of Research Needs to support MRWFD Case Study Flowsheet Advancement

In FY-13, a case study evaluation was performed of full recycle technologies for both the processing of light-water reactor (LWR) used nuclear fuels as well as fast reactor (FR) fuel in the full recycle option. This effort focused on the identification of the case study processes and the initial preparation of material balance flowsheets for the identified technologies. In identifying the case study flowsheets, it was decided that two cases would be developed: one which identifies the flowsheet as currently developed and another near-term target flowsheet which identifies the flowsheet as envisioned within two years, pending the results of ongoing research. The case study focus is on homogeneous aqueous recycle of the U/TRU resulting from the processing of LWR fuel as feed for metal fuel fabrication. The metal fuel is utilized in a sodium-cooled fast reactor, and the used fast reactor fuel is processed using electrochemical separations. The recovered U/TRU from electrochemical separations is recycled to fuel fabrication and the fast reactor. Waste streams from the aqueous and electrochemical processing are treated and prepared for disposition. Off-gas from the separations and waste processing are also treated. As part of the FY-13 effort, preliminary process unknowns and research needs to advance the near-term target flowsheets were identified. In FY-14, these research needs were updated, expanded and prioritized. This report again updates the prioritized list of research needs based upon results to date in FY-15. The research needs are listed for each of the main portions of the flowsheet: 1) Aqueous headend, 2) Headend tritium pretreatment off-gas, 3) Aqueous U/Pu/Np recovery, 4) Aqueous TRU product solidification, 5) Aqueous actinide/lanthanide separation, 6) Aqueous off-gas treatment, 7) Aqueous HLW management, 8) Treatment of aqueous process wastes, 9) E-chem actinide separations, 10) E-chem off-gas, 11) E-chem HLW management. The identified research needs were prioritized within each of these areas. No effort was made to perform an overall prioritization. This information will be used by the MRWFD Campaign leadership in research planning for FY-16. Additionally, this information will be incorporated into the next version of the Case Study Report scheduled to be issued September 2015.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Direct-Feed Low-Activity Waste: First Feed Flowsheet - 17108

In support of hot commissioning of the Low-Activity Waste Pretreatment System (LAWPS) and the Hanford Tank Waste Treatment and Immobilization Plant (WTP) LAW facility, the First Feed Flowsheet was initiated to project process stream characteristics and unit operation performance. The First Feed Flowsheet uses flowsheet models to simulate the preparation of the first feed campaign in Tank Farms and the subsequent feeding of the waste to the Low-Activity Waste Pretreatment System (LAWPS) and the WTP LAW facility. Three different models were used: a chemistry model of waste transfers developed in the Environmental Simulation Program (ESP), Washington River Protection Solutions' (WRPS') dynamic gPROMS{sup TM} LAWPS flowsheet model, and WTP's Dynamic G2 flowsheet model. Information from all three models showed favorable comparison of the waste feeds and secondary liquid wastes against waste compatibility assessment and waste acceptance criteria at multiple stages in the flowsheet. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Actinide Removal Process Modified Flowsheet Demonstration: Filter Only Operation - 17415

January 2016 marked the demonstration of the Actinide Removal Process (ARP) Filter-Only Flowsheet at the Savannah River Site's Liquid Waste Organization operated by Savannah River Remediation, LLC. This ongoing flowsheet demonstration has resulted in a marked increase in throughput over the historical Monosodium Titanate (MST) flowsheet through both increased facility availability and production rate. Not only has the Interim Salt Disposition Program been able to process salt waste more quickly, there has been more processing time between cleaning of the filter elements, which has the added benefit of reducing cesium-137 (Cs-137) loading to the Salt Disposal Facility (SDF). The new flowsheet is a promising improvement to salt waste processing, enhancing the mission of closing Waste Tanks at the Savannah River Site in Aiken, South Carolina. The demonstration of Filter-Only flowsheet has increased salt waste throughput, reduced Cs-137 load to the SDF, and may increase equipment longevity compared to the historical MST flowsheet. The gains in salt volume dispositioned through ARP/MCU benefits the safe, timely, and cost effective closure of legacy waste tanks at the Savannah River Site, reducing the most significant environmental risk in the State of South Carolina. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Mercury Phase II Study - Mercury Behavior in Salt Processing Flowsheet

