WorldWideScience

Sample records for coolant cleanup systems

  1. Improvements of primary coolant shutdown chemistry and reactor coolant system cleanup

    Energy Technology Data Exchange (ETDEWEB)

    Gaudard, G.; Gilles, B.; Mesnage, F. [EDF/GDL (France); Cattant, F. [EDF R and D (France)

    2002-07-01

    In the framework of a radiation exposure management program entitled <>, EDF aims at decreasing the mass dosimetry of nuclear power plants workers. So, the annual dose per unit, which has improved from 2.44 m.Sv in 1991 to 1.08 in 2000, should target 0.8 mSv in the year 2005 term in order to meet the results of the best nuclear operators. One of the guidelines for irradiation source term reduction is the optimization of operation parameters, including reactor coolant system (RCS) chemistry in operation, RCS shutdown chemistry and RCS cleanup improvement. This paper presents the EDF strategy for the shutdown and start up RCS chemistry optimization. All the shutdown modes have been reviewed and for each of them, the chemical specifications will be fine tuned. A survey of some US PWRs shutdown practices has been conducted for an acid and reducing shutdown chemistry implementation test at one EDF unit. This survey shows that deviating from the EPRI recommended practice for acid and reducing shutdown chemistry is possible and that critical path impact can be minimized. The paper also presents some investigations about soluble and insoluble species behavior and characterization; the study focuses here on {sup 110m}Ag, {sup 122}Sb, {sup 124}Sb and iodine contamination. Concerning RCS cleanup improvement, the paper presents two studies. The first one highlights some limited design modifications that are either underway or planned, for an increased flow rate during the most critical periods of the shutdown. The second one focuses on the strategy EDF envisions for filters and resins selection criteria. Matching the study on contaminants behavior with the study of filters and resins selection criteria should allow improving the cleanup efficiency. (authors)

  2. Environmentally Friendly Coolant System

    Energy Technology Data Exchange (ETDEWEB)

    David Jackson Principal Investigator

    2011-11-08

    Energy reduction through the use of the EFCS is most improved by increasing machining productivity. Throughout testing, nearly all machining operations demonstrated less land wear on the tooling when using the EFCS which results in increased tool life. These increases in tool life advance into increased productivity. Increasing productivity reduces cycle times and therefore reduces energy consumption. The average energy savings by using the EFCS in these machining operations with these materials is 9%. The advantage for end milling stays with flood coolant by about 6.6% due to its use of a low pressure pump. Face milling and drilling are both about 17.5% less energy consumption with the EFCS than flood coolant. One additional result of using the EFCS is improved surface finish. Certain machining operations using the EFCS result in a smoother surface finish. Applications where finishing operations are required will be able to take advantage of the improved finish by reducing the time or possibly eliminating completely one or more finishing steps and thereby reduce their energy consumption. Some machining operations on specific materials do not show advantages for the EFCS when compared to flood coolants. More information about these processes will be presented later in the report. A key point to remember though, is that even with equivalent results, the EFCS is replacing petroleum based coolants whose production produces GHG emissions and create unsafe work environments.

  3. Development of teleoperated cleanup system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ki Ho; Park, J. J.; Yang, M. S.; Kwon, H. J

    2005-01-01

    This report describes the development of a teleoperated cleanup system for use in a highly radioactive environment of DFDF(DUPIC Fuel Demonstration Facility) at KAERI where direct human access to the in-cell is strictly limited. The teleoperated cleanup system was designed to remotely remove contaminants placed or fixed on the floor surface of the hot-cell by mopping them with wet cloth. This cleanup system consists of a mopping slave, a mopping master and a control console. The mopping slave located at the in-cell comprises a mopping tool with a mopping cloth and a mobile platform, which were constructed in modules to facilitate maintenance. The mopping master that is an input device to control the mopping slave has kinematic dissimilarity with the mopping slave. The control console provides a means of bilateral control flows and communications between the mopping master and the mopping slave. In operation, the human operator from the out-of-cell performs a series of decontamination tasks remotely by manipulating the mopping slave located in-cell via a mopping master, having a sense of real mopping. The environmental and mechanical design considerations, and control systems of the developed teleoperated cleanup system are also described.

  4. Water coolant supply in relation to different ultrasonic scaler systems, tips and coolant settings

    NARCIS (Netherlands)

    Koster, T.J.G.; Timmerman, M.F.; Feilzer, A.J.; van der Velden, U.; van der Weijden, F.A.

    2009-01-01

    Objective: This study evaluated "in vitro" the consistency of the water coolant supply for five ultrasonic scaler systems in relation to the tip type and different coolant settings. Material and Methods: The systems were: EMS PM-400, EMS PM-600, Satelec P-max, Dürr Vector and Dentsply Cavitron. For

  5. Assessment, Cleanup and Redevelopment Exchange System (ACRES)

    Data.gov (United States)

    U.S. Environmental Protection Agency — The Assessment, Cleanup and Redevelopment Exchange System (ACRES) is an online database for Brownfields Grantees to electronically submit data directly to EPA.

  6. ISS Internal Active Thermal Control System (IATCS) Coolant Remediation Project

    Science.gov (United States)

    Morrison, Russell H.; Holt, Mike

    2005-01-01

    The IATCS coolant has experienced a number of anomalies in the time since the US Lab was first activated on Flight 5A in February 2001. These have included: 1) a decrease in coolant pH, 2) increases in inorganic carbon, 3) a reduction in phosphate buffer concentration, 4) an increase in dissolved nickel and precipitation of nickel salts, and 5) increases in microbial concentration. These anomalies represent some risk to the system, have been implicated in some hardware failures and are suspect in others. The ISS program has conducted extensive investigations of the causes and effects of these anomalies and has developed a comprehensive program to remediate the coolant chemistry of the on-orbit system as well as provide a robust and compatible coolant solution for the hardware yet to be delivered. The remediation steps include changes in the coolant chemistry specification, development of a suite of new antimicrobial additives, and development of devices for the removal of nickel and phosphate ions from the coolant. This paper presents an overview of the anomalies, their known and suspected system effects, their causes, and the actions being taken to remediate the coolant.

  7. SUBSTATIONS OF DISTRICT HEATING SYSTEMS WITH PULSE COOLANT CIRCULATION

    Directory of Open Access Journals (Sweden)

    Andrey N. Makeev

    2017-01-01

    Full Text Available Abstract. Objectives The aim of the study is to generalise the results of the application of technologies and means for organising pulse coolant flow within a district heating system in order to increase its energy efficiency based on the organisation of local hydraulic shocks and the subsequent use of their energy to ensure the purification of heat energy equipment, intensify the heat transfer process and realise the possibility of transforming the available head from one hydraulic circuit to another. Methods Substations connecting the thermal power installations of consumers with heat networks via dependent and independent schemes are analytically generalised. The use of pulse coolant circulation is proposed as a means of overcoming identified shortcomings. Results Principal schemes of substations with pulse coolant circulation for dependent and independent connection of thermal power installations are detailed. A detailed description of their operation is given. The advantages of using pulse coolant circulation in substations are shown. The materials reflecting the results of the technical implementation and practical introduction of this technology are presented. Conclusion Theoretical analysis of the operation of the basic schemes of substations with pulse coolant circulation and the results of their practical application, as well as the materials of scientific works devoted to the use of the energy of a hydraulic impact and the study of the effect of pulse coolant flow on thermal and hydrodynamic processes, have yielded a combination of factors reflecting technical and economic rationality of application of pulse coolant circulation. 

  8. Expert system for online surveillance of nuclear reactor coolant pumps

    Energy Technology Data Exchange (ETDEWEB)

    Gross, K.C.; Singer, R.M.; Humenik, K.E.

    1992-12-31

    This report describes an expert system for online surveillance of nuclear reactor coolant pumps. This system provides a means for early detection of pump or sensor degradation. Degradation is determined through the use of a statistical analysis technique, sequential probability ratio test, applied to information from several sensors which are responsive to differing physical parameters. The results of sequential testing of the data provide the operator with an early warning of possible sensor or pump failure.

  9. Design of Reactor Coolant Pump Seal Online Monitoring System

    Energy Technology Data Exchange (ETDEWEB)

    Ah, Sang Ha; Chang, Soon Heung [KAIST, Daejeon (Korea, Republic of); Lee, Song Kyu [Korea Power Engineering Co., Yongin (Korea, Republic of)

    2008-05-15

    As a part of a Department of Korea Power Engineering Co., (KOPEC) Project, Statistical Quality Control techniques have been applied to many aspects of industrial engineering. An application to nuclear power plant maintenance and control is also presented that can greatly improve plant safety. As a demonstration of such an approach, a specific system is analyzed: the reactor coolant pumps (RCPs) and the fouling resistance of heat exchanger. This research uses Shewart X-bar, R charts, Cumulative Sum charts (CUSUM), and Sequential Probability Ratio Test (SPRT) to analyze the process for the state of statistical control. And the Control Chart Analyzer (CCA) has been made to support these analyses that can make a decision of error in process. The analysis shows that statistical process control methods can be applied as an early warning system capable of identifying significant equipment problems well in advance of traditional control room alarm indicators. Such a system would provide operators with enough time to respond to possible emergency situations and thus improve plant safety and reliability. RCP circulates reactor coolant to transfer heat from the reactor to the steam generators. RCP seals are in the pressure part of reactor coolant system, so if it breaks, it can cause small break LOCA. And they are running on high pressure, and high temperature, so they can be easily broken. Since the reactor coolant pumps operate within the containment building, physical access to the pumps occurs only during refueling outages. Engineers depend on process variables transmitted to the control room and through the station's data historian to assess the pumps' condition during normal operation.

  10. System Study: High-Pressure Coolant Injection 1998-2014

    Energy Technology Data Exchange (ETDEWEB)

    Schroeder, John Alton [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk Assessment and Management Services Dept.

    2015-12-01

    This report presents an unreliability evaluation of the high-pressure coolant injection system (HPCI) at 25 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing or decreasing trends were identified in the HPCI results.

  11. System Study: High-Pressure Coolant Injection 1998–2013

    Energy Technology Data Exchange (ETDEWEB)

    Schroeder, John Alton [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk Assessment and Management Services Dept.

    2015-01-31

    This report presents an unreliability evaluation of the high-pressure coolant injection system (HPCI) at 25 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2013 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10-year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing or decreasing trends were identified in the HPCI results.

  12. Evaluation of beach cleanup effects using linear system analysis.

    Science.gov (United States)

    Kataoka, Tomoya; Hinata, Hirofumi

    2015-02-15

    We established a method for evaluating beach cleanup effects (BCEs) based on a linear system analysis, and investigated factors determining BCEs. Here we focus on two BCEs: decreasing the total mass of toxic metals that could leach into a beach from marine plastics and preventing the fragmentation of marine plastics on the beach. Both BCEs depend strongly on the average residence time of marine plastics on the beach (τ(r)) and the period of temporal variability of the input flux of marine plastics (T). Cleanups on the beach where τ(r) is longer than T are more effective than those where τ(r) is shorter than T. In addition, both BCEs are the highest near the time when the remnants of plastics reach the local maximum (peak time). Therefore, it is crucial to understand the following three factors for effective cleanups: the average residence time, the plastic input period and the peak time.

  13. Nonflammable coolants for space vehicle environmental control systems Compatibility of component materials with selected dielectric fluids.

    Science.gov (United States)

    Howard, R. T.; Korpolinski, T. S.; Mace, E. W.

    1971-01-01

    This paper summarizes a 4-year effort to evaluate and implement a nonflammable substitute coolant for application in the Saturn instrument unit (IU) environmental control system (ECS). Discussed are candidate material evaluations, detailed investigations of the properties of the coolant selected, and a summary of the implementation into a flight vehicle.

  14. Fuel cell cooling system using a non-dielectric coolant

    Energy Technology Data Exchange (ETDEWEB)

    Grevstad, P.E.; Gelting, R.L.

    1976-07-13

    A cooler for removing waste heat from a stack of fuel cells uses a non-dielectric coolant which is carried in a plurality of tubes passing through one or more separator plates in the stack. Preferably the coolant is water so that heat removal is by evaporation of the water within the tubes by boiling. The tubes are electrically insulated from the cells by a coating of dielectric material such as polytetrafluoroethylene. In one embodiment of the invention the cooler tubes are connected to the stack coolant supply conduits by dielectric hoses having a high length to diameter ratio to provide a several hundred thousand ohm impedance path in case of a flaw in the protective dielectric coating, in order that a short circuit of the stack does not occur.

  15. Purification of liquid metal systems with sodium coolant from oxygen using getters

    Science.gov (United States)

    Kozlov, F. A.; Konovalov, M. A.; Sorokin, A. P.

    2016-05-01

    For increasing the safety and economic parameters of nuclear power stations (NPSs) with sodium coolant, it was decided to install all systems contacting radioactive sodium, including purification systems of circuit I, in the reactor vessel. The performance and capacity of cold traps (CTs) (conventional element of coolant purification systems) in these conditions are limited by their volume. It was proposed to use hot traps (HTs) in circuit I for coolant purification from oxygen. It was demonstrated that, at rated parameters of the installation when the temperature of the coolant streamlining the getter (gas absorber) is equal to 550°C, the hot trap can provide the required coolant purity. In shutdown modes at 250-300°C, the performance of the hot trap is reduced by four orders of magnitude. Possible HT operation regimes for shutdown modes and while reaching rated parameters were proposed and analyzed. Basic attention was paid to purification modes at power rise after commissioning and accidental contamination of the coolant when the initial oxygen concentration in it reached 25 mln-1. It was demonstrated that the efficiency of purification systems can be increased using HTs with the getter in the form of a foil or granules. The possibility of implementing the "fast purification" mode in which the coolant is purified simultaneously with passing over from the shutdown mode to the rated parameters was substantiated.

  16. Selection of an Alternate Biocide for the ISS Internal Thermal Control System Coolant, Phase 2

    Science.gov (United States)

    Wilson, Mark E.; Cole, Harold; Weir, Natalee; Oehler, Bill; Steele, John; Varsik, Jerry; Lukens, Clark

    2004-01-01

    The ISS (International Space Station) ITCS (Internal Thermal Control System) includes two internal coolant loops that utilize an aqueous based coolant for heat transfer. A silver salt biocide had previously been utilized as an additive in the coolant formulation to control the growth and proliferation of microorganisms within the coolant loops. Ground-based and in-flight testing demonstrated that the silver salt was rapidly depleted, and did not act as an effective long-term biocide. Efforts to select an optimal alternate biocide for the ITCS coolant application have been underway and are now in the final stages. An extensive evaluation of biocides was conducted to down-select to several candidates for test trials and was reported on previously. Criteria for that down-select included: the need for safe, non-intrusive implementation and operation in a functioning system; the ability to control existing planktonic and biofilm residing microorganisms; a negligible impact on system-wetted materials of construction; and a negligible reactivity with existing coolant additives. Candidate testing to provide data for the selection of an optimal alternate biocide is now in the final stages. That testing has included rapid biocide effectiveness screening using Biolog MT2 plates to determine minimum inhibitory concentration (amount that will inhibit visible growth of microorganisms), time kill studies to determine the exposure time required to completely eliminate organism growth, materials compatibility exposure evaluations, coolant compatibility studies, and bench-top simulated coolant testing. This paper reports the current status of the effort to select an alternate biocide for the ISS ITCS coolant. The results of various test results to select the optimal candidate are presented.

  17. Shoreline oil cleanup, recovery and treatment evaluation system (SOCRATES)

    Energy Technology Data Exchange (ETDEWEB)

    Rusin, J.; Lunel, T.; Sommerville, M. [National Environmental Technology Centre, Culham (United Kingdom); Tyler, A.; Marshall, I. [BMT Marine Information Systems Ltd., Hampshire (United Kingdom)

    1996-09-01

    A beach cleanup computer system was developed to mitigate the impact of shoreline oiling. The program, entitled SOCRATES, was meant to determine the most suitable cleanup methodologies for a range of different spill scenarios. The development, operation and capabilities of SOCRATES was described, with recent examples of successful use during the Sea Empress spill. The factors which influenced decision making and which were central to the numerical solution were: (1) the volumetric removal rate of oil, (2) area removal rate of oil, (3) length of oil slick removed per hour, (4) volumetric removal rate of oily waste, (5) area of the oil slick, (6) length of the oil slick, (7) volume of liquid emulsion, and (8) length of beach. 14 figs.

  18. ISS Internal Active Thermal Control System (IATCS) Coolant Remediation Project -2006 Update

    Science.gov (United States)

    Morrison, Russell H.; Holt, Mike

    2006-01-01

    The IATCS coolant has experienced a number of anomalies in the time since the US Lab was first activated on Flight 5A in February 2001. These have included: 1) a decrease in coolant pH, 2) increases in inorganic carbon, 3) a reduction in phosphate concentration, 4) an increase in dissolved nickel and precipitation of nickel salts, and 5) increases in microbial concentration. These anomalies represent some risk to the system, have been implicated in some hardware failures and are suspect in others. The ISS program has conducted extensive investigations of the causes and effects of these anomalies and has developed a comprehensive program to remediate the coolant chemistry of the on-orbit system as well as provide a robust and compatible coolant solution for the hardware yet to be delivered. This paper presents a status of the coolant stability over the past year as well as results from destructive analyses of hardware removed from the on-orbit system and the current approach to coolant remediation.

  19. Membrane systems and their use in nuclear power plants. Treatment of primary coolant

    Energy Technology Data Exchange (ETDEWEB)

    Kus, Pavel; Bartova, Sarka; Skala, Martin; Vonkova, Katerina [Research Centre Rez, Husinec-Rez (Czech Republic). Technological Circuits Innovation Dept.; Zach, Vaclav; Kopa, Roman [CEZ a.s., Temelin (Czech Republic). Nuclear Power Plant Temelin

    2016-03-15

    In nuclear power plants, drained primary coolant containing boric acid is currently treated in the system of evaporators and by ion exchangers. Replacement of the system of evaporators by membrane system (MS) will result in lower operating cost mainly due to lower operation temperature. In membrane systems the feed primary coolant is separated into two output streams: retentate and permeate. Retentate stream consists of the concentrated boric acid solution together with other components, while permeate stream consists of purified water. Results are presented achieved by testing a pilot-plant unit of reverse osmosis in nuclear power plant (NPP) Temelin.

  20. The state of the art on zinc addition effect in the nuclear reactor coolant system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, U. C.; Sung, K. W.; Kim, K. R.; Paek, S.; Maeng, W. Y

    1999-12-01

    Zinc addition to the primary coolant appears promising with regard to reducing radiation dose rate, and is being used in several plants. Zinc acts to inhibit the corrosion of stainless steel by forming a thin protective film. This oxide film, with no associated increase in cobalt concentration within the film, thereby lowers the dose rate. This report on the state of art presents an overview of the zinc addition to the reactor coolant to reduce the primary system dose rate. This report discusses the effect of zinc addition for BWRs and PWRs, the thermodynamic of zinc chemistry, and the effect of zinc addition on material corrosion. (author)

  1. An Improved Design for Air Removal from Aerospace Fluid Loop Coolant Systems

    Science.gov (United States)

    Ritchie, Stephen M. C.; Holladay, Jon B.; Holt, J. Mike; Clark, Dallas W.

    2003-01-01

    Aerospace applications with requirements for large capacity heat removal (launch vehicles, platforms, payloads, etc.) typically utilize a liquid coolant fluid as a transport media to increase efficiency and flexibility in the vehicle design. An issue with these systems however, is susceptibility to the presence of noncondensable gas (NCG) or air. The presence of air in a coolant loop can have numerous negative consequences, including loss of centrifugal pump prime, interference with sensor readings, inhibition of heat transfer, and coolant blockage to remote systems. Hardware ground processing to remove this air is also cumbersome and time consuming which continuously drives recurring costs. Current systems for maintaining the system free of air are tailored and have demonstrated only moderate success. An obvious solution to these problems is the development and advancement of a passive gas removal device, or gas trap, that would be installed in the flight cooling system simplifying the initial coolant fill procedure and also maintaining the system during operations. The proposed device would utilize commercially available membranes thus increasing reliability and reducing cost while also addressing both current and anticipated applications. In addition, it maintains current pressure drop, water loss, and size restrictions while increasing tolerance for pressure increases due to gas build-up in the trap.

  2. Lead Coolant Test Facility Systems Design, Thermal Hydraulic Analysis and Cost Estimate

    Energy Technology Data Exchange (ETDEWEB)

    Soli Khericha; Edwin Harvego; John Svoboda; Ryan Dalling

    2012-01-01

    The Idaho National Laboratory prepared a preliminary technical and functional requirements (T&FR), thermal hydraulic design and cost estimate for a lead coolant test facility. The purpose of this small scale facility is to simulate lead coolant fast reactor (LFR) coolant flow in an open lattice geometry core using seven electrical rods and liquid lead or lead-bismuth eutectic coolant. Based on review of current world lead or lead-bismuth test facilities and research needs listed in the Generation IV Roadmap, five broad areas of requirements were identified as listed: (1) Develop and Demonstrate Feasibility of Submerged Heat Exchanger; (2) Develop and Demonstrate Open-lattice Flow in Electrically Heated Core; (3) Develop and Demonstrate Chemistry Control; (4) Demonstrate Safe Operation; and (5) Provision for Future Testing. This paper discusses the preliminary design of systems, thermal hydraulic analysis, and simplified cost estimate. The facility thermal hydraulic design is based on the maximum simulated core power using seven electrical heater rods of 420 kW; average linear heat generation rate of 300 W/cm. The core inlet temperature for liquid lead or Pb/Bi eutectic is 4200 C. The design includes approximately seventy-five data measurements such as pressure, temperature, and flow rates. The preliminary estimated cost of construction of the facility is $3.7M (in 2006 $). It is also estimated that the facility will require two years to be constructed and ready for operation.

  3. Lead coolant test facility systems design, thermal hydraulic analysis and cost estimate

    Energy Technology Data Exchange (ETDEWEB)

    Khericha, Soli, E-mail: slk2@inel.gov [Battelle Energy Alliance, LLC, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Harvego, Edwin; Svoboda, John; Evans, Robert [Battelle Energy Alliance, LLC, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Dalling, Ryan [ExxonMobil Gas and Power Marketing, Houston, TX 77069 (United States)

    2012-01-15

    The Idaho National Laboratory prepared a preliminary technical and functional requirements (T and FR), thermal hydraulic design and cost estimate for a lead coolant test facility. The purpose of this small scale facility is to simulate lead coolant fast reactor (LFR) coolant flow in an open lattice geometry core using seven electrical rods and liquid lead or lead-bismuth eutectic coolant. Based on review of current world lead or lead-bismuth test facilities and research needs listed in the Generation IV Roadmap, five broad areas of requirements were identified as listed below: Bullet Develop and demonstrate feasibility of submerged heat exchanger. Bullet Develop and demonstrate open-lattice flow in electrically heated core. Bullet Develop and demonstrate chemistry control. Bullet Demonstrate safe operation. Bullet Provision for future testing. This paper discusses the preliminary design of systems, thermal hydraulic analysis, and simplified cost estimated. The facility thermal hydraulic design is based on the maximum simulated core power using seven electrical heater rods of 420 kW; average linear heat generation rate of 300 W/cm. The core inlet temperature for liquid lead or Pb/Bi eutectic is 4200 Degree-Sign C. The design includes approximately seventy-five data measurements such as pressure, temperature, and flow rates. The preliminary estimated cost of construction of the facility is $3.7M (in 2006 $). It is also estimated that the facility will require two years to be constructed and ready for operation.

  4. Loss of Coolant Accident (LOCA) / Emergency Core Coolant System (ECCS Evaluation of Risk-Informed Margins Management Strategies for a Representative Pressurized Water Reactor (PWR)

    Energy Technology Data Exchange (ETDEWEB)

    Szilard, Ronaldo Henriques [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    A Risk Informed Safety Margin Characterization (RISMC) toolkit and methodology are proposed for investigating nuclear power plant core, fuels design and safety analysis, including postulated Loss-of-Coolant Accident (LOCA) analysis. This toolkit, under an integrated evaluation model framework, is name LOCA toolkit for the US (LOTUS). This demonstration includes coupled analysis of core design, fuel design, thermal hydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results.

  5. MATLAB/Simulink Framework for Modeling Complex Coolant Flow Configurations of Advanced Automotive Thermal Management Systems

    Energy Technology Data Exchange (ETDEWEB)

    Titov, Gene; Lustbader, Jason; Leighton, Daniel; Kiss, Tibor

    2016-04-05

    The National Renewable Energy Laboratory's (NREL's) CoolSim MATLAB/Simulink modeling framework was extended by including a newly developed coolant loop solution method aimed at reducing the simulation effort for arbitrarily complex thermal management systems. The new approach does not require the user to identify specific coolant loops and their flow. The user only needs to connect the fluid network elements in a manner consistent with the desired schematic. Using the new solution method, a model of NREL's advanced combined coolant loop system for electric vehicles was created that reflected the test system architecture. This system was built using components provided by the MAHLE Group and included both air conditioning and heat pump modes. Validation with test bench data and verification with the previous solution method were performed for 10 operating points spanning a range of ambient temperatures between -2 degrees C and 43 degrees C. The largest root mean square difference between pressure, temperature, energy and mass flow rate data and simulation results was less than 7%.

  6. Cold atoms as a coolant for levitated optomechanical systems

    CERN Document Server

    Ranjit, Gambhir; Geraci, Andrew A

    2014-01-01

    Optically trapped dielectric objects are well suited for reaching the quantum regime of their center of mass motion in an ultra-high vacuum environment. We show that ground state cooling of an optically trapped nanosphere is achievable when starting at room temperature, by sympathetic cooling of a cold atomic gas optically coupled to the nanoparticle. Unlike cavity cooling in the resolved sideband limit, this system requires only a modest cavity finesse and it allows the cooling to be turned off, permitting subsequent observation of strongly-coupled dynamics between the atoms and sphere. Nanospheres cooled to their quantum ground state could have applications in quantum information science or in precision sensing.

  7. System approach in the investigation of coolant parametrical oscillations in passive safety injection systems (PSIS)

    Energy Technology Data Exchange (ETDEWEB)

    Proskouriakov, K.N. [Moskovskij Ehnergeticheskij Inst., Moscow (Russian Federation)

    2001-07-01

    The use of thermal-hydraulic computer codes is an important part of the work programme for activities in the field of nuclear power plants (NPP) Safety Research as it will enable to define better the test configuration and parameter range extensions and to extrapolate the results of the small scale experiments towards full scale reactor applications. The CATHARE2, RELAP5, the WCOBRA/TRAC, and APROS codes are the estimate thermal hydraulic codes for the evaluation of large and small break loss of coolant accidents (LOCA). The relatively good agreement experimental data with the calculations have been presented. There was shown also some big mistakes in predicting distribution of flow when two phase are present. Model of parametrical oscillation (P.O.) worked out gives explanation for flow oscillations and indicates that the phenomenon of P.O. appears under certain combination of thermal-hydraulic parameters and structure of heat-removal system. (orig.)

  8. Design of coolant distribution system (CDS) for ITER PF AC/DC converter

    Energy Technology Data Exchange (ETDEWEB)

    Guo, Bin [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Song, Zhiquan, E-mail: zhquansong@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Fu, Peng; Xu, Xuesong; Li, Chuan [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Wang, Min; Dong, Lin [China International Nuclear Fusion Energy Program Execution Center, Beijing 100862 (China)

    2016-10-15

    Highlights: • System process and arrangement has been proposed to meet the multiple requirements from the converter system. • Thermal hydraulic analysis model has been developed to size and predict the system operation behavior. • Prototype test has been performed to validate the proposed design methodology. - Abstract: The Poloidal Field (PF) converter unit, playing an essential role in the plasma shape and position control in vertical and horizontal direction, which is an important part of ITER power supply system. As an important subsystem of the converter unit, the coolant distribution system has the function to distribute the cooling water from ITER component cooling water system (CCWS) to its main components at the required flow rate, pressure and temperature. This paper presents the thermal hydraulic design of coolant distribution system for the ITER PF converter unit. Different operational requirements of the PF converter unit regarding flow rate, temperature and pressure have been analyzed to design the system process and arrangement. A thermal-hydraulic analysis model has been built to size the system and predict the flow rate and temperature distribution of the system under the normal operation. Based on the system thermal-hydraulic analysis results, the system pressure profile has been plotted to evaluate the pressure behavior along each client flow path. A CDS prototype for the ITER PF converter has been constructed and some experiments have been performed on it. A good agreement of the flow distribution and temperature behavior between the simulated and test results validate the proposed design methodology.

  9. Lamp system with conditioned water coolant and diffuse reflector of polytetrafluorethylene(PTFE)

    Energy Technology Data Exchange (ETDEWEB)

    Zapata, Luis E. (Livermore, CA); Hackel, Lloyd (Livermore, CA)

    1999-01-01

    A lamp system with a very soft high-intensity output is provided over a large area by water cooling a long-arc lamp inside a diffuse reflector of polytetrafluorethylene (PTFE) and titanium dioxide (TiO.sub.2) white pigment. The water is kept clean and pure by a one micron particulate filter and an activated charcoal/ultraviolet irradiation system that circulates and de-ionizes and biologically sterilizes the coolant water at all times, even when the long-arc lamp is off.

  10. Hot-gas cleanup system model development. Volume I. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Ushimaru, K.; Bennett, A.; Bekowies, P.J.

    1982-11-01

    This two-volume report summarizes the state of the art in performance modeling of advanced high-temperature, high-pressure (HTHP) gas cleanup devices. Volume I contains the culmination of the research effort carried over the past 12 months and is a summary of research achievements. Volume II is the user's manual for the computer programs developed under the present research project. In this volume, Section 2 presents background information on pressurized, fluidized-bed combustion concepts, a description of the role of the advanced gas cleanup systems, and a list of advanced gas cleanup systems that are currently in development under DOE sponsorship. Section 3 describes the methodology for the software architecture that forms the basis of the well-disciplined and structured computer programs developed under the present project. Section 4 reviews the fundamental theories that are important in analyzing the cleanup performance of HTHP gas filters. Section 5 discusses the effect of alkali agents in HTHP gas cleanup. Section 6 evaluates the advanced HTHP gas cleanup models based on their mathematical integrity, availability of supporting data, and the likelihood of commercialization. As a result of the evaluation procedure detailed in Section 6, five performance models were chosen to be incorporated into the overall system simulation code, ASPEN. These five models (the electrocyclone, ceramic bag filter, moving granular bed filter, electrostatic granular bed filter, and electrostatic precipitator) are described in Section 7. The method of cost projection for these five models is discussed in Section 8. The supporting data and validation of the computer codes are presented in Section 9, and finally the conclusions and recommendations for the HTHP gas cleanup system model development are given in Section 10. 72 references, 19 figures, 25 tables.

  11. Final safety analysis addendum to hazard summary report, experimental breeder reactor No. II (EBR-II): the EBR-II cover-gas cleanup system

    Energy Technology Data Exchange (ETDEWEB)

    Fryer, R M; Monson, L R; Price, C C; Hooker, D W

    1979-04-01

    This report evaluates abnormal and accident conditions postulated for the EBR-II cover-gas cleanup system (CGCS). Major considerations include loss of CGCS function with a high level of cover-gas activity, loss of the liquid-nitrogen coolant required for removing fission products from the cover gas, contamination of the cover gas from sources other than the reactor, and loss of system pressure boundary. Calculated exposures resulting from the maximum hypothetical accident (MHA) are less than 2% of the 25-Rem limit stipulated in U.S. Regulation 10 CFR 100; i.e., a person standing at any point on an exclusion boundary (area radius of 600 m) for 2 h following onset of the postulated release would receive less than 0.45 Rem whole-body dose. The on-site whole-body dose (10 m from the source) would be less than 16 Rem.

  12. Systems engineering functions and requirements for the Hanford cleanup mission. First issue, Addendum 2

    Energy Technology Data Exchange (ETDEWEB)

    Holmes, J.J.

    1994-01-01

    This addendum provides the technical detail of a systems engineering functional analysis for the Hanford cleanup mission. Details of the mission analysis including mission statement, scope, problem statement, initial state definition, and final state definition are provided in the parent document. The functional analysis consists of Input Computer Automated Manufacturing Definition (IDEFO) diagrams an definitions, which will be understood by systems engineers, but which may be difficult for others to comprehend. For a more complete explanation of this work, refer to the parent document. The analysis covers the total Hanford cleanup mission including the decomposition levels at which the various Hanford programs or integrated activities are encountered.

  13. Fixed-bed gasifier and cleanup system engineering summary report through Test Run No. 100

    Energy Technology Data Exchange (ETDEWEB)

    Pater, K. Jr.; Headley, L.; Kovach, J.; Stopek, D.

    1984-06-01

    The state-of-the-art of high-pressure, fixed-bed gasification has been advanced by the many refinements developed over the last 5 years. A novel full-flow gas cleanup system has been installed and tested to clean coal-derived gases. This report summarizes the results of tests conducted on the gasifier and cleanup system from its inception through 1982. Selected process summary data are presented along with results from complementary programs in the areas of environmental research, process simulation, analytical methods development, and component testing. 20 references, 32 figures, 42 tables.

  14. Numerical and experimental investigation of surface vortex formation in coolant reservoirs of reactor safety systems

    Energy Technology Data Exchange (ETDEWEB)

    Pandazis, Peter [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany); Babcsany, Boglarka [Budapest Univ. of Technology and Economics (Hungary). Inst. of Nuclear Techniques

    2016-11-15

    The reliable operation of the emergency coolant pumps and passive gravitational injection systems are an important safety issue during accident scenarios with coolant loss in pressurized water reactors. Because of the pressure drop and flow disturbances surface vortices develops at the pump intakes if the water level decreasing below a critical value. The induced swirling flow and gas entrainment lead to flow limitation and to pump failures and damages. The prediction of the critical submergence to avoid surface vortex building is difficult because it depends on many geometrical and fluid dynamical parameters. An alternative and new method has been developed for the investigation of surface vortices. The method based on the combination of CFD results with the analytical vortex model of Burgers and Rott. For further investigation the small scale experiments from the Institute of Nuclear Techniques of the Budapest University of Technology and Economics are used which were inspired from flow limitation problems during the draining of the bubble condenser trays at a VVER type nuclear power plants.

  15. Enhanced Control of PWR Primary Coolant Water Chemistry Using Selective Separation Systems for Recovery and Recycle of Enriched Boric Acid

    Energy Technology Data Exchange (ETDEWEB)

    Ken Czerwinski; Charels Yeamans; Don Olander; Kenneth Raymond; Norman Schroeder; Thomas Robison; Bryan Carlson; Barbara Smit; Pat Robinson

    2006-02-28

    The objective of this project is to develop systems that will allow for increased nuclear energy production through the use of enriched fuels. The developed systems will allow for the efficient and selective recover of selected isotopes that are additives to power water reactors' primary coolant chemistry for suppression of corrosion attack on reactor materials.

  16. Enhanced Control of PWR Primary Coolant Water Chemistry Using Selective Separation Systems for Recovery and Recycle of Enriched Boric Acid

    Energy Technology Data Exchange (ETDEWEB)

    Ken Czerwinski; Charels Yeamans; Don Olander; Kenneth Raymond; Norman Schroeder; Thomas Robison; Bryan Carlson; Barbara Smit; Pat Robinson

    2006-02-28

    The objective of this project is to develop systems that will allow for increased nuclear energy production through the use of enriched fuels. The developed systems will allow for the efficient and selective recover of selected isotopes that are additives to power water reactors' primary coolant chemistry for suppression of corrosion attack on reactor materials.

  17. VICTORIA: A mechanistic model of radionuclide behavior in the reactor coolant system under severe accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Heames, T.J. (Science Applications International Corp., Albuquerque, NM (USA)); Williams, D.A.; Johns, N.A.; Chown, N.M. (UKAEA Atomic Energy Establishment, Winfrith (UK)); Bixler, N.E.; Grimley, A.J. (Sandia National Labs., Albuquerque, NM (USA)); Wheatley, C.J. (UKAEA Safety and Reliability Directorate, Culcheth (UK))

    1990-10-01

    This document provides a description of a model of the radionuclide behavior in the reactor coolant system (RCS) of a light water reactor during a severe accident. This document serves as the user's manual for the computer code called VICTORIA, based upon the model. The VICTORIA code predicts fission product release from the fuel, chemical reactions between fission products and structural materials, vapor and aerosol behavior, and fission product decay heating. This document provides a detailed description of each part of the implementation of the model into VICTORIA, the numerical algorithms used, and the correlations and thermochemical data necessary for determining a solution. A description of the code structure, input and output, and a sample problem are provided. The VICTORIA code was developed upon a CRAY-XMP at Sandia National Laboratories in the USA and a CRAY-2 and various SUN workstations at the Winfrith Technology Centre in England. 60 refs.

  18. Source Term Analysis for Reactor Coolant System with Consideration of Fuel Burnup

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yu Jong; Ahn, Joon Gi; Hwang, Hae Ryong [KEPCO EnC, Daejeon (Korea, Republic of)

    2015-10-15

    The radiation source terms in reactor coolant system (RCS) of pressurized water reactor (PWR) are basic design information for ALARA design such as radiation protection and shielding. Usually engineering companies own self-developed computer codes to estimate the source terms in RCS. DAMSAM and FIPCO are the codes developed by engineering companies. KEPCO E and C has developed computer code, RadSTAR, for use in the Radiation Source Term Analysis for Reactor coolant system during normal operation. The characteristics of RadSTAR are as follows. (1) RadSTAR uses fuel inventory data calculated by ORIGEN, such as ORIGEN2 or ORIGEN-S to consider effects of the fuel burnup. (2) RadSTAR estimates fission products by using finite differential method and analytic method to minimize numerical error. (3) RadSTAR enhances flexibility by adding the function to build the nuclide data library (production pathway library) for user-defined nuclides from ORIGEN data library. (4) RadSTAR consists of two modules. RadSTAR-BL is to build the nuclide data library. RadSTAR-ST is to perform numerical analysis on source terms. This paper includes descriptions on the numerical model, the buildup of nuclide data library, and the sensitivity analysis and verification of RadSTAR. KEPCO E and C developed RadSTAR to calculate source terms in RCS during normal operation. Sensitivity analysis and accuracy verification showed that RadSTAR keeps stability at Δt of 0.1 day and gives more accurate results in comparison with DAMSAM. After development, RadSTAR will replace DAMSAM. The areas, necessary to further development of RadSTAR, are addition of source term calculations for activation products and for shutdown operation.

  19. Interfacing systems LOCAs (Loss of Coolant Accidents) at boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chu, Tsong-Lun; Fitzpatrick, R.; Stoyanov, S.

    1987-01-01

    The work presented in this paper was performed by Brookhaven National Laboratory (BNL) in support of Nuclear Regulatory Commission's (NRC) effort towards the resolution of Generic Issue 105 ''Interfacing System Loss of Coolant Accidents (LOCAs) at Boiling Water Reactors (BWRs).'' For BWRs, intersystem LOCA have typically either not been considered in probabilistic risk analyses, or if considered, were judged to contribute little to the risk estimates because of their perceived low frequency of occurrence. However, recent operating experience indicates that the pressure isolation valves (PIVs) in BWRs may not adequately protect against overpressurization of low pressure systems. The objective of this paper is to present the results of a study which analyzed interfacing system LOCA at several BWRs. The BWRs were selected to best represent a spectrum of BWRs in service using industry operating event experience and plant-specific information/configurations. The results presented here include some possible changes in test requirements/practices as well as an evaluation of their reduction potential in terms of core damage frequency (CDF).

  20. An Analysis of an Automatic Coolant Bypass in the International Space Station Node 2 Internal Active Thermal Control System

    Science.gov (United States)

    Clanton, Stephen E.; Holt, James M.; Turner, Larry D. (Technical Monitor)

    2001-01-01

    A challenging part of International Space Station (ISS) thermal control design is the ability to incorporate design changes into an integrated system without negatively impacting performance. The challenge presents itself in that the typical ISS Internal Active Thermal Control System (IATCS) consists of an integrated hardware/software system that provides active coolant resources to a variety of users. Software algorithms control the IATCS to specific temperatures, flow rates, and pressure differentials in order to meet the user-defined requirements. What may seem to be small design changes imposed on the system may in fact result in system instability or the temporary inability to meet user requirements. The purpose of this paper is to provide a brief description of the solution process and analyses used to implement one such design change that required the incorporation of an automatic coolant bypass in the ISS Node 2 element.

  1. Proceedings of the seventh annual gasification and gas stream cleanup systems contractors review meeting: Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    Ghate, M.R.; Markel, K.E. Jr.; Jarr, L.A.; Bossart, S.J. (eds.)

    1987-08-01

    On June 16 through 19, 1987, METC sponsored the Seventh Annual Gasification and Gas Stream Cleanup Systems Contractors Review Meeting which was held at the Sheraton Lakeview Conference Center in Morgantown, West Virginia. The primary purpose of the meeting was threefold: to review the technical progress and current status of the gasification and gas stream cleanup projects sponsored by the Department of Energy; to foster technology exchange among participating researchers and other technical communities; to facilitate interactive dialogues which would identify research needs that would make coal-based gasification systems more attractive economically and environmentally. More than 310 representatives of Government, academia, industry, and foreign energy research organizations attended the 4-day meeting. Fifty-three papers and thirty poster dsplays were presented summarizing recent developments in the gasification and gas stream cleanup programs. Volume II covers papers presented at sessions 5 and 6 on system for the production of synthesis gas, and on system for the production of power. All papers have been processed for inclusion in the Energy Data Base.

  2. Systems Engineering functions and requirements for the Hanford Cleanup mission: First issue

    Energy Technology Data Exchange (ETDEWEB)

    Holmes, J.J.

    1994-01-01

    This report documents the top-level SE mission analysis, functions analysis, and requirements analysis for the Hanford Site cleanup mission. Because SE is an iterative process, this document will be continuously updated as the mission evolves. This first issue will be subject to change as lower-level work is conducted or primary system architecture is changed as a result of public involvement, NEPA processes, or changes in DOE/HQ direction.

  3. Systems engineering product description report for the Hanford Cleanup Mission: First issue

    Energy Technology Data Exchange (ETDEWEB)

    Holmes, J.J.; Bailey, K.B. [Westinghouse Hanford Co., Richland, WA (United States); Collings, J.L.; Hubbard, A.B.; Niepke, T.M. [Science Applications International Corp. (United States)

    1994-06-01

    This document describes the upper level physical and administrative (nonphysical) products that, when delivered, complete the Hanford Cleanup Mission. Development of product descriptions is a continuation of the Sitewide Systems Engineering work described in the Sitewide functional analysis, the architecture synthesis, and is consistent with guidance contained in the mission plan. This document provides a bridge between all three documents and the products required to complete the mission of cleaning up the Hanford Site.

  4. The development of robotic system for inspecting and repairing NPP primary coolant system of high-level radioactive environment

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seung Ho; Kim, Ki Ho; Jung, Seung Ho; Kim, Byung Soo; Hwang, Suk Yeoung; Kim, Chang Hoi; Seo, Yong Chil; Lee, Young Kwang; Lee, Yong Bum; Cho, Jai Wan; Lee, Jae Kyung; Lee, Yong Deok

    1997-07-01

    This project aims at developing a robotic system to automatically handle inspection and maintenance of NPP safety-related facilities in high-level radioactive environment. This robotic system under development comprises two robots depending on application fields - a mobile robot and multi-functional robot. The mobile robot is designed to be used in the area of primary coolant system during the operation of NPP. This robot enables to overcome obstacles and perform specified tasks in unstructured environment. The multi-functional robot is designed for performing inspection and maintenance tasks of steam generator and nuclear reactor vessel during the overhaul periods of NPP. Nuclear facilities can be inspected and repaired all the time by use of both the mobile robot and the multi-functional robot. Human operator, by teleoperation, monitors the movements of such robots located at remote task environment via video cameras and controls those remotely generating desired commands via master manipulator. We summarize the technology relating to the application of the mobile robot to primary coolant system environment, the applicability of the mobile robot through 3D graphic simulation, the design of the mobile robot, the design of its radiation-hardened controller. We also describe the mechanical design, modeling, and control system of the multi-functional robot. Finally, we present the design of the force-reflecting master and the modeling of virtual task environment for a training simulator. (author). 47 refs., 16 tabs., 43 figs.

  5. Proceedings of the seventh annual gasification and gas stream cleanup systems contractors review meeting: Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Ghate, M.R.; Markel, K.E. Jr.; Jarr, L.A.; Bossart, S.J. (eds.)

    1987-08-01

    On June 16 through 19, 1987, METC sponsored the Seventh Annual Gasification and Gas Stream Cleanup Systems Contractors Review Meeting which was held at the Sheraton Lakeview Conference Center in Morgantown, West Virginia. The primary purpose of the meeting was threefold: to review the technical progress and current status of the gasification and gas stream cleanup projects sponsored by the Department of Energy; to foster technology exchange among participating researchers and other technical communities; to facilitate interactive dialogues which would identify research needs that would make coal-based gasification systems more attractive economically and environmentally. More than 310 representatives of Government, academia, industry, and foreign energy research organizations attended the 4-day meeting. Fifty-three papers and thirty poster displays were presented summarizing recent developments in the gasification and gas stream cleanup programs. Volume I covers information presented at sessions 1 through 4 on systems for the production of Co-products and industrial fuel gas, environmental projects, and components and materials. Individual papers have been processed for the Energy Data Base.

  6. Effects of coatings on storability of carrot under evaporative coolant system.

    Directory of Open Access Journals (Sweden)

    Adetunji Charles Oluwaseun

    2013-09-01

    Full Text Available Four different coatings were developed from the mucilage of Cactus and their effects were investigated on the quality and storability of carrot fruits. The four experimental coatings were: Pure mucilage extracts (ME, Mucilage extract mixed with 5ml glycerol (MEG, Mucilage extract mixed 5ml soy oil ( MESO, Mucilage extract mixed with 5ml olive oil(MEOO the addition of oil served as plasticizer. The following parameters were measured: weight loss, ascorbic acid content, pH, firmness and microbial qualities. Four hundred and eighty (480 carrot were arranged randomly into five treatments, the control (untreated and four coating treatments were stored for seven weeks under Evaporative Coolant System (ECS. Prior to storage, the carrot samples were surface sterilized using 100mg/L sodium hypochlorite. Results showed that Cactus mucilage was effective in extending the shelf-life of carrotwhen compared to untreated control in the following order: MESO>MEOO>MEG>ME>Control. Results revealed that coatings hindered the growth of microorganisms significantly

  7. Remedial action assessment system: Decision support for environmental cleanup

    Energy Technology Data Exchange (ETDEWEB)

    Pennock, K.A.; Bohn, S.; Franklin, A.L.

    1991-11-01

    A large number of hazardous waste sites across the United States await treatment. Waste sites can be physically complex entities composed of multiple, possibly interacting contaminants distributed throughout one or more media. The sites may be active as well with contaminants escaping through one or more potential escape paths. Treatment of these sites requires a long and costly commitment involving the coordination of activities among several waste treatment professionals. In order to reduce the cost and time required for the specification of treatment at these waste sites. The Remedial Action Assessment System (RAAS) was proposed. RAAS is an automated information management system which utilizes a combination of expert reasoning and numerical models to produce the combinations of treatment technologies, known as treatment trains, which satisfy the treatment objectives of a particular site. In addition, RAAS supports the analysis of these trains with regard to effectiveness and cost so that the viable treatment trains can be measured against each other. The Remedial Action Assessment System is a hybrid system designed and constructed using object-oriented tools and techniques. RAAS is advertised as a hybrid system because it combines, in integral fashion, numerical computing (primarily quantitative models) with expert system reasoning. An object-oriented approach was selected due to many of its inherent advantages, among these the naturalness of modeling physical objects and processes.

  8. Integrated low emission cleanup system for direct coal-fueled turbines (electrostatic agglomeration)

    Energy Technology Data Exchange (ETDEWEB)

    Quimby, J.M.; Kumar, K.S.

    1992-01-01

    The objective of this contract was to investigate the removal of SO[sub x] and particulate matter from direct coal fired combustion gas streams at high temperature and high pressure conditions. This investigation was to be accomplished through a bench scale testing and evaluation program for SO[sub x] removal and the innovative particulate collection concept of particulate growth through electrostatic agglomeration followed by high efficiency mechanical collection. The process goal was to achieve control better than that required by 1979 New Source Performance Standards. During Phase I, the designs of the combustor and gas cleanup apparatus were successfully completed. Hot gas cleanup was designed to be accomplished at temperature levels between 1800[degrees] and 2500[degrees]F at pressures up to 15 atmospheres. The combustor gas flow rate could be varied between 0.2--0.5 pounds per second. The electrostatic agglomerator residence time could be varied between 0.25 to 3 seconds. In Phase II, all components were fabricated, and erected successfully. Test data from shakedown testing was obtained. Unpredictable difficulties in pilot plant erection and shakedown consumed more budget resources than was estimated and as a consequence DOE, METC, decided ft was best to complete the contract at the end of Phase II. Parameters studied in shakedown testing revealed that high-temperature high pressure electrostatics offers an alternative to barrier filtration in hot gas cleanup but more research is needed in successful system integration between the combustor and electrostatic agglomerator.

  9. Toyota's heat management system - coolant heat storage for mass production today, new technologies for the future

    Energy Technology Data Exchange (ETDEWEB)

    Ichinose, Hiroki; Takaoka, Toshifumi; Kobayashi, Hideo [Toyota Motor Corporation (Japan)

    2004-07-01

    There has been pressing needs for the protection of metropolitan environment and the challenge of global warming. A heat management system prevails to meet such requirements. In actual driving condition, only about 30% of the total fuel energy is consumed for propulsion and air conditioner. At the same time 60% of fuel energy is wasted as exhaust gas, thermal loss and warm up loss. It is important to manage total thermal energy as a whole vehicle to improve thermal efficiency. The principle is to reduce heat loss in order to increase exhaust gas temperature and recover heat energy for pre-heating at the next cold start. Further developed versions may include thermal management strategies including turbocharger and thermal exchanger using exhaust gas heat energy. Toyota has developed the Coolant Heat Storage system (CHSS) for one of heat management systems to reduce cold emission and improve cabin comfort. The system enables to store hot coolant at the warmed up condition in a heat storage tank. At the next cold start, it is possible for CHSS to reduce unburned hydrocarbon by preheating intake port quickly with the hot coolant in the tank. CHSS was adopted in hybrid vehicle Prius for the US model in 2003. This vehicle achieved to meet the ATPZEV,the most stringent emission regulation in the US. This paper describes a total heat management focusing on the newly developed CHSS.

  10. MATLAB/Simulink Framework for Modeling Complex Coolant Flow Configurations of Advanced Automotive Thermal Management Systems: Preprint

    Energy Technology Data Exchange (ETDEWEB)

    Titov, Eugene; Lustbader, Jason; Leighton, Daniel; Kiss, Tibor

    2016-03-22

    The National Renewable Energy Laboratory's (NREL's) CoolSim MATLAB/Simulink modeling framework was extended by including a newly developed coolant loop solution method aimed at reducing the simulation effort for arbitrarily complex thermal management systems. The new approach does not require the user to identify specific coolant loops and their flow. The user only needs to connect the fluid network elements in a manner consistent with the desired schematic. Using the new solution method, a model of NREL's advanced combined coolant loop system for electric vehicles was created that reflected the test system architecture. This system was built using components provided by the MAHLE Group and included both air conditioning and heat pump modes. Validation with test bench data and verification with the previous solution method were performed for 10 operating points spanning a range of ambient temperatures between -2 degrees C and 43 degrees C. The largest root mean square difference between pressure, temperature, energy and mass flow rate data and simulation results was less than 7%.

  11. Equipment Design and Cost Estimation for Small Modular Biomass Systems, Synthesis Gas Cleanup, and Oxygen Separation Equipment; Task 2: Gas Cleanup Design and Cost Estimates -- Wood Feedstock

    Energy Technology Data Exchange (ETDEWEB)

    Nexant Inc.

    2006-05-01

    As part of Task 2, Gas Cleanup and Cost Estimates, Nexant investigated the appropriate process scheme for treatment of wood-derived syngas for use in the synthesis of liquid fuels. Two different 2,000 metric tonne per day gasification schemes, a low-pressure, indirect system using the gasifier, and a high-pressure, direct system using gasification technology were evaluated. Initial syngas conditions from each of the gasifiers was provided to the team by the National Renewable Energy Laboratory. Nexant was the prime contractor and principal investigator during this task; technical assistance was provided by both GTI and Emery Energy.

  12. Investigations of the VVER-1000 coolant transient benchmark phase 1 with the coupled code system RELAP5/PARCS

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez-Espinoza, Victor Hugo

    2008-07-15

    As part of the reactor dynamics activities of FZK/IRS, the qualification of best-estimate coupled code systems for reactor safety evaluations is a key step toward improving their prediction capability and acceptability. The VVER-1000 Coolant Transient Benchmark Phase 1 represents an excellent opportunity to validate the simulation capability of the coupled code system RELAP5/PACRS regarding both the thermal hydraulic plant response (RELAP5) using measured data obtained during commissioning tests at the Kozloduy nuclear power plant unit 6 and the neutron kinetics models of PARCS for hexagonal geometries. The Phase 1 is devoted to the analysis of the switching on of one main coolant pump while the other three pumps are in operation. It includes the following exercises: (a) investigation of the integral plant response using a best-estimate thermal hydraulic system code with a point kinetics model (b) analysis of the core response for given initial and transient thermal hydraulic boundary conditions using a coupled code system with 3D-neutron kinetics model and (c) investigation of the integral plant response using a best-estimate coupled code system with 3D-neutron kinetics. Already before the test, complex flow conditions exist within the RPV e.g. coolant mixing in the upper plenum caused by the reverse flow through the loop-3 with the stopped pump. The test is initiated by switching on the main coolant pump of loop-3 that leads to a reversal of the flow through the respective piping. After about 13 s the mass flow rate through this loop reaches values comparable with the one of the other loops. During this time period, the increased primary coolant flow causes a reduction of the core averaged coolant temperature and thus an increase of the core power. Later on, the power stabilizes at a level higher than the initial power. In this analysis, special attention is paid on the prediction of the spatial asymmetrical core cooling during the test and its effects on the

  13. Investigation of austenitic alloys for advanced heat recovery and hot gas cleanup systems

    Energy Technology Data Exchange (ETDEWEB)

    Swindeman, R.W.; Ren, W.

    1996-06-01

    The objective of the research is to provide databases and design criteria to assist in the selection of optimum alloys for construction of components needed to contain process streams in advanced heat recovery and hot-gas cleanup systems. Typical components include: steam line piping and superheater tubing for low emission boilers (600 to 700{degrees}C), heat exchanger tubing for advanced steam cycles and topping cycle systems (650 to 800{degrees}C), foil materials for recuperators, on advanced turbine systems (700 to 750{degrees}C), and tubesheets for barrier filters, liners for piping, cyclones, and blowback system tubing for hot-gas cleanup systems (850 to 1000{degrees}C). The materials being examined fall into several classes, depending on which of the advanced heat recovery concepts is of concern. These classes include martensitic steels for service to 650{degrees}C, lean stainless steels and modified 25Cr-30Ni steels for service to 700{degrees}C, modified 25Cr-20Ni steels for service to 900{degrees}C, and high Ni-Cr-Fe or Ni-Cr-Co-Fe alloys for service to 1000{degrees}C.

  14. Coupling a system code with computational fluid dynamics for the simulation of complex coolant reactivity effects

    Energy Technology Data Exchange (ETDEWEB)

    Bertolotto, D.

    2011-11-15

    The current doctoral research is focused on the development and validation of a coupled computational tool, to combine the advantages of computational fluid dynamics (CFD) in analyzing complex flow fields and of state-of-the-art system codes employed for nuclear power plant (NPP) simulations. Such a tool can considerably enhance the analysis of NPP transient behavior, e.g. in the case of pressurized water reactor (PWR) accident scenarios such as Main Steam Line Break (MSLB) and boron dilution, in which strong coolant flow asymmetries and multi-dimensional mixing effects strongly influence the reactivity of the reactor core, as described in Chap. 1. To start with, a literature review on code coupling is presented in Chap. 2, together with the corresponding ongoing projects in the international community. Special reference is made to the framework in which this research has been carried out, i.e. the Paul Scherrer Institute's (PSI) project STARS (Steady-state and Transient Analysis Research for the Swiss reactors). In particular, the codes chosen for the coupling, i.e. the CFD code ANSYS CFX V11.0 and the system code US-NRC TRACE V5.0, are part of the STARS codes system. Their main features are also described in Chap. 2. The development of the coupled tool, named CFX/TRACE from the names of the two constitutive codes, has proven to be a complex and broad-based task, and therefore constraints had to be put on the target requirements, while keeping in mind a certain modularity to allow future extensions to be made with minimal efforts. After careful consideration, the coupling was defined to be on-line, parallel and with non-overlapping domains connected by an interface, which was developed through the Parallel Virtual Machines (PVM) software, as described in Chap. 3. Moreover, two numerical coupling schemes were implemented and tested: a sequential explicit scheme and a sequential semi-implicit scheme. Finally, it was decided that the coupling would be single

  15. Cleanup and Dismantling of Highly Contaminated Ventilation Systems Using Robotic Tools - 13162

    Energy Technology Data Exchange (ETDEWEB)

    Chambon, Frederic [AREVA FEDERAL SERVICES, Columbia MD (United States); CIZEL, Jean-Pierre [AREVA BE/NV, Marcoule (France); Blanchard, Samuel [CEA DEN/DPAD, Marcoule (France)

    2013-07-01

    The UP1 plant reprocessed nearly 20,000 tons of used natural uranium gas cooled reactor fuel coming from the first generation of civil nuclear reactors in France. Following operating incidents in the eighties, the ventilation system of the continuous dissolution line facility was shut down and replaced. Two types of remote controlled tool carriers were developed to perform the decontamination and dismantling operations of the highly contaminated ventilation duct network. The first one, a dedicated small robot, was designed from scratch to retrieve a thick powder deposit within a duct. The robot, managed and confined by two dedicated glove boxes, was equipped for intervention inside the ventilation duct and used for carrying various cleanup and inspection tools. The second type, consisting of robotic tools developed on the base of an industrial platform, was used for the clean-up and dismantling of the ventilation duct system. Depending on the type of work to be performed, on the shape constraints of the rooms and any equipment to be dismantled, different kinds of robotic tools were developed and installed on a Brokk 40 carrier. After more than ten years of ventilation duct D and D operations at the UP1 plant, a lot of experience was acquired about remote operations. The three main important lessons learned in terms of remote controlled operation are: characterizing the initial conditions as much as reasonably possible, performing non-radioactive full scale testing and making it as simple and modular as possible. (authors)

  16. Sorption Mechanisms for Mercury Capture in Warm Post-Gasification Gas Clean-Up Systems

    Energy Technology Data Exchange (ETDEWEB)

    Jost Wendt; Sung Jun Lee; Paul Blowers

    2008-09-30

    The research was directed towards a sorbent injection/particle removal process where a sorbent may be injected upstream of the warm gas cleanup system to scavenge Hg and other trace metals, and removed (with the metals) within the warm gas cleanup process. The specific objectives of this project were to understand and quantify, through fundamentally based models, mechanisms of interaction between mercury vapor compounds and novel paper waste derived (kaolinite + calcium based) sorbents (currently marketed under the trade name MinPlus). The portion of the research described first is the experimental portion, in which sorbent effectiveness to scavenge metallic mercury (Hg{sup 0}) at high temperatures (>600 C) is determined as a function of temperature, sorbent loading, gas composition, and other important parameters. Levels of Hg{sup 0} investigated were in an industrially relevant range ({approx} 25 {micro}g/m{sup 3}) although contaminants were contained in synthetic gases and not in actual flue gases. A later section of this report contains the results of the complementary computational results.

