Heat integrated ethanol dehydration flowsheets

Energy Technology Data Exchange (ETDEWEB)

Hutahaean, L.S.; Shen, W.H.; Brunt, V. Van [Univ. of South Carolina, Columbia, SC (United States)

1995-04-01

zA theoretical evaluation of heat-integrated heterogeneous-azeotropic ethanol-water distillation flowsheets is presented. Simulations of two column flowsheets using several different hydrocarbon entrainers reveal a region of potential heat integration and substantial reduction in operating energy. In this paper, methods for comparing hydrocarbon entrainers are shown. Two aspects of entrainers are related to operating and capital costs. The binary azeotropic composition of the entrainer-ethanol mixture is related to the energy requirements of the flowsheet. A temperature difference in the azeotrophic column is related to the size of the column and overall process staging requirements. Although the hydrophobicity of an entrainer is essential for specification of staging in the dehydration column, no substantial increase in operating energy results from an entrainer that has a higher water content. Likewise, liquid-liquid equilibria between several entrainer-ethanol-water mixtures have no substantial effect on either staging or operation. Rather, increasing the alcohol content of the entrainer-ethanol azeotrope limits its recovery in the dehydration column, and increases the recycle and reflux streams. These effects both contribute to increasing the separation energy requirements and reducing the region of potential heat integration. A cost comparison with a multieffect extractive distillation flowsheet reveals that the costs are comparable; however, the extractive distillation flowsheet is more cost effective as operating costs increase.

Recommendation of ruthenium source for sludge batch flowsheet studies

Energy Technology Data Exchange (ETDEWEB)

Woodham, W. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2017-09-13

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.

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

Energy Technology Data Exchange (ETDEWEB)

Lambert, Dan P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Woodham, Wesley H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Williams, Matthew S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Newell, J. David [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Luther, Michelle C. [Auburn Univ., AL (United States); Brandenburg, Clayton H. [Univ.of South Carolina, Columbia, SC (United States)

2016-09-27

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.

Sludge batch 9 simulant runs using the nitric-glycolic acid flowsheet

Energy Technology Data Exchange (ETDEWEB)

Lambert, D. P. [Savannah River Site (SRS), Aiken, SC (United States); Williams, M. S. [Savannah River Site (SRS), Aiken, SC (United States); Brandenburg, C. H. [Savannah River Site (SRS), Aiken, SC (United States); Luther, M. C. [Savannah River Site (SRS), Aiken, SC (United States); Newell, J. D. [Savannah River Site (SRS), Aiken, SC (United States); Woodham, W. H. [Savannah River Site (SRS), Aiken, SC (United States)

2016-11-01

Testing was completed to develop a Sludge Batch 9 (SB9) nitric-glycolic acid chemical process flowsheet for the Defense Waste Processing Facility’s (DWPF) Chemical Process Cell (CPC). CPC simulations were completed using SB9 sludge simulant, Strip Effluent Feed Tank (SEFT) simulant and Precipitate Reactor Feed Tank (PRFT) simulant. Ten sludge-only Sludge Receipt and Adjustment Tank (SRAT) cycles and four SRAT/Slurry Mix Evaporator (SME) cycles, and one actual SB9 sludge (SRAT/SME cycle) were completed. As has been demonstrated in over 100 simulations, the replacement of formic acid with glycolic acid virtually eliminates the CPC’s largest flammability hazards, hydrogen and ammonia. Recommended processing conditions are summarized in section 3.5.1. Testing demonstrated that the interim chemistry and Reduction/Oxidation (REDOX) equations are sufficient to predict the composition of DWPF SRAT product and SME product. Additional reports will finalize the chemistry and REDOX equations. Additional testing developed an antifoam strategy to minimize the hexamethyldisiloxane (HMDSO) peak at boiling, while controlling foam based on testing with simulant and actual waste. Implementation of the nitric-glycolic acid flowsheet in DWPF is recommended. This flowsheet not only eliminates the hydrogen and ammonia hazards but will lead to shorter processing times, higher elemental mercury recovery, and more concentrated SRAT and SME products. The steady pH profile is expected to provide flexibility in processing the high volume of strip effluent expected once the Salt Waste Processing Facility starts up.

Reprocessing flowsheet and material balance for MEU spent fuel

International Nuclear Information System (INIS)

Abraham, L.

1978-10-01

In response to nonproliferation concerns, the high-temperature gas-cooled reactor (HTGR) Fuel Recycle Development Program is investigating the processing requirements for a denatured medium-enriched uranium--thorium (MEU/Th) fuel cycle. Prior work emphasized the processing requirements for a high-enriched uranium--thorium (HEU/Th) fuel cycle. This report presents reprocessing flowsheets for an HTGR/MEU fuel recycle base case. Material balance data have been calculated for reprocessing of spent MEU and recycle fuels in the HTGR Recycle Reference Facility (HRRF). Flowsheet and mass flow effects in MEU-cycle reprocessing are discussed in comparison with prior HEU-cycle flowsheets

Method for innovative synthesis-design of chemical process flowsheets

DEFF Research Database (Denmark)

Kumar Tula, Anjan; Gani, Rafiqul

Chemical process synthesis-design involve the identification of the processing route to reach a desired product from a specified set of raw materials, design of the operations involved in the processing route, the calculations of utility requirements, the calculations of waste and emission...... to the surrounding and many more. Different methods (knowledge-based [1], mathematical programming [2], hybrid, etc.) have been proposed and are also currently employed to solve these synthesis-design problems. D’ Anterroches [3] proposed a group contribution based approach to solve the synthesis-design problem...... of chemical processes, where, chemical process flowsheets could be synthesized in the same way as atoms or groups of atoms are synthesized to form molecules in computer aided molecular design (CAMD) techniques [4]. That, from a library of building blocks (functional process-groups) and a set of rules to join...

FRACTIONAL CRYSTALLIZATION FLOWSHEET TESTS WITH ACTUAL TANK WASTE

International Nuclear Information System (INIS)

HERTING, D.L.

2006-01-01

Laboratory-scale flowsheet tests of the fractional crystallization process were conducted with actual tank waste samples in a hot cell at the 222-S Laboratory. The process is designed to separate medium-curie liquid waste into a low-curie stream for feeding to supplemental treatment and a high-curie stream for double-shell tank storage. Separations criteria (for Cs-137 sulfate, and sodium) were exceeded in all three of the flowsheet tests that were performed

Hydrogen generation in SRAT with nitric acid and late washing flowsheets

International Nuclear Information System (INIS)

Hsu, C.W.

1992-01-01

Melter feed preparation processes, incorporating a final wash of the precipitate slurry feed to Defense Waste Processing Facility (DWPF) and a partial substitution of the SRAT formic acid requirement with nitric acid, should not produce peak hydrogen generation rates during Cold Chemical Runs (CCR's) and radioactive operation greater than their current, respective hydrogen design bases of 0.024 lb/hr and 1.5 lb/hr. A single SRAT bench-scale process simulation for CCR-s produced a DWPF equivalent peak hydrogen generation rate of 0.004 lb/hr. During radioactive operation, the peak hydrogen generation rate will be dependent on the extent DWPF deviates from the nominal precipitate hydrolysis and melter feed preparation process operating parameters. Two actual radioactive sludges were treated according to the new flowsheets. The peak hydrogen evolution rates were equivalent to 0.038 and 0.20 lb/hr (DWPF scale) respectively. Compared to the formic acid -- HAN hydrolysis flowsheets, these peak rates were reduced by a factor of 2.5 and 3.4 for Tank 15 and Tank 11 sludges, respectively

Nitric-glycolic flowsheet testing for maximum hydrogen generation rate

Energy Technology Data Exchange (ETDEWEB)

Martino, C. J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Newell, J. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Williams, M. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-03-01

The Defense Waste Processing Facility (DWPF) at the Savannah River Site is developing for implementation a flowsheet with a new reductant to replace formic acid. Glycolic acid has been tested over the past several years and found to effectively replace the function of formic acid in the DWPF chemical process. The nitric-glycolic flowsheet reduces mercury, significantly lowers the chemical generation of hydrogen and ammonia, allows purge reduction in the Sludge Receipt and Adjustment Tank (SRAT), stabilizes the pH and chemistry in the SRAT and the Slurry Mix Evaporator (SME), allows for effective adjustment of the SRAT/SME rheology, and is favorable with respect to melter flammability. The objective of this work was to perform DWPF Chemical Process Cell (CPC) testing at conditions that would bound the catalytic hydrogen production for the nitric-glycolic flowsheet.

The development and testing of the new flowsheets for the plutonium purification of the Purex process

Energy Technology Data Exchange (ETDEWEB)

Bugrov, K.V.; Korotaev, V.G.; Korchenkin, K.K.; Logunov, M.V.; Ludin, S.A.; Mashkin, A.N.; Melentev, A.B.; Samarina, N.S. [FSUE ' PAMayak' , Lenin st., 35, Ozersk 456780 (Russian Federation)

2016-07-01

In order to improve the extraction flowsheet of RT-1 Plant two versions of plutonium purification unit flowsheet were developed: a flowsheet with stabilization of Pu(IV)-Np(IV) valence pair and Pu, Np co-recovery, and a flowsheet with stabilization of Pu(IV)-Np(V) valence pair and Pu recovery. The task related to stabilization of the valence pair of the target components in the required state was solved with the use of reactants already applied at RT-1 Plant, namely, hydrogen peroxide, hydrazine nitrate and catalyst (Fe). Both flowsheets were adapted for the plant purification facility with minimum modifications of the equipment, and passed the full scale industrial testing. As a result of this work, reduction in volume and salt content of the raffinate was achieved. (authors)

DWPF nitric-glycolic flowsheet chemical process cell chemistry. Part 1

Energy Technology Data Exchange (ETDEWEB)

Zamecnik, J. R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Edwards, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-02-01

The conversions of nitrite to nitrate, the destruction of glycolate, and the conversion of glycolate to formate and oxalate were modeled for the Nitric-Glycolic flowsheet using data from Chemical Process Cell (CPC) simulant runs conducted by SRNL from 2011 to 2015. The goal of this work was to develop empirical correlations for these variables versus measureable variables from the chemical process so that these quantities could be predicted a-priori from the sludge composition and measurable processing variables. The need for these predictions arises from the need to predict the REDuction/OXidation (REDOX) state of the glass from the Defense Waste Processing Facility (DWPF) melter. This report summarizes the initial work on these correlations based on the aforementioned data. Further refinement of the models as additional data is collected is recommended.

Hot Experimental Facility reference flowsheet

International Nuclear Information System (INIS)

North, E.D.

1982-01-01

This paper is a useful set of background information of HEF flowsheets, although many changes have been made in the past three years. The HEF reference flowsheet is a modified high-acid PUREX flowsheet capable of operating in the coprocessing mode or with full partitioning of U and Pu. Adequate decontamination factors are provided to purify high-burnup, fast breeder-reactor fuels to levels required for recycle back to a fuel fabrication facility. Product streams are mixed U-Pu oxide and uranium oxide. No contaminated liquid wastes are intentionally discharged to the environment. All wastes are solidified and packaged for appropriate disposal. Acid and water are recovered for internal recycle. Excess water is treated and discharged from the plant stack. Several changes have been made in the reference flowsheet since that time, and these are noted briefly

Technetium removal: preliminary flowsheet options

International Nuclear Information System (INIS)

Eager, K.M.

1995-01-01

This document presents the results of a preliminary investigation into options for preliminary flowsheets for 99Tc removal from Hanford Site tank waste. A model is created to show the path of 99Tc through pretreatment to disposal. The Tank Waste Remediation (TWRS) flowsheet (Orme 1995) is used as a baseline. Ranges of important inputs to the model are developed, such as 99Tc inventory in the tanks and important splits through the TWRS flowsheet. Several technetium removal options are discussed along with sensitivities of the removal schemes to important model parameters

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

Energy Technology Data Exchange (ETDEWEB)

Law, Jack Douglas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Soelberg, Nicholas Ray [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-06-17

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

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

International Nuclear Information System (INIS)

Law, Jack Douglas; Soelberg, Nicholas Ray

2015-01-01

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

GLYCOLIC-NITRIC ACID FLOWSHEET DEMONSTRATION OF THE DWPF CHEMICAL PROCESS CELL WITH SLUDGE AND SUPERNATE SIMULANTS

Energy Technology Data Exchange (ETDEWEB)

Lambert, D.; Stone, M.; Newell, J.; Best, D.; Zamecnik, J.

2012-08-28

Savannah River Remediation (SRR) is evaluating changes to its current Defense Waste Processing Facility (DWPF) flowsheet to improve processing cycle times. This will enable the facility to support higher canister production while maximizing waste loading. Higher throughput is needed in the Chemical Process Cell (CPC) since the installation of the bubblers into the melter has increased melt rate. Due to the significant maintenance required for the DWPF gas chromatographs (GC) and the potential for production of flammable quantities of hydrogen, reducing or eliminating the amount of formic acid used in the CPC is being developed. Earlier work at Savannah River National Laboratory has shown that replacing formic acid with an 80:20 molar blend of glycolic and formic acids has the potential to remove mercury in the SRAT without any significant catalytic hydrogen generation. This report summarizes the research completed to determine the feasibility of processing without formic acid. In earlier development of the glycolic-formic acid flowsheet, one run (GF8) was completed without formic acid. It is of particular interest that mercury was successfully removed in GF8, no formic acid at 125% stoichiometry. Glycolic acid did not show the ability to reduce mercury to elemental mercury in initial screening studies, which is why previous testing focused on using the formic/glycolic blend. The objective of the testing detailed in this document is to determine the viability of the nitric-glycolic acid flowsheet in processing sludge over a wide compositional range as requested by DWPF. This work was performed under the guidance of Task Technical and Quality Assurance Plan (TT&QAP). The details regarding the simulant preparation and analysis have been documented previously.

Hanford Low-Activity Waste Processing: Demonstration of the Off-Gas Recycle Flowsheet - 13443

Energy Technology Data Exchange (ETDEWEB)

Ramsey, William G.; Esparza, Brian P. [Washington River Protection Solutions, LLC, Richland, WA 99532 (United States)

2013-07-01

Vitrification of Hanford Low-Activity Waste (LAW) is nominally the thermal conversion and incorporation of sodium salts and radionuclides into borosilicate glass. One key radionuclide present in LAW is technetium-99. Technetium-99 is a low energy, long-lived beta emitting radionuclide present in the waste feed in concentrations on the order of 1-10 ppm. The long half-life combined with a high solubility in groundwater results in technetium-99 having considerable impact on performance modeling (as potential release to the environment) of both the waste glass and associated secondary waste products. The current Hanford Tank Waste Treatment and Immobilization Plant (WTP) process flowsheet calls for the recycle of vitrification process off-gas condensates to maximize the portion of technetium ultimately immobilized in the waste glass. This is required as technetium acts as a semi-volatile specie, i.e. considerable loss of the radionuclide to the process off-gas stream can occur during the vitrification process. To test the process flowsheet assumptions, a prototypic off-gas system with recycle capability was added to a laboratory melter (on the order of 1/200 scale) and testing performed. Key test goals included determination of the process mass balance for technetium, a non-radioactive surrogate (rhenium), and other soluble species (sulfate, halides, etc.) which are concentrated by recycling off-gas condensates. The studies performed are the initial demonstrations of process recycle for this type of liquid-fed melter system. This paper describes the process recycle system, the waste feeds processed, and experimental results. Comparisons between data gathered using process recycle and previous single pass melter testing as well as mathematical modeling simulations are also provided. (authors)

Preliminary flowsheet: Ion exchange for separation of cesium from Hanford tank waste using resorcinol-formaldehyde resin

International Nuclear Information System (INIS)

Penwell, D.L.

1994-01-01

This preliminary flowsheet document describes an ion exchange process which uses resorcinol-formaldehyde (R-F) resin to remove cesium from Hanford tank waste. The flowsheet describes one possible equipment configuration, and contains mass balances based on that configuration with feeds of Neutralized Current Acid Waste, and Double Shell Slurry Feed. The flowsheet also discusses process alternatives, unresolved issues, and development needs associated with the ion exchange process. It is expected that this flowsheet will evolve as open issues are resolved and progress is made on development needs. This is part of the Tank Waste Remediation Program at Hanford. 26 refs, 6 figs, 25 tabs

Defense Waste Processing Facility Nitric- Glycolic Flowsheet Chemical Process Cell Chemistry: Part 2

Energy Technology Data Exchange (ETDEWEB)

Zamecnik, J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Edwards, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2017-06-06

The conversions of nitrite to nitrate, the destruction of glycolate, and the conversion of glycolate to formate and oxalate were modeled for the Nitric-Glycolic flowsheet using data from Chemical Process Cell (CPC) simulant runs conducted by Savannah River National Laboratory (SRNL) from 2011 to 2016. The goal of this work was to develop empirical correlation models to predict these values from measureable variables from the chemical process so that these quantities could be predicted a-priori from the sludge or simulant composition and measurable processing variables. The need for these predictions arises from the need to predict the REDuction/OXidation (REDOX) state of the glass from the Defense Waste Processing Facility (DWPF) melter. This report summarizes the work on these correlations based on the aforementioned data. Previous work on these correlations was documented in a technical report covering data from 2011-2015. This current report supersedes this previous report. Further refinement of the models as additional data are collected is recommended.

A technetium rejection flowsheet

International Nuclear Information System (INIS)

Baker, R.; Miles, J.H.; Roberts, P.T.

1990-01-01

A single contactor unit has been designed which enables Tc to be removed from a TBP/diluent stream bearing U and Pu, by means of a 5M HNO 3 wash. A Tc waste stream is produced which is virtually free from U and Pu. The flowsheet has been tested firstly with U and Tc, then with U, Pu and Tc, and finally in a highly active facility with real PWR fuel solution. About 97% of the Tc was removed from the organic phase and U and Pu levels in the Tc waste have usually been below 0.04% of those in the product stream. (author)

GLYCOLIC-NITRIC ACID FLOWSHEET DEMONSTRATION OF THE DWPF CHEMICAL PROCESSING CELL WITH MATRIX SIMULANTS AND SUPERNATE

Energy Technology Data Exchange (ETDEWEB)

Lambert, D.; Stone, M.; Newell, J.; Best, D.

2012-05-07

Savannah River Remediation (SRR) is evaluating changes to its current DWPF flowsheet to improve processing cycle times. This will enable the facility to support higher canister production while maximizing waste loading. Higher throughput is needed in the CPC since the installation of the bubblers into the melter has increased melt rate. Due to the significant maintenance required for the DWPF gas chromatographs (GC) and the potential for production of flammable quantities of hydrogen, reducing or eliminating the amount of formic acid used in the CPC is being developed. Earlier work at Savannah River National Laboratory has shown that replacing formic acid with an 80:20 molar blend of glycolic and formic acids has the potential to remove mercury in the SRAT without any significant catalytic hydrogen generation. This report summarizes the research completed to determine the feasibility of processing without formic acid. In earlier development of the glycolic-formic acid flowsheet, one run (GF8) was completed without formic acid. It is of particular interest that mercury was successfully removed in GF8, no formic acid at 125% stoichiometry. Glycolic acid did not show the ability to reduce mercury to elemental mercury in initial screening studies, which is why previous testing focused on using the formic/glycolic blend. The objective of the testing detailed in this document is to determine the viability of the nitric-glycolic acid flowsheet in processing sludge over a wide compositional range as requested by DWPF. This work was performed under the guidance of Task Technical and Quality Assurance Plan (TT and QAP). The details regarding the simulant preparation and analysis have been documented previously.

Mixed Waste Treatment Project: Computer simulations of integrated flowsheets

International Nuclear Information System (INIS)

Dietsche, L.J.

1993-12-01

The disposal of mixed waste, that is waste containing both hazardous and radioactive components, is a challenging waste management problem of particular concern to DOE sites throughout the United States. Traditional technologies used for the destruction of hazardous wastes need to be re-evaluated for their ability to handle mixed wastes, and in some cases new technologies need to be developed. The Mixed Waste Treatment Project (MWTP) was set up by DOE's Waste Operations Program (EM30) to provide guidance on mixed waste treatment options. One of MWTP's charters is to develop flowsheets for prototype integrated mixed waste treatment facilities which can serve as models for sites developing their own treatment strategies. Evaluation of these flowsheets is being facilitated through the use of computer modelling. The objective of the flowsheet simulations is to provide mass and energy balances, product compositions, and equipment sizing (leading to cost) information. The modelled flowsheets need to be easily modified to examine how alternative technologies and varying feed streams effect the overall integrated process. One such commercially available simulation program is ASPEN PLUS. This report contains details of the Aspen Plus program

A new flowsheeting tool for flue gas treating

NARCIS (Netherlands)

van Elk, E. P.; Arendsen, A. R. J.; Versteeg, G. F.

2009-01-01

A new flowsheeting tool, specifically designed for steady-state simulation of acid gas treating processes, has been developed. The models implemented in the new tool combine all issues relevant for the design, optimization and analysis of acid gas treating processes, including post-combustion and

Report on the flowsheet model for the electrochemical treatment of liquid radioactive wastes

International Nuclear Information System (INIS)

Hobbs, D.T.

1995-01-01

The objective of this report is to describe the modeling and optimization procedure for the electrochemical removal of nitrates and nitrites from low level radioactive wastes. The simulation is carried out in SPEEDUP trademark, which is a state of the art flowsheet modeling package. The flowsheet model will provide a better understanding of the process and aid in the scale-up of the system. For example, the flowsheet model has shown that the electrochemical cell must be operated in batch mode to achieve 95% destruction. The present status of the flowsheet model is detailed in this report along with a systematic description of the batch optimization of the electrochemical cell. Results from two batch runs and one optimization run are also presented

Modeling and flowsheet design of an Am separation process using TODGA and H{sub 4}TPAEN

Energy Technology Data Exchange (ETDEWEB)

Vanel, V.; Marie, C.; Montuir, M.; Boubals, N.; Sorel, C. [CEA, Centre de Marcoule, Nuclear Energy Division, Radiochemistry and Processes Department, Bagnols-sur-Ceze, F-30207 (France); Kaufholz, P.; Modolo, G. [Forschungszentrum Juelich GmbH, Institute of Energy and Climate Research, Nuclear Waste Management and Reactor Safety, D-52428 (Germany); Geist, A. [Karlsruher Institut fuer Technologie - KIT, Institut fuer Nukleare Entsorgung - INE, Karlsruhe (Germany)

2016-07-01

Recycling americium from spent fuels is an important consideration for the future nuclear fuel cycle, as americium is the main contributor to the long-term radiotoxicity and heat power of the final waste, after separation of uranium and plutonium using the PUREX process. The separation of americium alone from a PUREX raffinate can be achieved by co-extracting lanthanide (Ln(III)) and actinide (An(III)) cations into an organic phase containing the diglycolamide extractant TODGA, and then stripping Am(III) with selectivity towards Cm(III) and lanthanides. The water soluble ligand H{sub 4}TPAEN was tested to selectively strip Am from a loaded organic phase. Based on experimental data obtained by Juelich, NNL and CEA laboratories since 2013, a phenomenological model has been developed to simulate the behavior of americium, curium and lanthanides during their extraction by TODGA and their complexation by H{sub 4}TPAEN (complex stoichiometry, extraction and complexation constants, kinetics). The model was gradually implemented in the PAREX code and helped to narrow down the best operating conditions. Thus, the following 2 modifications of initial operating conditions were proposed: -) an increase in the concentration of TPAEN as much as the solubility limit allows, and -) an improvement of the lanthanide scrubbing from the americium flow by adding nitrates to the aqueous phase. A qualification of the model was begun by comparing on the one hand constants determined with the model to those measured experimentally, and on the other hand, simulation results and experimental data on new independent batch experiments. A first sensitivity analysis identified which parameter has the most dominant effect on the process. A flowsheet was proposed for a spiked test in centrifugal contactors performed with a simulated PUREX raffinate with trace amounts of Am and Cm. If the feasibility of the process is confirmed, the results of this test will be used to consolidate the model and to

Impact of Salt Waste Processing Facility Streams on the Nitric-Glycolic Flowsheet in the Chemical Processing Cell

Energy Technology Data Exchange (ETDEWEB)

Martino, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2017-08-08

An evaluation of the previous Chemical Processing Cell (CPC) testing was performed to determine whether the planned concurrent operation, or “coupled” operations, of the Defense Waste Processing Facility (DWPF) with the Salt Waste Processing Facility (SWPF) has been adequately covered. Tests with the nitricglycolic acid flowsheet, which were both coupled and uncoupled with salt waste streams, included several tests that required extended boiling times. This report provides the evaluation of previous testing and the testing recommendation requested by Savannah River Remediation. The focus of the evaluation was impact on flammability in CPC vessels (i.e., hydrogen generation rate, SWPF solvent components, antifoam degradation products) and processing impacts (i.e., acid window, melter feed target, rheological properties, antifoam requirements, and chemical composition).

FY13 GLYCOLIC-NITRIC ACID FLOWSHEET DEMONSTRATIONS OF THE DWPF CHEMICAL PROCESS CELL WITH SIMULANTS

Energy Technology Data Exchange (ETDEWEB)

Lambert, D.; Zamecnik, J.; Best, D.

2014-03-13

Savannah River Remediation is evaluating changes to its current Defense Waste Processing Facility flowsheet to replace formic acid with glycolic acid in order to improve processing cycle times and decrease by approximately 100x the production of hydrogen, a potentially flammable gas. Higher throughput is needed in the Chemical Processing Cell since the installation of the bubblers into the melter has increased melt rate. Due to the significant maintenance required for the safety significant gas chromatographs and the potential for production of flammable quantities of hydrogen, eliminating the use of formic acid is highly desirable. Previous testing at the Savannah River National Laboratory has shown that replacing formic acid with glycolic acid allows the reduction and removal of mercury without significant catalytic hydrogen generation. Five back-to-back Sludge Receipt and Adjustment Tank (SRAT) cycles and four back-to-back Slurry Mix Evaporator (SME) cycles were successful in demonstrating the viability of the nitric/glycolic acid flowsheet. The testing was completed in FY13 to determine the impact of process heels (approximately 25% of the material is left behind after transfers). In addition, back-to-back experiments might identify longer-term processing problems. The testing was designed to be prototypic by including sludge simulant, Actinide Removal Product simulant, nitric acid, glycolic acid, and Strip Effluent simulant containing Next Generation Solvent in the SRAT processing and SRAT product simulant, decontamination frit slurry, and process frit slurry in the SME processing. A heel was produced in the first cycle and each subsequent cycle utilized the remaining heel from the previous cycle. Lower SRAT purges were utilized due to the low hydrogen generation. Design basis addition rates and boilup rates were used so the processing time was shorter than current processing rates.

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

Energy Technology Data Exchange (ETDEWEB)

Williams, M. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Miller, D. H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2017-08-03

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.

Integration of SWPF into the DWPF Flowsheet: Gap Analysis and Test Matrix Development

Energy Technology Data Exchange (ETDEWEB)

Peeler, D. K. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Edwards, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2014-12-10

Based on Revision 19 of the High Level Waste (HLW) System Plan, it is anticipated that the Salt Waste Processing Facility (SWPF) will be integrated into the Defense Waste Processing Facility (DWPF) flowsheet in October 2018 (or with Sludge Batch 11 (SB11)). Given that, Savannah River Remediation (SRR) has requested a technical basis be developed that validates the current Product Composition Control System (PCCS) models for use during the processing of the SWPF-based coupled flowsheet or that leads to the refinements of or modifications to the models that are needed so that the models may be used during the processing of the SWPF-based coupled flowsheet. To support this objective, Savannah River National Laboratory (SRNL) has completed three key interim activities prior to validation of the current or development of refined PCCS models over the anticipated glass composition region for SWPF processing. These three key activities include: (1) defining the glass compositional region over which SWPF is anticipated to be processed, (2) comparing the current PCCS model validation ranges to the SWPF glass compositional region from which compositional gaps can be identified, and (3) developing a test matrix to cover the compositional gaps.

Baseline Flowsheet Generation for the Treatment and Disposal of Idaho National Engineering and Environmental Laboratory Sodium Bearing Waste

International Nuclear Information System (INIS)

Barnes, C.M.; Lauerhass, L.; Olson, A.L.; Taylor, D.D.; Valentine, J.H.; Lockie, K.A.

2002-01-01

The High-Level Waste (HLW) Program at the Idaho National Engineering and Environmental Laboratory (INEEL) must implement technologies and processes to treat and qualify radioactive wastes located at the Idaho Nuclear Technology and Engineering Center (INTEC) for permanent disposal. This paper describes the approach and accomplishments to date for completing development of a baseline vitrification treatment flowsheet for sodium-bearing waste (SBW), including development of a relational database used to manage the associated process assumptions. A process baseline has been developed that includes process requirements, basis and assumptions, process flow diagrams, a process description, and a mass balance. In the absence of actual process or experimental results, mass and energy balance data for certain process steps are based on assumptions. Identification, documentation, validation, and overall management of the flowsheet assumptions are critical to ensuring an integrated, focused program. The INEEL HLW Program initially used a roadmapping methodology, developed through the INEEL Environmental Management Integration Program, to identify, document, and assess the uncertainty and risk associated with the SBW flowsheet process assumptions. However, the mass balance assumptions, process configuration and requirements should be accessible to all program participants. This need resulted in the creation of a relational database that provides formal documentation and tracking of the programmatic uncertainties related to the SBW flowsheet

First-cycle studies of coprocessing flowsheets

International Nuclear Information System (INIS)

Gray, J.H.

1981-06-01

Selected portions of two coprocessing flowsheets developed for use at the Barnwell Nuclear Fuel Plant (BNFP) have been tested in the laboratory with uranium, plutonium, and fission products. Processing conditions and stream compositions for first cycle extraction and uranium-plutonium partitioning in an electropulse column were controlled to examine the behavior of nitric acid, uranium, plutonium, and fission products during coprocessing. The ability to adapt coprocessing technology for use in the BNFP reprocessing facility was successful for first cycle extraction and partition. The only process adjustment involved a reduction in nitric acid concentration to attain proper uranium to plutonium ratios

Development and Testing of an Americium/Lanthanide Separation Flowsheet Using Sodium Bismuthate

Energy Technology Data Exchange (ETDEWEB)

Jack Law; Bruce Mincher; Troy Garn; Mitchell Greenhalgh; Nicholas Schmitt; Veronica Rutledge

2014-04-01

The separation of Am from the lanthanides and curium is a key step in proposed advanced fuel cycle scenarios. The partitioning and transmutation of Am is desirable to minimize the long-term heat load of material interred in a future high-level waste repository. A separation process amenable to process scale-up remains elusive. Given only subtle chemistry differences within and between the ions of the trivalent actinide and lanthanide series this separation is challenging ; however, higher oxidation states of americium can be prepared using sodium bismuthate and separated via solvent extraction using diamylamylphosphonate (DAAP) extraction. Among the other trivalent metals only Ce is also oxidized and extracted. Due to the long-term instability of Am(VI) , the loaded organic phase is readily selectively stripped to partition the actinide to a new acidic aqueous phase. Batch extraction distribution ratio measurements were used to design a flowsheet to accomplish this separation. Additionally, crossflow filtration was investigated as a method to filter the bismuthate solids from the feed solution prior to extraction. Results of the filtration studies, flowsheet development work and flowsheet performance testing using a centrifugal contactor are detailed.

Design of preconcentration flow-sheet for processing Bhimunipatnam beach sands using pilot plant experiments and computer simulation

International Nuclear Information System (INIS)

Padmanabhan, N.P.H.; Sridhar, U.

1993-01-01

Simulation was carried out using a beach sand beneficiation plant simulator software, SANDBEN, currently being developed in Indian School of Mines, Dhanbad, and the results were compared and analyzed with those obtained by actual pilot plant experiments on a beach sand sample from Bhimunipatnam deposit. The software is discussed and its capabilities and limitations are highlighted. An optimal preconcentrator flow-sheet for processing Bhimunipatnam beach sand was developed by simulation and using the results of the pilot plant experiments. (author). 13 refs., 2 tabs., 3 figs

Processing flowsheet for the accelerator transmutation of waste (ATW) program

International Nuclear Information System (INIS)

Dewey, H.; Walker, R.; Yarbro, S.

1992-01-01

At Los Alamos, an innovative approach to transmuting long-lived radioactive waste is under investigation. The concept is to use a linear proton accelerator coupled to a solid target to produce an intense neutron flux. The intense stream of neutrons can then be used to fission or transmute long-lived radionuclides to either stable or shorter-lived isotopes. For the program to be successful, robust chemical separations with high efficiencies (>10 5 ) are required. The actual mission, either defense or commercial, will determine what suite of unit operations will be needed. If the mission is to process commercial spent fuel, there are several options available for feed preparation and blanket processing. The baseline option would be an improved PUREX system with the main alternative being the current ATW actinide blanket processing flowsheet. 99 Tc and 129 I are more likely to reach the biosphere than the actinides. Many models have been developed for predicting how the radionuclides will behave in a repository over long time periods. The general conclusion is that the actinides will be sorbed by the soil. Therefore, over a long time period, e.g., a million years their hazard will be lessened because of radioactive decay and dispersion. However, some of the long-lived fission products are not sorbed and could potentially reach the environment over a few thousand year period. Hence, they could present a significant safety hazard. Because of limited resources, most of the priority has been focused on the actinide and technetium blanket assemblies

Effects of solvent-extraction contactor selection on flowsheet and facility design

International Nuclear Information System (INIS)

Whatley, M.E.

1982-01-01

The notion is developed that the selection of a solvent extraction contactor is part of a more general development of principles and philosophy guiding the overall plant design. Specifically, the requirements and constraints placed on the plant by the solvent extraction system must be consistent with those imposed by the other operations, which generally are more expensive and more complicated. Were a conservative philosophy employed throughout the plant, the choice of pulsed columns seem correct. Were the plant intended to employ modern techniques and state-of-the-art technology, particularly in remote maintenance and process control, the selection of centrifugal contactors seems appropriate. The process improvements attainable from employing more stages in a more tightly controlled solvent extraction system seem marginal at present when applied to conventional flowsheets, although the cost-benefit may be attractive in a modern plant. The potential for improvement through major flowsheet modification can not presently be assessed quantitatively

Preliminary flowsheet for the conversion of Hanford high-level waste to glass

International Nuclear Information System (INIS)

Beary, M.M.; Chick, L.A.; Ely, P.C.; Gott, S.A.

1977-06-01

The flowsheets describe a process for converting waste removed from the Hanford underground waste tanks to more immobile form. The process involves a chemical separation of the radionuclides from industrial chemicals, and then making glass from the resulting small volume of highly radioactive waste. Removal of Sr, actinides, cesium, and technetium is discussed

Low temperature dissolution flowsheet for plutonium metal

Energy Technology Data Exchange (ETDEWEB)

Daniel, W. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Almond, P. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Rudisill, T. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-05-01

The H-Canyon flowsheet used to dissolve Pu metal for PuO₂ production utilizes boiling HNO₃. SRNL was requested to develop a complementary dissolution flowsheet at two reduced temperature ranges. The dissolution and H₂ generation rates of Pu metal were investigated using a dissolving solution at ambient temperature (20-30 °C) and for an intermediate temperature of 50-60 °C. Additionally, the testing included an investigation of the dissolution rates and characterization of the off-gas generated from the ambient temperature dissolution of carbon steel cans and the nylon bags that contain the Pu metal when charged to the dissolver.

HYBRID SULFUR CYCLE FLOWSHEETS FOR HYDROGEN PRODUCTION USING HIGH-TEMPERATURE GAS-COOLED REACTORS

Energy Technology Data Exchange (ETDEWEB)

Gorensek, M.

2011-07-06

Two hybrid sulfur (HyS) cycle process flowsheets intended for use with high-temperature gas-cooled reactors (HTGRs) are presented. The flowsheets were developed for the Next Generation Nuclear Plant (NGNP) program, and couple a proton exchange membrane (PEM) electrolyzer for the SO2-depolarized electrolysis step with a silicon carbide bayonet reactor for the high-temperature decomposition step. One presumes an HTGR reactor outlet temperature (ROT) of 950 C, the other 750 C. Performance was improved (over earlier flowsheets) by assuming that use of a more acid-tolerant PEM, like acid-doped poly[2,2'-(m-phenylene)-5,5'-bibenzimidazole] (PBI), instead of Nafion{reg_sign}, would allow higher anolyte acid concentrations. Lower ROT was accommodated by adding a direct contact exchange/quench column upstream from the bayonet reactor and dropping the decomposition pressure. Aspen Plus was used to develop material and energy balances. A net thermal efficiency of 44.0% to 47.6%, higher heating value basis is projected for the 950 C case, dropping to 39.9% for the 750 C case.

Towards an optimized flow-sheet for a SANEX demonstration process using centrifugal contactors

International Nuclear Information System (INIS)

Magnusson, D.; Christiansen, B.; Glatz, J.P.; Malmbeck, R.; Serrano-Purroy, D.; Modolo, G.; Sorel, C.

2008-01-01

The design of an efficient process flow-sheet requires accurate extraction data for the experimental set-up used. Often this data is provided as equilibrium data. Due to the small hold-up volume compared to the flow rate in centrifugal contactors the time for extraction is often too short to reach the equilibrium D-ratios. In this work single stage kinetics experiments have been carried out to investigate the D-ratio dependence of the flow rate and also to compare with equilibrium batch experiments for CyMe 4 - BTBP. The first centrifuge experiment was run with spiked solutions while in the second a genuine actinide/lanthanide fraction from a TODGA process was used. Three different flow rates were tested with each set-up. The results show that even with low flow rates, around 8% of the equilibrium D-ratio (Am) was reached for the extraction in the spiked test and around 16% in the hot test (the difference is due to the size of the centrifuges). The general conclusion is that the development of a process flow sheet needs investigation of the kinetic behaviour in the actual equipment used. (authors)

Decontamination flowsheet development for a waste oil containing mixed radioactive contaminants

International Nuclear Information System (INIS)

Vijayan, S.; Buckley, L.P.

1993-01-01

The majority of waste oils contaminated with both radioactive and hazardous components are generated in nuclear power plant, research lab. and uranium-refinery operations. The waste oils are complex, requiring a detailed examination of the waste management strategies and technology options. It may appear that incineration offers a total solution, but this may not be true in all cases. An alternative approach is to decontaminate the waste oils to very low contaminant levels, so that the treated oils can be reused, burned as fuel in boilers, or disposed of by commercial incineration. This paper presents selected experimental data and evaluation results gathered during the development of a decontamination flowsheet for a specific waste oil stores at Chalk River Labs. (CRL). The waste oil contains varying amounts of lube oils, grease, paint, water, particulates, sludge, light chloro- and fluoro-solvents, polychlorinated biphenyls (PCB), complexing chemicals, uranium, chromium, iron, arsenic and manganese. To achieve safe management of this radioactive and hazardous waste, several treatment and disposal methods were screened. Key experiments were performed at the laboratory-scale to confirm and select the most appropriate waste-management scheme based on technical, environmental and economic criteria. The waste-oil-decontamination flowsheet uses a combination of unit operations, including prefiltration, acid scrubbing, and aqueous-leachage treatment by precipitation, microfiltration, filter pressing and carbon adsorption. The decontaminated oil containing open-quotes de minimisclose quotes levels of contaminants will undergo chemical destruction of PCBs and final disposal by incineration. The recovered uranium will be recycled to a uranium milling process

Flowsheet development for HTGR fuel reprocessing

International Nuclear Information System (INIS)

Baxter, B.; Benedict, G.E.; Zimmerman, R.D.

1976-01-01

Development studies to date indicate that the HTGR fuel blocks can be effectively crushed with two stages of eccentric jaw crushing, followed by a double-roll crusher, a screener and an eccentrically mounted single-roll crusher for oversize particles. Burner development results indicate successful long-term operation of both the primary and secondary fluidized-bed combustion systems can be performed with the equipment developed in this program. Aqueous separation development activities have centered on adapting known Acid-Thorex processing technology to the HTGR reprocessing task. Significant progress has been made on dissolution of burner ash, solvent extraction feed preparation, slurry transfer, solids drying and solvent extraction equipment and flowsheet requirements

Significance of mineralogy in the development of flowsheets for processing uranium ores

International Nuclear Information System (INIS)

1980-01-01

This report has been prepared from material developed at and subsequent to a consultants' meeting held in Vienna in January 1978. The main purpose of the meeting was to prepare a document in the form of a guide for planning and developing treatment flowsheets for uranium ore processing. It was apparent that ore mineralogy, analysed, described and interpreted in ways most meaningful to the metallurgist, is the most essential information required for forming the basis of such planning. This topic, here termed metallurgical mineralogy, is therefore a major theme of this publication. In preparing the report the Agency has borne in mind the important need to impart the experience and knowledge gained in the more developed countries to those who are in the early stages of exploiting their uranium resources. The contents may be criticized as lacking, in some respects, the requisite depth and detail of treatment. The Agency and the consultants are conscious of the need to expand the information in a number of ways. However, the report is presented in its present form in the belief that, as the first attempt to correlate, on a world-wide basis, ore type with processing, it will be considered as a useful basis for future development of these themes

DWPF Flowsheet Studies with Simulants to Determine Modular Caustic Side Solvent Extraction Unit Solvent Partitioning and Verify Actinide Removal Process Incorporation Strategy

International Nuclear Information System (INIS)

Herman, C

2006-01-01

The Actinide Removal Process (ARP) facility and the Modular Caustic Side Solvent Extraction Unit (MCU) are scheduled to begin processing salt waste in fiscal year 2007. A portion of the streams generated in the salt processing facilities will be transferred to the Defense Waste Processing Facility (DWPF) to be incorporated in the glass matrix. Before the streams are introduced, a combination of impact analyses and research and development studies must be performed to quantify the impacts on DWPF processing. The Process Science and Engineering (PS and E) section of the Savannah River National Laboratory (SRNL) was requested via Technical Task Request (TTR) HLW/DWPF/TTR-2004-0031 to evaluate the impacts on DWPF processing. Simulant Chemical Process Cell (CPC) flowsheet studies have been performed using previous composition and projected volume estimates for the ARP sludge/monosodium titanate (MST) stream. Due to changes in the flammability control strategy for DWPF for salt processing, the incorporation strategy for ARP has changed and additional ARP flowsheet tests were necessary to validate the new processing strategy. The last round of ARP testing included the incorporation of the MCU stream and identified potential processing issues with the MCU solvent. The identified issues included the potential carry-over and accumulation of the MCU solvent components in the CPC condensers and in the recycle stream to the Tank Farm. Therefore, DWPF requested SRNL to perform additional MCU flowsheet studies to better quantify the organic distribution in the CPC vessels. The previous MCU testing used a Sludge Batch 4 (SB4) simulant since it was anticipated that both of these facilities would begin salt processing during SB4 processing. The same sludge simulant recipe was used in this round of ARP and MCU testing to minimize the number of changes between the two phases of testing so a better comparison could be made. ARP and MCU stream simulants were made for this phase of

Spent Nuclear Fuel Reprocessing Flowsheet. A Report by the WPFC Expert Group on Chemical Partitioning of the NEA Nuclear Science Committee

International Nuclear Information System (INIS)

Na, Chan; Yamagishi, Isao; Choi, Yong-Joon; Glatz, Jean-Paul; Hyland, Bronwyn; Uhlir, Jan; Baron, Pascal; Warin, Dominique; De Angelis, Giorgio; Luce, Alfredo; INOUE, Tadashi; Morita, Yasuji; Minato, Kazuo; Lee, Han Soo; Ignatiev, Victor V.; Kormilitsyn, Mikhail V.; Caravaca, Concepcion; Lewin, Robert G.; Taylor, Robin J.; Collins, Emory D.; Laidler, James J.

2012-06-01

Under the auspices of the NEA Nuclear Science Committee (NSC), the Working Party on Scientific Issues of the Fuel Cycle (WPFC) has been established to co-ordinate scientific activities regarding various existing and advanced nuclear fuel cycles, including advanced reactor systems, associated chemistry and flowsheets, development and performance of fuel and materials, and accelerators and spallation targets. The WPFC has different expert groups to cover a wide range of scientific fields in the nuclear fuel cycle. The Expert Group on Chemical Partitioning was created in 2001 to (1) perform a thorough technical assessment of separations processes in application to a broad set of partitioning and transmutation (P and T) operating scenarios and (2) identify important research, development and demonstration necessary to bring preferred technologies to a deployable stage and (3) recommend collaborative international efforts to further technological development. This report aims to collect spent nuclear fuel reprocessing flowsheet of various processes developed by member states: aqueous, pyro and fluoride volatility. Contents: 1 - Hydrometallurgy process: Standard PUREX, Extended PUREX, UREX+3, Grind/Leach; 2 - Pyrometallurgy process: pyro-process (CRIEPI - Japan), 4-group partitioning process, pyro-process (KAERI - Korea), Direct electrochemical processing of metallic fuel, PyroGreen (reduce radiotoxicity to the level of low and intermediate level waste - LILW); 3 - Fluoride volatility process: Fluoride volatility process, Uranium and protactinium removal from fuel salt compositions by fluorine bubbling, Flowsheet studies on non-aqueous reprocessing of LWR/FBR spent nuclear fuel; Appendix A: Flowsheet studies of RIAR (Russian Federation), List of contributors, Members of the expert group

Actual waste demonstration of the nitric-glycolic flowsheet for sludge batch 9 qualification

Energy Technology Data Exchange (ETDEWEB)

Newell, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Pareizs, J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Martino, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Reboul, S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Coleman, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Edwards, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Johnson, F. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2017-03-09

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 is processable. Based on the results of this actual-waste qualification and previous simulant studies, SRNL recommends implementation of the nitric-glycolic acid flowsheet in DWPF. Other recommendations resulting from this demonstration are reported in section 5.0.

Process synthesis, design and analysis using a process-group contribution method

DEFF Research Database (Denmark)

Kumar Tula, Anjan; Eden, Mario R.; Gani, Rafiqul

2015-01-01

) techniques. The fundamental pillars of this framework are the definition and use of functional process-groups (building blocks) representing a wide range of process operations, flowsheet connectivity rules to join the process-groups to generate all the feasible flowsheet alternatives and flowsheet property...... models like energy consumption, atom efficiency, environmental impact to evaluate the performance of the generated alternatives. In this way, a list of feasible flowsheets are quickly generated, screened and selected for further analysis. Since the flowsheet is synthesized and the operations......This paper describes the development and application of a process-group contribution method to model, simulate and synthesize chemical processes. Process flowsheets are generated in the same way as atoms or groups of atoms are combined to form molecules in computer aided molecular design (CAMD...

Impact of scaling on the nitric-glycolic acid flowsheet

Energy Technology Data Exchange (ETDEWEB)

Lambert, D. [Savannah River Site (SRS), Aiken, SC (United States)

2016-02-01

Savannah River Remediation (SRR) is considering using glycolic acid as a replacement for formic acid in Sludge Receipt and Adjustment Tank (SRAT) processing in the Defense Waste Processing Facility (DWPF). Catalytic decomposition of formic acid is responsible for the generation of hydrogen, a potentially flammable gas, during processing. To prevent the formation of a flammable mixture in the offgas, an air purge is used to dilute the hydrogen concentration below the 60% of the Composite Lower Flammability Limit (CLFL). The offgas is continuously monitored for hydrogen using Gas Chromatographs (GCs). Since formic acid is much more volatile and toxic than glycolic acid, a formic acid spill would lead to the release of much larger quantities to the environment. Switching from formic acid to glycolic acid is expected to eliminate the hydrogen flammability hazard leading to lower air purges, thus downgrading of Safety Significant GCs to Process Support GCs, and minimizing the consequence of a glycolic acid tank leak in DWPF. Overall this leads to a reduction in process operation costs and an increase in safety margin. Experiments were completed at three different scales to demonstrate that the nitric-glycolic acid flowsheet scales from the 4-L lab scale to the 22-L bench scale and 220-L engineering scale. Ten process demonstrations of the sludge-only flowsheet for SRAT and Slurry Mix Evaporator (SME) cycles were performed using Sludge Batch 8 (SB8)-Tank 40 simulant. No Actinide Removal Process (ARP) product or strip effluent was added during the runs. Six experiments were completed at the 4-L scale, two experiments were completed at the 22-L scale, and two experiments were completed at the 220-L scale. Experiments completed at the 4-L scale (100 and 110% acid stoichiometry) were repeated at the 22-L and 220-L scale for scale comparisons.

Development of computer code SIMPSEX for simulation of FBR fuel reprocessing flowsheets: II. additional benchmarking results

International Nuclear Information System (INIS)

Shekhar Kumar; Koganti, S.B.

2003-07-01

Benchmarking and application of a computer code SIMPSEX for high plutonium FBR flowsheets was reported recently in an earlier report (IGC-234). Improvements and recompilation of the code (Version 4.01, March 2003) required re-validation with the existing benchmarks as well as additional benchmark flowsheets. Improvements in the high Pu region (Pu Aq >30 g/L) resulted in better results in the 75% Pu flowsheet benchmark. Below 30 g/L Pu Aq concentration, results were identical to those from the earlier version (SIMPSEX Version 3, code compiled in 1999). In addition, 13 published flowsheets were taken as additional benchmarks. Eleven of these flowsheets have a wide range of feed concentrations and few of them are β-γ active runs with FBR fuels having a wide distribution of burnup and Pu ratios. A published total partitioning flowsheet using externally generated U(IV) was also simulated using SIMPSEX. SIMPSEX predictions were compared with listed predictions from conventional SEPHIS, PUMA, PUNE and PUBG. SIMPSEX results were found to be comparable and better than the result from above listed codes. In addition, recently reported UREX demo results along with AMUSE simulations are also compared with SIMPSEX predictions. Results of the benchmarking SIMPSEX with these 14 benchmark flowsheets are discussed in this report. (author)

Sludge batch 9 follow-on actual-waste testing for the nitric-glycolic flowsheet

Energy Technology Data Exchange (ETDEWEB)

Martino, C. J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Newell, J. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Crawford, C. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Pareizs, J. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Williams, M. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2017-03-23

An actual-waste Sludge Batch 9 qualification run with the nitric-glycolic flowsheet (SC-18) was performed in FY16. In order to supplement the knowledge base for the nitric-glycolic flowsheet, additional testing was performed on the product slurries, condensates, and intermediate samples from run SC-18.

Enriched uranium recovery flowsheet improvements

International Nuclear Information System (INIS)

Holt, D.L.

1986-01-01

Savannah River uses 7.5% TBP to recover and purify enriched uranium. Adequate decontamination from fission products is necessary to reduce personnel exposure and to ensure that the enriched uranium product meets specifications. Initial decontamination of the enriched uranium from the fission products is carried out in the 1A bank, 16 stages of mixer-settlers. Separation of the enriched uranium from the fission product, 95 Zr, has been adequate, but excessive solvent degradation caused by the long phase contact times in the mixer-settlers has limited the 95 Zr decontamination factor (DF). An experimental program is investigating the replacement of the current 1A bank with either centrifugal contactors or a combination of centrifugal contactors and mixer-settlers. Experimental work completed has compared laboratory-scale centrifugal contactors and mixer-settlers for 95 Zr removal efficiencies. Feed solutions spiked with actual plant solutions were used. The 95 Zr DF was significantly better in the mixer-settlers than in the centrifugal contactors. As a result of this experimental study, a hybrid equipment flowsheet has been proposed for plant use. The hybrid equipment flowsheet combines the advantages of both types of solvent extraction equipment. Centrifugal contactors would be utilized in the extraction and initial scrub sections, followed by additional scrub stages of mixer-settlers

Validation and Refinement of a Pain Information Model from EHR Flowsheet Data.

Science.gov (United States)

Westra, Bonnie L; Johnson, Steven G; Ali, Samira; Bavuso, Karen M; Cruz, Christopher A; Collins, Sarah; Furukawa, Meg; Hook, Mary L; LaFlamme, Anne; Lytle, Kay; Pruinelli, Lisiane; Rajchel, Tari; Settergren, Theresa Tess; Westman, Kathryn F; Whittenburg, Luann

2018-01-01

Secondary use of electronic health record (EHR) data can reduce costs of research and quality reporting. However, EHR data must be consistent within and across organizations. Flowsheet data provide a rich source of interprofessional data and represents a high volume of documentation; however, content is not standardized. Health care organizations design and implement customized content for different care areas creating duplicative data that is noncomparable. In a prior study, 10 information models (IMs) were derived from an EHR that included 2.4 million patients. There was a need to evaluate the generalizability of the models across organizations. The pain IM was selected for evaluation and refinement because pain is a commonly occurring problem associated with high costs for pain management. The purpose of our study was to validate and further refine a pain IM from EHR flowsheet data that standardizes pain concepts, definitions, and associated value sets for assessments, goals, interventions, and outcomes. A retrospective observational study was conducted using an iterative consensus-based approach to map, analyze, and evaluate data from 10 organizations. The aggregated metadata from the EHRs of 8 large health care organizations and the design build in 2 additional organizations represented flowsheet data from 6.6 million patients, 27 million encounters, and 683 million observations. The final pain IM has 30 concepts, 4 panels (classes), and 396 value set items. Results are built on Logical Observation Identifiers Names and Codes (LOINC) pain assessment terms and extend the need for additional terms to support interoperability. The resulting pain IM is a consensus model based on actual EHR documentation in the participating health systems. The IM captures the most important concepts related to pain. Schattauer GmbH Stuttgart.

Energy consumption analysis of integrated flowsheets for production of fuel ethanol from lignocellulosic biomass

International Nuclear Information System (INIS)

Cardona Alzate, C.A.; Sanchez Toro, O.J.

2006-01-01

Fuel ethanol is considered one of the most important renewable fuels due to the economic and environmental benefits of its use. Lignocellulosic biomass is the most promising feedstock for producing bioethanol due to its global availability and to the energy gain that can be obtained when non-fermentable materials from biomass are used for cogeneration of heat and power. In this work, several process configurations for fuel ethanol production from lignocellulosic biomass were studied through process simulation using Aspen Plus. Some flowsheets considering the possibilities of reaction-reaction integration were taken into account among the studied process routes. The flowsheet variants were analyzed from the energy point of view utilizing as comparison criterion the energy consumption needed to produce 1 L of anhydrous ethanol. Simultaneous saccharification and cofermentation process with water recycling showed the best results accounting an energy consumption of 41.96 MJ/L EtOH. If pervaporation is used as dehydration method instead of azeotropic distillation, further energy savings can be obtained. In addition, energy balance was estimated using the results from the simulation and literature data. A net energy value of 17.65-18.93 MJ/L EtOH was calculated indicating the energy efficiency of the lignocellulosic ethanol

Energy consumption analysis of integrated flowsheets for production of fuel ethanol from lignocellulosic biomass

Energy Technology Data Exchange (ETDEWEB)

Cardona Alzate, C.A. [Department of Chemical Engineering, National University of Colombia at Manizales, Cra. 27 No. 64-60, Manizales (Colombia)]. E-mail: ccardonaal@unal.edu.co; Sanchez Toro, O.J. [Department of Chemical Engineering, National University of Colombia at Manizales, Cra. 27 No. 64-60, Manizales (Colombia); Department of Engineering, University of Caldas, Calle 65 No. 26-10, Manizales (Colombia)

2006-10-15

Fuel ethanol is considered one of the most important renewable fuels due to the economic and environmental benefits of its use. Lignocellulosic biomass is the most promising feedstock for producing bioethanol due to its global availability and to the energy gain that can be obtained when non-fermentable materials from biomass are used for cogeneration of heat and power. In this work, several process configurations for fuel ethanol production from lignocellulosic biomass were studied through process simulation using Aspen Plus. Some flowsheets considering the possibilities of reaction-reaction integration were taken into account among the studied process routes. The flowsheet variants were analyzed from the energy point of view utilizing as comparison criterion the energy consumption needed to produce 1 L of anhydrous ethanol. Simultaneous saccharification and cofermentation process with water recycling showed the best results accounting an energy consumption of 41.96 MJ/L EtOH. If pervaporation is used as dehydration method instead of azeotropic distillation, further energy savings can be obtained. In addition, energy balance was estimated using the results from the simulation and literature data. A net energy value of 17.65-18.93 MJ/L EtOH was calculated indicating the energy efficiency of the lignocellulosic ethanol.

Actual Waste Demonstration of the Nitric-Glycolic Flowsheet for Sludge Batch 9 Qualification

Energy Technology Data Exchange (ETDEWEB)

Newell, J. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Pareizs, J. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Martino, C. J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Reboul, S. H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Coleman, C. J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Edwards, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Johnson, F. C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-09-01

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 is processable. Testing performed by the Savannah River National Laboratory has shown glycolic acid to be effective in replacing the function of formic acid in the DWPF chemical process. The nitric-glycolic flowsheet reduces mercury, significantly lowers the catalytic generation of hydrogen and ammonia which could allow purge reduction in the Sludge Receipt and Adjustment Tank (SRAT), stabilizes the pH and chemistry in the SRAT and the Slurry Mix Evaporator (SME), allows for effective rheology adjustment, and is favorable with respect to melter flammability. In order to implement the new flowsheet, SRAT and SME cycles, designated SC-18, were performed using a Sludge Batch (SB) 9 slurry blended from SB8 Tank 40H and Tank 51H samples. The SRAT cycle involved adding nitric and glycolic acids to the sludge, refluxing to steam strip mercury, and dewatering to a targeted solids concentration. Data collected during the SRAT cycle included offgas analyses, process temperatures, heat transfer, and pH measurements. The SME cycle demonstrated the addition of glass frit and the replication of six canister decontamination additions. The demonstration concluded with dewatering to a targeted solids concentration. Data collected during the SME cycle included offgas analyses, process temperatures, heat transfer, and pH measurements. Slurry and condensate samples were collected for subsequent analysis

Evaluation of conceptual flowsheets for incorporating Light Water Reactor (LWR) fuel materials in an advanced nuclear fuel cycle

International Nuclear Information System (INIS)

Bell, J.T.; Burch, W.D.; Collins, E.D.; Forsberg, C.W.; Prince, B.E.; Bond, W.D.; Campbell, D.O.; Delene, J.G.; Mailen, J.C.

1990-08-01

A preliminary study by a group of experts at ORNL has generated and evaluated a number of aqueous and non-aqueous flowsheets for recovering transuranium actinides from LWR fuel for use as fuel in an LMR and, at the same time, for transmutation of the wastes to less hazardous materials. The need for proliferation resistance was a consideration in the flowsheets. The current state of development of the flowsheets was evaluated and recommendations for additional study were made. 3 refs., 6 figs

An analysis of alternative New Technical Strategy flowsheets for tank waste remediation system

International Nuclear Information System (INIS)

Booker, C.P.

1994-01-01

The Hanford Tank remediation plans have gone through a few revisions for the best waste processing system. Some designs have been complex while others have been fairly simple. One of the key means in understanding and selecting among the various proposed systems is a discrete events modeling of the system. This modeling provides insight into (1) The total required size of the system; (2) The amount of material, such as reagents and other added materials that must be supplied; (3) The final mass of waste that must be stored; and (4) Areas within the system where a small change can greatly effect the total system. Discrete events modeling also provides the means by which various proposed systems may be compared. It is the framework in which variations within a particular system may be explored and compared to other instantiations. This study examines the current New Technical Strategy flowsheet system with discrete event modeling. Some of the possible variations within that system are examined and compared. Further, an previously proposed, more complex system is examined

Tank SY-102 remediation project: Flowsheet and conceptual design report

Energy Technology Data Exchange (ETDEWEB)

Yarbro, S.L.; Punjak, W.A.; Schreiber, S.B.; Dunn, S.L.; Jarvinen, G.D.; Marsh, S.F.; Pope, N.G.; Agnew, S.; Birnbaum, E.R.; Thomas, K.W.; Ortic, E.A.

