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Sample records for extraction purex facility

  1. PUREX facility hazards assessment

    International Nuclear Information System (INIS)

    Sutton, L.N.

    1994-01-01

    This report documents the hazards assessment for the Plutonium Uranium Extraction Plant (PUREX) located on the US Department of Energy (DOE) Hanford Site. Operation of PUREX is the responsibility of Westinghouse Hanford Company (WHC). This hazards assessment was conducted to provide the emergency planning technical basis for PUREX. DOE Order 5500.3A requires an emergency planning hazards assessment for each facility that has the potential to reach or exceed the lowest level emergency classification. In October of 1990, WHC was directed to place PUREX in standby. In December of 1992 the DOE Assistant Secretary for Environmental Restoration and Waste Management authorized the termination of PUREX and directed DOE-RL to proceed with shutdown planning and terminal clean out activities. Prior to this action, its mission was to reprocess irradiated fuels for the recovery of uranium and plutonium. The present mission is to establish a passively safe and environmentally secure configuration at the PUREX facility and to preserve that condition for 10 years. The ten year time frame represents the typical duration expended to define, authorize and initiate follow-on decommissioning and decontamination activities

  2. PUREX facility preclosure work plan

    International Nuclear Information System (INIS)

    Engelmann, R.H.

    1997-01-01

    This preclosure work plan presents a description of the PUREX Facility, the history of the waste managed, and addresses transition phase activities that position the PUREX Facility into a safe and environmentally secure configuration. For purposes of this documentation, the PUREX Facility does not include the PUREX Storage Tunnels (DOE/RL-90/24). Information concerning solid waste management units is discussed in the Hanford Facility Dangerous Waste Permit Application, General Information Portion (DOE/RL-91-28, Appendix 2D)

  3. Surveillance and Maintenance Plan for the Plutonium Uranium Extraction (PUREX) Facility

    International Nuclear Information System (INIS)

    Woods, P.J.

    1998-05-01

    This document provides a plan for implementing surveillance and maintenance (S ampersand M) activities to ensure the Plutonium Uranium Extraction (PUREX) Facility is maintained in a safe, environmentally secure, and cost-effective manner until subsequent closure during the final disposition phase of decommissioning. This plan has been prepared in accordance with the guidelines provided in the U.S. Department of Energy (DOE), Office of Environmental Management (EM) Decommissioning Resource Manual (DOE/EM-0246) (DOE 1995), and Section 8.6 of TPA change form P-08-97-01 to the Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) (Ecology, et al. 1996). Specific objectives of the S ampersand M program are: Ensure adequate containment of remaining radioactive and hazardous material. Provide security control for access into the facility and physical safety to surveillance personnel. Maintain the facility in a manner that will minimize potential hazards to the public, the environment, and surveillance personnel. Provide a plan for the identification and compliance with applicable environmental, safety, health, safeguards, and security requirements

  4. PUREX/UO3 facilities deactivation lessons learned history

    International Nuclear Information System (INIS)

    Gerber, M.S.

    1997-01-01

    In May 1997, a historic deactivation project at the PUREX (Plutonium URanium EXtraction) facility at the Hanford Site in south-central Washington State concluded its activities (Figure ES-1). The project work was finished at $78 million under its original budget of $222.5 million, and 16 months ahead of schedule. Closely watched throughout the US Department of Energy (DOE) complex and by the US Department of Defense for the value of its lessons learned, the PUREX Deactivation Project has become the national model for the safe transition of contaminated facilities to shut down status

  5. PUREX/UO{sub 3} facilities deactivation lessons learned: History

    Energy Technology Data Exchange (ETDEWEB)

    Gerber, M.S.

    1997-11-25

    In May 1997, a historic deactivation project at the PUREX (Plutonium URanium EXtraction) facility at the Hanford Site in south-central Washington State concluded its activities (Figure ES-1). The project work was finished at $78 million under its original budget of $222.5 million, and 16 months ahead of schedule. Closely watched throughout the US Department of Energy (DOE) complex and by the US Department of Defense for the value of its lessons learned, the PUREX Deactivation Project has become the national model for the safe transition of contaminated facilities to shut down status.

  6. PUREX/UO{sub 3} facilities deactivation lessons learned history

    Energy Technology Data Exchange (ETDEWEB)

    Hamrick, D.G.; Gerber, M.S.

    1995-01-01

    The Plutonium-Uranium Extraction (PUREX) Facility operated from 1956-1972, from 1983-1988, and briefly during 1989-1990 to produce for national defense at the Hanford Site in Washington State. The Uranium Trioxide (UO{sub 3}) Facility operated at the Hanford Site from 1952-1972, 1984-1988, and briefly in 1993. Both plants were ordered to permanent shutdown by the U.S. Department of Energy (DOE) in December 1992, thus initiating their deactivation phase. Deactivation is that portion of a facility`s life cycle that occurs between operations and final decontamination and decommissioning (D&D). This document details the history of events, and the lessons learned, from the time of the PUREX Stabilization Campaign in 1989-1990, through the end of the first full fiscal year (FY) of the deactivation project (September 30, 1994).

  7. PUREX/UO3 facilities deactivation lessons learned history

    International Nuclear Information System (INIS)

    Hamrick, D.G.; Gerber, M.S.

    1995-01-01

    The Plutonium-Uranium Extraction (PUREX) Facility operated from 1956-1972, from 1983-1988, and briefly during 1989-1990 to produce for national defense at the Hanford Site in Washington State. The Uranium Trioxide (UO 3 ) Facility operated at the Hanford Site from 1952-1972, 1984-1988, and briefly in 1993. Both plants were ordered to permanent shutdown by the U.S. Department of Energy (DOE) in December 1992, thus initiating their deactivation phase. Deactivation is that portion of a facility's life cycle that occurs between operations and final decontamination and decommissioning (D ampersand D). This document details the history of events, and the lessons learned, from the time of the PUREX Stabilization Campaign in 1989-1990, through the end of the first full fiscal year (FY) of the deactivation project (September 30, 1994)

  8. Project C-018H, 242-A Evaporator/PUREX Plant Process Condensate Treatment Facility, functional design criteria. Revision 3

    International Nuclear Information System (INIS)

    Sullivan, N.

    1995-01-01

    This document provides the Functional Design Criteria (FDC) for Project C-018H, the 242-A Evaporator and Plutonium-Uranium Extraction (PUREX) Plant Condensate Treatment Facility (Also referred to as the 200 Area Effluent Treatment Facility [ETF]). The project will provide the facilities to treat and dispose of the 242-A Evaporator process condensate (PC), the Plutonium-Uranium Extraction (PUREX) Plant process condensate (PDD), and the PUREX Plant ammonia scrubber distillate (ASD)

  9. PUREX/UO3 Facilities deactivation lessons learned history

    Energy Technology Data Exchange (ETDEWEB)

    Gerber, M.S.

    1996-09-19

    Disconnecting the criticality alarm permanently in June 1996 signified that the hazards in the PUREX (plutonium-uranium extraction) plant had been so removed and reduced that criticality was no longer a credible event. Turning off the PUREX criticality alarm also marked a salient point in a historic deactivation project, 1 year before its anticipated conclusion. The PUREX/UO3 Deactivation Project began in October 1993 as a 5-year, $222.5- million project. As a result of innovations implemented during 1994 and 1995, the project schedule was shortened by over a year, with concomitant savings. In 1994, the innovations included arranging to send contaminated nitric acid from the PUREX Plant to British Nuclear Fuels, Limited (BNFL) for reuse and sending metal solutions containing plutonium and uranium from PUREX to the Hanford Site tank farms. These two steps saved the project $36.9- million. In 1995, reductions in overhead rate, work scope, and budget, along with curtailed capital equipment expenditures, reduced the cost another $25.6 million. These savings were achieved by using activity-based cost estimating and applying technical schedule enhancements. In 1996, a series of changes brought about under the general concept of ``reengineering`` reduced the cost approximately another $15 million, and moved the completion date to May 1997. With the total savings projected at about $75 million, or 33.7 percent of the originally projected cost, understanding how the changes came about, what decisions were made, and why they were made becomes important. At the same time sweeping changes in the cultural of the Hanford Site were taking place. These changes included shifting employee relations and work structures, introducing new philosophies and methods in maintaining safety and complying with regulations, using electronic technology to manage information, and, adopting new methods and bases for evaluating progress. Because these changes helped generate cost savings and were

  10. Hanford facility dangerous waste permit application, PUREX storage tunnels

    International Nuclear Information System (INIS)

    Price, S.M.

    1997-01-01

    The Hanford Facility Dangerous Waste Permit Application is considered to be a single application organized into a General Information Portion (document number DOE/RL-91-28) and a Unit-Specific Portion. The scope of the Unit-Specific Portion is limited to Part B permit application documentation submitted for individual, operating treatment, storage, and/or disposal units, such as the PUREX Storage Tunnels (this document, DOE/RL-90-24). Both the General Information and Unit-Specific portions of the Hanford Facility Dangerous Waste Permit Application address the content of the Part B permit application guidance prepared by the Washington State Department of Ecology (Ecology 1996) and the US Environmental Protection Agency (40 Code of Federal Regulations 270), with additional information needs defined by the Hazardous and Solid Waste Amendments and revisions of Washington Administrative Code 173-303. For ease of reference, the Washington State Department of Ecology alpha-numeric section identifiers from the permit application guidance documentation (Ecology 1996) follow, in brackets, the chapter headings and subheadings. A checklist indicating where information is contained in the PUREX Storage Tunnels permit application documentation, in relation to the Washington State Department of Ecology guidance, is located in the Contents Section. Documentation contained in the General Information Portion is broader in nature and could be used by multiple treatment, storage, and/or disposal units (e.g., the glossary provided in the General Information Portion). Wherever appropriate, the PUREX Storage Tunnels permit application documentation makes cross-reference to the General Information Portion, rather than duplicating text. Information provided in this PUREX Storage Tunnels permit application documentation is current as of April 1997

  11. Purex process extraction cycles: a potential for progress today

    Energy Technology Data Exchange (ETDEWEB)

    Boullis, B.; Germain, M.; Goumondy, J.P.; Rouyer, H.

    1994-12-31

    The Purex process very quickly and very widely supplanted the other concepts considered for nuclear fuel reprocessing after the presentation made at the Geneva Conference in 1955. The selectivity and radiolytic stability of tributylphosphate (T.B.P) clearly appeared to augur an extremely attractive process for completing the separation of valuable elements in the irradiated fuel. The concept has confirmed its validity, and subsequently its ability to adapt to changing requirements or constraints. Its industrial viability is in fact unquestioned today: Purex process is the basis of all the reprocessing plants in operation or planned throughout the world, and recent commissioning of the UP3 plant in France, in remarkable conditions, attests to such a level of maturity that one is tempted to ask the question: ``What remains to be proved, discovered or improved in the core of the Purex process?``. (authors). 7 refs., 4 tabs.

  12. Purex process extraction cycles: a potential for progress today

    International Nuclear Information System (INIS)

    Boullis, B.; Germain, M.; Goumondy, J.P.; Rouyer, H.

    1994-01-01

    The Purex process very quickly and very widely supplanted the other concepts considered for nuclear fuel reprocessing after the presentation made at the Geneva Conference in 1955. The selectivity and radiolytic stability of tributylphosphate (T.B.P) clearly appeared to augur an extremely attractive process for completing the separation of valuable elements in the irradiated fuel. The concept has confirmed its validity, and subsequently its ability to adapt to changing requirements or constraints. Its industrial viability is in fact unquestioned today: Purex process is the basis of all the reprocessing plants in operation or planned throughout the world, and recent commissioning of the UP3 plant in France, in remarkable conditions, attests to such a level of maturity that one is tempted to ask the question: ''What remains to be proved, discovered or improved in the core of the Purex process?''. (authors). 7 refs., 4 tabs

  13. Delisting strategy for the Hanford Site 242-A Evaporator PUREX Plant Condensate Treatment Facility

    International Nuclear Information System (INIS)

    1992-04-01

    This document describes the strategy that the US Department of Energy, Richland Field Office intends to use in preparing the delisting petition for the 242-A Evaporator/PUREX Plant Condensate Treatment Facility. Because the 242-A Evaporator/PUREX Plant Condensate Treatment Facility will not be operational until 1994, the delisting petition will be structured as an up-front petition based on the ''multiple waste treatment facility'' approach outline in the 1985 US Environmental Protection Agency's Petitions to Delist Hazardous Waste. The 242-A evaporator/PUREX Plant Condensate Treatment Facility effluent characterization data will not be available to support the delisting petition, because the delisting petition will be submitted to the US Environmental Protection Agency before start-up of the 242-A Evaporator/PUREX Plant Condensate Treatment Facility. Therefore, the delisting petition will be based on data collected during the pilot plant testing for the 242-A Evaporator/PUREX Plant Condensate Treatment Facility. This pilot plant testing will be conducted on synthetic waste. The composition of the synthetic waste will be based on: (1) constituents of regulatory concern, and (2) on process knowledge. The pilot plant testing will be performed to determine the removal efficiencies of the process equipment at concentrations greater than reasonably could be expected in the actual waste. This strategy document also describes the logic used to develop the synthetic waste, to develop the pilot plant testing program, and to prepare the delisting petition. This strategy document also described how full-scale operating data will be collected during initial operation of the 242-A Evaporator/PUREX Plant Condensate Treatment Facility to verify information presented in the delisting petition

  14. The Necessary and Sufficient Closure Process Completion Report for Purex FacilitySurveillance and Maintenance

    International Nuclear Information System (INIS)

    Gerald, J.W.

    1997-10-01

    This document completes the U.S. Department of Energy Closure Process for Necessary and Sufficient Sets of Standards process for the Plutonium Uranium Extraction facility located at the Hanford Site in Washington State. This documentation is provided to support the Work Smart Standards set identified for the long-term surveillance and maintenance of PUREX. This report is organized into two volumes. Volume 1 contains the following sections: Section 1: Provides an introduction for the document Section 2: Provides a basis for initiating the N ampersand S process Section 3: Defines the work and hazards to be addressed Section 4: Identifies the N ampersand S set of standards and requirements Section 5: Provides the justification for adequacy of the work smart standards Section 6: Shows the criteria and qualifications of the teams Section 7: Describes the stakeholder participation and concerns Section 8: Provides a list of references used within the document

  15. Disposition of PUREX facility tanks D5 and E6 uranium and plutonium solutions

    International Nuclear Information System (INIS)

    Harty, D.P.

    1993-12-01

    Approximately 9 kilograms of plutonium and 5 metric tons of uranium in a 1 molar nitric acid solution are being stored in two PUREX facility vessels, tanks D5 and E6. The plutonium was accumulated during cleanup activities of the plutonium product area of the PUREX facility. Personnel at PUREX recently completed a formal presentation to the Surplus Materials Peer Panel (SMPP) regarding disposition of the material currently in these tanks. The peer panel is a group of complex-wide experts who have been chartered by EM-64 (Office of Site and Facility Transfer) to provide a third party independent review of disposition decisions. The information presented to the peer panel is provided in the first section of this report. The panel was generally receptive to the information provided at that time and the recommendations which were identified

  16. Data quality objectives for PUREX deactivation flushing

    International Nuclear Information System (INIS)

    Bhatia, R.K.

    1995-01-01

    This Data Quality Objection (DQO) defines the sampling and analysis requirements necessary to support the deactivation of the Plutonium-Uranium Extraction (PUREX) facility vessels that are regulated by WAC 173-303. Specifically, sampling and analysis requirements are identified for the flushing operations that are a major element of PUREX deactivation

  17. Functional design criteria for the 242-A evaporator and PUREX [Plutonium-Uranium Extraction] Plant condensate interim retention basin

    International Nuclear Information System (INIS)

    Cejka, C.C.

    1990-01-01

    This document contains the functional design criteria for a 26- million-gallon retention basin and 10 million gallons of temporary storage tanks. The basin and tanks will be used to store 242-A Evaporator process condensate, the Plutonium-Uranium Extraction (PUREX) Plant process distillate discharge stream, and the PUREX Plant ammonia scrubber distillate stream. Completion of the project will allow both the 242-A Evaporator and the PUREX Plant to restart. 4 refs

  18. Evaluation of Proposed New LLW Disposal Activity: Disposal of Aqueous PUREX Waste Stream in the Saltstone Disposal Facility

    International Nuclear Information System (INIS)

    Cook, J.R.

    2003-01-01

    The Aqueous PUREX waste stream from Tanks 33 and 35, which have been blended in Tank 34, has been identified for possible processing through the Saltstone Processing Facility for disposal in the Saltstone Disposal Facility

  19. Purex process

    International Nuclear Information System (INIS)

    Starks, J.B.

    1977-01-01

    The following aspects of the Purex Process are discussed: head end dissolution, first solvent extraction cycle, second plutonium solvent extraction cycle, second uranium solvent extraction cycle, solvent recovery systems, primary recovery column for high activity waste, low activity waste, laboratory waste evaporation, vessel vent system, airflow and filtration, acid recovery unit, fume recovery, and discharges to seepage basin

  20. 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

  1. Plutonium uranium extraction (PUREX) end state basis for interim operation (BIO) for surveillance and maintenance

    International Nuclear Information System (INIS)

    DODD, E.N.

    1999-01-01

    This Basis for Interim Operation (BIO) was developed for the PUREX end state condition following completion of the deactivation project. The deactivation project has removed or stabilized the hazardous materials within the facility structure and equipment to reduce the hazards posed by the facility during the surveillance and maintenance (S and M) period, and to reduce the costs associated with the S and M. This document serves as the authorization basis for the PUREX facility, excluding the storage tunnels, railroad cut, and associated tracks, for the deactivated end state condition during the S and M period. The storage tunnels, and associated systems and areas, are addressed in WHC-SD-HS-SAR-001, Rev. 1, PUREX Final Safety Analysis Report. During S and M, the mission of the facility is to maintain the conditions and equipment in a manner that ensures the safety of the workers, environment, and the public. The S and M phase will continue until the final decontamination and decommissioning (D and D) project and activities are begun. Based on the methodology of DOE-STD-1027-92, Hazards Categorization and Accident Analysis Techniques for Compliance with DOE Order 5480.23, Nuclear Safety Analysis Reports, the final facility hazards category is identified as hazards category This considers the remaining material inventories, form and distribution of the material, and the energies present to initiate events of concern. Given the current facility configuration, conditions, and authorized S and M activities, there are no operational events identified resulting in significant hazard to any of the target receptor groups (e.g., workers, public, environment). The only accident scenarios identified with consequences to the onsite co-located workers were based on external natural phenomena, specifically an earthquake. The dose consequences of these events are within the current risk evaluation guidelines and are consistent with the expectations for a hazards category 2

  2. Plutonium uranium extraction (PUREX) end state basis for interim operation (BIO) for surveillance and maintenance

    Energy Technology Data Exchange (ETDEWEB)

    DODD, E.N.

    1999-05-12

    This Basis for Interim Operation (BIO) was developed for the PUREX end state condition following completion of the deactivation project. The deactivation project has removed or stabilized the hazardous materials within the facility structure and equipment to reduce the hazards posed by the facility during the surveillance and maintenance (S and M) period, and to reduce the costs associated with the S and M. This document serves as the authorization basis for the PUREX facility, excluding the storage tunnels, railroad cut, and associated tracks, for the deactivated end state condition during the S and M period. The storage tunnels, and associated systems and areas, are addressed in WHC-SD-HS-SAR-001, Rev. 1, PUREX Final Safety Analysis Report. During S and M, the mission of the facility is to maintain the conditions and equipment in a manner that ensures the safety of the workers, environment, and the public. The S and M phase will continue until the final decontamination and decommissioning (D and D) project and activities are begun. Based on the methodology of DOE-STD-1027-92, Hazards Categorization and Accident Analysis Techniques for Compliance with DOE Order 5480.23, Nuclear Safety Analysis Reports, the final facility hazards category is identified as hazards category This considers the remaining material inventories, form and distribution of the material, and the energies present to initiate events of concern. Given the current facility configuration, conditions, and authorized S and M activities, there are no operational events identified resulting in significant hazard to any of the target receptor groups (e.g., workers, public, environment). The only accident scenarios identified with consequences to the onsite co-located workers were based on external natural phenomena, specifically an earthquake. The dose consequences of these events are within the current risk evaluation guidelines and are consistent with the expectations for a hazards category 2

  3. PUREX transition project case study

    International Nuclear Information System (INIS)

    Jasen, W.G.

    1996-01-01

    In December 1992, the US Department of Energy (DOE) directed that the Plutonium-Uranium Extraction (PUREX) Plant be shut down and deactivated because it was no longer needed to support the nation's production of weapons-grade plutonium. The PUREX/UO 2 Deactivation Project will establish a safe and environmentally secure configuration for the facility and preserve that configuration for 10 years. The 10-year span is used to predict future maintenance requirements and represents the estimated time needed to define, authorize, and initiate the follow-on decontamination and decommissioning activities. Accomplishing the deactivation project involves many activities. Removing major hazards, such as excess chemicals, spent fuel, and residual plutonium are major goals of the project. The scope of the PUREX Transition Project is described within

  4. 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)

  5. PUREX Storage Tunnels dangerous waste permit application

    International Nuclear Information System (INIS)

    1991-12-01

    This report is part of a dangerous waste permit application for the storage of wastes from the Purex process at Hanford. Appendices are presented on the following: construction drawings; HSW-5638, specifications for disposal facility for failed equipment, Project CA-1513-A; HWS-8262, specification for Purex equipment disposal, Project CGC 964; storage tunnel checklist; classification of residual tank heels in Purex storage tunnels; emergency plan for Purex facility; training course descriptions; and the Purex storage tunnels engineering study

  6. PUBG; purex solvent extraction process model. [IBM3033; CDC CYBER175; FORTRAN IV

    Energy Technology Data Exchange (ETDEWEB)

    Geldard, J.F.; Beyerlein, A.L.

    PUBG is a chemical model of the Purex solvent extraction system, by which plutonium and uranium are recovered from spent nuclear fuel rods. The system comprises a number of mixer-settler banks. This discrete stage structure is the basis of the algorithms used in PUBG. The stages are connected to provide for countercurrent flow of the aqueous and organic phases. PUBG uses the common convention that has the aqueous phase enter at the lowest numbered stage and exit at the highest one; the organic phase flows oppositely. The volumes of the mixers are smaller than those of the settlers. The mixers generate a fine dispersion of one phase in the other. The high interfacial area is intended to provide for rapid mass transfer of the plutonium and uranium from one phase to the other. The separation of this dispersion back into the two phases occurs in the settlers. The species considered by PUBG are Hydrogen (1+), Plutonium (4+), Uranyl Oxide (2+), Plutonium (3+), Nitrate Anion, and reductant in the aqueous phase and Hydrogen (1+), Uranyl Oxide (2+), Plutonium (4+), and TBP (tri-n-butylphosphate) in the organic phase. The reductant used in the Purex process is either Uranium (4+) or HAN (hydroxylamine nitrate).IBM3033;CDC CYBER175; FORTRAN IV; OS/MVS or OS/MVT (IBM3033), NOS 1.3 (CDC CYBER175); The IBM3033 version requires 150K bytes of memory for execution; 62,000 (octal) words are required by the CDC CYBER175 version..

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

    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

    Counter-extraction experiments were carried out under the conditions relevant to the partitioning column (IBX) in the purex process to know the path of neptunium present as Np (VI) the organic phase during the partitioning step. The results obtained show that when ferrous sulphamates is used as the reducing agent, most of the neptunium continues to remain with uranium in the organic stream while with hydrazine stabilized uranous nitrate as the reducing agent, a major fraction of neptunium follows the aqueous stream. Mixer-settler experiments were also carried out under the conditions relevant to the uranium purification cycle (2D) to establish the conditions for forcing neptunium to the aqueous raffinate or for partitioning it from uranium if both neptunium and uranium are co-extracted in this cycle and the results obtained are reported here. (auth)

  8. A development and an application of Mixset-X computer code for simulating the Purex solvent extraction system

    International Nuclear Information System (INIS)

    Shida, M.; Naito, M.; Suto, T.; Omori, E.; Nojiri, T.

    2001-01-01

    MIXSET is a FORTRAN code developed to simulate the Purex solvent extraction system using mixer-settler extractors. Japan Nuclear Cycle Development Institute (JNC) has been developing the MIXSET code since the years 1970 to analyze the behavior of nuclides in the solvent extraction processes in Tokai Reprocessing Plant (TRP). This paper describes the history of MIXSET code development, the features of the latest version, called MIXSET-X and the application of the code for safety evaluation work. (author)

  9. Sampling and Analysis Plan for PUREX canyon vessel flushing

    International Nuclear Information System (INIS)

    Villalobos, C.N.

    1995-01-01

    A sampling and analysis plan is necessary to provide direction for the sampling and analytical activities determined by the data quality objectives. This document defines the sampling and analysis necessary to support the deactivation of the Plutonium-Uranium Extraction (PUREX) facility vessels that are regulated pursuant to Washington Administrative Code 173-303

  10. Alternatives for the disposition of PUREX organic solution

    International Nuclear Information System (INIS)

    Nelson, D.W.

    1995-01-01

    This Supporting Document submits options and recommendations for final management of Tank 40 Plutonium-Uranium Extraction (PUREX) Plant organic solution per Tri-Party Agreement Milestorm Number M-80-00-T03. Hanford is deactivating the PUREX Plant for the US DOE. One the key element of this Deactivation is disposition of approximately 81,300 liters (21,500 gallons) of slightly radioactively contaminated organic solution to reduce risk to the environment, reduce cost of long-term storage, and assure regulatory compliance. An announcement in the Commerce Business Daily (CBD) on October 14, 1994 has resulted in the submission of proposals from two facilities capabLe of receiving and thermally destroying the solution. Total decomposition by thermal destruction is the recommended option for the disposition of the PUREX organic solution and WHC is evaluating the proposals from the two facilities

  11. Purex optimization by computer simulation

    International Nuclear Information System (INIS)

    Campbell, T.G.; McKibben, J.M.

    1980-08-01

    For the past 2 years computer simulation has been used to study the performance of several solvent extraction banks in the Purex facility at the Savannah River Plant in Aiken, South Carolina. Individual process parameters were varied about their normal base case values to determine their individual effects on concentration profiles and end-stream compositions. The data are presented in graphical form to show the extent to which product losses, decontamination factors, solvent extraction bank inventories of fissile materials, and other key properties are affected by process changes. Presented in this way, the data are useful for adapting flowsheet conditions to a particular feed material or product specification, and for evaluating nuclear safety as related to bank inventories

  12. DOE Richland readiness review for PUREX

    International Nuclear Information System (INIS)

    Zamorski, M.J.

    1984-01-01

    For ten months prior to the November 1983 startup of the Plutonium and URanium EXtraction (PUREX) Plant, the Department of Energy's Richland Operations Office conducted an operational readiness review of the facility. This review was performed consistent with DOE and RL Order 5481.1 and in accordance with written plans prepared by the program and safety divisions. It involved personnel from five divisions within the office. The DOE review included two tasks: (1) overview and evaluation of the operating contractor's (Rockwell Hanford) readiness review for PUREX, and (2) independent assessment of 25 significant aspects of the startup effort. The RL reviews were coordinated by the program division and were phased in succession with the contractor's readiness review. As deficiencies or concerns were noted in the course of the review they were documented and required formal response from the contractor. Startup approval was given in three steps as the PUREX Plant began operation. A thorough review was performed and necessary documentation was prepared to support startup authorization in November 1983, before the scheduled startup date

  13. Partitioning of actinides from high active waste solution of Purex origin counter-current extraction studies using TBP and CMPO

    International Nuclear Information System (INIS)

    Chitnis, R.R.; Dhami, P.S.; Gopalkrishnan, V.; Wattal, P.K.; Ramanujam, A.; Murali, M.S.; Mathur, J.N.; Bauri, A.K.; Chattopadhyay, S.

    2000-10-01

    A solvent extraction scheme has been formulated for the partitioning of actinides from Purex high level waste (HLW). The scheme is based on the results of earlier studies carried out with simulated waste solutions. In the present studies, the scheme was tested with high active waste (HAW) solution generated during the reprocessing of spent fuel from research reactors using laboratory scale mixer-settlers. The proposed process involved two-step extraction using tri-n-butyl phosphate (TBP) and octyl (phenyl)-N,N-diisobutylcarbamolylmethylphosphine oxide (CMPO). In the first step, uranium, neptunium and plutonium were removed from the waste using TBP as extractant. The minor actinides left in the raffinate were extracted using a mixture of CMPO and TBP in the second step. The results showed complete extraction of actinides from the waste solution. Plutonium and neptunium extracted in TBP, were stripped together using a mixture of hydrogen peroxide and ascorbic acid in 2 M nitric acid medium, leaving uranium in the organic phase. Uranium can later be stripped using dilute nitric acid. Actinides extracted in CMPO-TBP phase were stripped using a mixture of formic acid, hydrazine, hydrate and citric acid. The stripping was quantitative in both the stripping runs. An additional extraction step for the preferential recovery of uranium, neptunium and plutonium from the waste solution using TBP is a modification over the conventional Truex process. Selective stripping of neptunium and plutonium from large quantities of uranium. The extraction of uranium using TBP eliminates the possibility of third phase and undesired loading of CMPO-TBP in the following step. Use of citrate-containing strippant allows the recovery of actinides from loaded CMPO-TBP mixture without causing any reflux of the actinides during stripping. The process has been developed with due consideration to minimising the generation of secondary wastes. The proposed strippants are effective even in presence of

  14. Detection of uranium extraction zone by axial temperature profiles in a pulsed column for Purex process

    International Nuclear Information System (INIS)

    Tsukada, T.; Takahashi, K.

    1991-01-01

    A new method was presented for detecting uranium extraction zone in a pulsed column by means of measuring axial temperature profile originated from reaction heat during uranium extraction. Key parameters of the temperature profiles were estimated with a code developed for calculating temperature profiles in a direct-contact heat exchanger such as a pulsed column, and were verified using data from a small pulsed column simulating reaction heat with injecting hot water. Finally, the results were compared with those from an actual uranium extraction tests, indicating that the method presented was promising for detecting uranium extraction zone in a pulsed column. (author)

  15. Modifications of the SEPHIS computer code for calculating the Purex solvent extraction system

    International Nuclear Information System (INIS)

    Watson, S.B.; Rainey, R.H.

    1975-12-01

    The SEPHIS computer program was developed to simulate the countercurrent solvent extraction. This report gives modifications in the program which result in improved fit to experimental data, a decrease in computer storage requirements, and a decrease in execution time. Methods for applying the computer program to practical solvent extraction problems are explained

  16. Purex process solvent: literature review

    Energy Technology Data Exchange (ETDEWEB)

    Geier, R.G.

    1979-10-01

    This document summarizes the data on Purex process solvent presently published in a variety of sources. Extracts from these various sources are presented herein and contain the work done, the salient results obtained, and the original, unaltered conclusions of the author of each paper. Three major areas are addressed: solvent stability, solvent quality testing, and solvent treatment processes. 34 references, 44 tables.

  17. Purex process solvent: literature review

    International Nuclear Information System (INIS)

    Geier, R.G.

    1979-10-01

    This document summarizes the data on Purex process solvent presently published in a variety of sources. Extracts from these various sources are presented herein and contain the work done, the salient results obtained, and the original, unaltered conclusions of the author of each paper. Three major areas are addressed: solvent stability, solvent quality testing, and solvent treatment processes. 34 references, 44 tables

  18. Characteristics and behaviour of interface sludges formed in the first extraction cycles of the purex process

    International Nuclear Information System (INIS)

    Gue, J.P.; Philippe, M.; Masson, M.

    1990-01-01

    The importance of clarification for the volume of sludges has been demonstrated. However, even very good clarification does not totally eliminate the extraction sludges. If their volume is considerably increased by the presence of dissolution fines that have escaped clarification, this only slightly alters the hydrodynamic behaviour of the pulsed column in the first cycles. The sludges obtained with and without feed clarification have been characterized and their origin recalled. The management of these bulky remains problematic for reprocessing

  19. PUREX storage tunnels waste analysis plan

    International Nuclear Information System (INIS)

    Haas, C.R.

    1995-01-01

    Washington Administrative Code 173-303-300 requires that a facility develop and follow a written waste analysis plan which describes the procedures that will be followed to ensure that its dangerous waste is managed properly. This document covers the activities at the PUREX Storage Tunnels used to characterize and designate waste that is generated within the PUREX plant, as well as waste received from other on-site sources

  20. PUREX storage tunnels waste analysis plan

    International Nuclear Information System (INIS)

    Haas, C.R.

    1996-01-01

    Washington Administrative Code 173-303-300 requires that a facility develop and follow a written waste analysis plan which describes the procedures that will be followed to ensure that its dangerous waste is managed properly. This document covers the activities at the PUREX Storage Tunnels used to characterize and designate waste that is generated within the PUREX Plant, as well as waste received from other on-site sources

  1. 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

  2. PUREX Storage Tunnels dangerous waste permit application

    International Nuclear Information System (INIS)

    1991-12-01

    The PUREX Storage Tunnels are a mixed waste storage unit consisting of two underground railroad tunnels: Tunnel Number 1 designated 218-E-14 and Tunnel Number 2 designated 218-E-15. The two tunnels are connected by rail to the PUREX Plant and combine to provide storage space for 48 railroad cars (railcars). The PUREX Storage Tunnels provide a long-term storage location for equipment removed from the PUREX Plant. Transfers into the PUREX Storage Tunnels are made on an as-needed basis. Radioactively contaminated equipment is loaded on railcars and remotely transferred by rail into the PUREX Storage Tunnels. Railcars act as both a transport means and a storage platform for equipment placed into the tunnels. This report consists of part A and part B. Part A reports on amounts and locations of the mixed water. Part B permit application consists of the following: Facility Description and General Provisions; Waste Characteristics; Process Information; Groundwater Monitoring; Procedures to Prevent Hazards; Contingency Plan; Personnel Training; Exposure Information Report

  3. Integrating safety and health during deactiviation: With lessons learned from PUREX

    International Nuclear Information System (INIS)

    1995-01-01

    This report summarizes an integrated safety and health approach used during facility deactivation activities at the Department of Energy (DOE) Plutonium-Uranium Extraction (PUREX) Facility in Hanford, Washington. Resulting safety and health improvements and the potential, complex-wide application of this approach are discussed in this report through a description of its components and the impacts, or lessons-learned, of its use during the PUREX deactivation project. As a means of developing and implementing the integrated safety and health approach, the PUREX technical partnership was established in 1993 among the Office of Environment, Safety and Health's Office of Worker Health and Safety (EH-5); the Office of Environmental Management's Offices of Nuclear Material and Facility Stabilization (EM-60) and Compliance and Program Coordination (EM-20); the DOE Richland Operations Office; and the Westinghouse Hanford Company. It is believed that this report will provide guidance for instituting an integrated safety and health approach not only for deactivation activities, but for decommissioning and other clean-up activities as well. This confidence is based largely upon the rationality of the approach, often termed as common sense, and the measurable safety and health and project performance results that application of the approach produced during actual deactivation work at the PUREX Facility

  4. Regulatory Support of Treatment of Savannah River Site Purex Waste

    International Nuclear Information System (INIS)

    Reid, L.T.

    2009-01-01

    This paper describes the support given by federal and state regulatory agencies to Savannah River Site (SRS) during the treatment of an organic liquid mixed waste from the Plutonium Extraction (Purex) process. The support from these agencies allowed (SRS) to overcome several technical and regulatory barriers and treat the Purex waste such that it met LDR treatment standards. (authors)

  5. Prediction equations for corrosion rates of a A-537 and A-516 steels in Double Shell Slurry, Future PUREX, and Hanford Facilities Wastes

    International Nuclear Information System (INIS)

    Divine, J.R.; Bowen, W.M.; Mackey, D.B.; Bates, D.J.; Pool, K.H.

    1985-06-01

    Even though the interest in the corrosion of radwaste tanks goes back to the mid-1940's when waste storage was begun, and a fair amount of corrosion work has been done since then, the changes in processes and waste types have outpaced the development of new data pertinent to the new double shell tanks. As a consequence, Pacific Northwest Laboratory (PNL) began a development of corrosion data on a broad base of waste compositions in 1980. The objective of the program was to provide operations personnel with corrosion rate data as a function of waste temperature and composition. The work performed in this program examined A-537 tank steel in Double Shell Slurry and Future PUREX Wastes, at temperatures between 40 and 180 0 C as well as in Hanford Facilities Waste at 25 and 50 0 C. In general, the corrosion rates were less than 1 mpy (0.001 in./y) and usually less than 0.5 mpy. Excessive corrosion rates (>1 mpy) were only found in dilute waste compositions or in concentrated caustic compositions at temperatures above 140 0 C. Stress corrosion cracking was only observed under similar conditions. The results are presented as polynomial prediction equations with examples of the output of existing computer codes. The codes are not provided in the text but are available from the authors. 12 refs., 5 figs., 19 tabs

  6. Engineering Phase 2 and Phase 3 certification programs -- PUREX deactivation

    International Nuclear Information System (INIS)

    Walser, R.L.

    1994-01-01

    This document describes the training programs required to become a Phase 2 and Phase 3 certified engineer at PUREX during deactivation. With the change in mission, the PUREX engineering/certification training program is being revamped as discussed below. The revised program will be administered by PUREX Technical Training using existing courses and training materials. The program will comply with the requirements of the Department of Energy (DOE) order 5480.20A, ''Personnel Selection, Qualification, Training, and Staffing Requirements at DOE Reactor and Non-Reactor Nuclear Facilities.''

  7. Engineering Phase 2 and Phase 3 certification programs -- PUREX deactivation

    Energy Technology Data Exchange (ETDEWEB)

    Walser, R.L.

    1994-12-13

    This document describes the training programs required to become a Phase 2 and Phase 3 certified engineer at PUREX during deactivation. With the change in mission, the PUREX engineering/certification training program is being revamped as discussed below. The revised program will be administered by PUREX Technical Training using existing courses and training materials. The program will comply with the requirements of the Department of Energy (DOE) order 5480.20A, ``Personnel Selection, Qualification, Training, and Staffing Requirements at DOE Reactor and Non-Reactor Nuclear Facilities.``

  8. Electrical/instrumentation acceptance test report for Project C-018H, 242-A Evaporator/PUREX Plant condensate treatment facility

    International Nuclear Information System (INIS)

    Compau, R.A. Jr.

    1995-01-01

    This project is part of the 200 Area Effluent Treatment Facility. The acceptance test procedure describes test methods for leak detection units, pump flow switches, pump level control valves, room air temperature monitor, leachate pump status contacts, basin pump status contacts, catch basin leak detector, leachate level monitors, and basin level monitors. These are all components of the C-018H Collection System

  9. Acceptance test procedure for C-018H, 242-A evaporator/PUREX plant process condensate treatment facility

    International Nuclear Information System (INIS)

    Parrish, D.E.

    1994-01-01

    This Acceptance Test Procedure (ATP) has been prepared to demonstrate that the Electrical/Instrumentation system function as required for this facility. Each company or organization participating in this ATP will designate personnel to assume the responsibilities and duties as defined herein for their respective roles

  10. PUREX Storage Tunnels waste analysis plan. Revision 1

    International Nuclear Information System (INIS)

    Stephenson, M.J.

    1995-11-01

    Washington Administrative Code 173-303-300 requires that a facility develop and follow a written waste analysis plan which describes the procedures that will be followed to ensure that its dangerous waste is managed properly. This document covers the activities at the PUREX Storage Tunnels used to characterize and designate waste that is generated within the PUREX Plant, as well as waste received from other on-site sources

  11. 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

  12. PUREX new substation ATR

    International Nuclear Information System (INIS)

    Nelson, D.E.

    1997-01-01

    This document is the acceptance test report (ATR) for the New PUREX Main and Minisubstations. It covers the factory and vendor acceptance and commissioning test reports. Reports are presented for the Main 5 kV substation building, the building fire system, switchgear, and vacuum breaker; the minisubstation control building and switch gear; commissioning test; electrical system and loads inspection; electrical utilities transformer and cable; and relay setting changes based on operational experience

  13. 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

  14. Neptunium determination in PUREX process

    International Nuclear Information System (INIS)

    Rawat, Neetika; Kar, Aishwarya S.; Tomar, B.S.; Pandey, M.P.; Umadevi, K.

    2016-10-01

    237 Np is one of the most important minor actinides present in nuclear spent fuel both from environmental and application point of view. The routing of neptunium to the particular stream of PUREX process is necessary for its separation and purification as 237 Np is the target nuclide for production of 238 Pu. The routing of neptunium to a particular PUREX stream will also help in better nuclear waste management, which in turn, will impart less bearing on the environment considering its long half life, alpha emitting properties and mobile nature. In order to route Neptunium to a particular stream of PUREX process, it is imperative to understand the distribution of neptunium in various process streams. Owing to high dose of actual samples, the neptunium distribution was studied using 239 Np tracer by simulating actual column conditions of PUREX streams in lab scale. The present study deals with neptunium determination in actual PUREX streams samples also. (author)

  15. 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.)

  16. Radioactive air emissions notice of construction for deactivation of the PUREX storage tunnel number 2; FINAL

    International Nuclear Information System (INIS)

    JOHNSON, R.E.

    1999-01-01

    The Plutonium-Uranium Extraction (PUREX) Plant Storage Tunnel Number 2 (hereafter referred to as the PUREX Tunnel) was built in 1964. Since that time, the PUREX Tunnel has been used for storage of radioactive and mixed waste. In 1991, the PUREX Plant ceased operations and was transitioned to deactivation. The PUREX Tunnel continued to receive PUREX Plant waste material for storage during transition activities. Before 1995, a decision was made to store radioactive and mixed waste in the PUREX Tunnel generated from other onsite sources, on a case-by-case basis. This notice of construction (NOC) describes the activities associated with the reactivation of the PUREX Tunnel ventilation system and the transfer of up to 3.5 million curies (MCi) of radioactive waste to the PUREX Tunnel from any location on the Hanford Site. The unabated total effective dose equivalent (TEDE) estimated for the hypothetical offsite maximally exposed individual (MEI) is 5.6 E-2 millirem (mrem). The abated TEDE conservatively is estimated to account for 1.9 E-5 mrem to the MEI. The following text provides information requirements of Appendix A of Washington Administrative Code (WAC) 246-247 (requirements 1 through 18)

  17. PUREX source Aggregate Area management study report

    International Nuclear Information System (INIS)

    1993-03-01

    This report presents the results of an aggregate area management study (AAMS) for the PUREX Plant Aggregate Area in the 200 Areas of the US Department of Energy (DOE)Hanford Site in Washington State. This scoping level study provides the basis for initiating Remedial Investigation/Feasibility Study (RI/FS) activities under the comprehensive Environmental Response, Compensation, and Liability Act of 1980 (CERCLA) or Resource Conservation and Recovery Act (RCRA) Facility Investigations (RFI) and Corrective Measures Studies (CMS) under RCRA. This report also integrates select RCRA treatment, storage, or disposal (TSD) closure activities with CERCLA and RCRA past-practice investigations

  18. 1997 project of the year, PUREX deactivation project

    International Nuclear Information System (INIS)

    Bailey, R.W.

    1998-01-01

    At the end of 1992, the PUREX and UO 3 plants were deemed no longer necessary for the defense needs of the United States. Although no longer necessary, they were very costly to maintain in their post-operation state. The DOE embarked on a deactivation strategy for these plants to reduce the costs of providing continuous surveillance of the facilities and their hazards. Deactivation of the PUREX and UO 3 plants was estimated to take 5 years and cost $222.5 million and result in an annual surveillance and maintenance cost of $2 million. Deactivation of the PUREX/UO 3 plants officially began on October 1, 1993. The deactivation was 15 months ahead of the original schedule and $75 million under the original cost estimate. The annual cost of surveillance and maintenance of the plants was reduced to less than $1 million

  19. TBP and diluent mass balances in the PUREX Plant at Hanford, 1955--1991

    International Nuclear Information System (INIS)

    Sederburg, J.P.; Reddick, J.A.

    1994-12-01

    The purpose of this report is to develop an estimate of the quantities of tributyl phosphate and diluent discharged in aqueous waste streams to the tank farms from the Hanford Purex Plant over its operating life. Purex was not the sole source of organics in the tank farms, but was a major contributor. Tributyl phosphate (TBP) and diluent, which changed from Shell E-2342 reg-sign to Soltrol-170 reg-sign and then to normal paraffin hydrocarbon (NPH), were organic chemicals used in the Purex solvent extraction process at Hanford to separate plutonium and uranium from spent nuclear fuels. This report is an estimate of the material balances for these chemicals in the Purex Plant at Hanford over its entire operating life. The Purex Plant had cold start up in November 1955 and shut down in 1990. It's process used a solution of 30 vol% TBP in diluent

  20. Some plutonium IV polymers properties in Purex process

    International Nuclear Information System (INIS)

    Scoazec, H.; Pasquiou, J.Y.; Germain, M.

    1990-01-01

    The metabolism of plutonium polymers in fuel reprocessing using the Purex process with tributylphosphate as solvent, and its practical consequence in real operation conditions are examined. Precipitation with dibutylphosphoric acid, a solvent degradation product, occurs both in extraction and stripping units when polymers are present. (author)

  1. Review of the Tritium Extraction Facility design

    International Nuclear Information System (INIS)

    Barton, R.W.; Bamdad, F.; Blackman, J.

    2000-01-01

    The Defense Nuclear Facilities Safety Board (DNFSB) is an independent executive branch agency responsible for technical safety oversight of the US Department of Energy's (DOE's) defense nuclear facilities. One of DNFSB's responsibilities is the review of design and construction projects for DOE's defense nuclear facilities to ensure that adequate health and safety requirements are identified and implemented. These reviews are performed with the expectation that facility designs are being developed within the framework of a site's Integrated Safety Management (ISM) program. This paper describes the application of ISM principles in DNFSB's ongoing review of the Tritium Extraction Facility (TEF) design/construction project

  2. Review of the Tritium Extraction Facility Design

    International Nuclear Information System (INIS)

    Ronald W. Barton; Farid Bamdad; Joel Blackman

    2000-01-01

    The Defense Nuclear Facilities Safety Board (DNFSB) is an independent executive branch agency responsible for technical safety oversight of the U.S. Department of Energy's (DOE's) defense nuclear facilities. One of DNFSB's responsibilities is the review of design and construction projects for DOE's defense nuclear facilities to ensure that adequate health and safety requirements are identified and implemented. These reviews are performed with the expectation that facility designs are being developed within the framework of a site's Integrated Safety Management (ISM) program. This paper describes the application of ISM principles in DNFSB's ongoing review of the Tritium Extraction Facility (TEF) design/construction project

  3. PUREX/UO3 deactivation project management plan

    International Nuclear Information System (INIS)

    Washenfelder, D.J.

    1993-12-01

    From 1955 through 1990, the Plutonium-Uranium Extraction Plant (PUREX) provided the United States Department of Energy Hanford Site with nuclear fuel reprocessing capability. It operated in sequence with the Uranium Trioxide (UO 3 ) Plant, which converted the PUREX liquid uranium nitrate product to solid UO 3 powder. Final UO 3 Plant operation ended in 1993. In December 1992, planning was initiated for the deactivation of PUREX and UO 3 Plant. The objective of deactivation planning was to identify the activities needed to establish a passively safe, environmentally secure configuration at both plants, and ensure that the configuration could be retained during the post-deactivation period. The PUREX/UO 3 Deactivation Project management plan represents completion of the planning efforts. It presents the deactivation approach to be used for the two plants, and the supporting technical, cost, and schedule baselines. Deactivation activities concentrate on removal, reduction, and stabilization of the radioactive and chemical materials remaining at the plants, and the shutdown of the utilities and effluents. When deactivation is completed, the two plants will be left unoccupied and locked, pending eventual decontamination and decommissioning. Deactivation is expected to cost $233.8 million, require 5 years to complete, and yield $36 million in annual surveillance and maintenance cost savings

  4. PUREX Plant deactivation mission analysis report

    International Nuclear Information System (INIS)

    Lund, D.P.

    1995-01-01

    The purpose of the PUREX Deactivation Project mission analysis is to define the problem to be addressed by the PUREX mission, and to lay the ground work for further system definition. The mission analysis is an important first step in the System Engineering (SE) process. This report presents the results of the PUREX Deactivation Project mission analysis. The purpose of the PUREX Deactivation Project is to prepare PUREX for Decontamination and Decommissioning within a five year time frame. This will be accomplished by establishing a passively safe and environmentally secure configuration of the PUREX Plant, that can be preserved for a 10-year horizon. During deactivation, appropriate portions of the safety envelop will be maintained to ensure deactivation takes place in a safe and regulatory compliant manner

  5. Control measurement system in purex process

    International Nuclear Information System (INIS)

    Mani, V.V.S.

    1985-01-01

    The dependence of a bulk facility handling Purex Process on the control measurement system for evaluating the process performance needs hardly be emphasized. process control, Plant control, inventory control and quality control are the four components of the control measurement system. The scope and requirements of each component are different and the measurement methods are selected accordingly. However, each measurement system has six important elements. These are described in detail. The quality assurance programme carried out by the laboratory as a mechanism through which the quality of measurements is regularly tested and stated in quantitative terms is also explained in terms of internal and external quality assurance, with examples. Suggestions for making the control measurement system more responsive to the operational needs in future are also briefly discussed. (author)

  6. Calculation code revised MIXSET for Purex process

    International Nuclear Information System (INIS)

    Gonda, Kozo; Oka, Koichiro; Fukuda, Shoji.

    1979-02-01

    Revised MIXSET is a FORTRAN IV calculation code developed to simulate steady and transient behaviors of the Purex extraction process and calculate the optimum operating condition of the process. Revised MIXSET includes all the functions of MIXSET code as shown below. a) Maximum chemical system of eight components can be handled with or without mutual dependence of the distribution of components. b) The flowrate and concentration of feed can be renewed successively at any state, transient or steady, for searching optimum operating conditions. c) Optimum inputs of feed concentrations and flowrates can be calculated to satisfy both of specification and recovery rate of a product. d) Radioactive decay reactions can be handled on each component. Besides these functions, the following chemical reactions concerned in Purex process are newly-included in Revised MIXSET code and the quantitative changes of components such as H + , U(IV), U(VI), Pu(III), Pu(IV), NH 2 OH, N 2 H 4 can be simulated. 1st Gr. (i) reduction of Pu(IV); U 4+ + 2Pu 4+ + 2H 2 O → UO 2 2+ + 2Pu 3+ + 4H + . (ii) oxidation of Pu(III); 2Pu 3+ + 3H + + NO 3 - → 2Pu 4+ + HNO 2 + H 2 O. (iii) oxidation of U(IV); U 4+ + NO 3 - + H 2 O → UO 2 2+ + H + + HNO 2 2U 4+ + O 2 + 2H 2 O → 2UO 2 2+ + 4H + . (iv) decomposition of HNO 2 ; HNO 2 + N 2 H 5 + → HN 3 + 2H 2 O + H + . (author)

  7. PUREX irradiated fuel recovery simulation

    International Nuclear Information System (INIS)

    Jaquish, W.R.

    1994-09-01

    This paper discusses the application of IGRIP (Interactive Graphical Robot Instruction Program) to assist environmental remediation efforts at the Department of Energy PUREX Plant at the Hanford Site. An IGRIP simulation was developed to plan, review, and verify proposed remediation activities. This simulation was designed to satisfy a number of unique purposes that each placed specific constraints and requirements on the design and implementation of the simulation. These purposes and their influence on the design of the simulation are presented. A discussion of several control code architectures for mechanical system simulations, including their advantages and limitations, is also presented

  8. Calibrations of a tritium extraction facility

    International Nuclear Information System (INIS)

    Bretscher, M.M.; Oliver, B.M.; Farrar, H. IV.

    1983-01-01

    A tritium extraction facility has been built for the purpose of measuring the absolute tritium concentration in neutron-irradiated lithium metal samples. Two independent calibration procedures have been used to determine what fraction, if any, of tritium is lost during the extraction process. The first procedure compares independently measured 4 He and 3 H concentrations from the 6 Li(n,α)T reaction. The second procedure compared measured 6 Li(n,α)T/ 197 Au (n,γ) 198 Au thermal neutron reaction rate ratios with those obtained from Monte Carlo calculations using well-known cross sections. Both calibration methods show that within experimental errors (approx. 1.5%) no tritium is lost during the extraction process

  9. 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.)

  10. TRU decontamination of high-level Purex waste by solvent extraction using a mixed octyl(phenyl)-N,N-diisobutyl-carbamoylmethylphosphine oxide/TBP/NPH (TRUEX) solvent

    International Nuclear Information System (INIS)

    Horwitz, E.P.; Kalina, D.G.; Diamond, H.; Kaplan, L.; Vandegrift, G.F.; Leonard, R.A.; Steindler, M.J.; Schulz, W.W.

    1984-01-01

    The TRUEX (transuranium extraction) process was tested on a simulated high-level dissolved sludge waste (DSW). A batch counter-current extraction mode was used for seven extraction and three scrub stages. One additional extraction stage and two scrub stages and all strip stages were performed by batch extraction. The TRUEX solvent consisted of 0.20 M octyl(phenyl)-N,N-diisobutylcarbamoyl-methylphosphine oxide-1.4 M TBP in Conoco (C 12 -C 14 ). The feed solution was 1.0 M in HNO 3 , 0.3 M in H 2 C 2 O 4 and contained mixed (stable) fission products, U, Np, Pu, and Am, and a number of inert constituents, e.g., Fe and Al. The test showed that the process is capable of reducing the TRU concentration in the DSW by a factor of 4 x 10 4 (to <100 nCi/g of disposed form) and reducing the quantity of TRU waste by two orders of magnitude

  11. PUREX Deactivation Health and Safety documentation

    International Nuclear Information System (INIS)

    Dodd, E.N. III.

    1995-01-01

    The purpose of the PUREX Deactivation Project is to establish a passively safe and environmentally secure configuration of PUREX at the Hanford Site, and to preserve that configuration for a 10-year horizon. The 10-year horizon is used to predict future maintenance requirements and represents they typical time duration expended to define, authorize, and initiate the follow-on Decontamination and Decommissioning (D ampersand D) activities. This document was prepared to increase attention to worker safety issues during the deactivation project and, as such, identifies the documentation and programs associated with PUREX Deactivation Health and Safety

  12. PUREX Deactivation Health and Safety documentation

    Energy Technology Data Exchange (ETDEWEB)

    Dodd, E.N. III

    1995-01-01

    The purpose of the PUREX Deactivation Project is to establish a passively safe and environmentally secure configuration of PUREX at the Hanford Site, and to preserve that configuration for a 10-year horizon. The 10-year horizon is used to predict future maintenance requirements and represents they typical time duration expended to define, authorize, and initiate the follow-on Decontamination and Decommissioning (D&D) activities. This document was prepared to increase attention to worker safety issues during the deactivation project and, as such, identifies the documentation and programs associated with PUREX Deactivation Health and Safety.

  13. Pretreatment of Hanford purex plant first-cycle waste

    International Nuclear Information System (INIS)

    Gibson, M.W.; Gerboth, D.M.; Peters, B.B.

    1987-01-01

    A process has been developed to pretreat neutralized, first-cycle high-level waste from the fuels reprocessing facility (PUREX Plant) at the Hanford Site. The process separates solids from the supernate liquid, which contains soluble salts. The solids, including most of the fission products and transuranic elements, may then be vitrified for disposal, while the low-level supernate stream may be processed into a less expensive grout waste form. The process also includes ion exchange treatment of the separated supernate stream to remove radiocesium. A flow sheet based on these operations was completed to support a planned demonstration of the process in the Hanford site B Plant canyon facility

  14. Characterization of past and present solid waste streams from the Plutonium-Uranium Extraction Plant

    International Nuclear Information System (INIS)

    Pottmeyer, J.A.; Weyns, M.I.; Lorenzo, D.S.; Vejvoda, E.J.; Duncan, D.R.