  17. System Assessment of Carbon Dioxide Used as Gas Oxidant and Coolant in Vanadium-Extraction Converter

    Science.gov (United States)

    Du, Wei Tong; Wang, Yu; Liang, Xiao Ping

    2017-07-01

    With the aim of reducing carbon dioxide (CO2) emissions and of using waste resources in steel plants, the use of CO2 as a gas oxidant and coolant in the converter to increase productivity and energy efficiency was investigated in this study. Experiments were performed in combination with thermodynamic theory on vanadium-extraction with CO2 and oxygen (O2) mixed injections. The results indicate that the temperature of the hot metal bath decreased as the amount of CO2 introduced into O2 increased. At an injection of 85 vol.% O2 and 15 vol.% CO2, approximately 12% of additional carbon was retained in the hot metal. Moreover, the content of vanadium trioxide in the slag was higher. In addition, the O2 consumption per ton of hot metal was reduced by 8.5% and additional chemical energy was recovered by the controlled injection of CO2 into the converter. Therefore, using CO2 as a gas coolant was conducive to vanadium extraction, and O2 consumption was reduced.

  18. System Assessment of Carbon Dioxide Used as Gas Oxidant and Coolant in Vanadium-Extraction Converter

    Science.gov (United States)

    Du, Wei Tong; Wang, Yu; Liang, Xiao Ping

    2017-10-01

    With the aim of reducing carbon dioxide (CO2) emissions and of using waste resources in steel plants, the use of CO2 as a gas oxidant and coolant in the converter to increase productivity and energy efficiency was investigated in this study. Experiments were performed in combination with thermodynamic theory on vanadium-extraction with CO2 and oxygen (O2) mixed injections. The results indicate that the temperature of the hot metal bath decreased as the amount of CO2 introduced into O2 increased. At an injection of 85 vol.% O2 and 15 vol.% CO2, approximately 12% of additional carbon was retained in the hot metal. Moreover, the content of vanadium trioxide in the slag was higher. In addition, the O2 consumption per ton of hot metal was reduced by 8.5% and additional chemical energy was recovered by the controlled injection of CO2 into the converter. Therefore, using CO2 as a gas coolant was conducive to vanadium extraction, and O2 consumption was reduced.

  19. Investigation of austenitic alloys for advanced heat recovery and hot gas cleanup systems

    Energy Technology Data Exchange (ETDEWEB)

    Swindeman, R.W.; Ren, W. [Oak Ridge National Lab., TN (United States)

    1996-08-01

    Materials properties were collected for the design and construction of structural components for use in advanced heat recovery and hot gas cleanup systems. Alloys systems included 9Cr-1Mo-V steel, modified 316 stainless steel, modified type 310 stainless steel, modified 20Cr-25Ni-Nb stainless steel, modified alloy 800, and two sulfidation resistant alloys: HR160 and HR120. Experimental work was undertaken to expand the databases for potentially useful alloys. Types of testing included creep, stress-rupture, creep-crack growth, fatigue, and post-exposure short-time tensile tests. Because of the interest in relatively inexpensive alloys for service at 700{degrees}C and higher, research emphasis was placed on a modified type 310 stainless steel and a modified 20Cr-25Ni-Nb stainless steel. Both steels were found to have useful strength to 925{degrees}C with good weldability and ductility.

  20. Investigation of austenitic alloys for advanced heat recovery and hot-gas cleanup systems

    Energy Technology Data Exchange (ETDEWEB)

    Swindeman, R.W. [Oak Ridge National Lab., TN (United States)

    1997-12-01

    Materials properties were collected for the design and construction of structural components for use in advanced heat recovery and hot gas cleanup systems. Alloys systems included 9Cr-1Mo-V steel, modified 316 stainless steel, modified type 310 stainless steel, modified 20Cr-25Ni-Nb stainless steel, and modified alloy 800. Experimental work was undertaken to expand the databases for potentially useful alloys. Types of testing included creep, stress-rupture, creep-crack growth, fatigue, and post-exposure short-time tensile tests. Because of the interest in relatively inexpensive alloys for service at 700 C and higher, research emphasis was placed on a modified type 310 stainless steel and a modified 20Cr-25Ni-Nb stainless steel. Both steels were found to have useful strength to 925 C with good weldability and ductility.

  1. Investigation of austenitic alloys for advanced heat recovery and hot gas cleanup systems

    Energy Technology Data Exchange (ETDEWEB)

    Swindeman, R.W.; Ren, W. [Oak Ridge National Lab., TN (United States)

    1995-08-01

    Alloys for design and construction of structural components needed to contain process streams and provide internal structures in advanced heat recovery and hot gas cleanup systems were examined. Emphasis was placed on high-strength, corrosion-resistant alloys for service at temperatures above 1000 {degrees}F (540{degrees}C). Data were collected that related to fabrication, joining, corrosion protection, and failure criteria. Alloys systems include modified type 310 and 20Cr-25Ni-Nb steels and sulfidation-resistance alloys HR120 and HR160. Types of testing include creep, stress-rupture, creep crack growth, fatigue, and post-exposure short-time tensile. Because of the interest in relatively inexpensive alloys for high temperature service, a modified type 310 stainless steel was developed with a target strength of twice that for standard type 310 stainless steel.

  2. Extended Life Coolant Testing

    Science.gov (United States)

    2016-06-06

    ELC – Extended Life Coolant SCA – Supplemental Coolant Additive SOW – Scope of Work SwRI – Southwest Research Institute TARDEC – Tank Automotive...ethylene or propylene glycol and 35% extended life coolant #1 (ELC1) with a balance of water. At a higher ELC1 content of 45% or 50%, the mass loss...UNCLASSIFIED TABLE OF CONTENTS EXTENDED LIFE COOLANT TESTING INTERIM REPORT TFLRF No. 478 by Gregory A. T. Hansen Edwin A

  3. Experimental study on thermal-hydraulic behaviors of a pressure balanced coolant injection system for a passive safety light water reactor JPSR

    Energy Technology Data Exchange (ETDEWEB)

    Satoh, Takashi; Watanabe, Hironori; Araya, Fumimasa; Nakajima, Katsutoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Iwamura, Takamichi; Murao, Yoshio

    1998-02-01

    A conceptual design study of a passive safety light water reactor JPSR has been performed at Japan Atomic Energy Research Institute JAERI. A pressure balanced coolant injection experiment has been carried out, with an objective to understand thermal-hydraulic characteristics of a passive coolant injection system which has been considered to be adopted to JPSR. This report summarizes experimental results and data recorded in experiment run performed in FY. 1993 and 1994. Preliminary experiments previously performed are also briefly described. As the results of the experiment, it was found that an initiation of coolant injection was delayed with increase in a subcooling in the pressure balance line. By inserting a separation device which divides the inside of core make-up tank (CMT) into several small compartments, a diffusion of a high temperature region formed just under the water surface was restrained and then a steam condensation was suppressed. A time interval from an uncovery of the pressure balance line to the initiation of the coolant injection was not related by a linear function with a discharge flow rate simulating a loss-of-coolant accident (LOCA) condition. The coolant was injected intermittently by actuation of a trial fabricated passive valve actuated by pressure difference for the present experiment. It was also found that the trial passive valve had difficulties in setting an actuation set point and vibrations noises and some fraction of the coolant was remained in CMT without effective use. A modification was proposed for resolving these problems by introducing an anti-closing mechanism. (author)

  4. Cleanup worker exposures to hazardous chemicals at a former nuclear weapons plant: piloting of an exposure surveillance system.

    Science.gov (United States)

    LaMontagne, A D; Van Dyke, M V; Martyny, J W; Ruttenber, A J

    2001-02-01

    Cleanup of former U.S. Department of Energy (DOE) nuclear weapons production facilities involves potential exposures to various hazardous chemicals. We have collaboratively developed and piloted an exposure database and surveillance system for cleanup worker hazardous chemical exposure data with a cleanup contractor at the Rocky Flats Environmental Technology Site (RFETS). A unique system feature is the incorporation of a 34-category work task-coding scheme. This report presents an overview of the data captured by this system during development and piloting from March 1995 through August 1998. All air samples collected were entered into the system. Of the 859 breathing zone samples collected, 103 unique employees and 39 unique compounds were represented. Breathing zone exposure levels were usually low (86% of breathing zone samples were below analytical limits of detection). The use of respirators and other exposure controls was high (87 and 88%, respectively). Occasional high-level excursions did occur. Detailed quantitative summaries are provided for the six most monitored compounds: asbestos, beryllium, carbon tetrachloride, chromium, lead, and methylene chloride. Task and job title data were successfully collected for most samples, and showed specific cleanup activities by pipe fitters to be the most commonly represented in the database. Importantly, these results demonstrate the feasibility of the implementation of integrated exposure database and surveillance systems by practicing industrial hygienists employed in industry as well as the preventive potential and research uses of such systems. This exposure database and surveillance system--the central features of which are applicable in any industrial work setting--has enabled one of the first systematic quantitative characterizations of DOE cleanup worker exposures to hazardous chemicals.

  5. Survey of tracking systems and rotary joints for coolant piping. Final report, August 15, 1978-August 14, 1978. [Includes patents

    Energy Technology Data Exchange (ETDEWEB)

    Furaus, J P; Gruchalla, M E; Sower, G D

    1980-01-01

    Problems were surveyed and evaluated with respect to solar tracking mechanisms and rotary joints for coolant piping. An analytical development of celestial mechanics, one- and two-axis tracking configurations and the effect of tracking accuracy versus collector efficiency are reported. Daily operational requirements and tracking modes were defined and evaluated. A literature and patent search on solar tracking technology was performed. Tracking system and control system performance specifications were determined. Alternative conceptual tracking approaches were defined and a cost and performance evaluation of a mechanical tracking concept was performed. Fluid coupling service specifications were determined. The cost and performance of several types of actuators and error detectors were evaluated with respect to solar tracking mechanisms.

  6. VICTORIA: A mechanistic model of radionuclide behavior in the reactor coolant system under severe accident conditions. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Heams, T J [Science Applications International Corp., Albuquerque, NM (United States); Williams, D A; Johns, N A; Mason, A [UKAEA, Winfrith, (England); Bixler, N E; Grimley, A J [Sandia National Labs., Albuquerque, NM (United States); Wheatley, C J [UKAEA, Culcheth (England); Dickson, L W [Atomic Energy of Canada Ltd., Chalk River, ON (Canada); Osborn-Lee, I [Oak Ridge National Lab., TN (United States); Domagala, P; Zawadzki, S; Rest, J [Argonne National Lab., IL (United States); Alexander, C A [Battelle, Columbus, OH (United States); Lee, R Y [Nuclear Regulatory Commission, Washington, DC (United States)

    1992-12-01

    The VICTORIA model of radionuclide behavior in the reactor coolant system (RCS) of a light water reactor during a severe accident is described. It has been developed by the USNRC to define the radionuclide phenomena and processes that must be considered in systems-level models used for integrated analyses of severe accident source terms. The VICTORIA code, based upon this model, predicts fission product release from the fuel, chemical reactions involving fission products, vapor and aerosol behavior, and fission product decay heating. Also included is a detailed description of how the model is implemented in VICTORIA, the numerical algorithms used, and the correlations and thermochemical data necessary for determining a solution. A description of the code structure, input and output, and a sample problem are provided.

  7. Assessment of coal gasification/hot gas cleanup based advanced gas turbine systems

    Energy Technology Data Exchange (ETDEWEB)

    1990-12-01

    The major objectives of the joint SCS/DOE study of air-blown gasification power plants with hot gas cleanup are to: (1) Evaluate various power plant configurations to determine if an air-blown gasification-based power plant with hot gas cleanup can compete against pulverized coal with flue gas desulfurization for baseload expansion at Georgia Power Company's Plant Wansley; (2) determine if air-blown gasification with hot gas cleanup is more cost effective than oxygen-blown IGCC with cold gas cleanup; (3) perform Second-Law/Thermoeconomic Analysis of air-blown IGCC with hot gas cleanup and oxygen-blown IGCC with cold gas cleanup; (4) compare cost, performance, and reliability of IGCC based on industrial gas turbines and ISTIG power island configurations based on aeroderivative gas turbines; (5) compare cost, performance, and reliability of large (400 MW) and small (100 to 200 MW) gasification power plants; and (6) compare cost, performance, and reliability of air-blown gasification power plants using fluidized-bed gasifiers to air-blown IGCC using transport gasification and pressurized combustion.

  8. Assessment and Accommodation of Thermal Expansion of the Internal Active Thermal Control System Coolant During Launch to On-Orbit Activation of International Space Station Elements

    Science.gov (United States)

    Edwards, Darryl; Ungar, Eugene K.; Holt, James M.

    2002-01-01

    The International Space Station (ISS) employs an Internal Active Thermal Control System (IATCS) comprised of several single-phase water coolant loops. These coolant loops are distributed throughout the ISS pressurized elements. The primary element coolant loops (i.e. U.S. Laboratory module) contain a fluid accumulator to accomodate thermal expansion of the system. Other element coolant loops are parasitic (i.e. Airlock), have no accumulator, and require an alternative approach to insure that the system maximum design pressure (MDP) is not exceeded during the Launch to Activation (LTA) phase. During this time the element loops is a stand alone closed system. The solution approach for accomodating thermal expansion was affected by interactions of system components and their particular limitations. The mathematical solution approach was challenged by the presence of certain unknown or not readily obtainable physical and thermodynamic characteristics of some system components and processes. The purpose of this paper is to provide a brief description of a few of the solutions that evolved over time, a novel mathematical solution to eliminate some of the unknowns or derive the unknowns experimentally, and the testing and methods undertaken.

  9. Machine coolant waste reduction by optimizing coolant life. Project summary

    Energy Technology Data Exchange (ETDEWEB)

    Pallansch, J.

    1995-08-01

    The project was designed to study the following: A specific water-soluble coolant (Blasocut 2000 Universal) in use with a variety of machines, tools, and materials; Coolant maintenance practices associated with three types of machines; Health effects of use and handling of recycled coolant; Handling practices for chips and waste coolant; Chip/coolant separation; and Oil/water separation.

  10. Technology of high temperature organic coolant

    Energy Technology Data Exchange (ETDEWEB)

    Makin, R.S.; Vorobei, M.P.; Kuprienko, V.A.; Starkov, V.A.; Tsykanov, V.A.; Checketkin, Y.V. [Research Institute of Atomic Reactors, Ulyanovsk (Russian Federation)

    1993-12-31

    Research has been performed on the problems related to the use of high temperature organic coolants in small and medium nuclear power plants. The work performed and also the experience of operating the ARBUS reactor confirmed the inherent safety features, reliability, and enhanced safety margins of the plants with this type of coolants. The advantages of this system and research highlights are presented.

  11. Design and fabrication of magnetic coolant filter

    Science.gov (United States)

    Prashanth, B. N.

    2017-07-01

    Now a day's use of coolants in industry has become dominant because of high production demands. Coolants not only help in speeding up the production but also provide many advantages in the metal working operation. As the consumption of coolants is very high a system is badly in need, so as to recirculate the used coolant. Also the amount of hazardous waste generated by industrial plants has become an increasingly costly problem for the manufactures and an additional stress on the environment. Since the purchase and disposal of the spent cutting fluids is becoming increasingly expensive, fluid recycling is a viable option for minimizing the cost. Separation of metallic chips from the coolants by using magnetic coolant separation has proven a good management and maintenance of the cutting fluid. By removing the metallic chips, the coolant life is greatly extended, increases the machining quality and reduces downtime. Above being the case, a magnetic coolant filter is developed which utilizes high energy permanent magnets to develop a dense magnetic field along a narrow flow path into which the contaminated coolant is directed. The ferromagnetic particles captured and aligned by the dense magnetic field, from the efficient filter medium. This enables the unit to remove ferromagnetic particles from the coolant. Magnetic coolant filters use the principle of magnetic separation to purify the used coolant. The developed magnetic coolant separation has the capability of purifying 40 litres per minute of coolant with the size of the contaminants ranging from 1 µm to 30 µm. The filter will be helpful in saving the production cost as the cost associated with the proposed design is well justified by the cost savings in production. The magnetic field produced by permanent magnets will be throughout the area underneath the reservoir. This produces magnetic field 30mm above the coolant reservoir. Very fine particles are arrested without slip. The magnetic material used will not

  12. Carbon formation and metal dusting in hot-gas cleanup systems of coal gasifiers

    Energy Technology Data Exchange (ETDEWEB)

    Tortorelli, P.F.; DeVan, H.J.; Judkins, R.R. [and others

    1995-06-01

    The product gas resulting from the partial oxidation of carboniferous materials in a gasifier consists predominantly of CO, CO{sub 2}, H{sub 2}, H{sub 2}O, CH{sub 4}, and, for air-blown units, N{sub 2} in various proportions at temperatures ranging from about 400 to 1000{degree}C. Depending on the source of the fuel, smaller concentrations of H{sub 2}S, COS, and NH{sub 3} can also be present. The gas phase is typically characterized by high carbon and sulfur, but low oxygen, activities and, consequently, severe degradation of the structural and functional materials used in the gasifier can occur. Therefore, there are numerous concerns about materials performance in coal gasification systems, particularly at the present time when demonstration-scale projects are in or nearing the construction and operation phases. This study focused on the subset of materials degradation phenomena resulting from carbon formation and carburization processes, which are related to potential operating problems in certain gasification components and subsystems. More specifically, it examined the current state of knowledge regarding carbon deposition and a carbon-related degradation phemonenon known as metal dusting as they affect the long-term operation of the gas clean-up equipment downstream of the gasifier and addressed possible means to mitigate the degradation processes. These effects would be primarily associated with the filtering and cooling of coal-derived fuel gases from the gasifier exit temperature to as low as 400{degree}C. However, some of the consideratins are sufficiently general to cover conditions relevant to other parts of gasification systems.

  13. Experimental investigations of flow distribution in coolant system of Helium-Cooled-Pebble-Bed Test Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Ilić, M.; Schlindwein, G., E-mail: georg.schlindwein@kit.edu; Meyder, R.; Kuhn, T.; Albrecht, O.; Zinn, K.

    2016-02-15

    Highlights: • Experimental investigations of flow distribution in HCPB TBM are presented. • Flow rates in channels close to the first wall are lower than nominal ones. • Flow distribution in central chambers of manifold 2 is close to the nominal one. • Flow distribution in the whole manifold 3 agrees well with the nominal one. - Abstract: This paper deals with investigations of flow distribution in the coolant system of the Helium-Cooled-Pebble-Bed Test Blanket Module (HCPB TBM) for ITER. The investigations have been performed by manufacturing and testing of an experimental facility named GRICAMAN. The facility involves the upper poloidal half of HCPB TBM bounded at outlets of the first wall channels, at outlet of by-pass pipe and at outlets of cooling channels in breeding units. In this way, the focus is placed on the flow distribution in two mid manifolds of the 4-manifold system: (i) manifold 2 to which outlets of the first wall channels and inlet of by-pass pipe are attached and (ii) manifold 3 which supplies channels in breeding units with helium coolant. These two manifolds are connected with cooling channels in vertical/horizontal grids and caps. The experimental facility has been built keeping the internal structure of manifold 2 and manifold 3 exactly as designed in HCPB TBM. The cooling channels in stiffening grids, caps and breeding units are substituted by so-called equivalent channels which provide the same hydraulic resistance and inlet/outlet conditions, but have significantly simpler geometry than the real channels. Using the conditions of flow similarity, the air pressurized at 0.3 MPa and at ambient temperature has been used as working fluid instead of HCPB TBM helium coolant at 8 MPa and an average temperature of 370 °C. The flow distribution has been determined by flow rate measurements at each of 28 equivalent channels, while the pressure distribution has been obtained measuring differential pressure at more than 250 positions. The

  14. Equipment Design and Cost Estimation for Small Modular Biomass Systems, Synthesis Gas Cleanup, and Oxygen Separation Equipment; Task 2.3: Sulfur Primer

    Energy Technology Data Exchange (ETDEWEB)

    Nexant Inc.

    2006-05-01

    This deliverable is Subtask 2.3 of Task 2, Gas Cleanup Design and Cost Estimates, of NREL Award ACO-5-44027, ''Equipment Design and Cost Estimation for Small Modular Biomass Systems, Synthesis Gas Cleanup and Oxygen Separation Equipment''. Subtask 2.3 builds upon the sulfur removal information first presented in Subtask 2.1, Gas Cleanup Technologies for Biomass Gasification by adding additional information on the commercial applications, manufacturers, environmental footprint, and technical specifications for sulfur removal technologies. The data was obtained from Nexant's experience, input from GTI and other vendors, past and current facility data, and existing literature.

  15. TRACG-CFD analysis of ESBWR reactor water cleanup shutdown cooling system mixing coefficient

    Energy Technology Data Exchange (ETDEWEB)

    Gallardo, J. [UNAM, Facultad de Ingenieria, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Marquino, W.; Mistreanu, A.; Yang, J., E-mail: euqrop@hotmail.com [General Electric Hitachi Nuclear Energy, Wilmington, 28401 North Carolina (United States)

    2015-09-15

    The ESBWR is a 1520 nominal [M We] Generation III+ natural circulation boiling water reactor designed to high levels of safety utilizing features that have been successfully used before in operating BWRs, as well as standard features common to A BWR. In September of 2014, the US NRC has certified the ESBWR design for use in the USA. The RWCU/Sdc is an auxiliary system for the ESBWR nuclear island. Basic functions it performs include purifying the reactor coolant during normal operation and shutdown and providing shutdown cooling and cooldown to cold shutdown conditions. The performance of the RWCU system during shutdown cooling is directly related to the temperature of the water removed through the outlets, which is coupled with the vessel and F W temperatures through a thermal mixing coefficient. The complex three-dimensional (3-D) geometry of the BWR downcomer and lower plenum has a great impact on the flow mixing. Only a fine mesh technique like CFD can predict the 3-D temperature distribution in the RPV during shutdown and provide the RWCU/Sdc system inlet temperature. Plant shutdown is an unsteady event by nature and was modeled as a succession of CFD steady-state simulations. It is required to establish the mixing coefficient (which is a function of the heat balance and the core flow) during the operation of the RWCU system in the multiple shutdown cooling modes, and therefore a range of core flows needs to be estimated using quasi steady states obtained with TRACG. The lower end of that range is obtained from a system with minimal power decay heat and core flow; while the higher end corresponds to the power at the beginning of RWCU/Sdc operation when the cooldown is transferred to the RWCU/Sdc after the initial depressurization via the turbine bypass valves. Because the ESBWR RWCU/Sdc return and suction designs provide good mixing, the uniform mixing energy balance was found to be an adequate alternative for deriving the mixing coefficient. The CFD mass flow

  16. Transmutation performance analysis on coolant options in a hybrid reactor system design for high level waste incineration

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Seong-Hee; Siddique, Muhammad Tariq; Kim, Myung Hyun, E-mail: mhkim@khu.ac.kr

    2015-11-15

    Highlights: • Waste transmutation performance was compared and analyzed for seven different coolant options. • Reactions of fission and capture showed big differences depending on coolant options. • Moderation effect significantly affects on energy multiplication, tritium breeding and waste transmutation. • Reduction of radio-toxicities of TRUs showed different trend to coolant choice from performance of waste transmutation. - Abstract: A fusion–fission hybrid reactor (FFHR) is one of the most attractive candidates for high level waste transmutation. The selection of coolant affects the transmutation performance of a FFHR. LiPb coolant, as a conventional coolant for a FFHR, has problems such as reduction in neutron economic and magneto-hydro dynamics (MHD) pressure drop. Therefore, in this work, transmutation performance is evaluated and compared for various coolant options such as LiPb, H{sub 2}O, D{sub 2}O, Na, PbBi, LiF-BeF{sub 2} and NaF-BeF{sub 2} applicable to a hybrid reactor for waste transmutation (Hyb-WT). Design parameters measuring performance of a hybrid reactor were evaluated by MCNPX. They are k{sub eff}, energy multiplication factor, neutron absorption ratio, tritium breeding ratio, waste transmutation ratio, support ratio and radiotoxicity reduction. Compared to LiPb, H{sub 2}O and D{sub 2}O are not suitable for waste transmutation because of neutron moderation effect. Waste transmutation performances with Na and PbBi are similar to each other and not different much from LiPb. Even though molten salt such as LiF-BeF{sub 2} and NaF-BeF{sub 2} is good for avoiding MHD pressure drop problem, waste transmutation performance is dropped compared with LiPb.

  17. Investigation of a hydrogen mitigation system during large break loss-of-coolant accident for a two-loop pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dehjourian, Mehdi; Rahgoshay, Mohmmad; Jahanfamia, Gholamreza [Dept. of Nuclear Engineering, Science and Research Branch, Islamic Azad University of Tehran, Tehran (Iran, Islamic Republic of); Sayareh, Reza [Faculty of Electrical and Computer Engineering, Kerman Graduate University of Technology, Kerman (Iran, Islamic Republic of); Shirani, Amir Saied [Faculty of Engineering, Shahid Beheshti University, Tehran (Iran, Islamic Republic of)

    2016-10-15

    Hydrogen release during severe accidents poses a serious threat to containment integrity. Mitigating procedures are necessary to prevent global or local explosions, especially in large steel shell containments. The management of hydrogen safety and prevention of over-pressurization could be implemented through a hydrogen reduction system and spray system. During the course of the hypothetical large break loss-of-coolant accident in a nuclear power plant, hydrogen is generated by a reaction between steam and the fuel-cladding inside the reactor pressure vessel and also core concrete interaction after ejection of melt into the cavity. The MELCOR 1.8.6 was used to assess core degradation and containment behavior during the large break loss-of-coolant accident without the actuation of the safety injection system except for accumulators in Beznau nuclear power plant. Also, hydrogen distribution in containment and performance of hydrogen reduction system were investigated.

  18. Integrated low emission cleanup system for direct coal-fueled turbines (electrostatic agglomeration). Draft final technical report

    Energy Technology Data Exchange (ETDEWEB)

    Quimby, J.M.; Kumar, K.S.

    1992-12-31

    The objective of this contract was to investigate the removal of SO{sub x} and particulate matter from direct coal fired combustion gas streams at high temperature and high pressure conditions. This investigation was to be accomplished through a bench scale testing and evaluation program for SO{sub x} removal and the innovative particulate collection concept of particulate growth through electrostatic agglomeration followed by high efficiency mechanical collection. The process goal was to achieve control better than that required by 1979 New Source Performance Standards. During Phase I, the designs of the combustor and gas cleanup apparatus were successfully completed. Hot gas cleanup was designed to be accomplished at temperature levels between 1800{degrees} and 2500{degrees}F at pressures up to 15 atmospheres. The combustor gas flow rate could be varied between 0.2--0.5 pounds per second. The electrostatic agglomerator residence time could be varied between 0.25 to 3 seconds. In Phase II, all components were fabricated, and erected successfully. Test data from shakedown testing was obtained. Unpredictable difficulties in pilot plant erection and shakedown consumed more budget resources than was estimated and as a consequence DOE, METC, decided ft was best to complete the contract at the end of Phase II. Parameters studied in shakedown testing revealed that high-temperature high pressure electrostatics offers an alternative to barrier filtration in hot gas cleanup but more research is needed in successful system integration between the combustor and electrostatic agglomerator.

  19. Development and demonstration of a mobile reverse osmosis adsorption treatment system for environmental emergency clean-ups. Report No. EE-102

    Energy Technology Data Exchange (ETDEWEB)

    1988-01-01

    This study is concerned with the remedation of contaminated water resulting from the release of organic chemicals. Of particular concern is the cleanup of water contaminated by accidental spill of chemicals or petroleum products and the cleanup of water contaminated by oil chemicals disposed of in inadequately sealed or improperly designed landfills. This report presents the results of a project undertaken to develop and demonstrate a mobile reverse osmosis/adsorption system for treating water contaminated by organic chemicals.

  20. Thermal modeling in an engine cooling system to control coolant flow for fuel consumption improvement

    Science.gov (United States)

    Park, Sangki; Woo, Seungchul; Kim, Minho; Lee, Kihyung

    2016-09-01

    The design and evaluation of engine cooling and lubrication systems is generally based on real vehicle tests. Our goal here was to establish an engine heat balance model based on mathematical and interpretive analysis of each element of a passenger diesel engine cooling system using a 1-D numerical model. The purpose of this model is to determine ways of optimizing the cooling and lubrication components of an engine and then to apply these methods to actual cooling and lubrication systems of engines that will be developed in the future. Our model was operated under the New European Driving Cycle (NEDC) mode conditions, which represent the fuel economy evaluation mode in Europe. The flow rate of the cooling system was controlled using a control valve. Our results showed that the fuel efficiency was improved by as much as 1.23 %, cooling loss by 1.35 %, and friction loss by 2.21 % throughout NEDC modes by modification of control conditions.

  1. Bypass valve and coolant flow controls for optimum temperatures in waste heat recovery systems

    Science.gov (United States)

    Meisner, Gregory P

    2013-10-08

    Implementing an optimized waste heat recovery system includes calculating a temperature and a rate of change in temperature of a heat exchanger of a waste heat recovery system, and predicting a temperature and a rate of change in temperature of a material flowing through a channel of the waste heat recovery system. Upon determining the rate of change in the temperature of the material is predicted to be higher than the rate of change in the temperature of the heat exchanger, the optimized waste heat recovery system calculates a valve position and timing for the channel that is configurable for achieving a rate of material flow that is determined to produce and maintain a defined threshold temperature of the heat exchanger, and actuates the valve according to the calculated valve position and calculated timing.

  2. Interfacing systems LOCA (loss-of-coolant accidents): Pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bozoki, G.; Kohut, P.; Fitzpatrick, R.

    1989-02-01

    This report summarizes a study performed by Brookhaven National Laboratory for the Office of Nuclear Regulatory Research, Reactor and Plant Safety Issues Branch, Division of Reactor and Plant Systems, US Nuclear Regulatory Commission. This study was requested by the NRC in order to provide a technical basis for the resolution of Generic Issue 105 ''Interfacing LOCA at LWRs.'' This report deals with pressurized water reactors (PWRs). A parallel report was also accomplished for boiling water reactors. This study focuses on three representative PWRs and extrapolates the plant-specific findings for their generic applicability. In addition, a generic analysis was performed to investigate the cost-benefit aspects of imposing a testing program that would require some minimum level of leak testing of the pressure isolation valves on plants that presently have no such requirements. 28 refs., 31 figs., 64 tabs.

  3. Equipment Design and Cost Estimation for Small Modular Biomass Systems, Synthesis Gas Cleanup, and Oxygen Separation Equipment; Task 2: Gas Cleanup Design and Cost Estimates -- Black Liquor Gasification

    Energy Technology Data Exchange (ETDEWEB)

    Nexant Inc.

    2006-05-01

    As part of Task 2, Gas Cleanup and Cost Estimates, Nexant investigated the appropriate process scheme for removal of acid gases from black liquor-derived syngas for use in both power and liquid fuels synthesis. Two 3,200 metric tonne per day gasification schemes, both low-temperature/low-pressure (1100 deg F, 40 psi) and high-temperature/high-pressure (1800 deg F, 500 psi) were used for syngas production. Initial syngas conditions from each of the gasifiers was provided to the team by the National Renewable Energy Laboratory and Princeton University. Nexant was the prime contractor and principal investigator during this task; technical assistance was provided by both GTI and Emery Energy.

  4. Gas stream cleanup

    Energy Technology Data Exchange (ETDEWEB)

    Bossart, S.J.; Cicero, D.C.; Zeh, C.M.; Bedick, R.C.

    1990-08-01

    This report describes the current status and recent accomplishments of gas stream cleanup (GSCU) projects sponsored by the Morgantown Energy Technology Center (METC) of the US Department of Energy (DOE). The primary goal of the Gas Stream Cleanup Program is to develop contaminant control strategies that meet environmental regulations and protect equipment in advanced coal conversion systems. Contaminant control systems are being developed for integration into seven advanced coal conversion processes: Pressurized fludized-bed combustion (PFBC), Direct coal-fueled turbine (DCFT), Intergrated gasification combined-cycle (IGCC), Gasification/molten carbonate fuel cell (MCFC), Gasification/solid oxide fuel cell (SOFC), Coal-fueled diesel (CFD), and Mild gasification (MG). These advanced coal conversion systems present a significant challenge for development of contaminant control systems because they generate multi-contaminant gas streams at high-pressures and high temperatures. Each of the seven advanced coal conversion systems incorporates distinct contaminant control strategies because each has different contaminant tolerance limits and operating conditions. 59 refs., 17 figs., 5 tabs.

  5. Gas stream cleanup

    Energy Technology Data Exchange (ETDEWEB)

    Bossart, S.J.; Cicero, D.C.; Zeh, C.M.; Bedick, R.C.

    1990-08-01

    This report describes the current status and recent accomplishments of gas stream cleanup (GSCU) projects sponsored by the Morgantown Energy Technology Center (METC) of the US Department of Energy (DOE). The primary goal of the Gas Stream Cleanup Program is to develop contaminant control strategies that meet environmental regulations and protect equipment in advanced coal conversion systems. Contaminant control systems are being developed for integration into seven advanced coal conversion processes: Pressurized fludized-bed combustion (PFBC), Direct coal-fueled turbine (DCFT), Intergrated gasification combined-cycle (IGCC), Gasification/molten carbonate fuel cell (MCFC), Gasification/solid oxide fuel cell (SOFC), Coal-fueled diesel (CFD), and Mild gasification (MG). These advanced coal conversion systems present a significant challenge for development of contaminant control systems because they generate multi-contaminant gas streams at high-pressures and high temperatures. Each of the seven advanced coal conversion systems incorporates distinct contaminant control strategies because each has different contaminant tolerance limits and operating conditions. 59 refs., 17 figs., 5 tabs.

  6. ENVIRONMENTALLY REDUCING OF COOLANTS IN METAL CUTTING

    Directory of Open Access Journals (Sweden)

    Veijo KAUPPINEN

    2012-11-01

    Full Text Available Strained environment is a global problem. In metal industries the use of coolant has become more problematic in terms of both employee health and environmental pollution. It is said that the use of coolant forms approximately 8 - 16 % of the total production costs.The traditional methods that use coolants are now obviously becoming obsolete. Hence, it is clear that using a dry cutting system has great implications for resource preservation and waste reduction. For this purpose, a new cooling system is designed for dry cutting. This paper presents the new eco-friendly cooling innovation and the benefits gained by using this method. The new cooling system relies on a unit for ionising ejected air. In order to compare the performance of using this system, cutting experiments were carried out. A series of tests were performed on a horizontal turning machine and on a horizontal machining centre.

  7. Accelerating cleanup: Paths to closure

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-06-01

    This report describes the status of Environmental Management`s (EM`s) cleanup program and a direction forward to complete achievement of the 2006 vision. Achieving the 2006 vision results in significant benefits related to accomplishing EM program objectives. As DOE sites accelerate cleanup activities, risks to public health, the environment, and worker safety and health are all reduced. Finding more efficient ways to conduct work can result in making compliance with applicable environmental requirements easier to achieve. Finally, as cleanup activities at sites are completed, the EM program can focus attention and resources on the small number of sites with more complex cleanup challenges. Chapter 1 describes the process by which this report has been developed and what it hopes to accomplish, its relationship to the EM decision-making process, and a general background of the EM mission and program. Chapter 2 describes how the site-by-site projections were constructed, and summarizes, for each of DOE`s 11 Operations/Field Offices, the projected costs and schedules for completing the cleanup mission. Chapter 3 presents summaries of the detailed cleanup projections from three of the 11 Operations/Field Offices: Rocky Flats (Colorado), Richland (Washington), and Savannah River (South Carolina). The remaining eight Operations/Field Office summaries are in Appendix E. Chapter 4 reviews the cost drivers, budgetary constraints, and performance enhancements underlying the detailed analysis of the 353 projects that comprise EM`s accelerated cleanup and closure effort. Chapter 5 describes a management system to support the EM program. Chapter 6 provides responses to the general comments received on the February draft of this document.

  8. Analytical study on creep behavior of a tube of coolant piping system in nuclear power plant. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Miyazaki, Noriyuki [Kyushu Univ., Fukuoka (Japan); Hagihara, Seiya [Saga Univ., Saga (Japan); Chino, Eiichi; Maeda, Akio [MRI Systems Inc., Tokyo (Japan); Maruyama, Yu; Hashimoto, Kazuichiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-10-01

    During severe accident of a light water reactor (LWR), reactor coolant piping would be damaged when the piping is subjected to high internal pressure and very high temperature due to heat transfer from high-temperature gas and decay heat from wall-deposited fission product (FP), both from degraded core. In such a case, high-temperature fast creep deformation could be the main cause for the pipe failure. For the evaluation of piping integrity during severe accidents, a method to predict such high-temperature fast creep deformation should be developed, using a creep constitutive equation considering tertiary creep behavior which has not been considered well in the pipe failure analyses. In this study, a creep constitutive equation was developed first based on the Kachanov-Ravotnov isotropic damage rule that considers the tertiary creep behavior. JAERI creep tensile test data for both nuclear-grade cold-drawn SUS316N and hot-extruded SUS316 materials were used to determine coefficients of the developed constitutive equation. Using the developed constitutive equation, finite element analyses were performed for local creep deformation of coolant piping under two temperature conditions: uniform temperature and temperature gradient. The analytical results indicated the damage variable being integrated following the evolution of creep damage can indicate pipe wall internal damage condition quantitatively. The damage variable was confirmed further to be able to reproduce the observation in JAERI piping failure tests, that is, pipe failure from the wall outside. (author)

  9. Hot-gas cleanup system model development. Volume II. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Ushimaru, K.; Bennett, A.; Bekowies, P.J.

    1982-11-01

    Under Contract to the Department of Energy (DOE) through the Morgantown Energy Technology Center (METC), Flow Industries, Inc., has developed computer models to simulate the physical performance of five hot-gas cleanup devices for pressurized, fluidized-bed combustion (PFBC), combined-cycle power plants. Separate cost models have also been developed to estimate the cost of each device. The work leading to the development of these models is described in Volume I of this report. This volume contains the user's manuals for both the physical and cost models. The manuals for the physical models are given first followed by those for the cost models. Each manual is a complete and separate document. The model names and devices and their respective subroutine names are: (1) Moving Granular Bed Filter by Combustion Power Company, USRCGB, QFCOST; (2) Ceramic Bag Filter by Acurex, USRACB, QDCOST; (3) Electrostatic Granular Bed Filter by General Electric, USRGGB, QACOST; (4) Electrostatic Precipitator by Research Cottrell, USRCEP, QECOST; and (5) Electrocyclone by General Electric, USRGCY, QBCOST.

  10. On-Line Coolant Chemistry Analysis

    Energy Technology Data Exchange (ETDEWEB)

    LM Bachman

    2006-07-19

    Impurities in the gas coolant of the space nuclear power plant (SNPP) can provide valuable indications of problems in the reactor and an overall view of system health. By monitoring the types and amounts of these impurities, much can be implied regarding the status of the reactor plant. However, a preliminary understanding of the expected impurities is important before evaluating prospective detection and monitoring systems. Currently, a spectroscopy system is judged to hold the greatest promise for monitoring the impurities of interest in the coolant because it minimizes the number of entry and exit points to the plant and provides the ability to detect impurities down to the 1 ppm level.

  11. Development and piloting of an exposure database and surveillance system for DOE cleanup operations. Department of Energy.

    Science.gov (United States)

    LaMontagne, Anthony D; Van Dyke, Michael V; Martyny, John W; Simpson, Mark W; Holwager, Lee Ann; Clausen, Bret M; Ruttenber, A James

    2002-01-01

    An industrial hygiene exposure database and surveillance system was developed in partnership between National Institute for Occupational Safety and Health (NIOSH)-funded independent investigators and practicing industrial hygienists at the Rocky Flats Environmental Technology Site (RFETS) in Golden, Colo. RFETS is a former U.S. Department of Energy nuclear weapons plant that is now in cleanup phase. This project is presented as a case study in the development of an exposure database and surveillance system in terms that are generalizable to most other industries and work contexts. Steps include gaining organizational support; defining system purpose and scope; defining database elements and coding; planning practical and efficient analysis strategies; incorporating reporting capabilities; and anticipating communication strategies that maximize the probability that surveillance findings will feed back to preventive applications. For each of these topics, the authors describe both general considerations as well as the specific choices made for this system. An important feature of the system is a two-tier task-coding scheme comprising 33 categories of task groups. Examples of grouped analyses of exposure data captured during the system pilot period demonstrate applications to exposure control, medical surveillance, and other preventive measures.

  12. Application of the Severe Accident Code ATHLET-CD. Coolant injection to primary circuit of a PWR by mobile pump system in case of SBLOCA severe accident scenario

    Energy Technology Data Exchange (ETDEWEB)

    Jobst, Matthias; Wilhelm, Polina; Kliem, Soeren; Kozmenkov, Yaroslav [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Reactor Safety

    2017-06-01

    The improvement of the safety of nuclear power plants is a continuously on-going process. The analysis of transients and accidents is an important research topic, which significantly contributes to safety enhancements of existing power plants. In case of an accident with multiple failures of safety systems, core uncovery and heat-up can occur. In order to prevent the accident to turn into a severe one or to mitigate the consequences of severe accidents, different accident management measures can be applied. By means of numerical analyses performed with the compute code ATHLET-CD, the effectiveness of coolant injection with a mobile pump system into the primary circuit of a PWR was studied. According to the analyses, such a system can stop the melt progression if it is activated prior to 10 % of total core is molten.

  13. Effect of Coolant Inventories and Parallel Loop Interconnections on the Natural Circulation in Various Heat Transport Systems of a Nuclear Power Plant during Station Blackout

    Directory of Open Access Journals (Sweden)

    Avinash J. Gaikwad

    2008-01-01

    Full Text Available Provision of passive means to reactor core decay heat removal enhances the nuclear power plant (NPP safety and availability. In the earlier Indian pressurised heavy water reactors (IPHWRs, like the 220 MWe and the 540 MWe, crash cooldown from the steam generators (SGs is resorted to mitigate consequences of station blackout (SBO. In the 700 MWe PHWR currently being designed an additional passive decay heat removal (PDHR system is also incorporated to condense the steam generated in the boilers during a SBO. The sustainability of natural circulation in the various heat transport systems (i.e., primary heat transport (PHT, SGs, and PDHRs under station blackout depends on the corresponding system's coolant inventories and the coolant circuit configurations (i.e., parallel paths and interconnections. On the primary side, the interconnection between the two primary loops plays an important role to sustain the natural circulation heat removal. On the secondary side, the steam lines interconnections and the initial inventory in the SGs prior to cooldown, that is, hooking up of the PDHRs are very important. This paper attempts to open up discussions on the concept and the core issues associated with passive systems which can provide continued heat sink during such accident scenarios. The discussions would include the criteria for design, and performance of such concepts already implemented and proposes schemes to be implemented in the proposed 700 MWe IPHWR. The designer feedbacks generated, and critical examination of performance analysis results for the added passive system to the existing generation II & III reactors will help ascertaining that these safety systems/inventories in fact perform in sustaining decay heat removal and augmenting safety.

  14. Design, construction, and operation of a life-cycle test system for the evaluation of flue gas cleanup processes

    Energy Technology Data Exchange (ETDEWEB)

    Pennline, H.W.; Yeh, James T.; Hoffman, J.S. [USDOE Pittsburgh Energy Technology Center, PA (United States); Longton, E.J.; Vore, P.A.; Resnik, K.P.; Gromicko, F.N. [Gilbert/Commonwealth, Inc., Library, PA (United States)

    1995-12-01

    The Pittsburgh Energy Technology Center of the US Department of Energy has designed, constructed, and operated a Life-Cycle Test Systems (LCTS) that will be used primarily for the investigation of dry, regenerable sorbent flue gas cleanup processes. Sorbent continuously cycles from an absorber reactor where the pollutants are removed from the flue gas, to a regenerator reactor where the activity of the spent sorbent is restored and a usable by-product stream of gas is produced. The LCTS will initially be used to evaluate the Moving-Bed Copper Oxide Process by determining the effects of various process parameters on SO{sub 2} and NO{sub x} removals. The purpose of this paper is to document the design rationale and details, the reactor/component/instrument installation, and the initial performance of the system. Although the Moving-Bed Copper Oxide Process will be investigated initially, the design of the LCTS evolved to make the system a multipurpose, versatile research facility. Thus, the unit can be used to investigate various other processes for pollution abatement of SO{sub 2}, NO{sub x}, particulates, air toxics, and/or other pollutants.

  15. A MODEL FOR PREDICTING FISSION PRODUCT ACTIVITIES IN REACTOR COOLANT: APPLICATION OF MODEL FOR ESTIMATING I-129 LEVELS IN RADIOACTIVE WASTE

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, B.J.; Husain, A.

    2003-02-27

    A general model was developed to estimate the activities of fission products in reactor coolant and hence to predict a value for the I-129/Cs-137 scaling factor; the latter can be applied along with measured Cs-137 activities to estimate I-129 levels in reactor waste. The model accounts for fission product release from both defective fuel rods and uranium contamination present on in-core reactor surfaces. For simplicity, only the key release mechanisms were modeled. A mass balance, considering the two fuel source terms and a loss term due to coolant cleanup was solved to estimate fission product activity in the primary heat transport system coolant. Steady state assumptions were made to solve for the activity of shortlived fission products. Solutions for long-lived fission products are time-dependent. Data for short-lived radioiodines I-131, I-132, I-133, I-134 and I-135 were analyzed to estimate model parameters for I-129. The estimated parameter values were then used to determine I-1 29 coolant activities. Because of the chemical affinity between iodine and cesium, estimates of Cs-137 coolant concentrations were also based on parameter values similar to those for the radioiodines; this assumption was tested by comparing measured and predicted Cs-137 coolant concentrations. Application of the derived model to Douglas Point and Darlington Nuclear Generating Station plant data yielded estimates for I-129/I-131 and I-129/Cs-137 which are consistent with values reported for pressurized water reactors (PWRs) and boiling water reactors (BWRs). The estimated magnitude for the I-129/Cs-137 ratio was 10-8 - 10-7.

  16. Improvements made to the Bruce A upgraders and heavy water cleanup system as part of the Bruce A Units 1 and 2 restart project and commissioning results

    Energy Technology Data Exchange (ETDEWEB)

    Davloor, R., E-mail: ram.davloor@brucepower.com [Bruce Power, Tiverton, Ontario (Canada); Steinberg, G.; Boddy, C. [SNC Lavalin Nuclear, Oakville, Ontario (Canada); Rocci, D. [Aecon Nuclear, Cambridge, Ontario (Canada)

    2013-07-01

    As part of the Bruce A Units 1 and 2 Restart Project, major modifications and maintenance were completed on the heat transport and moderator upgraders and the heavy water cleanup system. This represents the first time that major rehabilitation has been done to such systems in a CANDU nuclear station for the purpose of life extension. Prior to shutdown in 1997, the upgraders and cleanup system significantly underperformed against the stated design. The rehabilitation, which included major design changes and implementation of new systems, resulted in the upgraders exceeding design throughput and making product with quality much better than specified. This paper describes the work done, results from inspections and follow-up, and performance data from commissioning. (author)

  17. A system for automatic biocide treatment of coolant cycles with concentration; Ein System einer automatisierten Biozidbehandlung von Kuehlkreislaeufen mit Eindickung

    Energy Technology Data Exchange (ETDEWEB)

    Lutat, A.; Pflug, H.D.; Schoenfelder, T. [KNG, Rostock (Germany)

    2000-07-01

    Regular, discontinuous treatment of industrial-scale recirculation cooling systems with microcide solutions must be carried out in view of the following aspects: Microbial growth inside the whole cooling systems during the interval times must be weakened so that no operational disturbances will occur and no additional measures for removal of organic depositions will be required. - Limiting values for microcide concentrations in the effluents must be observed. - Microcide treatment must be possible without load reduction of the power plant. - Consumption of microcide must be minimised. This means that both the time for treatment and the microcide concetration are limited. The contribution describes the automatic shock chlorination process employed at Rostock power station since a few years, i.e. water treatment and measuring instruments, investigation of microbial growth in the cooling water of a pilot plant, concept of biocide treatment of Rostock power station, cooling tower operation and chlorine concentrations during chlorination. [German] Bei der regelmaessigen, diskontinuierlichen Behandlung grosser Rueckkuehlsysteme mit Mikroziden, im allgemeinen chloraktiven Loesungen, sind mehrere Forderungen zu beruecksichtigen, naemlich: - Das mikrobielle Wachstum im gesamten Kuehlsystem muss in den Intervallzeiten soweit geschwaecht sein, dass es zu keinen betrieblichen Stoerungen kommt und moeglichst auch keine zusaetzlichen Massnahmen zur Entfernung von organischem Bewuchs noetig werden. - Die behoerdlichen Auflagen bezueglich des tolerierbaren Restgehalts an Mikrozid in der Abflut sind einzuhalten. - Eine Mikrozidbehandlung sollte ohne Lasteinschraenkung des Kraftwerks durchgefuehrt werden koennen. - Der Verbrauch an Mikrozid ist zu minimieren. Das bedeutet, dass nur eine begrenzte Betriebszeit bei geschlossener Kuehlturmabflut zur Verfuegung steht, um die wirksame Konzentration an Mikrozid bis zu dem behoerdlich zugelassenen Grenzwert abzubauen. Eine Ueberdosis

  18. Major weapon system environmental life-cycle cost estimating for Conservation, Cleanup, Compliance and Pollution Prevention (C3P2)

    Science.gov (United States)

    Hammond, Wesley; Thurston, Marland; Hood, Christopher

    1995-01-01

    The Titan 4 Space Launch Vehicle Program is one of many major weapon system programs that have modified acquisition plans and operational procedures to meet new, stringent environmental rules and regulations. The Environmental Protection Agency (EPA) and the Department of Defense (DOD) mandate to reduce the use of ozone depleting chemicals (ODC's) is just one of the regulatory changes that has affected the program. In the last few years, public environmental awareness, coupled with stricter environmental regulations, has created the need for DOD to produce environmental life-cycle cost estimates (ELCCE) for every major weapon system acquisition program. The environmental impact of the weapon system must be assessed and budgeted, considering all costs, from cradle to grave. The Office of the Secretary of Defense (OSD) has proposed that organizations consider Conservation, Cleanup, Compliance and Pollution Prevention (C(sup 3)P(sup 2)) issues associated with each acquisition program to assess life-cycle impacts and costs. The Air Force selected the Titan 4 system as the pilot program for estimating life-cycle environmental costs. The estimating task required participants to develop an ELCCE methodology, collect data to test the methodology and produce a credible cost estimate within the DOD C(sup 3)P(sup 2) definition. The estimating methodology included using the Program Office weapon system description and work breakdown structure together with operational site and manufacturing plant visits to identify environmental cost drivers. The results of the Titan IV ELCCE process are discussed and expanded to demonstrate how they can be applied to satisfy any life-cycle environmental cost estimating requirement.

  19. Phased Array Ultrasonic Examination of Reactor Coolant System (Carbon Steel-to-CASS) Dissimilar Metal Weld Mockup Specimen

    Energy Technology Data Exchange (ETDEWEB)

    Crawford, S. L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Cinson, A. D. [US Nuclear Regulatory Commission (NRC), Washington, DC (United States); Diaz, A. A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Anderson, M. T. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-11-23

    In the summer of 2009, Pacific Northwest National Laboratory (PNNL) staff traveled to the Electric Power Research Institute (EPRI) NDE Center in Charlotte, North Carolina, to conduct phased-array ultrasonic testing on a large bore, reactor coolant pump nozzle-to-safe-end mockup. This mockup was fabricated by FlawTech, Inc. and the configuration originated from the Port St. Lucie nuclear power plant. These plants are Combustion Engineering-designed reactors. This mockup consists of a carbon steel elbow with stainless steel cladding joined to a cast austenitic stainless steel (CASS) safe-end with a dissimilar metal weld and is owned by Florida Power & Light. The objective of this study, and the data acquisition exercise held at the EPRI NDE Center, were focused on evaluating the capabilities of advanced, low-frequency phased-array ultrasonic testing (PA-UT) examination techniques for detection and characterization of implanted circumferential flaws and machined reflectors in a thick-section CASS dissimilar metal weld component. This work was limited to PA-UT assessments using 500 kHz and 800 kHz probes on circumferential flaws only, and evaluated detection and characterization of these flaws and machined reflectors from the CASS safe-end side only. All data were obtained using spatially encoded, manual scanning techniques. The effects of such factors as line-scan versus raster-scan examination approaches were evaluated, and PA-UT detection and characterization performance as a function of inspection frequency/wavelength, were also assessed. A comparative assessment of the data is provided, using length-sizing root-mean-square-error and position/localization results (flaw start/stop information) as the key criteria for flaw characterization performance. In addition, flaw signal-to-noise ratio was identified as the key criterion for detection performance.

  20. Nuclear radiation cleanup and uranium prospecting

    Energy Technology Data Exchange (ETDEWEB)

    Mariella, Jr., Raymond P.; Dardenne, Yves M.

    2017-01-03

    Apparatus, systems, and methods for nuclear radiation cleanup and uranium prospecting include the steps of identifying an area; collecting samples; sample preparation; identification, assay, and analysis; and relating the samples to the area.

  1. Novel Cleanup Agents Designed Exclusively for Oil Field Membrane Filtration Systems Low Cost Field Demonstrations of Cleanup Agents in Controlled Experimental Environments

    Energy Technology Data Exchange (ETDEWEB)

    David Burnett; Harold Vance

    2007-08-31

    The goal of our project is to develop innovative processes and novel cleaning agents for water treatment facilities designed to remove fouling materials and restore micro-filter and reverse osmosis (RO) membrane performance. This project is part of Texas A&M University's comprehensive study of the treatment and reuse of oilfield brine for beneficial purposes. Before waste water can be used for any beneficial purpose, it must be processed to remove contaminants, including oily wastes such as residual petroleum hydrocarbons. An effective way of removing petroleum from brines is the use of membrane filters to separate oily waste from the brine. Texas A&M and its partners have developed highly efficient membrane treatment and RO desalination for waste water including oil field produced water. We have also developed novel and new cleaning agents for membrane filters utilizing environmentally friendly materials so that the water from the treatment process will meet U.S. EPA drinking water standards. Prototype micellar cleaning agents perform better and use less clean water than alternate systems. While not yet optimized, the new system restores essentially complete membrane flux and separation efficiency after cleaning. Significantly the amount of desalinated water that is required to clean the membranes is reduced by more than 75%.

  2. Carbon Formation and Metal Dusting in Hot-Gas Cleanup Systems of Coal Gasifiers

    Energy Technology Data Exchange (ETDEWEB)

    Tortorelli, Peter F.; Judkins, Roddie R.; DeVan, Jackson H.; Wright, Ian G.

    1995-12-31

    There are several possible materials/systems degradation modes that result from gasification environments with appreciable carbon activities. These processes, which are not necessarily mutually exclusive, include carbon deposition, carburization, metal dusting, and CO disintegration of refractories. Carbon formation on solid surfaces occurs by deposition from gases in which the carbon activity (a sub C) exceeds unity. The presence of a carbon layer CO can directly affect gasifier performance by restricting gas flow, particularly in the hot gas filter, creating debris (that may be deposited elsewhere in the system or that may cause erosive damage of downstream components), and/or changing the catalytic activity of surfaces.

  3. Development of tool for implementation of energy economic cooling systems with natural coolants. Part 1: Installation and calculation of annual energy consumption; Udvikling af vaerktoej til implementering af energioekonomiske koeleanlaeg med naturlige koelemidler. Del 1: Opstilling og beregning af aersenergiforbrug

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2006-02-15

    This report presents the results reached during the project 'Development of tools for implementation of energy economic cooling systems with natural coolants. Part 1: Installation and calculation of annual energy consumption'. The report is primarily directed at grant-awarding authorities, as the results only propose recommendations concerning further research. The aim of making a tool is, quite simply, to compare the economy for alternative designs of cooling systems when both investments and operational costs during the system's service life are taken into consideration. Thus the tool gives a buyer of a cooling system the possibility of choosing the most energy economic system, and hereby spread the most energy economic systems in connection with the system replacement that will take place due to out phasing of CFC/HCFC/HFC coolants. The project has been divided into two phases. Phase no. 1, this phase, contains the development of the tool. (BA)

  4. Cleanups in My Community

    Data.gov (United States)

    U.S. Environmental Protection Agency — Cleanups in My Community (CIMC) is a public web application that enables integrated access through maps, lists and search filtering to site-specific information EPA...

  5. Carbon formation and metal dusting in hot-gas cleanup systems of coal gasifiers

    Energy Technology Data Exchange (ETDEWEB)

    Judkins, R.R.; Tortorelli, P.F.; Judkins, R.R.; DeVan, J.H.; Wright, I.G. [Oak Ridge National Lab., TN (United States). Metals and Ceramics Div.

    1995-11-01

    The product gas resulting from the partial oxidation of Carboniferous materials in a gasifier is typically characterized by high carbon and sulfur, but low oxygen, activities and, consequently, severe degradation of the structural and functional materials can occur. The objective of this task was to establish the potential risks of carbon deposition and metal dusting in advanced coal gasification processes by examining the current state of knowledge regarding these phenomena, making appropriate thermochemical calculations for representative coal gasifiers, and addressing possible mitigation methods. The paper discusses carbon activities, iron-based phase stabilities, steam injection, conditions that influence kinetics of carbon deposition, and influence of system operating parameters on carbon deposition and metal dusting.

  6. Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

    Directory of Open Access Journals (Sweden)

    Hwang Bae

    2017-08-01

    Full Text Available Three small-break loss-of-coolant accident (SBLOCA tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor, i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

  7. Equipment Design and Cost Estimation for Small Modular Biomass Systems, Synthesis Gas Cleanup, and Oxygen Separation Equipment; Task 9: Mixed Alcohols From Syngas -- State of Technology

    Energy Technology Data Exchange (ETDEWEB)

    Nexant Inc.

    2006-05-01

    This deliverable is for Task 9, Mixed Alcohols from Syngas: State of Technology, as part of National Renewable Energy Laboratory (NREL) Award ACO-5-44027, ''Equipment Design and Cost Estimation for Small Modular Biomass Systems, Synthesis Gas Cleanup and Oxygen Separation Equipment''. Task 9 supplements the work previously done by NREL in the mixed alcohols section of the 2003 technical report Preliminary Screening--Technical and Economic Assessment of Synthesis Gas to Fuels and Chemicals with Emphasis on the Potential for Biomass-Derived Syngas.