1994-01-01

The US Department of Energy established the Tank Waste Remediation System (TWRS) to safely manage and dispose of radioactive waste stored in underground tanks on the Hanford Site. A major program in TWRS is pretreatment which was established to process the waste prior to disposal. Pretreatment is needed to resolve tank safety issues and to separate wastes into high-level and low-level fractions for subsequent immobilization and disposal. There is a fixed inventory of actinides and fission products in the tank which must be prepared for disposal. By segregating the actinides and fission products from the bulk of the waste, the tank`s contents can be effectively managed. Due to the high public visibility and environmental sensitivity of this problem, real progress and demonstrated efforts toward addressing it must begin as soon as possible. As a part of this program, personnel at the Los Alamos National Laboratory (LANL) have developed and demonstrated a flowsheet to remediate tank SY-102 which is located in the 200 West Area and contains high-level radioactive waste. This report documents the results of the flowsheet demonstrations performed with simulated, but radioactive, wastes using an existing glovebox line at the Los Alamos Plutonium Facility. The tank waste was characterized using both a tank history approach and an exhaustive evaluation of the available core sample analyses. This report also presents a conceptual design complete with a working material flow model, a major equipment list, and cost estimates.

Tank SY-102 remediation project: Flowsheet and conceptual design report

International Nuclear Information System (INIS)

Yarbro, S.L.; Punjak, W.A.; Schreiber, S.B.; Dunn, S.L.; Jarvinen, G.D.; Marsh, S.F.; Pope, N.G.; Agnew, S.; Birnbaum, E.R.; Thomas, K.W.; Ortic, E.A.

1994-01-01

The US Department of Energy established the Tank Waste Remediation System (TWRS) to safely manage and dispose of radioactive waste stored in underground tanks on the Hanford Site. A major program in TWRS is pretreatment which was established to process the waste prior to disposal. Pretreatment is needed to resolve tank safety issues and to separate wastes into high-level and low-level fractions for subsequent immobilization and disposal. There is a fixed inventory of actinides and fission products in the tank which must be prepared for disposal. By segregating the actinides and fission products from the bulk of the waste, the tank's contents can be effectively managed. Due to the high public visibility and environmental sensitivity of this problem, real progress and demonstrated efforts toward addressing it must begin as soon as possible. As a part of this program, personnel at the Los Alamos National Laboratory (LANL) have developed and demonstrated a flowsheet to remediate tank SY-102 which is located in the 200 West Area and contains high-level radioactive waste. This report documents the results of the flowsheet demonstrations performed with simulated, but radioactive, wastes using an existing glovebox line at the Los Alamos Plutonium Facility. The tank waste was characterized using both a tank history approach and an exhaustive evaluation of the available core sample analyses. This report also presents a conceptual design complete with a working material flow model, a major equipment list, and cost estimates

Antifoam Degradation Products in Off Gas and Condensate of Sludge Batch 9 Simulant Nitric-Formic Flowsheet Testing for the Defense Waste Processing Facility

Energy Technology Data Exchange (ETDEWEB)

Smith, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-04-14

Ten chemical processing cell (CPC) experiments were performed using simulant to evaluate Sludge Batch 9 for sludge-only and coupled processing using the nitric-formic flowsheet in the Defense Waste Processing Facility (DWPF). Sludge Receipt and Adjustment Tank (SRAT) and Slurry Mix Evaporator (SME) cycles were performed on eight of the ten. The other two were SRAT cycles only. Samples of the condensate, sludge, and off gas were taken to monitor the chemistry of the CPC experiments. The Savannah River National Laboratory (SRNL) has previously shown antifoam decomposes to form flammable organic products, (hexamethyldisiloxane (HMDSO), trimethylsilanol (TMS), and propanal), that are present in the vapor phase and condensate of the CPC vessels. To minimize antifoam degradation product formation, a new antifoam addition strategy was implemented at SRNL and DWPF to add antifoam undiluted.

Sludge Washing And Demonstration Of The DWPF Flowsheet In The SRNL Shielded Cells For Sludge Batch 8 Qualification

Energy Technology Data Exchange (ETDEWEB)

Pareizs, J. M.; Crawford, C. L.

2013-04-26

The current Waste Solidification Engineering (WSE) practice is to prepare sludge batches in Tank 51 by transferring sludge from other tanks to Tank 51. Tank 51 sludge is washed and transferred to Tank 40, the current Defense Waste Processing Facility (DWPF) feed tank. Prior to transfer of Tank 51 to Tank 40, the Savannah River National Laboratory (SRNL) typically simulates the Tank Farm and DWPF processes using a Tank 51 sample (referred to as the qualification sample). WSE requested the SRNL to perform characterization on a Sludge Batch 8 (SB8) sample and demonstrate the DWPF flowsheet in the SRNL shielded cells for SB8 as the final qualification process required prior to SB8 transfer from Tank 51 to Tank 40. A 3-L sample from Tank 51 (the SB8 qualification sample; Tank Farm sample HTF-51-12-80) was received by SRNL on September 20, 2012. The as-received sample was characterized prior to being washed. The washed material was further characterized and used as the material for the DWPF process simulation including a Sludge Receipt and Adjustment Tank (SRAT) cycle, a Slurry Mix Evaporator (SME) cycle, and glass fabrication and chemical durability measurements.

Nitric-glycolic flowsheet evaluation with the slurry-fed melt rate furnace

Energy Technology Data Exchange (ETDEWEB)

Williams, M. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Miller, D. H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Fowley, M. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Zamecnik, J. R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2017-03-01

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 supports Deliverable 4 of the DWPF & Saltstone Facility Engineering Technical Task Request (TTR)1 and is supplemental to the Cold Cap Evaluation Furnace (CEF) testing conducted in 2014.2 The Slurry-fed Melt Rate Furnace (SMRF) was selected for the supplemental testing as it requires significantly less resources than the CEF and could provide a tool for more rapid analysis of melter feeds in the future. The SMRF platform has been used previously to evaluate melt rate behavior of DWPF glasses, but was modified to accommodate analysis of the offgas stream. Additionally, the Melt Rate Furnace (MRF) and Quartz Melt Rate Furnace (QMRF) were utilized for evaluations. MRF data was used exclusively for melt behavior observations and REDuction/OXidation (REDOX) prediction comparisons and will be briefly discussed in conjunction with its support of the SMRF testing. The QMRF was operated similarly to the SMRF for the same TTR task, but will be discussed in a separate future report. The overall objectives of the SMRF testing were to; 1) Evaluate the efficacy of the SMRF as a platform for steady state melter testing with continuous feeding and offgas analysis; and 2) Generate supplemental melter offgas flammability data to support the melter offgas flammability modelling effort for DWPF implementation of the NG flowsheet.

Flowsheet development studies for the decontamination of high-activity-level water at Three Mile Island Unit 2

International Nuclear Information System (INIS)

Collins, E.D.; Bigelow, J.E.; Campbell, D.O.; King, L.J.; Knauer, J.B.

1980-01-01

Several chemical processing flowsheets were considered for the decontamination of high-activity-level water at the Three Mile Island (TMI) Unit 2. A zeolite ion exchange process was evaluated and recommended for absorption of the bulk of the highly radioactive cesium and strontium. Standard organic ion-exchange resins were selected to remove the remaining traces of radioactive nuclides (except tritium which cannot be removed by any practical process). Process conditions were evaluated using both synthetic, tracer-level solutions and samples of actual, high-activity level water from TMI Unit 2

Flowsheet for 63Ni production

International Nuclear Information System (INIS)

Williams, D.F.; Knauer, J.B.; O'Kelley, G.D.; Wiggins, J.T.; Porter, C.E.

1992-01-01

The production of large quantities of high specific activity 63 Ni (>10Ci/g) requires both a highly enriched 62 Ni target and a long irradiation period at high neutron flux. Trace impurities in the nickel and associated target materials are also activated and account for a significant fraction of the discharged activity and essentially all of the gamma activity. While most of these undesirable activation products (mainly transition metals) can be easily removed as chloride complexes during anion exchange, chromium, present as 51 Cr, and manganese, present as 54 Mn, are exceptions and require solvent extraction of the in-cell product to achieve the desired purity. In addition to summarizing the current development and production experience, optimized flowsheets are discussed

The conceptual flowsheet of effluent treatment during preparing spherical fuel elements of HTR

Energy Technology Data Exchange (ETDEWEB)

Ying, Quan, E-mail: quanying@tsinghua.edu.cn; Xiao-tong, Chen; Bing, Liu; Gen-na, Fu; Yang, Wang; You-lin, Shao; Zhen-ming, Lu; Ya-ping, Tang; Chun-he, Tang

2014-05-01

High temperature gas-cooled reactor (HTR) is one of the advanced nuclear reactors owing to its inherent safety and broad applications. For HTR, one of the key components is the ceramic fuel element. During the preparation of spherical fuel elements, the radioactive effluent treatment is necessary. Referring to the current treatment technologies and methods, the conceptual flowsheet of low-level radioactive effluent treatment during preparing spherical fuel elements was established. According to the above treatment process, the uranium concentration was decreased from 200 mg/l to the level of discharged standard.

MRPP: multiregion processing plant code

International Nuclear Information System (INIS)

Kee, C.W.; McNeese, L.E.

1976-09-01

The report describes the machine solution of a large number (approximately 52,000) of simultaneous linear algebraic equations in which the unknowns are the concentrations of nuclides in the fuel salt of a fluid-fueled reactor (MSBR) having a continuous fuel processing plant. Most of the equations define concentrations at various points in the processing plant. The code allows as input a generalized description of a processing plant flowsheet; it also performs the iterative adjustment of flowsheet parameters for determination of concentrations throughout the flowsheet, and the associated effect of the specified processing mode on the overall reactor operation

A flowsheet for a wave power unit

Energy Technology Data Exchange (ETDEWEB)

Sobierajski, E.; Kasperowicz, Z.

1984-01-01

A flowsheet is examined for a wave power unit designed to produce electricity, for flooding or drying a coastal zone, cleaning or protecting water areas of ports from sand deposits. The unit includes a vertical cylinder attached to the sea floor with input and output water ducts and valves. The cylinder has a rod with piston that is actuated through a flexible cable by float arranged next to the cylinder. The water injected under pressure into the pressure pipe can be used directly or as an intermediate energy source.

Sludge Washing and Demonstration of the DWPF Nitric/Formic Flowsheet in the SRNL Shielded Cells for Sludge Batch 9 Qualification

Energy Technology Data Exchange (ETDEWEB)

Pareizs, J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Newell, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Martino, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Crawford, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Johnson, F. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-11-01

Savannah River National Laboratory (SRNL) was requested by Savannah River Remediation (SRR) to qualify the next batch of sludge – Sludge Batch 9 (SB9). Current practice is to prepare sludge batches in Tank 51 by transferring sludge to Tank 51 from other tanks. The sludge is washed and transferred to Tank 40, the current Defense Waste Process Facility (DWPF) feed tank. Prior to sludge transfer from Tank 51 to Tank 40, the Tank 51 sludge must be qualified. SRNL qualifies the sludge in multiple steps. First, a Tank 51 sample is received, then characterized, washed, and again characterized. SRNL then demonstrates the DWPF Chemical Process Cell (CPC) flowsheet with the sludge. The final step of qualification involves chemical durability measurements of glass fabricated in the DWPF CPC demonstrations. In past sludge batches, SRNL had completed the DWPF demonstration with Tank 51 sludge. For SB9, SRNL has been requested to process a blend of Tank 51 and Tank 40 at a targeted ratio of 44% Tank 51 and 56% Tank 40 on an insoluble solids basis.

Plutonium--uranium partitioning; alternate flowsheet Plutonium Reclamation Facility. [SEPHIS

Energy Technology Data Exchange (ETDEWEB)

Fort, L.A.

1975-12-01

The SEPHIS computer program was used to predict the transient and steady-state concentrations in a stage-wise scheme for the Pu reclamation solvent extraction system. With the aid of the computer an alternative flowsheet for Pu--U partitioning was constructed. The goal of the alternative program is to reduce Pu losses from the initial stripping column and reduce the quantity of Pu-bearing wastes from the solvent extraction system. (JSR)

Evaluation of a dry process for conversion of U-AVLIS product to UF6. Milestone U361

International Nuclear Information System (INIS)

1992-05-01

A technical and engineering evaluation has been completed for a dry UF 6 production system to convert the product of an initial two-line U-AVLIS plant. The objective of the study has been to develop a better understanding of process design requirements, capital and operating costs, and demonstration requirements for this alternate process. This report summarizes the results of the study and presents various comparisons between the baseline and alternate processes, building on the information contained in UF 6 Product Alternatives Review Committee -- Final Report. It also provides additional information on flowsheet variations for the dry route which may warrant further consideration. The information developed by this study and conceptual design information for the baseline process will be combined with information to be developed by the U-AVLIS program and by industrial participants over the next twelve months to permit a further comparison of the baseline and alternate processes in terms of cost, risk, and compatibility with U-AVLIS deployment schedules and strategies. This comparative information will be used to make a final process flowsheet selection for the initial U-AVLIS plant by March 1993. The process studied is the alternate UF 6 production flowsheet. Process steps are (1) electron-beam distillation to reduce enriched product iron content from about 10 wt % or less, (2) hydrofluorination of the metal to UF 4 , (3) fluorination of UF 4 to UF 6 , (4) cold trap collection of the UF 6 product, (5) UF 6 purification by distillation, and (6) final blending and packaging of the purified UF 6 in cylinders. A preliminary system design has been prepared for the dry UF 6 production process based on currently available technical information. For some process steps, such information is quite limited. Comparisons have been made between this alternate process and the baseline plant process for UF 6 production

An equation oriented approach to steady state flowsheeting of methanol synthesis loop

International Nuclear Information System (INIS)

Fathikalajahi, J.; Baniadam, M.; Rahimpour, M.R.

2008-01-01

An equation-oriented approach was developed for steady state flowsheeting of a commercial methanol plant. The loop consists of fixed bed reactor, flash separator, preheater, coolers, and compressor. For steady sate flowsheeting of the plant mathematical model of reactor and other units are needed. Reactor used in loop is a Lurgi type and its configuration is rather complex. Previously reactor and flash separator are modeled as two important units of plant. The model is based on mass and energy balances in each equipment and utilizing some auxiliary equations such as rate of reaction and thermodynamics model for activity coefficients of liquid. In order to validate the mathematical model for the synthesis loop, some simulation data were performed using operating conditions and characteristics of the commercial plant. The good agreement between the steady state simulation results and the plant data shows the validity of the model

Development of a SREX flowsheet for the separation of strontium from dissolved INEEL zirconium calcine

International Nuclear Information System (INIS)

Law, J.D.; Wood, D.J.; Todd, T.A.

1999-01-01

Laboratory experimentation has indicated that the SREX process is effective for partitioning 90 Sr from acidic radioactive waste solutions located at the Idaho Nuclear Technology and Engineering Center. These laboratory results were used to develop a flowsheet for countercurrent testing of the SREX process with dissolved pilot plant calcine. Testing was performed using 24 stages of 2-cm diameter centrifugal contactors which are installed in the Remote Analytical Laboratory hot cell. Dissolved Run No.64 pilot plant calcine spiked with 85 Sr was used as feed solution for the testing. The flowsheet tested consisted of an extraction section (0.15 M 4prime,4prime(5prime)-di-(tert-butylcyclohexo)-18-crown-6 and 1.5 M TBP in Isopar-L.), a 1.0 M NaNO 3 scrub section to remove extracted K from the SREX solvent, a 0.01 M HNO 3 strip section for the removal of Sr from the SREX solvent, a 0.25 M Na2CO 3 wash section to remove degradation products from the solvent, and a 0.1 M HNO 3 rinse section. The behavior of 85 Sr, Na, K, Al, B, Ca, Cr, Fe, Ni, and Zr was evaluated. The described flowsheet successfully extracted 85 Sr from the dissolved pilot plant calcine with a removal efficiency of 99.6%. Distribution coefficients for 85 Sr ranged from 3.6 to 4.5 in the extraction section. With these distribution coefficients a removal efficiency of approximately >99.99% was expected. It was determined that the lower than expected removal efficiency can be attributed to a stage efficiency of only 60% in the extraction section. Extracted K was effectively scrubbed from the SREX solvent with the 1.0 M NaNO 3 resulting in only 6.4% of the K in the HLW strip product. Sodium was not extracted from the dissolved calcine by the SREX solvent; however, the use of a 1.0 M NaNO 3 scrub solution resulted in a Na concentration of 70 mg/L (12.3% of the feed concentration) in the HLW strip product. Al, B, Ca, Cr, Fe, Ni, and Zr were determined to be essentially inextractable

Recovery of transuranics from process residues

International Nuclear Information System (INIS)

Gray, J.H.; Gray, L.W.

1987-01-01

Process residues are generated at both the Rocky Flats Plant (RFP) and the Savannah River Plant (SRP) during aqueous chemical and pyrochemical operations. Frequently, process operations will result in either impure products or produce residues sufficiently contaminated with transuranics to be nondiscardable as waste. Purification and recovery flowsheets for process residues have been developed to generate solutions compatible with subsequent Purex operations and either solid or liquid waste suitable for disposal. The ''scrub alloy'' and the ''anode heel alloy'' are examples of materials generated at RFP which have been processed at SRP using the developed recovery flowsheets. Examples of process residues being generated at SRP for which flowsheets are under development include LECO crucibles and alpha-contaminated hydraulic oil

Computer Aided Methodology for Simultaneous Synthesis, Design & Analysis of Chemical Products-Processes

DEFF Research Database (Denmark)

d'Anterroches, Loïc; Gani, Rafiqul

2006-01-01

A new combined methodology for computer aided molecular design and process flowsheet design is presented. The methodology is based on the group contribution approach for prediction of molecular properties and design of molecules. Using the same principles, process groups have been developed...... a wide range of problems. In this paper, only the computer aided flowsheet design related features are presented....... together with their corresponding flowsheet property models. To represent the process flowsheets in the same way as molecules, a unique but simple notation system has been developed. The methodology has been converted into a prototype software, which has been tested with several case studies covering...

Flowsheets and source terms for radioactive waste projections

International Nuclear Information System (INIS)

Forsberg, C.W.

1985-03-01

Flowsheets and source terms used to generate radioactive waste projections in the Integrated Data Base (IDB) Program are given. Volumes of each waste type generated per unit product throughput have been determined for the following facilities: uranium mining, UF 6 conversion, uranium enrichment, fuel fabrication, boiling-water reactors (BWRs), pressurized-water reactors (PWRs), and fuel reprocessing. Source terms for DOE/defense wastes have been developed. Expected wastes from typical decommissioning operations for each facility type have been determined. All wastes are also characterized by isotopic composition at time of generation and by general chemical composition. 70 references, 21 figures, 53 tables

Development and testing of a SREX flowsheet for the partitioning of strontium and lead from simulated ICPP sodium-bearing waste

International Nuclear Information System (INIS)

Law, J.D.; Wood, D.J.

1996-11-01

Laboratory experimentation has indicated that the SREX process is effective for partitioning 90 Sr from acidic radioactive waste solutions located at the Idaho Chemical Processing Plant. Previous countercurrent flowsheet testing of the SREX process with simulated waste resulted in 99.98% removal of Sr. With this previous test, however, Pb was extracted by the SREX solvent and was not back-extracted in the dilute nitric acid strip section. The Pb concentration increased in the recycled solvent and in the aqueous phase of the strip section, resulting in the formation of a Pb precipitate. Subsequently, studies were initiated to identify alternative stripping agents which will selectively strip Sr and Pb from the SREX solvent. Based on the results of these studies, a countercurrent flow sheet was developed and tested in the 5.5-cm Centrifugal Contactor Mockup using simulated waste. The flowsheet tested consisted of an extraction section (0.15 M 4',4'(5)-di-(tert-butyldicyclohexo)-18-crown-6 and 1.2 M TBP in Isopar-L reg-sign), a 0.05 M nitric acid strip section for the removal of Sr from the SREX solvent, a 0.1 M ammonium citrate strip section for the removal of Pb from the SREX solvent, and a 2.0 M nitric acid equilibration section. The behavior of Sr, Pb, Al, Ca, Hg, Na, Zr, and H + was evaluated. The described flowsheet successfully extracted and selectively stripped Sr and Pb from the SBW simulant. Removal efficiencies of 97.9% and 99.91% were obtained for Sr and Pb, respectively. Essentially all of the extracted Sr (99.998%) and 1.9% of extracted Pb exited with the 0.05 M nitric acid strip product; whereas, 0.002% of the extracted Sr and 97.9% of the extracted Pb existed with the 0.1 M ammonium citrate strip product. Also, 95% of the Hg and 63% of the Zr were extracted by the SREX solvent

Computer-Aided Sustainable Process Synthesis-Design and Analysis

DEFF Research Database (Denmark)

Kumar Tula, Anjan

-groups is that, the performance of the entire process can be evaluated from the contributions of the individual process-groups towards the selected flowsheet property (for example, energy consumed). The developed flowsheet property models include energy consumption, carbon footprint, product recovery, product......Process synthesis involves the investigation of chemical reactions needed to produce the desired product, selection of the separation techniques needed for downstream processing, as well as taking decisions on sequencing the involved separation operations. For an effective, efficient and flexible...... focuses on the development and application of a computer-aided framework for sustainable synthesis-design and analysis of process flowsheets by generating feasible alternatives covering the entire search space and includes analysis tools for sustainability, LCA and economics. The synthesis method is based...

Evaluation of a dry process for conversion of U-AVLIS product to UF{sub 6}. Milestone U361

Energy Technology Data Exchange (ETDEWEB)

NONE

1992-05-01

A technical and engineering evaluation has been completed for a dry UF{sub 6} production system to convert the product of an initial two-line U-AVLIS plant. The objective of the study has been to develop a better understanding of process design requirements, capital and operating costs, and demonstration requirements for this alternate process. This report summarizes the results of the study and presents various comparisons between the baseline and alternate processes, building on the information contained in UF{sub 6} Product Alternatives Review Committee -- Final Report. It also provides additional information on flowsheet variations for the dry route which may warrant further consideration. The information developed by this study and conceptual design information for the baseline process will be combined with information to be developed by the U-AVLIS program and by industrial participants over the next twelve months to permit a further comparison of the baseline and alternate processes in terms of cost, risk, and compatibility with U-AVLIS deployment schedules and strategies. This comparative information will be used to make a final process flowsheet selection for the initial U-AVLIS plant by March 1993. The process studied is the alternate UF{sub 6} production flowsheet. Process steps are (1) electron-beam distillation to reduce enriched product iron content from about 10 wt % or less, (2) hydrofluorination of the metal to UF{sub 4}, (3) fluorination of UF{sub 4} to UF{sub 6}, (4) cold trap collection of the UF{sub 6} product, (5) UF{sub 6} purification by distillation, and (6) final blending and packaging of the purified UF{sub 6} in cylinders. A preliminary system design has been prepared for the dry UF{sub 6} production process based on currently available technical information. For some process steps, such information is quite limited. Comparisons have been made between this alternate process and the baseline plant process for UF{sub 6} production.

Flowsheet Validation For The Permanganate Digestion Of REILLEX(trademark) HPQ Anion Resin

International Nuclear Information System (INIS)

Kyser, E.

2009-01-01

The flowsheet for the digestion of Reillex(trademark) HPQ was validated both under the traditional alkaline conditions and under strongly acidic conditions. Due to difficulty in performing a pH adjustment in the large tank where this flowsheet must be performed, the recommended digestion conditions were changed from pH 8-10 to 8 M HNO 3 . Thus, no pH adjustment of the solution is required prior to performing the permanganate addition and digestion and the need to sample the digestion tank to confirm appropriate pH range for digestion may be avoided. Neutralization of the acidic digestion solution will be performed after completion of the resin digestion cycle. The amount of permanganate required for this type of resin (Reillex(trademark) HPQ) was increased from 1 kg/L resin to 4 kg/L resin to reduce the amount of residual resin solids to a minimal amount ( 2 ) solids (1.71 kg/L resin) and involves the generation of a significant liquid volume due to the low solubility of permanganate. However, since only two batches of resin (40 L each) are expected to be digested, the total waste generated is limited.

Flowsheet finalisation for immobilisation of SGHWR wastes

International Nuclear Information System (INIS)

Lee, D.J.

1984-09-01

This report summarises research and development work carried out during the year ended March 1983 on the programme for cementing the Winfrith Reactor (SGHWR) sludge. Further results from the characterisation programme are reported, together with data from the cementation programme. Formulations based on Ordinary Portland Cement (OPC), ground granulated blast furnace slag (BFS) and Pulverised Fuel Ash (PFA) have been tested. The results show that a blend of 90% BFS/10% OPC by weight, gives the best properties. Chemical pretreatment as a method for producing a stable waste form is discussed. A dewatering pretreatment to provide a sludge suitable for direct cementation is also outlined. A flowsheet for cementing the SGHWR sludge is proposed based on these laboratory and pilot scale studies. The major components required for the active plant are identified and provisional plant layouts are given. (author)

Process synthesis and intensification of hybrid separations

DEFF Research Database (Denmark)

Errico, Massimiliano

2017-01-01

Hybrid flowsheets are defined, in the context of process intensification, as alternatives suitable for replacing energy-intensive separation methods through the combination of more than one unit operation. Distillation is one of the first options considered for achieving a required separation...... and commented on. The corresponding distillation-based processes are considered for comparison. Synthesis of the possible hybrid flowsheets appears to be important, especially when multicomponent mixtures are considered. This aspect is discussed for the combination of liquid-liquid extraction and distillation...... as applied to the separation of biobutanol from its fermentation broth. The synthesis of alternative hybrid flowsheets is reported, showing that one configuration can realize a 43% reduction in the total annual cost. Bioalcohol production by fermentation perfectly represents the casewhere distillation alone...

/sup 238/Pu fuel form processes quarterly report, April-June 1980

Energy Technology Data Exchange (ETDEWEB)

Folger, R. L.

1980-06-01

Savannah River Laboratory (SRL) completed the development of a production process to fabricate /sup 238/PuO/sub 2/ fuel forms for the GPHS. The fabrication flowsheet was based on a flowsheet originally developed at Los Alamos National Scientific Laboratory (LANSL). A summary report of the SRL process development effort is presented.

Corrosion Testing of Monofrax K-3 Refractory in Defense Waste Processing Facility (DWPF) Alternate Reductant Feeds

Energy Technology Data Exchange (ETDEWEB)

Williams, M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Jantzen, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Burket, P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-04-06

The Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) uses a combination of reductants and oxidants while converting high level waste (HLW) to a borosilicate waste form. A reducing flowsheet is maintained to retain radionuclides in their reduced oxidation states which promotes their incorporation into borosilicate glass. For the last 20 years of processing, the DWPF has used formic acid as the main reductant and nitric acid as the main oxidant. During reaction in the Chemical Process Cell (CPC), formate and formic acid release measurably significant H₂ gas which requires monitoring of certain vessel’s vapor spaces. A switch to a nitric acid-glycolic acid (NG) flowsheet from the nitric-formic (NF) flowsheet is desired as the NG flowsheet releases considerably less H₂ gas upon decomposition. This would greatly simplify DWPF processing from a safety standpoint as close monitoring of the H₂ gas concentration could become less critical. In terms of the waste glass melter vapor space flammability, the switch from the NF flowsheet to the NG flowsheet showed a reduction of H₂ gas production from the vitrification process as well. Due to the positive impact of the switch to glycolic acid determined on the flammability issues, evaluation of the other impacts of glycolic acid on the facility must be examined.

Process Flow Sheet Generation & Design through a Group Contribution Approach

DEFF Research Database (Denmark)

d'Anterroches, Loïc

2006-01-01

Denne afhandling beskriver udviklingen af et framework til opstilling og design af proces flowsheet ved hjælp af en systematisk strategi for Computer Aided Flowsheet Design (CAFD). Det udviklede framework omfatter formulering, løsning og analyse af CAFD problemer baseret på et koncept med...... forbindelsesregler samt deres bidrag til specifikke flowsheet egenskaber på samme måde som kemiske molekyler bliver syntetiseret og testet for deres egenskaber. Hertil er simple og effektive metoder til processyntese og design blevet udviklet. Alternative flowsheet for kemiske processer opstilles baglæns ved...... at kombinere procesgrupper således at der dannes flowsheet strukturer som har de ønskede egenskaber. Derefter udvælges de mest lovende flowsheetalternativer til design hvorved de enkelte enhedsoperationer beregnes baglæns udfra specifikationerne for deres ind- og udgangsstrømme svarende til procesgruppernes...

Vitrification process testing for reference HWVP waste

International Nuclear Information System (INIS)

Perez, J.M. Jr.; Goles, R.W.; Nakaoka, R.K.; Kruger, O.L.

1991-01-01

The Hanford Waste Vitrification Plant (HWVP) is being designed to vitrify high-level radioactive wastes stored on the Hanford site. The vitrification flow-sheet is being developed to assure the plant will achieve plant production requirements and the glass product will meet all waste form requirements for final geologic disposal. The first Hanford waste to be processed by the HWVP will be a neutralized waste resulting from PUREX fuel reprocessing operations. Testing is being conducted using representative nonradioactive simulants to obtain process and product data required to support design, environmental, and qualification activities. Plant/process criteria, testing requirements and approach, and results to date will be presented

Development of a solvent extraction process for cesium removal from SRS tank waste

International Nuclear Information System (INIS)

Leonard, R.A.; Conner, C.; Liberatore, M.W.; Sedlet, J.; Aase, S.B.; Vandegrift, G.F.; Delmau, L.H.; Bonnesen, P.V.; Moyer, B.A.

2001-01-01

An alkaline-side solvent extraction process was developed for cesium removal from Savannah River Site (SRS) tank waste. The process was invented at Oak Ridge National Laboratory and developed and tested at Argonne National Laboratory using singlestage and multistage tests in a laboratory-scale centrifugal contactor. The dispersion number, hydraulic performance, stage efficiency, and general operability of the process flowsheet were determined. Based on these tests, further solvent development work was done. The final solvent formulation appears to be an excellent candidate for removing cesium from SRS tank waste.

A computer-aided approach for achieving sustainable process design by process intensification

DEFF Research Database (Denmark)

Anantasarn, Nateetorn; Suriyapraphadilok, Uthaiporn; Babi, Deenesh Kavi

2017-01-01

to generate flowsheet alternatives that satisfy the design targets thereby, minimizing and/or eliminating the process hot-spots. The application of the framework is highlighted through the production of para-xylene via toluene methylation where more sustainable flowsheet alternatives that consist of hybrid......Process intensification can be applied to achieve sustainable process design. In this paper, a systematic, 3-stage synthesis-intensification framework is applied to achieve more sustainable design. In stage 1, the synthesis stage, an objective function and design constraints are defined and a base...... case is synthesized. In stage 2, the design and analysis stage, the base case is analyzed using economic and environmental analyses to identify process hot-spots that are translated into design targets. In stage 3, the innovation design stage, phenomena-based process intensification is performed...

Development of some operations in technological flowsheet for spent VVER fuel reprocessing at a pilot plant

International Nuclear Information System (INIS)

Lazarev, L.N.; Galkin, B.Ya; Lyubtsev, R.I.; Romanovskii, V.N.; Velikhov, E.P.

1981-01-01

The fuel reprocessing pilot plants for high active materials would permit the study and development or particular processing steps and flowsheet variations; in some cases, these experimental installations realize on a small scale practically all technological chains of large reprocessing plants. Such a fuel reprocessing pilot plant with capacity of 3 kg U/d has been built at V. G. Khlopin Radium Institute. The pilot plant is installed in the hot cell of radiochemical compartment, and is composed of the equipments for fuel element cutting and dissolving, the preparation of feed solution (clarification, correction), extraction reprocessing and the production of uranium, plutonium and neptunium concentrates, the complex processing of liquid and solid wastes and a special unit for gas purification and analysis. In the last few years, a series of experiments have been carried out on the reprocessing of spent VVER fuel. (J.P.N.)

An evaluation of five flowsheets for the recovery of uranium from Wits leach pulps

International Nuclear Information System (INIS)

Boydell, D.W.; Viljoen, E.B.

1978-01-01

This article evaluates five flowsheets for uranium recovery and an incremental net present value is calculated for each by the discounting of cash flows at 25 per cent per year over a projected life of 15 years. The highest net present value results from the circuit that employs belt filtration followed by continuous ion exchange, plus solvent extraction, in the particular case of the material used in the examples

Technical feasibility of the Diamex process

International Nuclear Information System (INIS)

Sorel, Ch

2007-01-01

Full text of publication follows. The DIAMEX process was developed to facilitate the separation of the trivalent actinides from the trivalent lanthanides. It consists in co-extracting the trivalent actinides and lanthanides using a diamide extractant: Di-Methyl Di-Octyl Hexyl Ethoxy Malonamide (DMDOHEMA). The flow-sheet comprises: Co-extraction at high acidity (3 M HNO 3 ) of the trivalent actinides and lanthanides by the diamide; scrubbing of some fission products (Zr, Mo, Fe, Pd) by a mixture of oxalic acid and HEDTA, followed by de-acidification to prepare for the next step; stripping of the actinides + lanthanides at low acidity; solvent treatment prior to recycling. This flow-sheet was successfully tested at laboratory scale from 1999 to 2003 in mixer-settlers and subsequently in ECLHA centrifugal extractors on active solutions from the dissolution of actual spent fuel samples. Actinide recovery factors above 99.9% were obtained with high purification factors for spurious fission products. The main objectives of the final ''technical feasibility'' demonstration tests at the end of 2005 with a PUREX raffinate solution were to test continuous solvent recycling (not included during the earlier tests) and to carry out essential operations in continuous contactors representative of pulsed columns that could be used at industrial scale. We therefore decided to carry out the demonstration in the shielded process line (CBP) with some of the devices already used for a PUREX test. During these tests the first two steps in the flow-sheet were therefore carried out in pulsed columns 4 meters high; An+Ln stripping was performed in mixer-settlers and the solvent treatment in ECRAN. The americium and curium recovery yield exceeded 99.9% and the decontamination factors obtained at the end of the test with respect to the fission products Zr, Mo and Fe were 800, 100 and 10, respectively. (author)

Characterization of Neptunium Oxide Generated Using the HB-Line Phase II Flowsheet

Energy Technology Data Exchange (ETDEWEB)

Duffey, J

2003-08-29

Approximately 98 grams of neptunium(IV) oxide (NpO{sub 2}) were produced at the Savannah River Technology Center (SRTC) for use in gas generation tests to support the neptunium stabilization program at the Savannah River Site (SRS). The NpO{sub 2} was produced according to the anticipated HB-Line flowsheet consisting of anion exchange, oxalate precipitation, filtration, and calcination. Characterization of the NpO{sub 2} product to be used in gas generation tests included bulk and tap density measurements, X-ray diffraction, particle size distribution, specific surface area measurements, and moisture analysis.

Development of an integrated MOX-scrap recycling flow-sheet by dry and wet routes using microwave heating techniques

Energy Technology Data Exchange (ETDEWEB)

Mallik, G K; Malav, R K; Karande, A P; Bhargava, V K; Kamath, H S [Advanced Fuel Fabrication Facility, Bhabha Atomic Research Centre, Tarapur (India)

1999-01-01

A simple, short and efficient scrap, recycling flow-sheet, which is exclusively based on microwave heating techniques and, includes both dry and wet routes, for (U,Pu)O{sub 2} fuel scrap recycling has been developed and evaluated. (author) 6 refs., 1 tab.

Dissolution flowsheet for high flux isotope reactor fuel

Energy Technology Data Exchange (ETDEWEB)

Foster, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-09-27

As part of the Spent Nuclear Fuel (SNF) processing campaign, H-Canyon is planning to begin dissolving High Flux Isotope Reactor (HFIR) fuel in late FY17 or early FY18. Each HFIR fuel core contains inner and outer fuel elements which were fabricated from uranium oxide (U₃O₈) dispersed in a continuous Al phase using traditional powder metallurgy techniques. Fuels fabricated in this manner, like other SNF’s processed in H-Canyon, dissolve by the same general mechanisms with similar gas generation rates and the production of H₂. The HFIR fuel cores will be dissolved and the recovered U will be down-blended into low-enriched U. HFIR fuel was previously processed in H-Canyon using a unique insert in both the 6.1D and 6.4D dissolvers. Multiple cores will be charged to the same dissolver solution maximizing the concentration of dissolved Al. The objective of this study was to identify flowsheet conditions through literature review and laboratory experimentation to safely and efficiently dissolve the HFIR fuel in H-Canyon. Laboratory-scale experiments were performed to evaluate the dissolution of HFIR fuel using both Al 1100 and Al 6061 T6 alloy coupons. The Al 1100 alloy was considered a representative surrogate which provided an upper bound on the generation of flammable (i.e., H₂) gas during the dissolution process. The dissolution of the Al 6061 T6 alloy proceeded at a slower rate than the Al 1100 alloy and was used to verify that the target Al concentration in solution could be achieved for the selected Hg concentration. Mass spectrometry and Raman spectroscopy were used to provide continuous monitoring of the concentration of H₂ and other permanent gases in the dissolution offgas allowing the development of H₂ generation rate profiles. The H₂ generation rates were subsequently used to evaluate if a full HFIR core could be dissolved in an H-Canyon dissolver without exceeding 60% of the

Development of the SREX process for the treatment of ICPP liquid wastes

International Nuclear Information System (INIS)

Wood, D.J.; Law, J.D.; Garn, T.G.; Tillotson, R.D.; Tullock, P.A.; Todd, T.A.

1997-10-01

The removal of 90 Sr from actual and simulated wastes at the Idaho Chemical Processing Plant (ICPP) at the Idaho National Engineering and Environmental Laboratory (INEEL) has been demonstrated with the SREX process. This solvent extraction process employs the extractant 4',4'(5') di-(t-butylcyclohexano)-18-crown-6 in 1-octanol or a mixture of tributyl phosphate and a hydrocarbon diluent called Isopar L reg-sign. Process flowsheets have been designed for testing in countercurrent experiments with centrifugal contractors. The flowsheets have been designed using batch contract solvent extraction methods. The extraction of Sr as well as other interfering ions has been studied. The effect of various parameters including nitric acid dependence, extractant concentration dependence, hydronium ion concentration, and interferent concentrations upon the extraction efficiency of the process has been evaluated. The radiolysis of the SREX solvent has also been investigated as a function of absorbed gamma radiation. The extraction efficiency of the solvent has been shown to be only slightly dependent upon absorbed dose in the range 0--1,000 kGy. The decontamination of actual sodium-bearing waste and dissolved calcine solutions has been accomplished in batch contact flowsheets. Decontamination factors as high as 10E3 have been obtained with sequential batch contacts. Flowsheets have been developed to accomplish decontamination of the liquid wastes with respect to 90 Sr as well as the removal of Pb and Hg. Pb may be partitioned from the Sr fraction in a separate stripping procedure using ammonium citrate. This work has led to the formulation of countercurrent flowsheets which have been tested in centrifugal contractors with actual waste and reported in the document INEEL/EXT-97-00832

LANDFILL GAS CONVERSION TO LNG AND LCO{sub 2}. PHASE 1, FINAL REPORT FOR THE PERIOD MARCH 1998-FEBRUARY 1999

Energy Technology Data Exchange (ETDEWEB)

COOK,W.J.; NEYMAN,M.; SIWAJEK,L.A.; BROWN,W.R.; VAN HAUWAERT,P.M.; CURREN,E.D.

1998-02-25

Process designs and economics were developed to produce LNG and liquid carbon dioxide (CO{sub 2}) from landfill gas (LFG) using the Acrion CO{sub 2} wash process. The patented Acrion CO{sub 2} wash process uses liquid CO{sub 2} to absorb contaminants from the LFG. The process steps are compression, drying, CO{sub 2} wash contaminant removal and CO{sub 2} recovery, residual CO{sub 2} removal and methane liquefaction. Three flowsheets were developed using different residual CO{sub 2} removal schemes. These included physical solvent absorption (methanol), membranes and molecular sieves. The capital and operating costs of the flowsheets were very similar. The LNG production cost was around ten cents per gallon. In parallel with process flowsheet development, the business aspects of an eventual commercial project have been explored. The process was found to have significant potential commercial application. The business plan effort investigated the economics of LNG transportation, fueling, vehicle conversion, and markets. The commercial value of liquid CO{sub 2} was also investigated. This Phase 1 work, March 1998 through February 1999, was funded under Brookhaven National laboratory contract 725089 under the research program entitled ``Liquefied Natural Gas as a Heavy Vehicle Fuel.'' The Phase 2 effort will develop flowsheets for the following: (1) CO{sub 2} and pipeline gas production, with the pipeline methane being liquefied at a peak shaving site, (2) sewage digester gas as an alternate feedstock to LFG and (3) the use of mixed refrigerants for process cooling. Phase 2 will also study the modification of Acrion's process demonstration unit for the production of LNG and a market site for LNG production.

Enriched uranium processing with 7-1/2% TBP

International Nuclear Information System (INIS)

Orth, D.A.; Martin, W.H.; Pickett, C.E.

1983-01-01

The 7-1/2% TBP flowsheet gives adequate recovery of uranium and neptunium or plutonium, with reduced waste volume as compared to the prior aluminum-salted 3-1/2% TBP flowsheet. Decontamination from fission products is sensitive to numerous variables, including aluminum nitrate concentration in the feed, impeller speeds, and prior treatment of the fuel solution in head end operations. The impeller speed in the 1A bank also influences uranium losses as well as the fission product decontamination. The magnitudes of these effects suggest that stage efficiency is poor with this flowsheet in this mixer settler unit. The existing continuous solvent washers give evidence of low washing efficiency that limits permissible feed activity and that may be related to low contact time between the solvent and the carbonate wash solution. The most general conclusion is that satisfactory operation can be obtained with all projected domestic and foreign fuels under consideration for processing, by suitable adjustment of operating conditions. Also, possible flowsheet and equipment changes are known that could improve operations with these fuels further. 7 references

Conceptual design of distillation-based hybrid separation processes.

Science.gov (United States)

Skiborowski, Mirko; Harwardt, Andreas; Marquardt, Wolfgang

2013-01-01

Hybrid separation processes combine different separation principles and constitute a promising design option for the separation of complex mixtures. Particularly, the integration of distillation with other unit operations can significantly improve the separation of close-boiling or azeotropic mixtures. Although the design of single-unit operations is well understood and supported by computational methods, the optimal design of flowsheets of hybrid separation processes is still a challenging task. The large number of operational and design degrees of freedom requires a systematic and optimization-based design approach. To this end, a structured approach, the so-called process synthesis framework, is proposed. This article reviews available computational methods for the conceptual design of distillation-based hybrid processes for the separation of liquid mixtures. Open problems are identified that must be addressed to finally establish a structured process synthesis framework for such processes.

Defense Waste Processing Facility Simulant Chemical Processing Cell Studies for Sludge Batch 9

International Nuclear Information System (INIS)

Smith, Tara E.; Newell, J. David; Woodham, Wesley H.

2016-01-01

The Savannah River National Laboratory (SRNL) received a technical task request from Defense Waste Processing Facility (DWPF) and Saltstone Engineering to perform simulant tests to support the qualification of Sludge Batch 9 (SB9) and to develop the flowsheet for SB9 in the DWPF. These efforts pertained to the DWPF Chemical Process Cell (CPC). CPC experiments were performed using SB9 simulant (SB9A) to qualify SB9 for sludge-only and coupled processing using the nitric-formic flowsheet in the DWPF. Two simulant batches were prepared, one representing SB8 Tank 40H and another representing SB9 Tank 51H. The simulant used for SB9 qualification testing was prepared by blending the SB8 Tank 40H and SB9 Tank 51H simulants. The blended simulant is referred to as SB9A. Eleven CPC experiments were run with an acid stoichiometry ranging between 105% and 145% of the Koopman minimum acid equation (KMA), which is equivalent to 109.7% and 151.5% of the Hsu minimum acid factor. Three runs were performed in the 1L laboratory scale setup, whereas the remainder were in the 4L laboratory scale setup. Sludge Receipt and Adjustment Tank (SRAT) and Slurry Mix Evaporator (SME) cycles were performed on nine of the eleven. The other two were SRAT cycles only. One coupled flowsheet and one extended run were performed for SRAT and SME processing. Samples of the condensate, sludge, and off-gas were taken to monitor the chemistry of the CPC experiments.

Defense Waste Processing Facility Simulant Chemical Processing Cell Studies for Sludge Batch 9

Energy Technology Data Exchange (ETDEWEB)

Smith, Tara E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Newell, J. David [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Woodham, Wesley H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-08-10

The Savannah River National Laboratory (SRNL) received a technical task request from Defense Waste Processing Facility (DWPF) and Saltstone Engineering to perform simulant tests to support the qualification of Sludge Batch 9 (SB9) and to develop the flowsheet for SB9 in the DWPF. These efforts pertained to the DWPF Chemical Process Cell (CPC). CPC experiments were performed using SB9 simulant (SB9A) to qualify SB9 for sludge-only and coupled processing using the nitric-formic flowsheet in the DWPF. Two simulant batches were prepared, one representing SB8 Tank 40H and another representing SB9 Tank 51H. The simulant used for SB9 qualification testing was prepared by blending the SB8 Tank 40H and SB9 Tank 51H simulants. The blended simulant is referred to as SB9A. Eleven CPC experiments were run with an acid stoichiometry ranging between 105% and 145% of the Koopman minimum acid equation (KMA), which is equivalent to 109.7% and 151.5% of the Hsu minimum acid factor. Three runs were performed in the 1L laboratory scale setup, whereas the remainder were in the 4L laboratory scale setup. Sludge Receipt and Adjustment Tank (SRAT) and Slurry Mix Evaporator (SME) cycles were performed on nine of the eleven. The other two were SRAT cycles only. One coupled flowsheet and one extended run were performed for SRAT and SME processing. Samples of the condensate, sludge, and off-gas were taken to monitor the chemistry of the CPC experiments.

Technology Evaluation of Process Configurations for Second Generation Bioethanol Production using Dynamic Model-based Simulations

DEFF Research Database (Denmark)

Morales Rodriguez, Ricardo; Meyer, Anne S.; Gernaey, Krist

2011-01-01

An assessment of a number of different process flowsheets for bioethanol production was performed using dynamic model-based simulations. The evaluation employed diverse operational scenarios such as, fed-batch, continuous and continuous with recycle configurations. Each configuration was evaluated...... against the following benchmark criteria, yield (kg ethanol/kg dry-biomass), final product concentration and number of unit operations required in the different process configurations. The results has shown the process configuration for simultaneous saccharification and co-fermentation (SSCF) operating...... in continuous mode with a recycle of the SSCF reactor effluent, results in the best productivity of bioethanol among the proposed process configurations, with a yield of 0.18 kg ethanol /kg dry-biomass....

The monitoring and control of TRUEX processes

International Nuclear Information System (INIS)

Regalbuto, M.C.; Misra, B.; Chamberlain, D.B.; Leonard, R.A.; Vandegrift, G.F.

1992-04-01

The Generic TRUEX Model (GTM) was used to design a flowsheet for the TRUEX solvent extraction process that would be used to determine its instrumentation and control requirements. Sensitivity analyses of the key process variables, namely, the aqueous and organic flow rates, feed compositions, and the number of contactor stages, were carried out to assess their impact on the operation of the TRUEX process. Results of these analyses provide a basis for the selection of an instrument and control system and the eventual implementation of a control algorithm. Volume Two of this report is an evaluation of the instruments available for measuring many of the physical parameters. Equations that model the dynamic behavior of the TRUEX process have been generated. These equations can be used to describe the transient or dynamic behavior of the process for a given flowsheet in accordance with the TRUEX model. Further work will be done with the dynamic model to determine how and how quickly the system responds to various perturbations. The use of perturbation analysis early in the design stage will lead to a robust flowsheet, namely, one that will meet all process goals and allow for wide control bounds. The process time delay, that is, the speed with which the system reaches a new steady state, is an important parameter in monitoring and controlling a process. In the future, instrument selection and point-of-variable measurement, now done using the steady-state results reported here, will be reviewed and modified as necessary based on this dynamic method of analysis

Preconceptual design of a salt splitting process using ceramic membranes

Energy Technology Data Exchange (ETDEWEB)

Kurath, D.E.; Brooks, K.P.; Hollenberg, G.W.; Clemmer, R. [Pacific Northwest National Lab., Richland, WA (United States); Balagopal, S.; Landro, T.; Sutija, D.P. [Ceramatec, Inc., Salt Lake City, UT (United States)

1997-01-01

Inorganic ceramic membranes for salt splitting of radioactively contaminated sodium salt solutions are being developed for treating U. S. Department of Energy tank wastes. The process consists of electrochemical separation of sodium ions from the salt solution using sodium (Na) Super Ion Conductors (NaSICON) membranes. The primary NaSICON compositions being investigated are based on rare- earth ions (RE-NaSICON). Potential applications include: caustic recycling for sludge leaching, regenerating ion exchange resins, inhibiting corrosion in carbon-steel tanks, or retrieving tank wastes; reducing the volume of low-level wastes volume to be disposed of; adjusting pH and reducing competing cations to enhance cesium ion exchange processes; reducing sodium in high-level-waste sludges; and removing sodium from acidic wastes to facilitate calcining. These applications encompass wastes stored at the Hanford, Savannah River, and Idaho National Engineering Laboratory sites. The overall project objective is to supply a salt splitting process unit that impacts the waste treatment and disposal flowsheets and meets user requirements. The potential flowsheet impacts include improving the efficiency of the waste pretreatment processes, reducing volume, and increasing the quality of the final waste disposal forms. Meeting user requirements implies developing the technology to the point where it is available as standard equipment with predictable and reliable performance. This report presents two preconceptual designs for a full-scale salt splitting process based on the RE-NaSICON membranes to distinguish critical items for testing and to provide a vision that site users can evaluate.

Preconceptual design of a salt splitting process using ceramic membranes

International Nuclear Information System (INIS)

Kurath, D.E.; Brooks, K.P.; Hollenberg, G.W.; Clemmer, R.; Balagopal, S.; Landro, T.; Sutija, D.P.

1997-01-01

Inorganic ceramic membranes for salt splitting of radioactively contaminated sodium salt solutions are being developed for treating U. S. Department of Energy tank wastes. The process consists of electrochemical separation of sodium ions from the salt solution using sodium (Na) Super Ion Conductors (NaSICON) membranes. The primary NaSICON compositions being investigated are based on rare- earth ions (RE-NaSICON). Potential applications include: caustic recycling for sludge leaching, regenerating ion exchange resins, inhibiting corrosion in carbon-steel tanks, or retrieving tank wastes; reducing the volume of low-level wastes volume to be disposed of; adjusting pH and reducing competing cations to enhance cesium ion exchange processes; reducing sodium in high-level-waste sludges; and removing sodium from acidic wastes to facilitate calcining. These applications encompass wastes stored at the Hanford, Savannah River, and Idaho National Engineering Laboratory sites. The overall project objective is to supply a salt splitting process unit that impacts the waste treatment and disposal flowsheets and meets user requirements. The potential flowsheet impacts include improving the efficiency of the waste pretreatment processes, reducing volume, and increasing the quality of the final waste disposal forms. Meeting user requirements implies developing the technology to the point where it is available as standard equipment with predictable and reliable performance. This report presents two preconceptual designs for a full-scale salt splitting process based on the RE-NaSICON membranes to distinguish critical items for testing and to provide a vision that site users can evaluate

Solvent-extraction purification of neptunium

International Nuclear Information System (INIS)

Kyser, E.A.; Hudlow, S.L.

2008-01-01

The Savannah River Site (SRS) has recovered 237 Np from reactor fuel that is currently being processed into NpO 2 for future production of 238 Pu. Several purification flowsheets have been utilized. An oxidizing solvent-extraction (SX) flowsheet was used to remove Fe, sulfate ion, and Th while simultaneously 237 Np, 238 Pu, u, and nonradioactive Ce(IV) was extracted into the tributyl phosphate (TBP) based organic solvent. A reducing SX flowsheet (second pass) removed the Ce and Pu and recovered both Np and U. The oxidizing flowsheet was necessary for solutions that contained excessive amounts of sulfate ion. Anion exchange was used to perform final purification of Np from Pu, U, and various non-actinide impurities. The Np(IV) in the purified solution was then oxalate-precipitated and calcined to an oxide for shipment to other facilities for storage and future target fabrication. Performance details of the SX purification and process difficulties are discussed. (authors)

Flowchart evaluations of irradiated fuel treatment process of low burnup thorium

International Nuclear Information System (INIS)

Linardi, M.

1987-01-01

A literature survey has been carried out, on some versions of the acid-thorex process. Flowsheets of the different parts of the process were evaluated with mixer-settlers experiments. A low burnup thorium fuel (mass ratio Th/U∼100/1), proposed for Brazilian fast breeder reactor initial program, was considered. The behaviour of some fission products was studied by irradiated tracers techniques. Modifications in some of the process parameters were necessary to achieve low losses of 233 U and 232 U and 232 Th. A modified acid-thorex process flowsheet, evaluated in a complete operational cycle, for the treatment of low burnup thorium fuels, is presented. High decontamination factors of thorium in uranium, with reasonable decontamination of uranium in thorium, were achieved. (author) [pt

Steady-State Process Modelling

DEFF Research Database (Denmark)

Cameron, Ian; Gani, Rafiqul

2011-01-01

illustrate the “equation oriented” approach as well as the “sequential modular” approach to solving complex flowsheets for steady state applications. The applications include the Williams-Otto plant, the hydrodealkylation (HDA) of toluene, conversion of ethylene to ethanol and a bio-ethanol process....

40 CFR 61.134 - Standard: Naphthalene processing, final coolers, and final-cooler cooling towers.

Science.gov (United States)

2010-07-01

... coolers, and final-cooler cooling towers. 61.134 Section 61.134 Protection of Environment ENVIRONMENTAL... Standard: Naphthalene processing, final coolers, and final-cooler cooling towers. (a) No (“zero”) emissions are allowed from naphthalene processing, final coolers and final-cooler cooling towers at coke by...

Inductive classification of operating data from a fluidized bed calciner

International Nuclear Information System (INIS)

O'Brien, B.H.

1990-01-01

A process flowsheet expert system for a fluidized bed calciner which solidifies high-level radioactive liquid waste was developed from pilot-plant data using a commercial, inductive classification program. After initial classification of the data, the resulting rules were inspected and adjusted to match existing knowledge of process chemistry. The final expert system predicts performance of process flowsheets based upon the chemical composition of the calciner feed and has been successfully used to identify potential operational problems prior to calciner pilot-plant testing of new flowsheets and to provide starting parameters for pilot-plant tests. By using inductive classification techniques to develop the initial rules from the calciner pilot-plant data and using existing process knowledge to verify the accuracy of these rules, an effective expert system was developed with a minimum amount of effort. This method may be applied for developing expert systems for other processes where numerous operating data are available and only general process chemistry effects are known

The conceptual flowsheet of effluent treatment during total gelation of uranium process for preparing ceramic UO2 particles of high temperature gas-cooled reactors

International Nuclear Information System (INIS)

Quan Ying; Chen Xiaotong; Wang Yang; Liu Bing; Tang Yaping; Tang Chunhe

2014-01-01

Today, more and more people pay attention to the environmental protection and ecological environment. Along with the development of nuclear industry, many radioactive effluents may be discharged into environment, which can lead to the pollutions of water, atmosphere and soil. So radioactive effluents including low-activity and medium-level wastes solution treatments have been becoming one of significant subjects. High temperature gas-cooled reactor (HTR) is one of advanced nuclear reactors owing to its reliability, security and broad application in which the fabrication of spherical fuel element is a key technology. During the production of spherical fuel elements, the radioactive effluent treatment is necessary. Referring to the current treatment technologies and methods, the conceptual flowsheet of low-level radioactive effluent treatment during preparing spherical fuel elements was summarized which met the 'Zero Emission' demand. (authors)

Process modeling for the Integrated Nonthermal Treatment System (INTS) study

Energy Technology Data Exchange (ETDEWEB)

Brown, B.W.

1997-04-01

This report describes the process modeling done in support of the Integrated Nonthermal Treatment System (INTS) study. This study was performed to supplement the Integrated Thermal Treatment System (ITTS) study and comprises five conceptual treatment systems that treat DOE contract-handled mixed low-level wastes (MLLW) at temperatures of less than 350{degrees}F. ASPEN PLUS, a chemical process simulator, was used to model the systems. Nonthermal treatment systems were developed as part of the INTS study and include sufficient processing steps to treat the entire inventory of MLLW. The final result of the modeling is a process flowsheet with a detailed mass and energy balance. In contrast to the ITTS study, which modeled only the main treatment system, the INTS study modeled each of the various processing steps with ASPEN PLUS, release 9.1-1. Trace constituents, such as radionuclides and minor pollutant species, were not included in the calculations.

Demonstration of the Defense Waste Processing Facility vitrification process for Tank 42 radioactive sludge -- Glass preparation and characterization

International Nuclear Information System (INIS)

Bibler, N.E.; Fellinger, T.L.; Marshall, K.M.; Crawford, C.L.; Cozzi, A.D.; Edwards, T.B.

1999-01-01

The Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) is currently processing and immobilizing the radioactive high level waste sludge at SRS into a durable borosilicate glass for final geological disposal. The DWPF has recently finished processing the first radioactive sludge batch, and is ready for the second batch of radioactive sludge. The second batch is primarily sludge from Tank 42. Before processing this batch in the DWPF, the DWPF process flowsheet has to be demonstrated with a sample of Tank 42 sludge to ensure that an acceptable melter feed and glass can be made. This demonstration was recently completed in the Shielded Cells Facility at SRS. An earlier paper in these proceedings described the sludge composition and processes necessary for producing an acceptable melter fee. This paper describes the preparation and characterization of the glass from that demonstration. Results substantiate that Tank 42 sludge after mixing with the proper amount of glass forming frit (Frit 200) can be processed to make an acceptable glass

Report for Treating Hanford LAW and WTP SW Simulants: Pilot Plant Mineralizing Flowsheet

International Nuclear Information System (INIS)

Olson, Arlin

2012-01-01

The US Department of Energy is responsible for managing the disposal of radioactive liquid waste in underground storage tanks at the Hanford site in Washington State. The Hanford waste treatment and immobilization plant (WPT) will separate the waste into a small volume of high level waste (HLW), containing most of the radioactive constituents, and a larger volume of low activity waste (LAW), containing most of the non-radioactive chemical and hazardous constituents. The HLW and LAW will be converted into immobilized waste forms for disposal. Currently there is inadequate LAW vitrification capacity planned at the WTP to complete the mission within the required timeframe. Therefore additional LAW capacity is required. One candidate supplemental treatment technology is the fluidized bed steam reformer process (FBSR). This report describes the demonstration testing of the FBSR process using a mineralizing flowsheet for treating simulated Hanford LAW and secondary waste from the WTP (WTP SW). The FBSR testing project produced leach-resistant solid products and environmentally compliant gaseous effluents. The solid products incorporated normally soluble ions into an alkali alumino-silicate (NaS) mineral matrix. Gaseous emissions were found to be within regulatory limits. Cesium and rhenium were captured in the mineralized products with system removal efficiencies of 99.999% and 99.998 respectively. The durability and leach performance of the FBSR granular solid were superior to the low activity reference material (LMR) glass standards. Normalized product consistency test (PCT) release rates for constituents of concern were approximately 2 orders of magnitude less than that of sodium in the Hanford glass [standard].