    1993-04-01

    During the next two decades the transuranic wastes, now stored in the burial trenches and storage facilities at the Hanford Site, are to be retrieved, processed at the Waste Receiving and Processing Facility, and shipped to the Waste Isolation Pilot Plant near Carlsbad, New Mexico for final disposal. Over 7% of the transuranic waste to be retrieved for shipment to the Waste Isolation Pilot Plant has been generated at the Plutonium-Uranium Extraction (PUREX) Plant. The purpose of this report is to characterize the radioactive solid wastes generated by PUREX using process knowledge, existing records, and oral history interviews. The PUREX Plant is currently operated by the Westinghouse Hanford Company for the US Department of Energy and is now in standby status while being prepared for permanent shutdown. The PUREX Plant is a collection of facilities that has been used primarily to separate plutonium for nuclear weapons from spent fuel that had been irradiated in the Hanford Site's defense reactors. Originally designed to reprocess aluminum-clad uranium fuel, the plant was modified to reprocess zirconium alloy clad fuel elements from the Hanford Site's N Reactor. PUREX has provided plutonium for research reactor development, safety programs, and defense. In addition, the PUREX was used to recover slightly enriched uranium for recycling into fuel for use in reactors that generate electricity and plutonium. Section 2.0 provides further details of the PUREX's physical plant and its operations. The PUREX Plant functions that generate solid waste are as follows: processing operations, laboratory analyses and supporting activities. The types and estimated quantities of waste resulting from these activities are discussed in detail

  15. PUREX Plant deactivation function analysis report

    International Nuclear Information System (INIS)

    Lund, D.P.

    1995-09-01

    The document contains the functions, function definitions, function interfaces, function interface definitions, Input Computer Automated Manufacturing Definition (IDEFO) diagrams, and a function hierarchy chart that describe what needs to be performed to deactivate PUREX

  16. Continuous plutonium(IV) oxalate precipitation, filtration, and calcination process. [From product streams from Redox, Purex, or Recuplex solvent extraction plants

    Energy Technology Data Exchange (ETDEWEB)

    Beede, R L

    1956-09-27

    A continuous plutonium (IV) oxalate precipitation, filtration, and calcination process has been developed. Continuous and batch decomposition of the oxalate in the filtrates has been demonstrated. The processes have been demonstrated in prototype equipment. Plutonium (IV) oxalate was precipitated continuously at room temperature by the concurrent addition of plutonium (IV) nitrate feed and oxalic acid into the pan of a modified rotary drum filter. The plutonium (IV) oxalate was calcined to plutonium dioxide, which could be readily hydrofluorinated. Continuous decomposition of the oxalate in synthetic plutonium (IV) oxalate filtrates containing plutonium (IV) oxalate solids was demonstrated using co-current flow in a U-shaped reactor. Feeds containing from 10 to 100 g/1 Pu, as plutonium (IV) nitrate, and 1.0 to 6.5 M HNO/sub 3/, respectively, can be processed. One molar oxalic acid is used as the precipitant. Temperatures of 20 to 35/sup 0/C for the precipitation and filtration are satisfactory. Plutonium (IV) oxalate can be calcined at 300 to 400/sup 0/C in a screw-type drier-calciner to plutonium dioxide and hydrofluorinated at 450 to 550/sup 0/C. Plutonium dioxide exceeding purity requirements has been produced in the prototype equipment. Advantages of continuous precipitation and filtration are: uniform plutonium (IV) oxalate, improved filtration characteristics, elimination of heating and cooling facilities, and higher capacities through a single unit. Advantages of the screw-type drier-calciner are the continuous production of an oxide satisfactory for feed for the proposed plant vibrating tube hydrofluorinator, and ease of coupling continuous precipitation and filtration to this proposed hydrofluorinator. Continuous decomposition of oxalate in filtrates offers advantages in decreasing filtrate storage requirements when coupled to a filtrate concentrator. (JGB)

  17. 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)

  18. Ion exchange flowsheet for recovery of cesium from purex sludge supernatant at B Plant

    International Nuclear Information System (INIS)

    Carlstrom, R.F.

    1977-01-01

    Purex Sludge Supernatant (PSS) contains significant amounts of 137 Cs left after removal of strontium from fission product bearing Purex wastes. To remove cesium from PSS, an Ion Exchange Recovery system has been set up in Cells 17-21 at B Plant. The cesium that is recovered is stored within B Plant for eventual purification through the Cesium Purification process in Cell 38 and eventual encapsulation and storage in a powdered form at the Waste Encapsulation Storage Facility. Cesium depleted waste streams from the Ion Exchange processes are transferred to underground storage

  19. Disposition of PUREX contaminated nitric acid the role of stakeholder involvement

    International Nuclear Information System (INIS)

    Jasen, W.G.; Duncan, R.A.

    1996-01-01

    What does the United States space shuttle and the Hanford PUREX facility's contaminated nitric acid have in common. Both are reusable. The PUREX Transition Project has achieved success and, minimized project expenses and waste generation by looking at excess chemicals not as waste but as reusable substitutes for commercially available raw materials. This philosophy has helped PUREX personnel to reuse or recycle more than 2.5 million pounds of excess chemicals, a portion of which is the slightly contaminated nitric acid. After extensive public review, the first shipment of contaminated acid was made in May 1995. Removal of the acid was completed on November 6, 1995 when the fiftieth shipment left the Hanford site. This activity, which avoided dispositioning the contaminated acid as a waste, generated significantly more public input and concern than was expected. One of the lessons learned from this process is to not underestimate public perceptions regarding the reuse of contaminated materials

  20. Canyon Facilities

    Data.gov (United States)

    Federal Laboratory Consortium — B Plant, T Plant, U Plant, PUREX, and REDOX (see their links) are the five facilities at Hanford where the original objective was plutonium removal from the uranium...

  1. PUREX exhaust ventilation system installation test report

    International Nuclear Information System (INIS)

    Blackaby, W.B.

    1997-01-01

    This Acceptance Test Report validates the testing performed, the exceptions logged and resolved and certifies this portion of the SAMCONS has met all design and test criteria to perform as an operational system. The proper installation of the PUREX exhaust ventilation system components and wiring was systematically evaluated by performance of this procedure. Proper operation of PUREX exhaust fan inlet, outlet, and vortex damper actuators and limit switches were verified, using special test equipment, to be correct and installed wiring connections were verified by operation of this equipment

  2. Determination of hydroxylamine in purex process solutions

    International Nuclear Information System (INIS)

    Ertel, D.; Weindel, P.

    1984-05-01

    In PUREX process solutions hydroxylamine or HAN (hydrolammonium nitrate) respectively, can be oxidized specifically to give nitrous acid, HNO 2 , which by sybsequent GRIESS reaction forms the well-known reddish azo-dye. Its absorbance is spectrophotometrically measured at 520 nm and results in linear calibration graphs covering the analytical range of 10 -5 to 10 -6 M NH 2 OH. The influence of other reductants (N 2 H 4 , Pu-III) as well as of further PUREX main constituents like U-VI, HNO 3 etc. was checked-up and determined quantitatively. There are no analytical limitations in case of HAN concentrations > 10 -2 M. (orig.) [de

  3. An advanced purex process based on salt-free reductants

    Energy Technology Data Exchange (ETDEWEB)

    He, Hui; Ye, Guoan; Tang, Hongbin; Zheng, Weifang; Li, Gaoliang; Lin, Rushan [China Institute of Atomic Energy, Beijing (China). Dept. of Radiochemistry

    2014-04-01

    An advanced plutonium and uranium recovery process has been established based on two organic reductants, N,N-dimethylhydroxylamine (DMHAN) and methylhydrazine (MH), as U/Pu separation reagents. This Advanced Purex process based on Organic Reductants (APOR) is composed of three cycles, including U/Pu co-decontamination/separation cycle, uranium purification cycle and plutonium purification cycle. Using DMHAN and MH as plutonium stripping reagents in the U/Pu co-decontamination/separation cycle and plutonium purification cycle, the APOR process exhibits high performance with following highlights: (1) the process is much simpler because of the elimination of Tc scrubbing operation and the supplement extraction operation, (2) high efficiency of U/Pu separation can be achieved in the first cycle, (3) plutonium product solution of high concentration can be obtained in the Pu purification cycle with a simple extraction operation instead of circumfluent extraction or evaporation of the plutonium solution. (orig.)

  4. Extraction and beam transfer for the SHiP facility

    CERN Document Server

    Goddard, Brennan; Borburgh, Jan; Balhan, Bruno; Le Godec, Gilles; Zerlauth, Markus; Tommasini, Davide; Kain, Verena; Cornelis, Karel; Wenninger, Jorg; Jensen, Lars; Todd, Benjamin; Bauche, Jeremie; Puccio, Bruno

    2015-01-01

    This document summarises the key feasibility issues associated with the SPS extraction and beam transfer systems required for the SHiP facility. It describes the expected performance limits of the electrostatic septa, the expected beam losses during extraction and consequences, the design of the new beamline geometry and equipment systems and the expected extracted spill structure.

  5. Facility effluent monitoring plan for the plutonium uranium extraction facility

    Energy Technology Data Exchange (ETDEWEB)

    Wiegand, D.L.

    1994-09-01

    A facility effluent monitoring plan is required by the US Department of Energy in DOE Order 5400.1 for any operations that involve hazardous materials and radioactive substances that could impact employee or public safety or the environment. This document is prepared using the specific guidelines identified in A Guide for Preparing Hanford Site Facility Effluent Monitoring Plans, WHC-EP-0438-01. This facility effluent monitoring plan assesses effluent monitoring systems and evaluates whether they are adequate to ensure the public health and safety as specified in applicable federal, state, and local requirements. This facility effluent monitoring plan shall ensure long-range integrity of the effluent monitoring systems by requiring an update whenever a new process or operation introduces new hazardous materials or significant radioactive materials. This document must be reviewed annually even if there are no operational changes, and it must be updated at a minimum of every three years.

  6. Facility effluent monitoring plan for the plutonium uranium extraction facility

    International Nuclear Information System (INIS)

    Wiegand, D.L.

    1994-09-01

    A facility effluent monitoring plan is required by the US Department of Energy in DOE Order 5400.1 for any operations that involve hazardous materials and radioactive substances that could impact employee or public safety or the environment. This document is prepared using the specific guidelines identified in A Guide for Preparing Hanford Site Facility Effluent Monitoring Plans, WHC-EP-0438-01. This facility effluent monitoring plan assesses effluent monitoring systems and evaluates whether they are adequate to ensure the public health and safety as specified in applicable federal, state, and local requirements. This facility effluent monitoring plan shall ensure long-range integrity of the effluent monitoring systems by requiring an update whenever a new process or operation introduces new hazardous materials or significant radioactive materials. This document must be reviewed annually even if there are no operational changes, and it must be updated at a minimum of every three years

  7. Facility effluent monitoring plan for the Plutonium Uranium Extraction Facility

    International Nuclear Information System (INIS)

    Greager, E.M.

    1997-01-01

    A facility effluent monitoring plan is required by the US Department of Energy in DOE Order 5400.1 for any operations that involve hazardous materials and radioactive substances that could impact employee or public safety or the environment. This document is prepared using the specific guidelines identified in A Guide for Preparing Hanford Site Facility Effluent Monitoring Plans, WHC-EP-0438-01. This facility effluent monitoring plan assesses effluent monitoring systems and evaluates whether these systems are adequate to ensure the public health and safety as specified in applicable federal, state, and local requirements. This facility effluent monitoring plan will ensure long-range integrity of the effluent monitoring systems by requiring an update whenever a new process or operation introduces new hazardous materials or significant radioactive materials. This document must be reviewed annually even if there are no operational changes, and it must be updated, at a minimum, every 3 years

  8. A new concept for product refining in the Purex process

    International Nuclear Information System (INIS)

    Henrich, E.; Bauder, U.; Marquardt, R.

    1986-01-01

    In actual Purex plants the products are refined in additional solvent extraction cycles. Crystallization of uranyl and plutonyl nitrate from aqueous nitric acid solution is proposed as a potentially simpler product refining concept. Suitable crystallization conditions are being investigated in the laboratory using simulated and actual process solutions. A thorough removal of mother liquor is an essential purification step and well washed crystals usually contain less than 1% of an individual impurity. Crystallization simultaneously comprises a product concentration step. Hexavalent uranium can be separated from lower-valent plutonium. An outline of an integrated processing concept is given. Product refining by crystallization is compact; recycling of mother liquor plus wash acid prevents product loss and the generation of additional waste streams. (orig.) [de

  9. A process to remove ammonia from PUREX plant effluents

    International Nuclear Information System (INIS)

    Moore, J.D.

    1990-01-01

    Zirconium-clad nuclear fuel from the Hanford N-Reactor is reprocessed in the PUREX (Plutonium Uranium Extraction) Plant operated by the Westinghouse Hanford Comapny. Before dissolution, cladding is chemically removed from the fuel elements with a solution of ammonium fluoride-ammonium nitrate (AFAN). a solution batch with an ammonia equivalent of about 1,100 kg is added to each fuel batch of 10 metric tons. This paper reports on this decladding process, know as the 'Zirflex' process which produces waste streams containing ammonia and ammonium slats. Waste stream treatment, includes ammonia scrubbing, scrub solution evaporation, residual solids dissolution, and chemical neutralization. These processes produce secondary liquid and gaseous waste streams containing varying concentrations of ammonia and low-level concentrations of radionuclides. Until legislative restrictions were imposed in 1987, these secondary streams were released to the soil in a liquid disposal 'crib' and to the atmosphere

  10. Testing and economical evaluation of U(IV) in Purex

    International Nuclear Information System (INIS)

    Hoisington, J.E.; Hsu, T.C.

    1983-01-01

    The use of uranous nitrate, U(IV), as a plutonium reductant in the Purex solvent extraction process could significantly reduce the waste generation at the Savannah River Plant. The current reductant is a ferrous sulfamate (FS)/hydroxylamine nitrate (HAN) mixture. The iron and sulfate in the FS are major contributors to waste generation. The U(IV) reductant oxidizes to U(VI) producing no waste. The Savannah River Laboratory has developed an efficient electrochemical cell for U(IV) production and has demonstrated the effectiveness of U(IV) as a plutonium reductant. Plant tests and economic analyses are currently being conducted to determine the cost effectiveness of U(IV) implementation. The results of recent studies are presented

  11. Purex Process Improvements for Pu and NP Control in Total Actinide Recycle Flowsheets

    International Nuclear Information System (INIS)

    Birkett, J.E.; Carrott, M.J.; Crooks, G.; Fox, O.D.; Maher, C.J.; Taylor, R.J.; Woodhead, D.A.

    2006-01-01

    Significant improvements are required in the Purex process to optimise it for Advanced Fuel Cycles. Two key challenges we have identified are, firstly, developing more efficient methods for U/Pu separations especially at elevated Pu concentrations and, secondly, improving recovery, control and routing of Np in a modified Purex process. A series of Purex-like flowsheets for improved Pu separations based on hydroxamic acids and are reported. Purex-like flowsheets have been tested on a glovebox-housed 30-stage miniature centrifugal contactor train. A series of trials have been performed to demonstrate the processing of feeds with varying Pu contents ranging from 7 - 40% by weight. These flowsheets have demonstrated hydroxamic acids are excellent reagents for complexant stripping of Pu being able to achieve high decontamination factors (DF) on both the U and Pu product streams and co - recover Np with Pu. The advantages of a complexant-based approach are shown to be especially relevant when AFC scenarios are considered, where the Pu content of the fuel is expected to b e significantly higher. Recent results towards modifying the Purex process to improve recovery and control of Np in short residence time contactors are reported. Work on the development of chemical and process models to describe the complicated behaviour of Np under primary separation conditions (i.e. the HA extraction contactor) is described. To test the performance of the model a series of experiments were performed including testing of flowsheets on a fume-hood housed miniature centrifugal contactor train. The flowsheet was designed to emulate the conditions of a primar y separations contactor with the Np split between the U-solvent product and aqueous raffinate. In terms of Np routing the process model showed good agreement with flowsheet trial however much further work is required to fully understand this complex system. (authors)

  12. Production and remediation of low-sludge, simulated Purex waste glasses, 1: Effects of sludge oxide additions on melter operation

    International Nuclear Information System (INIS)

    Ramsey, W.G.

    1993-01-01

    Glass produced during the Purex 4 campaigns of the Integrated Defense Waste Processing Facility (DWPF) Melter System (IDMS) and the 774 Research Melter contained a lower fraction of sludge components than targeted by the Product Composition Control System (PCCS). Purex 4 glass was more durable than the benchmark (EA) glass, but less durable than most simulated SRS high-level waste glasses. Also, Purex 4 glass was considerably less durable than predicted by the algorithm which will be used to control production of DWPF glass. A melter run was performed using the 774 Research Melter to determine if the initial PCCS target composition determined for Purex 4 would produce acceptable glass whose durability could be accurately modeled by Hydration Thermodynamics. Reagent grade oxides and carbonates were added to Purex 4 melter feed stock to simulate a higher sludge loading. Each canister of glass produced was sampled and the composition, crystallinity, and durability was determined. This document details the melter operation and composition and crystallinity analyses

  13. Separation of radio cesium from PUREX feed solution by sorption on composite ammonium molybdo phosphate (AMP)

    International Nuclear Information System (INIS)

    Singh, I.J.; Achuthan, P.V.; Jain, S.; Janardanan, C.; Gopalakrishnan, V.; Wattal, P.K.; Ramanujam, A.

    2001-01-01

    Composite AMP exchanger was developed and evaluated for separation of radio cesium from dissolver solutions of PUREX process using a column experiment. The composite shows excellent sorption of radio cesium from dissolver solutions without any loss of plutonium and uranium. The removal of radio cesium from dissolver solutions will help in lowering the degradation of tri-n-butyl phosphate (TBP) in the solvent extraction process and will also help in reducing the radiation related problems. (author)

  14. Environmental report of Purex Plant and Uranium Oxide Plant - Hanford reservation

    International Nuclear Information System (INIS)

    1979-04-01

    A description of the site, program, and facilities is given. The data and calculations indicate that there will be no significant adverse environmental impact from the resumption of full-scale operations of the Purex and Uranium Oxide Plants. All significant pathways of radionuclides in Purex Plant effluents are evaluated. This includes submersion in the airborne effluent plumes, consumption of drinking water and foodstuffs irrigated with Columbia River water, ingestion of radioactive iodine through the cow-to-milk pathway, consumption of fish, and other less significant pathways. A summary of research and surveillance programs designed to assess the possible changes in the terresstrial and aquatic environments on or near the Hanford Reservation is presented. The nonradiological discharges to the environment of prinicpal interest are chemicals, sewage, and solid waste. These discharges will not lead to any significant adverse effects on the environment

  15. Structural acceptance criteria Remote Handling Building Tritium Extraction Facility

    Energy Technology Data Exchange (ETDEWEB)

    Mertz, G.

    1999-12-16

    This structural acceptance criteria contains the requirements for the structural analysis and design of the Remote Handling Building (RHB) in the Tritium Extraction Facility (TEF). The purpose of this acceptance criteria is to identify the specific criteria and methods that will ensure a structurally robust building that will safely perform its intended function and comply with the applicable Department of Energy (DOE) structural requirements.

  16. Structural acceptance criteria Remote Handling Building Tritium Extraction Facility

    International Nuclear Information System (INIS)

    Mertz, G.

    1999-01-01

    This structural acceptance criteria contains the requirements for the structural analysis and design of the Remote Handling Building (RHB) in the Tritium Extraction Facility (TEF). The purpose of this acceptance criteria is to identify the specific criteria and methods that will ensure a structurally robust building that will safely perform its intended function and comply with the applicable Department of Energy (DOE) structural requirements

  17. Hazards classification determination for PUREX fuel transfer to K-Basins

    International Nuclear Information System (INIS)

    Dodd, E.N. III.

    1995-01-01

    The PUREX Plant presently contains 2.9 metric tons of an aluminum clad Single Pass Reactor (SPR) fuel which is stored under water in four open top buckets in the PUREX slug storage basin. The PUREX dissolver cells contain approximately 0.5 metric tons of zirconium clad N Reactor fuel which was inadvertently placed into the process cell during charging operations. The dissolver N reactor elements will be recovered from the process floors using new crane operated tools. When the fuel shipment(s) is scheduled, the cask cars will be positioned into the PUREX rail tunnel and the overhead door will be opened. All the SPR fuel will be loaded into two cask rail cars inside four casks. The N Reactor fuel will be loaded into a separate rail car inside two or three casks. The car loading is initiated by opening the rail car lid and removing the cask lids. Prior to loading the canisters of N Reactor fuel, the canisters will be refilled with water (as needed) and a lid will be installed. The baskets of SPR fuel or canisters of N Reactor fuel will then be loaded into the casks. The lids to the casks will then be reinstalled and the car lids closed. The rail cars will then be decontaminated as necessary. The cask cars will be shipped either in two shipments or a combined single shipment using the rail route between PUREX and the K Basins. At the basin, the cask car will be positioned in the loadout area. The cask car lid will be opened and a single cask moved into the loadout pit, which is a lowered section of the basin. The cask lid is removed while the cask is lower into the pit. The fuel is then removed from the cask and stored in the basin. The cask is then removed, the lid reinstalled during removal, and the cask replaced into the cask car. This document identifies the hazard classification of the Fuel Transfer from the PUREX facility to K-Basins

  18. Interface control document between PUREX/UO3 Plant Transition and Solid Waste Disposal Division

    International Nuclear Information System (INIS)

    Duncan, D.R.

    1994-01-01

    This interface control document (ICD) between PUREX/UO 3 Plant Transition (PPT) and Solid Waste Disposal Division (SWD) establishes at a top level the functional responsibilities of each division where interfaces exist between the two divisions. Since the PUREX Transition and Solid Waste Disposal divisions operate autonomously, it is important that each division has a clear understanding of the other division's expectations regarding these interfaces. This ICD primarily deals with solid wastes generated by the PPT. In addition to delineating functional responsibilities, the ICD includes a baseline description of those wastes that will require management as part of the interface between the divisions. The baseline description of wastes includes waste volumes and timing for use in planning the proper waste management capabilities: the primary purpose of this ICD is to ensure defensibility of expected waste stream volumes and Characteristics for future waste management facilities. Waste descriptions must be as complete as-possible to ensure adequate treatment, storage, and disposal capability will exist. The ICD also facilitates integration of existing or planned waste management capabilities of the PUREX. Transition and Solid Waste Disposal divisions. The ICD does not impact or affect the existing processes or procedures for shipping, packaging, or approval for shipping wastes by generators to the Solid Waste Division

  19. Literature Review: Crud Formation at the Liquid/Liquid Interface of TBP-Based Solvent-Extraction Processes

    Energy Technology Data Exchange (ETDEWEB)

    Delegard, Calvin H.; Casella, Amanda J.

    2016-09-30

    This report summarizes the literature reviewed on crud formation at the liquid:liquid interface of solvent extraction processes. The review is focused both on classic PUREX extraction for industrial reprocessing, especially as practiced at the Hanford Site, and for those steps specific to plutonium purification that were used at the Plutonium Reclamation Facility (PRF) within the Plutonium Finishing Plant (PFP) at the Hanford Site.

  20. Facility Operations 1993 fiscal year work plan: WBS 1.3.1

    Energy Technology Data Exchange (ETDEWEB)

    1992-11-01

    The Facility Operations program is responsible for the safe, secure, and environmentally sound management of several former defense nuclear production facilities, and for the nuclear materials in those facilities. As the mission for Facility Operations plants has shifted from production to support of environmental restoration, each plant is making a transition to support the new mission. The facilities include: K Basins (N Reactor fuel storage); N Reactor; Plutonium-Uranium Reduction Extraction (PUREX) Plant; Uranium Oxide (UO{sub 3}) Plant; 300 Area Fuels Supply (N Reactor fuel supply); Plutonium Finishing Plant (PFP).

  1. Facility Operations 1993 fiscal year work plan: WBS 1.3.1

    International Nuclear Information System (INIS)

    1992-11-01

    The Facility Operations program is responsible for the safe, secure, and environmentally sound management of several former defense nuclear production facilities, and for the nuclear materials in those facilities. As the mission for Facility Operations plants has shifted from production to support of environmental restoration, each plant is making a transition to support the new mission. The facilities include: K Basins (N Reactor fuel storage); N Reactor; Plutonium-Uranium Reduction Extraction (PUREX) Plant; Uranium Oxide (UO 3 ) Plant; 300 Area Fuels Supply (N Reactor fuel supply); Plutonium Finishing Plant (PFP)

  2. Purex process operation and performance, 1970 Thoria Campaign

    International Nuclear Information System (INIS)

    Jackson, R.R.; Walser, R.L.

    1977-03-01

    The Hanford Purex Plant fulfilled a 1970 commitment to the Atomic Energy Commission to produce 360 kilograms of high purity 233 U as uranyl nitrate solution. Overall plant performance during both 1970 and 1966 confirmed the suitability of Purex for processing thorium on a campaign basis. The 1970 processing campaign, including flushing operations, is discussed with particular emphasis on problem areas. Background information on the process and equipment used is also presented. The organizations and their designations described are those existing in 1970

  3. Fission products control by gamma spectrometry in purex process solutions

    International Nuclear Information System (INIS)

    Goncalves, Maria Augusta

    1982-01-01

    This paper deals with a radiometric method for fission products analysisby gamma spectrometry. This method will be applied for fission productscontrol at the irradiated material processing facility, under construction inthe Instituto de Pesquisas Energeticas e Nucleares, SP, Brazil. Countinggeometry was defined taking into account the activities of process solutionsto be analysed, the remotely operated aliquotation device of analytical celland the available detection system. Natural and 19,91% enriched uraniumsamples were irradiated at IEAR-1 reactor in order to simulate thecomposition of Purex process solutions. After a short decay time, the sampleswere dissolved with HNO 3 and then, conditioned in standard flasks withdefined geometry. The spectra were obtained by a Ge(Li) semiconductordetector and analysed by the GELIGAM software system, losing a floppy-diskconnected to a PDP-11/05 computer. Libraries were prepared and calibrationswere made with standard sources to fit the programs to the analysis offission products in irradiated uranium solutions. It was possible to choosethe best program to be used in routine analysis with the obtained data.(author)

  4. Fisson product control by gamma spectrometry in Purex process solutions

    International Nuclear Information System (INIS)

    Goncalves, Maria A.; Matsuda, H.T.

    1982-01-01

    A radiometric method for fission product analysis by gamma spectrometry, to be applied for fission product control at an irradiated material processing facility, is described. Counting geometry was defined taking into account the activities of process solutions to be analysed, the remotely operated aliquotation device of the analytical cell and the available detection system. Natural and 19,91% enriched uranium samples were irradiated in order to simulate the composition of Purex process solutions. After a short decay time the samples were dissolved with HNO 3 and then conditioned in standard flasks with defined geometry. The spectra were obtained by a Ge(Li) semiconductor detector and analysed by the GELIGAM software system, using a floppy-disk connected to a PDP-11/05 computer. Libraries were prepared and calibrations were made with standard sources to fit the analysis of fission products in irradiated uranium solutions. It was possible to choose the best program to be used in routine analysis with the obtained data. (Author) [pt

  5. High-Activity ICP-AES Measurements in the ATALANTE Facility Applied to Analytical Monitoring of an Extraction Test

    Energy Technology Data Exchange (ETDEWEB)

    Esbelin, E.; Boyer-Deslys, V.; Beres, A.; Viallesoubranne, C. [CEA Marcoule, DEN/DRCP/SE2A/LAMM, BP17171, 30207 Bagnols-sur-Ceze (France)

    2008-07-01

    The Material Analysis and Metrology Laboratory (LAMM) of the Cea's Atalante complex ensures analytical monitoring of enhanced separation tests. Certain fission products, actinides and lanthanides were assayed by ICP-AES (Inductively Coupled Plasma-Atomic Emission Spectroscopy) in the CBA shielded analysis line. These analyses were particularly effective for controlling the Diamex test, and contributed to its success. The Diamex process consists in extracting the actinides and lanthanides from a Purex raffinate using a diamide, DMDOHEMA, followed by stripping at low acidity. The major elements analyzed during the test were Am, Nd, Mo, Fe, and Zr.

  6. High-Activity ICP-AES Measurements in the ATALANTE Facility Applied to Analytical Monitoring of an Extraction Test

    International Nuclear Information System (INIS)

    Esbelin, E.; Boyer-Deslys, V.; Beres, A.; Viallesoubranne, C.

    2008-01-01

    The Material Analysis and Metrology Laboratory (LAMM) of the Cea's Atalante complex ensures analytical monitoring of enhanced separation tests. Certain fission products, actinides and lanthanides were assayed by ICP-AES (Inductively Coupled Plasma-Atomic Emission Spectroscopy) in the CBA shielded analysis line. These analyses were particularly effective for controlling the Diamex test, and contributed to its success. The Diamex process consists in extracting the actinides and lanthanides from a Purex raffinate using a diamide, DMDOHEMA, followed by stripping at low acidity. The major elements analyzed during the test were Am, Nd, Mo, Fe, and Zr

  7. Fast Extraction Kicker for the Accelerator Test Facility

    International Nuclear Information System (INIS)

    De Santis, Stefano; Urakawa, Junji; Naito, Takashi

    2007-01-01

    We present the results of a study for the design of a fast extraction kicker to be installed in the Accelerator Test Facility ring at KEK. This activity is carried on in the framework of the ATF2 project, which will be built on the KEK Tsukuba campus as an extension of the existing ATF, taking advantage of the worlds smallest normalized emittance achieved there. ATF2's primary goal is to operate as a test facility and establish the hardware and beam handling technologies envisaged for the International Linear Collider. In particular, the fast extraction kicker object of the present paper is an important component of the ILC damping rings, since its rise and fall time define the minimum distance between bunches and ultimately the damping rings length itself. Building on the initial results presented at EPAC '06, we report on the present status of the kicker design and define the minimum characteristics for pulsers and other subsystems. In addition to the original scheme with multiple stripline modules producing a total deflection of 5 mrad, we also investigated a scheme with a single kicker module for a reduced deflection of 1 mrad placed inside a closed orbit bump, which takes the electron closer to the extraction septum

  8. Colorimetric determination of reducing normality in the Purex process

    International Nuclear Information System (INIS)

    Baumann, E.W.

    1983-07-01

    Adjustment of the valence state of plutonium from extractable Pu(IV) to nonextractable Pu(III) in the Purex process is accomplished by addition of reductants such as Fe(II), hydroxylamine nitrate (HAN), or U(IV). To implement on-line monitoring of this reduction step for improved process control at the Savannah River Plant, a simple colorimetric method for determining excess reductant (reducing normality) was developed. The method is based on formation of a colored complex of Fe(II) with FerroZine (Hach Chemical Company). The concentration of Fe(II) is determined directly. The concentration of HAN or U(IV), in addition to Fe(II), is determined indirectly as Fe(II), produced through reduction of Fe(III). Experimental conditions for a HAN-Fe(III) reaction of known stoichiometry were established. The effect of hydrazine, which stabilizes U(IV), was also determined. Real-time measurements of color development were made that simulated on-line performance. A laboratory analytical procedure is included. 5 references, 8 figures

  9. Calculation code PULCO for Purex process in pulsed column

    International Nuclear Information System (INIS)

    Gonda, Kozo; Matsuda, Teruo

    1982-03-01

    The calculation code PULCO, which can simulate the Purex process using a pulsed column as an extractor, has been developed. The PULCO is based on the fundamental concept of mass transfer that the mass transfer within a pulsed column occurs through the interface of liquid drops and continuous phase fluid, and is the calculation code different from conventional ones, by which various phenomena such as the generation of liquid drops, their rising and falling, and the unification of liquid drops actually occurring in a pulsed column are exactly reflected and can be correctly simulated. In the PULCO, the actually measured values of the fundamental quantities representing the extraction behavior of liquid drops in a pulsed column are incorporated, such as the mass transfer coefficient of each component, the diameter and velocity of liquid drops in a pulsed column, the holdup of dispersed phase, and axial turbulent flow diffusion coefficient. The verification of the results calculated with the PULCO was carried out by installing a pulsed column of 50 mm inside diameter and 2 m length with 40 plate stage in a glove box for unirradiated uranium-plutonium mixed system. The results of the calculation and test were in good agreement, and the validity of the PULCO was confirmed. (Kako, I.)

  10. Studies in support of an SNM cutoff agreement: The PUREX exercise

    International Nuclear Information System (INIS)

    Stanbro, W.D.; Libby, R.; Segal, J.

    1995-01-01

    On September 23, 1993, President Clinton, in a speech before the United Nations General Assembly, called for an international agreement banning the production of plutonium and highly enriched uranium for nuclear explosive purposes. A major element of any verification regime for such an agreement would probably involve inspections of reprocessing plants in Nuclear Nonproliferation Treaty weapons states. Many of these are large facilities built in the 1950s with no thought that they would be subject to international inspection. To learn about some of the problems that might be involved in the inspection of such large, old facilities, the Department of Energy, Office of Arms Control and Nonproliferation, sponsored a mock inspection exercise at the PUREX plant on the Hanford Site. This exercise examined a series of alternatives for inspections of the PUREX as a model for this type of facility at other locations. A series of conclusions were developed that can be used to guide the development of verification regimes for a cutoff agreement at reprocessing facilities

  11. Deactivating a major nuclear fuels reprocessing facility

    International Nuclear Information System (INIS)

    LeBaron, G.J.

    1997-01-01

    This paper describes three key processes used in deactivating the Plutonium Uranium Extraction (PUREX) Facility, a large, complex nuclear reprocessing facility, 15 months ahead of schedule and $77 million under budget. The organization was reengineered to refine its business processes and more effectively organize around the deactivation work scope. Multi-disciplined work teams were formed to be self-sufficient and empowered to make decisions and perform work. A number of benefits were realized by reengineering. A comprehensive process to develop end points which clearly identified specific results and the post-project facility configuration was developed so all areas of a facility were addressed. Clear and specific end points allowed teams to focus on completing deactivation activities and helped ensure there were no unfulfilled end-of-project expectations. The RCRA regulations require closure of permitted facilities within 180 days after cessation of operations which may essentially necessitate decommissioning. A more cost effective approach was adopted which significantly reduced risk to human health and the environment by taking the facility to a passive, safe, inexpensive-to-maintain surveillance and maintenance condition (deactivation) prior to disposition. PUREX thus became the first large reprocessing facility with active TSD [treatment, storage, and disposal] units to be deactivated under the RCRA regulations

  12. Idaho Chemical Processing Plant and Plutonium-Uranium Extraction Plant phaseout/deactivation study

    International Nuclear Information System (INIS)

    Patterson, M.W.; Thompson, R.J.

    1994-01-01

    The decision to cease all US Department of Energy (DOE) reprocessing of nuclear fuels was made on April 28, 1992. This study provides insight into and a comparison of the management, technical, compliance, and safety strategies for deactivating the Idaho Chemical Processing Plant (ICPP) at Westinghouse Idaho Nuclear Company (WINCO) and the Westinghouse Hanford Company (WHC) Plutonium-Uranium Extraction (PUREX) Plant. The purpose of this study is to ensure that lessons-learned and future plans are coordinated between the two facilities

  13. Advance purex process for the new reprocessing plants in France and in Japan

    International Nuclear Information System (INIS)

    Viala, M.

    1991-01-01

    In the early Eighties, Japanese utilities formed the Japan Nuclear Fuel Service Co (JNFS), which is in charge of the construction and the operation of the first commercial reprocessing plant in Japan to be erected in Rokkasho Village, Aomori Prefecture. Following a thorough worldwide examination of available processes and technologies, JNFS selected the French technology developed for UP3 and UP2 800 for the plants' main facilities. For these three new plants, the 40-year old PUREX process which is used worldwide for spent fuel reprocessing, has been significantly improved. This paper describes some of the innovative features of the selected processes

  14. Removal of fission product ruthenium from purex process solutions: thiourea as complexing agent

    International Nuclear Information System (INIS)

    Floh, B.; Abrao, A.

    1980-01-01

    A new method for the treatment of spent uranium fuel is presented. It is based on the Purex Process using thiourea to increase the ruthenium decontamination factor. Thiourea exhibits a strong tendency for the formation of coordination compounds in acidic media. This tendency serves as a basis to transform nitrosyl-ruthenium species into Ru /SC(NH)(NH 2 )/ 2+ and Ru /SC(NH)(NH 2 )/ 3 complexes which are unextractable by TBP-varsol. The best conditions for the ruthenium-thiourea complex formation were found to be: thiourea-ruthenium ratio (mass/mass) close to 42, at 75 0 C, 30 minutes reaction time and aging period of 60 minutes. The ruthenium decontamination factor for a single uranium extraction are ca. 80-100, not interfering with extraction of actinides. These values are rather high in comparison to those obtained using the conventional Purex Process (e.g. F.D. sub(Ru)=10). By this reason the method developed here is suitable for the treatment of spent uranium fuels. Thiourea (100g/l) scrubbing experiments of ruthenium, partially co-extracted with actinides, confirmed the possibility of its removal from the extract. A decontamination greater than 83,5% for ruthenium as fission product is obtained in two stages with this procedure. (Author) [pt

  15. Standardization of a method to study the distribution of Americium in purex process

    International Nuclear Information System (INIS)

    Dapolikar, T.T.; Pant, D.K.; Kapur, H.N.; Kumar, Rajendra; Dubey, K.

    2017-01-01

    In the present work the distribution of Americium in PUREX process is investigated in various process streams. For this purpose a method has been standardized for the determination of Am in process samples. The method involves extraction of Am with associated actinides using 30% TRPO-NPH at 0.3M HNO 3 followed by selective stripping of Am from the organic phase into aqueous phase at 6M HNO 3 . The assay of aqueous phase for Am content is carried out by alpha radiometry. The investigation has revealed that 100% Am follows the HLLW route. (author)

  16. 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)

  17. PUREX Plant aggregate area management study technical baseline report

    International Nuclear Information System (INIS)

    DeFord, D.H.; Carpenter, R.W.

    1995-05-01

    The PUREX aggregate area is made up of six operable units; 200-PO-1 through 200-PO-6 and consists of liquid and solid waste disposal sites in the vicinity of, and related to, PUREX Plant operations. This report describes PUREX and its waste sites, including cribs, french drains, septic tanks and drain fields, trenches and ditches, ponds, catch tanks, settling tanks, diversion boxes, underground tank farms, and the lines and encasements that connect them. Each waste site in the aggregate area is described separately. Close relationships between waste units, such as overflow from one to another, are also discussed. This document provides a technical baseline of the aggregate area and results from an environmental investigation. This document is based upon review and evaluation of numerous Hanford Site current and historical reports, drawings and photographs, supplemented with site inspections and employee interviews. No intrusive field investigations or sampling were conducted

  18. PUREX SAMCONS uninterruptible power supply (UPS) acceptance test report

    International Nuclear Information System (INIS)

    Blackaby, W.B.

    1997-01-01

    This Acceptance Test Report for the PUREX Surveillance and Monitoring and Control System (SAMCONS) Uninterruptible Power Supply (UPS) Acceptance Test Procedure validates the operation of the UPS, all alarming and display functions and the ability of the UPS to supply power to the SAMCONS as designed. The proper installation of the PUREX SAMCONS Trailer UPS components and wiring will be systematically evaluated by performance of this procedure. Proper operation of the SAMCONS computer UPS will be verified by performance of a timed functional load test, and verification of associated alarms and trouble indications. This test procedure will be performed in the SAMCONS Trailer and will include verification of receipt of alarms at the SAMCONS computer stations. This test may be performed at any time after the completion of HNF-SD-CP-ATP-083, PUREX Surveillance and Monitoring and Control System (SAMCONS) Acceptance Test Procedure, when computer display and alarm functions have been proven to operate correctly

  19. Commercial Light Water Reactor Tritium Extraction Facility Geotechnical Summary Report

    International Nuclear Information System (INIS)

    Lewis, M.R.

    2000-01-01

    A geotechnical investigation program has been completed for the Circulating Light Water Reactor - Tritium Extraction Facility (CLWR-TEF) at the Savannah River Site (SRS). The program consisted of reviewing previous geotechnical and geologic data and reports, performing subsurface field exploration, field and laboratory testing and geologic and engineering analyses. The purpose of this investigation was to characterize the subsurface conditions for the CLWR-TEF in terms of subsurface stratigraphy and engineering properties for design and to perform selected engineering analyses. The objectives of the evaluation were to establish site-specific geologic conditions, obtain representative engineering properties of the subsurface and potential fill materials, evaluate the lateral and vertical extent of any soft zones encountered, and perform engineering analyses for slope stability, bearing capacity and settlement, and liquefaction potential. In addition, provide general recommendations for construction and earthwork

  20. Effect of Entrainment and Overflow Occurrences on Concentration Profile in PUREX Flow Sheet

    International Nuclear Information System (INIS)

    Ueda, Yoshinori; Ishii, Junichi; Matsumoto, Shiro

    2003-01-01

    A deviation in the operational condition of a mixer settler and a centrifugal contactor causes an entrainment or an overflow, which affects the concentration profile. Although there has been no quantitative study about the effect of such abnormal flows on the concentration profile, the occurrence of such abnormal flows has been severely restricted for a PUREX flow sheet. However, the restriction of abnormal flows can be relaxed when the effect of such flows is limited within the allowable range such that the concentration of the product does not deviate from its specification. This relaxation could serve to benefit a continuous operation under a certain degree of deviation from the operational condition and a smaller design load of a solvent extractor. From this viewpoint, the relationship between the magnitude of abnormal flows and the effect of them on the process was studied quantitatively using a specially developed code in a wide range of PUREX flow sheet conditions, and the possibility of this relaxation was investigated. The results showed that the effect of the abnormal flow on the concentration in the organic outflow or aqueous raffinate was dominated by the leakage fraction under normal conditions regardless of each specific flow sheet condition. The common correlations were found between the leakage fraction of uranium and plutonium under the occurrence of abnormal flows and that under no abnormal flow for the stripping and extracting conditions, respectively. Comparing the given correlations and the usual specification of the leakage fraction of uranium and plutonium suggested that the restriction of the abnormal flows could be relaxed for a usual PUREX flow sheet

  1. Chemical processing of HTR fuels applying either THOREX or PUREX flow sheets

    Energy Technology Data Exchange (ETDEWEB)

    Zimmer, E; Merz, E [Kernforschungsanlage, Juelich GmbH, Institut fuer Chemische Technologie der Nuklearen Entsorgung, Juelich (Germany)

    1985-07-01

    Two fuel cycles are considered for utilization in high temperature gas-cooled reactors (HTRs): the high-enriched thorium-uranium (HEU 93% U-235) and the low-enriched uranium (LEU 8-12% U-235) fuel concept. For both fuel compositions suitable reprocessing procedures are required which are capable to separate the actinides thorium, uranium and plutonium from fission products and from each other. In any case, the processes under consideration utilize Tri-n-butylphosphate (TBP) together with a straight-chain paraffinic diluent (C{sub 8}-C{sub 14}, to day usually dodecane) as extractant in an aqueous nitrate system; most commonly, the related processes are known by the acronyms PUREX and THOREX. The PUREX process has become the reprocessing procedure quite generally used for all fuel types containing natural, slightly or highly enriched uranium together with lower or higher contents of plutonium. The THOREX process on the other hand has been developed to separate thorium, uranium and fission products from thorium based irradiated fuel. Generally, the utilization of the thorium fuel cycle is most attractive for High Temperature Reactors. On the other hand, the strong recommendation of INFCE to abandon the use of high-enriched uranium for nuclear energy applications virtually rules out the thorium fuel cycle, since economic utilization of thorium as a fertile material requires the use of high-enriched U-235. Thus, it was decided in the Federal Republic of Germany to switch over, at least for the foreseeable future, to the low enrichment uranium-plutonium fuel cycle, well aware of its economic shortcomings. In this paper various THOREX flowsheets as well as a PUREX variant suitable for LEU fuel reprocessing are described. Both processes have in common that the main stream is always presented by the fertile material, that means thorium and U-238, respectively.

  2. 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

  3. Waste Feed Delivery Purex Process Connector Design Pressure

    International Nuclear Information System (INIS)

    BRACKENBURY, P.J.

    2000-01-01

    The pressure retaining capability of the PUREX process connector is documented. A context is provided for the connector's current use within existing Projects. Previous testing and structural analyses campaigns are outlined. The deficient condition of the current inventory of connectors and assembly wrenches is highlighted. A brief history of the connector is provided. A bibliography of pertinent references is included

  4. Cement waste form qualification report: WVDP [West Valley Demonstration Project] PUREX decontaminated supernatant

    International Nuclear Information System (INIS)

    McVay, C.W.; Stimmel, J.R.; Marchetti, S.

    1988-08-01

    This report provides a summary of work performed to develop a cement-based, low-level waste formulation suitable for the solidification of decontaminated high-level waste liquid produced as a by-product of PUREX spent fuel reprocessing. The resultant waste form is suitable for interim storage and is intended for ultimate disposal as low-level Class C waste; it also meets the stability requirements of the NRC Branch Technical Position on Waste Form Qualification, May 1983 and the requirements of 10 CFR 61. A recipe was developed utilizing only Portland Type I cement based on an inorganic salts simulant of the PUREX supernatant. The qualified recipe was tested full scale in the production facility and was observed to produce a product with entrained air, low density, and lower-than-expected compressive strength. Further laboratory scale testing with actual decontaminated supernatant revealed that set retarders were present in the supernatant, precluding setting of the product and allowing the production of ''bleed water.'' Calcium nitrate and sodium silicate were added to overcome the set retarding effect and produced a final product with improved performance when compared to the original formulation. This report describes the qualification process and qualification test results for the final product formulation. 7 refs., 38 figs., 21 tabs

  5. On the identification of complexing radiolysis products in the Purex system. (20%TBP - Dodecane - HNO3)

    International Nuclear Information System (INIS)

    Becker, R.; Baumgartner, F.; Steiglitz, L.

    1978-09-01

    The lifetime of the extraction system TBP Dodecane-aqueous HNO. In the Purex process is limited by radiolytic and hydrolytic decomposition of the extracting and diluting agent which is indicated by an increased retention of fission products, especially zirconium. In this work, the radiolytically formed complexing agents responsible for this are enriched (molecular distillation) and separated in several fractions by liquid chromatography. The chemical composition of these fractions was identified by a combination of gas chromatography and mass spectrometry, supplemented by infra-red spectroscopy. As for doubtful complexing agents, they are mainly long-chain phosphoric acid esters, and, to a lesser extent, the existence of polycarbonyl compounds is suspected. The high molecular weight components of the phosphate ester fraction could be separated by gas chromatography and identified as oligomeric phosphate esters. (author)

  6. The isolation of lutetium from gadolinium contained in Purex process solutions

    International Nuclear Information System (INIS)

    Bostick, D.T.; Vick, D.O.; May, M.P.; Walker, R.L.

    1992-09-01

    A chemical separation procedure has been devised to isolate Lu from Purex dissolver solutions containing the neutron poison, Gd. The isolation procedure involves the removal of U and >Pu from a dissolver solution using tributylphosphate solvent extraction. If required, solvent extraction using di-(2-ethylhexyl) phosphoric acid can be employed to further purify the sample be removing alkali and alkali earth elements. Finally, Lu is chromatographically separated from Gd and rare earth fission products on a Dowex 50W-X8 resin column using an alpha-hydroxyisobutyrate eluant. The success of the chemical separation procedure has been demonstrated in the quantitative recovery of as little as 1.4 ng Lu from solutions containing a 5000-fold excess of Gd. Additionally, Lu has been isolated from synthetic dissolver samples containing U, Ba, Cs, and Gd. Thermal emission MS data indicated that the Lu fraction of the synthetic sample was free of Gd interference

  7. Purex process operation and performance: 1970 thoria campaign

    International Nuclear Information System (INIS)

    Walser, R.L.

    1978-02-01

    The Hanford Purex Plant has demonstrated suitability for reprocessing irradiated thoria (ThO 2 ) target elements on a campaign basis. A 1965 process test and major production campaigns conducted in 1966 and 1970 recovered nitrate solution form products totaling approximately 565 tons of thorium and 820 kilograms of 233 U. The overall recoveries for the 1970 campaign based on reactor input data were 94.9 percent for thorium and 95.2 percent for uranium. The primary function of the Hanford Purex Plant is reprocessing of irradiated uranium fuel elements to separate and purify uranium, plutonium and neptunium. Converting the plant to thoria reprocessing required major process development work and equipment modifications. The operation and performance of the Plant during the 1970 thoria reprocessing campaign is discussed in this report. The discussion includes background information on the process and equipment, problems encountered, and changes recommended for future campaigns

  8. Hanford Site Near-Facility Environmental Monitoring Data Report for Calendar Year 2007- Appendix 2

    Energy Technology Data Exchange (ETDEWEB)

    Perkins, Craig J.; Dorsey, Michael; Mckinney, Stephen M.; Wilde, Justin W.; Duncan, Joanne P.

    2008-10-13

    Near-facility environmental monitoring is defined as monitoring near facilities that have the potential to discharge or have discharged, stored, or disposed of radioactive or hazardous materials. Monitoring locations are associated with nuclear facilities such as the Plutonium Finishing Plant (PFP), Canister Storage Building (CSB), and the K Basins; inactive nuclear facilities such as N Reactor and the Plutonium-Uranium Extraction (PUREX) Facility; and waste storage or disposal facilities such as burial grounds, cribs, ditches, ponds, tank farms, and trenches. Much of the monitoring consists of collecting and analyzing environmental samples and methodically surveying areas near facilities. The program is also designed to evaluate acquired analytical data, determine the effectiveness of facility effluent monitoring and controls, assess the adequacy of containment at waste disposal units, and detect and monitor unusual conditions.

  9. Hanford Site Near-Facility Environmental Monitoring Data Report for Calendar Year 2008

    Energy Technology Data Exchange (ETDEWEB)

    Perkins, Craig J.; Dorsey, Michael C.; Mckinney, Stephen M.; Wilde, Justin W.; Poston, Ted M.

    2009-09-15

    Near-facility environmental monitoring is defined as monitoring near facilities that have the potential to discharge or have discharged, stored, or disposed of radioactive or hazardous materials. Monitoring locations are associated with nuclear facilities such as the Plutonium Finishing Plant, Canister Storage Building, and the K Basins; inactive nuclear facilities such as N Reactor and the Plutonium-Uranium Extraction (PUREX) Facility; and waste storage or disposal facilities such as burial grounds, cribs, ditches, ponds, tank farms, and trenches. Much of the monitoring consists of collecting and analyzing environmental samples and methodically surveying areas near facilities. The program is also designed to evaluate acquired analytical data, determine the effectiveness of facility effluent monitoring and controls, assess the adequacy of containment at waste disposal units, and detect and monitor unusual conditions.

  10. Di-hydroxyurea-a Promising Reducing Reagent for the U/Pu split in the PUREX process

    Energy Technology Data Exchange (ETDEWEB)

    Taihong, Yan; Weifang, Zheng; Guoan, Ye; Yu, Zhang; Liang, Xian; Ying, Di; Xiaoyan, Bian [Department of Radiochemistry, China Institute of Atomic Energy - CIAE, Beijing 102413 (China)

    2009-06-15

    In the reprocessing of spent nuclear fuel by the Purex process, the separation of U and Pu is a major stage. This is commonly achieved by a redox process, in which a reducing agent (e.g. U(IV) or (FeII)) and a stabiliser (e.g. N{sub 2}H{sub 4} or NH{sub 2}SO{sub 3}H) are added to reduce extractable Pu{sup 4+} to un-extractable Pu{sup 3+}. The stabiliser prevents the nitrous acid catalysed re-oxidation of Pu(III) back to Pu(IV). One of the key objectives is to reduce both the number of solvent extraction cycles and the waste stream volumes [1]. One option for Advanced Purex flowsheets is to adopt a new salt-free reductant in the U/Pu split. Di-hydroxyurea(DHU)-a new Reducing reagent was synthesized with tri-associated solid phosgene (Bis(trichloromethyl)Carbonate) solved in dioxane and hydroxylamine hydrochloride solved in potassium acetate solution. The Reduction of Pu(IV) by DHU was investigated using UV-Vis spectrophotometer. The reduction back-extraction behavior of Pu(IV) in 30%TBP /OK was firstly investigated under conditions of different temperature, different concentration of DHU and HNO{sub 3} and various phase contract time respectively.The results showed that Pu(IV) in organic phase can be stripped rapidly to aqueous phase by DHU. Simulating the 1B contactor of the Purex process by DHU with nitric acid solution as the stripping agent,the separation factors of uranium/plutonium can reach 2.1 10{sup 4}. This indicates that DHU is a promising salt free agent for uranium/plutonium separation. (authors)

  11. PUREX (SAMCONS) uninterruptible power supply (UPS) acceptance test procedure

    International Nuclear Information System (INIS)

    Blackaby, W.B.

    1997-01-01

    This Acceptance Test Procedure for the PUREX Surveillance and Monitoring and Control System (SAMCONS) Uninterruptible Power Supply (UPS) provides for testing and verifying the proper operation of the control panel alarms and trouble functions, the 6roper functioning of the AC inverter, ability of the battery supply to maintain the SAMCONS load for a minimum of two hours , and proper interaction with the SAMCONS Video graphic displays for alarm displays

  12. Solvent distillation studies for a purex reprocessing plant

    International Nuclear Information System (INIS)

    Ginisty, C.; Guillaume, B.

    1990-01-01

    A distillation system has been developed for regeneration of Purex solvent and will be implemented for the first time in a reprocessing plant. The results are described and analyzed, with emphasis on laboratory experiments which were made with a radioactive plant solvent. Particularly the distillation provides a good separation of solvent degradation products, which was verified by measurements of interfacial tension and plutonium or ruthenium retention. 16 refs., 3 figs., 5 tabs

  13. Separation of neptunium from uranium and plutonium in the Purex process

    International Nuclear Information System (INIS)

    Kolarik, Z.; Schuler, R.

    1984-01-01

    The possibility of removing neptunium from the Purex process in the first extraction cycle was investigated. Butyraldehyde was found to reduce Np(VI) to Np(V), but not Pu(IV) to Pu(III). Up to 99.7% Np can be separated from uranium and plutonium in the 1A extractor or, much more favourably, in an additional partitioning extractor. Hydroxylamine nitrate can be used for reducing Np(VI) to Np(V) in a uranium purification cycle at a high U concentration in the feed solution. Here the decontamination factor for Np can be as high as 2300 and is lowered if iron is present in the feed. (author)

  14. Combination RCRA groundwater monitoring plan for the 216-A-10, 216-A-36B, and 216-A-37-1 PUREX cribs

    International Nuclear Information System (INIS)

    Lindberg, J.W.

    1997-06-01

    This document presents a groundwater quality assessment monitoring plan, under Resource Conservation and Recovery Act of 1976 (RCRA) regulatory requirements for three RCRA sites in the Hanford Site's 200 East Area: 216-A-10, 216-A-36B, and 216-A-37-1 cribs (PUREX cribs). The objectives of this monitoring plan are to combine the three facilities into one groundwater quality assessment program and to assess the nature, extent, and rate of contaminant migration from these facilities. A groundwater quality assessment plan is proposed because at least one downgradient well in the existing monitoring well networks has concentrations of groundwater constituents indicating that the facilities have contributed to groundwater contamination. The proposed combined groundwater monitoring well network includes 11 existing near-field wells to monitor contamination in the aquifer in the immediate vicinity of the PUREX cribs. Because groundwater contamination from these cribs is known to have migrated as far away as the 300 Area (more than 25 km from the PUREX cribs), the plan proposes to use results of groundwater analyses from 57 additional wells monitored to meet environmental monitoring requirements of US Department of Energy Order 5400.1 to supplement the near-field data. Assessments of data collected from these wells will help with a future decision of whether additional wells are needed

  15. 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

  16. 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.