  8. Development and application of an information-analytic system on the problem of flow accelerated corrosion of pipeline elements in the secondary coolant circuit of VVER-440-based power units at the Novovoronezh nuclear power plant

    Science.gov (United States)

    Tomarov, G. V.; Povarov, V. P.; Shipkov, A. A.; Gromov, A. F.; Kiselev, A. N.; Shepelev, S. V.; Galanin, A. V.

    2015-02-01

    Specific features relating to development of the information-analytical system on the problem of flow-accelerated corrosion of pipeline elements in the secondary coolant circuit of the VVER-440-based power units at the Novovoronezh nuclear power plant are considered. The results from a statistical analysis of data on the quantity, location, and operating conditions of the elements and preinserted segments of pipelines used in the condensate-feedwater and wet steam paths are presented. The principles of preparing and using the information-analytical system for determining the lifetime to reaching inadmissible wall thinning in elements of pipelines used in the secondary coolant circuit of the VVER-440-based power units at the Novovoronezh NPP are considered.

  9. Tornado wind-loading requirements based on risk assessment techniques (For specific reactor safety Class 1 coolant system features)

    Science.gov (United States)

    Deobald, Theodore L.; Coles, Garill A.; Smith, Gary L.

    1992-01-01

    Regulations require that nuclear power plants be protected from tornado winds. If struck by a tornado, a plant must be capable of safely shutting down and removing decay heat. Probabilistic techniques are used to show that risk to the public from the U.S. Department of Energy SP-100 reactor is acceptable without tornado hardening parts of the secondary system. Relaxed requirements for design wind loadings will result in significant cost savings. To demonstrate an acceptable level of risk, this document examines tornado-initiated accidents. The two tornado-initiated accidents examined in detail are loss of cooling resulting in core damage and loss of secondary system boundary integrity leading to sodium release. Loss of core cooling is analyzed using fault/event tree models. Loss of secondary system boundary integrity is analyzed by comparing the consequences to acceptance criteria for the release of radioactive material or alkali metal aerosol.

  10. APT Blanket System Loss-of-Coolant Accident (LOCA) Based on Initial Conceptual Design - Case 2: with Beam Shutdown Only

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

    This report is one of a series of reports that document normal operation and accident simulations for the Accelerator Production of Tritium (APT) blanket heat removal system. These simulations were performed for the Preliminary Safety Analysis Report. This report documents the results of simulations of a Loss-of-Flow Accident (LOFA) where power is lost to all of the pumps that circulate water in the blanket region, the accelerator beam is shut off and neither the residual heat removal nor cavity flood systems operate.

  11. KSTAR Severe Accident Analysis using MELCOR : Ex-vessel Coolant Pipe Break with Failure of Fusion Power Termination System

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Sung Bo; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of)

    2015-10-15

    To investigate the consequence of severe accidents in fusion reactor, a number of thermal hydraulics simulation codes were used (ECART, INTRA, ATHENA/RELAP and so on). MELCOR is chosen as the thermal hydraulics code to simulate the consequence of radioactive material release from accident in preliminary safety report. Capability of the simulation code for fusion reactor severe accident analysis is ability to simulate the hydraulic system in ITER and the transport phenomenon of radionuclides. MELCOR is a fully integrated code that models the accidents in Light Water Reactor (LWR). There are three kinds of radioactive materials in fusion reactor; tritium (or Tiritiated water: HTO), activation products (AP) of divertor or first-wall and activated corrosion products(ACP). In generic Site Safety Report (GSSR), the release guidelines for tritium and activation products are listed for normal operation, incidents, and accidents. And this guidelines presented in Table 1. Not only ITER, the KSTAR (Korea Superconducting Tokamak Advanced Research) is also developing fusion research reactor. The scale of facility is smaller than ITER but this small scale of facility offers the experimental flexibility to develop fusion technology. The major differences between KSTAR and ITER systems are presented in Table 2. Fusion source difference between KSTAR and ITER is D-D fusion reaction (Deuterium-Deuterium fusion reaction) and D-T fusion reaction (Deuterium-Tritium fusion reaction). This D-D fusion makes one tritium by 50 percent chance. The radioactivity of tritium is small to consider compared to radioactive materials in nuclear fission reactor. This reaction is presented in equation (1) In the present work, conservatively estimated tritium inventory amount in KSTAR is used with one of the most severe accident in ITER; Ex-vessel pipe break with Fusion Power Termination System (FPTS). The MELCOR KSTAR input is made by scaling down the ITER input deck. So, the detail system is not same

  12. Environmental compliance and cleanup

    Energy Technology Data Exchange (ETDEWEB)

    Black, D.G.

    1995-06-01

    This section of the 1994 Hanford Site Environmental Report summarizes the roles of the principal agencies, organizations, and public in environmental compliance and cleanup of the Hanford Site. Regulatory oversight, the Federal Facility Agreement and Consent Order, the role of Indian tribes, public participation, and CERCLA Natural Resource Damage Assessment Trustee Activities are all discussed.

  13. ANALYSIS OF THE IMPACT PROPERTIES OF THE COOLANT RECOVERY SYSTEM HEAT LOSSES OF COMBINED COMPRESSOR-POWER PLANT ON ITS CHARACTERISTICS

    Directory of Open Access Journals (Sweden)

    Yusha V.L.

    2012-12-01

    Full Text Available The paper presents results of theoretical analysis of the effectiveness of an ideal thermodynamic cycle internal combustion engine combined with an external utilization of exhaust heat. The influence of the properties of the coolant circuit of utilization on its operational parameters and characteristics of the power plant.

  14. Parameters influencing the transgranular stress corrosion cracking behaviour of austenitic stainless steels in systems conveying reactor coolant

    Energy Technology Data Exchange (ETDEWEB)

    Kilian, R.; Wesseling, U. [Framatome ANP (Germany); Wachter, O. [E.ON Kernkraft (Germany); Widera, M. [RWE Power (Germany); Brummer, G. [HEW - (Germany); Ilg, U. [EnBW - (Germany)

    2002-07-01

    During replacement of an auxiliary system in the German PWR KKS (NPP Stade) a damage was detected in a valve housing and in the connected piping both made from stabilised austenitic stainless steel. During operation stagnant conditions are present in this area. Based on the failure analysis chloride induced stress corrosion cracking (SCC) was found as the dominating root cause. In the open literature many cases of corrosion observed in the water/steam interface in valve components as well as in adjacent portions of auxiliary circuits made of un-stabilized stainless steels are mentioned. A common feature of the reported cases is that transgranular cracking was found. Extensive laboratory investigations revealed that non-stabilised austenitic stainless steels are also sensitive to transgranular cracking in boric acid solutions particularly in concentrated solutions. Often these solutions are contaminated with chlorides and/or oxygen is present. Taking into account the literature data the question could arise whether the above mentioned cracking may be also caused by boric acid attack. Thus, for stabilised stainless steels laboratory exposure tests at 80 C in saturated aerated boric acid solution and at 300 C in (at 100 C) saturated, oxygen free boric acid solution have been performed. Double-U-bend specimens and wedge loaded 1T-CT specimens made of Ti- and Nb-stabilised austenitic stainless steels were used. The results revealed no evidence of crack initiation and crack growth. Based on the laboratory results and the literature data an attempt is undertaken to separate parameters influencing chloride induced SCC from the effect of boric acid. (authors)

  15. Exploring new coolants for nuclear breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lafuente, A., E-mail: anlafuente@etsii.upm.e [ETSII-UPM, c/Jose Gutierrez Abascal, 2, 28006 Madrid (Spain); Piera, M. [ETSII:UNED, c/Juan del Rosal, 12, 28040 Madrid (Spain)

    2010-06-15

    Breeder reactors are considered a unique tool for fully exploiting natural nuclear resources. In current Light Water Reactors (LWR), only 0.5% of the primary energy contained in the nuclei removed from a mine is converted into useful heat. The rest remains in the depleted uranium or spent fuel. The need to improve resource-efficiency has stimulated interest in Fast-Reactor-based fuel cycles, which can exploit a much higher fraction of the energy content of mined uranium by burning U-238, mainly after conversion into Pu-239. Thorium fuel cycles also offer several potential advantages over a uranium fuel cycle. The coolant initially selected for most of the FBR programs launched in the 1960s was sodium, which is still considered the best candidate for these reactors. However, Na-cooled FBRs have a positive void reactivity coefficient. Among other factors, this fundamental drawback has resulted in the canceled deployment of these reactors. Therefore, it seems reasonable to explore new options for breeder coolants. In this paper, a proposal is presented for a new molten salt (F{sub 2}Be) coolant that could overcome the safety issues related to the positive void reactivity coefficient of molten metal coolants. Although it is a very innovative proposal that would require an extensive R and D program, this paper presents the very appealing properties of this salt when using a specific type of fuel that is similar to that of pebble bed reactors. The F{sub 2}Be concept was studied over a typical MOX composition and extended to a thorium-based cycle. The general analysis took into account the requirements for criticality (opening the option of hybrid subcritical systems); the requirements for breeding; and the safety requirement of having a negative coolant void reactivity coefficient. A design window was found in the definition of a F{sub 2}Be cooled reactor where the safety requirement was met, unlike for molten metal-cooled reactors, which always have positive void

  16. Exploring new coolants for nuclear breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lafuente, A. [ETSI Industriales-Universidad Politecnica de Madrid, C/Jose Gutierrez Abascal, 2. 28006 Madrid (Spain)

    2010-07-01

    Breeder reactors are considered the unique tool for fully exploiting the natural nuclear resources. In current LWR, only a 0.5% of the primary energy contained in the nuclei removed from the mine is converted into useful heat, with the rest remaining in the depleted uranium or in the spent fuel. The objective of resource-efficiency stimulated the interest in Fast- Reactor-based fuel cycles which can exploit a much higher fraction of the energy content of the mined uranium by burning U-238, mainly after conversion into Pu-239. Thorium fuel cycles would also offers several potential advantages over a uranium fuel cycle. The coolant initially chosen for most of the FBR programs launched in the 60's was sodium, which still is considered the best candidate for these reactors. However, Na-cooled FBR have a positive void reactivity coefficient, which has been among others, a fundamental drawback that has cancelled the deployment of these reactors. Therefore, it seems reasonable to explore totally new options on coolants for breeders. In this paper, a proposal is presented on a new molten salt (F{sub 2}Be) coolant that could overcome the safety issues related to the positive void reactivity coefficient of molten metal coolants. Although it is a very innovative proposal that would need an extensive R and D programme, this paper presents the very appealing properties of this salt, in the case of using a specific type of fuel, similar to that of pebble bed reactors. The concept will be studied over a typical MOX composition and extended to a Thorium-based cycle. The general analysis takes into account requirements for criticality (opening the option of hybrid subcritical systems); requirements for breeding; and the safety requirement of having a negative coolant void reactivity coefficient. A design window is found in the definition of a F{sub 2}Be cooled reactor where the safety requirement is met, unlike for molten metal cooled reactors which always have positive void

  17. Rotor dynamic analysis of main coolant pump

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chong Won; Seo, Jeong Hwan; Kim, Choong Hwan; Shin, Jae Chul; Wang, Lei Tian [Korea Advanced Institute of Science and Technology, Taejon (Korea)

    1999-03-01

    A rotor dynamic analysis program DARBS/MCP, for the main coolant pump of the integral reactor, has been developed. The dynamic analysis model of the main coolant pump includes a vertical shaft, three grooved radial journal bearings and gaps that represent the structure-fluid interaction effects between the rotor and the lubricant fluid. The electromagnetic force from the motor and the hydro-dynamic force induced by impeller are the major sources of vibration that may affect the rotor system stability. DARBS/MCP is a software that is developed to effectively analyze the dynamics of MCP rotor systems effectively by applying powerful numerical algorithms such as FEM with modal truncation and {lambda}-matrix method for harmonic analysis. Main design control parameters, that have much influence to the dynamic stability, have been found by Taguchi's sensitivity analysis method. Design suggestions to improve the stability of MCP rotor system have been documented. The dynamic bearing parameters of the journal bearings used for main coolant pump have been determined by directly solving the Reynolds equation using FDM method. Fluid-structure interaction effect that occurs at the small gaps between the rotor and the stator were modeled as equivalent seals, the electromagnetic force effect was regarded as a linear negative radial spring and the impeller was modeled as a rigid disk with hydrodynamic and static radial force. Although there exist critical speeds in the range of operational speeds for type I and II rotor systems, the amplitude of vibration appears to be less than the vibration limit set by the API standards. Further more, it has been verified that the main design parameters such as the clearance and length of journal bearings, and the static radial force of impeller should be properly adjusted, in order to the improve dynamic stability of the rotor system. (author). 39 refs., 81 figs., 17 tabs.

  18. Equipment Design and Cost Estimation for Small Modular Biomass Systems, Synthesis Gas Cleanup, and Oxygen Separation Equipment; Task 1: Cost Estimates of Small Modular Systems

    Energy Technology Data Exchange (ETDEWEB)

    Nexant Inc.

    2006-05-01

    This deliverable is the Final Report for Task 1, Cost Estimates of Small Modular Systems, as part of NREL Award ACO-5-44027, ''Equipment Design and Cost Estimation for Small Modular Biomass Systems, Synthesis Gas Cleanup and Oxygen Separation Equipment''. Subtask 1.1 looked into processes and technologies that have been commercially built at both large and small scales, with three technologies, Fluidized Catalytic Cracking (FCC) of refinery gas oil, Steam Methane Reforming (SMR) of Natural Gas, and Natural Gas Liquids (NGL) Expanders, chosen for further investigation. These technologies were chosen due to their applicability relative to other technologies being considered by NREL for future commercial applications, such as indirect gasification and fluidized bed tar cracking. Research in this subject is driven by an interest in the impact that scaling has on the cost and major process unit designs for commercial technologies. Conclusions from the evaluations performed could be applied to other technologies being considered for modular or skid-mounted applications.

  19. Louisiana's statewide beach cleanup

    Science.gov (United States)

    Lindstedt, Dianne M.; Holmes, Joseph C.

    1989-01-01

    Litter along Lousiana's beaches has become a well-recognized problem. In September 1987, Louisiana's first statewide beach cleanup attracted about 3300 volunteers who filled 16,000 bags with trash collected along 15 beaches. An estimated 800,173 items were gathered. Forty percent of the items were made of plastic and 11% were of polystyrene. Of all the litter collected, 37% was beverage-related. Litter from the oil and gas, commercial fishing, and maritime shipping industries was found, as well as that left by recreational users. Although beach cleanups temporarily rid Louisiana beaches of litter, the real value of the effort is in public participation and education. Civic groups, school children, and individuals have benefited by increasing their awareness of the problems of trash disposal.

  20. Molten Fuel-Coolant Interactions induced by coolant injection into molten fuel

    Energy Technology Data Exchange (ETDEWEB)

    Park, H.S.; Yamano, Norihiko; Maruyama, Yu; Moriyama, Kiyofumi; Yang, Y.; Sugimoto, Jun [Severe Accident Research Laboratory, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan)

    1999-07-01

    To investigate Molten Fuel-Coolant Interactions (MFCIs) in various contact geometries, an experimental program, called MUSE (MUlti-configurations in Steam Explosions), has been initiated under the ALPHA program at JAERI in Japan. The first series of MUSE test has been focused on the coolant injection (CI) and stratified modes of FCIs using water as coolant and molten thermite as molten fuel. The effects of water jet subcooling, jet dynamics, jet shape and system constraint on FCIs energetic in these modes were experimentally investigated by precisely measuring their mechanical energy release in the MUSE facility. It was observed that measured mechanical energy increased with increasing of jet subcooling in a weakly constraint system but decreased in a strongly constraint system. FCI energetic also increased with increasing of water jet velocity. These results suggested that the penetration and dispersion phenomena of a water jet inside a melt determined the mixing conditions of FCIs in these contact modes and consequently played important roles on FCI energetics. To understand fundamental physics of these phenomena and possible mixing conditions in the MUSE tests, a set of visualization tests with several pairs of jet-pool liquids in non-boiling and isothermal conditions were carried out. Numerical simulations of a water jet penetrating into a water pool at non-boiling conditions showed similar behaviors to those observed in the visualization tests. (author)

  1. Data center coolant switch

    Energy Technology Data Exchange (ETDEWEB)

    Iyengar, Madhusudan K.; Parida, Pritish R.; Schultz, Mark D.

    2015-10-06

    A data center cooling system is operated in a first mode; it has an indoor portion wherein heat is absorbed from components in the data center, and an outdoor heat exchanger portion wherein outside air is used to cool a first heat transfer fluid (e.g., water) present in at least the outdoor heat exchanger portion of the cooling system during the first mode. The first heat transfer fluid is a relatively high performance heat transfer fluid (as compared to the second fluid), and has a first heat transfer fluid freezing point. A determination is made that an appropriate time has been reached to switch from the first mode to a second mode. Based on this determination, the outdoor heat exchanger portion of the data cooling system is switched to a second heat transfer fluid, which is a relatively low performance heat transfer fluid, as compared to the first heat transfer fluid. It has a second heat transfer fluid freezing point lower than the first heat transfer fluid freezing point, and the second heat transfer fluid freezing point is sufficiently low to operate without freezing when the outdoor air temperature drops below a first predetermined relationship with the first heat transfer fluid freezing point.

  2. Coolants for a closed-cycle cooling system of device for removing the condensation heat of teh waste steam from steam-turbine power plants

    Energy Technology Data Exchange (ETDEWEB)

    Richter, D.; Heeren, H.

    1976-03-25

    A series of demands are made on the coolant for the purpose mentioned. They should (a) have a large heat of evaporation, a low freezing point, a high vapor density at evaporation-condensation temperature, high heat transfer numbers with bubble vapors and condensation and a low saturated vapor pressure at operational temperature, (b) not be poisonous and inflammable, and (c) be economical compared to forced-water circulation. According to the invention, an azeotropic mixture of water and methyl-propylene glycol fulfills these requirements.

  3. Loss of Coolant Accident Analysis Methodology for SMART-P

    Energy Technology Data Exchange (ETDEWEB)

    Bae, K. H.; Lee, G. H.; Yang, S. H.; Yoon, H. Y.; Kim, S. H.; Kim, H. C

    2006-02-15

    The analysis methodology on the Loss-of-coolant accidents (LOCA's) for SMART-P is described in this report. SMART-P is an advanced integral type PWR producing a maximum thermal power of 65.5 MW with metallic fuel. LOCA's are hypothetical accidents that would result from the loss of reactor coolant, at a rate in excess of the capability of the reactor coolant makeup system, from breaks in pipes in the reactor coolant pressure boundary up to and including a break equivalent in size to the double-ended rupture of the largest pipe in the reactor coolant system. Since SMART-P contains the major primary circuit components in a single Reactor Pressure Vessel (RPV), the possibility of a large break LOCA (LBLOCA) is inherently eliminated and only the small break LOCA is postulated. This report describes the outline and acceptance criteria of small break LOCA (SBLOCA) for SMART-P and documents the conservative analytical model and method and the analysis results using the TASS/SMR code. This analysis method is applied in the SBLOCA analysis performed for the ECCS performance evaluation which is described in the section 6.3.3 of the safety analysis report. The prediction results of SBLOCA analysis model of SMART-P for the break flow, system's pressure and temperature distributions, reactor coolant distribution, single and two-phase natural circulation phenomena, and the time of major sequence of events, etc. should be compared and verified with the applicable separate and integral effects test results. Also, it is required to set-up the feasible acceptance criteria applicable to the metallic fueled integral reactor of SMART-P. The analysis methodology for the SBLOCA described in this report will be further developed and validated as the design and licensing status of SMART-P evolves.

  4. Effect of emergency core cooling system flow reduction on channel temperature during recirculation phase of large break loss-of-coolant accident at Wolsong unit 1

    Directory of Open Access Journals (Sweden)

    Seon Oh Yu

    2017-08-01

    Full Text Available The feasibility of cooling in a pressurized heavy water reactor after a large break loss-of-coolant accident has been analyzed using Multidimensional Analysis of Reactor Safety-KINS Standard code during the recirculation phase. Through evaluation of sensitivity of the fuel channel temperature to various effective recirculation flow areas, it is determined that proper cooling of the fuel channels in the broken loop is feasible if the effective flow area remains above approximately 70% of the nominal flow area. When the flow area is reduced by more than approximately 25% of the nominal value, however, incipience of boiling is expected, after which the thermal integrity of the fuel channel can be threatened. In addition, if a dramatic reduction of the recirculation flow occurs, excursions and frequent fluctuations of temperature in the fuel channels are likely to be unavoidable, and thus damage to the fuel channels would be anticipated. To resolve this, emergency coolant supply through the newly installed external injection path can be used as one alternative means of cooling, enabling fuel channel integrity to be maintained and permanently preventing severe accident conditions. Thus, the external injection flow required to guarantee fuel channel coolability has been estimated.

  5. Lithium as a blanket coolant

    Energy Technology Data Exchange (ETDEWEB)

    Wells, W.M.

    1977-01-01

    Recent re-assessment of tokamak reactors which move towards smaller size and lower required field strength (higher beta)/sup 2/ change the picture as regards the magnitude of MHD effects on flow resistance for lithium coolant. Perhaps the most important consequence of this as regards use of this coolant is that of clear acceptability of such effects when the flow is predominantly transverse to the magnetic field. This permits defining a blanket that consists entirely of round tubes containing the circulated lithium with voids between the tubes. Required thermal-hydraulic calculations are then on bases which are well established, especially in view of recent results dealing with perturbations of ducts and magnetic fields. Mitigation of MHD effects is feasible through tapering of tube wall thickness or use of insulated layers, but their use was not mandatory for the assumed conditions. Blanket configurations utilizing flowing lithium in round tubes immersed in static lithium may be suitable, but calculational methods do not now exist for this situation. Use of boiling potassium or cesium appears to be prohibitive in terms of vapor flow area when temperature levels are consistent with stainless steel. Liquid sodium, in addition to not being a breeding material, requires higher velocity than lithium for the same heat removal.

  6. Improvement of Measurement Accuracy of Coolant Flow in a Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jintae; Kim, Jong-Bum; Joung, Chang-Young; Ahn, Sung-Ho; Heo, Sung-Ho; Jang, Seoyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this study, to improve the measurement accuracy of coolant flow in a coolant flow simulator, elimination of external noise are enhanced by adding ground pattern in the control panel and earth around signal cables. In addition, a heating unit is added to strengthen the fluctuation signal by heating the coolant because the source of signals are heat energy. Experimental results using the improved system shows good agreement with the reference flow rate. The measurement error is reduced dramatically compared with the previous measurement accuracy and it will help to analyze the performance of nuclear fuels. For further works, out of pile test will be carried out by fabricating a test rig mockup and inspect the feasibility of the developed system. To verify the performance of a newly developed nuclear fuel, irradiation test needs to be carried out in the research reactor and measure the irradiation behavior such as fuel temperature, fission gas release, neutron dose, coolant temperature, and coolant flow rate. In particular, the heat generation rate of nuclear fuels can be measured indirectly by measuring temperature variation of coolant which passes by the fuel rod and its flow rate. However, it is very difficult to measure the flow rate of coolant at the fuel rod owing to the narrow gap between components of the test rig. In nuclear fields, noise analysis using thermocouples in the test rig has been applied to measure the flow velocity of coolant which circulates through the test loop.

  7. Trivalent copper chelate-luminol chemiluminescence system for highly sensitive CE detection of dopamine in biological sample after clean-up using SPE.

    Science.gov (United States)

    Wang, Lin; Liu, Ying; Xie, Haoyue; Fu, Zhifeng

    2012-06-01

    A transition metal chelate unstable at a high oxidation state, diperiodatocuprate (III) (K₅[Cu(HIO₆)₂], DPC), was synthesized and applied in the luminol-based chemiluminescence (CL) system for highly sensitive CE end-column detection of dopamine (DA). This method was based on the fact that DA enhanced the CL emission resulting from the reaction between luminol and DPC in alkaline medium. The DPC-luminol-DA CL system showed very intensive emission and very fast kinetic characteristics, thus resulting in a high sensitivity in flow-through detection mode for CE. Under optimal conditions, the linear range was 1.0 × 10⁻⁸-5.0 × 10⁻⁵ g/mL (R² = 0.9984) with a limit of detection of 6.0 × 10⁻⁹ g/mL (S/N = 3). The RSDs of the peak height and the migration time were about 4.2 and 2.4% for a standard sample at 3.0 × 10⁻⁶ g/mL (n = 5), respectively. The presented method has been successfully used for the determination of DA in commercial preparation and human urine samples after clean-up using SPE.

  8. Dynamic Analysis of Coolant Channel and Its Internals of Indian 540 MWe PHWR Reactor

    Directory of Open Access Journals (Sweden)

    N. Dharmaraju

    2008-01-01

    Full Text Available The horizontal coolant channel is one of the important parts of primary heat transport system in PHWR type of reactors. There are in all 392 channels in the core of Indian 540 MWe reactor. Each channel houses 13 natural uranium fuel bundles and shielding and sealing plugs one each on either side of the channel. The heavy water coolant flows through the coolant channel and carries the nuclear heat to outside the core for steam generation and power production in the turbo-generator. India has commissioned one 540 MWe PHWR reactor in September 2005 and another similar unit will be going into operation very shortly. For a complete dynamic study of the channel and its internals under the influence of high coolant flow, experimental and modeling studies have been carried out. A good correlation has been achieved between the results of experimental and analytical models. The operating life of a typical coolant channel typically ranges from 10 to 15 full-power years. Towards the end of its operating life, its health monitoring becomes an important activity. Vibration diagnosis plays an important role as a tool for life management of coolant. Through the study of dynamic characteristics of the coolant channel under simulated loading condition, an attempt has been made to develop a diagnostics to monitor the health of the coolant channel over its operating life. A study has been also carried out to characterize the fuel vibration under different flow condition.

  9. Flow boiling test of GDP replacement coolants

    Energy Technology Data Exchange (ETDEWEB)

    Park, S.H. [comp.

    1995-08-01

    The tests were part of the CFC replacement program to identify and test alternate coolants to replace CFC-114 being used in the uranium enrichment plants at Paducah and Portsmouth. The coolants tested, C{sub 4}F{sub 10} and C{sub 4}F{sub 8}, were selected based on their compatibility with the uranium hexafluoride process gas and how well the boiling temperature and vapor pressure matched that of CFC-114. However, the heat of vaporization of both coolants is lower than that of CFC-114 requiring larger coolant mass flow than CFC-114 to remove the same amount of heat. The vapor pressure of these coolants is higher than CFC-114 within the cascade operational range, and each coolant can be used as a replacement coolant with some limitation at 3,300 hp operation. The results of the CFC-114/C{sub 4}F{sub 10} mixture tests show boiling heat transfer coefficient degraded to a minimum value with about 25% C{sub 4}F{sub 10} weight mixture in CFC-114 and the degree of degradation is about 20% from that of CFC-114 boiling heat transfer coefficient. This report consists of the final reports from Cudo Technologies, Ltd.

  10. Computing Flows Of Coolants In Turbomachines

    Science.gov (United States)

    Meitner, P. L.

    1994-01-01

    Coolant Passage Flow (CPF) computer code developed to predict accurately coolant flow and heat transfer inside turbomachinery cooling passages (either radial or axial blading). Computes flow in one-inlet/one-outlet passage of any shape. Calculates rate of flow of coolant, temperature, pressure, velocity, and heat-transfer coefficients along passage. Integrates one-dimensional momentum and energy equations along defined flow path, taking into account change in area, addition or subtraction of mass, pumping, friction, and transfer of heat. Written in FORTRAN IV.

  11. New Configurations of Micro Plate-Fin Heat Sink to Reduce Coolant Pumping Power

    DEFF Research Database (Denmark)

    Kolaei, Alireza Rezania; Rosendahl, Lasse

    2012-01-01

    The thermal resistance of heat exchangers has a strong influence on the electric power produced by a thermoelectric generator (TEG). In this work, a real TEG device is applied to three configurations of micro plate-fin heat sink. The distance between certain microchannels is varied to find...... the optimum heat sink configuration. The particular focus of this study is to reduce the coolant mass flow rate by considering the thermal resistances of the heat sinks and, thereby, to reduce the coolant pumping power in the system. The threedimensional governing equations for the fluid flow and the heat...... heat sink configurations reduces the coolant pumping power in the system....

  12. Innovative technologies for soil cleanup

    Energy Technology Data Exchange (ETDEWEB)

    Yow, J.L. Jr.

    1992-09-01

    These notes provide a broad overview of current developments in innovative technologies for soil cleanup. In this context, soil cleanup technologies include site remediation methods that deal primarily with the vadose zone and with relatively shallow, near-surface contamination of soil or rock materials. This discussion attempts to emphasize approaches that may be able to achieve significant improvements in soil cleanup cost or effectiveness. However, since data for quantitative performance and cost comparisons of new cleanup methods are scarce, preliminary comparisons must be based on the scientific approach used by each method and on the sits-specific technical challenges presented by each sold contamination situation. A large number of technical alternatives that are now in research, development, and testing can be categorized by the scientific phenomena that they employ and by the site contamination situations that they treat. After cataloging a representative selection of these technologies, one of the new technologies, Dynamic Underground Stripping, is discussed in more detail to highlight a promising soil cleanup technology that is now being field tested.

  13. Behavior of primary coolant pump shaft seals during station blackout conditions

    Energy Technology Data Exchange (ETDEWEB)

    Hill, R.C.; Rhodes, D.B.

    1986-09-12

    An assessment is made of the ability of typical Reactor Coolant Pump (RCP) Shaft Seals to withstand the conditions predicted for a station blackout (loss of all alternating current power) at a nuclear power station. Several factors are identified that are key to seal stability including inlet fluid conditions, pressure downstream of the seal, and geometrical details of the seal rings. Limits for stable seal operation are determined for various combinations of these factors, and the conclusion is drawn that some RPC seals would be near the threshold of instability during a station blackout. If the threshold were exceeded, significant leakage of coolant from the primary coolant system could be expected.

  14. Parameters important to reactor coolant pump seal stability during station blackout

    Energy Technology Data Exchange (ETDEWEB)

    Hill, R.C.; Rhodes, D.B.

    1986-10-24

    An assessment is made of the ability of typical Reactor Coolant Pump (RCP) Shaft Seals to withstand the conditions predicted for a station blackout (loss of all alternating current power) at a nuclear power station. Several factors are identified that are key to seal stability including inlet fluid conditions, pressure downstream of the seal, and geometrical details of the seal rings. Limits for stable seal operation are determined for various combinations of these factors, and the conclusion is drawn that some RPC seals would be near or over the threshold of instability during a station blackout. If the threshold were exceeded, significant leakage of coolant from the primary coolant system could be expected.

  15. Performance of Helical Coil Heat Recovery Exchanger using Nanofluid as Coolant

    Directory of Open Access Journals (Sweden)

    Navid Bozorgan

    2015-07-01

    Full Text Available Nanofluids are expected to be a promising coolant condidate in chemical processes for heat transfer system size reduction. This paper focuses on reducing the number of turns in a helical coil heat recovery exchanger with a given heat exchange capacity in a biomass heating plant using γ-Al2O3/n-decane nanofluid as coolant. The nanofluid flows through the tubes and the hot n-hexane flows through the shell. The numerical results show that using nanofluid as coolant in a helical coil heat exchanger can reduce the manufacturing cost of the heat exchanger and pumping power by reducing the number of turns of the coil.

  16. Development and demonstration of a mobile reverse osmosis adsorption treatment system for environmental emergency clean-ups

    Energy Technology Data Exchange (ETDEWEB)

    1988-03-01

    A project was undertaken to develop and demonstrate a mobile reverse osmosis/adsorption system for treating water contaminated by organic chemicals. The system has two primary unit operations. The contaminated water is processed by reverse osmosis to produce a clean stream for discharge and a stream for further processing in which the organic contaminants have been concentrated up to 10 times their original concentration. The latter stream is treated in granular adsorbent columns where the contaminants are removed and an effluent suitable for discharge is produced. The contaminated water can be treated on-site and the contaminants can be removed from the site adsorbed on a relatively small amount of carbon. Field tests were conducted at two sites, one contaminated by leachate from a former chemical waste landfill and the other by drainage water from a petroleum and petrochemical transfer station. The 60-day demonstrations showed that reverse osmosis technology can be successfully used for treatment of water contaminated by toxic volatile organics and that granular activated carbon adsorption columns can be successfully used to remove those organics from the concentrate produced by reverse osmosis processing. However, the study also showed that the presence of significant quantities of suspended materials or Fe cause operational problems which limit the success of reverse osmosis processing under these conditions. These problems can be effectively addressed by adding an ultrafiltration pretreatment. 13 refs., 60 figs., 56 tabs.

  17. SUPERFUND CLEANUPS AND INFANT HEALTH

    Science.gov (United States)

    Currie, Janet; Greenstone, Michael; Moretti, Enrico

    2013-01-01

    We are the first to examine the effect of Superfund cleanups on infant health rather than focusing on proximity to a site. We study singleton births to mothers residing within 5km of a Superfund site between 1989–2003 in five large states. Our “difference in differences” approach compares birth outcomes before and after a site clean-up for mothers who live within 2,000 meters of the site and those who live between 2,000– 5,000 meters of a site. We find that proximity to a Superfund site before cleanup is associated with a 20 to 25% increase in the risk of congenital anomalies. PMID:25152535

  18. Thermal transfer structures coupling electronics card(s) to coolant-cooled structure(s)

    Science.gov (United States)

    David, Milnes P; Graybill, David P; Iyengar, Madhusudan K; Kamath, Vinod; Kochuparambil, Bejoy J; Parida, Pritish R; Schmidt, Roger R

    2014-12-16

    Cooling apparatuses and coolant-cooled electronic systems are provided which include thermal transfer structures configured to engage with a spring force one or more electronics cards with docking of the electronics card(s) within a respective socket(s) of the electronic system. A thermal transfer structure of the cooling apparatus includes a thermal spreader having a first thermal conduction surface, and a thermally conductive spring assembly coupled to the conduction surface of the thermal spreader and positioned and configured to reside between and physically couple a first surface of an electronics card to the first surface of the thermal spreader with docking of the electronics card within a socket of the electronic system. The thermal transfer structure is, in one embodiment, metallurgically bonded to a coolant-cooled structure and facilitates transfer of heat from the electronics card to coolant flowing through the coolant-cooled structure.

  19. World Record Earned Value Management System Certification for Cleanup of the East Tennessee Technology Park, Oak Ridge, Tennessee, USA - 13181

    Energy Technology Data Exchange (ETDEWEB)

    Haynes, Ray; Hirschy, Anita [URS - CH2M Oak Ridge, LLC (UCOR), East Tennessee Technology Park D and D and Environmental Remediation Project, Oak Ridge, Tennessee 37830 (United States)

    2013-07-01

    On projects that require Earned Value Management (EVMS) Certification, it is critical to quickly prepare for and then successfully obtain certification. This is especially true for government contracts. Projects that do poorly during the review are subject to financial penalties to their company and they lose creditability with their customer creating problems with the project at the outset. At East Tennessee Technology Park (ETTP), we began preparing for Department of Energy (DOE) certification early during proposal development. Once the contract was awarded, while still in transition phase from the previous contractor to our new company, we immediately began reviewing the project controls systems that were in place on the project and determined if any replacements needed to be made immediately. The ETTP contract required the scheduling software to be upgraded to Primavera P6 and we determined that no other software changes would be done prior to certification. Next, preparation of the Project Controls System Description (PCSD) and associated procedures began using corporate standards as related to the project controls systems. During the transition phase, development was started on the Performance Measurement Baseline which is the resource loaded schedule used to measure our performance on the project and which is critical to good Earned Value Management of the project. Early on, and throughout the baseline review, there was positive feedback from the Department of Energy that the quality of the new baseline was good. Having this superior baseline also contributed to our success in EVMS certification. The combined companies of URS and CH2M Hill had recent experience with certifications at other Department of Energy sites and we were able to capitalize on that knowledge and experience. Generic PCSD and procedures consistent with our co-operations approach to Earned Value Management were available to us and were easily tailorable to the specifics of our contract

  20. Corrosion problems with aqueous coolants, final report

    Energy Technology Data Exchange (ETDEWEB)

    Diegle, R B; Beavers, J A; Clifford, J E

    1980-04-11

    The results of a one year program to characterize corrosion of solar collector alloys in aqueous heat-transfer media are summarized. The program involved a literature review and a laboratory investigation of corrosion in uninhibited solutions. It consisted of three separate tasks, as follows: review of the state-of-the-art of solar collector corrosion processes; study of corrosion in multimetallic systems; and determination of interaction between different waters and chemical antifreeze additives. Task 1 involved a comprehensive review of published literature concerning corrosion under solar collector operating conditions. The reivew also incorporated data from related technologies, specifically, from research performed on automotive cooling systems, cooling towers, and heat exchangers. Task 2 consisted of determining the corrosion behavior of candidate alloys of construction for solar collectors in different types of aqueous coolants containing various concentrations of corrosive ionic species. Task 3 involved measuring the degradation rates of glycol-based heat-transfer media, and also evaluating the effects of degradation on the corrosion behavior of metallic collector materials.

  1. Recycling Facilities - Land Recycling Cleanup Locations

    Data.gov (United States)

    NSGIC GIS Inventory (aka Ramona) — Land Recycling Cleanup Location Land Recycling Cleanup Locations (LRCL) are divided into one or more sub-facilities categorized as media: Air, Contained Release or...

  2. Recycling Facilities - Land Recycling Cleanup Locations

    Data.gov (United States)

    NSGIC Education | GIS Inventory — Land Recycling Cleanup Location Land Recycling Cleanup Locations (LRCL) are divided into one or more sub-facilities categorized as media: Air, Contained Release or...

  3. Power Module Cooling for Future Electric Vehicle Applications: A Coolant Comparison of Oil and PGW

    Science.gov (United States)

    2006-11-01

    POWER MODULE COOLING FOR FUTURE ELECTRIC VEHICLE APPLICATIONS: A COOLANT COMPARISON OF OIL AND PGW T. E. Salem U. S. Naval Academy 105...and efficient power converters are being developed to support the needs of future ground vehicle systems. This progress is being driven by...2006 2. REPORT TYPE N/A 3. DATES COVERED - 4. TITLE AND SUBTITLE Power Module Cooling For Future Electric Vehicle Applications: A Coolant

  4. Steam as coolant and lubricant in turning of metal matrix composites

    Institute of Scientific and Technical Information of China (English)

    Raviraj SHETTY; Raghuvir PAI; Vasanth KAMATH; Shrikanth S.RAO

    2008-01-01

    Green cutting has become focus of attention in ecological and environmental protection.Steam is cheap.pollution-free and eco-friendly,and then is a good and economical coolant and lubricant.Steam generator and steam feeding system were developed to generate and feed steam.Comparative experiments were carried out in cutting AA6061-15 v0l.%SiC(25 μm particle size),with cubic boron nitride(CBN)insert KB-90 grade under the conditions of compressed air,oil water emulsion,steam as coolant and lubricant,and dry cutting,respectively.The experimental results show that,with steam as coolant and lubricant,gradual reduction in the cutting force,friction coefficient,surface roughness and cutting temperature values were observed.Further,there was reduction in built up edge formation.1t is proved that use of water steam as coolant and lubricant is environmentally friendly.

  5. Application of damage function analysis to reactor coolant circuits

    Energy Technology Data Exchange (ETDEWEB)

    MacDonald, D.D. [Center for Electrochemical Science and Technology, Pennsylvania State Univ., University Park, PA (United States)

    2002-07-01

    The application of deterministic models for simulating stress corrosion cracking phenomena in Boiling Water Reactor primary coolant circuits is described. The first generation code, DAMAGE-PREDICTOR, has been used to model the radiolysis of the coolant, to estimate the electrochemical corrosion potential (ECP), and to calculate the crack growth rate (CGR) at fixed state points during reactor operation in about a dozen plants worldwide. This code has been validated in ''double-blind'' comparisons between the calculated and measured hydrogen concentration, oxygen concentration, and ECP in the recirculation system of the Leibstadt BWR in Switzerland, as well as through less formal comparisons with data from other plants. Second generation codes have now been developed, including REMAIN for simulating BWRs with internal coolant pumps and the ALERT series for modeling reactors with external pumps. One of this series, ALERT, yields the integrated damage function (IDF), which is the crack length versus time, on a component-by-component basis for a specified future operating scenario. This code therefore allows one to explore proposed future operating protocols, with the objective of identifying those that are most cost-effective and which minimizes the risk of failure of components in the coolant circuit by stress corrosion cracking. The application of this code is illustrated by exploring the benefits of partial hydrogen water chemistry (HWC) for an actual reactor, in which hydrogen is added to the feedwater over only limited periods during operation. The simulations show that the benefits, in terms of reduction in the IDFs for various components, are sensitive to when HWC was initiated in the plant life and to the length of time over which it is applied. (author)

  6. Technology of high-temperature organic coolant

    Energy Technology Data Exchange (ETDEWEB)

    Vorobei, M.P.; Makin, R.S.; Kuprienko, V.A. [and others

    1993-12-31

    A wide range of studies were carried out in RIAR on the problems connected with the use of high-temperature organic coolant at nuclear power plants. The work performed and successful experience gained in persistent operation of the ARBUS reactor confirmed the inherent safety characteristics, high operational reliability, as well as improved safety of stations with similar reactors. A large scope of studies were carried out at the ARBUS pilot reactor and loop with the organic coolant of the MIR reactor and a wide range of problems were solved. The studies are described.

  7. Comparison between MAAP and ECART predictions of radionuclide transport throughout a French standard PWR reactor coolant system; Transport des radionucleides dans le circuit primaire d`un REP: comparaison des codes MAAP et ECART

    Energy Technology Data Exchange (ETDEWEB)

    Hervouet, C.; Ranval, W. [Electricite de France (EDF), 92 - Clamart (France); Parozzi, F.; Eusebi, M. [Ente Nazionale per l`Energia Elettrica, Rome (Italy)

    1996-04-01

    In the framework of a collaboration agreement between EDF and ENEL, the MAAP (Modular Accident Analysis Program) and ECART (ENEL Code for Analysis of radionuclide Transport) predictions about the fission product retention inside the reactor cooling system of a French PWR 1300 MW during a small Loss of Coolant Accident were compared. The volatile fission products CsI, CsOH, TeO{sub 2} and the structural materials, all of them released early by the core, are more retained in MAAP than in ECART. On the other hand, the non-volatile fission products, released later, are more retained in ECART than in MAAP, because MAAP does not take into account diffusion-phoresis: in fact, this deposition phenomenon is very significant when the molten core vaporizes the water of the vessel lower plenum. Centrifugal deposition in bends, that can be modeled only with ECART, slightly increases the whole retention in the circuit if it is accounted for. (authors). 18 refs., figs., tabs.

  8. NGNP Reactor Coolant Chemistry Control Study

    Energy Technology Data Exchange (ETDEWEB)

    Brian Castle

    2010-11-01

    The main focus of this paper is to identify the most desirable ranges of impurity levels in the primary coolant to optimize component life in the primary circuit of the Next Generation Nuclear Plant (NGNP), which will either be a prismatic block or pebble bed reactor.

  9. Army Environmental Cleanup Strategic Plan

    Science.gov (United States)

    2009-05-01

    New success indicators are all definable, measurable, and achievable MAY 200918 of 29 Emerging Issues  Emerging contaminants  MMRP progress  NCP...programmatic expectations  NDNODS  Operational range program  Vapor Intrusion MAY 200919 of 29 Emerging Contaminants – Hexavalent Chromium...regulatory standards  Several emerging contaminants have been assessed and judged to have a significant potential impact to Army cleanup programs

  10. Oil Spill Cleanup

    Science.gov (United States)

    1994-01-01

    Petroleum Remediation Product (PRP) is a new way of cleaning up oil spills. It consists of thousands of microcapsules, tiny balls of beeswax with hollow centers, containing live microorganisms and nutrients to sustain them. As oil flows through the microcapsule's shell, it is consumed and digested by the microorganisms. Pressure buildup causes the PRP to explode and the enzymes, carbon dioxide and water are released into the BioBoom used in conjunction with PRP, preventing contaminated water from spreading. The system incorporates technology originally developed at the Jet Propulsion Laboratory and Marshall Space Flight Center.

  11. APT Blanket System Loss-of-Coolant Accident (LOCA) Based on Initial Conceptual Design - Case 4: External Pressurizer Surge Line Break Near Inlet Header

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

    This report is one of a series of reports documenting accident scenario simulations for the Accelerator Production of Tritium (APT) blanket heat removal systems. The simulations were performed in support of the Preliminary Safety Analysis Report (PSAR) for the APT.

  12. APT Blanket System Loss-of-Coolant Accident Based on Initial Conceptual Design - Case 5: External RHR Break Near Inlet Header

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

    This report is one of a series of reports that document normal operation and accident simulations for the Accelerator Production of Tritium (APT) blanket heat removal system. These simulations were performed for the Preliminary Safety Analysis Report.

  13. APT Blanket System Loss-of-Coolant Accident (LOCA) Analysis Based on Initial Conceptual Design - Case 3: External HR Break at Pump Outlet without Pump Trip

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

    This report is one of a series of reports that document normal operation and accident simulations for the Accelerator Production of Tritium (APT) blanket heat removal (HR) system. These simulations were performed for the Preliminary Safety Analysis Report.

  14. APT Blanket System Loss-of-Coolant Analysis Based on Initial Conceptual Design - Case 2: External HR Break HR Break at Pump Outlet with Pump Trip

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

    This report is one of a series of reports that document normal operation and accident simulations for the Accelerator Production of Tritium (APT) blanket heat removal system. These simulations were performed for the Preliminary Safety Analysis Report.

  15. APT Blanket System Loss-of-Coolant Accident (LOCA) Based on Initial Conceptual Design - Case 1: External HR Break Near Inlet Header

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

    The APT blanket system has about 57 MW of thermal energy deposited within the blanket region under normal operating conditions from the release of neutrons and the interaction of the High energy particles with the blanket materials. This corresponds to about 48 percent of total thermal energy deposited in the APT target/blanket system. The deposited thermal energy under normal operation conditions is an important input parameter used in the thermal-hydraulic design and accident analysis.

  16. Research Progress of Decontamination Process and its Corrosion Effect on Primary Coolant Systems of Nuclear Reactor%反应堆一回路系统去污工艺及其对结构材料腐蚀的影响

    Institute of Scientific and Technical Information of China (English)

    谭昭怡; 李烨; 孙宇; 汪小琳; 张东

    2012-01-01

    采用化学去污工艺可降低反应堆一回路冷却系统周围辐射场.总结了近年来反应堆一回路冷却系统去污工艺和去污试剂对结构材料的腐蚀影响的研究成果,并建议后续研究方向.%Radiation field intensity in the primary coolant system of water-cooled reactors could be reduced by chemical decontamination process.Thus,the recent research progresses of the decontamination process and its corrosion effect on structural materials of the primary coolant systems were summarized in this paper.

  17. A tritium vessel cleanup experiment in TFTR

    Energy Technology Data Exchange (ETDEWEB)

    Caorlin, M.; Kamperschroer, J.; Owens, D.K.; Voorhees, D.; Mueller, D.; Ramsey, A.T.; La Marche, P.H. [Princeton Univ., NJ (United States). Plasma Physics Lab.; Barnes, C.W. [Los Alamos National Lab., NM (United States); Loughlin, M.J. [JET Joint Undertaking, Abingdon (United Kingdom)

    1995-03-01

    A simple tritium cleanup experiment was carried out in TFTR following the initial high power deuterium-tritium discharges in December 1993. A series of 34 ohmic and deuterium neutral beam fueled shots was used to study the removal of tritium implanted into the wall and limiters. A very large plasma was created in each discharge to ``scrub`` an area as large as possible. Beam-fueled shots at 2.5 to 7.5 MW of injected power were used to monitor tritium concentration levels in the plasma by detection of DT-neutrons. The neutron signal decreased by a factor of 4 during the experiment, remaining well above the expected T-burnup level. The amount of tritium recovered at the end of the cleanup was about 8% of the amount previously injected with high power DT discharges. The experience gained suggests that measurements of tritium inventory in the torus are very difficult to execute and require dedicated systems with overall accuracy of 1%.

  18. Accelerated cleanup risk reduction

    Energy Technology Data Exchange (ETDEWEB)

    Knapp, R.B.; Aines, R.M.; Blake, R.G.; Copeland, A.B.; Newmark, R.L.; Tompson, A.F.B.

    1998-02-01

    period in which the well was `capped`. Our results show the formation of an inclined gas phase during injection and a fast collapse of the steam zone within an hour of terminating steam injection. The majority of destruction occurs during the collapse phase, when contaminant laden water is drawn back towards the well. Little to no noncondensible gasses are created in this process, removing any possibility of sparging processes interfering with contaminant destruction. Our models suggest that the thermal region should be as hot and as large as possible. To have HPO accepted, we need to demonstrate the in situ destruction of contaminants. This requires the ability to inexpensively sample at depth and under high temperatures. We proved the ability to implies monitoring points at depths exceeding 150 feet in highly heterogeneous soils by use of cone penetrometry. In addition, an extractive system has been developed for sampling fluids and measuring their chemistry under the range of extreme conditions expected. We conducted a collaborative field test of HPO at a Superfund site in southern California where the contaminant is mainly creosote and pentachlorophenol. Field results confirm the destruction of contaminants by HPO, validate our field design from simulations, demonstrate that accurate field measurements of the critical fluid parameters can be obtained using existing monitoring wells (and minimal capital cost) and yield reliable cost estimates for future commercial application. We also tested the in situ microbial filter technology as a means to intercept and destroy the accelerated flow of contaminants caused by the injection of steam. A series of laboratory and field tests revealed that the selected bacterial species effectively degrades trichloroethene in LLNL Groundwater and under LLNL site conditions. In addition, it was demonstrated that the bacteria effectively attach to the LLNL subsurface media. An in-well treatability study indicated that the bacteria

  19. Accelerated cleanup Initiatives Putting the Acceleration Plans into Action

    Energy Technology Data Exchange (ETDEWEB)

    TYREE, G.T.

    2003-01-01

    This paper describes project successes during the last year and presents strategies for accomplishing work required to accelerate waste retrieval, treatment and closure of 177 large underground waste tanks at the Hanford Site. The tanks contain approximately 53 million gallons of liquid, sludge, and solid waste resulting from decades of national defense production. The Hanford Site is a 560 square-mile area in southeastern Washington State. One of the nation's largest rivers, the Columbia River, flows through the site and within seven miles of the waste tanks. The US. Department of Energy (DOE) Office of River Protection and CH2M HILL Hanford Group, Inc. (CH2M HILL) drew upon the recommendations in the DOE's Top-To-Bottom Review and the ideas that emerged from the Cleanup Challenges and Constraints Team (C3T) when creating new initiatives last fall in accelerated tank cleanup. The initiatives reflect discussions and planning during the last year by the DOE, regulatory,agencies, Hanford stakeholders, and CH2M HILL on how to accelerate tank cleanup and closure. The initiatives focus on near-term risk reduction, deployment of proven cleanup technologies, and completing the feed delivery and waste storage systems needed to support Hanford's Waste Treatment Plant. Working with the Office of River Protection, CH2M HILL is changing the way it does business to align with the new focus on accelerated tank cleanup initiatives. A key concept of this new approach is to deploy simple, proven technologies whenever possible to accomplish program goals. Finding existing technologies and evaluating whether they can be applied to or adapted to Hanford tank cleanup provide the best chance for success in achieving treatment of all of Hanford's tank waste by 2028.

  20. Recovery studies for plutonium machining oil coolant

    Energy Technology Data Exchange (ETDEWEB)

    Navratil, J. D.; Baldwin, C. E.

    1977-04-27

    Lathe coolant oil, contaminated with plutonium and having a carbon tetrachloride diluent, is generated in plutonium machining areas at Rocky Flats. A research program was initiated to determine the nature of plutonium in this mixture of oil and carbon tetrachloride. Appropriate methods then could be developed to remove the plutonium and to recycle the oil and carbon tetrachloride. Studies showed that the mixtures of spent oil and carbon tetrachloride contained particulate plutonium and plutonium species that are soluble in water or in oil and carbon tetrachloride. The particulate plutonium was removed by filtration; the nonfilterable plutonium was removed by adsorption on various materials. Laboratory-scale tests indicated the lathe-coolant oil mixture could be separated by distilling the carbon tetrachloride to yield recyclable products.

  1. An approach for IC engine coolant energy recovery based on low-temperature organic Rankine cycle

    Institute of Scientific and Technical Information of China (English)

    付建勤; 刘敬平; 徐政欣; 邓帮林; 刘琦

    2015-01-01

    To promote the fuel utilization efficiency of IC engine, an approach was proposed for IC engine coolant energy recovery based on low-temperature organic Rankine cycle (ORC). The ORC system uses IC engine coolant as heat source, and it is coupled to the IC engine cooling system. After various kinds of organic working media were compared, R124 was selected as the ORC working medium. According to IC engine operating conditions and coolant energy characteristics, the major parameters of ORC system were preliminary designed. Then, the effects of various parameters on cycle performance and recovery potential of coolant energy were analyzed via cycle process calculation. The results indicate that cycle efficiency is mainly influenced by the working pressure of ORC, while the maximum working pressure is limited by IC engine coolant temperature. At the same working pressure, cycle efficiency is hardly affected by both the mass flow rate and temperature of working medium. When the bottom cycle working pressure arrives at the maximum allowable value of 1.6 MPa, the fuel utilization efficiency of IC engine could be improved by 12.1%. All these demonstrate that this low-temperature ORC is a useful energy-saving technology for IC engine.

  2. Enhancing resistance to burnout via coolant chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Tu, J. P.; Dinh, T. N.; Theofanous, T. G. [Univ. of California, Santa Barbara (United States)

    2003-07-01

    Boiling Crisis (BC) on horizontal, upwards-facing copper and steel surfaces under the influence of various coolant chemistries relevant to reactor containment waters is considered. In addition to Boric Acid (BA) and TriSodium Phosphate (TSP), pure De-Ionized Water (DIW) and Tap Water (TW) are included in experiments carried out in the BETA facility. The results are related to a companion paper on the large scale ULPU facility.

  3. Development of a risk-based approach to Hanford Site cleanup

    Energy Technology Data Exchange (ETDEWEB)

    Hesser, W.A.; Daling, P.M. [Pacific Northwest Lab., Richland, WA (United States); Baynes, P.A. [Westinghouse Hanford Co., Richland, WA (United States)] [and others

    1995-06-01

    In response to a request from Mr. Thomas Grumbly, Assistant Secretary of Energy for Environmental Management, the Hanford Site contractors developed a conceptual set of risk-based cleanup strategies that (1) protect the public, workers, and environment from unacceptable risks; (2) are executable technically; and (3) fit within an expected annual funding profile of 1.05 billion dollars. These strategies were developed because (1) the US Department of Energy and Hanford Site budgets are being reduced, (2) stakeholders are dissatisfied with the perceived rate of cleanup, (3) the US Congress and the US Department of Energy are increasingly focusing on risk and riskreduction activities, (4) the present strategy is not integrated across the Site and is inconsistent in its treatment of similar hazards, (5) the present cleanup strategy is not cost-effective from a risk-reduction or future land use perspective, and (6) the milestones and activities in the Tri-Party Agreement cannot be achieved with an anticipated funding of 1.05 billion dollars annually. The risk-based strategies described herein were developed through a systems analysis approach that (1) analyzed the cleanup mission; (2) identified cleanup objectives, including risk reduction, land use, and mortgage reduction; (3) analyzed the existing baseline cleanup strategy from a cost and risk perspective; (4) developed alternatives for accomplishing the cleanup mission; (5) compared those alternatives against cleanup objectives; and (6) produced conclusions and recommendations regarding the current strategy and potential risk-based strategies.

  4. Corrosion of magnesium alloys in commercial engine coolants

    Energy Technology Data Exchange (ETDEWEB)

    Song, G.; StJohn, D.H. [CRC for Cast Metals Manufacturing (CAST), Division of Materials, School of Engineering, The University of Queensland, Brisbane, QLD 4072 (Australia)

    2005-01-01

    A number of magnesium alloys show promise as engine block materials. However, a critical issue for the automotive industry is corrosion of the engine block by the coolant and this could limit the use of magnesium engine blocks. This work assesses the corrosion performance of conventional magnesium alloy AZ91D and a recently developed engine block magnesium alloy AM-SC1 in several commercial coolants. Immersion testing, hydrogen evolution measurement, galvanic current monitoring and the standard ASTM D1384 test were employed to reveal the corrosion performance of the magnesium alloys subjected to the coolants. The results show that the tested commercial coolants are corrosive to the magnesium alloys in terms of general and galvanic corrosion. The two magnesium alloys exhibited slightly different corrosion resistance to the coolants with AZ91D being more corrosion resistant than AM-SC1. The corrosivity varied from coolant to coolant. Generally speaking, an organic-acid based long life coolant was less corrosive to the magnesium alloys than a traditional coolant. Among the studied commercial coolants, Toyota long life coolant appeared to be the most promising one. In addition, it was found that potassium fluoride effectively inhibited corrosion of the magnesium alloys in the studied commercial coolants. Both general and galvanic corrosion rates were significantly decreased by addition of KF, and there were no evident side effects on the other engine block materials, such as copper, solder, brass, steel and aluminium alloys, in terms of their corrosion performance. The ASTM D 1384 test further confirmed these results and suggested that Toyota long life coolant with 1%wt KF addition is a promising coolant for magnesium engine blocks. (Abstract Copyright [2005], Wiley Periodicals, Inc.)

  5. Flatness Control Using Roll Coolant Based on Predicted Flatness Variation in Cold Rolling Mills

    Science.gov (United States)

    Dohmae, Yukihiro; Okamura, Yoshihide

    Flatness control for cold rolling mills is one of the important technologies for improving of product quality and productivity. In particular, poor flatness leads to strip tearing in the extreme case and, moreover, it significantly reduces productivity. Therefore, various flatness control system has been developed. The main actuators for flatness control are classified into two types; one is mechanical equipment such as roll bender, the other is roll coolant, which controls thermal expansion of roll. Flatness variation such as center buckle or edge wave is mainly controlled by mechanical actuator which has high response characteristics. On another front, flatness variation of local zone can be controlled by roll coolant although one's response is lower than the response of mechanical actuator. For accomplishing good flatness accuracy in cold rolling mills, it is important to improve the performance of coolant control moreover. In this paper, a new coolant control method based on flatness variation model is described. In proposed method, the state of coolant spray on or off is selected to minimize the flatness deviation by using predicted flatness variation. The effectiveness of developed system has been demonstrated by application in actual plant.

  6. Experience in operation of the experimental atomic power plant ''ARBUS'' with the high-boiling organic coolant-moderator ditolylmethane

    Energy Technology Data Exchange (ETDEWEB)

    Tzikanov, V.A.; Aleksenko, Yu.N.; Tetyukov, V.D.; Kuprienko, V.A.; Kobzar, I.G.; Khramchenkov, V.A.; Mexcheryakov, M.P.; Zinoviev, V.I.