Report for Treating Hanford LAW and WTP SW Simulants: Pilot Plant Mineralizing Flowsheet

Energy Technology Data Exchange (ETDEWEB)

Arlin Olson

2012-02-28

The US Department of Energy is responsible for managing the disposal of radioactive liquid waste in underground storage tanks at the Hanford site in Washington State. The Hanford waste treatment and immobilization plant (WPT) will separate the waste into a small volume of high level waste (HLW), containing most of the radioactive constituents, and a larger volume of low activity waste (LAW), containing most of the non-radioactive chemical and hazardous constituents. The HLW and LAW will be converted into immobilized waste forms for disposal. Currently there is inadequate LAW vitrification capacity planned at the WTP to complete the mission within the required timeframe. Therefore additional LAW capacity is required. One candidate supplemental treatment technology is the fluidized bed steam reformer process (FBSR). This report describes the demonstration testing of the FBSR process using a mineralizing flowsheet for treating simulated Hanford LAW and secondary waste from the WTP (WTP SW). The FBSR testing project produced leach-resistant solid products and environmentally compliant gaseous effluents. The solid products incorporated normally soluble ions into an alkali alumino-silicate (NaS) mineral matrix. Gaseous emissions were found to be within regulatory limits. Cesium and rhenium were captured in the mineralized products with system removal efficiencies of 99.999% and 99.998 respectively. The durability and leach performance of the FBSR granular solid were superior to the low activity reference material (LMR) glass standards. Normalized product consistency test (PCT) release rates for constituents of concern were approximately 2 orders of magnitude less than that of sodium in the Hanford glass [standard].

Conceptual design of heterogeneous azeotropic distillation process for ethanol dehydration using 1-butanol as entrainer

Directory of Open Access Journals (Sweden)

Paritta Prayoonyong

2014-12-01

Full Text Available The synthesis of a heterogeneous azeotropic distillation process for ethanol dehydration using 1-butanol as entrainer is presented. The residue curve map of the ethanol/water/1-butanol mixture is computationally generated using non-random twoliquid thermodynamic model. It is found that 1-butanol leads to a residue curve map topological structure different from that generated by typical entrainers used in ethanol dehydration. Synthesised by residue curve map analysis, the distillation flowsheet for ethanol dehydration by 1-butanol comprises a double-feed column integrated with an overhead decanter and a simple column. The double-feed column is used to recover water as the top product, whereas the simple column is used for recovering ethanol and 1-butanol. The separation feasibility and the economically near-optimal designs of distillation columns in the flowsheet are evaluated and identified by using the boundary value design method. The distillation flowsheet using 1-butanol is compared with the conventional process using benzene as entrainer. Based on their total annualised costs, the ethanol dehydration process using 1-butanol is less economically attractive than the process using benzene. However, 1-butanol is less toxic than benzene.

Improvements mineral dressing and extraction processes of gold-silver ores from San Pedro Frio Mining District, Colombia

International Nuclear Information System (INIS)

Yanez Traslavina, J. J.; Vargas Avila, M. A.; Garcia Paez, I. H.; Pedraza Rosas, J. E.

2005-01-01

The San Pedro Frio district mining, Colombia, is a rich region production gold-silver ores. Nowadays, the extraction processes used are amalgamation, percolation cyanidation and precipitation with zinc wood. Due to the ignorance of the ore characteristics, gold and silver treatment processes are inadequate and not efficient. In addition the inappropriate use of mercury and cyanide cause environmental contamination. In this research the ore characterization was carried out obtained fundamental parameters for the technical selection of more efficient gold and silver extraction processes. Experimental work was addressed to the study of both processes the agitation cyanidation and the adsorption on activated carbon in pulp. As a final result proposed a flowsheet to improve the precious metals recovery and reduce the environment contamination. (Author)

Dynamic Model-Based Evaluation of Process Configurations for Integrated Operation of Hydrolysis and Co-Fermentation for Bioethanol Production from Lignocellulose

DEFF Research Database (Denmark)

Morales Rodriguez, Ricardo; Meyer, Anne S.; Gernaey, Krist

2011-01-01

In this study a number of different process flowsheets were generated and their feasibility evaluated using simulations of dynamic models. A dynamic modeling framework was used for the assessment of operational scenarios such as, fed-batch, continuous and continuous with recycle configurations. E......) operating in continuous mode with a recycle of the SSCF reactor effluent, results in the best productivity of bioethanol among the proposed process configurations, with a yield of 0.18 kg ethanol/kg dry-biomass........ Each configuration was evaluated against the following benchmark criteria, yield (kg ethanol/kg dry-biomass), final product concentration and number of unit operations required in the different process configurations. The results show that simultaneous saccharification and co-fermentation (SSCF...

FLOWSHEET FOR ALUMINUM REMOVAL FROM SLUDGE BATCH 6

International Nuclear Information System (INIS)

Pike, J.; Gillam, J.

2008-01-01

Samples of Tank 12 sludge slurry show a substantially larger fraction of aluminum than originally identified in sludge batch planning. The Liquid Waste Organization (LWO) plans to formulate Sludge Batch 6 (SB6) with about one half of the sludge slurry in Tank 12 and one half of the sludge slurry in Tank 4. LWO identified aluminum dissolution as a method to mitigate the effect of having about 50% more solids in High Level Waste (HLW) sludge than previously planned. Previous aluminum dissolution performed in a HLW tank in 1982 was performed at approximately 85 C for 5 days and dissolved nearly 80% of the aluminum in the sludge slurry. In 2008, LWO successfully dissolved 64% of the aluminum at approximately 60 C in 46 days with minimal tank modifications and using only slurry pumps as a heat source. This report establishes the technical basis and flowsheet for performing an aluminum removal process in Tank 51 for SB6 that incorporates the lessons learned from previous aluminum dissolution evolutions. For SB6, aluminum dissolution process temperature will be held at a minimum of 65 C for at least 24 days, but as long as practical or until as much as 80% of the aluminum is dissolved. As planned, an aluminum removal process can reduce the aluminum in SB6 from about 84,500 kg to as little as 17,900 kg with a corresponding reduction of total insoluble solids in the batch from 246,000 kg to 131,000 kg. The extent of the reduction may be limited by the time available to maintain Tank 51 at dissolution temperature. The range of dissolution in four weeks based on the known variability in dissolution kinetics can range from 44 to more than 80%. At 44% of the aluminum dissolved, the mass reduction is approximately 1/2 of the mass noted above, i.e., 33,300 kg of aluminum instead of 66,600 kg. Planning to reach 80% of the aluminum dissolved should allow a maximum of 81 days for dissolution and reduce the allowance if test data shows faster kinetics. 47,800 kg of the dissolved

A Hybrid MPC-PID Control System Design for the Continuous Purification and Processing of Active Pharmaceutical Ingredients

Directory of Open Access Journals (Sweden)

Maitraye Sen

2014-05-01

Full Text Available In this work, a hybrid MPC (model predictive control-PID (proportional-integral-derivative control system has been designed for the continuous purification and processing framework of active pharmaceutical ingredients (APIs. The specific unit operations associated with the purification and processing of API have been developed from first-principles and connected in a continuous framework in the form of a flowsheet model. These integrated unit operations are highly interactive along with the presence of process delays. Therefore, a hybrid MPC-PID is a promising alternative to achieve the desired control loop performance as mandated by the regulatory authorities. The integrated flowsheet model has been simulated in gPROMSTM (Process System Enterprise, London, UK. This flowsheet model has been linearized in order to design the control scheme. The ability to track the set point and reject disturbances has been evaluated. A comparative study between the performance of the hybrid MPC-PID and a PID-only control scheme has been presented. The results show that an enhanced control loop performance can be obtained under the hybrid control scheme and demonstrate that such a scheme has high potential in improving the efficiency of pharmaceutical manufacturing operations.

Case study on chemical plant accidents for flow-sheet design of the HTTR-IS system

International Nuclear Information System (INIS)

Homma, Hiroyuki; Sato, Hiroyuki; Kasahara, Seiji; Hara, Teruo; Kato, Ryoma; Sakaba, Nariaki; Ohashi, Hirofumi

2007-02-01

At the present time, we are alarmed by depletion of fossil energy and adverse effect of rapid increase in fossil fuel burning on environment such as climate changes and acid rain, because our lives depend still heavily upon fossil energy. It is thus widely recognized that hydrogen is one of important future energy carriers in which it is used without emission of carbon dioxide greenhouse gas and atmospheric pollutants and that hydrogen demand will increase greatly as fuel cells are developed and applied widely in the near future. To meet massive demand of hydrogen, hydrogen production from water utilizing nuclear, especially by thermochemical water-splitting Iodine-Sulphur (IS) process utilizing heat from High-Temperature Gas-cooled Reactors (HTGRs), offers one of the most attractive zero-emission energy strategies and the only one practical on a substantial scale. However, to establish a technology based for the HTGR hydrogen production by the IS process, we should close several technology gaps through R and D with the High-Temperature Engineering Test Reactor (HTTR), which is the only Japanese HTGR built and operated at the Oarai Research and Development Centre of Japan Atomic Energy Agency (JAEA). We have launched design studies of the IS process hydrogen production system coupled with the HTTR (HTTR-IS system) to demonstrate HTGR hydrogen production. In designing the HTTR-IS system, it is necessary to consider preventive and breakdown maintenance against accidents occurred in the IS process as a chemical plant. This report describes case study on chemical plant accidents relating to the IS process plant and shows a proposal of accident protection measures based on above case study, which is necessary for flow-sheet design of the HTTR-IS system. (author)

Am/Cm Vitrification Process: Vitrification Material Balance Calculations

International Nuclear Information System (INIS)

Smith, F.G.

2000-01-01

This report documents material balance calculations for the Americium/Curium vitrification process and describes the basis used to make the calculations. The material balance calculations reported here start with the solution produced by the Am/Cm pretreatment process as described in ``Material Balance Calculations for Am/Cm Pretreatment Process (U)'', SRT-AMC-99-0178 [1]. Following pretreatment, small batches of the product will be further treated with an additional oxalic acid precipitation and washing. The precipitate from each batch will then be charged to the Am/Cm melter with glass cullet and vitrified to produce the final product. The material balance calculations in this report are designed to provide projected compositions of the melter glass and off-gas streams. Except for decanted supernate collected from precipitation and precipitate washing, the flowsheet neglects side streams such as acid washes of empty tanks that would go directly to waste. Complete listings of the results of the material balance calculations are provided in the Appendices to this report

Basic TRUEX process for Rocky Flats Plant

International Nuclear Information System (INIS)

Leonard, R.A.; Chamberlain, D.B.; Dow, J.A.; Farley, S.E.; Nunez, L.; Regalbuto, M.C.; Vandegrift, G.F.

1994-08-01

The Generic TRUEX Model was used to develop a TRUEX process flowsheet for recovering the transuranics (Pu, Am) from a nitrate waste stream at Rocky Flats Plant. The process was designed so that it is relatively insensitive to changes in process feed concentrations and flow rates. Related issues are considered, including solvent losses, feed analysis requirements, safety, and interaction with an evaporator system for nitric acid recycle

Modelling Template for the Development of the Process Flowsheet

DEFF Research Database (Denmark)

Fedorova, Marina; Gani, Rafiqul

2015-01-01

Models are playing important roles in design and analysis of chemicals/bio-chemicals based products and the processes that manufacture them. Model-based methods and tools have the potential to decrease the number of experiments, which can be expensive and time consuming, and point to candidates...... in connection to other modelling tools within the modelling framework are forming a user-friendly system, which will make the model development process easier and faster and provide the way for unified and consistent model documentation. The modeller can use the template for their specific problem or to extend...... models systematically, efficiently and reliably. In this way, development of products and processes can be faster, cheaper and very efficient. The developed modelling framework involves three main parts: 1) a modelling tool, that includes algorithms for model generation; 2) a template library, which...

Perspectives on Multienzyme Process Technology

DEFF Research Database (Denmark)

Santacoloma, Paloma A.; Woodley, John M.

2014-01-01

. One consequence is that decisions about the format of the biocatalyst and reactor type as well as the process flowsheet require more extensive knowledge. In this chapter, some of the background to these decisions and decision-making tools to help establish effective multienzyme processes in a timely......There is little doubt that chemical processing of the future will involve an increasing number of biocatalytic processes using more than one enzyme. There are good reasons for developing such innovative biocatalytic processes and interesting new biocatalyst and process options will be introduced...

Process developments in gasoil hydrotreating

Energy Technology Data Exchange (ETDEWEB)

Lawrence, R.C.; Kinley, D.H.; Wood, M.A. [Davy Process Technology Limited, London (United Kingdom)

1997-07-01

Changing demand patterns and legislation increase the pressure upon hydrotreating capacities at many refineries. To meet these pressures, improvements have been and will be necessary not only in catalysts, but also in the hydrotreating process. On the basis of its hydrogenation experience, Davy Process Technology has developed and tested a number of concepts aimed at improving the effectiveness of the basic process - enabling economic deep desulfurisation and opening up the potential for an integrated HDS/HDA flowsheet using sulphur tolerant HDA Catalysts.

Dissolution Flowsheet for High Flux Isotope Reactor Fuel

Energy Technology Data Exchange (ETDEWEB)

Daniel, W. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Rudisill, T. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); O' Rourke, P. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Karay, N. S [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-09-27

As part of the Spent Nuclear Fuel (SNF) processing campaign, H-Canyon is planning to begin dissolving High Flux Isotope Reactor (HFIR) fuel in late FY17 or early FY18. Each HFIR fuel core contains inner and outer fuel elements which were fabricated from uranium oxide (U₃O₈) dispersed in a continuous Al phase using traditional powder metallurgy techniques. Fuels fabricated in this manner, like other SNF’s processed in H-Canyon, dissolve by the same general mechanisms with similar gas generation rates and the production of H₂. The HFIR fuel cores will be dissolved and the recovered U will be down-blended into low-enriched U. HFIR fuel was previously processed in H-Canyon using a unique insert in both the 6.1D and 6.4D dissolvers. Multiple cores will be charged to the same dissolver solution maximizing the concentration of dissolved Al. The objective of this study was to identify flowsheet conditions through literature review and laboratory experimentation to safely and efficiently dissolve the HFIR fuel in H-Canyon. Laboratory-scale experiments were performed to evaluate the dissolution of HFIR fuel using both Al 1100 and Al 6061 T6 alloy coupons. The Al 1100 alloy was considered a representative surrogate which provided an upper bound on the generation of flammable (i.e., H₂) gas during the dissolution process. The dissolution of the Al 6061 T6 alloy proceeded at a slower rate than the Al 1100 alloy, and was used to verify that the target Al concentration in solution could be achieved for the selected Hg concentration. Mass spectrometry and Raman spectroscopy were used to provide continuous monitoring of the concentration of H₂ and other permanent gases in the dissolution offgas, allowing the development of H₂ generation rate profiles. The H₂ generation rates were subsequently used to evaluate if a full HFIR core could be dissolved in an H-Canyon dissolver without exceeding 60% of the

Determination of fission product and heavy metals inventories in FTE-4 fuel rods by a grind-burn-leach flowsheet

International Nuclear Information System (INIS)

Fitzgerald, C.L.; Vaughen, V.C.A.; Lamb, C.E.

1977-07-01

Experiments using High-Temperature Gas-Cooled Reactor (HTGR) fuel material, TRISO-coated (2.75 Th/U)C 2 --TRISO-coated ThC 2 and TRISO-coated UO 2 --BISO-coated ThO 2 , were performed in Building 4507 (the High-Level Chemical Development Facility) to determine the inventory and transport behavior of fission products and heavy metals from a grind-burn-leach process flowsheet. In addition, values calculated by the ORNL Isotope Generation and Depletion Code (ORIGEN, a computer program used for predicting quantities of activation products, actinides, and fission products from irradiation data and nuclear data libraries) are compared with values derived by chemical analyses (CA) and those measured by a gamma-scan nondestructive analytical (NDA) technique. Reasonable agreement was obtained between ORIGEN and NDA results for one of the tests, but the values obtained by chemical analysis were lower than either of the two other sets of values. With the exception of 234 U, isotopic uranium values determined by chemical analysis (mass spectrometry) agreed within 15 percent of the ORIGEN prediction

Zirconium and technetium recovery and partitioning in the presence of actinides in modified Purex process for ATW program. Final report

International Nuclear Information System (INIS)

Dzekun, E.G.; Fedorov, Y.S.; Galkin, B.Y.; Lyubtsev, R.I.; Mashkin, A.N.; Mishin, E.N.; Zilberman, B.Y.

1994-01-01

The modified Purex process flowsheet is based on combination of all irradiated materials, their joint dissolution and reprocessing as a NPP spent fuel solution with abnormal Pu content after addition of recycled depleted U concentrate. Some groups of long-lived radionuclides could be completely recovered and localized at the stage of extraction reprocessing using 30% TBP. Studies were conducted for 10 y to develop the process for recovery, concentration, and localization of U, Pu, Np, Tc, and Zr within 1st extraction cycle. Actinides are recovered from high-level raffinate of this cycle after evaporation and feed adjustment. Results in this report show that combined deep recovery of several elements from highly irradiated materials by TBP extraction, for further transmutation, is possible. Selective stripping of Zr from solvent phase containing U, Pu, Np, and Tc is quite effective. Development of the modified Purex process is not complete; main problem to be solved should be oxide separation from the loop and permissible storage duration before reprocessing and reuse in the loop

Zirconium and technetium recovery and partitioning in the presence of actinides in modified Purex process for ATW program. Final report

Energy Technology Data Exchange (ETDEWEB)

Dzekun, E.G.; Fedorov, Y.S.; Galkin, B.Y.; Lyubtsev, R.I.; Mashkin, A.N.; Mishin, E.N.; Zilberman, B.Y. [Radievyj Inst., Leningrad (Russian Federation)

1994-12-31

The modified Purex process flowsheet is based on combination of all irradiated materials, their joint dissolution and reprocessing as a NPP spent fuel solution with abnormal Pu content after addition of recycled depleted U concentrate. Some groups of long-lived radionuclides could be completely recovered and localized at the stage of extraction reprocessing using 30% TBP. Studies were conducted for 10 y to develop the process for recovery, concentration, and localization of U, Pu, Np, Tc, and Zr within 1st extraction cycle. Actinides are recovered from high-level raffinate of this cycle after evaporation and feed adjustment. Results in this report show that combined deep recovery of several elements from highly irradiated materials by TBP extraction, for further transmutation, is possible. Selective stripping of Zr from solvent phase containing U, Pu, Np, and Tc is quite effective. Development of the modified Purex process is not complete; main problem to be solved should be oxide separation from the loop and permissible storage duration before reprocessing and reuse in the loop.

Separation of technetium and rare earth metals for co-decontamination process

International Nuclear Information System (INIS)

Riddle, Catherine; Martin, Leigh

2015-01-01

Poster. In the US there are several technologies under consideration for the separation of the useful components in used nuclear fuel. One such process is the co-decontamination process to separate U, Np and Pu in a single step and produce a Np/ Pu and a U product stream. Although the behavior of the actinide elements is reasonably well defined in this system, the same is not true for the fission products, mainly Zr, Mo, Ru and Tc. As these elements are cationic and anionic they may interact with each other to extract in a manner not predicted by empirical models such as AMUSE. This poster presentation will discuss the initial results of batch contact testing under flowsheet conditions and as a function of varying acidity and flowsheet conditions to optimize recovery of Tc and minimize extraction of Mo, Zr and Ru with the goal of developing a better understanding of the behavior of these elements in the co-decontamination process.

Simulation-Assisted Evaluation of Grinding Circuit Flowsheet Design Alternatives: Aghdarreh Gold Ore Processing Plant / Ocena Alternatywnych Schematów Technologicznych Procesu Rozdrabniania W Zakładach Przeróbki Rud Złota W Aghdarreh, Z Wykorzystaniem Metod Symulacji

Science.gov (United States)

Farzanegan, A.; Ghalaei, A. Ebtedaei

2015-03-01

The run of mine ore from Aghdarreh gold mine must be comminuted to achieve the desired degree of liberation of gold particles. Currently, comminution circuits include a single-stage crushing using a jaw crusher and a single-stage grinding using a Semi-Autogenous Grinding (SAG) mill in closed circuit with a hydrocyclone package. The gold extraction is done by leaching process using cyanidation method through a series of stirred tanks. In this research, an optimization study of Aghdarreh plant grinding circuit performance was done to lower the product particle size (P80) from 70 μm to approximately 40 μm by maintaining current throughput using modeling and simulation approach. After two sampling campaigns from grinding circuit, particle size distribution data were balanced using NorBal software. The first and second data sets obtained from the two sampling campaigns were used to calibrate necessary models and validate them prior to performing simulation trials using MODSIM software. Computer simulations were performed to assess performance of two proposed new circuit flowsheets. The first proposed flowsheet consists of existing SAG mill circuit and a new proposed ball mill in closed circuit with a new second hydrocyclone package. The second proposed flowsheet consists of existing SAG mill circuit followed by a new proposed ball mill in closed circuit with the existing hydrocyclone package. In all simulations, SAGT, CYCL and MILL models were selected to simulate SAG mill, Hydrocyclone packages and ball mill units. SAGT and MILL models both are based on population balance model of grinding process. CYCL model is based on Plitt's empirical model of classification process in hydrocyclone units. It was shown that P80 can be reduced to about 40 μm and 42 μm for the first and second proposed circuits, respectively. Based on capital and operational costs, it can be concluded that the second proposed circuit is a more suitable option for plant grinding flowsheet

Process Simulation of Biobutanol Production from Lignocellulosic Feedstocks

NARCIS (Netherlands)

Procentese, A.; Guida, T.; Raganati, F.; Olivieri, G.; Salatino, P.; Marzocchella, A.

2014-01-01

A potential flowsheet to produce butanol production by conversion of a lignocellulosic biomass has been simulated by means of the software Aspen Plus®. The flowsheet has included upstream, fermentation, and downstream sections and the attention has been focused on the upstream section. The proposed

A flowsheet model of a coal-fired MHD/steam combined electricity generating cycle, using the access computer model

International Nuclear Information System (INIS)

Davison, J.E.; Eldershaw, C.E.

1992-01-01

This document forms the final report on a study of a coal-fired magnetohydrodynamic (MHD)/steam electric power generation system carried out by British Coal Corporation for the Commission of the European Communities. The study objective was to provide mass and energy balances and overall plant efficiency predictions for MHD to assist the Commission in their evaluation of advanced power generation technologies. In early 1990 the British Coal Corporation completed a study for the Commission in which a computer flowsheet modelling package was used to predict the performance of a conceptual air blown MHD plant. Since that study was carried out increasing emphasis has been placed on the possible need to reduce CO 2 emissions to counter the so-called greenhouse effect. Air blown MHD could greatly reduce CO 2 emissions per KWh by virtue of its high thermal efficiency. However, if even greater reductions in CO 2 emissions were required the CO 2 produced by coal combustion may have to be disposed of, for example into the deep ocean or underground caverns. To achieve this at minimum cost a concentrated CO 2 flue gas would be required. This could be achieved in an MHD plant by using a mixture of high purity oxygen and recycled CO 2 flue gas in the combustor. To assess this plant concept the European Commission awarded British Coal a contract to produce performance predictions using the access computer program

Adaptation of U(IV) reductant to Savannah River Plant Purex processes

International Nuclear Information System (INIS)

Orebaugh, E.G.

1986-04-01

Partitioning of uranium and plutonium in the Purex process requires the reduction of the extracted Pu(IV) to the less extractable Pu(III). This valence adjustment at SRP has historically been performed by the addition of ferrous ion, which eventually constitutes a major component of high-level waste solids requiring costly permanent disposal. Uranous nitrate, U(IV), is a kinetically fast reductant which may be substituted for Fe(II) without contributing to waste solids. This report documents U(IV) flowsheet development in the miniature mixer-settler equipment at SRL and provides an insight into the mechanisms responsible for the successful direct substitution of U(IV) for Fe(II) in 1B bank extractant. U(IV) will be the reductant of choice when its fast reduction kinetics are required in centrifugal-contactor-based processing. The flowsheets investigated here should transfer to such equipment with minimal modifications

Final-state interaction in processes of deuteron breaking

International Nuclear Information System (INIS)

Thome Filho, Z.D.

1974-12-01

Interaction between particles in the final state of reactions can strongly affect the experimental angular distributions, as in the scattering processes with the breaking of the deuteron target, where the final state interaction is responsible for the disappearance of the differential cross section in the front direction. It is then necessary to include the contribution of the final state interaction to small angles of incoherent processes particle-deuteron. In this work line, an analysis is made of the process πd → πpn for different values of the incident energy. The data obtained are compared with existing experimental data. The hypothesis is also considered of the nucleon which collides with the incident particle being outside the mass layer. An analytical extension of the resonant amplitude πN outwards the mass layer is also used

Hadron final states in deep inelastic processes

International Nuclear Information System (INIS)

Bjorken, J.D.

1976-05-01

Lectures are presented dealing mainly with the description and discussion of hadron final states in electroproduction, colliding beams, and neutrino reactions from the point of view of the simple parton model. Also the space-time evolution of final states in the parton model is considered. It is found that the picture of space-time evolution of hadron final states in deep inelastic processes isn't totally trivial and that it can be made consistent with the hypotheses of the parton model. 39 references

Evaluation of alternative chemical additives for high-level waste vitrification feed preparation processing

International Nuclear Information System (INIS)

Seymour, R.G.

1995-01-01

During the development of the feed processing flowsheet for the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS), research had shown that use of formic acid (HCOOH) could accomplish several processing objectives with one chemical addition. These objectives included the decomposition of tetraphenylborate, chemical reduction of mercury, production of acceptable rheological properties in the feed slurry, and controlling the oxidation state of the glass melt pool. However, the DEPF research had not shown that some vitrification slurry feeds had a tendency to evolve hydrogen (H 2 ) and ammonia (NH 3 ) as the result of catalytic decomposition of CHOOH with noble metals (rhodium, ruthenium, palladium) in the feed. Testing conducted at Pacific Northwest Laboratory and later at the Savannah River Technical Center showed that the H 2 and NH 3 could evolve at appreciable rates and quantities. The explosive nature of H 2 and NH 3 (as ammonium nitrate) warranted significant mitigation control and redesign of both facilities. At the time the explosive gas evolution was discovered, the DWPF was already under construction and an immediate hardware fix in tandem with flowsheet changes was necessary. However, the Hanford Waste Vitrification Plant (HWVP) was in the design phase and could afford to take time to investigate flowsheet manipulations that could solve the problem, rather than a hardware fix. Thus, the HWVP began to investigate alternatives to using HCOOH in the vitrification process. This document describes the selection, evaluation criteria, and strategy used to evaluate the performance of the alternative chemical additives to CHOOH. The status of the evaluation is also discussed

Nitric-glycolic flowsheet reduction/oxidation (redox) model for the defense waste processing facility (DWPF)

Energy Technology Data Exchange (ETDEWEB)

Jantzen, C. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Williams, M. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Edwards, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Trivelpiece, C. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Ramsey, W. G. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2017-06-14

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, retains radionuclides in the melt and thus the final glass. Specifically, long-lived radioactive ⁹⁹Tc species are less volatile in the reduced Tc⁴⁺ state as TcO₂ than as NaTcO₄ or Tc₂O₇, and ruthenium radionuclides in the reduced Ru⁴⁺ state are insoluble RuO₂ in the melt which are not as volatile as NaRuO₄ where the Ru is in the +7 oxidation state. Similarly, hazardous volatile Cr⁶⁺ occurs in oxidized melt pools as Na₂CrO₄ or Na₂Cr₂O₇, while the Cr⁺³ state is less volatile and remains in the melt as NaCrO₂ or precipitates as chrome rich spinels. The melter REDOX control balances the oxidants and reductants from the feed and from processing additives such as antifoam.

Radioactive demonstration of the ''late wash'' Precipitate Hydrolysis Process

International Nuclear Information System (INIS)

Bibler, N.E.; Ferrara, D.M.; Ha, B.C.

1992-01-01

This report presents results of the radioactive demonstration of the DWPF Precipitate Hydrolysis Process as it would occur in the ''late wash'' flowsheet in the absence of hydroxylamine nitrate. Radioactive precipitate containing Cs-137 from the April, 1983, in-tank precipitation demonstration in Tank 48 was used for these tests

Modeling a novel glass immobilization waste treatment process using flow

International Nuclear Information System (INIS)

Ferrada, J.J.; Nehls, J.W. Jr.; Welch, T.D.; Giardina, J.L.

1996-01-01

One option for control and disposal of surplus fissile materials is the Glass Material Oxidation and Dissolution System (GMODS), a process developed at ORNL for directly converting Pu-bearing material into a durable high-quality glass waste form. This paper presents a preliminary assessment of the GMODS process flowsheet using FLOW, a chemical process simulator. The simulation showed that the glass chemistry postulated ion the models has acceptable levels of risks

Electrochemical processing of spent nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Williamson, M. A.; Willit, J. L.; Barnes, L. A.; Figueroa, J.; Limmer, S. L.; Blaskovitz, R. [Argonne National Laboratory, Argonne (United States)

2008-08-15

Our work in developing the fuel cycles and electrochemical technologies needed for the treatment of spent light water reactor and spent fast reactor fuel is progressing well. Baseline flowsheets along with a theoretical material balance have been developed for treatment of each type of fuel. A discussion about the flowsheets provides the opportunity to present the status of our technology development activities and future research and development directions.

Electrochemical processing of spent nuclear fuel

International Nuclear Information System (INIS)

Williamson, M. A.; Willit, J. L.; Barnes, L. A.; Figueroa, J.; Limmer, S. L.; Blaskovitz, R.

2008-01-01

Our work in developing the fuel cycles and electrochemical technologies needed for the treatment of spent light water reactor and spent fast reactor fuel is progressing well. Baseline flowsheets along with a theoretical material balance have been developed for treatment of each type of fuel. A discussion about the flowsheets provides the opportunity to present the status of our technology development activities and future research and development directions

Pretreatment Engineering Platform Phase 1 Final Test Report

Energy Technology Data Exchange (ETDEWEB)

Kurath, Dean E.; Hanson, Brady D.; Minette, Michael J.; Baldwin, David L.; Rapko, Brian M.; Mahoney, Lenna A.; Schonewill, Philip P.; Daniel, Richard C.; Eslinger, Paul W.; Huckaby, James L.; Billing, Justin M.; Sundar, Parameshwaran S.; Josephson, Gary B.; Toth, James J.; Yokuda, Satoru T.; Baer, Ellen BK; Barnes, Steven M.; Golovich, Elizabeth C.; Rassat, Scot D.; Brown, Christopher F.; Geeting, John GH; Sevigny, Gary J.; Casella, Amanda J.; Bontha, Jagannadha R.; Aaberg, Rosanne L.; Aker, Pamela M.; Guzman-Leong, Consuelo E.; Kimura, Marcia L.; Sundaram, S. K.; Pires, Richard P.; Wells, Beric E.; Bredt, Ofelia P.

2009-12-23

Pacific Northwest National Laboratory (PNNL) was tasked by Bechtel National Inc. (BNI) on the River Protection Project, Hanford Tank Waste Treatment and Immobilization Plant (RPP-WTP) project to conduct testing to demonstrate the performance of the WTP Pretreatment Facility (PTF) leaching and ultrafiltration processes at an engineering-scale. In addition to the demonstration, the testing was to address specific technical issues identified in Issue Response Plan for Implementation of External Flowsheet Review Team (EFRT) Recommendations - M12, Undemonstrated Leaching Processes.( ) Testing was conducted in a 1/4.5-scale mock-up of the PTF ultrafiltration system, the Pretreatment Engineering Platform (PEP). Parallel laboratory testing was conducted in various PNNL laboratories to allow direct comparison of process performance at an engineering-scale and a laboratory-scale. This report presents and discusses the results of those tests.

Pretreatment Engineering Platform Phase 1 Final Test Report

International Nuclear Information System (INIS)

Kurath, Dean E.; Hanson, Brady D.; Minette, Michael J.; Baldwin, David L.; Rapko, Brian M.; Mahoney, Lenna A.; Schonewill, Philip P.; Daniel, Richard C.; Eslinger, Paul W.; Huckaby, James L.; Billing, Justin M.; Sundar, Parameshwaran S.; Josephson, Gary B.; Toth, James J.; Yokuda, Satoru T.; Baer, Ellen B.K.; Barnes, Steven M.; Golovich, Elizabeth C.; Rassat, Scot D.; Brown, Christopher F.; Geeting, John G.H.; Sevigny, Gary J.; Casella, Amanda J.; Bontha, Jagannadha R.; Aaberg, Rosanne L.; Aker, Pamela M.; Guzman-Leong, Consuelo E.; Kimura, Marcia L.; Sundaram, S.K.; Pires, Richard P.; Wells, Beric E.; Bredt, Ofelia P.

2009-01-01

Pacific Northwest National Laboratory (PNNL) was tasked by Bechtel National Inc. (BNI) on the River Protection Project, Hanford Tank Waste Treatment and Immobilization Plant (RPP-WTP) project to conduct testing to demonstrate the performance of the WTP Pretreatment Facility (PTF) leaching and ultrafiltration processes at an engineering-scale. In addition to the demonstration, the testing was to address specific technical issues identified in Issue Response Plan for Implementation of External Flowsheet Review Team (EFRT) Recommendations - M12, Undemonstrated Leaching Processes. Testing was conducted in a 1/4.5-scale mock-up of the PTF ultrafiltration system, the Pretreatment Engineering Platform (PEP). Parallel laboratory testing was conducted in various PNNL laboratories to allow direct comparison of process performance at an engineering-scale and a laboratory-scale. This report presents and discusses the results of those tests.

Phenomena based Methodology for Process Synthesis incorporating Process Intensification

DEFF Research Database (Denmark)

Lutze, Philip; Babi, Deenesh Kavi; Woodley, John

2013-01-01

at processes at the lowest level of aggregation which is the phenomena level. In this paper, a phenomena based synthesis/design methodology incorporating process intensification is presented. Using this methodology, a systematic identification of necessary and desirable (integrated) phenomena as well......Process intensification (PI) has the potential to improve existing as well as conceptual processes, in order to achieve a more sustainable production. PI can be achieved at different levels. That is, the unit operations, functional and/or phenomena level. The highest impact is expected by looking...... as generation and screening of phenomena based flowsheet options are presented using a decomposition based solution approach. The developed methodology as well as necessary tools and supporting methods are highlighted through a case study involving the production of isopropyl-acetate....

Development of a new miniature short-residence-time annular centrifugal solvent extraction contactor for tests of process flowsheets in hot cells

International Nuclear Information System (INIS)

Lanoe, J.Y.; Rivalier, P.

2000-01-01

Researches undertaken on new nuclear fuel reprocessing extraction processes need tests of process flowsheets in hot cells. To this goal, a new miniature short residence-time annular centrifugal solvent extraction contactor was conceived and developed at Marcoule. This single stage contactor is composed of an outer stationary cylinder (made of transparent plexiglas on prototype and of stainless steel on models for hot cells) and a suspended inner rotating cylinder of stainless steel; the inside diameter of the rotor is 12 mm. Aqueous and organic phases are fed into the gap between the two cylinders. The mixture flows down the annular space and then up through an orifice at the bottom of the rotor. Into the rotor, the emulsion breaks rapidly under the centrifugal force (up to 600 g with rotor speed of 10,000 rpm). The separated phases flow over their weirs and discharge at the top in their collector rings. The liquid hold-up of this centrifugal contactor is approximately 6 mL. The use in hots cells needed original designs for: - the assembly of a single-stage contactor: every part (motor, rotor, stationary housing) is simply inserted on the other one without screws and nuts; - the assembly of multistage group: every stage is stacking in two rails and an intermediate part (supported on the two rails) links exit ports and their corresponding inlet ports. All the parts are pressed and sealed against a terminal plate with a screw. Separating capacity tests with. a prototype were conducted using water as the aqueous phase and hydrogenated tetra-propylene (TPH) as the organic phase with aqueous to organic (A/O) flow ratio equal to 1. The best performances were obtained with rotor speed ranging from 4000 to 5000 rpm; the total throughput was then up to 2 L.h -1 . For a total throughput of 300 mL.h -1 , the hold-up in the annular mixing zone varied from 0.5 to 1.5 mL according to the A/O ratio and the starting mode. A number of tests were also performed to measure the

Late Wash/Nitric Acid flowsheet hydrogen generation bases for simulation of a deflagration/detonation in the DWPF CPC

International Nuclear Information System (INIS)

Ritter, J.A.

1993-01-01

Hydrogen generation data obtained from IDMS runs PX4 and PX5 will be used to determine a bases for a deflagration/detonation simulation in the DWPF CPC. This simulation is necessary due to the new chemistry associated with the Late Wash/ Nitric Acid flowsheet and process modifications associated with the presence of H 2 in the offgas. The simulation will be performed by Professor Van Brunt from the University of South Carolina. The scenario which leads up to the deflagration/detonation simulation will be chosen such that the following conditions apply. The SRAT is filled to its maximum operating level with 9,600 gal of sludge, which corresponds to the minimum vapor space above the sludge. The SRAT is at the boiling point, producing H 2 at a very low rate (about 10 % of the peak) and 15 scfm of air inleakage is entering the SRAT. Then, the H 2 generation rate will be allowed to increase exponentially (catalyst activation) until it readies the peak H 2 generation rate of the IDMS run, after which the H 2 generation rate will be allowed to decay exponentially (catalyst deactivation) until the total amount of H2 produced is between 85 and 100% of that produced during the IDMS run

Flowsheet for shear/leach processing of N Reactor fuel at PUREX

International Nuclear Information System (INIS)

Enghusen, M.B.

1995-01-01

This document was originally prepared to support the restart of the PUREX plant using a new Shear/Leach head end process. However, the PUREX facility was shutdown and processing of the remaining N Reactor fuel is no longer considered an alternative for fuel disposition. This document is being issued for reference only to document the activities which were investigated to incorporate the shear/leach process in the PUREX plant

Counter-current extraction studies for the recovery of neptunium by the Purex process. Part I

Energy Technology Data Exchange (ETDEWEB)

Srinivasan, N.; Nadkarni, M. N.; Kumar, S. V.; Kartha, P. K.S.; Sonavane, R. R.; Ramaniah, M. V.; Patil, S. K.

1974-07-01

It is proposed to recover neptunium-237, along with uranium and plutonium, during the fuel reprocessing in the PREFRE plant at Tarapur. Counter-current extraction studies, relevant to the code contamination (HA) and partitioning (IA) cycles of the purex process, were carried out to arrive at suitable chemical flowsheet conditions which would enable the co-extraction of neptunium along with uranium and plutonium. The results of the studies carried out using a laboratory mixer-settler unit and synthetic mixtures of neptunium and uranium are reported here. Based on these results, the chemical flowsheet conditions are proposed for the co-extraction of neptunium even if it exists as Np(V) in the aqueous feed solution. (auth)

The Michelin uranium project, Labrador, Canada metallurgical testwork, economic studies and process design

Energy Technology Data Exchange (ETDEWEB)

Goode, J.R., E-mail: jrgoode@sympatico.ca [Aurora Energy Resources Inc., Toronto, ON (Canada); Brown, J.A. [SGS Mineral Services, Lakefield, ON (Canada)

2010-07-01

Aurora Energy Resources Inc. is proposing to build and operate a 10,000 t/d process plant to produce 97 million pounds of U{sub 3}O{sub 8} over a seventeen-year project life from deposits in coastal Labrador. This paper summarizes the testwork, generally done by SGS Mineral Services in Lakefield, Ontario, and the economic studies that support flowsheet selection. The selected flowsheet includes SAG and ball milling, acid leaching using air/SO{sub 2} as an oxidant, and resin-in-pulp (RIP) extraction of uranium from the leached slurry. Other unit operations examined include ore sorting, heap leaching, liquid-solid separation, solvent extraction, and nanofiltration for eluate upgrading. We also review the extensive programs of environmental testwork and studies that were completed. (author)

Report on the Behavior of Fission Products in the Co-decontamination Process

International Nuclear Information System (INIS)

Martin, Leigh Robert; Riddle, Catherine Lynn

2015-01-01

This document was prepared to meet FCT level 3 milestone M3FT-15IN0302042, 'Generate Zr, Ru, Mo and Tc data for the Co-decontamination Process.' This work was carried out under the auspices of the Lab-Scale Testing of Reference Processes FCT work package. This document reports preliminary work in identifying the behavior of important fission products in a Co-decontamination flowsheet. Current results show that Tc, in the presence of Zr alone, does not behave as the Argonne Model for Universal Solvent Extraction (AMUSE) code would predict. The Tc distribution is reproducibly lower than predicted, with Zr distributions remaining close to the AMUSE code prediction. In addition, it appears there may be an intricate relationship between multiple fission product metals, in different combinations, that will have a direct impact on U, Tc and other important fission products such as Zr, Mo, and Rh. More extensive testing is required to adequately predict flowsheet behavior for these variances within the fission products.

Electronic equipment for spectrometric data processing

International Nuclear Information System (INIS)

Antonov, L.J.; Trenev, A.M.; Todorova, E.I.; Dimitrov, V.D.

1978-01-01

Electronic equipment carrying out logical operations and a full set of the arithmetic operations was developed for spectrometric data processing. The flowsheet of the computing part of the device, made on the basis of a specialized integral circuit, is given. The device includes input registers, multiplexor, matrix commutator, arithmetic unit and indication unit. The equipment is rated to carry out calculations according to comparatively complex formulae in several seconds

Improvements mineral dressing and extraction processes of gold-silver ores from San Pedro Frio Mining District, Colombia; Mejora de los procesos de beneficio y extraccion de minerales auroargentiferos del asentamiento minero de San Pedro Frio, Colombia

Energy Technology Data Exchange (ETDEWEB)

Yanez Traslavina, J. J.; Vargas Avila, M. A.; Garcia Paez, I. H.; Pedraza Rosas, J. E.

2005-07-01

The San Pedro Frio district mining, Colombia, is a rich region production gold-silver ores. Nowadays, the extraction processes used are amalgamation, percolation cyanidation and precipitation with zinc wood. Due to the ignorance of the ore characteristics, gold and silver treatment processes are inadequate and not efficient. In addition the inappropriate use of mercury and cyanide cause environmental contamination. In this research the ore characterization was carried out obtained fundamental parameters for the technical selection of more efficient gold and silver extraction processes. Experimental work was addressed to the study of both processes the agitation cyanidation and the adsorption on activated carbon in pulp. As a final result proposed a flowsheet to improve the precious metals recovery and reduce the environment contamination. (Author)

Chemical flowsheet conditions for preparing urania spheres by internal gelation

International Nuclear Information System (INIS)

Haas, P.A.; Begovich, J.M.; Ryon, A.D.; Vavruska, J.S.

1979-01-01

Small, ceramic urania spheres can be prepared for use as nuclear fuel by internal chemical gelation of uranyl nitrate solution droplets. Decomposition of hexamethylenetetramine (HMTA) dissolved in the uranyl nitrate solution releases ammonia to precipitate hydrated UO 3 . Previously established flowsheet conditions have been improved and modified at ORNL and have been applied to prepare dense UO 2 spheres with average diameters of 1200, 300, and 30 μm. Acid-deficient uranyl nitrate (ADUN) solutions up to 3.4 M in uranium with NO 3 - /U mole ratios of 1.5 to 1.7 are prepared by dissolution of U 3 O 8 or UO 3 . Continuous mixing of metered, cooled ADUN containing urea and HMTA solutions provides a smooth, regulated flow of the temperature-sensitive feed solution. The gelation times for solution drops in organic liquids at 45 to 95 0 C depend on both the chemical reaction rates and the rates of heat transfer. The gel properties vary with temperature and other gelation variables. Gelation conditions were determined which allow easy washing, drying, firing, and sintering to produce dense UO 2 spheres of all three sizes. The 1200- and 300-μm UO 2 spheres were pepared by gelation in trichloroethylene at 50 to 65 0 C; 2-ethyl-l-hexanol was used as the gelation medium to prepare 30-μm UO 2 spheres. Washing and drying requirements were determined. The gel dried to 225 0 C contains about 95% UO 3 ; the remaining components are H 2 O, NH 3 - , which are volatilized during firing to UO 2

Continuous precipitation process of plutonium salts

International Nuclear Information System (INIS)

Richard, P.

1967-03-01

This work concerns the continuous precipitation process of plutonium oxalate. Investigations about the solubility of different valence states in nitric-oxalic and in nitric-sulfuric-oxalic medium lead to select the precipitation process of tetravalent plutonium oxalate. Settling velocity and granulometry of tetravalent oxalate plutonium have been studied with variation of several precipitation parameters such as: temperature, acidity, excess of oxalic acid and aging time. Then are given test results of some laboratory continuous apparatus. Conditions of operation with adopted tubular apparatus are defined in conclusion. A flow-sheet is given for a process at industrial scale. (author) [fr

14 CFR 11.31 - How does FAA process direct final rules?

Science.gov (United States)

2010-01-01

... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false How does FAA process direct final rules? 11... PROCEDURAL RULES GENERAL RULEMAKING PROCEDURES Rulemaking Procedures General § 11.31 How does FAA process direct final rules? (a) A direct final rule will take effect on a specified date unless FAA receives an...

Sustainable Process Synthesis-Intensification

DEFF Research Database (Denmark)

Babi, Deenesh Kavi; Holtbruegge, Johannes; Lutze, Philip

2014-01-01

Sustainable process design can be achieved by performing process synthesis and process intensification together. This approach first defines a design target through a sustainability analysis and then finds design alternatives that match the target through process intensification. A systematic......, multi-stage framework for process synthesis- intensification that identifies more sustainable process designs has been developed. At stages 1-2, the working scale is at the level of unit operations, where a base case design is identified and analyzed with respect to sustainability metrics. At stages 3......, a phenomena-based process synthesis method is applied, where the phenomena involved in each tasks are identified, manipulated and recombined to generate new and/or existing unit operations configured into flowsheets that are more sustainable from those found in the previous levels. An overview of the key...

Consolidation of the EXAm process: towards the reprocessing of a concentrated PUREX raffinate

Energy Technology Data Exchange (ETDEWEB)

Vanel, V.; Bollesteros, M.J.; Marie, C.; Montuir, M.; Pacary, V.; Antegnard, F.; Costenoble, S.; Boyer-Deslys, V. [CEA Marcoule, Nuclear Energy Division, Radiochemistry and Processes Department, Bagnols-sur-Ceze, F-30207 (France)

2016-07-01

Recycling americium alone from the spent fuel is an important issue currently studied for the future nuclear cycle (Generation IV systems) as Am is one of the main contributors to the long-term radiotoxicity and heat power of final waste. The solvent extraction process called EXAm has been developed by the CEA to enable the recovery of Am alone from a PUREX raffinate (with U, Np and Pu already removed). A mixture of DMDOHEMA and HDEHP diluted in TPH is used as the solvent and the Am/Cm selectivity is improved using TEDGA as a selective complexing agent to maintain Cm and the heavier lanthanides in the acidic aqueous phase (HNO{sub 3} 5-6 M). Americium is then selectively stripped from the light lanthanides at low acidity (pH 2.5-3) with a poly-aminocarboxylic acid (DTPA). An additional step is necessary before Am recovery, in order to strip molybdenum which would otherwise be complexed by DTPA and contaminate the Am raffinate. In order to make the process and its associated future plant more compact, the objective is now to adapt the EXAm process to a concentrated raffinate. With a concentrated PUREX raffinate, the process operates under conditions close to saturation both for the solvent and complexing agent TEDGA during the Am extraction step. Consequently, some changes were needed to adapt the flowsheet to higher concentrations of cations. Before the test on a real PUREX raffinate in the CBP shielded line at ATALANTE (at the end of 2015), the EXAm flowsheet had to be consolidated and achievable target performances ensured. A series of experiments and tests was performed: on laboratory scale (batch experiments), to identify the good operating conditions and to simulate the main phenomena involved (2010-2014); first on an inactive surrogate feed solution at G1 facility (2011-2013), and then on a surrogate feed solution with trace amounts of americium and curium (spiked test) in the C17 shielded line at ATALANTE (2014). (authors)

Simulation, integration, and economic analysis of gas-to-liquid processes

International Nuclear Information System (INIS)

Bao, Buping; El-Halwagi, Mahmoud M.; Elbashir, Nimir O.

2010-01-01

Gas-to-liquid (GTL) involves the chemical conversion of natural gas into synthetic crude that can be upgraded and separated into different useful hydrocarbon fractions including liquid transportation fuels. Such technology can also be used to convert other abundant natural resources such as coal and biomass to fuels and value added chemicals (referred to as coal-to-liquid (CTL) and biomass-to-liquid (BTL)). A leading GTL technology is the Fischer-Tropsch (FT) process. The objective of this work is to provide a techno-economic analysis of the GTL process and to identify optimization and integration opportunities for cost saving and reduction of energy usage while accounting for the environmental impact. First, a base-case flowsheet is synthesized to include the key processing steps of the plant. Then, a computer-aided process simulation is carried out to determine the key mass and energy flows, performance criteria, and equipment specifications. Next, energy and mass integration studies are performed to address the following items: (a) heating and cooling utilities, (b) combined heat and power (process cogeneration), (c) management of process water, (c) optimization of tail gas allocation, and (d) recovery of catalyst-supporting hydrocarbon solvents. Finally, these integration studies are conducted and the results are documented in terms of conserving energy and mass resources as well as providing economic impact. Finally, an economic analysis is undertaken to determine the plant capacity needed to achieve the break-even point and to estimate the return on investment for the base-case study. (author)

Indicators for Building Process without Final Defects -

DEFF Research Database (Denmark)

Jørgensen, Kirsten; Rasmussen, Grane Mikael Gregaard; Thuesen, Christian Langhoff

2011-01-01

This article introduces the preliminary data analysis, as well as the underlying theories and methods for identifying the indicators for building process without final defects. Since 2004, the Benchmark Centre for the Danish Construction Sector (BEC) has collected information about legal defects...

Process synthesis for natural products from plants based on PAT methodology

DEFF Research Database (Denmark)

Malwade, Chandrakant Ramkrishna; Qu, Haiyan; Rong, Ben-Guang

2017-01-01

(QbD) approach, has been included at various steps to obtain molecular level information of process streams and thereby, support the rational decision making. The formulated methodology has been used to isolate and purify artemisinin, an antimalarial drug, from dried leaves of the plant Artemisia...... generates different process flowsheet alternatives consisting of multiple separation techniques. Decision making is supported by heuristics as well as basic process information already available from previous studies. In addition, process analytical technology (PAT) framework, a part of Quality by Design...

New Waste Calciner High Temperature Operation

International Nuclear Information System (INIS)

Swenson, M.C.

2000-01-01

A new Calciner flowsheet has been developed to process the sodium-bearing waste (SBW) in the INTEC Tank Farm. The new flowsheet increases the normal Calciner operating temperature from 500 C to 600 C. At the elevated temperature, sodium in the waste forms stable aluminates, instead of nitrates that melt at calcining temperatures. From March through May 2000, the new high-temperature flowsheet was tested in the New Waste Calcining Facility (NWCF) Calciner. Specific test criteria for various Calciner systems (feed, fuel, quench, off-gas, etc.) were established to evaluate the long-term operability of the high-temperature flowsheet. This report compares in detail the Calciner process data with the test criteria. The Calciner systems met or exceeded all test criteria. The new flowsheet is a visible, long-term method of calcining SBW. Implementation of the flowsheet will significantly increase the calcining rate of SBW and reduce the amount of calcine produced by reducing the amount of chemical additives to the Calciner. This will help meet the future waste processing milestones and regulatory needs such as emptying the Tank Farm

Purex: process and equipment performance

International Nuclear Information System (INIS)

Orth, D.A.

1986-01-01

The Purex process is the solvent extraction system that uses tributyl phosphate as the extractant for separating uranium and plutonium from irradiated reactor fuels. Since the first flowsheet was proposed at Oak Ridge National Laboratory in 1950, the process has endured for over 30 years with only minor modifications. The spread of the technology was rapid, and worldwide use or research on Purex-type processes was reported by the time of the 1955 Geneva Conference. The overall performance of the process has been so good that there are no serious contenders for replacing it soon. This paper presents: process description; equipment performance (mixer-settlers, pulse columns, rapid contactors); fission product decontamination; solvent effects (solvent degradation products); and partitioning of uranium and plutonium

Modelling of innovative SANEX process mal-operations

International Nuclear Information System (INIS)

McLachlan, F.; Taylor, R.; Whittaker, D.; Woodhead, D.; Geist, A.

2016-01-01

The innovative (i-) SANEX process for the separation of minor actinides from PUREX highly active raffinate is expected to employ a solvent phase comprising 0.2 M TODGA with 5 v/v% 1-octanol in an inert diluent. An initial extract / scrub section would be used to extract trivalent actinides and lanthanides from the feed whilst leaving other fission products in the aqueous phase, before the loaded solvent is contacted with a low acidity aqueous phase containing a sulphonated bis-triazinyl pyridine ligand (BTP) to effect a selective strip of the actinides, so yielding separate actinide (An) and lanthanide (Ln) product streams. This process has been demonstrated in lab scale trials at Juelich (FZJ). The SACSESS (Safety of Actinide Separation processes) project is focused on the evaluation and improvement of the safety of such future systems. A key element of this is the development of an understanding of the response of a process to upsets (mal-operations). It is only practical to study a small subset of possible mal-operations experimentally and consideration of the majority of mal-operations entails the use of a validated dynamic model of the process. Distribution algorithms for HNO_3, Am, Cm and the lanthanides have been developed and incorporated into a dynamic flowsheet model that has, so far, been configured to correspond to the extract-scrub section of the i-SANEX flowsheet trial undertaken at FZJ in 2013. Comparison is made between the steady state model results and experimental results. Results from modelling of low acidity and high temperature mal-operations are presented. (authors)

Interpolation of final geometry and result fields in process parameter space

NARCIS (Netherlands)

Misiun, Grzegorz Stefan; Wang, Chao; Geijselaers, Hubertus J.M.; van den Boogaard, Antonius H.; Saanouni, K.

2016-01-01

Different routes to produce a product in a bulk forming process can be described by a limited set of process parameters. The parameters determine the final geometry as well as the distribution of state variables in the final shape. Ring rolling has been simulated using different parameter settings.

Uranium decontamination in Purex second plutonium cycle: An example of solvent extraction modeling

International Nuclear Information System (INIS)

Hsu, T.C.

1986-01-01

The existing Purex flowsheet used in the second plutonium cycle at the Savannah River Plant (SRP) does not remove uranium from the plutonium stream. To develop new flowsheets for the Purex second plutonium cycle, computer simulation using SEPHIS was used. SEPHIS is an ORNL-developed solvent extraction simulation code. Box-Wilson experimental design was used to select the minimum set of process conditions simulated. The calculated results were plotted into three-dimensional response surfaces by SAS/Graph (statistical analysis systems). These surfaces provide a broad and complete overview of the responses. Specific ranges of key variables were then investigated. The second series of process simulations identified flowsheets that provide high uranium decontamination while meeting all other key process requirements. The proposed flowsheet consists of modifying the existing 2B bank flowsheet by relocating the feed, increasing the extractant acidity, and adding a scrub stream. The nuclear safety issue was also examined

The process system analysis for advanced spent fuel management technology (I)

International Nuclear Information System (INIS)

Lee, H. H.; Lee, J. R.; Kang, D. S.; Seo, C. S.; Shin, Y. J.; Park, S. W.

1997-12-01

Various pyrochemical processes were evaluated, and viable options were selected in consideration of the proliferation safety, technological feasibility and compatibility to the domestic nuclear power system. Detailed technical analysis were followed on the selected options such as unit process flowsheet including physico-chemical characteristics of the process systems, preliminary concept development, process design criteria and materials for equipment. Supplementary analysis were also carried out on the support technologies including sampling and transport technologies of molten salt, design criteria and equipment for glove box systems, and remote operation technologies. (author). 40 refs., 49 tabs., 37 figs

Evaluation of a modified Zirflex process to minimize high-level waste generation at the Idaho Chemical Processing Plant

International Nuclear Information System (INIS)

Chipman, N.A.; Carleson, T.E.

1987-06-01

Extensive laboratory experimentation was conducted to develop a dissolvent suitable for Zircaloy based fuels having a surface oxide coating. Other laboratory experimentation was conducted on the precipitation and solids separation steps of the process. Computer simulation was used to determine the stability and uranium extractability of the output stream, and potential waste volume reduction. From these studies a conceptual flowsheet was developed which could potentially reduce HLW volumes by about 30%. Other process alternatives being investigated achieve equal HLW volume reduction and potentially improve safety of operation. Therefore, the Modified Zirflex process is not presently being considered for further development. 22 refs., 21 figs., 3 tabs

Wet Chemical Oxidation of Organic Waste Using Nitric-Phosphoric Acid Technology

Energy Technology Data Exchange (ETDEWEB)

Pierce, R.A.

1998-10-06

Experimental progress has been made in a wide range of areas which support the continued development of the nitric-phosphoric acid oxidation process for combustible, solid organic wastes. An improved understanding of the overall process operation has been obtained, acid recovery and recycle systems have been studied, safety issues have been addressed, two potential final waste forms have been tested, preliminary mass flow diagrams have been prepared, and process flowsheets have been developed. The flowsheet developed is essentially a closed-loop system which addresses all of the internally generated waste streams. The combined activities aim to provide the basis for building and testing a 250-400 liter pilot-scale unit. Variations of the process now must be evaluated in order to address the needs of the primary customer, SRS Solid Waste Management. The customer is interested in treating job control waste contaminated with Pu-238 for shipment to WIPP. As a result, variations for feed preparation, acid recycle, and final form manufacturing must be considered to provide for simpler processing to accommodate operations in high radiation and contamination environments. The purpose of this program is to demonstrate a nitric-phosphoric acid destruction technology which can treat a heterogeneous waste by oxidizing the solid and liquid organic compounds while decontaminating noncombustible items.

Computer aided process control equipment at the Karlsruhe reprocessing pilot plant, WAK

International Nuclear Information System (INIS)

Winter, R.; Finsterwalder, L.; Gutzeit, G.; Reif, J.; Stollenwerk, A.H.; Weinbrecht, E.; Weishaupt, M.

1991-01-01

A computer aided process control system has been installed at the Karlsruhe Spent Fuel Reprocessing Plant, WAK. All necessary process control data of the first extraction cycle is collected via a data collection system and is displayed in suitable ways on a screen for the operator in charge of the unit. To aid verification of displayed data, various measurements are associated to each other using balance type process modeling. Thus, deviation of flowsheet conditions and malfunctioning of measuring equipment are easily detected. (orig.) [de

Wastewater Triad Project: Final Summary Report

Energy Technology Data Exchange (ETDEWEB)

Walker, J.F.