  17. DIST: a computer code system for calculation of distribution ratios of solutes in the purex system

    Energy Technology Data Exchange (ETDEWEB)

    Tachimori, Shoichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1996-05-01

    Purex is a solvent extraction process for reprocessing the spent nuclear fuel using tri n-butylphosphate (TBP). A computer code system DIST has been developed to calculate distribution ratios for the major solutes in the Purex process. The DIST system is composed of database storing experimental distribution data of U(IV), U(VI), Pu(III), Pu(IV), Pu(VI), Np(IV), Np(VI), HNO{sub 3} and HNO{sub 2}: DISTEX and of Zr(IV), Tc(VII): DISTEXFP and calculation programs to calculate distribution ratios of U(IV), U(VI), Pu(III), Pu(IV), Pu(VI), Np(IV), Np(VI), HNO{sub 3} and HNO{sub 2}(DIST1), and Zr(IV), Tc(VII)(DITS2). The DIST1 and DIST2 determine, by the best-fit procedures, the most appropriate values of many parameters put on empirical equations by using the DISTEX data which fulfill the assigned conditions and are applied to calculate distribution ratios of the respective solutes. Approximately 5,000 data were stored in the DISTEX and DISTEXFP. In the present report, the following items are described, 1) specific features of DIST1 and DIST2 codes and the examples of calculation 2) explanation of databases, DISTEX, DISTEXFP and a program DISTIN, which manages the data in the DISTEX and DISTEXFP by functions as input, search, correction and delete. and at the annex, 3) programs of DIST1, DIST2, and figure-drawing programs DIST1G and DIST2G 4) user manual for DISTIN. 5) source programs of DIST1 and DIST2. 6) the experimental data stored in the DISTEX and DISTEXFP. (author). 122 refs.

  18. Modelling of uranium/plutonium splitting in purex process

    International Nuclear Information System (INIS)

    Boullis, B.; Baron, P.

    1987-06-01

    A mathematical model simulating the highly complex uranium/plutonium splitting operation in PUREX process has been achieved by the french ''Commissariat a l'Energie Atomique''. The development of such a model, which includes transfer and redox reactions kinetics for all the species involved, required an important experimental work in the field of basis chemical data acquisition. The model has been successfully validated by comparison of its results with those of specific trials achieved (at laboratory scale), and with the available results of the french reprocessing units operation. It has then been used for the design of french new plants splitting operations

  19. Advanced Purex process for the new French reprocessing plants

    International Nuclear Information System (INIS)

    Viala, M.; Ledermann, P.; Pradel, P.

    1993-01-01

    The paper describes the main process innovations of the new Cogema reprocessing plants of La Hague (UP3 and UP2 800). Major improvements of process like the use of rotary dissolvers and annular columns, and also entirely new processes like solvent distillation and plutonium oxidizing dissolution, yield an advanced Purex process. The results of these innovations are significant improvements for throughput, end-products purification performances and waste minimization. They contribute also to limit personnel exposure. The main results of the first three years of operation are described. (author). 3 refs., 5 figs

  20. Facility effluent monitoring plan for the plutonium-uranium extraction facility

    International Nuclear Information System (INIS)

    Nickels, J.M.; Geiger, J.L.

    1992-11-01

    A facility effluent monitoring plan is required by the US Department of Energy in DOE Order 5400.1 for any operations that involve hazardous materials and radioactive substances that could impact employee or public safety or the environment. A facility effluent monitoring plan determination was performed during Calendar Year 1991 and the evaluation requires the need for a facility effluent monitoring plan. This document is prepared using the specific guidelines identified. in. A Guide for Preparing Hanford Site Facility Effluent Monitoring Plans, WHC-EP-0438. This facility effluent monitoring plan assesses effluent monitoring systems and evaluates whether they are adequate to ensure the public health and safety as specified in applicable federal, state, and local requirements

  1. Facility effluent monitoring plan for the plutonium-uranium extraction facility

    International Nuclear Information System (INIS)

    Lohrasbi, J.; Johnson, D.L.; De Lorenzo, D.S.

    1993-12-01

    A facility effluent monitoring plan is required by the US Department of Energy in DOE Order 5400.1 for any operations that involve hazardous materials and radioactive substances that could impact employee or public safety or the environment. This document is prepared using the specific guidelines identified in A Guide for Preparing Hanford Site Facility Effluent Monitoring Plans, WHC-EP-0438-01. This facility effluent monitoring plan assesses effluent monitoring systems and evaluates whether they are adequate to ensure the public health and safety as specified in applicable federal, state, and local requirements. This facility effluent monitoring plan shall ensure long-range integrity of the effluent monitoring systems by requiring an update whenever a new process or operation introduces new hazardous materials or significant radioactive materials. This document must be reviewed annually even if there are no operational changes, and it must be updated at a minimum of every three years

  2. Facility effluent monitoring plan for the plutonium-uranium extraction facility

    International Nuclear Information System (INIS)

    Thompson, R.J.; Sontage, S.

    1991-11-01

    A facility effluent monitoring plan is required by the US Department of Energy in DOE Order 5400.1 for any operations that involve hazardous materials and radioactive substances that could impact employee or public safety or the environment. This document is prepared using the specific guidelines identified in A Guide for Preparing Hanford Site Facility Effluent Monitoring Plans, WHC-EP-0438. This facility effluent monitoring plan assesses effluent monitoring systems and evaluates whether they are adequate to ensure the public health and safety as specified in applicable federal, state, and local requirements. This facility effluent monitoring plan is the first annual report. It shall ensure long-range integrity of the effluent monitoring systems by requiring an update whenever a new process or operation introduces new hazardous materials or significant radioactive materials. This document must be reviewed annually even if there are no operational changes, and it must be updated as a minimum every three years

  3. Facility effluent monitoring plan for the plutonium-uranium extraction facility

    Energy Technology Data Exchange (ETDEWEB)

    Lohrasbi, J.; Johnson, D.L. [Westinghouse Hanford Co., Richland, WA (United States); De Lorenzo, D.S. [Los Alamos Technical Associates, NM (United States)

    1993-12-01

    A facility effluent monitoring plan is required by the US Department of Energy in DOE Order 5400.1 for any operations that involve hazardous materials and radioactive substances that could impact employee or public safety or the environment. This document is prepared using the specific guidelines identified in A Guide for Preparing Hanford Site Facility Effluent Monitoring Plans, WHC-EP-0438-01. This facility effluent monitoring plan assesses effluent monitoring systems and evaluates whether they are adequate to ensure the public health and safety as specified in applicable federal, state, and local requirements. This facility effluent monitoring plan shall ensure long-range integrity of the effluent monitoring systems by requiring an update whenever a new process or operation introduces new hazardous materials or significant radioactive materials. This document must be reviewed annually even if there are no operational changes, and it must be updated at a minimum of every three years.

  4. Extraction, Scrub, and Strip Test Results for the Salt Waste Processing Facility Caustic Side Solvent Extraction Solvent Sample

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-10-06

    An Extraction, Scrub, and Strip (ESS) test was performed on a sample of Salt Waste Processing Facility (SWPF) Caustic-Side Solvent Extraction (CSSX) solvent and salt simulant to determine cesium distribution ratios (D(Cs)), and cesium concentration in the strip effluent (SE) and decontaminated salt solution (DSS) streams; this data will be used by Parsons to help determine if the solvent is qualified for use at the SWPF. The ESS test showed acceptable performance of the solvent for extraction, scrub, and strip operations. The extraction D(Cs) measured 12.5, exceeding the required value of 8. This value is consistent with results from previous ESS tests using similar solvent formulations. Similarly, scrub and strip cesium distribution ratios fell within acceptable ranges. This revision was created to correct an error. The previous revision used an incorrect set of temperature correction coefficients which resulted in slight deviations from the correct D(Cs) results.

  5. Extraction, scrub, and strip test results for the salt waste processing facility caustic side solvent extraction solvent example

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-08-01

    An Extraction, Scrub, and Strip (ESS) test was performed on a sample of Salt Waste Processing Facility (SWPF) Caustic-Side Solvent Extraction (CSSX) solvent and salt simulant to determine cesium distribution ratios (D(Cs)), and cesium concentration in the strip effluent (SE) and decontaminated salt solution (DSS) streams; this data will be used by Parsons to help determine if the solvent is qualified for use at the SWPF. The ESS test showed acceptable performance of the solvent for extraction, scrub, and strip operations. The extraction D(Cs) measured 12.9, exceeding the required value of 8. This value is consistent with results from previous ESS tests using similar solvent formulations. Similarly, scrub and strip cesium distribution ratios fell within acceptable ranges.

  6. Filtration and Leach Testing for PUREX Cladding Sludge and REDOX Cladding Sludge Actual Waste Sample Composites

    Energy Technology Data Exchange (ETDEWEB)

    Shimskey, Rick W.; Billing, Justin M.; Buck, Edgar C.; Casella, Amanda J.; Crum, Jarrod V.; Daniel, Richard C.; Draper, Kathryn E.; Edwards, Matthew K.; Hallen, Richard T.; Kozelisky, Anne E.; MacFarlan, Paul J.; Peterson, Reid A.; Swoboda, Robert G.

    2009-03-02

    A testing program evaluating actual tank waste was developed in response to Task 4 from the M-12 External Flowsheet Review Team (EFRT) issue response plan (Barnes and Voke 2006). The test program was subdivided into logical increments. The bulk water-insoluble solid wastes that are anticipated to be delivered to the Hanford Waste Treatment and Immobilization Plant (WTP) were identified according to type such that the actual waste testing could be targeted to the relevant categories. Under test plan TP RPP WTP 467 (Fiskum et al. 2007), eight broad waste groupings were defined. Samples available from the 222S archive were identified and obtained for testing. Under this test plan, a waste testing program was implemented that included: • Homogenizing the archive samples by group as defined in the test plan. • Characterizing the homogenized sample groups. • Performing parametric leaching testing on each group for compounds of interest. • Performing bench-top filtration/leaching tests in the hot cell for each group to simulate filtration and leaching activities if they occurred in the UFP2 vessel of the WTP Pretreatment Facility. This report focuses on a filtration/leaching test performed using two of the eight waste composite samples. The sample groups examined in this report were the plutonium-uranium extraction (PUREX) cladding waste sludge (Group 3, or CWP) and reduction-oxidation (REDOX) cladding waste sludge (Group 4, or CWR). Both the Group 3 and 4 waste composites were anticipated to be high in gibbsite, thus requiring caustic leaching. WTP RPT 167 (Snow et al. 2008) describes the homogenization, characterization, and parametric leaching activities before benchtop filtration/leaching testing of these two waste groups. Characterization and initial parametric data in that report were used to plan a single filtration/leaching test using a blend of both wastes. The test focused on filtration testing of the waste and caustic leaching for aluminum, in the form

  7. Purex pulse column designs for capacity factor of 3.0 to 3.5

    Energy Technology Data Exchange (ETDEWEB)

    Richardson, G.L.

    1955-04-12

    This memorandum indicates the Purex-Plant pulse-column and pulse- generator revisions which would be required to assure an instantaneous capacity of 25 tons U/day with a 20% capacity safety margin under Purex HW {number_sign}3 Flowsheet conditions. (The use of the Purex HW {number_sign}4 Flowsheet (6) with the revised columns would be expected to increase the capacity to 29 or 30 tons U/day.) The indicated design changes are recorded here for study and for possible reference if need for increased production capacity should arise. No recommendation for adoption at this time is made.

  8. Chemical-technology investigation of modified purex process for reprocessing of spent nuclear fuel, Annex 1; Prilog 1: Hemijsko-tehnolosko ispitivanje modifikovanog 'purex proces' za preradu isluzenog nuklearnog goriva

    Energy Technology Data Exchange (ETDEWEB)

    Tolic, A; Stefanovic, M [Institute of Nuclear Sciences Boris Kidric, Laboratorija za visoku aktivnost, Vinca, Beograd (Serbia and Montenegro)

    1963-12-15

    The objective of the task in this year was to verify the first part of the modified purex process which covers the operation of the two most important extractors HA and HS. Special attention was paid to the fact that the testing results in laboratory conditions must be identical to the results in the industrial process. The experimental part of the task was divided in the following phases: preparation of the uranium solution; preparation of the equipment; testing of the uranium extraction and nitric acid; testing the decontamination of the organic phase; testing of plutonium extraction and HNO{sub 3}. A high number of control chemical and radiochemical analyses had to be completed, as well as a number of preliminary calculations, which are presented in this report.

  9. Advanced Purex process and waste minimization at La Hague

    International Nuclear Information System (INIS)

    Masson, H.; Nouguier, H.; Bernard, C.; Runge, S.

    1993-01-01

    After a brief recall of the different aspects of the commercial irradiated fuel reprocessing, this paper presents the achievements of the recently commissioned UP3 plant at La Hague. The advanced Purex process implemented with a total waste management results in important waste volume minimization, so that the total volume of high-level and transuranic waste is lower than what it would be in a once-through cycle. Moreover, further minimization is still possible, based on an improved waste management. Cogema has launched the necessary program, which will lead to an overall volume of HLW and TRU wastes of less than 1 m 3 /t by the end of the decade, the maximum possible activity being concentrated in the glass

  10. Extraction, scrub, and strip test results for the solvent transfer to salt waste processing facility

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-09-07

    The Savannah River National Laboratory (SRNL) prepared approximately 240 gallons of Caustic-Side Solvent Extraction (CSSX) solvent for use at the Salt Waste Processing Facility (SWPF). An Extraction, Scrub, and Strip (ESS) test was performed on a sample of the prepared solvent using a salt solution prepared by Parsons to determine cesium distribution ratios (D(Cs)), and cesium concentration in the strip effluent (SE) and decontaminated salt solution (DSS) streams. This data will be used by Parsons to help qualify the solvent for use at the SWPF. The ESS test showed acceptable performance of the solvent for extraction, scrub, and strip operations. The extraction D(Cs) measured 15.5, exceeding the required value of 8. This value is consistent with results from previous ESS tests using similar solvent formulations. Similarly, scrub and strip cesium distribution ratios fell within acceptable ranges.

  11. Investigation on clean-up of Zr and HDBP in PUREX process with UDMH oxalate

    International Nuclear Information System (INIS)

    Zhang Youzhi; Wang Xuanjun; Li Zhengli; Liu Xiangxuan

    2007-01-01

    It is generally accepted that the interracial crud formation is related to the complex formation of Zr with degradation products of TBP, such as DBP and MBP, in PUREX process, especially in the first cycle. The crud seriously deteriorates the operation of extraction column and therefore must be properly cleared up. Various clear up methods were studied and those with salt-free washing agents were recently focused. In this paper a new scrubbing agent 1,1- dimethylhydrazine (UDMH) oxalate was proposed, the optimized experimental conditions were described, and the possible mechanism was discussed. The influence of different factors, including reaction temperature, UDMH oxalate concentration, organic-to-aqueous phase ratio, and free UDMH concentration, on the decontamination factors were examined with simulated Zr- and/or DBP-loaded solvents. The optical experimental parameters are found as follows: temperature 40-60 degree C, phase ratio V (o) /V (a) =1, concentration of UDMH oxalate solution 0.4-0.6 mol/L. Especialy some UDMH was added into the UDMH oxalate queues solution to make the concentration of free UDMH 0.2-0.3 mol/L. Under these conditions, the decontaminator factor of Zr from the corresponding simulated solvent with UDMH oxalate is up to 143, slightly higher than that with sodium carbonate. The decontamination factor of HDBP from the corresponding simulated solvent with UDMH oxalate is up to 100, similar to sodium carbonate. (authors)

  12. Simplified nuclear fuel reprocessing flowsheet: a single-cycle Purex process

    International Nuclear Information System (INIS)

    Montuir, M.; Dinh, B.; Baron, P.

    2004-01-01

    A simplified flowsheet with only one purification cycle instead of three is proposed for reprocessing spent nuclear fuel using the Purex process. A single-cycle flowsheet minimizes the process equipment required, the number of control points before transfer between process units, and the solvent and effluent quantities. For the uranium stream, an alpha barrier is used to strip any residual contaminants (Np, Th, Pu) from the uranium-loaded solvent. This additional step eliminates the need for a second uranium cycle. For the plutonium stream, an additional βγ co-decontamination step and a higher plutonium concentration are required before the oxalate conversion step; a plutonium 'half-cycle' is added downstream. The unloaded solvent from this half-cycle is returned to the selective plutonium stripping step, allowing significant plutonium half-cycle losses. It should be possible to reduce the number of stages in the half-cycle extraction step by recycling the raffinate to the upstream separation process. (authors)

  13. Strategy and current state of research on enhanced iodine separation during spent fuel reprocessing by the Purex process

    International Nuclear Information System (INIS)

    Devisme, F.; Juvenelle, A.; Touron, E.

    2001-01-01

    An enhanced separation process designed to recover and purify molecular iodine desorbed during dissolution is described in the context of 129 I management in the Purex process for transmutation or interim storage. It involves reducing acid scrubbing with hydroxyl-ammonium nitrate followed by oxidation with hydrogen peroxide to obtain selective desorption. The stoichiometry and kinetics are determined for each step and an experimental validation program is now in progress using a small pilot facility equipped with a scrubbing column. The technical feasibility of the process has already been demonstrated: room-temperature scrubbing with a HAN solution (0,5 mol.L -1 ) at a pH of about 5 results in 99% iodine trapping efficiency; the subsequent desorption yield is 99,5%. (author)

  14. Strategy and current state of research on enhanced iodine separation during spent fuel reprocessing by the Purex process

    Energy Technology Data Exchange (ETDEWEB)

    Devisme, F.; Juvenelle, A.; Touron, E. [CEA Valrho, Dir. de l' Energie Nucleaire, DEN/DRCP, 30 - Marcoule (France)

    2001-07-01

    An enhanced separation process designed to recover and purify molecular iodine desorbed during dissolution is described in the context of {sup 129}I management in the Purex process for transmutation or interim storage. It involves reducing acid scrubbing with hydroxyl-ammonium nitrate followed by oxidation with hydrogen peroxide to obtain selective desorption. The stoichiometry and kinetics are determined for each step and an experimental validation program is now in progress using a small pilot facility equipped with a scrubbing column. The technical feasibility of the process has already been demonstrated: room-temperature scrubbing with a HAN solution (0,5 mol.L{sup -1}) at a pH of about 5 results in 99% iodine trapping efficiency; the subsequent desorption yield is 99,5%. (author)

  15. Spectrophotometric determination of nitrite in simulated Purex Process solutions

    Energy Technology Data Exchange (ETDEWEB)

    Araujo, I.daC. de; Matsuda, H T; Araujo, B.F. de; Araujo, J.A. de

    1984-01-01

    A spectrophotometric method for nitrite determination in simulated Purex Process solutions is presented, utilizing the Griess reagent for the formation of the coloured azocompound with an absorption maximum at 525 nm. Molar absortivity was 36,262 and the sensitivity of the method 10/sup -6/M for nitrite. The calibration curve is linear in the range of 2 to 30..mu..g NO/sup -//sub 2//25 ml in cells of 1 cm optical path. The method can be used in the presence of uranium up to limits of an U/NO/sup -//sub 2/ ratio of 150. Test solutions were prepared to simulate composition and concentrations as obtained by irradiating standard fuel with a neutro flux of 3.2 x 10/sup 13/ n.s/sup -1/.cm/sup -2/, with a burn-up value of 33,000 Mwd/T and cooling time of two years. Nitrite determinations in these solutions were accurate within limits of 5%.

  16. An automatic device for sample insertion and extraction to/from reactor irradiation facilities

    International Nuclear Information System (INIS)

    Alloni, L.; Venturelli, A.; Meloni, S.

    1990-01-01

    At the previous European Triga Users Conference in Vienna,a paper was given describing a new handling tool for irradiated samples at the L.E.N.A plant. This tool was the first part of an automatic device for the management of samples to be irradiated in the TRIGA MARK ii reactor and successively extracted and stored. So far sample insertion and extraction to/from irradiation facilities available on reactor top (central thimble,rotatory specimen rack and channel f),has been carried out manually by reactor and health-physics operators using the ''traditional'' fishing pole provided by General Atomic, thus exposing reactor personnel to ''unjustified'' radiation doses. The present paper describes the design and the operation of a new device, a ''robot''type machine,which, remotely operated, takes care of sample insertion into the different irradiation facilities,sample extraction after irradiation and connection to the storage pits already described. The extraction of irradiated sample does not require the presence of reactor personnel on the reactor top and,therefore,radiation doses are strongly reduced. All work from design to construction has been carried out by the personnel of the electronic group of the L.E.N.A plant. (orig.)

  17. Effect of di-butyl phosphate on flash point of PUREX solvent

    International Nuclear Information System (INIS)

    Srivastav, Ravi Kant; Kumar, Shekhar; Balasubramonian, S.; Kamachi Mudali, U.; Natarajan, R.

    2015-01-01

    30% Tri-n-butyl phosphate (TBP) in a aliphatic diluent is used as a solvent for PUREX process. This diluent is essentially equivalent to commercial dodecane. The radiolytic and acidic degradation of TBP forms di-butyl phosphate (DBP) which is detrimental to the performance of the solvent during nuclear fuel reprocessing operations. To study the possible effect of DBP on the flashpoint of PUREX solvent, synthetic solutions were made by adding DBP and flashpoints of resultant mixtures were determined with an automatic flashpoint tester as per ASTM procedures. Experimental results indicated virtually no effect of DBP on flash point of PUREX solvent in the concentration ranges of 0-16 g/L DBP. (author)

  18. Forefront of PUREX system engineering. Chemistry and engineering of ruthenium, technetium and neptunium

    International Nuclear Information System (INIS)

    2004-07-01

    The paper reports the activity of the research committee organized by the Atomic Energy Society of Japan on 'Ruthenium and Technetium Chemistry in the PUREX System', with focusing on basic behaviors of ruthenium, technetium and neptunium in the PUREX process, the principles of plant design, and behaviors during the final waste treatment. The scope of the work includes the following major topics: (1) basic solution and solid-state chemistry; (2) basic solution and solid-state chemistry of minor actinides in particular, Np; (3) partitioning chemistry in the PUREX system and environmental behavior of the components; (4) processes of recovery, purification, and utilization of rare metal fission products; (5) field data on plant design, operation, decontamination, and decommissioning; (6) numerical process simulations and process control technologies; (7) compilation of a data base for process chemistry and plant engineering. (S. Ohno)

  19. Commercial Light Water Reactor -Tritium Extraction Facility Process Waste Assessment (Project S-6091)

    Energy Technology Data Exchange (ETDEWEB)

    Hsu, R.H.; Delley, A.O.; Alexander, G.J.; Clark, E.A.; Holder, J.S.; Lutz, R.N.; Malstrom, R.A.; Nobles, B.R. [Westinghouse Savannah River Co., Aiken, SC (United States); Carson, S.D. [Sandia National Laboratories, New Mexico, NM (United States); Peterson, P.K. [Sandia National Laboratories, New Mexico, NM (United States)

    1997-11-30

    The Savannah River Site (SRS) has been tasked by the Department of Energy (DOE) to design and construct a Tritium Extraction Facility (TEF) to process irradiated tritium producing burnable absorber rods (TPBARs) from a Commercial Light Water Reactor (CLWR). The plan is for the CLWR-TEF to provide tritium to the SRS Replacement Tritium Facility (RTF) in Building 233-H in support of DOE requirements. The CLWR-TEF is being designed to provide 3 kg of new tritium per year, from TPBARS and other sources of tritium (Ref. 1-4).The CLWR TPBAR concept is being developed by Pacific Northwest National Laboratory (PNNL). The TPBAR assemblies will be irradiated in a Commercial Utility light water nuclear reactor and transported to the SRS for tritium extraction and processing at the CLWR-TEF. A Conceptual Design Report for the CLWR-TEF Project was issued in July 1997 (Ref. 4).The scope of this Process Waste Assessment (PWA) will be limited to CLWR-TEF processing of CLWR irradiated TPBARs. Although the CLWR- TEF will also be designed to extract APT tritium-containing materials, they will be excluded at this time to facilitate timely development of this PWA. As with any process, CLWR-TEF waste stream characteristics will depend on process feedstock and contaminant sources. If irradiated APT tritium-containing materials are to be processed in the CLWR-TEF, this PWA should be revised to reflect the introduction of this contaminant source term.

  20. Commercial Light Water Reactor -Tritium Extraction Facility Process Waste Assessment (Project S-6091)

    International Nuclear Information System (INIS)

    Hsu, R.H.; Delley, A.O.; Alexander, G.J.; Clark, E.A.; Holder, J.S.; Lutz, R.N.; Malstrom, R.A.; Nobles, B.R.; Carson, S.D.; Peterson, P.K.

    1997-01-01

    The Savannah River Site (SRS) has been tasked by the Department of Energy (DOE) to design and construct a Tritium Extraction Facility (TEF) to process irradiated tritium producing burnable absorber rods (TPBARs) from a Commercial Light Water Reactor (CLWR). The plan is for the CLWR-TEF to provide tritium to the SRS Replacement Tritium Facility (RTF) in Building 233-H in support of DOE requirements. The CLWR-TEF is being designed to provide 3 kg of new tritium per year, from TPBARS and other sources of tritium (Ref. 1-4).The CLWR TPBAR concept is being developed by Pacific Northwest National Laboratory (PNNL). The TPBAR assemblies will be irradiated in a Commercial Utility light water nuclear reactor and transported to the SRS for tritium extraction and processing at the CLWR-TEF. A Conceptual Design Report for the CLWR-TEF Project was issued in July 1997 (Ref. 4).The scope of this Process Waste Assessment (PWA) will be limited to CLWR-TEF processing of CLWR irradiated TPBARs. Although the CLWR- TEF will also be designed to extract APT tritium-containing materials, they will be excluded at this time to facilitate timely development of this PWA. As with any process, CLWR-TEF waste stream characteristics will depend on process feedstock and contaminant sources. If irradiated APT tritium-containing materials are to be processed in the CLWR-TEF, this PWA should be revised to reflect the introduction of this contaminant source term

  1. EXTRA·M: a computing code system for analysis of the Purex process with mixer settlers for reprocessing

    International Nuclear Information System (INIS)

    Tachimori, Shoichi

    1994-03-01

    A computer code system EXTRA·M, for simulation of transient behavior of the solutes in a multistage countercurrent extraction process, was developed aiming to predict the distribution and chemical behaviors of actinide elements, i.e., U, Pu, Np, and of technetium in the Purex process of fuel reprocessing. The mathematical model is applicable to a complete mixing stagewise contactor such as mixer settler and to the Purex, with tri-n-butylphosphate (TBP) and nitric acid system. The main characteristics of the EXTRA·M are as follows; i) Calculation of distribution ratios of the solutes is based on numerical equations of which parameter values are to be determined by a best fit method with a number of experimental data. ii) Total of 18 solutes; U(IV), U(VI), Pu(III), Pu(IV), Pu(V), Pu(VI), Np(IV), Np(V), Np(VI), Tc(IV), Tc(V), Tc(VI), Tc(VII), Zr(IV), HNO 3 , hydrazine, hydroxylamine nitrate and nitrous acid, are treated and rate equations of total 40 chemical reactions involving these solutes are incorporated. iii) Instantaneous change of flow conditions, i.e., concentration of the solutes and flow rate of the feeding solutions, is contrived by computation. iv) Reflux or bypass mode calculation, in which an aqueous raffinate stream is transferred to the preceding bank or stage, is possible. The present report explains the concept, assumptions and characteristics of the model, the material balance equations including distribution and reaction rate equations and their solution method, and the usefulness of the model by showing some examples of the verification results. A description and source program of EXTRA·M1, as an example, are listed in the annex. (J.P.N.) 63 refs

  2. Automated DNA extraction platforms offer solutions to challenges of assessing microbial biofouling in oil production facilities.

    Science.gov (United States)

    Oldham, Athenia L; Drilling, Heather S; Stamps, Blake W; Stevenson, Bradley S; Duncan, Kathleen E

    2012-11-20

    The analysis of microbial assemblages in industrial, marine, and medical systems can inform decisions regarding quality control or mitigation. Modern molecular approaches to detect, characterize, and quantify microorganisms provide rapid and thorough measures unbiased by the need for cultivation. The requirement of timely extraction of high quality nucleic acids for molecular analysis is faced with specific challenges when used to study the influence of microorganisms on oil production. Production facilities are often ill equipped for nucleic acid extraction techniques, making the preservation and transportation of samples off-site a priority. As a potential solution, the possibility of extracting nucleic acids on-site using automated platforms was tested. The performance of two such platforms, the Fujifilm QuickGene-Mini80™ and Promega Maxwell®16 was compared to a widely used manual extraction kit, MOBIO PowerBiofilm™ DNA Isolation Kit, in terms of ease of operation, DNA quality, and microbial community composition. Three pipeline biofilm samples were chosen for these comparisons; two contained crude oil and corrosion products and the third transported seawater. Overall, the two more automated extraction platforms produced higher DNA yields than the manual approach. DNA quality was evaluated for amplification by quantitative PCR (qPCR) and end-point PCR to generate 454 pyrosequencing libraries for 16S rRNA microbial community analysis. Microbial community structure, as assessed by DGGE analysis and pyrosequencing, was comparable among the three extraction methods. Therefore, the use of automated extraction platforms should enhance the feasibility of rapidly evaluating microbial biofouling at remote locations or those with limited resources.

  3. Photochemical technique for reduction of uranium and subsequently plutonium in the Purex process

    International Nuclear Information System (INIS)

    Goldstein, M.; Barker, J.J.; Gangwer, T.

    1976-09-01

    A photochemical modification of the Purex process is described in which a purified side stream of UO 2 ++ ion is reduced to U +4 outside the radioactive area of the reprocessing plant. The U +4 is then cycled back to step 2 of the Purex process to reduce the plutonium and effect separation within the partitioning column. This process is shown to be very energy efficient and compatible with existing conventional lamp technology. Preliminary cost estimates of the energy requirements for photon production are essentially negligible. Conceptual systems and photochemical reactor designs are presented. Potential benefits of this system are discussed

  4. Evaluation of consequence due to higher hydrazine content in partitioning stream of PUREX process

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, K. Suresh [Bhabha Atomic Research Centre, Mumbai (India). Special Nuclear Recycle Facility

    2016-07-01

    Hydrazine nitrate is being used as a stabilizer for U(IV) as well as Pu(III) during partitioning of Pu in PUREX process by scavenging the nitrous acid present along with nitric acid. As hydrazine hydrate as well as its salts have been successfully used for scrubbing of degradation products of TBP to aqueous phase, experiments were conducted to evaluate the consequence of hydrazine content during Pu partitioning. It was observed that higher amount of hydrazine nitrate along with uranous nitrate in the partitioning stream of PUREX process leads to build up of DBP in aqueous phase and resulted in precipitation of Pu.

  5. A method of neptunium recovery into the product stream of the Purex 1st codecontamination step for LWR fuel reprocessing

    International Nuclear Information System (INIS)

    Tsuboya, Takao; Nemoto, Shinichi; Hoshino, Tadaya; Segawa, Takeshi

    1973-01-01

    An improved nitrous acid method was applied for recovering neptunium in spent fuel. Counter-current solvent extraction has been performed to find out its recovery conditions. The nitrous acid in the form of sodium salt solution was fed to the 1st stage of extraction section, and hydrazine nitrate was fed to some stages near feed point. Flow rate and the concentration of additives were altered for finding out optimum condition. Laboratory scale mixer-settlers having 6 ml of mixing volume and 17 ml of settling volume for each stage were used. The nitrous acid method was improved so that the reduction reaction in scrub section can be eliminated by the decomposition of the nitrous acid using a reagent such as sulfamic acid, urea, or hydrazine. In operation, the feed rate of the nitrous acid was about 3 mM/hr, and about 61% of neptunium charged was discharged in the product stream of Purex-1st codecontamination step designed for the LWR fuel reprocessing plant of Power Reactor and Nuclear Fuel Development Corporation. The calculated value of Δx/x for extraction section agreed with the experimental value, where Δx is the quantity of oxidation, and x is the inventory for neptunium in each stage. In conclusion, the improved nitrous acid method is effective for the neptunium discharge in product stream, and the difference of neptunium extraction between estimate and experiment is caused by some of reduction reaction in scrub section. (Iwakiri, K.)

  6. A facile and green preparation of reduced graphene oxide using Eucalyptus leaf extract

    Science.gov (United States)

    Li, Chengyang; Zhuang, Zechao; Jin, Xiaoying; Chen, Zuliang

    2017-11-01

    In this paper, a green and facile synthesis of reduced graphene oxide (GO) by Eucalyptus leaf extract (EL-RGO) was investigated, which was characterized with ultraviolet-visible spectroscopy (UV), Raman spectroscopy, X-ray diffraction (XRD), scanning electron microscope (SEM), atomic force microscopy (AFM), X-ray photoelectron spectroscopy (XPS) and Thermal gravimetric analysis (TG). Eucalyptus leaf extract also play both reducing and capping stabilizing agents prepared EL-RGO as shown a good stability and electrochemical properties. This approach could provide an alternative method to prepare EL-RGO in large-scale production. Moreover, the good electrochemical property and biocompatibility can be used in various applications. In addition, the merit of this study is that both the oxidized products and the reducing agents are environmental friendly by green reduction.

  7. Potato extract as reducing agent and stabiliser in a facile green one-step synthesis of ZnO nanoparticles

    DEFF Research Database (Denmark)

    Buazar, Foad; Bavi, M.; Kroushawi, Feisal

    2016-01-01

    A facile green recipe was developed to synthesise highly pure, safe and durable zinc oxide nanoparticles (ZnO Nps) using homemade starch-rich potato extract. The ZnO Nps were synthesised using zinc nitrate and potato extract, and the whole reaction is carried out for 30 min at 80 °C...

  8. 324 Facility special-case waste assessment in support of 324 closure (TPA milestone M-89-05)

    International Nuclear Information System (INIS)

    Hobart, R.L.

    1998-01-01

    Hanford Federal Facility Agreement and Consent Order, also known as the Tri-Party Agreement Milestone M-89-05, requires US Department of Energy, Richland Operations Office to complete a 324 Facility Special-Case Waste Assessment in Support of 324 Closure. This document, HNF-1270, has been prepared with the intent of meeting this regulatory commitment. Alternatives for the special-case wastes located in the 324 Building were defined and analyzed. Based on the criteria of safety, environmental, complexity of interfaces, risk, cost, schedule, and long-term operability and maintainability, the best alternative was chosen. Waste packaging and transportation options are also included in the recommendations. The waste disposition recommendations for the B-Cell dispersibles/tank heels and High-Level Vault packaged residuals are to direct them to the Plutonium Uranium Extraction Facility (PUREX) Number 2 storage tunnel

  9. Criticality prevention specifications thorium--uranium-233 separations in the Purex Plant

    International Nuclear Information System (INIS)

    Matheison, W.E.; Oberg, G.C.; Ritter, G.L.

    1970-01-01

    The specifications in this document define the limits or restrictions required to maintain an acceptably low probability of the occurrence of a nuclear chain reaction in the Purex Plant while processing irradiated thoria targets. These criticality prevention specifications do not stipulate the system, procedures, or mechanisms to permit operation within the limits or restrictions

  10. Process specifications and standards for the 1970 thorium campaign in the Purex Plant

    International Nuclear Information System (INIS)

    Van der Cook, R.E.; Ritter, G.L.

    1970-01-01

    The process specifications and standards for thorium processing operations in the Purex Plant are presented. These specifications represent currently known limits within which plant processing conditions must be maintained to meet defined product requirements safely and with minimum effect on equipment service life. These specifications cover the general areas of feed, essential materials, and chemical hazards

  11. Production and remediation of low sludge simulated Purex waste glasses, 2: Effects of sludge oxide additions on glass durability

    International Nuclear Information System (INIS)

    Ramsey, W.G.

    1993-01-01

    Glass produced during the Purex 4 campaigns of the Integrated DWPF Melter System (IDMS) and the 774 Research Melter contained a lower fraction of sludge components than targeted by the Product Composition Control System (PCCS). Purex 4 glass was more durable than the benchmark (EA) glass, but was less durable than most other simulated SRS high-level waste glasses. Further, the measured durability of Purex 4 glass was not as well correlated with the durability predicted from the DWPF process control algorithm, probably because the algorithm was developed to predict the durability of SRS high-level waste glasses with higher sludge content than Purex 4. A melter run, designated Purex 4 Remediation, was performed using the 774 Research Melter to determine if the initial PCCS target composition determined for Purex 4 would produce acceptable glass whose durability could be accurately modeled by the DWPF glass durability algorithm. Reagent grade oxides and carbonates were added to Purex 4 melter feed stock to simulate a higher sludge loading. Each canister of glass produced was sampled and the glass durability was determined by the Product Consistency Test method. This document details the durability data and subsequent analysis

  12. The beam slow extraction from a magnetic ring of Moscow meson facility

    International Nuclear Information System (INIS)

    Gusev, O.A.; Malitsky, N.D.; Severgin, Yu.P.; Titov, V.A.; Shukeilo, I.A.; Aseev, V.N.; Grachev, M.I.; Lobashev, V.M.; Ostroumov, P.N.; Ponomaryov, O.V.

    1990-01-01

    The beam slow extraction from the circular accelerators or stretcher rings is generally realized by the resonant excitation of betratron oscillations. A precise betatron frequency control is proved to be quite necessary for high-efficient slow ejection. The Coulomb field turns out to have a significant influence upon the slow extraction from the high-current medium energy proton storage rings. It prevents resonant excitation at a reasonable rate and reduces the ejection efficiency. The proton storage ring of Moscow meson facility is an example of a stretcher with a noticeable beam space charge. The detailed investigation of the resonant ejection, having been performed for our stretcher, resulted in the conclusion that extracted beam average current should be limited by the value of 50 mA, which is only 10% of the linac design current. The search for the alternative version to the resonant ejection made us to analyze in details and to develop an old-fashioned method, based on the radial betatron oscillation excitation while the beam is being gradually shifted onto the thin target. (author) 5 refs., 4 figs

  13. Facility effluent monitoring plan for the uranium trioxide facility

    International Nuclear Information System (INIS)

    Thompson, R.J.; Sontag, S.

    1991-11-01

    A facility effluent monitoring plant is required by the US Department of Energy in DOE Order 5400.1 for any operations that involve hazardous materials and radioactive substances that could impact employee or public safety or the environment. This document is prepared using the specific guidelines identified in A Guide for Preparing Hanford Site Facility Effluent Monitoring Plans, WHC-EP-0438. This facility effluent monitoring plan assesses effluent monitoring systems and evaluates whether they are adequate to ensure the public health and safety as specified in applicable federal, state, and local requirements. This facility effluent monitoring plan is the first annual report. It shall ensure long-range integrity of the effluent monitoring systems by requiring an update whenever a new process or operation introduces new hazardous materials or significant radioactive materials. This document must be reviewed annually even if there are no operational changes, and it must be updated as a minimum every three years. The UO 3 Plant is located in the south-central portion of the 200 West Area of the Hanford Site. The plant consists of two primary processing buildings and several ancillary facilities. The purpose of the UO 3 Plant is to receive uranyl nitrate hexahydrate (UNH) from the Plutonium-Uranium Extraction (PUREX) Plant, concentrate it, convert the UNH to uranium trioxide (UO 3 ) powder by calcination and package it for offsite shipment. The UO 3 Plant has been placed in a standby mode. There are two liquid discharges, and three gaseous exhaust stacks, and seven building exhausters that are active during standby conditions

  14. Commercial Light Water Reactor Tritium Extraction Facility. Geotechnical Summary report (U)

    International Nuclear Information System (INIS)

    McHood, M.D.

    2000-09-01

    A geotechnical investigation program has been completed for the Commercial Light Water Reactor - Tritium Extraction Facility (CLWR-TEF) at the Savannah River Site (SRS). The program consisted of reviewing previous geotechnical and geologic data and reports, performing subsurface field exploration, field and laboratory testing, and geologic and engineering analyses. The purpose of this investigation was to characterize the subsurface conditions for the CLWR-TEF in terms of subsurface stratigraphy and engineering properties for design and to perform selected engineering analyses. The objectives of the evaluation were to establish site-specific geologic conditions, obtain representative engineering properties of the subsurface and potential fill materials, evaluate the lateral and vertical extent of any soft zones encountered, and perform engineering analyses for slope stability, bearing capacity and settlement, and liquefaction potential. In addition, provide general recommendations for construction and earthwork

  15. Method of neptunium recovery into the product stream of the Purex second codecontamination step for LWR fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Tsuboya, T; Nemoto, S; Hoshino, T; Segawa, T [Power Reactor and Nuclear Fuel Development Corp., Tokyo (Japan)

    1973-04-01

    The neptunium behavior in the second codecontamination step in Purex process of Power Reactor and Nuclear Fuel Development Corporation was experimentally studied, and the conditions for discharging neptunium in product stream were examined. Improved nitrous acid method was applied to the second codecontamination step. Nitrous acid (NaNO/sub 2/) was supplied to the 1st stage of extraction section at feed rate of 7.5 mM/hr, and hydrazine (hydrazine nitrate) was supplied to some stages near feed point at feed rate of 1.6 mM/hr, by using laboratory scale mixer-settlers having 6 ml of mixing volume and 17 ml of settling volume. Neptunium extraction behavior was analyzed by the code NEPTUN-I simulating neptunium concentration profile and by the code NEPTUN-II for calculating Np (V) and Np (VI) concentration. Batch experiments were performed for explaining the reduction reaction of Np (VI) in organic phase. After shaking the aqueous solution containing Np (VI) in 3 M nitric acid with the various volume ratios of TBP, both phases were separated, and the neptunium concentration was determined. In conclusion, the improved nitrous acid method was effective for the neptunium discharge in product stream when the flow ratio of organic phase to aqueous phase was increased to about three times.

  16. Analysis Methods for Extracting Knowledge from Large-Scale WiFi Monitoring to Inform Building Facility Planning

    DEFF Research Database (Denmark)

    Ruiz-Ruiz, Antonio; Blunck, Henrik; Prentow, Thor Siiger

    2014-01-01

    realistic data to inform facility planning. In this paper, we propose analysis methods to extract knowledge from large sets of network collected WiFi traces to better inform facility management and planning in large building complexes. The analysis methods, which build on a rich set of temporal and spatial......The optimization of logistics in large building com- plexes with many resources, such as hospitals, require realistic facility management and planning. Current planning practices rely foremost on manual observations or coarse unverified as- sumptions and therefore do not properly scale or provide....... Spatio-temporal visualization tools built on top of these methods enable planners to inspect and explore extracted information to inform facility-planning activities. To evaluate the methods, we present results for a large hospital complex covering more than 10 hectares. The evaluation is based on Wi...

  17. Simulation of solvent extraction in reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Shekhar; Koganti, S B [Reprocessing Group, Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    1994-06-01

    A SIMulation Program for Solvent EXtraction (SIMPSEX) has been developed for simulation of PUREX process used in nuclear fuel reprocessing. This computer program is written in double precision structured FORTRAN77 and at present it is used in DOS environment on a PC386. There is a plan to port it to ND supermini computers in future. (author). 5 refs., 3 figs.

  18. The application of integrated safety management principles to the Tritium Extraction Facility project

    International Nuclear Information System (INIS)

    Hickman, M.O.; Viviano, R.R.

    2000-01-01

    The DOE has developed a program that is accomplishing a heightened safety posture across the complex. The Integrated Safety Management (ISM) System (ISMS) program utilizes five core functions and seven guiding principles as the basis for implementation. The core functions define the work scope, analyze the hazards, develop and implement hazard controls, perform the work, and provide feedback for improvement. The guiding principles include line management responsibility, clear roles and responsibilities, competence per responsibilities, identification of safety standards/requirements, tailored hazard control, balanced priorities, and operations authorization. There exists an unspecified eighth principle, that is, worker involvement. A program requiring the direct involvement of the employees who are actually performing the work has been shown to be quite an effective method of communicating safety requirements, controlling work in a safe manner, and reducing safety violations and injuries. The Tritium Extraction Facility (TEF) projects, a component of the DOE's Commercial Light Water Reactor Tritium Production program, has taken the ISM principles and core functions and applied them to the project's design. The task of the design team is to design a facility and systems that will meet the production requirements of the DOE tritium mission as well as a design that minimizes the workers' exposure to adverse safety situations and hazards/hazardous materials. During the development of the preliminary design for the TEF, design teams consisted of not only designers but also personnel who had operational experience in the existing tritium and personnel who had operational experience in the existing tritium and personnel who had specialized experience from across the DOE complex. This design team reviewed multiple documents associated with the TEF operation in order to identify and document the hazards associated with the tritium process. These documents include hazards

  19. Chemical reactor for a PUREX reprocessing plant of 200Kg U/day capacity

    International Nuclear Information System (INIS)

    Oliveria Lopes, M.J. de.

    1974-03-01

    Dissolution of spent reactor fuels in Purex process is studied. Design of a chemical reactor for PWR elements, 3% enriched uranium dioxide with zircaloy cladding, for a 200Kg/day uranium plant is the main objective. Chop-leach process is employed and 7.5M nitric acid is used. Non-criticality was obtained by safe geometry and checked by spectrum homogeneous calculus and diffusion codes. Fuel cycle is considered and decladding and dissolution are treated more accurately

  20. RAMI modeling of plant systems for proposed tritium production and extraction facilities

    International Nuclear Information System (INIS)

    Blanchard, A.

    2000-01-01

    The control of life-cycle cost is a primary concern during the development, construction, operation, and decommissioning of DOE systems and facilities. An effective tool that can be used to control these costs, beginning with the design stage, is called a reliability, availability, maintainability, and inspectability analysis or, simply, RAMI for short. In 1997, RAMI technology was introduced to the Savannah River Site with applications at the conceptual design stage beginning with the Accelerator Production of Tritium (APT) Project and later extended to the Commercial Light Water Reactor (CLWR) Tritium Extraction Facility (TEF) Project. More recently it has been applied to the as-build Water Treatment Facilities designed for ground water environmental restoration. This new technology and database was applied to the assessment of balance-of-plant systems for the APT Conceptual Design Report. Initial results from the Heat Removal System Assessment revealed that the system conceptual design would cause the APT to fall short of its annual production goal. Using RAM technology to immediately assess this situation, it was demonstrated that the product loss could be gained back by upgrading the system's chiller unit capacity at a cost of less than $1.3 million. The reclaimed production is worth approximately $100 million. The RAM technology has now been extended to assess the conceptual design for the CLWR-TEF Project. More specifically, this technology and database is being used to translate high level availability goals into lower level system design requirements that will ensure the TEF meets its production goal. Results, from the limited number of system assessments performed to date, have already been used to modify the conceptual design for a remote handling system, improving its availability to the point that a redundant system, with its associated costs of installation and operation may no longer be required. RAMI results were also used to justify the elimination

  1. Demonstration of Minor Actinide separation from a genuine PUREX raffinate by TODGA/TBP and SANEX reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Magnusson, D. [European Commission, Joint Research Center, Institute for Transuranium Elements, Postfach 2340 D-76125 Karlsruhe (Germany); Chalmers University of Technology, Nuclear Chemistry, Deparment of Chemical and Biological Engineering, Gothenburg (Sweden); Christiansen, B.; Glatz, J.P.; Malmbeck, R.; Serrano-Purroy, D. [European Commission, Joint Research Center, Institute for Transuranium Elements, Postfach 2340 D-76125 Karlsruhe (Germany); Modolo, G. [Forschungszentrum Juelich, Institute for Energy Research, Safety Research and Reactor Technology, D-52425 Juelich (Germany); Sorel, C. [Commissariat a l' Energie Atomique Valrho (CEA), DRCP/SCPS, BP17171, 30207 Bagnols-sur-Ceze (France)

    2008-07-01

    A genuine High Active Raffinate was produced from small scale Purex reprocessing of a UO{sub 2} spent fuel solution and used as feed for a subsequent TODGA/TBP process. In this process, efficient recovery of the trivalent Minor Actinides (MA) actinides could be demonstrated using a hot cell set-up of 32 centrifugal contactor stages. The feed decontamination factors obtained for Am and Cm were in the range of 4.10{sup 4} which corresponds to a recovery of more than 99.99 % in the product fraction. Trivalent lanthanides and Y were co-extracted, otherwise only a small part of the Ru ended up in the product. The collected actinide/lanthanide fraction was later used as feed for a Sanex (separation of actinides from lanthanides) process based on the CyMe{sub 4}-BTBP ligand. Preliminary results show recoveries of more than 99.9 % of Am, Cm and less than 0.1 % of the major lanthanides in the product. (authors)

  2. Application of biomass for the sorption of radionuclides from low level Purex aqueous wastes

    Energy Technology Data Exchange (ETDEWEB)

    Ramanujam, A; Gopalakrishnan, V; Dhami, P S; Kannan, R [Fuel Reprocessing Div., Bhabha Atomic Research Centre, Mumbai (India); Udupa, S R; Salvi, N A [Bio-Organic Div., Bhabha Atomic Research Centre, Mumbai (India)

    1997-05-01

    Microbial biomass have been found to be good biological adsorbents for radioactive nuclides such as uranium and thorium with comparatively easy desorption and recovery. Based on this, sorption studies have been carried out to assess the feasibility of using biomass Rhizopus arrhizus (RA) for the removal of radionuclides present in Purex low level waste streams. Biomass Rhizopus arrhizus (RA) appears effective for the removal of actinides and fission products from low level Purex plant waste/effluent solutions. Maximum sorption for uranium and plutonium is observed at 6-7 pH whereas for Am, Eu, Pm, Ce and Zr the sorption is maximum at pH 2 with high D values and fast kinetics in both cases. Sorption for Ru and Cs are negligible. Sorbed nuclides are recoverable by elution with 1 M HNO{sub 3}, on once through basis. The method can be used for treating the evaporator condensates from the plant and the hold-up delay tank solution. The sodium nitrate salt concentration in the aqueous solution beyond 0.14 M seriously affects the metal uptake. The results from column experiments indicate a limited loading capacity in terms of mg of Am/U/Pu etc. per gm of RA. However, as the Purex low level effluents contain only trace level activities whose absolute ionic concentrations are much lower, the capacities observed with the present form of biomass may still be satisfactory. 15 refs., 12 tabs.

  3. Application of biomass for the sorption of radionuclides from low level Purex aqueous wastes

    International Nuclear Information System (INIS)

    Ramanujam, A.; Gopalakrishnan, V.; Dhami, P.S.; Kannan, R.; Udupa, S.R.; Salvi, N.A.

    1997-05-01

    Microbial biomass have been found to be good biological adsorbents for radioactive nuclides such as uranium and thorium with comparatively easy desorption and recovery. Based on this, sorption studies have been carried out to assess the feasibility of using biomass Rhizopus arrhizus (RA) for the removal of radionuclides present in Purex low level waste streams. Biomass Rhizopus arrhizus (RA) appears effective for the removal of actinides and fission products from low level Purex plant waste/effluent solutions. Maximum sorption for uranium and plutonium is observed at 6-7 pH whereas for Am, Eu, Pm, Ce and Zr the sorption is maximum at pH 2 with high D values and fast kinetics in both cases. Sorption for Ru and Cs are negligible. Sorbed nuclides are recoverable by elution with 1 M HNO 3 , on once through basis. The method can be used for treating the evaporator condensates from the plant and the hold-up delay tank solution. The sodium nitrate salt concentration in the aqueous solution beyond 0.14 M seriously affects the metal uptake. The results from column experiments indicate a limited loading capacity in terms of mg of Am/U/Pu etc. per gm of RA. However, as the Purex low level effluents contain only trace level activities whose absolute ionic concentrations are much lower, the capacities observed with the present form of biomass may still be satisfactory

  4. Experimental studies and tests on An(III)/Ln(III) separation using the TODGA extractant

    Energy Technology Data Exchange (ETDEWEB)

    Heres, Xavier; Sorel, Christian; Miguirditchian, Manuel; Cames, Beatrice; Hill, Clement; Bisel, Isabelle; Espinoux, Denis; Viallesoubranne, Carole; Baron, Pascal; Lorrain, Brigitte [CEA/DEN/MAR/DRCP, Marcoule, BP17171, 30207 Bagnols/Ceze (France)

    2009-06-15

    Minor actinide recycling by separation and transmutation is worldwide considered as one of the most promising strategies to reduce the inventory of radioactive waste, thus contributing to make nuclear energy more sustainable. One of the different options investigated at the CEA Marcoule and within the ACSEPT project (a European collaborative project partly funded by the 7. EURATOM Framework Program) to separate trivalent minor actinide (Am(III)-Cf(III)) from the fission and activation products contained in PUREX raffinates is the TODGA process, which consists in: 1. Co-extracting trivalent 4f and 5f elements from highly acidic PUREX raffinates by a mixture of TODGA (tetraoctyl-diglycolamide) and TBP (tributyl-phosphate), dissolved in HTP (hydrogenated tetra-propene). 2. Selectively stripping the trivalent minor actinides by a hydrophilic poly-aminocarboxylic acid used as a complexing agent in a buffered aqueous solution, while the trivalent lanthanides are kept in the organic solvent thanks to a sodium nitrate salting-out effect. 3. Stripping the lanthanides in a diluted nitric acid solution. The major difficulty of this TODGA separation process is to tune the pH in a very narrow range of operating conditions in the second step, because of the high sensitivity of the performances of the flow-sheet vs pH. This difficulty was however overcome. This paper describes the development of the TODGA process from experimental studies to hot test implementation in shielded cells of the ATALANTE facility, including (i) the optimization of the extraction system (both the formulation of the organic solvent and those of the aqueous scrubbing and stripping solutions), (ii) the implementation of a cold test in small scale mixer-settlers in the G1 facility (MARCEL loop), using a surrogate feed composed of major fission products, (iii) the validation of some steps of the process, using a surrogate feed, spiked with Am-241 and Eu-152, and similar laboratory contactors (medium activity

  5. Facile synthesis of high strength hot-water wood extract films with oxygen-barrier performance

    Science.gov (United States)

    Chen, Ge-Gu; Fu, Gen-Que; Wang, Xiao-Jun; Gong, Xiao-Dong; Niu, Ya-Shuai; Peng, Feng; Yao, Chun-Li; Sun, Run-Cang

    2017-01-01

    Biobased nanocomposite films for food packaging with high mechanical strength and good oxygen-barrier performance were developed using a hot-water wood extract (HWE). In this work, a facile approach to produce HWE/montmorillonite (MMT) based nanocomposite films with excellent physical properties is described. The focus of this study was to determine the effects of the MMT content on the structure and mechanical properties of nanocomposites and the effects of carboxymethyl cellulose (CMC) on the physical properties of the HWE-MMT films. The experimental results suggested that the intercalation of HWE and CMC in montmorillonite could produce compact, robust films with a nacre-like structure and multifunctional characteristics. This results of this study showed that the mechanical properties of the film designated FCMC0.05 (91.5 MPa) were dramatically enhanced because the proportion of HWE, MMT and CMC was 1:1.5:0.05. In addition, the optimized films exhibited an oxygen permeability below 2.0 cm3 μm/day·m2·kPa, as well as good thermal stability due to the small amount of CMC. These results provide a comprehensive understanding for further development of high-performance nanocomposites which are based on natural polymers (HWE) and assembled layered clays (MMT). These films offer great potential in the field of sustainable packaging.

  6. Pretreatment of Hanford PUREX Plant first-cycle waste

    International Nuclear Information System (INIS)

    Gibson, M.W.; Gerboth, D.M.; Peters, B.B.

    1987-04-01

    A process has been developed to pretreat neutralized, first-cycle high-level waste from the fuels reprocessing facility at the Hanford Site. The process separates solids from the supernate liquid, which contains soluble salts. The solids, including most of the fission products and transuranic elements, may then be vitrified for disposal, while the low-level supernate stream may be processed into a less expensive grout waste form. The process also includes ion exchange treatment of the separated supernate stream to remove radiocesium. A flow sheet based on these operations was completed to support a planned demonstration of the process in the Hanford Site B Plant canyon facility. 5 refs., 2 figs., 5 tabs

  7. Future proton and mixed-field irradiation facilities with slow extraction for LHC operation phase and for LHC upgrades

    CERN Document Server

    Assmann, Ralph Wolfgang; Brugger, Markus; Efthymiopoulos, Ilias; Feldbaumer, Eduard; Garrido, Mar Capeans; Glaser, Maurice; Kramer, Daniel; Linssen, Lucie; Losito, Roberto; Moll, Michael; Rembser, Christoph; Silari, Marco; Thurel, Yves; Tsesmelis, Emmanuel; Vincke, Helmut; CERN. Geneva. The LHC experiments Committee; LHCC

    2010-01-01

    In the present proposal we present the need for improved proton and mixed-field irradiation facilities with slow beam extraction at CERN. Strong needs are expressed by both the detector and accelerator communities and concern the LHC operation era as well as the upgrades of machine and experiments. The current facilities and test areas have a number of limitations and drawbacks. Preliminary studies indicate that there are possibilities for a coherent and cost-effective approach towards improved facilities for the future. The aim of this document is to inform the LHCC and seek its recognition for the need of such facilities. In addition we would appreciate the support of the LHCC for pursuing further implementation studies at a PS East Hall location.