    1978-04-01

    Radiolytic damage to the ditolylmethane organic coolant-moderator of the ARBUS reactor was removed by vacuum distillation. The majority of the degraded ditolylmethane formed gaseous and high-boiling materials, which were easily removed by the vacuum distillation. Unsaturated hydrocarbons and low-boiling residues were a minor contribution to the impurities produced by radiolysis in the primary coolant loop. Radioactivity in the primary coolant loop was found to be caused primarily from corrosion products of the system, /sup 16/N from dissolved oxygen, and impurities in the coolant-moderator. These also were significantly reduced in the vacuum distillation process.

  7. Development of an administrative record system and information repository system to support environmental cleanup at the Hanford Site in Richland, Washington

    Energy Technology Data Exchange (ETDEWEB)

    Sprouse, B.S.

    1991-09-01

    The purpose of this paper is to describe the development of an administrative record (AR) file system for the Hanford Site in Richland, Washington. This paper will focus on the background of the AR system and its implementation at the Hanford Site; the types of documents to be included in an AR file; the management system used to develop and implement the AR system; the unique characteristics of the AR system and the role of the information repositories. The objective of this session is to present the methodology used in developing an AR and information repository system so that common hurdles from various sites can be addressed and resolved similarly. Therefore unnecessary effort does not have to be put forth to resolve the same issues over and over again. The concepts described can be applied to all federal facility sites with a minimum amount of modification and revision. 9 refs.

  8. HANFORD SITE RIVER CORRIDOR CLEANUP

    Energy Technology Data Exchange (ETDEWEB)

    BAZZELL, K.D.

    2006-02-01

    In 2005, the US Department of Energy (DOE) launched the third generation of closure contracts, including the River Corridor Closure (RCC) Contract at Hanford. Over the past decade, significant progress has been made on cleaning up the river shore that bordes Hanford. However, the most important cleanup challenges lie ahead. In March 2005, DOE awarded the Hanford River Corridor Closure Contract to Washington Closure Hanford (WCH), a limited liability company owned by Washington Group International, Bechtel National and CH2M HILL. It is a single-purpose company whose goal is to safely and efficiently accelerate cleanup in the 544 km{sup 2} Hanford river corridor and reduce or eliminate future obligations to DOE for maintaining long-term stewardship over the site. The RCC Contract is a cost-plus-incentive-fee closure contract, which incentivizes the contractor to reduce cost and accelerate the schedule. At $1.9 billion and seven years, WCH has accelerated cleaning up Hanford's river corridor significantly compared to the $3.2 billion and 10 years originally estimated by the US Army Corps of Engineers. Predictable funding is one of the key features of the new contract, with funding set by contract at $183 million in fiscal year (FY) 2006 and peaking at $387 million in FY2012. Another feature of the contract allows for Washington Closure to perform up to 40% of the value of the contract and subcontract the balance. One of the major challenges in the next few years will be to identify and qualify sufficient subcontractors to meet the goal.

  9. Cryogenic-coolant He-4-superconductor interaction

    Science.gov (United States)

    Caspi, S.; Lee, J. Y.; Kim, Y. I.; Allen, R. J.; Frederking, T. H. K.

    1978-01-01

    The thermodynamic and thermal interaction between a type 2 composite alloy and cryo-coolant He4 was studied with emphasis on post quench phenomena of formvar coated conductors. The latter were investigated using a heater simulation technique. Overall heat transfer coefficients were evaluated for the quench onset point. Heat flux densities were determined for phenomena of thermal switching between a peak and a recovery value. The study covered near saturated liquid, pressurized He4, both above and below the lambda transition, and above and below the thermodynamic critical pressure. In addition, friction coefficients for relative motion between formvar insulated conductors were determined.

  10. Station blackout with reactor coolant pump seal leakage

    Energy Technology Data Exchange (ETDEWEB)

    Evinay, A. (Southern California Edison, Irvine, CA (United States))

    1993-01-01

    The U.S. Nuclear Regulatory Commission (NRC) amended its regulations in 10CFR50 with the addition of a new section, 50.63, [open quotes]Loss of All Alternating Current Power.[close quotes] The objective of these requirements is to ensure that all nuclear plants have the capability to withstand a station blackout (SBO) and maintain adequate reactor core cooling and containment integrity for a specified period of time. The NRC also issued Regulatory Guide (RG) 1.155, [open quotes]Station Blackout,[close quotes] to provide guidance for meeting the requirements of 10CFR50.63. Concurrent with RG-1.155, the Nuclear Utility Management and Resources Council (NUMARC) has developed NUMARC 87-00 to address SBO-coping duration and capabilities at light water reactors. Licensees are required to submit a topical report based on NUMARC 87-00 guidelines, to demonstrate compliance with the SBO rule. One of the key compliance criteria is the ability of the plant to maintain adequate reactor coolant system (RCS) inventory to ensure core cooling for the required coping duration, assuming a leak rate of 25 gal/min per reactor coolant pump (RCP) seal in addition to technical specification (TS) leak rate.

  11. Coolant and ambient temperature control for chillerless liquid cooled data centers

    Energy Technology Data Exchange (ETDEWEB)

    Chainer, Timothy J.; David, Milnes P.; Iyengar, Madhusudan K.; Parida, Pritish R.; Simons, Robert E.

    2017-08-29

    Cooling control methods and systems include measuring a temperature of air provided to one or more nodes by an air-to-liquid heat exchanger; measuring a temperature of at least one component of the one or more nodes and finding a maximum component temperature across all such nodes; comparing the maximum component temperature to a first and second component threshold and comparing the air temperature to a first and second air threshold; and controlling a proportion of coolant flow and a coolant flow rate to the air-to-liquid heat exchanger and the one or more nodes based on the comparisons.

  12. Experimental investigation of thermoelectric power generation versus coolant pumping power in a microchannel heat sink

    DEFF Research Database (Denmark)

    Kolaei, Alireza Rezania; Rosendahl, Lasse; Andreasen, Søren Juhl

    2012-01-01

    The coolant heat sinks in thermoelectric generators (TEG) play an important role in order to power generation in the energy systems. This paper explores the effective pumping power required for the TEGs cooling at five temperature difference of the hot and cold sides of the TEG. In addition......, the temperature distribution and the pressure drop in sample microchannels are considered at four sample coolant flow rates. The heat sink contains twenty plate-fin microchannels with hydraulic diameter equal to 0.93 mm. The experimental results show that there is a unique flow rate that gives maximum net...

  13. Deposition of hematite particles on alumina seal faceplates of nuclear reactor coolant pumps: Laboratory experiments and industrial feedback

    OpenAIRE

    Lefèvre Grégory; Živković Ljiljana S.; Jaubertie Anne

    2012-01-01

    In the primary circuit of pressurized water reactors (PWR), the dynamic sealing system in reactor coolant pumps is ensured by mechanical seals whose ceramic parts are in contact with the cooling solution. During the stretch-out phase in reactor operation, characterized by low boric acid concentration, the leak-off flow has been observed to abnormally evolve in industrial plants. The deposition of hematite particles, originating from corrosion, on alumina seals of coolant pumps is suspec...

  14. Always at the correct temperature. Thermal management with electric coolant pump; Immer richtig temperiert. Thermomanagement mit elektrischer Kuehlmittelpumpe

    Energy Technology Data Exchange (ETDEWEB)

    Genster, A.; Stephan, W. [Pierburg GmbH, Neuss (Germany)

    2004-11-01

    Through the use of the electric coolant pump it has become possible for the first time to attain a cooling performance which is adapted precisely to the engine load and which is independent of engine speed. For cooling the new BMW six cylinder in-line Otto engine with an engine power rating of 190 kW, the electric coolant pump by Pierburg requires only 200 W of electrical power from the onboard electrical system. (orig.)

  15. Modeling of melt-coolant mixing by bottom injection

    Energy Technology Data Exchange (ETDEWEB)

    Kazachkov, I.V.; Paladino, D.; Sehgal, B.R. [Royal Inst. of Tech., Div. of Nuclear Power Safety, Stockholm (Sweden)

    2001-07-01

    In this paper, the flow characteristics during the coolant injection, with submerged nozzles, at the bottom of a molten pool are studied. The flow pattern developed by the rising coolant is considered for the case of complete coolant vaporization, and the pool-coolant phase distributions are assessed by a modeling approach delivered from literature for a heterogeneous turbulent jet. To calculate the basic characteristics of such flow, integral relationships are proposed for the two-phase boundary layer. The results of numerical computations and approximate solution are compared with the experimental data obtained in the low temperature experiments, conducted in the DECOBI (debris coolability by bottom injection) facility. (authors)

  16. Efficiency of water coolant for DEMO divertor

    Energy Technology Data Exchange (ETDEWEB)

    Fetzer, Renate, E-mail: renate.fetzer@kit.edu; Igitkhanov, Yuri; Bazylev, Boris

    2015-10-15

    Up to now, water-cooled divertor concepts have been developed for limited incident fluxes without taking into account transient power loadings. In this paper we analyzed the efficiency of water as a coolant for the particular PFC tungsten monoblock shield with a cooling tube made from Cu alloy (Cu OFHC) as a laminate adjacent to W and a low activation martensitic steel (Eurofer) as inner tube contacting the coolant. Thermal analysis is carried out by using the code MEMOS, which simulates W armour damage under the repetitive ELM heat loads. We consider cooling conditions which allow one to keep relatively high material temperatures (in the range 300–600 °C) thus minimizing Eurofer embrittlement under neutron irradiation. Expected DEMO I and DEMO II heat loads including type I ELMs are found to cause melting of the W surface during unmitigated ELMs. By mitigation of the ELMs melting of W is avoided. DEMO I operation under these conditions is save for cooling at water pressure 15.5 MPa and temperature 325 °C, while for DEMO II with mitigated ELMs the critical heat flux is exceeded and safe operation is not provided.

  17. Fusion Blanket Coolant Section Criteria, Methodology, and Results

    Energy Technology Data Exchange (ETDEWEB)

    DeMuth, J. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Meier, W. R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Jolodosky, A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Frantoni, M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Reyes, S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-10-02

    The focus of this LDRD was to explore potential Li alloys that would meet the tritium breeding and blanket cooling requirements but with reduced chemical reactivity, while maintaining the other attractive features of pure Li breeder/coolant. In other fusion approaches (magnetic fusion energy or MFE), 17Li- 83Pb alloy is used leveraging Pb’s ability to maintain high TBR while lowering the levels of lithium in the system. Unfortunately this alloy has a number of potential draw-backs. Due to the high Pb content, this alloy suffers from very high average density, low tritium solubility, low system energy, and produces undesirable activation products in particular polonium. The criteria considered in the selection of a tritium breeding alloy are described in the following section.

  18. Mathematical Model-Based Temperature Preparation of Liquid-Propellant Components Cooled by Liquid Nitrogen in the Heat Exchanger with a Coolant

    Directory of Open Access Journals (Sweden)

    S. K. Pavlov

    2014-01-01

    Full Text Available Before fuelling the tanks of missiles, boosters, and spacecraft with liquid-propellant components (LPC their temperature preparation is needed. The missile-system ground equipment performs this operation during prelaunch processing of space-purpose missiles (SPM. Usually, the fuel cooling is necessary to increase its density and provide heat compensation during prelaunch operation of SPM. The fuel temperature control systems (FTCS using different principles of operation and types of coolants are applied for fuel cooling.To determine parameters of LPC cooling process through the fuel heat exchange in the heat exchanger with coolant, which is cooled by liquid nitrogen upon contact heat exchange in the coolant reservoir, a mathematical model of this process and a design technique are necessary. Both allow us to determine design parameters of the cooling system and the required liquid nitrogen reserve to cool LPC to the appropriate temperature.The article presents an overview of foreign and domestic publications on cooling processes research and implementation using cryogenic products such as liquid nitrogen. The article draws a conclusion that it is necessary to determine the parameters of LPC cooling process through the fuel heat exchange in the heat exchanger with coolant, which is liquid nitrogen-cooled upon contact heat exchange in the coolant reservoir allowing to define rational propellant cooling conditions to the specified temperature.The mathematical model describes the set task on the assumption that a heat exchange between the LPC and the coolant in the heat exchanger and with the environment through the walls of tanks and pipelines of circulation loops is quasi-stationary.The obtained curves allow us to calculate temperature changes of LPC and coolant, cooling time and liquid nitrogen consumption, depending on the process parameters such as a flow rate of liquid nitrogen, initial coolant temperature, pump characteristics, thermal

  19. Increased leukemia risk in Chernobyl cleanup workers

    Science.gov (United States)

    A new study found a significantly elevated risk for chronic lymphocytic leukemia among workers who were engaged in recovery and clean-up activities following the Chernobyl power plant accident in 1986.

  20. Streamlining Site Cleanup in New York City

    Science.gov (United States)

    This joint effort, supported by the New York State Department of Environmental Conservation (NYS DEC), advances the environmental cleanup goals of PlaNYC 2030, the city's comprehensive sustainability plan.

  1. A Citizen's Guide to Drycleaner Cleanup

    Science.gov (United States)

    The State Coalition for Remediation of Drycleaners (SCRD) has prepared an easy-to-read guide explaining the drycleaner cleanup process and describing the technologies that are most commonly used to clean up contaminated drycleaner sites.

  2. Bioavailability: implications for science/cleanup policy

    Energy Technology Data Exchange (ETDEWEB)

    Denit, Jeffery; Planicka, J. Gregory

    1998-12-01

    This paper examines the role of bioavailability in risk assessment and cleanup decisions. Bioavailability refers to how chemicals ''behave'' and their ''availability'' to interact with living organisms. Bioavailability has significant implications for exposure risks, cleanup goals, and site costs. Risk to human health and the environment is directly tied to the bioavailability of the chemicals of concern.

  3. Research on physical and chemical parameters of coolant in Light-Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Reis, Isabela C.; Mesquita, Amir Z., E-mail: icr@cdtn.br, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEM-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    The coolant radiochemical monitoring of light-water reactors, both power reactor as research reactors is one most important tasks of the system safe operation. The last years have increased the interest in the coolant chemical studying to optimize the process, to minimize the corrosion, to ensure the primary system materials integrity, and to reduce the workers exposure radiation. This paper has the objective to present the development project in Nuclear Technology Development Center (CDTN), which aims to simulate the primary water physical-chemical parameters of light-water-reactors (LWR). Among these parameters may be cited: the temperature, the pressure, the pH, the electric conductivity, and the boron concentration. It is also being studied the adverse effects that these parameters can result in the reactor integrity. The project also aims the mounting of a system to control and monitoring of temperature, electric conductivity, and pH of water in the Installation of Test in Accident Conditions (ITCA), located in the Thermal-Hydraulic Laboratory at CDTN. This facility was widely used in the years 80/90 for commissioning of several components that were installed in Angra 2 containment. In the test, the coolant must reproduce the physical and chemical conditions of the primary. It is therefore fundamental knowledge of the main control parameters of the primary cooling water from PWR reactors. Therefore, this work is contributing, with the knowledge and the reproduction with larger faithfulness of the reactors coolant in the experimental circuits. (author)

  4. Coolant rate distribution in horizontal steam generator under natural circulation

    Energy Technology Data Exchange (ETDEWEB)

    Blagovechtchenski, A.; Leontieva, V.; Mitrioukhin, A. [St. Petersburg State Technical Univ. (Russian Federation)

    1997-12-31

    In the presentation the major factors determining the conditions of NCC (Natural Coolant Circulation) in the primary circuit and in particular conditions of coolant rate distribution on the horizontal tubes of PGV-1000 in NPP with VVER-1000 under NCC are considered. 5 refs.

  5. Coolant mixing in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hoehne, T.; Grunwald, G.

    1998-10-01

    The behavior of PWRs during cold water or boron dilution transients is strongly influenced by the distribution of coolant temperature and boron concentration at the core inlet. This distribution is the needed input to 3-dimensional neutron kinetics to calculate the power distribution in the core. It mainly depends on how the plugs of cold or unborated water formed in a single loop are mixed in the downcomer and in the lower plenum. To simulate such mixture phenomena requires the application of 3-dimensional CFD (computational fluid dynamics) codes. The results of the simulation have to be validated against mixture experiments at scaled facilities. Therefore, in the framework of a research project funded by BMBF, the institute creates a 1:5 mixture facility representing first the geometry of a German pressurized water reactor and later the European Pressurized Water Reactor (EPR) geometry. The calculations are based on the CFD Code CFX-4. (orig.)

  6. Power module assemblies with staggered coolant channels

    Science.gov (United States)

    Herron, Nicholas Hayden; Mann, Brooks S; Korich, Mark D

    2013-07-16

    A manifold is provided for supporting a power module assembly with a plurality of power modules. The manifold includes a first manifold section. The first face of the first manifold section is configured to receive the first power module, and the second face of the first manifold section defines a first cavity with a first baseplate thermally coupled to the first power module. The first face of the second manifold section is configured to receive the second power module, and the second face of the second manifold section defines a second cavity with a second baseplate thermally coupled to the second power module. The second face of the first manifold section and the second face of the second manifold section are coupled together such that the first cavity and the second cavity form a coolant channel. The first cavity is at least partially staggered with respect to second cavity.

  7. SIMMER-III applications to fuel-coolant interactions

    Energy Technology Data Exchange (ETDEWEB)

    Morita, K.; Kondo, Sa.; Tobita, Y.; Brear, D.J. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-01-01

    The main purpose of the SIMMER-III code is to provide a numerical simulation of complex multiphase, multicomponent flow problems essential to investigate core disruptive accidents in liquid-metal fast reactors (LMFRs). However, the code is designed to be sufficiently flexible to be applied to a variety of multiphase flows, in addition to LMFR safety issues. In the present study, some typical experiments relating to fuel-coolant interactions (FCIs) have been analyzed by SIMMER-III to demonstrate that the code is applicable to such complex and highly transient multiphase flow situations. It is shown that SIMMER-III can reproduce the premixing phase both in water and sodium systems as well as the propagation of steam explosion. It is thus demonstrated the code is basically capable of simulating integral multiphase thermal-hydraulic problems included in FCI experiments. (author)

  8. Cleanup of a jet fuel spill

    Science.gov (United States)

    Fesko, Steve

    1996-11-01

    Eaton operates a corporate aircraft hanger facility in Battle Creek, Michigan. Tests showed that two underground storage tanks leaked. Investigation confirmed this release discharged several hundred gallons of Jet A kerosene into the soil and groundwater. The oil moved downward approximately 30 feet and spread laterally onto the water table. Test results showed kerosene in the adsorbed, free and dissolved states. Eaton researched and investigated three clean-up options. They included pump and treat, dig and haul and bioremediation. Jet fuel is composed of readily biodegradable hydrocarbon chains. This fact coupled with the depth to groundwater and geologic setting made bioremediation the low cost and most effective alternative. A recovery well was installed at the leading edge of the dissolved contamination. A pump moved water from this well into a nutrient addition system. Nutrients added included nitrogen, phosphorous and potassium. Additionally, air was sparged into the water. The water was discharged into an infiltration gallery installed when the underground storage tanks were removed. Water circulated between the pump and the infiltration basin in a closed loop fashion. This oxygenated, nutrient rich water actively and aggressively treated the soils between the bottom of the gallery and the top of the groundwater and the groundwater. The system began operating in August of 1993 and reduced jet fuel to below detection levels. In August of 1995 The State of Michigan issued a clean closure declaration to the site.

  9. Barriers to the Application of High-Temperature Coolants in Hybrid Electric Vehicles

    Energy Technology Data Exchange (ETDEWEB)

    Staunton, Robert H [ORNL; Hsu, John S [ORNL; Starke, Michael R [ORNL

    2006-09-01

    This study was performed by the Oak Ridge National Laboratory (ORNL) to identify practical approaches, technical barriers, and cost impacts to achieving high-temperature coolant operation for certain traction drive subassemblies and components of hybrid electric vehicles (HEV). HEVs are unique in their need for the cooling of certain dedicated-traction drive subassemblies/components that include the electric motor(s), generators(s), inverter, dc converter (where applicable), and dc-link capacitors. The new coolant system under study would abandon the dedicated 65 C coolant loop, such as used in the Prius, and instead rely on the 105 C engine cooling loop. This assessment is important because automotive manufacturers are interested in utilizing the existing water/glycol engine cooling loop to cool the HEV subassemblies in order to eliminate an additional coolant loop with its associated reliability, space, and cost requirements. In addition, the cooling of power electronic devices, traction motors, and generators is critical in meeting the U.S. Department of Energy (DOE) FreedomCAR and Vehicle Technology (FCVT) goals for power rating, volume, weight, efficiency, reliability, and cost. All of these have been addressed in this study. Because there is high interest by the original equipment manufacturers (OEMs) in reducing manufacturing cost to enhance their competitive standing, the approach taken in this analysis was designed to be a positive 'can-do' approach that would be most successful in demonstrating the potential or opportunity of relying entirely on a high-temperature coolant system. Nevertheless, it proved to be clearly evident that a few formidable technical and cost barriers exist and no effective approach for mitigating the barriers was evident in the near term. Based on comprehensive thermal tests of the Prius reported by ORNL in 2005 [1], the continuous ratings at base speed (1200 rpm) with different coolant temperatures were projected from

  10. Barriers to the Application of High-Temperature Coolants in Hybrid Electric Vehicles

    Energy Technology Data Exchange (ETDEWEB)

    Hsu, J.S.; Staunton, M.R.; Starke, M.R.

    2006-09-30

    This study was performed by the Oak Ridge National Laboratory (ORNL) to identify practical approaches, technical barriers, and cost impacts to achieving high-temperature coolant operation for certain traction drive subassemblies and components of hybrid electric vehicles (HEV). HEVs are unique in their need for the cooling of certain dedicated-traction drive subassemblies/components that include the electric motor(s), generators(s), inverter, dc converter (where applicable), and dc-link capacitors. The new coolant system under study would abandon the dedicated 65 C coolant loop, such as used in the Prius, and instead rely on the 105 C engine cooling loop. This assessment is important because automotive manufacturers are interested in utilizing the existing water/glycol engine cooling loop to cool the HEV subassemblies in order to eliminate an additional coolant loop with its associated reliability, space, and cost requirements. In addition, the cooling of power electronic devices, traction motors, and generators is critical in meeting the U.S. Department of Energy (DOE) FreedomCAR and Vehicle Technology (FCVT) goals for power rating, volume, weight, efficiency, reliability, and cost. All of these have been addressed in this study. Because there is high interest by the original equipment manufacturers (OEMs) in reducing manufacturing cost to enhance their competitive standing, the approach taken in this analysis was designed to be a positive 'can-do' approach that would be most successful in demonstrating the potential or opportunity of relying entirely on a high-temperature coolant system. Nevertheless, it proved to be clearly evident that a few formidable technical and cost barriers exist and no effective approach for mitigating the barriers was evident in the near term. Based on comprehensive thermal tests of the Prius reported by ORNL in 2005 [1], the continuous ratings at base speed (1200 rpm) with different coolant temperatures were projected from

  11. Steam as turbine blade coolant: Experimental data generation

    Energy Technology Data Exchange (ETDEWEB)

    Wilmsen, B.; Engeda, A.; Lloyd, J.R. [Michigan State Univ., East Lansing, MI (United States)

    1995-10-01

    Steam as a coolant is a possible option to cool blades in high temperature gas turbines. However, to quantify steam as a coolant, there exists practically no experimental data. This work deals with an attempt to generate such data and with the design of an experimental setup used for the purpose. Initially, in order to guide the direction of experiments, a preliminary theoretical and empirical prediction of the expected experimental data is performed and is presented here. This initial analysis also compares the coolant properties of steam and air.

  12. Lubricant-coolant fluid for machining metals

    Energy Technology Data Exchange (ETDEWEB)

    Berlin, A.A.; Epshtein, V.R.; Pastunov, V.A.; Sherle, A.I.; Shpin' kov, V.A.; Sladkova, T.A.

    1981-03-10

    For improving the antiwear and anticorrosion properties, the lubricant-coolant fluid (LCF) based on water, triethanolamine, and NaNO/sub 2/ contains additionally the sodium salt of an acid ester of maleic acid and substituted oligooxyethylenes (NMO) with the following proportions of the components: triethanolamine 0.3-0.5%, NaNO/sub 2/ 0.3-0.5%, NMO 0.5-2.0%, and water the remainder. In the case of using the proposed LCF on high-speed machine tools, it can contain additionally a foam suppressor in an amount of 0.005-0.1%. For preventing microbiological contamination of the LCF, bactericides of the type furacillin, formalin, vazin (transliteration), and others in an amount of 0.005-0.1% can be added to its composition. Introduction of the NMO additive ensures high wetting and lubricating characteristics in the LCF, which is characterized by stability during storage and service and good anticorrosion properties. Use of the proposed LCF makes it possible to increase the life of the cutting tool by a factor of 2.2 in machining Steel 40Kh and by a factor of 1.3 in machining corroding steel by comparison with the prototype; at the same time the service life of the LCF is increased twofold. The LCF can be used in machining parts of alloyed construction and corrosionresistant steels with cutting-edge and abrasive tools.

  13. Investigating Liquid CO2 as a Coolant for a MTSA Heat Exchanger Design

    Science.gov (United States)

    Paul, Heather L.; Padilla, Sebastian; Powers, Aaron; Iacomini, Christie

    2009-01-01

    Metabolic heat regenerated Temperature Swing Adsorption (MTSA) technology is being developed for thermal and carbon dioxide (CO 2) control for a future Portable Life Support System (PLSS), as well as water recycling. CO 2 removal and rejection is accomplished by driving a sorbent through a temperature swing of approximately 210 K to 280 K . The sorbent is cooled to these sub-freezing temperatures by a Sublimating Heat Exchanger (SHX) with liquid coolant expanded to sublimation temperatures. Water is the baseline coolant available on the moon, and if used, provides a competitive solution to the current baseline PLSS schematic. Liquid CO2 (LCO2) is another non-cryogenic coolant readily available from Martian resources which can be produced and stored using relatively low power and minimal infrastructure. LCO 2 expands from high pressure liquid (5800 kPa) to Mars ambient (0.8 kPa) to produce a gas / solid mixture at temperatures as low as 156 K. Analysis and experimental work are presented to investigate factors that drive the design of a heat exchanger to effectively use this sink. Emphasis is given to enabling efficient use of the CO 2 cooling potential and mitigation of heat exchanger clogging due to solid formation. Minimizing mass and size as well as coolant delivery are also considered. The analysis and experimental work is specifically performed in an MTSA-like application to enable higher fidelity modeling for future optimization of a SHX design. In doing so, the work also demonstrates principles and concepts so that the design can be further optimized later in integrated applications (including Lunar application where water might be a choice of coolant).

  14. HARVESTING EMSP RESEARCH RESULTS FOR WASTE CLEANUP

    Energy Technology Data Exchange (ETDEWEB)

    Guillen, Donna Post; Nielson, R. Bruce; Phillips, Ann Marie; Lebow, Scott

    2003-02-27

    The extent of environmental contamination created by the nuclear weapons legacy combined with expensive, ineffective waste cleanup strategies at many U.S. Department of Energy (DOE) sites prompted Congress to pass the FY96 Energy and Water Development Appropriations Act, which directed the DOE to: ''provide sufficient attention and resources to longer-term basic science research, which needs to be done to ultimately reduce cleanup costs'', ''develop a program that takes advantage of laboratory and university expertise, and'' ''seek new and innovative cleanup methods to replace current conventional approaches which are often costly and ineffective.'' In response, the DOE initiated the Environmental Management Science Program (EMSP)-a targeted, long-term research program intended to produce solutions to DOE's most pressing environmental problems. EMSP funds basic research to lower cleanup cost and reduce risk to workers, the public, and the environment; direct the nation's scientific infrastructure towards cleanup of contaminated waste sites; and bridge the gap between fundamental research and technology development activities. EMSP research projects are competitively awarded based on the project's scientific, merit coupled with relevance to addressing DOE site needs. This paper describes selected EMSP research projects with long, mid, and short-term deployment potential and discusses the impacts, focus, and results of the research. Results of EMSP research are intended to accelerate cleanup schedules, reduce cost or risk for current baselines, provide alternatives for contingency planning, or provide solutions to problems where no solutions exist.

  15. Advances in Forecasting and Prevention of Resonances Between Coolant Acoustical Oscillations and Fuel Rod Vibrations

    Energy Technology Data Exchange (ETDEWEB)

    Proskuryakov, Konstantin Nicolaevich [NPP, NPEI, 14, Krasnokazarmennaya str. Moscow, 111250 (Russian Federation)

    2009-06-15

    To prevent the appearance of the conditions for resonance interaction between the fluid flow and the reactor internals (RI), fuel rod (FR ) and fuel assemblies (FA) it is necessary to de-tune Eigen frequency of coolant pressure oscillations (EFCPO) and natural frequency of mechanical element's oscillations and also of the system which is formed by the comprising of these elements. Other words it is necessary to de-tune acoustic resonance frequency and natural frequencies of RI, FR and FA. While solving these problems it is necessary to have a theoretical and settlement substantiation of an oscillation frequency band of the coolant outside of which there is no resonant interaction with structure vibrations. The presented work is devoted to finding the solution of this problem. There are results of an estimation of width of such band as well as the examples of a preliminary quantitative estimation of Q - factors of coolant acoustic oscillatory circuit formed by the equipment of the NPP. Abnormal growth of intensity of pressure pulsations in a mode with definite value of reactor capacity have been found out by measurements on VVER - 1000 reactor. This phenomenon has been found out casually and its original reason had not been identified. Paper shows that disappearance of this effect could be reached by realizing outlet of EFCPO from so-called, pass bands of frequencies (PBF). PBF is located symmetrical on both parties from frequency of own oscillations of FA. Methods, algorithms of calculations and quantitative estimations are developed for EFCPO, Q and PBF in various modes of operation NPP with VVER-1000. Results of calculations allow specifying area of resonant interaction EFCPO with vibrations of FR, FA and a basket of reactor core. For practical realization of the received results it is offered to make corresponding additions to the design documentation and maintenance instructions of the equipment of the NPP with VVER-1000. The improvement of these documents

  16. Transient two-phase performance of LOFT reactor coolant pumps

    Energy Technology Data Exchange (ETDEWEB)

    Chen, T.H.; Modro, S.M.

    1983-01-01

    Performance characteristics of Loss-of-Fluid Test (LOFT) reactor coolant pumps under transient two-phase flow conditions were obtained based on the analysis of two large and small break loss-of-coolant experiments conducted at the LOFT facility. Emphasis is placed on the evaluation of the transient two-phase flow effects on the LOFT reactor coolant pump performance during the first quadrant operation. The measured pump characteristics are presented as functions of pump void fraction which was determined based on the measured density. The calculated pump characteristics such as pump head, torque (or hydraulic torque), and efficiency are also determined as functions of pump void fractions. The importance of accurate modeling of the reactor coolant pump performance under two-phase conditions is addressed. The analytical pump model, currently used in most reactor analysis codes to predict transient two-phase pump behavior, is assessed.

  17. PIV measurements of coolant flow field in a diesel engine cylinder head

    Science.gov (United States)

    Ma, Hongwei; Zhang, Zhenyang; Xue, Cheng; Huang, Yunlong

    2015-04-01

    This paper presents experimental measurements of coolant flow field in the water jacket of a diesel engine cylinder head. The test was conducted at three different flow rates using a 2-D PIV system. Appropriate tracing particles were selected and delivery device was designed and manufactured before the test. The flow parameters, such as velocity, vorticity and turbulence, were used to analyze the flow field. The effects of vortex which was located between the intake valve and the exhaust valve were discussed. The experimental results showed an asymmetric distribution of velocity in the water jacket. This led to an asymmetric thermal distribution, which would shorten the service life of the cylinder head. The structure optimization to the water jacket of cylinder head was proposed in this paper. The experimental system, especially the 2-D PIV system, is a great help to study the coolant flow structure and analyze cooling mechanism in the diesel engine cylinder head.

  18. Optimized Coolant-Flow Diverter For Increased Bearing Life

    Science.gov (United States)

    Subbaraman, Maria R.; Butner, Myles F.

    1995-01-01

    Coolant-flow diverter for rolling-element bearings in cryogenic turbopump designed to enhance cooling power of flow in contact with bearings and thereby reduce bearing wear. Delivers jets of coolant as close as possible to hot spots at points of contact between balls and race. Also imparts swirl that enhances beneficial pumping effect. Used with success in end ball bearing of high-pressure-oxidizer turbopump.

  19. Steam as turbine blade coolant: Experimental data generation

    Energy Technology Data Exchange (ETDEWEB)

    Wilmsen, B.; Engeda, A.; Lloyd, J.R. [Michigan State Univ., East Lansing, MI (United States). Dept. of Mechanical Engineering

    1995-12-31

    Steam as a coolant is a possible option to cool blades in high temperature gas turbines; however there is practically no experimental data. This work deals with an attempt to generate such data and with the design of an experimental setup used for the purpose. Initially, in order to guide the direction of experiments, a preliminary theoretical and empirical prediction of the expected experimental data is performed and is presented here. This initial analysis also compares the coolant properties of steam and air.

  20. Analysis of Coolant Options for Advanced Metal Cooled Nuclear Reactors

    Science.gov (United States)

    2006-12-01

    1992) PFR UK 250 MWe - 14 Shut Down (1994) Rapsodie France 40 MWe - 40 Shut Down (1983) Phenix France 233 MWe - 22 In Operation BOR-60 Russia...107years.98 • Problems with radioactive waste management and coolant disposal during decommissioning .99 O th er • Lead is abundantly available in...is high due to Bi-210, half-life 3.6 106years.102 • Problems with radioactive waste management and coolant disposal during decommissioning . 103 O

  1. Renewable Natural Gas Clean-up Challenges and Applications

    Science.gov (United States)

    2011-01-13

    removed including CO, CO2,CH4) Illustrative Process Flow Diagram for On-site H2 Supply System & SOFC Power Generation...day 1.5 to SOFC ) 13.2 scfm . 8.0 scfm Flow rate: ~ 2.9 scfm ( PSA: ~ 31.7scfm) Usable Heat Electricity 2 CO: ~0.5% H2: ~73.5 Total Flow...34.6 scfm SOFC : 17 t 18 Example Gas Cl eanup Sys em for WWDG > Configured a gas cleanup system utilizing a membrane module for CO

  2. Flood Cleanup to Protect Indoor Air Quality

    Science.gov (United States)

    During a flood cleanup, the indoor air quality in your home or office may appear to be the least of your problems. However, failure to remove contaminated materials and to reduce moisture and humidity can present serious long-term health risks.

  3. Assessment of synfuel spill cleanup options

    Energy Technology Data Exchange (ETDEWEB)

    Petty, S.E.; Wakamiya, W.; English, C.J.; Strand, J.A.; Mahlum, D.D.

    1982-04-01

    Existing petroleum-spill cleanup technologies are reviewed and their limitations, should they be used to mitigate the effects of synfuels spills, are discussed. The six subsections of this report address the following program goals: synfuels production estimates to the year 2000; possible sources of synfuel spills and volumes of spilled fuel to the year 2000; hazards of synfuels spills; assessment of existing spill cleanup technologies for oil spills; assessment of cleanup technologies for synfuel spills; and disposal of residue from synfuel spill cleanup operations. The first goal of the program was to obtain the most current estimates on synfuel production. These estimates were then used to determine the amount of synfuels and synfuel products likely to be spilled, by location and by method of transportation. A review of existing toxicological studies and existing spill mitigation technologies was then completed to determine the potential impacts of synthetic fuel spills on the environment. Data are presented in the four appendixes on the following subjects: synfuel production estimates; acute toxicity of synfuel; acute toxicity of alcohols.

  4. Estimation of the coolant flow through a natural circulation BWR fuel channel applying and equivalent electrical model

    Energy Technology Data Exchange (ETDEWEB)

    Valle H, J.; Morales S, J. B. [UNAM, DEPFI, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, Campus Morelos en IMTA, Jiutepec, Morelos (Mexico); Espinosa P, G., E-mail: julfi_ig@yahoo.com.m [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Av. San Rafael Atlixco No. 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)

    2010-10-15

    conditions, the model was scaled and compared to nominal values of a 4500 MWt ESBWR. The design of the reduced order model controller was mathematically based on the theory of an extended Kalman filter, whose algorithm allows to carry out an advanced control of the system. The estimator uses the equivalent electrical model, which was developed from the system. The estimator uses the equivalent electrical model, which was developed from the analogies between electrical current and voltage with coolant flows and pressure drops. (Author)

  5. Development of additional module to neutron-physic and thermal-hydraulic computer codes for coolant acoustical characteristics calculation

    Energy Technology Data Exchange (ETDEWEB)

    Proskuryakov, K.N.; Bogomazov, D.N.; Poliakov, N. [Moscow Power Engineering Institute (Technical University), Moscow (Russian Federation)

    2007-07-01

    The new special module to neutron-physic and thermal-hydraulic computer codes for coolant acoustical characteristics calculation is worked out. The Russian computer code Rainbow has been selected for joint use with a developed module. This code system provides the possibility of EFOCP (Eigen Frequencies of Oscillations of the Coolant Pressure) calculations in any coolant acoustical elements of primary circuits of NPP. EFOCP values have been calculated for transient and for stationary operating. The calculated results for nominal operating were compared with results of measured EFOCP. For example, this comparison was provided for the system: 'pressurizer + surge line' of a WWER-1000 reactor. The calculated result 0.58 Hz practically coincides with the result of measurement (0.6 Hz). The EFOCP variations in transients are also shown. The presented results are intended to be useful for NPP vibration-acoustical certification. There are no serious difficulties for using this module with other computer codes.

  6. Cleanups In My Community (CIMC) - Recovery Act Funded Cleanups, National Layer

    Data.gov (United States)

    U.S. Environmental Protection Agency — This data layer provides access to Recovery Act Funded Cleanup sites as part of the CIMC web service. The American Recovery and Reinvestment Act was signed into law...

  7. Fast-Track Cleanup at Closing DoD Installations

    Science.gov (United States)

    The Fast-Track Cleanup program strives to make parcels available for reuse as quickly as possible by the transfer of uncontaminated or remediated parcels, the lease of contaminated parcels where cleanup is underway, or the 'early transfer' of contaminated property undergoing cleanup.

  8. Development of core design and analysis technology for integral reactor; development of coolant activity and dose evaluation program

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Chang Sun; Kim, Byeong Soo; Go, Hyun Seok; Lee, Young Wook; Jang, Mee [Seoul National University, Seoul (Korea)

    2002-03-01

    SMART, small- medium-sized integral reactor, is different from the customary electricity-generation PWR in design concepts and structures. The conventional coolant activity evaluation codes used in customary PWRs cannot be applied to SMART. In this study, SAEP(Specific Activity Evaluation Program) is developed that can be applied to both customary PWR and advanced reactor such as SMART. SAEP uses three methods(SAEP Ver.02, Ver.05, Ver.06) to evaluate coolant activity. They solve inhomogeneous linearly-coupled differential equations generated by considering nuclear system as N sub-components. Coolant activities of customary PWR are evaluated by use of SAEP. The results show good agreement with FSAR data. SAEP is used to evaluate coolant activities for SMART and the results are proposed in this study. These results show that SAEP is able to perform coolant activity evaluation for both customary PWR and advanced reactor such as SMART. In addition, with respect to radiation shielding optimization, conventional optimization methods and their characteristics related to radiation shielding are reviewed and analyzed. Strategies for proper usage of conventional methods are proposed to agree with the shielding design cases. 30 refs., 25 figs., 6 tabs. (Author)

  9. Development of core design and analysis technology for integral reactor; development of coolant activity and dose evaluation program

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Chang Sun; Kim, Byeong Soo; Go, Hyun Seok; Lee, Young Wook; Jang, Mee [Seoul National University, Seoul (Korea)

    2002-03-01

    SMART, small- medium-sized integral reactor, is different from the customary electricity-generation PWR in design concepts and structures. The conventional coolant activity evaluation codes used in customary PWRs cannot be applied to SMART. In this study, SAEP(Specific Activity Evaluation Program) is developed that can be applied to both customary PWR and advanced reactor such as SMART. SAEP uses three methods(SAEP Ver.02, Ver.05, Ver.06) to evaluate coolant activity. They solve inhomogeneous linearly-coupled differential equations generated by considering nuclear system as N sub-components. Coolant activities of customary PWR are evaluated by use of SAEP. The results show good agreement with FSAR data. SAEP is used to evaluate coolant activities for SMART and the results are proposed in this study. These results show that SAEP is able to perform coolant activity evaluation for both customary PWR and advanced reactor such as SMART. In addition, with respect to radiation shielding optimization, conventional optimization methods and their characteristics related to radiation shielding are reviewed and analyzed. Strategies for proper usage of conventional methods are proposed to agree with the shielding design cases. 30 refs., 25 figs., 6 tabs. (Author)

  10. Hydrodynamics of heavy liquid metal coolant processes and filtering apparatus

    Energy Technology Data Exchange (ETDEWEB)

    Albert K Papovyants; Yuri I Orlov; Pyotr N Martynov; Yuri D Boltoev [Institute for Physics and Power Engineering named after A.I. Leypunsky Bondarenko sq. 1, 249033, Obninsk, Kaluga region (Russian Federation)

    2005-07-01

    Full text of publication follows: To optimize the design of filters for cleaning heavy liquid metal coolant (HLMC) from suspended impurities and choose appropriate filter material, the contribution is considered of different mechanisms of delivery and retention of these impurities from the coolant flow, which is governed by its specificity as a thermodynamically instable disperse system to a large extent. It is shown that the buildup of deposits in the filter is favored by the hydrodynamic regime with minimum filtration rates being due to the predominance in the suspension of the fine-dispersed solid phase (oxides Fe{sub 3}O{sub 4}, Cr{sub 2}O{sub 3} and so on). With concentrating the last mentioned phase in filter material pores or stagnant zones, coagulation structuration is possible, which is accompanied by sharp local increase in the viscosity and strength of the solid phase medium being built from liquid metal, i.e. slag sedimentary deposits. In rather extended pores, disintegration of such structures is possible, which is accompanied by sedimentation of large particles produced due to sticking together at coagulation. The analytical solution of the problem of particle sedimentation due to diffusion indicated that in the case under consideration, this mechanism takes place for particles less than {approx} 0,05 {mu}m in size, which is specified by the fact that the time of their delivery to the filter material surface is longer than that of the coolant being in the filter. The London-Van-der-Waals molecular forces play a crucial role in the stage of retention of a separate particle. The constant of the molecular interaction between a spherical particle and the flat surface has been estimated for the chosen value of the gap between the contacting bodies, being dependent on the wetting angle. The sufficient condition for d{sub p}-diameter particle capture by the adhesion force field (with a gap of H {approx_equal} 30 nm) is that it be brought by the appropriate

  11. A study on safety measure of LMR coolant

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Sung Tai; Choi, Y. D.; Choi, J. H.; Kim, T. J.; Jeong, K. C.; Kwon, S. W.; Kim, B. H.; Jeong, J. Y.; Park, J. H.; Kim, K. R.; Jo, B. R.

    1997-08-01

    A study on safety measures of LMR coolant showed the results as follows: 1. Sodium fire characteristics. A. Sodium pool temp., gas temp., oxygen concentration calculated by flame combustion model were generally higher than those calculated by surface combustion model. B. Basic and detail designs for medium sodium fire test facility were carried out and medium sodium fire test facility was constructed. 2. Sodium/Cover gas purification technology. A. Construction and operation of calibration loop. B. Purification analysis and conceptual design of the packing for a cold trap. 3. Analysis of sodium-water reaction characteristics. We have investigated the characteristics analysis for micro and small leaks phenomena, development of the computer code for analysis of initial and quasi steady-state spike pressures to analyze large leak accident. Also, water mock-up test facility for the analysis of large leak accident phenomena was designed and manufactured. 4. Development of water leak detection technology. Detection signals were appeared when the hydrogen detector is operated to Ar-H{sub 2} gas system. The technology for the passive acoustic detection with respect to large leakage of water into sodium media was reviewed. And water mock-up test equipment and instrument system were designed and constructed. (author). 19 refs., 45 tabs., 52 figs.

  12. Primary coolant sampling for activated corrosion product studies at Hanford N Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bechtold, D.B.

    1985-01-31

    A special system for sampling primary coolant at N Reactor during operation has been constructed and operated from 1977 to 1983. The basic criteria and design for solving the difficult problem of getting representative samples have been presented; this report details how the instrumentation was configured and sampling was done. Equipment and procedures were put together to allow one person to enter a radiation zone, check on 5 monitoring instruments, operate two batch instruments, gather five partitioned samples, record 26 pieces of information, annotate a strip chart and leave the zone in 30 minutes while expending 10 mRem of exposure. Additionally, the reduction of the samples' analysis, digitization of strip chart information and storage of all data on data management systems is maintained. As built, the system provides 0.3 to 1.0 gpm streams of coolant from upstream and downstream of a steam generator. The streams are cooled to 50 to 60/sup 0/C. The radiation environment averages 20 to 50 mR/hr to the worker. Instruments and special equipment for data gathering at the sampler include pH, conductance, dissolved oxygen, dissolved hydrogen and nitrogen, hot leg and cold leg coolant temperatures, particle sizing, turbidimetry, filtration, and continuous strip chart recording.

  13. Methodology for determining of the weighted mean coolant temperature in the primary circuit hot legs of WWER-1000 reactor plants

    Energy Technology Data Exchange (ETDEWEB)

    Saunin, Yuri V.; Dobrotvorski, Alexander N.; Semenikhin, Alexander V. [JSC ' Atomtechenergo' , Filial ' Novovoronezhatomtechenergo' , Novovorenezh (Russian Federation); Ryasny, Sergei I. [JSC ' Atomtechenergo' , Mytishi (Russian Federation)

    2016-09-15

    At WWER-1000 NPPs, as well as at PWR NPPs, there is a problem of determining the correct weighted mean coolant temperature in the primary circuit hot legs based on the measuring channels information. The problem is caused by the coolant temperature stratification. The technical documentation for engineering support and maintenance of I and C systems does not provide any regulatory guidelines to consider this effect. Therefore, it is very important to represent a new methodology for determining the weighted mean coolant temperature in the primary circuit hot legs of the WWER-1000 reactor plants. The given paper presents the basic preconditions and approaches applied during the methodology development. They were worked out on the basis of the executed numerical and experimental research taking into account the analysis of the extensive material obtained by the authors from full-scale tests during the commissioning of WWER-1000 power units, as well as operational data obtained from several power units with different fuel loadings.

  14. A Frequency Transfer and Cleanup System for Ultra-High Stability at Both Long and Short Times for the Cassini Ka-Band Experiment

    Science.gov (United States)

    Calhoun, M. D.; Dick, G. J.; Wang, R. T.

    1999-01-01

    New radio science experiments, including a gravitational wave search and several atmospheric occultation studies, are planned for the Cassini Ka-band experiment. These experiments are made possible by reduced solar-induced phase fluctuations at the high-frequency (32 GHZ) of the radio link between the earth and the spacecraft. In order to match the improved link performance, a significant upgrade is under way to improve the frequency stability capabilities of NASA's Deep Space Network (DSN). Significant improvements are being undertaken in many areas, including antenna vibration and (wet) tropospheric calibration, in addition to frequency generation and distribution. We describe here the design and development of a system to provide a reference signal with the highest possible frequency stability for both long-term, short-term, and phase noise, at an antenna (DSS 25) that is remote from the frequency standards room at SPC-10 at the Goldstone site. The new technologies were developed in order to meet the very tight requirements. They are: 1) a Stabilized Fiber-Optic Distribution Assembly (SFODA) that includes active compensation of thermal variations to transfer long-term stability over 16 km of ordinary fiber-optic cable, and 2) a Compensated Sapphire Oscillator (CSO) that provides short-term performance in a cryocooled sapphire oscillator with ultra-high short-term stability and low phase noise.

  15. Small-break loss-of-coolant accidents in the updated PIUS 600 advanced reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Boyack, B.E.; Steiner, J.L.; Harmony, S.C. [Los Alamos National Lab., Albuquerque, NM (United States)] [and others

    1995-09-01

    The PIUS advanced reactor is a 640-MWe pressurized water reactor developed by Asea Brown Boveri (ABB). A unique feature of the PIUS concept is the absence of mechanical control and shutdown rods. Reactivity is normally controlled by coolant boron concentration and the temperature of the moderator coolant. ABB submitted the PIUS design to the US Nuclear Regulatory Commission (NRC) for preapplication review, and Los Alamos supported the NRC`s review effort. Baseline analyses of small-break initiators at two locations were performed with the system neutronic and thermal-hydraulic analysis code TRAC-PF1/MOD2. In addition, sensitivity studies were performed to explore the robustness of the PIUS concept to severe off-normal conditions having a very low probability of occurrence.

  16. DRUCKFLAMM - Investigation on combustion and hot gas cleanup in pulverized coal combustion systems. Final report; DRUCKFLAMM - Untersuchungen zur Verbrennung und Heissgasreinigung bei der Druckkohlenstaubfeuerung. Schlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Hein, K.R.G.; Benoehr, A.; Schuermann, H.; Stroehle, J.; Klaiber, C.; Kuhn, R.; Maier, J.; Schnell, U.; Unterberger, S.

    2001-07-01

    The ambitions of making energy supply more efficient and less polluting brought forth the development of coal based combined cycle power plants allowing considerable increases in net efficiencies. One of the regarded firing concepts for a coal based combined cycle power plant is represented by the pressurised pulverised coal combustion process which has the highest efficiency potential compared with the other coal based concepts. The fundamental purpose of the project was to gain firm knowledge concerning firing behaviour of coal in a pressurised pulverised coal combustion system. Detailed investigations were carried out in a pressurised entrained flow reactor taking into account fuel conversion and particle behaviour, pollutant formation and material behaviour under conditions of a pressurised pulverised coal firing. During the project's investigations several different measurement techniques were tested and partially also acquired (e.g. a two-colour-pyrometry system to measure simultaneous particle surface temperature and particle diameter of burning fuel particles). Calculation models under pressurised conditions for pressure vessel simulation and better scale-up were developed synchronously with the experimental investigations. The results gained using the pressurised entrained flow reactor show that many combustion mechanisms are influenced by increased pressure, for instance the fuel conversion is intensified and at the same time pollutant emissions decreased. The material investigations show that the ceramic materials used due to the very high combustion temperatures are very sensitive versus slagging and fast temperature changes, therefore further development requirements are needed to fully realise the high durability of ceramics in the pressurised furnace. Concerning the improvement of existing models for furnace simulation under pressurised conditions, a good resemblance can be observed when considering the actual measurement results from the test

  17. Development of an Integrated Multi-Contaminant Removal Process Applied to Warm Syngas Cleanup for Coal-Based Advanced Gasification Systems

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, Howard

    2010-11-30

    This project met the objective to further the development of an integrated multi-contaminant removal process in which H2S, NH3, HCl and heavy metals including Hg, As, Se and Cd present in the coal-derived syngas can be removed to specified levels in a single/integrated process step. The process supports the mission and goals of the Department of Energy's Gasification Technologies Program, namely to enhance the performance of gasification systems, thus enabling U.S. industry to improve the competitiveness of gasification-based processes. The gasification program will reduce equipment costs, improve process environmental performance, and increase process reliability and flexibility. Two sulfur conversion concepts were tested in the laboratory under this project, i.e., the solventbased, high-pressure University of California Sulfur Recovery Process High Pressure (UCSRP-HP) and the catalytic-based, direct oxidation (DO) section of the CrystaSulf-DO process. Each process required a polishing unit to meet the ultra-clean sulfur content goals of <50 ppbv (parts per billion by volume) as may be necessary for fuel cells or chemical production applications. UCSRP-HP was also tested for the removal of trace, non-sulfur contaminants, including ammonia, hydrogen chloride, and heavy metals. A bench-scale unit was commissioned and limited testing was performed with simulated syngas. Aspen-Plus®-based computer simulation models were prepared and the economics of the UCSRP-HP and CrystaSulf-DO processes were evaluated for a nominal 500 MWe, coal-based, IGCC power plant with carbon capture. This report covers the progress on the UCSRP-HP technology development and the CrystaSulf-DO technology.

  18. Simulation of a large break loss of coolant (LBLOCA), without actuation of the emergency injection systems (ECCS) for a BWR-5; Simulacion de un escenario de perdida de refrigerante grande (LBLOCA), sin actuacion de los sistemas de inyeccion de emergencia (ECCS) para un reactor BWR-5

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas V, J.; Mugica R, C. A.; Lopez M, R., E-mail: jaime.cardenas@cnsns.gob.mx [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779, Col. Narvarte, 03020 Ciudad de Mexico (Mexico)

    2015-09-15

    In this paper the analysis of scenario for the loss of coolant case was realized with break at the bottom of a recirculation loop of a BWR-5 with containment type Mark II and a thermal power of 2317 MWt considering that not have coolant injection. This in order to observe the speed of progression of the accident, the phenomenology of the scenario, the time to reach the limit pressure of containment venting and the amount of radionuclides released into the environment. This simulation was performed using the MELCOR code version 2.1. The scenario posits a break in one of the shear recirculation loops. The emergency core cooling system (ECCS) and the reactor core isolation cooling (Rcic) have not credit throughout the event, which allowed achieve greater severity on scenario. The venting of the primary containment was conducted via valve of 30 inches instead of the line of 24 inches of wet well, this in order to have a larger area of exhaust of fission products directly to the reactor building. The venting took place when the pressure in the primary containment reached the 4.5 kg/cm{sup 2} and remained open for the rest of the scenario to maximize the amount released of radionuclides to the atmosphere. The safety relief valves were considered functional they do not present mechanical failure or limit their ability to release pressure due to the large number of performances in safety mode. The results of the analysis covers about 48 hours, time at which the accident evolution was observed; behavior of level, pressure in the vessel and the fuel temperature profile was analyzed. For progression of the scenario outside the vessel, the pressure and temperature of the primary containment, level and temperature of the suppression pool, the hydrogen accumulation in the container and the radionuclides mass released into the atmosphere were analyzed. (Author)

  19. Radioactive Waste and Clean-up: Introduction

    Energy Technology Data Exchange (ETDEWEB)

    Collard, G

    2000-07-01

    SCK-CEN's Radioactive Waste and Clean-up Division performs studies and develops strategies, techniques and technologies in the area of radioactive waste management, the decontamination and decommissioning of nuclear installations and the remediation of radioactive-contaminated sites. These activities are performed in the context of our responsibility towards the safety of present and future generations and contribute to achieve intrageneration equity.

  20. Evaluation of containment failure and cleanup time for Pu shots on the Z machine.

    Energy Technology Data Exchange (ETDEWEB)

    Darby, John L.

    2010-02-01

    Between November 30 and December 11, 2009 an evaluation was performed of the probability of containment failure and the time for cleanup of contamination of the Z machine given failure, for plutonium (Pu) experiments on the Z machine at Sandia National Laboratories (SNL). Due to the unique nature of the problem, there is little quantitative information available for the likelihood of failure of containment components or for the time to cleanup. Information for the evaluation was obtained from Subject Matter Experts (SMEs) at the Z machine facility. The SMEs provided the State of Knowledge (SOK) for the evaluation. There is significant epistemic- or state of knowledge- uncertainty associated with the events that comprise both failure of containment and cleanup. To capture epistemic uncertainty and to allow the SMEs to reason at the fidelity of the SOK, we used the belief/plausibility measure of uncertainty for this evaluation. We quantified two variables: the probability that the Pu containment system fails given a shot on the Z machine, and the time to cleanup Pu contamination in the Z machine given failure of containment. We identified dominant contributors for both the time to cleanup and the probability of containment failure. These results will be used by SNL management to decide the course of action for conducting the Pu experiments on the Z machine.

  1. Fracture mechanics evaluation for at typical PWR primary coolant pipe

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, T. [Kansai Electric Power Company, Osaka (Japan); Shimizu, S.; Ogata, Y. [Mitsubishi Heavy Industries, Ltd., Kobe (Japan)

    1997-04-01

    For the primary coolant piping of PWRs in Japan, cast duplex stainless steel which is excellent in terms of strength, corrosion resistance, and weldability has conventionally been used. The cast duplex stainless steel contains the ferrite phase in the austenite matrix and thermal aging after long term service is known to change its material characteristics. It is considered appropriate to apply the methodology of elastic plastic fracture mechanics for an evaluation of the integrity of the primary coolant piping after thermal aging. Therefore we evaluated the integrity of the primary coolant piping for an initial PWR plant in Japan by means of elastic plastic fracture mechanics. The evaluation results show that the crack will not grow into an unstable fracture and the integrity of the piping will be secured, even when such through wall crack length is assumed to equal the fatigue crack growth length for a service period of up to 60 years.

  2. Influence of coolant motion on structure of hardened steel element

    Directory of Open Access Journals (Sweden)

    A. Kulawik

    2008-08-01

    Full Text Available Presented paper is focused on volumetric hardening process using liquid low melting point metal as a coolant. Effect of convective motion of the coolant on material structure after hardening is investigated. Comparison with results obtained for model neglecting motion of liquid is executed. Mathematical and numerical model based on Finite Element Metod is described. Characteristic Based Split (CBS method is used to uncouple velocities and pressure and finally to solve Navier-Stokes equation. Petrov-Galerkin formulation is employed to stabilize convective term in heat transport equation. Phase transformations model is created on the basis of Johnson-Mehl and Avrami laws. Continuous cooling diagram (CTPc for C45 steel is exploited in presented model of phase transformations. Temporary temperatures, phases participation, thermal and structural strains in hardening element and coolant velocities are shown and discussed.

  3. Actively controlling coolant-cooled cold plate configuration

    Energy Technology Data Exchange (ETDEWEB)

    Chainer, Timothy J.; Parida, Pritish R.

    2015-07-28

    A method is provided to facilitate active control of thermal and fluid dynamic performance of a coolant-cooled cold plate. The method includes: monitoring a variable associated with at least one of the coolant-cooled cold plate or one or more electronic components being cooled by the cold plate; and dynamically varying, based on the monitored variable, a physical configuration of the cold plate. By dynamically varying the physical configuration, the thermal and fluid dynamic performance of the cold plate are adjusted to, for example, optimally cool the one or more electronic components, and at the same time, reduce cooling power consumption used in cooling the electronic component(s). The physical configuration can be adjusted by providing one or more adjustable plates within the coolant-cooled cold plate, the positioning of which may be adjusted based on the monitored variable.

  4. Experimental study of high temperature particle dropping in coolant liquid

    Institute of Scientific and Technical Information of China (English)

    LI Tianshu; YANG Yanhua; LI Xiaoyan; HU Zhihua

    2007-01-01

    A series of experiments of the premixing stage of fuel-coolant interactions (FCI), namely the particles falling into water, were carried out. The force on the particles during the course of falling has been studied. The dropping character of hot particle was influenced by three main parameters, i.e., particle temperature, particle diameter and coolant subcooling that varied over a wide range. A high-speed camera recorded the falling speed of the particle and the moving curves were obtained. The experimental results showed that for the film boiling on the surface of particle and water, the temperature increase of either particle or coolant would slow down the particle falling velocity. The falling velocity of particle in small diameter is lower than that of the bigger particle. The present work can provide an experimental foundation for further investigation of high-speed transient evaporation heat transfer.