2001-12-27

The U.S. Department of Energy (DOE) facilities have performed nuclear energy research and radiochemical production since the early 1940s. Currently, millions of gallons of legacy radioactive liquid and sludge wastes are contained in over 300 large underground storage tanks, located primarily at Hanford, the Savannah River Site (SRS), Idaho National Engineering and Environmental Laboratory (INEEL), and Oak Ridge National Laboratory (ORNL). Plans for tank waste retrieval, treatment, and immobilization are being developed and implemented throughout the DOE complex In order to meet regulatory requirements for remediation of underground storage tanks, ORNL has developed an integrated approach to the management of its waste that has applications across the DOE complex. The integrated approach consolidates plans for remediation of inactive tanks; upgrade of the active waste collection, storage, and treatment systems; and treatment of transuranic (TRU) tank waste for disposal. Important elements of this integrated approach to tank waste management include waste retrieval of sludges from tanks, conditioning and transport of retrieved waste to active storage tanks or treatment facilities, solid/liquid separations for supernatant recycle and/or waste treatment, removal of cesium from the supernatant, volume reduction of the supernatant, and solidification of sludges and supernatant for disposal. Each unit operation of the flowsheet is interconnected and impacts the overall efficiency of the entire flowsheet. ORNL has implemented innovative but proven technologies for each of the major unit operations to accelerate clean-up. ORNL used the integrated plan to determine where developing technologies were required to create an optimized flowsheet to (1) accelerate clean-out and remediation of underground storage tanks; (2) provide significant cost avoidance and schedule reductions; (3) consolidate wastes for private-sector immobilization; (4) facilitate regulatory compliance with

Nitrogen Trifluoride-Based Fluoride- Volatility Separations Process: Initial Studies

Energy Technology Data Exchange (ETDEWEB)

McNamara, Bruce K.; Scheele, Randall D.; Casella, Andrew M.; Kozelisky, Anne E.

2011-09-28

This document describes the results of our investigations on the potential use of nitrogen trifluoride as the fluorinating and oxidizing agent in fluoride volatility-based used nuclear fuel reprocessing. The conceptual process uses differences in reaction temperatures between nitrogen trifluoride and fuel constituents that produce volatile fluorides to achieve separations and recover valuable constituents. We provide results from our thermodynamic evaluations, thermo-analytical experiments, kinetic models, and provide a preliminary process flowsheet. The evaluations found that nitrogen trifluoride can effectively produce volatile fluorides at different temperatures dependent on the fuel constituent.

Techno-economic analysis of organosolv pretreatment process from lignocellulosic biomass

DEFF Research Database (Denmark)

Rodrigues Gurgel da Silva, Andrè; Errico, Massimiliano; Rong, Ben-Guang

2018-01-01

data, we propose a feasible process flowsheet for organosolv pretreatment. Simulation of the pretreatment process provided mass and energy balances for a techno-economic analysis, and the values were compared with the most prevalent and mature pretreatment method: diluted acid. Organosolv pretreatment...... required more energy, 578.1 versus 213.8 MW for diluted acid pretreatment, but resulted in a higher ethanol concentration after the biomass fermentation, 11.1% compared to 5.4%. Total annual costs (TACs) calculations showed advantages for diluted acid pretreatment, but future improvements explored...

Minimum energy consumption process synthesis for energy saving

Energy Technology Data Exchange (ETDEWEB)

Xiao-Ping, Jia [Institute for Petroleum and Chemical Industry, Qingdao University of Science and Technology, Qingdao 266042, Shandong (China); Department of Environmental Science and Engineering, Tsinghua University, Beijing 100084 (China); Fang, Wang; Shu-Guang, Xiang; Xin-Sun, Tan; Fang-Yu, Han [Institute for Petroleum and Chemical Industry, Qingdao University of Science and Technology, Qingdao 266042, Shandong (China)

2008-05-15

The paper presents a synthesis strategy for the chemical processes with energy saving. The concept of minimum energy consumption process (MECP) is proposed. Three characteristics of MECP are introduced, including thermodynamic minimum energy demand, energy consumption efficiency and integration degree. These characteristics are evaluated according to quantitative thermodynamic analysis and qualitative knowledge rules. The procedure of synthesis strategy is proposed to support the generation of MECP alternatives, which combine flowsheet integration and heat integration. The cases studies will focus on how integration degrees of a process affect the energy-saving results. The separation sequences of the hydrodealkylation of toluene (HDA) process and ethanol distillation process as case studies are used to illustrate. (author)

IS process for thermochemical hydrogen production

International Nuclear Information System (INIS)

Onuki, Kaoru; Nakajima, Hayato; Ioka, Ikuo; Futakawa, Masatoshi; Shimizu, Saburo

1994-11-01

The state-of-the-art of thermochemical hydrogen production by IS process is reviewed including experimental data obtained at JAERI on the chemistry of the Bunsen reaction step and on the corrosion resistance of the structural materials. The present status of laboratory scale demonstration at JAERI is also included. The study on the chemistry of the chemical reactions and the products separations has identified feasible methods to function the process. The flowsheeting studies revealed a process thermal efficiency higher than 40% is achievable under efficient process conditions. The corrosion resistance of commercially available structural materials have been clarified under various process conditions. The basic scheme of the process has been realized in a laboratory scale apparatus. R and D requirements to proceed to the engineering demonstration coupled with HTTR are briefly discussed. (author)

Operability and flexibility of a milk production line

DEFF Research Database (Denmark)

Cheng, Hongyuan; Friis, Alan

2007-01-01

The operability and flexibility of an existing milk treatment process are investigated through flowsheet modelling and simulation. From the flowsheet simulation, a process operating region was determined using incoming milk flow viscosity and heat exchanger pressure drop as characteristic...

The reprocessing of irradiated fuels improvement and extension of the solvent extraction process

International Nuclear Information System (INIS)

Faugeras, P.; Chesne, A.

1964-01-01

Improvements made in the conventional tri-butylphosphate process are described, in particular. the concentration and the purification of plutonium by one extraction cycle using tri-butyl-phosphate with reflux; and the use of an apparatus working continuously for precipitating plutonium oxalate, for calcining the oxalate, and for fluorinating the oxide. The modifications proposed for the treatment of irradiated uranium - molybdenum alloys are described, in particular, the dissolution of the fuel, and the concentration of the fission product solutions. The solvent extraction treatment is used also for the plutonium fuels utilized for the fast breeder reactor (Rapsodie) An outline of the process is presented and discussed, as well as the first experimental results and the plans for a pilot plant having a capacity of 1 kg/day. The possible use of tn-lauryl-amine in the plutonium purification cycle is now under consideration for the processing plant at La Hague. The flowsheet for this process and its performance are presented. The possibility of vitrification is considered for the final treatment of the concentrated radioactive wastes from the Marcoule (irradiated uranium) and La Hague (irradiated uranium-molybdenum) Centers. Three possible processes are described and discussed, as well as the results obtained from the operation of the corresponding experimental units using tracers. (authors) [fr

Cement-based processes for the immobilization of intermediate level radioactive waste

International Nuclear Information System (INIS)

Brown, D.J.; Lee, D.J.; Price, M.S.T.; Smith, D.L.G.

1985-01-01

Increasing attention is being paid to the use of cement-based materials for the immobilisation of intermediate level wastes. Various cementitious materials are surveyed and the use of blast furnace slag is shown to be advantageous. The properties of cemented wastes are surveyed both during processing and as solid products. The application of Winfrith Cementation Laboratory technology to plant and flowsheet development for Winfrith Reactor sludge immobilisation is described. (author)

Phenomena-based Process Synthesis and Design to achieve Process Intensification

DEFF Research Database (Denmark)

Lutze, Philip; Babi, Deenesh Kavi; Woodley, John

2012-01-01

at the lowest level of aggregation: phenomena. Therefore, in this paper, a phenomena-based synthesis/design methodology is presented. Using this methodology, a systematic identification of necessary and desirable (integrated) phenomena as well as generation and screening of phenomena-based flowsheet options...

Final Rule for Industrial Process Cooling Towers: Fact Sheet

Science.gov (United States)

Fact sheet concerning a final rule to reduce air toxics emissions from industrial process cooling towers. Air toxics are those pollutants known or suspected of causing cancer or other serious health effects.

ATAC Process Proof of Concept Final Report

Energy Technology Data Exchange (ETDEWEB)

Bri Rolston; Sarah Freeman

2014-03-01

Researchers at INL with funding from the Department of Energy’s Office of Electricity Delivery and Energy Reliability (DOE-OE) evaluated a novel approach for near real-time consumption of threat intelligence. Demonstration testing in an industry environment supported the development of this new process to assist the electric sector in securing their critical networks. This report provides the reader with an understanding of the methods used during this proof of concept project. The processes and templates were further advanced with an industry partner during an onsite assessment. This report concludes with lessons learned and a roadmap for final development of these materials for use by industry.

Systematic sustainable process design and analysis of biodiesel processes

DEFF Research Database (Denmark)

Mansouri, Seyed Soheil; Ismail, Muhammad Imran; Babi, Deenesh Kavi

2013-01-01

Biodiesel is a promising fuel alternative compared to traditional diesel obtained from conventional sources such as fossil fuel. Many flowsheet alternatives exist for the production of biodiesel and therefore it is necessary to evaluate these alternatives using defined criteria and also from...... a biodiesel production case study....

Fuel processing

International Nuclear Information System (INIS)

Allardice, R.H.

1990-01-01

The technical and economic viability of the fast breeder reactor as an electricity generating system depends not only upon the reactor performance but also on a capability to recycle plutonium efficiently, reliably and economically through the reactor and fuel cycle facilities. Thus the fuel cycle is an integral and essential part of the system. Fuel cycle research and development has focused on demonstrating that the challenging technical requirements of processing plutonium fuel could be met and that the sometimes conflicting requirements of the fuel developer, fuel fabricator and fuel reprocessor could be reconciled. Pilot plant operation and development and design studies have established both the technical and economic feasibility of the fuel cycle but scope for further improvement exists through process intensification and flowsheet optimization. These objectives and the increasing processing demands made by the continuing improvement to fuel design and irradiation performance provide an incentive for continuing fuel cycle development work. (author)

Technology of Polish copper ore beneficiation – perspectives from the past experience

Directory of Open Access Journals (Sweden)

Wieniewski Andrzej

2016-01-01

Full Text Available The paper describes the main types of existing copper deposits in the world and the most common enrichment technologies. The characteristic elements of the current flowsheet of the Polish ore beneficiation were discussed together with the reasons for using them. The paper presents a perspective flowsheet based on more than 50-years of experience in beneficiation of Polish copper ore. The main elements of proposed technology are: leaving in justified cases the division of ore into fractions of increased sandstone and carbonate ores content, flotation in the milling circuit as part of the effective preparation of the ore to rougher flotation, intensive rougher flotation, classic cleaning system, processing of the middlings with grinding system in new type mills, classification and flotation with outlet of final tailings.

Purex process modelling - do we really need speciation data?

International Nuclear Information System (INIS)

Taylor, R.J.; May, I.

2001-01-01

The design of reprocessing flowsheets has become a complex process requiring sophisticated simulation models, containing both chemical and engineering features. Probably the most basic chemical data needed is the distribution of process species between solvent and aqueous phases at equilibrium, which is described by mathematical algorithms. These algorithms have been constructed from experimentally determined distribution coefficients over a wide range of conditions. Distribution algorithms can either be empirical fits of the data or semi-empirical equations, which describe extraction as functions of process variables such as temperature, activity coefficients, uranium loading, etc. Speciation data is not strictly needed in the accumulation of distribution coefficients, which are simple ratios of analyte concentration in the solvent phase to that in the aqueous phase. However, as we construct process models of increasing complexity, speciation data becomes much more important both to raise confidence in the model and to understand the process chemistry at a more fundamental level. UV/vis/NIR spectrophotometry has been our most commonly used speciation method since it is a well-established method for the analysis of actinide ion oxidation states in solution at typical process concentrations. However, with the increasing availability to actinide science of more sophisticated techniques (e.g. NMR; EXAFS) complementary structural information can often be obtained. This paper will, through examples, show how we have used spectrophotometry as a primary tool in distribution and kinetic experiments to obtain data for process models, which are then validated through counter-current flowsheet trials. It will also discuss how spectrophotometry and other speciation methods are allowing us to study the link between molecular structure and extraction behaviour, showing how speciation data really is important in PUREX process modelling. (authors)

Safe, secure, and clean disposal of final nuclear wastes using 'PyroGreen' strategies

International Nuclear Information System (INIS)

Jung, HyoSook; Choi, Sungyeol; Hwang, Il Soon

2011-01-01

Spent nuclear fuels (SNFs) present global challenges that must be overcome to pave way for safe, secure, peaceful and clean nuclear energy. As one of innovative solutions, we have proposed an innovative partitioning, transmutation, and disposal approach named as 'PyroGreen' that is designed to eliminate the need for high-level waste repositories. A flowsheet of pyrochemical partitioning process with technically achievable values of decontamination factors on long-living radionuclides has been established to enable all the final wastes to be disposed of as low and intermediate level wastes. The long-term performance of a geological repository was assessed by SAFE-ROCK code for the final wastes from the PyroGreen processing of entire 26,000 MTHM of SNFs arising from lifetime operation of 24 pressurized water reactors. The assessment results agree well with an earlier study in the fact that most harmful radionuclides dominating groundwater migration risk are shown to be long-living fission products including C-14, Cl-36, Se-79, I-129, and Cs-135, whereas most actinides including U, Pu, Np, Am, and Cm are shown to remain near the repository. It is shown that the final wastes can meet the radiological dose limit of current Korean regulation on the low and intermediate level waste repository. Long-living actinide concentration in wastes is comparable with those in wastes in Waste Isolation Pilot Plant that has proved adequately low risk of human intrusion. Overall decontamination factors required for PyroGreen are finally determined as 20,000 for uranium and all transuranic elements whereas much lower values in the range of 10-50 are required for important fission products including Se, Tc, I, Sr, and Cs in order to eliminate the need for any high-level waste repository. It has been shown that experimentally demonstrated recovery rate data for key process steps positively support the feasibility of PyroGreen. SAFE-ROCK code was used to evaluate the long-term performance

Design of sustainable chemical processes: Systematic retrofit analysis, generation and evaluation alternatives

DEFF Research Database (Denmark)

Carvalho, Ana; Gani, Rafiqul; Matos, Henrique

2008-01-01

eliminating the need to identify trade-off-based solutions. These indicators are also able to reduce (where feasible) a set of safety indicators. An indicator sensitivity analysis algorithm has been added to the methodology to define design targets and to generate sustainable process alternatives. A computer-aided...... tool has been developed to facilitate the calculations needed for the application of the methodology. The application of the indicator-based methodology and the developed software are highlighted through a process flowsheet for the production of vinyl chlorine monomer (VCM)....

SME Acceptability Determination For DWPF Process Control (U)

Energy Technology Data Exchange (ETDEWEB)

Edwards, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2017-06-12

The statistical system described in this document is called the Product Composition Control System (PCCS). K. G. Brown and R. L. Postles were the originators and developers of this system as well as the authors of the first three versions of this technical basis document for PCCS. PCCS has guided acceptability decisions for the processing at the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) since the start of radioactive operations in 1996. The author of this revision to the document gratefully acknowledges the firm technical foundation that Brown and Postles established to support the ongoing successful operation at the DWPF. Their integration of the glass propertycomposition models, developed under the direction of C. M. Jantzen, into a coherent and robust control system, has served the DWPF well over the last 20+ years, even as new challenges, such as the introduction into the DWPF flowsheet of auxiliary streams from the Actinide Removal Process (ARP) and other processes, were met. The purpose of this revision is to provide a technical basis for modifications to PCCS required to support the introduction of waste streams from the Salt Waste Processing Facility (SWPF) into the DWPF flowsheet. An expanded glass composition region is anticipated by the introduction of waste streams from SWPF, and property-composition studies of that glass region have been conducted. Jantzen, once again, directed the development of glass property-composition models applicable for this expanded composition region. The author gratefully acknowledges the technical contributions of C.M. Jantzen leading to the development of these glass property-composition models. The integration of these models into the PCCS constraints necessary to administer future acceptability decisions for the processing at DWPF is provided by this sixth revision of this document.

Process Options Description for Vitrification Flowsheet Model of INEEL Sodium Bearing Waste

Energy Technology Data Exchange (ETDEWEB)

Nichols, Todd Travis; Taylor, Dean Dalton; Lauerhass, Lance; Barnes, Charles Marshall

2001-02-01

The purpose of this document is to provide the technical information to Savannah River Site (SRS) personnel that is required for the development of a basic steady-state process simulation of the vitrification treatment train of sodium bearing waste (SBW) at Idaho National Engineering and nvironmental Laboratory (INEEL). INEEL considers simulation to have an important role in the integration/optimization of treatment process trains for the High Level Waste (HLW) Program. This project involves a joint Technical Task Plan (TTP ID77WT31, Subtask C) between SRS and INEEL. The work scope of simulation is different at the two sites. This document addresses only the treatment of SBW at INEEL. The simulation model(s) is to be built by SRS for INEEL in FY-2001.

Ferrocyanide safety program: Final report on adiabatic calorimetry and tube propagation tests with synthetic ferrocyanide materials

International Nuclear Information System (INIS)

Fauske, H.F.; Meacham, J.E.; Cash, R.J.

1995-01-01

Based on Fauske and Associates, Inc. Reactive System Screening Tool tests, the onset or initiation temperature for a ferrocyanide-nitrate propagating reaction is about 250 degrees Celcius. This is at about 200 degrees Celcius higher than current waste temperatures in the highest temperature ferrocyanide tanks. Furthermore, for current ambient waste temperatures, the tube propagation tests show that a ferrocyanide concentration of 15.5 wt% or more is required to sustain a propagation reaction in the complete absence of free water. Ignoring the presence of free water, this finding rules out propagating reactions for all the Hanford flowsheet materials with the exception of the ferrocyanide waste produced by the original In Farm flowsheet

FINAL REPORT: Transformational electrode drying process

Energy Technology Data Exchange (ETDEWEB)

Claus Daniel, C.; Wixom, M.(A123 Systems, Inc.)

2013-12-19

This report includes major findings and outlook from the transformational electrode drying project performance period from January 6, 2012 to August 1, 2012. Electrode drying before cell assembly is an operational bottleneck in battery manufacturing due to long drying times and batch processing. Water taken up during shipment and other manufacturing steps needs to be removed before final battery assembly. Conventional vacuum ovens are limited in drying speed due to a temperature threshold needed to avoid damaging polymer components in the composite electrode. Roll to roll operation and alternative treatments can increase the water desorption and removal rate without overheating and damaging other components in the composite electrode, thus considerably reducing drying time and energy use. The objective of this project was the development of an electrode drying procedure, and the demonstration of processes with no decrease in battery performance. The benchmark for all drying data was an 80°C vacuum furnace treatment with a residence time of 18 – 22 hours. This report demonstrates an alternative roll to roll drying process with a 500-fold improvement in drying time down to 2 minutes and consumption of only 30% of the energy compared to vacuum furnace treatment.

Uranium extraction from ores with lemon juice I,b-uranium recovery from pregnant lemon juice liquors obtained by attacking phosphate ores and suggested flowsheet

International Nuclear Information System (INIS)

EL-Sayed, M.H.

1992-01-01

In order to recover uranium from the pregnant liquors obtained by attacking safaga phosphate and qatrani phosphatic sandstone ore materials with lemon juice, methylation for acidic fraction-salt separation has been carried out. Afterwards, separation of uranium from the associated calcium (mainly present in lemon juice liquors as citrate) has been performed by making-use of the wide difference in their water solubility. The solutions containing the separated uranium were then subjected to evaporation till dryness whereby the precipitated uranyl citrate was calcined at 500 degree C to obtain the yellow orange oxide powder (U o 3 ). On the basis of one ton ore treatment, a flowsheet for uranium recovery from the two ore materials has been suggested

Uranium extraction from ores with lemon juice; II,b. uranium recovery from pregnant lemon juice liquors obtained by attacking phosphate ore and suggested flowsheet

International Nuclear Information System (INIS)

Hussein, E.M.

1997-01-01

In order to recover uranium from the pregnant liquors obtained by attacking Safaga phosphate and Qatrani phosphatic sandstone ore materials with lemon juice, methylation for acidic fraction-salt separation has been carried out. Afterwards, separation of uranium from the associated calcium (mainly present in lemon juice liquors as citrate) has been performed by making-use of the wide difference in their water solubility. The solutions containing the separated uranium were then subjected to evaporation till dryness whereby the precipitated uranyl citrate was calcined at 500 degree C to obtain the yellow orange oxide powder (UO 3 ). On the basis of one ton ore treatment, a flowsheet for uranium recovery from the two ore materials has been suggested

Space Processing Applications rocket project SPAR III. Final report

International Nuclear Information System (INIS)

Reeves, F.

1978-01-01

This document presents the engineering report and science payload III test report and summarizes the experiment objectives, design/operational concepts, and final results of each of five scientific experiments conducted during the third Space Processing Applications Rocket (SPAR) flight flown by NASA in December 1976. The five individual SPAR experiments, covering a wide and varied range of scientific materials processing objectives, were entitled: Liquid Mixing, Interaction of Bubbles with Solidification Interfaces, Epitaxial Growth of Single Crystal Film, Containerless Processing of Beryllium, and Contact and Coalescence of Viscous Bodies

Thermochemical water-splitting cycle, bench-scale investigations and process engineering. Annual report, October 1, 1978-September 30, 1979

Energy Technology Data Exchange (ETDEWEB)

Caprioglio, G.; McCorkle, K.H.; Besenbruch, G.E.; Rode, J.S.

1980-03-01

A program to investigate thermochemical water splitting has been under way at General Atomic Company (GA) since October 1972. This document is an annual progress report of Department of Energy (DOE) sponsored process development work on the GA sulfur-iodine thermochemical water splitting cycle. The work consisted of laboratory bench-scale investigations, demonstration of the process in a closed-loop cycle demonstrator, and process engineering design studies. A bench-scale system, consisting of three subunits, has been designed to study the cycle under continuous flow conditions. The designs of subunit I, which models the main solution reaction and product separation, and subunit II, which models the concentration and decomposition of sulfuric acid, were presented in an earlier annual report. The design of subunit III, which models the purification and decomposition of hydrogen iodide, is given in this report. Progress on the installation and operation of subunits I and II is described. A closed-loop cycle demonstrator was installed and operated based on a DOE request. Operation of the GA sulfur-iodine cycle was demonstrated in this system under recycle conditions. The process engineering addresses the flowsheet design of a large-scale production process consisting of four chemical sections (I through IV) and one helium heat supply section (V). The completed designs for sections I through V are presented. The thermal efficiency of the process calculated from the present flowsheet is 47%.

A review of the demonstration of innovative solvent extraction processes for the recovery of trivalent minor actinides from PUREX raffinate

International Nuclear Information System (INIS)

Modolo, G.; Wilden, A.; Geist, A.; Magnusson, D.; Malmbeck, R.

2012-01-01

The selective partitioning (P) of long-lived minor actinides from highly active waste solutions and their transmutation (T) to short-lived or stable isotopes by nuclear reactions will reduce the long-term hazard of the high-level waste and significantly shorten the time needed to ensure their safe confinement in a repository. The present paper summarizes the on-going research activities at Forschungszentrum Juelich (FZJ), Karlsruher Institut fuer Technologie (KIT) and Institute for Transuranium Elements (ITU) in the field of actinide partitioning using innovative solvent extraction processes. European research over the last few decades, i.e. in the NEWPART, PARTNEW and EUROPART programmes, has resulted in the development of multi-cycle processes for minor actinide partitioning. These multi-cycle processes are based on the co-separation of trivalent actinides and lanthanides (e.g. by the DIAMEX process), followed by the subsequent actinide(III)/lanthanide(III) group separation in the SANEX process. The current direction of research for the development of innovative processes within the recent European ACSEPT project is discussed additionally. This paper is focused on the development of flow-sheets for recovery of americium and curium from highly active waste solutions. The flow-sheets are verified by demonstration processes, in centrifugal contactors, using synthetic or genuine fuel solutions. The feasibility of the processes is also discussed. (orig.)

Organics Characterization Of DWPF Alternative Reductant Simulants, Glycolic Acid, And Antifoam 747

International Nuclear Information System (INIS)

White, T. L.; Wiedenman, B. J.; Lambert, D. P.; Crump, S. L.; Fondeur, F. F.; Papathanassiu, A. E.; Kot, W. K.; Pegg, I. L.

2013-01-01

The present study examines the fate of glycolic acid and other organics added in the Chemical Processing Cell (CPC) of the Defense Waste Processing Facility (DWPF) as part of the glycolic alternate flowsheet. Adoption of this flowsheet is expected to provide certain benefits in terms of a reduction in the processing time, a decrease in hydrogen generation, simplification of chemical storage and handling issues, and an improvement in the processing characteristics of the waste stream including an increase in the amount of nitrate allowed in the CPC process. Understanding the fate of organics in this flowsheet is imperative because tank farm waste processed in the CPC is eventually immobilized by vitrification; thus, the type and amount of organics present in the melter feed may affect optimal melt processing and the quality of the final glass product as well as alter flammability calculations on the DWPF melter off gas. To evaluate the fate of the organic compounds added as the part of the glycolic flowsheet, mainly glycolic acid and antifoam 747, samples of simulated waste that was processed using the DWPF CPC protocol for tank farm sludge feed were generated and analyzed for organic compounds using a variety of analytical techniques at the Savannah River National Laboratory (SRNL). These techniques included Ion Chromatography (IC), Gas Chromatography-Mass Spectrometry (GC-MS), Inductively Coupled Plasma-Atomic Emission Spectroscopy (ICP-AES), and Nuclear Magnetic Resonance (NMR) Spectroscopy. A set of samples were also sent to the Catholic University of America Vitreous State Laboratory (VSL) for analysis by NMR Spectroscopy at the University of Maryland, College Park. Analytical methods developed and executed at SRNL collectively showed that glycolic acid was the most prevalent organic compound in the supernatants of Slurry Mix Evaporator (SME) products examined. Furthermore, the studies suggested that commercially available glycolic acid contained minor amounts

Organics Characterization Of DWPF Alternative Reductant Simulants, Glycolic Acid, And Antifoam 747

Energy Technology Data Exchange (ETDEWEB)

White, T. L. [Savannah River Site (SRS), Aiken, SC (United States); Wiedenman, B. J. [Savannah River Site (SRS), Aiken, SC (United States); Lambert, D. P. [Savannah River Site (SRS), Aiken, SC (United States); Crump, S. L. [Savannah River Site (SRS), Aiken, SC (United States); Fondeur, F. F. [Savannah River Site (SRS), Aiken, SC (United States); Papathanassiu, A. E. [Catholic University of America Vitreous State Laboratory, Washington, DC (United States); Kot, W. K. [Catholic University of America Vitreous State Laboratory, Washington, DC (United States); Pegg, I. L. [Catholic University of America Vitreous State Laboratory, Washington, DC (United States)

2013-10-01

The present study examines the fate of glycolic acid and other organics added in the Chemical Processing Cell (CPC) of the Defense Waste Processing Facility (DWPF) as part of the glycolic alternate flowsheet. Adoption of this flowsheet is expected to provide certain benefits in terms of a reduction in the processing time, a decrease in hydrogen generation, simplification of chemical storage and handling issues, and an improvement in the processing characteristics of the waste stream including an increase in the amount of nitrate allowed in the CPC process. Understanding the fate of organics in this flowsheet is imperative because tank farm waste processed in the CPC is eventually immobilized by vitrification; thus, the type and amount of organics present in the melter feed may affect optimal melt processing and the quality of the final glass product as well as alter flammability calculations on the DWPF melter off gas. To evaluate the fate of the organic compounds added as the part of the glycolic flowsheet, mainly glycolic acid and antifoam 747, samples of simulated waste that was processed using the DWPF CPC protocol for tank farm sludge feed were generated and analyzed for organic compounds using a variety of analytical techniques at the Savannah River National Laboratory (SRNL). These techniques included Ion Chromatography (IC), Gas Chromatography-Mass Spectrometry (GC-MS), Inductively Coupled Plasma-Atomic Emission Spectroscopy (ICP-AES), and Nuclear Magnetic Resonance (NMR) Spectroscopy. A set of samples were also sent to the Catholic University of America Vitreous State Laboratory (VSL) for analysis by NMR Spectroscopy at the University of Maryland, College Park. Analytical methods developed and executed at SRNL collectively showed that glycolic acid was the most prevalent organic compound in the supernatants of Slurry Mix Evaporator (SME) products examined. Furthermore, the studies suggested that commercially available glycolic acid contained minor amounts

Preliminary technical data summary No. 3 for the Defense Waste Processing Facility

International Nuclear Information System (INIS)

Landon, L.F.

1980-05-01

This document presents an update on the best information presently available for the purpose of establishing the basis for the design of a Defense Waste Processing Facility. Objective of this project is to provide a facility to fix the radionuclides present in Savannah River Plant (SRP) high-level liquid waste in a high-integrity form (glass). Flowsheets and material balances reflect the alternate CAB case including the incorporation of low-level supernate in concrete

The acid aging as alternative process for uranium recovery from silicated ores

International Nuclear Information System (INIS)

Cipriani, M.; Della Testa, A.

1984-01-01

The influence of different variables on the extraction uranium efficiency and on the silicate solubility by means of acid aging is studied. The variables studied in bench scale were: acid/ore, oxidizing/ore and liquid/solid relationships; reaction time; temperature and recovery time. The results are discussed and compared with the ones of continuous operation of a semi-pilot plant. A flowsheet of the industrial process application is presented. (M.A.C.) [pt

Separation of carbohydrate and protein from wheat for the production of energy and food: conventional and proposed process

Energy Technology Data Exchange (ETDEWEB)

Hunwick, R J

1980-09-01

Historically, wheat has been wet-fractionated to produce starch and gluten, items of value for a broad range of industries as diverse as baking, paper manufacture and sweetener production. In Australia wheat flour has traditionally been the raw material for starch and gluten production with demand for gluten largely dictating starch production. Although this industry is of considerable economic significance in this country, plant throughputs are quite small in a global context. This situation could change dramatically if alcohol derived from wheat were to make a significant contribution to Australia's transport fuel requirements. This paper examines in general terms the impact such a trend could have on starch production in Australia. Traditional flowsheets based upon wheat flour as the raw material are discussed, the most important being the Martin process in which a thick dough is made which is repeatedly washed to liberate starch, bran and solubles as a starch 'milk' from the gluten mass. The starch milk is refined to fractionate its components into relatively pure materials. Recent efforts to improve this technology have been directed towards lowering water consumption mainly to simplify effluent disposal. These have led to the various batter processes which are briefly described. When the object is to produce large quantities of alcohol it is questioned whether it is justified to commence with flour. Whole wheat may be a better feedstock whence wheat could be wet-milled in a manner similar to that employed on a massive scale in North America, in particular for corn (maize). Current corn wet-milling practice is mentioned as an introduction to a summary of novel wet wheat milling flowsheets. Equipment generally used in these flowsheets is described.

Process Design and Evaluation for Chemicals Based on Renewable Resources

DEFF Research Database (Denmark)

Fu, Wenjing

. In addition, another characteristic of chemicals based on renewable feedstocks is that many alternative technologies and possible routes exist, resulting in many possible process flowsheets. The challenge for process engineers is then to choose between possible process routes and alternative technologies...... development of chemicals based on renewable feedstocks. As an example, this thesis especially focuses on applying the methodology in process design and evaluation of the synthesis of 5-hydroxymethylfurfural (HMF) from the renewable feedstock glucose/fructose. The selected example is part of the chemoenzymatic......One of the key steps in process design is choosing between alternative technologies, especially for processes producing bulk and commodity chemicals. Recently, driven by the increasing oil prices and diminishing reserves, the production of bulk and commodity chemicals from renewable feedstocks has...

Process control guidelines for CY 70 thorium campaign

International Nuclear Information System (INIS)

Jackson, R.R.

1970-01-01

The report comprises five parts, with part I being an introduction. Part II consists of a general treatment of process control methods. Parts III through V discuss, in the flowsheet sequence, those problems pertinent to each equipment piece or system and provide operating guidelines. Specific operations that are somewhat different from those normally encountered in Purex are discussed at length. Operations routine to Purex can be found in the pertinent standard operating procedures. Part VI describes in general terms the sequence to be followed in initiating and completing a variety of transient conditions

Process Options Description for Vitrification Flowsheet Model of INEEL Sodium Bearing Waste

International Nuclear Information System (INIS)

Nichols, T.T.; Taylor, D.D.; Lauerhass, L.; Barnes, C.M.

2002-01-01

The technical information required for the development of a basic steady-state process simulation of the vitrification treatment train of sodium bearing waste (SBW) at Idaho National Engineering and Environmental Laboratory (INEEL) is presented. The objective of the modeling effort is to provide the predictive capability required to optimize an entire treatment train and assess system-wide impacts of local changes at individual unit operations, with the aim of reducing the schedule and cost of future process/facility design efforts. All the information required a priori for engineers to construct and link unit operation modules in a commercial software simulator to represent the alternative treatment trains is presented. The information is of a mid- to high-level nature and consists of the following: (1) a description of twenty-four specific unit operations--their operating conditions and constraints, primary species and key outputs, and the initial modeling approaches that will be used in the first year of the simulation's development; (2) three potential configurations of the unit operations (trains) and their interdependencies via stream connections; and (3) representative stream compositional makeups

Process Options Description for Vitrification Flowsheet Model of INEEL Sodium Bearing Waste

Energy Technology Data Exchange (ETDEWEB)

Nichols, T.T.; Taylor, D.D.; Lauerhass, L.; Barnes, C.M.

2002-02-21

The technical information required for the development of a basic steady-state process simulation of the vitrification treatment train of sodium bearing waste (SBW) at Idaho National Engineering and Environmental Laboratory (INEEL) is presented. The objective of the modeling effort is to provide the predictive capability required to optimize an entire treatment train and assess system-wide impacts of local changes at individual unit operations, with the aim of reducing the schedule and cost of future process/facility design efforts. All the information required a priori for engineers to construct and link unit operation modules in a commercial software simulator to represent the alternative treatment trains is presented. The information is of a mid- to high-level nature and consists of the following: (1) a description of twenty-four specific unit operations--their operating conditions and constraints, primary species and key outputs, and the initial modeling approaches that will be used in the first year of the simulation's development; (2) three potential configurations of the unit operations (trains) and their interdependencies via stream connections; and (3) representative stream compositional makeups.

Continuous precipitation process of plutonium salts; Procede continu de precipitation des sels de plutonium

Energy Technology Data Exchange (ETDEWEB)

Richard, P [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1967-03-01

This work concerns the continuous precipitation process of plutonium oxalate. Investigations about the solubility of different valence states in nitric-oxalic and in nitric-sulfuric-oxalic medium lead to select the precipitation process of tetravalent plutonium oxalate. Settling velocity and granulometry of tetravalent oxalate plutonium have been studied with variation of several precipitation parameters such as: temperature, acidity, excess of oxalic acid and aging time. Then are given test results of some laboratory continuous apparatus. Conditions of operation with adopted tubular apparatus are defined in conclusion. A flow-sheet is given for a process at industrial scale. (author) [French] Cette etude porte sur la precipitation continue de l'oxalate de plutonium. L'etude de la solubilite des differentes valences du plutonium dans des milieux acides nitrique-oxalique, puis nitrique-sulfurique-oxalique conduit a choisir la precipitation de l'oxalate de plutonium tetravalent. L'etude porte ensuite sur la sedimentation et la granulometrie de l'oxalate de Pu{sup 4+} obtenue en faisant varier differents parametres de la precipitation : la temperature, l'acidite, l'exces oxalique et le temps de murissement. La derniere partie traite des resultats obtenus avec plusieurs types d'appareils continus essayes au laboratoire. En conclusion sont donnees les conditions de marche de l'appareil tubulaire adopte, ainsi qu'une extrapolation a l'echelle industrielle sous forme d'un flow-sheet. (auteur)

A systematic methodology for the design of continuous active pharmaceutical ingredient production processes

DEFF Research Database (Denmark)

Cervera Padrell, Albert Emili; Gani, Rafiqul; Kiil, Søren

2011-01-01

Continuous pharmaceutical manufacturing (CPM) has emerged as a powerful technology to obtain higher reaction yields and improved separation efficiencies, potentially leading to simplified process flowsheets, reduced total costs, lower environmental impacts, and safer and more flexible production...... and representation, as well as on how to employ this knowledge for process (re-)design. The aim of this paper is to introduce a methodology that systematically identifies already existing PSE methods and tools which can assist in the design of CPM processes. This methodology has been applied to a process...... for the production of an API developed by H. Lundbeck A/S, demonstrating the mentioned potential benefits that CPM can offer....

Thin film silicon solar cells: advanced processing and characterization - Final report

Energy Technology Data Exchange (ETDEWEB)

Ballif, Ch.

2008-04-15

This final report elaborated for the Swiss Federal Office of Energy (SFOE) takes a look at the results of a project carried out at the photovoltaics laboratory at the University of Neuchatel in Switzerland. The project aimed to demonstrate the production of high-efficiency thin-film silicon devices on flexible substrates using low cost processes. New ways of improving processing and characterisation are examined. The process and manufacturing know-how necessary to provide support for industrial partners within the framework of further projects is discussed. The authors state that the efficiency of most devices was significantly improved, both on glass substrates and on flexible plastic foils. The process reproducibility was also improved and the interactions between the different layers in the device are now said to be better understood. The report presents the results obtained and discusses substrate materials, transparent conductors, defect analyses and new characterisation tools. Finally, the laboratory infrastructure is described.

Uranium milling costs

International Nuclear Information System (INIS)

Coleman, R.B.

1980-01-01

Basic process flowsheets are reviewed for conventional milling of US ores. Capital costs are presented for various mill capacities for one of the basic processes. Operating costs are shown for various mill capacities for all of the basic process flowsheets. The number of mills using, or planning to use, a particular process is reviewed. A summary of the estimated average milling costs for all operating US mills is shown

Space Processing Applications Rocket project, SPAR 1. Final report

International Nuclear Information System (INIS)

Reeves, F.; Chassay, R.

1976-12-01

The experiment objectives, design/operational concepts, and final results of each of nine scientific experiments conducted during the first Space Processing Applications Rocket (SPAR) flight are summarized. The nine individual SPAR experiments, covering a wide and varied range of scientific materials processing objectives, were entitled: solidification of Pb-Sb eutectic, feasibility of producing closed-cell metal foams, characterization of rocket vibration environment by measurement of mixing of two liquids, uniform dispersions of crystallization processing, direct observation of solidification as a function of gravity levels, casting thoria dispersion-strengthened interfaces, contained polycrystalline solidification, and preparation of a special alloy for manufacturing of magnetic hard superconductor under zero-g environment

Hydrogen generation during melter feed preparation of Tank 42 sludge and salt washed loaded CST in the Defense Waste Processing Facility

International Nuclear Information System (INIS)

Daniel, W.E.

1999-01-01

The main objective of these scoping tests was to measure the rate of hydrogen generation in a series of experiments designed to duplicate the expected SRAT and SME processing conditions in laboratory scale vessels. This document details the testing performed to determine the maximum hydrogen generation expected with a coupled flowsheet of sludge, loaded CST [crystalline silicotitanate], and frit

TAME (tertiary-amyl-methyl ether) gasoline additive production process development; Desenvolvimento do processo de producao de TAME-aditivo para gasolina

Energy Technology Data Exchange (ETDEWEB)

Lovisi, Humberto [Petroflex Industria e Comercio S.A., Duqye de Caxias, RJ (Brazil); Piccoli, Ricardo [COPESUL, Companhia Petroquimica do Sul, Triunfo, RS (Brazil)

1992-12-31

PETROFLEX and COPESUL jointly developed a TAME production process. Tertiary-amyl-methyl ether (TAME) is obtained by the methoxylation of isoamylenes (2-methyl-1-butene and 2-methyl-2-butene) in a C{sub s} cut over a sulfonic acid resin. Process was developed on the basis of pilot plant and batch experiments. A simplified process flow-sheet and pilot plant data are presented. Isoamylenes conversions higher than 70% were achieved with low by-products formation. (author) 22 refs., 2 figs., 2 tabs.

TAME (tertiary-amyl-methyl ether) gasoline additive production process development; Desenvolvimento do processo de producao de TAME-aditivo para gasolina

Energy Technology Data Exchange (ETDEWEB)

Lovisi, Humberto [Petroflex Industria e Comercio S.A., Duqye de Caxias, RJ (Brazil); Piccoli, Ricardo [COPESUL, Companhia Petroquimica do Sul, Triunfo, RS (Brazil)

1993-12-31

PETROFLEX and COPESUL jointly developed a TAME production process. Tertiary-amyl-methyl ether (TAME) is obtained by the methoxylation of isoamylenes (2-methyl-1-butene and 2-methyl-2-butene) in a C{sub s} cut over a sulfonic acid resin. Process was developed on the basis of pilot plant and batch experiments. A simplified process flow-sheet and pilot plant data are presented. Isoamylenes conversions higher than 70% were achieved with low by-products formation. (author) 22 refs., 2 figs., 2 tabs.

Final processing vessel for radioactive waste

International Nuclear Information System (INIS)

Tejima, Takaya; Hiraki, Akimitsu.

1989-01-01

An inorganic inner layer comprising dense inorganic material such as organic polymer-impregnated concretes is formed to about 10 - 50 mm in average thickness at the inside of a metal vessel. Further, the surface of the vessel is formed as a flat surface with no or only small reinforcing protrusions. Thus, if the final processing vessel should be dropped during transportation or handling by mistake, since impact shocks do not concentrate to protrusions as usual, no local stress concentration occurs to the inorganic inner liner layer. Accordingly, the risk of rapture can be reduced greatly. Further, since impact shock resistance layer put between the metal vessel and the inorganic inner liner layer absorbs shocks, a further sufficient strength can be obtained against dropping accident. (T.M.)

Alternate Reductant Cold Cap Evaluation Furnace Phase II Testing

Energy Technology Data Exchange (ETDEWEB)

Johnson, F. C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Stone, M. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Miller, D. H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2014-09-03

Savannah River Remediation (SRR) conducted a Systems Engineering Evaluation (SEE) to determine the optimum alternate reductant flowsheet for the Defense Waste Processing Facility (DWPF). Specifically, two proposed flowsheets (nitric–formic–glycolic and nitric–formic–sugar) were evaluated based upon results from preliminary testing. Comparison of the two flowsheets among evaluation criteria indicated a preference towards the nitric–formic–glycolic flowsheet. Further research and development of this flowsheet eliminated the formic acid, and as a result, the nitric–glycolic flowsheet was recommended for further testing. Based on the development of a roadmap for the nitric–glycolic acid flowsheet, Waste Solidification Engineering (WS-E) issued a Technical Task Request (TTR) to address flammability issues that may impact the implementation of this flowsheet. Melter testing was requested in order to define the DWPF flammability envelope for the nitric-glycolic acid flowsheet. The Savannah River National Laboratory (SRNL) Cold Cap Evaluation Furnace (CEF), a 1/12^th scale DWPF melter, was selected by the SRR Alternate Reductant project team as the melter platform for this testing. The overall scope was divided into the following sub-tasks as discussed in the Task Technical and Quality Assurance Plan (TTQAP): Phase I - A nitric–formic acid flowsheet melter test (unbubbled) to baseline the CEF cold cap and vapor space data to the benchmark melter flammability models; Phase II - A nitric–glycolic acid flowsheet melter test (unbubbled and bubbled) to: Define new cold cap reactions and global kinetic parameters in support of the melter flammability model development; Quantify off-gas surging potential of the feed; Characterize off-gas condensate for complete organic and inorganic carbon species. After charging the CEF with cullet from Phase I CEF testing, the melter was slurry-fed with glycolic flowsheet based SB6-Frit 418 melter feed at 36% waste

Potential dispositioning flowsheets for ICPP SNF and wastes

Energy Technology Data Exchange (ETDEWEB)

Olson, A.L. [ed.; Anderson, P.A.; Bendixsen, C.L. [and others

1995-11-01

The Idaho Chemical Processing Plant (ICPP), located at the Idaho National Laboratory (INEL), has reprocessed irradiated nuclear fuels for the US Department of Energy (DOE) since 1953. This activity resulted mainly in the recovery of uranium and the management of the resulting wastes. The acidic radioactive high-level liquid waste was routinely stored in stainless steel tanks and then calcined to form a dry granular solid. The calcine is stored in stainless steel bins that are housed in underground concrete vaults. In April 1992, the DOE discontinued the practice of reprocessing irradiated nuclear fuels. This decision has left a legacy of 1.8 million gallons of radioactive liquid wastes (1.5 million gallons of radioactive sodium-bearing liquid wastes and 0.3 million gallons of high-level liquid waste), 3800 cubic meters of calcine waste, and 289 metric tons of heavy metal within unprocessed spent nuclear fuel (SNF) left in inventory at the ICPP. The nation`s radioactive waste policy has been established by the Nuclear Waste Policy Act (NWPA), which requires the final disposal of SNF and radioactive waste in accordance with US Environmental Protection Agency (EPA) and Nuclear Regulatory Commission (NRC) standards. In accordance with these regulations and other legal agreements between the State of Idaho and the DOE, the DOE must, among other requirements, (1) complete a final Environmental Impact Statement by April 30, 1995, (2) evaluate and test sodium-bearing waste pre-treatment technologies, (3) select the sodium-bearing and calcine waste pre-treatment technology, if necessary, by June 1, 1995, and (4) select a technology for converting calcined waste into an appropriate disposal form by June 1, 1995.

Potential dispositioning flowsheets for ICPP SNF and wastes

International Nuclear Information System (INIS)

Olson, A.L.; Anderson, P.A.; Bendixsen, C.L.

1995-11-01

The Idaho Chemical Processing Plant (ICPP), located at the Idaho National Laboratory (INEL), has reprocessed irradiated nuclear fuels for the US Department of Energy (DOE) since 1953. This activity resulted mainly in the recovery of uranium and the management of the resulting wastes. The acidic radioactive high-level liquid waste was routinely stored in stainless steel tanks and then calcined to form a dry granular solid. The calcine is stored in stainless steel bins that are housed in underground concrete vaults. In April 1992, the DOE discontinued the practice of reprocessing irradiated nuclear fuels. This decision has left a legacy of 1.8 million gallons of radioactive liquid wastes (1.5 million gallons of radioactive sodium-bearing liquid wastes and 0.3 million gallons of high-level liquid waste), 3800 cubic meters of calcine waste, and 289 metric tons of heavy metal within unprocessed spent nuclear fuel (SNF) left in inventory at the ICPP. The nation's radioactive waste policy has been established by the Nuclear Waste Policy Act (NWPA), which requires the final disposal of SNF and radioactive waste in accordance with US Environmental Protection Agency (EPA) and Nuclear Regulatory Commission (NRC) standards. In accordance with these regulations and other legal agreements between the State of Idaho and the DOE, the DOE must, among other requirements, (1) complete a final Environmental Impact Statement by April 30, 1995, (2) evaluate and test sodium-bearing waste pre-treatment technologies, (3) select the sodium-bearing and calcine waste pre-treatment technology, if necessary, by June 1, 1995, and (4) select a technology for converting calcined waste into an appropriate disposal form by June 1, 1995

Evaluation of high-level waste vitrification feed preparation chemistry for an NCAW simulant, FY 1994: Alternate flowsheets (DRAFT)

International Nuclear Information System (INIS)

Smith, H.D.; Merz, M.D.; Wiemers, K.D.; Smith, G.L.

1996-02-01

High-level radioactive waste stored in tanks at the U.S. Department of Energy's (DOE's) Hanford Site will be pretreated to concentrate radioactive constituents and fed to the vitrification plant A flowsheet for feed preparation within the vitrification plant (based on the Hanford Waste Vitrification Plant (HWVP) design) called for HCOOH addition during the feed preparation step to adjust rheology and glass redox conditions. However, the potential for generating H 2 and NH 3 during treatment of high-level waste (HLW) with HCOOH was identified at Pacific Northwest Laboratory (PNL). Studies at the University of Georgia, under contract with Savannah River Technology Center (SRTC) and PNL, have verified the catalytic role of noble metals (Pd, Rh, Ru), present in the waste, in the generation of H 2 and NH 3 . Both laboratory-scale and pilot-scale studies at SRTC have documented the H 2 and NH 3 generation phenomenal Because H 2 and NH 3 may create hazardous conditions in the vessel vapor space and offgas system of a vitrification plant, reducing the H 2 generation rate and the NH 3 generation to the lowest possible levels consistent with desired melter feed characteristics is important. The Fiscal Year 1993 and 1994 studies were conducted with simulated (non-radioactive), pre-treated neutralized current acid waste (NCAW). Neutralized current acid waste is a high-level waste originating from the plutonium/uranium extraction (PUREX) plant that has been partially denitrated with sugar, neutralized with NaOH, and is presently stored in double-shell tanks. The non-radioactive simulant used for the present study includes all of the trace components found in the waste, or substitutes a chemically similar element for radioactive or very toxic species. The composition and simulant preparation steps were chosen to best simulate the chemical processing characteristics of the actual waste

Separation of carbohydrate and protein from wheat for the production of energy and food: conventional and proposed process

Energy Technology Data Exchange (ETDEWEB)

Hunwick, R.J.

1980-09-01

Historically, wheat has been wet-fractionated to produce starch and gluten, items of value for a broad range of industries as diverse as baking, paper manufacture and sweetener production. In Australia wheat flour has traditionally been the raw material for starch and gluten production with demand for gluten largely dictating starch production. Although this industry is of considerable economic significance in this country, plant throughputs are quite small in a global context. This situation could change dramatically if alcohol derived from wheat were to make a significant contribution to Australia's transport fuel requirements. This paper examines in general terms the impact such a trend could have on starch production in Australia. Traditional flowsheets based upon wheat flour as the raw material are discussed, the most important being the Martin process in which a thick dough is made which is repeatedly washed to liberate starch, bran and solubles as a starch 'milk' from the gluten mass. The starch milk is refined to fractionate its components into relatively pure materials. Recent efforts to improve this technology have been directed towards lowering water consumption mainly to simplify effluent disposal. These have led to the various batter processes which are briefly described. When the object is to produce large quantities of alcohol it is questioned whether it is justified to commence with flour. Whole wheat may be a better feedstock whence wheat could be wet-milled in a manner similar to that employed on a massive scale in North America, in particular for corn (maize). Current corn wet-milling practice is mentioned as an introduction to a summary of novel wet wheat milling flowsheets. Equipment generally used in these flowsheets is described.

Analysis and optimal process development of the iodine-Sulfur cycle for nuclear hydrogen production

International Nuclear Information System (INIS)

Lee, Byung Jin

2009-02-01

solution, a flowsheet of I-S cycle was devised to generate a highly enriched hydrogen-iodide gas through a series of processes of liquid-liquid separation of product mixture from Bunsen reaction and flash of over-azeotropic HI solution. Operating temperature and pressure for HI enrichment need not to be increased as high as those for existing flowsheets; as a result, the operating conditions become less corrosive. Chance of pipe clogging due to iodine solidification is low because there is no process where iodine is concentrated that high. Enrichment of HI through spontaneous L-L phase separation and simple flash processes avoiding complicated separate process is considered to be an additional benefit. Analysis of overall and component material balances showed that excess amount of feed to each process to get a desired output depends on the efficiency of flash and decomposition processes. Compared to previous ones, the proposed flowsheet requires more recirculation flows throughout the whole cycle mainly because only a small portion of HI content exceeding the azeotrope is allowed to evaporate in the flash drum without employing a separate HI enrichment process. Thermal efficiency of the proposed flowsheet was evaluated, together with a series of parametric analyses for the sensitivity to key operating parameters and component performances. We show that thermal efficiency of higher than 60% is feasible if the system and operating conditions are optimized

Hydrogen generation in SRAT with nitric acid and late washing flowsheets

International Nuclear Information System (INIS)

Hsu, C.W.

1992-01-01

Recently, SRTC recommended two process changes: (1) a final wash of the tetraphenylborate precipitate feed slurry and (2) the use of nitric acid to neutralize the sludge in the SRAT. The first change produced an aqueous hydrolysis product (PHA) with higher formic acid/formate and copper concentration, and reduced the nitrate content in the PHA by an order of magnitude. The second change is to substitute part of formic acid added to the SRAT with nitric acid, and therefore may reduce the hydrogen generated in the SRAT as well as provide nitrate as an oxidant to balance the redox state of the melter feed. The purpose of this report is to determine the pertinent variables that could affect the hydrogen generation rate with these process changes

Fundamental Processes in Plasmas. Final report

International Nuclear Information System (INIS)

O'Neil, Thomas M.; Driscoll, C. Fred

2009-01-01

This research focuses on fundamental processes in plasmas, and emphasizes problems for which precise experimental tests of theory can be obtained. Experiments are performed on non-neutral plasmas, utilizing three electron traps and one ion trap with a broad range of operating regimes and diagnostics. Theory is focused on fundamental plasma and fluid processes underlying collisional transport and fluid turbulence, using both analytic techniques and medium-scale numerical simulations. The simplicity of these systems allows a depth of understanding and a precision of comparison between theory and experiment which is rarely possible for neutral plasmas in complex geometry. The recent work has focused on three areas in basic plasma physics. First, experiments and theory have probed fundamental characteristics of plasma waves: from the low-amplitude thermal regime, to inviscid damping and fluid echoes, to cold fluid waves in cryogenic ion plasmas. Second, the wide-ranging effects of dissipative separatrices have been studied experimentally and theoretically, finding novel wave damping and coupling effects and important plasma transport effects. Finally, correlated systems have been investigated experimentally and theoretically: UCSD experients have now measured the Salpeter correlation enhancement, and theory work has characterized the 'guiding center atoms of antihydrogen created at CERN

An evaluation of foaming potential in the IDMS melter

International Nuclear Information System (INIS)

Hutson, N.D.

1992-01-01

The present DWPF flowsheet calls for the chemical treatment of waste sludge with 90 wt% formic acid prior to the addition of the Precipitate Hydrolysis Aqueous (PHA) product. An alternative processing methodology, denoted the ''Nitric Acid Flowsheet'', has been proposed. In the application of this flowsheet, nitric acid would be used to neutralize sludge base components (hydroxides and carbonates) prior to the addition of late wash PHA. The late wash PHA will contain sufficient quantities of formic acid to adequately complete necessary reduction-oxidation (REDOX) reactions. The use of this flowsheet may result in a change in the nominal concentrations of two of the major REDOX reaction participants: formate (HCOO minus ) and nitrate (NO 3 minus )

Design of environmentally benign processes

DEFF Research Database (Denmark)

Hostrup, Martin; Harper, Peter Mathias; Gani, Rafiqul

1999-01-01

because of environmental constraints are particularly suited for solution with the hybrid method. Application of the hybrid method is highlighted through two illustrative examples. The first example involves the determination of an optimal flowsheet for the removal of a chemical species from an azeotropic...

Capturing connectivity and causality in complex industrial processes

CERN Document Server

Yang, Fan; Shah, Sirish L; Chen, Tongwen

2014-01-01

This brief reviews concepts of inter-relationship in modern industrial processes, biological and social systems. Specifically ideas of connectivity and causality within and between elements of a complex system are treated; these ideas are of great importance in analysing and influencing mechanisms, structural properties and their dynamic behaviour, especially for fault diagnosis and hazard analysis. Fault detection and isolation for industrial processes being concerned with root causes and fault propagation, the brief shows that, process connectivity and causality information can be captured in two ways: ·      from process knowledge: structural modeling based on first-principles structural models can be merged with adjacency/reachability matrices or topology models obtained from process flow-sheets described in standard formats; and ·      from process data: cross-correlation analysis, Granger causality and its extensions, frequency domain methods, information-theoretical methods, and Bayesian ne...

Process development study on production of uranium metal from monazite sourced crude uranium tetra-fluoride

International Nuclear Information System (INIS)

Chowdhury, S; Satpati, S.K.; Hareendran, K.N.; Roy, S.B.

2014-01-01

Development of an economic process for recovery, process flow sheet development, purification and further conversion to nuclear grade uranium metal from the crude UF 4 has been a technological challenge and the present paper, discusses the same.The developed flow-sheet is a combination of hydrometallurgical and pyrometallurgical processes. Crude UF 4 is converted to uranium di-oxide (UO 2 ) by chemical conversion route and UO 2 produced is made fluoride-free by repeated repulping, followed by solid liquid separation. Uranium di-oxide is then purified by two stages of dissolution and suitable solvent extraction methods to get uranium nitrate pure solution (UNPS). UNPS is then precipitated with air diluted ammonia in a leak tight stirred vessel under controlled operational conditions to obtain ammonium di-uranate (ADU). The ADU is then calcined and reduced to produce metal grade UO 2 followed by hydro-fluorination using anhydrous hydrofluoric acid to obtain metal grade UF 4 with ammonium oxalate insoluble (AOI) content of 4 is essential for critical upstream conversion process. Nuclear grade uranium metal ingot is finally produced by metallothermic reduction process at 650℃ in a closed vessel, called bomb reactor. In the process, metal-slag separation plays an important role for attaining metal purity as well as process yield. Technological as well economic feasibility of indigenously developed process for large scale production of uranium metal from the crude UF 4 has been established in Bhabha Atomic Research Centre (BARC), India

Processing and solidification of Savannah River Plant high-level waste

International Nuclear Information System (INIS)

Kelley, J.A.

1981-01-01

The entire flowsheet for processing and solidification of Savannah River Plant (SRP) high-level wastes has been demonstrated. A new small-scale integrated pilot plant is operating with actual radioactive wastes, and large-scale equipment is being demonstrated with nonradioactive simulated wastes. Design of a full-scale waste solidification plant is in progress. Plant construction is expected to begin in 1983, and startup is anticipated in 1988. The plant will poduce about 500 cans of glass per year with each can containing about 1.5 tons of glass

A review of United States yellow cake precipitation practice

International Nuclear Information System (INIS)

Litz, J.E.; Coleman, R.B.

1980-01-01

The various process flowsheets used to produce concentrated uranium solutions are reviewed. The choices of flowsheets are affected by ore alkalinity, uranium mineralization, and the impurities solubilized during leaching. The techniques used to precipitate yellow cake from concentrated uranium solutions are reviewed. Consideration is given to precipitation chemistry, reagent requirements, and process equipment and costs for precipitation, dewatering, drying and calcining. (author)

Gold-copper ores processing-Case study: optimization of flotation residue cyanidation

International Nuclear Information System (INIS)

McMullen, J.; Pelletier, P.; Breau, Y.; Pelletier, D.

1999-01-01

Typically, economic optimization of Gold-Copper ore processing presents challenges. Barrick's Bousquet II ore is a good example where many processing units such as gravity, flotation and cyanidation are required to efficiently recover the metals from the ore. Flowsheet criterion, operating strategy selections, cyanidation process optimization and its inter-dependence with flotation are discussed in detail. Real-time conservation integration of the cyanide control strategy has allowed a reduction of cyanide consumption of over 40% since 1994, while maintaining or improving metals recovery. A detailed analysis of the cyanidation control strategy such as key process sensor reliability and accuracy, process monitoring and fault detection is presented. This robust and efficient control strategy is a key building block that enhances the overall economic return of the process plant. (author)

Uranium processing developments

International Nuclear Information System (INIS)

Jones, J.Q.