  8. A Study on the Ion Beam Extraction using Duo-PiGatron Ion source for Vertical Type Ion Beam Facility

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Bom Sok; Lee, Chan young; Lee, Jae Sang [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    In Korea Multipurpose Accelerator Complex (KOMAC), we have started ion beam service in the new beam utilization building since March this year. For various ion beam irradiation services, we are developed implanters such as metal (150keV/1mA), gaseous (200keV/5mA) and high current ion beam facility (20keV/150mA). One of the new one is a vertical type ion beam facility without acceleration tube (60keV/20mA) which is easy to install the sample. After the installation is complete, it is where you are studying the optimal ion beam extraction process. Detailed experimental results will be presented. Vertical Type Ion Beam Facility without acceleration tube of 60keV 20mA class was installed. We successfully extracted 60keV 20mA using Duo- PiGatron Ion source for Vertical Type Ion Beam Facility. Use the BPM and Faraday-cup, is being studied the optimum conditions of ion beam extraction.

  9. Management of Purex spent solvents by the alkaline hydrolysis process

    International Nuclear Information System (INIS)

    Srinivas, C.; Manohar, Smitha; Vincent, Tessy; Wattal, P.K.; Theyyunni, T.K.

    1995-01-01

    Various treatment processes were evaluated on a laboratory scale for the management of the spent solvent from the extraction of nuclear materials. Based on the lab scale evaluation it is proposed to adopt the alkaline hydrolysis process as the treatment mode for the spent solvent. The process has advantages over the other processes in terms of simplicity, low cost and ease of disposal of the secondary waste generated. (author)

  10. Neptunium control in co-decontamination step of purex process

    International Nuclear Information System (INIS)

    Zhang Zefu; He Jianyu; Zhu Zhaowu; Ye Guoan; Zhao Zhiqiang

    2002-01-01

    A new alternative method for separation of Np in the first co-decontamination step is proposed. It comprises two steps, namely, preconditioning of Np valence state in the dissolved solution of spent fuel by NO gas bubbling in HNO 3 medium to produce HNO 2 , which is considered as salt-free process to convert Np(VI) to Np(V) and stabilization of Np(V) with urea, finally, the demonstrative counter current cascade extraction of Np(IV) and Np(V) in a miniature mixer-settler was carried out. The batch experiments show that Np(V) produced after conditioning may be slowly oxidized again to Np(VI) during standing time. Addition of urea in the HNO 3 solution might enhance the stability of Np(V). On the other hand, the solvent extraction by 30% TBP/kerosene could greatly accelerate the oxidation rate of Np(V). The chemical flow sheet study at 25degC shows that, more than 98% of Np could be routed into HLLW if urea is added in the HNO 3 solution. The operating temperature has great influence on the kinetics of Np(V) oxidation. If operation temperature races to 36degC and urea is not added, about 38% of Np will go along with U and Pu into organic phase. The behavior of Np(IV) during extraction shows great accumulation in the middle stages of battery. (author)

  11. Facile Separation of 5-O-Galloylquinic Acid from Chinese Green Tea Extract using Mesoporous Zirconium Phosphate.

    Science.gov (United States)

    Ma, Yilong; Shang, Yafang; Zhu, Danye; Wang, Caihong; Zhong, Zhifeng; Xu, Ziyang

    2016-05-01

    5-O-Galloylquinic acid from green tea and other plants is attracting increasing attention for its antioxidant and antileishmanial bioactivities. It is always isolated using a silica column, a Sephadex column and high-performance liquid chromatography (HPLC) methods, which are either laborious or instrument dependent. To develop a new method to easily separate 5-O-galloylquinic acid. Mesoporous zirconium phosphate (m-ZrP) was prepared to conveniently separate 5-O-galloylquinic acid from Chinese green tea extract, and the target compound was easily obtained by simple steps of adsorption, washing and desorption. The effects of the green tea extraction conditions, extract concentrations, and m-ZrP adsorption/desorption dynamics on the 5-O-galloylquinic acid separation were evaluated. 5-O-Galloylquinic acid that was separated from a 70% ethanol extract of green tea was of moderate HPLC purity (92%) and recovery (88%), and an increased non-specific binding of epigallocatechin gallate (EGCG) on m-ZrP was observed in the diluted tea extract. The times for maximal adsorption of 5-O-galloylquinic acid in 70% ethanol extract and maximal desorption of 5-O-galloylquinic acid in 0.4% phosphoric acid solution were confirmed as 7 h and 5 h, respectively. A facile method to separate 5-O-galloylquinic acid from Chinese green tea extract using m-ZrP was established. Copyright © 2016 John Wiley & Sons, Ltd. Copyright © 2016 John Wiley & Sons, Ltd.

  12. Uranous nitrate production for purex process applications using PtO2 catalyst and H2/H2-gas mixtures

    International Nuclear Information System (INIS)

    Sreenivasa Rao, K.; Shyamali, R.; Narayan, C.V.; Patil, A.R.; Jambunathan, U.; Ramanujam, A.; Kansara, V.P.

    2003-04-01

    In the Purex process of spent fuel reprocessing. the twin objectives- decontamination and partitioning are achieved by extracting uranium (VI) and plutonium (IV) together in the solvent 30% TBP-dodecane and then selectively reducing Pu (IV) to Pu (III) in which valency it is least extractable in the solvent. Uranous nitrate stabilized with hydrazine nitrate is the widely employed partitioning agent. The conventional method of producing U(IV) is by the electrolytic reduction of uranyl nitrate with hydrazine nitrate as uranous ion stabilizer. Tre percentage conversion of U(VI) to U(IV) obtained in this method is 50 -60 %. The use of this solution as partitioning agent leads not only to the dilution of the plutonium product but also to increase in uranium processing load by each externally fed uranous nitrate batch. Also the oxide coating of the anode, TSIA (Titanium Substrate Insoluble Anode) wears out after a certain period of operation. This necessitates recoating which is quite cumbersome considering the amount of the decontamination involved. An alternative to the conventional electrolytic method of reduction of uranyl nitrate to uranous nitrate was explored at FRD laboratory .The studies have revealed that near 100% uranous nitrate can be produced by reducing uranyl nitrate with H 2 gas or H 2 (8%)- Ar/N 2 gas mixture in presence of PtO 2 catalyst. This report describes the laboratory scale studies carried out to optimize the various parameters. Based on these studies reduction of uranyl nitrate on a pilot plant scale was carried out. The design and operation of the reductor column and also the various studies carried out in the pilot plant studies are discussed. Near 100% conversion of uranyl nitrate to uranous nitrate and also the redundancy of supply of electrical energy make this process a viable alternative to the existing electrolytic method. (author)

  13. Separation of 90Sr from Purex high level waste and development of a 90Sr-90Y generator

    International Nuclear Information System (INIS)

    Ramanujam, A.; Dhami, P.S.; Chitnis, R.R.; Achuthan, P.V.; Kannan, R.; Gopalakrishnan, V.; Balu, K.

    2000-04-01

    90 Y (T 1/2 =64.2 h) finds several applications in nuclear medicine. It is formed from the decay of 90 Sr which has a long half-life of 28.8 years. 90 Sr can be used as a long-lasting source for the production of carrier-free 90 Y. 90 Sr itself is abundantly available in high level waste (HLW) of PUREX origin. The present studies deal with the separation of pure 90 Sr from HLW and the subsequent separation of 90 Y from 90 Sr. Actinides and some of the fission products like lanthanides, zirconium, molybdenum and cesium were first removed from the HLW using methods based on solvent extraction and ion-exchange studied earlier in our laboratory. The resulting waste solution was used as a feed for the present process. The separation of 90 Sr from HLW was based on radiochemical method which involved a repeated scavenging with ferric hydroxide followed by strontium carbonate precipitation. The separation of 90 Y from 90 Sr was achieved by membrane separation technique. A compact generator is developed for this separation using a commercially available polytetrafluoroethylene (PTFE) membrane, impregnated with indigenously synthesised 2-ethylhexyl 2-ethylhexyl phosphonic acid (KSM-17). Generator system overcomes the drawbacks associated with conventional solvent extraction and ion-exchange based generators. The product is in chloride form and is suitable for complexation studies. After gaining an operating experience of ∼3 years in generating carrier-free 90 Y at 2 mCi level for initial studies in radiotherapeutic applications, the process was scaled up for the production of about 12 mCi of 90 Y to be used for animal studies before its application to patients. Radiochemical and chemical purity of the product was critically assayed by radiometry, ICP-AES, etc. The process is amenable for further scaling up. (author)

  14. Innovative SANEX process for trivalent actinides separation from PUREX raffinate

    International Nuclear Information System (INIS)

    Sypula, Michal

    2013-01-01

    Recycling of nuclear spent fuel and reduction of its radiotoxicity by separation of long-lived radionuclides would definitely help to close the nuclear fuel cycle ensuring sustainability of the nuclear energy. Partitioning of the main radiotoxicity contributors followed by their conversion into short-lived radioisotopes is known as partitioning and transmutation strategy. To ensure efficient transmutation of the separated elements (minor actinides) the content of lanthanides in the irradiation targets has to be minimised. This objective can be attained by solvent extraction using highly selective ligands that are able to separate these two groups of elements from each other. The objective of this study was to develop a novel process allowing co-separation of minor actinides and lanthanides from a high active acidic feed solution with subsequent actinide recovery using just one cycle, so-called innovative SANEX process. The conditions of each step of the process were optimised to ensure high actinide separation efficiency. Additionally, screening tests of several novel lipophilic and hydrophilic ligands provided by University of Twente were performed. These tests were aiming in better understanding the influence of the extractant structural modifications onto An(III)/Ln(III) selectivity and complexation properties. Optimal conditions for minor actinides separation were found and a flow-sheet of a new innovative SANEX process was proposed. Tests using a single centrifugal contactor confirmed high Eu(III)/Am(III) separation factor of 15 while the lowest SF Ln/Am obtained was 6,5 (for neodymium). In addition, a new masking agent for zirconium was found as a substitution for oxalic acid. This new masking agent (CDTA) was also able to mask palladium without any negative influence on An(III)/Ln(III). Additional tests showed no influence of CDTA on plutonium present in the feed solution unlike oxalic acid which causes Pu precipitation. Therefore, CDTA was proposed as a Zr

  15. Innovative SANEX process for trivalent actinides separation from PUREX raffinate

    Energy Technology Data Exchange (ETDEWEB)

    Sypula, Michal

    2013-07-01

    Recycling of nuclear spent fuel and reduction of its radiotoxicity by separation of long-lived radionuclides would definitely help to close the nuclear fuel cycle ensuring sustainability of the nuclear energy. Partitioning of the main radiotoxicity contributors followed by their conversion into short-lived radioisotopes is known as partitioning and transmutation strategy. To ensure efficient transmutation of the separated elements (minor actinides) the content of lanthanides in the irradiation targets has to be minimised. This objective can be attained by solvent extraction using highly selective ligands that are able to separate these two groups of elements from each other. The objective of this study was to develop a novel process allowing co-separation of minor actinides and lanthanides from a high active acidic feed solution with subsequent actinide recovery using just one cycle, so-called innovative SANEX process. The conditions of each step of the process were optimised to ensure high actinide separation efficiency. Additionally, screening tests of several novel lipophilic and hydrophilic ligands provided by University of Twente were performed. These tests were aiming in better understanding the influence of the extractant structural modifications onto An(III)/Ln(III) selectivity and complexation properties. Optimal conditions for minor actinides separation were found and a flow-sheet of a new innovative SANEX process was proposed. Tests using a single centrifugal contactor confirmed high Eu(III)/Am(III) separation factor of 15 while the lowest SF{sub Ln/Am} obtained was 6,5 (for neodymium). In addition, a new masking agent for zirconium was found as a substitution for oxalic acid. This new masking agent (CDTA) was also able to mask palladium without any negative influence on An(III)/Ln(III). Additional tests showed no influence of CDTA on plutonium present in the feed solution unlike oxalic acid which causes Pu precipitation. Therefore, CDTA was proposed as

  16. Transfer of Plutonium-Uranium Extraction Plant and N Reactor irradiated fuel for storage at the 105-KE and 105-KW fuel storage basins, Hanford Site, Richland Washington

    International Nuclear Information System (INIS)

    1995-07-01

    The U.S. Department of Energy (DOE) needs to remove irradiated fuel from the Plutonium-Uranium Extraction (PUREX) Plant and N Reactor at the Hanford Site, Richland, Washington, to stabilize the facilities in preparation for decontamination and decommissioning (D ampersand D) and to reduce the cost of maintaining the facilities prior to D ampersand D. DOE is proposing to transfer approximately 3.9 metric tons (4.3 short tons) of unprocessed irradiated fuel, by rail, from the PUREX Plant in the 200 East Area and the 105 N Reactor (N Reactor) fuel storage basin in the 100 N Area, to the 105-KE and 105-KW fuel storage basins (K Basins) in the 100 K Area. The fuel would be placed in storage at the K Basins, along with fuel presently stored, and would be dispositioned in the same manner as the other existing irradiated fuel inventory stored in the K Basins. The fuel transfer to the K Basins would consolidate storage of fuels irradiated at N Reactor and the Single Pass Reactors. Approximately 2.9 metric tons (3.2 short tons) of single-pass production reactor, aluminum clad (AC) irradiated fuel in four fuel baskets have been placed into four overpack buckets and stored in the PUREX Plant canyon storage basin to await shipment. In addition, about 0.5 metric tons (0.6 short tons) of zircaloy clad (ZC) and a few AC irradiated fuel elements have been recovered from the PUREX dissolver cell floors, placed in wet fuel canisters, and stored on the canyon deck. A small quantity of ZC fuel, in the form of fuel fragments and chips, is suspected to be in the sludge at the bottom of N Reactor's fuel storage basin. As part of the required stabilization activities at N Reactor, this sludge would be removed from the basin and any identifiable pieces of fuel elements would be recovered, placed in open canisters, and stored in lead lined casks in the storage basin to await shipment. A maximum of 0.5 metric tons (0.6 short tons) of fuel pieces is expected to be recovered

  17. Construction and operation of a tritium extraction facility at the Savannah River Site. Final environmental impact statement

    International Nuclear Information System (INIS)

    1999-03-01

    DOE proposes to construct and operate a Tritium Extraction Facility (TEF) at H Area on the Savannah River Site (SRS) to provide the capability to extract tritium from commercial light water reactor (CLWR) targets and from targets of similar design. The proposed action is also DOE's preferred alternative. An action alternative is to construct and operate TEF at the Allied General Nuclear Services facility, which is adjacent to the eastern side of the SRS. Under the no-action alternative DOE could incorporate tritium extraction capabilities in the accelerator for production of tritium. This EIS is linked to the Final Programmatic Environmental Impact Statement for Tritium Supply and Recycling, from which DOE determined that it would produce tritium either in an accelerator or in a commercial light water reactor. The purpose of the proposed action and alternatives evaluated in this EIS is to provide tritium extraction capability to support either tritium production technology. The EIS assesses the environmental impacts from the proposed action and the alternatives, including the no action alternative

  18. Uranium/plutonium and uranium/neptunium separation by the Purex process using hydroxyurea

    International Nuclear Information System (INIS)

    Zhu Zhaowu; He Jianyu; Zhang Zefu; Zhang Yu; Zhu Jianmin; Zhen Weifang

    2004-01-01

    Hydroxyurea dissolved in nitric acid can strip plutonium and neptunium from tri-butyl phosphate efficiently and has little influence on the uranium distribution between the two phases. Simulating the 1B contactor of the Purex process by hydroxyurea with nitric acid solution as a stripping agent, the separation factors of uranium/plutonium and uranium/neptunium can reach values as high as 4.7 x 10 4 and 260, respectively. This indicates that hydroxyurea is a promising salt free agent for uranium/plutonium and uranium/neptunium separations. (author)

  19. Destruction of nitric acid in purex process streams by formaldehyde treatment

    International Nuclear Information System (INIS)

    Kumar, S.V.; Nadkarni, M.N.; Mayankutty, P.C.; Pillai, N.S.; Shinde, S.S.

    1974-01-01

    Efficiency of destruction of nitric acid in purex process streams with formaldehyde has been studied as a function of initial acidity, uranium concentration, rate of addition of formaldehyde and temperature in the range 6 - 0.5M acid. Guidelines are suggested for the accurate calculations of the volume of formaldehyde needed to effect the required change of acidity at 100degC. Sodium nitrite has been established as a 'key' to initiate the reaction and water as an effective scrubber for collecting the acid fumes emanating from the reaction vessel. (author)

  20. Uranium removal from organic solutions of PUREX process

    International Nuclear Information System (INIS)

    Dell'Occhio, L.A.; Dupetit, G.A.; Pascale, A.A.; Vicens, H.E.

    1987-01-01

    During the uranium extraction process with tributyl phosphate (TBP) in nitric medium, a bi solvated, non hydrated complex is formed, of formula UO2(NO3)2TBP, which is soluble in the diluent, a paraffin hydrocarbon. As it is known that some uranium salts, for instance the nitrate, when dissolved in organic solvents, like isopropanol, can be discharged as complex molecules at the cathode of an electrodeposition cell, it was decided to apply this technique to uranium loaded TBP solutions. From preliminary experiments resulted a practical possibility for the analytical control through the alpha measurement of electro deposits. This technique could be applied as well to the treatment of depleted organic streams carrying undesirable alpha activity, because the so treated solutions become deprived of uranium. This work presents the curves obtained working at constant voltage with uranium-loaded TBP solutions, the determination of the optimal operation voltage in these conditions, the electrodeposition yield for electro polished copper and stainless steel cathodes and the tests of reproducibility of deposits. A summary of the results obtained operating the high voltage supply at constant power is also presented. (Author)

  1. Stability and modification of passive films of new PUREX-materials

    International Nuclear Information System (INIS)

    Schultze, J.W.; Siemensmeyer, B.; Patzelt, T.

    1991-10-01

    The valve metals Ti, Zr and others and their alloys can be used in nitric acid solutions of the Purex process. They are protected by passive films which are stable at least at low temperatures and concentrations. Electrochemical investigations and corrosion tests are applied to check improvements of the materials. Niobium can be used to substitute the very expensive tantalum. Electrochemical and analytical investigations show the formation of the corrosion stable oxide film. Special problems are treated, such as the stability of welded joints or the influence of radioactive irradiation. α-radiation and hot atoms are simulated by ion implantation, β- and γ-radiation are simulated by laser light. In both types of experiments no decrease of stability is indicated. The alloy Ti5Ta is more stable than Ti, but it is not as good as Ta. Other alloys of Ti were investigated, but they are not suitable for the Purex process. New protection layers are tested. With respect to their preparation as well as their corrosion stability, ANOF-films are promising, but TiN-films are not stable enough. (orig.) With 71 refs., 7 tabs., 71 figs [de

  2. Vegetable-origin foam employed in dye extraction in tanning and leather processing facilities

    Directory of Open Access Journals (Sweden)

    José M. Cangemi

    2009-01-01

    Full Text Available This study addressed the use of conventional and vegetable origin polyurethane foams to extract C. I. Acid Orange 61 dye. The quantitative determination of the residual dye was carried out with an UV/Vis absorption spectrophotometer. The extraction of the dye was found to depend on various factors such as pH of the solution, foam cell structure, contact time and dye and foam interactions. After 45 days, better results were obtained for conventional foam when compared to vegetable foam. Despite presenting a lower percentage of extraction, vegetable foam is advantageous as it is considered a polymer with biodegradable characteristics.

  3. Reprocessing of spent nuclear fuel, Annex 2: Chemical-technology study of the modified 'Purex' process Chemical and radiochemical control analyses; Prerada isluzenog nuklearnog goriva, Prilog 2: Hemijsko tehnolosko ispitivanje modifikovanog 'purex' procesa

    Energy Technology Data Exchange (ETDEWEB)

    Gal, I [Institute of Nuclear Sciences Boris Kidric, Laboratorija za hemiju visoke aktivnosti, Vinca, Beograd (Serbia and Montenegro)

    1964-12-15

    The objective of this task was testing of the modified purex process in the constructed separation cell, and verification of the reliability and efficiency of the process. Extractors used were 1BX, 1BS and 1C. testing was done with syntetic solutions.

  4. Shielding calculation of slow extracted beam facility at KEK proton synchrotron

    International Nuclear Information System (INIS)

    Hirabayashi, Hiromi; Katoh, Kazuaki

    1978-01-01

    The KEK proton synchrotron has two external beam lines, i.e. a fast extracted beam line for a bubble chamber and a slow extracted beam line for counter experiments. The maximum total intensity of the slow beam is estimated as 5 x 10 12 protons per sec. For beam losses along the line, shielding calculation was made, and on the basis of these results, adequacy of the current shielding construction plans was discussed. (Mori, K.)

  5. Facile and eco-friendly fabrication of AgNPs coated silk for antibacterial and antioxidant textiles using honeysuckle extract.

    Science.gov (United States)

    Zhou, Yuyang; Tang, Ren-Cheng

    2018-01-01

    Recently, there is a growing trend towards the functionalization of silk through nanotechnology for the prevention of fiber damage from microbial attack and the enhancement of hygienic aspects. Considering sustainable development and environmental protection, the eco-friendly fabrication of silver nanoparticles (AgNPs)-modified silk using natural extracts has currently become a hot research area. This study presents a facile strategy for the fabrication of colorful and multifunctional silk fabric using biogenic AgNPs prepared by honeysuckle extract as natural reductant and stabilizing agents. The influences of pH and reactant concentrations on the AgNPs synthesis were investigated. The color characteristics and functionalities of AgNPs treated silk were evaluated. The results revealed that the particle size of AgNPs decreased with increasing pH. The diameter of AgNPs decreased with increasing amount of honeysuckle extract and reducing amount of silver nitrate. The transmission electron microscopy image showed that the AgNPs were spherical in shape with a narrow size distribution. The treated silk showed excellent antibacterial activities against E. coli and S. aureus, and certain antioxidant activity. Both of the antibacterial and antioxidant activities were well maintained even after 30 washing cycles. This work provides a sustainable and eco-friendly approach to the fabrication of AgNPs coated silk for colorful and long-term multifunctional textiles using honeysuckle extract. Copyright © 2017 Elsevier B.V. All rights reserved.

  6. 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

  7. Removal of radionuclides from radioactive effluents of Purex origin using biomass banana pith as sorbant

    International Nuclear Information System (INIS)

    Ramanujam, A.; Dhami, P.S.; Kannan, R.; Das, S.K.; Naik, P.W.; Gopalakrishnan, V.; Kansra, V.P.; Balu, K.

    1998-06-01

    Investigations have been carried out on the applicability of dried banana pith (inner stem) for the sorption of various radionuclides viz. U, Pu, 241 Am, 144 Ce, 147 Pm, 152+154 Eu and 137 Cs which are generally present at trace level in Purex process waste effluents. The sorption of trivalent radionuclides as well as tetravalent plutonium was found to be high at pH 2, whereas sorption of uranium was found to be maximum at pH 6. Cesium was not found to be sorbed. 241 Am sorption was investigated in detail as a representative element of trivalent actinides and fission products to study the general trend. Though its sorption was kinetically slow, near-quantitative sorption was observed on prolonged contact. 241 Am sorption was studied in presence of NaNO 3 (up to 1 M) and Nd(III) up to 500 mg/l. Whereas no significant change in distribution ratios (D) was observed in the presence of NaNO 3 , it increased with neodymium concentration in the range tested. This indicates the effectiveness of the biomass as sorbent even in presence of sodium salts. Sorbed metal ions could be recovered by leaching with 2 M nitric acid. The dried biomass samples prepared from different sources were found to be stable for months and gave similar results on testing. The biomass was tested for its applicability for sorbing radionuclides present in Purex evaporator condensate and diluted high level waste solution on once through basis. The sorption capacity of banana pith for trivalent actinide-lanthanide is in the range of 60 mg/g banana pith. The results indicate that the biomass can be used effectively for the treatment of Purex Waste effluents for the removal of strontium, tri- and tetravalent actinides and fission products. The biomass was also tested for the sorption of toxic metal ions viz. Sr, Hg, Pb, Cr, Cd, and As from a nitrate solution at pH 2 and 4. D values followed the order Hg>Sr>Cd>Pb at pH 2, with Cr and As showing no uptake. These results indicate the potential of this

  8. Subcontracting strategy for the decontamination and decommissioning of Savannah River Site's First Tritium Extraction Facility, 232-F

    International Nuclear Information System (INIS)

    Smith, C.W. Jr.; Dowd, A.S. Jr.; Hinds, S.S.; Johnson, S.V.

    1994-01-01

    The Savannah River Site (SRS) has been actively proceeding with the decontamination and decommissioning (D and D) of various facilities and structures which were instrumental in the success of past missions at the site. The most ambitious of these efforts involves the subcontracting of the complete D and D of the first SRS Tritium Extraction Facility, identified as building 232-F. This facility operated in the mid 1950's and discontinued operations permanently in 1958. The approach utilized for this effort attempts to invoke the novel principle of open-quotes As Commercial As Reasonably Achievableclose quotes or open-quotes ACARAclose quotes. This concept of ACARA applies only the minimum essential requirements necessary to successfully perform the D and D task. Integral to this approach is the subcontractor provision for maximum flexibility in the identification of and adherence to the requirements of applicable DOE Orders, federal, state and local laws and regulations, as well as site specific procedures without violating the site contractual requirements. The technical specification prepared for this effort provides the basis for a competitively bid contract to perform the entire D and D evolution, including initial facility characterization, waste stream characterization and certification, D and D and waste disposal. Preparation and development of this specification and the subsequent Request For Proposal (RFP) was a successful team oriented endeavor. The schedule for this fast-track undertaking took three months to complete. Successful initiation of this task will be the first D and D of a facility containing both radioactive and hazardous material at an operating site within the DOE Weapons Complex. The strategy for preparing the D and D subcontract for the 232-F structure was facilitated by applying the ACARA principle. This approach resulted in the accelerated development of the specification and RFP documents, as well as minimized the complexities of

  9. Theme 1: fuel cycle and waste management. 1.3 the nuclear fuel cycle in the future. 1.3.1. thermal recycle of plutonium ''Ongoing industrialization of Purex'

    International Nuclear Information System (INIS)

    Wakem, M.J.

    2001-01-01

    The Purex process has been developed over many years from a process supporting military programmes in the years 1940 with the emphasis on production of a single product to today sophisticated large scale commercial plants designed to separate Uranium and Plutonium as high quality products. The plants have been designed, and are operated so as to discharge minimal aerial and liquid effluents whilst at the same time minimising arisings of liquid and solid waste. The scope of the facilities includes treatment of such wastes to create a form that is suitable for interim storage prior to final disposal. Typical of such plants are Thorp at Sellafield and UP3 at Cap La Hague, where plutonium dioxide is separated for the production of Mixed Oxide Fuel (MOX). The paper demonstrates the practical application of improvements to the Purex process at an industrial scale with the constraints imposed by technical, regulatory and commercial requirements. Successful examples will be addressed which illustrate the logical progression from technical concept, strategic decision and option taking, front end engineering definition, design and initial safety approval, regulatory approval leading to effective plant implementation and proving. (author)

  10. Results Of The Extraction-Scrub-Strip Testing Using An Improved Solvent Formulation And Salt Waste Processing Facility Simulated Waste

    International Nuclear Information System (INIS)

    Peters, T.; Washington, A.; Fink, S.

    2012-01-01

    The Office of Waste Processing, within the Office of Technology Innovation and Development, is funding the development of an enhanced solvent - also known as the next generation solvent (NGS) - for deployment at the Savannah River Site to remove cesium from High Level Waste. The technical effort is a collaborative effort between Oak Ridge National Laboratory (ORNL) and Savannah River National Laboratory (SRNL). As part of the program, the Savannah River National Laboratory (SRNL) has performed a number of Extraction-Scrub-Strip (ESS) tests. These batch contact tests serve as first indicators of the cesium mass transfer solvent performance with actual or simulated waste. The test detailed in this report used simulated Tank 49H material, with the addition of extra potassium. The potassium was added at 1677 mg/L, the maximum projected (i.e., a worst case feed scenario) value for the Salt Waste Processing Facility (SWPF). The results of the test gave favorable results given that the potassium concentration was elevated (1677 mg/L compared to the current 513 mg/L). The cesium distribution value, DCs, for extraction was 57.1. As a comparison, a typical D Cs in an ESS test, using the baseline solvent formulation and the typical waste feed, is ∼15. The Modular Caustic Side Solvent Extraction Unit (MCU) uses the Caustic-Side Solvent Extraction (CSSX) process to remove cesium (Cs) from alkaline waste. This process involves the use of an organic extractant, BoBCalixC6, in an organic matrix to selectively remove cesium from the caustic waste. The organic solvent mixture flows counter-current to the caustic aqueous waste stream within centrifugal contactors. After extracting the cesium, the loaded solvent is stripped of cesium by contact with dilute nitric acid and the cesium concentrate is transferred to the Defense Waste Processing Facility (DWPF), while the organic solvent is cleaned and recycled for further use. The Salt Waste Processing Facility (SWPF), under

  11. Potentiometric determination of uranium in simulated Purex Process solutions by acidiometry

    International Nuclear Information System (INIS)

    Cohen, V.H.; Matsuda, H.T.; Araujo, B.F. de; Araujo, J.A. de

    1983-01-01

    A potentiometric methods for sequential free acidity and uranium determination in simulated Purex Process solutions is described. An oxalate solution or a mixture of fluoride-oxalate pellets were used as complexing agent for free titration. Following this first equivalent point, uranium is determined-by indirect titration of H + liberated in the peruanate reaction. Some elements present in the standard fuel elements with a burn-up of 33.000 Mwd/t, neutron flux of 3,2 x 10 13 n.cm -2 .s -1 and cooling time of two years were considered as interfering elements in uranium analyses. As a substitute of Pu-IV, Th(NO 3 ) 4 solution was used. The method can be applied to aqueous and organic (TBP/diluent) solutions with 2% precision and 2% accuracy. (Autor) [pt

  12. The study of reductive reextraction of plutonium in the Purex process

    International Nuclear Information System (INIS)

    Poczynajlo, A.

    1985-01-01

    The methods of separation of U and Pu in the Purex process and the thermodynamic and kinetic properties of Pu(4) reductants are discussed. The kinetic equation of the process of reductive reextraction of plutonium for the first order reaction with respect to Pu(4) is derived. The kinetics of plutonium reextraction with the use of uranium (4), ascorbic acid and other reductants has been studied. The necessity of application of the stoichiometric excess of reductant has been explained by simultaneously occured reoxidation process of plutonium. The method of calculation of the steady- state plutonium concentration profiles has been elaborated for counter-current separation of U and Pu in multistage contactor. 90 refs., 20 tabs., 29 figs. (author)

  13. Recent studies related to head-end fuel processing at the Hanford PUREX plant

    Energy Technology Data Exchange (ETDEWEB)

    Swanson, J.L.

    1988-08-01

    This report presents the results of studies addressing several problems in the head-end processing (decladding, metathesis, and core dissolution) of N Reactor fuel elements in the Hanford PUREX plant. These studies were conducted over 2 years: FY 1986 and FY 1987. The studies were divided into three major areas: 1) differences in head-end behavior of fuels having different histories, 2) suppression of /sup 106/Ru volatilization when the ammonia scrubber solution resulting from decladding is decontaminated by distillation prior to being discharged, and 3) suitability of flocculating agents for lowering the amount of transuranic (TRU) element-containing solids that accompany the decladding solution to waste. 16 refs., 43 figs.

  14. Potentiometric determination of uranium in simulated Purex Process solutions by acidiometry

    Energy Technology Data Exchange (ETDEWEB)

    Cohen, V H; Matsuda, H T; Araujo, B.F. de; Araujo, J.A. de

    1984-01-01

    A potentiometric methods for sequential free acidity and uranium determination in simulated Purex Process solutions is described. An oxalate solution or a mixture of fluoride-oxalate pellets were used as complexing agent for free titration. Following this first equivalent point, uranium is determined-by indirect titration of H/sup +/ liberated in the peruanate reaction. Some elements present in the standard fuel elements with a burn-up of 33.000 Mwd/t, neutron flux of 3,2 x 10/sup 13/n.cm/sup -2/.s/sup -1/ and cooling time of two years were considered as interfering elements in uranium analyses. As a substitute of Pu-IV, Th(NO/sub 3/)/sub 4/ solution was used. The method can be applied to aqueous and organic (TBP/diluent) solutions with 2% precision and 2% accuracy. (Autor).

  15. Alkaline hydrolysis process for treatment and disposal of Purex solvent waste

    International Nuclear Information System (INIS)

    Srinivas, C.; Venkatesh, K.A.; Wattal, P.K.; Theyyunni, T.K.; Kartha, P.K.S.; Tripathi, S.C.

    1994-01-01

    Treatment of spent Purex solvent (30% TBP-70% n-dodecane mixture) from reprocessing plants by alkaline hydrolysis process was investigated using inactive 30% TBP solvent as well as actual radioactive spent solvent. Complete conversion of TBP to water-soluble reaction products was achieved in 7 hours reaction time at 130 deg C using 50%(w/v) NaOH solution at NaOH to TBP mole ratio of 3:2. Addition of water to the product mixture resulted in the complete separation of diluent containing less than 2 and 8 Bg./ml. of α and β activity respectively. Silica gel and alumina were found effective for purification of the separated diluent. Aqueous phase containing most of the original radioactivity was found compatible with cement matrix for further conditioning and disposal. (author). 17 refs., 10 tabs., 1 fig

  16. 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

  17. Initiating events study of the first extraction cycle process in a model reprocessing plant

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Renze; Zhang, Jian Gang; Zhuang, Dajie; Feng, Zong Yang [China Institute for Radiation Protection, Taiyuan (China)

    2016-06-15

    Definition and grouping of initiating events (IEs) are important basics for probabilistic safety assessment (PSA). An IE in a spent fuel reprocessing plant (SFRP) is an event that probably leads to the release of dangerous material to jeopardize workers, public and environment. The main difference between SFRPs and nuclear power plants (NPPs) is that hazard materials spread diffusely in a SFRP and radioactive material is just one kind of hazard material. Since the research on IEs for NPPs is in-depth around the world, there are several general methods to identify IEs: reference of lists in existence, review of experience feedback, qualitative analysis method, and deductive analysis method. While failure mode and effect analysis (FMEA) is an important qualitative analysis method, master logic diagram (MLD) method is the deductive analysis method. IE identification in SFRPs should be consulted with the experience of NPPs, however the differences between SFRPs and NPPs should be considered seriously. The plutonium uranium reduction extraction (Purex) process is adopted by the technics in a model reprocessing plant. The first extraction cycle (FEC) is the pivotal process in the Purex process. Whether the FEC can function safely and steadily would directly influence the production process of the whole plant-production quality. Important facilities of the FEC are installed in the equipment cells (ECs). In this work, IEs in the FEC process were identified and categorized by FMEA and MLD two methods, based on the fact that ECs are containments in the plant. The results show that only two ECs in the FEC do not need to be concerned particularly with safety problems, and criticality, fire and red oil explosion are IEs which should be emphatically analyzed. The results are accordant with the references.

  18. Results from Commissioning of the Energy Extraction Facilities of the LHC Machine

    CERN Document Server

    Coelingh, G J; Mess, K H

    2008-01-01

    The risk of damage to the superconducting magnets, bus bars and current leads of the LHC machine in case of a resistive transition (quench) is being minimized by adequate protection. The protection is based on early quench detection, bypassing the quenching magnets by cold diodes, energy density dilution in the quenching magnets using heaters and, eventually, energy extraction. For two hundred and twenty-six LHC circuits (600 A and 13 kA) extraction of the stored magnetic energy to external dump resistors was required. All these systems are now installed in the machine and the final hardware commissioning has been undertaken. After a short description of the topology and definitive features, layouts and parameters of these systems the paper will focus on the results from their successful commissioning and an analysis of the system performance.

  19. Upgrading a 1944 plutonium-extraction plant to a modern decontamination facility

    International Nuclear Information System (INIS)

    Wills, C.E.; Millikin, R.M.; Cruz, E.A.

    1993-10-01

    The Hanford Site, located in south-central Washington State, is currently undergoing extensive modifications as its mission changes from defense material production to one of waste management and environmental restoration. Starting in World War II, Hanford's mission for over four decades was the production of plutonium for defense needs. With the removal of such defense requirements over the last several years, the Hanford Site has refocused its efforts on the issues of cleanup and safety. The T Plant Complex is the first of the existing facilities to begin conversion from the old mission to the new. This conversion process and associated problems are described

  20. Engineering study: Fast Flux Test Facility fuel reprocessing

    International Nuclear Information System (INIS)

    Beary, M.M.; Raab, G.J.; Reynolds, W.R. Jr.; Yoder, R.A.

    1974-01-01

    Several alternatives were studied for reprocessing FFTF fuels at Hanford. Alternative I would be to decontaminate and trim the fuel at T Plant and electrolytically dissolve the fuel at Purex. Alternative II would be to decontaminate and shear leach the fuels in a new facility near Purex. Alternative III would be to decontaminate and store fuel elements indefinitely at T Plant for subsequent offsite shipment. Alternative I, 8 to 10 M$ and 13 quarter-years; for Alternative II, 24 to 28 M$ and 20 quarter-years; for Alternative III, 3 to 4 M$ and 8 quarter-years. Unless there is considerable slippage in the FFTF shipping schedule, it would not be possible to build a new facility as described in Alternative II in time without building temporary storage facilities at T Plant, as described in Alternative III

  1. Analytical control of reducing agents on uranium/plutonium partitioning at purex process; Controle analitico dos agentes redutores na particao uranio/plutonio no processo purex

    Energy Technology Data Exchange (ETDEWEB)

    Araujo, Izilda da Cruz de

    1995-07-01

    Spectrophotometric methods for uranium (IV), hydrazine (N{sub 2}H{sub 4}) and its decomposition product hydrazoic acid(HN{sub 3}), and hydroxylamine (NH{sub 2} OH) determinations were developed aiming their applications for the process control of CELESTE I installation at IPEN/CNEN-SP. These compounds are normally present in the U/Pu partitioning phase of the spent nuclear treatment via PUREX process. The direct spectrophotometry was used for uranium (IV) analysis in nitric acid-hydrazine solutions based on the absorption measurement at 648 nm. The azomethine compound formed by reaction of hydrazine and p-dimethylamine benzaldehyde with maximum absorption at 457 nm was the basis for the specific analytical method for hydrazine determination. The hydrazoic acid analysis was performed indirectly by its conversion into ferric azide complex with maximum absorption at 465 nm. The hydroxylamine detection was accomplished based on its selective oxidation to nitrous acid which is easily analyzed by the reaction with Griess reagent. The resulted azocompound gas a maximum absorption at 520 nm. The sensibility of 1,4x10{sup -6}M for U(IV) with 0,8% of precision, 1,6x10{sup -6}M for hydrazine with 0,8% of precision, 2,3x10{sup -6}M hydrazoic acid with 0,9% of precision and 2,5x10{sup -6}M for hydroxylamine with 0,8% of precision were achieved. The interference studies have shown that each reducing agent can be determined in the presence of each other without any interference. Uranium(VI) and plutonium have also shown no interference in these analysis. The established methods were adapted to run inside glove-boxes by using an optical fiber colorimetry and applied to process control of the CELESTE I installation. The results pointed out that the methods are reliable and safety in order to provide just-in-time information about process conditions. (author)

  2. Leaching study of heavy and radioactive elements present in wastes discarded by a uranium extraction and processing facility

    International Nuclear Information System (INIS)

    Pihlak, A.; Lippmaa, E.; Maremaee, E.; Sirk, A; Uustalu, E.

    1995-08-01

    The present report provides a systematic leaching study of the waste depository at the Sillamaee metallurgical plant 'Silmet' (former uranium extraction and processing facility), its construction and environmental impact. The following data are presented: γ-activity data of the depository and two drill cores, chemical composition and physical properties of depository material and leaching waters, results of γ- and α-spectrometric studies, leaching (with demineralized and sea water) intensities of loparite and uranium ore processing waste components. Environmental danger presented by the Sillamaee waste dump to the Gulf of Finland and the surrounding environment in Estonia is mainly due to uranium leaching and the presence of a large array of chemically poisonous substances

  3. Fluorescent carbon nanodots facilely extracted from Coca Cola for temperature sensing

    Science.gov (United States)

    Li, Feiming; Chen, Qiaoling; Cai, Zhixiong; Lin, Fangyuan; Xu, Wei; Wang, Yiru; Chen, Xi

    2017-12-01

    A novel method for the fabrication of carbon nanodots (CDs) is introduced: extracting CDs from the well-known soft drink Coca Cola via dialysis. The obtained CDs are of good monodispersity with a narrow size distribution (average diameter of 3.0 nm), good biocompatibility, high solubility (about 180 mg ml-1) and stable fluorescence even at a high salt concentration. Furthermore, they are sensitive to the temperature change with a linear relationship between the fluorescence intensity and temperature from 5 °C-95 °C. The CDs have been applied in high stable temperature sensing. This protocol is quite simple, green, cost-effective and technologically simple, which might be used for a range of applications including sensing, catalysts, drug and gene delivery, and so on.

  4. EXTRACT

    DEFF Research Database (Denmark)

    Pafilis, Evangelos; Buttigieg, Pier Luigi; Ferrell, Barbra

    2016-01-01

    The microbial and molecular ecology research communities have made substantial progress on developing standards for annotating samples with environment metadata. However, sample manual annotation is a highly labor intensive process and requires familiarity with the terminologies used. We have the...... and text-mining-assisted curation revealed that EXTRACT speeds up annotation by 15-25% and helps curators to detect terms that would otherwise have been missed.Database URL: https://extract.hcmr.gr/......., organism, tissue and disease terms. The evaluators in the BioCreative V Interactive Annotation Task found the system to be intuitive, useful, well documented and sufficiently accurate to be helpful in spotting relevant text passages and extracting organism and environment terms. Comparison of fully manual...

  5. Occupational radiological aspects related to the presence of natural radionuclides at extraction and production petroleum facilities

    International Nuclear Information System (INIS)

    Matta, Luiz Ernesto Santos de Carvalho

    2001-01-01

    This work presents an evaluation of exposure to natural radiation, after samples collected at the Campos Basin region, incurred to workers involved in the actions of exploration and production of petroleum, E and P, may be submitted to. It also evaluates the national standards of controlling and registration of practices involving radioactive and nuclear material, leading to the conclusion that it is not possible to classify the practice of exploration and production of petroleum, following the controlling criteria proposed by the standards. An occupational radiological protection program is made available to be immediately applied to E and P facilities. This program intends that in spite of the high values of activity of mass unity obtained from the samples, it is possible to maintain the occupational doses to levels lower than those proposed by the Safety Series 115. Moreover, it adds that the workers must be considered as occupationally exposed and not as member of the public. A proposal of standard so that the regulatory authority may classify and control a practice of E and P petroleum is also done. (author)

  6. The reduction of Np(VI) and Np(V) by tit dihydroxyurea and its application to the U/Np separation in the PUREX process

    Energy Technology Data Exchange (ETDEWEB)

    Yan, T.H.; Zheng, W.F.; Zuo, C.; Xian, L.; Zhang, Y.; Bian, X.Y.; Li, R.X.; Di, Y. [Dept. of Radiochemistry, China Inst. of Atomic Energy, BJ (China)

    2010-07-01

    The reduction of Np(VI) and Np(V) by Dihydroxyurea (DHU) was studied by spectrophotometry. The results show that the reduction of Np(VI) to Np(V) by DHU is particularly fast. The apparent rate constant is 1.86s{sup -1} at 4 C as [HNO{sub 3}] = 0.44 M and [DHU] = 7.5 x 10{sup -2} M. While further reduction of Np(V) to Np(IV) is so slow that no Np(IV) is observed in 2 h. The reduction back-extraction behavior of Np(VI) in 30% tri-butyl phosphate/kerosene was firstly investigated under conditions of different temperature, different concentrations of DHU and HNO{sub 3} and various phase contact time, respectively. The results show that 98% of Np(VI) in the organic phase can be stripped rapidly to the aqueous phase by DHU under the given experimental conditions. As the concentration of HNO{sub 3} in the aqueous phase increases, the stripping efficiency decreases. While the stripping efficiency increases with the increase of the concentration of DHU. Simulating the 1B contactor of the PUREX process using DHU as the stripping agent, the SF{sub U}/Np equals to 183 under the given experimental conditions. It indicates that Np will follow with Pu in the U/Pu separation stage in the reprocessing of spent fuels. (orig.)

  7. Summary of the first step of active test of separation facility at Rokkasho Reprocessing Plant

    International Nuclear Information System (INIS)

    Iseki, T.; Takahashi, N.; Tanaka, Y.

    2006-01-01

    Full text: Full text: 1. Purpose: The aim of this presentation is to explain and discuss the first step of active test results in the Separation facility. We had previously performed uranium tests with depleted uranium from February of 2005 to January of 2006. Then, the first step of active test has been in progress since March of 2006 toward the start of commercial operation. 2. Outline of Separation facility: PUREX process is selected for the separation facility that has 6 pulsed columns (PC) and 6 mixer-settlers banks (MS). The dissolution solution is fed to extraction PC where U and Pu are extracted to solvent. The loaded solvent is washed with nitric acid in FP scrubbing PC and T/Tc scrubbing PC and then the aqueous raffinates from T/Tc scrubbing PC flows to complementary extraction MS to recover slightly back extracted U and Pu. The loaded solvent flows to Pu stripping PC where U remains in solvent and Pu is stripped. And U in Pu stream is washed with solvent. A Pu decontamination of U loaded solvent is carried out in Pu barrier MS, and U is back extracted from solvent with acidified water at higher temperature in U stripping MS. The outlet aqueous phases are washed with n-dodecane to remove the TBP. 3. Main active test contents: The active test is performed with the spent fuel elements. The main purpose of active test is (1) Checking of the separation and stripping performances, (2) Checking of uranium and plutonium losses into wastes, (3) Checking of diluent washing efficiency and (4) Checking of the throughout performance. 4. Results: The test results of the first step of active test are presented: (a) Plutonium stripping performance, (b) decontamination factor of fission products and (c) uranium and plutonium losses into wastes

  8. Solvent extraction in the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Eccles, H.; Naylor, A.

    1987-01-01

    Solvent extraction techniques have been used in the uranium nuclear fuel cycle in three main areas; concentration of uranium from ore leach liquor, purification of ore concentrates and fuel reprocessing. Solvent extraction has been extended to the removal of transuranic elements from active waste liquor, the recovery of uranium from natural sources and the recovery of noble metals from active waste liquor. Schemes are presented for solvent extraction of uranium using the Amex or Dapex process; spent fuel reprocessing and the Purex process. Recent and future developments of the techniques are outlined. (UK)

  9. 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)

  10. NMR characterization of segmental dynamics in poly(alkyl methacrylate) using CODEX and PUREX exchange techniques

    International Nuclear Information System (INIS)

    Becker-Guedes, Fabio; Azevedo, Eduardo R. de; Bonagamba, Tito J.; Schmidt-Rohr, Klaus

    2001-01-01

    Slow side group dynamics in a series of five poly(alkyl methacrylate)s with varying side group sizes (PMAA, PMMA, PEMA, PiBMA, and PcHMA, with -H, -CH 3 , -CH 2 CH 3 , -CH 2 CH(CH 3 ) 2 , and -cyclohexyl alkyl substituents, respectively) have been studied quantitatively by center band-only detection of exchange (CODEX) and pure exchange (PUREX) 13 C solid-state nuclear magnetic resonance (NMR). Flips and small-angle motions of the ester groups associated with the β-relaxation are observed distinctly, and the fraction of slowly flipping groups has been measured with 3% precision. In PMMA, 34% of side groups flip, while the fraction is 31% in PEMA at 25 C. Even the large isobutyl ether and cyclohexylester side groups can flip in the glassy state, although the flipping fraction is reduced to 22% and ∼10%, respectively. In poly methacrylic acid, no slow side group flips are detected. In PMMA, the flipping fraction is temperature-independent between 25 C and 80 C, while in Pemal it increases continuously from 31 to 60% between 25 C and 60 C. A similar doubling is also observed in Pi BMA. (author)

  11. Spectrophotometric determination of dissolved tri n-butyl phosphate in aqueous streams of Purex process

    International Nuclear Information System (INIS)

    Ganesh, S.; Velavendan, P.; Pandey, N.K.; Ahmed, M.K.; Kamachi Mudali, U.; Natarajan, R.

    2012-01-01

    A spectrophotometric method is developed for the determination of dissolved tri-n butyl phosphate (TBP) in aqueous streams of Purex process used in nuclear fuel reprocessing. The method is based on the formation of phosphomolybdate with added ammonium molybdate followed by reduction with hydrazine sulphate in acid medium. Orthophosphate and molybdate ions combine in acidic solution to give molybdophosphoric (phosphomolybdic) acid, which upon selective reduction (with hydrazinium sulphate) produces a blue colour, due to molybdenum blue. The intensity of blue colour is proportional to the amount of phosphate. If the acidity at the time of reduction is 0.5 M in sulphuric acid and hydrazinium sulphate is the reductant, the resulting blue complex exhibits maximum absorption at 810-840 nm. The system obeys Lambert-Beer's law at 830 nm in the concentration range of 0.1-1.0 μg/mol of phosphate. Molar Absorptivity was determined to be 3.1 x 10 4 L mol -1 cm -1 at 830 nm. The results obtained are reproducible with standard deviation of 1 % and relative error less than 2 % and are in good agreement with those obtained by ion chromatographic technique. (author)

  12. Characteristics and mechanism of explosive reactions of Purex solvents with Nitric Acid at elevated temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Miyata, Teijiro [Radiation Application Development Association, Tokai, Ibaraki (Japan); Takada, Junichi; Koike, Tadao; Tsukamoto, Michio; Watanabe, Koji [Department of Fuel Cycle Safety Research, Nuclear Safety Research Center, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan); Ida, Masaaki [JGC PLANTECH CO., LTD (Japan); Nakagiri, Naotaka [JGC Corp., Tokyo (Japan); Nishio, Gunji [Research Organization for Information Science and Technology, Tokai, Ibaraki (Japan)

    2000-03-01

    This investigation was undertaken to make clear the energetic properties and mechanism of explosive decomposition of Purex solvent systems (TBP/n-Dodecane/HNO{sub 3}) by Nitric Acid at elevated temperatures using a calorimetric technique (DSC, ARC) and a chromatographic technique (GC, GC/MS). The measurement of exothermic events of solvent-HNO{sub 3} reactions using DSC with a stainless steel sealed cell showed distinct two peaks with maxima at around 170 and 320degC, respectively. The peak at around 170degC was mainly attributed to the reactions of dealkylation products (n-butyl nitrate) of TBP and the solvent with nitric acid, and the peak at around 320degC was attributed to the exothermic decomposition of nitrated dodecanes formed in the foregoing exothermic reaction of dodecane with nitric acid. By using the data obtained in ARC experiments, activation energies of 123.2 and 152.5 kJ/mol were determined for the exothermic reaction of TBP with nitric acid and for the exothermic pyrolysis of n-butyl nitrate, respectively. Some possible pathways were considered for the explosive decomposition of TBP by nitric acid at elevated temperatures. (author)

  13. Facility stabilization project, fiscal year 1998 Multi-Year Workplan (MYWP) for WBS 1.4

    International Nuclear Information System (INIS)

    Floberg, W.C.

    1997-01-01

    The primary Facility Stabilization mission is to provide minimum safe surveillance and maintenance of facilities and deactivate facilities on the Hanford Site, to reduce risks to workers, the public and environment, transition the facilities to a low cost, long term surveillance and maintenance state, and to provide safe and secure storage of special nuclear materials, nuclear materials, and nuclear fuel. Facility Stabilization will protect the health and safety of the public and workers, protect the environment and provide beneficial use of the facilities and other resources. Work will be in accordance with the Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement), local, national, international and other agreements, and in compliance with all applicable Federal, state, and local laws. The stakeholders will be active participants in the decision processes including establishing priorities, and in developing a consistent set of rules, regulations, and laws. The work will be leveraged with a view of providing positive, lasting economic impact in the region. Effectiveness, efficiency, and discipline in all mission activities will enable Hanford Site to achieve its mission in a continuous and substantive manner. As the mission for Facility Stabilization has shifted from production to support of environmental restoration, each facility is making a transition to support the Site mission. The mission goals include the following: (1) Achieve deactivation of facilities for transfer to EM-40, using Plutonium Uranium Extraction (PUREX) plant deactivation as a model for future facility deactivation; (2) Manage nuclear materials in a safe and secure condition and where appropriate, in accordance with International Atomic Energy Agency (IAEA) safeguards rules; (3) Treat nuclear materials as necessary, and store onsite in long-term interim safe storage awaiting a final disposition decision by US Department of Energy; (4) Implement nuclear materials

  14. Analytical control of reducing agents on uranium/plutonium partitioning at purex process

    International Nuclear Information System (INIS)

    Araujo, Izilda da Cruz de

    1995-01-01

    Spectrophotometric methods for uranium (IV), hydrazine (N 2 H 4 ) and its decomposition product hydrazoic acid(HN 3 ), and hydroxylamine (NH 2 OH) determinations were developed aiming their applications for the process control of CELESTE I installation at IPEN/CNEN-SP. These compounds are normally present in the U/Pu partitioning phase of the spent nuclear treatment via PUREX process. The direct spectrophotometry was used for uranium (IV) analysis in nitric acid-hydrazine solutions based on the absorption measurement at 648 nm. The azomethine compound formed by reaction of hydrazine and p-dimethylamine benzaldehyde with maximum absorption at 457 nm was the basis for the specific analytical method for hydrazine determination. The hydrazoic acid analysis was performed indirectly by its conversion into ferric azide complex with maximum absorption at 465 nm. The hydroxylamine detection was accomplished based on its selective oxidation to nitrous acid which is easily analyzed by the reaction with Griess reagent. The resulted azocompound gas a maximum absorption at 520 nm. The sensibility of 1,4x10 -6 M for U(IV) with 0,8% of precision, 1,6x10 -6 M for hydrazine with 0,8% of precision, 2,3x10 -6 M hydrazoic acid with 0,9% of precision and 2,5x10 -6 M for hydroxylamine with 0,8% of precision were achieved. The interference studies have shown that each reducing agent can be determined in the presence of each other without any interference. Uranium(VI) and plutonium have also shown no interference in these analysis. The established methods were adapted to run inside glove-boxes by using an optical fiber colorimetry and applied to process control of the CELESTE I installation. The results pointed out that the methods are reliable and safety in order to provide just-in-time information about process conditions. (author)

  15. Behavior of mercury and iodine during vitrification of simulated alkaline Purex waste

    International Nuclear Information System (INIS)

    Holton, L.K.

    1981-09-01

    Current plans indicate that the high-level wastes stored at the Savannah River Plant will be solidified by vitrification. The behavior of mercury and iodine during the vitrification process is of concern because: mercury is present in the waste in high concentrations (0.1 to 2.8 wt%); mercury will react with iodine and the other halogens present in the waste during vitrification and; the mercury compounds formed will be volatilized from the vitrification process placing a high particulate load in the vitrification system off-gas. Twelve experiments were completed to study the behavior of mercury during vitrification of simulated SRP Purex waste. The mercury was completely volatized from the vitrification system in all experiments. The mercury reacted with iodine, chlorine and oxygen to form a fine particulate solid. Quantitative recovery of mercury compounds formed in the vitrification system off-gas was not possible due to high (37 to 90%) deposition of solids in the off-gas piping. The behavior of mercury and iodine was most strongly influenced by the vitrification system atmosphere. During experiments performed in which the oxygen content of the vitrification system atmosphere was low (< 1 vol%); iodine retention in the glass product was 27 to 55%, the mercury composition of the solids recovered from the off-gas scrub solutions was 75 to 85 wt%, and a small quantity of metallic mercury was recovered from the off-gas scrub solution. During experiments performed in which the oxygen content of the vitrification system atmosphere was high (20 vol%), iodide retention in the glass product was 3 to 15%, the mercury composition of the solids recovered from the off-gas scrub solutions was 60 to 80 wt%, and very little metallic mercury was recovered from the off-gas scrub solution

  16. Studies on non dispersive solvent extraction for removal of dissolved di-butyl phosphate (DBP) from aqueous medium using hollow fiber membrane contactor

    International Nuclear Information System (INIS)

    Singh, Suman Kumar; Bindu, M.; Tripathi, S.C.; Gandhi, P.M.