  5. Beam cleanup of a 532-nm pulsed solid-state laser using a bimorph mirror

    Institute of Scientific and Technical Information of China (English)

    Xiang Lei; Bing Xu; Ping Yang; Lizhi Dong; Wenjin Liu; Hu Yan

    2012-01-01

    A successful beam cleanup of a 5-mJ/200-μs pulsed solid-state laser system operating at 532-nm wavelength is demonstrated. In this beam cleanup system, a wave-front sensor-less adaptive optics (AO) system is set up with a 20-element bimorph mirror (BM), a high-voltage amplifier, a charge-coupled device camera, and a control software implementing the stochastic parallel gradient descent (SPGD) algorithm. The brightness of the laser focal spot is improved because the wave-front distortions have been compensated. The performance of this system is presented and the experimental results are analyzed.%A successful beam cleanup of a 5-mJ/200-μs pulsed solid-state laser system operating at 532-nm wavelength is demonstrated.In this beam cleanup system,a wave-front sensor-less adaptive optics (AO) system is set up with a 20-element bimorph mirror (BM),a high-voltage amplifier,a charge-coupled device camera,and a control software implementing the stochastic parallel gradient descent (SPGD) algorithm.The brightness of the laser focal spot is improved because the wave-front distortions have been compensated.The performance of this system is presented and the experimental results are analyzed.

  6. Fuel, Structural Material and Coolant for an Advanced Fast Micro-Reactor

    Science.gov (United States)

    Do Nascimento, J. A.; Duimarães, L. N. F.; Ono, S.

    The use of nuclear reactors in space, seabed or other Earth hostile environment in the future is a vision that some Brazilian nuclear researchers share. Currently, the USA, a leader in space exploration, has as long-term objectives the establishment of a permanent Moon base and to launch a manned mission to Mars. A nuclear micro-reactor is the power source chosen to provide energy for life support, electricity for systems, in these missions. A strategy to develop an advanced micro-reactor technologies may consider the current fast reactor technologies as back-up and the development of advanced fuel, structural and coolant materials. The next generation reactors (GEN-IV) for terrestrial applications will operate with high output temperature to allow advanced conversion cycle, such as Brayton, and hydrogen production, among others. The development of an advanced fast micro-reactor may create a synergy between the GEN-IV and space reactor technologies. Considering a set of basic requirements and materials properties this paper discusses the choice of advanced fuel, structural and coolant materials for a fast micro-reactor. The chosen candidate materials are: nitride, oxide as back-up, for fuel, lead, tin and gallium for coolant, ferritic MA-ODS and Mo alloys for core structures. The next step will be the neutronic and burnup evaluation of core concepts with this set of materials.

  7. Technological status of reactor coolant pumps in generation III+ pressurized nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Brecht, Bernhard; Bross, Stephan [KSB Aktiengesellschaft, Frankenthal (Germany)

    2016-05-15

    KSB has been developing and producing pumps for thermal power plants for nearly 90 years. Consequently, KSB also started to develop and manufacture pumps for all kinds of nuclear power plants from the very beginning of the civil use of nuclear energy. This is especially true for reactor coolant pumps for pressurized water reactors. For the generation of advanced evolutionary reactors (Generation III+ reactors), KSB developed an advanced shaft seal system which is also able to fulfill the requirements of station blackout conditions. The tests in the KSB test rigs, which were successfully completed in December 2015, proved the full functionality of the new design. For generation III+ passive plant reactors KSB developed a new reactor coolant pump type called RUV, which is based on the experience of classic reactor coolant pumps and reactor internal pumps. It is a very compact, hermetically sealed vertical pump-motor unit with a wet winding motor. A full scale prototype successfully passed the 1st stage qualification test program in October 2015.

  8. Neutronic analysis of a high power density hybrid reactor using innovative coolants

    Indian Academy of Sciences (India)

    Senay Yalçin; Mustafa Übeylı; Adem Acir

    2005-08-01

    In this study, neutronic investigation of a deuterium–tritium (DT) driven hybrid reactor using ceramic uranium fuels, namely UC, UO2 or UN under a high neutron wall load (NWL) of 10 MW/m2 at the first wall is conducted over a period of 24 months for fissile fuel breeding for light water reactors (LWRs). New substances, namely, Flinabe or Li20Sn80 are used as coolants in the fuel zone to facilitate heat transfer out of the blanket. Natural lithium is also utilized for comparison to these two innovative coolants. Neutron transport calculations are performed on a simple experimental hybrid blanket with cylindrical geometry with the help of the SCALE 4·3 System by solving the Boltzmann transport equation with the XSDRNPM code in 238 neutron groups and an S8-P3 approximation. The investigated blanket using Flinabe or Li20Sn80 shows better fissile fuel breeding and fuel enrichment characteristics compared to that with natural lithium which shows that these two innovative coolants can be used in hybrid reactors for higher fissile fuel breeding performance. Furthermore, using a high NWL of 10 MW/m2 at the first wall of the investigated blanket can decrease the time for fuel rods to reach the level for charging in LWRs.

  9. Inverse design of a proper number, shapes, sizes, and locations of coolant flow passages

    Science.gov (United States)

    Dulikravich, George S.

    1992-01-01

    During the past several years we have developed an inverse method that allows a thermal cooling system designer to determine proper sizes, shapes, and locations of coolant passages (holes) in, say, an internally cooled turbine blade, a scram jet strut, a rocket chamber wall, etc. Using this method the designer can enforce a desired heat flux distribution on the hot outer surface of the object, while simultaneously enforcing desired temperature distributions on the same hot outer surface as well as on the cooled interior surfaces of each of the coolant passages. This constitutes an over-specified problem which is solved by allowing the number, sizes, locations and shapes of the holes to adjust iteratively until the final internally cooled configuration satisfies the over-specified surface thermal conditions and the governing equation for the steady temperature field. The problem is solved by minimizing an error function expressing the difference between the specified and the computed hot surface heat fluxes. The temperature field analysis was performed using our highly accurate boundary integral element code with linearly varying temperature along straight surface panels. Examples of the inverse design applied to internally cooled turbine blades and scram jet struts (coated and non-coated) having circular and non-circular coolant flow passages will be shown.

  10. Reactor coolant pump shaft seal behavior during blackout conditions

    Energy Technology Data Exchange (ETDEWEB)

    Mings, W.J.

    1985-01-01

    The United States Nuclear Regulatory Commission has classified the problem of reactor coolant pump seal failures as an unresolved safety issue. This decision was made in large part due to experimental results obtained from a research program developed to study shaft seal performance during station blackout and reported in this paper. Testing and analysis indicated a potential for pump seal failure under postulated blackout conditions leading to a loss of primary coolant with a concomitant danger of core uncovery. The work to date has not answered all the concerns regarding shaft seal failure but it has helped scope the problem and focus future research needed to completely resolve this issue.

  11. 反应堆一回路系统优化设计方案的可行性验证%Feasibility Test for Reactor Coolant System Optimized Design Scheme

    Institute of Scientific and Technical Information of China (English)

    陈磊; 阎昌琪; 王建军

    2014-01-01

    采用优选运行参数和结构参数的方法,可达到降低核动力装置尺寸的目的。在优化设计方案投入制造前,有必要研究其在设计基准事故下的响应特性,以检验优化方案的可行性。采用 REL A P5/M OD3.2程序研究现有一回路系统优化方案在完全失去厂外电、主给水丧失和小破口失水事故下的响应特性,并将安全设计准则参数与母型对比。结果表明:针对所研究的3种设计基准事故,优化方案各主要安全准则参数满足设计要求;优化方案可成功抵御这3类设计基准事故。%The size of a nuclear component could be reduced by optimum selections of the operational and structural parameters .Before an optimized design scheme is manu‐factured ,it is necessary to obtain its transient behaviors and verify its feasibility under design basis accidents .In this work ,the RELAP5/MOD3.2 code was employed to simulate the transient characteristics of a proposed optimized scheme under the complete loss of off‐site power ,loss of feedwater and small break loss of coolant accidents ,and the safety criteria were compared with the prototype reactor design . The simulation results indicate that the safety criteria of the optimized scheme satisfy the design requirements ,and the safety of the optimized scheme can be guaranteed in those three accidents .

  12. Reactor coolant pump testing using motor current signatures analysis

    Energy Technology Data Exchange (ETDEWEB)

    Burstein, N.; Bellamy, J.

    1996-12-01

    This paper describes reactor coolant pump motor testing carried out at Florida Power Corporation`s Crystal River plant using Framatome Technologies` new EMPATH (Electric Motor Performance Analysis and Trending Hardware) system. EMPATH{trademark} uses an improved form of Motor Current Signature Analysis (MCSA), technology, originally developed at Oak Ridge National Laboratories, for detecting deterioration in the rotors of AC induction motors. Motor Current Signature Analysis (MCSA) is a monitoring tool for motor driven equipment that provides a non-intrusive means for detecting the presence of mechanical and electrical abnormalities in the motor and the driven equipment. The base technology was developed at the Oak Ridge National Laboratory as a means for determining the affects of aging and service wear specifically on motor-operated valves used in nuclear power plant safety systems, but it is applicable to a broad range of electric machinery. MCSA is based on the recognition that an electric motor (ac or dc) driving a mechanical load acts as an efficient and permanently available transducer by sensing mechanical load variations, large and small, long-term and rapid, and converting them into variations in the induced current generated in the motor windings. The motor current variations, resulting from changes in load caused by gears, pulleys, friction, bearings, and other conditions that may change over the life of the motor, are carried by the electrical cables powering the motor and are extracted at any convenient location along the motor lead. These variations modulate the 60 Hz carrier frequency and appear as sidebands in the spectral plot.

  13. Correct numerical simulation of a two-phase coolant

    Science.gov (United States)

    Kroshilin, A. E.; Kroshilin, V. E.

    2016-02-01

    Different models used in calculating flows of a two-phase coolant are analyzed. A system of differential equations describing the flow is presented; the hyperbolicity and stability of stationary solutions of the system is studied. The correctness of the Cauchy problem is considered. The models' ability to describe the following flows is analyzed: stable bubble and gas-droplet flows; stable flow with a level such that the bubble and gas-droplet flows are observed under and above it, respectively; and propagation of a perturbation of the phase concentration for the bubble and gas-droplet media. The solution of the problem about the breakdown of an arbitrary discontinuity has been constructed. Characteristic times of the development of an instability at different parameters of the flow are presented. Conditions at which the instability does not make it possible to perform the calculation are determined. The Riemann invariants for the nonlinear problem under consideration have been constructed. Numerical calculations have been performed for different conditions. The influence of viscosity on the structure of the discontinuity front is studied. Advantages of divergent equations are demonstrated. It is proven that a model used in almost all known investigating thermohydraulic programs, both in Russia and abroad, has significant disadvantages; in particular, it can lead to unstable solutions, which makes it necessary to introduce smoothing mechanisms and a very small step for describing regimes with a level. This does not allow one to use efficient numerical schemes for calculating the flow of two-phase currents. A possible model free from the abovementioned disadvantages is proposed.

  14. AUTOMOTIVE AND HEAVY-DUTY ENGINE COOLANT RECYCLING BY DISTILLATION

    Science.gov (United States)

    This evaluation addresses the product quality, waste reduction, and economic issues involved in recycling automotive and heavy-duty engine coolants for a facility such as the New Jersey Department of Transportation garage in Ewing, New Jersey. he specific recycling evaluated is b...

  15. EVALUATION OF FILTRATION AND DISTILLATION METHODS FOR RECYCLING AUTOMOTIVE COOLANT.

    Science.gov (United States)

    This evaluation addresses the product quality, waste reduction, and economic issues involved in recycling automotive and heavy-duty engine coolants at a New Jersey Department of Transportation garage. The specific recycling units evaluated are based on the technologies of filtrat...

  16. Fuels, Lubricants, and Coolants. FOS: Fundamentals of Service.

    Science.gov (United States)

    John Deere Co., Moline, IL.

    This manual on fuels, lubricants, and coolants is one of a series of power mechanics tests and visual aids on automotive and off-the-road agricultural and construction equipment. Materials present basic information with illustrations for use by vocational students and teachers as well as shop servicemen and laymen. Focusing on fuels, the first of…

  17. Integral coolant channels supply made by melt-out method

    Science.gov (United States)

    Escher, W. J. D.

    1964-01-01

    Melt-out method of constructing strong, pressure-tight fluid coolant channels for chambers is accomplished by cementing pins to the surface and by depositing a melt-out material on the surface followed by two layers of epoxy-resin impregnated glass fibers. The structure is heated to melt out the low-melting alloy.

  18. EVALUATION OF FILTRATION AND DISTILLATION METHODS FOR RECYCLING AUTOMOTIVE COOLANT.

    Science.gov (United States)

    This evaluation addresses the product quality, waste reduction, and economic issues involved in recycling automotive and heavy-duty engine coolants at a New Jersey Department of Transportation garage. The specific recycling units evaluated are based on the technologies of filtrat...

  19. AUTOMOTIVE AND HEAVY-DUTY ENGINE COOLANT RECYCLING BY DISTILLATION

    Science.gov (United States)

    This evaluation addresses the product quality, waste reduction, and economic issues involved in recycling automotive and heavy-duty engine coolants for a facility such as the New Jersey Department of Transportation garage in Ewing, New Jersey. he specific recycling evaluated is b...

  20. Correlation between Ni base alloys surface conditioning and cation release mitigation in primary coolant

    Energy Technology Data Exchange (ETDEWEB)

    Clauzel, M.; Guillodo, M.; Foucault, M. [AREVA NP SAS, Technical Centre, Le Creusot (France); Engler, N.; Chahma, F.; Brun, C. [AREVA NP SAS, Chemistry and Radiochemistry Group, Paris La Defense (France)

    2010-07-01

    The mastering of the reactor coolant system radioactive contamination is a real stake of performance for operating plants and new builds. The reduction of activated corrosion products deposited on RCS surfaces allows minimizing the global dose integrated by workers which supports the ALARA approach. Moreover, the contamination mastering limits the volumic activities in the primary coolant and thus optimizes the reactor shutdown duration and environment releases. The main contamination sources on PWR are due to Co-60 and Co-58 nuclides which come respectively Co-59 and Ni-58, naturally present in alloys used in the RCS. Co is naturally present as an impurity in alloys or as the main component of hardfacing materials (Stellites™). Ni is released mainly by SG tubes which represent the most important surface of the RCS. PWR steam generators (SG), due to the huge wetted surface are the main source of corrosion products release in the primary coolant circuit. As corrosion products may be transported throughout the whole circuit, activated in the core, and redeposited all over circuit surfaces, resulting in an increase of activity buildup, it is of primary importance to gain a better understanding of phenomenon leading to corrosion product release from SG tubes before setting up mitigation measures. Previous studies have shown that SG tubing made of the same material had different release rates. To find the origin of these discrepancies, investigations have been performed on tubes at the as-received state and after exposure to a nominal primary chemistry in titanium recirculating loop. These investigations highlighted the existence of a correlation between the inner surface metallurgical properties and the release of corrosion products in primary coolant. Oxide films formed in nominal primary chemistry are always protective, their morphology and their composition depending strongly on the geometrical, metallurgical and physico-chemical state of the surface on which they

  1. Firms vie to offer DOE a prize-winning recipe for cleanup

    Energy Technology Data Exchange (ETDEWEB)

    Powers, M.B.

    1994-04-25

    Eager to get the most bang for its waste cleanup bucks, the US Department of Energy is conducting its own version of the Pillsbury bake-off. DOE is pitting two environmental contractors, Rust International Corp. and Lockheed Environmental Systems and Technologies Co., against each other to come up with the prize-winning recipe for cleaning up some nasty waste problems.

  2. Shoreline clean-up methods : biological treatments

    Energy Technology Data Exchange (ETDEWEB)

    Massoura, S.T. [Oil Spill Response Limited, Southampton (United Kingdom)

    2009-07-01

    The cleanup of oil spills in shoreline environments is a challenging issue worldwide. Oil spills receive public and media attention, particularly in the event of a coastal impact. It is important to evaluate the efficiency and effectiveness of cleanup methods when defining the level of effort and consequences that are appropriate to remove or treat different types of oil on different shoreline substrates. Of the many studies that have compared different mechanical, chemical and biological treatments for their effectiveness on various types of oil, biological techniques have received the most attention. For that reason, this paper evaluated the effectiveness and effects of shoreline cleanup methods using biological techniques. It summarized data from field experiments and oil spill incidents, including the Exxon Valdez, Sea Empress, Prestige, Grand Eagle, Nakhodka, Guanabara Bay and various Gulf war oil spills. Five major shoreline types were examined, notably rocky intertidal, cobble/pebble/gravel, sand/mud, saltmarsh, and mangrove/sea-grass. The biological techniques that were addressed were nutrient enrichment, hydrocarbon-utilizing bacteria, vegetable oil biosolvents, plants, surf washing, oil-particle interactions and natural attenuation. The study considered the oil type, volume and fate of stranded oil, location of coastal materials, extent of pollution and the impact of biological techniques. The main factors that affect biodegradation of hydrocarbons are the volume, chemical composition and weathering state of the petroleum product as well as the temperature, oxygen availability of nutrients, water salinity, pH level, water content, and microorganisms in the shoreline environment. The interaction of these factors also affect the biodegradation of oil. It was concluded that understanding the fate of stranded oil can help in the development of techniques that improve the weathering and degradation of oil on complex shoreline substrates. 39 refs.

  3. Oil spill cleanup method and apparatus

    Energy Technology Data Exchange (ETDEWEB)

    Mayes, F.M.

    1980-06-24

    A method for removing oil from the surface of water where an oil spill has occurred, particularly in obstructed or shallow areas, which comprises partially surrounding a hovercraft with a floating oil-collecting barrier, there being no barrier at the front of the hovercraft, moving the oil-barrier-surrounded-hovercraft into oil contaminated water, and collecting oil gathered within the barrier behind the hovercraft through a suction line which carries the oil to a storage tank aboard the hovercraft. The invention also embodies the hovercraft adapted to effect an oil spill cleanup.

  4. Data center cooling system

    Energy Technology Data Exchange (ETDEWEB)

    Chainer, Timothy J; Dang, Hien P; Parida, Pritish R; Schultz, Mark D; Sharma, Arun

    2015-03-17

    A data center cooling system may include heat transfer equipment to cool a liquid coolant without vapor compression refrigeration, and the liquid coolant is used on a liquid cooled information technology equipment rack housed in the data center. The system may also include a controller-apparatus to regulate the liquid coolant flow to the liquid cooled information technology equipment rack through a range of liquid coolant flow values based upon information technology equipment temperature thresholds.

  5. Directly connected heat exchanger tube section and coolant-cooled structure

    Energy Technology Data Exchange (ETDEWEB)

    Chainer, Timothy J.; Coico, Patrick A.; Graybill, David P.; Iyengar, Madhusudan K.; Kamath, Vinod; Kochuparambil, Bejoy J.; Schmidt, Roger R.; Steinke, Mark E.

    2015-09-15

    A method is provided for fabricating a cooling apparatus for cooling an electronics rack, which includes an air-to-liquid heat exchanger, one or more coolant-cooled structures, and a tube. The heat exchanger is associated with the electronics rack and disposed to cool air passing through the rack, includes a plurality of coolant-carrying tube sections, each tube section having a coolant inlet and outlet, one of which is coupled in fluid communication with a coolant loop to facilitate flow of coolant through the tube section. The coolant-cooled structure(s) is in thermal contact with an electronic component(s) of the rack, and facilitates transfer of heat from the component(s) to the coolant. The tube connects in fluid communication one coolant-cooled structure and the other of the coolant inlet or outlet of the one tube section, and facilitates flow of coolant directly between that coolant-carrying tube section of the heat exchanger and the coolant-cooled structure.

  6. Numerical investigation on thermal striping conditions for a tee junction of LMFBR coolant pipes. 3. Investigation on diameter ratio between the coolant pipes

    Energy Technology Data Exchange (ETDEWEB)

    Muramatsu, Toshiharu [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1999-04-01

    This report presents numerical results on thermal striping characteristics at a tee junction of LMFBR coolant pipe, carried out using a direct numerical simulation code DINUS-3. In the numerical investigations, it was considered a tee junction system consisted of a main pipe (1.33 cm{sup I.D.}) with a 90deg elbow and a branch pipe having various inner diameters, and five diameter ratio conditions between both the pipes, i.e., (D{sub main}/D{sub branch}) = 1.0, 2.0, 3.0, 5.0 and 10.0. From the numerical investigations, the following characteristics were obtained: (1) Maximum sodium temperature fluctuation amplitude in the downstream region of the tee junction were decreased with increasing of the diameter ratio (decreasing of the branch pipe diameter). One of the main reasons for this behavior was considered to be that the affects of the branch pipe jet for the main pipe flows was decreased with decreasing of the branch pipe diameter. (2) Auto-power spectral density levels were decreased by the increasing of the diameter ratio. It was indicated that coolant mixing characteristics in the downstream region were controlled by locally random turbulence processes. (3) To suppress sodium temperature fluctuations in the downstream region, it is a suitable combination for larger velocity ratio and larger diameter ratio. (author)

  7. Review of State Soil Cleanup Levels for Dioxin (December 2009)

    Science.gov (United States)

    This final report summarizes a survey of state soil cleanup levels for dioxin and characterizes the science underlying these values. The objective of this project was to summarize existing state cleanup levels for dioxin in soil, together with their scientific bases where availa...

  8. Experimental approach to investigate the dynamics of mixing coolant flow in complex geometry using PIV and PLIF techniques

    Directory of Open Access Journals (Sweden)

    Hutli Ezddin

    2015-01-01

    Full Text Available The aim of this work is to investigate experimentally the increase of mixing phenomenon in a coolant flow in order to improve the heat transfer, the economical operation and the structural integrity of Light Water Reactors-Pressurized Water Reactors (LWRs-PWRs. Thus the parameters related to the heat transfer process in the system will be investigated. Data from a set of experiments, obtained by using high precision measurement techniques, Particle Image Velocimetry and Planar Laser-Induced Fluorescence (PIV and PLIF, respectively are to improve the basic understanding of turbulent mixing phenomenon and to provide data for CFD code validation. The coolant mixing phenomenon in the head part of a fuel assembly which includes spacer grids has been investigated (the fuel simulator has half-length of a VVER 440 reactor fuel. The two-dimensional velocity vector and temperature fields in the area of interest are obtained by PIV and PLIF technique, respectively. The measurements of the turbulent flow in the regular tube channel around the thermocouple proved that there is rotation and asymmetry in the coolant flow caused by the mixing grid and the geometrical asymmetry of the fuel bundle. Both PIV and PLIF results showed that at the level of the core exit thermocouple the coolant is homogeneous. The discrepancies that could exist between the outlet average temperature of the coolant and the temperature at in-core thermocouple were clarified. Results of the applied techniques showed that both of them can be used as good provider for data base and to validate CFD results.

  9. Hot gas clean-up with ceramic filter elements

    Energy Technology Data Exchange (ETDEWEB)

    Christ, A.; Gross, R.; Renz, U. [Rheinisch-Westfaelische Technische Hochschule, Aachen (Germany). Lehrstuhl fuer Waermeuebertragung und Klimatechnik

    1998-12-31

    Hot gas cleanup is necessary during the combined cycle combustion of coal, ceramic filters are frequently used for filtration. Pressure and velocity measurements inside the filter elements during pulse cleaning of a single ceramic candle filter element were carried out. The experimental set-up is described and the results for filter element cleaning at {var_theta} = 25,500 and 900{degree}C and cleaning pressures of P{sub B} = 1.4, 1.9 and 2.8 bar are presented and discussed. Numerical simulations of filter element cleaning for the experimental conditions are presented and discussed as well. Regarding the good agreement of experimental results with numerical predictions it is proven that numerical simulations of the back-pulse cleaning should be employed as design tool for filter cleaning systems. 32 figs., 2 tabs.

  10. Hybrid method for numerical modelling of LWR coolant chemistry

    Science.gov (United States)

    Swiatla-Wojcik, Dorota

    2016-10-01

    A comprehensive approach is proposed to model radiation chemistry of the cooling water under exposure to neutron and gamma radiation at 300 °C. It covers diffusion-kinetic processes in radiation tracks and secondary reactions in the bulk coolant. Steady-state concentrations of the radiolytic products have been assessed based on the simulated time dependent concentration profiles. The principal reactions contributing to the formation of H2, O2 and H2O2 were indicated. Simulation was carried out depending on the amount of extra hydrogen dissolved in the coolant to reduce concentration of corrosive agents. High sensitivity to the rate of reaction H+H2O=OH+H2 is shown and discussed.

  11. Effect of coolant inhibitors on AZ91D

    Institute of Scientific and Technical Information of China (English)

    I.M. Baghni; WU Yinshun; ZHANG Wei; LI Jiuqing

    2004-01-01

    The inhibition effects of sodium vanadate along with inorganic coolant inhibitors were examined on corrosion of AZ91D in ASTM D1384-80 corrosive water by polarization measurements. The galvanic corrosion of AZ91D coupled to 3003, 6063, and 356 Al alloys were also tested. An effective combination of inhibitors containing (but not limited to) sodium vanadate, silicate, and nitrate was proposed for inhibition of AZ91D and prevention of galvanic corrosion.

  12. Crack stability analysis of low alloy steel primary coolant pipe

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, T.; Kameyama, M. [Kansai Electric Power Company, Osaka (Japan); Urabe, Y. [Mitsubishi Heavy Industries, Ltd., Takasago (Japan)] [and others

    1997-04-01

    At present, cast duplex stainless steel has been used for the primary coolant piping of PWRs in Japan and joints of dissimilar material have been applied for welding to reactor vessels and steam generators. For the primary coolant piping of the next APWR plants, application of low alloy steel that results in designing main loops with the same material is being studied. It means that there is no need to weld low alloy steel with stainless steel and that makes it possible to reduce the welding length. Attenuation of Ultra Sonic Wave Intensity is lower for low alloy steel than for stainless steel and they have advantageous inspection characteristics. In addition to that, the thermal expansion rate is smaller for low alloy steel than for stainless steel. In consideration of the above features of low alloy steel, the overall reliability of primary coolant piping is expected to be improved. Therefore, for the evaluation of crack stability of low alloy steel piping to be applied for primary loops, elastic-plastic future mechanics analysis was performed by means of a three-dimensioned FEM. The evaluation results for the low alloy steel pipings show that cracks will not grow into unstable fractures under maximum design load conditions, even when such a circumferential crack is assumed to be 6 times the size of the wall thickness.

  13. Liquid metal reactor development -Studies on safety measure of LMR coolant

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Sung Tae; Choi, Yoon Dong; Park, Jin Hoh; Kwon, Sun Kil; Choi, Jong Hyun; Cho, Byung Ryul; Kim, Tae Joon; Kwon, Sang Woon; Jung, Kyung Chae; Kim, Byung Hoh; Hong, Soon Bok; Jung, Ji Yung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    A study on the safety measures of LMR coolant showed the results as follows; 1. LMR coolant safety measure. A. Analysis and improvement of sodium fire code. B. Analysis of sodium fire phenomena. 2. Sodium fire aerosol characteristics. It was carried out conceptual design and basic design for sodium fire facility of medium size composed of sodium supply tank, sodium reactor vessel, sodium fire aerosol filter system and scrubbing column, and drain tank etc. 3. Sodium purification technology. A. Construction of calibration loop. (1) Design of sodium loop for the calibration of the equipment. (2) Construction of sodium loop including test equipments and other components. B. Na-analysis technology. (1) Oxygen concentration determination by the wet method. (2) Cover gas purification preliminary experiment. 4. The characteristics of sodium-water reaction. A. Analysis of the micro and small leak phenomena. (1) Manufacture of the micro-leak test apparatus. B. Analysis of large leak events. (1) Development of preliminary code for analysis of initial spike pressure. (2) Sample calculation and comparison with previous works. C. Development of test facility for large leak event evaluation. (1) Conceptional and basic design for the water and sodium-water test facility. D. Technology development for water leak detection system. (1) Investigations for the characteristics of active acoustic detection system. (2) Testing of the characteristics of hydrogen leak detection system. 171 figs, 29 tabs, 3 refs. (Author).

  14. Performance of water and diluted ethylene glycol as coolants for electronic cooling

    Directory of Open Access Journals (Sweden)

    M. Gayatri,

    2015-05-01

    Full Text Available As the number of transistors increases with new generation of microprocessor chips, the power draw and heat load to dissipate during operation increases. As a result of increasing the heat loads and heat fluxes the Conventional cooling technologies such as fan, heat sinks are unable to absorb and heat transfer excess heat dissipated by these new microprocessor. So, new technologies are needed to improve the heat removal capacity. In the present work single phase liquid cooling system with mini channel is analyzed and experimentally investigated. Mini channels are chosen as to provide higher heat transfer co-efficient than conventional channel. Copper pipes of 0.36 mm diameter are taken to fabricate heat sink and heat exchanger. A pump is used to circulate the fluid through heat sink and heat exchanger. A solid heated aluminium block to simulate heat generated electronic component is used and electrical input is supplied to the heated aluminium block and cooling system is placed over the heated block. The performance of the cooling system is analyzed from the experimental data obtained. It is experimentally observed that the mini channel liquid cooling system with water as a coolant has better performance than diluted ethylene glycol as coolant at different flow rates. The surface temperature of the heated aluminium block with convective heat transfer co-efficient is observed

  15. Nanoporous polystyrene fibers for oil spill cleanup.

    Science.gov (United States)

    Lin, Jinyou; Shang, Yanwei; Ding, Bin; Yang, Jianmao; Yu, Jianyong; Al-Deyab, Salem S

    2012-02-01

    The development of oil sorbents with high sorption capacity, low cost, scalable fabrication, and high selectivity is of great significance for water environmental protection, especially for oil spillage on seawater. In this work, we report nanoporous polystyrene (PS) fibers prepared via a one-step electrospinning process used as oil sorbents for oil spill cleanup. The oleophilic-hydrophobic PS oil sorbent with highly porous structures shows a motor oil sorption capacity of 113.87 g/g, approximately 3-4 times that of natural sorbents and nonwoven polypropylene fibrous mats. Additionally, the sorbents also exhibit a relatively high sorption capacity for edible oils, such as bean oil (111.80 g/g) and sunflower seed oil (96.89 g/g). The oil sorption mechanism of the PS sorbent and the sorption kinetics were investigated. Our nanoporous material has great potential for use in wastewater treatment, oil accident remediation and environmental protection.

  16. Waste Heat Recovery from the Advanced Test Reactor Secondary Coolant Loop

    Energy Technology Data Exchange (ETDEWEB)

    Donna Post Guillen

    2012-11-01

    This study investigated the feasibility of using a waste heat recovery system (WHRS) to recover heat from the Advanced Test Reactor (ATR) secondary coolant system (SCS). This heat would be used to preheat air for space heating of the reactor building, thus reducing energy consumption, carbon footprint, and energy costs. Currently, the waste heat from the reactor is rejected to the atmosphere via a four-cell, induced-draft cooling tower. Potential energy and cost savings are 929 kW and $285K/yr. The WHRS would extract a tertiary coolant stream from the SCS loop and pump it to a new plate and frame heat exchanger, from which the heat would be transferred to a glycol loop for preheating outdoor air supplied to the heating and ventilation system. The use of glycol was proposed to avoid the freezing issues that plagued and ultimately caused the failure of a WHRS installed at the ATR in the 1980s. This study assessed the potential installation of a new WHRS for technical, logistical, and economic feasibility.

  17. DESIK final report. Energy wise design and regulation of the secondary side of indirect cooling systems with natural coolants; DESIK slutrapportering. Energirigtig design og regulering af sekundaersiden pae indirekte koeleanlaeg med naturlige koelemidler

    Energy Technology Data Exchange (ETDEWEB)

    Jakobsen, Arne

    2006-02-15

    The project's aim was to produce knowledge and tools to facilitate the process of implementing energy efficient secondary cooling systems, or merely to avoid overconsumption of energy on account of insufficient relevant professional background. The project has been communicated as a PC tool, which can be ordered from aj(commercial at)ipu.dk. Project focus has been on some general aspects of secondary systems as well as two scopes of application: supermarket cooling systems and air conditioning of office buildings. (BA)

  18. Effect of Coolant Temperature and Mass Flow on Film Cooling of Turbine Blades

    Science.gov (United States)

    Garg, Vijay K.; Gaugler, Raymond E.

    1997-01-01

    A three-dimensional Navier Stokes code has been used to study the effect of coolant temperature, and coolant to mainstream mass flow ratio on the adiabatic effectiveness of a film-cooled turbine blade. The blade chosen is the VKI rotor with six rows of cooling holes including three rows on the shower head. The mainstream is akin to that under real engine conditions with stagnation temperature = 1900 K and stagnation pressure = 3 MPa. Generally, the adiabatic effectiveness is lower for a higher coolant temperature due to nonlinear effects via the compressibility of air. However, over the suction side of shower-head holes, the effectiveness is higher for a higher coolant temperature than that for a lower coolant temperature when the coolant to mainstream mass flow ratio is 5% or more. For a fixed coolant temperature, the effectiveness passes through a minima on the suction side of shower-head holes as the coolant to mainstream mass flow, ratio increases, while on the pressure side of shower-head holes, the effectiveness decreases with increase in coolant mass flow due to coolant jet lift-off. In all cases, the adiabatic effectiveness is highly three-dimensional.

  19. Tritium research laboratory cleanup and transition project final report

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, A.J.

    1997-02-01

    This Tritium Research Laboratory Cleanup and Transition Project Final Report provides a high-level summary of this project`s multidimensional accomplishments. Throughout this report references are provided for in-depth information concerning the various topical areas. Project related records also offer solutions to many of the technical and or administrative challenges that such a cleanup effort requires. These documents and the experience obtained during this effort are valuable resources to the DOE, which has more than 1200 other process contaminated facilities awaiting cleanup and reapplication or demolition.

  20. Experimental approach to investigate the dynamics of mixing coolant flow in complex geometry using PIV and PLIF techniques

    OpenAIRE

    Hutli Ezddin; Gottlasz Valer; Tar Dániel; Ezsol Gyorgy; Baranyai Gabor

    2015-01-01

    The aim of this work is to investigate experimentally the increase of mixing phenomenon in a coolant flow in order to improve the heat transfer, the economical operation and the structural integrity of Light Water Reactors-Pressurized Water Reactors (LWRs-PWRs). Thus the parameters related to the heat transfer process in the system will be investigated. Data from a set of experiments, obtained by using high precision measurement techniques, Particle Image V...

  1. 高温气冷堆数字化规程系统的总体设计%Study on General Design of the Digital Procedure System of High Temperature Gas Coolant Reactor

    Institute of Scientific and Technical Information of China (English)

    冯静阁; 黄晓津

    2015-01-01

    This thesis analyzed digital procedures of the domestic and international nuclear power plant, combined with the own characteristics of the high temperature reactor demonstra-tion project (HTR-PM). The general design of digital procedure system (DPS) of the demonstra-tion project was studied by comprehensively utilizing the multi-disciplinary knowledge including the digital control system (DCS) technology, human factors engineering theory, and computer technology. It provides basis for further study of the system.%文章分析了国内外核电站数字化规程研究现状,结合高温气冷堆示范工程(H T R-P M)的自身特点,综合运用数字化仪控系统(D C S)技术、人因工程理论、计算机技术等多学科知识,对示范工程数字化规程系统的总体设计进行了研究,为该系统的进一步研发提供了基础.

  2. Performance Investigation of Automobile Radiator Operated with ZnFe2O4 Nano Fluid based Coolant

    Directory of Open Access Journals (Sweden)

    Tripathi Ajay

    2015-01-01

    Full Text Available The cooling system of an Automobile plays an important role in its performance, consists of two main parts, known as radiator and fan. Improving thermal efficiency of engine leads to increase the engine's performance, decline the fuel consumption and decrease the pollution emissions. Water and ethylene glycol as conventional coolants have been widely used in radiators of an automotive industry for many years. These heat transfer fluids offer low thermal conductivity. With the advancement of nanotechnology, the new generation of heat transfer fluids called, “nanofluids” have been developed and researchers found that these fluids offer higher thermal conductivity compared to that of conventional coolants. This study focused on the preparation of Zinc based nanofluids (ZnFe2O4 using chemical co-precipitation method and its application in an automotive cooling system along with mixture of ethylene glycol and water (50:50. Relevant input data, nanofluids properties and empirical correlations were obtained from literatures to investigate the heat transfer enhancement of an automotive car radiator operated with nano fluid-based coolants. It was observed that, overall heat transfer coefficient and heat transfer rate in engine cooling system increased with the usage of nanofluids (with ethylene glycol the base-fluid compared to ethylene glycol (i.e. base-fluid alone. It is observed that, about 78% of heat transfer enhancement could be achieved with the addition of 1% ZnFe2O4 particles in a base fluid at the Reynolds number of 84.4x103 and 39.5x103 for air and coolant respectively

  3. Reactor coolant pump shaft seal stability during station blackout

    Energy Technology Data Exchange (ETDEWEB)

    Rhodes, D B; Hill, R C; Wensel, R G

    1987-05-01

    Results are presented from an investigation into the behavior of Reactor Coolant Pump shaft seals during a potential station blackout (loss of all ac power) at a nuclear power plant. The investigation assumes loss of cooling to the seals and focuses on the effect of high temperature on polymer seals located in the shaft seal assemblies, and the identification of parameters having the most influence on overall hydraulic seal performance. Predicted seal failure thresholds are presented for a range of station blackout conditions and shaft seal geometries.

  4. Numerical Investigation of Urea Freezing and Melting Characteristics Using Coolant Heater

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Seung Yeop; Kim, Nam Il; Kim, Man Young [Chounbuk Nat' l Univ., Jeonju (Korea, Republic of); Park, Yun Beom [Jeju College of Technology, Jeju (Korea, Republic of)

    2013-08-15

    UREA-SCR technology is known as one of the powerful NOx reduction systems for vehicles as well as stationary applications. For its consistent and reliable operation in vehicle applications, however, the freezing and melting of the urea solution in cold environments have to be resolved. In this study, therefore, a numerical study of three-dimensional unsteady problems was analyzed to understand the urea freezing and heating phenomena and heat transfer characteristics in terms of urea liquid volume fraction, temperature profiles, and phase change behavior in urea solutions with time by using the commercial software Fluent 6.3. As a result, it was found that the freezing phenomenon proceeds with a phase change from the tank wall to the center, whereas the melting phenomenon occurs faster in the upper part of the storage tank by natural convection and in the adjacent part of the coolant pipe than in other parts. Furthermore, approximately 190s were required to obtain 1a of urea solution using a 4-coiled coolant heater under conditions of 70 .deg. C and 200 L/h.

  5. Cleanup Verification Package for the 300-8 Waste Site

    Energy Technology Data Exchange (ETDEWEB)

    J. M. Capron

    2005-11-07

    This cleanup verification package documents completion of remedial action for the 300-8 waste site. This waste site was formerly used to stage scrap metal from the 300 Area in support of a program to recycle aluminum.

  6. Engineering Forum Issue Paper: Online Hazardous Waste Cleanup Technical Resources

    Science.gov (United States)

    This issue paper is intended to give the reader examples of some online technical resources that can assist with hazardous waste cleanups in the Superfund, Resource Conservation and Recovery Act (RCRA), and Brownfields programs.

  7. EPA proposes St. Regis Paper Co. Cleanup Plan, accepting comments

    Science.gov (United States)

    For Immediate Release No.16-OPA006 EPA proposes St. Regis Paper Co. Cleanup Plan; accepting comments CHICAGO (March 30, 2016) -- U.S. Environmental Protection Agency is issuing a proposed plan to clean up soil contamin

  8. 二代改进型核电厂严重事故下一回路卸压时机敏感性研究%Sensitivity Analysis on Time of Reactor Coolant System Depressurization under Severe Accident for Generation II+ Nuclear Power Plants

    Institute of Scientific and Technical Information of China (English)

    种毅敏; 杨志义; 石雪垚; 张佳佳; 李春; 倪曼; 徐雨婷

    2015-01-01

    Reactor coolant system (RCS)depressurization is necessary measure for nuclear power plant to mitigate the severe accident,as well as a significant part of the severe accident management guidelines (SAMG).Difference may exist on the time of RCS depressurization in different NPPs.In this paper,based on MAAP4,the sensitivity analysis of time to implement RCS depressurization is performed.A typical integrated computer program,and different effects on mitigation of severe accident are compared.The simulation scenario is typical drill situation of generation II + NPPs,and conclusions can be reference for similar NPPs to implement severe accident management strategy.%一回路卸压是核电厂缓解严重事故的必要手段,也是严重事故管理导则(SAMG)的重要内容,国内核电厂严重事故管理中对一回路卸压的要求并不相同,本文基于典型二代改进型核电厂 SAMG 演练的场景,使用一体化计算程序 MAAP4,对一回路卸压时机进行敏感性分析,比较不同卸压时机对缓解严重事故效果的影响,所给出的结论可为相同类型核电厂制定严重事故管理策略时提供参考。

  9. Hot particulate removal and desulfurization results from the METC integrated gasification and hot gas cleanup facility

    Energy Technology Data Exchange (ETDEWEB)

    Rockey, J.M.

    1995-06-01

    The Morgantown Energy Technology Center (METC) is conducting experimental testing using a 10-inch diameter fluid-bed gasifier (FBG) and modular hot gas cleanup rig (MGCR) to develop advanced methods for removing contaminants in hot coal gasifier gas streams for commercial development of integrated gasification combined-cycle (IGCC) power systems. The program focus is on hot gas particulate removal and desulfurization technologies that match the temperatures and pressures of the gasifier, cleanup system, and power generator. The purpose of this poster is to present the program objectives and results of the work conducted in cooperation with industrial users and vendors to meet the vision for IGCC of reducing the capital cost per kilowatt to $1050 and increasing the plant efficiency to 52% by the year 2010.

  10. Simulation of 3D Flow in Turbine Blade Rows including the Effects of Coolant Ejection

    Institute of Scientific and Technical Information of China (English)

    Jian-Jun LIU; Bai-Tao AN; Yun-Tao ZENG

    2008-01-01

    This paper describes the numerical simulation of three-dimensional viscous flows in air-cooled turbine blade rows with the effects of coolant ejection. A TVD Navier-Stokes flow solver incorporated with Baldwin-Lomax turbulence model and multi-grid convergence acceleration algorithm are used for the simulation. The influences of coolant ejection on the main flow are accounted by volumetric coolant source terms. Numerical results for a four-stage turbine are presented and discussed.

  11. Experimental Investigation of Coolant Boiling in a Half-Heated Circular Tube - Final CRADA Report

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Wenhua [Argonne National Lab. (ANL), Argonne, IL (United States); Singh, Dileep [Argonne National Lab. (ANL), Argonne, IL (United States); France, David M. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-11-01

    Coolant subcooled boiling in the cylinder head regions of heavy-duty vehicle engines is unavoidable at high thermal loads due to high metal temperatures. However, theoretical, numerical, and experimental studies of coolant subcooled flow boiling under these specific application conditions are generally lacking in the engineering literature. The objective of this project was to provide such much-needed information, including the coolant subcooled flow boiling characteristics and the corresponding heat transfer coefficients, through experimental investigations.

  12. Central cortical cleanup and zonular deficiency

    Science.gov (United States)

    Mansour, Ahmad M; Antonios, Rafic S; Ahmed, Iqbal Ike K

    2016-01-01

    Background Complete removal of the cortex has been advocated to prevent posterior capsular opacification but carries the risk of zonular dehiscence, hence there is a need for a safe maximal cortical cleanup technique in eyes with severe diffuse zonulopathy in subjects above age 90. Methods We used bimanual central cortical cleaning by elevating central fibers and aspirating them toward the periphery. Peripheral cortical fibers were removed passively only when they became loose due to copious irrigation. A one-piece foldable implant was inserted without a capsular tension ring. Postoperative corticosteroid drops were used. Results This technique was safely performed in a dozen eyes with severe pseudo-exfoliation or brunescent cataract with weak zonules. Posterior capsular rupture, iritis, vitreous loss, and lens subluxation were not observed. Moderate capsular phimosis occurred but with maintained central vision. Conclusion The dogma of “complete cortical cleanup” in severe zonulopathy needs to be revisited in favor of a clear visual axis with maximal preservation of the damaged zonules. This technique is ideal in patients above age 90 where posterior capsular opacification and late dislocation of intraocular lens–capsule bag complex are unlikely to occur until several years postoperatively. PMID:27784979

  13. Central cortical cleanup and zonular deficiency

    Directory of Open Access Journals (Sweden)

    Mansour AM

    2016-10-01

    Full Text Available Ahmad M Mansour,1,2 Rafic S Antonios,1 Iqbal Ike K Ahmed3 1Department of Ophthalmology, American University of Beirut, Beirut, Lebanon; 2Department of Ophthalmology, Rafic Hariri University Hospital, Beirut, Lebanon; 3Department of Ophthalmology, University of Toronto, Toronto, ON, Canada Background: Complete removal of the cortex has been advocated to prevent posterior capsular opacification but carries the risk of zonular dehiscence, hence there is a need for a safe maximal cortical cleanup technique in eyes with severe diffuse zonulopathy in subjects above age 90. Methods: We used bimanual central cortical cleaning by elevating central fibers and aspirating them toward the periphery. Peripheral cortical fibers were removed passively only when they became loose due to copious irrigation. A one-piece foldable implant was inserted without a capsular tension ring. Postoperative corticosteroid drops were used. Results: This technique was safely performed in a dozen eyes with severe pseudo-exfoliation or brunescent cataract with weak zonules. Posterior capsular rupture, iritis, vitreous loss, and lens subluxation were not observed. Moderate capsular phimosis occurred but with maintained central vision. Conclusion: The dogma of “complete cortical cleanup” in severe zonulopathy needs to be revisited in favor of a clear visual axis with maximal preservation of the damaged zonules. This technique is ideal in patients above age 90 where posterior capsular opacification and late dislocation of intraocular lens–capsule bag complex are unlikely to occur until several years postoperatively. Keywords: brunescent cataract, cortex aspiration, phacoemulsification, pseudo-exfoliation, weak zonules

  14. Cleanup Verification Package for the 118-F-3, Minor Construction Burial Ground

    Energy Technology Data Exchange (ETDEWEB)

    M. J. Appel

    2007-01-04

    This cleanup verification package documents completion of remedial action for the 118-F-3, Minor Construction Burial Ground waste site. This site was an open field covered with cobbles, with no vegetation growing on the surface. The site received irradiated reactor parts that were removed during conversion of the 105-F Reactor from the Liquid 3X to the Ball 3X Project safety systems and received mostly vertical safety rod thimbles and step plugs.

  15. Aqueous Nanofluid as a Two-Phase Coolant for PWR

    Directory of Open Access Journals (Sweden)

    Pavel N. Alekseev

    2012-01-01

    Full Text Available Density fluctuations in liquid water consist of two topological kinds of instant molecular clusters. The dense ones have helical hydrogen bonds and the nondense ones are tetrahedral clusters with ice-like hydrogen bonds of water molecules. Helical ordering of protons in the dense water clusters can participate in coherent vibrations. The ramified interface of such incompatible structural elements induces clustering impurities in any aqueous solution. These additives can enhance a heat transfer of water as a two-phase coolant for PWR due to natural forming of nanoparticles with a thermal conductivity higher than water. The aqueous nanofluid as a new condensed matter has a great potential for cooling applications. It is a mixture of liquid water and dispersed phase of extremely fine quasi-solid particles usually less than 50 nm in size with the high thermal conductivity. An alternative approach is the formation of gaseous (oxygen or hydrogen nanoparticles in density fluctuations of water. It is possible to obtain stable nanobubbles that can considerably exceed the molecular solubility of oxygen (hydrogen in water. Such a nanofluid can convert the liquid water in the nonstoichiometric state and change its reduction-oxidation (RedOx potential similarly to adding oxidants (or antioxidants for applying 2D water chemistry to aqueous coolant.

  16. Fitness for service assessment of coolant channels of Indian PHWRs

    Science.gov (United States)

    Sinha, R. K.; Sinha, S. K.; Madhusoodanan, K.

    2008-12-01

    A typical coolant channel assembly of pressurised heavy water reactors mainly consists of pressure tube, calandria tube, garter spring spacers, all made of zirconium alloys and end fittings made of SS 403. The pressure tube is rolled at both its ends to the end fittings and is located concentrically inside the calandria tube with the help of garter spring spacers. Pressure tube houses the fuel bundles, which are cooled by means of pressurised heavy water. It, thus, operates under the environment of high pressure and temperature (typically 10 MPa and 573 K), and fast neutron flux (typically 3 × 10 17 n/m 2 s, E > 1 MeV neutrons). Under this operating environment, the material of the pressure tube undergoes degradation over a period of time, and eventually needs to be assessed for fitness for continued operation, without jeopardising the safety of the reactor. The other components of the coolant channel assembly, which are inaccessible for any in-service inspection, are assessed for their fitness, whenever a pressure tube is removed for either surveillance purpose or any other reasons. This paper, while describing the latest developments taking place to address the issue of fitness for service of the Zr-2.5 wt% Nb pressure tubes, also dwells briefly upon the developments taken place, to address the issues of life management and extension of zircaloy-2 pressure tubes in the earlier generation of Indian pressurised heavy water reactors.

  17. Fitness for service assessment of coolant channels of Indian PHWRs

    Energy Technology Data Exchange (ETDEWEB)

    Sinha, R.K.; Sinha, S.K. [Reactor Design and Development Group, Bhabha Atomic Research Centre, Mumbai 400 085 (India); Madhusoodanan, K. [Reactor Design and Development Group, Bhabha Atomic Research Centre, Mumbai 400 085 (India)], E-mail: kmadhu@barc.gov.in

    2008-12-15

    A typical coolant channel assembly of pressurised heavy water reactors mainly consists of pressure tube, calandria tube, garter spring spacers, all made of zirconium alloys and end fittings made of SS 403. The pressure tube is rolled at both its ends to the end fittings and is located concentrically inside the calandria tube with the help of garter spring spacers. Pressure tube houses the fuel bundles, which are cooled by means of pressurised heavy water. It, thus, operates under the environment of high pressure and temperature (typically 10 MPa and 573 K), and fast neutron flux (typically 3 x 10{sup 17} n/m{sup 2} s, E > 1 MeV neutrons). Under this operating environment, the material of the pressure tube undergoes degradation over a period of time, and eventually needs to be assessed for fitness for continued operation, without jeopardising the safety of the reactor. The other components of the coolant channel assembly, which are inaccessible for any in-service inspection, are assessed for their fitness, whenever a pressure tube is removed for either surveillance purpose or any other reasons. This paper, while describing the latest developments taking place to address the issue of fitness for service of the Zr-2.5 wt% Nb pressure tubes, also dwells briefly upon the developments taken place, to address the issues of life management and extension of zircaloy-2 pressure tubes in the earlier generation of Indian pressurised heavy water reactors.

  18. A Preliminary Study of Banana Stem Juice as a Plant-Based Coagulant for Treatment of Spent Coolant Wastewater

    Directory of Open Access Journals (Sweden)

    Habsah Alwi

    2013-01-01

    Full Text Available The effectiveness of banana stem juice as a natural coagulant for treatment of spent coolant wastewater was investigated . Three main parameters were studied, namely, chemical oxygen demand (COD, suspended solids (SSs, and turbidity of effluent. Coagulation experiments using jar test were performed with a flocculation system where the effects of spent coolant wastewater pH as well as banana stem juice dosage on coagulation effectiveness were examined. The highest recorded COD, SS, and turbidity removal percentages by banana stem juice were 80.1%, 88.6%, and 98.5%, respectively, observed for effluent at pH 7 using 90 mL dosage. The inulin concentration in the banana stem was examined to be 1.22016 mg/mL. It could be concluded that banana stem juice showed tremendous potential as a natural coagulant for water treatment purposes and could be applied in the pretreatment stage of Malaysian spent coolant wastewater prior to secondary treatment.

  19. Investigation on two-phase critical flow for loss-of-coolant accident of pressurized water reactor

    Institute of Scientific and Technical Information of China (English)

    1996-01-01

    The previous investigations were mainly conducted under the condition of low pressure,however,the steam-water specific volume and the interphase evaporation rate in high pressure are much different from those in low pressure,Therefore,the new experimental and theoretical investigation are performed in Xi'an Jiaotong University.The investigation results could be directly applied to the analysis of loss-of -coolant accident for pressurized water reacor.The system transition characteristics of cold leg and hot leg break loss-of -coolant tests are described for convective circulation test loop.Two types of loss-of-coolant accident are identified for :hot leg” break,while three types for “cold leg”break and the effect parameters on the break geometries.Tests indicate that the mass flow rate with convergent-divergent nozzle reaches the maximum value among the different break sections at the same inlet fluid condition because the fluid separation does not occur.A wall surface cavity nucleation model is developed for prediction of the critical mass flow rate with water flowing in convergentdivergent nozzles.

  20. Analysis of a hypothetical loss of coolant accident in a Konvoi type NPP by GASFLOW and COCOSYS

    Energy Technology Data Exchange (ETDEWEB)

    Benz, Stefan; Royl, Peter [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Band, Sebastian [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Koeln (Germany)

    2013-07-01

    The 3D computational fluid dynamics code GASFLOW and the German containment code system COCOSYS, which is based on a lumped-parameter approach, are used to simulate the hydrogen-air-steam distribution and hydrogen mitigation in a Konvoi type nuclear power plant in a postulated hypothetical core melt accident. A break in a coolant loop and the subsequent loss of the coolant causes a strong heat-up of the core. As a consequence hydrogen is produced by oxidation of cladding tubes. The residual steam and the produced hydrogen are released into the containment through the break in the coolant loop. Without suitable counter measures, sensitive mixtures can build up with a combustion potential which could threaten the integrity of the containment. A model of a Konvoi type nuclear power plant which is equipped with passive autocatalytic recombiners is used to simulate such accident scenario. COCOSYS allows comprehensive simulation of all relevant processes of severe accidents, whereas GASFLOW is primarily designed to simulate the distribution of steam and hydrogen within the containment. This paper presents the comparison of GASFLOW and COCOSYS simulation results for the in-vessel phase of the selected accident. (orig.)

  1. Radiological assessment in case of an incident at the hot cells clean-up

    Directory of Open Access Journals (Sweden)

    Dragolici Cristian A.

    2014-01-01

    Full Text Available The clean-up and decontamination of the hot cells will be performed in the second phase of the WWR-S research reactor decommissioning. Identification of possible incidents or accidents is the key element in radiological assessment and prevention. As major incident it was considered a fire burst that occurred during the progress of the clean-up operations. The postulated incident has, as a consequence, thick smoke generation from the burned radioactive material and the dispersion of this material in the environment through the technological ventilation system and the evacuation chimney. From the performed analysis it can be seen that in the case of an incident to the reactor hot cells, an operator engaged in intervention operations could take an effective dose of 5.29 Sv per event, coming from both external and internal exposure. Such an incident, if it happens, would be classified of level 3 on the INES scale.

  2. Safety analysis of the US dual coolant liquid lead lithium ITER test blanket module

    Science.gov (United States)

    Merrill, Brad; Reyes, Susana; Sawan, Mohamed; Wong, Clement

    2007-07-01

    The US is proposing a prototype of a dual coolant liquid lead-lithium DEMO blanket concept for testing in the International Thermonuclear Experimental Reactor (ITER) as an ITER test blanket module (TBM). Because safety considerations are an integral part of the design process to ensure that this TBM does not adversely impact the safety of ITER, a safety assessment has been conducted for this TBM and its ancillary systems as requested by the ITER project. Four events were selected by the ITER international team (IT) to address specific reactor safety concerns, such as vaccum vessel (VV) pressurization, confinement building pressure build-up, TBM decay heat removal capability, tritium and activation products release from the TBM system and hydrogen and heat production from chemical reactions. This paper summarizes the results of this safety assessment conducted with the MELCOR computer code.

  3. SIMMER-III Analyses of Local Fuel-Coolant Interactions in a Simulated Molten Fuel Pool: Effect of Coolant Quantity

    Directory of Open Access Journals (Sweden)

    Songbai Cheng

    2015-01-01

    Full Text Available Studies on local fuel-coolant interactions (FCI in a molten pool are important for the analyses of severe accidents that could occur for sodium-cooled fast reactors (SFRs. To clarify the mechanisms underlying this interaction, in recent years, several experimental tests, with comparatively larger difference in coolant volumes, were conducted at the Japan Atomic Energy Agency by delivering a given quantity of water into a molten pool formed with a low-melting-point alloy. In this study, to further understand this interaction, interaction characteristics including the pressure buildup as well as mechanical energy release and its conversion efficiency are investigated using the SIMMER-III, an advanced fast reactor safety analysis code. It is found that the SIMMER-III code not only reasonably simulates the transient pressure and temperature variations during local FCIs, but also supports the limited tendency of pressurization and resultant mechanical energy release as observed from experiments when the volume of water delivered into the pool increases. The performed analyses also suggest that the most probable reason leading to such limited tendency should be primarily due to an isolation effect of vapor bubbles generated at the water-melt interface.

  4. CRADA opportunities with METC`s gasification and hot gas cleanup facility

    Energy Technology Data Exchange (ETDEWEB)

    Galloway, E.N.; Rockey, J.M.; Tucker, M.S.

    1995-06-01

    Opportunities exist for Cooperative Research and Development Agreements (CRADA) at the Morgantown Energy Technology Center (METC) to support commercialization of IGCC power systems. METC operates an integrated gasifier and hot gas cleanup facility for the development of gasification and hot gas cleanup technologies. The objective of our program is to gather performance data on gasifier operation, particulate removal, desulfurization and regeneration technologies. Additionally, slip streams are provided for developing various technologies such as; alkali monitoring, particulate measuring, chloride removal, and contaminate recovery processes. METC`s 10-inch diameter air blown Fluid Bed Gasifier (FBG) provides 300 lb/hr of coal gas at 1100{degrees}F and 425 psig. The particulate laden gas is transported to METC`s Modular Gas Cleanup Rig (MGCR). The gas pressure is reduced to 285 psig before being fed into a candle filter vessel. The candle filter vessel houses four candle filters and multiple test coupons. The particulate free gas is then desulfurized in a sorbent reactor. Starting in 1996 the MGCR system will be able to regenerate the sorbent in the same vessel.

  5. Cooling system with automated seasonal freeze protection

    Energy Technology Data Exchange (ETDEWEB)

    Campbell, Levi A.; Chu, Richard C.; David, Milnes P.; Ellsworth, Jr., Michael J.; Iyengar, Madhusudan K.; Simons, Robert E.; Singh, Prabjit; Zhang, Jing

    2016-05-24

    An automated multi-fluid cooling system and method are provided for cooling an electronic component(s). The cooling system includes a coolant loop, a coolant tank, multiple valves, and a controller. The coolant loop is at least partially exposed to outdoor ambient air temperature(s) during normal operation, and the coolant tank includes first and second reservoirs containing first and second fluids, respectively. The first fluid freezes at a lower temperature than the second, the second fluid has superior cooling properties compared with the first, and the two fluids are soluble. The multiple valves are controllable to selectively couple the first or second fluid into the coolant in the coolant loop, wherein the coolant includes at least the second fluid. The controller automatically controls the valves to vary first fluid concentration level in the coolant loop based on historical, current, or anticipated outdoor air ambient temperature(s) for a time of year.