1977-01-01

The basic methods for processing ore to recover the contained uranium have not changed significantly since the 1954-62 period. Improvements in mill operations have been the result of better or less expensive reagents, changes in equipment, and in the successful resolvement of many environmental matters. There is also an apparent trend toward large mills that can profitably process lower grade ores. The major thrust in the near future will not be on process technology but on the remaining environmental constraints associated with milling. At this time the main ''spot light'' is on tailings dam and impoundment area construction and reclamation. Plans must provide for an adequate safety factor for stability, no surface or groundwater contamination, and minimal discharge of radionuclides to unrestricted areas, as may be required by law. Solution mining methods must also provide for plans to restore the groundwater back to its original condition as defined by local groundwater regulations. Basic flowsheets (each to finished product) plus modified versions of the basic types are shown

Chemical analysis used in nuclear fuels reprocessing of uranium and thorium

International Nuclear Information System (INIS)

Schvartzman, M.M.A.M.

1986-01-01

An overall review of the analytical chemistry in nuclear fuel reprocessing is done. In Purex and Thorex process flowsheets, the analyses required to the control of the process, balance and accountability of fissile and fertile materials, and final product specification are pointed out. Some analytical methods applied to the determination of uranium, plutonium, thorium, nitric acid, tributylphosphate and fission products are described. Specific features of the analytical laboratories are presented. The radioactivity level of the samples requires facilities as shielded cells and glove boxes, and handling by remote control. Finally it is reported an application of one analytical method to evaluate thorium content in organic and aqueous solutions, in cold tests of Thorex process. These tests were performed at CDTN/NUCLEBRAS. (author) [pt

Waste Minimization Study on Pyrochemical Reprocessing Processes

International Nuclear Information System (INIS)

Boussier, H.; Conocar, O.; Lacquement, J.

2006-01-01

' new block diagram allowing internal solvent recycling, and self eliminating reactants. This new flowsheet minimizes the quantity of inactive inlet flows that would have inevitably to be incorporated in a final waste form. The study identifies all knowledge gaps to be filled and suggest some possible R and D issues to confirm or infirm the feasibility of the proposed process fittings. (authors)

Dynamic (G2) Model Design Document, 24590-WTP-MDD-PR-01-002, Rev. 12

Energy Technology Data Exchange (ETDEWEB)

Deng, Yueying; Kruger, Albert A.

2013-12-16

The Hanford Tank Waste Treatment and Immobilization Plant (WTP) Statement of Work (Department of Energy Contract DE-AC27-01RV14136, Section C) requires the contractor to develop and use process models for flowsheet analyses and pre-operational planning assessments. The Dynamic (G2) Flowsheet is a discrete-time process model that enables the project to evaluate impacts to throughput from eventdriven activities such as pumping, sampling, storage, recycle, separation, and chemical reactions. The model is developed by the Process Engineering (PE) department, and is based on the Flowsheet Bases, Assumptions, and Requirements Document (24590-WTP-RPT-PT-02-005), commonly called the BARD. The terminologies of Dynamic (G2) Flowsheet and Dynamic (G2) Model are interchangeable in this document. The foundation of this model is a dynamic material balance governed by prescribed initial conditions, boundary conditions, and operating logic. The dynamic material balance is achieved by tracking the storage and material flows within the plant as time increments. The initial conditions include a feed vector that represents the waste compositions and delivery sequence of the Tank Farm batches, and volumes and concentrations of solutions in process equipment before startup. The boundary conditions are the physical limits of the flowsheet design, such as piping, volumes, flowrates, operation efficiencies, and physical and chemical environments that impact separations, phase equilibriums, and reaction extents. The operating logic represents the rules and strategies of running the plant.

Source term development for the 300 Area Treated Effluent Disposal Facility

International Nuclear Information System (INIS)

Bendixsen, R.B.

1994-04-01

A novel method for developing a source term for radiation and hazardous material content of sludge processing equipment and barrels in a new waste water treatment facility is presented in this paper. The 300 Area Treated Effluent Disposal Facility (TEDF), located at the Hanford Site near Richland, Washington, will treat process sewer waste water from the 300 Area and discharge a permittable effluent flow into the Columbia River. A process information and hazards analysis document needed a process flowsheet detailing the concentrations of radionuclides, inorganics, and organics throughout the process, including the sludge effluent flow. A hazards analysis for a processing facility usually includes a flowsheet showing the process, materials, heat balances, and instrumentation for that facility. The flow sheet estimates stream flow quantities, activities, compositions, and properties. For the 300 Area TEDF, it was necessary to prepare the flow sheet with all of the information so that radiation doses to workers could be estimated. The noble method used to develop the 300 Area TEDF flowsheet included generating recycle factors. To prepare each component in the flowsheet, precipitation, destruction, and two recycle factors were developed. The factors were entered into a spreadsheet and provided a method of estimating the steady-state concentrations of all of the components in the facility. This report describes how the factors were developed, explains how they were used in developing the flowsheet, and presents the results of using these values to estimate radiation doses for personnel working in the facility. The report concludes with a discussion of the effect of estimates of radioactive and hazardous material concentrations on shielding design and the need for containment features for equipment in the facility

Electrochemical processing of carbon dioxide.

Science.gov (United States)

Oloman, Colin; Li, Hui

2008-01-01

With respect to the negative role of carbon dioxide on our climate, it is clear that the time is ripe for the development of processes that convert CO(2) into useful products. The electroreduction of CO(2) is a prime candidate here, as the reaction at near-ambient conditions can yield organics such as formic acid, methanol, and methane. Recent laboratory work on the 100 A scale has shown that reduction of CO(2) to formate (HCO(2)(-)) may be carried out in a trickle-bed continuous electrochemical reactor under industrially viable conditions. Presuming the problems of cathode stability and formate crossover can be overcome, this type of reactor is proposed as the basis for a commercial operation. The viability of corresponding processes for electrosynthesis of formate salts and/or formic acid from CO(2) is examined here through conceptual flowsheets for two process options, each converting CO(2) at the rate of 100 tonnes per day.

Possibilities for reusing the waste from the process of Zn-Pb ore beneficiation

Directory of Open Access Journals (Sweden)

Cichy Krystian

2017-01-01

Full Text Available This paper discusses the areas of storage, resources, and granulometric and chemical characteristics of old Zn-Pb tailings stored in heaps in the city of Bytom area. It presents the results of laboratory tests for development of the technological flowsheet for transformation of the material into Zn- Pb sulfide concentrates and the results of trials in an experimental system of the beneficiation flowsheet which was developed. In the further part of the paper, the results of the research work on preparation of the tailings with reduced metal content for further use are presented.

Experimental studies of an optimal operating condition for the Bunsen process in the I-S thermochemical cycle

International Nuclear Information System (INIS)

Yoon, Ho Joon; No, Hee Cheon; Kim, Young Soo; Jin, Hyung Gon; Lee, Jeong Ik; Lee, Byung Jin

2009-01-01

Conventional I-S cycles have critical limitations in material integrity and thermal efficiency. The HIx and sulfuric acids in high temperature and pressure cause serious material corrosions. They also carry too much water and iodine over the entire processes. To try to find a solution to these problems, KAIST proposed an optimal operating condition of Bunsen section through a parametric study of existing experimental data, and, based on it, devised a new flowsheet. When the contents of water and I 2 in the feed are controlled within the optimal band, HI concentration in HIx phase becomes strongly over-azeotropic. By simple flashing of the over-azeotropic HI solution, highly enriched HI vapor can be obtained, which leads to improved energy efficiency of the cycle. Since the cycle is operable under low pressures, the corrosivity of the operating condition can also be alleviated. In order to validate the previous experimental data and enhance the feasibility of the newly proposed flowsheet, KAIST is performing experiments. Procedure and results of early stage of experiments are introduced in this paper. (author)

Aqueous biphasic extraction of uranium and thorium from contaminated soils. Final report

International Nuclear Information System (INIS)

Chaiko, D.J.; Gartelmann, J.; Henriksen, J.L.; Krause, T.R.; Deepak; Vojta, Y.; Thuillet, E.; Mertz, C.J.

1995-07-01

The aqueous biphasic extraction (ABE) process for soil decontamination involves the selective partitioning of solutes and fine particulates between two immiscible aqueous phases. The biphase system is generated by the appropriate combination of a water-soluble polymer (e.g., polyethlene glycol) with an inorganic salt (e.g., sodium carbonate). Selective partitioning results in 99 to 99.5% of the soil being recovered in the cleaned-soil fraction, while only 0.5 to 1% is recovered in the contaminant concentrate. The ABE process is best suited to the recovery of ultrafine, refractory material from the silt and clay fractions of soils. During continuous countercurrent extraction tests with soil samples from the Fernald Environmental Management Project site (Fernald, OH), particulate thorium was extracted and concentrated between 6- and 16-fold, while the uranium concentration was reduced from about 500 mg/kg to about 77 mg/kg. Carbonate leaching alone was able to reduce the uranium concentration only to 146 mg/kg. Preliminary estimates for treatment costs are approximately $160 per ton of dry soil. A detailed flowsheet of the ABE process is provided

Evaluation of interim and final waste forms for the newly generated liquid low-level waste flowsheet

International Nuclear Information System (INIS)

Abotsi, G.M.K.; Bostick, D.T.; Beck, D.E.

1996-05-01

The purpose of this review is to evaluate the final forms that have been proposed for radioactive-containing solid wastes and to determine their application to the solid wastes that will result from the treatment of newly generated liquid low-level waste (NGLLLW) and Melton Valley Storage Tank (MVST) supernate at the Oak Ridge National Laboratory (ORNL). Since cesium and strontium are the predominant radionuclides in NGLLLW and MVST supernate, this review is focused on the stabilization and solidification of solid wastes containing these radionuclides in cement, glass, and polymeric materials-the principal waste forms that have been tested with these types of wastes. Several studies have shown that both cesium and strontium are leached by distilled water from solidified cement, although the leachabilities of cesium are generally higher than those of strontium under similar conditions. The situation is exacerbated by the presence of sulfates in the solution, as manifested by cracking of the grout. Additives such as bentonite, blast-furnace slag, fly ash, montmorillonite, pottery clay, silica, and zeolites generally decrease the cesium and strontium release rates. Longer cement curing times (>28 d) and high ionic strengths of the leachates, such as those that occur in seawater, also decrease the leach rates of these radionuclides. Lower cesium leach rates are observed from vitrified wastes than from grout waste forms. However, significant quantities of cesium are volatilized due to the elevated temperatures required to vitrify the waste. Hence, vitrification will generally require the use of cleanup systems for the off-gases to prevent their release into the atmosphere

Computer Aided Design and Analysis of Separation Processes with Electrolyte Systems

DEFF Research Database (Denmark)

A methodology for computer aided design and analysis of separation processes involving electrolyte systems is presented. The methodology consists of three main parts. The thermodynamic part "creates" the problem specific property model package, which is a collection of pure component and mixture...... property models. The design and analysis part generates process (flowsheet) alternatives, evaluates/analyses feasibility of separation and provides a visual operation path for the desired separation. The simulation part consists of a simulation/calculation engine that allows the screening and validation...... of process alternatives. For the simulation part, a general multi-purpose, multi-phase separation model has been developed and integrated to an existing computer aided system. Application of the design and analysis methodology is highlighted through two illustrative case studies....

Computer Aided Design and Analysis of Separation Processes with Electrolyte Systems

DEFF Research Database (Denmark)

Takano, Kiyoteru; Gani, Rafiqul; Kolar, P.

2000-01-01

A methodology for computer aided design and analysis of separation processes involving electrolyte systems is presented. The methodology consists of three main parts. The thermodynamic part 'creates' the problem specific property model package, which is a collection of pure component and mixture...... property models. The design and analysis part generates process (flowsheet) alternatives, evaluates/analyses feasibility of separation and provides a visual operation path for the desired separation. The simulation part consists of a simulation/calculation engine that allows the screening and validation...... of process alternatives. For the simulation part, a general multi-purpose, multi-phase separation model has been developed and integrated to an existing computer aided system. Application of the design and analysis methodology is highlighted through two illustrative case studies, (C) 2000 Elsevier Science...

High-level waste processing at the Savannah River Site: An update

International Nuclear Information System (INIS)

Marra, J.E.; Bennett, W.M.; Elder, H.H.; Lee, E.D.; Marra, S.L.; Rutland, P.L.

1997-01-01

The Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) in Aiken, SC mg began immobilizing high-level radioactive waste in borosilicate glass in 1996. Currently, the radioactive glass is being produced as a ''sludge-only'' composition by combining washed high-level waste sludge with glass frit. The glass is poured in stainless steel canisters which will eventually be disposed of in a permanent, geological repository. To date, DWPF has produced about 100 canisters of vitrified waste. Future processing operations will, be based on a ''coupled'' feed of washed high-level waste sludge, precipitated cesium, and glass frit. This paper provides an update of the processing activities completed to date, operational/flowsheet problems encountered, and programs underway to increase production rates

Computer application in coal preparation industry in China

Energy Technology Data Exchange (ETDEWEB)

Lu, M.; Wu, L.; Ni, Q. (China Univ. of Mining and Technology, Xuzhou (China))

1990-01-01

This paper describes several packages of microcomputer programs developed for designing and managing the coal preparation plants. Three parts are included: Coal Cleaning Package (CCP), Coal Preparation Optimization Program (CPO) and Coal Preparation Computer Aided Design System (CPCAD). The function of CCP is: evaluating and predicting coal cleaning result. Coal presentation process modelling and optimization; coal preparation flowsheet design and optimization. The CPO is a nonlinear optimization program. It can simulate and optimize the profit for different flowsheet to get the best combination of the final products. The CPCAD was developed based upon AutoCAD and makes full use of AutoLISP, digitizer menus and AutoCAD commands, combining the functions provided by AutoCAD and the principle used in conventional coal preparation plant design, forming a designer-oriented CPCAD system. These packages have proved to be reliable, flexible and easy to learn and use. They are a powerful tool for coal preparation plant design and management. (orig.).

EPR design tools. Integrated data processing tools

International Nuclear Information System (INIS)

Kern, R.

1997-01-01

In all technical areas, planning and design have been supported by electronic data processing for many years. New data processing tools had to be developed for the European Pressurized Water Reactor (EPR). The work to be performed was split between KWU and Framatome and laid down in the Basic Design contract. The entire plant was reduced to a logical data structure; the circuit diagrams and flowsheets of the systems were drafted, the central data pool was established, the outlines of building structures were defined, the layout of plant components was planned, and the electrical systems were documented. Also building construction engineering was supported by data processing. The tasks laid down in the Basic Design were completed as so-called milestones. Additional data processing tools also based on the central data pool are required for the phases following after the Basic Design phase, i.e Basic Design Optimization; Detailed Design; Management; Construction, and Commissioning. (orig.) [de

SELECTION AND PRELIMINARY EVALUATION OF ALTERNATIVE REDUCTANTS FOR SRAT PROCESSING

Energy Technology Data Exchange (ETDEWEB)

Stone, M.; Pickenheim, B.; Peeler, D.

2009-06-30

Defense Waste Processing Facility - Engineering (DWPF-E) has requested the Savannah River National Laboratory (SRNL) to perform scoping evaluations of alternative flowsheets with the primary focus on alternatives to formic acid during Chemical Process Cell (CPC) processing. The reductants shown below were selected for testing during the evaluation of alternative reductants for Sludge Receipt and Adjustment Tank (SRAT) processing. The reductants fall into two general categories: reducing acids and non-acidic reducing agents. Reducing acids were selected as direct replacements for formic acid to reduce mercury in the SRAT, to acidify the sludge, and to balance the melter REDuction/OXidation potential (REDOX). Non-acidic reductants were selected as melter reductants and would not be able to reduce mercury in the SRAT. Sugar was not tested during this scoping evaluation as previous work has already been conducted on the use of sugar with DWPF feeds. Based on the testing performed, the only viable short-term path to mitigating hydrogen generation in the CPC is replacement of formic acid with a mixture of glycolic and formic acids. An experiment using glycolic acid blended with formic on an 80:20 molar basis was able to reduce mercury, while also targeting a predicted REDuction/OXidation (REDOX) of 0.2 expressed as Fe{sup 2+}/{Sigma}Fe. Based on this result, SRNL recommends performing a complete CPC demonstration of the glycolic/formic acid flowsheet followed by a design basis development and documentation. Of the options tested recently and in the past, nitric/glycolic/formic blended acids has the potential for near term implementation in the existing CPC equipment providing rapid throughput improvement. Use of a non-acidic reductant is recommended only if the processing constraints to remove mercury and acidify the sludge acidification are eliminated. The non-acidic reductants (e.g. sugar) will not reduce mercury during CPC processing and sludge acidification would

Sustainable Approach for Landfill Management at Final Processing Site Cikundul in Sukabumi City, Indonesia

OpenAIRE

Sri Darwati

2012-01-01

The main problem of landfill management in Indonesia is the difficulty in getting a location for Final Processing Sites (FPS) due to limited land and high land prices. Besides, about 95% of existing landfills are uncontrolled dumping sites, which could potentially lead to water, soil and air pollution. Based on data from the Ministry of Environment (2010), The Act of the Republic of Indonesia Number 18 Year 2008 Concerning Solid Waste Management, prohibits open dumping at final processing sit...

Process modeling study of the CIF incinerator

International Nuclear Information System (INIS)

Hang, T.

1995-01-01

The Savannah River Site (SRS) plans to begin operating the Consolidated Incineration Facility (CIF) in 1996. The CIF will treat liquid and solid low-level radioactive, mixed and RCRA hazardous wastes generated at SRS. In addition to experimental test programs, process modeling was applied to provide guidance in areas of safety, environmental regulation compliances, process improvement and optimization. A steady-state flowsheet model was used to calculate material/energy balances and to track key chemical constituents throughout the process units. Dynamic models were developed to predict the CIF transient characteristics in normal and abnormal operation scenarios. Predictions include the rotary kiln heat transfer, dynamic responses of the CIF to fluctuations in the solid waste feed or upsets in the system equipments, performance of the control system, air inleakage in the kiln, etc. This paper reviews the modeling study performed to assist in the deflagration risk assessment

Next Generation Solvent Performance in the Modular Caustic Side Solvent Extraction Process - 15495

Energy Technology Data Exchange (ETDEWEB)

Smith, Tara E. [Savannah River Remediation, LLC., Aiken, SC (United States); Scherman, Carl [Savannah River Remediation, LLC., Aiken, SC (United States); Martin, David [Savannah River Remediation, LLC., Aiken, SC (United States); Suggs, Patricia [Savannah River Site (SRS), Aiken, SC (United States)

2015-01-14

Changes to the Modular Caustic Side Solvent Extraction Unit (MCU) flow-sheet were implemented in the facility. Implementation included changing the scrub and strip chemicals and concentrations, modifying the O/A ratios for the strip, scrub, and extraction contactor banks, and blending the current BoBCalixC6 extractant-based solvent in MCU with clean MaxCalix extractant-based solvent. During the successful demonstration period, the MCU process was subject to rigorous oversight to ensure hydraulic stability and chemical/radionuclide analysis of the key process tanks (caustic wash tank, solvent hold tank, strip effluent hold tank, and decontaminated salt solution hold tank) to evaluate solvent carryover to downstream facilities and the effectiveness of cesium removal from the liquid salt waste. Results indicated the extraction of cesium was significantly more effective with an average Decontamination Factor (DF) of 1,129 (range was 107 to 1,824) and that stripping was effective. The contactor hydraulic performance was stable and satisfactory, as indicated by contactor vibration, contactor rotational speed, and flow stability; all of which remained at or near target values. Furthermore, the Solvent Hold Tank (SHT) level and specific gravity was as expected, indicating that solvent integrity and organic hydraulic stability were maintained. The coalescer performances were in the range of processing results under the BOBCalixC6 flow sheet, indicating negligible adverse impact of NGS deployment. After the Demonstration period, MCU began processing via routine operations. Results to date reiterate the enhanced cesium extraction and stripping capability of the Next Generation Solvent (NGS) flow sheet. This paper presents process performance results of the NGS Demonstration and continued operations of MCU utilizing the blended BobCalixC6-MaxCalix solvent under the NGS flowsheet.

A computer-aided software-tool for sustainable process synthesis-intensification

DEFF Research Database (Denmark)

Kumar Tula, Anjan; Babi, Deenesh K.; Bottlaender, Jack

2017-01-01

and determine within the design space, the more sustainable processes. In this paper, an integrated computer-aided software-tool that searches the design space for hybrid/intensified more sustainable process options is presented. Embedded within the software architecture are process synthesis...... operations as well as reported hybrid/intensified unit operations is large and can be difficult to manually navigate in order to determine the best process flowsheet for the production of a desired chemical product. Therefore, it is beneficial to utilize computer-aided methods and tools to enumerate, analyze...... constraints while also matching the design targets, they are therefore more sustainable than the base case. The application of the software-tool to the production of biodiesel is presented, highlighting the main features of the computer-aided, multi-stage, multi-scale methods that are able to determine more...

Assessment of environmental aspects of uranium mining and milling. Final report, 12 February--7 July 1976

International Nuclear Information System (INIS)

Reed, A.K.; Meeks, H.C.; Pomeroy, S.E.; Hale, V.Q.

1976-12-01

This research program was initiated with the basic objective of making a preliminary assessment of the potential environmental impacts associated with the mining and milling of domestic uranium ores. All forms of pollution except radiation were considered. The program included a review of the characteristics and locations of domestic uranium ore reserves and a review of the conventional methods for mining and milling these ores. Potential environmental impacts associated with the entire cycle from exploration and mining to recovery and production of yellowcake are identified and discussed. Land reclamation aspects are also discussed. The methods currently used for production of yellowcake were divided into four categories - open pit mining-acid leach process, underground mining-acid leach process, underground mining-alkaline leach process, and in-situ mining. These are discussed from the standpoint of typical active mills which were visited during the program. Flowsheets showing specific environmental impacts for each category are provided

SustainPro - A tool for systematic process analysis, generation and evaluation of sustainable design alternatives

DEFF Research Database (Denmark)

Carvalho, Ana; Matos, Henrique A.; Gani, Rafiqul

2013-01-01

the user through the necessary steps according to work-flow of the implemented methodology. At the end the design alternatives, are evaluated using environmental impact assessment tools and safety indices. The extended features of the methodology incorporate Life Cycle Assessment analysis and economic....... The software tool is based on the implementation of an extended systematic methodology for sustainable process design (Carvalho et al. 2008 and Carvalho et al. 2009). Using process information/data such as the process flowsheet, the associated mass / energy balance data and the cost data, SustainPro guides...... analysis. The application and the main features of SustainPro are illustrated through a case study of ß-Galactosidase production....

An Improved Plutonium Trifluoride Precipitation Flowsheet

Energy Technology Data Exchange (ETDEWEB)

Harmon, H.D.

2001-06-26

This report discusses results of the plutonium trifluoride two-stage precipitation study. A series of precipitation experiments was used to identify the significant process variables affecting precipitation performance. A mathematical model of the precipitation process was developed which is based on the formation of plutonium fluoride complexes. The precipitation model relates all process variables, in a single equation, to a single parameter which can be used to control the performance of the plutonium trifluoride precipitation process. Recommendations have been made which will optimize the FB-Line plutonium trifluoride precipitation process.

An Improved Plutonium Trifluoride Precipitation Flowsheet

International Nuclear Information System (INIS)

Harmon, H.D.

2001-01-01

This report discusses results of the plutonium trifluoride two-stage precipitation study. A series of precipitation experiments was used to identify the significant process variables affecting precipitation performance. A mathematical model of the precipitation process was developed which is based on the formation of plutonium fluoride complexes. The precipitation model relates all process variables, in a single equation, to a single parameter which can be used to control the performance of the plutonium trifluoride precipitation process. Recommendations have been made which will optimize the FB-Line plutonium trifluoride precipitation process

Scaled Vitrification System III (SVS III) Process Development and Laboratory Tests at the West Valley Demonstration Project

International Nuclear Information System (INIS)

Jain, V.; Barnes, S.M.; Bindi, B.G.; Palmer, R.A.

2000-01-01

At the West Valley Demonstration Project (WVDP),the Vitrification Facility (VF)is designed to convert the high-level radioactive waste (HLW)stored on the site to a stable glass for disposal at a Department of Energy (DOE)-specified federal repository. The Scaled Vitrification System III (SVS-III)verification tests were conducted between February 1995 and August 1995 as a supplemental means to support the vitrification process flowsheet, but at only one seventh the scale.During these tests,the process flowsheet was refined and optimized. The SVS-III test series was conducted with a focus on confirming the applicability of the Redox Forecasting Model, which was based on the Index of Feed Oxidation (IFO)developed during the Functional and Checkout Testing of Systems (FACTS)and SVS-I tests. Additional goals were to investigate the prototypical feed preparation cycle and test the new target glass composition. Included in this report are the basis and current designs of the major components of the Scale Vitrification System and the results of the SVS-III tests.The major subsystems described are the feed preparation and delivery, melter, and off-gas treatment systems. In addition,the correlation between the melter's operation and its various parameters;which included feed rate,cold cap coverage,oxygen reduction (redox)state of the glass,melter power,plenum temperature,and airlift analysis;were developed

Americium-curium vitrification process development

International Nuclear Information System (INIS)

Fellinger, A.P.; Baich, M.A.; Hardy, B.J

1999-01-01

The successful demonstration of sequentially drying, calcining and vitrifying an oxalate slurry in the Drain Tube Test Stand (DTTS) vessel provided the process basis for testing on a larger scale in a cylindrical induction heated melter. A single processing issue, that of batch volume expansion, was encountered during the initial stage of testing. The increase in batch volume centered on a sintered frit cap and high temperature bubble formation. The formation of a sintered frit cap expansion was eliminated with the use of cullet. Volume expansions due to high temperature bubble formation (oxygen liberation from cerium reduction) were mitigated in the DTTS melter vessel through a vessel temperature profile that effectively separated the softening point of the glass cullet and the evolving oxygen from cerium reduction. An increased processing temperature of 1,470 C and a two hour hold time to find any remaining bubbles successfully reduced bubbles in the poured glass to an acceptable level. The success of the preliminary process demonstrations provided a workable process basis that was directly applicable to the newly installed Cylindrical Induction Melter (CIM) system, making the batch flowsheet the preferred option for vitrification of the americium-curium surrogate feed stream

ACHEMA '85: Process control systems

International Nuclear Information System (INIS)

Rosskopf, E.

1985-01-01

The strategy obviously adopted by the well-established manufacturers is to offer 'easy-to-handle' equipment to gain new customers, and there is a variety of new compact systems or personal computers being put on the market. The changes and improvements within the processing sector proceed more or less in silence; high-capacity storage devices and multiprocessor configurations are obtainable at a moderate price, offering a greater variety of basic functions and enhanced control possibilities. Redundancy problems are handled with greater flexibility, and batch programs are advancing. Data communication has become a common feature, transmission speed and bus length have been improved. Important improvements have been made with regard to data display; even medium-sized equipment now offer the possibility of making dynamic flow-sheets and reserving space for process history display, and the hierarchy of displays has been considerably simplified. The user software also has been made more easy, 'fill-in-the-blancs' is the prevailing motto for dialog configurations, and such big terms as process computer' or 'programming skill' are passing into oblivion. (orig./HP) [de

Main results obtained in France in the development of the gaseous diffusion process for uranium isotope separation

International Nuclear Information System (INIS)

Frejacques, C.; Bilous, O.; Dixmier, J.; Massignon, D.; Plurien, P.

1958-01-01

The main problems which occur in the study of uranium isotope separation by the gaseous diffusion process, concern the development of the porous barrier, the corrosive nature of uranium hexafluoride and also the chemical engineering problems related to process design and the choice of best plant and stage characteristics. Porous barriers may be obtained by chemical attack of non porous media or by agglomeration of very fine powders. Examples of these two types of barriers are given. A whole set of measurement techniques were developed for barrier structure studies, to provide control and guidance of barrier production methods. Uranium hexafluoride reactivity and corrosive properties are the source of many difficult technological problems. A high degree of plant leak tightness must be achieved. This necessity creates a special problem in compressor bearing design. Barrier lifetime is affected by the corrosive properties of the gas, which may lead to a change of barrier structure with time. Barrier hexafluoride permeability measurements have helped to make a systematic study of this point. Finally an example of a plant flowsheet, showing stage types and arrangements and based on a minimisation of enriched product costs is also given as an illustration of some of the chemical engineering problems present. (author) [fr

Caustic-Side Solvent Extraction Chemical and Physical Properties Progress in FY 2000 and FY 2001.

Energy Technology Data Exchange (ETDEWEB)

Moyer, BA

2002-04-17

The purpose of this work was to provide chemical- and physical-property data addressing the technical risks of the Caustic-Side Solvent Extraction (CSSX) process as applied specifically to the removal of cesium from alkaline high-level salt waste stored at the US Department of Energy Savannah River Site. As part of the overall Salt Processing Project, this effort supported decision-making in regards to selecting a preferred technology among three alternatives: (1) CSSX, (2) nonelutable ion-exchange with an inorganic silicotitanate material and (3) precipitation with tetraphenylborate. High risks, innate to CSSX, that needed specific attention included: (1) chemical stability of the solvent matrix, (2) radiolytic stability of the solvent matrix, (3) proof-of-concept performance of the proposed process flowsheet with simulated waste, and (4) performance of the CSSX flowsheet with actual SRS high-level waste. This body of work directly addressed the chemical-stability risk and additionally provided supporting information that served to plan, carry out, and evaluate experiments conducted by other CSSX investigators addressing the other high risks. Information on cesium distribution in extraction, scrubbing, and stripping served as input for flowsheet design, provided a baseline for evaluating solvent performance under numerous stresses, and contributed to a broad understanding of the effects of expected process variables. In parallel, other measurements were directed toward learning how other system components distribute in the flowsheet. Such components include the solvent components themselves, constituents of the waste, and solvent-degradation products. Upon understanding which components influence flowsheet performance, it was then possible to address in a rational fashion how to clean up the solvent and maintain its stable function.

Hadronic final states and sum rules in deep inelastic processes

International Nuclear Information System (INIS)

Pal, B.K.

1977-01-01

In order to get maximum information on the hadronic final states and sum rules in deep inelastic processes, Regge phenomenology and quarks parton model have been used. The unified picture for the production of hadrons of type i as a function of Bjorken and Feyman variables with only one adjustable parameter is formulated. The results of neutrino experiments and the production of charm particles are discussed in sum rules. (author)

Alternative calcination development status report

International Nuclear Information System (INIS)

Boardman, R.D.

1997-12-01

The Programmatic Spent Nuclear Fuel and (INEEL) Environmental Restoration and Waste Management Programs Environmental Impact Statement Record of Decision, dated June 1, 1995, specifies that high-level waste stored in the underground tanks at the ICPP continue to be calcined while other options to treat the waste are studied. Therefore, the High-Level Waste Program has funded a program to develop new flowsheets to increase the liquid waste processing rate. Simultaneously, a radionuclide separation process, as well as other options, are also being developed, which will be compared to the calcination treatment option. Two alternatives emerged as viable candidates; (1) elevated temperature calcination (also referred to as high temperature calcination), and (2) sugar-additive calcination. Both alternatives were determined to be viable through testing performed in a lab-scale calcination mockup. Subsequently, 10-cm Calciner Pilot Plant scoping tests were successfully completed for both flowsheets. The results were compared to the standard 500 C, high-ANN flow sheet (baseline flowsheet). The product and effluent streams were characterized to help elucidate the process chemistry and to investigate potential environmental permitting issues. Several supplementary tests were conducted to gain a better understanding of fine-particles generation, calcine hydration, scrub foaming, feed makeup procedures, sugar/organic elimination, and safety-related issues. Many of the experiments are only considered to be scoping tests, and follow-up experiments will be required to establish a more definitive understanding of the flowsheets. However, the combined results support the general conclusion that flowsheet improvements for the NWCF are technically viable

Liquid low-level waste (LLLW) solidification at Oak Ridge National Laboratory

International Nuclear Information System (INIS)

Schultz, R.M.; Monk, T.H.; duMont, S.P.; Helms, R.E.; Keigan, M.V.; Morris, M.I.

1987-01-01

In general, the presentation describes the disposal of liquid, low-level (radioactive) waste (LLLW) by the hydrofracture process at Oak Ridge National Laboratory until 1984, when it was shut down due to regulatory concerns and operational anomalies. As a result of this, about 400,000 gallons of concentrated LLLW and 50,000 gallons of transuranic waste-bearing sludges have accumulated in the active, double-contained tank system which is reaching its operational capacity. A major initiative to develop an alternative means of LLLW treatment and disposal was begun about two years ago. This presentation summarizes the implementation strategy of the most likely process options. The strategy is being developed in two phases; a near-term flowsheet and a long-term or reference flowsheet. First, reliable and fully demonstrated commercial, cement solidification systems are being assessed for execution of an initial 50,000 gallon campaign in 1988. Second, development is under way to determine viable sludge separation, LLLW decontamination and solidification alternatives. A flowsheet analysis and cost study is being conducted by a consultant to ensure proper consideration of process developments at other sites. It is estimated that, depending upon funding requirements, it could take up to six years to implement the reference flowsheet

Transuranium processing plant

International Nuclear Information System (INIS)

King, L.J.

1983-01-01

The Transuranium Processing Plant (TRU) is a remotely operated, hot-cell, chemical processing facility of advanced design. The heart of TRU is a battery of nine heavily shielded process cells housed in a two-story building. Each cell, with its 54-inch-thick walls of a special high-density concrete, has enough shielding to stop the neutrons and gamma radiation from 1 gram of 252/sub Cf/ and associated fission products. Four cells contain chemical processing equipment, three contain equipment for the preparation and inspection of HFIR targets, and two cells are used for analytical chemistry operations. In addition, there are eight laboratories used for process development, for part of the process-control analyses, and for product finishing operations. Although the Transuranium Processing Plant was built for the purpose of recovering transuranium elements from targets irradiated in the High Flux Isotope Reactor (HFIR), it is also a highly versatile facility which has extensive provisions for changing and modifying equipment. Thus, it was a relatively simple matter to install a Solvent Extraction Test Facility (SETF) in one of the TRU chemical processing cells for use in the evaluation and demonstration of solvent extraction flowsheets for the recovery of fissile and fertile materials from irradiated reactor fuels. The equipment in the SETF has been designed for process development and demonstrations and the particular type of mixer-settler contactors was chosen because it is easy to observe and sample

Solvent extraction process development for high plutonium fuel cycles

Energy Technology Data Exchange (ETDEWEB)

Anil Kumar, R; Selvaraj, P G; Natarajan, R; Raman, V R [Reprocessing Group, Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

1994-06-01

The purification of high plutonium bearing irradiated fuels using 30% TBP in dodecane diluent requires precise determination of concentration profiles during steady state, transient and process upset conditions. Mathematical models have been developed and a computer code is in use for determining Pu-U concentration profiles in a solvent extraction equipment in a typical reprocessing plant. The process parameters have been optimised for recovery of U and Pu and decontamination from the fission products. This computer code is used to analyse the extraction flow sheets of fuels of two typical Pu-U compositions encountered in Indian fast breeder programme. The analysis include the effect of uncertainty in equilibrium condition prediction by the model and the variation of flows of streams during plant operation. The studies highlight the margin available to avoid second organic phase formation and adjustments required in the process flowsheet. (author). 7 refs., 7 figs., 2 tabs.

SCIX IMPACT ON DWPF CPC

Energy Technology Data Exchange (ETDEWEB)

Koopman, D.

2011-07-14

A program was conducted to systematically evaluate potential impacts of the proposed Small Column Ion Exchange (SCIX) process on the Defense Waste Processing Facility (DWPF) Chemical Processing Cell (CPC). The program involved a series of interrelated tasks. Past studies of the impact of crystalline silicotitanate (CST) and monosodium titanate (MST) on DWPF were reviewed. Paper studies and material balance calculations were used to establish reasonable bounding levels of CST and MST in sludge. Following the paper studies, Sludge Batch 10 (SB10) simulant was modified to have both bounding and intermediate levels of MST and ground CST. The SCIX flow sheet includes grinding of the CST which is larger than DWPF frit when not ground. Nominal ground CST was not yet available, therefore a similar CST ground previously in Savannah River National Laboratory (SRNL) was used. It was believed that this CST was over ground and that it would bound the impact of nominal CST on sludge slurry properties. Lab-scale simulations of the DWPF CPC were conducted using SB10 simulants with no, intermediate, and bounding levels of CST and MST. Tests included both the Sludge Receipt and Adjustment Tank (SRAT) and Slurry Mix Evaporator (SME) cycles. Simulations were performed at high and low acid stoichiometry. A demonstration of the extended CPC flowsheet was made that included streams from the site interim salt processing operations. A simulation using irradiated CST and MST was also completed. An extensive set of rheological measurements was made to search for potential adverse consequences of CST and MST and slurry rheology in the CPC. The SCIX CPC impact program was conducted in parallel with a program to evaluate the impact of SCIX on the final DWPF glass waste form and on the DWPF melter throughput. The studies must be considered together when evaluating the full impact of SCIX on DWPF. Due to the fact that the alternant flowsheet for DWPF has not been selected, this study did not

The TRUEX [TRansUranium EXtraction] process and the management of liquid TRU [transuranic] waste

International Nuclear Information System (INIS)

Schulz, W.W.; Horwitz, E.P.

1987-01-01

The TRUEX process is a new generic liquid-liquid extraction process for removal of all actinides from acidic nitrate or chloride nuclear waste solutions. Because of its high efficiency and great flexibility, the TRUEX process appears destined to be widely used in the US and possibly in other countries for cost-effective management and disposal of transuranic (TRU) wastes. In the US, TRU wastes are those that contain ≥3.7 x 10 6 Bq/kg) of TRU elements with half-lives greater than 20 y. This paper gives a brief review of the relevant chemistry and summarizes the current status of development and deployment of the TRUEX (TRansUranium EXtraction) process flowsheets to treat specific acidic waste solutions at several US Department of Energy sites. 19 refs., 4 figs., 4 tabs

Characterize and Model Final Waste Formulations and Offgas Solids from Thermal Treatment Processes - FY-98 Final Report for LDRD 2349

Energy Technology Data Exchange (ETDEWEB)

Kessinger, Glen Frank; Nelson, Lee Orville; Grandy, Jon Drue; Zuck, Larry Douglas; Kong, Peter Chuen Sun; Anderson, Gail

1999-08-01

The purpose of LDRD #2349, Characterize and Model Final Waste Formulations and Offgas Solids from Thermal Treatment Processes, was to develop a set of tools that would allow the user to, based on the chemical composition of a waste stream to be immobilized, predict the durability (leach behavior) of the final waste form and the phase assemblages present in the final waste form. The objectives of the project were: • investigation, testing and selection of thermochemical code • development of auxiliary thermochemical database • synthesis of materials for leach testing • collection of leach data • using leach data for leach model development • thermochemical modeling The progress toward completion of these objectives and a discussion of work that needs to be completed to arrive at a logical finishing point for this project will be presented.

Calculation code MIXSET for Purex process

International Nuclear Information System (INIS)

Gonda, Kozo; Fukuda, Shoji.

1977-09-01

MIXSET is a FORTRAN IV calculation code for Purex process that simulate the dynamic behavior of solvent extraction processes in mixer-settlers. Two options permit terminating dynamic phase by time or by achieving steady state. These options also permit continuing calculation successively using new inputs from a arbitrary phase. A third option permits artificial rapid close to steady state and a fourth option permits searching optimum input to satisfy both of specification and recovery rate of product. MIXSET handles maximum chemical system of eight components with or without mutual dependence of the distribution of the components. The chemical system in MIXSET includes chemical reactions and/or decaying reaction. Distribution data can be supplied by third-power polynominal equations or tables, and kinetic data by tables or given constants. The fluctuation of the interfacial level height in settler is converted into the flow rate changes of organic and aqueous stream to follow dynamic behavior of extraction process in detail. MIXSET can be applied to flowsheet study, start up and/or shut down procedure study and real time process management in countercurrent solvent extraction processes. (auth.)

Dynamics of process at the final stage of nuclear fission

International Nuclear Information System (INIS)

Koljari, I.G.; Mavlitov, N.D.

2005-01-01

Numerous experimental data show, that the final stage of nuclear fission near to a scission point plays an essential role at formation of characteristics of fission products. At the description of a final stage of fission there is a number of problems: Definition of the form of the nuclear near the scission point and definition forms of a fission fragments; The account of dynamic processes in compound nuclear directly before of fission. The condition of the quasistatic al adiabatic process - dS/dt=0 - is applied in a point of transition from the uniform compound nuclei to several forms for definition of generalized coordinates and speeds. Calculation of dependence of post neutrons from nuclear mass of fission fragments for reactions is α+ 83 Bi 209 → 85 At 213 (E lab = 45 MeV); α+ 92 U 242 → 94 Pu 242 (E lab = 45 MeV); 8 O 18 + 79 Au 197 → 97 Fr 215 (E lab = 159 MeV). System of equations, which describes behaviour of system in a point of nuclear fission-transition from the uniform form to system of a two (and, probably more) fission fragments is given. The system of the equations allows in a fission point to define the generalized coordinates, and the generalized speeds for each of the generalized coordinates of collective deformation variables

Study on designing a complete pilot plant for processing sandstone ores in Palua-Parong area

International Nuclear Information System (INIS)

Le Quang Thai; Tran Van Son; Tran The Dinh; Trinh Nguyen Quynh; Vu Khac Tuan

2015-01-01

Design work is the first step of the construction and operation of pilot plant. Thus, the project Study on designing a complete pilot plant for processing sandstone ores in Palua - Parong area was conducted to design a pilot plant for testing entire technological process to obtain yellowcake. Based on a literature review of uranium ore processing technology in the world, information of ore and previous research results of uranium ore in PaLua - PaRong area at the Institute for Technology of Radioactive and Rare Elements, a suitable technological flowsheet for processing this ore has been selected. The size, location of the pilot plant and planed experiments has been selected during the implementation of this project, in which basic parameters, designed system of equipment, buildings, ect. were also calculated. (author)

Enhanced separation of rare earth elements

Energy Technology Data Exchange (ETDEWEB)

Lyon, K. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Greenhalgh, M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Herbst, R. S. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Garn, T. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Welty, A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Soderstrom, M. D. [Cytec Solvay Group, Tempe, AZ (United States); Jakovljevic, B. [Cytec Solvay Group, Niagara Falls, ON (Canada)

2016-09-01

Industrial rare earth separation processes utilize PC88A, a phosphonic acid ligand, for solvent extraction separations. The separation factors of the individual rare earths, the equipment requirements, and chemical usage for these flowsheets are well characterized. Alternative ligands such as Cyanex® 572 and the associated flowsheets are being investigated at the pilot scale level to determine if significant improvements to the current separation processes can be realized. These improvements are identified as higher separation factors, reduced stage requirements, or reduced chemical consumption. Any of these improvements can significantly affect the costs associated with these challenging separation proccesses. A mid/heavy rare earth element (REE) separations flowsheet was developed and tested for each ligand in a 30 stage mixer-settler circuit to compare the separation performance of PC88A and Cyanex® 572. The ligand-metal complex strength of Cyanex® 572 provides efficient extraction of REE while significantly reducing the strip acid requirements. Reductions in chemical consumption have a significant impact on process economics for REE separations. Partitioning results summarized Table 1 indicate that Cyanex® 572 offers the same separation performance as PC88A while reducing acid consumption by 30% in the strip section for the mid/heavy REE separation. Flowsheet Effluent Compositions PC88A Cyanex® 572 Raffinate Mid REE Heavy REE 99.40% 0.60% 99.40% 0.60% Rich Mid REE Heavy REE 2.20% 97.80% 0.80% 99.20% Liquor Strip Acid Required 3.4 M 2.3 M Table 1 – Flowsheet results comparing separation performance of PC88A and Cyanex® 572 for a mid/heavy REE separation.

Study of the processes of ion pairs formation by the method of ion-ion coincidence: I2 and chlorine-containing hydrocarbons

International Nuclear Information System (INIS)

Golovin, A.V.

1991-01-01

A method of ion-ion coincidences was suggested to study the process of ion pairs formation during molecule photoionization. The principle of action of ion-ion coincidence method is based on recording of only the negative and positive ions that formed as a result of a molecule decomposition. The flowsheet of the facility of ion-ion coincidences was presented. The processes of ion pairs formation in iodine, chloroform, propyl-, n-propenyl-, tert.butyl- and benzyl-chlorides were studied. A simple model permitting to evaluate the dependence of quantum yield of ion pair formation on excitation energy was suggested

Treatability studies in support of the nonradiological wastewater treatment project

Energy Technology Data Exchange (ETDEWEB)

Begovich, J.M.; Brown, C.H. Jr.; Villiers-Fisher, J.F.; Fowler, V.L.

1986-07-01

The Nonradiological Wastewater Treatment Project (NRWTP) will treat nonradiological wastewaters generated at the Oak Ridge National Laboratory (ORNL) to pollutant levels acceptable under restrictions imposed by the effluent limits of best available technology (BAT) regulations of the US Environmental Protection Agency (EPA), according to the goals established by the Clean Water Act. A three-phase treatability study was conducted to resolve many of the uncertainties facing the NRWTP. The first phase consisted of batch simulation of the proposed NRWTP flowsheet in the laboratory. The Phase I results revealed no major problems with the proposed flowsheet. Phase II consisted of more-detailed parametric studies of the flowsheet processes at a bench-scale level in the laboratory. The Phase II results were used to guide the planning and design of the Phase III study, which consisted of flowsheet simulation on a continuous basis using a mini-pilot plant (MPP) facility. This facility is contained within two connected semitrailer vans and an analytical trailer.

Phenomena Based Process Intensification of Toluene Methylation for Sustainable Para-xylene Production

DEFF Research Database (Denmark)

Anantasarn, Nateetorn; Babi, Deenesh Kavi; Suriyapraphadilok, Uthaiporn

2016-01-01

The objective of this work is to generate more sustainable intensified process designs for the production of important chemicals in the petrochemical sector. A 3-stage approach is applied. In stage 1, the base case design is generated or selected from literature. In stage 2, the base case design...... is analysed in terms of economics, sustainability and LCA factors in order to identify process hot-spots that are translated into design targets. In stage 3, intensified flowsheet alternatives are generated that match the targets and thereby eliminate and/or minimize the process hot-spots using a phenomena...... operations to generate more sustainable designs. An overview of the key concepts and framework are presented together with the results from a case study highlighting the application of the framework to the sustainable design of a production process for para-xylene, which is an important chemical utilized...

Process design and economic evaluation of green extraction methods for recovery of astaxanthin from shrimp waste

DEFF Research Database (Denmark)

Razi Parjikolaei, Behnaz; Errico, Massimiliano; El-Houri, Rime Bahij

2017-01-01

(ASX) from shrimp processing waste. The feasibility of commercial use of the green solvents under plausible process conditions is compared to extraction with a mixture of hexane: isopropanol (Hex:IPA). The process flowsheets describing these processes were modelled by means of SuperPro Designer...... processes with SF or the methyl ester of SF (ME-SF) was 2.5 and 153 ppm with a production cost of 0.06 and 0.16 $/mg of ASX, respectively. In addition, shrimp feed production was considered as a feasible application of the low concentration ASX obtained by SF extraction. A combination of ASX extracted...... with SF and synthetic ASX yielded a shrimp feed production cost comparable to the current market price. The calculated feed price based on the ASX production cost of the other green processes, ME-SF and SCFE, resulted in a significantly higher production cost....

Monazite upgradation and production of high pure rare earths

International Nuclear Information System (INIS)

Asnani, C.K.; Mohanty, D.; Kumar, S.S.

2014-01-01

Rare earth extraction from monazite and further processing of mixed rare earth chlorides for producing individual high pure rare earths involves a complex flowsheet based on solvent extraction process. Apart from involving multiple extractions, scrubbing and stripping operations, the flowsheet requires optimization of critical parameters such as solvent molarity, solvent saponification level and recycling of product solutions as reflux to ensure preferential upload of required rare earths to generate high purity product. This paper tracks monazite flow from the raw sand feed through to the monazite product and its processing to generate rare earths of internationally acceptable quality

DWPF SB6 Initial CPC Flowsheet Testing SB6-1 TO SB6-4L Tests Of SB6-A And SB6-B Simulants

International Nuclear Information System (INIS)

Lambert, D.; Pickenheim, B.; Best, D.

2009-01-01

The Defense Waste Processing Facility (DWPF) will transition from Sludge Batch 5 (SB5) processing to Sludge Batch 6 (SB6) processing in late fiscal year 2010. Tests were conducted using non-radioactive simulants of the expected SB6 composition to determine the impact of varying the acid stoichiometry during the Sludge Receipt and Adjustment Tank (SRAT) and Slurry Mix Evaporator (SME) processes. The work was conducted to meet the Technical Task Request (TTR) HLW/DWPF/TTR-2008-0043, Rev.0 and followed the guidelines of a Task Technical and Quality Assurance Plan (TT and QAP). The flowsheet studies are performed to evaluate the potential chemical processing issues, hydrogen generation rates, and process slurry rheological properties as a function of acid stoichiometry. These studies were conducted with the estimated SB6 composition at the time of the study. This composition assumed a blend of 101,085 kg of Tank 4 insoluble solids and 179,000 kg of Tank 12 insoluble solids. The current plans are to subject Tank 12 sludge to aluminum dissolution. Liquid Waste Operations assumed that 75% of the aluminum would be dissolved during this process. After dissolution and blending of Tank 4 sludge slurry, plans included washing the contents of Tank 51 to ∼1M Na. After the completion of washing, the plan assumes that 40 inches on Tank 40 slurry would remain for blending with the qualified SB6 material. There are several parameters that are noteworthy concerning SB6 sludge: (1) This is the second batch DWPF will be processing that contains sludge that has had a significant fraction of aluminum removed through aluminum dissolution; (2) The sludge is high in mercury, but the projected concentration is lower than SB5; (3) The sludge is high in noble metals, but the projected concentrations are lower than SB5; and(4) The sludge is high in U and Pu - components that are not added in sludge simulants. Six DWPF process simulations were completed in 4-L laboratory-scale equipment using

On-Line Monitoring for Process Control and Safeguarding of Radiochemical Streams at Spent Fuel Reprocessing Plants

International Nuclear Information System (INIS)

Bryan, S.; Levitskaia, T.; Casella, A.

2015-01-01

The International Atomic Energy Agency (IAEA) has established international safe- guards standards for fissionable material at spent nuclear fuel reprocessing plants to ensure that significant quantities of weapons-grade nuclear material are not diverted from these facilities. Currently, methods to verify material control and accountancy (MC&A) at these facilities require time-consuming and resource-intensive destructive assay (DA). Leveraging new on-line non-destructive assay (NDA) techniques in conjunction with the traditional and highly precise DA methods may provide a more timely, cost-effective and resource-efficient means for MC&A verification at such facilities. Pacific Northwest National Laboratory (PNNL) is developing on-line NDA process monitoring technologies, including a spectroscopy-based monitoring system, to potentially reduce the time and re- source burden associated with current techniques. The spectroscopic monitor continuously measures chemical compositions of the process streams including actinide metal ions (U, Pu, Np), selected fission products, and major cold flowsheet chemicals using ultra-violet and visible, near infrared and Raman spectroscopy. This paper will provide an overview of the methods and report our on-going efforts to develop and demonstrate the technologies. Our ability to identify material intentionally diverted from a liquid-liquid solvent extraction contactor system was successfully tested using on-line process monitoring as a means to detect the amount of material diverted. A chemical diversion, and detection of that diversion, from a solvent extraction scheme was demonstrated using a centrifugal contactor system operating with the PUREX flowsheet. A portion of the feed from a counter-current extraction system was diverted while a continuous extraction experiment was underway. The amount observed to be diverted by on-line spectroscopic process monitoring was in excellent agreement with values based from the known mass of

Development and demonstration of a new SANEX Partitioning Process for selective actinide(III)/lanthanide(III) separation using a mixture of CyMe{sub 4}BTBP and TODGA

Energy Technology Data Exchange (ETDEWEB)

Modolo, G.; Wilden, A.; Daniels, H. [Forschungszentrum Juelich GmbH (Germany). Institute for Energy and Climate Research, IEK-6, Nuclear Waste Management and Reactor Safety; Geist, A.; Magnusson, D. [Karlsruher Institut fuer Technologie, Karlsruhe (Germany). Inst. fuer Nukleare Entsorgung; Malmbeck, R. [European Commission, JRC, Karlsruhe (Germany). Inst. for Transuranium Elements (ITU)

2013-05-01

Within the framework of the European collaborative project ACSEPT, a new SANEX partitioning process was developed at Forschungszentrum Juelich for the separation of the trivalent minor actinides americium, curium and californium from lanthanide fission products in spent nuclear fuels. The development is based on batch solvent extraction studies, single-centrifugal contactor tests and on flow-sheet design by computer code calculations. The used solvent is composed of 6,6{sup '}-bis(5,5,8,8-tetramethyl-5,6,7,8-tetrahydrobenzo-[1,2-4]trizazin-3-yl)-[2,2{sup '}]-bipyridine (CyMe{sub 4}BTBP) and N,N,N{sup '},N{sup '}-tetraoctyldiglycolamide (TODGA) dissolved in n-octanol. A spiked continuous counter-current test was carried out in miniature centrifugal contactors with the aid of a 20-stage flow-sheet consisting of 12 extraction, 4 scrubbing and 4 stripping stages. A product fraction containing more than 99.9% of the trivalent actinides Am(III), Cm(III) and Cf(III) was obtained. High product/feed decontamination factors >1000 were achieved for these actinides. The trivalent lanthanides were directed to the raffinate of the process with the actinide (III) product stream being contaminated with less than 0.5 mass-% in the initial lanthanides. (orig.)

Determination of the impact of glycolate on ARP and MCU operations

Energy Technology Data Exchange (ETDEWEB)

Taylor-Pashow, K. M.L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Peters, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Fondeur, F. F. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Shehee, T. C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Washington, A. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2017-05-17

Savannah River Remediation (SRR) is evaluating an alternate flowsheet for the Defense Waste Processing Facility (DWPF) using glycolic acid as a reductant. An important aspect of the development of the glycolic acid flowsheet is determining if glycolate has any detrimental downstream impacts. Testing was performed to determine if there is any impact to the strontium and actinide sorption by monosodium titanate (MST) and modified monosodium titanate (mMST) or if there is an impact to the cesium removal, phase separation, or coalescer performance at the Modular Caustic-Side Solvent Extraction Processing Unit (MCU).

Hydrogen production processes

International Nuclear Information System (INIS)

2003-01-01

The goals of this first Gedepeon workshop on hydrogen production processes are: to stimulate the information exchange about research programs and research advances in the domain of hydrogen production processes, to indicate the domains of interest of these processes and the potentialities linked with the coupling of a nuclear reactor, to establish the actions of common interest for the CEA, the CNRS, and eventually EDF, that can be funded in the framework of the Gedepeon research group. This document gathers the slides of the 17 presentations given at this workshop and dealing with: the H 2 question and the international research programs (Lucchese P.); the CEA's research program (Lucchese P., Anzieu P.); processes based on the iodine/sulfur cycle: efficiency of a facility - flow-sheets, efficiencies, hard points (Borgard J.M.), R and D about the I/S cycle: Bunsen reaction (Colette S.), R and D about the I/S cycle: the HI/I 2 /H 2 O system (Doizi D.), demonstration loop/chemical engineering (Duhamet J.), materials and corrosion (Terlain A.); other processes under study: the Westinghouse cycle (Eysseric C.), other processes under study at the CEA (UT3, plasma,...) (Lemort F.), database about thermochemical cycles (Abanades S.), Zn/ZnO cycle (Broust F.), H 2 production by cracking, high temperature reforming with carbon trapping (Flamant G.), membrane technology (De Lamare J.); high-temperature electrolysis: SOFC used as electrolyzers (Grastien R.); generic aspects linked with hydrogen production: technical-economical evaluation of processes (Werkoff F.), thermodynamic tools (Neveu P.), the reactor-process coupling (Aujollet P.). (J.S.)

The extraction of uranium from wet process phosphoric acid using a liquid surfactant membrane system

International Nuclear Information System (INIS)

Dickens, N.; Davies, G.A.

1984-01-01

A liquid membrane extraction process is examined for the extraction of uranium from wet process phosphoric acid. Uranium is present in the acid in concentrations up to 100 ppm which in principle makes it ideal for treatment with a membrane process. The membrane system studied is based on extraction using DEHPA-TOPO reagents which are contained within the organic phase of a water in oil emulsion. Formulations of the emulsion membrane system have been studied, the limitations of acid temperature, P 2 O 5 concentration and solid dispersed impurities in the acid have been studied in laboratory batch experiments and in a continuous pilot plant unit capable of treating 5l of concentrated acid per minute. Data from the pilot plant work has been used to develop a flowsheet for a commercial unit based on this process. (author)

Plan for advanced microelectronics processing technology application. Final report

Energy Technology Data Exchange (ETDEWEB)

Goland, A.N.

1990-10-01

The ultimate objective of the tasks described in the research agreement was to identify resources primarily, but not exclusively, within New York State that are available for the development of a Center for Advanced Microelectronics Processing (CAMP). Identification of those resources would enable Brookhaven National Laboratory to prepare a program plan for the CAMP. In order to achieve the stated goal, the principal investigators undertook to meet the key personnel in relevant NYS industrial and academic organizations to discuss the potential for economic development that could accompany such a Center and to gauge the extent of participation that could be expected from each interested party. Integrated of these discussions was to be achieved through a workshop convened in the summer of 1990. The culmination of this workshop was to be a report (the final report) outlining a plan for implementing a Center in the state. As events unfolded, it became possible to identify the elements of a major center for x-ray lithography on Lone Island at Brookhaven National Laboratory. The principal investigators were than advised to substitute a working document based upon that concept in place of a report based upon the more general CAMP workshop originally envisioned. Following that suggestion from the New York State Science and Technology Foundation, the principals established a working group consisting of representatives of the Grumman Corporation, Columbia University, the State University of New York at Stony Brook, and Brookhaven National Laboratory. Regular meetings and additional communications between these collaborators have produced a preproposal that constitutes the main body of the final report required by the contract. Other components of this final report include the interim report and a brief description of the activities which followed the establishment of the X-ray Lithography Center working group.

Used nuclear fuel separations process simulation and testing

International Nuclear Information System (INIS)

Pereira, C.; Krebs, J.F.; Copple, J.M.; Frey, K.E.; Maggos, L.E.; Figueroa, J.; Willit, J.L.; Papadias, D.D.