Mercury (Hg) in the Savannah River Site Liquid Waste System (LWS) originated from decades of canyon processing where it was used as a catalyst for dissolving the aluminum cladding of reactor fuel. Approximately 60 metric tons of mercury is currently present throughout the LWS. Mercury has long been a consideration in the LWS, from both hazard and processing perspectives. In February 2015, a Mercury Program Team was established at the request of the Department of Energy to develop a comprehensive action plan for long-term management and removal of mercury. Evaluation was focused in two Phases. Phase I activities assessed the Liquid Waste inventory and chemical processing behavior using a system-by-system review methodology, and determined the speciation of the different mercury forms (Hg+, Hg++, elemental Hg, organomercury, and soluble versus insoluble mercury) within the LWS. Phase II activities are building on the Phase I activities, and results of the LWS flowsheet evaluations will be summarized in three reports: Mercury Behavior in the Salt Processing Flowsheet (i.e. this report); Mercury Behavior in the Defense Waste Processing Facility (DWPF) Flowsheet; and Mercury behavior in the Tank Farm Flowsheet (Evaporator Operations). The evaluation of the mercury behavior in the salt processing flowsheet indicates, inter alia, the following: (1) In the assembled Salt Batches 7, 8 and 9 in Tank 21, the total mercury is mostly soluble with methylmercury (MHg) contributing over 50% of the total mercury. Based on the analyses of samples from 2H Evaporator feed and drop tanks (Tanks 38/43), the source of MHg in Salt Batches 7, 8 and 9 can be attributed to the 2H evaporator concentrate used in assembling the salt batches. The 2H Evaporator is used to evaporate DWPF recycle water. (2) Comparison of data between Tank 21/49, Salt Solution Feed Tank (SSFT), Decontaminated Salt Solution Hold Tank (DSSHT), and Tank 50 samples suggests that the total mercury as well as speciated forms in the assembled salt batches in Tanks 21/49 pass through the Actinide Removal Process (ARP) / Modular Caustic Side Solvent Extraction Unit (MCU) process to Tank 50 with no significant change in the mercury chemistry. (3) In Tank 50, Decontaminated Salt Solution (DSS) from ARP/MCU is the major contributor to the total mercury including MHg. (4) Speciation analyses of TCLP leached solutions of the grout samples prepared from Tank 21, as well as Tank 50 samples, show the majority of the mercury released in the solution is MHg.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Iodine Flowsheet Status Report

A simplified flowsheet was developed to evaluate the impact of varied assumptions of I-129 decontamination factors on the overall I-129 material balance in the Direct Feed Low-Activity Waste (DFLAW) facility. The flowsheet allows inputs for I-129 feed and Decontamination Factors (DFs) for multiple DFLAW unit operations and provides information on I-129 partitioning to output streams. A series of 6 DFs were evaluated, including baseline DFs from a prior Waste Treatment and Immobilization Plant (WTP) study and refined values based on a Savannah River National Laboratory (SRNL) review. The Microsoft Excel flowsheet spreadsheet will be provided to WRPS to allow further evaluation of the impact of DF assumptions on output streams. The spreadsheet is not specific to iodine, the partitioning of any species can be estimated by entering the assumed DFs for each unit operation. Flowsheet outputs matched results from the prior WTP study and show that different assumptions for the DFs can greatly impact iodine partitioning to primary and secondary waste streams. Results from SRNL's baseline DF values suggest less than 50% of I-129 would be retained in glass versus the 96% when using the BNI DF values. In addition, the flowsheet demonstrated that more than 5 DFLAW operational cycles are required to achieve a steady-state of I-129 to the primary waste stream. It is not possible with the data available to determine whether the BNI approach or the approach taken in this study will be a better match for the iodine partitioning once the LAW facility begins operation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Radioactive Waste Sludge Washing and Demonstration of the Nitric-Glycolic Acid Flowsheet for Sludge Batch 10 Qualification