    2013-01-01

    PUREX process is based on the principle of mass transfer by liquid liquid dispersion. Tri-n-butyl phosphate (TBP) is universal extractant for PUREX process which is employed for reprocessing the irradiated nuclear fuels for separation and recovery of fissile and fertile materials. The multi cycle solvent extraction processes encompass continuous extraction and stripping operations that are invariably carried out in pulsed columns. The continuous exposure of organic solvent (TBP) to high acidic and radioactive medium leads to decrease the solvent extraction efficiency as it degraded to different level producing di-butyl phosphate and mono-butyl phosphate in significant quantities. Efficiency of purex process decreases as di-butyl phosphate forms aqueous soluble complexes with uranium. Removal of such dissolved DBP from aqueous medium is of direct interest in reprocessing processes as this would enable to sustain the better efficiency of the process and also control the loss of fissile and fertile materials. The non-dispersive solvent extraction is a configuration of the conventional solvent-extraction process where a microporous membrane separates both the immiscible phases, one of which impregnates the membrane, thus bringing the liquid-liquid interface to one side of the membrane. This study is a preliminary evaluation of microporous hollow fiber membrane modules for the removal of dissolved DBP from acidic medium. The performance of the proposed system can be improved by optimizing controlling parameters of the process for quantitative transport of dissolved DBP from acidic medium in the purex process context

  17. An improved facile method for extraction and determination of steroidal saponins in Tribulus terrestris by focused microwave-assisted extraction coupled with GC-MS.

    Science.gov (United States)

    Li, Tianlin; Zhang, Zhuomin; Zhang, Lan; Huang, Xinjian; Lin, Junwei; Chen, Guonan

    2009-12-01

    An improved fast method for extraction of steroidal saponins in Tribulus terrestris based on the use of focus microwave-assisted extraction (FMAE) is proposed. Under optimized conditions, four steroidal saponins were extracted from Tribulus terrestris and identified by GC-MS, which are Tigogenin (TG), Gitogenin (GG), Hecogenin (HG) and Neohecogenin (NG). One of the most important steroidal saponins, namely TG was quantified finally. The recovery of TG was in the range of 86.7-91.9% with RSDTribulus terrestris from different areas of occurrence. The difference in chromatographic characteristics of steroidal saponins was proved to be related to the different areas of occurrence. The results showed that FMAE-GC-MS is a simple, rapid, solvent-saving method for the extraction and determination of steroidal saponins in Tribulus terrestris.

  18. Crud in the solvent extraction process for spent fuel reprocessing

    International Nuclear Information System (INIS)

    Chen Jing

    2004-01-01

    The crud occurred in Purex process is caused by the degradations of extractant and solvent and the existence of insoluble solid particle in the nuclear fuel reprocessing. The crud seriously affects the operation of the extraction column. The present paper reviews the study status on the crud in the Purex process. It is generally accepted that in the Purex process, particularly in the first cycle, the crud occurrence is related to the capillary chemistry phenomena resulting from the deposits of Zr with TBP degradation products HDBP, H 2 MBP, H 3 PO 4 and the insoluble particle RuO 2 and Pd. The occurrence of deposits and the type of crud are tightly related to the molar ratio of HDBP and Zr, and the aqueous pH. In addition, the effect of degradation products from the diluent, such as kerosene, is an unnegligible factor to cause the crud. The crud can be discharged from the extraction equipment with Na 2 CO 3 or oxalic acid. In the study on simulating the crud, the effects of the deposits of Zr with TBP degradation products HDBP, H 2 MBP and H 2 PO 4 , and the insoluble particle RuO 2 and Pd should be considered at the same time. (authors)

  19. Project C-018H, 242-A evaporator/PUREX Plant Process Condensate Treatment Facility Instrumentation and Control (I ampersand C)

    International Nuclear Information System (INIS)

    Dupuis, A.

    1995-01-01

    This Acceptance Test Procedure (ATP) has been prepared to demonstrate that the Collection System Instrumentation ampersand Control System for Project C-018H performs according to design. Specifically, this ATP is designed to verify the following overall system requirements: The input and outputs properly connected to the LCU terminal strips. The control system software conforms to the configuration specified by the logic diagrams, piping and instrumentation diagrams (P ampersand ID), and the LERF operating philosophy. Testing will be performed using actual signals. If actual signals are not available, then simulated signals will be used to complete the tests

  20. Standby status report Hot Semiworks facility

    Energy Technology Data Exchange (ETDEWEB)

    Cooley, C.R.

    1957-09-01

    This report is written to provide information concerning the status of the Hot Semiworks facility as it is placed in stand-by on July 1, 1957. The plant was constructed in 1951 and early 1952. It vas operated on Redox type investigations until the last of 1953. The plant was then converted to the Purex flowsheet under Project CA 513 D. Operations on the Purex type investigations were started in early 1955 and continued until early in 1956. At that time a maintenance program for plant improvement and repair was initiated. This program was completed on July 1, 1957. Statements are contained in this report which pertain to the present status of physical equipment and facilities and the adequacy, operating experience, recommendations for improvement, previous work, and future considerations of the plant. However, the primary intent of the report is to provide pertinent information to personnel associated with a future start-up. For this reason, certain parts of the report are quite detailed. Only statements concerning the existing or previous state of the facility and equipment are factual. Others are opinions or experiences of plant operating personnel. Emphasis has also been placed on the faults encountered rather than the good features of the plant, in order that these faults might be corrected in the future.

  1. Facility Modeling Capability Demonstration Summary Report

    International Nuclear Information System (INIS)

    Key, Brian P.; Sadasivan, Pratap; Fallgren, Andrew James; Demuth, Scott Francis; Aleman, Sebastian E.; Almeida, Valmor F. de; Chiswell, Steven R.; Hamm, Larry; Tingey, Joel M.

    2017-01-01

    A joint effort has been initiated by Los Alamos National Laboratory (LANL), Oak Ridge National Laboratory (ORNL), Savanah River National Laboratory (SRNL), Pacific Northwest National Laboratory (PNNL), sponsored by the National Nuclear Security Administration's (NNSA's) office of Proliferation Detection, to develop and validate a flexible framework for simulating effluents and emissions from spent fuel reprocessing facilities. These effluents and emissions can be measured by various on-site and/or off-site means, and then the inverse problem can ideally be solved through modeling and simulation to estimate characteristics of facility operation such as the nuclear material production rate. The flexible framework called Facility Modeling Toolkit focused on the forward modeling of PUREX reprocessing facility operating conditions from fuel storage and chopping to effluent and emission measurements.

  2. Facility Modeling Capability Demonstration Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Key, Brian P. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Sadasivan, Pratap [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Fallgren, Andrew James [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Demuth, Scott Francis [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Aleman, Sebastian E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); de Almeida, Valmor F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Chiswell, Steven R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hamm, Larry [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Tingey, Joel M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2017-02-01

    A joint effort has been initiated by Los Alamos National Laboratory (LANL), Oak Ridge National Laboratory (ORNL), Savanah River National Laboratory (SRNL), Pacific Northwest National Laboratory (PNNL), sponsored by the National Nuclear Security Administration’s (NNSA’s) office of Proliferation Detection, to develop and validate a flexible framework for simulating effluents and emissions from spent fuel reprocessing facilities. These effluents and emissions can be measured by various on-site and/or off-site means, and then the inverse problem can ideally be solved through modeling and simulation to estimate characteristics of facility operation such as the nuclear material production rate. The flexible framework called Facility Modeling Toolkit focused on the forward modeling of PUREX reprocessing facility operating conditions from fuel storage and chopping to effluent and emission measurements.

  3. A simple approach for facile synthesis of Ag, anisotropic Au and bimetallic (Ag/Au) nanoparticles using cruciferous vegetable extracts

    International Nuclear Information System (INIS)

    Jacob, Jasmine; Mukherjee, Tulsi; Kapoor, Sudhir

    2012-01-01

    We present a simple and straightforward approach for the synthesis and stabilization of relatively monodisperse Ag, Au and bimetallic (Ag/Au) nanoparticles by using cruciferous vegetable (green/red) extracts by simply adjusting the pH environment in the aqueous medium. The vegetable extracts act both as reducing and capping agents. The monometallic and bimetallic nanoparticles of Ag and Au so obtained were characterized by UV–visible spectroscopy, X-ray diffraction (XRD), dynamic light scattering (DLS) and transmission electron microscopy (TEM). It is shown that red cabbage extract can be used for the preparation of anisotropic Au nanoparticles. The formation of Au anisotropic nanoparticles was found to depend on a number of environmental factors, such as the pH of the reaction medium, reaction time, and initial reactant concentrations. Additionally, it is shown that these extract-stabilized Au and Ag nanoparticles can be used as a seed for preparation of bimetallic Au/Ag nanoparticles. For bimetallic alloy nanoparticles the absorption peak was observed between the two maxima of the corresponding metallic particles. The surface plasmon absorption maxima for bimetallic nanoparticles changed linearly with increasing Au mole ratio content in various alloy compositions. It has been shown that the formation of hollow Au spheres depends on the experimental conditions. - Graphical abstract: TEM image of gold nanoparticles at pH 3.27 formed by red cabbage extract. Highlights: ► First report on the reactivity of the extracts toward metal ions using a spectrophotometric technique. ► Red cabbage extract has better reducing properties than green cabbage extract. ► Red cabbage extract can reduce metal ions at any pH. ► Reduction of metal ions can have important consequences in the study of soil chemistry.

  4. A simple approach for facile synthesis of Ag, anisotropic Au and bimetallic (Ag/Au) nanoparticles using cruciferous vegetable extracts

    Energy Technology Data Exchange (ETDEWEB)

    Jacob, Jasmine; Mukherjee, Tulsi; Kapoor, Sudhir, E-mail: sudhirk@barc.gov.in

    2012-10-01

    We present a simple and straightforward approach for the synthesis and stabilization of relatively monodisperse Ag, Au and bimetallic (Ag/Au) nanoparticles by using cruciferous vegetable (green/red) extracts by simply adjusting the pH environment in the aqueous medium. The vegetable extracts act both as reducing and capping agents. The monometallic and bimetallic nanoparticles of Ag and Au so obtained were characterized by UV-visible spectroscopy, X-ray diffraction (XRD), dynamic light scattering (DLS) and transmission electron microscopy (TEM). It is shown that red cabbage extract can be used for the preparation of anisotropic Au nanoparticles. The formation of Au anisotropic nanoparticles was found to depend on a number of environmental factors, such as the pH of the reaction medium, reaction time, and initial reactant concentrations. Additionally, it is shown that these extract-stabilized Au and Ag nanoparticles can be used as a seed for preparation of bimetallic Au/Ag nanoparticles. For bimetallic alloy nanoparticles the absorption peak was observed between the two maxima of the corresponding metallic particles. The surface plasmon absorption maxima for bimetallic nanoparticles changed linearly with increasing Au mole ratio content in various alloy compositions. It has been shown that the formation of hollow Au spheres depends on the experimental conditions. - Graphical abstract: TEM image of gold nanoparticles at pH 3.27 formed by red cabbage extract. Highlights: Black-Right-Pointing-Pointer First report on the reactivity of the extracts toward metal ions using a spectrophotometric technique. Black-Right-Pointing-Pointer Red cabbage extract has better reducing properties than green cabbage extract. Black-Right-Pointing-Pointer Red cabbage extract can reduce metal ions at any pH. Black-Right-Pointing-Pointer Reduction of metal ions can have important consequences in the study of soil chemistry.

  5. Solvent extraction for spent nuclear fuel reprocessing plant

    International Nuclear Information System (INIS)

    Masui, Jinichi

    1986-01-01

    The purex process provides a solvent extraction method widely used for separating uranium and plutonium from nitric acid solution containing spent fuel. The Tokai Works has adopted the purex process with TPB-n dodecane as the extraction agent and a mixer settler as the solvent extraction device. The present article outlines the solvent extraction process and discuss the features of various extraction devices. The chemical principle of the process is described and a procedure for calculating the number of steps for countercurrent equilibrium extraction is proposed. Discussion is also made on extraction processes for separating and purifying uranium and plutonium from fission products and on procedures for managing these processes. A small-sized high-performance high-reliability device is required for carrying out solvent extraction in reprocessing plants. Currently, mixer settler, pulse column and centrifugal contactor are mainly used in these plants. Here, mixer settler is comparted with pulse column with respect to their past achievements, design, radiation damage to solvent, operation halt, controllability and maintenance. Processes for co-extraction, partition, purification and solvent recycling are described. (Nogami, K.)

  6. Solvent extraction studies in miniature centrifugal contactors

    International Nuclear Information System (INIS)

    Siczek, A.A.; Meisenhelder, J.H.; Bernstein, G.J.; Steindler, M.J.

    1980-01-01

    A miniature short-residence-time centrifugal solvent extraction contactor and an eight-stage laboratory minibank of centrifugal contactors were used for testing the possibility of utilizing kinetic effects for improving the separation of uranium from ruthenium and zirconium in the Purex process. Results of these tests showed that a small improvement found in ruthenium and zirconium decontamination in single-stage solvent extraction tests was lost in the multistage extraction tests- in fact, the extent of saturation of the solvent by uranium, rather than the stage residence time, controlled the extent of ruthenium and zirconium extraction. In applying the centrifugal contactor to the Purex process, the primary advantages would be less radiolytic damage to the solvent, high troughput, reduced solvent inventory, and rapid attainment of steady-state operating conditions. The multistage mini contactor was also tested to determine the suitability of short-residence-time contactors for use with the Civex and Thorex processes and was found to be compatible with the requirements of these processes. (orig.) [de

  7. Fuel conditioning facility material accountancy

    International Nuclear Information System (INIS)

    Yacout, A.M.; Bucher, R.G.; Orechwa, Y.

    1995-01-01

    The operation of the Fuel conditioning Facility (FCF) is based on the electrometallurgical processing of spent metallic reactor fuel. It differs significantly, therefore, from traditional PUREX process facilities in both processing technology and safeguards implications. For example, the fissile material is processed in FCF only in batches and is transferred within the facility only as solid, well-characterized items; there are no liquid steams containing fissile material within the facility, nor entering or leaving the facility. The analysis of a single batch lends itself also to an analytical relationship between the safeguards criteria, such as alarm limit, detection probability, and maximum significant amount of fissile material, and the accounting system's performance, as it is reflected in the variance associated with the estimate of the inventory difference. This relation, together with the sensitivity of the inventory difference to the uncertainties in the measurements, allows a thorough evaluation of the power of the accounting system. The system for the accountancy of the fissile material in the FCF has two main components: a system to gather and store information during the operation of the facility, and a system to interpret this information with regard to meeting safeguards criteria. These are described and the precision of the inventory closure over one batch evaluated

  8. 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

  9. Facile synthesis of size-tunable gold nanoparticles by pomegranate (Punica granatum) leaf extract: Applications in arsenate sensing

    Energy Technology Data Exchange (ETDEWEB)

    Rao, Ashit; Mahajan, Ketakee; Bankar, Ashok [Institute of Bioinformatics and Biotechnology, University of Pune, Pune 411007 (India); Srikanth, Rapole [Proteomics Laboratory, National Centre for Cell Science, Pune 411007 (India); Kumar, Ameeta Ravi [Institute of Bioinformatics and Biotechnology, University of Pune, Pune 411007 (India); Gosavi, Suresh, E-mail: swg@physics.unipune.ac.in [Department of Physics, University of Pune, Pune 411007 (India); Centre for Sensor Studies, University of Pune, Pune 411007 (India); Zinjarde, Smita, E-mail: smita@unipune.ac.in [Institute of Bioinformatics and Biotechnology, University of Pune, Pune 411007 (India); Centre for Sensor Studies, University of Pune, Pune 411007 (India)

    2013-03-15

    Highlights: ► Pomegranate leaf extracts mediated rapid gold nanoparticle (AuNP) synthesis. ► The phyto-inspired AuNPs were size-tuned and characterized. ► The reducing and capping agents in the extract were identified. ► The nanoparticles reacted specifically with arsenate (V) ions. - Abstract: When pomegranate leaf extracts were incubated with chloroauric acid (HAuCl{sub 4}), gold nanoparticles (AuNPs) were synthesized. These were characterized by a variety of techniques. With an increasing content of the leaf extract, a gradual decrease in size and an increase in monodispersity were observed. Transmission electron microscope (TEM) images showed that the phyto-fabricated AuNPs were surrounded by an amorphous layer. Gallic acid in the extract mediated the reduction and a natural decapeptide capped the nanostructures. Blocking of thiol groups in the decapeptide cysteine residues caused the nanoparticles to aggregate. On interaction with arsenate (V) ions, the UV–vis spectra of the nanoparticles showed a decrease in intensity and a red-shift. Energy dispersive spectra confirmed the presence of arsenate associated with the AuNPs. Thus, by using these AuNPs, a method for sensing the toxic arsenate ions could be developed.

  10. Facile synthesis of size-tunable gold nanoparticles by pomegranate (Punica granatum) leaf extract: Applications in arsenate sensing

    International Nuclear Information System (INIS)

    Rao, Ashit; Mahajan, Ketakee; Bankar, Ashok; Srikanth, Rapole; Kumar, Ameeta Ravi; Gosavi, Suresh; Zinjarde, Smita

    2013-01-01

    Highlights: ► Pomegranate leaf extracts mediated rapid gold nanoparticle (AuNP) synthesis. ► The phyto-inspired AuNPs were size-tuned and characterized. ► The reducing and capping agents in the extract were identified. ► The nanoparticles reacted specifically with arsenate (V) ions. - Abstract: When pomegranate leaf extracts were incubated with chloroauric acid (HAuCl 4 ), gold nanoparticles (AuNPs) were synthesized. These were characterized by a variety of techniques. With an increasing content of the leaf extract, a gradual decrease in size and an increase in monodispersity were observed. Transmission electron microscope (TEM) images showed that the phyto-fabricated AuNPs were surrounded by an amorphous layer. Gallic acid in the extract mediated the reduction and a natural decapeptide capped the nanostructures. Blocking of thiol groups in the decapeptide cysteine residues caused the nanoparticles to aggregate. On interaction with arsenate (V) ions, the UV–vis spectra of the nanoparticles showed a decrease in intensity and a red-shift. Energy dispersive spectra confirmed the presence of arsenate associated with the AuNPs. Thus, by using these AuNPs, a method for sensing the toxic arsenate ions could be developed

  11. Transportation impact analysis for the shipment of low specific activity nitric acid. Revisison 1

    Energy Technology Data Exchange (ETDEWEB)

    Green, J.R.

    1995-05-16

    This is in support of the Plutonium-Uranium Extraction (PUREX) Facility Low Specific Activity (LSA) Nitric Acid Shipment Environmental Assessment. It analyzes potential toxicological and radiological risks associated with transportation of PUREX Facility LSA Nitric Acid from the Hanford Site to Portsmouth VA, Baltimore MD, and Port Elizabeth NJ.

  12. Transportation impact analysis for the shipment of low specific activity nitric acid. Revisison 1

    International Nuclear Information System (INIS)

    Green, J.R.

    1995-01-01

    This is in support of the Plutonium-Uranium Extraction (PUREX) Facility Low Specific Activity (LSA) Nitric Acid Shipment Environmental Assessment. It analyzes potential toxicological and radiological risks associated with transportation of PUREX Facility LSA Nitric Acid from the Hanford Site to Portsmouth VA, Baltimore MD, and Port Elizabeth NJ

  13. Transportation impact analysis for the shipment of Low Specific Activity Nitric Acid

    International Nuclear Information System (INIS)

    Green, J.R.

    1994-01-01

    This document was written in support of the Plutonium-Uranium Extraction (PUREX) Facility Low Specific Activity (LSA) Nitric Acid Shipment Environmental Assessment. It analyzes the potential toxicological and radiological risks associated with the transportation of PUREX Facility LSA Nitric Acid from the Hanford Site in Washington State to three Eastern ports

  14. Facile green synthesis of graphene-Au nanorod nanoassembly for on-line extraction and sensitive stripping analysis of methyl parathion

    International Nuclear Information System (INIS)

    Zhu, Wenxin; Liu, Wei; Li, Tianbao; Yue, Xiaoyue; Liu, Tao; Zhang, Wentao; Yu, Shaoxuan; Zhang, Daohong; Wang, Jianlong

    2014-01-01

    Graphical abstract: Schematic illustration for the facile green fabrication of GN-AuNRs/GCE and its application for the extraction and electroanalysis of MP. - Highlights: • This paper described a facile green electrochemical approach to synthesize graphene-AuNRs nanocomposite. • The as-synthesized sensor shows low LOD and wide linear concentration range towards MP. • The sensor can be well used for the determination of MP in water and kiwi fruits samples. • This paper further enlarges the scope of facile green synthetic methods of GN-based hybrids. - Abstract: This paper described a facile green electrochemical approach to synthesize graphene-AuNRs nanocomposite (GN-AuNRs) onto glassy carbon electrode (GCE) for electrocatalytic analysis of methyl parathion (MP). This electrochemical synthesis of GN-AuNRs hybrid is environmentally friendly for not involving the chemical reduction of graphene oxide (GO) and facile for just on the basis of electrostatic interaction between GO and AuNRs, as well as electrochemical reduction of GO-AuNRs to GN-AuNRs. Combined the high conductivity, large surface area, good adsorption capacity towards aromatic rings and high catalytic ability of graphene with the excellent electronic properties and adsorption capacity of AuNRs, the high sensitive methyl parathion sensor was fabricated with the GN-AuNRs nanocomposite. The limit of detection (LOD) of the proposed sensor was calculated to be 0.82 ng/mL, which was lower than many previously reported enzyme or nonenzyme-based sensors. In the meantime, the linear detection range of this sensor was from 10 to 500 ng/mL and 750 to 4000 ng/mL, which was also wider than many other enzyme or enzymeless sensors. Furthermore, the facile and green electrochemical reduction strategy provided here could also be used to construct more GN-based hybrids. And the GN-based hybrid might be a new and highly efficient SPE factor, which opens new opportunities for green, facile and sensitive analysis of

  15. Facile preparation of water soluble curcuminoids extracted from turmeric (Curcuma longa L.) powder by using steviol glucosides.

    Science.gov (United States)

    Nguyen, Thi Thanh Hanh; Si, Jinbeom; Kang, Choongil; Chung, Byoungsang; Chung, Donghwa; Kim, Doman

    2017-01-01

    Curcuminoids from rhizomes of Curcuma longa possess various biological activities. However, low aqueous solubility and consequent poor bioavailability of curcuminoids are major limitations to their use. In this study, curcuminoids extracted from turmeric powder using stevioside (Ste), rebaudioside A (RebA), or steviol glucosides (SG) were solubilized in water. The optimum extraction condition by Ste, RebA, or SG resulted in 11.3, 9.7, or 6.7mg/ml water soluble curcuminoids. Curcuminoids solubilized in water showed 80% stability at pH from 6.0 to 10.0 after 1week of storage at 25°C. The particle sizes of curcuminoids prepared with Ste, RebA, and SG were 110.8, 95.7, and 32.7nm, respectively. The water soluble turmeric extracts prepared with Ste, RebA, and SG showed the 2,2-diphenyl-1-picrylhydrazyl radical scavenging (SC50) activities of 127.6, 105.4, and 109.8μg/ml, and the inhibition activities (IC50) against NS2B-NS3(pro) from dengue virus type IV of 14.1, 24.0 and 15.3μg/ml, respectively. Copyright © 2016 The Author(s). Published by Elsevier Ltd.. All rights reserved.

  16. Lithium recycling and cathode material regeneration from acid leach liquor of spent lithium-ion battery via facile co-extraction and co-precipitation processes.

    Science.gov (United States)

    Yang, Yue; Xu, Shengming; He, Yinghe

    2017-06-01

    A novel process for extracting transition metals, recovering lithium and regenerating cathode materials based on facile co-extraction and co-precipitation processes has been developed. 100% manganese, 99% cobalt and 85% nickel are co-extracted and separated from lithium by D2EHPA in kerosene. Then, Li is recovered from the raffinate as Li 2 CO 3 with the purity of 99.2% by precipitation method. Finally, organic load phase is stripped with 0.5M H 2 SO 4 , and the cathode material LiNi 1/3 Co 1/3 Mn 1/3 O 2 is directly regenerated from stripping liquor without separating metal individually by co-precipitation method. The regenerative cathode material LiNi 1/3 Co 1/3 Mn 1/3 O 2 is miro spherical morphology without any impurities, which can meet with LiNi 1/3 Co 1/3 Mn 1/3 O 2 production standard of China and exhibits good electrochemical performance. Moreover, a waste battery management model is introduced to guarantee the material supply for spent battery recycling. Copyright © 2017 Elsevier Ltd. All rights reserved.

  17. Gas chromatographic analysis of extractive solvent in reprocessing plants

    International Nuclear Information System (INIS)

    Marlet, B.

    1984-01-01

    Operation of a reprocessing plant using the Purex process is recalled and analytical controls for optimum performance are specified. The aim of this thesis is the development of analytical methods using gas chromatography required to follow the evolution of the extraction solvent during spent fuel reprocessing. The solvent at different concentrations, is analysed along the reprocessing lines in organic or aqueous phases. Solvent degradation interferes with extraction and decomposition products are analysed. The solvent becomes less and less efficient, also it is distilled and quality is checked. Traces of solvent should also be checked in waste water. Analysis are made as simple as possible to facilitate handling of radioactive samples [fr

  18. U.S. Environmental Protection Agency Clear Air Act notice of construction for the spent nuclear fuel project - Cold Vacuum Drying Facility, project W-441

    International Nuclear Information System (INIS)

    Turnbaugh, J.E.

    1996-01-01

    This document provides information regarding the source and the estimated quantity of potential airborne radionuclide emissions resulting from the operation of the Cold Vacuum Drying (CVD) Facility. The construction of the CVD Facility is scheduled to commence on or about December 1996, and will be completed when the process begins operation. This document serves as a Notice of Construction (NOC) pursuant to the requirements of 40 Code of Federal Regulations (CFR) 61 for the CVD Facility. About 80 percent of the U.S. Department of Energy's spent nuclear fuel (SNF) inventory is stored under water in the Hanford Site K Basins. Spent nuclear fuel in the K West Basin is contained in closed canisters, while the SNF in the K East Basin is in open canisters, which allow release of corrosion products to the K East Basin water. Storage of the current inventory in the K Basins was originally intended to be on an as-needed basis to sustain operation of the N Reactor while the Plutonium-Uranium Extraction (PUREX) Plant was refurbished and restarted. The decision in December 1992 to deactivate the PURF-X Plant left approximately 2,100 MT (2,300 tons) of uranium as part of the N Reactor SNF in the K Basins with no means for near-term removal and processing. The CVD Facility will be constructed in the 100 Area northwest of the 190 K West Building, which is in close proximity to the K East and K West Basins (Figures 1 and 08572). The CVD Facility will consist of five processing bays, with four of the bays fully equipped with processing equipment and the fifth bay configured as an open spare bay. The CVD Facility will have a support area consisting of a control room, change rooms, and other functions required to support operations

  19. Calculation code of mass and heat transfer in a pulsed column for Purex process

    International Nuclear Information System (INIS)

    Tsukada, Takeshi; Takahashi, Keiki

    1993-01-01

    A calculation code for extraction behavior analysis in a pulsed column employed at an extraction process of a reprocessing plant was developed. This code was also combined with our previously developed calculation code for axial temperature profiles in a pulsed column. The one-dimensional dispersion model was employed for both of the extraction behavior analysis and the axial temperature profile analysis. The reported values of the fluid characteristics coefficient, the transfer coefficient and the diffusivities in the pulsed column were used. The calculated concentration profiles of HNO 3 , U and Pu for the steady state have a good agreement with the reported experimental results. The concentration and temperature profiles were calculated under the operation conditions which induce the abnormal U extraction behavior, i.e. U extraction zone is moved to the bottom of the column. Thought there is slight difference between calculated and experimental value, it is appeared that our developed code can be applied to the simulation under the normal operation condition and the relatively slowly transient condition. Pu accumulation phenomena was analyzed with this code and the accumulation tendency is similar to the reported analysis results. (author)

  20. Facile biosynthesis of Ag-NPs using Otostegia limbata plant extract: Physical characterization and auspicious biological activities

    Directory of Open Access Journals (Sweden)

    Rizwan Kausar

    2016-09-01

    Full Text Available Silver nanoparticles (Ag-NPs synthesized through reduction by Otostegia limbata green extract are, hereby, reported for the first time. It is very interesting to observe that in this case, O. limbata plant extract acts as a strong chelating agent in Ag-NPs formation through AgNO3. Scanning electron microscope (SEM studies expose that Ag-NPs formation is highly homogenous and spherical with mean particle size of 32±0.8 nm. A typical Ag absorption peak has been observed at 419 nm by ultra violet (UV-visible spectroscopy which have endorsed the successful formation of single phase Ag-NPs. X-ray diffraction (XRD and Fourier transform infrared spectroscopy (FTIR examination further validates the crystalline pure phase structure of Ag-NPs. Promising results have been recorded against protein kinase inhibition assay and antibacterial assay having prominent pathogenic strains. Our present study explores that biosynthesized eco-friendly Ag-NPs have great potential, in the future, for anticancer drug development with wide range pharmaceutical applications.

  1. Facile and rapid DNA extraction and purification from food matrices using IFAST (immiscible filtration assisted by surface tension).

    Science.gov (United States)

    Strotman, Lindsay N; Lin, Guangyun; Berry, Scott M; Johnson, Eric A; Beebe, David J

    2012-09-07

    Extraction and purification of DNA is a prerequisite to detection and analytical techniques. While DNA sample preparation methods have improved over the last few decades, current methods are still time consuming and labor intensive. Here we demonstrate a technology termed IFAST (Immiscible Filtration Assisted by Surface Tension), that relies on immiscible phase filtration to reduce the time and effort required to purify DNA. IFAST replaces the multiple wash and centrifugation steps required by traditional DNA sample preparation methods with a single step. To operate, DNA from lysed cells is bound to paramagnetic particles (PMPs) and drawn through an immiscible fluid phase barrier (i.e. oil) by an external handheld magnet. Purified DNA is then eluted from the PMPs. Here, detection of Clostridium botulinum type A (BoNT/A) in food matrices (milk, orange juice), a bioterrorism concern, was used as a model system to establish IFAST's utility in detection assays. Data validated that the DNA purified by IFAST was functional as a qPCR template to amplify the bont/A gene. The sensitivity limit of IFAST was comparable to the commercially available Invitrogen ChargeSwitch® method. Notably, pathogen detection via IFAST required only 8.5 μL of sample and was accomplished in five-fold less time. The simplicity, rapidity and portability of IFAST offer significant advantages when compared to existing DNA sample preparation methods.

  2. Microwave assisted facile green synthesis of silver and gold nanocatalysts using the leaf extract of Aerva lanata

    Science.gov (United States)

    Joseph, Siby; Mathew, Beena

    2015-02-01

    Herein, we report a simple microwave assisted method for the green synthesis of silver and gold nanoparticles by the reduction of aqueous metal salt solutions using leaf extract of the medicinal plant Aerva lanata. UV-vis., FTIR, XRD, and HR-TEM studies were conducted to assure the formation of nanoparticles. XRD studies clearly confirmed the crystalline nature of the synthesized nanoparticles. From the HR-TEM images, the silver nanoparticles (AgNPs) were found to be more or less spherical and gold nanoparticles (AuNPs) were observed to be of different morphology with an average diameter of 18.62 nm for silver and 17.97 nm for gold nanoparticles. In order to evaluate the effect of microwave heating upon rate of formation, the synthesis was also conducted under ambient condition without the assistance of microwave radiation and the former method was found to be much faster than the later. The synthesized nanoparticles were used as nanocatalysts in the reduction of 4-nitrophenol to 4-aminophenol by NaBH4.

  3. Analysis of phthalate esters in soils near an electronics manufacturing facility and from a non-industrialized area by gas purge microsyringe extraction and gas chromatography

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Wei [MOE Key Laboratory of Environment and Health, Institute of Environmental Medicine, School of Public Health, Tongji Medical College, Huazhong University of Science and Technology, Wuhan, Hubei (China); Hu, Jia [Suzhou Center for Disease Prevention and Control, Suzhou, Jiangsu (China); Wang, Jinqi; Chen, Xuerong; Yao, Na [MOE Key Laboratory of Environment and Health, Institute of Environmental Medicine, School of Public Health, Tongji Medical College, Huazhong University of Science and Technology, Wuhan, Hubei (China); Tao, Jing, E-mail: jingtao1982@126.com [MOE Key Laboratory of Environment and Health, Institute of Environmental Medicine, School of Public Health, Tongji Medical College, Huazhong University of Science and Technology, Wuhan, Hubei (China); Zhou, Yi-Kai, E-mail: zhouyk@mails.tjmu.edu.cn [MOE Key Laboratory of Environment and Health, Institute of Environmental Medicine, School of Public Health, Tongji Medical College, Huazhong University of Science and Technology, Wuhan, Hubei (China)

    2015-03-01

    Here, a novel technique is described for the extraction and quantitative determination of six phthalate esters (PAEs) from soils by gas purge microsyringe extraction and gas chromatography. Recovery of PAEs ranged from 81.4% to 120.3%, and the relative standard deviation (n = 6) ranged from 5.3% to 10.5%. Soil samples were collected from roadsides, farmlands, residential areas, and non-cultivated areas in a non-industrialized region, and from the same land-use types within 1 km of an electronics manufacturing facility (n = 142). Total PAEs varied from 2.21 to 157.62 mg kg{sup −1} in non-industrialized areas and from 8.63 to 171.64 mg kg{sup −1} in the electronics manufacturing area. PAE concentrations in the non-industrialized area were highest in farmland, followed (in decreasing order) by roadsides, residential areas, and non-cultivated soil. In the electronics manufacturing area, PAE concentrations were highest in roadside soils, followed by residential areas, farmland, and non-cultivated soils. Concentrations of dimethyl phthalate (DMP), diethyl phthalate (DEP), and di-n-butyl phthalate (DnBP) differed significantly (P < 0.01) between the industrial and non-industrialized areas. Principal component analysis indicated that the strongest explanatory factor was related to DMP and DnBP in non-industrialized soils and to butyl benzyl phthalate (BBP) and DMP in soils near the electronics manufacturing facility. Congener-specific analysis confirmed that diethylhexyl phthalate (DEHP) was a predictive indication both in the non-industrialized area (r{sup 2} = 0.944, P < 0.01) and the industrialized area (r{sup 2} = 0.860, P < 0.01). The higher PAE contents in soils near the electronics manufacturing facility are of concern, considering the large quantities of electronic wastes generated with ongoing industrialization. - Highlights: • A new method for determining phthalate esters in soil samples was developed. • Investigate six phthalates near an industry and a

  4. Facile synthesis of silver nanoparticles using Euphorbia antiquorum L. latex extract and evaluation of their biomedical perspectives as anticancer agents

    Directory of Open Access Journals (Sweden)

    Chandrasekaran Rajkuberan

    2017-12-01

    Full Text Available This study reveals the rapid biosynthesis of silver nanoparticles (EAAgNPs using aqueous latex extract of Euphorbia antiquorum L as a potential bioreductant. Synthesized EAAgNPs generate the surface plasmonic resonance peak at 438 nm in UV–Vis spectrophotometer. Size and shape of EAAgNPs were further characterized through transmission electron microscope (TEM which shows well-dispersed spherical nanoparticles with size ranging from 10 to 50 nm. Energy dispersive X-ray spectroscopic analysis (EDAX confirms the presence of silver (Ag as the major constituent element. X-ray diffraction (XRD pattern of EAAgNPs corresponding to (111, (200, (220 and (311 planes, reveals that the generated nanoparticles were face centered cubic crystalline in nature. Interestingly, fourier-transform infrared spectroscopy (FTIR analysis shows the major role of active phenolic constituents in reduction and stabilization of EAAgNPs. Phyto-fabricated EAAgNPs exhibits significant antimicrobial and larvicidal activity against bacterial human pathogens as well as disease transmitting blood sucking parasites such as Culex quinquefasciatus and Aedes aegypti (IIIrd instar larvae. On the other hand, in vitro cytotoxicity assessment of bioformulated EAAgNPs has shown potential anticancer activity against human cervical carcinoma cells (HeLa. The preliminary biochemical (MTT assay and microscopic studies depict that the synthesized EAAgNPs at minimal dosage (IC50 = 28 μg triggers cellular toxicity response. Hence, the EAAgNPs can be considered as an environmentally benign and non-toxic nanobiomaterial for biomedical applications. Keywords: Crystal structure, Euphorbia antiquorum L., Silver nanoparticles, Anticancer, Human pathogens

  5. Partitioning and recovery of neptunium from high level waste streams of PUREX origin using 30% TBP

    International Nuclear Information System (INIS)

    Mathur, J.N.; Murali, M.S.; Balarama Krishna, M.V.; Iyer, R.H.; Chitnis, R.R.; Wattal, P.K.; Theyyunni, T.K.; Ramanujam, A.; Dhami, P.S.; Gopalakrishnan, V.

    1995-01-01

    237 Np is one of the longest-lived nuclides among the actinides present in the high level waste solutions of reprocessing origin. Its separation, recovery and transmutation can reduce the problem of long term storage of the vitrified waste to a great extent. With this objective, the present work was initiated to study the extraction of neptunium into TBP under the conditions relevant to high level waste, along with uranium and plutonium by oxidising it to hexavalent state using potassium dichromate and subsequently recovering it by selective stripping. Three types of simulated HLW solutions namely sulphate bearing (SB), with an acidity of ∼ 0.3 M and non-sulphate wastes originating from the reprocessing of fuels from pressurised heavy water reactor (PHWR) and fast breeder reactor (FBR) with acidities of 3.0 M HNO 3 were employed in these studies. The extraction of U(VI), Np(VI) and Pu(VI) was very high for PHWR- and FBR-HLW solutions, whereas for the SB-HLW solution, these values were less but reasonably high. Quantitative recovery of neptunium and plutonium was achieved using a stripping solution containing 0.1 M H 2 O 2 and 0.01 M ascorbic acid at an acidity of 2.0 M. Since, cerium present in the waste solutions is expected to undergo oxidation in presence of K 2 Cr 2 O 7 , its extraction behaviour was also studied under similar conditions. Based on the results, a scheme was formulated for the recovery of neptunium along with plutonium and was successfully applied to actual high level waste solution originating from the reprocessing of research reactor fuels. (author). 19 refs., 2 figs., 17 tabs

  6. Application of low bitrate image coding to surveillance of electric power facilities. Part 1. Proposal of low bitrate coding for surveillance of electric power facilities and examination of facilities region extraction method; Denryoku setsubi kanshi eno tei rate fugoka hoshiki no tekiyo. 1. Setsubi kanshiyo fugoka hoshiki no teian to setsubi ryoiki chushutsuho no kento

    Energy Technology Data Exchange (ETDEWEB)

    Murata, H.; Ishino, R. [Central Research Institute of Electric Power Industry, Tokyo (Japan)

    1996-03-01

    Current status of low bitrate image coding has been investigated, and a low bitrate coding suitable for the surveillance of electric power facilities has been proposed, to extract its problems to be solved. For the conventional image coding, the waveform coding has been used by which the images are processed as signals. While, for the MPEG-4, a coding method with considering the image information has been proposed. For these coding methods, however, image information lacks details primarily, when lowering the bitrate. Accordingly, these methods can not be applied when the details in the images are important, such as in the case of surveillance of facilities. Then, the coding method has been proposed by expanding the partially detailed coding, and by separating constituent images of facilities, such as power cables and steel towers, designated by operators. It is the special feature of this method that the method can easily respond to the low bitrate and the detailed information can be conserved by using the structure extraction coding for the designated partial image which is generally processed by the low bitrate waveform coding. 29 refs., 17 figs., 1 tab.

  7. Unsymmetrical phosphate as extractant for the extraction of nitric acid

    International Nuclear Information System (INIS)

    Gaikwad, R.H.; Jayaram, R.V.

    2016-01-01

    Tri-n-butyl phosphate (TBP) was first used as an extractant in 1944, during Manhattan project for the separation of actinides and further explored by Warf in 1949 for the extraction of Ce(IV) from aqueous nitric acid. TBP was further used as an extractant in the Plutonium Uranium Recovery by Extraction (PUREX) process. To meet the stringent requirements of the nuclear industry TBP has been extensively investigated. In spite of its wide applicability, TBP suffers from various disadvantages such as high aqueous solubility, third phase formation, chemical and radiation degradation leading to the formation of undesired products. It also suffers from incomplete decontamination of the actinides from fission products. Various attempts have been made to overcome the problems associated with TBP by way of using higher homologues of TBP such as Tri-iso amyl phosphate (TiAP), Tri-secondary butyl phosphate (TsBP), Tri amyl phosphate (TAP). It was found that in some cases the results were considerably better than those obtained with TBP for uranium/thorium extraction. The extraction of nitric acid by TBP and its higher homologues which are symmetrical are well documented. However, no solvent has emerged clearly superior than TBP. Here in we report the extraction of nitric acid with neutral unsymmetrical phosphates and study them as extractants for the extraction of nitric acid. Dibutyl secbutyl phosphate, dibutyl pentyl phosphate and dibutyl heptyl phosphate were synthesised for this purpose and the extraction of nitric acid was studied in n-dodecane. The results indicate that the substitution of one of the alkyl groups of the symmetrical phosphate adjacent to the phosphoryl (P=O) group of the phosphate does not have any pronounced effect on the extraction capacity of nitric acid. (author)

  8. Palladium behavior in the presence of irradiated diluent in the PUREX process

    Energy Technology Data Exchange (ETDEWEB)

    Sio, S. de; Vigier, N. [AREVA NC/DOR/RDP, 1 place Jean Millier, 92084 Paris La Defense (France); Klur, I. [AREVA NC/DT/EP/P, La Hague (France); Tison, E. [AREVA NC/DT/EP/EL, La Hague (France); Bouyer, C.; Eysseric, C. [CEA, Centre de Marcoule, /DEN/DRCP, BP 17171, 30207 Bagnols-sur-Ceze Cedex (France); Lebeau, D.; Goutelard, F. [CEA, Centre de Saclay, /DEN/DPC, 91191 Gif-sur-Yvette Cedex (France); Sejourne, L. [CEA, Centre de Saclay, /DEN/DMN, 91191 Gif-sur-Yvette (France)

    2016-07-01

    AREVA La Hague plants UP3 and UP2-800 started operations to reprocess spent nuclear fuel in 1990 and 1994 respectively. Aging equipment in these plants is a cause for concern as it could lead to process dysfunctions or production rate decrease. A few years ago, several columns had to be replaced in UP3-T4 plutonium purification facility because of clogging. Analyses revealed that TPH degradation products could be responsible for precipitating palladium compounds. 1 M NaOH solutions proved to be efficient to dissolve most of the precipitate. Therefore, several columns in both UP3 and UP2-800 are from now on washed periodically with 1 M NaOH solutions to avoid further clogging and to dissolve current precipitates. (authors)

  9. The fission products palladium and rhodium: Their state in solutions, their behavior in the regeneration of fuel of atomic power stations, and the search for selective extraction techniques

    International Nuclear Information System (INIS)

    Arseenkov, L.V.; Zakharkin, B.S.; Lunichkina, K.P.; Renard, E.V.; Rogozhkin, V.Yu.; Shorokhov, N.A.

    1992-01-01

    At the present time many research centers are working on the extraction of noble metals in the form of fission fragments. Consistent data has been obtained on the mass accumulation of noble metals in various forms of processed nuclear fuel. Requirements are noted that must be met for obtaining industrial and economic efficiency in the extraction of noble metals by the Purex process. Presently there is a lack of information on the extraction of noble metals from spent fuel, particularly as far as the nitric acid media of the Purex process are concerned. The authors will discuss individual test observations on simulating systems and real systems with noble metals. The investigations focused on the noble metals of lowest radioactivity, namely palladium and rhodium. The complexity of the chemistry of ruthenium, on the one hand, and the possible selective, clearing distillation of ruthenium tetroxide from nitric acid solutions, on the other hand, make it necessary to focus the attention on the unresolved problems of the extraction of palladium and rhodium. The article further includes discussion on the following topics: noble metals in solutions of purex process, electrochemical operations involving noble metals, extraction systems for rhodium and palladium, separation of palladium from real solutions

  10. Treatment of tributyl phosphate wastes by extraction cum pyrolysis process

    International Nuclear Information System (INIS)

    Deshingkar, D.S.; Ramaswamy, M.; Kartha, P.K.S.; Kutty, P.V.E.; Ramanujam, A.

    1989-01-01

    For the treatment of spent tri n-butyl phospate (TBP) wastes from Purex process, a method involving extraction of TBP with phosphoric acid followed by pyrolysis of TBP - phosphoric acid phase was investigated. The process was examined with respect to simulated waste, process solvent wastes and aged organic waste samples. These studies seem to offer a simple treatment method for the separation of bulk of diluent from spent solvent wastes. The diluent phase needs further purification for reuse in reprocessing plant; otherwise it can be incinerated. (author). 18 refs., 3 tabs., 6 figs

  11. Variations of uranium and plutonium coprocessing as proliferation-resistant alternatives to the classical purex process

    International Nuclear Information System (INIS)

    Buckham, J.A.; Sumner, W.B.

    1979-08-01

    Evaluation of these alternatives for processing LWR fuel has led to the following conclusions: (1) None of the alternaives provide a pure, technical solution which completely eliminates the potential for proliferation of nuclear weapons by utilizing plutonium from the light water reactors. (2) The heat spike alternative appears feasible and provides the most effective method of rendering the LWR plutonim unattractive for weapons use. (3) The low-DF process alternate would require demonstration to: (a) determine the reliability of the in-cell recycle streams which are used to prevent reversion of the process for purification of plutonium, and (b) verify the fission product decontamination factors. (4) The alternates evaluated have no significant impacts on the design of waste treatment facilities, although the required capacities of high-level solid waste processing and high-level liquid waste storage can be significantly altered. (5) The impact of these alternate processes on fuel fabrication and other aspects of the fuel cycle requires additional evaluation

  12. Separation of An(III) from PUREX raffinate as an innovative SANEX process based on a mixture of TODGA/TBP

    International Nuclear Information System (INIS)

    Sypula, Michal; Wilden, Andreas; Schreinemachers, Christian; Modolo, Giuseppe

    2010-01-01

    Within the ACSEPT project, an innovative SANEX process based on TODGA/TBP for selective An(III) separation from PUREX raffinate was studied. Oxalic acid usually used for Zr complexation is considered a weak point. An investigation to substitute oxalic acid with a different masking agent was carried out. A new masking agent already studied in FZJ was applied and showed good complexation properties towards Zr and Pd. Re-investigation of the formula of the actinide stripping solution was also performed. Good separation of Ln over Am was obtained by means of DTPA and malic acid. Glycine appeared to be the strongest within the tested buffers. (authors)

  13. Fast quantification of short chain fatty acids and ketone bodies by liquid chromatography-tandem mass spectrometry after facile derivatization coupled with liquid-liquid extraction.

    Science.gov (United States)

    Zeng, Mingfei; Cao, Huachuan

    2018-04-15

    Short chain fatty acids (SCFA) and ketone bodies recently emerged as important physiological relevant metabolites because of their association with microbiota, immunology, obesity and other metabolic states. They were commonly analyzed by GC-MS with long run time and laborious sample preparation. In this study we developed a novel LC-MS/MS method using fast derivatization coupled with liquid-liquid extraction to detect SCFA and ketone bodies in plasma and feces. Several different derivatization reagents were evaluated to compare the efficiency, the sensitivity and chromatographic separation of structural isomers. O‑benzylhydroxylamine was selected for its superior overall performance in reaction time and isomeric separation that allowed the measurement of each SCFAs and ketone bodies free from interferences. The derivatization procedure is facile and reproducible in aqueous-organic medium, which abolished the evaporation procedure hampering the analysis of volatile short chain acids. Enhancement in sensitivity remarkably improved the detection limit of SCFA and ketone bodies to sub-fmol level. This novel method was applied to quantify these metabolites in fecal and plasma samples from lean and DIO mouse. Copyright © 2018 Elsevier B.V. All rights reserved.

  14. Heating, ventilating, and air conditioning deactivation thermal analysis of PUREX Plant

    Energy Technology Data Exchange (ETDEWEB)

    Chen, W.W.; Gregonis, R.A. [Westinghouse Hanford Company, Richland, WA (United States)

    1997-08-01

    Thermal analysis was performed for the proposed Plutonium Uranium Extraction Plant exhaust system after deactivation. The purpose of the analysis was to determine if enough condensation will occur to plug or damage the filtration components. A heat transfer and fluid flow analysis was performed to evaluate the thermal characteristics of the underground duct system, the deep-bed glass fiber filter No. 2, and the high-efficiency particulate air filters in the fourth filter building. The analysis is based on extreme variations of air temperature, relative humidity, and dew point temperature using 15 years of Hanford Site weather data as a basis. The results will be used to evaluate the need for the electric heaters proposed for the canyon exhaust to prevent condensation. Results of the analysis indicate that a condition may exist in the underground ductwork where the duct temperature can lead or lag changes in the ambient air temperature. This condition may contribute to condensation on the inside surfaces of the underground exhaust duct. A worst case conservative analysis was performed assuming that all of the water is removed from the moist air over the inside surface of the concrete duct area in the fully developed turbulent boundary layer while the moist air in the free stream will not condense. The total moisture accumulated in 24 hours is negligible. Water puddling would not be expected. The results of the analyses agree with plant operating experiences. The filters were designed to resist high humidity and direct wetting, filter plugging caused by slight condensation in the upstream duct is not a concern. 19 refs., 2 figs.

  15. State waste discharge permit application: 200 Area Treated Effluent Disposal Facility (Project W-049H)

    International Nuclear Information System (INIS)

    1994-08-01

    As part of the original Hanford Federal Facility Agreement and Concent Order negotiations, US DOE, US EPA and the Washington State Department of Ecology agreed that liquid effluent discharges to the ground to the Hanford Site are subject to permitting in the State Waste Discharge Permit Program (SWDP). This document constitutes the SWDP Application for the 200 Area TEDF stream which includes the following streams discharged into the area: Plutonium Finishing Plant waste water; 222-S laboratory Complex waste water; T Plant waste water; 284-W Power Plant waste water; PUREX chemical Sewer; B Plant chemical sewer, process condensate, steam condensate; 242-A-81 Water Services waste water

  16. State waste discharge permit application: 200 Area Treated Effluent Disposal Facility (Project W-049H)

    Energy Technology Data Exchange (ETDEWEB)

    1994-08-01

    As part of the original Hanford Federal Facility Agreement and Concent Order negotiations, US DOE, US EPA and the Washington State Department of Ecology agreed that liquid effluent discharges to the ground to the Hanford Site are subject to permitting in the State Waste Discharge Permit Program (SWDP). This document constitutes the SWDP Application for the 200 Area TEDF stream which includes the following streams discharged into the area: Plutonium Finishing Plant waste water; 222-S laboratory Complex waste water; T Plant waste water; 284-W Power Plant waste water; PUREX chemical Sewer; B Plant chemical sewer, process condensate, steam condensate; 242-A-81 Water Services waste water.

  17. RCRA facility investigation report for the 200-PO-1 operable unit. Revision 1

    International Nuclear Information System (INIS)

    1997-05-01

    This Resource Conservation and Recovery Act (RCRA) Facility Investigation (RFI) report is prepared in support of the RFI/corrective measures study process for the 200-PO-1 Groundwater Operable Unit in the 200 East Area of the Hanford Site. This report summarizes existing information on this operable unit presented in the 200 East and PUREX Aggregate Area Management Study Reports, contaminant specific studies, available modeling data, and groundwater monitoring data summary reports. Existing contaminant data are screened against current regulatory limits to determine contaminants of potential concern (COPC). Each identified COPC is evaluated using well-specific and plume trend analyses

  18. Purex diluent degradation

    International Nuclear Information System (INIS)

    Tallent, O.K.; Mailen, J.C.; Pannell, K.D.

    1984-02-01

    The chemical degradation of normal paraffin hydrocarbon (NPH) diluents both in the pure state and mixed with 30% tributyl phosphate (TBP) was investigated in a series of experiments. The results show that degradation of NPH in the TBP-NPH-HNO 3 system is consistent with the active chemical agent being a radical-like nitrogen dioxide (NO 2 ) molecule, not HNO 3 as such. Spectrophotometric, gas chromatographic, mass spectrographic, and titrimetric methods were used to identify the degradation products, which included alkane nitro and nitrate compounds, alcohols, unsaturated alcohols, nitro alcohols, nitro alkenes, ketones, and carboxylic acids. The degradation rate was found to increase with increases in the HNO 3 concentration and the temperature. The rate was decreased by argon sparging to remove NO 2 and by the addition of butanol, which probably acts as a NO 2 scavenger. 13 references, 11 figures

  19. Separation of americium by liquid-liquid extraction using diglycol-amides water-soluble complexing agents

    Energy Technology Data Exchange (ETDEWEB)

    Chapron, S.; Marie, C.; Pacary, V.; Duchesne, M.T.; Miguirditchian, M. [CEA, Centre de Marcoule, Nuclear Energy Division, RadioChemistry and Processses Departement, 30207 Bagnols-sur-Ceze (France); Arrachart, G.; Pellet-Rostaing, S. [Institut de Chimie Separative de Marcoule, LTSM, Bat 426, F-30207 Bagnols-sur- Ceze (France)

    2016-07-01

    Recycling americium (Am) alone from spent nuclear fuels is an important option studied for the future nuclear cycle (Generation IV systems) since Am belongs to the main contributors of the long-term radiotoxicity and heat power of final waste. Since 2008, a liquid-liquid extraction process called EXAm has been developed by the CEA to allow the recovery of Am alone from a PUREX raffinate (a dissolution solution already cleared from U, Np and Pu). A mixture of DMDOHEMA (N,N'-dimethyl-N,N'-dioctyl-2-(2-(hexyloxy)ethyl)-malonamide) and HDEHP (di-2-ethylhexylphosphoric acid) in TPH is used as the solvent and the Am/Cm selectivity is improved using TEDGA (N,N,N',N'-tetraethyl-diglycolamide) as a selective complexing agent to maintain Cm and heavier lanthanides in the acidic aqueous phase (5 M HNO{sub 3}). Americium is then stripped selectively from light lanthanides at low acidity (pH=3) with a poly-aminocarboxylic acid. The feasibility of sole Am recovery was already demonstrated during hot tests in ATALANTE facility and the EXAm process was adapted to a concentrated raffinate to optimize the process compactness. The speciation of TEDGA complexes formed in the aqueous phase with Am, Cm and lanthanides was studied to better understand and model the behavior of TEDGA in the process. Some Ln-TEDGA species are extracted into the organic phase and this specific chemistry might play a role in the Am/Cm selectivity improvement. Hence the hydrophilicity-lipophilicity balance of the complexing agent is an important parameter. In this comprehensive study, new analogues of TEDGA were synthesized and tested in the EXAm process conditions to understand the relationship between their structure and selectivity. New derivatives of TEDGA with different N-alkyl chain lengths and ramifications were synthesized. The impact of lipophilicity on ligand partitioning and Am/Cm selectivity was investigated. (authors)

  20. Partitioning of actinides from high level waste of PUREX origin using octylphenyl-N,N'-diisobutylcarbamoylmethyl phosphine oxide (CMPO)-based supported liquid membrane

    International Nuclear Information System (INIS)

    Ramanujam, A.; Dhami, P.S.; Gopalakrishnan, V.; Dudwadkar, N.L.; Chitnis, R.R.; Mathur, J.N.