2013-01-01

Recent efforts in separations process simulation at Argonne have expanded from the traditional focus on solvent extraction flowsheet design in order to capture process dynamics and to simulate other components, processing and systems of a used nuclear fuel reprocessing plant. For example, the Argonne Model for Universal Solvent Extraction (AMUSE) code has been enhanced to make it both more portable and more readily extensible. Moving away from a spreadsheet environment makes the addition of new species and processes simpler for the expert user, which should enable more rapid implementation of chemical models that simulate evolving processes. The dyAMUSE (dynamic AMUSE) version allows the simulation of transient behavior across an extractor. Electrochemical separations have now been modeled using spreadsheet codes that simulate the electrochemical recycle of fast reactor fuel. The user can follow the evolution of the salt, products, and waste compositions in the electro-refiner, cathode processors, and drawdown as a function of fuel batches treated. To further expand capabilities in integrating multiple unit operations, a platform for linking mathematical models representing the different operations that comprise a reprocessing facility was adapted to enable systems-level analysis and optimization of facility functions. (authors)

Used nuclear fuel separations process simulation and testing

Energy Technology Data Exchange (ETDEWEB)

Pereira, C.; Krebs, J.F.; Copple, J.M.; Frey, K.E.; Maggos, L.E.; Figueroa, J.; Willit, J.L.; Papadias, D.D. [Argonne National Laboratory: 9700 South Cass Avenue, Argonne, IL 60439 (United States)

2013-07-01

Recent efforts in separations process simulation at Argonne have expanded from the traditional focus on solvent extraction flowsheet design in order to capture process dynamics and to simulate other components, processing and systems of a used nuclear fuel reprocessing plant. For example, the Argonne Model for Universal Solvent Extraction (AMUSE) code has been enhanced to make it both more portable and more readily extensible. Moving away from a spreadsheet environment makes the addition of new species and processes simpler for the expert user, which should enable more rapid implementation of chemical models that simulate evolving processes. The dyAMUSE (dynamic AMUSE) version allows the simulation of transient behavior across an extractor. Electrochemical separations have now been modeled using spreadsheet codes that simulate the electrochemical recycle of fast reactor fuel. The user can follow the evolution of the salt, products, and waste compositions in the electro-refiner, cathode processors, and drawdown as a function of fuel batches treated. To further expand capabilities in integrating multiple unit operations, a platform for linking mathematical models representing the different operations that comprise a reprocessing facility was adapted to enable systems-level analysis and optimization of facility functions. (authors)

Chemical Engineering Division Fuel Cycle Programs. Quarterly progress report, April-June 1978. [Advanced solvent extraction; accidents; pyrochemical; radwaste in metal matrix; waste migration

Energy Technology Data Exchange (ETDEWEB)

Steindler, M. J.; Ader, M.; Barletta, R. E.

1979-12-01

Fuel cycle studies reported include development of centrifugal contactors for Purex processes. Tricaprylmethyl-ammonium nitrate and di-n-amyl-n-amylphosphonate are being evaluated as Thorex extractants. Dispersion of uranium and plutonium by fires, and mechanisms for subdividing and dispersing liquids and solids were reviewed. In the pyrochemical and dry processing program, a facility for testing containment materials is under construction; a flowsheet for carbide fuel processing has been designed and studies of carbide reactions in bismuth are underway; salt transport processes are being studied; process-size refractory metal vessels are being fabricated; the feasibility of AIROX reprocessing is being determined; the solubility of UO/sub 2/, UO/sub 2/ + fission products, and PuO/sub 2/ in molten alkali metal nitrates, has been investigated; a flowsheet was developed for reprocessing actinide oxides in molten salts; preparation of Th-U carbide from the oxide is being studied; new flowsheets based on the Dow Aluminum Pyrometallurgical process for reprocessing of spent uranium metal fuel have been prepared; the chloride volitility processing of thorium-based fuels is being studied; the reprocessing of (Th,U)O/sub 2/ solid solution in KCl-LiCl-ThCl/sub 4/-Th is being studied; and a flowsheet for processing spent nuclear fuel in molten tin has been constructed. Leach rates of simulated encapsulated waste forms in a metal matrix were studied. Nine criteria for handling waste cladding hulls were established. Strontium and tin migration in glauconite columns was measured. Radioactive Sr in a stream of water moved through oolitic limestone as rapidly as water, but in a stream of water equilibrated with the limestone, Sr moved through the limestone one-tenth as fast. Migration of trace quantities of Cs and I through kaolinite was studied. 88 figures, 53 tables.

Final Technical Report - Advanced Optical Sensors to Minimize Energy Consumption in Polymer Extrusion Processes

Energy Technology Data Exchange (ETDEWEB)

Susan J. Foulk

2012-07-24

Project Objective: The objectives of this study are to develop an accurate and stable on-line sensor system to monitor color and composition on-line in polymer melts, to develop a scheme for using the output to control extruders to eliminate the energy, material and operational costs of off-specification product, and to combine or eliminate some extrusion processes. Background: Polymer extrusion processes are difficult to control because the quality achieved in the final product is complexly affected by the properties of the extruder screw, speed of extrusion, temperature, polymer composition, strength and dispersion properties of additives, and feeder system properties. Extruder systems are engineered to be highly reproducible so that when the correct settings to produce a particular product are found, that product can be reliably produced time after time. However market conditions often require changes in the final product, different products or grades may be processed in the same equipment, and feed materials vary from lot to lot. All of these changes require empirical adjustment of extruder settings to produce a product meeting specifications. Optical sensor systems that can continuously monitor the composition and color of the extruded polymer could detect process upsets, drift, blending oscillations, and changes in dispersion of additives. Development of an effective control algorithm using the output of the monitor would enable rapid corrections for changes in materials and operating conditions, thereby eliminating most of the scrap and recycle of current processing. This information could be used to identify extruder systems issues, diagnose problem sources, and suggest corrective actions in real-time to help keep extruder system settings within the optimum control region. Using these advanced optical sensor systems would give extruder operators real-time feedback from their process. They could reduce the amount of off-spec product produced and

Economic assessment of advanced flue gas desulfurization processes. Final report

Energy Technology Data Exchange (ETDEWEB)

Bierman, G. R.; May, E. H.; Mirabelli, R. E.; Pow, C. N.; Scardino, C.; Wan, E. I.

1981-09-01

This report presents the results of a project sponsored by the Morgantown Energy Technology Center (METC). The purpose of the study was to perform an economic and market assessment of advanced flue gas desulfurization (FGD) processes for application to coal-fired electric utility plants. The time period considered in the study is 1981 through 1990, and costs are reported in 1980 dollars. The task was divided into the following four subtasks: (1) determine the factors affecting FGD cost evaluations; (2) select FGD processes to be cost-analyzed; (3) define the future electric utility FGD system market; and (4) perform cost analyses for the selected FGD processes. The study was initiated in September 1979, and separate reports were prepared for the first two subtasks. The results of the latter two subtasks appear only in this final reprot, since the end-date of those subtasks coincided with the end-date of the overall task. The Subtask 1 report, Criteria and Methods for Performing FGD Cost Evaluations, was completed in October 1980. A slightly modified and condensed version of that report appears as appendix B to this report. The Subtask 2 report, FGD Candidate Process Selection, was completed in January 1981, and the principal outputs of that subtask appear in Appendices C and D to this report.

Removal of actinides from high activity wastes by solvent extraction: outline of the research work at Ispra J.R.C. laboratories

International Nuclear Information System (INIS)

Mannone, F.

1976-07-01

The development of an advanced waste management alternative such as the actinide nuclear incineration requires an almost quantitative removal of actinides from waste streams. Within the framework of the Ispra JRC Waste Disposal R and D programme, actinide separation studies were directed towards solvent extraction and precipitation methods. To develop a tentative waste partitioning flow-sheet based on solvent extraction, two conceptual process flow-sheet for actinide removal were evaluated on the basis of the currently used actinide recovery processes, i.e. removal after waste adjustment to low-acidity conditions and direct actinide removal from acidic wastes, as they are generated in actual reprocessing plants. No improvements have been devised for actinide recoveries within the conventional Purex reprocessing operations and a currently agreed value has been assumed for neptunium recovery (90%). According to these basic orientations some organic extractants have been selected for testing as promising candidates for waste partitioning and laboratory studies, designed to develop a satisfactory partitioning flow-sheet, have been proposed and described

Computer simulation of transitional process to the final stable Brayton cycle in magnetic refrigeration

International Nuclear Information System (INIS)

Numasawa, T.; Hashimoto, T.

1981-01-01

The final working cycle in the magnetic refrigeration largely depends on the heat transfer coefficient β in the system, the parameter γ of the heat inflow from the outer system to this cycle and the period tau of the cycle. Therefore, so as to make clear this dependence, the time variation of the Brayton cycle with β, γ and tau has been investigated. In the present paper the transitional process of this cycle and the dependence of the final cooling temperature of the heat load on β, γ and tau have all been shown. (orig.)

Using solvent extraction to process nitrate anion exchange column effluents

International Nuclear Information System (INIS)

Yarbro, S.L.

1987-10-01

Octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO), a new organophosphorous extractant, and a new centrifugal mixer-settler both recently developed at Argonne were evaluated for their potential use in the recovery of actinides from nitrate anion exchange column effluents. The performance of the extractant was evaluated by measuring the extraction coefficient values as a function of acid and salt concentration. Additional performance parameters include extraction coefficient behavior as a function of the total metal concentration in the organic phase, and comparison of different stripping and organic scrubbing techniques. A simulated effluent stream was used to evaluate the performance of the centrifugal mixer-settlers by comparing experimental and calculated interstage concentration profiles. Both the CMPO extractant and the centrifugal mixer-settlers have potential for processing nitrate column effluents, particularly if the stripping behavior can be improved. Details of the proposed process are presented in the flowsheet and contactor design analyses

Using solvent extraction to process nitrate anion exchange column effluents

Energy Technology Data Exchange (ETDEWEB)

Yarbro, S.L.

1987-10-01

Octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO), a new organophosphorous extractant, and a new centrifugal mixer-settler both recently developed at Argonne were evaluated for their potential use in the recovery of actinides from nitrate anion exchange column effluents. The performance of the extractant was evaluated by measuring the extraction coefficient values as a function of acid and salt concentration. Additional performance parameters include extraction coefficient behavior as a function of the total metal concentration in the organic phase, and comparison of different stripping and organic scrubbing techniques. A simulated effluent stream was used to evaluate the performance of the centrifugal mixer-settlers by comparing experimental and calculated interstage concentration profiles. Both the CMPO extractant and the centrifugal mixer-settlers have potential for processing nitrate column effluents, particularly if the stripping behavior can be improved. Details of the proposed process are presented in the flowsheet and contactor design analyses.

Computational model for a high temperature electrolyzer coupled to a HTTR for efficient nuclear hydrogen production

International Nuclear Information System (INIS)

Gonzalez, Daniel; Rojas, Leorlen; Rosales, Jesus; Castro, Landy; Gamez, Abel; Brayner, Carlos; Garcia, Lazaro; Garcia, Carlos; Torre, Raciel de la; Sanchez, Danny

2015-01-01

High temperature electrolysis process coupled to a very high temperature reactor (VHTR) is one of the most promising methods for hydrogen production using a nuclear reactor as the primary heat source. However there are not references in the scientific publications of a test facility that allow to evaluate the efficiency of the process and other physical parameters that has to be taken into consideration for its accurate application in the hydrogen economy as a massive production method. For this lack of experimental facilities, mathematical models are one of the most used tools to study this process and theirs flowsheets, in which the electrolyzer is the most important component because of its complexity and importance in the process. A computational fluid dynamic (CFD) model for the evaluation and optimization of the electrolyzer of a high temperature electrolysis hydrogen production process flowsheet was developed using ANSYS FLUENT®. Electrolyzer's operational and design parameters will be optimized in order to obtain the maximum hydrogen production and the higher efficiency in the module. This optimized model of the electrolyzer will be incorporated to a chemical process simulation (CPS) code to study the overall high temperature flowsheet coupled to a high temperature accelerator driven system (ADS) that offers advantages in the transmutation of the spent fuel. (author)

Dynamic behaviour of solvent contactors in fuel reprocessing plants- an analysis

Energy Technology Data Exchange (ETDEWEB)

Raju, R P; Siddiqui, H R [Nuclear Waste Management Group, Bhabha Atomic Research Centre, Mumbai (India); Murthy, K K; Kansra, V P [Fuel Reprocessing Group, Bhabha Atomic Research Centre, Mumbai (India)

1994-06-01

Fuel reprocessing plants carry out separation of useful fissile and fertile materials from spent nuclear fuels by isolating highly radioactive fission products using solvent extraction method. In the fuel reprocessing step of nuclear fuel cycle, optimisation of process parameters in the PUREX flowsheet design is of great importance particularly on account of the need to realize high degree of recovery of fissile and fertile materials and to ensure proper control on concentrations of fissile element in process streams for avoidance of criticality. In counter-current solvent contactors of PUREX flowsheet there are a variety of processes conditions which may cause plutonium accumulations that requires attention to ascertain safe Pu concentrations within the contactors. A study was carried out using the PUREX process mathematical model Solvent Extraction Program Having Interacting Solutes (SEPHIS) for pulsed solvent contactors in PREFRE-1, Tarapur and PREFRE-2, Kalpakkam flowsheets for optimising the process parameters in plutonium purification cycles. The study was extended to predict the behaviour of contactors handling plutonium bearing solutions under certain anticipated deviations in the process parameters. Modifications wherever necessary were carried out to the original SEPHIS code. This paper discusses the results obtained during this analysis. (author). 2 figs., 2 tabs.

Industrial wastewater reuse in petroleum refinery using the WSD for regeneration systems

Directory of Open Access Journals (Sweden)

Lídia Yokoyama

2011-12-01

Full Text Available Wastewater reuse practices in the industry require an adequate understanding of the characteristics of the manufacture processes, to minimize the water consumption and the generation of effluent. The objective of this work was to apply the WSD method, used to defining the target of minimum process water consumption in a case study of oil refinery, by means of the reuse and recycling operations, including regeneration processes. The importance and influence of the wastewater treatment plant in the regeneration quality, including intermediate process streams, for the reuse and the recycling operations, were evaluated. Furthermore, centralized and distributed treatment flowsheet configurations were tested. Thus, this work presented the solution of a case study with three contaminants in water streams processes, different interconnections approaches, used to illustrate the application of this procedure showing the reduction of water flow rate and total costs compared to the original flowsheet. The scenarios revealed to be greatly promising, and flowsheet configurations were reached with higher than 4 % and 20 % of reduction in the water flow rate consumption and the total costs, respectively. Regarding the ecoefficiency processes, the results demonstrate that the applied technique is successful when the minimum water consumption is the main goal in the industry.

Computational model for a high temperature electrolyzer coupled to a HTTR for efficient nuclear hydrogen production

Energy Technology Data Exchange (ETDEWEB)

Gonzalez, Daniel; Rojas, Leorlen; Rosales, Jesus; Castro, Landy; Gamez, Abel; Brayner, Carlos, E-mail: danielgonro@gmail.com [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil); Garcia, Lazaro; Garcia, Carlos; Torre, Raciel de la, E-mail: lgarcia@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas (InSTEC), La Habana (Cuba); Sanchez, Danny [Universidade Estadual de Santa Cruz (UESC), Ilheus, BA (Brazil)

2015-07-01

High temperature electrolysis process coupled to a very high temperature reactor (VHTR) is one of the most promising methods for hydrogen production using a nuclear reactor as the primary heat source. However there are not references in the scientific publications of a test facility that allow to evaluate the efficiency of the process and other physical parameters that has to be taken into consideration for its accurate application in the hydrogen economy as a massive production method. For this lack of experimental facilities, mathematical models are one of the most used tools to study this process and theirs flowsheets, in which the electrolyzer is the most important component because of its complexity and importance in the process. A computational fluid dynamic (CFD) model for the evaluation and optimization of the electrolyzer of a high temperature electrolysis hydrogen production process flowsheet was developed using ANSYS FLUENT®. Electrolyzer's operational and design parameters will be optimized in order to obtain the maximum hydrogen production and the higher efficiency in the module. This optimized model of the electrolyzer will be incorporated to a chemical process simulation (CPS) code to study the overall high temperature flowsheet coupled to a high temperature accelerator driven system (ADS) that offers advantages in the transmutation of the spent fuel. (author)

APPRAISAL OF FINAL TAILINGS APPLICABILITY FOR PROCESSING AND PRODUCTION OF MODIFIERS OF IRON-CARBON ALLOYS

Directory of Open Access Journals (Sweden)

A. S. Panasugin

2011-01-01

Full Text Available The methodology of rating of the galvanic final tailings applicability for further processing in the interests of needs of metallurgical production of the Republic Belarus is offered.

A low-temperature process for the denitration of Hanford single-shell tank, nitrate-based waste utilizing the nitrate to ammonia and ceramic (NAC) or nitrate to ammonia and glass (NAG) process: Phase 2 report

International Nuclear Information System (INIS)

Mattus, A.J.; Walker, J.F. Jr.; Youngblood, E.L.; Farr, L.L.; Lee, D.D.; Dillow, T.A.; Tiegs, T.N.

1994-12-01

Continuing benchtop studies using Hanford single-shell tank (SST) simulants and actual Oak Ridge National Laboratory (ORNL) low-level waste (LLW), employing a new denitration process for converting nitrate to ammonia and ceramic (NAC), have conclusively shown that between 85 and 99% of the nitrate can be readily converted to gaseous ammonia. In this process, aluminum powders can be used to convert alkaline, nitrate-based supernate to ammonia and an aluminum oxide-sodium aluminate-based solid. The process may be able to use contaminated aluminum scrap metal from DOE sites to effect the conversion. The final, nitrate-free ceramic product can be pressed and sintered like other ceramics or silica and/or fluxing agents can be added to form a glassy ceramic or a flowable glass product. Based upon the starting volumes of 6.2 and 3.1 M sodium nitrate solution, volume reductions of 50 to 70% were obtained for the waste form produced. Sintered pellets produced from supernate from Melton Valley Storage Tanks (MVSTs) have been leached in accordance with the 16.1 leach test for the radioelements 85 Sr and 137 Cs. Despite lengthy counting times, 85 Sr could not be detected in the leachates. 137 Cs was only slightly above background and corresponded to a leach index of 12.2 to 13.7 after 8 months of leaching. Leach testing of unsintered and sintered reactor product spiked with hazardous metals proved that both sintered and unsintered product passed the Toxicity Characteristic Leaching Procedure (TCLP) test. Design of the equipment and flowsheet for a pilot demonstration-scale system to prove the nitrate destruction portion of the NAC process and product formation is under way

DISSOLUTION OF IRRADIATED MURR FUEL ASSEMBLIES

Energy Technology Data Exchange (ETDEWEB)

Kyser, E.

2010-06-17

A literature survey on the dissolution of spent nuclear fuel from the University of Missouri Research Reactor (MURR) has been performed. This survey encompassed both internal and external literature sources for the dissolution of aluminum-clad uranium alloy fuels. The most limiting aspect of dissolution in the current facility configuration involves issues related to the control of the flammability of the off-gas from this process. The primary conclusion of this work is that based on past dissolution of this fuel in H-Canyon, four bundles of this fuel (initial charge) may be safely dissolved in a nitric acid flowsheet catalyzed with 0.002 M mercuric nitrate using a 40 scfm purge to control off-gas flammability. The initial charge may be followed by a second charge of up to five bundles to the same dissolver batch depending on volume and concentration constraints. The safety of this flowsheet relies on composite lower flammability limits (LFL) estimated from prior literature, pilot-scale work on the dissolution of site fuels, and the proposed processing flowsheet. Equipment modifications or improved LFL data offer the potential for improved processing rates. The fuel charging sequence, as well as the acid and catalyst concentrations, will control the dissolution rate during the initial portion of the cycle. These parameters directly impact the hydrogen and off-gas generation and, along with the purge flowrate determine the number of bundles that may be charged. The calculation approach within provides Engineering a means to determine optimal charging patterns. Downstream processing of this material should be similar to that of recent processing of site fuels requiring only minor adjustments of the existing flowsheet parameters.

Next Generation Solvent Development for Caustic-Side Solvent Extraction of Cesium

Energy Technology Data Exchange (ETDEWEB)

Moyer, Bruce A. [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); Birdwell, Joseph F. [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); Bonnesen, Peter V. [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); Bruffey, Stephanie [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)

2014-03-01

This report summarizes the FY 2010 and 2011 accomplishments at Oak Ridge National Laboratory (ORNL) in developing the Next Generation Caustic-Side Solvent Extraction (NG-CSSX) process, referred to commonly as the Next Generation Solvent (NGS), under funding from the U.S. Department of Energy, Office of Environmental Management (DOE-EM), Office of Technology Innovation and Development. The primary product of this effort is a process solvent and preliminary flowsheet capable of meeting a target decontamination factor (DF) of 40,000 for worst-case Savannah River Site (SRS) waste with a concentration factor of 15 or higher in the 18-stage equipment configuration of the SRS Modular Caustic-Side Solvent Extraction Unit (MCU). In addition, the NG-CSSX process may be readily adapted for use in the SRS Salt Waste Processing Facility (SWPF) or in supplemental tank-waste treatment at Hanford upon appropriate solvent or flowsheet modifications. Efforts in FY 2010 focused on developing a solvent composition and process flowsheet for MCU implementation. In FY 2011 accomplishments at ORNL involved a wide array of chemical-development activities and testing up through single-stage hydraulic and mass-transfer tests in 5-cm centrifugal contactors. Under subcontract from ORNL, Argonne National Laboratory (ANL) designed a preliminary flowsheet using ORNL cesium distribution data, and Tennessee Technological University confirmed a chemical model for cesium distribution ratios (DCs) as a function of feed composition. Inter laboratory efforts were coordinated with complementary engineering tests carried out (and reported separately) by personnel at Savannah River National Laboratory (SRNL) and Savannah River Remediation (SRR) with helpful advice by Parsons Engineering and General Atomics on aspects of possible SWPF implementation.

Next Generation Solvent (NGS): Development for Caustic-Side Solvent Extraction of Cesium

Energy Technology Data Exchange (ETDEWEB)

Moyer, Bruce A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Birdwell, Jr, Joseph F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bonnesen, Peter V. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bruffey, Stephanie H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Delmau, Laetitia Helene [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Duncan, Nathan C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ensor, Dale [Tennessee Technological Univ., Cookeville, TN (United States); Hill, Talon G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Lee, Denise L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Rajbanshi, Arbin [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Roach, Benjamin D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Szczygiel, Patricia L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Sloop, Jr., Frederick V. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Stoner, Erica L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Williams, Neil J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2014-03-01

This report summarizes the FY 2010 and 2011 accomplishments at Oak Ridge National Laboratory (ORNL) in developing the Next Generation Caustic-Side Solvent Extraction (NG-CSSX) process, referred to commonly as the Next Generation Solvent (NGS), under funding from the U.S. Department of Energy, Office of Environmental Management (DOE-EM), Office of Technology Innovation and Development. The primary product of this effort is a process solvent and preliminary flowsheet capable of meeting a target decontamination factor (DF) of 40,000 for worst-case Savannah River Site (SRS) waste with a concentration factor of 15 or higher in the 18-stage equipment configuration of the SRS Modular Caustic-Side Solvent Extraction Unit (MCU). In addition, the NG-CSSX process may be readily adapted for use in the SRS Salt Waste Processing Facility (SWPF) or in supplemental tank-waste treatment at Hanford upon appropriate solvent or flowsheet modifications. Efforts in FY 2010 focused on developing a solvent composition and process flowsheet for MCU implementation. In FY 2011 accomplishments at ORNL involved a wide array of chemical-development activities and testing up through single-stage hydraulic and mass-transfer tests in 5-cm centrifugal contactors. Under subcontract from ORNL, Argonne National Laboratory (ANL) designed a preliminary flowsheet using ORNL cesium distribution data, and Tennessee Technological University confirmed a chemical model for cesium distribution ratios (DCs) as a function of feed composition. Interlaboratory efforts were coordinated with complementary engineering tests carried out (and reported separately) by personnel at Savannah River National Laboratory (SRNL) and Savannah River Remediation (SRR) with helpful advice by Parsons Engineering and General Atomics on aspects of possible SWPF implementation.

The Process Synthesis Pyramid: Conceptual design of a Liquefied Energy Chain using Pinch Analysis,Exergy Analysis,Deterministic Optimization and Metaheuristic Searches

International Nuclear Information System (INIS)

Aspelund, Audun

2012-01-01

Process Synthesis (PS) is a term used to describe a class of general and systematic methods for the conceptual design of processing plants and energy systems. The term also refers to the development of the process flowsheet (structure or topology), the selection of unit operations and the determination of the most important operating conditions.In this thesis an attempt is made to characterize some of the most common methodologies in a PS pyramid and discuss their advantages and disadvantages as well as where in the design phase they could be used most efficiently. The thesis shows how design tools have been developed for subambient processes by combining and expanding PS methods such as Heuristic Rules, sequential modular Process Simulations, Pinch Analysis, Exergy Analysis, Mathematical Programming using Deterministic Optimization methods and optimization using Stochastic Optimization methods. The most important contributions to the process design community are three new methodologies that include the pressure as an important variable in heat exchanger network synthesis (HENS).The methodologies have been used to develop a novel and efficient energy chain based on stranded natural gas including power production with carbon capture and sequestration (CCS). This Liquefied Energy Chain consists of an offshore process a combined gas carrier and an onshore process. This energy chain is capable of efficiently exploiting resources that cannot be utilized economically today with minor Co2 emissions. Finally, a new Stochastic Optimization approach based on a Tabu Search (TS), the Nelder Mead method or Downhill Simplex Method (NMDS) and the sequential process simulator HYSYS is used to search for better solutions for the Liquefied Energy Chain with respect to minimum cost or maximum profit. (au)

The Process Synthesis Pyramid: Conceptual design of a Liquefied Energy Chain using Pinch Analysis,Exergy Analysis,Deterministic Optimization and Metaheuristic Searches

Energy Technology Data Exchange (ETDEWEB)

Aspelund, Audun

2012-07-01

Process Synthesis (PS) is a term used to describe a class of general and systematic methods for the conceptual design of processing plants and energy systems. The term also refers to the development of the process flowsheet (structure or topology), the selection of unit operations and the determination of the most important operating conditions.In this thesis an attempt is made to characterize some of the most common methodologies in a PS pyramid and discuss their advantages and disadvantages as well as where in the design phase they could be used most efficiently. The thesis shows how design tools have been developed for subambient processes by combining and expanding PS methods such as Heuristic Rules, sequential modular Process Simulations, Pinch Analysis, Exergy Analysis, Mathematical Programming using Deterministic Optimization methods and optimization using Stochastic Optimization methods. The most important contributions to the process design community are three new methodologies that include the pressure as an important variable in heat exchanger network synthesis (HENS).The methodologies have been used to develop a novel and efficient energy chain based on stranded natural gas including power production with carbon capture and sequestration (CCS). This Liquefied Energy Chain consists of an offshore process a combined gas carrier and an onshore process. This energy chain is capable of efficiently exploiting resources that cannot be utilized economically today with minor Co2 emissions. Finally, a new Stochastic Optimization approach based on a Tabu Search (TS), the Nelder Mead method or Downhill Simplex Method (NMDS) and the sequential process simulator HYSYS is used to search for better solutions for the Liquefied Energy Chain with respect to minimum cost or maximum profit. (au)

A systematic framework for CAFD and resources allocation optimisation using MINLP in vegetable oil processing

DEFF Research Database (Denmark)

Quaglia, Alberto; Sarup, Bent; Sin, Gürkan

Although being a mature and well established industry segment, over the last few decades the vegetable oil industry has been facing many important new challenges due to emerging new products (such as biodiesel and nutraceuticals compounds), as well as new trends and regulations with regards....... In this paper, a systematic framework for Computer-Aided Flowsheet Syntesis and Design (CAFD) and resources allocation for the vegetable oil sector is presented. In the framework a Mixed Integer Non Linear Programming (MINLP) problem is formulated and solved for a soybean processing case study, to determine...... the optimal processing network for vegetable oil extraction and refining (including biodiesel production and various options for byproducts valorization), as well as the optimal material flows to each processing step. In order to optimize the resources needed to solve such a large and complex problem...

Processing of FRG mixed oxide fuel elements at General Atomic under the US/FRG cooperative agreement for spent fuel elements

International Nuclear Information System (INIS)

Holder, N.D.; Strand, J.B.; Schwarz, F.A.; Tischer, H.E.

1980-11-01

The Federal Republic of Germany (FRG) and the United States (US) are cooperating on certain aspects gas-cooled reactor technology under an umbrella agreement. Under the spent fuel treatment section of the agreement, FRG fuel spheres were recently sent for processing in the Department of Energy sponsored cold pilot plant for High-Temperature Gas-Cooled Reactor (HTGR) fuel processing at General Atomic Company in San Diego, California. The FRG fuel spheres were crushed and burned to recover coated fuel particles. These particles were in turn crushed and burned to recover the fuel-bearing kernels for further treatment for uranium recovery. Successful completion of the tests described in this paper demonstrated the applicability of the US HTGR fuel treatment flowsheet to FRG fuel processing. 10 figures

Pyrometallurgical processing of Integral Fast Reactor metal fuels

International Nuclear Information System (INIS)

Battles, J.E.; Miller, W.E.; Gay, E.C.

1991-01-01

The pyrometallurgical process for recycling spent metal fuels from the Integral Fast Reactor is now in an advanced state of development. This process involves electrorefining spent fuel with a cadmium anode, solid and liquid cathodes, and a molten salt electrolyte (LiCl-KCl) at 500 degrees C. The initial process feasibility and flowsheet verification studies have been conducted in a laboratory-scale electrorefiner. Based on these studies, a dual cathode approach has been adopted, where uranium is recovered on a solid cathode mandrel and uranium-plutonium is recovered in a liquid cadmium cathode. Consolidation and purification (salt and cadmium removal) of uranium and uranium-plutonium products from the electrorefiner have been successful. The process is being developed with the aid of an engineering-scale electrorefiner, which has been successfully operated for more than three years. In this electrorefiner, uranium has been electrotransported from the cadmium anode to a solid cathode in 10 kg quantities. Also, anodic dissolution of 10 kg batches of chopped, simulated fuel (U--10% Zr) has been demonstrated. Development of the liquid cadmium cathode for recovering uranium-plutonium is under way

Final audit report of remedial action construction at the UMTRA Project, Grand Junction, Colorado, processing site

International Nuclear Information System (INIS)

1995-02-01

This final audit report (FAR) for remedial action at the Grand Junction, Colorado, Uranium Mill Tailings Remedial Action (UMTRA) Project processing site consists of a summary of the radiological surveillances/ audits, the quality assurance (QA) in-process surveillances, and the QA final close-out inspection performed by the US Department of Energy (DOE) and Technical Assistance Contractor (TAC). The FAR also summarizes other surveillances performed by the US Nuclear Regulatory Commission (NRC). To summarize, a total of one finding and 127 observations were noted during DOE/TAC audit and surveillance activities. The NRC noted general site-related observations during the OSCRs. Follow-up to responses required from MK-Ferguson for the DOE/TAC finding and observations indicated that all issues related to the Grand Junction processing site were resolved and closed out to the DOE's satisfaction. The NRC OSCRs resulted in no issues related to the Grand Junction processing site requiring a response from MK-Ferguson

Physico - chemical Up-grading studies of uraniferous xenotime mineral from ferruginous siltstone sediments of Um hamd, Um bogma area, sinai, Egypt

International Nuclear Information System (INIS)

Abdel-Monem, H.M.; Elassy, I.E.; Bishay, A.F.; El-Kammar, A.M.

1998-01-01

The exposed paleozoic deposits at Um Bogma area in west central sinai contain three main units namely; the lower sandstone, the middle carbonate and the upper sandstone. These units are composed of intercalation of clay minerals, sand, carbonate and iron oxides.The upper layers of the lower sandstone and the base of the middle carbonate units contain possible resources of anomalously uraniferous xenotime (YPO 4 ). Xenotime, being preferentially concentrated with the very fine grain sizes, is more abundant in the argillaceous sediments. The associated clay minerals are represented by kaolinite the non-clay minerals include hematite, goethite and quartz.The up-grading of uraniferous xenotime from a bulk head sample assaying 1.44% Y 2 O 3 was performed by collaborative technological processes that include; liquid/liquid separation using two immiscible liquids (CCI 4 and H 2 O) and finally acid leaching. By applying the proposed flowsheet, more than 90% of xenotime is recovered with an assay value of more than 45% Y 2 O 3 . The simplicity of the proposed flowsheet allows it to be adopted to semi-industrial or even industrial purposes

Characterization Of The As-Received Sludge Batch 9 Qualification Sample (Htf-51-15-81)

Energy Technology Data Exchange (ETDEWEB)

Pareizs, J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2015-09-30

Savannah River National Laboratory (SRNL) personnel have been requested to qualify the next sludge batch (Sludge Batch 9 – SB9) for processing at the Defense Waste Processing Facility (DWPF). To accomplish this task, Savannah River Remediation (SRR) has sent SRNL a 3-L slurried sample of Tank 51H (HTF-51-15-81) to be characterized, washed, and then used in a lab-scale demonstration of the DWPF flowsheet (potentially after combining with Tank 40H sludge). This report documents the first steps of the qualification process – characterization of the as-received Tank 51H qualification sample. These results will be used to support a reprojection of SB9 by SRR from which final Tank 51H washing, frit development, and Chemical Processing Cell (CPC) activities will be based.

Significant volume reduction of tank waste by selective crystallization: 1994 Annual report

International Nuclear Information System (INIS)

Herting, D.L.; Lunsford, T.R.

1994-01-01

The objective of this technology task plan is to develop and demonstrate a scaleable process of reclaim sodium nitrate (NaNO 3 ) from Hanford waste tanks as a clean nonradioactive salt. The purpose of the so-called Clean Salt Process is to reduce the volume of low level waste glass by as much as 70%. During the reporting period of October 1, 1993, through May 31, 1994, progress was made on four fronts -- laboratory studies, surrogate waste compositions, contracting for university research, and flowsheet development and modeling. In the laboratory, experiments with simulated waste were done to explore the effects of crystallization parameters on the size and crystal habit of product NaNO 3 crystals. Data were obtained to allows prediction of decontamination factor as a function of solid/liquid separation parameters. Experiments with actual waste from tank 101-SY were done to determine the extent of contaminant occlusions in NaNO 3 crystals. In preparation for defining surrogate waste compositions, single shell tanks were categorized according to the weight percent NaNO 3 in each tank. A detailed process flowsheet and computer model were created using the ASPENPlus steady state process simulator. This is the same program being used by the Tank Waste Remediation System (TWRS) program for their waste pretreatment and disposal projections. Therefore, evaluations can be made of the effect of the Clean Salt Process on the low level waste volume and composition resulting from the TWRS baseline flowsheet. Calculations, using the same assumptions as used for the TWRS baseline where applicable indicate that the number of low level glass vaults would be reduced from 44 to 16 if the Clean Salt Process were incorporated into the baseline flowsheet

Advanced Multi-Product Coal Utilization By-Product Processing Plant

Energy Technology Data Exchange (ETDEWEB)

Thomas Robl; John Groppo

2009-06-30

The overall objective of this project is to design, construct, and operate an ash beneficiation facility that will generate several products from coal combustion ash stored in a utility ash pond. The site selected is LG&E's Ghent Station located in Carroll County, Kentucky. The specific site under consideration is the lower ash pond at Ghent, a closed landfill encompassing over 100 acres. Coring activities revealed that the pond contains over 7 million tons of ash, including over 1.5 million tons of coarse carbon and 1.8 million tons of fine (<10 {micro}m) glassy pozzolanic material. These potential products are primarily concentrated in the lower end of the pond adjacent to the outlet. A representative bulk sample was excavated for conducting laboratory-scale process testing while a composite 150 ton sample was also excavated for demonstration-scale testing at the Ghent site. A mobile demonstration plant with a design feed rate of 2.5 tph was constructed and hauled to the Ghent site to evaluate unit processes (i.e. primary classification, froth flotation, spiral concentration, secondary classification, etc.) on a continuous basis to determine appropriate scale-up data. Unit processes were configured into four different flowsheets and operated at a feed rate of 2.5 tph to verify continuous operating performance and generate bulk (1 to 2 tons) products for product testing. Cementitious products were evaluated for performance in mortar and concrete as well as cement manufacture process addition. All relevant data from the four flowsheets was compiled to compare product yields and quality while preliminary flowsheet designs were generated to determine throughputs, equipment size specifications and capital cost summaries. A detailed market study was completed to evaluate the potential markets for cementitious products. Results of the study revealed that the Ghent local fly ash market is currently oversupplied by more than 500,000 tpy and distant markets (i

Integration of thermodynamic insights and MINLP optimisation for the synthesis, design and analysis of process flowsheets

DEFF Research Database (Denmark)

Hostrup, Martin; Gani, Rafiqul; Kravanja, Zdravko

1999-01-01

This paper presents an integrated approach to the solution of process synthesis, design and analysis problems. Integration is achieved by combining two different techniques, synthesis based on thermodynamic insights and structural optimization together with a simulation engine and a properties pr...

Glycolic Acid Physical Properties, Impurities, And Radiation Effects Assessment

International Nuclear Information System (INIS)

Pickenheim, B.; Bibler, N.

2010-01-01

series of tests to determine whether the polymer will be formed is currently being outlined. The first phase will be a simple experiment where a simulated SRAT supernatant containing the 80:20 blend of glycolic - formic acid could be irradiated in the Co-60 gamma source at SRNL to a very large dose resembling the dose received by the radioactive SRAT solution after several weeks. The resulting solution could then be heated to simulate refluxing in the SRAT process. Finally a radioactive demonstration of the SRAT process should be performed in the SRNL Shielded Cells to confirm successful execution of the glycolic - formic acid flowsheet.

Thorium base fuels reprocessing at the L.P.R. (Radiochemical Processes Laboratory) experimental plant

International Nuclear Information System (INIS)

Almagro, J.C.; Dupetit, G.A.; Deandreis, R.A.

1987-01-01

The availability of the LPR (Radiochemical Processes Laboratory) plant offers the possibility to demonstrate and create the necessary technological basis for thorium fuels reprocessing. To this purpose, the solvents extraction technique is used, employing TBP (at 30%) as solvent. The process is named THOREX, a one-cycle acid, which permits an adequate separation of Th 232 and U 233 components and fission products. For thorium oxide elements dissolution, the 'chopp-leach' process (installed at LPR) is used, employing a NO 3 H 13N, 0.05M FH and 0.1M Al (NO 3 ) 3 , as solvent. To adapt the pilot plant to the flow-sheet requirements proposed, minor modifications must be carried out in the interconnection of the existing decanting mixers. The input of the plant has been calculated by Origin Code modified for irradiations in reactors of the HWR type. (Author)

Am/Cm Vitrification Process: Pretreatment Material Balance Calculations

International Nuclear Information System (INIS)

Smith, F.G.

2001-01-01

This report documents material balance calculations for the pretreatment steps required to prepare the Americium/Curium solution currently stored in Tank 17.1 in the F-Canyon for vitrification. The material balance uses the latest analysis of the tank contents to provide a best estimate calculation of the expected plant operations during the pretreatment process. The material balance calculations primarily follow the material that directly leads to melter feed. Except for vapor products of the denitration reactions and treatment of supernate from precipitation and precipitate washing, the flowsheet does not include side streams such as acid washes of the empty tanks that would go directly to waste. The calculation also neglects tank heels. This report consolidates previously reported results, corrects some errors found in the spreadsheet and provides a more detailed discussion of the calculation basis

Tank Farm Contractor Operation and Utilization Plan [SEC 1 Thru 3

International Nuclear Information System (INIS)

KIRKBRIDE, R.A.

2000-01-01

This document updates the operating scenario and plans for feed delivery to BNFL Inc. of retrieval and waste from single-shell tanks, and the overall process flowsheets for Phases 1 and 2 of the River Protection Project. The plans and flowsheets are updated with the most recent guidance from ORP and tank-by-tank inventory. The results provide the technical basis for the RTP-2 planning effort. Sensitivity cases were run to evaluate the effect of changes on key parameters

Development of a New Process for the Selective Extraction of Uranium from Phosphate Rocks

International Nuclear Information System (INIS)

Bernier, G.; Miguirditchian, M.; Pacary, V.; Balaguer, C.; Bertrand, M.; Camès, B.; Hérès, X.; Mokhtari, H.

2014-01-01

Conclusions: • Amido-phosphonate DEHCNPB is very promising for U extraction from phosphoric acid – Selective extraction of U even among high concentrations of iron in genuine 5 M H3PO4 solution : [Fe]/[U] reduced from 30 (feed) to 4.3 10-4 (product); • Parametric laboratory studies: – Characterize the stoichiometry of extracted complex in organic phase; – Propose extraction mechanisms; • Model developed and implemented in CEA PAREX code: – Modelling and flowsheet design; • Counter-current test on PROUST platform in G1 facility with genuine industrial phosphoric solution: – Treatment of an industrial phosphoric acid in laboratory-scale mixer-settlers; – Very promising performances of the run with good U recovery (91%) and good decontamination ([Fe] / [U] < ASTM specifications); – Experimental and calculated profiles of U and Fe in good agreement; • Experiments in progress to optimize this flowsheet

Estimating the transmission potential of supercritical processes based on the final size distribution of minor outbreaks.

Science.gov (United States)

Nishiura, Hiroshi; Yan, Ping; Sleeman, Candace K; Mode, Charles J

2012-02-07

Use of the final size distribution of minor outbreaks for the estimation of the reproduction numbers of supercritical epidemic processes has yet to be considered. We used a branching process model to derive the final size distribution of minor outbreaks, assuming a reproduction number above unity, and applying the method to final size data for pneumonic plague. Pneumonic plague is a rare disease with only one documented major epidemic in a spatially limited setting. Because the final size distribution of a minor outbreak needs to be normalized by the probability of extinction, we assume that the dispersion parameter (k) of the negative-binomial offspring distribution is known, and examine the sensitivity of the reproduction number to variation in dispersion. Assuming a geometric offspring distribution with k=1, the reproduction number was estimated at 1.16 (95% confidence interval: 0.97-1.38). When less dispersed with k=2, the maximum likelihood estimate of the reproduction number was 1.14. These estimates agreed with those published from transmission network analysis, indicating that the human-to-human transmission potential of the pneumonic plague is not very high. Given only minor outbreaks, transmission potential is not sufficiently assessed by directly counting the number of offspring. Since the absence of a major epidemic does not guarantee a subcritical process, the proposed method allows us to conservatively regard epidemic data from minor outbreaks as supercritical, and yield estimates of threshold values above unity. Crown Copyright © 2011. Published by Elsevier Ltd. All rights reserved.

Prototype equipment status monitor for plant operational configuration management

International Nuclear Information System (INIS)

DeVerno, M.; Trask, D.; Groom, S.

1998-01-01

CANDU plants, such as the Point Lepreau GS, have tens of thousands of operable devices. The status of each operable device must be immediately available to plan and execute future changes to the plant. Historically, changes to the plant's operational configuration have been controlled using manual and administrative methods where the status of each operable device is maintained on operational flowsheets located in the work control area of the main control room. The operational flowsheets are used to plan and develop Operating Orders (OOs) or Order-to-Operate (OTOs) and the control centre work processes are used to manage their execution. After performing each OO procedure, the operational flowsheets are updated to reflect the new plant configuration. This process can be very time consuming, and due to the manual processes, can lead to the potential for time lags and errors in the recording of the current plant configuration. Through a cooperative research and development program, Canadian CANDU utilities and Atomic Energy of Canada Limited, the design organization, have applied modern information technologies to develop a prototype Equipment Status Monitor (ESM) to address processes and information flow for efficient operational configuration management. The ESM integrates electronic operational flowsheets, equipment databases, engineering and work management systems, and computerized procedures to assess, plan, execute, track, and record changes to the plant's operational configuration. This directly leads to improved change control, more timely and accurate plant status information, fewer errors, and better decision making regarding future changes. These improvements to managing the plant's operational configuration are essential to increasing plant safety, achieving a high plant availability, and maintaining high capability and capacity factors. (author)

Study of Advanced Reactor Mixed Oxide Fuel Production of (U,Th)O2

International Nuclear Information System (INIS)

Busron-Masduki; Damunir; Pristi-Hartati; R-Sukarsono; Bangun-Wasito

2000-01-01

The high price and starting scarcity of reserved of oil drive the people to drill the alternative nuclear energy. Accelerator-driven Transmutation Waste (ATW) is a prospective technology to solve the problem of used fuel waste, to reduce the anxiety of long term disposal waste, to increase the public acceptance of nuclear energy enter into the third millennium. The future of large nuclear energy appears in many-branched industry will depend on the capability to generate relatively low priced fuel on the basis of commercial nuclear energy. Utilization of uranium-233 -thorium cycle insures long-term fuel supply, makes the nuclear energy production more flexible and enables the self-provision regime to be realized in future. Flowsheet of mixed oxide fuel production for advanced reactor of (U,Th)O 2 is a combination of existing manufacturing equipment and quality assurance program from commercial LWR and HTR. The front-end of flowsheet using sol-gel process. The external sol-gel process is chosen due to simple equipment can anticipate refabrication of U-233 which always contains a few hundred ppm of U-232 and its gamma-emitting daughters, besides yielding smaller waste. The decision to choose external sol-gel process encourages to develop External Gelation Thorium (EGT). In order to get higher density and relatively low compaction pressures (i.e. for advanced LWR) adopted flowsheet EGT is developed to be Sol-Gel Microsphere Pelletization (SGMP). Using the optimal parameters, SGMP become established flowsheet for producing mixed oxide fuel of (U,Th)O 2 for advanced reactor. (author)

Physicochemical aspects of decomposition of silica-alumina ores of argillites and green clays of Chashma-Sang Deposit of the Republic of Tajikistan by hydrochloric and nitric acids

International Nuclear Information System (INIS)

Kayumov, A.M.

2018-01-01

The purpose of work is to study the processes of decomposition of silica-alumina ores of argillites and green clays of Chashma-Sang Deposit of the Republic of Tajikistan by hydrochloric and nitric acids in temperature interval 20-98 deg C with the using of methods of selective extraction of valuable materials; elaboration of rational conditions of decomposition of raw material. Physicochemical properties of initial aluminium comprising ores, intermediate and final products of processing of argillites and green clays have been studied. Kinetic parameters of processes at acidic decomposition of argillites and green clays have been studied as well. The kinetic parameters of processes of decomposition of green clays and argillites by nitric and hydrochloric acids have been calculated. The flowsheet of complex processing of green clays and argillites of Chashma-Sang Deposit has been elaborated.

Assessment of thermochemical hydrogen production. Project 8994 mid-contract progress report, July 1--November 1, 1977. [Iron chloride and copper sulfate cycles

Energy Technology Data Exchange (ETDEWEB)

Dafler, J.R.; Foh, S.E.; Schreiber, J.D.

1977-12-01

We have completed the base-case (first-cut) flowsheet analysis for two thermochemical water-splitting cycles that have been under study at the Institute of Gas Technology: a four-step iron chloride cycle (denoted B-1) and a four-step copper sulfate cycle (denoted H-5). In the case of Cycle B-1, an energy balance has located the worst problem areas in the cycle, and flowsheet modifications have begun. Calculations of equilibrium effects due to the hydrolysis of ferrous chloride at pressures high enough to interface with projected hydrogen transmission systems will, apparently, necessitate higher temperature process heat input for this step. Higher pressure operation of some critical separation processes yields more favorable heat balances. For Cycle H-5, the unmodified (base-case) flowsheet indicates that reaction product separations will be relatively simple with respect to Cycle B-1. Work of Schuetz and others dealing with the electrolysis and thermodynamics of HBr/H/sub 2/O/SO/sub 2/ systems is being extensively reviewed. Work plans for this part of the contract are currently being reviewed.

Heat transfer enhanced microwave process for stabilization of liquid radioactive waste slurry. Final report

International Nuclear Information System (INIS)

White, T.L.

1995-01-01

The objectve of this CRADA is to combine a polymer process for encapsulation of liquid radioactive waste slurry developed by Monolith Technology, Inc. (MTI), with an in-drum microwave process for drying radioactive wastes developed by Oak Ridge National Laboratory (ORNL), for the purpose of achieving a fast, cost-effectve commercial process for solidification of liquid radioactive waste slurry. Tests performed so far show a four-fold increase in process throughput due to the direct microwave heating of the polymer/slurry mixture, compared to conventional edge-heating of the mixer. We measured a steady-state throughput of 33 ml/min for 1.4 kW of absorbed microwave power. The final waste form is a solid monolith with no free liquids and no free particulates

Spent fuel reprocessing and minor actinide partitioning safety related research at the UK National Nuclear Laboratory

International Nuclear Information System (INIS)

Carrott, Michael; Flint, Lauren; Gregson, Colin; Griffiths, Tamara; Hodgson, Zara; Maher, Chris; Mason, Chris; McLachlan, Fiona; Orr, Robin; Reilly, Stacey; Rhodes, Chris; Sarsfield, Mark; Sims, Howard; Shepherd, Daniel; Taylor, Robin; Webb, Kevin; Woodall, Sean; Woodhead, David

2015-01-01

The development of advanced separation processes for spent nuclear fuel reprocessing and minor actinide recycling is an essential component of international R and D programmes aimed at closing the nuclear fuel cycle around the middle of this century. While both aqueous and pyrochemical processes are under consideration internationally, neither option will gain broad acceptance without significant advances in process safety, waste minimisation, environmental impact and proliferation resistance; at least when compared to current reprocessing technologies. The UK National Nuclear Laboratory (NNL) is developing flowsheets for innovative aqueous separation processes. These include advanced PUREX options (i.e. processes using tributyl phosphate as the extractant for uranium, plutonium and possibly neptunium recovery) and GANEX (grouped actinide extraction) type processes that use diglycolamide based extractants to co-extract all transuranic actinides. At NNL, development of the flowsheets is closely linked to research on process safety, since this is essential for assessing prospects for future industrialisation and deployment. Within this context, NNL is part of European 7. Framework projects 'ASGARD' and 'SACSESS'. Key topics under investigation include: hydrogen generation from aqueous and solvent phases; decomposition of aqueous phase ligands used in separations prior to product finishing and recycle of nitric acid; dissolution of carbide fuels including management of organics generated. Additionally, there is a strong focus on use of predictive process modelling to assess flowsheet sensitivities as well as engineering design and global hazard assessment of these new processes. (authors)

Waste minimisation in a hard chromiun plating Small Medium Enterprise (SME).

Science.gov (United States)

Viguri, J R; Andrés, A; Irabien, A

2002-01-01

The high potential of waste stream minimisation in the metal finishing sector justifies specific studies of Small and Medium Enterprises (SME). In this work, the minimisation options of the wastes generated in a hard chromium plating activity have been analysed. The study has been performed in a small job shop company, which works in batch mode with big pieces. A process flowsheet after connecting the unit operations and determining the process inputs (raw and secondary materials) and outputs (waste streams) has been carried out. The main properties, quantity and current management of the waste streams have been shown. The obvious lack of information has been identified and finally the waste minimisation options that could be adopted by the company have been recorded.

U.S. Department of Energy integrated manufacturing & processing predoctoral fellowships. Final Report

Energy Technology Data Exchange (ETDEWEB)

Petrochenkov, Margaret

2003-03-31

The objective of this program was threefold: to create a pool of PhDs trained in the integrated approach to manufacturing and processing, to promote academic interest in the field, and to attract talented professionals to this challenging area of engineering. It was anticipated that the program would result in the creation of new manufacturing methods that would contribute to improved energy efficiency, to better utilization of scarce resources, and to less degradation of the environment. Emphasis in the competition was on integrated systems of manufacturing and the integration of product design with manufacturing processes. Research addressed such related areas as aspects of unit operations, tooling and equipment, intelligent sensors, and manufacturing systems as they related to product design. This is the final report to close out the contract.

An evaluation of reactor cooling and coupled hydrogen production processes using the modular helium reactor

International Nuclear Information System (INIS)

Harvego, E.A.; Reza, S.M.M.; Richards, M.; Shenoy, A.

2006-01-01

The high-temperature characteristics of the modular helium reactor (MHR) make it a strong candidate for producing hydrogen using either thermochemical or high-temperature electrolysis (HTE) processes. Using heat from the MHR to drive a sulfur-iodine (SI) thermochemical hydrogen production process has been the subject of a U.S. Department of Energy sponsored Nuclear Engineering Research Initiative (NERI) project led by General Atomics, with participation from the Idaho National Laboratory (INL) and Texas A and M University. While the focus of much of the initial work was on the SI thermochemical production of hydrogen, recent activities included development of a preconceptual design for an integral HTE hydrogen production plant driven by the process heat and electricity produced by a 600 MW MHR. This paper describes ATHENA analyses performed to evaluate alternative primary system cooling configurations for the MHR to minimize peak reactor vessel and core temperatures while achieving core helium outlet temperatures in the range of 900-1000 deg. C that are needed for the efficient production of hydrogen using either the SI or HTE process. The cooling schemes investigated are intended to ensure peak fuel temperatures do not exceed specified limits under normal or transient upset conditions, and that reactor vessel temperatures do not exceed American Society of Mechanical Engineers (ASME) code limits for steady-state or transient conditions using standard light water reactor vessel materials. Preconceptual designs for SI and HTE hydrogen production plants driven by one or more 600 MW MHRs at helium outlet temperatures in the range of 900-1000 deg. C are described and compared. An initial SAPHIRE model to evaluate the reliability, maintainability, and availability of the SI hydrogen production plant is also described. Finally, a preliminary flowsheet for a conceptual design of an HTE hydrogen production plant coupled to a 600 MW modular helium reactor is presented and

The CAM-ICU has now a French "official" version. The translation process of the 2014 updated Complete Training Manual of the Confusion Assessment Method for the Intensive Care Unit in French (CAM-ICU.fr).

Science.gov (United States)

Chanques, Gérald; Garnier, Océane; Carr, Julie; Conseil, Matthieu; de Jong, Audrey; Rowan, Christine M; Ely, E Wesley; Jaber, Samir

2017-10-01

Delirium is common in Intensive-Care-Unit (ICU) patients but under-recognized by bed-side clinicians when not using validated delirium-screening tools. The Confusion-Assessment-Method for the ICU (CAM-ICU) has demonstrated very good psychometric properties, and has been translated into many different languages though not into French. We undertook this opportunity to describe the translation process. The translation was performed following recommended guidelines. The updated method published in 2014 including introduction letters, worksheet and flowsheet for bed-side use, the method itself, case-scenarios for training and Frequently-Asked-Questions (32 pages) was translated into French language by a neuropsychological researcher who was not familiar with the original method. Then, the whole method was back-translated by a native English-French bilingual speaker. The new English version was compared to the original one by the Vanderbilt University ICU-delirium-team. Discrepancies were discussed between the two teams before final approval of the French version. The entire process took one year. Among the 3692 words of the back-translated version of the method itself, 18 discrepancies occurred. Eight (44%) lead to changes in the final version. Details of the translation process are provided. The French version of CAM-ICU is now available for French-speaking ICUs. The CAM-ICU is provided with its complete training-manual that was challenging to translate following recommended process. While many such translations have been done for other clinical tools, few have published the details of the process itself. We hope that the availability of such teaching material will now facilitate a large implementation of delirium-screening in French-speaking ICUs. Copyright © 2017 Société française d'anesthésie et de réanimation (Sfar). All rights reserved.

Tank Farm Contractor Waste Remediation System and Utilization Plan

International Nuclear Information System (INIS)

KIRKBRIDE, R.A.

1999-01-01

The Tank Waste Remediation System Operation and Utilization Plan updates the operating scenario and plans for the delivery of feed to BNFL Inc., retrieval of waste from single-shell tanks, and the overall process flowsheets for Phases I and II of the privatization of the Tank Waste Remediation System. The plans and flowsheets are updated with the most recent tank-by-tank inventory and sludge washing data. Sensitivity cases were run to evaluate the impact or benefits of proposed changes to the BNFL Inc. contract and to evaluate a risk-based SST retrieval strategy

Tank Farm Contractor Operation and Utilization Plan [SEC 1 Thru 3

Energy Technology Data Exchange (ETDEWEB)

KIRKBRIDE, R.A.

1999-05-04

The Tank Waste Remediation System Operation and Utilization Plan updates the operating scenario and plans for the delivery of feed to BNFL Inc., retrieval of waste from single-shell tanks, and the overall process flowsheets for Phases I and II of the privatization of the Tank Waste Remediation System. The plans and flowsheets are updated with the most recent tank-by-tank inventory and sludge washing data. Sensitivity cases were run to evaluate the impact or benefits of proposed changes to the BNFL Inc. contract and to evaluate a risk-based SST retrieval strategy.

Study of assessing aqueous reprocessing process for the pipeless reprocessing plant

International Nuclear Information System (INIS)

Hanzawa, Masatoshi; Morioka, Nobuo; Fumoto, Hiromichi; Nishimura, Kenji; Chikazawa, Takahiro

2000-02-01

The purpose of this study is to investigate the possibility of new reprocessing process for the purpose of introducing pipeless plant concept, where aqueous separation methods other than solvent extraction method are adopted in order to develop more economical FBR fuel (MOX fuel) reprocessing process. At it's first stage, literature survey on precipitation method, crystallization method and ion-exchange method was performed. Based on the results, following processes were candidated for pipeless reprocessing plant. (1) The process adopting crystallization method and peroxide precipitation method (2) The process adopting oxalate precipitation method (3) The process under mild aqueous conditions (crystallization method and precipitation method) (4) The process adopting crystallization method and ion-exchange method (5) The process adopting crystallization method and solvent extraction method. The processes (1)-(5) were compared with each others in terms of competitiveness to the conventional reference process, and merits and demerits were evaluated from the viewpoint of applicability to pipeless reprocessing plant, safety, economy, Efficiencies in consumption of Resources, non-proliferation, and, Operation and Maintenance. As a result, (1) The process adopting crystallization method and peroxide precipitation method was selected as the most reasonable process to pipeless plant. Preliminary criticality safety analyses, main process chemical flowsheet, main equipment list and layout of mobile vessels and stations were reported for the (1) process. (author)

Development and demonstration of innovative partitioning processes (i-SANEX and 1-cycle SANEX) for actinide partitioning

International Nuclear Information System (INIS)

Wilden, Andreas; Modolo, Giuseppe; Geist, Andreas

2015-01-01

For the recovery of the trivalent actinides Am(III) and Cm(III) from PUREX raffinate, two innovative partitioning processes were developed within the European project ACSEPT. In the 'innovative-SANEX' concept, trivalent actinides (An(III)) and lanthanides (Ln(III)) are co-extracted by a TODGA-based solvent, which is then subjected to several stripping steps: selective stripping of An(III) with the hydrophilic ligand SO 3 -Ph-BTP, followed by subsequent stripping of Ln(III). A more challenging route studied also within our laboratories is the direct An(III) separation using a mixture of CyMe 4 BTBP and TODGA, the so-called '1-cycle SANEX' process. Both processes have been successfully demonstrated using spiked simulate solutions in laboratory-scale miniature annular centrifugal contactors using 32-stages flowsheets. The process development and results of the demonstration tests will be presented and discussed. Both processes showed a high recovery of An(III) with high fission-product decontamination factors. The safety of these processes is studied within the current European project SACSESS. (authors)

Dehydration of ethanol with salt extractive distillation-a comparative analysis between processes with salt recovery

Energy Technology Data Exchange (ETDEWEB)

Ligero, E.L.; Ravagnani, T.M.K. [Departamento de Engenharia de Sistemas Qumicos, Faculdade de Engenharia Qumica, Universidade Estadual de Campinas, Campinas, Sao Paulo (Brazil)

2003-07-01

Anhydrous ethanol can be obtained from a dilute aqueous solution of ethanol via extractive distillation with potassium acetate. Two process flowsheets with salt recovery were proposed. In the first, dilute ethanol is directly fed to a salt extractive distillation column and, after that, the salt is recovered in a multiple effect evaporator followed by a spray dryer. In the second, the concentrated ethanol from conventional distillation is fed to a salt extractive distillation column. In this case, salt is recovered in a single spray dryer. In both processes the recovered salt is recycled to be used in the extractive distillation column. Every component of each process was rigorously modeled and its behavior was simulated for a wide range of operating conditions. A global simulation was then carried out. The results show that the second process is more interesting in terms of energy consumption than the first. Furthermore, it would be easier to implement changes on existing benzene extractive anhydrous ethanol plants to convert them to more ecologically attractive concentrated ethanol feed processes. (author)

The Theory of High Energy Collision Processes - Final Report DOE/ER/40158-1

Energy Technology Data Exchange (ETDEWEB)

Wu, Tai, T.