  6. IED Cleanup: A Cooperative Classroom Robotics Challenge--The Benefits and Execution of a Cooperative Classroom Robotics Challenge

    Science.gov (United States)

    Piotrowski, Mark; Kressly, Rich

    2009-01-01

    This article describes a cooperative classroom robotics challenge named "IED Cleanup". This classroom challenge was created to incorporate a humanitarian project with the use of a robotics design system in order to remove simulated IEDs (Improvised Explosive Devices) to a detonation zone within a specified amount of time. Throughout the activity,…

  7. Reactor coolant pump shaft seal behavior during station blackout

    Energy Technology Data Exchange (ETDEWEB)

    Kittmer, C.A.; Wensel, R.G.; Rhodes, D.B.; Metcalfe, R.; Cotnam, B.M.; Gentili, H.; Mings, W.J.

    1985-04-01

    A testing program designed to provide fundamental information pertaining to the behavior of reactor coolant pump (RCP) shaft seals during a postulated nuclear power plant station blackout has been completed. One seal assembly, utilizing both hydrodynamic and hydrostatic types of seals, was modeled and tested. Extrusion tests were conducted to determine if seal materials could withstand predicted temperatures and pressures. A taper-face seal model was tested for seal stability under conditions when leaking water flashes to steam across the seal face. Test information was then used as the basis for a station blackout analysis. Test results indicate a potential problem with an elastomer material used for O-rings by a pump vendor; that vendor is considering a change in material specification. Test results also indicate a need for further research on the generic issue of RCP seal integrity and its possible consideration for designation as an unresolved safety issue.

  8. Actively controlling coolant-cooled cold plate configuration

    Energy Technology Data Exchange (ETDEWEB)

    Chainer, Timothy J.; Parida, Pritish R.

    2016-04-26

    Cooling apparatuses are provided to facilitate active control of thermal and fluid dynamic performance of a coolant-cooled cold plate. The cooling apparatus includes the cold plate and a controller. The cold plate couples to one or more electronic components to be cooled, and includes an adjustable physical configuration. The controller dynamically varies the adjustable physical configuration of the cold plate based on a monitored variable associated with the cold plate or the electronic component(s) being cooled by the cold plate. By dynamically varying the physical configuration, the thermal and fluid dynamic performance of the cold plate are adjusted to, for example, optimally cool the electronic component(s), and at the same time, reduce cooling power consumption used in cooling the electronic component(s). The physical configuration can be adjusted by providing one or more adjustable plates within the cold plate, the positioning of which may be adjusted based on the monitored variable.

  9. Leak rate analysis of the Westinghouse Reactor Coolant Pump

    Energy Technology Data Exchange (ETDEWEB)

    Boardman, T.; Jeanmougin, N.; Lofaro, R.; Prevost, J.

    1985-07-01

    An independent analysis was performed by ETEC to determine what the seal leakage rates would be for the Westinghouse Reactor Coolant Pump (RCP) during a postulated station blackout resulting from loss of ac electric power. The object of the study was to determine leakage rates for the following conditions: Case 1: All three seals function. Case 2: No. 1 seal fails open while Nos. 2 and 3 seals function. Case 3: All three seals fail open. The ETEC analysis confirmed Westinghouse calculations on RCP seal performance for the conditions investigated. The leak rates predicted by ETEC were slightly lower than those predicted by Westinghouse for each of the three cases as summarized below. Case 1: ETEC predicted 19.6 gpm, Westinghouse predicted 21.1 gpm. Case 2: ETEC predicted 64.7 gpm, Westinghouse predicted 75.6 gpm. Case 3: ETEC predicted 422 gpm, Westinghouse predicted 480 gpm. 3 refs., 22 figs., 6 tabs.

  10. Diesel engine coolant analysis, new application for established instrumentation

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, D.P.; Lukas, M.; Lynch, B.K. [Spectro Incorporated, Littleton, MA (United States)

    1997-12-31

    Rotating disk electrode (RDE) arc emission spectrometers are user` many commercial, industrial and military laboratories throughout the world to analyze millions of oil and fuel samples each year. In fact, RDE spectrometers have been used exclusively for oil and fuel analysis for so long that it has nearly been forgotten by most practitioners that when RDE spectrometers were first introduced more than 40 years ago, they were routinely used for aqueous samples as well. This presentation reviews early methods of aqueous sample analysis using RDE technology. This presentation also describes recent work to calibrate an RDE spectrometer for both water samples and for engine coolant samples which are a mixture of approximately 50 % water and 50 % ethylene or propylene glycol. Limits of detection determined for aqueous standards are comparable to limits of detection for oil standards. Repeatability of aqueous samples is comparable to the repeatability achieved for oil samples. A comparison of results for coolant samples measured by both inductively coupled plasma (ICP) and rotating disk electrode (RDE) spectrometers is presented. Not surprisingly, RDE results are significantly higher for samples containing particles larger than a few micrometers. Although limits of detection for aqueous samples are not as low as can be achieved using the more modern ICP spectrometric method or the more cumbersome atomic absorption (AA) method, this presentation suggests that RDE spectrometers may be appropriate for certain types of aqueous samples in situations where the more sensitive ICP or AA spectrometers and the laboratory environment and skilled personnel needed for them to operate are not conveniently available. (orig.) 4 refs.

  11. Efficiency of different protocols for enamel clean-up after bracket debonding: an in vitro study

    Science.gov (United States)

    Sigilião, Lara Carvalho Freitas; Marquezan, Mariana; Elias, Carlos Nelson; Ruellas, Antônio Carlos; Sant'Anna, Eduardo Franzotti

    2015-01-01

    Objective: This study aimed to assess the efficiency of six protocols for cleaning-up tooth enamel after bracket debonding. Methods: A total of 60 premolars were divided into six groups, according to the tools used for clean-up: 12-blade bur at low speed (G12L), 12-blade bur at high speed (G12H), 30-blade bur at low speed (G30L), DU10CO ORTHO polisher (GDU), Renew System (GR) and Diagloss polisher (GD). Mean roughness (Ra) and mean roughness depth (Rz) of enamel surface were analyzed with a profilometer. Paired t-test was used to assess Ra and Rz before and after enamel clean-up. ANOVA/Tukey tests were used for intergroup comparison. The duration of removal procedures was recorded. The association between time and variation in enamel roughness (∆Ra, ∆Rz) were evaluated by Pearson's correlation test. Enamel topography was assessed by scanning electron microscopy (SEM). Results: In Groups G12L and G12H, original enamel roughness did not change significantly. In Groups G30L, GDU, GR and GD, a smoother surface (p < 0.05) was found after clean-up. In Groups G30L and GD, the protocols used were more time-consuming than those used in the other groups. Negative and moderate correlation was observed between time and (∆Ra, ∆Rz); Ra and (∆Ra, ∆Rz); Rz (r = - 0.445, r = - 0.475, p < 0.01). Conclusion: All enamel clean-up protocols were efficient because they did not result in increased surface roughness. The longer the time spent performing the protocol, the lower the surface roughness. PMID:26560825

  12. Efficiency of different protocols for enamel clean-up after bracket debonding: an in vitro study

    Directory of Open Access Journals (Sweden)

    Lara Carvalho Freitas Sigilião

    2015-10-01

    Full Text Available Objective: This study aimed to assess the efficiency of six protocols for cleaning-up tooth enamel after bracket debonding.Methods:A total of 60 premolars were divided into six groups, according to the tools used for clean-up: 12-blade bur at low speed (G12L, 12-blade bur at high speed (G12H, 30-blade bur at low speed (G30L, DU10CO ORTHO polisher (GDU, Renew System (GR and Diagloss polisher (GD. Mean roughness (Ra and mean roughness depth (Rz of enamel surface were analyzed with a profilometer. Paired t-test was used to assess Ra and Rz before and after enamel clean-up. ANOVA/Tukey tests were used for intergroup comparison. The duration of removal procedures was recorded. The association between time and variation in enamel roughness (∆Ra, ∆Rz were evaluated by Pearson's correlation test. Enamel topography was assessed by scanning electron microscopy (SEM.Results:In Groups G12L and G12H, original enamel roughness did not change significantly. In Groups G30L, GDU, GR and GD, a smoother surface (p < 0.05 was found after clean-up. In Groups G30L and GD, the protocols used were more time-consuming than those used in the other groups. Negative and moderate correlation was observed between time and (∆Ra, ∆Rz; Ra and (∆Ra, ∆Rz; Rz (r = - 0.445, r = - 0.475, p < 0.01.Conclusion:All enamel clean-up protocols were efficient because they did not result in increased surface roughness. The longer the time spent performing the protocol, the lower the surface roughness.

  13. Validation of advanced NSSS simulator model for loss-of-coolant accidents

    Energy Technology Data Exchange (ETDEWEB)

    Kao, S.P.; Chang, S.K.; Huang, H.C. [Nuclear Training Branch, Northeast Utilities, Waterford, CT (United States)

    1995-09-01

    The replacement of the NSSS (Nuclear Steam Supply System) model on the Millstone 2 full-scope simulator has significantly increased its fidelity to simulate adverse conditions in the RCS. The new simulator NSSS model is a real-time derivative of the Nuclear Plant Analyzer by ABB. The thermal-hydraulic model is a five-equation, non-homogeneous model for water, steam, and non-condensible gases. The neutronic model is a three-dimensional nodal diffusion model. In order to certify the new NSSS model for operator training, an extensive validation effort has been performed by benchmarking the model performance against RELAP5/MOD2. This paper presents the validation results for the cases of small-and large-break loss-of-coolant accidents (LOCA). Detailed comparisons in the phenomena of reflux-condensation, phase separation, and two-phase natural circulation are discussed.

  14. Discrete element method study of fuel relocation and dispersal during loss-of-coolant accidents

    Science.gov (United States)

    Govers, K.; Verwerft, M.

    2016-09-01

    The fuel fragmentation, relocation and dispersal (FFRD) during LOCA transients today retain the attention of the nuclear safety community. The fine fragmentation observed at high burnup may, indeed, affect the Emergency Core Cooling System performance: accumulation of fuel debris in the cladding ballooned zone leads to a redistribution of the temperature profile, while dispersal of debris might lead to coolant blockage or to debris circulation through the primary circuit. This work presents a contribution, by discrete element method, towards a mechanistic description of the various stages of FFRD. The fuel fragments are described as a set of interacting particles, behaving as a granular medium. The model shows qualitative and quantitative agreement with experimental observations, such as the packing efficiency in the balloon, which is shown to stabilize at about 55%. The model is then applied to study fuel dispersal, for which experimental parametric studies are both difficult and expensive.

  15. Influence of coolant temperature and pressure on destructive forces at fuel failure in the NSRR experiment

    Energy Technology Data Exchange (ETDEWEB)

    Kusagaya, Kazuyuki [Global Nuclear Fuel - Japan Co., Ltd., Yokosuka, Kanagawa (Japan); Sugiyama, Tomoyuki; Nakamura, Takehiko; Uetsuka, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-01-01

    In order to design a new experimental capsule to be used in the NSRR (Nuclear Safety Research Reactor) experiment with the temperature and pressure conditions in a typical commercial BWR, coolant temperature and pressure influence is estimated for destructive forces during fuel rod failure in the experiment simulating reactivity-initiated accident (RIA). Considering steam property dependence on temperature and pressure, it is qualitatively shown that the destructive forces in the BWR operation condition are smaller than those in the room temperature and atmospheric pressure condition. Water column velocity, which determines impact by water hammer, is further investigated quantitatively by modeling the experimental system and water hammer phenomenon. As a result, the maximum velocity of the water column in the BWR operation conditions is calculated to be only about 10% of that in the room temperature and atmospheric pressure condition. (author)

  16. Vapor pressures of mixtures of CFC-114 with the potential replacement coolants C{sub 4}F{sub 10} and c-C{sub 4}F{sub 8}

    Energy Technology Data Exchange (ETDEWEB)

    Trowbridge, L.D. [Oak Ridge K-25 Site, TN (United States); Otey, M.G. [Paducah Gaseous Diffusion Plant, KY (United States)

    1994-09-01

    The U.S. Enrichment Corporation`s production of isotopically enriched uranium depends solely on two plants which utilize the gaseous diffusion process. This process uses large quantities of CFC-114 as an evaporative coolant. CFC-114, however, will be phased out of production at the end of 1995 due to its potential to deplete stratospheric ozone. A search has been underway for substitutes for a number of years. The initial search (1988-89) for an ozone-friendly, commercially available, chemically compatible substitute yielded two candidates, FC-c318 (c-C{sub 4}F{sub 8}) and FC-3110 (C{sub 4}F{sub 10}). The intended mode of replacing coolant was to stage the new coolant into independent subsystems of the plants, so that some systems would continue to operate on CFC-114, and an increasing number would operate on the new coolant. During that changeover process, the possibility of coolant mixing arises in variety of scenarios. This work was intended to generate sufficient experimental information to be able to predict the vapor pressure of coolant mixtures over the range of operating conditions likely to be found in the diffusion plants. Specifically, vapor pressures were measured over the temperature range 322 to 355 K (120{degrees}F to 180{degrees}F) and over the full range of mole fractions for binary mixtures of CFC-114 with FC-3110, and of CFC-114 with FC-c318.

  17. Debonding and adhesive remnant cleanup: an in vitro comparison of bond quality, adhesive remnant cleanup, and orthodontic acceptance of a flash-free product.

    Science.gov (United States)

    Grünheid, Thorsten; Sudit, Geoffrey N; Larson, Brent E

    2015-10-01

    A new flash-free adhesive promises to eliminate the need to clean up excess adhesive upon orthodontic bracket bonding. This study evaluated this new adhesive with regard to microleakage at the enamel-bracket interface, amount of adhesive remaining on the tooth after bracket debonding, time required for adhesive remnant cleanup, and clinical practitioners' preference in comparison to a conventional adhesive. A total of 184 bovine incisors were bonded with ceramic brackets using either the flash-free adhesive (APC Flash-Free Adhesive Coated Appliance System, 3M Unitek [3M], Monrovia, California, USA) or a conventional adhesive (APCII Adhesive Coated Appliance System, 3M). Twenty-four of the teeth were scanned using microcomputed tomography to quantify microleakage into the adhesive layer. Twenty orthodontists debonded the brackets, removed the remaining adhesive, and then completed a survey regarding their preference for one of the two adhesives. The adhesive remnant was quantified and the time required for its removal recorded. Differences between the adhesives were tested for statistical significance. For both adhesives, the microleakage was minimal with no significant differences between the two adhesives. The adhesive remnant was significantly larger for the flash-free adhesive, whereas there was no significant difference in adhesive cleanup time. Fourteen out of the 20 orthodontists preferred the flash-free adhesive over the conventional adhesive. In vitro testing cannot replicate the actual clinical situation during in vivo debonding. With regard to bond quality and adhesive remnant cleanup, the new flash-free adhesive performs just as well as the conventional adhesive, and, of the two products, is the one preferred by most orthodontists. © The Author 2014. Published by Oxford University Press on behalf of the European Orthodontic Society. All rights reserved. For permissions, please email: journals.permissions@oup.com.

  18. Hot gas cleanup test facility for gasification and pressurized combustion project. Quarterly report, October--December 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-02-01

    The objective of this project is to evaluate hot gas particle control technologies using coal-derived gas streams. This will entail the design, construction, installation, and use of a flexible test facility which can operate under realistic gasification and combustion conditions. The conceptual design of the facility was extended to include a within scope, phased expansion of the existing Hot Gas Cleanup Test Facility Cooperative Agreement to also address systems integration issues of hot particulate removal in advanced coal-based power generation systems. This expansion included the consideration of the following modules at the test facility in addition to the original Transport Reactor gas source and Hot Gas Cleanup Units: Carbonizer/pressurized circulating fluidized bed gas source; hot gas cleanup units to mate to all gas streams; combustion gas turbine; and fuel cell and associated gas treatment. This expansion to the Hot Gas Cleanup Test Facility is herein referred to as the Power Systems Development Facility (PSDF). The major emphasis during this reporting period was continuing the detailed design of the facility towards completion and integrating the balance-of-plant processes and particulate control devices (PCDs) into the structural and process designs. Substantial progress in construction activities was achieved during this quarter.

  19. Dosimetric impact evaluation of primary coolant chemistry of the internal tritium breeding cycle of a fusion reactor DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Velarde, M. [Instituto de Fusion Nuclear (DENIM), ETSII, Universidad Politecnica Madrid UPM, J. Gutierrez Abascal 2, Madrid 28006 (Spain); Sedano, L. A. [Asociacion Euratom-Ciematpara Fusion, Av. Complutense 22, 28040 Madrid (Spain); Perlado, J. M. [Instituto de Fusion Nuclear (DENIM), ETSII, Universidad Politecnica Madrid UPM, J. Gutierrez Abascal 2, Madrid 28006 (Spain)

    2008-07-15

    Tritium will be responsible for a large fraction of the environmental impact of the first generation of DT fusion reactors. Today, the efforts of conceptual development of the tritium cycle for DEMO are mainly centred in the so called Inner Breeding Tritium Cycle, conceived as guarantee of reactor fuel self-sufficiency. The EU Fusion Programme develops for the short term of fusion power technology two breeding blanket conceptual designs both helium cooled. One uses Li-ceramic material (HCPB, Helium-Cooled Pebble Bed) and the other a liquid metal eutectic alloy (Pb15.7Li) (HCLL, Helium-Cooled Lithium Lead). Both are Li-6 enriched materials. At a proper scale designs will be tested as Test Blanket Modules in ITER. The tritium cycles linked to both blanket concepts are similar, with some different characteristics. The tritium is recovered from the He purge gas in the case of HCPB, and directly from the breeding alloy through a carrier gas in HCLL. For a 3 GWth self-sufficient fusion reactor the tritium breeding need is few hundred grams of tritium per day. Safety and environmental impact are today the top priority design criteria. Dose impact limits should determine the key margins and parameters in its conception. Today, transfer from the cycle to the environment is conservatively assumed to be operating in a 1-enclosure scheme through the tritium plant power conversion system (intermediate heat exchangers and helium blowers). Tritium loss is caused by HT and T{sub 2} permeation and simultaneous primary coolant leakage through steam generators. Primary coolant chemistry appears to be the most natural way to control tritium permeation from the breeder into primary coolant and from primary coolant through SG by H{sub 2} tritium flux isotopic swamping or steel (EUROFER/INCOLOY) oxidation. A primary coolant chemistry optimization is proposed. Dynamic flow process diagrams of tritium fluxes are developed ad-hoc and coupled with tritiated effluents dose impact evaluations

  20. Implications of the KONVERGENCE Model for Difficult Cleanup Decisions

    Energy Technology Data Exchange (ETDEWEB)

    Piet, Steven James; Dakins, Maxine Ellen; Gibson, Patrick Lavern; Joe, Jeffrey Clark; Kerr, Thomas A; Nitschke, Robert Leon

    2002-08-04

    Abstract—Some cleanup decisions, such as cleanup of intractable contaminated sites or disposal of spent nuclear fuel, have proven difficult to make. Such decisions face high resistance to agreement from stakeholders possibly because they do not trust the decision makers, view the consequences of being wrong as too high, etc. Our project’s goal is to improve sciencebased cleanup decision-making. This includes diagnosing intractable situations, as a step to identifying a path toward sustainable solutions. Companion papers describe the underlying philosophy of the KONVERGENCE Model for Sustainable Decisions,1 and the overall framework and process steps.2 Where knowledge, values, and resources converge (the K, V, and R in KONVERGENCE), you will find a sustainable decision – a decision that works over time. For intractable cases, serious consideration of the adaptable class of alternatives is warranted – if properly implemented and packaged.

  1. Deriving cleanup guidelines for radionuclides at Brookhaven National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Meinhold, A.F.; Morris, S.C.; Dionne, B.; Moskowitz, P.D.

    1997-01-01

    Past activities at Brookhaven National Laboratory (BNL) resulted in soil and groundwater contamination. As a result, BNL was designated a Superfund site under the Comprehensive Environmental Response Compensation and Liability Act (CERCLA). BNL`s Office of Environmental Restoration (OER) is overseeing environmental restoration activities at the Laboratory. With the exception of radium, there are no regulations or guidelines to establish cleanup guidelines for radionuclides in soils at BNL. BNL must derive radionuclide soil cleanup guidelines for a number of Operable Units (OUs) and Areas of Concern (AOCs). These guidelines are required by DOE under a proposed regulation for radiation protection of public health and the environment as well as to satisfy the requirements of CERCLA. The objective of this report is to propose a standard approach to deriving risk-based cleanup guidelines for radionuclides in soil at BNL. Implementation of the approach is briefly discussed.

  2. Joint venture to apply biotechnology to waste cleanup

    Energy Technology Data Exchange (ETDEWEB)

    1988-11-01

    ENSR Corporation and Celgene Corporation have announced the signing of a letter of intent to form ENSR-Celgene, a joint venture that will develop and apply biotechnology to the cleanup of hazardous wastes. ENSR-Celgene will perform all bioremediation projects contracted by ENSR or Celgene, with technical support provided by both ENSR and Celgene. ENSR recently completed a successful demonstration project using bioremediation to cleanup petrochemical sludges in a seven-acre lagoon at a Superfund site. The EPA has not approved this technology for the site's final remediation. Bioremediation technology is expected to save responsible parties $75 million in cleanup costs and be completed nearly four years sooner than by using incineration, according to Zoch.

  3. Use of microPCM fluids as enhanced liquid coolants in automotive EV and HEV vehicles. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Mulligan, James C.; Gould, Richard D.

    2001-10-31

    Proof-of-concept experiments using a specific microPCM fluid that potentially can have an impact on the thermal management of automotive EV and HEV systems have been conducted. Samples of nominally 20-micron diameter microencapsulated octacosane and glycol/water coolant were prepared for testing. The melting/freezing characteristics of the fluid, as well as the viscosity, were determined. A bench scale pumped-loop thermal system was used to determine heat transfer coefficients and wall temperatures in the source heat exchanged. Comparisons were made which illustrate the enhancements of thermal performance, reductions of pumping power, and increases of heat transfer which occur with the microPCM fluid.

  4. Biotechnologies for Marine Oil Spill Cleanup: Indissoluble Ties with Microorganisms

    KAUST Repository

    Mapelli, Francesca

    2017-05-13

    The ubiquitous exploitation of petroleum hydrocarbons (HCs) has been accompanied by accidental spills and chronic pollution in marine ecosystems, including the deep ocean. Physicochemical technologies are available for oil spill cleanup, but HCs must ultimately be mineralized by microorganisms. How environmental factors drive the assembly and activity of HC-degrading microbial communities remains unknown, limiting our capacity to integrate microorganism-based cleanup strategies with current physicochemical remediation technologies. In this review, we summarize recent findings about microbial physiology, metabolism and ecology and describe how microbes can be exploited to create improved biotechnological solutions to clean up marine surface and deep waters, sediments and beaches.

  5. Simulating experimental investigation on the safety of nuclear heating reactor in loss-of-coolant accidents

    Science.gov (United States)

    Xu, Zhanjie

    1996-12-01

    The 5MW low temperature nuclear heating reactor (NHR-5) is a new and advanced type of nuclear reactor developed by Institute of Nuclear Energy Technology (INET) of Tsinghua University of China in 1989. Its main loop is a thermal-hydraulic system with natural circulation. This paper studies the safety of NHR under the condition of loss-of-coolant accidents (LOCAs) by means of simulant experiments. First, the background and necessity of the experiments are presented, then the experimental system, including the thermal-hydraulic system and the data collection system, and similarity criteria are introduced. Up to now, the discharge experiments with the residual heating power (20% rated heating power) have been carried out on the experimental system. The system parameters including circulation flow rate, system pressure, system temperature, void fraction, discharge mass and so on have been recorded and analyzed. Based on the results of the experiments, the conclusions are shown as follos: on the whole, the reactor is safe under the condition of LOCAs, but the thermal vacillations resulting from the vibration of the circulation flow rate are disadvantageous to the internal parts of the reactor core.

  6. Integrity of the reactor coolant boundary of the European pressurized water reactor (EPR)

    Energy Technology Data Exchange (ETDEWEB)

    Goetsch, D.; Bieniussa, K.; Schulz, H.; Jalouneix, J.

    1997-04-01

    This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branching pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications.

  7. NONUNIFORMITIES OF TWO-PHASE COOLANT DISTRIBUTION IN A HEAT GENERATING PARTICLES BED

    Directory of Open Access Journals (Sweden)

    V. V. Sorokin

    2014-01-01

    Full Text Available Sufficient atomic power generation safety increase may be done with microfuel adapting to reactor plants with water coolant. Microfuel particle is a millimeter size grain containing fission material core in a protecting coverage. The coverage protects fuel contact with coolant and provides isolation of fission products inside. Well thermophysical properties of microfuel bed in a direct contact with water coolant excludes fuel overheating when accidents. Microfuel use was suggested for a VVER, а direct flow reactor for superheat steam generation, a reactor with neutron spectra adjustment by the steam partial content varying in the coolant.Nonuniformities of two-phase coolant distribution in a heat generating particles bed are predicted by calculations in this text. The one is due to multiple-valuedness of pressure drop across the bed on the steam quality dependency. The nonuniformity decreases with flow rate and particle size growths absolute pressure diminishing while porosity effect is weak. The worse case is for pressure quality of order of one. Some pure steam filled pores appears parallel to steam water mixture filled pores, latter steam quality is less than the mean of the bed. Considering this regime for the direct flow reactor for superheat steam generation we predict some water drops at the exit flow. The two-phase coolant filtration with subcooled water feed is unstable to strong disturbance effects are found. Uniformity of two-phase coolant distribution is worse than for one-phase in the same radial type reactor.

  8. Assessment of Candidate Molten Salt Coolants for the NGNP/NHI Heat-Transfer Loop

    Energy Technology Data Exchange (ETDEWEB)

    Williams, D. F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2006-06-30

    This report provides an assessment of candidate salts proposed as the coolant for the loop that shuttles heat from the Next Generation Nuclear Plant (NGNP) to the Nuclear Hydrogen Initiative (NHI) hydrogen-production plant. The physical properties most relevant for coolant service were reviewed, and key chemical factors that influence material compatibility were also analyzed for the purpose of screening candidate salts. A preliminary assessment of the cost of the raw materials required to produce the coolant is also presented. Salts that are suitable for use as the primary coolant in a high-temperature nuclear reactor were previously analyzed. Some of the fluoride salts identified in the previous study are also appropriate for consideration as the secondary coolant in a heat-transfer loop; therefore, results from the previous report are used in this document. However, alternative coolant salts (i.e., chlorides and fluoroborates) that were not considered in the previous report should be considered for service in the heat-transfer loop. These alternative coolants are considered in this report.

  9. Process-information definition for evaluation of gasification and gas-cleanup processes for use in molten-carbonate fuel-cell power plants. Task A topical report

    Energy Technology Data Exchange (ETDEWEB)

    Vidt, E.J.

    1981-11-01

    This report satisfies the requirements for DOE contract DE-AC21-81MC16220 to list coal gasifiers and gas cleanup systems suitable for supplying fuel to molten carbonate fuel cells (MCFC) in industrial and utility power plants. The process information and data necessary for this study were extracted from sources in the public domain, including reports from DOE, EPRI, and EPA; work sponsored in whole or in part by federal agencies; and from trade journals, MCFC developers, and manufacturers. The listings included data on the state of development, operating characteristics, effluents, and effectiveness of the gasifiers and coal gas cleanup systems, to the extent that such information is available in the public domain. Information available in the public domain on the effects of contaminants on MCFC performance and on the design constraints on heat recovery equipment used to adjust coal gas temperatures to levels appropriate for available cleanup systems was also provided. Cleanup systems not chosen by DOE's MCFC contractors, General Electric and United Technologies, Inc., for their MCFC power plant work, by virtue of the resource requirements of those systems for commercial development, were extensively characterized. Such characterization is included in Appendix B, principally for the hot gas cleanup processes listed therein. One of those processes, using zinc ferrite for coal gas desulfurization, is now under active development by METC and has the potential for effective use in MCFC power plants.

  10. Warm Cleanup of Coal-Derived Syngas: Multicontaminant Removal Process Demonstration

    Energy Technology Data Exchange (ETDEWEB)

    Spies, Kurt A.; Rainbolt, James E.; Li, Xiaohong S.; Braunberger, Beau; Li, Liyu; King, David L.; Dagle, Robert A.

    2017-02-15

    Warm cleanup of coal- or biomass-derived syngas requires sorbent and catalytic beds to protect downstream processes and catalysts from fouling. Sulfur is particularly harmful because even parts-per-million amounts are sufficient to poison downstream synthesis catalysts. Zinc oxide (ZnO) is a conventional sorbent for sulfur removal; however, its operational performance using real gasifier-derived syngas and in an integrated warm cleanup process is not well reported. In this paper, we report the optimal temperature for bulk desulfurization to be 450oC, while removal of sulfur to parts-per-billion levels requires a lower temperature of approximately 350oC. Under these conditions, we found that sulfur in the form of both hydrogen sulfide and carbonyl sulfide could be absorbed equally well using ZnO. For long-term operation, sorbent regeneration is desirable to minimize process costs. Over the course of five sulfidation and regeneration cycles, a ZnO bed lost about a third of its initial sulfur capacity, however sorbent capacity stabilized. Here, we also demonstrate, at the bench-scale, a process and materials used for warm cleanup of coal-derived syngas using five operations: 1) Na2CO3 for HCl removal, 2) regenerable ZnO beds for bulk sulfur removal, 3) a second ZnO bed for trace sulfur removal, 4) a Ni-Cu/C sorbent for multi-contaminant inorganic removal, and 5) a Ir-Ni/MgAl2O4 catalyst employed for ammonia decomposition and tar and light hydrocarbon steam reforming. Syngas cleanup was demonstrated through successful long-term performance of a poison-sensitive, Cu-based, water-gas-shift catalyst placed downstream of the cleanup process train. The tar reformer is an important and necessary operation with this particular gasification system; its inclusion was the difference between deactivating the water-gas catalyst with carbon deposition and successful 100-hour testing using 1 LPM of coal-derived syngas.

  11. Natural convection heat transfer characteristics of the molten metal pool with solidification by boiling coolant

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jae Seon; Suh, Kune Yull; Chung, Chang Hyun [Seoul National University, Seoul (Korea, Republic of); Paark, Rae Joon; Kim, Sang Baik [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    This paper presents results of experimental studies on the heat transfer and solidification of the molten metal pool with overlying coolant with boiling. The metal pool is heated from the bottom surface and coolant is injected onto the molten metal pool. Ad a result, the crust, which is a solidified layer, may form at the top of the molten metal pool. Heat transfer is accomplished by a conjugate mechanism, which consists of the natural convection of the molten metal pool, the conduction in the crust layer and the convective boiling heat transfer in the coolant. This work examines the crust formation and the heat transfer rate on the molten metal pool with boiling coolant. The simulant molten pool material is tin (Sn) with the melting temperature of 232 deg C. Demineralized water is used as the working coolant. The crust layer thickness was ostensibly varied by the heated bottom surface temperature of the test section, but not much affected by the coolant injection rate. The correlation between the Nusselt number and the Rayleigh number in the molten metal pool region of this study is compared against the crust formation experiment without coolant boiling and the literature correlations. The present experimental results are higher than those from the experiment without coolant boiling, but show general agreement with the Eckert correlation, with some deviations in the high and low ends of the Rayleigh number. This discrepancy is currently attributed to concurrent rapid boiling of the coolant on top of the metal layer. 10 refs., 4 figs., 1 tab. (Author)

  12. HANDBOOK ON THE BENEFITS, COSTS, AND IMPACTS OF LAND CLEANUP AND REUSE

    Science.gov (United States)

    Summarizes the theoretical and empirical literature addressing benefit-cost and impact assessment of the land cleanup and reuse scenario. When possible, recommendations are provided for conducting economic analysis of land cleanup and reuse sites and programs. The knowledge base ...

  13. Numerical investigation on thermal striping conditions for a tee junction of LMFBR coolant pipes. 1. Investigation on velocity ratio between the coolant pipes

    Energy Technology Data Exchange (ETDEWEB)

    Muramatsu, Toshiharu [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-02-01

    This report presents numerical results on thermal striping characteristics at a tee junction of LMFBR coolant pipe, carried out using a direct numerical simulation code DINUS-3. In the numerical investigations, it was considered a tee junction system consisted of a main pipe (1.33 cm{sup I.D.}) with a 90deg elbow and a branch pipe having same inner diameter to the main pipe, and five velocity ratio conditions between both the pipes, i.e., (V{sub main}/V{sub branch}) = 0.25; 0.5; 1.0; 2.0 and 4.0. From the numerical investigations the following characteristics were obtained: (1) Temperature fluctuations in the downstream region of the tee junction were formulated by lower frequency components (<7.0 Hz) due to the interactions between main pipe flows and jet flows from the branch pipe, and higher frequency components (>10.0 Hz) generated by the vortex released frequency from the outer edge of the branch pipe jet flows. (2) On the top plane of the main pipe, peak values of the temperature fluctuation amplitude was decreased with increasing flow velocity in the main pipe, and its position was shifted to downstream direction of the main pipe by the increase of the main pipe flow velocity. (3) On the bottom plane of the main pipe, contrary to (2), peak values of the temperature fluctuation amplitude was increased with increasing flow velocity in the main pipe. (author)

  14. Computation of Space Shuttle high-pressure cryogenic turbopump ball bearing two-phase coolant flow

    Science.gov (United States)

    Chen, Yen-Sen

    1990-01-01

    A homogeneous two-phase fluid flow model, implemented in a three-dimensional Navier-Stokes solver using computational fluid dynamics methodology is described. The application of the model to the analysis of the pump-end bearing coolant flow of the high-pressure oxygen turbopump of the Space Shuttle main engine is studied. Results indicate large boiling zones and hot spots near the ball/race contact points. The extent of the phase change of the liquid oxygen coolant flow due to the frictional and viscous heat fluxes near the contact areas has been investigated for the given inlet conditions of the coolant.

  15. Single-beam thermal lens measurement of thermal diffusivity of engine coolants

    Science.gov (United States)

    George, Nibu A.; Thomas, Nibu B.; Chacko, Kavya; T, Neethu V.; Hussain Moidu, Haroon; Piyush, K.; David, Nitheesh M.

    2015-04-01

    Automobile engine coolant liquids are commonly used for efficient heat transfer from the engine to the surroundings. In this work we have investigated the thermal diffusivity of various commonly available engine coolants in Indian automobile market. We have used single beam laser induced thermal lens technique for the measurements. Engine coolants are generally available in concentrated solution form and are recommended to use at specified dilution. We have investigated the samples in the entire recommended concentration range for the use in radiators. While some of the brands show an enhanced thermal diffusivity compared to pure water, others show slight decrease in thermal diffusivity.

  16. Cleanup at Los Alamos National Laboratory - the challenges - 9493

    Energy Technology Data Exchange (ETDEWEB)

    Stiger, Susan G [Los Alamos National Laboratory; Hargis, Kenneth M [Los Alamos National Laboratory; Graham, Michael J [Los Alamos National Laboratory; Rael, George J [NNSL/LASO

    2008-01-01

    This paper provides an overview of environmental cleanup at the Los Alamos National Laboratory (LANL) and some of the unique aspects and challenges. Cleanup of the 65-year old Department of Energy Laboratory is being conducted under a RCRA Consent Order with the State of New Mexico. This agreement is one of the most recent cleanup agreements signed in the DOE complex and was based on lessons learned at other DOE sites. A number of attributes create unique challenges for LANL cleanup -- the proximity to the community and pueblos, the site's topography and geology, and the nature of LANL's on-going missions. This overview paper will set the stage for other papers in this session, including papers that present: Plans to retrieve buried waste at Material Disposal Area B, across the street from oen of Los Alamos' commercial districts and the local newspaper; Progress to date and joint plans with WIPP for disposal of the remaining inventory of legacy transuranic waste; Reviews of both groundwater and surface water contamination and the factors complicating both characterization and remediation; Optimizing the disposal of low-level radioactive waste from ongoing LANL missions; A stakeholder environmental data transparency project (RACER), with full public access to all available information on contamination at LANL, and A description of the approach to waste processing cost recovery from the programs that generate hazardous and radioactive waste at LANL.

  17. Cleanup Verification Package for the 118-F-6 Burial Ground

    Energy Technology Data Exchange (ETDEWEB)

    H. M. Sulloway

    2008-10-02

    This cleanup verification package documents completion of remedial action for the 118-F-6 Burial Ground located in the 100-FR-2 Operable Unit of the 100-F Area on the Hanford Site. The trenches received waste from the 100-F Experimental Animal Farm, including animal manure, animal carcasses, laboratory waste, plastic, cardboard, metal, and concrete debris as well as a railroad tank car.

  18. Cleanup Verification Package for the 618-2 Burial Ground

    Energy Technology Data Exchange (ETDEWEB)

    W. S. Thompson

    2006-12-28

    This cleanup verification package documents completion of remedial action for the 618-2 Burial Ground, also referred to as Solid Waste Burial Ground No. 2; Burial Ground No. 2; 318-2; and Dry Waste Burial Site No. 2. This waste site was used primarily for the disposal of contaminated equipment, materials and laboratory waste from the 300 Area Facilities.

  19. Chemical and radiolytical characterization of perfluorocarbon fluids used as coolants for LHC experiments : radiolysis effects in perfluorohexane fluids.

    CERN Document Server

    Ilie, Soran; Teissandier, B; CERN. Geneva. TS Department

    2007-01-01

    Perfluorohexane fluids, used as coolants within High Energy Physics Detectors in the Large Hadrons Collider (LHC) at CERN, were irradiated using gammas 60Co and characterized using different analytical techniques. The aim of this work was the assessment of radiation induced effects as a function of the chemical nature of these fluids and their impurity content. Were evidenced the radioinduced polymers and acidity, as well as different chemical by-products. Purification tests and measurements were carried out on different irradiated fluid samples to assess the efficiency of such purification treatments in view of their re-use in the HEP detector cooling systems.

  20. PROGRESS & CHALLENGES IN CLEANUP OF HANFORDS TANK WASTES

    Energy Technology Data Exchange (ETDEWEB)

    HEWITT, W.M.; SCHEPENS, R.

    2006-01-23

    The River Protection Project (RPP), which is managed by the Department of Energy (DOE) Office of River Protection (ORP), is highly complex from technical, regulatory, legal, political, and logistical perspectives and is the largest ongoing environmental cleanup project in the world. Over the past three years, ORP has made significant advances in its planning and execution of the cleanup of the Hartford tank wastes. The 149 single-shell tanks (SSTs), 28 double-shell tanks (DSTs), and 60 miscellaneous underground storage tanks (MUSTs) at Hanford contain approximately 200,000 m{sup 3} (53 million gallons) of mixed radioactive wastes, some of which dates back to the first days of the Manhattan Project. The plan for treating and disposing of the waste stored in large underground tanks is to: (1) retrieve the waste, (2) treat the waste to separate it into high-level (sludge) and low-activity (supernatant) fractions, (3) remove key radionuclides (e.g., Cs-137, Sr-90, actinides) from the low-activity fraction to the maximum extent technically and economically practical, (4) immobilize both the high-level and low-activity waste fractions by vitrification, (5) interim store the high-level waste fraction for ultimate disposal off-site at the federal HLW repository, (6) dispose the low-activity fraction on-site in the Integrated Disposal Facility (IDF), and (7) close the waste management areas consisting of tanks, ancillary equipment, soils, and facilities. Design and construction of the Waste Treatment and Immobilization Plant (WTP), the cornerstone of the RPP, has progressed substantially despite challenges arising from new seismic information for the WTP site. We have looked closely at the waste and aligned our treatment and disposal approaches with the waste characteristics. For example, approximately 11,000 m{sup 3} (2-3 million gallons) of metal sludges in twenty tanks were not created during spent nuclear fuel reprocessing and have low fission product concentrations. We

  1. FILM-30: A Heat Transfer Properties Code for Water Coolant

    Energy Technology Data Exchange (ETDEWEB)

    MARSHALL, THERON D.

    2001-02-01

    A FORTRAN computer code has been written to calculate the heat transfer properties at the wetted perimeter of a coolant channel when provided the bulk water conditions. This computer code is titled FILM-30 and the code calculates its heat transfer properties by using the following correlations: (1) Sieder-Tate: forced convection, (2) Bergles-Rohsenow: onset to nucleate boiling, (3) Bergles-Rohsenow: partially developed nucleate boiling, (4) Araki: fully developed nucleate boiling, (5) Tong-75: critical heat flux (CHF), and (6) Marshall-98: transition boiling. FILM-30 produces output files that provide the heat flux and heat transfer coefficient at the wetted perimeter as a function of temperature. To validate FILM-30, the calculated heat transfer properties were used in finite element analyses to predict internal temperatures for a water-cooled copper mockup under one-sided heating from a rastered electron beam. These predicted temperatures were compared with the measured temperatures from the author's 1994 and 1998 heat transfer experiments. There was excellent agreement between the predicted and experimentally measured temperatures, which confirmed the accuracy of FILM-30 within the experimental range of the tests. FILM-30 can accurately predict the CHF and transition boiling regimes, which is an important advantage over current heat transfer codes. Consequently, FILM-30 is ideal for predicting heat transfer properties for applications that feature high heat fluxes produced by one-sided heating.

  2. Hot Gas Cleanup Test Facility for gasification and pressurized combustion. Quarterly report, October--December 1994

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-02-01

    The objective of this project is to evaluate hot gas particle control technologies using coal-derived gas streams. This will entail the design, construction, installation, and use of a flexible test facility which can operate under realistic gasification and combustion conditions. The major particulate control device issues to be addressed include the integration of the particulate control devices into coal utilization systems, on-line cleaning techniques, chemical and thermal degradation of components, fatigue or structural failures, blinding, collection efficiency as a function of particle size, and scale-up of particulate control systems to commercial size. The conceptual design of the facility was extended to include a within scope, phased expansion of the existing Hot Gas Cleanup Test Facility Cooperative Agreement to also address systems integration issues of hot particulate removal in advanced coal-based power generation systems. This expansion included the consideration of the following modules at the test facility in addition to the original Transport Reactor gas source and Hot Gas Cleanup Units: carbonizer/pressurized circulating fluidized bed gas source; hot gas cleanup units to mate to all gas streams; combustion gas turbine; and fuel cell and associated gas treatment. The major emphasis during this reporting period was continuing the detailed design of the facility and integrating the particulate control devices (PCDs) into structural and process designs. Substantial progress in underground construction activities was achieved during the quarter. Delivery and construction of coal handling and process structural steel began during the quarter. Delivery and construction of coal handling and process structural steel began during the quarter. MWK equipment at the grade level and the first tier are being set in the structure.

  3. Non-cancer morbidity among Estonian Chernobyl cleanup workers: a register-based cohort study.

    Science.gov (United States)

    Rahu, Kaja; Bromet, Evelyn J; Hakulinen, Timo; Auvinen, Anssi; Uusküla, Anneli; Rahu, Mati

    2014-05-14

    To examine non-cancer morbidity in the Estonian Chernobyl cleanup workers cohort compared with the population sample with special attention to radiation-related diseases and mental health disorders. Register-based cohort study. Estonia. An exposed cohort of 3680 men (cleanup workers) and an unexposed cohort of 7631 men (population sample) were followed from 2004 to 2012 through the Population Registry and Health Insurance Fund database. Morbidity in the exposed cohort compared with the unexposed controls was estimated in terms of rate ratio (RR) with 95% CIs using Poisson regression models. Elevated morbidity in the exposed cohort was found for diseases of the nervous system, digestive system, musculoskeletal system, ischaemic heart disease and for external causes. The most salient excess risk was observed for thyroid diseases (RR=1.69; 95% CI 1.38 to 2.07), intentional self-harm (RR=1.47; 95% CI 1.04 to 2.09) and selected alcohol-related diagnoses (RR=1.25; 95% CI 1.12 to 1.39). No increase in morbidity for stress reactions, depression, headaches or sleep disorders was detected. No obvious excess morbidity consistent with biological effects of radiation was seen in the exposed cohort, with the possible exception of benign thyroid diseases. Increased alcohol-induced morbidity may reflect alcohol abuse, and could underlie some of the higher morbidity rates. Mental disorders in the exposed cohort were probably under-reported. The future challenge will be to study mental and physical comorbidities in the Chernobyl cleanup workers cohort. Published by the BMJ Publishing Group Limited. For permission to use (where not already granted under a licence) please go to http://group.bmj.com/group/rights-licensing/permissions.

  4. Use of a PKZh-902 instrument for monitoring solid phases in an organic coolant

    Energy Technology Data Exchange (ETDEWEB)

    Gavrillin, A.I.; Gagarin, S.I.; Sokolov, V.E.; Zabelin, A.I.

    1986-11-01

    Preliminary tests have been performed with a PKZh-902 in checking feed and circulating coolant, and also in evaluating the performance of the cleaning devices. The test program involved determining the stability of the esnsor materials in ditolymethane, examining the effects of radioactive products and those of the optical characteristics of impurities in the first-loop coolant on the readings and errors of measurement, and checking the scope for using the instrument for continuous monitoring of particle concentrations. Results confirm that the PKZh-902 enables one to monitor the composition and concentration of the solid dispersed phase reliably and with adequate accuracy in the feed coolant. The use for monitoring the loop coolant requires additonal research.

  5. Study on diesel cylinder-head cooling using nanofluid coolant with jet impingement

    Directory of Open Access Journals (Sweden)

    Su Zhong-Gen

    2015-01-01

    Full Text Available To improve the heat-transfer performance of a diesel-engine cylinder head, nanofluid coolant as a new fluid was investigated, and jet impingement technology was then used to study on how to better improve heat-transfer coefficient at the nose bridge area in the diesel-engine cylinder head. Computational fluid dynamic simulation and experiments results demonstrated that using the same jet impingement parameters, the different volume shares of nanofluids showed better cooling effect than traditional coolant, but the good effect of the new cooling method was unsuitable for high volume share of nanofluid. At the same volume share of nanofluid, different jet impingement parameters such as jet angles showed different heat-transfer performance. This result implies that a strong association exists between jet impingement parameters and heat-transfer coefficient. The increase in coolant viscosity of the nanofluid coolant using jet impingement requires the expense of more drive-power cost.

  6. Heat transfer and fluid flow aspects of fuel--coolant interactions. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, M L

    1978-09-01

    A major portion of the safety analysis effort for the LMFBR is involved in assessing the consequences of a Hypothetical Core Disruptive Accident (HCDA). The thermal interaction of the hot fuel and the sodium coolant during the HCDA is investigated in two areas. A postulated loss of flow transient may produce a two-phase fuel at high pressures. The thermal interaction phenomena between fuel and coolant as the fuel is ejected into the upper plenum are investigated. A postulated transient overpower accident may produce molten fuel being released into sodium coolant in the core region. An energetic coolant vapor explosion for these reactor materials does not seem likely. However, experiments using other materials (e.g., Freon/water, tin/water) have demonstrated the possibility of this phenomenon.

  7. Turbulent Dispersion of Film Coolant and Hot Streaks in a Turbine Vane Cascade

    Science.gov (United States)

    2015-01-18

    configuration due to the large amounts of turning in the test section geometry and measurement techniques such as hot wire anemometry or temperature probe...Approved for Public Release; Distribution Unlimited Final Report: Turbulent Dispersion of Film Coolant and Hot Streaks in a Turbine Vane Cascade The...reviewed journals: Final Report: Turbulent Dispersion of Film Coolant and Hot Streaks in a Turbine Vane Cascade Report Title Magnetic resonance

  8. Optimal estimate of the coolant flow in the assemblies of a BWR of natural circulation in real time; Estimacion optima del flujo de refrigerante en los ensambles de un BWR de circulacion natural en tiempo real

    Energy Technology Data Exchange (ETDEWEB)

    Valle H, J.; Morales S, J. B. [UNAM, Facultad de Ingenieria, Division de Estudios de Posgrado, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, Col. Progreso, 62550 Jiutepec, Morelos (Mexico); Espinosa P, G., E-mail: julfi_jg@yahoo.com.mx [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)

    2012-10-15

    The present work exposes the design and the implementation of an advanced controller that allows estimating the coolant flow in the fuel assemblies of a BWR reactor of natural circulation in real time. To be able to reduce the penalizations that are established in the calculations of the operation limits due to the magnitude of the uncertainties in the coolant flows of a natural circulation reactor, is the objective of this research. In this work the construction of the optimal controller that allows estimating the coolant flows in a fuel channels group of the reactor is shown, as well as the operation of this applied to a reduced order model that simulates the dynamics of a natural circulation reactor. The controller design required of an estimator of the valuation variables not directly of the plant and of the estimates use of the local distributions of the coolant flow. The controller construction of the estimator was based mathematically in the filter Kalman whose algorithm allows to be carried out an advanced control of the system. To prove the estimator operation was development a simplified model that reproduces the basic dynamics of the flowing coolant in the reactor, which works as observer of the system, this model is coupled by means of the estimator controller to a detail model of the plant. The results are presented by means of graphics of the interest variables and the estimate flow, and they are documented in the chart that is presented at the end of this article. (Author)

  9. Experimental Study on the Effect of Late-Phase Coolant Injection on the Metallic Layer

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kyoung Ho; Park, Rae Joon; Cho, Young Ro; Kim, Sang Baik; Hong, Seong Wan; Kim, Hee Dong

    2007-04-15

    Sustained heating experiments, named ELIAS (Experiments on Late-phase coolant Injection to ASsess the mitigation of focusing effect of metallic layer), were performed to quantify the boiling heat removal rate at the upper surface of a metallic layer for precise evaluations on the effect of a late in-vessel coolant injection. Heat fluxes from the melt layer to the water pool varied from 250 to 550kW/m2 depending on the experimental conditions. Comparison of boiling heat fluxes between the ELIAS experiments and the calculation using the Berenson's film boiling correlation shows that effective heat removal was accomplished via late-phase coolant injection in the ELIAS experiments. In this study, simple model was developed to evaluate the mitigation of focusing effect in the metallic layer via late-phase coolant injection. The ELIAS experimental data on the heat transfer rate at the upper surface of the metallic layer were used as input data in the simple model. The calculation results for the large break loss of coolant accident in the APR1400 show that the risk induced by the focusing effect is highly dependent on the metallic layer thickness and the integrity of the reactor pressure vessel can be enhanced via late-phase coolant injection.

  10. The electrochemistry of IGSCC mitigation in BWR coolant circuits

    Energy Technology Data Exchange (ETDEWEB)

    Macdonald, D.D. [Center for Electrochemical Science and Technology, The Pennsylvania State Univ., University Park, PA (United States)

    2002-07-01

    A brief review is presented of the electrochemical mitigation of IGSCC in water-cooled reactor heat transport circuit structural materials. Electrochemical control and mitigation is possible, because of the existence of a critical potential for IGSCC and by the feasibility of modifying the environment to displace the corrosion potential (ECP) to a value that is more negative than the critical value. However, even in cases where the ECP cannot be displaced sufficiently in the negative direction to become more negative than the critical potential, considerable advantage is accrued, because of the roughly exponential dependence of crack growth rate on potential. The most important parameters in affecting electrochemical control over the ECP and crack growth rate are the kinetic parameters (exchange current densities and Tafel constants) for the redox reactions involving the principal radiolysis products of water (O{sub 2}, H{sub 2}, H{sub 2}O{sub 2}), external solution composition (concentrations of O{sub 2}, H{sub 2}O{sub 2}, and H{sub 2}), flow velocity, and the conductivity of the bulk environment. The kinetic parameters for the redox reactions essentially determine the charge transfer impedance of the steel surface, which is shown to be one of the key parameters in affecting the magnitude of the coupling current and hence the crack growth rate. The exchange current densities, in particular, are amenable to control by catalysis or inhibition, with the result that surface modification techniques are highly effective in controlling and mitigating IGSCC in reactor coolant circuit materials. (authors)

  11. Analysis of Loss-of-Coolant Accidents in the NBSR

    Energy Technology Data Exchange (ETDEWEB)

    Baek J. S.; Cheng L.; Diamond, D.

    2014-05-23

    This report documents calculations of the fuel cladding temperature during loss-of-coolant accidents in the NBSR. The probability of a pipe failure is small and procedures exist to minimize the loss of water and assure emergency cooling water flows into the reactor core during such an event. Analysis in the past has shown that the emergency cooling water would provide adequate cooling if the water filled the flow channels within the fuel elements. The present analysis is to determine if there is adequate cooling if the water drains from the flow channels. Based on photographs of how the emergency water flows into the fuel elements from the distribution pan, it can be assumed that this water does not distribute uniformly across the flow channels but rather results in a liquid film flowing downward on the inside of one of the side plates in each fuel element and only wets the edges of the fuel plates. An analysis of guillotine breaks shows the cladding temperature remains below the blister temperature in fuel plates in the upper section of the fuel element. In the lower section, the fuel plates are also cooled by water outside the element that is present due to the hold-up pan and temperatures are lower than in the upper section. For small breaks, the simulation results show that the fuel elements are always cooled on the outside even in the upper section and the cladding temperature cannot be higher than the blister temperature. The above results are predicated on assumptions that are examined in the study to see their influence on fuel temperature.

  12. Fuel-Coolant Interaction visualization in TROI test facility

    Energy Technology Data Exchange (ETDEWEB)

    Na, Young Su; Hong, Seong-Ho; Song, Jin Ho; Hong, Seong-Wan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    It is necessary to observe the FCI (Fuel-Coolant Interaction) phenomena at the condition of vessel failure to IVR. We carried out a visualization test on the interaction of a corium melt and water to observe the premixing phase without a free fall of a melt jet in a gas phase before contacting the cooling water. This paper is based on the previous study presented at Ninth Korea-Japan Symposium on Nuclear Hydraulics and Safety, we added the results on sieved debris distribution. The visualization test on the FCI without a free fall of a corium melt jet in a gas phase was conducted carefully in the TROI test facility. A prototypic corium consisting of uranium oxide and zirconium oxide with a weight ratio of UO{sub 2} to ZrO{sub 2} of 80 to 20, respectively, was heated up using the induction heating method. It was observed that a corium melt jet penetrated into water with 1000 mm in depth, and it took about 0.6 seconds from opening the releasing valve, which was confirmed by the sequential variation of the temperature measured by the sacrificial thermocouples installed in the direction of a falling melt jet. The cumulative mass fraction of the debris smaller than 1.0 mm was 15%, and the mass mean diameter of the debris was 2.9 mm. This visualization test can generate the valuable information such as the behavior of the corium melt jet and the size of mixing zone for validating the computer code.

  13. Cladding embrittlement during postulated loss-of-coolant accidents.

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M.; Yan, Y.; Burtseva, T.; Daum, R.; Nuclear Engineering Division

    2008-07-31

    The effect of fuel burnup on the embrittlement of various cladding alloys was examined with laboratory tests conducted under conditions relevant to loss-of-coolant accidents (LOCAs). The cladding materials tested were Zircaloy-4, Zircaloy-2, ZIRLO, M5, and E110. Tests were performed with specimens sectioned from as-fabricated cladding, from prehydrided (surrogate for high-burnup) cladding, and from high-burnup fuel rods which had been irradiated in commercial reactors. The tests were designed to determine for each cladding material the ductile-to-brittle transition as a function of steam oxidation temperature, weight gain due to oxidation, hydrogen content, pre-transient cladding thickness, and pre-transient corrosion-layer thickness. For short, defueled cladding specimens oxidized at 1000-1200 C, ring compression tests were performed to determine post-quench ductility at {le} 135 C. The effect of breakaway oxidation on embrittlement was also examined for short specimens oxidized at 800-1000 C. Among other findings, embrittlement was found to be sensitive to fabrication processes--especially surface finish--but insensitive to alloy constituents for these dilute zirconium alloys used as cladding materials. It was also demonstrated that burnup effects on embrittlement are largely due to hydrogen that is absorbed in the cladding during normal operation. Some tests were also performed with longer, fueled-and-pressurized cladding segments subjected to LOCA-relevant heating and cooling rates. Recommendations are given for types of tests that would identify LOCA conditions under which embrittlement would occur.

  14. Elevated-pressure mixed-coolants Joule Thomson cryocooling

    Science.gov (United States)

    Maytal, B.-Z.; Nellis, G. F.; Klein, S. A.; Pfotenhauer, J. M.

    2006-01-01

    This paper explores the potential of mixed coolants at elevated pressures for Joule-Thomson cryocooling. A numerical model of a Joule-Thomson cryocooler is developed that is capable of simulating operation with mixtures of up to 9 components consisting of hydrocarbons, non-flammable halogenated refrigerants, and inert gases. The numerical model is integrated with a genetic optimization algorithm, which has a high capability for convergence in an environment of discontinuities, constraints and local optima. The genetic optimization algorithm is used to select the optimal mixture compositions that separately maximizes following two objective functions at each elevated pressure for 80, 90 and 95 K cryocooling: the molar specific cooling capacity (the highest attainable is 3200 J/mol) and the produced cooling capacity per thermal conductance which is a measure of the compactness of the recuperator. The optimized cooling capacity for a non-flammable halogenated refrigerant mixture is smaller than for a hydrocarbon mixture; however, the cooling capacity of the two types of mixtures approach one another as pressure becomes higher. The coefficient of performance, the required heat transfer area and the effect of the number of components in the mixture is investigated as a function of the pressure. It is shown that mixtures with more components provide a higher cooling capacity but require larger recuperative heat exchangers. Optimized mixtures for 90 K cryocooling have similar cooling capacity as those for 80 K. Optimized compactness for 80 K is about 50% higher than can be achieved by pure nitrogen. For 90 K, no mixture provides a more compact recuperator than can be achieved using pure argon. The results are discussed in the context of potential applications for closed and open cycle cryocoolers.

  15. Sustainable Materials Management in Site Cleanup

    Science.gov (United States)

    In 2006, the management of materials accounted for 42 of the United States’ greenhouse gas (GHG) emissions, based on a systems analysis (U.S. EPA; 2009). The systems view of materials management represents U.S. emissions related to the...

  16. Liquid-lithium nitrate: candidate fusion reactor coolant or chemical curiosity

    Energy Technology Data Exchange (ETDEWEB)

    Adamson, M.G.

    1986-01-01

    On the basis of the recent Blanket Comparison and Selection Study, 700 K was selected as the threshold for useful high-temperature operation of a liquid nitrate coolant and 800 K was assumed as a realistic upper operating limit. Both standard Second Law equilibrium calculations and computerized Gibbs energy minimization methods were used to determine equilibrium compositions of multiphase, multicomponent salt systems at specified temperatures under certain condensed were performed on the LiNO/sub 3/-LiNO/sub 2/, NaNO/sub 3/- NaNO/sub 2/, and KNO/sub 3/-KNO/sub 2/ systems, and then predicted decomposition pressures were compared for equivalent degrees of decomposition at temperatures ranging from 600 to 900K. Two approaches were taken in calculating decomposition pressures over MNO/sub 3/-MNO/sub 2/ systems: (a) allowing the formation of molecular N/sub 2/ as a gaseous reaction product and (b) not allowing its formation. In calculations of MNO/sub 2/-M/sub 2/O-MOH-H/sub 2/O equilibria, which were used to evaluate the reversibility of tritium dissolution and release, the activity of hydroxide reaction product was determined as a function of water activity at two representative temperatures. Preliminary results and conclusions are summarized.