For each sludge batch that is processed in the Defense Waste Processing Facility (DWPF), the Savannah River National Laboratory (SRNL) performs qualification testing to demonstrate that the sludge batch (SB) is processible. During processing of SB9, DWPF will be transitioning from the Nitric-Formic Acid (NFA) flowsheet to the Nitric-Glycolic Acid (NGA) flowsheet. Thus, the qualification of SB10 was requested to only be performed using the NGA flowsheet. In order to qualify the batch for the NGA flowsheet, Sludge Receipt and Adjustment Tank (SRAT) and Slurry Mix Evaporator (SME) cycles, designated SC-19, were performed using SB10 Tank 51 sample material. SRNL received Tank 51 material in the midst of Tank Farm washing. SRNL continued the washing in the SRNL Shielded Cells. The SRNL process included the addition of Sodium Reactor Experiment (SRE) material from H Canyon, simulating the transfer of SRE from H Canyon to Tank 51 during Tank Farm washing. The washed SB10 Tank 51 material, with SRE, was characterized prior to flowsheet qualification testing.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Interim glycol flowsheet reduction/oxidation (redox) model for the Defense Waste Processing Facility (DWPF)

Control of the REDuction/OXidation (REDOX) state of glasses containing high concentrations of transition metals, such as High Level Waste (HLW) glasses, is critical in order to eliminate processing difficulties caused by overly reduced or overly oxidized melts. Operation of a HLW melter at Fe +2 /ΣFe ratios of between 0.09 and 0.33, a range which is not overly oxidizing or overly reducing, helps retain radionuclides in the melt, i.e. long-lived radioactive 99 Tc species in the less volatile reduced Tc 4+ state, 104 Ru in the melt as reduced Ru +4 state as insoluble RuO 2 , and hazardous volatile Cr 6+ in the less soluble and less volatile Cr +3 state in the glass. The melter REDOX control balances the oxidants and reductants from the feed and from processing additives such as antifoam. Currently, the Defense Waste Processing Facility (DWPF) is running a formic acid-nitric acid (FN) flowsheet where formic acid is the main reductant and nitric acid is the main oxidant. During decomposition formate and formic acid releases H 2 gas which requires close control of the melter vapor space flammability. A switch to a nitric acid-glycolic acid (GN) flowsheet is desired as the glycolic acid flowsheet releases considerably less H 2 gas upon decomposition. This would greatly simplify DWPF processing. Development of an EE term for glycolic acid in the GN flowsheet is documented in this study.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Implementation of flowsheet change to minimize hydrogen and ammonia generation during chemical processing of high level waste in the Defense Waste Processing Facility

Testing was completed to develop a chemical processing flowsheet for the Defense Waste Processing Facility (DWPF), designed to vitrify and stabilize high level radioactive waste. DWPF processing uses a reducing acid (formic acid) and an oxidizing acid (nitric acid) to rheologically thin the slurry and complete the necessary acid base and reduction reactions (primarily mercury and manganese). Formic acid reduces mercuric oxide to elemental mercury, allowing the mercury to be removed during the boiling phase of processing through steam stripping. In runs with active catalysts, formic acid can decompose to hydrogen and nitrate can be reduced to ammonia, both flammable gases, due to rhodium and ruthenium catalysis. Replacement of formic acid with glycolic acid eliminates the generation of rhodium- and ruthenium-catalyzed hydrogen and ammonia. In addition, mercury reduction is still effective with glycolic acid. Hydrogen, ammonia and mercury are discussed in the body of the report. Ten abbreviated tests were completed to develop the operating window for implementation of the flowsheet and determine the impact of changes in acid stoichiometry and the blend of nitric and glycolic acid as it impacts various processing variables over a wide processing region. Three full-length 4-L lab-scale simulations demonstrated the viability of the flowsheet under planned operating conditions. The flowsheet is planned for implementation in early 2017.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Evaluation of quartz melt rate furnace with the nitric-glycolic flowsheet