    1999-01-01

    The present studies deal with the application of the supported liquid membrane (SLM) technique for partitioning of actinides from high level waste of PUREX origin. The process uses a solution of octylphenyl-N,N'-diisobutylcarbamoylmethyl phosphine oxide (CMPO) in n-dodecane as a carrier with a polytetrafluoroethylene support and a mixture of citric acid, formic acid, and hydrazine hydrate as the receiving phase. The studies involve the investigation of such parameters as carrier concentration in SLM, acidity of the feed, and the feed composition. The studies indicated good transport of actinides like neptunium, americium, and plutonium across the membrane from nitric acid medium. A high concentration of uranium in the feed retards the transport of americium, suggesting the need for prior removal of uranium from the waste. The separation of actinides from uranium-lean simulated samples as well as actual high level waste has been found to be feasible using the above technique

  1. Operator care and eco-concerned development of a fast, facile and economical assay for basic nitrogenous drugs based on simplified ion-pair mini-scale extraction using safer solvent combined with drop-based spectrophotometry.

    Science.gov (United States)

    Plianwong, Samarwadee; Sripattanaporn, Areerut; Waewsa-nga, Kwanrutai; Buacheen, Parin; Opanasopit, Praneet; Ngawhirunpat, Tanasait; Rojanarata, Theerasak

    2012-08-30

    A fast, facile, and economical assay for basic nitrogenous drugs has been developed based on the mini-scale extraction of the drug-dye ion pair complex combined with the use of safe-for-analyst and eco-friendlier organic extractant and drop-based micro-spectrophotometry. Instead of using large volume devices, the extraction was simply carried out in typical 1.5 mL microcentrifuge tubes along with the use of micropipettes for accurate transfer of liquids, vortex mixer for efficient partitioning of solutes and benchtop centrifuge for rapid phase separation. In the last step, back-extraction was performed by using the microvolume of acidic solution in order to concentrate the colored species into a confined aqueous microdrop and to keep the analyst away from unwanted contact and inhalation of organic solvents during the quantitation step which was achieved by using cuvetteless UV-vis micro-spectrophotometry without any prior dilutions. Using chlorpheniramine maleate as a representative analyte and n-butyl acetate as a less toxic and non-ozone depleting extractant, the miniaturized method was less laborious and much faster. It was accurate, precise and insensitive to the interferences from common excipients. Notably, it gave the assay results of drug in tablets and oral solution comparable to the large-scale pharmacopeial method while the consumption of organic solvents and the release of wastes were lowered by 200-400 folds. Copyright © 2012 Elsevier B.V. All rights reserved.

  2. Study on the uranium-cerium extraction and his application to the treatment of irradiated uranium

    International Nuclear Information System (INIS)

    Lobao, Afonso dos Santos Tome

    1979-01-01

    It was made a study on the behavior of uranium and cerium(IV) extraction, using the latter element as a plutonium simulator in a flowsheet of the treatment of irradiated uranium. Cerium(IV) was used under the same conditions as a plutonium in the Purex process because the admitted similar properties. An experimental work was initiated to determine the equilibrium curves of uranium, under the following conditions: concentration of 1 to 20 g U/1 and acidity varying from 1 to 5M in HNO 3 . Other parameters studied were the volumetric ratio of the phases and the influence of the concentration of TBP (tri-n-butyl phosphate). To guarantee the cerium(IV) extraction, the diluent (varsol) was previously treated with 10% potassium dichromate in perchloric acid, potassium permanganate in 1M sulphuric acid and concentrated sulphuric acid at 70 deg to eliminate reducing compounds. The results obtained for cerium extraction, allowed a better understanding of its behavior in solution. The results permitted to conclude that the decontamination for cerium are very high in the first Purex extraction cycle. The easy as cerium(IV) is reduced to the trivalent state contributes a great deal to its decontamination. (author)

  3. Theoretical treatment of equilibrium data and evaluation of diffusion coefficients in extraction of uranium

    Energy Technology Data Exchange (ETDEWEB)

    Manohar, Smitha; Theyyunni, T K [Process Engineering and Systems Development Division, Bhabha Atomic Research Centre, Mumbai (India); Ragunathan, T S [Department of Chemical Engineering, Indian Inst. of Tech., Mumbai (India)

    1994-06-01

    A meaningful approach to the calculation of the performance of solvent extraction contactors in the PUREX process requires a good understanding of the equilibrium distribution of the important constituents, namely uranyl nitrate and nitric acid. Published literature refers to the empirical correlation of the distribution data, generally in the form of polynomials. Attempts are made to describe the distribution data in a form which is specially convenient for numerical computations along with its theoretical significance. Attempts are also made to derive suitable equations which would aid in estimation of diffusion coefficients in the uranyl nitrate-nitric acid-TBP/diluent system. (author). 2 tabs.

  4. Evaluation of a novel task specific ionic liquid for actinide ion extraction

    International Nuclear Information System (INIS)

    Paramanik, M.; Ghosh, S.K.; Raut, D.R.; Mohapatra, P.K.

    2016-01-01

    Separation of U and Pu from nuclear waste is of great relevance for a sustainable closed fuel cycle point of view. Spent fuel reprocessing by the well known PUREX process is done world wide for the recovery of U and Pu using TBP as the extractant. Room temperature ionic liquids (RTILs) have shown significantly higher extraction of metal ions, particularly at lower acidity as compared to the molecular diluents. Functionalization of ionic liquids has resulted in highly efficient task specific ionic liquids (TSILs) with superior extraction properties than the analogous extractants dissolved in ionic liquids. The present paper reports the evaluation of a novel task specific ionic liquid (TSIL) containing >P=O functional group for the extraction of actinides like U(VI) and Pu(IV)

  5. Study of the extraction mechanisms by TBP saturated by uranyl nitrate; Etude des mecanismes d'extraction du TBP sature par le nitrate d'uranyle

    Energy Technology Data Exchange (ETDEWEB)

    Meze, F

    2004-02-15

    This work deals with a particular phenomenon likely to occur in the nuclear waste reprocessing process PUREX. It was shown earlier by Russian works that the extractant molecule, tributyl phosphate (TBP), saturated by uranyl nitrate keeps its extraction capacities for nitric acid and tetravalent actinides. This study is composed of three parts. Firstly, some liquid-liquid extraction experiments were conducted to verify the ability of TBP saturated by uranyl nitrate to conserve its extraction capacities for nitric acid. Then, during these experiments, the UV and infrared spectra of both phases were recorded to obtain the organic phase speciation. At last, the informations gathered during the experimental part were used to build a general species distribution model of the H{sub 2}O/HNO{sub 3}/UO{sub 2}(NO{sub 3}){sub 2}/TBP system. (author)

  6. Facile one step synthesis of novel TiO2 nanocoral by sol-gel method using Aloe vera plant extract

    Science.gov (United States)

    Venkatesh, K. S.; Krishnamoorthi, S. R.; Palani, N. S.; Thirumal, V.; Jose, Sujin P.; Wang, Fu-Ming; Ilangovan, R.

    2015-05-01

    Titanium oxide (TiO2) nanoparticles (NPs) were synthesized by sol gel method using Aloe vera plant extract as a biological capping agent and a cauliflower-nanocoral morphology was observed in this technique. The assynthesized TiO2 nanopowder was calcined at a range of temperatures (300-600 °C) for 1 h. The influence of A. vera plant extract on the thermal, structural and morphological properties of TiO2 nanopowder was evaluated. Thermogravimetric analysis/differential thermal analysis was employed to study the thermal properties of the assynthesized TiO2 nanopowder. The crystallinity, phase transformation and the crystallite size of the calcined samples were studied by X-ray diffraction technique. XRD result confirmed the presence of TiO2 with anatase phase. FT Raman spectra showed the Raman active modes pertaining to the TiO2 anatase phase and Raman band shift was also observed with respect to particle size variation. The different functional group vibrations of as dried pure A. vera plant extract were compared with the mixture of TiO2 and A. vera plant extract by FT-IR analysis. The scanning electron microscopy images apparently showed the formation of spherical shaped NPs and also it demonstrated the effect of A. vera plant extract on the reduction of particles size. The surface area of the TiO2 NPs was measured through Brunauer-Emmett-Teller analysis. Transmission electron microscopy images ascertained that the spherical shaped TiO2 NPs were formed with cauliflower-nanocoral morphology decorated with nanopolyps with the size range between 15 and 30 nm.

  7. Distribution and identification of Plutonium(IV) species in tri-n-butyl phosphate/HNO3 extraction system containing acetohydroxamic acid

    International Nuclear Information System (INIS)

    Tkac, P.; Paulenova, A.; Vandegrift, G.F.; Krebs, J.F.

    2009-01-01

    There was a significant research progress achieved with the aim to modify conventional PUREX process by stripping of plutonium from the tri-n-butyl phosphate (TBP) extraction product in the form of non-extractable complexes upon addition of back-hold complexation agents. The present paper reports effects of such salt-free complexant, acetohydroxamic acid (HAHA), on distribution ratio of Pu(IV) under wide concentration of nitric acid and additional nitrate. General formula of plutonium species present in the organic phase can be described as Pu(OH) x (AHA) y (NO3) 4-x-y x 2TBP x wHNO 3 . (author)

  8. Irradiated uranium reprocessing, Final report I-VI, Part V - report on development of laboratory extraction procedure for separation of U, Pu, and FP on the tracer level; Prerada ozracenog urana. Zavrani izvestaj - I-VI, V Deo - Izvestaj o razradi laboratorijskog procesa ekstrakcije za odvajanje U, Pu i FP na nivou obelezavaca

    Energy Technology Data Exchange (ETDEWEB)

    Gal, I [Institute of Nuclear Sciences Boris Kidric, Odeljenje za eksploataciju nuklearnog goriva, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    A laboratory extraction procedure was developed for separation of uranium, plutonium and fission products from the nitric solution. The procedure would be applied for uranium and spent fuel from the RA reactor in Vinca. This is a Purex type of procedure adapted for laboratory purposes. Experimental data are obtained by using syntetic nitric uranium solutions with Pu and fission products additions as tracers. A device for completing the process was constructed.

  9. Calculations of high-power production target and beamdump for the GSI future Super-FRS for a fast extraction scheme at the FAIR Facility

    International Nuclear Information System (INIS)

    Tahir, N A; Weick, H; Iwase, H

    2005-01-01

    A superconducting fragment separator (Super-FRS) is being designed for the production and separation of radioactive isotopes at the future FAIR (Facility for Antiprotons and Ion Research) facility at Darmstadt. This paper discusses various aspects and requirements for the high-power production target that will be used in the Super-FRS experiments. The production target must survive over an extended period of time as it will be used during the course of many experiments. The specific power deposited by the high intensity beam that will be generated at the future FAIR facility will be high enough to destroy the target in most of the cases as a result of a single shot from the new heavy ion synchrotrons SIS100/300. By using an appropriate beam intensity and focal spot parameters, the target would survive after being irradiated once. However, the heat should be dissipated efficiently before the same target area is irradiated again. We have considered a wheel shaped solid carbon target that rotates around its axis so that different areas of the target are irradiated successively. This allows for cooling of the beam heated region by thermal conduction before the same part of the target is irradiated a second time. Another attractive option is to use a liquid jet target at the Super-FRS. First calculations of a possible liquid lithium target are also presented in this paper. One of the advantages of using lithium as a target is that it will survive even if one uses a smaller focal spot, which has half the area of that used for a solid carbon target. This will significantly improve the isotope resolution. A similar problem associated with these experiments will be safe deposition of the beam energy in a beamdump after its interaction with the production target. We also present calculations to study the suitability of a proposed beamdump

  10. Effect of abnormal outflow from end stages on concentration profile in uranium-stripping bank of PUREX flowsheet

    International Nuclear Information System (INIS)

    Ueda, Yoshinori; Matsumoto, Shiro

    2002-01-01

    The effect of the abnormal outflow from the end stages on the concentration profile was studied for the uranium-stripping bank to consider the design and the operation of the solvent extraction process, which eases the undesirable effects due to such abnormal flow. The abnormal outflow affected the concentration profile in the same manner as the decrease in the rate of the corresponding liquid flow rate entering the bank. The results suggested that the solvent extractor at the aqueous inlet stage in stripping banks and the solvent extractor at the organic inlet stage in extraction banks should be carefully designed to restrict the respective abnormal aqueous and organic outflows within the normal operational liquid flow rate range. Combining the result and the inherent phase separation behavior of the extractor suggested the possibility of designing the process with the self-controlled function of throughput, which eases the change of the concentration profile due to the undesirable increase in the rate of liquid flow rate entering the bank. Basically the proposed approaches are probably applicable to other general extraction and stripping processes. (author)

  11. Partitioning of Minor Actinides from High Active Raffinates using Bis-Diglycol-amides (BisDGA) as new efficient Extractants

    Energy Technology Data Exchange (ETDEWEB)

    Modolo, G.; Vijgen, H. [Forschungszentrum Juelich GmbH, Institute for Energy Research, Safety Research and Reactor Technology, 52425 Juelich (Germany); Espartero, A.G. [Centro de Investigaciones Energeticas, Medioambientales y Tecnologicas (CIEMAT), Avda. Complutense 22, 28040-Madrid (Spain); Prados, P. [Departamento de Quimica Organica, Facultad de Ciencias, Universidad Autonoma de Madrid - UAM, carretera de Colmenar Viejo km 15.3, 28049-Madrid (Spain); Mendoza, J. de [Departamento de Quimica Organica, Facultad de Ciencias, Universidad Autonoma de Madrid - UAM, carretera de Colmenar Viejo km 15.3, 28049-Madrid (Spain); Institut Catala d' Investigacio Quimica (ICIQ) Av. Paisos Catalans 16, 43007-Tarragona (Spain)

    2008-07-01

    Two new polyamide extractants has been selected, namely UAM-069 and UAM-081, both synthesized at the University of Madrid (UAM), to develop a new separation process. These two ligands are bis-diglycol-amides, consisting of two diglycol-amides moieties grafted on an aromatic platform (UAM-069) or on an aliphatic linker (UAM-081), respectively. The extraction of actinides and fission products was studied from synthetic PUREX raffinate. Actinides(III) and lanthanides(III) are highly extracted from acidities > 1 mol/L HNO{sub 3}. The extraction of Zr, Mo and Pd could be suppressed with complexing agents such as oxalic acid and HEDTA. In the present paper the results of the batch extraction results are presented which serve for the development of a new continuous counter current process to be tested in centrifugal contactors. (authors)

  12. Facility transition instruction

    International Nuclear Information System (INIS)

    Morton, M.R.

    1997-01-01

    The Bechtel Hanford, Inc. facility transition instruction was initiated in response to the need for a common, streamlined process for facility transitions and to capture the knowledge and experience that has accumulated over the last few years. The instruction serves as an educational resource and defines the process for transitioning facilities to long-term surveillance and maintenance (S and M). Generally, these facilities do not have identified operations missions and must be transitioned from operational status to a safe and stable configuration for long-term S and M. The instruction can be applied to a wide range of facilities--from process canyon complexes like the Plutonium Uranium Extraction Facility or B Plant, to stand-alone, lower hazard facilities like the 242B/BL facility. The facility transition process is implemented (under the direction of the US Department of Energy, Richland Operations Office [RL] Assistant Manager-Environmental) by Bechtel Hanford, Inc. management, with input and interaction with the appropriate RL division and Hanford site contractors as noted in the instruction. The application of the steps identified herein and the early participation of all organizations involved are expected to provide a cost-effective, safe, and smooth transition from operational status to deactivation and S and M for a wide range of Hanford Site facilities

  13. Facile synthesis of magnetic carbon nitride nanosheets and its application in magnetic solid phase extraction for polycyclic aromatic hydrocarbons in edible oil samples.

    Science.gov (United States)

    Zheng, Hao-Bo; Ding, Jun; Zheng, Shu-Jian; Zhu, Gang-Tian; Yuan, Bi-Feng; Feng, Yu-Qi

    2016-01-01

    In this study, we proposed a method to fabricate magnetic carbon nitride (CN) nanosheets by simple physical blending. Low-cost CN nanosheets prepared by urea possessed a highly π-conjugated structure; therefore the obtained composites were employed as magnetic solid-phase extraction (MSPE) sorbent for extraction of polycyclic aromatic hydrocarbons (PAHs) in edible oil samples. Moreover, sample pre-treatment time could be carried out within 10 min. Thus, a simple and cheap method for the analysis of PAHs in edible oil samples was established by coupling magnetic CN nanosheets-based MSPE with gas chromatography-mass spectrometry (GC/MS) analysis. Limits of quantitation (LOQs) for eight PAHs ranged from 0.4 to 0.9 ng/g. The intra- and inter-day relative standard deviations (RSDs) were less than 15.0%. The recoveries of PAHs for spiked soybean oil samples ranged from 91.0% to 124.1%, with RSDs of less than 10.2%. Taken together, the proposed method offers a simple and cost-effective option for the convenient analysis of PAHs in oil samples. Copyright © 2015 Elsevier B.V. All rights reserved.

  14. A facile biomimetic preparation of highly stabilized silver nanoparticles derived from seed extract of Vigna radiata and evaluation of their antibacterial activity

    Science.gov (United States)

    Choudhary, Manoj Kumar; Kataria, Jyoti; Cameotra, Swaranjit Singh; Singh, Jagdish

    2016-01-01

    The significant antibacterial activity of silver nanoparticles draws the major attention toward the present nanobiotechnology. Also, the use of plant material for the synthesis of metal nanoparticles is considered as a green technology. In this context, a non-toxic, eco-friendly, and cost-effective method has been developed for the synthesis of silver nanoparticles using seed extract of mung beans ( Vigna radiata). The synthesized nanoparticles have been characterized by UV-visible spectroscopy (UV-Vis), Fourier transform infrared spectroscopy (FT-IR), transmission electron microscopy (TEM), atomic absorption spectroscopy (AAS), and X-ray diffraction (XRD). The UV-visible spectrum showed an absorption peak at around 440 nm. The different types of phytochemicals present in the seed extract synergistically reduce the Ag metal ions, as each phytochemical is unique in terms of its structure and antioxidant function. The colloidal silver nanoparticles were observed to be highly stable, even after 5 months. XRD analysis showed that the silver nanoparticles are crystalline in nature with face-centered cubic geometry and the TEM micrographs showed spherical particles with an average size of 18 nm. Further, the antibacterial activity of silver nanoparticles was evaluated by well-diffusion method and it was observed that the biogenic silver nanoparticles have an effective antibacterial activity against Escherichia coli and Staphylococcus aureus. The outcome of this study could be useful for nanotechnology-based biomedical applications.

  15. Facilities & Leadership

    Data.gov (United States)

    Department of Veterans Affairs — The facilities web service provides VA facility information. The VA facilities locator is a feature that is available across the enterprise, on any webpage, for the...

  16. Facility effluent monitoring plan determinations for the 200 Area facilities

    International Nuclear Information System (INIS)

    Nickels, J.M.

    1991-11-01

    The following facility effluent monitoring plan determinations document the evaluations conducted for the Westinghouse Hanford Company 200 Area facilities (chemical processing, waste management, 222-S Laboratory, and laundry) on the Hanford Site in south central Washington State. These evaluations determined the need for facility effluent monitoring plans for the 200 Area facilities. The facility effluent monitoring plan determinations have been prepared in accordance with A Guide for Preparing Hanford Site Facility Effluent Monitoring Plans, WHC-EP-0438 (WHC 1991). The Plutonium/Uranium Extraction Plant and UO 3 facility effluent monitoring plan determinations were prepared by Los Alamos Technical Associates, Richland, Washington. The Plutonium Finishing Plant, Transuranic Waste Storage and Assay Facility, T Plant, Tank Farms, Low Level Burial Grounds, and 222-S Laboratory determinations were prepared by Science Applications International Corporation of Richland, Washington. The B Plant Facility Effluent Monitoring Plan Determination was prepared by ERCE Environmental Services of Richland, Washington

  17. Biochemistry Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Biochemistry Facility provides expert services and consultation in biochemical enzyme assays and protein purification. The facility currently features 1) Liquid...

  18. Thermal neutron calibration of a tritium extraction facility using the 6Li(n,t)4He/197Au(n,γ)198Au cross section ratio for standardization

    International Nuclear Information System (INIS)

    Bretscher, M.M.; Smith, D.L.

    1980-08-01

    Absolute tritium activities in a neutron-activated metallic lithium samples have been measured by liquid scintillation methods to provide data needed for the determination of capture-to-fission ratios in fast breeder reactor spectra and for recent measurements of the 7 Li(n,n't) 4 He cross section. The tritium extraction facility used for all these experiments has now been calibrated by measuring the 6 Li(n,t) 4 He/ 197 Au/n,γ) 198 Au activity ratio for thermal neutrons and comparing the result with the well-known cross sections. The calculated-to-measured activity ratio was found to be 1.033 +- 0.018. 2 figures, 20 tables

  19. 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.

  20. Facile synthesis of multifunctional attapulgite/Fe{sub 3}O{sub 4}/polyaniline nanocomposites for magnetic dispersive solid phase extraction of benzoylurea insecticides in environmental water samples

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Xiaoling; Qiao, Kexin; Ye, Yiren; Yang, Miyi; Li, Jing; Gao, Haixiang; Zhang, Sanbing; Zhou, Wenfeng; Lu, Runhua, E-mail: rhlu@cau.edu.cn

    2016-08-31

    In this study, the superparamagnetic attapulgite/Fe{sub 3}O{sub 4}/polyaniline (ATP/Fe{sub 3}O{sub 4}/PANI) nanocomposites were successfully synthesized by a one-pot method. Fe (III) was applied as both the oxidant for the oxidative polymerization of aniline and the single iron source of Fe{sub 3}O{sub 4} formed by the redox reaction between aniline and Fe (III). The ATP/Fe{sub 3}O{sub 4}/PANI was used as sorbent for magnetic dispersive solid phase extraction (MDSPE) of benzoylurea insecticides (BUs) in environmental water samples. The as-prepared nanocomposite sorbents were characterized by Fourier transform infrared spectra (FT-IR), X Ray diffraction (XRD), scanning electron microscopy(SEM), transmission electron microscopy (TEM), and vibrating sample magnetometry. Various experimental parameters affecting the ATP/Fe{sub 3}O{sub 4}/PANI-based MDSPE procedure, including the composition of the nanocomposite sorbents, amount of ATP/Fe{sub 3}O{sub 4}/PANI nanocomposites, vortex time, pH, and desorption conditions were investigated. Under the optimal conditions, a good linearity was observed for all target analytes, with correlation coefficients (r{sup 2}) ranging from 0.9985 to 0.9997; the limits of detection (LOD) were in the range of 0.02–0.43 μg L{sup −1}, and the recoveries of analytes using the proposed method ranged between 77.37% and 103.69%. The sorbents exhibited an excellent reproducibility in the range of 1.52–5.27% in extracting the five target analytes. In addition, the intra-day and inter-day precision values were found to be in the range of 0.78–6.86% and 1.66–8.41%, respectively. Finally, the proposed ATP/Fe{sub 3}O{sub 4}/PANI-based MDSPE method was successfully applied to analyze river water samples by rapid preconcentration of BUs. - Highlights: • A novel superparamagnetic ATP/Fe{sub 3}O{sub 4}/PANI nanocomposite was first introduced in MDSPE. • ATP/Fe{sub 3}O{sub 4}/PANI nanocomposites exhibited fast adsorption and desorption

  1. Facile synthesis of multifunctional attapulgite/Fe3O4/polyaniline nanocomposites for magnetic dispersive solid phase extraction of benzoylurea insecticides in environmental water samples

    International Nuclear Information System (INIS)

    Yang, Xiaoling; Qiao, Kexin; Ye, Yiren; Yang, Miyi; Li, Jing; Gao, Haixiang; Zhang, Sanbing; Zhou, Wenfeng; Lu, Runhua

    2016-01-01

    In this study, the superparamagnetic attapulgite/Fe 3 O 4 /polyaniline (ATP/Fe 3 O 4 /PANI) nanocomposites were successfully synthesized by a one-pot method. Fe (III) was applied as both the oxidant for the oxidative polymerization of aniline and the single iron source of Fe 3 O 4 formed by the redox reaction between aniline and Fe (III). The ATP/Fe 3 O 4 /PANI was used as sorbent for magnetic dispersive solid phase extraction (MDSPE) of benzoylurea insecticides (BUs) in environmental water samples. The as-prepared nanocomposite sorbents were characterized by Fourier transform infrared spectra (FT-IR), X Ray diffraction (XRD), scanning electron microscopy(SEM), transmission electron microscopy (TEM), and vibrating sample magnetometry. Various experimental parameters affecting the ATP/Fe 3 O 4 /PANI-based MDSPE procedure, including the composition of the nanocomposite sorbents, amount of ATP/Fe 3 O 4 /PANI nanocomposites, vortex time, pH, and desorption conditions were investigated. Under the optimal conditions, a good linearity was observed for all target analytes, with correlation coefficients (r 2 ) ranging from 0.9985 to 0.9997; the limits of detection (LOD) were in the range of 0.02–0.43 μg L −1 , and the recoveries of analytes using the proposed method ranged between 77.37% and 103.69%. The sorbents exhibited an excellent reproducibility in the range of 1.52–5.27% in extracting the five target analytes. In addition, the intra-day and inter-day precision values were found to be in the range of 0.78–6.86% and 1.66–8.41%, respectively. Finally, the proposed ATP/Fe 3 O 4 /PANI-based MDSPE method was successfully applied to analyze river water samples by rapid preconcentration of BUs. - Highlights: • A novel superparamagnetic ATP/Fe 3 O 4 /PANI nanocomposite was first introduced in MDSPE. • ATP/Fe 3 O 4 /PANI nanocomposites exhibited fast adsorption and desorption kinetics. • An excellent sorbent-to-sorbent reproducibility was demonstrated in the

  2. Partitioning of minor actinides from HLLW using the DIAMEX process. Pt. 1. Demonstration of extraction performances and hydraulic behaviour of the solvent in a continuous process

    International Nuclear Information System (INIS)

    Courson, O.; Lebrun, M.; Malmbeck, R.; Pagliosa, G.; Roemer, K.; Saetmark, B.; Glatz, J.P.

    2000-01-01

    The French DIAMEX process shows very promising capabilities in separating minor actinides from HLLW. A counter-current centrifugal extractor experiment has been conducted to investigate the capabilities and possibilities of the DIAMEX process (hydraulic and extraction behaviour), for the separation of lanthanides from a simulated high level liquid waste (HLLW), corresponding in concentration to a raffinate from the PUREX process. A ''hot'' batch test, using genuine HLLW, and a continuous counter-current experiment have verified the excellent extraction and hydraulic behaviour, respectively. With only four extraction stages in the cold experiment, lanthanide decontamination factors were higher than 2000, except for europium. Co-extraction of molybdenum and zirconium was efficiently prevented using oxalic acid in the feed solution. The back-extraction was very efficient, yielding in 4 stages more than 99% recovery of lanthanides. Palladium and ruthenium were more difficult to back-extract and for these elements further investigations are needed. (orig.)

  3. Dance Facilities.

    Science.gov (United States)

    Ashton, Dudley, Ed.; Irey, Charlotte, Ed.

    This booklet represents an effort to assist teachers and administrators in the professional planning of dance facilities and equipment. Three chapters present the history of dance facilities, provide recommended dance facilities and equipment, and offer some adaptations of dance facilities and equipment, for elementary, secondary and college level…

  4. Liquid-liquid extraction kinetics of uranyl nitrate and actinides (III)-lanthanides nitrates by extractants with amide function

    International Nuclear Information System (INIS)

    Toulemonde, V.

    1995-01-01

    Nowadays, the most important part of electric power is generated by fission energy. But spent fuels have then to be reprocessed. The production of these reprocessed materials separately and with a high purity level is done according to a liquid-liquid extraction process (Purex process) with the use of tributyl phosphate as solvent. Optimization studies concerning the extracting agent have been undertaken. This work gives the results obtained for the uranyl nitrate and the actinides (III)-lanthanides (III) nitrates extraction by extractants with amide function (monoamide for U(VI) and diamide for actinides (III) and lanthanides (III)). The extraction kinetics have been studied in the case of a metallic specie transfer from the aqueous phase towards the organic phase. The experiments show that the nitrates extraction kinetics is limited by the complexation chemical reaction of the species at the interface between the two liquids. An adsorption-desorption interfacial reactional mechanism (Langmuir theory) is proposed for the uranyl nitrate. (O.M.)

  5. PERMCAT experiments with tritium at high helium flow rates relevant for the tritium extraction systems using the CAPER facility at TLK

    Energy Technology Data Exchange (ETDEWEB)

    Bükki-Deme, András, E-mail: andras.buekki-deme@kit.edu; Demange, David; Le, Thanh-Long; Fanghänel, Eleonore; Simon, Karl-Heinz

    2016-11-01

    Highlights: • We examined PERMCAT reactor efficiency processing tritiated water at high Helium carrier flow rates. • We have found that – as expected from previous studies – that the swamping ratio (ratio between the impurity and purge side flow rates) has a key effect on the decontamination factors. • On the other hand, some rather unexpected effects tend to show that the limiting phenomena of such specific operation of PERMCAT reactors (at high impurity flow rates, thus short residence time) lies on the kinetics of the isotope exchange reactions. - Abstract: Experiments are still necessary to consolidate the processes retained for the Tritium Extraction Systems of the European ITER Test Blanket Modules (TBM). A PERMCAT reactor combines a catalyst promoting isotope exchange reactions and a Pd/Ag membrane allowing tritium recovery from complex gaseous mixtures containing tritium in different chemical forms. Originally developed for the Tokamak Exhaust Processing, the PERMCAT process is also candidate to detritiate the water arising from an adsorption column installed in the TBM ancillary systems. We discuss the results of an extensive experimental campaign using a PERMCAT reactor to process Q{sub 2}O containing impurity gas mixtures at high flow rates. Two different experimental configurations were studied, namely PERMCAT stand-alone, and PERMCAT in combination with a zeolite molecular sieve bed (MSB, previously loaded with Q{sub 2}O) under regeneration. On the one hand, many expected behaviors were observed, such as the key influence of the swamping ratio (ratio between the impurity and purge side flow rates) on the decontamination factors. On the other hand, some rather unexpected effects tend to show that the limiting phenomena of such specific operation of PERMCAT reactors (at high flow rates, thus short residence time) lies on the kinetics of the isotope exchange reactions.

  6. Facile green synthesis of zinc oxide nanoparticles using Ulva lactuca seaweed extract and evaluation of their photocatalytic, antibiofilm and insecticidal activity.

    Science.gov (United States)

    Ishwarya, Ramachandran; Vaseeharan, Baskaralingam; Kalyani, Subramanian; Banumathi, Balan; Govindarajan, Marimuthu; Alharbi, Naiyf S; Kadaikunnan, Shine; Al-Anbr, Mohammed N; Khaled, Jamal M; Benelli, Giovanni

    2018-01-01

    The bioactivity of semiconductor nanocomplexes has been poorly studied in the field of pesticide science. In this research, the synthesis of zinc nanoparticles was accomplished through new effortless green chemistry process, using the Ulva lactuca seaweed extract as a reducing and capping agent. The production of U. lactuca-fabricated ZnO nanoparticles (Ul-ZnO Nps) was characterized by powder X-ray diffraction (XRD), UV-visible, Fourier transform infrared (FTIR) spectroscopy, selected area electron diffraction (SAED) analysis and transmission electron microscopy (TEM). The U. lactuca-fabricated ZnO NPs were tested for their photodegradative action against organic dyes, as well as for antibiofilm and larvicidal activities. The UV visible absorbance spectrum of Ul-ZnO NPs exhibited the absorbance band at 325nm and TEM highlighted average crystallite sizes of nanoparticles of 10-50nm. Methylene blue (MB) dye was efficiently corrupted under sunlight in presence of Ul-ZnO NPs. Excellent bactericidal activity was shown by the Ul-ZnO Nps on Gram positive (Bacillus licheniformis and Bacillus pumilis) and Gram negative (Escherichia coliand Proteus vulgaris) bacteria. High antibiofilm potential was noted under both dark and sunlight conditions. The impact of a single treatment with Ul-ZnO NPs on biofilm architecture was also analyzed by confocal laser scanning microscopy (CLSM) on both Gram positive and Gram negative bacteria. Moreover, Ul-ZnO NPs led to 100% mortality of Aedes aegypti fourth instar larvae at the concentration of 50μg/ml within 24h. The effects of ZnO nanoparticle-based treatment on mosquito larval morphology and histology were monitored. Overall, based on our results, we believe that the synthesis of multifunctional Ul-ZnO Nps using widely available seaweed products can be promoted as a potential eco-friendly option to chemical methods currently used for nanosynthesis of antimicrobials and insecticides. Copyright © 2017 Elsevier B.V. All rights reserved.

  7. Fiber optic adaptation of the interference filter photometer SPECTRAN for in-line measurements in PUREX process control

    International Nuclear Information System (INIS)

    Buerck, J.; Kraemer, K.; Koenig, W.

    1990-02-01

    The multicomponent version of the interference filter photometer SPECTRAN was adapted by radiation resistant quartz glass optical fibers to in-line flow cells in the aqueous and organic bypass stream of an uranium laboratory extraction column. A combined photometric/electrolytical conductivity measurement allows this modified process instrument to be used as uranium/plutonium in-line monitor in radioactive process streams. By applying a high performance 100 W quartz halogen lamp and suitable light focussing optics the light intensity, attenuated by coupling losses, could be increased to the desired level even when 1000 μm-single strand fibers (2x18 m) were used to transmit the light. In a series of calibration experiments the U(VI)- and U(IV)-extinction coefficients were determined as a function of nitric acid molarity (for U(VI) also in TBP/kerosene). Furthermore the validity of Lambert-Beer's law was examined for both oxidation states at different optical path lengths and nitric acid/electrolytical conductivity calibration functions between 0-100 g/l U(VI) and 0-4 mol/l HNO 3 were set up. (orig./EF) [de

  8. Alpha-contained laboratory scale pulse column facility for SRL

    International Nuclear Information System (INIS)

    Reif, D.J.; Cadieux, J.R.; Fauth, D.J.; Thompson, M.C.

    1980-01-01

    For studying solvent extraction processes, a laboratory-sized pulse column facility was constructed at the Savannah River Laboratory. This facility, in conjunction with existing miniature mixer-settler equipment and the centrifugal contactor facility currently under construction at SRL, provides capability for cross comparison of solvent extraction technology. This presentation describes the design and applications of the Pulse Column Facility at SRL

  9. Waste Facilities

    Data.gov (United States)

    Vermont Center for Geographic Information — This dataset was developed from the Vermont DEC's list of certified solid waste facilities. It includes facility name, contact information, and the materials...

  10. Health Facilities

    Science.gov (United States)

    Health facilities are places that provide health care. They include hospitals, clinics, outpatient care centers, and specialized care centers, ... psychiatric care centers. When you choose a health facility, you might want to consider How close it ...

  11. Fabrication Facilities

    Data.gov (United States)

    Federal Laboratory Consortium — The Fabrication Facilities are a direct result of years of testing support. Through years of experience, the three fabrication facilities (Fort Hood, Fort Lewis, and...

  12. TiO2/SiO2 prepared via facile sol-gel method as an ideal support for green synthesis of Ag nanoparticles using Oenothera biennis extract and their excellent catalytic performance in the reduction of 4-nitrophenol

    Directory of Open Access Journals (Sweden)

    Bahar Khodadadi

    2017-01-01

    Full Text Available In the present study, the extract of the plant of Oenothera biennis was used to green synthesis of silver nanoparticles (Ag NPs as an environmentally friendly, simple and low cost method. And Additionally, TiO2/SiO2 was prepared via facile sol-gel method using starch as an important, naturally abundant organic polymer as an ideal support. The Ag NPs/TiO2/SiO2 as an effective catalyst was prepared through reduction of Ag+ ions using Oenothera biennis extract as the reducing and stabilizing agent and Ag NPs immobilization on TiO2/SiO2 surface in the absence of any stabilizer or surfactant. Several techniques such as FT-IR spectroscopy, UV-Vis spectroscopy, X-ray Diffraction (XRD, sScanning eElectron mMicroscopy (FE-SEM, Eenergy dDispersive X-ray sSpectroscopy (EDS, and Ttransmission Eelectron Mmicroscopy (TEM were used to characterize TiO2/SiO2, silver nanoparticles (Ag NPs, and Ag NPs/TiO2/SiO2. Moreover, the catalytic activity of the Ag NPs/ TiO2/SiO2 was investigated in the reduction of 4-nitrophenol (4-NP at room temperature. On the basis of the results, the Ag NPs/TiO2/SiO2 was found to be high catalytic activity highly active catalyst according to the experimental results in this study. In addition, Ag NPs/TiO2/SiO2 can be recovered and reused several times in the reduction of 4-NP with no significant loss of catalytic activity.

  13. Development of polymer-extractant composite beads for separation of radionuclides

    International Nuclear Information System (INIS)

    Kumar, Manmohan; Singh, Krishankant; Bajaj, P.N.

    2009-01-01

    A novel micro porous polymer-extractant composite bead system, containing liquid extractant encapsulated in the core of a polymeric shell, with required porosity and hydrophilicity, to allow exchange of radionuclides without, any significant leaching out of the encapsulated extractant, has been developed for solid-liquid extraction of radionuclides from acidic waste solutions. The reuse of the beads is possible as there is practically no change in its radionuclide extraction efficiency, after repeated extraction second time. The high porosity and hydrophilicity of the synthesized TBP-encapsulated polymeric beads is evident from the presence of ∼ 78% water in the swollen condition. Evaluation of the synthesized beads for extraction of uranium and plutonium from aqueous acidic waste solutions, indicated possibility of their use under the conditions of PUREX process. The aliquate 336-encapsulated polymeric beads showed selective extraction of plutonium from aqueous nitrate solutions in the presence of uranium, and back extraction of the loaded plutonium, using dilute nitric acid or ascorbic acid. (author)

  14. CLIC Test Facility 3

    CERN Multimedia

    Kossyvakis, I; Faus-golfe, A

    2007-01-01

    The design of CLIC is based on a two-beam scheme, where short pulses of high power 30 GHz RF are extracted from a drive beam running parallel to the main beam. The 3rd generation CLIC Test Facility (CTF3) will demonstrate the generation of the drive beam with the appropriate time structure, the extraction of 30 GHz RF power from this beam, as well as acceleration of a probe beam with 30 GHz RF cavities. The project makes maximum use of existing equipment and infrastructure of the LPI complex, which became available after the closure of LEP.

  15. Animal facilities

    International Nuclear Information System (INIS)

    Fritz, T.E.; Angerman, J.M.; Keenan, W.G.; Linsley, J.G.; Poole, C.M.; Sallese, A.; Simkins, R.C.; Tolle, D.

    1981-01-01

    The animal facilities in the Division are described. They consist of kennels, animal rooms, service areas, and technical areas (examining rooms, operating rooms, pathology labs, x-ray rooms, and 60 Co exposure facilities). The computer support facility is also described. The advent of the Conversational Monitor System at Argonne has launched a new effort to set up conversational computing and graphics software for users. The existing LS-11 data acquisition systems have been further enhanced and expanded. The divisional radiation facilities include a number of gamma, neutron, and x-ray radiation sources with accompanying areas for related equipment. There are five 60 Co irradiation facilities; a research reactor, Janus, is a source for fission-spectrum neutrons; two other neutron sources in the Chicago area are also available to the staff for cell biology studies. The electron microscope facilities are also described

  16. The separation of extractants implemented in the DIAMEX-SANEX process

    International Nuclear Information System (INIS)

    Heres, Xavier; Baron, P.; Hill, C.; Ameil, E.; Martinez, I.; Rivalier, P.

    2008-01-01

    DIAMEX-SANEX is a process developed at the Cea to recover selectively the actinides(III) after a COEX TM or a PUREX process, in order to significantly decrease the radiotoxicity of the ultimate waste produced by the nuclear industry. This liquid-liquid extraction process is based on the DIAMEX process, using a malonamide supplemented by an acidic extractant. Besides an actinide extraction step and a lanthanide stripping step are implemented an actinide(III) stripping step and an extractant splitting step. The latter is carried out to avoid interactions between these two extractants during the first co-extraction step of the actinides and the lanthanides. This paper gives some results obtained with di-n-hexyl phosphoric acid (HDHP), which fulfills the required criteria for process development. Batch experiments or cold counter-current tests showed that it is possible to separate this extractant from DMDOHEMA. HDHP can moreover maintain the lanthanides(III) in the organic phase when the actinides(III) are back extracted from the organic phase. (authors)

  17. The separation of extractants implemented in the DIAMEX-SANEX process

    Energy Technology Data Exchange (ETDEWEB)

    Heres, Xavier [CEA-Marcoule, DEN/MAR/DRCP/SCPS, BP 17171, 30207 Bagnols-sur-Ceze Cedex (France); Baron, P.; Hill, C.; Ameil, E.; Martinez, I. [CEA-Marcoule, DEN/MAR/DRCP/SCPS, BP 17171, 30207 Bagnols-sur-Ceze Cedex (France); Rivalier, P. [CEA-Marcoule, DEN/MAR/DTEC/SGCS, BP 17171, 30207 Bagnols-sur-Ceze Cedex (France)

    2008-07-01

    DIAMEX-SANEX is a process developed at the Cea to recover selectively the actinides(III) after a COEX{sup TM} or a PUREX process, in order to significantly decrease the radiotoxicity of the ultimate waste produced by the nuclear industry. This liquid-liquid extraction process is based on the DIAMEX process, using a malonamide supplemented by an acidic extractant. Besides an actinide extraction step and a lanthanide stripping step are implemented an actinide(III) stripping step and an extractant splitting step. The latter is carried out to avoid interactions between these two extractants during the first co-extraction step of the actinides and the lanthanides. This paper gives some results obtained with di-n-hexyl phosphoric acid (HDHP), which fulfills the required criteria for process development. Batch experiments or cold counter-current tests showed that it is possible to separate this extractant from DMDOHEMA. HDHP can moreover maintain the lanthanides(III) in the organic phase when the actinides(III) are back extracted from the organic phase. (authors)

  18. Improved Purex solvent scrubbing methods

    International Nuclear Information System (INIS)

    Mailen, J.C.; Tallent, O.K.

    1984-01-01

    Studies of hydrazine and hydroxylamine salts as solvent scrubbing agents that can be decomposed into gases are summarized. Results from testing of countercurrent scrubbers and solid sorber columns that produce lesser amounts of permanent salts are reported. The status of studies of the acid-degradation of paraffin diluent and the options for removal of long-chain organic acids is given

  19. Facilities Programming.

    Science.gov (United States)

    Bullis, Robert V.

    1992-01-01

    A procedure for physical facilities management written 17 years ago is still worth following today. Each of the steps outlined for planning, organizing, directing, controlling, and evaluating must be accomplished if school facilities are to be properly planned and constructed. However, lessons have been learned about energy consumption and proper…

  20. Nuclear facilities

    International Nuclear Information System (INIS)

    Anon.

    2000-01-01

    Here is given the decree (2000-1065) of the 25. of October 2000 reporting the publication of the convention between the Government of the French Republic and the CERN concerning the safety of the LHC (Large Hadron Collider) and the SPS (Proton Supersynchrotron) facilities, signed in Geneva on July 11, 2000. By this convention, the CERN undertakes to ensure the safety of the LHC and SPS facilities and those of the operations of the LEP decommissioning. The French legislation and regulations on basic nuclear facilities (concerning more particularly the protection against ionizing radiations, the protection of the environment and the safety of facilities) and those which could be decided later on apply to the LHC, SPS and auxiliary facilities. (O.M.)

  1. Partnew - New solvent extraction processes for minor actinides - final report

    International Nuclear Information System (INIS)

    Madic, C.; Testard, F.; Hudson, M.J.; Liljenzin, J.O.; Christiansen, B.; Ferrando, M.; Facchini, A.; Geist, A.; Modolo, G.; Gonzalez-Espartero, A.; Mendoza, J. de

    2004-01-01

    The objectives of the European project PARTNEW were to define solvent extraction processes for the partitioning of the minor actinides, Am and Cm, from the aqueous high active raffinate or high active concentrate issuing the reprocessing of nuclear spent fuels by the PUREX process. Eleven laboratories participated to the research: 1/ CEA-DEN (Marcoule), 2/ CEA-DSM (Saclay), 3/ UREAD (U.K.), 4/ CTU (Sweden), 5/ ITU (Germany), 6/ ENEA (Italy), 7/ PoliMi (Italy), 8/ FZK-INE (Germany), 9/ FZJ-ISR (Germany), 10/ CIEMAT (Spain) and 11/ UAM (Spain). The research was organised into eight work packages (WP): Basic and applied DIAMEX studies, using diamide extractants for the co-extraction of actinides(III) (An(III)) and lanthanides(III) (Ln(III)) nitrates (WP1 and WP2), Basic and applied SANEX studies based on the use of polydentate N-ligands for the An(III)/Ln(III) separation (WP3 and WP4), Basic and applied SANEX studies based on the use of synergistic mixtures made of bis-(chloro-phenyl)-di-thio-phosphinic acid + neutral O-bearing ligand, (WP5 and WP6), Basic SANEX studies for the An(III)/Ln(III) separation, based on the use of new S-bearing ligands, Basic and applied studies for the Am(III)/Cm(III) separation. The work done in the fundamental and applied domains was very fruitful. Several processes have been successfully tested with genuine high active raffinates and concentrate. (authors)

  2. Liquid-solid extraction of cationic metals by cationic amphiphiles

    International Nuclear Information System (INIS)

    Muller, W.

    2010-01-01

    In the field of selective separation for recycling of spent nuclear fuel, liquid-liquid extraction processes are widely used (PUREX, DIAMEX..) in industrial scale. In order to guarantee a sustainable nuclear energy for the forthcoming generations, alternative reprocessing techniques are under development. One of them bases on the studies from Heckmann et al in the 80's and consists in selectively precipitating actinides from aqueous waste solutions by cationic surfactants (liquid-solid extraction). This technique has some interesting advantages over liquid-liquid extraction techniques, because several steps are omitted like stripping or solvent washing. Moreover, the amount of waste is decreased considerably, since no contaminated organic solvent is produced. In this thesis, we have carried out a physico-chemical study to understand the specific interactions between the metallic cations with the cationic surfactant. First, we have analysed the specific effect of the different counter-ions (Cl - , NO 3 - , C 2 O 4 2- ) and then the effect of alkaline cations on the structural properties of the surfactant aggregation in varying thermodynamical conditions. Finally, different multivalent cations (Cu 2+ , Zn 2+ , UO 2 2+ , Fe 3+ , Nd 3+ , Eu 3+ , Th 4+ ) were considered; we have concluded that depending on the anionic complex of these metals formed in acidic media, we can observe either an adsorption at the micellar interface or not. This adsorption has a large influence of the surfactant aggregation properties and determines the limits of the application in term of ionic strength, temperature and surfactant concentration. (author) [fr

  3. Empirical equations of the solvent extraction of the energetic inputs, uranium and plutonium, calculated by using the program Microsoft Excel

    International Nuclear Information System (INIS)

    Bento, Dercio Lopes

    2006-01-01

    PUREX is one of the purification process for irradiated nuclear fuel. In the flowchart the program uses various uranium and plutonium extraction phases by using organic solvent contained in the aqueous phase obtained in the dissolution of the fuel element. A posterior extraction U and Pu are changed to the aqueous phase. So it is fundamental to know the distribution coefficient (dS), at the temperature (tc), of the substances among the two immiscible phases, for better calculation the suitable flowchart. A mathematical model was elaborated based on experimental data, for the calculation of the dS and applied to a referential band of substance concentrations in the aqueous phase (xS) and organic (yS). By using the program Excel, we personalized the empirical equations calculated by the root mean square. The relative deviation, among the calculated values and the experimental ones are the standards

  4. Chemical process developments in reprocessing from 1965--1975 in the Institute for Hot Chemistry

    International Nuclear Information System (INIS)

    Baumgaertner, F.

    Work on the aqueous reprocessing of fuels is described. The following are discussed: LABEX (laboratory-scale extraction), MILLI facility (1 kg/day), problems of aqueous reprocessing, centrifugal extractor development, radiolytic products from Purex process, and TAMARA facility. Results of the MILLI operation are reviewed. Solutions to problems are discussed

  5. Mammography Facilities

    Data.gov (United States)

    U.S. Department of Health & Human Services — The Mammography Facility Database is updated periodically based on information received from the four FDA-approved accreditation bodies: the American College of...

  6. Relativistic heavy ion facilities: worldwide

    International Nuclear Information System (INIS)

    Schroeder, L.S.

    1986-05-01

    A review of relativistic heavy ion facilities which exist, are in a construction phase, or are on the drawing boards as proposals is presented. These facilities span the energy range from fixed target machines in the 1 to 2 GeV/nucleon regime, up to heavy ion colliders of 100 GeV/nucleon on 100 GeV/nucleon. In addition to specifying the general features of such machines, an outline of the central physics themes to be carried out at these facilities is given, along with a sampling of the detectors which will be used to extract the physics. 22 refs., 17 figs., 3 tabs

  7. Modeling of Pu(IV) extraction and HNO3 speciation in nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    De-Sio, S.

    2012-01-01

    The PUREX process is a solvent extraction method dedicated to the reprocessing of irradiated nuclear fuel in order to recover pure uranium and plutonium from aqueous solutions of concentrated nitric acid. The tri-n-butylphosphate (TBP) is used as the extractant in the organic phase. The aim of this thesis work was to improve the modeling of liquid-liquid extraction media in nuclear fuel reprocessing. First, Raman and 14 N NMR measurements, coupled with theoretical calculations based on simple solutions theory and BIMSA modeling, were performed in order to get a better understanding of nitric acid dissociation in binary and ternary solutions. Then, Pu(IV) speciation in TBP after extraction from low nitric acid concentrations was investigated by EXAFS and vis-NIR spectroscopies. We were able to show evidence of the extraction of Pu(IV) hydrolyzed species into the organic phase. A new structural study was conducted on An(VI)/TBP and An(IV)/TBP complexes by coupling EXAFS measurements with DFT calculations. Finally, extraction isotherms modeling was performed on the Pu(IV)/HNO 3 /H 2 O/TBP 30%/dodecane system (with Pu at tracer scale) by taking into account deviation from ideal behaviour in both organic and aqueous phases. The best modeling was obtained when considering three plutonium (IV) complexes in the organic phase: Pu(OH) 2 (NO 3 ) 2 (TBP) 2 , Pu(NO 3 ) 4 (TBP) 2 and Pu(NO 3 ) 4 (TBP) 3 . (author) [fr

  8. The method for simultaneous extraction and back extraction in liquid three-phase system and equipment for simultaneous extraction and back extraction in liquid three-phase system

    International Nuclear Information System (INIS)

    Palyska, W.; Chmielewski, A.G.

    1992-01-01

    The method for simultaneous extraction and back extraction in liquid three-phase system has been worked out. The equipment designed for that process has been also subject of the patent. The interesting component is extracted first to intermediate phase consists of magnetic solvent keeping two extracting phases separately. The intermediate magnetic liquid has been kept in its position using a stable magnet maintained on the surface of the extraction vessel. Then the component pass from intermediate phase to the third phase as a result of back extraction. Mixing in the extraction and back extraction zones is organized by means of rotating shaft going along the whole apparatus. The extraction and back extraction processes occur simultaneously as a result of continuous flow of solvent in their zones. The single extraction back extraction facilities can be joined in larger batteries. 3 figs

  9. Extraction process

    International Nuclear Information System (INIS)

    Rendall, J.S.; Cahalan, M.J.

    1979-01-01

    A process is described for extracting at least two desired constituents from a mineral, using a liquid reagent which produces the constituents, or compounds thereof, in separable form and independently extracting those constituents, or compounds. The process is especially valuable for the extraction of phosphoric acid and metal values from acidulated phosphate rock, the slurry being contacted with selective extractants for phosphoric acid and metal (e.g. uranium) values. In an example, uranium values are oxidized to uranyl form and extracted using an ion exchange resin. (U.K.)

  10. Contribution of molecular modeling and of structure-activity relations to the liquid-liquid extraction. Application to the case of U(VI) extraction by monoamides

    International Nuclear Information System (INIS)

    Rabbe, C.

    1996-01-01

    In France, spent fuels are in most cases reprocessed. The aim of the reprocessing is to separate the recyclable fissile materials (for instance, uranium and plutonium) of radioactive wastes. The industrial process used until now is the Purex (Plutonium Uranium Refining by EXtraction) process. Recently (1991), the CEA has undertaken researches on the fields of separation and transmutation of long-lived radionuclides as minor actinides. Some molecules with an amide function have been at first considered especially for the uranium extraction. In order to rationalize the research of new extracting molecules, some molecular modeling methods (quantum chemistry calculations, molecular mechanics) have been used. In fact, there are three determining parameters for a molecule to be a good extractant: it has to own: 1) one or several sites which present a sufficient electron density in order that the metallic cation be complexed 2) the smallest possible substituents to avoid interferences with the complexation 3) a sufficient lipophilic effect. (O.M.). 139 refs., 43 figs., 36 tabs

  11. Solvent extraction

    Energy Technology Data Exchange (ETDEWEB)

    Coombs, D.M.; Latimer, E.G.

    1988-01-05

    It is an object of this invention to provide for the demetallization and general upgrading of heavy oil via a solvent extracton process, and to improve the efficiency of solvent extraction operations. The yield and demetallization of product oil form heavy high-metal content oil is maximized by solvent extractions which employ either or all of the following techniques: premixing of a minor amount of the solvent with feed and using countercurrent flow for the remaining solvent; use of certain solvent/free ratios; use of segmental baffle tray extraction column internals and the proper extraction column residence time. The solvent premix/countercurrent flow feature of the invention substantially improves extractions where temperatures and pressures above the critical point of the solvent are used. By using this technique, a greater yield of extract oil can be obtained at the same metals content or a lower metals-containing extract oil product can be obtained at the same yield. Furthermore, the premixing of part of the solvent with the feed before countercurrent extraction gives high extract oil yields and high quality demetallization. The solvent/feed ratio features of the invention substanially lower the captial and operating costs for such processes while not suffering a loss in selectivity for metals rejection. The column internals and rsidence time features of the invention further improve the extractor metals rejection at a constant yield or allow for an increase in extract oil yield at a constant extract oil metals content. 13 figs., 3 tabs.

  12. Support facilities

    International Nuclear Information System (INIS)

    Williamson, F.S.; Blomquist, J.A.; Fox, C.A.

    1977-01-01

    Computer support is centered on the Remote Access Data Station (RADS), which is equipped with a 1000 lpm printer, 1000 cpm reader, and a 300 cps paper tape reader with 500-foot spools. The RADS is located in a data preparation room with four 029 key punches (two of which interpret), a storage vault for archival magnetic tapes, card files, and a 30 cps interactive terminal principally used for job inquiry and routing. An adjacent room provides work space for users, with a documentation library and a consultant's office, plus file storage for programs and their documentations. The facility has approximately 2,600 square feet of working laboratory space, and includes two fully equipped photographic darkrooms, sectioning and autoradiographic facilities, six microscope cubicles, and five transmission electron microscopes and one Cambridge scanning electron microscope equipped with an x-ray energy dispersive analytical system. Ancillary specimen preparative equipment includes vacuum evaporators, freeze-drying and freeze-etching equipment, ultramicrotomes, and assorted photographic and light microscopic equipment. The extensive physical plant of the animal facilities includes provisions for holding all species of laboratory animals under controlled conditions of temperature, humidity, and lighting. More than forty rooms are available for studies of the smaller species. These have a potential capacity of more than 75,000 mice, or smaller numbers of larger species and those requiring special housing arrangements. There are also six dog kennels to accommodate approximately 750 dogs housed in runs that consist of heated indoor compartments and outdoor exercise areas

  13. Design of a hot pilot plant facility for demonstration of the pot calcination process

    Energy Technology Data Exchange (ETDEWEB)

    Buckham, J A

    1962-01-01

    A facility was designed for demonstration of the pot calcination process with wastes from processing aluminum alloy fuels, Darex or electrolytic processing of stainless-steel fuels, and Purex processes. This facility will also permit determination of procedures required for economical production of low-porosity, relatively nonleachable materials by addition of suitable reagents to the wastes fed to the calciner. The process consists of concentration by evaporation and thermal decomposition in situ in pots which also serve as the final disposal containers. This unit permits determination of pot loading and density, leachability, melting point, volatile material content, heat release, and thermal conductivity of the calcine. Also to be determined are transient calcine temperature distributions, fission product behavior during calcination, deentrainment obtained in the various parts of the system, decontamination achieved on all liquid and gaseous effluent streams, need for venting of stored pots, optimum means of remotely sealing the pots, and methods required for production of a minimum volume of noncondensable off-gas. This facility will employ nominal full-scale pots 8 and 12 in. in diameter and 8 ft long. A unique evaporator design was evolved to permit operation either with close-coupled continuous feed preparation or with bath feed preparation. Provisions were made to circumvent possible explosions due to organic material in feed solutions and other suspected hazards.