  17. Biotechnologies for Marine Oil Spill Cleanup: Indissoluble Ties with Microorganisms.

    Science.gov (United States)

    Mapelli, Francesca; Scoma, Alberto; Michoud, Grégoire; Aulenta, Federico; Boon, Nico; Borin, Sara; Kalogerakis, Nicolas; Daffonchio, Daniele

    2017-09-01

    The ubiquitous exploitation of petroleum hydrocarbons (HCs) has been accompanied by accidental spills and chronic pollution in marine ecosystems, including the deep ocean. Physicochemical technologies are available for oil spill cleanup, but HCs must ultimately be mineralized by microorganisms. How environmental factors drive the assembly and activity of HC-degrading microbial communities remains unknown, limiting our capacity to integrate microorganism-based cleanup strategies with current physicochemical remediation technologies. In this review, we summarize recent findings about microbial physiology, metabolism and ecology and describe how microbes can be exploited to create improved biotechnological solutions to clean up marine surface and deep waters, sediments and beaches. Copyright © 2017 Elsevier Ltd. All rights reserved.

  18. Cleanup Verification Package for the 618-8 Burial Ground

    Energy Technology Data Exchange (ETDEWEB)

    M. J. Appel

    2006-08-10

    This cleanup verification package documents completion of remedial action for the 618-8 Burial Ground, also referred to as the Solid Waste Burial Ground No. 8, 318-8, and the Early Solid Waste Burial Ground. During its period of operation, the 618-8 site is speculated to have been used to bury uranium-contaminated waste derived from fuel manufacturing, and construction debris from the remodeling of the 313 Building.

  19. Cleanup delays at hazardous waste sites: an incomplete information game

    OpenAIRE

    Rausser, Gordon C.; Simon, Leo K.; Zhao, Jinhua

    1999-01-01

    This paper studies the incentives facing Potentially Responsible Parties at a hazardous waste site to promote excessive investigation of the site and thus postpone the beginning of the remediation phase of the cleanup. We model the problem as an incomplete information, simultaneous-move game between PRPs. We assume that PRP's liability shares are predetermined. Each PRP's type is its private information about the precision of its own records relating to the site. A strategy for a PRP is a fun...

  20. Liability of Defense Contractors for Hazardous Waste Cleanup Costs

    Science.gov (United States)

    1989-04-01

    laws against toxic waste dumpers . I want faster cleanups and tougher enforcement of penalties against polluters. [Address by President Bush to joint...excludes coverage for "[a]ny loss, cost, or expense arising out of any governmental direction or request that [the insured] test for, monitor , clean...test for, monitor , clean up, remove, contain, treat, detoxify, or neutralize the pollutants.Ř 63 ENDNOTES 1. Text of President Bush’s Address to

  1. Numerical investigation on thermal striping conditions for a tee junction of LMFBR coolant pipes. 4. Investigation on second-order moments in coolant mixing region

    Energy Technology Data Exchange (ETDEWEB)

    Muramatsu, Toshiharu [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

    2000-02-01

    This report presents numerical results on thermal striping characteristics at a tee junction of LMFBR coolant pipe, carried out using a direct numerical simulation code DINUS-3. In the numerical investigations, it was considered a tee junction system consisted of a main pipe (1.33 cm{sup I.D.}) with a 90deg elbow and a brunch pipe, and four parameters, i.e., (1) diameter ratio {alpha} between both the pipes, (2) flow velocity ratio {beta} between both the pipes, (3) angle {gamma} between both the pipes, and (4) Reynolds number Re. From the numerical investigations, the following characteristics were obtained: (1) According to the decreasing of the diameter ratio, significant area of second-order moments was expanded in the fixed condition of {beta}=1.0. (2) Significant second-order moments area was expanded for the increasing of the flow velocity ratio {beta} specified by varying of the main pipe velocity in the case of a {alpha}=1.0 constant condition. On the other hand, the area was expanded for the decreasing of the velocity ratio {beta} defined by varying of the branch pipe velocity in the case of a {alpha}=1.0 constant condition. (3) Maximum second-order moments values were generated in the case of {gamma}=180deg due to the influence of interactions between main pipe flows and jet flows from the branch pipe. (4) According to the increase of Reynolds number, significant area of second-order moments was expanded due to the activation of turbulence mixing in the main pipe. (author)

  2. Strategies for reactor safety: Preventing loss of coolant accidents. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Lydell, B.O.Y. [RSA Technoligies, Vista (United States)

    1997-12-01

    This final report on the NKS/RAK-1.2 summarizes the main features of the PIFRAP PC-program and its intended implementation. Regardless of the preferred technical approach to LOCA frequency estimation, the analysis approach must include recognition of the following technical issues: (a) Degradation and failure mechanisms potentially affecting piping systems within the reactor coolant pressure boundary (RCPB) and the potential consequences; (b) In-service inspection practices and how they influence piping reliability; and (c) The service experience with piping systems. The report consists of six sections and one appendix. A Nordic perspective on LOCA and nuclear safety is given. It includes summaries of results from research in material sciences and current regulatory philosophies regarding piping reliability. A summary of the LOCA concept is applied in Nordic PSA studies. It includes a discussion on deterministic and probabilistic views on LOCA. The R and D on piping reliability by SKI and the PIFRAP model is summarized. Next, Section 6 presents conclusion and recommendations. Finally, Appendix A contains a list of abbreviations and acronyms, together with a glossary of technical terms. (EG) 16 refs.

  3. Cleanup of contaminated soil -- Unreal risk assumptions: Contaminant degradation

    Energy Technology Data Exchange (ETDEWEB)

    Schiffman, A. [New Jersey Department of Environmental Protection, Ewing, NJ (United States)

    1995-12-31

    Exposure assessments for development of risk-based soil cleanup standards or criteria assume that contaminant mass in soil is infinite and conservative (constant concentration). This assumption is not real for most organic chemicals. Contaminant mass is lost from soil and ground water when organic chemicals degrade. Factors to correct for chemical mass lost by degradation are derived from first-order kinetics for 85 organic chemicals commonly listed by USEPA and state agencies. Soil cleanup criteria, based on constant concentration, are then corrected for contaminant mass lost. For many chemicals, accounting for mass lost yields large correction factors to risk-based soil concentrations. For degradation in ground water and soil, correction factors range from greater than one to several orders of magnitude. The long exposure durations normally used in exposure assessments (25 to 70 years) result in large correction factors to standards even for carcinogenic chemicals with long half-lives. For the ground water pathway, a typical soil criterion for TCE of 1 mg/kg would be corrected to 11 mg/kg. For noncarcinogens, correcting for mass lost means that risk algorithms used to set soil cleanup requirements are inapplicable for many chemicals, especially for long periods of exposure.

  4. Restructuring of an Event Tree for a Loss of Coolant Accident in a PSA model

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Ho-Gon; Han, Sang-Hoon; Park, Jin-Hee; Jang, Seong-Chul [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    Conventional risk model using PSA (probabilistic Safety Assessment) for a NPP considers two types of accident initiators for internal events, LOCA (Loss of Coolant Accident) and transient event such as Loss of electric power, Loss of cooling, and so on. Traditionally, a LOCA is divided into three initiating event (IE) categories depending on the break size, small, medium, and large LOCA. In each IE group, safety functions or systems modeled in the accident sequences are considered to be applicable regardless of the break size. However, since the safety system or functions are not designed based on a break size, there exist lots of mismatch between safety system/function and an IE, which may make the risk model conservative or in some case optimistic. Present paper proposes new methodology for accident sequence analysis for LOCA. We suggest an integrated single ET construction for LOCA by incorporating a safety system/function and its applicable break spectrum into the ET. Integrated accident sequence analysis in terms of ET for LOCA was proposed in the present paper. Safety function/system can be properly assigned if its applicable range is given by break set point. Also, using simple Boolean algebra with the subset of the break spectrum, final accident sequences are expressed properly in terms of the Boolean multiplication, the occurrence frequency and the success/failure of safety system. The accident sequence results show that the accident sequence is described more detailed compared with the conventional results. Unfortunately, the quantitative results in terms of MCS (minimal Cut-Set) was not given because system fault tree was not constructed for this analysis and the break set points for all 7 point were not given as a specified numerical quantity. Further study may be needed to fix the break set point and to develop system fault tree.

  5. Use of Distribution Devices for Hydraulic Profiling of Coolant Flow in Core Gas-cooled Reactors

    Directory of Open Access Journals (Sweden)

    A. A. Satin

    2014-01-01

    Full Text Available In setting up a reactor plant for the transportation-power module of the megawatt class an important task is to optimize the path of flow, i.e. providing moderate hydraulic resistance, uniform distribution of the coolant. Significant contribution to the hydraulic losses makes one selected design of the coolant supplies. It is, in particular, hemispherical or semi-elliptical shape of the supply reservoir, which is selected to reduce its mass, resulting in the formation of torusshaped vortex in the inlet manifold, that leads to uneven coolant velocity at the inlet into the core, the flow pulsations, hydraulic losses.To control the flow redistribution in the core according to the level of energy are used the switchgear - deflectors installed in a hemispherical reservoir supplying coolant to the fuel elements (FE of the core of gas-cooled reactor. This design solution has an effect on the structure of the flow, rate in the cooling duct, and the flow resistance of the collector.In this paper we present the results of experiments carried out on the gas dynamic model of coolant paths, deflectors, and core, comprising 55 fuel rod simulators. Numerical simulation of flow in two-parameter model, using the k-ε turbulence model, and the software package ANSYS CFX v14.0 is performed. The paper demonstrates that experimental results are in compliance with calculated ones.The results obtained suggest that the use of switchgear ensures a coolant flow balance directly at the core inlet, thereby providing temperature reduction of fuel rods with a uniform power release in the cross-section. Considered options to find constructive solutions for deflectors give an idea to solve the problem of reducing hydraulic losses in the coolant paths, to decrease pulsation components of flow in the core and length of initial section of flow stabilization.

  6. Deposition of hematite particles on alumina seal faceplates of nuclear reactor coolant pumps: Laboratory experiments and industrial feedback

    Directory of Open Access Journals (Sweden)

    Lefèvre Grégory

    2012-01-01

    Full Text Available In the primary circuit of pressurized water reactors (PWR, the dynamic sealing system in reactor coolant pumps is ensured by mechanical seals whose ceramic parts are in contact with the cooling solution. During the stretch-out phase in reactor operation, characterized by low boric acid concentration, the leak-off flow has been observed to abnormally evolve in industrial plants. The deposition of hematite particles, originating from corrosion, on alumina seals of coolant pumps is suspected to be the cause. As better understanding of the adhesion mechanism is the key factor in the prevention of fouling and particle removal, an experimental study was carried out using a laboratory set-up. With model materials, hematite and sintered alumina, the adhesion rate and surface potentials of the interacting solids were measured under different chemical conditions (solution pH and composition in analogy with the PWR ones. The obtained results were in good agreement with the DLVO (Derjaguin-Landau-Verwey- Overbeek theory and used as such to interpret this industrial phenomenon.

  7. Bacteria Provide Cleanup of Oil Spills, Wastewater

    Science.gov (United States)

    2010-01-01

    Through Small Business Innovation Research (SBIR) contracts with Marshall Space Flight Center, Micro-Bac International Inc., of Round Rock, Texas, developed a phototrophic cell for water purification in space. Inside the cell: millions of photosynthetic bacteria. Micro-Bac proceeded to commercialize the bacterial formulation it developed for the SBIR project. The formulation is now used for the remediation of wastewater systems and waste from livestock farms and food manufacturers. Strains of the SBIR-derived bacteria also feature in microbial solutions that treat environmentally damaging oil spills, such as that resulting from the catastrophic 2010 Deepwater Horizon oil rig explosion in the Gulf of Mexico.

  8. Generic study on the relation between contamination if primary coolants and occupational radiation exposure in nuclear power plants with PWR. Final report; Generische Studie zum Zusammenhang zwischen Kontamination von Primaerkreislaufmedien und beruflicher Strahlenexposition bei Kernkraftwerken mit Druckwasserreaktor. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Artmann, Andreas; Bruhn, Gerd; Schneider, Sebastian [Gesellschaft fuer Anlagen- und Reaktorsicherheit, Koeln (Germany); Strub, Erik [Koeln Univ. (Germany)

    2016-01-15

    A generic model for the primary cooling system contamination in pressurized water reactors and the resulting radiological consequences has been developed. The functional capability was demonstrated by means of three examples concerning manipulation procedures during revision outages. Activities at the main reactor coolant pumps were studied and the influence of the coolant contamination on the resulting dose rates and collective doses were calculated. The effect of a Co-90 hot spot in a more remote area on the radiation exposure during the specific action at the reactor pumps was considered.

  9. Effects of staggered blades on the hydraulic characteristics of a 1400-MW canned nuclear coolant pump

    Directory of Open Access Journals (Sweden)

    Fang-Ming Zhou

    2016-08-01

    Full Text Available A canned nuclear coolant pump is used in an advanced third-generation pressurized water reactor. Impeller is a key component of a canned nuclear coolant pump. Usually, the blade is installed between the hub and the shroud as an entire part. The blade is divided into two parts and is staggered in the circumferential direction is an approach of blade design. To understand the effects of staggered blades on a canned nuclear coolant pump, this article numerically investigated different types of staggering. The validity of the numerical simulation was confirmed by comparing the numerical and experimental results. The performance change of a canned nuclear coolant pump with staggered blades was acquired. Hydraulic performance curves, axial force curves, static pressure distributions at the impeller outlet, and static pressure pulsations were performed to investigate the performance changes caused by the staggered blades. The results show that the staggered blade has an important influence on the performance of canned nuclear coolant pumps. A staggered blade does not improve hydraulic performance but does improve the axial force and pressure pulsation. Specifically, the staggered blades can significantly reduce the pressure pulsation amplitude on the impeller pass frequency.

  10. Aging and loss-of-coolant accident (LOCA) testing of electrical connections

    Energy Technology Data Exchange (ETDEWEB)

    Nelson, C.F. [Sandia National Labs., Albuquerque, NM (United States)

    1998-01-01

    This report presents the results of an experimental program to determine the aging and loss-of-coolant accident (LOCA) behavior of electrical connections in order to obtain an initial scoping of their performance. Ten types of connections commonly used in nuclear power plants were tested. These included 3 types of conduit seals, 2 types of cable-to-device connectors, 3 types of cable-to-cable connectors, and 2 types of in-line splices. The connections were aged for 6 months under simultaneous thermal (99 C) and radiation (46 Gy/hr) conditions. A simulated LOCA consisting of sequential high dose-rate irradiation (3 kGy/hr) and high-temperature steam exposures followed the aging. Connection functionality was monitored using insulation resistance measurements during the aging and LOCA exposures. Because only 5 of the 10 connection types passed a post-LOCA, submerged dielectric withstand test, further detailed investigation of electrical connections and the effects of cable jacket integrity on the cable-connection system is warranted.

  11. Prevention of unacceptable material fatigue considering the coolant. Principles and application

    Energy Technology Data Exchange (ETDEWEB)

    Kraetschmer, Daniel; Herter, Karl-Heinz; Schuler, Xaver [Stuttgart Univ. (Germany). MPA

    2010-07-01

    For the construction, design and operation of nuclear systems, structures and components the appropriate technical codes and standards provide material data, detailed stress analysis procedures and a design philosophy which guarantees a reliable behavior of the structural components throughout the specified lifetime. Especially for cyclic stress evaluation the different codes and standards provide different fatigue analyses procedures to be performed considering the various mechanical and thermal loading histories and geometric complexities of the components. For the fatigue design curves used as limiting criteria the influence of different factors like e.g., environment, surface finish and temperature must be taken into consideration in an appropriate way. A general numerical calculation procedure was developed to calculate equivalent stress- and strain ranges, according to different technical codes and standards. The additional implementation of already published environmental correction factors, depending on actual temperature, strain rate and dissolved-oxygen level, allows the practical and fast application of a strain-based approach to evaluate fatigue at varying temperatures and strain rates for specimens and components exposed to coolant environment. Proposed new design curves with and without the incorporation of environmental effects as well as design curves according to current technical codes and standards are considered. The application of this procedure is demonstrated and discussed by the example of a pressurizer nozzle under transient stratification loads, measured by in-service monitoring. (orig.)

  12. Review of experimental data for modelling LWR fuel cladding behaviour under loss of coolant accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Massih, Ali R. [Quantum Technologies AB, Uppsala Science Park (Sweden)

    2007-02-15

    Extensive range of experiments has been conducted in the past to quantitatively identify and understand the behaviour of fuel rod under loss-of-coolant accident (LOCA) conditions in light water reactors (LWRs). The obtained experimental data provide the basis for the current emergency core cooling system acceptance criteria under LOCA conditions for LWRs. The results of recent experiments indicate that the cladding alloy composition and high burnup effects influence LOCA acceptance criteria margins. In this report, we review some past important and recent experimental results. We first discuss the background to acceptance criteria for LOCA, namely, clad embrittlement phenomenology, clad embrittlement criteria (limitations on maximum clad oxidation and peak clad temperature) and the experimental bases for the criteria. Two broad kinds of test have been carried out under LOCA conditions: (i) Separate effect tests to study clad oxidation, clad deformation and rupture, and zirconium alloy allotropic phase transition during LOCA. (ii) Integral LOCA tests, in which the entire LOCA sequence is simulated on a single rod or a multi-rod array in a fuel bundle, in laboratory or in a tests and results are discussed and empirical correlations deduced from these tests and quantitative models are conferred. In particular, the impact of niobium in zirconium base clad and hydrogen content of the clad on allotropic phase transformation during LOCA and also the burst stress are discussed. We review some recent LOCA integral test results with emphasis on thermal shock tests. Finally, suggestions for modelling and further evaluation of certain experimental results are made.

  13. Analysis of an AP600 intermediate-size loss-of-coolant accident

    Energy Technology Data Exchange (ETDEWEB)

    Boyack, B.E.; Lime, J.F. [Los Alamos National Lab., NM (United States)

    1995-09-01

    A postulated double-ended guillotine break of an AP600 direct-vessel-injection line has been analyzed. This event is characterized as an intermediate-break loss-of-coolant accident. Most of the insights regarding the response of the AP600 safety systems to the postulated accident are derived from calculations preformed with the TRAC-PF1/MOD2 code. However, complementary insights derived from a scaled experiment conducted in the ROSA facility, as well as insights based upon calculations by other codes, are also presented. Based upon the calculated and experimental results, the AP600 will not experience a core heat up and will reach a safe shutdown state using only safety-class equipment. Only the early part of the long-term cooling period initiated by In-containment Refueling Water Storage Tank injection was evaluated. Thus, the observation that the core is continuously cooled should be verified for the later phase of the long-term cooling period when sump injection and containment cooling processes are important.

  14. Simulation of a transient with loss of primary coolant due to a small rupture in Angra 2 nuclear power plant with RELAP5/MOD3.2.2G code; Simulacao de um acidente postulado de perda de refrigerante primario por pequena ruptura na usina de Angra 2 com o codigo RELAP5/MOD3.2.2G

    Energy Technology Data Exchange (ETDEWEB)

    Sabundjian, Gaiane; Andrade, Delvonei Alves de [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)

    2002-07-01

    This paper presents a nodalization for Angra 2 Nuclear Power Plant, as well as the results obtained for a Small Break Loss of Coolant Accident (SBLOCA), simulated with RELAP5/MOD3.2G code. This accident consists in a small break (380 m{sup 2}) in the line of the Emergency Core Coolant System (ECCS) in loop 20 of Angra 2. Results are not as expected, however they are satisfactory regarding the nodalization used. (author)

  15. Analysis of polycyclic aromatic hydrocarbons in vegetable oils combining gel permeation chromatography with solid-phase extraction clean-up

    DEFF Research Database (Denmark)

    Fromberg, Arvid; Højgård, A.; Duedahl-Olesen, Lene

    2007-01-01

    of benzo[a]pyrene levels in foods laid down by the Commission of the European Communities. A survey of 69 vegetable oils sampled from the Danish market included olive oil as well as other vegetable oils such as rapeseed oil, sunflower oil, grape seed oil and sesame oil. Levels of benzo[a]pyrene in all......A semi-automatic method for the determination of polycyclic aromatic hydrocarbons (PAHs) in edible oils using a combined gel permeation chromatography/solid-phase extraction (GPC/SPE) clean-up is presented. The method takes advantage of automatic injections using a Gilson ASPEC XL sample handling...... system equipped with a GPC column (S-X3) and pre-packed silica SPE columns for the subsequent clean-up and finally gas chromatography-mass spectrometry (GC-MS) determination. The method was validated for the determination of PAHs in vegetable oils and it can meet the criteria for the official control...

  16. CALCULATING ECONOMIC RISK AFTER HANFORD CLEANUP

    Energy Technology Data Exchange (ETDEWEB)

    Scott, M.J.

    2003-02-27

    Since late 1997, researchers at the Hanford Site have been engaged in the Groundwater Protection Project (formerly, the Groundwater/Vadose Zone Project), developing a suite of integrated physical and environmental models and supporting data to trace the complex path of Hanford legacy contaminants through the environment for the next thousand years, and to estimate corresponding environmental, human health, economic, and cultural risks. The linked set of models and data is called the System Assessment Capability (SAC). The risk mechanism for economics consists of ''impact triggers'' (sequences of physical and human behavior changes in response to, or resulting from, human health or ecological risks), and processes by which particular trigger mechanisms induce impacts. Economic impacts stimulated by the trigger mechanisms may take a variety of forms, including changes in either costs or revenues for economic sectors associated with the affected resource or activity. An existing local economic impact model was adapted to calculate the resulting impacts on output, employment, and labor income in the local economy (the Tri-Cities Economic Risk Model or TCERM). The SAC researchers ran a test suite of 25 realization scenarios for future contamination of the Columbia River after site closure for a small subset of the radionuclides and hazardous chemicals known to be present in the environment at the Hanford Site. These scenarios of potential future river contamination were analyzed in TCERM. Although the TCERM model is sensitive to river contamination under a reasonable set of assumptions concerning reactions of the authorities and the public, the scenarios show low enough future contamination that the impacts on the local economy are small.

  17. On the fundamentals of thermal treatment for the cleanup of contaminated soils

    Energy Technology Data Exchange (ETDEWEB)

    Lighty, J.S.; Silcox, G.D.; Pershing, D.W. (Utah Univ., Salt Lake City, UT (USA)); Cundy, V.A. (Louisiana State Univ., Baton Rouge, LA (USA))

    1988-01-01

    Considerable research has focused on air emissions from the afterburner, mainly as a result of the regulations regarding destruction and removal efficiency of a principle organic hazardous constituent (POHC) -99.99% of the POHC must be destroyed in the system based on gas measurements from the afterburner. Research focusing on the primary desorber environment, the evolution of contaminants from solids and the resulting quality of the ash, is limited. The primary desorber is often operated at high temperatures which is costly, particularly for the cleanup of contaminated solid, due to high auxiliary fuel requirements. A more desirable option would be to desorb the contaminants from the soil at lower temperatures and then expose the off-gas to a high-temperature afterburner for decomposition of the hazardous compounds. In addition, the ability to predict the quality of the resulting soil is desirable for delisting purposes. To understand the desorption process, research must explore the rate controlling processes that are occurring. The overall goal of this research is to develop an understanding of the fundamental transport phenomena associated with the evolution of hazardous materials from soils in the primary desorber environment. As well, the rate information obtained can be used to model the thermal desorption of contaminants under a variety of experimental conditions; from these results large-scale operating parameters can be determined for optimum cleanup conditions.

  18. Biomass Gas Cleanup Using a Therminator

    Energy Technology Data Exchange (ETDEWEB)

    Dayton, David C; Kataria, Atish; Gupta, Rabhubir

    2012-03-06

    The objective of the project is to develop and demonstrate a novel fluidized-bed process module called a Therminator to simultaneously destroy and/or remove tar, NH3 and H2S from raw syngas produced by a fluidized-bed biomass gasifier. The raw syngas contains as much as 10 g/m3 of tar, 4,000 ppmv of NH3 and 100 ppmv of H2S. The goal of the Therminator module would be to use promising regenerable catalysts developed for removing tar, ammonia, and H2S down to low levels (around 10 ppm). Tars are cracked to a non-condensable gas and coke that would deposit on the acid catalyst. We will deposit coke, much like a fluid catalytic cracker (FCC) in a petroleum refinery. The deposited coke fouls the catalyst, much like FCC, but the coke would be burned off in the regenerator and the regenerated catalyst would be returned to the cracker. The rapid circulation between the cracker and regenerator would ensure the availability of the required amount of regenerated catalyst to accomplish our goal. Also, by removing sulfur down to less than 10 ppmv, NH3 decomposition would also be possible in the cracker at 600-700°C. In the cracker, tar decomposes and lays down coke on the acid sites of the catalyst, NH3 is decomposed using a small amount of metal (e.g., nickel or iron) catalyst incorporated into the catalyst matrix, and H2S is removed by a small amount of a metal oxide (e.g. zinc oxide or zinc titanate) by the H2S-metal oxide reaction to form metal sulfide. After a tolerable decline in activity for these reactions, the catalyst particles (and additives) are transported to the regenerator where they are exposed to air to remove the coke and to regenerate the metal sulfide back to metal oxide. Sulfate formation is avoided by running the regeneration with slightly sub-stoichiometric quantity of oxygen. Following regeneration, the catalyst is transported back to the cracker and the cycling continues. Analogous to an FCC reactor system, rapid cycling will allow the use of very

  19. Measurement of Coolant in a Flat Heat Pipe Using Neutron Radiography

    Science.gov (United States)

    Mizuta, Kei; Saito, Yasushi; Goshima, Takashi; Tsutsui, Toshio

    A newly developed flat heat pipe FGHPTM (Morex Kiire Co.) was experimentally investigated by using neutron radiography. The test sample of the FGHP heat spreader was 65 × 65 × 2 mm3 composed of several etched copper plates and pure water was used as the coolant. Neutron radiography was performed at the E-2 port of the Kyoto University Research Reactor (KUR). The coolant distributions in the wick area of the FGHP and its heat transfer characteristics were measured at heating conditions. Experimental results show that the coolant distributions depend slightly on its installation posture and that the liquid thickness in the wick region remains constant with increasing heat input to the FGHP. In addition, it is found that the wick surface does not dry out even in the vertical posture at present experimental conditions.

  20. Simulating the corrosion of zirconium alloys in the water coolant of VVER reactors

    Science.gov (United States)

    Kritskii, V. G.; Berezina, I. G.; Motkova, E. A.

    2013-07-01

    A model for predicting the corrosion of cladding zirconium alloys depending on their composition and operating conditions is proposed. Laws of thermodynamics and chemical kinetics of the reactions through which the multicomponent zirconium alloy is oxidized in the reactor coolant constitute the physicochemical heart of the model. The developed version of the model is verified against the results obtained from tests of fuel rod claddings made of commercial-grade and experimental zirconium alloys carried out by different researchers under autoclave and reactor conditions. It is shown that the proposed model adequately describes the corrosion of alloys in coolants used at nuclear power stations. It is determined that, owing to boiling of coolant and its acidification in a VVER-1200 reactor, Zr-1% Nb alloys with additions of iron and oxygen must be more resistant to corrosion than the commercial-grade alloy E110.

  1. Compatibility of structural materials with fusion reactor coolant and breeder fluids

    Energy Technology Data Exchange (ETDEWEB)

    DeVan, J.H.

    1979-01-01

    Fusion reactors are characterized by a lithium-containing blanket, a heat transfer medium that is integral with the blanket and first wall, and a heat engine that couples to the heat transfer medium. A variety of lithium-containing substances have been identified as potential blanket materials, including molten lithium metal, molten LiF--BeF/sub 2/, Pb--Li alloys, and solid ceramic compounds such as Li/sub 2/O. Potential heat transfer media include liquid lithium, liquid sodium, molten nitrates, water, and helium. Each of these coolants and blankets requires a particular set of chemical and mechanical properties with respect to the associated reactor and heat engine structural materials. This paper discusses the materials factors that underlie the selection of workable combinations of blankets and coolants. It also addresses the materials compatibility problems generic to those blanket-coolant combinations currently being considered in reactor design studies.

  2. Performance Evaluation of AI2O3/Water Nanofluid as Coolant in a Double-Tube Heat Exchanger Flowing under a Turbulent Flow Regime

    Directory of Open Access Journals (Sweden)

    Navid Bozorgan

    2012-01-01

    Full Text Available Nanofluids are expected to be a promising coolant candidate in chemical processes for water waste remediation and heat transfer system size reduction. This paper focuses on the potential mass flowrate reduction in exchanger with a given heat exchange capacity using nanofluids. Al2O3 nanoparticles with diameters of 7 nm dispersed in water with volume concentrations up to 2% are selected as a coolant, and their performance in a horizontal double-tube counterflow heat exchanger under turbulent flow conditions is numerically studied. The results show that the flowrate of nanofluid coolant decreases with the increase of concentration of nanoparticles in the exchanger with a given heat exchange capacity. The mass flowrate of the nanofluid at a volume concentration of 2 vol.% is approximately 24.5% lower than that of pure water (base fluid for given conditions. For the pressure drop, the results show that the pressure drop of nanofluid is slightly higher than water and increases with increase of volume concentrations. In addition, the reduction of wall temperature and heat transfer area is estimated.

  3. ROSA-III base test series for a large break loss-of-coolant accident in a boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tasaka, K.; Abe, N.; Anoda, Y.; Koizumi, Y.; Shiba, M.

    1982-05-01

    The rig of safety assessment (ROSA)-III facility is a volumetrically scaled (1/424) boiling water reactor (BWR) system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency core cooling system (ECCS) tests. It is confirmed from the experimental results obtained so far that the ROSA-III test facility can simulate major aspects of a BWR LOCA, such as boiling transition by lowering of the mixture level in the core, rewetting by the lower plenum flashing, and final quenching by the ECCS. The overall agreement between the calculated results by the RELAP5/ MOD0 code and the experimental results is good; however, the calculated lower plenum flashing rewetted the whole core and the calculated cladding temperature considerably underpredicts the measured value at the upper part of the core.

  4. Effects of molten material temperatures and coolant temperatures on vapor explosion

    Institute of Scientific and Technical Information of China (English)

    LI Tianshu; YANG Yanhua; YUAN Minghao; HU Zhihua

    2007-01-01

    An observable experiment facility for low-temperature molten materials to be dropped into water was set up in this study to investigate the mechanism of the vapor explosion. The effect of the fuel and coolant interaction(FCI) on the vapor explosion during the severe accidents of a fission nuclear reactor has been studied. The experiment results showed that the molten material temperature has an important effect on the vapor explosion behavior and pressure. The increase of the coolant temperature would decrease the pressure of the vapor explosion.

  5. Integrated Fuel-Coolant Interaction (IFCI 7.0) Code User's Manual

    Energy Technology Data Exchange (ETDEWEB)

    Young, Michael F.

    1999-05-01

    The integrated fuel-coolant interaction (IFCI) computer code is being developed at Sandia National Laboratories to investigate the fuel-coolant interaction (FCI) problem at large scale using a two-dimensional, three-field hydrodynamic framework and physically based models. IFCI will be capable of treating all major FCI processes in an integrated manner. This document is a description of IFCI 7.0. The user's manual describes the hydrodynamic method and physical models used in IFCI 7.0. Appendix A is an input manual provided for the creation of working decks.

  6. Coolant and ambient temperature control for chillerless liquid cooled data centers

    Energy Technology Data Exchange (ETDEWEB)

    Chainer, Timothy J.; David, Milnes P.; Iyengar, Madhusudan K.; Parida, Pritish R.; Simons, Robert E.

    2016-02-02

    Cooling control methods include measuring a temperature of air provided to a plurality of nodes by an air-to-liquid heat exchanger, measuring a temperature of at least one component of the plurality of nodes and finding a maximum component temperature across all such nodes, comparing the maximum component temperature to a first and second component threshold and comparing the air temperature to a first and second air threshold, and controlling a proportion of coolant flow and a coolant flow rate to the air-to-liquid heat exchanger and the plurality of nodes based on the comparisons.

  7. Surface Waviness in Grinding of Thin Mould Insert Using Chilled Air as Coolant

    Institute of Scientific and Technical Information of China (English)

    Yeo; S; H; K; Ramesh

    2002-01-01

    On going trend of miniaturization in electronic rel at ed parts, which is an average of two times in every 5~7 years introduce grindin g challenges. In grinding process, the surface waviness control of thin parts is an ardent task due to its warpage, induced by the high specific grinding energy (2~10 J/mm 3). Therefore, coolant is often used to avoid thermal damage, obtai n better surface integrity and to prolong wheel life. However coolant, the incomp ressibility media introduce high forces at the gri...

  8. Thermostat-controlled coolant pump - a new concept for fuel saving

    Energy Technology Data Exchange (ETDEWEB)

    Etemad, S. [Volvo Car Components Corp., Gothenburg (Sweden); Anderson, A. [Volvo Truck Corp., Gothenburg (Sweden)

    1999-07-01

    A new coolant pump concept has been developed for better fuel economy. The flow returning from the radiator is fed coaxially into the pump. The by-pass flow is fed tangentially into the pump, generating a pre-swirl with the same direction of rotation as the coolant pump impeller. The relative velocity between the flow and the impeller decreases. This reduces the transferred momentum from the impeller to the fluid, reducing the power consumption. The flow split between the radiator and the by-pass channel is controlled by the ordinary thermostat. Results from analysis and measurements are presented. (author)

  9. Land Use and Land Cover - MO 2008 Brownfields Voluntary Cleanup Program Sites (SHP)

    Data.gov (United States)

    NSGIC GIS Inventory (aka Ramona) — The Brownfields/Voluntary Cleanup Program (BVCP) provides property buyers, sellers, developers, bankers, development agencies, local government and other voluntary...

  10. Memorandum of the Establishment of Cleanup Levels for CERCLA Sites with Radioactive Contamination

    Science.gov (United States)

    This memorandum presents clarifying guidance for establishing protective cleanup levels for radioactive contamination at Comprehensive Environmental Response, Compensation, and Liability Act of 1980 (CERCLA) sites.

  11. Modern coolant additives. Environmental friendly and light metal compatible coolant additives for modern combustion engines; Moderne Kuehlmittelzusaetze. Umwelt- und leichtmetallvertraegliche Kuehlmittelzusaetze fuer moderne Verbrennungskraftmaschinen. Abschlussbericht. Vorhaben Nr. 777

    Energy Technology Data Exchange (ETDEWEB)

    Gugau, M.; Kaiser, M.

    2004-01-31

    The authors of the contribution under consideration report on the influence of the enhanced thermal stress on the impact of environmental friendly and light metal compatible coolant additives. The application and advancement of new research methods under mechanism-oriented objective led to a validation of a new guideline to the examination of the suitability of coolant additives for the coolant of internal combustion engines. Moreover, the authors create a knowledge base, on which a purposeful development can take place from suitable formulations of inhibitor for magnesium. For aluminium with silicate containing corrosion anti-freezes a close relationship between the surface temperature and the impoverishment of silicate exists. During the excess of limit temperatures, cooling agent-specific damage features arise reproducibly. The comparison of the different methods for the investigation of cavitation showed that one cannot dispense with both methods in order to evaluate a demand of insulating cavitation and a cavitative / corrosive complex regarding to the development of a test guideline. By the comprehensive electro-chemical and cavitative investigations for the magnesium alloy AZ91hp, a broad knowledge base could be formed, on which a purposeful development and evaluation of inhibitors under the use can take place from different glycols.

  12. Union job fight boiling at DOE cleanup sites

    Energy Technology Data Exchange (ETDEWEB)

    Setzer, S.W.

    1993-11-15

    The US DOE is facing a growing jurisdictional dispute over which unions will perform the majority of clean-up work at its facilities. Unions affiliated with the AFL-CIO Metal Trades Council representing operations employees at the sites believe they have a fundamental right to work. Unions in the AFL-CIO's Building and Construction Trades Dept. insist that they have a clear mandate under federal labor law and the Davis-Bacon Act. The issue has heated up in recent weeks at the policy level and is boiling in a contentious dispute at DOE's Fernald site in Ohio.

  13. Cleanup Verification Package for the 116-K-2 Effluent Trench

    Energy Technology Data Exchange (ETDEWEB)

    J. M. Capron

    2006-04-04

    This cleanup verification package documents completion of remedial action for the 116-K-2 effluent trench, also referred to as the 116-K-2 mile-long trench and the 116-K-2 site. During its period of operation, the 116-K-2 site was used to dispose of cooling water effluent from the 105-KE and 105-KW Reactors by percolation into the soil. This site also received mixed liquid wastes from the 105-KW and 105-KE fuel storage basins, reactor floor drains, and miscellaneous decontamination activities.

  14. Cleanup Verification Package for the 118-F-1 Burial Ground

    Energy Technology Data Exchange (ETDEWEB)

    E. J. Farris and H. M. Sulloway

    2008-01-10

    This cleanup verification package documents completion of remedial action for the 118-F-1 Burial Ground on the Hanford Site. This burial ground is a combination of two locations formerly called Minor Construction Burial Ground No. 2 and Solid Waste Burial Ground No. 2. This waste site received radioactive equipment and other miscellaneous waste from 105-F Reactor operations, including dummy elements and irradiated process tubing; gun barrel tips, steel sleeves, and metal chips removed from the reactor; filter boxes containing reactor graphite chips; and miscellaneous construction solid waste.

  15. Fernald restoration: ecologists and engineers integrate restoration and cleanup

    Energy Technology Data Exchange (ETDEWEB)

    Woods, Eric; Homer, John

    2002-07-15

    As cleanup workers excavate pits and tear down buildings at the Fernald site in southwest Ohio, site ecologists are working side-by-side to create thriving wetlands and develop the early stages of forest, prairie, and savanna ecosystems to restore natural resources that were impacted by years of site operations. In 1998, the U.S. Department of Energy-Fernald Office (DOE-FN) and its cleanup contractor, Fluor Fernald, Inc., initiated several ecological restoration projects in perimeter areas of the site (e.g., areas not used for or impacted by uranium processing or waste management). The projects are part of Fernald's final land use plan to restore natural resources over 904 acres of the 1,050-acre site. Pete Yerace, the DOE-FN Natural Resource Trustee representative is working with the Fernald Natural Resource Trustees in an oversight role to resolve the state of Ohio's 1986 claim against DOE for injuries to natural resources. Fluor Fernald, Inc., and DOE-FN developed the ''Natural Resource Restoration Plan'', which outlines 15 major restoration projects for the site and will restore injured natural resources at the site. In general, Fernald's plan includes grading to maximize the formation of wetlands or expanded floodplain, amending soil where topsoil has been removed during excavation, and establishing native vegetation throughout the site. Today, with cleanup over 35 percent complete and site closure targeted for 2006, Fernald is entering a new phase of restoration that involves heavily remediated areas. By working closely with engineers and cleanup crews, site ecologists can take advantage of remediation fieldwork (e.g., convert an excavated depression into a wetland) and avoid unnecessary costs and duplication. This collaboration has also created opportunities for relatively simple and inexpensive restoration of areas that were discovered during ongoing remediation. To ensure the survival of the plant material in heavily

  16. Evaluation of gasification and gas cleanup processes for use in molten carbonate fuel cell power plants. Final report. [Contains lists and evaluations of coal gasification and fuel gas desulfurization processes

    Energy Technology Data Exchange (ETDEWEB)

    Jablonski, G.; Hamm, J.R.; Alvin, M.A.; Wenglarz, R.A.; Patel, P.

    1982-01-01

    This report satisfies the requirements for DOE Contract AC21-81MC16220 to: List coal gasifiers and gas cleanup systems suitable for supplying fuel to molten carbonate fuel cells (MCFC) in industrial and utility power plants; extensively characterize those coal gas cleanup systems rejected by DOE's MCFC contractors for their power plant systems by virtue of the resources required for those systems to be commercially developed; develop an analytical model to predict MCFC tolerance for particulates on the anode (fuel gas) side of the MCFC; develop an analytical model to predict MCFC anode side tolerance for chemical species, including sulfides, halogens, and trace heavy metals; choose from the candidate gasifier/cleanup systems those most suitable for MCFC-based power plants; choose a reference wet cleanup system; provide parametric analyses of the coal gasifiers and gas cleanup systems when integrated into a power plant incorporating MCFC units with suitable gas expansion turbines, steam turbines, heat exchangers, and heat recovery steam generators, using the Westinghouse proprietary AHEAD computer model; provide efficiency, investment, cost of electricity, operability, and environmental effect rankings of the system; and provide a final report incorporating the results of all of the above tasks. Section 7 of this final report provides general conclusions.

  17. Natural microbial system for heavy metals cleanup application

    African Journals Online (AJOL)

    compq

    2012-05-24

    May 24, 2012 ... biomineralization, intracellular accumulation and enzyme- catalyzed ..... source for the heterotrophic marine bacterium Sagittula stellata. ... Defining biominerals and organominerals: direct and indirect indicators of life.

  18. Thermal hydraulic characteristics during ingress of coolant and loss of vacuum events in fusion reactors

    Science.gov (United States)

    Takase, K.; Kunugi, T.; Seki, Y.; Akimoto, H.

    2000-03-01

    The thermal hydraulic characteristics in the vacuum vessel (VV) of a fusion reactor under an ingress of coolant event (ICE) and a loss of vacuum event (LOVA) were investigated quantitatively using preliminary experimental apparatuses. In the ICE experiments, pressure rise characteristics in the VV were clarified for experimental parameters of the wall temperature and water temperature and for cases with and without a blowdown tank. In addition, the functional performance of a blowdown tank with and without a water cooling system was examined and it was confirmed that the blowdown tank with a water cooling system is effective for suppressing the pressure rise during the ICE. In the LOVA experiments, the saturation time in the VV from vacuum to atmosphere was investigated for various breach sizes and it was found that the saturation time is in inverse proportion to the breach size. In addition, the characteristics of exchange flow through breaches were clarified for the different breach positions on the VV. It was proven from the experimental results that the exchange flow became a counter-current flow when the breach was positioned on the top of the VV and a stratified flow when it was formed on the side wall of the VV, and that the exchange flow under the stratified flow condition was smoother than that of counter-current flow. On the basis of these results, the severest breach condition in ITER was changed from the top-break case to the side-break case. To predict with high accuracy the thermal hydraulic characteristics during ICEs and LOVAs under ITER conditions, a large scale test facility will be necessary. The current conceptual design of the combined ICE-LOVA test facility with a scaling factor of 1/1000 in comparison with the ITER volume is presented.

  19. High-temperature high-pressure gas cleanup with ceramic bag filters. Draft final report

    Energy Technology Data Exchange (ETDEWEB)

    Shackleton, M.; Chang, R.; Sawyer, J.; Kuby, W.; Turner-Tamiyasu, E.

    1982-12-06

    Advanced processes designed for the efficient use of coal in the production of energy will benefit from, or even depend upon, highly efficient, economical, high-temperature removal systems for fine particulates. In the case of pressurized fluidized-bed combustion (PFBC), the hot gas cleanup device must operate at approximately 1600/sup 0/F. Existing commercial filter systems are temperature limited due to the filter material, but ceramic fibers intended for refractory insulation offer the promise of a practical high-temperature filter media if they can be incorporated into a design which combines filter performance with acceptable durability. The current work was initiated to further develop and demonstrate on a larger-scale basis, a ceramic fiber filtration system for application to coal-fired PFBC's. The development effort centered around the need to replace the knit metal wire scrim, used in earlier designs as support for the fine fiber ceramic mat filtration medium, with a corrosion-resistant material. This led to the selection of woven ceramic cloth for support of the mat layer. Because of the substantial difference in strength and other material properties between the metal and ceramic cloth, tests were necessary to optimize the filter; pulse parameters such as pulse duration, pulse pressure, and pulse injection orifice size; woven cloth mesh configuration; the technique for clamping the bag to the support; and similar structural, fluid, and control parameters. The demonstration effort included both tests to prove this concept in a real application and a systems analysis to show commercial feasibility of the ceramic filtration approach for hot gas cleanup in PFBC's. 12 references, 57 figures, 23 tables.

  20. Cleanup of an urban site contaminated by monazite processing plant

    Energy Technology Data Exchange (ETDEWEB)

    Lauria, Dejanira C.; Zenaro, Rozangela; Sachett, Ivanor A. [Instituto de Radioprotecao e Dosimetria (IRD), Rio de Janeiro, RJ (Brazil). Dept. de Radioprotecao Ambiental

    2001-07-01

    For half a century the Santo Amaro Mill processed monazite sand in order to isolate rare earth elements. At the beginning of its operation, the mill was located far from the centre of Sao Paulo city. However, over the years the city spread and engulfed the mill, which, together with economical and radiological problems, led to its being shutdown and later decommissioned. Based on a future residential occupation scenario complying with a dose limit of 1 mSv/y, a concentration guideline level of 0.65 Bq/g of {sup 228} Ra activity concentration in the soil was derived. The cleanup actions led for removing of about 2300 m{sup 3} of soil from the area, of which 60 m{sup 3} was sent to a repository and 2240 m{sup 3} to a municipal landfill. This paper address to present the criteria for the establishment of the derived concentration guideline level of radionuclides in soil and the studies carried out for establishment of measurement procedures for on-site radiation measurements aiming speed-up of the analyses during the cleanup actions. (author)

  1. ENGINEERING A NEW MATERIAL FOR HOT GAS CLEANUP

    Energy Technology Data Exchange (ETDEWEB)

    T.D. Wheelock; L.K. Doraiswamy; K.P. Constant

    2003-09-01

    The overall purpose of this project was to develop a superior, regenerable, calcium-based sorbent for desulfurizing hot coal gas with the sorbent being in the form of small pellets made with a layered structure such that each pellet consists of a highly reactive lime core enclosed within a porous protective shell of strong but relatively inert material. The sorbent can be very useful for hot gas cleanup in advanced power generation systems where problems have been encountered with presently available materials. An economical method of preparing the desired material was demonstrated with a laboratory-scale revolving drum pelletizer. Core-in-shell pellets were produced by first pelletizing powdered limestone or other calcium-bearing material to make the pellet cores, and then the cores were coated with a mixture of powdered alumina and limestone to make the shells. The core-in-shell pellets were subsequently calcined at 1373 K (1100 C) to sinter the shell material and convert CaCO{sub 3} to CaO. The resulting product was shown to be highly reactive and a very good sorbent for H{sub 2}S at temperatures in the range of 1113 to 1193 K (840 to 920 C) which corresponds well with the outlet temperatures of some coal gasifiers. The product was also shown to be both strong and attrition resistant, and that it can be regenerated by a cyclic oxidation and reduction process. A preliminary evaluation of the material showed that while it was capable of withstanding repeated sulfidation and regeneration, the reactivity of the sorbent tended to decline with usage due to CaO sintering. Also it was found that the compressive strength of the shell material depends on the relative proportions of alumina and limestone as well as their particle size distributions. Therefore, an extensive study of formulation and preparation conditions was conducted to improve the performance of both the core and shell materials. It was subsequently determined that MgO tends to stabilize the high

  2. ENGINEERING A NEW MATERIAL FOR HOT GAS CLEANUP

    Energy Technology Data Exchange (ETDEWEB)

    T.D. Wheelock; L.K. Doraiswamy; K.P. Constant

    2003-09-01

    The overall purpose of this project was to develop a superior, regenerable, calcium-based sorbent for desulfurizing hot coal gas with the sorbent being in the form of small pellets made with a layered structure such that each pellet consists of a highly reactive lime core enclosed within a porous protective shell of strong but relatively inert material. The sorbent can be very useful for hot gas cleanup in advanced power generation systems where problems have been encountered with presently available materials. An economical method of preparing the desired material was demonstrated with a laboratory-scale revolving drum pelletizer. Core-in-shell pellets were produced by first pelletizing powdered limestone or other calcium-bearing material to make the pellet cores, and then the cores were coated with a mixture of powdered alumina and limestone to make the shells. The core-in-shell pellets were subsequently calcined at 1373 K (1100 C) to sinter the shell material and convert CaCO{sub 3} to CaO. The resulting product was shown to be highly reactive and a very good sorbent for H{sub 2}S at temperatures in the range of 1113 to 1193 K (840 to 920 C) which corresponds well with the outlet temperatures of some coal gasifiers. The product was also shown to be both strong and attrition resistant, and that it can be regenerated by a cyclic oxidation and reduction process. A preliminary evaluation of the material showed that while it was capable of withstanding repeated sulfidation and regeneration, the reactivity of the sorbent tended to decline with usage due to CaO sintering. Also it was found that the compressive strength of the shell material depends on the relative proportions of alumina and limestone as well as their particle size distributions. Therefore, an extensive study of formulation and preparation conditions was conducted to improve the performance of both the core and shell materials. It was subsequently determined that MgO tends to stabilize the high

  3. Use of ethanolamine for alkalization of secondary coolant. First experience at VVER reactor

    Energy Technology Data Exchange (ETDEWEB)

    Smiesko, I. [NPP Jaslovske Bohunice (Slovakia); Bystriansky, J. [TEDIS-KOR, Dobra (Czech Republic); Szalo, A. [NPPRI Trnava (Slovakia)

    2002-07-01

    The paper summarises preparatory work and results of six-week plant trial aimed at use of ethanolamine for alkalization of secondary coolant. Operational data in pre-test and test period are given and outage inspection results are commented. Future plans are outlined. (authors)

  4. Study on effects of mixing vane grids on coolant temperature distribution by subchannel analysis

    Energy Technology Data Exchange (ETDEWEB)

    Mao, H.; Yang, B.W.; Han, B. [Xi' an Jiaotong Univ., Shaanxi (China). Science and Technology Center for Advanced Nuclear Fuel Research

    2016-07-15

    Mixing vane grids (MVG) have great influence on coolant temperature field in the rod bundle. The MVG could enhance convective heat transfer between the fuel rod wall and the coolant, and promote inter-subchannel mixing at the same time. For the influence of the MVG on convective heat transfer enhancement, many experiments have been done and several correlations have been developed based on the experimental data. However, inter-subchannel mixing promotion caused by the MVG is not well estimated in subchannel analysis because the information of mixing vanes is totally missing in most subchannel codes. This paper analyzes the influence of mixing vanes on coolant temperature distribution using the improved MVG model in subchannel analysis. The coolant temperature distributions with the MVG are analyzed, and the results show that mixing vanes lead to a more uniform temperature distribution. The performances of split vane grids under different power conditions are evaluated. The results are compared with those of spacer grids without mixing vanes and some conclusions are obtained.

  5. Partial Discharge Measurements in HV Rotating Machines in Dependence on Pressure of Coolant

    Directory of Open Access Journals (Sweden)

    I. Kršňák

    2002-01-01

    Full Text Available The influence of the pressure of the coolant used in high voltage rotating machines on partial discharges occurring in stator insulation is discussed in this paper. The first part deals with a theoretical analysis of the topic. The second part deals with the results obtained on a real generator in industrial conditions. Finally, theoretical assumptions and obtained results are compared.

  6. Modeling Film-Coolant Flow Characteristics at the Exit of Shower-Head Holes

    Science.gov (United States)

    Garg, Vijay K.; Gaugler, R. E. (Technical Monitor)

    2000-01-01

    The coolant flow characteristics at the hole exits of a film-cooled blade are derived from an earlier analysis where the hole pipes and coolant plenum were also discretized. The blade chosen is the VKI rotor with three staggered rows of shower-head holes. The present analysis applies these flow characteristics at the shower-head hole exits. A multi-block three-dimensional Navier-Stokes code with Wilcox's k-omega model is used to compute the heat transfer coefficient on the film-cooled turbine blade. A reasonably good comparison with the experimental data as well as with the more complete earlier analysis where the hole pipes and coolant plenum were also gridded is obtained. If the 1/7th power law is assumed for the coolant flow characteristics at the hole exits, considerable differences in the heat transfer coefficient on the blade surface, specially in the leading-edge region, are observed even though the span-averaged values of h (heat transfer coefficient based on T(sub o)-T(sub w)) match well with the experimental data. This calls for span-resolved experimental data near film-cooling holes on a blade for better validation of the code.

  7. Vibration signal analysis of main coolant pump flywheel based on Hilbert–Huang transform

    Directory of Open Access Journals (Sweden)

    Meiru Liu

    2015-03-01

    In this paper, we present a Hilbert–Huang transform (HHT algorithm for flywheel vibration analysis. The simulation indicated that the proposed flywheel vibration signal analysis method performs well, which means that the method can lay the foundation for the detection and diagnosis in a reactor main coolant pump.

  8. Contribution to the diagnosis of mixed friction in the bearings of a reactor coolant pump

    Energy Technology Data Exchange (ETDEWEB)

    Gaev, G.P.; Shilejko, P.G.; Kail, I.T.; Proskuryakov, K.N. (Moskovskij Ehnergeticheskij Inst. (USSR)); Hippmann, N.; Kinsky, D.; Sturm, A.; Uhlemann, S. (Ingenieurhochschule Zittau (German Democratic Republic))

    1984-10-01

    Theoretical and experimental investigations have been performed to study the vibrational behaviour of a vertical, slide-bearing, fully encapsulated reactor coolant pump at various operational conditions. Magnetical and mechanical noise is interpreted as a function of pump delivery, pressure, volume flow, and temperature, and an example of an inadmissible operational condition (mixed friction in the bearings) is diagnosed.

  9. Control of oxidizing potential of Pb and Pb-Bi coolants

    Directory of Open Access Journals (Sweden)

    Vladimir Vladimirovich Ulyanov

    2015-12-01

    Full Text Available Analytical and experimental data on formation of oxygen oxidizing potential in heavy liquid metal coolants (Pb and Pb-Bi eutectic was considered. It was revealed that oxygen could be both dissolved in these coolants and included in various thermodynamically unstable oxide compounds. In case of heavy liquid metal coolant (HLMC flowing in non-isothermal circuit, these compounds are broken down with oxygen release or formed fixing dissolved oxygen. The amount of oxygen, which is present in HLMC and exhibits its activity with temperature, could be much greater than the value detected by oxygen sensor. That is why HLMC possess internal oxygen reserves inhibiting corrosion in the circuits. Presence of thermodynamically unstable oxide phases in the above coolants, non-isoconcentration distribution of active oxygen, and impossibility to currently obtain the analytical relationship showing dissolved oxygen distribution make it necessary to use at least three oxygen sensors for studying processes of formation of HLMC oxidizing potential. These sensors should be located in the zones of max and min temperatures (tmax, tmin and in that at t=450-550°С. In order to assure the most accurate estimate it is reasonable to provide additional two or more sensors in the zone at t=450-550°С.

  10. OPAL REACTOR: Calculation/Experiment comparison of Neutron Flux Mapping in Flux Coolant Channels

    Energy Technology Data Exchange (ETDEWEB)

    Barbot, L.; Domergue, C.; Villard, J. F.; Destouches, C. [CEA, Paris (France); Braoudakis, G.; Wassink, D.; Sinclair, B.; Osborn, J. C.; Huayou, Wu [ANSTO, Syeney (Australia)

    2013-07-01

    The measurement and calculation of the neutron flux mapping of the OPAL research reactor are presented. Following an investigation of fuel coolant channels using sub-miniature fission chambers to measure thermal neutron flux profiles, neutronic calculations were performed. Comparison between calculation and measurement shows very good agreement.

  11. The upgrade of intense pulsed neutron source (IPNS) through the change of coolant and reflector

    CERN Document Server

    Baek, I C; Iverson, E B

    2002-01-01

    The current intense pulsed neutron source (IPNS) depleted uranium target is cooled by light water. The inner reflector material is graphite and the outer reflector material is beryllium. The presence of H sub 2 O in the target moderates neutrons and leads to a higher absorption loss in the target than is necessary. D sub 2 O coolant in the small quantities required minimizes this effect. We have studied the possible improvement in IPNS beam fluxes that would result from changing the coolant from H sub 2 O to D sub 2 O and the inner reflector from graphite to beryllium. Neutron intensities were calculated for directions normal to the viewed surface of each moderator for four different cases of combinations of target coolant and reflector materials. The simulations reported here were performed using the MCNPX (version 2.1.5) computer program. Our results show that substantial gains in neutron beam intensities can be achieved by appropriate combination of target coolant and reflector materials. The combination o...

  12. An experimental investigation of the post-CHF enhancement factor for a prototypical ITER divertor plate with water coolant

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, T.D. [Rensselaer Polytechnic Institute, Troy, NY (United States); Watson, R.D.; McDonald, J.M. [Sandia National Lab., Albuquerque, NM (United States)] [and others

    1995-09-01

    In an off-normal event, water-cooled copper divertor plates in the International Thermonuclear Experimental Reactor (ITER) may either experience heat loads beyond their design basis, or the normal heat loads may be accompanied by low coolant pressure and velocity. The purpose of this experiment was to illustrate that during one-sided heating, as in ITER, a copper divertor plate with the proper side wall thickness, at low system pressure and velocity can absorb without failing an incident heat flux, q{sub i}, that significantly exceed the value, q{sub i}{sup CHF}, which is associated with local CHF at the wall of the coolant channel. The experiment was performed using a 30 kW electron beam test system for heating of a square cross-section divertor heat sink with a smooth circular channel of 7.63 mm diameter. The heated width, length, and wall thickness were 16, 40, and 3 mm, respectively. Stable surface temperatures were observed at incident heat fluxes greater than the local CHF point, presumably due to circumferential conduction around the thick tube walls when q{sub i}{sup CHF} was exceeded. The Post-CHF enhancement factor, {eta}, is defined as the ratio of the incident burnout heat flux, q{sub i}{sup BO}, to q{sub i}{sup CHF}. For this experiment with water at inlet conditions of 70{degrees}C, 1 m/s, and 1 MPa, q{sub i}{sup CHF} and q{sub i}{sup BO} were 600 and 1100 W/cm{sup 2}, respectively, which gave an {eta} of 1.8.