The Savannah River National Laboratory (SRNL) was tasked to support validation of the Defense Waste Processing Facility (DWPF) melter offgas flammability model for the Nitric-Glycolic (NG) flowsheet. The work is supplemental to the Cold Cap Evaluation Furnace (CEF) testing conducted in 20141 and the Slurry-fed Melt Rate Furnace (SMRF) testing conducted in 20162 that supported Deliverable 4 of the DWPF & Saltstone Facility Engineering Technical Task Request (TTR).3 The Quartz Melt Rate Furnace (QMRF) was evaluated as a bench-scale scoping tool to potentially be used in lieu of or simply prior to the use of the larger-scale SMRF or CEF. The QMRF platform has been used previously to evaluate melt rate behavior and offgas compositions of DWPF glasses prepared from the Nitric-Formic (NF) flowsheet but not for the NG flowsheet and not with continuous feeding.4 The overall objective of the 2016-2017 testing was to evaluate the efficacy of the QMRF as a lab-scale platform for steady state, continuously fed melter testing with the NG flowsheet as an alternative to more expensive and complex testing with the SMRF or CEF platforms.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Flowsheet Definition for the Tank Side Cesium Removal Unit and the Importance of Early Communication - 19142

The U.S. Department of Energy (DOE) tank operations contractor Washington River Protection Solutions, LLC (WRPS) is working toward the start of low-activity waste (LAW) treatment in direct feed mode as soon as 2022. WRPS is responsible for providing the pre-treated feed for this direct feed mode and had originally planned to use the Low Activity Waste Pretreatment System (LAWPS) to supply the initial feed in 2021. In September 2017, in response to a DOE-sponsored LAWPS program review, a new system was proposed to provide the initial feed during direct feed mode until a supplemental pretreatment capability is ready to begin processing feed. This system, known as the Tank Side Cesium Removal (TSCR) system, is based on the Savannah River Tank Closure Cesium Removal (TCCR) system and technology used to separate cesium out of reactor water at Fukushima. A preliminary flowsheet was needed to form the initial basis for future decisions and allow discussions between operations, project management, and the customer. The main goal of the TSCR flowsheet was to form a process flow diagram, process description, and much of the preliminary information to be used to help inform future decisions. When flowsheet development first began, little to no information was readily known and what was known needed to be compiled in one place and digestible by several audiences. The impacts of various parameters on the system also needed a vehicle for clear communication to varying audiences. While a more complicated model was being completed, a model of the process was created using Microsoft Excela which could have several parameters changed while maintaining basic logical requirements (such as treated waste volume cannot exceed the volume of the Interim LAW Storage Tank [ILST], must shutdown for so many days for column changeout, etc.). This program allowed the unit downtime, start dates, treatment rates, and several other variables to be modified and reflect the ILST storage volume as a function of mission progression. The model with several different parameters gathered from the various groups was combined into an early flowsheet of the TSCR process and created one of the first citable documents about TSCR implementation at Hanford. (authors)

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Refining the Approach to Process Flowsheet Development and Maturation at Hanford - 20049

Rigorous process flowsheet management strategies form the basis of an enduring approach to refining the River Protection Project Integrated Process Flowsheet. Incorporation of lessons learned, recognizing upstream and downstream impacts and understanding barriers to implementation of process flowsheets from the outset of the decision-making process will help future revisions to the River Protection Project Integrated Process Flowsheet be more robust, technically defensible, cost effective, and time efficient, ultimately leading to an improved environmental stewardship of the Hanford site. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Simulant Testing of a Co-Decontamination (CoDCon) Flowsheet for a Product with a Controlled Uranium-to-Plutonium Ratio

Multiple tests of a co-decontamination (CoDCon) tributyl phosphate (TBP) based solvent extraction flowsheet are being performed to determine the uncertainties associated with maintaining a target U/Pu ratio in the Pu-containing product. The first two flowsheet tests used simple dissolved fuel simulants containing only U and Pu, but increasingly more complex simulants will but used in subsequent tests, with the ultimate goal of performing tests with irradiated fuel. The flowsheet is designed so as not to create a purified Pu stream, but rather a mixture of U and Pu with a U/Pu mass ratio of 7/3. Uranium and Pu are first co-extracted into the TBP solvent, and then U(IV) is used to strip the Pu from the loaded solvent. A fresh TBP solvent stream is introduced at the same stage the Pu-containing aqueous stream exits. The flowrate of the fresh TBP solvent stream is adjusted so that sufficient U bleeds over into the Pu product to produce the desired U/Pu ratio. The system is monitored in real time using optical spectroscopy, which allows for control of the U/Pu ratio through adjustment of the fresh TBP solvent flowrate. This paper will describe the results of the first two CoDCon flowsheet tests.

CoDCon, fuel recycle, fuel reprocessing, on-line m↗