  14. Pilot studies of an extraction process for reprocessing of spent fuel from fast reactors: Hardware and process details of extractor selection

    International Nuclear Information System (INIS)

    Anisimov, V.I.; Pavlovich, V.B.; Smetanin, E.Ya.; Glazunov, N.V.; Shklyar, L.I.; Dubrovskii, V.G.; Serov, A.V.; Zakharkin, B.S.; Konorchenko, V.D.; Korotkov, I.A.; Neumoev, N.V.; Renard, E.V.

    1992-01-01

    While acknowledging the bold and persistent efforts of U.S. and Russian specialists to develop the concept of pyrochemical reprocessing of spent nuclear fuel from fast reactors on remote-controlled equipment for removal of actinides from the fission products one should recognize that the tasks of reprocessing such fuel can be handled only by using water-extraction technology, especially since the known Purex process continues to be improved to the point that a single-cycle scheme may be developed. This article presents results of pilot studies conducted in hot cells using multistage extractors in continuous counterflow operation; data on various extractor types used in reprocessing spent mixed oxide nuclear fuel; advantages and disadvantages of centrifugal and pulsed column extractor; comparison of column-type and centrifugal extractors; and extraction process

  15. Extraction method

    International Nuclear Information System (INIS)

    Stary, J.; Kyrs, M.; Navratil, J.; Havelka, S.; Hala, J.

    1975-01-01

    Definitions of the basic terms and of relations are given and the knowledge is described of the possibilities of the extraction of elements, oxides, covalent-bound halogenides and heteropolyacids. Greatest attention is devoted to the detailed analysis of the extraction of chelates and ion associates using diverse agents. For both types of compounds detailed conditions are given of the separation and the effects of the individual factors are listed. Attention is also devoted to extractions using mixtures of organic agents, the synergic effects thereof, and to extractions in non-aqueous solvents. The effects of radiation on extraction and the main types of apparatus used for extractions carried out in the laboratory are described. (L.K.)

  16. Facile synthesis of new nano sorbent for magnetic solid-phase extraction by self assembling of bis-(2,4,4-trimethyl pentyl)-dithiophosphinic acid on Fe3O4-Ag core-shell nanoparticles: Characterization and application

    International Nuclear Information System (INIS)

    Tahmasebi, Elham; Yamini, Yadollah

    2012-01-01

    Graphical abstract: Self assembling of bis-(2,4,4-trimethylpentyl)-dithiophosphinic acid on Fe 3 O 4 -Ag core-shell nanoparticles and application of it for solid phase extraction of PAHs. Highlights: ► A novel sorbent for magnetic solid-phase extraction of PAHs was introduced. ► Silver was coated on Fe 3 O 4 nanoparticles (MNPs) by reduction of AgNO 3 with NaBH 4 . ► Bis-(2,4,4-trimethylpentyl)-dithiophosphinic acid self-assembled on silver coated MNPs. ► Size, morphology, composition and properties of the nanoparticles were characterized. ► Extraction efficiency of the sorbent was investigated by extraction of five PAHs. - Abstract: A novel sorbent for magnetic solid-phase extraction by self-assembling of organosulfur compound, (bis-(2,4,4-trimethylpentyl)-dithiophosphinic acid), onto the silver-coated Fe 3 O 4 nanoparticles was introduced. Due to the formation of covalent bond of S-Ag, the new coating on the silver surface was very stable and showed high thermal stability (up to 320 °C). The size, morphology, composition, and properties of the prepared nanoparticles have also been characterized and determined using scanning electron microscopy (SEM), energy-dispersive X-ray analyzer (EDX), dynamic light scattering (DLS), Fourier transform-infrared (FT-IR) spectroscopy, and thermal gravimetric analysis (TGA). Extraction efficiency of the new sorbent was investigated by extraction of five polycyclic aromatic hydrocarbons (PAHs) as model compounds. The optimum extraction conditions for PAHs were obtained as of extraction time, 20 min; 50 mg sorbent from 100 mL of the sample solution, and elution with 100 μL of 1-propanol under fierce vortex for 2 min. Under the optimal conditions, the calibration curves were obtained in the range of 0.05–100 μg L −1 (R 2 > 0.9980) and the LODs (S/N = 3) were obtained in the range of 0.02–0.10 μg L −1 . Relative standard deviations (RSDs) for intra- and inter-day precision were 2.6–4.2% and 3.6–8

  17. Determination of species activities in organic phase. Modelling of liquid-liquid extraction system using uniquac and unifac models; Determination des activites des especes en phase organique. Application d`uniquac et unifac a la modelisation des systemes d`extraction liquide-liquide

    Energy Technology Data Exchange (ETDEWEB)

    Rat, B. [CEA Saclay, 91 - Gif-sur-Yvette (France). Dept. de Recherche en Retraitement et en Vitrification]|[Paris-6 Univ., 75 (France)

    1998-12-31

    The aim of nuclear fuel reprocessing is to separate reusable elements, uranium and plutonium from the other elements, fission products and minor actinides. PUREX process uses liquid-liquid extraction as separation method. Numerical codes for modelling the extraction operations of PUREX process use a semi-empirical model to represent the partition of species. To improve the precision and precision and predictive nature of the models, we looked for a theoretical tool which permits to quantify medium effects, especially in the organic phase, for which few models are available. The Sergeivskii-Dannus model permits to quantify deviations from ideality in organic phase equilibrated with aqueous phase, but with parameters depending on extractant/diluent ratio. We decided to investigate UNIQUAC and UNIFAC models which permit to estimate activity coefficients in non-electrolytic phases taking account of the mutual interactions of molecules and their morphology. UNIFAC is based on UNIQUAC but molecules are considered as structural groups assemblies. Before applying these model to extraction systems, we investigate their abilities to describe simple systems, binary and ternary systems. UNIQUAC has been applied to TBP/diluent mixtures and permits to estimate activity coefficients for diluents whose interactions with TPB are very different in nature and strength. Group contribution (UNIFAC) applied to TBP/alkane mixtures permits to represent the effect of lengthening alkane chain but not the effect of branching. UNIQUAC fails to describe the TBP/diluent/water/non-extractable-salt systems in case of strong TBP diluent interactions. In order to obtain a correct description of these systems, we used the Chem-UNIFAC model, where the INIFAC equation is supplemented with chemical equilibria allowing explicitly for complexes formation and where group contribution is used to describes complexes. We have with Chem-UNIFAC a model available which can take the effect of the diluent into

  18. Emission Facilities - Erosion & Sediment Control Facilities

    Data.gov (United States)

    NSGIC Education | GIS Inventory — An Erosion and Sediment Control Facility is a DEP primary facility type related to the Water Pollution Control program. The following sub-facility types related to...

  19. Contribution of molecular modeling and of structure-activity relations to the liquid-liquid extraction. Application to the case of U(VI) extraction by monoamides; Apport de la modelisation moleculaire et des relations structure -activite a l`extraction liquide-liquide. Application au cas de l`extraction d`U(VI) par les monoamides

    Energy Technology Data Exchange (ETDEWEB)

    Rabbe, C.

    1996-06-07

    In France, spent fuels are in most cases reprocessed. The aim of the reprocessing is to separate the recyclable fissile materials (for instance, uranium and plutonium) of radioactive wastes. The industrial process used until now is the Purex (Plutonium Uranium Refining by EXtraction) process. Recently (1991), the CEA has undertaken researches on the fields of separation and transmutation of long-lived radionuclides as minor actinides. Some molecules with an amide function have been at first considered especially for the uranium extraction. In order to rationalize the research of new extracting molecules, some molecular modeling methods (quantum chemistry calculations, molecular mechanics) have been used. In fact, there are three determining parameters for a molecule to be a good extractant: it has to own: (1) one or several sites which present a sufficient electron density in order that the metallic cation be complexed (2) the smallest possible substituents to avoid interferences with the complexation (3) a sufficient lipophilic effect. (O.M.). 139 refs., 43 figs., 36 tabs.

  20. Vacuum extraction

    DEFF Research Database (Denmark)

    Maagaard, Mathilde; Oestergaard, Jeanett; Johansen, Marianne

    2012-01-01

    Objectives. To develop and validate an Objective Structured Assessment of Technical Skills (OSATS) scale for vacuum extraction. Design. Two-part study design: Primarily, development of a procedure-specific checklist for vacuum extraction. Hereafter, validation of the developed OSATS scale for vac...

  1. Electromembrane extraction

    DEFF Research Database (Denmark)

    Huang, Chuixiu; Chen, Zhiliang; Gjelstad, Astrid

    2017-01-01

    Electromembrane extraction (EME) was inspired by solid-phase microextraction and developed from hollow fiber liquid-phase microextraction in 2006 by applying an electric field over the supported liquid membrane (SLM). EME provides rapid extraction, efficient sample clean-up and selectivity based...

  2. Partnew - New solvent extraction processes for minor actinides - final report; Partnew - Nouveaux procedes d'extraction par solvant pour les actinides mineurs - rapport final

    Energy Technology Data Exchange (ETDEWEB)

    Madic, C.; Testard, F.; Hudson, M.J.; Liljenzin, J.O.; Christiansen, B.; Ferrando, M.; Facchini, A.; Geist, A.; Modolo, G.; Gonzalez-Espartero, A.; Mendoza, J. de

    2004-07-01

    The objectives of the European project PARTNEW were to define solvent extraction processes for the partitioning of the minor actinides, Am and Cm, from the aqueous high active raffinate or high active concentrate issuing the reprocessing of nuclear spent fuels by the PUREX process. Eleven laboratories participated to the research: 1/ CEA-DEN (Marcoule), 2/ CEA-DSM (Saclay), 3/ UREAD (U.K.), 4/ CTU (Sweden), 5/ ITU (Germany), 6/ ENEA (Italy), 7/ PoliMi (Italy), 8/ FZK-INE (Germany), 9/ FZJ-ISR (Germany), 10/ CIEMAT (Spain) and 11/ UAM (Spain). The research was organised into eight work packages (WP): Basic and applied DIAMEX studies, using diamide extractants for the co-extraction of actinides(III) (An(III)) and lanthanides(III) (Ln(III)) nitrates (WP1 and WP2), Basic and applied SANEX studies based on the use of polydentate N-ligands for the An(III)/Ln(III) separation (WP3 and WP4), Basic and applied SANEX studies based on the use of synergistic mixtures made of bis-(chloro-phenyl)-di-thio-phosphinic acid + neutral O-bearing ligand, (WP5 and WP6), Basic SANEX studies for the An(III)/Ln(III) separation, based on the use of new S-bearing ligands, Basic and applied studies for the Am(III)/Cm(III) separation. The work done in the fundamental and applied domains was very fruitful. Several processes have been successfully tested with genuine high active raffinates and concentrate. (authors)

  3. Reprocessing of spent nuclear fuel; Prerada isluzenog nuklearnog goriva

    Energy Technology Data Exchange (ETDEWEB)

    Gal, I [Institute of Nuclear Sciences Boris Kidric, Laboratorija za hemiju visoke aktivnosti, Vinca, Beograd (Serbia and Montenegro)

    1964-12-15

    This volume contains the following reports: Experimental facility for testing and development of pulsed columns and auxiliary devices; Chemical-technology study of the modified 'Purex' process; Chemical and radiometric control analyses; Chromatographic separation of rare earth elements on paper treated by di-n butylphosphate; Preliminary study of some organic nitrogen extracts significant in fuel reprocessing.

  4. Reprocessing of spent nuclear fuel

    International Nuclear Information System (INIS)

    Gal, I.

    1964-12-01

    This volume contains the following reports: Experimental facility for testing and development of pulsed columns and auxiliary devices; Chemical-technology study of the modified 'Purex' process; Chemical and radiometric control analyses; Chromatographic separation of rare earth elements on paper treated by di-n butylphosphate; Preliminary study of some organic nitrogen extracts significant in fuel reprocessing

  5. Air Quality Facilities

    Data.gov (United States)

    Iowa State University GIS Support and Research FacilityFacilities with operating permits for Title V of the Federal Clean Air Act, as well as facilities required to submit an air emissions inventory, and other facilities...

  6. Identification of organic iodocompounds in the PUREX process with the help of methods for chromatographic separation and spectrometric detection as well as characterization of their behavior during extraction. Final report

    International Nuclear Information System (INIS)

    Gaul, G.; Gibau, F.; Knoechel, A.

    1993-01-01

    In the system HNO 3 KI, Dodecan and TBP the radiolytic reactive behaviour of the described compounds during the dissolution of nuclear fuel elements was simulated. External γ-irradiation gave the best informations. As a consequence of radiolyticly induced reactions several volatile and non-volatile iodoorganic compounds like iodoalkanes, iodonitroalkanes and iodonitroalkylphosphates are formed. They were separated by gaschromatography (GC) and high performance liquid chromatography (HPLC). For HPLC a special photolytic/electrochemical detector with comparable sensitivity like the electroncapture detector in the GC-field was developed. With the help of the two described chromatographic techniques the different iodoorganic compounds could be separated singularily or in groups of isomers. The separation of all compounds demands two-dimensional chromatographies including the capillary-SFC, which was not available for this project. Most of the iodoorganic compounds could be identified by comparison of the retention times of well known compounds. In the other cases, the compounds were studied mass-spectrometrically. Unfortunately all available ionisation techniques (EI; CI; FAB) were too hard for the labile C-I-bond. Therefore an identification of these compounds was not possible. In any case, instructive fingerprint spectra are available enabling relationships between the generation of the iodoorganic compounds and the reaction conditions during their formation. (orig.) [de

  7. Bevalac extraction

    International Nuclear Information System (INIS)

    Kalnins, J.G.; Krebs, G.; Tekawa, M.; Cowles, D.; Byrne, T.

    1992-02-01

    This report will describe some of the general features of the Bevatron extraction system, primarily the dependence of the beam parameters and extraction magnet currents on the Bevalac field. The extraction magnets considered are: PFW, XPl, XP2, XS1, XS2, XM1, XM2, XM3, XQ3A and X03B. This study is based on 84 past tunes (from 1987 to the present) of various ions (p,He,O,Ne,Si,S,Ar,Ca,Ti,Fe,Nb,La,Au and U), for Bevalac fields from 1.749 to 12.575 kG, where all tunes included a complete set of beam line wire chamber pictures. The circulating beam intensity inside the Bevalac is measured with Beam Induction Electrodes (BIE) in the South Tangent Tank. The extracted beam intensity is usually measured with the Secondary Emission Monitor (SEM) in the F1-Box. For most of the tunes the extraction efficiency, as given by the SEM/BIE ratio, was not recorded in the MCR Log Book, but plotting the available Log Book data as a function of the Bevalac field, see Fig.9, we find that the extraction efficiency is typically between 30->60% with feedback spill

  8. Purex Plant comparison with 40 CFR 61, subpart H, and other referenced guidelines for the Product Removal (PR) (296-A-1) stack

    International Nuclear Information System (INIS)

    Lohrasbi, J.

    1994-08-01

    Dose calculations for atmospheric radionuclide releases from the Hanford Site for calendar year (CY) 1992 were performed by Pacific Northwest Laboratory (PNL) using the approved US Environmental Protection Agency (EPA) CAP-88 computer model. Emissions from discharge points in the Hanford Site 100, 200, 300, 400, and 600 areas were calculated based on results of analyses of continuous and periodic sampling conducted at the discharge points. These calculated emissions were provided for inclusion in the CAP-88 model by area and by individual facility for those facilities having the potential to contribute more than 10 percent of the Hanford Site total or to result in an impact of greater than 0.1 mrem per year to the maximally exposed individual (MEI). Also included in the assessment of offsite dose modeling are the measured radioactive emissions from all Hanford Site stacks that have routine monitoring performed. Record sampling systems have been installed on all stacks and vents that use exhaust fans to discharge air that potentially may carry airborne radioactivity. Estimation of activity from ingrowth of long-lived radioactive progeny is not included in the CAP-88 model; therefore, the Hanford Site GENII code (Napier et al. 1988) was used to supplement the CAP-88 dose calculations. When the dose to the MEI located in the Ringold area was calculated, the effective dose equivalent (EDE) from combined Hanford Site radioactive airborne emissions was shown to be 3.7E-03 mrem. This value was reported in the annual air emission report prepared for the Hanford Site (RL 1993)

  9. Reactor facility

    International Nuclear Information System (INIS)

    Suzuki, Hiroaki; Murase, Michio; Yokomizo, Osamu.

    1997-01-01

    The present invention provides a BWR type reactor facility capable of suppressing the amount of steams generated by the mutual effect of a failed reactor core and coolants upon occurrence of an imaginal accident, and not requiring spacial countermeasures for enhancing the pressure resistance of the container vessel. Namely, a means for supplying cooling water at a temperature not lower by 30degC than the saturated temperature corresponding to the inner pressure of the containing vessel upon occurrence of an accident is disposed to a lower dry well below the pressure vessel. As a result, upon occurrence of such an accident that the reactor core should be melted and flown downward of the pressure vessel, when cooling water at a temperature not lower than the saturated temperature, for example, cooling water at 100degC or higher is supplied to the lower dry well, abrupt generation of steams by the mutual effect of the failed reactor core and cooling water is scarcely caused compared with a case of supplying cooling water at a temperature lower than the saturation temperature by 30degC or more. Accordingly, the amount of steams to be generated can be suppressed, and special countermeasure is no more necessary for enhancing the pressure resistance of the container vessel is no more necessary. (I.S.)

  10. Nuclear facilities

    International Nuclear Information System (INIS)

    Anon.

    2002-01-01

    During September and October 2001, 15 events were recorded on the first grade and 1 on the second grade of the INES scale. The second grade event is in fact a re-classification of an incident that occurred on the second april 2001 at Dampierre power plant. This event happened during core refueling, a shift in the operation sequence led to the wrong positioning of 113 assemblies. A preliminary study of this event shows that this wrong positioning could have led, in other circumstances, to the ignition of nuclear reactions. Even in that case, the analysis made by EDF shows that the consequences on the staff would have been limited. Nevertheless a further study has shown that the existing measuring instruments could not have detected the power increase announcing the beginning of the chain reaction. The investigation has shown that there were deficiencies in the control of the successive operations involved in refueling. EDF has proposed a series of corrective measures to be implemented in all nuclear power plants. The other 15 events are described in the article. During this period 121 inspections have been made in nuclear facilities. (A.C.)

  11. National Ignition Facility site requirements

    International Nuclear Information System (INIS)

    1996-07-01

    The Site Requirements (SR) provide bases for identification of candidate host sites for the National Ignition Facility (NIF) and for the generation of data regarding potential actual locations for the facilities. The SR supplements the NIF Functional Requirements (FR) with information needed for preparation of responses to queries for input to HQ DOE site evaluation. The queries are to include both documents and explicit requirements for the potential host site responses. The Sr includes information extracted from the NIF FR (for convenience), data based on design approaches, and needs for physical and organization infrastructure for a fully operational NIF. The FR and SR describe requirements that may require new construction or may be met by use or modification of existing facilities. The SR do not establish requirements for NIF design or construction project planning. The SR document does not constitute an element of the NIF technical baseline

  12. Coupling a Transient Solvent Extraction Module with the Separations and Safeguards Performance Model

    Energy Technology Data Exchange (ETDEWEB)

    de Almeida, Valmor F [ORNL; Birdwell Jr, Joseph F [ORNL; DePaoli, David W [ORNL; Gauld, Ian C [ORNL

    2009-10-01

    A past difficulty in safeguards design for reprocessing plants is that no code existed for analysis and evaluation of the design. A number of codes have been developed in the past, but many are dated, and no single code is able to cover all aspects of materials accountancy, process monitoring, and diversion scenario analysis. The purpose of this work was to integrate a transient solvent extraction simulation module developed at Oak Ridge National Laboratory, with the SSPM Separations and Safeguards Performance Model, developed at Sandia National Laboratory, as a first step toward creating a more versatile design and evaluation tool. The SSPM was designed for materials accountancy and process monitoring analyses, but previous versions of the code have included limited detail on the chemical processes, including chemical separations. The transient solvent extraction model is based on the ORNL SEPHIS code approach to consider solute build up in a bank of contactors in the PUREX process. Combined, these capabilities yield a much more robust transient separations and safeguards model for evaluating safeguards system design. This coupling and the initial results are presented. In addition, some observations toward further enhancement of separations and safeguards modeling based on this effort are provided, including: items to be addressed in integrating legacy codes, additional improvements needed for a fully functional solvent extraction module, and recommendations for future integration of other chemical process modules.

  13. Extraction-wet oxidation process using sulphuric acid for treatment of TBP-dodecane wastes

    International Nuclear Information System (INIS)

    Deshingkar, D.S.; Kartha, P.K.S.

    1998-03-01

    In the nuclear fuel reprocessing plants, 30% n-tributyl phosphate in hydrocarbon diluent is used for extraction of uranium and plutonium from the spent fuel by Purex process. When TBP-dodecane can no longer be purified from its degradation products, it is discarded as alpha bearing, intermediate level wastes containing plutonium and ruthenium-106. To overcome shortcomings of extraction-pyrolysis and saponification processes, studies were undertaken to find the suitability of H 2 SO 4 as an alternative extractant for TBP. Oxidation of TBP to H 3 PO 4 using H 2 O 2 was also explored as H 3 PO 4 can be treated by known procedures for removal of plutonium and ruthenium-106. The experiments were conducted with aged spent solvent wastes discharged from reprocessing plant at Trombay using H 2 SO 4 and H 2 SO 4 - H 3 PO 4 mixture. The decontamination factors (DFs) for alpha activity were found to be satisfactory. The DFs for ruthenium were lower as compared to those obtained in experiments with simulated degraded waste. The gas chromatographic analysis of separated diluent revealed high branched alkane content and low n-dodecane content of separated diluent. It is very much different from that of diluent currently in use. Hence incineration of separated diluent is recommended. (author)

  14. Coupling a transient solvent extraction module with the separations and safeguards performance model.

    Energy Technology Data Exchange (ETDEWEB)

    DePaoli, David W. (Oak Ridge National Laboratory, Oak Ridge, TN); Birdwell, Joseph F. (Oak Ridge National Laboratory, Oak Ridge, TN); Gauld, Ian C. (Oak Ridge National Laboratory, Oak Ridge, TN); Cipiti, Benjamin B.; de Almeida, Valmor F. (Oak Ridge National Laboratory, Oak Ridge, TN)

    2009-10-01

    A number of codes have been developed in the past for safeguards analysis, but many are dated, and no single code is able to cover all aspects of materials accountancy, process monitoring, and diversion scenario analysis. The purpose of this work was to integrate a transient solvent extraction simulation module developed at Oak Ridge National Laboratory, with the Separations and Safeguards Performance Model (SSPM), developed at Sandia National Laboratory, as a first step toward creating a more versatile design and evaluation tool. The SSPM was designed for materials accountancy and process monitoring analyses, but previous versions of the code have included limited detail on the chemical processes, including chemical separations. The transient solvent extraction model is based on the ORNL SEPHIS code approach to consider solute build up in a bank of contactors in the PUREX process. Combined, these capabilities yield a more robust transient separations and safeguards model for evaluating safeguards system design. This coupling and initial results are presented. In addition, some observations toward further enhancement of separations and safeguards modeling based on this effort are provided, including: items to be addressed in integrating legacy codes, additional improvements needed for a fully functional solvent extraction module, and recommendations for future integration of other chemical process modules.

  15. ASACUSA facility

    CERN Multimedia

    Maximilien Brice

    2011-01-01

    Photo 1-6 : view of the RFQ - RFQ of the ASACUSA experiment. It allows to slow down antiprotons coming from the AD from 5 MeV to 100 KeV with high efficiency. -------------- Photo 7 - 16 : view of the TRAP - The ASACUSA Cusp trap. Thanks to its special magnetic field configuration, it enables the extraction of an anti-hydrogen beam, thus allowing a high precision microwave spectroscopy outside the magnetic field of the trap. This new method opens a new path to make a stringent test of CPT symmetry between matter and antimatter. #mypanoviewer { height:480px; width: 800px; margin:auto} var viewer=new PTGuiViewer(); viewer.setSwfUrl("/record/1331558/files/PTGuiViewer.swf"); viewer.preferFlashViewer(); viewer.setVars({ pano: "/record/1331558/files/panoA_", format: "14faces", pan: 0, minpan: -180, maxpan: 180, tilt:0, mintilt: -75.60468140442133, maxtilt: 75.60468140442133, fov: 90, minfov: 10, maxfov: 120, autorotatespeed: 5, autorotatedelay: 1...

  16. Assessment of proliferation resistances of aqueous reprocessing techniques using the TOPS methodology

    International Nuclear Information System (INIS)

    Åberg Lindell, M.; Grape, S.; Håkansson, A.; Jacobsson Svärd, S.

    2013-01-01

    Highlights: • Proliferation resistances of three possible LFR fuel cycles are assessed. • The TOPS methodology has been chosen for the PR assessment. • Reactor operation, reprocessing and fuel fabrication are examined. • Purex, Ganex, and a combination of Purex, Diamex and Sanex, are compared. • The safeguards analysis speaks in favor of Ganex as opposed to the Purex process. - Abstract: The aim of this study is to assess and compare the proliferation resistances (PR) of three possible Generation IV lead-cooled fast reactor fuel cycles, involving the reprocessing techniques Purex, Ganex and a combination of Purex, Diamex and Sanex, respectively. The examined fuel cycle stages are reactor operation, reprocessing and fuel fabrication. The TOPS methodology has been chosen for the PR assessment, and the only threat studied is the case where a technically advanced state diverts nuclear material covertly. According to the TOPS methodology, the facilities have been divided into segments, here roughly representing the different forms of nuclear material occurring in each examined fuel cycle stage. For each segment, various proliferation barriers have been assessed. The results make it possible to pinpoint where the facilities can be improved. The results show that the proliferation resistance of a fuel cycle involving recycling of minor actinides is higher than for the traditional Purex reprocessing cycle. Furthermore, for the purpose of nuclear safeguards, group actinide extraction should be preferred over reprocessing options where pure plutonium streams occur. This is due to the fact that a solution containing minor actinides is less attractive to a proliferator than a pure Pu solution. Thus, the safeguards analysis speaks in favor of Ganex as opposed to the Purex process

  17. Irradiation Facilities at CERN

    CERN Document Server

    Gkotse, Blerina; Carbonez, Pierre; Danzeca, Salvatore; Fabich, Adrian; Garcia, Alia, Ruben; Glaser, Maurice; Gorine, Georgi; Jaekel, Martin, Richard; Mateu,Suau, Isidre; Pezzullo, Giuseppe; Pozzi, Fabio; Ravotti, Federico; Silari, Marco; Tali, Maris

    2017-01-01

    CERN provides unique irradiation facilities for applications in many scientific fields. This paper summarizes the facilities currently operating for proton, gamma, mixed-field and electron irradiations, including their main usage, characteristics and information about their operation. The new CERN irradiation facilities database is also presented. This includes not only CERN facilities but also irradiation facilities available worldwide.

  18. Research Facilities | Wind | NREL

    Science.gov (United States)

    Research Facilities Research Facilities NREL's state-of-the-art wind research facilities at the Research Facilities Photo of five men in hard hards observing the end of a turbine blade while it's being tested. Structural Research Facilities A photo of two people silhouetted against a computer simulation of

  19. RETRACTED: Facile, eco-friendly and template free photosynthesis of cauliflower like ZnO nanoparticles using leaf extract of Tamarindus indica (L.) and its biological evolution of antibacterial and antifungal activities.

    Science.gov (United States)

    Elumalai, K; Velmurugan, S; Ravi, S; Kathiravan, V; Ashokkumar, S

    2015-02-05

    In the present investigation, we chose the very simple and eco-friendly chemical method for synthesis of zinc oxide nanoparticles from leaf extract of Tamarindus indica (L.) (T. indica) and developed the new green route for synthesis of nanoparticles. Formed product has been studied by UV-vis absorption spectroscopy, Photoluminescence (PL) spectroscopy, X-ray diffraction (XRD), Fourier transform infrared spectroscopy (FT-IR), Field emission scanning electron microscopy (FE-SEM) and with corresponding energy dispersive X-ray analysis (EDX). Mainly, the present results depicted that the synthesized nanoproducts are moderately stable, hexagonal phase, roughly spherical with maximum particles in size range within 19-37 nm in diameter. The antibacterial and fungal activities of aqueous extracts of T. indica were ended with corresponding disk diffusion and Minimum Inhibitory Concentration (MIC). The highest mean zones of inhibition were observed in the ZnO NPs (200 μg/mL) against Staphylococcus aureus (13.1±0.28). Finally, it can be concluded that microbial activity of ZnO NPs has more susceptible S. aureus than the other micro organisms. Further, the present investigation suggests that ZnO NPs has the potential applications for various medical and industrial fields. Copyright © 2014 Elsevier B.V. All rights reserved.

  20. North Slope, Alaska ESI: FACILITY (Facility Points)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This data set contains data for oil field facilities for the North Slope of Alaska. Vector points in this data set represent oil field facility locations. This data...

  1. EXPANDING EXTRACTIONS

    NARCIS (Netherlands)

    Dietzenbacher, Erik; Lahr, Michael L.

    2013-01-01

    In this paper, we generalize hypothetical extraction techniques. We suggest that the effect of certain economic phenomena can be measured by removing them from an input-output (I-O) table and by rebalancing the set of I-O accounts. The difference between the two sets of accounts yields the

  2. Jupiter Laser Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Jupiter Laser Facility is an institutional user facility in the Physical and Life Sciences Directorate at LLNL. The facility is designed to provide a high degree...

  3. Basic Research Firing Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Basic Research Firing Facility is an indoor ballistic test facility that has recently transitioned from a customer-based facility to a dedicated basic research...

  4. RETRACTED: Facile, eco-friendly and template free photosynthesis of cauliflower like ZnO nanoparticles using leaf extract of Tamarindus indica (L.) and its biological evolution of antibacterial and antifungal activities

    Science.gov (United States)

    Elumalai, K.; Velmurugan, S.; Ravi, S.; Kathiravan, V.; Ashokkumar, S.

    2015-02-01

    This article has been retracted at the request of the Editor. The article contains an image (Fig. 5B) which was published later again in "Green synthesis of zinc oxide nanoparticles using Moringa oleifera leaf extract and evaluation of its antimicrobial activity" by K. Elumalai et al. in Spectrochimica Acta Part A: Molecular and Biomolecular Spectroscopy 143 (2015) 158-164, http://dx.doi.org/10.1016/j.saa.2015.02.011 despite being attributed to different nanoparticles. Even though this represents the first publication of this image, the attribution to different sets of nanoparticles makes the dataset suspect. The scientific community takes a very strong view on this scientific misbehavior and apologies are offered to readers of the journal that this was not detected during the submission process.

  5. Aperture area measurement facility

    Data.gov (United States)

    Federal Laboratory Consortium — NIST has established an absolute aperture area measurement facility for circular and near-circular apertures use in radiometric instruments. The facility consists of...

  6. High Throughput Facility

    Data.gov (United States)

    Federal Laboratory Consortium — Argonne?s high throughput facility provides highly automated and parallel approaches to material and materials chemistry development. The facility allows scientists...

  7. Licensed Healthcare Facilities

    Data.gov (United States)

    California Natural Resource Agency — The Licensed Healthcare Facilities point layer represents the locations of all healthcare facilities licensed by the State of California, Department of Health...

  8. Facility Registry Service (FRS)

    Data.gov (United States)

    U.S. Environmental Protection Agency — The Facility Registry Service (FRS) provides an integrated source of comprehensive (air, water, and waste) environmental information about facilities across EPA,...

  9. Photochemical separation and extraction device

    International Nuclear Information System (INIS)

    Wada, Yukio; Morimoto, Kyoichi.

    1998-01-01

    The present invention concerns a device for separating neptunium and plutonium from highly radioactive liquid wastes, in which valance control by irradiation of UV rays and extraction operation by using an organic solvent can be conducted simultaneously in the same reaction vessel. Namely, a step of irradiating UV rays to the liquid in the reaction vessel to control the valence of predetermined materials and a step of separating the materials by conducting solvent-extraction while stirring with a solvent are conducted simultaneously or successively. Then, facilities for the separation method can be reduced and the operation steps can be simplified. (N.H.)

  10. Guide to research facilities

    Energy Technology Data Exchange (ETDEWEB)

    1993-06-01

    This Guide provides information on facilities at US Department of Energy (DOE) and other government laboratories that focus on research and development of energy efficiency and renewable energy technologies. These laboratories have opened these facilities to outside users within the scientific community to encourage cooperation between the laboratories and the private sector. The Guide features two types of facilities: designated user facilities and other research facilities. Designated user facilities are one-of-a-kind DOE facilities that are staffed by personnel with unparalleled expertise and that contain sophisticated equipment. Other research facilities are facilities at DOE and other government laboratories that provide sophisticated equipment, testing areas, or processes that may not be available at private facilities. Each facility listing includes the name and phone number of someone you can call for more information.

  11. Facile biological synthetic strategy to morphologically aligned CeO2/ZrO2 core nanoparticles using Justicia adhatoda extract and ionic liquid: Enhancement of its bio-medical properties.

    Science.gov (United States)

    Pandiyan, Nithya; Murugesan, Balaji; Sonamuthu, Jegatheeswaran; Samayanan, Selvam; Mahalingam, Sundrarajan

    2018-01-01

    In this study, a typical green synthesis route has approached for CeO 2 /ZrO 2 core metal oxide nanoparticles using ionic liquid mediated Justicia adhatoda extract. This synthesis method is carried out at simple room temperature condition to obtain the core metal oxide nanoparticles. XRD, SEM and TEM studies employed to study the crystalline and surface morphological properties under nucleation, growth, and aggregation processes. CeO 2 /ZrO 2 core metal oxides display agglomerated nano stick-like structure with 20-45nm size. GC-MS spectroscopy confirms the presence of vasicinone and N,N-Dimethylglycine present in the plant extract, which are capable of converting the corresponding metal ion precursor to CeO 2 /ZrO 2 core metal oxide nanoparticles. In FTIR, the corresponding stretching for Ce-O and Zr-O bands indicated at 498 and 416cm -1 and Raman spectroscopy also supports typical stretching frequencies at 463 and 160cm -1 . Band gap energy of the CeO 2 /ZrO 2 core metal oxide is 3.37eV calculated from UV- DRS spectroscopy. The anti-bacterial studies performed against a set of bacterial strains the result showed that core metal oxide nanoparticles more susceptible to gram-positive (G+) bacteria than gram-negative (G-) bacteria. A unique feature of the antioxidant behaviors core metal oxides reduces the concentration of DPPH radical up to 89%. The CeO 2 /ZrO 2 core metal oxide nanoparticles control the S. marcescent bio-film formation and restrict the quorum sensing. The toxicology behavior of CeO 2 /ZrO 2 core metal oxide NPs is found due to the high oxygen site vacancies, ROS formation, smallest particle size and higher surface area. This type of green synthesis route may efficient and the core metal oxide nanoparticles will possess a good bio-medical agent in future. Copyright © 2017 Elsevier B.V. All rights reserved.

  12. Facile preparation of surface-exchangeable core@shell iron oxide@gold nanoparticles for magnetic solid-phase extraction: Use of gold shell as the intermediate platform for versatile adsorbents with varying self-assembled monolayers

    Energy Technology Data Exchange (ETDEWEB)

    Li, Yaping [Beijing National Laboratory of Molecular Sciences, Key Laboratory of Analytical Chemistry for Living Biosystems, Institute of Chemistry, Chinese Academy of Sciences, Beijing 100190 (China); Graduate School, University of Chinese Academy of Sciences, Beijing 100049 (China); Qi, Li, E-mail: qili@iccas.ac.cn [Beijing National Laboratory of Molecular Sciences, Key Laboratory of Analytical Chemistry for Living Biosystems, Institute of Chemistry, Chinese Academy of Sciences, Beijing 100190 (China); Shen, Ying [Beijing National Laboratory of Molecular Sciences, Key Laboratory of Analytical Chemistry for Living Biosystems, Institute of Chemistry, Chinese Academy of Sciences, Beijing 100190 (China); Graduate School, University of Chinese Academy of Sciences, Beijing 100049 (China); Ma, Huimin [Beijing National Laboratory of Molecular Sciences, Key Laboratory of Analytical Chemistry for Living Biosystems, Institute of Chemistry, Chinese Academy of Sciences, Beijing 100190 (China)

    2014-02-06

    Graphical abstract: -- Highlights: •The core@shell Fe{sub 3}O{sub 4}@Au nanoparticles functionalized with SAMs were successfully constructed. •The SAMs could be transformed from one kind to another via thiol exchange process. •The developed nanomaterials could be applied in mode switching MSPE. -- Abstract: The core@shell Fe{sub 3}O{sub 4}@Au nanoparticles (NPs) functionalized with exchangeable self-assembled monolayers have been developed for mode switching magnetic solid-phase extraction (MSPE) using high performance liquid chromatography with ultraviolet detection. The adsorbents were synthesized by chemical coprecipitation to prepare magnetic cores followed by sonolysis to produce gold shells. Functionalization of Fe{sub 3}O{sub 4}@Au NPs surface was realized through self-assembly of commercially available low molecular weight thiol-containing ligands using gold shells as intermediate platform and the dynamic nature of Au–S chemistry allowed substituent of one thiol-containing ligand with another simply by thiol exchange process. The resultant adsorbents were characterized by transmission electronic microscopy, Fourier transform infrared spectroscopy, elemental analysis, contact angle measurement, and vibrating sample magnetometry. To evaluate the versatile performance of the developed MSPE adsorbents, they were applied for normal-phase SPE followed by reversed-phase SPE. A few kinds of diphenols and polycyclic aromatic hydrocarbons (PAHs) were employed as model analytes, respectively. The predominant parameters affecting extraction efficiency were investigated and optimized. Under the optimum experimental conditions, wide dynamic linear range (6.25–1600 μg L{sup −1} for diphenols and 1.56–100 μg L{sup −1} for PAHs) with good linearity (r{sup 2} ≥ 0.989) and low detection limits (0.34–16.67 μg L{sup −1} for diphenols and 0.26–0.52 μg L{sup −1} for PAHs) were achieved. The advantage of the developed method is that the Fe{sub 3}O

  13. Advanced Extraction Methods for Actinide/Lanthanide Separations

    International Nuclear Information System (INIS)

    Scott, M.J.

    2005-01-01

    The separation of An(III) ions from chemically similar Ln(III) ions is perhaps one of the most difficult problems encountered during the processing of nuclear waste. In the 3+ oxidation states, the metal ions have an identical charge and roughly the same ionic radius. They differ strictly in the relative energies of their f- and d-orbitals, and to separate these metal ions, ligands will need to be developed that take advantage of this small but important distinction. The extraction of uranium and plutonium from nitric acid solution can be performed quantitatively by the extraction with the TBP (tributyl phosphate). Commercially, this process has found wide use in the PUREX (plutonium uranium extraction) reprocessing method. The TRUEX (transuranium extraction) process is further used to coextract the trivalent lanthanides and actinides ions from HLLW generated during PUREX extraction. This method uses CMPO [(N, N-diisobutylcarbamoylmethyl) octylphenylphosphineoxide] intermixed with TBP as a synergistic agent. However, the final separation of trivalent actinides from trivalent lanthanides still remains a challenging task. In TRUEX nitric acid solution, the Am(III) ion is coordinated by three CMPO molecules and three nitrate anions. Taking inspiration from this data and previous work with calix[4]arene systems, researchers on this project have developed a C3-symmetric tris-CMPO ligand system using a triphenoxymethane platform as a base. The triphenoxymethane ligand systems have many advantages for the preparation of complex ligand systems. The compounds are very easy to prepare. The steric and solubility properties can be tuned through an extreme range by the inclusion of different alkoxy and alkyl groups such as methyoxy, ethoxy, t-butoxy, methyl, octyl, t-pentyl, or even t-pentyl at the ortho- and para-positions of the aryl rings. The triphenoxymethane ligand system shows promise as an improved extractant for both tetravalent and trivalent actinide recoveries form

  14. Advanced Extraction Methods for Actinide/Lanthanide Separations

    Energy Technology Data Exchange (ETDEWEB)

    Scott, M.J.

    2005-12-01

    The separation of An(III) ions from chemically similar Ln(III) ions is perhaps one of the most difficult problems encountered during the processing of nuclear waste. In the 3+ oxidation states, the metal ions have an identical charge and roughly the same ionic radius. They differ strictly in the relative energies of their f- and d-orbitals, and to separate these metal ions, ligands will need to be developed that take advantage of this small but important distinction. The extraction of uranium and plutonium from nitric acid solution can be performed quantitatively by the extraction with the TBP (tributyl phosphate). Commercially, this process has found wide use in the PUREX (plutonium uranium extraction) reprocessing method. The TRUEX (transuranium extraction) process is further used to coextract the trivalent lanthanides and actinides ions from HLLW generated during PUREX extraction. This method uses CMPO [(N, N-diisobutylcarbamoylmethyl) octylphenylphosphineoxide] intermixed with TBP as a synergistic agent. However, the final separation of trivalent actinides from trivalent lanthanides still remains a challenging task. In TRUEX nitric acid solution, the Am(III) ion is coordinated by three CMPO molecules and three nitrate anions. Taking inspiration from this data and previous work with calix[4]arene systems, researchers on this project have developed a C3-symmetric tris-CMPO ligand system using a triphenoxymethane platform as a base. The triphenoxymethane ligand systems have many advantages for the preparation of complex ligand systems. The compounds are very easy to prepare. The steric and solubility properties can be tuned through an extreme range by the inclusion of different alkoxy and alkyl groups such as methyoxy, ethoxy, t-butoxy, methyl, octyl, t-pentyl, or even t-pentyl at the ortho- and para-positions of the aryl rings. The triphenoxymethane ligand system shows promise as an improved extractant for both tetravalent and trivalent actinide recoveries form

  15. Communication grounding facility

    International Nuclear Information System (INIS)

    Lee, Gye Seong

    1998-06-01

    It is about communication grounding facility, which is made up twelve chapters. It includes general grounding with purpose, materials thermal insulating material, construction of grounding, super strength grounding method, grounding facility with grounding way and building of insulating, switched grounding with No. 1A and LCR, grounding facility of transmission line, wireless facility grounding, grounding facility in wireless base station, grounding of power facility, grounding low-tenton interior power wire, communication facility of railroad, install of arrester in apartment and house, install of arrester on introduction and earth conductivity and measurement with introduction and grounding resistance.

  16. The ISAC facility at TRIUMF

    International Nuclear Information System (INIS)

    Dilling, J.

    2005-01-01

    ISAC at TRIUMF, Vancouver is one of the prime radioactive beam facilities worldwide. The isotopes are produced via the isol method and are extracted to typically 30-60 keV beams and subsequently mass selected. The beam can be further accelerated to 1.8 meV/u and with the completion of ISAC II (2005/6) up to 6.5 meV/u. One of the primary motivations at ISAC are nuclear astrophysics experiments. In addition to cross-section determination, Q-values are key parameters. The latter ones are accessible via mass measurements. The TITAN system at ISAC will allow to carry out such measurements with the very high precision (δm/m ≤ x 10 -8 ) on short-lived isotopes (T 1/2 ∼ 10 ms). An introduction to TITAN, together with an overview of the ISAC facility will be given. (author)

  17. AOV Facility Tool/Facility Safety Specifications -

    Data.gov (United States)

    Department of Transportation — Develop and maintain authorizing documents that are standards that facilities must follow. These standards are references of FAA regulations and are specific to the...

  18. THOREX processing and zeolite transfer for high-level waste stream processing blending

    International Nuclear Information System (INIS)

    Kelly, S. Jr.; Meess, D.C.

    1997-07-01

    The West Valley Demonstration Project (WVDP) completed the pretreatment of the high-level radioactive waste (HLW) prior to the start of waste vitrification. The HLW originated form the two million liters of plutonium/uranium extraction (PUREX) and thorium extraction (THOREX) wastes remaining from Nuclear Fuel Services' (NFS) commercial nuclear fuel reprocessing operations at the Western New York Nuclear Service Center (WNYNSC) from 1966 to 1972. The pretreatment process removed cesium as well as other radionuclides from the liquid wastes and captured these radioactive materials onto silica-based molecular sieves (zeolites). The decontaminated salt solutions were volume-reduced and then mixed with portland cement and other admixtures. Nineteen thousand eight hundred and seventy-seven 270-liter square drums were filled with the cement-wastes produced from the pretreatment process. These drums are being stored in a shielded facility on the site until their final disposition is determined. Over 6.4 million liters of liquid HLW were processed through the pretreatment system. PUREX supernatant was processed first, followed by two PUREX sludge wash solutions. A third wash of PUREX/THOREX sludge was then processed after the neutralized THOREX waste was mixed with the PUREX waste. Approximately 6.6 million curies of radioactive cesium-137 (Cs-137) in the HLW liquid were removed and retained on 65,300 kg of zeolites. With pretreatment complete, the zeolite material has been mobilized, size-reduced (ground), and blended with the PUREX and THOREX sludges in a single feed tank that will supply the HLW slurry to the Vitrification Facility

  19. Selective extraction of actinides from high level liquid wastes. Study of the possibilities offered by the Redox properties of actinides

    International Nuclear Information System (INIS)

    Adnet, J.M.

    1991-07-01

    Partitioning of high level liquid wastes coming from nuclear fuel reprocessing by the PUREX process, consists in the elimination of minor actinides (Np, Am, and traces of Pu and U). Among the possible processes, the selective extraction of actinides with oxidation states higher than three is studied. First part of this work deals with a preliminary step; the elimination of the ruthenium from fission products solutions using the electrovolatilization of the RuO4 compound. The second part of this work concerns the complexation and oxidation reactions of the elements U, Np, Pu and Am in presence of a compound belonging to the insaturated polyanions family: the potassium phosphotungstate. For actinide ions with oxidation state (IV) complexed with phosphotungstate anion the extraction mechanism by dioctylamine was studied and the use of a chromatographic extraction technic permitted successful separations between tetravalents actinides and trivalents actinides. Finally, in accordance with the obtained results, the basis of a separation scheme for the management of fission products solutions is proposed

  20. Nuclear power generation facility

    International Nuclear Information System (INIS)

    Kubo, Mitsuji.

    1996-01-01

    Main steams are introduced from a moisture separation device for removing moisture content of the main steams to a low pressure turbine passing through a cross-around pipe. A condensate desalter comprising a mixed floor-type desalting tower using granular ion exchange resins is disposed at the downstream of the main condensator by way of condensate pipelines, and a feedwater heater is disposed at the downstream. Structural members of the main condensator are formed by weather proof steels. Low alloy steels are used partially or entirely for the cross-around pipe, gas extraction pipelines, heat draining pipelines, inner structural members other than pipelines in the feedwater heater, and the body and the inner structural members of the moisture separator. Titanium or a titanium alloy is used for the pipelines in the main condensator. With such a constitution, BWR type reactor facilities, in which the concentration of cruds inflown to the condensate cleanup system is reduced to simplify the condensate cleanup device can be obtained. (I.N.)

  1. Gaseous waste processing facility

    International Nuclear Information System (INIS)

    Konno, Masanobu; Uchiyama, Yoshio; Suzuki, Kunihiko; Kimura, Masahiro; Kawabe, Ken-ichi.

    1992-01-01

    Gaseous waste recombiners 'A' and 'B' are connected in series and three-way valves are disposed at the upstream and the downstream of the recombiners A and B, and bypass lines are disposed to the recombiners A and B, respectively. An opening/closing controller for the three-way valves is interlocked with a hydrogen densitometer disposed to a hydrogen injection line. Hydrogen gas and oxygen gas generated by radiolysis in the reactor are extracted from a main condenser and caused to flow into a gaseous waste processing system. Gaseous wastes are introduced together with overheated steams to the recombiner A upon injection of hydrogen. Both of the bypass lines of the recombiners A and B are closed, and recombining reaction for the increased hydrogen gas is processed by the recombiners A and B connected in series. In an operation mode not conducting hydrogen injection, it is passed through the bypass line of the recombiner A and processed by the recombiner B. With such procedures, the increase of gaseous wastes due to hydrogen injection can be coped with existent facilities. (I.N.)

  2. Generalized plotting facility

    Energy Technology Data Exchange (ETDEWEB)

    Burris, R.D.; Gray, W.H.

    1978-01-01

    A command which causes the translation of any supported graphics file format to a format acceptable to any supported device was implemented on two linked DECsystem-10s. The processing of the command is divided into parsing and translating phases. In the parsing phase, information is extracted from the command and augmented by default data. The results of this phase are saved on disk, and the appropriate translating routine is invoked. Twenty-eight translating programs were implemented in this system. They support four different graphics file formats, including the DISSPLA and Calcomp formats, and seven different types of plotters, including Tektronix, Calcomp, and Versatec devices. Some of the plotters are devices linked to the DECsystem-10s, and some are driven by IBM System/360 computers linked via a communications network to the DECsystem-10s. The user of this facility can use any of the supported packages to create a file of graphics data, preview the file on an on-line scope, and, when satisfied, cause the same data to be plotted on a hard-copy device. All of the actions utilize a single simple command format. 2 figures.

  3. Extracting oils

    Energy Technology Data Exchange (ETDEWEB)

    Patart, G

    1926-03-15

    In the hydrogenation or extraction of by-products from organic substances at high temperatures and pressures, the gases or liquids, or both, used are those which are already heated and compressed during industrial operations such as exothermic synthesizing reactions such as the production of methanol from hydrogen and carbon monoxide in a catalytic process. Gases from this reaction may be passed upwardly through a digester packed with pine wood while liquid from the same catalytic process is passed downwardly through the material. The issuing liquid contains methanol, pine oil, acetone, isopropyl alcohol, and acetic acid. The gases contain additional hydrogen, carbon monoxide, methane, ethylene, and its homologs which are condensed upon the catalyser to liquid hydrocarbons. Petroleum oils and coal may be treated similarly.

  4. Depleted uranium processing and fluorine extraction

    International Nuclear Information System (INIS)

    Laflin, S.T.

    2010-01-01

    Since the beginning of the nuclear era, there has never been a commercial solution for the large quantities of depleted uranium hexafluoride generated from uranium enrichment. In the United States alone, there is already in excess of 1.6 billion pounds (730 million kilograms) of DUF_6 currently stored. INIS is constructing a commercial uranium processing and fluorine extraction facility. The INIS facility will convert depleted uranium hexafluoride and use it as feed material for the patented Fluorine Extraction Process to produce high purity fluoride gases and anhydrous hydrofluoric acid. The project will provide an environmentally friendly and commercially viable solution for DUF_6 tails management. (author)

  5. Advanced integrated solvent extraction systems

    Energy Technology Data Exchange (ETDEWEB)

    Horwitz, E.P.; Dietz, M.L.; Leonard, R.A. [Argonne National Lab., IL (United States)

    1997-10-01

    Advanced integrated solvent extraction systems are a series of novel solvent extraction (SX) processes that will remove and recover all of the major radioisotopes from acidic-dissolved sludge or other acidic high-level wastes. The major focus of this effort during the last 2 years has been the development of a combined cesium-strontium extraction/recovery process, the Combined CSEX-SREX Process. The Combined CSEX-SREX Process relies on a mixture of a strontium-selective macrocyclic polyether and a novel cesium-selective extractant based on dibenzo 18-crown-6. The process offers several potential advantages over possible alternatives in a chemical processing scheme for high-level waste treatment. First, if the process is applied as the first step in chemical pretreatment, the radiation level for all subsequent processing steps (e.g., transuranic extraction/recovery, or TRUEX) will be significantly reduced. Thus, less costly shielding would be required. The second advantage of the Combined CSEX-SREX Process is that the recovered Cs-Sr fraction is non-transuranic, and therefore will decay to low-level waste after only a few hundred years. Finally, combining individual processes into a single process will reduce the amount of equipment required to pretreat the waste and therefore reduce the size and cost of the waste processing facility. In an ongoing collaboration with Lockheed Martin Idaho Technology Company (LMITCO), the authors have successfully tested various segments of the Advanced Integrated Solvent Extraction Systems. Eichrom Industries, Inc. (Darien, IL) synthesizes and markets the Sr extractant and can supply the Cs extractant on a limited basis. Plans are under way to perform a test of the Combined CSEX-SREX Process with real waste at LMITCO in the near future.

  6. Lesotho - Health Facility Survey

    Data.gov (United States)

    Millennium Challenge Corporation — The main objective of the 2011 Health Facility Survey (HFS) was to establish a baseline for informing the Health Project performance indicators on health facilities,...

  7. Armament Technology Facility (ATF)

    Data.gov (United States)

    Federal Laboratory Consortium — The Armament Technology Facility is a 52,000 square foot, secure and environmentally-safe, integrated small arms and cannon caliber design and evaluation facility....

  8. Projectile Demilitarization Facilities

    Data.gov (United States)

    Federal Laboratory Consortium — The Projectile Wash Out Facility is US Army Ammunition Peculiar Equipment (APE 1300). It is a pilot scale wash out facility that uses high pressure water and steam...

  9. Rocketball Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — This test facility offers the capability to emulate and measure guided missile radar cross-section without requiring flight tests of tactical missiles. This facility...

  10. Wastewater Treatment Facilities

    Data.gov (United States)

    Iowa State University GIS Support and Research Facility — Individual permits for municipal, industrial, and semi-public wastewater treatment facilities in Iowa for the National Pollutant Discharge Elimination System (NPDES)...

  11. Materiel Evaluation Facility

    Data.gov (United States)

    Federal Laboratory Consortium — CRREL's Materiel Evaluation Facility (MEF) is a large cold-room facility that can be set up at temperatures ranging from −20°F to 120°F with a temperature change...

  12. Environmental Toxicology Research Facility

    Data.gov (United States)

    Federal Laboratory Consortium — Fully-equipped facilities for environmental toxicology researchThe Environmental Toxicology Research Facility (ETRF) located in Vicksburg, MS provides over 8,200 ft...

  13. Dialysis Facility Compare

    Data.gov (United States)

    U.S. Department of Health & Human Services — Dialysis Facility Compare helps you find detailed information about Medicare-certified dialysis facilities. You can compare the services and the quality of care that...

  14. Energetics Conditioning Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Energetics Conditioning Facility is used for long term and short term aging studies of energetic materials. The facility has 10 conditioning chambers of which 2...

  15. Explosive Components Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The 98,000 square foot Explosive Components Facility (ECF) is a state-of-the-art facility that provides a full-range of chemical, material, and performance analysis...

  16. Facilities for US Radioastronomy.

    Science.gov (United States)

    Thaddeus, Patrick

    1982-01-01

    Discusses major developments in radioastronomy since 1945. Topics include proposed facilities, very-long-baseline interferometric array, millimeter-wave telescope, submillimeter-wave telescope, and funding for radioastronomy facilities and projects. (JN)

  17. Neighbourhood facilities for sustainability

    CSIR Research Space (South Africa)

    Gibberd, Jeremy T

    2013-01-01

    Full Text Available . In this paper these are referred to as ‘Neighbourhood Facilities for Sustainability’. Neighbourhood Facilities for Sustainability (NFS) are initiatives undertaken by individuals and communities to build local sustainable systems which not only improve...

  18. Cold Vacuum Drying Facility

    Data.gov (United States)

    Federal Laboratory Consortium — Located near the K-Basins (see K-Basins link) in Hanford's 100 Area is a facility called the Cold Vacuum Drying Facility (CVDF).Between 2000 and 2004, workers at the...