2011-09-15

In 1984, DOE awarded Harvard University a new Grant DE-FG02-84ER40158 to continue their support of Tai Tsun Wu as Principal Investigator of research on the theory of high energy collision processes. This Grant was renewed and remained active continuously from June 1, 1984 through November 30, 2007. Topics of interest during the 23-year duration of this Grant include: the theory and phenomenology of collision and production processes at ever higher energies; helicity methods of QED and QCD; neutrino oscillations and masses; Yang-Mills gauge theory; Beamstrahlung; Fermi pseudopotentials; magnetic monopoles and dyons; cosmology; classical confinement; mass relations; Bose-Einstein condensation; and large-momentum-transfer scattering processes. This Final Report describes the research carried out on Grant DE-FG02-84ER40158 for the period June 1, 1984 through November 30, 2007. Two books resulted from this project and a total of 125 publications.

An efficient hybrid sulfur process using PEM electrolysis with a bayonet decomposition reactor - HTR2008-58207

International Nuclear Information System (INIS)

Gorensek, M. B.; Summers, W. A.; Lahoda, E. J.; Bolthrunis, C. O.; Greyvenstein, R.

2008-01-01

The Hybrid Sulfur (HyS) Process is being developed to produce hydrogen by water-splitting using heat from advanced nuclear reactors. It has the potential for high efficiency and competitive hydrogen production cost, and has been demonstrated at a laboratory scale. As a two-step process, the HyS is one of the simplest thermochemical cycles. The sulfuric acid decomposition reaction is common to all sulfur cycles, including the Sulfur-Iodine (SI) cycle. What distinguishes the HyS Process from the other sulfur cycles is the use of sulfur dioxide (SO 2 ) to depolarize the anode of a water electrolyzer. The two critical HyS Process components are the SO 2 - depolarized electrolyzer (SDE), and the high-temperature decomposition reactor. A proton exchange membrane (PEM)- type SDE and a silicon carbide bayonet-type high-temperature decomposition reactor are being developed for DOE's Nuclear Hydrogen Initiative (NHI) by Savannah River National Laboratory (SRNL) and by Sandia National Laboratories (SNL), respectively. The ultimate goal of the NHI-sponsored work is to couple the SDE and the reactor in an integrated laboratory scale experiment to prove the technical readiness of the HyS cycle for the NGNP demonstration. This paper describes the flowsheet that is being prepared to combine these two components into a viable process and presents the latest performance projections and economics for a HyS Process coupled to a PBMR heat source. The basic flowsheet for this process has been described elsewhere [4]. It requires an acid concentration section because the SDE product, which is limited to no more than 50% H 2 SO 4 by cell voltage considerations, is too dilute to be fed directly to the bayonet, which needs at least 65% H 2 SO 4 in the feed for acceptable performance. Optimization involved trade-offs between decomposition reaction and acid concentration heat requirements. The PBMR heat source can split its heat output between the decomposition reaction and either steam

Process integration and waste heat recovery in Lithuanian and Danish industry. Final report phase 1

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-08-01

The present document forms the Final Report for the first phase of the project `Process Integration and Waste Heat Recovery in Lithuanian and Danish Industry`. The project is carried out in the period 1995-1998 in a co-operation between the COWI offices in Lyngby and Vilnius, The Technical University of Denmark (Institute for Energetics), Kaunas University of Technology (CIPAI) and Vilnius Technical University, financed by The Danish Ministry of Energy`s EFP-95-programme, Lithuanian Energy Agency as well as the participants. The first phase of the project has comprised the establishment of the CIPAI centre (Centre for Industrial Process Analysis and Integration) at Kaunas University of Technology, training and knowledge transfer as well as elaboration of 6 industrial case-studies within the area of `Process Integration and waste Heat Recovery`. The second phase of the project has comprised R and D activities in this area in order to present general conclusions from the project as well as to present new and improved methods and tools for PI-analysis. The aim of the Final Report for the first phase of the project is to summarise project activities and the achieved results from case-studies and from the operation of the CIPAI-centre in general. (au)

Reducing the Entrainment of Gangue Fines in Low Grade Microcrystalline Graphite Ore Flotation Using Multi-Stage Grinding-Flotation Process

Directory of Open Access Journals (Sweden)

Xiaoqing Weng

2017-03-01

Full Text Available A suitable grinding fineness and flow-sheet could potentially reduce the mechanical entrainment of gangue minerals in the flotation process of microcrystalline graphite. In this study, the suitable grinding fineness of a commercial graphite ore was estimated by mineralogy analysis and laboratory grind-flotation tests. The target grind size of this ore should be 92% passing 74 μm based on the mineralogical evaluation and the flotation performance. A comparison of a single-stage and a three-stage grinding circuit was conducted. Experimental results demonstrated that the three-stage grinding circuit could effectively improve the separation effect, which was attributed to the reduction of slimes. In the end, a more desirable beneficiation result was obtained with the application of three-stage grinding-flotation process by minimizing gangue entrainment.

Membrane/distillation hybrid process research and development. Final report, phase II

Energy Technology Data Exchange (ETDEWEB)

Mazanec, T.J.

1997-07-01

This report covers work conducted under the grant awarded to BP by DOE in late 1991 entitled {open_quotes}Membrane/Distillation Hybrid Process Research and Development.{close_quotes} The program was directed towards development and commercialization of the BP process for separation of vapor phase olefins from non-olefins via facilitated transport using an aqueous facilitator. The program has come to a very successful conclusion, with formation of a partnership between BP and Stone and Webster Engineering Corporation (SWEC) to market and commercialize the technology. The focus of this report is the final portion of the program, during which engineering re-design, facilitator optimization, economic analysis, and marketing have been the primary activities. At the end of Phase II BP was looking to partner with an engineering firm to advance the selective olefin recovery (SOR) technology from the lab/demo stage to full commercialization. In August 1995 BP and SWEC reached an agreement to advance the technology by completing additional Phase III work with DOE and beginning marketing activities.

ALTERNATE REDUCTANT COLD CAP EVALUATION FURNACE PHASE I TESTING

Energy Technology Data Exchange (ETDEWEB)

Johnson, F.; Miller, D.; Zamecnik, J.; Lambert, D.

2014-04-22

Savannah River Remediation (SRR) conducted a Systems Engineering Evaluation (SEE) to determine the optimum alternate reductant flowsheet for the Defense Waste Processing Facility (DWPF). Specifically, two proposed flowsheets (nitric–formic–glycolic and nitric–formic–sugar) were evaluated based upon results from preliminary testing. Comparison of the two flowsheets among evaluation criteria indicated a preference towards the nitric–formic–glycolic flowsheet. Further evaluation of this flowsheet eliminated the formic acid1, and as a result, the nitric–glycolic flowsheet was recommended for further testing. Based on the development of a roadmap for the nitric–glycolic acid flowsheet, Waste Solidification Engineering (WS-E) issued a Technical Task Request (TTR) to address flammability issues that may impact the implementation of this flowsheet. Melter testing was requested in order to define the DWPF flammability envelope for the nitric glycolic acid flowsheet. The Savannah River National Laboratory (SRNL) Cold Cap Evaluation Furnace (CEF), a 1/12th scale DWPF melter, was selected by the SRR Alternate Reductant project team as the melter platform for this testing. The overall scope was divided into the following sub-tasks as discussed in the Task Technical and Quality Assurance Plan (TTQAP): Phase I - A nitric–formic acid flowsheet melter test (unbubbled) to baseline the Cold Cap Evaluation Furnace (CEF) cold cap and vapor space data to the benchmark melter flammability models Phase II - A nitric–glycolic acid flowsheet melter test (unbubbled and bubbled) to: o Define new cold cap reactions and global kinetic parameters for the melter flammability models o Quantify off-gas surging potential of the feed o Characterize off-gas condensate for complete organic and inorganic carbon species Prior to startup, a number of improvements and modifications were made to the CEF, including addition of cameras, vessel support temperature measurement, and a heating

Package Equivalent Reactor Networks as Reduced Order Models for Use with CAPE-OPEN Compliant Simulation

Energy Technology Data Exchange (ETDEWEB)

Meeks, E.; Chou, C. -P.; Garratt, T.

2013-03-31

Engineering simulations of coal gasifiers are typically performed using computational fluid dynamics (CFD) software, where a 3-D representation of the gasifier equipment is used to model the fluid flow in the gasifier and source terms from the coal gasification process are captured using discrete-phase model source terms. Simulations using this approach can be very time consuming, making it difficult to imbed such models into overall system simulations for plant design and optimization. For such system-level designs, process flowsheet software is typically used, such as Aspen Plus® [1], where each component where each component is modeled using a reduced-order model. For advanced power-generation systems, such as integrated gasifier/gas-turbine combined-cycle systems (IGCC), the critical components determining overall process efficiency and emissions are usually the gasifier and combustor. Providing more accurate and more computationally efficient reduced-order models for these components, then, enables much more effective plant-level design optimization and design for control. Based on the CHEMKIN-PRO and ENERGICO software, we have developed an automated methodology for generating an advanced form of reduced-order model for gasifiers and combustors. The reducedorder model offers representation of key unit operations in flowsheet simulations, while allowing simulation that is fast enough to be used in iterative flowsheet calculations. Using high-fidelity fluiddynamics models as input, Reaction Design’s ENERGICO® [2] software can automatically extract equivalent reactor networks (ERNs) from a CFD solution. For the advanced reduced-order concept, we introduce into the ERN a much more detailed kinetics model than can be included practically in the CFD simulation. The state-of-the-art chemistry solver technology within CHEMKIN-PRO allows that to be accomplished while still maintaining a very fast model turn-around time. In this way, the ERN becomes the basis for

The defense waste processing facility: the final processing step for defense high-level waste disposal

International Nuclear Information System (INIS)

Cowan, S.P.; Sprecher, W.M.; Walton, R.D.

1983-01-01

The policy of the U.S. Department of Energy is to pursue an aggressive and credible waste management program that advocates final disposal of government generated (defense) high-level nuclear wastes in a manner consistent with environmental, health, and safety responsibilities and requirements. The Defense Waste Processing Facility (DWPF) is an essential component of the Department's program. It is the first project undertaken in the United States to immobilize government generated high-level nuclear wastes for geologic disposal. The DWPF will be built at the Department's Savannah River Plant near Aiken, South Carolina. When construction is complete in 1989, the DWPF will begin processing the high-level waste at the Savannah River Plant into a borosilicate glass form, a highly insoluble and non-dispersable product, in easily handled canisters. The immobilized waste will be stored on site followed by transportation to and disposal in a Federal repository. The focus of this paper is on the DWPF. The paper discusses issues which justify the project, summarizes its technical attributes, analyzes relevant environmental and insitutional factors, describes the management approach followed in transforming technical and other concepts into concrete and steel, and concludes with observations about the future role of the facility

Development of an advanced continuous mild gasification process for the production of coproducts. Final report

Energy Technology Data Exchange (ETDEWEB)

Merriam, N.W.; Jha, M.C.

1991-11-01

This report is a final brief summary of development of a mild-gasification and char conversion process. Morgantown Energy Technology Center developed a concept called mild gasification. In this concept, devolatilization of coal under nonoxidizing and relatively mild temperature and pressure conditions can yield three marketable products: (1) a high-heating-value gas, (2) a high-aromatic coal liquid, and (3) a high-carbon char. The objective of this program is to develop an advanced, continuous, mild-gasification process to produce products that will make the concept economically and environmentally viable. (VC)

Safety assessment of the liquid-fed ceramic melter process

International Nuclear Information System (INIS)

Buelt, J.L.; Partain, W.L.

1980-08-01

As part of its development program for the solidification of high-level nuclear waste, Pacific Northwest Laboratory assessed the safety issues for a complete liquid-fed ceramic melter (LFCM) process. The LFCM process, an adaption of commercial glass-making technology, is being developed to convert high-level liquid waste from the nuclear fuel cycle into glass. This safety assessment uncovered no unresolved or significant safety problems with the LFCM process. Although in this assessment the LFCM process was not directly compared with other solidification processes, the safety hazards of the LFCM process are comparable to those of other processes. The high processing temperatures of the glass in the LFCM pose no additional significant safety concerns, and the dispersible inventory of dried waste (calcine) is small. This safety assessment was based on the nuclear power waste flowsheet, since power waste is more radioactive than defense waste at the time of solidification, and all accident conditions for the power waste would have greater radiological consequences than those for defense waste. An exhaustive list of possible off-standard conditions and equipment failures was compiled. These accidents were then classified according to severity of consequence and type of accident. Radionuclide releases to the stack were calculated for each group of accidents using conservative assumptions regarding the retention and decontamination features of the process and facility. Two recommendations that should be considered by process designers are given in the safety assessment

Waste Management: An integrated modeling approach for analyzing change in NWC production processes

International Nuclear Information System (INIS)

Christensen, D.C.; Sohn, C.L.; Helm, T.M.; Farish, T.J.; Reid, R.A.

1991-01-01

A problem-driven, integrated modeling, decision-support framework has been conceptualized to aid a team of experts determine the set of evolving technologies that should receive additional developmental support. This conceptual framework utilizes a variety of decision aiding models including Flowsheeting, Analytical Hierarchy Process, Linear and Goal Programming, and Object-Oriented Discrete Event Simulation. A number of the technologies under consideration are strong candidates to overcome current plutonium processing problems so that effective technology will be available for implementation in Complex 21. Complex 21 is a participatory, inter-installation planning effort sponsored by US DOE to consolidate and revitalize the nuclear weapons complex facilities by the 21st century. A computer-based dynamic simulation model has been constructed that will allow testing of alternative combinations of developing technologies. The modeling of new configurations of technologies under a number of different operating conditions and material flow assumptions provides information needed for effective decision making for Complex 21. 4 figs

Clean and efficient energy conversion processes (Cecon-project). Final report

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-12-31

The objectives of the work programme reported are the development and testing of two optimised energy conversion processes, both consisting of a radiant surface gas burner and a ceramic heat exchanger. The first sub-objective of the programme is related to industrial heating, drying and curing processes requireing low and medium heat fluxes. It is estimated that around one tenth of the total EC industrial energy use is associated with such processes. The majority of these processes currently use convection and conduction as the main heat transfer mechanisms and overall energy efficiencies are typically below 25%. For many drying and finishing processes (such as curing powder coatings and drying paints, varnishes, inks, and for the fabrication of paper and textiles), radiant heating can achieve much faster dyring rates and higher energy efficiency than convective heating. In the project new concepts of natural gas fired radiant heating have been investigated which would be much more efficient than the existing processes. One element of the programme was the evelopment of gas burners having enhanced radiant efficiencies. A second concerned the investigation of the safety of gas burners containing significant volumes of mixed gas and air. Finally the new gas burners were tested in combination with the high temperature heat exchanger to create highly efficient radiant heating systems. The second sub-objective concerned the development of a compact low cost heat exchanger capable of achieving high levels of heat recovery (up to 60%) which could be easily installed on industrial processes. This would make heat recovery a practical proposition on processes where existing heat recovery technology is currently not cost effective. The project will have an impact on industrial processes consuming around 80 MTOE of energy per year within EU countries (1 MTOE equals 41.8 PJ). The overall energy saving potential of the project is estimated to be around 22 MTOE which is around 10

Pyro-electrochemical reprocessing of irradiated MOX fast reactor fuel, testing of the reprocessing process with direct MOX fuel production

Energy Technology Data Exchange (ETDEWEB)

Kormilitzyn, M.V.; Vavilov, S.K.; Bychkov, A.V.; Skiba, O.V.; Chistyakov, V.M.; Tselichshev, I.V

2000-07-01

One of the advanced technologies for fast reactor fuel recycle is pyro-electrochemical molten salt technology. In 1998 we began to study the next phase of the irradiated oxide fuel reprocessing new process MOX {yields} MOX. This process involves the following steps: - Dissolution of irradiated fuel in molten alkaline metal chlorides, - Purification of melt from fission products that are co-deposited with uranium and plutonium oxides, - Electrochemical co-deposition of uranium and plutonium oxides under the controlled cathode potential, - Production of granulated MOX (crushing,salt separation and sizing), and - Purification of melt from fission products by phosphate precipitation. In 1998 a series of experiments were prepared and carried out in order to validate this process. It was shown that the proposed reprocessing flowsheet of irradiated MOX fuel verified the feasibility of its decontamination from most of its fission products (rare earths, cesium) and minor-actinides (americium, curium)

Pyro-electrochemical reprocessing of irradiated MOX fast reactor fuel, testing of the reprocessing process with direct MOX fuel production

International Nuclear Information System (INIS)

Kormilitzyn, M.V.; Vavilov, S.K.; Bychkov, A.V.; Skiba, O.V.; Chistyakov, V.M.; Tselichshev, I.V.

2000-01-01

One of the advanced technologies for fast reactor fuel recycle is pyro-electrochemical molten salt technology. In 1998 we began to study the next phase of the irradiated oxide fuel reprocessing new process MOX → MOX. This process involves the following steps: - Dissolution of irradiated fuel in molten alkaline metal chlorides, - Purification of melt from fission products that are co-deposited with uranium and plutonium oxides, - Electrochemical co-deposition of uranium and plutonium oxides under the controlled cathode potential, - Production of granulated MOX (crushing,salt separation and sizing), and - Purification of melt from fission products by phosphate precipitation. In 1998 a series of experiments were prepared and carried out in order to validate this process. It was shown that the proposed reprocessing flowsheet of irradiated MOX fuel verified the feasibility of its decontamination from most of its fission products (rare earths, cesium) and minor-actinides (americium, curium)

Pyrochemistry: from flowsheet to industrial facility

International Nuclear Information System (INIS)

Donaldson, N.; Thied, R.; Lamorlette, G.; Greneche, D.

2001-01-01

Challenges to any future commercial deployment of pyro-chemistry will be significant. The implications of industrial use must be well understood in technical, economic and social terms to gain commercial and regulatory acceptance. The broad base of knowledge necessary to support general commercial use of pyro-chemistry in the nuclear field is considered. Pyro-chemistry development is discussed in the context of a commercial application-based approach and issues to be addressed are outlined. A stepwise evolutionary development of pyro-chemical processing is anticipated which might allow industrialization in the absence of acceptance of evolutionary development at industrial scale which benefited Purex development. (author)

Analysis of reforming process of large distorted ring in final enlarging forging

International Nuclear Information System (INIS)

Miyazawa, Takeshi; Murai, Etsuo

2002-01-01

In the construction of reactors or pressure vessels for oil chemical plants and nuclear power stations, mono block open-die forging rings are often utilized. Generally, a large forged ring is manufactured by means of enlarging forging with reductions of the wall thickness. During the enlarging process the circular ring is often distorted and becomes an ellipse in shape. However the shape control of the ring is a complicated work. This phenomenon makes the matter still worse in forging of larger rings. In order to make precision forging of large rings, we have developed the forging method using a v-shape anvil. The v-shape anvil is geometrically adjusted to fit the distorted ring in the final circle and reform automatically the shape of the ring during enlarging forging. This paper has analyzed the reforming process of distorted ring by computer program based on F.E.M. and examined the effect on the precision of ring forging. (author)

Manual on laboratory testing for uranium ore processing

International Nuclear Information System (INIS)

1990-01-01

Laboratory testing of uranium ores is an essential step in the economic evaluation of uranium occurrences and in the development of a project for the production of uranium concentrates. Although these tests represent only a small proportion of the total cost of a project, their proper planning, execution and interpretation are of crucial importance. The main purposes of this manual are to discuss the objectives of metallurgical laboratory ore testing, to show the specific role of these tests in the development of a project, and to provide practical instructions for performing the tests and for interpreting their results. Guidelines on the design of a metallurgical laboratory, on the equipment required to perform the tests and on laboratory safety are also given. This manual is part of a series of Technical Reports on uranium ore processing being prepared by the IAEA's Division of Nuclear Fuel Cycle and Waste Management. A report on the Significance of Mineralogy in the Development of Flowsheets for Processing Uranium Ores (Technical Reports Series No. 196, 1980) and an instruction manual on Methods for the Estimation of Uranium Ore Reserves (No. 255, 1985) have already been published. 17 refs, 40 figs, 17 tabs

Flotation of zinc and lead oxide minerals from Olkusz region calamine ores

Directory of Open Access Journals (Sweden)

Cichy Krystian

2016-01-01

Full Text Available The paper presents chemical and mineralogical characteristics of calamine ore from the Pomorzany mine. A flowsheet for recovery of sulphide minerals of zinc and lead in the form of the Zn-Pb bulk concentrate was presented. In the following part, preparation of the feed for flotation of Zn-Pb oxide minerals and optimal conditions for separation from it iron sulphide minerals, represented by marcasite, were determined. In the final section the results of flotation of Zn-Pb oxide minerals with anionic collector AM2 belonging to the hydroxyamide group of collectors and a cationic collector in the form of a coconut amine, being a mixture of primary aliphatic amines, were presented. Basing on the obtained results, a technological flowsheet for the recovery of Zn-Pb sulphide and oxide minerals from the calamine ore of the Pomorzany mine was presented.

Process simulation of heavy water plants - a powerful analytical tool

International Nuclear Information System (INIS)

Miller, A.I.

1978-10-01

The commercially conscious designs of Canadian GS (Girdler-Sulphide) have proved sensitive to process conditions. That, combined with the large scale of our units, has meant that computer simulation of their behaviour has been a natural and profitable development. Atomic Energy of Canada Limited has developed a family of steady state simulations to describe all of the Canadian plants. Modelling of plant conditions has demonstrated that the simulation description is very precise and it has become an integral part of the industry's assessments of both plant operation and decisions on capital expenditures. The simulation technique has also found extensive use in detailed designing of both the rehabilitated Glace Bay and the new La Prade plants. It has opened new insights into plant design and uncovered a radical and significant flowsheet change for future designs as well as many less dramatic but valuable lesser changes. (author)

Underground Milling of High-Grade Uranium Ore

Energy Technology Data Exchange (ETDEWEB)

Edwards, C., E-mail: chuck.edwards@amec.com [AMEC Americas Limited, Saskatoon, Saskatchewan (Canada)

2014-05-15

There are many safety and technical issues involved in the mining and progressing of high grade uranium ores such as those exploited in Northern Canada at present. With more of this type of mine due to commence production in the near future, operators have been looking at ways to better manage the situation. The paper describes underground milling of high-grade uranium ore as a means of optimising production costs and managing safety issues. In addition the paper presents some examples of possible process flowsheets and plant layouts that could be applicable to such operations. Finally an assessment of potential benefits from underground milling from a variety of viewpoints is provided. (author)

The MOX Demonstration Facility - the stepping stone to commercial MOX production

International Nuclear Information System (INIS)

Macdonald, A.G.

1994-01-01

The paper provides an insight into MOX fuel and the economic benefits of its use in pressurized water reactors (PWRs). BNFL and AEA are collaborating in the design, construction and operation of a thermal MOX Demonstration Facility (MDF) on the AEA Windscale site in Cumbria. The process flowsheet and equipment employed in MDF are discussed and the special precautions required to handle plutonium bearing materials are highlighted. The process flowsheet includes the short binderless route which has been specially developed for use in MDF and results in fuel pellets with an homogeneous structure. MDF is the forerunner to the design and construction of a larger scale Sellafield MOX Plant and hence is the stepping-stone to commercial MOX production. (author)

Demonstration of a SANEX Process in Centrifugal Contactors using the CyMe{sub 4}-BTBP Molecule on a Genuine Fuel Solution

Energy Technology Data Exchange (ETDEWEB)

Magnusson, D.; Christiansen, B.; Glatz, J.P.; Malmbeck, R.; Serrano-Purroy, D. [European Commiss, Joint Res Ctr, Inst Transuranium Elements, D-76125 Karlsruhe, (Germany); Foreman, M.R.S. [Univ Reading, Dept Chem, Reading RG6 6AD, Berks, (United Kingdom); Geist, A. [Forschungszentrum Karlsruhe, Inst Nukl Entsorgung, D-76021 Karlsruhe, (Germany); Modolo, G. [Forschungszentrum Julich, Inst Energy Res Safety Res and Reactor Technol, D-52425 Julich, (Germany); Sorel, C. [Commissariat Energie Atom Valrho, CEA, DRCP SCPS, F-30207 Bagnols Sur Ceze, (France)

2009-07-01

Efficient recovery of minor actinides from a genuine spent fuel solution has been successfully demonstrated by the CyMe{sub 4}-BTBP/DMDOHEMA extractant mixture dissolved in octanol. The continuous countercurrent process, in which actinides(III) were separated from lanthanides(III), was carried out in laboratory centrifugal contactors using an optimized flow-sheet involving a total of 16 stages. The process was divided into 9 stages for extraction from a 2 M nitric acid feed solution, 3 stages for lanthanide scrubbing, and 4 stages for actinide back-extraction. Excellent feed decontamination factors for Am (7000) and Cm (1000) were obtained and the recoveries of these elements were higher than 99.9%. More than 99.9% of the lanthanides were directed to the raffinate except Gd for which 0.32% was recovered in the product. (authors)

Final workshop proceedings of the collaborative project ''Crystalline ROCK retention processes''

Energy Technology Data Exchange (ETDEWEB)

Rabung, Thomas; Garcia, David; Montoya Vanessa; Molinero, Jorge (eds.)

2014-07-01

The present document is the proceedings of the Final Workshop of the EURATOM FP7 Collaborative Project CROCK (Crystalline Rock Retention Processes). The key driver for initiation the CP CROCK, identified by national Waste Management Organizations, is the undesired high uncertainty and the associated conservatism with respect to the radionuclide transport in the crystalline host-rock far-field around geological disposal of high-level radioactive wastes.

Preparation of a technology development roadmap for the Accelerator Transmutation of Waste (ATW) System : report of the ATW separations technologies and waste forms technical working group

International Nuclear Information System (INIS)

Collins, E.; Duguid, J.; Henry, R.; Karell, E.J.; Laidler, J.J.; McDeavitt, S.M.; Thompson, M.; Toth, L.M.; Williamson, M.; Willit, J.L.

1999-01-01

In response to a Congressional mandate to prepare a roadmap for the development of Accelerator Transmutation of Waste (ATW) technology, a Technical Working Group comprised of members from various DOE laboratories was convened in March 1999 for the purpose of preparing that part of the technology development roadmap dealing with the separation of certain radionuclides for transmutation and the disposal of residual radioactive wastes from these partitioning operations. The Technical Working Group for ATW Separations Technologies and Waste Forms completed its work in June 1999, having carefully considered the technology options available. A baseline process flowsheet and backup process were identified for initial emphasis in a future research, development and demonstration program. The baseline process combines aqueous and pyrochemical processes to permit the efficient separation of the uranium, technetium, iodine and transuranic elements from the light water reactor (LWR) fuel in the head-end step. The backup process is an all- pyrochemical system. In conjunction with the aqueous process, the baseline flowsheet includes a pyrochemical process to prepare the transuranic material for fabrication of the ATW fuel assemblies. For the internal ATW fuel cycle the baseline process specifies another pyrochemical process to extract the transuranic elements, Tc and 1 from the ATW fuel. Fission products not separated for transmutation and trace amounts of actinide elements would be directed to two high-level waste forms, one a zirconium-based alloy and the other a glass/sodalite composite. Baseline cost and schedule estimates are provided for a RD and D program that would provide a full-scale demonstration of the complete separations and waste production flowsheet within 20 years

Preparation of a technology development roadmap for the Accelerator Transmutation of Waste (ATW) System : report of the ATW separations technologies and waste forms technical working group.

Energy Technology Data Exchange (ETDEWEB)

Collins, E.; Duguid, J.; Henry, R.; Karell, E.; Laidler, J.; McDeavitt, S.; Thompson, M.; Toth, M.; Williamson, M.; Willit, J.

1999-08-12

In response to a Congressional mandate to prepare a roadmap for the development of Accelerator Transmutation of Waste (ATW) technology, a Technical Working Group comprised of members from various DOE laboratories was convened in March 1999 for the purpose of preparing that part of the technology development roadmap dealing with the separation of certain radionuclides for transmutation and the disposal of residual radioactive wastes from these partitioning operations. The Technical Working Group for ATW Separations Technologies and Waste Forms completed its work in June 1999, having carefully considered the technology options available. A baseline process flowsheet and backup process were identified for initial emphasis in a future research, development and demonstration program. The baseline process combines aqueous and pyrochemical processes to permit the efficient separation of the uranium, technetium, iodine and transuranic elements from the light water reactor (LWR) fuel in the head-end step. The backup process is an all- pyrochemical system. In conjunction with the aqueous process, the baseline flowsheet includes a pyrochemical process to prepare the transuranic material for fabrication of the ATW fuel assemblies. For the internal ATW fuel cycle the baseline process specifies another pyrochemical process to extract the transuranic elements, Tc and 1 from the ATW fuel. Fission products not separated for transmutation and trace amounts of actinide elements would be directed to two high-level waste forms, one a zirconium-based alloy and the other a glass/sodalite composite. Baseline cost and schedule estimates are provided for a RD&D program that would provide a full-scale demonstration of the complete separations and waste production flowsheet within 20 years.

The processing of bed ashes of fluidized bed boilers to an applicable ingredient for building materials. Het bewerken van bedassen van wervelbedketels tot een geschikte grondstof voor toepassing in bouwprodukten

Energy Technology Data Exchange (ETDEWEB)

Dekker, W

1988-01-01

A study- and test program has been carried out to determine in what way bed ashes of fluidized bed boilers can be processed to applicate the products in building products. The program consisted of selecting applicable ashes; physical-chemical research; slack lime, present in the ashes; grinding and wind-sifting of the ashes; evaluation of the quality of the acquired samples for application in calcium-silicate brick and in mortar; the making of flow-sheets of the processing in the potential demonstration projects. The used sample was a bed ash with active CaO content of 21%. Conclusions were stated and recommendations were made. 6 figs., 6 refs., 9 tabs., 2 app.

Final Regulatory Determination for Special Wastes From Mineral Processing (Mining Waste Exclusion) - Federal Register Notice, June 13, 1991

Science.gov (United States)

This action presents the Agency's final regulatory determination required by section 3001(b)(3)(C) of the Resource Conservation and Recovery Act (RCRA) for 20 special wastes from the processing of ores and minerals.

Peculiarities of highly burned-up NPP SNF reprocessing and new approach to simulation of solvent extraction processes

Energy Technology Data Exchange (ETDEWEB)

Fedorov, Y.S.; Zilberman, B.Y.; Goletskiy, N.D.; Puzikov, E.A.; Ryabkov, D.V.; Rodionov, S.A.; Beznosyuk, V.I.; Petrov, Y.Y.; Saprykin, V.F.; Murzin, A.A.; Bibichev, B.A.; Aloy, A.S.; Kudinov, A.S.; Blazheva, I.V. [RPA ' V.G.Khlopin Radium Institute' , 28, 2 Murinsky av., St-Petersburg, 194 021 (Russian Federation); Kurenkov, N.V. [Institute of Industrial Nuclear Technology NRNU MEPHI, 31, Kashirskoye shosse, Moscow, 115409 (Russian Federation)

2013-07-01

Substantiation, general description and performance characteristics of a reprocessing flowsheet for WWER-1000 spent fuel with burn-up >60 GW*day/t U is given. Pu and U losses were <0.1%, separation factor > 10{sup 4}; their decontamination factor from γ-emitting fission products was 4*10{sup 4} and 3*10{sup 7}, respectively. Zr, Tc, Np removal was >98% at U and Pu losses <0.05%. A new approach to simulation of extraction equilibrium has been developed. It is based on a set of simultaneous chemical reactions characterized by apparent concentration constants. A software package was created for simulation of spent fuel component distribution in multistage countercurrent extraction processes in the presence of salting out agents. (authors)

Change of roles and attitudes in the Swedish localisation process for a final repository

International Nuclear Information System (INIS)

Hedberg, Bjoern

2001-01-01

Since the early research activities in the mid seventies related to a final repository for spent nuclear fuel and other radioactive wastes, much has changed in a direction that allows a more open and transparent decision-making process. Important changes have been noted in the legal framework - including EIA and financing - and in the roles of the Swedish authorities, local politicians, NGO's, and media. Trust and credibility is of course crucial for all actors in the decision-making process, but the ways to gain trust is different depending on which role to play in the process. A higher degree of trust in the different actors, and in the process itself, could be gained from a better distinction between facts and value judgements, but also if the roles of different actors are better clarified. To understand the roles of the different actors, it is important to define each actor's 'arena' in terms of responsibilities, goals, standpoints etc. in several dimensions. These dimensions could for example be geographic or the base for decisions (scientific - political)

Change of roles and attitudes in the Swedish localisation process for a final repository

Energy Technology Data Exchange (ETDEWEB)

Hedberg, Bjoern [Swedish Radiation Protection Inst., Stockholm (Sweden)

2001-07-01

Since the early research activities in the mid seventies related to a final repository for spent nuclear fuel and other radioactive wastes, much has changed in a direction that allows a more open and transparent decision-making process. Important changes have been noted in the legal framework - including EIA and financing - and in the roles of the Swedish authorities, local politicians, NGO's, and media. Trust and credibility is of course crucial for all actors in the decision-making process, but the ways to gain trust is different depending on which role to play in the process. A higher degree of trust in the different actors, and in the process itself, could be gained from a better distinction between facts and value judgements, but also if the roles of different actors are better clarified. To understand the roles of the different actors, it is important to define each actor's 'arena' in terms of responsibilities, goals, standpoints etc. in several dimensions. These dimensions could for example be geographic or the base for decisions (scientific - political)

Engineering evaluation of selective ion-exchange radioactive waste processing at Susquehanna Nuclear Power Plant: Final report

International Nuclear Information System (INIS)

Vance, J.N.

1989-01-01

This final report describes the work performed of an engineering feasibility evaluation of the use and benefits of a selective ion exchange treatment process in the Susquehanna radwaste system. The evaluation addressed operability and processing capability concerns, radiological impacts of operating in the radwaste discharge mode, required hardware modifications to the radwaste and plant make-up systems, impacts on plant water quality limits and impacts on higher waste classifications. An economic analysis is also reported showing the economic benefit of the use of selective ion exchange. 1 ref., 4 figs., 13 tabs

Preliminary technical data summary for the Defense Waste Processing Facility, Stage 1

International Nuclear Information System (INIS)

1980-09-01

This Preliminary Technical Data Summary presents the technical basis for design of Stage 1 of the Staged Defense Waste Processing Facility (DWPF), a process to efficiently immobilize the radionuclides in Savannah River Plant (SRP) high-level liquid waste. The radionuclides in SRP waste are present in sludge that has settled to the bottom of waste storage tanks and in crystallized salt and salt solution (supernate). Stage 1 of the DWPF receives washed, aluminum dissolved sludge from the waste tank farms and immobilizes it in a borosilicate glass matrix. The supernate is retained in the waste tank farms until completion of Stage 2 of the DWPF at which time it filtered and decontaminated by ion exchange in the Stage 2 facility. The decontaminated supernate is concentrated by evaporation and mixed with cement for burial. The radioactivity removed from the supernate is fixed in borosilicate glass along with the sludge. This document gives flowsheets, material, and curie balances, material and curie balance bases, and other technical data for design of the Stage 1 DWPF

Minimization of water consumption under uncertainty for PC process

Energy Technology Data Exchange (ETDEWEB)

Salazar, J.; Diwekar, U.; Zitney, S.

2009-01-01

Integrated gasification combined cycle (IGCC) technology is becoming increasingly important for the development of advanced power generation systems. As an emerging technology different process configurations have been heuristically proposed for IGCC processes. One of these schemes combines water-gas shift reaction and chemical-looping combustion for the CO2 removal prior the fuel gas is fed to the gas turbine reducing its size (improving economic performance) and producing sequestration-ready CO2 (improving its cleanness potential). However, these schemes have not been energetically integrated and process synthesis techniques can be used to obtain optimal flowsheets and designs. This work studies the heat exchange network synthesis (HENS) for the water-gas shift reaction train employing a set of alternative designs provided by Aspen energy analyzer (AEA) and combined in a process superstructure that was simulated in Aspen Plus (AP). For the alternative designs, large differences in the performance parameters (for instance, the utility requirements) predictions from AEA and AP were observed, suggesting the necessity of solving the HENS problem within the AP simulation environment and avoiding the AEA simplifications. A CAPE-OPEN compliant capability which makes use of a MINLP algorithm for sequential modular simulators was employed to obtain a heat exchange network that provided a cost of energy that was 27% lower than the base case.

Final disposal of spent nuclear fuel in Sweden. Some unresolved issues and challenges in the design and implementation of the forthcoming planning and EIA processes

International Nuclear Information System (INIS)

Bjarnadottir, H.; Hilding-Rydevik, T.

2001-06-01

The aim of the study is to highlight some unresolved and challenging issues in the forthcoming approximately six year long Environmental Impact Assessment (EIA) and planning process of the long-term disposal of spent nuclear fuel in Sweden. Different international and Nordic experiences of the processes for final disposal as well as from other development of similar scope, where experiences assumed to be of importance for final disposal of nuclear waste, have been described. Furthermore, issues relating to 'good EIA practice' as well as certain aspects of planning theory have also been presented. The current Swedish situation for the planning and EIA process of the final disposal of spent nuclear fuel was also been summarized. These different 'knowledge areas' have been compared and measured against our perception of the expectations towards the forthcoming process, put forward by different Swedish actors in the field. The result is a presentation of a number of questions and identification issues that the authors consider need special attention in the design and conduction of the planning and EIA process. The study has been realized through a literature survey and followed by reading and analysis of the written material. The main focus of the literature search was on material describing planning processes, actor perspectives and EIA. Material and literature on the technical and scientific aspects of spent nuclear fuel disposal was however deliberately avoided. There is a wealth of international and Swedish literature concerning final disposal of spent nuclear fuel - concerning both technical issues and issues concerning for example public participation and risk perception. But material of a more systematic and comparative nature (relating to both empirical and theoretical issues, and to practical experiences) in relation to EIA processes and communicative planning for final disposal of spent nuclear fuel seems to be more sparsely represented. Our perception of

Test Summary Report Vitrification Demonstration of an Optimized Hanford C-106/AY-102 Waste-Glass Formulation

International Nuclear Information System (INIS)

Goles, Ronald W.; Buchmiller, William C.; Hymas, Charles R.; MacIsaac, Brett D.

2002-01-01

In order to further the goal of optimizing Hanford?s HLW borosilicate flowsheet, a glass formulation effort was launched to develop an advanced high-capacity waste form exhibiting acceptable leach and crystal formation characteristics. A simulated C-106/AY-102 waste envelop inclusive of LAW pretreatment products was chosen as the subject of these nonradioactive optimization efforts. To evaluate this optimized borosilicate waste formulation under continuous dynamic vitrification conditions, a research-scale Joule-heated ceramic melter was used to demonstrate the advanced waste form?s flowsheet. The main objectives of this melter test was to evaluate (1) the processing characteristics of the newly formulated C-106/AY-102 surrogate melter-feed stream, (2) the effectiveness of sucrose as a glass-oxidation-state modifier, and (3) the impact of this reductant upon processing rates

Airborne effluent control for LMFBR fuel reprocessing plants

International Nuclear Information System (INIS)

Yarbro, O.O.; Groenier, W.S.; Stephenson, M.J.

1976-01-01

A significant part of the LMFBR fuel reprocessing development program has been devoted to the development of efficient removal systems for the volatile fission products, including 131 I, krypton, tritium, 129 I, and most recently 14 C. Flowsheet studies have indicated that very significant reductions of radioactive effluents can be achieved by integrating advanced effluent control systems with new concepts of containment and ventilation; however, the feasibility of such has not yet been established, nor have the economics been examined. This paper presents a flowsheet for the application of advanced containment systems to the processing of LMFBR fuels and summarizes the status and applicability of specific fission product removal systems

Development and evaluation of the process for final placement application: a review of the new student led allocation system

OpenAIRE

Mason, Rachael; Brackenbury, Debra; Broady, Sophie

2016-01-01

Background A process to facilitate nursing students to have more ownership of their final placement was introduced for this academic year by inviting them to apply for a specific placement they felt most appropriate. Whilst there has been significant research into preparing students for practice (Woods et al, 2015) and to explore the transition from student to graduate nurse (Kumaran and Carney, 2014), there is little to explore the effect of gaining preference for their final placement or...

Biomass steam gasification for production of SNG – Process design and sensitivity analysis

International Nuclear Information System (INIS)

Gröbl, Thomas; Walter, Heimo; Haider, Markus

2012-01-01

Highlights: ► A model for the SNG-production process from biomass to raw-SNG is prepared. ► A thermodynamic equilibrium model of the Biomass-Heatpipe-Reformer is developed. ► A sensitivity analysis on the most important operation parameters is carried out. ► Adopting the steam excess ratio a syngas ideally suitable for SNG production is generated. ► Thermodynamic equilibrium models are a useful tool for process design. -- Abstract: A process design for small-scale production of Substitute Natural Gas (SNG) by steam gasification of woody biomass is performed. In the course of this work, thermodynamic models for the novel process steps are developed and implemented into an already existing model library of commercial process simulation software IPSEpro. Mathematical models for allothermal steam gasification of biomass as well as for cleaning and methanation of product gas are provided by applying mass balances, energy balances and thermodynamic equilibrium equations. Using these models the whole process is integrated into the simulation software, a flowsheet for an optimum thermal integration of the single process steps is determined and energy savings are identified. Additionally, a sensitivity study is carried out in order to analyze the influence of various operation parameters. Their effects on amount and composition of the product gas and process efficiency are evaluated and discussed within this article.

Hydrogen production processes; Procedes de production d'hydrogene

Energy Technology Data Exchange (ETDEWEB)

NONE

2003-07-01

The goals of this first Gedepeon workshop on hydrogen production processes are: to stimulate the information exchange about research programs and research advances in the domain of hydrogen production processes, to indicate the domains of interest of these processes and the potentialities linked with the coupling of a nuclear reactor, to establish the actions of common interest for the CEA, the CNRS, and eventually EDF, that can be funded in the framework of the Gedepeon research group. This document gathers the slides of the 17 presentations given at this workshop and dealing with: the H{sub 2} question and the international research programs (Lucchese P.); the CEA's research program (Lucchese P., Anzieu P.); processes based on the iodine/sulfur cycle: efficiency of a facility - flow-sheets, efficiencies, hard points (Borgard J.M.), R and D about the I/S cycle: Bunsen reaction (Colette S.), R and D about the I/S cycle: the HI/I{sub 2}/H{sub 2}O system (Doizi D.), demonstration loop/chemical engineering (Duhamet J.), materials and corrosion (Terlain A.); other processes under study: the Westinghouse cycle (Eysseric C.), other processes under study at the CEA (UT3, plasma,...) (Lemort F.), database about thermochemical cycles (Abanades S.), Zn/ZnO cycle (Broust F.), H{sub 2} production by cracking, high temperature reforming with carbon trapping (Flamant G.), membrane technology (De Lamare J.); high-temperature electrolysis: SOFC used as electrolyzers (Grastien R.); generic aspects linked with hydrogen production: technical-economical evaluation of processes (Werkoff F.), thermodynamic tools (Neveu P.), the reactor-process coupling (Aujollet P.). (J.S.)

Proposal for a new CAPE-OPEN Object Model

Science.gov (United States)

Process simulation applications require the exchange of significant amounts of data between the flowsheet environment, unit operation model, and thermodynamic server. Packing and unpacking various data types and exchanging data using structured text-based architectures, including...

Computational techniques used in the development of coprocessing flowsheets

International Nuclear Information System (INIS)

Groenier, W.S.; Mitchell, A.D.; Jubin, R.T.

1979-01-01

The computer program SEPHIS, developed to aid in determining optimum solvent extraction conditions for the reprocessing of nuclear power reactor fuels by the Purex method, is described. The program employs a combination of approximate mathematical equilibrium expressions and a transient, stagewise-process calculational method to allow stage and product-stream concentrations to be predicted with accuracy and reliability. The possible applications to inventory control for nuclear material safeguards, nuclear criticality analysis, and process analysis and control are of special interest. The method is also applicable to other counntercurrent liquid--liquid solvent extraction processes having known chemical kinetics, that may involve multiple solutes and are performed in conventional contacting equipment

Conversion of highly active waste to solids

International Nuclear Information System (INIS)

Scheffler, K.

Borosilicate glasses were selected as matrix material for solidification of highly radioactive wastes. Current laboratory work on the VERA process is described. Goals were met by a five-component glass VG-38 and a glass-ceramic VC-15. The VERA process is described: flowsheet, denitration, calcinator, fusion facility

Pyrolysis of tyres. Influence of the final temperature of the process on emissions and the calorific value of the products recovered

International Nuclear Information System (INIS)

Diez, C.; Martinez, O.; Calvo, L.F.; Cara, J.; Moran, A.

2004-01-01

A study was made of the pyrolysis of tyre particles, with the aim of determining the possibilities of using the products resulting from the process as fuel. Three final temperatures were used, determined from thermogravimetric data. The design of the experiment was a horizontal oven containing a reactor into which particles of the original tyre were placed. After the process, a solid fraction (char) remained in the reactor, while the gases generated went through a set of scrubbers where most of the condensable fraction (oils) was retained. Finally, once free of this fraction, the gases were collected in glass ampoules. Solid and liquids fractions were subjected to thermogravimetric analyses in order to study their combustibility. The gas fraction was analysed by means of gas chromatography to establish the content of CO, CO 2 , H 2 and hydrocarbons present in the samples (mainly components of gases produced in the pyrolysis process). A special study was made of the sulphur and chlorine content of all the fractions, as the presence of these elements could be problematic if the products are used as fuel. Tyre pyrolysis engenders a solid carbon residue that concentrates sulphur and chorine, with a relatively high calorific value, although not so high as that of the original tyre. The liquid fraction produced by the process has a high calorific value, which rises with the final temperature, up to 40 MJ/kg. The chlorine content of this fraction is negligible. Over 95% of the gas fraction, regardless of the final temperature, is composed of hydrocarbons of a low molecular weight and hydrogen, this fraction also appearing to be free of chlorine

System for data acquisition and processing on the base of the minicomputers and CAMAC interfaces in the experiments on the L-2 stellarator

International Nuclear Information System (INIS)

Blokh, M.A.; Kamolova, T.I.; Kutsenko, A.V.; Kutsenko, V.A.; Nechaev, YU.I.; Fedyanin, O.I.; Shelobkov, V.I.

1983-01-01

The system for data acquisition and processing intended for automation of experiments on the L-2 stellarator is described. Hardware peculiarities and sofrware flowsheet are considered. The system is realized on the base of the TRAI minicomputer and CAMAC modules. The system provides data input from diagnostic sensors into the computer memory during the stellarator operational pulse and preliminary data processing in the interval between stellarator pulses, putout of the results a display device or a printer. For programming the Focal language is chosen. CAMAC module control and organization of the whole numbers massive for experimental data storage is realized by means of new functions written in Assembler. The system successfully operates since 1976. In 1978 the system is switched on through the CAMAC interfaces to the EC computer in order to provide the long-term information storage

Final disposal of spent nuclear fuel in Sweden. Some unresolved issues and challenges in the design and implementation of the forthcoming planning and EIA processes

Energy Technology Data Exchange (ETDEWEB)

Bjarnadottir, H.; Hilding-Rydevik, T. [Nordregio, Stockholm (Sweden)

2001-06-01

The aim of the study is to highlight some unresolved and challenging issues in the forthcoming approximately six year long Environmental Impact Assessment (EIA) and planning process of the long-term disposal of spent nuclear fuel in Sweden. Different international and Nordic experiences of the processes for final disposal as well as from other development of similar scope, where experiences assumed to be of importance for final disposal of nuclear waste, have been described. Furthermore, issues relating to 'good EIA practice' as well as certain aspects of planning theory have also been presented. The current Swedish situation for the planning and EIA process of the final disposal of spent nuclear fuel was also been summarized. These different 'knowledge areas' have been compared and measured against our perception of the expectations towards the forthcoming process, put forward by different Swedish actors in the field. The result is a presentation of a number of questions and identification issues that the authors consider need special attention in the design and conduction of the planning and EIA process. The study has been realized through a literature survey and followed by reading and analysis of the written material. The main focus of the literature search was on material describing planning processes, actor perspectives and EIA. Material and literature on the technical and scientific aspects of spent nuclear fuel disposal was however deliberately avoided. There is a wealth of international and Swedish literature concerning final disposal of spent nuclear fuel - concerning both technical issues and issues concerning for example public participation and risk perception. But material of a more systematic and comparative nature (relating to both empirical and theoretical issues, and to practical experiences) in relation to EIA processes and communicative planning for final disposal of spent nuclear fuel seems to be more sparsely represented

Progress report for 1983/84 from the Waste Treatment and Disposal Working Party covering joint BNFL/DOE funded work

International Nuclear Information System (INIS)

Higson, S.G.

1984-01-01

The subject is covered in paragraphs: introduction (arisings of intermediate-level radioactive waste); organisation and role of the Waste Treatment and Disposal Working Party; main objectives (to provide data on intermediate-level waste treatment systems and allow assessment of alternative processes); ILW process and flowsheeting studies; ILW product evaluation. (U.K.)

Alkaline-Side Extraction of Cesium from Savannah River Tank Waste Using a Calixarene-Crown Ether Extractant

Energy Technology Data Exchange (ETDEWEB)

Bonnesen, P.V.; Delmau, L.H.; Haverlock, T.J.; Moyer, B.A.

1998-12-01

Results are presented supporting the viability of the alkaline-side CSEX process as a potential replacement for the In-Tank Precipitation process for removal of cesium from aqueous high-level waste (HLW) at the Savannah River Site (SRS). Under funding from the USDOE Efficient Separations and Crosscutting program, a flowsheet was suggested in early June of 1998, and in the following four months, this flowsheet underwent extensive testing, both in batch tests at ORNL and ANL and in two centrifugal-contactor tests at ANL. To carry out these tests, the initial ESP funding was augmented by direct funds from Westinghouse Savannah River Corporation. The flowsheet employed a solvent containing a calixarene-crown hybrid compound called BoBCalixC6 that was invented at ORNL and can now be obtained commercially for government use from IBC Advanced Technologies. This special extractant is so powerful and selective that it can be used at only 0.01 M, compensating for its expense, but a modifier is required for use in an aliphatic diluent, primarily to increase the cesium distribution ratio D{sub Cs} in extraction. The modifier selected is a relatively economical fluorinated alcohol called Cs3, invented at ORNL and so far available. only from ORNL. For the flowsheet, the modifier is used at 0.2 M in the branched aliphatic kerosene Isopar{reg_sign} L. Testing at ORNL and ANL involved simulants of the SRS HLW. After extraction of the Cs from the waste simulant, the solvent is scrubbed with 0.05 M HNO{sub 3} and stripped with a solution comprised of 0.0005 M HNO{sub 3} and 0.0001 M CsNO{sub 3}. The selection of these conditions is justified in this report, both on the basis of experimental data and underlying theory.

Evaluation of the Hydraulic Performance and Mass Transfer Efficiency of the CSSX Process with the Optimized Solvent in a Single Stage of 5.5-Cm Diameter Centrifugal Contactor

International Nuclear Information System (INIS)

Law, J.D.; Tillotson, R.D.; Todd, T.A.

2002-01-01

The Caustic-Side Solvent Extraction (CSSX) process has been selected for the separation of cesium from Savannah River Site high-level waste. The solvent composition used in the CSSX process was recently optimized so that the solvent is no longer supersaturated with respect to the calixarene crown ether extractant. Hydraulic performance and mass transfer efficiency testing of a single stage of 5.5-cm ORNL-designed centrifugal contactor has been performed for the CSSX process with the optimized solvent. Maximum throughputs of the 5.5-cm centrifugal contactor, as a function of contactor rotor speed, have been measured for the extraction, scrub, strip, and wash sections of the CSSX flowsheet at the baseline organic/aqueous flow ratios (O/A) of the process, as well as at O/A's 20% higher and 20% lower than the baseline. Maximum throughputs are comparable to the design throughput of the contactor, as well as with throughputs obtained previously in a 5-cm centrifugal contactor with the non-optimized CSSX solvent formulation. The 20% variation in O/A had minimal effect on contactor throughput. Additionally, mass transfer efficiencies have been determined for the extraction and strip sections of the flowsheet. Efficiencies were lower than the process goal of greater than or equal to 80%, ranging from 72 to 75% for the extraction section and from 36 to 60% in the strip section. Increasing the mixing intensity and/or the solution level in the mixing zone of the centrifugal contactor (residence time) could potentially increase efficiencies. Several methods are available to accomplish this including (1) increasing the size of the opening in the bottom of the rotor, resulting in a contactor which is partially pumping instead of fully pumping, (2) decreasing the number of vanes in the contactor, (3) increasing the vane height, or (4) adding vanes on the rotor and baffles on the housing of the contactor. The low efficiency results obtained stress the importance of proper design of

LITERATURE REVIEW FOR OXALATE OXIDATION PROCESSES AND PLUTONIUM OXALATE SOLUBILITY

Energy Technology Data Exchange (ETDEWEB)

Nash, C.

2012-02-03

A literature review of oxalate oxidation processes finds that manganese(II)-catalyzed nitric acid oxidation of oxalate in precipitate filtrate is a viable and well-documented process. The process has been operated on the large scale at Savannah River in the past, including oxidation of 20 tons of oxalic acid in F-Canyon. Research data under a variety of conditions show the process to be robust. This process is recommended for oxalate destruction in H-Canyon in the upcoming program to produce feed for the MOX facility. Prevention of plutonium oxalate precipitation in filtrate can be achieved by concentrated nitric acid/ferric nitrate sequestration of oxalate. Organic complexants do not appear practical to sequester plutonium. Testing is proposed to confirm the literature and calculation findings of this review at projected operating conditions for the upcoming campaign. H Canyon plans to commence conversion of plutonium metal to low-fired plutonium oxide in 2012 for eventual use in the Mixed Oxide Fuel (MOX) Facility. The flowsheet includes sequential operations of metal dissolution, ion exchange, elution, oxalate precipitation, filtration, and calcination. All processes beyond dissolution will occur in HB-Line. The filtration step produces an aqueous filtrate that may have as much as 4 M nitric acid and 0.15 M oxalate. The oxalate needs to be removed from the stream to prevent possible downstream precipitation of residual plutonium when the solution is processed in H Canyon. In addition, sending the oxalate to the waste tank farm is undesirable. This report addresses the processing options for destroying the oxalate in existing H Canyon equipment.

Final Report on the Audit of the Administration of the Contract Closeout Process at the Defense Contract Management Region, Dallas

Science.gov (United States)

1990-09-18

This is our final report on the Audit of the Administration of the Contract Closeout Process at the Defense Contract Management Region, Dallas (DCMR... audit was made from January to October 1989. The objectives of the audit were to determine the timeliness of the contract closeout process, the validity...As part of the audit , we also evaluated internal controls over the contract closeout process. As of December 31, 1988, the Contract Administration

Development on the cryogenic hydrogen isotopes distillation process technology for tritium removal (Final report)

International Nuclear Information System (INIS)

Sung, Ki Woung; Kim, Yong Ik; Na, Jeong Won; Ku, Jae Hyu; Kim, Kwang Rak; Jeong, Yong Won; Lee, Han Soo; Cho, Young Hyun; Ahn, Do Hee; Baek, Seung Woo; Kang, Hee Seok; Kim, You Sun

1995-12-01

While tritium exposure to the site-workers in Wolsung NPP is up to about 40% of the total personnel exposure, Ministry of Science and Technology has asked tritium removal facility for requirement of post heavy-water reactor construction. For the purpose of essential removal of tritium from the Wolsung heavy-water reactor system, a preliminary study on the cryogenic Ar-N 2 and H 2 -D 2 distillation process for development of liquid-phase catalytic exchange cryogenic hydrogen distillation process technology. The Ar-N 2 distillation column showed good performance with approximately 97% of final Ar concentration, and a computer simulation code was modified using these data. A simulation code developed for cryogenic hydrogen isotopes (H 2 , HD, D 2 , HT, DT, T 2 ) distillation column showed good performance after comparison with the result of a JAERI code, and a H 2 -D 2 distillation column was made. Gas chromatography for hydrogen isotopes analysis was established using a vacuum sampling loop, and a schematic diagram of H 2 -D 2 distillation process was suggested. A feasibility on modification of H 2 -D 2 distillation process control system using Laser Raman Spectroscopy was studied, and the consideration points for tritium storage system for Wolsung tritium removal facility was suggested. 31 tabs., 79 figs., 68 refs. (Author)

Uranium recovery from phosphate rocks concentrated

International Nuclear Information System (INIS)

Azevedo, M.F. de.

1986-01-01

The reserves, geological data, chemical data and technical flowsheet from COPEBRAS and Goiasfertil ores are described, including the process of mining ore concentration. Samples of Goiasfertil ores are analysed by gravimetric analysis, for phosphate, and spectrofluorimetry for uranium. (author)

Coal demonstration plants. Quarterly report, April-June 1979

Energy Technology Data Exchange (ETDEWEB)

None

1980-04-01

The objective of the US DOE demonstration program is to demonstrate and verify second-generation technologies and validate the economic, environmental and productive capacity of a near commercial-size plant by integrating and operating a modular unit using commercial size equipment. These facilities are the final stage in the RD and D process aimed at accelerating and reducing the risks of industrial process implementation. Under the DOE program, contracts for the design, construction, and operation of the demonstration plants are awarded through competitive procedures and are cost shared with the industrial partner. The conceptual design phase is funded by the government, with the detailed design, procurement, construction, and operation phases being co-funded between industry and the government. The government share of the cost involved for a demonstration plant depends on the plant size, location, and the desirability and risk of the process to be demonstrated. The various plants and programs are discussed: Description and status, funding, history, flowsheet and progress during the current quarter. (LTN)

Review of thorium fuel reprocessing experience

International Nuclear Information System (INIS)

Brooksbank, R.E.; McDuffee, W.T.; Rainey, R.H.

1978-01-01

The review reveals that experience in the reprocessing of irradiated thorium materials is limited. Plants that have processed thorium-based fuels were not optimized for the operations. Previous demonstrations of several viable flowsheets provide a sound technological base for the development of optimum reprocessing methods and facilities. In addition to the resource benefit by using thorium, recent nonproliferation thrusts have rejuvenated an interest in thorium reprocessing. Extensive radiation is generated as the result of 232 U-contamination produced in the 233 U, resulting in the remote operation and fabrication operations and increased fuel cycle costs. Development of the denatured thorium flowsheet, which is currently of interest because of nonproliferation concerns, represents a difficult technological challenge

HTGR fuel reprocessing technology

International Nuclear Information System (INIS)

Brooks, L.H.; Heath, C.A.; Shefcik, J.J.