  19. Ouellette Thermal Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Thermal Test Facility is a joint Army/Navy state-of-the-art facility (8,100 ft2) that was designed to:Evaluate and characterize the effect of flame and thermal...

  20. Integrated Disposal Facility

    Data.gov (United States)

    Federal Laboratory Consortium — Located near the center of the 586-square-mile Hanford Site is the Integrated Disposal Facility, also known as the IDF.This facility is a landfill similar in concept...

  1. Facility design: introduction

    International Nuclear Information System (INIS)

    Unger, W.E.

    1980-01-01

    The design of shielded chemical processing facilities for handling plutonium is discussed. The TRU facility is considered in particular; its features for minimizing the escape of process materials are listed. 20 figures

  2. CLEAR test facility

    CERN Multimedia

    Ordan, Julien Marius

    2017-01-01

    A new user facility for accelerator R&D, the CERN Linear Electron Accelerator for Research (CLEAR), started operation in August 2017. CLEAR evolved from the former CLIC Test Facility 3 (CTF3) used by the Compact Linear Collider (CLIC). The new facility is able to host and test a broad range of ideas in the accelerator field.

  3. PROJECTIZING AN OPERATING NUCLEAR FACILITY

    International Nuclear Information System (INIS)

    Adams, N

    2007-01-01

    This paper will discuss the evolution of an operations-based organization to a project-based organization to facilitate successful deactivation of a major nuclear facility. It will describe the plan used for scope definition, staff reorganization, method estimation, baseline schedule development, project management training, and results of this transformation. It is a story of leadership and teamwork, pride and success. Workers at the Savannah River Site's (SRS) F Canyon Complex (FCC) started with a challenge--take all the hazardous byproducts from nearly 50 years of operations in a major, first-of-its-kind nuclear complex and safely get rid of them, leaving the facility cold, dark, dry and ready for whatever end state is ultimately determined by the United States Department of Energy (DOE). And do it in four years, with a constantly changing workforce and steadily declining funding. The goal was to reduce the overall operating staff by 93% and budget by 94%. The facilities, F Canyon and its adjoined sister, FB Line, are located at SRS, a 310-square-mile nuclear reservation near Aiken, S.C., owned by DOE and managed by Washington Group International subsidiary Washington Savannah River Company (WSRC). These facilities were supported by more than 50 surrounding buildings, whose purpose was to provide support services during operations. The radiological, chemical and industrial hazards inventory in the old buildings was significant. The historical mission at F Canyon was to extract plutonium-239 and uranium-238 from irradiated spent nuclear fuel through chemical processing. FB Line's mission included conversion of plutonium solutions into metal, characterization, stabilization and packaging, and storage of both metal and oxide forms. The plutonium metal was sent to another DOE site for use in weapons. Deactivation in F Canyon began when chemical separations activities were completed in 2002, and a cross-functional project team concept was implemented to successfully

  4. Facility or Facilities? That is the Question.

    Science.gov (United States)

    Viso, M.

    2018-04-01

    The management of the martian samples upon arrival on the Earth will require a lot of work to ensure a safe life detection and biohazard testing during the quarantine. This will induce a sharing of the load between several facilities.

  5. 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)

  6. Pyrometallurgical partitioning of uranium and transuranic elements from rare earth elements by electrorefining and reductive extraction

    International Nuclear Information System (INIS)

    Uozumi, Koichi; Kinoshita, Kensuke; Inoue, Tadashi; Storvick, T.S.; Krueger, C.L.; Nabelek, C.R.

    2001-01-01

    High-level liquid waste generated from PUREX reprocessing contains a small amount of transuranic elements, such as Np, Pu, Am, and Cm, with long-lived radioactivities. A pyrometallurgical partitioning process is being developed to recover transuranic elements from such waste. Small amounts of U contained in the high-level liquid waste are also recovered in the process. A key issue for developing the process is effective separation of U and the transuranic elements from the rare-earth elements, because the two elemental groups are chemically analogous. A series of process tests were carried out in the present study to demonstrate that a combination of electrorefining and reductive extraction is useful for separating U and transuranic elements from the rare-earth elements. The results indicate that 99% of U and each transuranic elements is recovered by the combination process as a product, and that the quantity of rare-earth elements contained in the product is smaller than the transuranic elements by weight. The overall mass balance of U and transuranic elements in the system ranged within the experimental errors assigned to sampling and analysis. (author)

  7. Advanced hybrid process with solvent extraction and pyro-chemical process of spent fuel reprocessing for LWR to FBR

    International Nuclear Information System (INIS)

    Fujita, Reiko; Mizuguchi, Koji; Fuse, Kouki; Saso, Michitaka; Utsunomiya, Kazuhiro; Arie, Kazuo

    2008-01-01

    Toshiba has been proposing a new fuel cycle concept of a transition from LWR to FBR. The new fuel cycle concept has better economical process of the LWR spent fuel reprocessing than the present Purex Process and the proliferation resistance for FBR cycle of plutonium with minor actinides after 2040. Toshiba has been developing a new Advanced Hybrid Process with Solvent Extraction and Pyrochemical process of spent fuel reprocessing for LWR to FBR. The Advanced Hybrid Process combines the solvent extraction process of the LWR spent fuel in nitric acid with the recovery of high pure uranium for LWR fuel and the pyro-chemical process in molten salts of impure plutonium recovery with minor actinides for metallic FBR fuel, which is the FBR spent fuel recycle system after FBR age based on the electrorefining process in molten salts since 1988. The new Advanced Hybrid Process enables the decrease of the high-level waste and the secondary waste from the spent fuel reprocessing plants. The R and D costs in the new Advanced Hybrid Process might be reduced because of the mutual Pyro-chemical process in molten salts. This paper describes the new fuel cycle concept of a transition from LWR to FBR and the feasibility of the new Advanced Hybrid Process by fundamental experiments. (author)

  8. Partitioning of actinide from simulated high level wastes arising from reprocessing of PHWR fuels: counter current extraction studies using CMPO

    International Nuclear Information System (INIS)

    Deshingkar, D.S.; Chitnis, R.R.; Wattal, P.K.; Theyyunni, T.K.; Nair, M.K.T.; Ramanujam, A.; Dhami, P.S.; Gopalakrishnan, V.; Rao, M.K.; Mathur, J.N.; Murali, M.S.; Iyer, R.H.; Badheka, L.P.; Banerji, A.

    1994-01-01

    High level wastes (HLW) arising from reprocessing of pressurised heavy water reactor (PHWR) fuels contain actinides like neptunium, americium and cerium which are not extracted in the Purex process. They also contain small quantities of uranium and plutonium in addition to fission products. Removal of these actinides prior to vitrification of HLW can effectively reduce the active surveillance period of final waste form. Counter current studies using indigenously synthesised octyl (phenyl)-N, N-diisobutylcarbamoylmethylphosphine oxide (CMPO) were taken up as a follow-up of successful runs with simulated sulphate bearing low acid HLW solutions. The simulated HLW arising from reprocessing of PHWR fuel was prepared based on presumed burnup of 6500 MWd/Te of uranium, 3 years cooling period and 800 litres of waste generation per tonne of fuel reprocessed. The alpha activity of the HLW raffinate after extraction with the CMPO-TBP mixture could be brought down to near background level. (author). 13 refs., 2 tabs., 12 figs

  9. Partitioning of actinide from simulated high level wastes arising from reprocessing of PHWR fuels: counter current extraction studies using CMPO

    Energy Technology Data Exchange (ETDEWEB)

    Deshingkar, D S; Chitnis, R R; Wattal, P K; Theyyunni, T K; Nair, M K.T. [Bhabha Atomic Research Centre, Bombay (India). Process Engineering and Systems Development Div.; Ramanujam, A; Dhami, P S; Gopalakrishnan, V; Rao, M K [Bhabha Atomic Research Centre, Bombay (India). Fuel Reprocessing Group; Mathur, J N; Murali, M S; Iyer, R H [Bhabha Atomic Research Centre, Bombay (India). Radiochemistry Div.; Badheka, L P; Banerji, A [Bhabha Atomic Research Centre, Bombay (India). Bio-organic Div.

    1994-12-31

    High level wastes (HLW) arising from reprocessing of pressurised heavy water reactor (PHWR) fuels contain actinides like neptunium, americium and cerium which are not extracted in the Purex process. They also contain small quantities of uranium and plutonium in addition to fission products. Removal of these actinides prior to vitrification of HLW can effectively reduce the active surveillance period of final waste form. Counter current studies using indigenously synthesised octyl (phenyl)-N, N-diisobutylcarbamoylmethylphosphine oxide (CMPO) were taken up as a follow-up of successful runs with simulated sulphate bearing low acid HLW solutions. The simulated HLW arising from reprocessing of PHWR fuel was prepared based on presumed burnup of 6500 MWd/Te of uranium, 3 years cooling period and 800 litres of waste generation per tonne of fuel reprocessed. The alpha activity of the HLW raffinate after extraction with the CMPO-TBP mixture could be brought down to near background level. (author). 13 refs., 2 tabs., 12 figs.

  10. Concept of a tritium extraction facility for a reprocessing plant

    International Nuclear Information System (INIS)

    Tunaboylu, K.; Paulovic, M.; Ulrich, D.

    1991-01-01

    There are several alternatives for reducing the release of tritium to the environment originating from the wastewater of a reprocessing plant. Such alternatives, which are applicable for sites not located by the sea or by large rivers, are limited to either injection of tritiated wastewater into suitable deep geological formations, or final disposal into a deep underground repository after adequate treatment similar to other low and intermediate active waste. Removal of tritium from the wastewater by enrichment represents a further feasible option of the second alternative, which allows reduction of the huge volume of tritiated water to be treated before disposal. A significant volume reduction increases the safety of the subsequent steps such as transport, interim storage and final disposal of tritiated waste, furthermore, decreases the corresponding overall waste management cost. The projected Wackersdorf reprocessing plant has been considered as a reference for assessing the permitted tritium releases and other site characteristics. (orig.)

  11. Facilities inventory protection for nuclear facilities

    International Nuclear Information System (INIS)

    Schmitt, F.J.

    1989-01-01

    The fact that shut-down applications have been filed for nuclear power plants, suggests to have a scrutinizing look at the scopes of assessment and decision available to administrations and courts for the protection of facilities inventories relative to legal and constitutional requirements. The paper outlines the legal bases which need to be observed if purposeful calculation is to be ensured. Based on the different actual conditions and legal consequences, the author distinguishes between 1) the legal situation of facilities licenced already and 2) the legal situation of facilities under planning during the licencing stage. As indicated by the contents and restrictions of the pertinent provisions of the Atomic Energy Act and by the corresponding compensatory regulation, the object of the protection of facilities inventor in the legal position of the facility owner within the purview of the Atomic Energy Act, and the licensing proper. Art. 17 of the Atomic Energy Act indicates the legislators intent that, once issued, the licence will be the pivotal point for regulations aiming at protection and intervention. (orig./HSCH) [de

  12. Design of extraction system in BRing at HIAF

    Science.gov (United States)

    Ruan, Shuang; Yang, Jiancheng; Zhang, Jinquan; Shen, Guodong; Ren, Hang; Liu, Jie; Shangguan, Jingbing; Zhang, Xiaoying; Zhang, Jingjing; Mao, Lijun; Sheng, Lina; Yin, Dayu; Wang, Geng; Wu, Bo; Yao, Liping; Tang, Meitang; Cai, Fucheng; Chen, Xiaoqiang

    2018-06-01

    The Booster Ring (BRing), which is the key part of HIAF (High Intensity heavy ion Accelerator Facility) complex at IMP (Institute of Modern Physics, Chinese Academy of Sciences), can provide uranium (A / q = 7) beam with a wide extraction energy range of 200-800 MeV/u. To fulfill a flexible beam extraction for multi-purpose experiments, both fast and slow extraction systems will be accommodated in the BRing. The fast extraction system is used for extracting short bunched beam horizontally in single-turn. The slow extraction system is used to provide quasi-continuous beam by the third order resonance and RF-knockout scheme. To achieve a compact structure, the two extraction systems are designed to share the same extraction channel. The general design of the fast and slow extraction systems and simulation results are discussed in this paper.

  13. Facilities projects performance measurement system

    International Nuclear Information System (INIS)

    Erben, J.F.

    1979-01-01

    The two DOE-owned facilities at Hanford, the Fuels and Materials Examination Facility (FMEF), and the Fusion Materials Irradiation Test Facility (FMIT), are described. The performance measurement systems used at these two facilities are next described

  14. The sodium process facility at Argonne National Laboratory - West

    International Nuclear Information System (INIS)

    Michelbacher, J.A.; Henslee, S.P.; McDermott, M.D.; Price, J.R.; Rosenberg, K.E.; Wells, P.B.

    1997-01-01

    Argonne National Laboratory - West (ANL-W) has approximately 680,000 liters (180,000 gallons) of raw sodium stored in facilities on site. As mandated by the State of Idaho and the United States Department of Energy (DOE), this sodium must be transformed into a stable condition for land disposal. To comply with this mandate, ANL-W designed and built the Sodium Process Facility (SPF) for the processing of this sodium into a dry, sodium carbonate powder. The major portion of the sodium stored at ANL-W is radioactively contaminated. The SPF was designed to react elemental sodium to sodium carbonate through two-stages involving caustic process and carbonate process steps. The sodium is first reacted to sodium hydroxide in the caustic process step. The caustic process step involves the injection of sodium into a nickel reaction vessel filled with a 50 wt% solution of sodium hydroxide. Water is also injected, controlling the boiling point of the solution. In the carbonate process, the sodium hydroxide is reacted with carbon dioxide to form sodium carbonate. This dry powder, similar in consistency to baking soda, is a waste form acceptable for burial in the State of Idaho as a non-hazardous, radioactive waste. The caustic process was originally designed and built in the 1980s for reacting the 290,000 liters (77,000 gallons) of primary sodium from the Fermi-1 Reactor to sodium hydroxide. The hydroxide was slated to be used to neutralize acid products from the PUREX process at the Hanford site. However, changes in the DOE mission precluded the need for hydroxide and the caustic process was never operated. With the shutdown of the Experimental Breeder Reactor-II (EBR-II), the necessity for a facility to react sodium was identified. In order to comply with Resource Conservation and Recovery Act (RCRA) requirements, the sodium had to be converted into a waste form acceptable for disposal in a Sub-Title D low-level radioactive waste disposal facility. Sodium hydroxide is a RCRA

  15. Status of the ELISE test facility

    International Nuclear Information System (INIS)

    Franzen, P.; Wünderlich, D.; Riedl, R.; Nocentini, R.; Fantz, U.; Fröschle, M.; Heinemann, B.; Martens, C.; Kraus, W.; Ruf, B.; Bonomo, F.; Pimazzoni, A.

    2015-01-01

    The test facility ELISE, equipped with a large radio frequency (RF) driven ion source (1×0.9 m2) of half the size of the ion source for the ITER neutral beam injection (NBI) system, is operational since beginning of 2013. The first experimental campaign was dedicated to a thorough qualification of the test facility and its diagnostic tools at low RF power (80 kW in total, i.e. 20 kW per driver) in volume operation, i.e. operation without cesium, where the negative hydrogen ion production is done in the plasma volume only. This paper reports on the main results of the second and third experimental campaigns, where Cs was inserted in the ion source for an enhancement of the negative ion production by the surface process. The second experimental campaign was done still with low RF power, both for hydrogen and deuterium, with pulse lengths of up to 500 s. The results of this campaign are rather encouraging, especially in hydrogen, where large current densities with respect to the low RF power could be achieved at a ratio of co-extracted electrons to extracted ions of 0.5-0.6 at the relevant source pressure of 0.3 Pa. Similar large extracted ion currents could be achieved also in deuterium, but with larger amounts of co-extracted electrons. The required ratio of co-extracted electrons to extracted ions of one could be achieved only in short pulses. The third experimental campaign aimed then for approaching the required ITER NBI parameters with respect to the ion and electron extracted currents, both for hydrogen and deuterium, by increasing the RF power with short pulses, i.e. beam-on times of up to 10 s and RF-on time up to 20 s. Current densities near the ITER NBI requirements could be achieved in hydrogen at a ratio of co-extracted electrons to extracted ions of 0.5-0.6 at the relevant source pressure of 0.3 Pa. As it was the case for the low RF operation, the required filter field was significantly lower than expected from the experience with the small

  16. First operations of the LNS heavy ions facility

    International Nuclear Information System (INIS)

    Calabretta, L.; Ciavola, G.; Cuttone, G.; Gammino, S.; Gmaj, P.; Migneco, E.; Raia, G.; Rifuggiato, D.; Rovelli, A.; Sura, J.; Scuderi, V.; Acerbi, E.; Alessandria, F.; Bellomo, G.; Bosotti, A.; Martinis, C. de; Giove, D.; Michelato, P.; Pagani, C.; Rossi, L.

    1996-01-01

    A heavy ion facility is now available at laboratorio nazionale del Sud (LNS) of Catania. It can deliver beams with an energy up to 100 MeV/amu. The facility is based on a 15 MV HVEC tandem and a K=800 superconducting cyclotron as booster. During the last year, the facility came into operation. A 58 Ni beam delivered by the tandem has been radially injected in the SC and then has been accelerated and extracted at 30 MeV/amu. In this paper the status of the facility together with the experience gained during the commissioning will be extensively reported. (orig.)

  17. First operations of the LNS heavy ions facility

    Energy Technology Data Exchange (ETDEWEB)

    Calabretta, L. [INFN-LNS, Catania (Italy); Ciavola, G. [INFN-LNS, Catania (Italy); Cuttone, G. [INFN-LNS, Catania (Italy); Gammino, S. [INFN-LNS, Catania (Italy); Gmaj, P. [INFN-LNS, Catania (Italy); Migneco, E. [INFN-LNS, Catania (Italy); Raia, G. [INFN-LNS, Catania (Italy); Rifuggiato, D. [INFN-LNS, Catania (Italy); Rovelli, A. [INFN-LNS, Catania (Italy); Sura, J. [INFN-LNS, Catania (Italy); Scuderi, V. [INFN-LNS, Catania (Italy); Acerbi, E. [INFN-sezione di Milano (Italy)]|[Univ. degli studi di Milano, Lab. LASA (Italy); Alessandria, F. [INFN-sezione di Milano (Italy)]|[Univ. degli studi di Milano, Lab. LASA (Italy); Bellomo, G. [INFN-sezione di Milano (Italy)]|[Univ. degli studi di Milano, Lab. LASA (Italy); Bosotti, A. [INFN-sezione di Milano (Italy)]|[Univ. degli studi di Milano, Lab. LASA (Italy); Martinis, C. de [INFN-sezione di Milano (Italy)]|[Univ. degli studi di Milano, Lab. LASA (Italy); Giove, D. [INFN-sezione di Milano (Italy)]|[Univ. degli studi di Milano, Lab. LASA (Italy); Michelato, P. [INFN-sezione di Milano (Italy)]|[Univ. degli studi di Milano, Lab. LASA (Italy); Pagani, C. [INFN-sezione di Milano (Italy)]|[Univ. degli studi di Milano, Lab. LASA (Italy); Rossi, L. [INFN-sezione di Milano (Italy)]|[Univ. degli studi di Milano, Lab. LASA (Italy)

    1996-11-11

    A heavy ion facility is now available at laboratorio nazionale del Sud (LNS) of Catania. It can deliver beams with an energy up to 100 MeV/amu. The facility is based on a 15 MV HVEC tandem and a K=800 superconducting cyclotron as booster. During the last year, the facility came into operation. A {sup 58}Ni beam delivered by the tandem has been radially injected in the SC and then has been accelerated and extracted at 30 MeV/amu. In this paper the status of the facility together with the experience gained during the commissioning will be extensively reported. (orig.).

  18. History of 232-F, tritium extraction processing

    International Nuclear Information System (INIS)

    Blackburn, G.W.

    1994-08-01

    In 1950 the Atomic Energy Commission authorized the Savannah River Project principally for the production of tritium and plutonium-239 for use in thermonuclear weapons. 232-F was built as an interim facility in 1953--1954, at a cost of $3.9M. Tritium extraction operations began in October, 1955, after the reactor and separations startups. In July, 1957 a larger tritium facility began operation in 232-H. In 1958 the capacity of 232-H was doubled. Also, in 1957 a new task was assigned to Savannah River, the loading of tritium into reservoirs that would be actual components of thermonuclear weapons. This report describes the history of 232-F, the process for tritium extraction, and the lessons learned over the years that were eventually incorporated into the new Replacement Tritium Facility

  19. 340 Facility compliance assessment

    International Nuclear Information System (INIS)

    English, S.L.

    1993-10-01

    This study provides an environmental compliance evaluation of the RLWS and the RPS systems of the 340 Facility. The emphasis of the evaluation centers on compliance with WAC requirements for hazardous and mixed waste facilities, federal regulations, and Westinghouse Hanford Company (WHC) requirements pertinent to the operation of the 340 Facility. The 340 Facility is not covered under either an interim status Part A permit or a RCRA Part B permit. The detailed discussion of compliance deficiencies are summarized in Section 2.0. This includes items of significance that require action to ensure facility compliance with WAC, federal regulations, and WHC requirements. Outstanding issues exist for radioactive airborne effluent sampling and monitoring, radioactive liquid effluent sampling and monitoring, non-radioactive liquid effluent sampling and monitoring, less than 90 day waste storage tanks, and requirements for a permitted facility

  20. Trauma facilities in Denmark

    DEFF Research Database (Denmark)

    Weile, Jesper; Nielsen, Klaus; Primdahl, Stine C

    2018-01-01

    Background: Trauma is a leading cause of death among adults aged challenge. Evidence supports the centralization of trauma facilities and the use multidisciplinary trauma teams. Because knowledge is sparse on the existing distribution of trauma facilities...... and the organisation of trauma care in Denmark, the aim of this study was to identify all Danish facilities that care for traumatized patients and to investigate the diversity in organization of trauma management. Methods: We conducted a systematic observational cross-sectional study. First, all hospitals in Denmark...... were identified via online services and clarifying phone calls to each facility. Second, all trauma care manuals on all facilities that receive traumatized patients were gathered. Third, anesthesiologists and orthopedic surgeons on call at all trauma facilities were contacted via telephone...

  1. Improving extraction efficiency of the third integer resonant extraction using higher order multipoles

    Energy Technology Data Exchange (ETDEWEB)

    Brown, K. A. [Brookhaven National Lab. (BNL), Upton, NY (United States); Schoefer, V. [Brookhaven National Lab. (BNL), Upton, NY (United States); Tomizawa, M. [High Energy Accelerator Research Organization (KEK), Tsukuba (Japan)

    2017-03-09

    The new accelerator complex at J-PARC will operate with both high energy and very high intensity proton beams. With a design slow extraction efficiency of greater than 99% this facility will still be depositing significant beam power onto accelerator components [2]. To achieve even higher efficiencies requires some new ideas. The design of the extraction system and the accelerator lattice structure leaves little room for improvement using conventional techniques. In this report we will present one method for improving the slow extraction efficiency at J-PARC by adding duodecapoles or octupoles to the slow extraction system. We will review the theory of resonant extraction, describe simulation methods, and present the results of detailed simulations. From our investigations we find that we can improve extraction efficiency and thereby reduce the level of residual activation in the accelerator components and surrounding shielding.

  2. Synchrotron radiation facilities

    CERN Multimedia

    1972-01-01

    Particularly in the past few years, interest in using the synchrotron radiation emanating from high energy, circular electron machines has grown considerably. In our February issue we included an article on the synchrotron radiation facility at Frascati. This month we are spreading the net wider — saying something about the properties of the radiation, listing the centres where synchrotron radiation facilities exist, adding a brief description of three of them and mentioning areas of physics in which the facilities are used.

  3. Facility of aerosol filtration

    Energy Technology Data Exchange (ETDEWEB)

    Duverger de Cuy, G; Regnier, J

    1975-04-18

    Said invention relates to a facility of aerosol filtration, particularly of sodium aerosols. Said facility is of special interest for fast reactors where sodium fires involve the possibility of high concentrations of sodium aerosols which soon clog up conventional filters. The facility intended for continuous operation, includes at the pre-filtering stage, means for increasing the size of the aerosol particles and separating clustered particles (cyclone separator).

  4. Textiles Performance Testing Facilities

    Data.gov (United States)

    Federal Laboratory Consortium — The Textiles Performance Testing Facilities has the capabilities to perform all physical wet and dry performance testing, and visual and instrumental color analysis...

  5. Geodynamics Research Facility

    Data.gov (United States)

    Federal Laboratory Consortium — This GSL facility has evolved over the last three decades to support survivability and protective structures research. Experimental devices include three gas-driven...

  6. Materials Characterization Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Materials Characterization Facility enables detailed measurements of the properties of ceramics, polymers, glasses, and composites. It features instrumentation...

  7. Mobile Solar Tracker Facility

    Data.gov (United States)

    Federal Laboratory Consortium — NIST's mobile solar tracking facility is used to characterize the electrical performance of photovoltaic panels. It incorporates meteorological instruments, a solar...

  8. Proximal Probes Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Proximal Probes Facility consists of laboratories for microscopy, spectroscopy, and probing of nanostructured materials and their functional properties. At the...

  9. Geospatial Data Analysis Facility

    Data.gov (United States)

    Federal Laboratory Consortium — Geospatial application development, location-based services, spatial modeling, and spatial analysis are examples of the many research applications that this facility...

  10. Facility Environmental Management System

    Data.gov (United States)

    Federal Laboratory Consortium — This is the Web site of the Federal Highway Administration's (FHWA's) Turner-Fairbank Highway Research Center (TFHRC) facility Environmental Management System (EMS)....

  11. Heated Tube Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Heated Tube Facility at NASA GRC investigates cooling issues by simulating conditions characteristic of rocket engine thrust chambers and high speed airbreathing...

  12. Magnetics Research Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Magnetics Research Facility houses three Helmholtz coils that generate magnetic fields in three perpendicular directions to balance the earth's magnetic field....

  13. Transonic Experimental Research Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Transonic Experimental Research Facility evaluates aerodynamics and fluid dynamics of projectiles, smart munitions systems, and sub-munitions dispensing systems;...

  14. Engine Test Facility (ETF)

    Data.gov (United States)

    Federal Laboratory Consortium — The Air Force Arnold Engineering Development Center's Engine Test Facility (ETF) test cells are used for development and evaluation testing of propulsion systems for...

  15. Target Assembly Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Target Assembly Facility integrates new armor concepts into actual armored vehicles. Featuring the capability ofmachining and cutting radioactive materials, it...

  16. Pavement Testing Facility

    Data.gov (United States)

    Federal Laboratory Consortium — Comprehensive Environmental and Structural AnalysesThe ERDC Pavement Testing Facility, located on the ERDC Vicksburg campus, was originally constructed to provide an...

  17. Composite Structures Manufacturing Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Composite Structures Manufacturing Facility specializes in the design, analysis, fabrication and testing of advanced composite structures and materials for both...

  18. GPS Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Global Positioning System (GPS) Test Facility Instrumentation Suite (GPSIS) provides great flexibility in testing receivers by providing operational control of...

  19. Manufacturing Demonstration Facility (MDF)

    Data.gov (United States)

    Federal Laboratory Consortium — The U.S. Department of Energy Manufacturing Demonstration Facility (MDF) at Oak Ridge National Laboratory (ORNL) provides a collaborative, shared infrastructure to...

  20. Surplus Facilities Management Program

    International Nuclear Information System (INIS)

    Coobs, J.H.

    1983-01-01

    This is the second of two programs that are concerned with the management of surplus facilities. The facilities in this program are those related to commercial activities, which include the three surplus experimental and test reactors [(MSRE, HRE-2, and the Low Intensity Test Reactor (LITR)] and seven experimental loops at the ORR. The program is an integral part of the Surplus Facilities Management Program, which is a national program administered for DOE by the Richland Operations Office. Very briefly reported here are routine surveillance and maintenance of surplus radioactively contaminated DOE facilities awaiting decommissioning

  1. Imagery Data Base Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Imagery Data Base Facility supports AFRL and other government organizations by providing imagery interpretation and analysis to users for data selection, imagery...

  2. Neutron Therapy Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Neutron Therapy Facility provides a moderate intensity, broad energy spectrum neutron beam that can be used for short term irradiations for radiobiology (cells)...

  3. Universal Drive Train Facility

    Data.gov (United States)

    Federal Laboratory Consortium — This vehicle drive train research facility is capable of evaluating helicopter and ground vehicle power transmission technologies in a system level environment. The...

  4. High Combustion Research Facility

    Data.gov (United States)

    Federal Laboratory Consortium — At NETL's High-Pressure Combustion Research Facility in Morgantown, WV, researchers can investigate new high-pressure, high-temperature hydrogen turbine combustion...

  5. Catalytic Fuel Conversion Facility

    Data.gov (United States)

    Federal Laboratory Consortium — This facility enables unique catalysis research related to power and energy applications using military jet fuels and alternative fuels. It is equipped with research...

  6. Flexible Electronics Research Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Flexible Electronics Research Facility designs, synthesizes, tests, and fabricates materials and devices compatible with flexible substrates for Army information...

  7. DUPIC facility engineering

    Energy Technology Data Exchange (ETDEWEB)

    Park, J. J.; Lee, H. H.; Kim, K. H. and others

    2000-03-01

    The objectives of this study are (1) the refurbishment for PIEF(Post Irradiation Examination Facility) and M6 hot-cell in IMEF(Irradiated Material Examination Facility), (2) the establishment of the compatible facility for DUPIC fuel fabrication experiments which is licensed by government organization, and (3) the establishment of the transportation system and transportation cask for nuclear material between facilities. The report for this project describes following contents, such as objectives, necessities, scope, contents, results of current step, R and D plan in future and etc.

  8. Liquid-liquid extraction of U(VI), Np(V) et Th(IV) ions by two calix[4]arene carboxyls, and effect of Na+ and K+ alkaline ions

    International Nuclear Information System (INIS)

    Montavon, Gilles

    1996-01-01

    As the process mainly used for the reprocessing of nuclear wastes was the Purex process, this research thesis first presents this process and outlines that it allows the residual fissile matter to be recovered and reused for the fabrication of new fuel elements, but is neither efficient nor safe enough to separate fission and activation products. Thus, this thesis reports the study of extraction and selectivity properties of two compounds derived from the p-tert-butyl-calix[4]arene with respect to actinide ions such as Th(IV), U(VI) and Np(VI). The liquid-liquid extraction technique has been used with chloroform and 1,2-dichloroethane as solvents. After some generalities on actinides, calixarenes and the liquid-liquid extraction technique, and a presentation of the experimental method, the author reports and discusses the extractive properties of the studied calixarenes with respect to Na + and K + ions. Structural studies by proton NMR have been performed. He reports and discusses the liquid-liquid extraction on actinide ions when they are alone or in presence on Na + and K + alkaline ions [fr

  9. Facility effluent monitoring plan determinations for the 400 Area facilities

    International Nuclear Information System (INIS)

    Nickels, J.M.

    1991-09-01

    This Facility Effluent Monitoring Plan determination resulted from an evaluation conducted for the Westinghouse Hanford Company 400 Area facilities on the Hanford Site. The Facility Effluent Monitoring Plan determinations have been prepared in accordance with A Guide for Preparing Hanford Site Facility Effluent Monitoring Plans. Two major Westinghouse Hanford Company facilities in the 400 Area were evaluated: the Fast Flux Test Facility and the Fuels Manufacturing and examination Facility. The determinations were prepared by Westinghouse Hanford Company. Of these two facilities, only the Fast Flux Test Facility will require a Facility Effluent Monitoring Plan. 7 refs., 5 figs., 4 tabs

  10. 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

  11. Green facility location

    NARCIS (Netherlands)

    Velázquez Martínez, J.C.; Fransoo, J.C.; Bouchery, Y.; Corbett, C.J.; Fransoo, J.C.; Tan, T.

    2017-01-01

    Transportation is one of the main contributing factors of global carbon emissions, and thus, when dealing with facility location models in a distribution context, transportation emissions may be substantially higher than the emissions due to production or storage. Because facility location models

  12. A Remote WIRELESS Facility

    Directory of Open Access Journals (Sweden)

    Kees Uiterwijk

    2007-10-01

    Full Text Available Continuing need for available distance learning facilities has led to the development of a remote lab facility focusing on wireless technology. In the field of engineering there is a student need of gaining experience in set-up, monitoring and maintenance of 802.11A/B/G based wireless LAN environments.

  13. Medical cyclotron facilities

    International Nuclear Information System (INIS)

    1984-09-01

    This report examines the separate proposals from the Austin Hospital and the Australian Atomic Energy Commission for a medical cyclotron facility. The proponents have argued that a cyclotron facility would benefit Australia in areas of patient care, availability and export of radioisotopes, and medical research. Positron emission tomography (PET) and neutron beam therapy are also examined

  14. Global Environment Facility |

    Science.gov (United States)

    environment Countries pledge US$4.1 billion to the Global Environment Facility Ringtail lemur mom with two of paradise Nations rally to protect global environment Countries pledge US$4.1 billion to the Global Environment Facility Stockholm, Sweden birds-eye view Events GEF-7 Replenishment Trung Truong Son Landscapes

  15. Samarbejdsformer og Facilities Management

    DEFF Research Database (Denmark)

    Storgaard, Kresten

    Resultater fra en surveyundersøgelse om fordele og ulemper ved forskellige samarbejdsformer indenfor Facilities Management fremlægges.......Resultater fra en surveyundersøgelse om fordele og ulemper ved forskellige samarbejdsformer indenfor Facilities Management fremlægges....

  16. DUPIC facility engineering

    International Nuclear Information System (INIS)

    Park, J. J.; Lee, H. H.; Kim, K. H.

    2002-03-01

    With starting DUPIC fuel fabrication experiment by using spent fuels, 1) operation and refurbishment for DFDF (DUPIC fuel development facility), and 2) operation and improvement of transportation equipment for radioactive materials between facilities became the objectives of this study. This report describes objectives of the project, necessities, state of related technology, R and D scope, R and D results, proposal for application etc

  17. Economics of reusable facilities

    International Nuclear Information System (INIS)

    Antia, D.D.J.

    1992-01-01

    In this paper some of the different economic development strategies that can be used for reusable facilities in the UK, Norway, Netherlands and in some production sharing contracts are outlined. These strategies focus on an integrated decision analysis approach which considers development phasing, reservoir management, tax planning and where appropriate facility purchase, leasing, or sale and leaseback decisions

  18. Nuclear physics accelerator facilities

    International Nuclear Information System (INIS)

    1988-12-01

    This paper describes many of the nuclear physics heavy-ion accelerator facilities in the US and the research programs being conducted. The accelerators described are: Argonne National Laboratory--ATLAS; Brookhaven National Laboratory--Tandem/AGS Heavy Ion Facility; Brookhaven National Laboratory--Relativistic Heavy Ion Collider (RHIC) (Proposed); Continuous Electron Beam Accelerator Facility; Lawrence Berkeley Laboratory--Bevalac; Lawrence Berkeley Laboratory--88-Inch Cyclotron; Los Alamos National Laboratory--Clinton P. Anderson Meson Physics Facility (LAMPF); Massachusetts Institute of Technology--Bates Linear Accelerator Center; Oak Ridge National Laboratory--Holifield Heavy Ion Research Facility; Oak Ridge National Laboratory--Oak Ridge Electron Linear Accelerator; Stanford Linear Accelerator Center--Nuclear Physics Injector; Texas AandM University--Texas AandM Cyclotron; Triangle Universities Nuclear Laboratory (TUNL); University of Washington--Tandem/Superconducting Booster; and Yale University--Tandem Van de Graaff

  19. Researches at hadron experiment facility

    International Nuclear Information System (INIS)

    Sawada, Shinya

    2006-01-01

    Some of the nuclear, hadron and elementary particle experiments proposed to hadron experiment facility to use the extracted slow proton beam at J-PARC are overviewed. Characteristic feature of the facility is the secondary beam obtained from the intense proton beam. Nuclear hadron physics experiments and kaon rare decay experiments are presented here as the typical ones. Hypernuclear spectroscopy with S=-2 state is expected to be started as soon as the beam becomes available. The kaon bound systems not only with three nucleons like K-pnn but also more numerous like Li and Be are to be studied systematically. Bound states of two kaons using (K - , K + ) reaction will be challenged. Pentaquark will be searched for and its properties will be studied if it really exists. Nuclear structure studies from the view point of large Bjorken x are planned to be studied by irradiating hydrogen, deuteron or heavier targets with primary proton beam and analyzing generated muon pairs. Properties of vector mesons in nuclear matter are to be studied with the primary beam. Neutral kaon rare decay will be investigated to study CP nonconservation. Large progress of elementary particle physics is anticipated by using the intense proton beam at J-PARC. (S. Funahashi)

  20. Altitude simulation facility for testing large space motors

    Science.gov (United States)

    Katz, U.; Lustig, J.; Cohen, Y.; Malkin, I.

    1993-02-01

    This work describes the design of an altitude simulation facility for testing the AKM motor installed in the 'Ofeq' satellite launcher. The facility, which is controlled by a computer, consists of a diffuser and a single-stage ejector fed with preheated air. The calculations of performance and dimensions of the gas extraction system were conducted according to a one-dimensional analysis. Tests were carried out on a small-scale model of the facility in order to examine the design concept, then the full-scale facility was constructed and operated. There was good agreement among the results obtained from the small-scale facility, from the full-scale facility, and from calculations.

  1. Outline of NUCEF facility

    International Nuclear Information System (INIS)

    Takeshita, Isao

    1996-01-01

    NUCEF is a multipurpose research facility in the field of safety and advanced technology of nuclear fuel cycle back-end. Various experiment facilities and its supporting installations, in which nuclear fuel materials, radio isotopes and TRU elements can be handled, are arranged in more than one hundred rooms of two experiment buildings. Its construction was completed in middle of 1994 and hot experiments have been started since then. NUCEF is located on the site (30,000 m 2 ) of southeastern part in the Tokai Research Establishment of JAERI facing to the Pacific Ocean. The base of Experiment Buildings A and B was directly founded on the rock existing at 10-15 m below ground level taking the aseismatic design into consideration. Each building is almost same sized and composed of one basement and three floors of which area is 17,500 m 2 in total. In the basement, there are exhaust facilities of ventilation system, treatment system of solution fuel and radioactive waste solution and storage tanks of them. Major experiment facilities are located on the first or the second floors in each building. An air-inlet facility of ventilation system for each building is equipped on the third floor. Most of experiment facilities for criticality safety research including two critical facilities: Static Experiment Critical Facility (STACY) and Transient Experiment Critical Facility (TRACY) are installed in Experiment Building A. Experiment equipments for research on advanced fuel reprocessing process and on TRU waste management, which are named BECKY (Back End Fuel Cycle Key Elements Research Facility), are installed in laboratories and a-g cells in Experiment Building B. (J.P.N.)

  2. Simulation of facility operations and materials accounting for a combined reprocessing/MOX fuel fabrication facility

    International Nuclear Information System (INIS)

    Coulter, C.A.; Whiteson, R.; Zardecki, A.

    1991-01-01

    We are developing a computer model of facility operations and nuclear materials accounting for a facility that reprocesses spent fuel and fabricates mixed oxide (MOX) fuel rods and assemblies from the recovered uranium and plutonium. The model will be used to determine the effectiveness of various materials measurement strategies for the facility and, ultimately, of other facility safeguards functions as well. This portion of the facility consists of a spent fuel storage pond, fuel shear, dissolver, clarifier, three solvent-extraction stages with uranium-plutonium separation after the first stage, and product concentrators. In this facility area mixed oxide is formed into pellets, the pellets are loaded into fuel rods, and the fuel rods are fabricated into fuel assemblies. These two facility sections are connected by a MOX conversion line in which the uranium and plutonium solutions from reprocessing are converted to mixed oxide. The model of the intermediate MOX conversion line used in the model is based on a design provided by Mike Ehinger of Oak Ridge National Laboratory (private communication). An initial version of the simulation model has been developed for the entire MOX conversion and fuel fabrication sections of the reprocessing/MOX fuel fabrication facility, and this model has been used to obtain inventory difference variance estimates for those sections of the facility. A significant fraction of the data files for the fuel reprocessing section have been developed, but these data files are not yet complete enough to permit simulation of reprocessing operations in the facility. Accordingly, the discussion in the following sections is restricted to the MOX conversion and fuel fabrication lines. 3 tabs

  3. DUPIC facility engineering

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J S; Choi, J W; Go, W I; Kim, H D; Song, K C; Jeong, I H; Park, H S; Im, C S; Lee, H M; Moon, K H; Hong, K P; Lee, K S; Suh, K S; Kim, E K; Min, D K; Lee, J C; Chun, Y B; Paik, S Y; Lee, E P; Yoo, G S; Kim, Y S; Park, J C

    1997-09-01

    In the early stage of the project, a comprehensive survey was conducted to identify the feasibility of using available facilities and of interface between those facilities. It was found out that the shielded cell M6 interface between those facilities. It was found out that the shielded cell M6 of IMEF could be used for the main process experiments of DUPIC fuel fabrication in regard to space adequacy, material flow, equipment layout, etc. Based on such examination, a suitable adapter system for material transfer around the M6 cell was engineered. Regarding the PIEF facility, where spent PWR fuel assemblies are stored in an annex pool, disassembly devices in the pool are retrofitted and spent fuel rod cutting and shipping system to the IMEF are designed and built. For acquisition of casks for radioactive material transport between the facilities, some adaptive refurbishment was applied to the available cask (Padirac) based on extensive analysis on safety requirements. A mockup test facility was newly acquired for remote test of DUPIC fuel fabrication process equipment prior to installation in the M6 cell of the IMEF facility. (author). 157 refs., 57 tabs., 65 figs.

  4. DUPIC facility engineering

    International Nuclear Information System (INIS)

    Lee, J. S.; Choi, J. W.; Go, W. I.; Kim, H. D.; Song, K. C.; Jeong, I. H.; Park, H. S.; Im, C. S.; Lee, H. M.; Moon, K. H.; Hong, K. P.; Lee, K. S.; Suh, K. S.; Kim, E. K.; Min, D. K.; Lee, J. C.; Chun, Y. B.; Paik, S. Y.; Lee, E. P.; Yoo, G. S.; Kim, Y. S.; Park, J. C.

    1997-09-01

    In the early stage of the project, a comprehensive survey was conducted to identify the feasibility of using available facilities and of interface between those facilities. It was found out that the shielded cell M6 interface between those facilities. It was found out that the shielded cell M6 of IMEF could be used for the main process experiments of DUPIC fuel fabrication in regard to space adequacy, material flow, equipment layout, etc. Based on such examination, a suitable adapter system for material transfer around the M6 cell was engineered. Regarding the PIEF facility, where spent PWR fuel assemblies are stored in an annex pool, disassembly devices in the pool are retrofitted and spent fuel rod cutting and shipping system to the IMEF are designed and built. For acquisition of casks for radioactive material transport between the facilities, some adaptive refurbishment was applied to the available cask (Padirac) based on extensive analysis on safety requirements. A mockup test facility was newly acquired for remote test of DUPIC fuel fabrication process equipment prior to installation in the M6 cell of the IMEF facility. (author). 157 refs., 57 tabs., 65 figs

  5. STAR facility tritium accountancy

    International Nuclear Information System (INIS)

    Pawelko, R. J.; Sharpe, J. P.; Denny, B. J.

    2008-01-01

    The Safety and Tritium Applied Research (STAR) facility has been established to provide a laboratory infrastructure for the fusion community to study tritium science associated with the development of safe fusion energy and other technologies. STAR is a radiological facility with an administrative total tritium inventory limit of 1.5 g (14,429 Ci) [1]. Research studies with moderate tritium quantities and various radionuclides are performed in STAR. Successful operation of the STAR facility requires the ability to receive, inventory, store, dispense tritium to experiments, and to dispose of tritiated waste while accurately monitoring the tritium inventory in the facility. This paper describes tritium accountancy in the STAR facility. A primary accountancy instrument is the tritium Storage and Assay System (SAS): a system designed to receive, assay, store, and dispense tritium to experiments. Presented are the methods used to calibrate and operate the SAS. Accountancy processes utilizing the Tritium Cleanup System (TCS), and the Stack Tritium Monitoring System (STMS) are also discussed. Also presented are the equations used to quantify the amount of tritium being received into the facility, transferred to experiments, and removed from the facility. Finally, the STAR tritium accountability database is discussed. (authors)

  6. Sustainable Facilities Management

    DEFF Research Database (Denmark)

    Nielsen, Susanne Balslev; Elle, Morten; Hoffmann, Birgitte

    2004-01-01

    The Danish public housing sector has more than 20 years of experience with sustainable facilities management based on user involvement. The paper outlines this development in a historical perspective and gives an analysis of different approaches to sustainable facilities management. The focus...... is on the housing departments and strateies for the management of the use of resources. The research methods used are case studies based on interviews in addition to literature studies. The paper explores lessons to be learned about sustainable facilities management in general, and points to a need for new...

  7. WORKSHOPS: Hadron facilities

    International Nuclear Information System (INIS)

    Anon.

    1987-01-01

    'Hadron facilities' – high intensity (typically a hundred microamps), medium energy (30-60 GeV) machines producing intense secondary beams of pions, kaons, etc., are being widely touted as a profitable research avenue to supplement what is learned through the thrust for higher and higher energies. This interest was reflected at an International Workshop on Hadron Facility Technology, held in Santa Fe, New Mexico. As well as invited talks describing the various projects being pushed in the US, Europe and Japan, the meeting included working groups covering linacs, beam dynamics, hardware, radiofrequency, polarized beams and experimental facilities

  8. Radioactive facilities classification criteria

    International Nuclear Information System (INIS)

    Briso C, H.A.; Riesle W, J.

    1992-01-01

    Appropriate classification of radioactive facilities into groups of comparable risk constitutes one of the problems faced by most Regulatory Bodies. Regarding the radiological risk, the main facts to be considered are the radioactive inventory and the processes to which these radionuclides are subjected. Normally, operations are ruled by strict safety procedures. Thus, the total activity of the radionuclides existing in a given facility is the varying feature that defines its risk. In order to rely on a quantitative criterion and, considering that the Annual Limits of Intake are widely accepted references, an index based on these limits, to support decisions related to radioactive facilities, is proposed. (author)

  9. Wind Energy Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Laurie, Carol

    2017-02-01

    This book takes readers inside the places where daily discoveries shape the next generation of wind power systems. Energy Department laboratory facilities span the United States and offer wind research capabilities to meet industry needs. The facilities described in this book make it possible for industry players to increase reliability, improve efficiency, and reduce the cost of wind energy -- one discovery at a time. Whether you require blade testing or resource characterization, grid integration or high-performance computing, Department of Energy laboratory facilities offer a variety of capabilities to meet your wind research needs.

  10. Test and User Facilities | NREL

    Science.gov (United States)

    Test and User Facilities Test and User Facilities Our test and user facilities are available to | L | M | N | O | P | Q | R | S | T | U | V | W | X | Y | Z B Battery Thermal and Life Test Facility Biochemical Conversion Pilot Plant C Controllable Grid Interface Test System D Dynamometer Test Facilities

  11. Radiation protection studies for the SHiP facility

    CERN Document Server

    Strabel, Claudia Christina; Vincke, Helmut

    2015-01-01

    The enlarged scope of the recently proposed experiment to search for Heavy Neutral Leptons, SPSC-EOI-010, is a general purpose fixed target facility which in the initial phase is aimed at a general Search for Hidden Particles (SHiP) as well as tau neutrino physics. This report summarizes radiation protection considerations for the SHiP facility and the primary beam extraction for SHiP.

  12. Aviation Flight Support Facility

    Data.gov (United States)

    Federal Laboratory Consortium — This facility consists of a 75' x 200' hanger with two adjacent helicopter pads located at Felker Army Airfield on Fort Eustis. A staff of Government and contractor...

  13. Airborne & Field Sensors Facilities

    Data.gov (United States)

    Federal Laboratory Consortium — RTTC facilities include an 800' x 60' paved UAV operational area, clearapproach/departure zone, concrete pads furnished with 208VAC, 3 phase,200 amp power, 20,000 sq...

  14. Field Research Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Field Research Facility (FRF) located in Duck, N.C. was established in 1977 to support the U.S. Army Corps of Engineers' coastal engineering mission. The FRF is...

  15. Air Data Calibration Facility

    Data.gov (United States)

    Federal Laboratory Consortium — This facility is for low altitude subsonic altimeter system calibrations of air vehicles. Mission is a direct support of the AFFTC mission. Postflight data merge is...

  16. Robotics Research Facility

    Data.gov (United States)

    Federal Laboratory Consortium — This 60 feet x 100 feet structure on the grounds of the Fort Indiantown Gap Pennsylvania National Guard (PNG) Base is a mixed-use facility comprising office space,...

  17. Ballistic Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Ballistic Test Facility is comprised of two outdoor and one indoor test ranges, which are all instrumented for data acquisition and analysis. Full-size aircraft...

  18. Concrete Research Facility

    Data.gov (United States)

    Federal Laboratory Consortium — This is a 20,000-sq ft laboratory that supports research on all aspects of concrete and materials technology. The staff of this facility offer wide-ranging expertise...

  19. Climatic Environmental Test Facilities

    Data.gov (United States)

    Federal Laboratory Consortium — RTTC has an extensive suite of facilities for supporting MIL-STD-810 testing, toinclude: Temperature/Altitude, Rapid Decompression, Low/High Temperature,Temperature...

  20. HNF - Helmholtz Nano Facility

    Directory of Open Access Journals (Sweden)

    Wolfgang Albrecht

    2017-05-01

    Full Text Available The Helmholtz Nano Facility (HNF is a state-of-the-art cleanroom facility. The cleanroom has ~1100 m2 with cleanroom classes of DIN ISO 1-3. HNF operates according to VDI DIN 2083, Good Manufacturing Practice (GMP and aquivalent to Semiconductor Industry Association (SIA standards. HNF is a user facility of Forschungszentrum Jülich and comprises a network of facilities, processes and systems for research, production and characterization of micro- and nanostructures. HNF meets the basic supply of micro- and nanostructures for nanoelectronics, fluidics. micromechanics, biology, neutron and energy science, etc.. The task of HNF is rapid progress in nanostructures and their technology, offering efficient access to infrastructure and equipment. HNF gives access to expertise and provides resources in production, synthesis, characterization and integration of structures, devices and circuits. HNF covers the range from basic research to application oriented research facilitating a broad variety of different materials and different sample sizes.

  1. Advanced Microscopy Facility

    Data.gov (United States)

    Federal Laboratory Consortium — FUNCTION: Provides a facility for high-resolution studies of complex biomolecular systems. The goal is an understanding of how to engineer biomolecules for various...

  2. Electra Laser Facility

    Data.gov (United States)

    Federal Laboratory Consortium — FUNCTION: The Electra Laser Facility is used to develop the science and technology needed to develop a reliable, efficient, high-energy, repetitively pulsed krypton...

  3. Mark 1 Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Mark I Test Facility is a state-of-the-art space environment simulation test chamber for full-scale space systems testing. A $1.5M dollar upgrade in fiscal year...

  4. Coastal Harbors Modeling Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Coastal Harbors Modeling Facility is used to aid in the planning of harbor development and in the design and layout of breakwaters, absorbers, etc.. The goal is...

  5. Corrosion Testing Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Corrosion Testing Facility is part of the Army Corrosion Office (ACO). It is a fully functional atmospheric exposure site, called the Corrosion Instrumented Test...

  6. Skilled Nursing Facility PPS

    Data.gov (United States)

    U.S. Department of Health & Human Services — Section 4432(a) of the Balanced Budget Act (BBA) of 1997 modified how payment is made for Medicare skilled nursing facility (SNF) services. Effective with cost...

  7. Frost Effects Research Facility

    Data.gov (United States)

    Federal Laboratory Consortium — Full-scale study in controlled conditionsThe Frost Effects Research Facility (FERF) is the largest refrigerated warehouse in the United States that can be used for a...

  8. GPS Satellite Simulation Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The GPS satellite simulation facility consists of a GPS satellite simulator controlled by either a Silicon Graphics Origin 2000 or PC depending upon unit under test...

  9. VT Telecommunication Facilities

    Data.gov (United States)

    Vermont Center for Geographic Information — (Link to Metadata) The UtilityTelecom_TELEFAC data layer contains points which are intended to represent the location of telecommunications facilities (towers and/or...

  10. Laser Guidance Analysis Facility

    Data.gov (United States)

    Federal Laboratory Consortium — This facility, which provides for real time, closed loop evaluation of semi-active laser guidance hardware, has and continues to be instrumental in the development...

  11. The Birmingham Irradiation Facility

    CERN Document Server

    Dervan, P; Hodgson, P; Marin-Reyes, H; Wilson, J

    2013-01-01

    At the end of 2012 the proton irradiation facility at the CERN PS [1] will shut down for two years. With this in mind, we have been developing a new ATLAS scanning facility at the University of Birmingham Medical Physics cyclotron. With proton beams of energy approximately 30 MeV, fluences corresponding to those of the upgraded Large Hadron Collider (HL-LHC) can be reached conveniently. The facility can be used to irradiate silicon sensors, optical components and mechanical structures (e.g. carbon fibre sandwiches) for the LHC upgrade programme. Irradiations of silicon sensors can be carried out in a temperature controlled cold box that can be scanned through the beam. The facility is described in detail along with the first tests carried out with mini (1 x 1 cm^2 ) silicon sensors.

  12. Advanced Microanalysis Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Advanced Microanalysis Facility fully integrates capabilities for chemical and structural analysis of electronic materials and devices for the U.S. Army and DoD....

  13. The Birmingham Irradiation Facility

    International Nuclear Information System (INIS)

    Dervan, P.; French, R.; Hodgson, P.; Marin-Reyes, H.; Wilson, J.

    2013-01-01

    At the end of 2012 the proton irradiation facility at the CERN PS will shut down for two years. With this in mind, we have been developing a new ATLAS scanning facility at the University of Birmingham Medical Physics cyclotron. With proton beams of energy approximately 30 MeV, fluences corresponding to those of the upgraded Large Hadron Collider (HL-LHC) can be reached conveniently. The facility can be used to irradiate silicon sensors, optical components and mechanical structures (e.g. carbon fibre sandwiches) for the LHC upgrade programme. Irradiations of silicon sensors can be carried out in a temperature controlled cold box that can be scanned through the beam. The facility is described in detail along with the first tests carried out with mini (1×1 cm 2 ) silicon sensors

  14. Decontamination of nuclear facilities

    International Nuclear Information System (INIS)

    1982-01-01

    Thirty-seven papers were presented at this conference in five sessions. Topics covered include regulation, control and consequences of decontamination; decontamination of components and facilities; chemical and non-chemical methods of decontamination; and TMI decontamination experience

  15. Pit Fragment Facility

    Data.gov (United States)

    Federal Laboratory Consortium — This facility contains two large (20 foot high by 20 foot diameter) double walled steel tubs in which experimental munitions are exploded while covered with sawdust....

  16. Joint Computing Facility

    Data.gov (United States)

    Federal Laboratory Consortium — Raised Floor Computer Space for High Performance ComputingThe ERDC Information Technology Laboratory (ITL) provides a robust system of IT facilities to develop and...

  17. Coastal Inlet Model Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Coastal Inlet Model Facility, as part of the Coastal Inlets Research Program (CIRP), is an idealized inlet dedicated to the study of coastal inlets and equipped...

  18. Wind Tunnel Testing Facilities

    Data.gov (United States)

    Federal Laboratory Consortium — NASA Ames Research Center is pleased to offer the services of our premier wind tunnel facilities that have a broad range of proven testing capabilities to customers...

  19. Space Power Facility (SPF)

    Data.gov (United States)

    Federal Laboratory Consortium — The Space Power Facility (SPF) houses the world's largest space environment simulation chamber, measuring 100 ft. in diameter by 122 ft. high. In this chamber, large...

  20. Airborne Evaluation Facility

    Data.gov (United States)

    Federal Laboratory Consortium — AFRL's Airborne Evaluation Facility (AEF) utilizes Air Force Aero Club resources to conduct test and evaluation of a variety of equipment and concepts. Twin engine...