1976-01-01

The following aspects of HTGR reprocessing technology are discussed: characteristics of HTGR fuels, criteria for a fuel reprocessing flowsheet; selection of a reference reprocessing flowsheet, and waste treatment

Validation Testing of the Nitric Acid Dissolution Step Within the K Basin Sludge Pretreatment Process

International Nuclear Information System (INIS)

AJ Schmidt; CH Delegard; KL Silvers; PR Bredt; CD Carlson; EW Hoppe; JC Hayes; DE Rinehart; SR Gano; BM Thornton

1999-01-01

The work described in this report involved comprehensive bench-scale testing of nitric acid (HNO 3 ) dissolution of actual sludge materials from the Hanford K East (KE) Basin to confirm the baseline chemical pretreatment process. In addition, process monitoring and material balance information was collected to support the development and refinement of process flow diagrams. The testing was performed by Pacific Northwest National Laboratory (PNNL)for the US Department of Energy's Office of Spent Fuel Stabilization (EM-67) and Numatec Hanford Corporation (NHC) to assist in the development of the K Basin Sludge Pretreatment Process. The baseline chemical pretreatment process for K Basin sludge is nitric acid dissolution of all particulate material passing a 1/4-in. screen. The acid-insoluble fraction (residual solids) will be stabilized (possibly by chemical leaching/rinsing and grouting), packaged, and transferred to the Hanford Environmental Restoration Disposal Facility (ERDF). The liquid fraction is to be diluted with depleted uranium for uranium criticality safety and iron nitrate for plutonium criticality safety, and neutralized with sodium hydroxide. The liquid fraction and associated precipitates are to be stored in the Hanford Tank Waste Remediation Systems (TWRS) pending vitrification. It is expected that most of the polychlorinated biphenyls (PCBs), associated with some K Basin sludges, will remain with the residual solids for ultimate disposal to ERDF. Filtration and precipitation during the neutralization step will further remove trace quantities of PCBs within the liquid fraction. The purpose of the work discussed in this report was to examine the dissolution behavior of actual KE Basin sludge materials at baseline flowsheet conditions and validate the.dissolution process step through bench-scale testing. The progress of the dissolution was evaluated by measuring the solution electrical conductivity and concentrations of key species in the dissolver

Validation Testing of the Nitric Acid Dissolution Step Within the K Basin Sludge Pretreatment Process

Energy Technology Data Exchange (ETDEWEB)

AJ Schmidt; CH Delegard; KL Silvers; PR Bredt; CD Carlson; EW Hoppe; JC Hayes; DE Rinehart; SR Gano; BM Thornton

1999-03-24

The work described in this report involved comprehensive bench-scale testing of nitric acid (HNO{sub 3}) dissolution of actual sludge materials from the Hanford K East (KE) Basin to confirm the baseline chemical pretreatment process. In addition, process monitoring and material balance information was collected to support the development and refinement of process flow diagrams. The testing was performed by Pacific Northwest National Laboratory (PNNL)for the US Department of Energy's Office of Spent Fuel Stabilization (EM-67) and Numatec Hanford Corporation (NHC) to assist in the development of the K Basin Sludge Pretreatment Process. The baseline chemical pretreatment process for K Basin sludge is nitric acid dissolution of all particulate material passing a 1/4-in. screen. The acid-insoluble fraction (residual solids) will be stabilized (possibly by chemical leaching/rinsing and grouting), packaged, and transferred to the Hanford Environmental Restoration Disposal Facility (ERDF). The liquid fraction is to be diluted with depleted uranium for uranium criticality safety and iron nitrate for plutonium criticality safety, and neutralized with sodium hydroxide. The liquid fraction and associated precipitates are to be stored in the Hanford Tank Waste Remediation Systems (TWRS) pending vitrification. It is expected that most of the polychlorinated biphenyls (PCBs), associated with some K Basin sludges, will remain with the residual solids for ultimate disposal to ERDF. Filtration and precipitation during the neutralization step will further remove trace quantities of PCBs within the liquid fraction. The purpose of the work discussed in this report was to examine the dissolution behavior of actual KE Basin sludge materials at baseline flowsheet conditions and validate the.dissolution process step through bench-scale testing. The progress of the dissolution was evaluated by measuring the solution electrical conductivity and concentrations of key species in the

Laboratory Tests on Post-Filtration Precipitation in the WTP Pretreatment Process

International Nuclear Information System (INIS)

Russell, Renee L.; Peterson, Reid A.; Rinehart, Donald E.; Crum, Jarrod V.

2009-01-01

Pacific Northwest National Laboratory (PNNL) has been tasked by Bechtel National Inc. (BNI) on the River Protection Project-Hanford Tank Waste Treatment and Immobilization Plant (RPP-WTP) project to perform research and development activities to resolve technical issues identified for the Pretreatment Facility (PTF). The Pretreatment Engineering Platform (PEP) was designed, constructed, and operated as part of a plan to respond to issue M12, 'Undemonstrated Leaching Processes,' of the External Flowsheet Review Team (EFRT) issue response plan (Barnes et al. 2006). The PEP is a 1/4.5-scale test platform designed to simulate the WTP pretreatment caustic leaching, oxidative leaching, ultrafiltration solids concentration, and slurry washing processes. The PEP replicates the WTP leaching processes using prototypic equipment and control strategies. A simplified flow diagram of the PEP system is shown in Figure 1.1. Two operating scenarios are currently being evaluated for the ultrafiltration process (UFP) and leaching operations. The first scenario has caustic leaching performed in the UFP-2 ultrafiltration feed vessels (i.e., vessel UFP-VSL-T02A in the PEP; and vessels UFP-VSL-00002A and B in the WTP PTF). The second scenario has caustic leaching conducted in the UFP-1 ultrafiltration feed preparation vessels (i.e., vessels UFP-VSL-T01A and B in the PEP; vessels UFP-VSL-00001A and B in the WTP PTF).

BCQ+: a body constitution questionnaire to assess Yang-Xu. Part I: establishment of a first final version through a Delphi process.

Science.gov (United States)

Su, Yi-Chang; Chen, Li-Li; Lin, Jun-Dai; Lin, Jui-Shan; Huang, Yi-Chia; Lai, Jim-Shoung

2008-12-01

Assessing an individual's level of Yang deficiency (Yang-Xu) by its manifestations is a frequent issue in traditional Chinese medicine (TCM) clinical trials. To this end, an objective, reliable and rigorous diagnostic tool is required. This study aimed to develop a first final version of the Yang-Xu Constitution Questionnaire. We conducted 3 steps to develop such an objective measurement tool: 1) the research team was formed and a panel of 26 experts was selected for the Delphi process; 2) items for the questionnaire were generated by literature review and a Delphi process; items were reworded into colloquial questions; face and content validity of the items were evaluated through a Delphi process again; 3) the difficulty of the questionnaire was evaluated in a pilot study with 81 subjects aged 20-60 years. The literature review retrieved 35 relevant items which matched the definition of 'constitution' and 'Yang-Xu'. After a first Delphi process, 22 items were retained and translated into colloquial questions. According to the second part of the Delphi process, the content validity index of each of the 22 questions ranged between 0.85-1. These 22 questions were evaluated by 81 subjects, 2 questions that were hard to tell the difference were combined; 3 questions were modified after the research team had discussed the participants' feedback. Finally, the questionnaire was established with 21 questions. This first final version of a questionnaire to assess Yang-Xu constitution with considerable face and content validity may serve as a basis to develop an advanced Yang-Xu questionnaire. 2008 S. Karger AG, Basel.

Utilities and offsites design baseline. Outside Battery Limits Facility 6000 tpd SRC-I Demonstration Plant. Volume 2

Energy Technology Data Exchange (ETDEWEB)

None

1984-05-25

Volume 2 contains flowsheets and equipment specifications for the following parts of the plant: cooling water systems, process water supply, potable water supply, nitrogen system, compressed air system, flares, incinerators, fuels and interconnecting systems (pipes). The instrumentation requirements are included. (LTN)

Coal Technology Program progress report for April 1976

Energy Technology Data Exchange (ETDEWEB)

1976-06-01

In the Hydrocarbonization Research program, two successful experiments were completed in the bench-scale hydrocarbonizer. A settling test at a lower temperature (390/sup 0/F) using 20 percent toluene in Solvent Refined Coal (SRC) Unfiltered Oil (UFO) produced a 30 percent clarified product in 2 hr. Characterization tests include distillation curves for Wilsonville's SRC-UFO and a particle size distribution of Pittsburg and Midway Coal Mining Company's (PAMCO) SRC-UFO. Studies of intermediate-temperature pyrolysis of large blocks have been maintained with char samples continuing to demonstrate pyrophoricity, even after heating to 700/sup 0/C. Simulated distillation analysis of tars produced by the last eight experiments are being compared with those performed at Laramie upon tars produced by the Hanna No. 2 experiment. In Coal-Fueled MIUS, stainless steel tubing to be used in one of the furnace tube bundles was ordered and the bid package for the furnace completed. Tests continued on the coal feed system and with the cold flow fluidized bed model. For the Synthoil process, flow diagrams, material balances, and utilities requirements were completed for the entire facility. For the Hydrocarbonization process, flowsheets were reviewed for compatibility; equipment lists were brought up to date; and utilities requirements were compiled from the individual flowsheets. The char recovery and storage subsystem flowsheet was completed. (auth)

Surface and subsurface cleanup protocol for radionuclides, Gunnison, Colorado, UMTRA project processing site: Final

International Nuclear Information System (INIS)

1994-01-01

Thorium 230 (Th-230) at the Gunnison, Colorado processing site will require remediation, however, a seasonally fluctuating groundwater table at the site significantly complicates conventional remedial action with respect to cleanup. Therefore, to effectively remediate the site with respect to Radium 226 (Ra-226) and Th-230, the following supplemental standard is proposed: In situ Ra-26 will be remediated to the EPA soil cleanup standards independent of groundwater considerations. In situ Th-230 concentrations will be remediated in the region above the encountered water table so the 1000-year projected Ra-226 concentration complies with the EPA soil cleanup concentration limits. If elevated Th-230 persists to the water table, an additional foot of excavation will be performed and the grid will be backfilled. Excavated grids will be backfilled to the final remedial action grade with clean cobbly soil. Final grid verification that is required below the water table will be performed by extracting and analyzing a single bulk soil sample with the bucket of a backhoe. Modeled surface radon flux values will be estimated and documented. A recommendation will be made that land records should be annotated to identify the presence of residual Th-230

Intensification of the Reverse Cationic Flotation of Hematite Ores with Optimization of Process and Hydrodynamic Parameters of Flotation Cell

Science.gov (United States)

Poperechnikova, O. Yu; Filippov, L. O.; Shumskaya, E. N.; Filippova, I. V.

2017-07-01

The demand of high grade iron ore concentrates is a major issue due to the depletion of rich iron-bearing ores and high competitiveness in the iron ore market. Iron ore production is forced out to upgrade flowsheets to decrease the silica content in the pelettes. Different types of ore have different mineral composition and texture-structural features which require different mineral processing methods and technologies. The paper presents a comparative study of the cationic and anionic flotation routes to process a fine-grain oxidized iron ore. The modified carboxymethyl cellulose was found as the most efficient depressant in reverse cationic flotation. The results of flotation optimization of hematite ores using matrix of second-order center rotatable uniform design allowed to define the collector concentration, impeller rotation speed and air flowrate as the main flotation parameters impacting on the iron ore concentrate quality and iron recovery in a laboratory flotation machine. These parameters have been selected as independent during the experiments.

Electrochemical Processes for In-Situ Treatment of Contaminated Soils - Final Report - 09/15/1996 - 01/31/2001

Energy Technology Data Exchange (ETDEWEB)

Huang, Chin-Pao

2001-05-31

This project will study electrochemical processes for the in situ treatment of soils contaminated by mixed wastes, i.e., organic and inorganic. Soil samples collected form selected DOE waste sites will be characterized for specific organic and metal contaminants and hydraulic permeability. The soil samples are then subject to desorption experiments under various physical-chemical conditions such as pH and the presence of surfactants. Batch electro-osmosis experiments will be conducted to study the transport of contaminants in the soil-water systems. Organic contaminants that are released from the soil substrate will be treated by an advanced oxidation process, i.e., electron-Fantan. Finally, laboratory reactor integrating the elector-osmosis and elector-Fantan processes will be used to study the treatment of contaminated soil in situ.

Simulation of the Press Hardening Process and Prediction of the Final Mechanical Material Properties

Science.gov (United States)

Hochholdinger, Bernd; Hora, Pavel; Grass, Hannes; Lipp, Arnulf

2011-08-01

Press hardening is a well-established production process in the automotive industry today. The actual trend of this process technology points towards the manufacturing of parts with tailored properties. Since the knowledge of the mechanical properties of a structural part after forming and quenching is essential for the evaluation of for example the crash performance, an accurate as possible virtual assessment of the production process is more than ever necessary. In order to achieve this, the definition of reliable input parameters and boundary conditions for the thermo-mechanically coupled simulation of the process steps is required. One of the most important input parameters, especially regarding the final properties of the quenched material, is the contact heat transfer coefficient (IHTC). The CHTC depends on the effective pressure or the gap distance between part and tool. The CHTC at different contact pressures and gap distances is determined through inverse parameter identification. Furthermore a simulation strategy for the subsequent steps of the press hardening process as well as adequate modeling approaches for part and tools are discussed. For the prediction of the yield curves of the material after press hardening a phenomenological model is presented. This model requires the knowledge of the microstructure within the part. By post processing the nodal temperature history with a CCT diagram the quantitative distribution of the phase fractions martensite, bainite, ferrite and pearlite after press hardening is determined. The model itself is based on a Hockett-Sherby approach with the Hockett-Sherby parameters being defined in function of the phase fractions and a characteristic cooling rate.

The PILO process: zeolites and titanates in the treatment of spent ion exchange resins

International Nuclear Information System (INIS)

Hultgren, Aa.; Thegerstroem, C.; Forberg, S.; Westermark, T.; Faelt, L.

1981-01-01

Spent ion exchange resins from power reactor operation contain more than 95% of the total radioactivity of wet reactor wastes. Cementation and bituminization are the two methods applied in Sweden up to now for the immobilization of spent resins. Over the last years, however, research and development work has resulted in a proposed process (PILO), where > 99.9 % of cesium and strontium and around 90 % of other radioactive nuclides are eluted from the spent resins and sorbed in zeolites and titanates in a chromatographic process. The inorganic sorbents are dried after loading and sintered to yield long-term stable products, while the treated resins may be incinerated to give ash residues of fairly short-lived activity. The development work has included production, characterization and testing of different zeolites and titanates, bench-scale optimization of the chromatographic process using actual spent resins, heat treatment of the loaded inorganic sorbents, and resin incineration. Over-all system design studies including transport requirements, integrated process flowsheets, and cost estimates are now in progress. The aim is to have a sufficient basis during spring 1982 to decide on the merits of a PILO plant at the planned repository for low and medium level waste (SFR), to be commissioned in 1988. (Auth.)

Techno-Economic Analysis of Magnesium Extraction from Seawater via a Catalyzed Organo-Metathetical Process

Energy Technology Data Exchange (ETDEWEB)

Liu, Jian; Bearden, Mark D.; Fernandez, Carlos A.; Fifield, Leonard S.; Nune, Satish K.; Motkuri, Radha K.; Koech, Philip K.; McGrail, B. Pete

2018-01-16

Magnesium (Mg) has many useful applications especially in various Mg alloys which can decrease weight while increasing strength. To increase the affordability and minimize environment consequence, a novel catalyzed organo-metathetical (COMET) process was proposed to extract Mg from seawater aiming to achieve significant reduction in total energy and production cost comparing with the melting salt electrolysis method currently adopted by US Mg LLC. A process flowsheet for a reference COMET process was set-up using Aspen Plus which included five key steps, anhydrous MgCl2 production, transmetallation, dibutyl Mg decomposition, n-BuLi regeneration, and LiCL electrolysis. The energy and production cost and CO2 emission were estimated based on the Aspen modeling using Aspen economic analyzer. Our results showed that it is possible to produce Mg from seawater with a production cost of $2.0/kg-Mg while consuming about 35.3 kWh/kg-Mg and releasing 7.0 kg CO2/kg-Mg. A simplified US Mg manufacturing process was also generated using Aspen and the cost and emission results were estimated for comparison purpose. Under our simulation conditions, the reference COMET process maintain a comparable CO2 emission rate and can save about 40% in production cost and save about 15% energy compared to the simplified US Mg process.

Development on the cryogenic hydrogen isotopes distillation process technology for tritium removal (Final report)

Energy Technology Data Exchange (ETDEWEB)

Sung, Ki Woung; Kim, Yong Ik; Na, Jeong Won; Ku, Jae Hyu; Kim, Kwang Rak; Jeong, Yong Won; Lee, Han Soo; Cho, Young Hyun; Ahn, Do Hee; Baek, Seung Woo; Kang, Hee Seok; Kim, You Sun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-12-01

While tritium exposure to the site-workers in Wolsung NPP is up to about 40% of the total personnel exposure, Ministry of Science and Technology has asked tritium removal facility for requirement of post heavy-water reactor construction. For the purpose of essential removal of tritium from the Wolsung heavy-water reactor system, a preliminary study on the cryogenic Ar-N{sub 2} and H{sub 2}-D{sub 2} distillation process for development of liquid-phase catalytic exchange cryogenic hydrogen distillation process technology. The Ar-N{sub 2} distillation column showed good performance with approximately 97% of final Ar concentration, and a computer simulation code was modified using these data. A simulation code developed for cryogenic hydrogen isotopes (H{sub 2}, HD, D{sub 2}, HT, DT, T{sub 2}) distillation column showed good performance after comparison with the result of a JAERI code, and a H{sub 2}-D{sub 2} distillation column was made. Gas chromatography for hydrogen isotopes analysis was established using a vacuum sampling loop, and a schematic diagram of H{sub 2}-D{sub 2} distillation process was suggested. A feasibility on modification of H{sub 2}-D{sub 2} distillation process control system using Laser Raman Spectroscopy was studied, and the consideration points for tritium storage system for Wolsung tritium removal facility was suggested. 31 tabs., 79 figs., 68 refs. (Author).

Preliminary technical data summary defense waste processing facility stage 2

International Nuclear Information System (INIS)

1980-12-01

This Preliminary Technical Data Summary presents the technical basis for design of Stage 2 of the Staged Defense Waste Processing Facility (DWPF). Process changes incorporated in the staged DWPF relative to the Alternative DWPF described in PTDS No. 3 (DPSTD-77-13-3) are the result of ongoing research and development and are aimed at reducing initial capital investment and developing a process to efficiently immobilize the radionuclides in Savannah River Plant (SRP) high-level liquid waste. The radionuclides in SRP waste are present in sludge that has settled to the bottom of waste storage tanks and in crystallized salt and salt solution (supernate). Stage 1 of the DWPF receives washed, aluminum dissolved sludge from the waste tank farms and immobilizes it in a borosilicate glass matrix. The supernate is retained in the waste tank farms until completion of Stage 2 of the DWPF at which time it is filtered and decontaminated by ion exchange in the Stage 2 facility. The decontaminated supernate is concentrated by evaporation and mixed with cement for burial. The radioactivity removed from the supernate is fixed in borosilicate glass along with the sludge. This document gives flowsheets, material and curie balances, material and curie balance bases, and other technical data for design of Stage 2 of the DWPF. Stage 1 technical data are presented in DPSTD-80-38

Development of once-through hybrid sulfur process for nuclear hydrogen production

International Nuclear Information System (INIS)

Jung, Yong Hun

2010-02-01

without concentration process depending on the acid concentration they need. It is reasonable to assume that nearly all the recovered sulfur is reportedly consumed after first converted to sulfuric acid, which is the leading sulfur end-use in all forms. Ot-HyS could meet the additionally rising sulfuric acid demand by feeding severely increasing sulfur surplus. Flowsheet for the sulfur combustion and SCHRS (Sulfur Combustion Heat Recovery System) including Rankine cycle, developed by referring to the existing facilities under some assumptions, was simulated using Aspen Plus with an ideal Henry model and STEAMNBS model. Other part of the flowsheet, modified from the SRNL's work, was simulated using Aspen Plus with OLI-MSE model. Acid concentration of sulfuric acid product was set to be 75 wt% and SDE was treated as a black box under the reasonable assumptions including a cell potential of 0.6 V versus current density of 500 mA/cm 2 , which is a development performance target of the SRNL. As the results, it was demonstrated that net thermal efficiency of Ot-HyS is 47.1 % (based on LHV) and 55.7 % (based on HHV) assuming 33.3 % thermal-to-electric conversion efficiency of nuclear power plant. Hydrogen produced through the energy-efficient Ot-HyS would be used as off-peak electricity storage, to relieve the burden of load-following and help to expand applications of nuclear energy, which is regarded as a 'sustainable development' technology. Further detailed economic feasibility study could help to show the feasibility of Ot-HyS

How does sustainability certification affect the design process? Mapping final design projects at an architectural office

DEFF Research Database (Denmark)

Landgren, Mathilde; Jensen, Lotte Bjerregaard

2017-01-01

process and informing the industry of them. This has led to optimised design processes such as Integrated Energy Design, in which many decisions related to energy consumption and indoor climate are made in the early design stages. The current tendency is to use an expanded notion of sustainability......, derived from the sustainability certification system itself, and to apply it even in the early design process. This perspective emphasises all phases of the life cycle of a building. The goal of the present study was to map how a Danish architectural office approached sustainability in the projects......The context of the study is the very strict regulation of energy consumption for operating buildings in Denmark. It is difficult to meet the requirements by system optimisation in the final design phase, so recent research has focused on ways of meeting the target by adapting the whole design...

32 CFR 536.64 - Final offers.

Science.gov (United States)

2010-07-01

... 32 National Defense 3 2010-07-01 2010-07-01 true Final offers. 536.64 Section 536.64 National... UNITED STATES Investigation and Processing of Claims § 536.64 Final offers. (a) When claims personnel... less than the amount claimed, a settlement authority will make a written final offer within his or her...

Conceptual design of cost-effective and environmentally-friendly configurations for fuel ethanol production from sugarcane by knowledge-based process synthesis.

Science.gov (United States)

Sánchez, Óscar J; Cardona, Carlos A

2012-01-01

In this work, the hierarchical decomposition methodology was used to conceptually design the production of fuel ethanol from sugarcane. The decomposition of the process into six levels of analysis was carried out. Several options of technological configurations were assessed in each level considering economic and environmental criteria. The most promising alternatives were chosen rejecting the ones with a least favorable performance. Aspen Plus was employed for simulation of each one of the technological configurations studied. Aspen Icarus was used for economic evaluation of each configuration, and WAR algorithm was utilized for calculation of the environmental criterion. The results obtained showed that the most suitable synthesized flowsheet involves the continuous cultivation of Zymomonas mobilis with cane juice as substrate and including cell recycling and the ethanol dehydration by molecular sieves. The proposed strategy demonstrated to be a powerful tool for conceptual design of biotechnological processes considering both techno-economic and environmental indicators. Copyright © 2011 Elsevier Ltd. All rights reserved.

Plutonium-uranium separation in the Purex process using mixtures of hydroxylamine nitrate and ferrous sulfamate

International Nuclear Information System (INIS)

McKibben, J.M.; Chostner, D.F.; Orebaugh, E.G.

1983-11-01

Laboratory studies, followed by plant operation, established that a mixture of hydroxylamine nitrate (HAN) and ferrous sulfamate (FS) is superior to FS used alone as a reductant for plutonium in the Purex first cycle. FS usage has been reduced by about 70% (from 0.12 to 0.04M) compared to the pre-1978 period. This reduced the volume of neutralized waste due to FS by 194 liters/metric ton of uranium (MTU) processed. The new flowsheet also gives lower plutonium losses to waste and at least comparable fission product decontamination. To achieve satisfactory performance at this low concentration of FS, the acidity in the 1B mixer-settler was reduced by using a split-scrub - a low acid scrub in stage one and a higher acid scrub in stage three - to remove acid from the solvent exiting the 1A centrifugal contactor. 8 references, 14 figures, 1 table

Advanced Analog Signal Processing for Fuzing Final Report CRADA No. TC-1306-96

Energy Technology Data Exchange (ETDEWEB)

Fu, C. Y. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Spencer, D. [Raymond Engineering, Middletown, CT (United States)

2018-01-24

The purpose of this CRADA between LLNL and Kaman Aerospace/Raymond Engineering Operations (Raymond) was to demonstrate the feasibility of using Analog/Digital Neural Network (ANN) Technology for advanced signal processing, fuzing, and other applications. This cooperation sought to Ieverage the expertise and capabilities of both parties--Raymond to develop the signature recognition hardware system, using Raymond’s extensive experience in the area of system development plus Raymond’s knowledge of military applications, and LLNL to apply ANN and related technologies to an area of significant interest to the United States government. This CRADA effort was anticipated to be a three-year project consisting of three phases: Phase I, Proof-of-Principle Demonstration; Phase II, Proof-of-Design, involving the development of a form-factored integrated sensor and ANN technology processo~ and Phase III, Final Design and Release of the integrated sensor and ANN fabrication process: Under Phase I, to be conducted during calendar year 1996, Raymond was to deliver to LLNL an architecture (design) for an ANN chip. LLNL was to translate the design into a stepper mask and to produce and test a prototype chip from the Raymond design.

Technology roadmaps

Energy Technology Data Exchange (ETDEWEB)

Pearson, B. [Natural Resources Canada, Ottawa, ON (Canada). CANMET Energy Technology Centre

2003-07-01

The purpose of a technology road map is to define the state of a current technology, relevant market issues, and future market needs; to develop a plan that industry can follow to provide these new products and services; and to map technology pathways and performance goals for bringing these products and services to market. The three stages (planning, implementation, and reviewing and updating), benefits, and status of the Clean Coal Technology Roadmap are outlined. Action Plan 2000, a $1.7 million 2000 Climate Change Technology and Innovation Program, which uses the technology roadmapping process, is described. The members of the management steering committee for the Clean Coal Technology Roadmap are listed. A flowsheet showing activities until November 2004, when the final clean coal road map is due, is included.

Pilot-Scale Silicone Process for Low-Cost Carbon Dioxide Capture. Final Scientific/Technical Report

Energy Technology Data Exchange (ETDEWEB)

Hancu, Dan [General Electric Company, Niskayuna, NY (United States); Wood, Benjamin [General Electric Company, Niskayuna, NY (United States); Genovese, Sarah [General Electric Company, Niskayuna, NY (United States); Westendorf, Tiffany [General Electric Company, Niskayuna, NY (United States); Perry, Robert [General Electric Company, Niskayuna, NY (United States); Spiry, Irina [General Electric Company, Niskayuna, NY (United States); Farnum, Rachael [General Electric Company, Niskayuna, NY (United States); Singh, Surinder [General Electric Company, Niskayuna, NY (United States); Wilson, Paul [General Electric Company, Niskayuna, NY (United States); Chen, Wei [General Electric Company, Niskayuna, NY (United States); McDermott, John [General Electric Company, Niskayuna, NY (United States); Doherty, Mark [General Electric Company, Niskayuna, NY (United States); Rainka, Matt [General Electric Company, Niskayuna, NY (United States); Miebach, Barbara [General Electric Company, Niskayuna, NY (United States)

2017-08-03

update the capture system process models, and the techno-economic analysis was performed for a 550 MW coal fired power plant. The 1st year CO₂ removal cost for the aminosilicone-based carbon-capture process was evaluated at $48/ton CO₂ using the steam stripper column. This is a 20% reduction compared to MEA, primarily due to lower overall capital cost. CO₂ cost using the CSTR desorber is dominated by the economics of the solvent make-up. The steam stripper desorber is the preferred unit operation due to a more efficient desorption, and reduced solvent make-up rate. Further reduction in CO₂ capture cost is expected by lowering the manufacturing cost of the solvent, implementing flowsheet optimization and/or implementing the next generation aminosilicone solvent with improved stability and increased CO₂ working capacity.

Survey of potential chlorine production processes. Final report

Energy Technology Data Exchange (ETDEWEB)

1979-04-01

This report is part of the ongoing study of industrial electrochemical processes for the purpose of identifying methods of improving energy efficiencies. A computerized literature search of past and current chlorine generation methods was performed to identify basic chlorine production processes. Over 200 pertinent references are cited involving 20 separate and distinct chlorine processes. Each basic process is evaluated for its engineering and economic viability and energy efficiency. A flow diagram is provided for each basic process. Four criteria are used to determine the most promising processes: raw material availability, type and amount of energy required, by-product demand/disposal and status of development. The most promising processes are determined to be the membrane process (with and without catalytic electrodes), Kel-Chlor, Mobay (direct electrolysis of hydrogen chloride), the Shell process (catalytic oxidation of hydrogen chloride) and oxidation of ammonium chloride. Each of these processes is further studied to determine what activities may be pursued.

Acid-curing and ferric-trickle leaching effluent used in closed circuit uranium extractive process

International Nuclear Information System (INIS)

Jin Suoqing; Xiang Qinfang; Guo Jianzheng; Lu Guizhu; Su Yanru

1998-01-01

The new uranium ore process consists of crushing ore, mixing crushed ore with strong acid in rotating drums and curing the mixture in piles, trickle-leaching the ore beds with ferric solution, extracting uranium from pregnant solution with tertiary amine, precipitating product and disposing residue tailings. All the process effluent is used in closed circuit. There will be no process water to be discharged in the flowsheet except the tailings carrying off 15% water because during leaching moisture content of the ore rises to 15%. Tailings produced by the process are moist and friable, and can be disposed of on a pile or returned to the mine. Main technical parameters of the process: (a) water consumption is 0.2∼0.3 m 3 /t ore, electric power consumption is 20∼30 kW·h/t ore; (b) ore crushing up to -5∼-7 mm, leaching period is 12∼45 d, U content of residue is 0.01%∼0.02%, producing pregnant solution is 0.3∼0.5 m 3 /t ore, which is 1/5∼1/8 that of conventional agitation leaching process; (c) organic agent consumption is 1/5∼1/8 that of the conventional agitation process. All the research results above are tested by the pilot-plant test and industrial test. The new process has been applied to recovery of uranium in the mine located at northeast of China

Technique of complex slime water treatment of coal-mining branch

OpenAIRE

Solodov, G. А.; Zhbyr, Е. V.; Papin, А. V.; Nevedrov, А. V.

2007-01-01

The possibility of complex slime water treatment at coal-mining and coal-treating plants producing marketable products: power-generating concentrate, coal-water fuel, magnetic fraction, industrial water is shown. A basic process flowsheet of slime water treatment presenting a united technological complex is suggested.

Laboratory services series: the utilization of scientific glassblowing in a national research and development laboratory

International Nuclear Information System (INIS)

Farnham, R.M.; Poole, R.W.

1976-04-01

Glassblowing services at a national research and development laboratory provide unique equipment tailored for specific research efforts, small-scale process items for flowsheet demonstrations, and solutions for unusual technical problems such as glass-ceramic unions. Facilities, equipment, and personnel necessary for such services are described

Materials, process, product analysis of coal process technology. Phase I final report

Energy Technology Data Exchange (ETDEWEB)

Saxton, J. C.; Roig, R. W.; Loridan, A.; Leggett, N. E.; Capell, R. G.; Humpstone, C. C.; Mudry, R. N.; Ayres, E.

1976-02-01

The purpose of materials-process-product analysis is a systematic evaluation of alternative manufacturing processes--in this case processes for converting coal into energy and material products that can supplement or replace petroleum-based products. The methodological steps in the analysis include: Definition of functional operations that enter into coal conversion processes, and modeling of alternative, competing methods to accomplish these functions; compilation of all feasible conversion processes that can be assembled from combinations of competing methods for the functional operations; systematic, iterative evaluation of all feasible conversion processes under a variety of economic situations, environmental constraints, and projected technological advances; and aggregative assessments (economic and environmental) of various industrial development scenarios. An integral part of the present project is additional development of the existing computer model to include: A data base for coal-related materials and coal conversion processes; and an algorithmic structure that facilitates the iterative, systematic evaluations in response to exogenously specified variables, such as tax policy, environmental limitations, and changes in process technology and costs. As an analytical tool, the analysis is intended to satisfy the needs of an analyst working at the process selection level, for example, with respect to the allocation of RDandD funds to competing technologies.

Medical Examination of Aliens--Revisions to Medical Screening Process. Final rule.

Science.gov (United States)

2016-01-26

The Centers for Disease Control and Prevention (CDC), within the Department of Health and Human Services (HHS), is issuing this final rule (FR) to amend its regulations governing medical examinations that aliens must undergo before they may be admitted to the United States. Based on public comment received, HHS/CDC did not make changes from the NPRM published on June 23, 2015. Accordingly, this FR will: Revise the definition of communicable disease of public health significance by removing chancroid, granuloma inguinale, and lymphogranuloma venereum as inadmissible health-related conditions for aliens seeking admission to the United States; update the notification of the health-related grounds of inadmissibility to include proof of vaccinations to align with existing requirements established by the Immigration and Nationality Act (INA); revise the definitions and evaluation criteria for mental disorders, drug abuse and drug addiction; clarify and revise the evaluation requirements for tuberculosis; clarify and revise the process for the HHS/CDC-appointed medical review board that convenes to reexamine the determination of a Class A medical condition based on an appeal; and update the titles and designations of federal agencies within the text of the regulation.

FY-2010 Process Monitoring Technology Final Report

Energy Technology Data Exchange (ETDEWEB)

Orton, Christopher R.; Bryan, Samuel A.; Casella, Amanda J.; Hines, Wes; Levitskaia, Tatiana G.; henkell, J.; Schwantes, Jon M.; Jordan, Elizabeth A.; Lines, Amanda M.; Fraga, Carlos G.; Peterson, James M.; Verdugo, Dawn E.; Christensen, Ronald N.; Peper, Shane M.

2011-01-01

During FY 2010, work under the Spectroscopy-Based Process Monitoring task included ordering and receiving four fluid flow meters and four flow visible-near infrared spectrometer cells to be instrumented within the centrifugal contactor system at Pacific Northwest National Laboratory (PNNL). Initial demonstrations of real-time spectroscopic measurements on cold-stream simulants were conducted using plutonium (Pu)/uranium (U) (PUREX) solvent extraction process conditions. The specific test case examined the extraction of neodymium nitrate (Nd(NO3)3) from an aqueous nitric acid (HNO3) feed into a tri-n-butyl phosphate (TBP)/ n-dodecane solvent. Demonstration testing of this system included diverting a sample from the aqueous feed meanwhile monitoring the process in every phase using the on-line spectroscopic process monitoring system. The purpose of this demonstration was to test whether spectroscopic monitoring is capable of determining the mass balance of metal nitrate species involved in a cross-current solvent extraction scheme while also diverting a sample from the system. The diversion scenario involved diverting a portion of the feed from a counter-current extraction system while a continuous extraction experiment was underway. A successful test would demonstrate the ability of the process monitoring system to detect and quantify the diversion of material from the system during a real-time continuous solvent extraction experiment. The system was designed to mimic a PUREX-type extraction process with a bank of four centrifugal contactors. The aqueous feed contained Nd(NO3)3 in HNO3, and the organic phase was composed of TBP/n-dodecane. The amount of sample observed to be diverted by on-line spectroscopic process monitoring was measured to be 3 mmol (3 x 10-3 mol) Nd3+. This value was in excellent agreement with the 2.9 mmol Nd3+ value based on the known mass of sample taken (i.e., diverted) directly from the system feed solution.

Development of the Falea polymetallic uranium project

International Nuclear Information System (INIS)

Ring, R.; Freeman, P.

2014-01-01

The Falea uranium, silver, copper deposit is located in south western Mali, West Africa and is owned by Denison Mines Corp. The current resource estimate is approximately 45 million pounds of U_3O_8 [~17,300 t U] at an average grade of ~ 0.07% U_3O_8. [~0.06% U].The deposit also contains ~37 million Oz Ag and ~70,000 t Cu. The dominant uranium mineral is uraninite, copper is present mainly as chalcopyrite and silver mainly as argentite, and in its native form. Only 5% of the property has been explored to date, and all zones remain open. This paper reports the results of several stages of metallurgical investigations to support ongoing economic studies for the project. The polymetallic nature of the Falea deposit dictates that there are a range of flowsheet options. The ore contains both carbonate and sulphide mineralisations, which have potential impacts on acid and alkaline leaching, respectively. There is also the need to recover both silver and copper. Two primary flowsheet options were considered: 1) Acid leach of ore to recover uranium / flotation of leach residue to recover sulphide concentrate, treatment of concentrate for Cu and Ag recovery; 2) Flotation of ore / alkaline leaching of flotation tails to recover uranium and treatment of flotation concentrate for Cu and Ag recovery. A number of sub-options were considered for each flowsheet. Test work showed that high recoveries of copper and silver to flotation concentrate were obtained for both flotation of ore or acid leach residue. Uranium extraction was also > 90% for both acid and alkaline leaching. The preferred flowsheet was selected after trade-off studies by DRA. This paper presents an overview of the various flowsheet options considered, an outline of the preferred flowsheet, and the results and conclusions of on-going engineering and laboratory/pilot studies to refine the preferred flowsheet. (author)

Dissolution of Material and Test reactor Fuel in an H-Canyon Dissolver

Energy Technology Data Exchange (ETDEWEB)

Daniel, W. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Rudisill, T. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); O' Rourke, P. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2017-01-26

In an amended record of decision for the management of spent nuclear fuel (SNF) at the Savannah River Site, the US Department of Energy has authorized the dissolution and recovery of U from 1000 bundles of Al-clad SNF. The SNF is fuel from domestic and foreign research reactors and is typically referred to as Material Test Reactor (MTR) fuel. Bundles of MTR fuel containing assemblies fabricated from U-Al alloys (or other U compounds) are currently dissolved using a Hg-catalyzed HNO3 flowsheet. Since the development of the existing flowsheet, improved experimental methods have been developed to more accurately characterize the offgas composition and generation rate during laboratory dissolutions. Recently, these new techniques were successfully used to develop a flowsheet for the dissolution of High Flux Isotope Reactor (HFIR) fuel. Using the data from the HFIR dissolution flowsheet development and necessary laboratory experiments, the Savannah River National Laboratory (SRNL) was requested to define flowsheet conditions for the dissolution of MTR fuels. With improved offgas characterization techniques, SRNL will be able define the number of bundles of fuel which can be charged to an H-Canyon dissolver with much less conservatism.

Separation processes for the pretreatment of high-level nuclear wastes at the Savannah River site - 59291

International Nuclear Information System (INIS)

Hobbs, David; Peters, Thomas; Taylor-Pashow, Kathryn; Fondeur, Fernando; Nash, Charles; Fink, Samuel; Herman, David; Marra, Jim

2012-01-01

Document available in abstract form only. Full text of publication follows: Separation methods for the pretreatment of the high-level nuclear wastes (HLW) at the Savannah River Site (SRS) include the Caustic Side Solvent Extraction (CSSX) process for cesium and adsorption/ion exchange for the removal of cesium, strontium and alpha-emitting actinides. The CSSX process uses a calixarene extractant in combination with phase modifiers in a hydrocarbon diluent. Monosodium titanate (MST), a hydrous metal oxide, is the baseline material for the removal of strontium and alpha-emitting radionuclides (principally Pu-238, Pu-239, Pu-240 and Np-237). Two pretreatment facilities, the Modular Caustic Side Solvent Extraction Unit (MCU) and the Actinide Removal Process (ARP) facility began radioactive operations at SRS in 2008. Together these facilities can treat approximately 4 million liters of waste per year. The same separation processes are also planned for the much larger Salt Waste Processing Facility (SWPF). The SWPF, which has a design throughput of about 27 million liters per year, is under construction and scheduled to begin radioactive operations in 2014. Current R and D activities for the CSSX process are focused on implementing a new solvent system and stripping flowsheet that offers enhanced extraction and stripping of cesium. This next generation solvent system features a different calixarene extractant, uses caustic instead of nitric acid

The production of fuels and chemicals from food processing wastes & cellulosics. Final research report

Energy Technology Data Exchange (ETDEWEB)

Dale, M.C.; Okos, M.; Burgos, N. [and others

1997-06-15

High strength food wastes of about 15-20 billion pounds solids are produced annually by US food producers. Low strength food wastes of 5-10 billion pounds/yr. are produced. Estimates of the various components of these waste streams are shown in Table 1. Waste paper/lignocellulosic crops could produce 2 to 5 billion gallons of ethanol per year or other valuable chemicals. Current oil imports cost the US about $60 billion dollars/yr. in out-going balance of trade costs. Many organic chemicals that are currently derived from petroleum can be produced through fermentation processes. Petroleum based processes have been preferred over biotechnology processes because they were typically cheaper, easier, and more efficient. The technologies developed during the course of this project are designed to allow fermentation based chemicals and fuels to compete favorably with petroleum based chemicals. Our goals in this project have been to: (1) develop continuous fermentation processes as compared to batch operations; (2) combine separation of the product with the fermentation, thus accomplishing the twin goals of achieving a purified product from a fermentation broth and speeding the conversion of substrate to product in the fermentation broth; (3) utilize food or cellulosic waste streams which pose a current cost or disposal problem as compared to high cost grains or sugar substrates; (4) develop low energy recovery methods for fermentation products; and finally (5) demonstrate successful lab scale technologies on a pilot/production scale and try to commercialize the processes. The scale of the wastes force consideration of {open_quotes}bulk commodity{close_quotes} type products if a high fraction of the wastes are to be utilized.

Final report on the public involvement process phase 1, Monitored Retrievable Storage Facility Feasibility Study

Energy Technology Data Exchange (ETDEWEB)

Moore, L.; Shanteau, C.

1992-12-01

This report summarizes the pubic involvement component of Phase 1 of the Monitored Retrievable Storage Facility (NM) Feasibility Study in San Juan County, Utah. Part of this summary includes background information on the federal effort to locate a voluntary site for temporary storage of nuclear waste, how San Juan County came to be involved, and a profile of the county. The heart of the report, however, summarizes the activities within the public involvement process, and the issues raised in those various forums. The authors have made every effort to reflect accurately and thoroughly all the concerns and suggestions expressed to us during the five month process. We hope that this report itself is a successful model of partnership with the citizens of the county -- the same kind of partnership the county is seeking to develop with its constituents. Finally, this report offers some suggestions to both county officials and residents alike. These suggestions concern how decision-making about the county's future can be done by a partnership of informed citizens and listening decision-makers. In the Appendix are materials relating to the public involvement process in San Juan County.

Final report on the public involvement process phase 1, Monitored Retrievable Storage Facility Feasibility Study

Energy Technology Data Exchange (ETDEWEB)

Moore, L.; Shanteau, C.

1992-12-01

This report summarizes the pubic involvement component of Phase 1 of the Monitored Retrievable Storage Facility (NM) Feasibility Study in San Juan County, Utah. Part of this summary includes background information on the federal effort to locate a voluntary site for temporary storage of nuclear waste, how San Juan County came to be involved, and a profile of the county. The heart of the report, however, summarizes the activities within the public involvement process, and the issues raised in those various forums. The authors have made every effort to reflect accurately and thoroughly all the concerns and suggestions expressed to us during the five month process. We hope that this report itself is a successful model of partnership with the citizens of the county -- the same kind of partnership the county is seeking to develop with its constituents. Finally, this report offers some suggestions to both county officials and residents alike. These suggestions concern how decision-making about the county`s future can be done by a partnership of informed citizens and listening decision-makers. In the Appendix are materials relating to the public involvement process in San Juan County.

Final report on the public involvement process phase 1, Monitored Retrievable Storage Facility Feasibility Study

International Nuclear Information System (INIS)

Moore, L.; Shanteau, C.

1992-12-01

This report summarizes the pubic involvement component of Phase 1 of the Monitored Retrievable Storage Facility (NM) Feasibility Study in San Juan County, Utah. Part of this summary includes background information on the federal effort to locate a voluntary site for temporary storage of nuclear waste, how San Juan County came to be involved, and a profile of the county. The heart of the report, however, summarizes the activities within the public involvement process, and the issues raised in those various forums. The authors have made every effort to reflect accurately and thoroughly all the concerns and suggestions expressed to us during the five month process. We hope that this report itself is a successful model of partnership with the citizens of the county -- the same kind of partnership the county is seeking to develop with its constituents. Finally, this report offers some suggestions to both county officials and residents alike. These suggestions concern how decision-making about the county's future can be done by a partnership of informed citizens and listening decision-makers. In the Appendix are materials relating to the public involvement process in San Juan County

Data breaches. Final rule.

Science.gov (United States)

2008-04-11

This document adopts, without change, the interim final rule that was published in the Federal Register on June 22, 2007, addressing data breaches of sensitive personal information that is processed or maintained by the Department of Veterans Affairs (VA). This final rule implements certain provisions of the Veterans Benefits, Health Care, and Information Technology Act of 2006. The regulations prescribe the mechanisms for taking action in response to a data breach of sensitive personal information.

Analyses of one-step liquid hydrogen production from methane and landfill gas

Energy Technology Data Exchange (ETDEWEB)

Huang, Cunping; T-Raissi, Ali [University of Central Florida, Florida Solar Energy Center, 1679 Clearlake Road, Cocoa, FL 32922-5703 (United States)

2007-11-15

Conventional liquid hydrogen (LH{sub 2}) production consists of two basic steps: (1) gaseous hydrogen (GH{sub 2}) production via steam methane reformation followed by purification by means of pressure swing adsorption (PSA), and (2) GH{sub 2} liquefaction. LH{sub 2} produced by the conventional processes is not carbon neutral because of the carbon dioxide (CO{sub 2}) emission from PSA operation. A novel concept is herein presented and flowsheeted for LH{sub 2} production with zero carbon emission using methane (CH{sub 4}) or landfill gas as feedstock. A cryogenic process is used for both H{sub 2} separation/purification and liquefaction. This one-step process can substantially increase the efficiency and reduce costs because no PSA step is required. Furthermore, the integrated process results in no CO{sub 2} emissions and minimal H{sub 2} losses. Of the five flowsheets presented, one that combines low and high temperature CO/CH{sub 4} reforming reactions in a single reactor shows the highest overall efficiency with the first and second law efficiencies of 85% and 56%, respectively. The latter figure assumes 10% overall energy loss and 30% efficiency for the cryogenic process. (author)

Development of processes for pilot plant production of purified uranyl nitrate solutions

Energy Technology Data Exchange (ETDEWEB)

Alfredson, P. G.; Charlton, B. G.; Ryan, R. K.; Vilkaitis, V. K.

1975-01-15

Nuclear purity uranyl nitrate solutions were produced from Rum Jungle yellow cake by dissolution in nitric acid and purification by solvent extraction with 20 vol. per cent tributyl phosphate in kerosene using pump - mix mixer-settler contactors. The design of the equipment, experimental studies and operating experience are described. Dissolution of yellow cake and recycled uranium oxide materials was readily carried out in a 100 l dissolver to give solutions containing 300 gU l{sup -1} and 0.5 to 4 M nitric acid. Filtration of silica from this solution prior to solvent extraction was not necessary in this work for yellow cake containing 0.25 per cent silica. A low acid flowsheet for uranium purification was developed in which the nitric acid consumption was reduced by 76 per cent and the throughput of the mixer-settler units was increased by 67 per cent compared with the initial design flowsheet. Nine extraction and seven scrubbing stages were used with a feed solution containing 300 gU l{sup -1} and 1.0 M nitric acid and with a portion of the product recycled as scrub solution. The loaded organic phase was stripped in 16 stages with 0.05 M nitric acid heated to 60 deg C to give a 120 gU l{sup -1} product. The uranium concentration in the raffinate was < 0.04 g l{sup -1}, corresponding to approximately 0.01 per cent of the feed. (author)

Zirconium behaviour in purex process solutions

International Nuclear Information System (INIS)

Shu, J.

1982-01-01

The extraction behaviour of zirconium, as fission product, in TBP/diluent- HNO 3 -H 2 O systems, simulating Purex solutions, is studied. The main purpose is to attain an increasing in the zirconium decontamination factor by adjusting the extraction parameters. Equilibrium diagram, TBP concentration, aqueous:organic ratio, salting-out effects and, uranium loading in the organic phase were the main factors studied. All these experiments had been made with zirconium in the 10 - 2 - 10 - 3 concentration range. The extractant degradation products influence uppon the zirconium behaviour was also verified. With the obtained data it was possible to introduce some modification in the standard Purex flow-sheet in order to obtain the uranium product with higher zirconium decontamination. (Author) [pt

Separation of actinides and long-lived fission products from high-level radioactive wastes (a review)

International Nuclear Information System (INIS)

Kolarik, Z.

1991-11-01

The management of high-level radioactive wastes is facilitated, if long-lived and radiotoxic actinides and fission products are separated before the final disposal. Especially important is the separation of americium, curium, plutonium, neptunium, strontium, cesium and technetium. The separated nuclides can be deposited separately from the bulk of the high-level waste, but their transmutation to short-lived nuclides is a muchmore favourable option. This report reviews the chemistry of the separation of actinides and fission products from radioactive wastes. The composition, nature and conditioning of the wastes are described. The main attention is paid to the solvent extraction chemistry of the elements and to the application of solvent extraction in unit operations of potential partitioning processes. Also reviewed is the behaviour of the elements in the ion exchange chromatography, precipitation, electrolysis from aqueous solutions and melts, and the distribution between molten salts and metals. Flowsheets of selected partitioning processes are shown and general aspects of the waste partitioning are shortly discussed. (orig.) [de

Permanent certification program for health information technology; revisions to ONC-Approved Accreditor processes. Final rule.

Science.gov (United States)

2011-11-25

Under the authority granted to the National Coordinator for Health Information Technology by section 3001(c)(5) of the Public Health Service Act (PHSA) as added by the Health Information Technology for Economic and Clinical Health (HITECH) Act, this final rule establishes a process for addressing instances where the ONC-Approved Accreditor (ONC-AA) engages in improper conduct or does not perform its responsibilities under the permanent certification program. This rule also addresses the status of ONC-Authorized Certification Bodies (ONC-ACBs) in instances where there may be a change in the accreditation organization serving as the ONC-AA and clarifies the responsibilities of the new ONC-AA.

Cavitational Hydrothermal Oxidation: A New Remediation Process - Final Report; FINAL

International Nuclear Information System (INIS)

Suslick, K. S.

2001-01-01

During the past year, we have continued to make substantial scientific progress on our understanding of cavitation phenomena in aqueous media and applications of cavitation to remediation processes. Our efforts have focused on three separate areas: sonoluminescence as a probe of conditions created during cavitational collapse in aqueous media, the use of cavitation for remediation of contaminated water, and an addition of the use of ultrasound in the synthesis of novel heterogeneous catalysts for hydrodehalogenation of halocarbons under mild conditions

Reprocessing of AHWR spent-fuel: Challenges and strategies

International Nuclear Information System (INIS)

Kant, S.

2005-01-01

Reprocessing of advanced heavy water reactor (AHWR) spent-fuel involves separation of Th, 233 U and Pu, from the fission products and from one another. A proper combination of Purex and Thorex processes is required. The technology development for a reprocessing facility is extremely complex owing to high fissile content, high levels of irradiation, presence high of levels of 232 U, difficulty in thoria dissolution, presence of thorium as the major constituent, problems due to third phase formation with Th, etc. It demands for development of suitable dissolution, solvent extraction, criticality control, U-Pu partitioning, and other equipments and/or techniques. Process modelling, simulation and optimisation are crucial in predicting behaviour of equipments/cycles, and in arriving at safe and optimum flowsheet. A significant success in this field has been achieved. This paper describes the reprocessing aspects pertaining to AHWR spent-fuel, indicating the major technological challenges, strategies to be followed and development requirements. A schematic flowsheet is proposed for Th- 233 U-Pu separation. (author)

Scale Up of Malonic Acid Fermentation Process: Cooperative Research and Development Final Report, CRADA Number CRD-16-612

Energy Technology Data Exchange (ETDEWEB)

Schell, Daniel J [National Renewable Energy Laboratory (NREL), Golden, CO (United States)

2018-04-16

The goal of this work is to use the large fermentation vessels in the National Renewable Energy Laboratory's (NREL) Integrated Biorefinery Research Facility (IBRF) to scale-up Lygos' biological-based process for producing malonic acid and to generate performance data. Initially, work at the 1 L scale validated successful transfer of Lygos' fermentation protocols to NREL using a glucose substrate. Outside of the scope of the CRADA with NREL, Lygos tested their process on lignocellulosic sugars produced by NREL at Lawrence Berkeley National Laboratory's (LBNL) Advanced Biofuels Process Development Unit (ABPDU). NREL produced these cellulosic sugar solutions from corn stover using a separate cellulose/hemicellulose process configuration. Finally, NREL performed fermentations using glucose in large fermentors (1,500- and 9,000-L vessels) to intermediate product and to demonstrate successful performance of Lygos' technology at larger scales.

Theoretical analysis of a biogas-fed PEMFC system with different hydrogen purifications: Conventional and membrane-based water gas shift processes

International Nuclear Information System (INIS)

Authayanun, Suthida; Aunsup, Pounyaporn; Patcharavorachot, Yaneeporn; Arpornwichanop, Amornchai

2014-01-01

Highlights: • Thermodynamic analysis of the biogas-fed PEMFC system is performed. • Conventional and membrane-based WGS processes for H 2 purification are studied. • A flowsheet model of the PEMFC system is developed. • Effect of key parameters on yields of H 2 and carbon in the biogas reformer is shown. • Performance of PEMFC systems with different H 2 purification processes is analyzed. - Abstract: This study presents a thermodynamic analysis of biogas reforming and proton electrolyte membrane fuel cell (PEMFC) integrated process with different hydrogen purifications: conventional and membrane-based water gas shift processes. The aim is to determine the optimal reforming process for hydrogen production from biogas in the PEMFC system. The formation of carbon is concerned in the hydrogen production. The simulation results show that increases in the steam-to-methane ratio and reformer temperature can improve the hydrogen yield and reduce the carbon formation. From the performance analysis, it is found that when the PEMFC is operated at high temperature and fuel utilization, the overall system efficiency enhances. The performance of the PEMFC system with the installation of a water gas shift membrane unit in the hydrogen purification step is slightly increased, compared with a conventional process

HYBRID SULFUR PROCESS REFERENCE DESIGN AND COST ANALYSIS

Energy Technology Data Exchange (ETDEWEB)

Gorensek, M.; Summers, W.; Boltrunis, C.; Lahoda, E.; Allen, D.; Greyvenstein, R.

2009-05-12

PBMR (Pty.) Ltd. in the RSA, with the Hybrid Sulfur (HyS) Process, under development by the Savannah River National Laboratory (SRNL) in the US as part of the NHI. This work was performed by SRNL, Westinghouse Electric Company, Shaw, PBMR (Pty) Ltd., and Technology Insights under a Technical Consulting Agreement (TCA). Westinghouse Electric, serving as the lead for the PBMR process heat application team, established a cost-shared TCA with SRNL to prepare an updated HyS thermochemical water-splitting process flowsheet, a nuclear hydrogen plant preconceptual design and a cost estimate, including the cost of hydrogen production. SRNL was funded by DOE under the NHI program, and the Westinghouse team was self-funded. The results of this work are presented in this Final Report. Appendices have been attached to provide a detailed source of information in order to document the work under the TCA contract.

Waste Feed Delivery Strategy for Tanks 241-AN-102 and 241-AN-107

International Nuclear Information System (INIS)

BLACKER, S.M.

2000-01-01

This engineering study establishes the detailed retrieval strategy, equipment requirements, and key parameters for preparing detailed process flowsheets; evaluates the technical and programmatic risks associated with processing, certifying, transferring, and delivering waste from Tanks 241-AN-102 and 241-AN-107 to BNFL; and provides a list of necessary follow-on actions so that program direction from ORP can be successfully implemented

Waste Feed Delivery Strategy for Tanks 241-AN-102 and 241-AN-107

Energy Technology Data Exchange (ETDEWEB)

BLACKER, S.M.

2000-04-13

This engineering study establishes the detailed retrieval strategy, equipment requirements, and key parameters for preparing detailed process flowsheets; evaluates the technical and programmatic risks associated with processing, certifying, transferring, and delivering waste from Tanks 241-AN-102 and 241-AN-107 to BNFL; and provides a list of necessary follow-on actions so that program direction from ORP can be successfully implemented.

Automated Solar Cell Assembly Teamed Process Research. Final subcontract report, 6 January 1993--31 October 1995

Energy Technology Data Exchange (ETDEWEB)

Nowlan, M. J.; Hogan, S. J.; Breen, W. F.; Murach, J. M.; Sutherland, S. F.; Patterson, J. S.; Darkazalli, G. [Spire Corp., Bedford, MA (US)

1996-02-01

This is the Final Technical Report for a program entitled ''Automated Solar Cell Assembly Teamed Process Research,'' funded by the US Department of Energy. This program was part of Phase 3A of the Photovoltaic Manufacturing Technology (PVMaT) project, which addressed the generic needs of the photovoltaic (PV) industry for improved quality, accelerated production scale-up, and substantially reduced manufacturing cost. Crystalline silicon solar cells (Czochralski monocrystalline, cast polycrystalline, and ribbon polycrystalline) are used in the great majority of PV modules produced in the US, accounting for 95% of all shipments in 1994. Spire's goal in this program was to reduce the cost of these modules by developing high throughput (5 MW per year) automated processes for interconnecting solar cells made from standard and thin silicon wafers. Spire achieved this goal by developing a completely new automated processing system, designated the SPI-ASSEMBLER{trademark} 5000, which is now offered as a commercial product to the PV industry. A discussion of the project and of the Assembler is provided.

Recovery of actinides from TBP-Na2Co3 scrub-waste solutions: the ARALEX process

International Nuclear Information System (INIS)

Horwitz, E.P.; Bloomquist, C.A.A.; Mason, G.W.; Leonard, R.A.; Ziegler, A.A.

1979-08-01

A flowsheet for the recovery of actinides from TBP-Na 2 CO 3 scrub-waste solutions has been developed, based on batch extraction data, and tested, using laboratory-scale countercurrent extraction techniques. The process, called the ARALEX process, uses 2-ethyl-1-hexanol (2-EHOH) to extract the TBP degradation products (HDBP and H 2 MBP) from acidified Na 2 CO 3 scrub waste leaving the actinides in the aqueous phase. Dibutyl and monobutyl phosphoric acids are attached to the 2-EHOH molecules through hydrogen bonds, which also diminish the ability of the HDBP and H 2 MBP to complex actinides. Thus all actinides remain in the aqueous raffinate. Dilute sodium hydroxide solutions can be used to back-extract the dibutyl and monobutyl phosphoric acid esters as their sodium salts. The 2-EHOH can then be recycled. After extraction of the acidified carbonate waste with 2-EHOH, the actinides may be readily extracted from the raffinate with DHDECMP or, in the case of tetra- and hexavalent actinides, with TBP. The ARALEX process can also be applied to other actinide waste streams which contain appreciable concentrations of polar organic compounds (e.g., detergents) that interfere with conventional actinide ion exchange and liquid-liquid extraction procedures. 20 figures, 6 tables