  13. Large-break loss-of-coolant accident phenomena identification and ranking table (PIRT) for the advanced Candu reactor

    Energy Technology Data Exchange (ETDEWEB)

    Popov, N.; Snell, V.G.; Sills, H.E.; Langman, V.J.; Boyack, B. [Atomic Energy of Canada Ltd (Canada)

    2004-07-01

    The Advanced Candu Reactor (ACR) is an evolutionary advancement of the current Candu-6 reactor, aimed at producing electrical power for a capital cost and unit-energy cost significantly less than that of current reactor designs. The ACR retains the modular concept of horizontal fuel channels surrounded by heavy water moderator, as with all Candu reactors. However, ACR uses slightly enriched uranium (SEU) fuel, compared to the natural uranium used in Candu 6. This achieves the twin goals of improved economics (e.g., via reductions in the heavy water requirements and the use of a light water coolant), as well as improved safety. This paper is focused on the double-ended guillotine critical inlet header break (CRIHB) loss-of-coolant accident (LOCA) in an ACR reactor, which is considered as a large break LOCA. Large Break LOCA in water-cooled reactors has been used historically as a design basis event by regulators, and it has attracted a very large share of safety analysis and regulatory review. The LBLOCA event covers a wide range of system behaviours and fundamental phenomena. The Phenomena Identification and Ranking Table (PIRT) for LBLOCA therefore provides a good understanding of many of the safety characteristics of the ACR design. The paper outlines the design characteristics of the ACR reactor that impact the PIRT process and computer code applicability. It also describes the LOCA phenomena, lists all components and systems that have an important role during the event, discusses the PIRT process and results, and presents the final PIRT summary table. (authors)

  14. Numerical and experimental hydrodynamic study of a coolant distributor for grinding applications

    Directory of Open Access Journals (Sweden)

    Tala Moussa

    2016-01-01

    Full Text Available In grinding, the high frictional energy is converted into heat, which may cause thermal damage and degradation of the wheel and the workpiece. Unwanted thermal effects must thus be reduced, often by external cooling using a curved-duct coolant distributor to match the wheel geometry. The performance of such a system depends strongly on the impinging jet flow properties to ensure efficient sprinkling of the hot spots. The fluid distributor, placed above the workpiece, is pierced with a certain number of identical nozzle fittings, providing multiple jets at the outlet of the nozzles. These jets sprinkle the solids over a given zone and remove the heat by convective transfer. The cooling is hence dependent on the flow structure, meaning the jet diameters, trajectories and velocities, determined up-flow by the distributor design. The present study is devoted to the hydrodynamics aspects of the fluid distributor, aiming to determine the flow-rate distribution at the different orifices and the flow-rate–pressure relationship, for a variety of nozzle diameters and feeding flow rates, under isothermal conditions. A simple hydraulic balance in the device was not able to predict with sufficient accuracy the actual measurements, even when the Venturi effect was accounted for. This discrepancy is due to the curvature of the distributor, inducing secondary flows in interaction with the nozzle outlets, which leads to a rather complex flow pattern. To overcome this issue, a computational fluid dynamics (CFD tool was used and compared with in situ experiments – global flow rate and pressure measurements were additionally taken with particle image velocimetry (PIV to gain insight into the local structure. Simulations were performed with a 3D turbulence model for Reynolds numbers up to 100,000. This model provides an efficient tool for coupling with the thermal study at a later step, allowing global sizing and energetic optimization of the grinding process.

  15. Assessment of the heat carrier movement in the primary coolant circuit by its own momentum

    Energy Technology Data Exchange (ETDEWEB)

    Kadalev, Stoyan, E-mail: kadalev@inrne.bas.bg

    2014-10-15

    Highlights: • We model the heat carrier flow alteration after the circulation pump(s) stop. • The general mathematical model used is described in details. • The model is adapted and applied to a particular example research reactor. • Assessment is presented in detail, step by step with references. • The information provided is enough to apply calculations to another facility. - Abstract: In the presented paper is considered the approach to an assessment of the heat carrier flow alteration in the primary water–water reactor coolant circuit after the circulation pump(s) stop. This topic is highly relevant trough advanced and increased nuclear safety requirements because such a process is observed in case of black-out accident or damaged pump(s). The general mathematical model used is described; enabling preparation of this evaluation adapted and applied to a particular example facility namely a pool type research reactor. The factors influencing to the heat carrier movement by its own momentum are examined. The evaluation measures and includes the factors influencing the heat carrier flow rate from the moment the pump(s) stops down to a negligible value. Assessment is presented in detail, step by step and where needed with references to specific data and/or formulae from reference books to allow repetition of the calculations and/or apply to another facility. The calculations are presented utilizing all necessary data according to the design and technological documentation. No account is given to the pressure of the natural circulation caused by the residual heat generation in the fuel after the reactor scram system extinction of the fission reaction.

  16. TACT1- TRANSIENT THERMAL ANALYSIS OF A COOLED TURBINE BLADE OR VANE EQUIPPED WITH A COOLANT INSERT

    Science.gov (United States)

    Gaugler, R. E.

    1994-01-01

    As turbine-engine core operating conditions become more severe, designers must develop more effective means of cooling blades and vanes. In order to design reliable, cooled turbine blades, advanced transient thermal calculation techniques are required. The TACT1 computer program was developed to perform transient and steady-state heat-transfer and coolant-flow analyses for cooled blades, given the outside hot-gas boundary condition, the coolant inlet conditions, the geometry of the blade shell, and the cooling configuration. TACT1 can analyze turbine blades, or vanes, equipped with a central coolant-plenum insert from which coolant-air impinges on the inner surface of the blade shell. Coolant-side heat-transfer coefficients are calculated with the heat transfer mode at each station being user specified as either impingement with crossflow, forced convection channel flow, or forced convection over pin fins. A limited capability to handle film cooling is also available in the program. The TACT1 program solves for the blade temperature distribution using a transient energy equation for each node. The nodal energy balances are linearized, one-dimensional, heat-conduction equations which are applied at the wall-outer-surface node, at the junction of the cladding and the metal node, and at the wall-inner-surface node. At the mid-metal node a linear, three-dimensional, heat-conduction equation is used. Similarly, the coolant pressure distribution is determined by solving the set of transfer momentum equations for the one-dimensional flow between adjacent fluid nodes. In the coolant channel, energy and momentum equations for one-dimensional compressible flow, including friction and heat transfer, are used for the elemental channel length between two coolant nodes. The TACT1 program first obtains a steady-state solution using iterative calculations to obtain convergence of stable temperatures, pressures, coolant-flow split, and overall coolant mass balance. Transient

  17. Models and numerical methods for the simulation of loss-of-coolant accidents in nuclear reactors

    Science.gov (United States)

    Seguin, Nicolas

    2014-05-01

    In view of the simulation of the water flows in pressurized water reactors (PWR), many models are available in the literature and their complexity deeply depends on the required accuracy, see for instance [1]. The loss-of-coolant accident (LOCA) may appear when a pipe is broken through. The coolant is composed by light water in its liquid form at very high temperature and pressure (around 300 °C and 155 bar), it then flashes and becomes instantaneously vapor in case of LOCA. A front of liquid/vapor phase transition appears in the pipes and may propagate towards the critical parts of the PWR. It is crucial to propose accurate models for the whole phenomenon, but also sufficiently robust to obtain relevant numerical results. Due to the application we have in mind, a complete description of the two-phase flow (with all the bubbles, droplets, interfaces…) is out of reach and irrelevant. We investigate averaged models, based on the use of void fractions for each phase, which represent the probability of presence of a phase at a given position and at a given time. The most accurate averaged model, based on the so-called Baer-Nunziato model, describes separately each phase by its own density, velocity and pressure. The two phases are coupled by non-conservative terms due to gradients of the void fractions and by source terms for mechanical relaxation, drag force and mass transfer. With appropriate closure laws, it has been proved [2] that this model complies with all the expected physical requirements: positivity of densities and temperatures, maximum principle for the void fraction, conservation of the mixture quantities, decrease of the global entropy… On the basis of this model, it is possible to derive simpler models, which can be used where the flow is still, see [3]. From the numerical point of view, we develop new Finite Volume schemes in [4], which also satisfy the requirements mentioned above. Since they are based on a partial linearization of the physical

  18. Corrosion and solubility in a TSP-buffered chemical environment following a loss of coolant accident: Part 1 – Aluminum

    Energy Technology Data Exchange (ETDEWEB)

    Howe, Kerry J., E-mail: howe@unm.edu [University of New Mexico, 210 University Blvd., Albuquerque, NM 87131 (United States); Mitchell, Lana, E-mail: lmitchell@alionscience.com [University of New Mexico, 210 University Blvd., Albuquerque, NM 87131 (United States); Kim, Seung-Jun, E-mail: skim@lanl.gov [University of New Mexico, 210 University Blvd., Albuquerque, NM 87131 (United States); Blandford, Edward D., E-mail: edb@unm.edu [University of New Mexico, 210 University Blvd., Albuquerque, NM 87131 (United States); Kee, Ernest J., E-mail: erniekee@gmail.com [South Texas Project Nuclear Operating Company, P.O. Box 270, Wadsworth, TX 77483 (United States)

    2015-10-15

    Highlights: • Trisodium phosphate (TSP) causes aluminum corrosion to cease after 24 h of exposure. • Chloride, iron, and copper have a minimal effect on the rate of aluminum corrosion when TSP is present. • Zinc can reduce the rate of aluminum corrosion when TSP is present. • Aluminum occasionally precipitates at concentrations lower than the calculated solubility for Al(OH){sub 3}. • Corrosion and solubility equations can be used to calculate the solids generated during a LOCA. - Abstract: Bench experiments were conducted to investigate the effect of the presence of trisodium phosphate (TSP) on the corrosion and release of aluminum from metallic aluminum surfaces under conditions representative of the containment pool following a postulated loss of coolant accident at a nuclear power generating facility. The experiments showed that TSP is capable of passivating the aluminum surface and preventing continued corrosion after about 24 h at the conditions tested. A correlation that describes the rate of corrosion including the passivation effect was developed from the bench experiments and validated with a separate set of experiments from a different test system. The saturation concentration of aluminum was shown to be well described by the solubility of amorphous aluminum hydroxide for the majority of cases, but instances have been observed when aluminum precipitates at concentrations lower than the calculated aluminum hydroxide solubility. Based on the experimental data and previous literature, an equation was developed to calculate the saturation concentration of aluminum as a function of pH and temperature under conditions representative of a loss of coolant accident (LOCA) in a TSP-buffered pressurized water reactor (PWR) containment. The corrosion equation and precipitation equation can be used in concert with each other to calculate the quantity of solids that would form as a function of time during a LOCA if the temperature and pH profiles were known.

  19. Estimate of coolant flow in assemblies of a natural circulation BWR applying and equivalent electric model; Estimacion del flujo de refrigerante en los ensambles de un BWR de circulacion natural aplicando un modelo electrico equivalente

    Energy Technology Data Exchange (ETDEWEB)

    Valle H, J.; Morales S, J. B. [UNAM, DEPFI, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico); Espinosa P, G. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)], e-mail: julfi_jg@yahoo.com.mx

    2009-10-15

    The present work exposes the design and implementation of an advanced controller that it allows to estimate the coolant flow in fuel assemblies of a natural circulation BWR in real time. the complete development of this study is part of a doctoral project in course. In this work the construction of optimal controller is shown that allows to estimate the coolant flows in reactor and its operation applied to an equivalent electric model to natural circulation ESBWR. The controller design that allows the completely automatic starter of natural circulation reactor, required of a variables estimator not meter directly of nuclear power plant and use of local distributions estimates of coolant flow, (this controller type at the moment is utilized in the A BWR and several BWR in operation in Japan). The construction of estimator controller is mathematically based in the theory referring to Kalman filter, whose algorithm provides an advanced control of system. To prove the estimator operation was developed a simplified model that reproduces the basic dynamic of coolant flowing in the ESBWR, a practice way and very interesting of representing this phenomenon is by means the use of an equivalent electric model, which was developed starting from analogies that there is among the relation that keep the pressure differences with the mass flow and differences of electric potential with electric current. A detailed analysis of equivalence among models will be presented in a later article. (Author)

  20. Simulating Experimental Investigation on the Safety of Nuclear Heating Reactor in Loss—of —Coolant Accidents

    Institute of Scientific and Technical Information of China (English)

    ZhanjieXu

    1996-01-01

    The 5MW low temperature nuclear heating reactor (NHR-5) is a new and advanced type of nuclear reactor developed by Institute of Nuclear Energy Technology(INET) of Tisinghuan University of CHina in 1989,Its main loop is a thermal-hydraulic system with natural circulation.This paper studies the safety of NHR under the condition of loss-of -coolant accidents(LOCAs) by means of simulant experiments.First,the Background and necessity of the experiments are presented.then the experimental system,including the thermal-hydraulic system and the data collection system,and similarity criteria are introduced.Up to now ,the discharge experiments with the residual heating power(20% rated heating power)have been carried out on the experimental system,The system prameters including circulation flow rate,system pressure,system temperature,void fraction,discharge mass and so on have been recorded and analyzed.Based on the results of the experiments,the conclusionas are shown as follos:on the whole,the reactor is safe under the condition of LOCAs,but the thermal vacillations resulting from the vibration of the circulation flow rate are disadvantageous to the internal parts of the reactor core.

  1. Hot Chili Peppers: Extraction, Cleanup, and Measurement of Capsaicin

    Science.gov (United States)

    Huang, Jiping; Mabury, Scott A.; Sagebiel, John C.

    2000-12-01

    Capsaicin, the pungent ingredient of the red pepper or Capsicum annuum, is widely used in food preparation. The purpose of this experiment was to acquaint students with the active ingredients of hot chili pepper (capsaicin and dihydrocapsaicin), the extraction, cleanup, and analysis of these chemicals, as a fun and informative analytical exercise. Fresh peppers were prepared and extracted with acetonitrile, removing plant co-extractives by addition to a C-18 solid-phase extraction cartridge. Elution of the capsaicinoids was accomplished with a methanol-acetic acid solution. Analysis was completed by reverse-phase HPLC with diode-array or variable wavelength detection and calibration with external standards. Levels of capsaicin and dihydrocapsaicin were typically found to correlate with literature values for a specific hot pepper variety. Students particularly enjoyed relating concentrations of capsaicinoids to their perceived valuation of "hotness".

  2. Experimental determination of coolant flow pattern in hot and cold pools of PFBR using a large scale model

    Energy Technology Data Exchange (ETDEWEB)

    Indranil Banerjee; Rajesh, K.; AnandaRaj, M.; Venkata Ramanan, J.; Gopal, C.A.; Padmakumar, G.; Prakash, V.; Vaidyanathan, G. [Indira Gandhi Center for Atomic Research, Kalpakkam, 603102 (India)

    2005-07-01

    Full text of publication follows: The construction of Prototype Fast Breeder Reactor (PFBR) to generate 500 MWe has commenced at Kalpakkam, India. PFBR is a liquid sodium cooled pool type reactor with two secondary loops. The primary sodium pool is divided into hot pool and cold pool by means of Inner vessel. Cold sodium at 670 K is pumped through the core subassemblies and after absorbing the fission heat in the core, the sodium comes out and mixes with the hot pool at 820 K. This hot sodium exchanges heat with secondary sodium in Intermediate Heat Exchangers (IHX) which in turn transfers the heat to water in the steam generator leading to production of superheated steam to generate power. All the components like Control Plug (CP), IHX, Decay Heat Exchangers (DHX), Pump etc., are immersed in the primary sodium pool. The presence of these components influence the flow and velocity patterns of the coolant, in the hot and cold pools. The coolant behaviour in the pool is an indicator of the temperature pattern in the pool and the mechanical and thermal stresses induced on the immersed structures during transients is of significance for the safe operation of the reactor, designed for a life span of 40 years. Hence it is essential to understand the pattern of coolant flow and velocity patterns in hot and cold pools, particularly near IHX and Control plug. A 1:4 scale down model in stainless steel is constructed, simulating all the internal structures of the PFBR primary circuit for investigating the various parameters experimentally in water, to enhance the confidence in design of the primary system. The velocity distribution in the hot pool and cold pool at different regions, around the control plug, around the IHX inlet window were studied experimentally. As the coolant flow path is mainly influenced by the gravity force and inertia force, the study is conducted using Froude similitude. The magnitude of the velocity of the fluid at different points on the selected

  3. Long term alterations of blood plasma albumin in Chernobyl clean-up workers

    Directory of Open Access Journals (Sweden)

    Inta Kalnina

    2014-09-01

    Full Text Available Albumin is the most generously represented protein in human blood plasma. Therefore it is important to follow and assess the transport function of albumin in clinic researches. Disturbances in structural/functional properties of albumin play an important role in the pathogenesis of various diseases and immune state in patients. Changes in albumin transformation can serve as a diagnostic and prognostic criterion in pathologies. ABM (3-aminobenzanthrone derivative developed at the Daugavpils University, Latvia has been previously shown as a potential biomarker for determination of the immune state of patients with different pathologies. The aim of this study was to determine the several aspects of plasma albumin alterations in the group of Chernobyl clean-up workers in long term period in relation with humans having no professional contact with radioactivity. The following parameters were examined: (1 spectral characteristics of ABM in blood plasma; (2 and #8216;effective and #8217; and total albumin (EA and TA concentration in blood plasma; (3 quantitative parameters of albumin auto-fluorescence; (4 albumin binding site characteristics. Screening of the individuals with a period of 25-26 years after the work in Chernobyl revealed two groups of patients differing in structural and functional albumin properties; first on conformations of plasma albumin, and second characteristics of tryptophanyl region of the molecule. The revealed structural modifications of albumin are dependent on radiation-induced factors. Concomitant diseases such as diabetes mellitus or cardio-vascular diseases reinforce radiation-induced effects. In conclusion, ABM is a sensitive probe for albumin alterations and can be used to elucidate the changes in protein systems. Significant differences in albumin dynamics exist between control (donors and groups of Chernobyl clean-up workers. [J Exp Integr Med 2014; 4(3.000: 165-170

  4. Investigation of the Compatibility Between ADS Target Material With Coolant

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    At present, the program of the compatibility study on ADS project of China Institute of Atomic Energy is focused on the compatibility tests for the tungsten with water and sodium. An ADS verification facility is proposed in next phase of ADS project, an existent swimming pool reactor will be repack as the subcritical reactor system, and tungsten will be used as the target. On the other hand, our CEFR is being constructed now, it may be one of the options as the subcritical reactor system of ADS, thus, it is necessary to understand the compatibility characteristics of tungsten with sodium and water.

  5. A Model for Molten Fuel-Coolant Interaction during Melt Slumping in a Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sohal, Manohar Singh; Siefken, Larry James

    1999-10-01

    This paper describes a simple fuel melt slumping model to replace the current parametric model in SCDAP/RELAP5. Specifically, a fuel-coolant interaction (FCI) model is developed to analyze the slumping molten fuel, molten fuel breakup, heat transfer to coolant, relocation of the molten droplets, size of a partially solidified particles that settle to the bottom of the lower plenum, and melt-plenum interaction, if any. Considering our objectives, the molten fuel jet breakup model, and fuel droplets Lagrangian model as included in a code TEXAS-V with Eulerian thermal hydraulics for water and steam from SCDAP/RELAP5 were used. The model was assessed with experimental data from MAGICO-2000 tests performed at University of California at Santa Barbara, and FARO Test L-08 performed at Joint Research Center, Ispra, Italy. The comparison was found satisfactory.

  6. Simulation of isothermal multi-phase fuel-coolant interaction using MPS method with GPU acceleration

    Energy Technology Data Exchange (ETDEWEB)

    Gou, W.; Zhang, S.; Zheng, Y. [Zhejiang Univ., Hangzhou (China). Center for Engineering and Scientific Computation

    2016-07-15

    The energetic fuel-coolant interaction (FCI) has been one of the primary safety concerns in nuclear power plants. Graphical processing unit (GPU) implementation of the moving particle semi-implicit (MPS) method is presented and used to simulate the fuel coolant interaction problem. The governing equations are discretized with the particle interaction model of MPS. Detailed implementation on single-GPU is introduced. The three-dimensional broken dam is simulated to verify the developed GPU acceleration MPS method. The proposed GPU acceleration algorithm and developed code are then used to simulate the FCI problem. As a summary of results, the developed GPU-MPS method showed a good agreement with the experimental observation and theoretical prediction.

  7. The effect of coolants on the performance of magnetic micro-refrigerators.

    Science.gov (United States)

    Silva, D J; Bordalo, B D; Pereira, A M; Ventura, J; Oliveira, J C R E; Araújo, J P

    2014-06-01

    Magnetic refrigeration is an alternative cooling technique with envisaged technological applications on micro- and opto-electronic devices. Here, we present a magnetic micro-refrigerator cooling device with embedded micro-channels and based on the magnetocaloric effect. We studied the influence of the coolant fluid in the refrigeration process by numerically simulating the heat transfer processes using the finite element method. This allowed us to calculate the cooling power of the device. Our results show that gallium is the most efficient coolant fluid and, when used with Gd5Si2Ge2, a maximum power of 11.2 W/mm3 at a working frequency of -5 kHz can be reached. However, for operation frequencies around 50 Hz, water is the most efficient fluid with a cooling power of 0.137 W/mm3.

  8. Development of Reactor Coolant Pump for APR1400

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Sang-Youn; Chu, Sung-Min; Chang, Jin-Young [Doosan Heavy Industries and Construction, Changwon (Korea, Republic of)

    2015-10-15

    The development was focused on the performance requirements for APR1400 and to achieve the goals of the safety, reliability and adaptability for APR1400 system design. In addition, APR1400 RCP design was customized considering convenience of installation, operation and maintainability. This paper describes the details of the development process, improved design feature and type test results. Based on development of core technology of RCP, DOOSAN supplies the localized and improved APR1400 RCP to Shin-Hanul 1 and 2 Project. This would be good experience that the RCP core technology can break foreign monopoly in supplying the domestic nuclear industry. Also, there expect APR1400 RCP can be sustainable revenue models in nuclear industry. Moreover, development of RCP will be a catalyst to enhance design capacity for equipment and system of nuclear power plant as well as evaluation and verification skills of Korean nuclear industry.

  9. Lead Coolant Test Facility Technical and Functional Requirements, Conceptual Design, Cost and Construction Schedule

    Energy Technology Data Exchange (ETDEWEB)

    Soli T. Khericha

    2006-09-01

    This report presents preliminary technical and functional requirements (T&FR), thermal hydraulic design and cost estimate for a lead coolant test facility. The purpose of this small scale facility is to simulate lead coolant fast reactor (LFR) coolant flow in an open lattice geometry core using seven electrical rods and liquid lead or lead-bismuth eutectic. Based on review of current world lead or lead-bismuth test facilities and research need listed in the Generation IV Roadmap, five broad areas of requirements of basis are identified: Develop and Demonstrate Prototype Lead/Lead-Bismuth Liquid Metal Flow Loop Develop and Demonstrate Feasibility of Submerged Heat Exchanger Develop and Demonstrate Open-lattice Flow in Electrically Heated Core Develop and Demonstrate Chemistry Control Demonstrate Safe Operation and Provision for Future Testing. These five broad areas are divided into twenty-one (21) specific requirements ranging from coolant temperature to design lifetime. An overview of project engineering requirements, design requirements, QA and environmental requirements are also presented. The purpose of this T&FRs is to focus the lead fast reactor community domestically on the requirements for the next unique state of the art test facility. The facility thermal hydraulic design is based on the maximum simulated core power using seven electrical heater rods of 420 kW; average linear heat generation rate of 300 W/cm. The core inlet temperature for liquid lead or Pb/Bi eutectic is 420oC. The design includes approximately seventy-five data measurements such as pressure, temperature, and flow rates. The preliminary estimated cost of construction of the facility is $3.7M. It is also estimated that the facility will require two years to be constructed and ready for operation.

  10. Additional requirements for leak-before-break application to primary coolant piping in Belgium

    Energy Technology Data Exchange (ETDEWEB)

    Roussel, G. [AIB Vincotte Nuclear, Brussels (Belgium)

    1997-04-01

    Leak-Before-Break (LBB) technology has not been applied in the first design of the seven Pressurized Water Reactors the Belgian utility is currently operating. The design basis of these plants required to consider the dynamic effects associated with the ruptures to be postulated in the high energy piping. The application of the LBB technology to the existing plants has been recently approved by the Belgian Safety Authorities but with a limitation to the primary coolant loop. LBB analysis has been initiated for the Doel 3 and Tihange 2 plants to allow the withdrawal of some of the reactor coolant pump snubbers at both plants and not reinstall some of the restraints after steam generator replacement at Doel 3. LBB analysis was also found beneficial to demonstrate the acceptability of the primary components and piping to the new conditions resulting from power uprating and stretch-out operation. LBB analysis has been subsequently performed on the primary coolant loop of the Tihange I plant and is currently being performed for the Doel 4 plant. Application of the LBB to the primary coolant loop is based in Belgium on the U.S. Nuclear Regulatory Commission requirements. However the Belgian Safety Authorities required some additional analyses and put some restrictions on the benefits of the LBB analysis to maintain the global safety of the plant at a sufficient level. This paper develops the main steps of the safety evaluation performed by the Belgian Safety Authorities for accepting the application of the LBB technology to existing plants and summarizes the requirements asked for in addition to the U.S. Nuclear Regulatory Commission rules.

  11. Experimental study of electroinsulating coatings in gallium coolant related to the divertor cooling loop

    Science.gov (United States)

    Beznosov, A. V.; Sherbakov, R. V.; Karatushina, I. V.; Romanov, P. V.

    1996-10-01

    Experimental investigation of electroinsulating coatings stability on the samples made of stainless stell, vanadium alloy and beryllium has been conducted at 80-350°C. The impact of gas pressure upon the liquid gallium open surface was studied. The stability of electroinsulating film parameters on divertor structure materials was confirmed for the divertor with open liquid metal coolant surface in the vacuum chamber.

  12. International Standard Problems and Small Break Loss-of-Coolant Accident (SBLOCA

    Directory of Open Access Journals (Sweden)

    N. Aksan

    2008-01-01

    Full Text Available Best-estimate thermal-hydraulic system codes are widely used to perform safety and licensing analyses of nuclear power plants and also used in the design of advance reactors. Evaluation of the capabilities and the performance of these codes can be accomplished by comparing the code predictions with measured experimental data obtained on different test facilities. OECD/NEA Committee on the Safety of Nuclear Installations (CSNI has promoted, over the last twenty-nine years, some forty-eight international standard problems (ISPs. These ISPs were performed in different fields as in-vessel thermal-hydraulic behaviour, fuel behaviour under accident conditions, fission product release and transport, core/concrete interactions, hydrogen distribution and mixing, containment thermal-hydraulic behaviour. 80% of these ISPs were related to the working domain of principal working group no.2 on coolant system behaviour (PWG2 and were one of the major PWG2 activities for many years. A global review and synthesis on the contribution that ISPs have made to address nuclear reactor safety issues was initiated by CSNI-PWG2 and an overview on the subject of small break LOCA ISPs is given in this paper based on a report prepared by a writing group. In addition, the relevance of small break LOCA in a PWR with relation to nuclear reactor safety and the reorientation of the reactor safety program after TMI-2 accident are shortly summarized. The experiments in four integral test facilities, LOBI, SPES, BETHSY, ROSA IV/LSTF and the recorded data during a steam generator tube rupture transient in the DOEL-2 PWR (Belgium were the basis of the five small break LOCA related ISP exercises, which deal with the phenomenon typical of small break LOCAs in Western design PWRs. Some lessons learned from these small break LOCA ISPs are identified in relation to code deficiencies and capabilities, progress in the code capabilities, possibility of scaling, and various additional aspects

  13. Simulation of fuel dispersion in the MYRRHA-FASTEF primary coolant with CFD and SIMMER-IV

    Energy Technology Data Exchange (ETDEWEB)

    Buckingham, Sophia, E-mail: sophia.buckingham@vki.ac.be [von Karman Institute, Chaussée de Waterloo 72, B-1640 Rhode-St-Genèse (Belgium); Planquart, Philippe [von Karman Institute, Chaussée de Waterloo 72, B-1640 Rhode-St-Genèse (Belgium); Eboli, Marica [University of Pisa, Largo Lucio Lazzarino 2, 56122 Pisa (Italy); Moreau, Vincent [CRS4, Science and Technology Park Polaris – Piscina Manna, 09010 Pula (Italy); Van Tichelen, Katrien [SCK-CEN, Boeretang 200, 2400 Mol (Belgium)

    2015-12-15

    Highlights: • A comparison between CFD and system codes applied to long-term dispersion of fuel particles inside the MYRRHA reactor is proposed. • Important accumulations at the free-surface level are to be expected. • The risk of core blockage should not be neglected. • Numerical approach and modeling assumptions have a strong influence on the simulation results and accuracy. - Abstract: The objective of this work is to assess the behavior of fuel redistribution in heavy liquid metal nuclear systems under fuel pin failure conditions. Two different modeling approaches are considered using Computational Fluid Dynamics (CFD) codes and a system code, applied to the MYRRHA facility primary coolant loop version 1.4. Two different CFD models are constructed: the first is a single-phase steady model prepared in ANSYS Fluent, while the second is a two-phase model based on the volume of fluid (VOF) method in STARCCM+ to capture the upper free-surface dynamics. Both use a Lagrangian tracking approach with oneway coupling to follow the particles throughout the reactor. The system code SIMMER-IV is used for the third model, without neutronic coupling. Although limited regarding the fluid dynamic aspects compared to the CFD codes, comparisons of particle distributions highlight strong similarities despite quantitative discrepancies in the size of fuel accumulations. These disparities should be taken into account while performing the safety analysis of nuclear systems and developing strategies for accident mitigation.

  14. Improved Traps for Removing Gases From Coolant Liquids

    Science.gov (United States)

    Holladay, John; Ritchie, Stephen

    2006-01-01

    Two documents discuss improvements in traps for removing noncondensable gases (e.g., air) from heat-transfer liquids (e.g., water) in spacecraft cooling systems. Noncondensable gases must be removed because they can interfere with operation. A typical trap includes a cylindrical hydrophobic membrane inside a cylindrical hydrophilic membrane, all surrounded by an outer cylindrical impermeable shell. The input mixture of gas bubbles and liquid flows into the annular volume between the membranes. Bubbles pass into the central hollow of the hydrophobic membrane and are vented. The liquid flows outward through the hydrophilic membrane and is recirculated.

  15. Effects of LWR coolant environments on fatigue design curves of carbon and low-alloy steels

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O.K.; Shack, W.J. [Argonne National Lab., IL (United States)

    1998-03-01

    The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components. Figures I-9.1 through I-9.6 of Appendix I to Section III of the code specify fatigue design curves for structural materials. While effects of reactor coolant environments are not explicitly addressed by the design curves, test data indicate that the Code fatigue curves may not always be adequate in coolant environments. This report summarizes work performed by Argonne National Laboratory on fatigue of carbon and low-alloy steels in light water reactor (LWR) environments. The existing fatigue S-N data have been evaluated to establish the effects of various material and loading variables such as steel type, dissolved oxygen level, strain range, strain rate, temperature, orientation, and sulfur content on the fatigue life of these steels. Statistical models have been developed for estimating the fatigue S-N curves as a function of material, loading, and environmental variables. The results have been used to estimate the probability of fatigue cracking of reactor components. The different methods for incorporating the effects of LWR coolant environments on the ASME Code fatigue design curves are presented.

  16. Comparative Evaluation of Coolant Mixing Experiments at the ROCOM, Vattenfall, and Gidropress Test Facilities

    Directory of Open Access Journals (Sweden)

    S. Kliem

    2007-01-01

    Full Text Available Coolant mixing is an important mitigative mechanism against reactivity accidents caused by local boron dilution. Experiments on coolant mixing were carried out at three different test facilities representing three different reactor types. These are the ROCOM test facility modelling a German KONVOI-type reactor, the Vattenfall test facility being a model of a Westinghouse three-loop PWR, and the Gidropress test facility modelling a VVER-1000 PWR. The scenario of the start-up of the first main coolant pump was investigated in all three facilities. The experiments were accompanied by velocity measurements in the downcomer for the same scenario in the ROCOM and the Vattenfall test facilities. A similar flow structure was found in these measurements in both cases. A maximum of the velocity is measured at the opposite side in regard to the position of the loop with the starting-up pump whilst a recirculation area was found just below this inlet nozzle in both facilities. The analysis of the slug mixing experiments showed also comparable flow behaviour. In accordance with the velocity measurements, the first part of the deboration is also found on the opposite side. In this region, the maximum deboration is measured in all three cases. These maximum values are in the same order of magnitude for nearly identical initial slug volumes.

  17. Assessment of fiber optic sensors for aging monitoring of industrial liquid coolants

    Science.gov (United States)

    Riziotis, Christos; El Sachat, Alexandros; Markos, Christos; Velanas, Pantelis; Meristoudi, Anastasia; Papadopoulos, Aggelos

    2015-03-01

    Lately the demand for in situ and real time monitoring of industrial assets and processes has been dramatically increased. Although numerous sensing techniques have been proposed, only a small fraction can operate efficiently under harsh industrial environments. In this work the operational properties of a proposed photonic based chemical sensing scheme, capable to monitor the ageing process and the quality characteristics of coolants and lubricants in industrial heavy machinery for metal finishing processes is presented. The full spectroscopic characterization of different coolant liquids revealed that the ageing process is connected closely to the acidity/ pH value of coolants, despite the fact that the ageing process is quite complicated, affected by a number of environmental parameters such as the temperature, humidity and development of hazardous biological content as for example fungi. Efficient and low cost optical fiber sensors based on pH sensitive thin overlayers, are proposed and employed for the ageing monitoring. Active sol-gel based materials produced with various pH indicators like cresol red, bromophenol blue and chorophenol red in tetraethylorthosilicate (TEOS), were used for the production of those thin film sensitive layers deposited on polymer's and silica's large core and highly multimoded optical fibers. The optical characteristics, sensing performance and environmental robustness of those optical sensors are presented, extracting useful conclusions towards their use in industrial applications.

  18. Experimental distribution of coolant in the IPR-R1 Triga nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Z., E-mail: amir@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Servico de Tecnologia de Reatores; Palma, Daniel A.P., E-mail: dapalma@cnen.gov.b [Comissao Nacional de Energia Nuclear (CNEN/RJ), Rio de Janeiro, RJ (Brazil); Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Reis, Patricia A.L., E-mail: claubia@nuclear.ufmg.b, E-mail: dora@nuclear.ufmg.b [Universidade Federal de Minas Gerais (DEN/UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear

    2011-07-01

    The IPR-R1 is a typical TRIGA Mark I light-water and open pool type reactor. The core has an annular configuration of six rings and is cooled by natural circulation. The core coolant channels extend from the bottom grid plate to the top grid plate. The cooling water flows through the holes in the bottom grid plate, passes through the lower unheated region of the element, flows upwards through the active region, passes through the upper unheated region, and finally leaves the channel through the differential area between a triangular spacer block on the top of the fuel element and a round hole in the grid. Direct measurement of the flow rate in a coolant channel is difficult because of the bulky size and low accuracy of flow meters. The flow rate through the channel may be determined indirectly from the heat balance across the channel using measurements of the water inlet and outlet temperatures. This paper presents the experiments performed in the IPR-R1 reactor to monitoring some thermo-hydraulic parameters in the core coolant channels, such as: the radial and axial temperature profile, temperature, velocity, mass flow rate, mass flux and Reynolds's number. Some results were compared with theoretical predictions, as it was expected the variables follow the power distribution (or neutron flux) in the core. (author)

  19. Heat transfer performance characteristics of hybrid nanofluids as coolant in louvered fin automotive radiator

    Science.gov (United States)

    Sahoo, Rashmi R.; Sarkar, Jahar

    2016-12-01

    Present study deals with the enhancement of convective heat transfer performance of EG brine based various hybrid nanofluids i.e. Ag, Cu, SiC, CuO and TiO2 in 0-1% volume fraction of Al2O3 nanofluid, as coolants for louvered fin automobile radiator. The effects of nanoparticles combination and operating parameters on thermo physical properties, heat transfer, effectiveness, pumping power and performance index of hybrid nanofluids have been evaluated. Comparison of studied hybrid nanofluids based on radiator size and pumping power has been made as well. Among all studied hybrid nanofluids, 1% Ag hybrid nanofluid (0.5% Ag and 0.5% Al2O3) yields highest effectiveness and heat transfer rate as well as pumping power. However, SiC + Al2O3 dispersed hybrid nanofluid yields maximum performance index and hence this can be recommended for best coolant. For the same radiator size and heat transfer rate, pumping power increases by using Ag hybrid nanofluids leading to increase in engine thermal efficiency and hence reduction in engine fuel consumption. For same coolant flow rate and heat transfer rate, the radiator size reduces and pumping power increases by using Ag hybrid nanofluids leading to reduction in radiator size, weight and cost.

  20. Use of Nitrogen Trifluoride To Purify Molten Salt Reactor Coolant and Heat Transfer Fluoride Salts

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-05-02

    Abstract: The molten salt cooled nuclear reactor is included as one of the Generation IV reactor types. One of the challenges with the implementation of this reactor is purifying and maintaining the purity of the various molten fluoride salts that will be used as coolants. The method used for Oak Ridge National Laboratory’s molten salt experimental test reactor was to treat the coolant with a mixture of H2 and HF at 600°C. In this article we evaluate thermal NF3 treatment for purifying molten fluoride salt coolant candidates based on NF3’s 1) past use to purify fluoride salts, 2) other industrial uses, 3) commercial availability, 4) operational, chemical, and health hazards, 5) environmental effects and environmental risk management methods, 6) corrosive properties, and 7) thermodynamic potential to eliminate impurities that could arise due to exposure to water and oxygen. Our evaluation indicates that nitrogen trifluoride is a viable and safer alternative to the previous method.

  1. The high-temperature sodium coolant technology in nuclear power installations for hydrogen power engineering

    Science.gov (United States)

    Kozlov, F. A.; Sorokin, A. P.; Alekseev, V. V.; Konovalov, M. A.

    2014-05-01

    In the case of using high-temperature sodium-cooled nuclear power installations for obtaining hydrogen and for other innovative applications (gasification and fluidization of coal, deep petroleum refining, conversion of biomass into liquid fuel, in the chemical industry, metallurgy, food industry, etc.), the sources of hydrogen that enters from the reactor plant tertiary coolant circuit into its secondary coolant circuit have intensity two or three orders of magnitude higher than that of hydrogen sources at a nuclear power plant (NPP) equipped with a BN-600 reactor. Fundamentally new process solutions are proposed for such conditions. The main prerequisite for implementing them is that the hydrogen concentration in sodium coolant is a factor of 100-1000 higher than it is in modern NPPs taken in combination with removal of hydrogen from sodium by subjecting it to vacuum through membranes made of vanadium or niobium. Numerical investigations carried out using a diffusion model showed that, by varying such parameters as fuel rod cladding material, its thickness, and time of operation in developing the fuel rods for high-temperature nuclear power installations (HT NPIs) it is possible to exclude ingress of cesium into sodium through the sealed fuel rod cladding. However, if the fuel rod cladding loses its tightness, operation of the HT NPI with cesium in the sodium will be unavoidable. Under such conditions, measures must be taken for deeply purifying sodium from cesium in order to minimize the diffusion of cesium into the structural materials.

  2. Heat transfer performance characteristics of hybrid nanofluids as coolant in louvered fin automotive radiator

    Science.gov (United States)

    Sahoo, Rashmi R.; Sarkar, Jahar

    2017-06-01

    Present study deals with the enhancement of convective heat transfer performance of EG brine based various hybrid nanofluids i.e. Ag, Cu, SiC, CuO and TiO2 in 0-1% volume fraction of Al2O3 nanofluid, as coolants for louvered fin automobile radiator. The effects of nanoparticles combination and operating parameters on thermo physical properties, heat transfer, effectiveness, pumping power and performance index of hybrid nanofluids have been evaluated. Comparison of studied hybrid nanofluids based on radiator size and pumping power has been made as well. Among all studied hybrid nanofluids, 1% Ag hybrid nanofluid (0.5% Ag and 0.5% Al2O3) yields highest effectiveness and heat transfer rate as well as pumping power. However, SiC + Al2O3 dispersed hybrid nanofluid yields maximum performance index and hence this can be recommended for best coolant. For the same radiator size and heat transfer rate, pumping power increases by using Ag hybrid nanofluids leading to increase in engine thermal efficiency and hence reduction in engine fuel consumption. For same coolant flow rate and heat transfer rate, the radiator size reduces and pumping power increases by using Ag hybrid nanofluids leading to reduction in radiator size, weight and cost.

  3. Hot gas cleanup test facility for gasification and pressurized combustion. Quarterly technical progress report, January 1--March 31, 1992

    Energy Technology Data Exchange (ETDEWEB)

    1992-12-01

    This quarterly technical progress report summarizes work completed during the Sixth Quarter of the First Budget Period, January 1 through March 31, 1992, under the Department of Energy (DOE) Cooperative Agreement No. DE-FC21-90MC25140 entitled ``Hot Gas Cleanup Test Facility for Gasification and Pressurized Combustion.`` The objective of this project is to evaluate hot gas particle control technologies using coal-derived gas streams. The major emphasis during this reporting period was expanding the test facility to address system integration issues of hot particulate removal in advanced power generation systems. The conceptual design of the facility was extended to include additional modules for the expansion of the test facility, which is referred to as the Power Systems Development Facility (PSOF). A letter agreement was negotiated between Southern Company Services (SCS) and Foster Wheeler (FW) for the conceptual design of the Advanced Pressurized Fluid-Bed Combustion (APFBC)/Topping Combustor/Gas Turbine System to be added to the facility. The expanded conceptual design also included modifications to the existing conceptual design for the Hot Gas Cleanup Test Facility (HGCTF), facility layout and balance of plant design for the PSOF. Southern Research Institute (SRI) began investigating the sampling requirements for the expanded facility and assisted SCS in contacting Particulate Control Device (PCD) vendors for additional information. SCS also contacted the Electric Power Research Institute (EPRI) and two molten carbonate fuel cell vendors for input on the fuel cell module for the PSDF.

  4. Advanced Thermal Storage for Central Receivers with Supercritical Coolants

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, Bruce D.

    2010-06-15

    The principal objective of the study is to determine if supercritical heat transport fluids in a central receiver power plant, in combination with ceramic thermocline storage systems, offer a reduction in levelized energy cost over a baseline nitrate salt concept. The baseline concept uses a nitrate salt receiver, two-tank (hot and cold) nitrate salt thermal storage, and a subcritical Rankine cycle. A total of 6 plant designs were analyzed, as follows: Plant Designation Receiver Fluid Thermal Storage Rankine Cycle Subcritical nitrate salt Nitrate salt Two tank nitrate salt Subcritical Supercritical nitrate salt Nitrate salt Two tank nitrate salt Supercritical Low temperature H2O Supercritical H2O Two tank nitrate salt Supercritical High temperature H2O Supercritical H2O Packed bed thermocline Supercritical Low temperature CO2 Supercritical CO2 Two tank nitrate salt Supercritical High temperature CO2 Supercritical CO2 Packed bed thermocline Supercritical Several conclusions have been drawn from the results of the study, as follows: 1) The use of supercritical H2O as the heat transport fluid in a packed bed thermocline is likely not a practical approach. The specific heat of the fluid is a strong function of the temperatures at values near 400 °C, and the temperature profile in the bed during a charging cycle is markedly different than the profile during a discharging cycle. 2) The use of supercritical CO2 as the heat transport fluid in a packed bed thermocline is judged to be technically feasible. Nonetheless, the high operating pressures for the supercritical fluid require the use of pressure vessels to contain the storage inventory. The unit cost of the two-tank nitrate salt system is approximately $24/kWht, while the unit cost of the high pressure thermocline system is nominally 10 times as high. 3) For the supercritical fluids, the outer crown temperatures of the receiver tubes are in the range of 700 to 800 °C. At temperatures of 700 °C and above

  5. Safety considerations regarding the use of propane and other liquefied gases as coolants for rapid freezing purposes.

    Science.gov (United States)

    Ryan, K P; Liddicoat, M I

    1987-09-01

    Liquid propane and similar coolants are used in the rapid freezing of biological specimens. These coolants form explosive gas mixtures with air, with a 14,000-fold increase in volume over that of the liquid. The liquefied gases have high vapour pressures and, unless they are maintained below their flashpoint, the vapour above them will reach ignitable concentrations. The flashpoint of liquid propane is -104 degrees C. Ethane has a higher vapour pressure, and vapour mixed with air above liquid ethane can be ignited at a coolant temperature of -130 degrees C. The danger is minimized if the coolant is maintained near its freezing point and under a nitrogen atmosphere, in a fume cupboard. Liquid nitrogen evaporates to a 690-fold increase in volume at room temperature. It is important to ventilate the working area, especially when cryo-sectioning in a small room, otherwise there is a possibility of asphyxiation.

  6. Automated Aflatoxin Analysis Using Inline Reusable Immunoaffinity Column Cleanup and LC-Fluorescence Detection.

    Science.gov (United States)

    Rhemrev, Ria; Pazdanska, Monika; Marley, Elaine; Biselli, Scarlett; Staiger, Simone

    2015-01-01

    A novel reusable immunoaffinity cartridge containing monoclonal antibodies to aflatoxins coupled to a pressure resistant polymer has been developed. The cartridge is used in conjunction with a handling system inline to LC with fluorescence detection to provide fully automated aflatoxin analysis for routine monitoring of a variety of food matrixes. The handling system selects an immunoaffinity cartridge from a tray and automatically applies the sample extract. The cartridge is washed, then aflatoxins B1, B2, G1, and G2 are eluted and transferred inline to the LC system for quantitative analysis using fluorescence detection with postcolumn derivatization using a KOBRA® cell. Each immunoaffinity cartridge can be used up to 15 times without loss in performance, offering increased sample throughput and reduced costs compared to conventional manual sample preparation and cleanup. The system was validated in two independent laboratories using samples of peanuts and maize spiked at 2, 8, and 40 μg/kg total aflatoxins, and paprika, nutmeg, and dried figs spiked at 5, 20, and 100 μg/kg total aflatoxins. Recoveries exceeded 80% for both aflatoxin B1 and total aflatoxins. The between-day repeatability ranged from 2.1 to 9.6% for aflatoxin B1 for the six levels and five matrixes. Satisfactory Z-scores were obtained with this automated system when used for participation in proficiency testing (FAPAS®) for samples of chilli powder and hazelnut paste containing aflatoxins.

  7. Effect of heat release in the coolant on the stability of a water-cooled-water-moderated reactor

    Energy Technology Data Exchange (ETDEWEB)

    Vdovin, S.I.; Sabaev, E.F.

    1985-10-01

    The authors use exact kinetic equations in order to estimate the effect of heat release on the coolant. The authors found that the instantaneous release of even an insignificant part of the heat in the coolant exerts a significant stabilizing effect on the stability of a boiling reactor, especially in the case of a high steam content at the core outlet, which must be taken into consideration when analyzing the dynamics of boiling reactors.

  8. Experimental simulation of low rate primary coolant leaks. For the case of vessel head penetrations affected by through wall cracking

    Energy Technology Data Exchange (ETDEWEB)

    You, D.; Feron, D. [CEA-Saclay - DEN/DPC/SCCME, 91 - Gif-sur-Yvette (France); Turluer, G. [CEA-Fontenay-aux-Roses - IPSN/DES/SAMS, 92 - Fontenay-aux-Roses (France)

    2002-07-01

    An experimental simulation of primary coolant leaks was carried out to determine how the composition of the leaking liquid would change. The experiment used the EVA experimental setup, specially designed for quantitatively investigating concentration phenomena driven by evaporation. The test showed that the final composition, obtained from a solution representative of the primary coolant at the beginning of the cycle, is highly concentrated and slightly acid. The experimental results are compared with those obtained using the MULTEQ software. (authors)

  9. From Cleanup to Stewardship. A companion report to Accelerating Cleanup: Paths to Closure and background information to support the scoping process required for the 1998 PEIS Settlement Study

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    1999-10-01

    Long-term stewardship is expected to be needed at more than 100 DOE sites after DOE's Environmental Management program completes disposal, stabilization, and restoration operations to address waste and contamination resulting from nuclear research and nuclear weapons production conducted over the past 50 years. From Cleanup to stewardship provides background information on the Department of Energy (DOE) long-term stewardship obligations and activities. This document begins to examine the transition from cleanup to long-term stewardship, and it fulfills the Secretary's commitment to the President in the 1999 Performance Agreement to provide a companion report to the Department's Accelerating Cleanup: Paths to Closure report. It also provides background information to support the scoping process required for a study on long-term stewardship required by a 1998 Settlement Agreement.

  10. Phenomena identification and ranking tables for Westinghouse AP600 small break loss-of-coolant accident, main steam line break, and steam generator tube rupture scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, G.E.; Fletcher, C.D.; Davis, C.B. [and others

    1997-06-01

    This report revision incorporates new experimental evidence regarding AP600 behavior during small break loss-of-coolant accidents. This report documents the results of Phenomena Identification and Ranking Table (PIRT) efforts for the Westinghouse AP600 reactor. The purpose of this PIRT is to identify important phenomena so that they may be addressed in both the experimental programs and the RELAP5/MOD3 systems analysis computer code. In Revision of this report, the responses of AP600 during small break loss-of-coolant accident, main steam line break, and steam generator tube rupture accident scenarios were evaluated by a committee of thermal-hydraulic experts. Committee membership included Idaho National Engineering and Environmental Laboratory staff and recognized thermal-hydraulic experts from outside of the laboratory. Each of the accident scenarios was subdivided into separate, sequential periods or phases. Within each phase, the plant behavior is controlled by, at most, a few thermal-hydraulic processes. The committee identified the phenomena influencing those processes, and ranked & influences as being of high, medium, low, or insignificant importance. The primary product of this effort is a series of tables, one for each phase of each accident scenario, describing the thermal-hydraulic phenomena judged by the committee to be important, and the relative ranking of that importance. The rationales for the phenomena selected and their rankings are provided. This document issue incorporates an update of the small break loss-of-coolant accident portion of the report. This revision is the result of the release of experimental evidence from AP600-related integral test facilities (ROSA/AP600, OSU, and SPES) and thermal-hydraulic expert review. The activities associated with this update were performed during the period from June 1995 through November 1996. 8 refs., 26 figs., 42 tabs.

  11. Assessment of alkali metal coolants for the ITER blanket

    Science.gov (United States)

    Natesan, K.; Reed, C. B.; Mattas, R. F.

    1994-06-01

    The blanket system is one of the most important components of a fusion reactor because it has a major impact on both the economics and safety of fusion energy. The primary functions of the blanket in a deuterium/tritium-fueled fusion reactor are to convert the fusion energy into sensible heat and to breed tritium for the fuel cycle. The blanket comparison and selection study, conducted earlier, described the overall comparative performance of different blanket concepts, including liquid metal, molten salt, water, and helium. This paper will discuss the ITER requirements for a self-cooled blanket concept with liquid lithium and for indirectly cooled concepts that use other alkali metals such as NaK. The paper addresses the thermodynamics of interactions between the liquid metals (e.g., lithium and NaK) and structural materials (e.g., V-base alloys), together with associated corrosion/compatibility issues. Available experimental data are used to assess the long-term performance of the first wall in a liquid metal environment. Other key issues include development of electrical insulator coatings on the first-wall structural material to MHD pressure drop, and tritium permeation/inventory in self-cooled and indirectly cooled concepts. Acceptable types of coatings (based on their chemical compatibility and physical properties) are identified, and surface-modification avenues to achieve these coatings on the first wall are discussed. The assessment examines the extent of our knowledge on structural materials performance in liquid metals and identifies needed research and development in several of the areas in order to establish performance envelopes for the first wall in a liquid-metal environment.

  12. Breakup of jet and drops during premixing phase of fuel coolant interactions

    Energy Technology Data Exchange (ETDEWEB)

    Haraldsson, Haraldur Oskar

    2000-05-01

    During the course of a hypothetical severe accident in a light water reactor, molten liquid may be introduced into a volatile coolant, which, under certain conditions, results in explosive interactions. Such fuel-coolant interactions (FCI) are characterised by an initial pre-mixing phase during which the molten liquid, metallic or oxidic in nature, undergoes a breakup (fragmentation) process which significantly increase the area available for melt-coolant contact, and thus energy transfer. Although substantial progress in the understanding of phenomenology of the FCI events has been achieved in recent years, there remain uncertainties in describing the primary and secondary breakup processes. The focus of this work is on the melt jet and drop breakup during the premixing phase of FCI. The objectives are to gain insight into the premixing phase of the FCI phenomena, to determine what fraction of the melt fragments and determine the size distribution. The approach is to perform experiments with various simulant materials, at different scales, different conditions and with variation of controlling parameters affecting jet and drop breakup processes. The analysis approach is to investigate processes at different level of detail and complexity to understand the physics, to rationalise experimental results and to develop and validate models. In the first chapter a brief introduction and review of the status of the FCI phenomena is performed. A review of previous and current experimental projects is performed. The status of the experimental projects and major findings are outlined. The first part of the second chapter deals with experimental investigation of jet breakup. Two series of experiments were performed with low and high temperature jets. The low temperature experiments employed cerrobend-70 as jet liquid. A systematic investigation of thermal hydraulic conditions and melt physical properties on the jet fragmentation and particle debris characteristics was

  13. Answers to frequently asked questions about cleanup activities at Three Mile Island, Unit 2

    Energy Technology Data Exchange (ETDEWEB)

    1984-03-01

    This question-and-answer report provides answers in nontechnical language to frequently asked questions about the status of cleanup activities at Three Mile Island, Unit 2. The answers update information first prepared in 1981, shortly after the cleanup got under way. Since then, a variety of important developments in the cleanup has occurred. The information in the report should be read in conjunction with NUREG 1060, a discussion of increased occupational exposure estimates for the cleanup. The questions and answers in this report cover purpose and community involvement, decontamination of water and reactor, fuel removal, radwaste transport, environmental impact, social and economic effects, worker exposures and safety, radiation monitoring, potential for accidents, and schedule and funding.

  14. Government Districts, Other, City trash cleanup zones, Published in 2006, Freelance.

    Data.gov (United States)

    NSGIC GIS Inventory (aka Ramona) — This Government Districts, Other dataset, was produced all or in part from Hardcopy Maps information as of 2006. It is described as 'City trash cleanup zones'. Data...

  15. Investigation of the moving-bed copper oxide process for flue gas cleanup

    Energy Technology Data Exchange (ETDEWEB)

    Pennline, H.W.; Hoffman, J.S.; Yeh, J.T. [Dept. of Energy, Pittsburgh, PA (United States). Pittsburgh Energy Technology Center; Resnik, K.P.; Vore, P.A. [Parsons Power Group, Inc., Pittsburgh, PA (United States)

    1996-12-31

    The Moving-Bed Copper Oxide Process is a dry, regenerable sorbent technique that uses supported copper oxide sorbent to simultaneously remove SO{sub 2} and NO{sub x} emissions from flue gas generated by coal combustion. The process can be integrated into the design of advanced power systems, such as the Low-Emission Boiler System (LEBS) or the High-Performance Power System (HIPPS). This flue gas cleanup technique is currently being evaluated in a life-cycle test system (LCTS) with a moving-bed flue gas contactor at DOE`s Pittsburgh Energy Technology Center. An experimental data base being established will be used to verify reported technical and economic advantages, optimize process conditions, provide scaleup information, and validate absorber and regenerator mathematical models. In this communication, the results from several process parametric test series with the LCTS are discussed. The effects of various absorber and regenerator parameters on sorbent performance (e.g., SO{sub 2} removal) were investigated. Sorbent spheres of 1/8-in diameter were used as compared to 1/16-in sized sorbent of a previous study. Also discussed are modifications to the absorber to improve the operability of the LCTS when fly ash is present during coal combustion.

  16. Thermal management systems and methods

    Science.gov (United States)

    Gering, Kevin L.; Haefner, Daryl R.

    2006-12-12

    A thermal management system for a vehicle includes a heat exchanger having a thermal energy storage material provided therein, a first coolant loop thermally coupled to an electrochemical storage device located within the first coolant loop and to the heat exchanger, and a second coolant loop thermally coupled to the heat exchanger. The first and second coolant loops are configured to carry distinct thermal energy transfer media. The thermal management system also includes an interface configured to facilitate transfer of heat generated by an internal combustion engine to the heat exchanger via the second coolant loop in order to selectively deliver the heat to the electrochemical storage device. Thermal management methods are also provided.

  17. Major clean-up effort in the ATLAS cavern

    CERN Multimedia

    Marzio Nessi

    On Tuesday 10 October, 58 ATLAS collaborators volunteered to give a hand for a major clean-up of the ATLAS detector prior to the toroid magnet ramp-up. This special task monopolised all of the technical coordination team and eight supervisors to oversee the volunteers who were assigned to two separate five-hour shifts. The volunteers removed all sorts of loose material inside and outside the detector, focusing mainly on potentially magnetic material lost inside the detector and dirt accumulated over several months, not to mention zillions of clipped cable ties! The technical crew provided 120 garbage bags and all were used. All sorts of material that had been lost inside the detector by various people was retrieved, in particular small tools which could potentially damage the detector, as well as metallic fillings hazardous for the electronics once the magnet will be ramped up. A more detailed inspection followed for all the inside of the detector, making sure the current on the magnet could be raised to 5KA ...

  18. Thermal cleanups using dynamic underground stripping and hydrous pyrolysis oxidation

    Energy Technology Data Exchange (ETDEWEB)

    Aines, R D; Knauss, K; Leif, R; Newmark, R L

    1999-05-01

    In the early 1990s, in collaboration with the School of Engineering at the University of California, Berkeley, Lawrence Livermore National Laboratory developed dynamic underground stripping (DUS), a method for treating subsurface contaminants with heat that is much faster and more effective than traditional treatment methods. more recently, Livermore scientists developed hydrous pyrolysis/oxidation (HPO), which introduces both heat and oxygen to the subsurface to convert contaminants in the ground to such benign products as carbon dioxide, chloride ion, and water. This process has effectively destroyed all contaminants it encountered in laboratory tests. With dynamic underground stripping, the contaminants are vaporized and vacuumed out of the ground, leaving them still to be destroyed elsewhere. Hydrous pyrolysis/oxidation technology takes the cleanup process one step further by eliminating the treatment, handling, and disposal requirements and destroying the contamination in the ground. When used in combination, HPO is especially useful in the final polishing of a site containing significant free-product contaminant, once the majority of the contaminant has been removed.

  19. Non-cancer morbidity among Estonian Chernobyl cleanup workers: a register-based cohort study

    OpenAIRE

    Rahu, Kaja; Bromet, Evelyn J.; Hakulinen, Timo; Auvinen, Anssi; Uusküla, Anneli; Rahu, Mati

    2014-01-01

    Objective To examine non-cancer morbidity in the Estonian Chernobyl cleanup workers cohort compared with the population sample with special attention to radiation-related diseases and mental health disorders. Design Register-based cohort study. Setting Estonia. Participants An exposed cohort of 3680 men (cleanup workers) and an unexposed cohort of 7631 men (population sample) were followed from 2004 to 2012 through the Population Registry and Health Insurance Fund database. Methods Morbidity ...

  20. Spreading, retention and clean-up of oil spills. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, Jr, M P

    1976-05-01

    This study reviews and assesses the technology of oil spill spreading, retention and cleanup and proposes research needs in these areas. Sources of oil spills are analyzed and the difficulty of gathering meaningful statistics is discussed. Barrier technology is reviewed and problem areas analyzed. Natural and forced biodegradation and natural and chemical dispersion of oil spills are considered. Research recommendations are categorized under the following two headings (1) Preventive techniques and (2) Containment, Cleanup and Dispersion.

  1. Predicting the conditions under which vibroacoustic resonances with external periodic loads occur in the primary coolant circuits of VVER-based NPPs

    Science.gov (United States)

    Proskuryakov, K. N.; Fedorov, A. I.; Zaporozhets, M. V.

    2015-08-01

    The accident at the Japanese Fukushima Daiichi nuclear power plant (NPP) caused by an earthquake showed the need of taking further efforts aimed at improving the design and engineering solutions for ensuring seismic resistance of NPPs with due regard to mutual influence of the dynamic processes occurring in the NPP building structures and process systems. Resonance interaction between the vibrations of NPP equipment and coolant pressure pulsations leads to an abnormal growth of dynamic stresses in structural materials, accelerated exhaustion of equipment service life, and increased number of sudden equipment failures. The article presents the results from a combined calculation-theoretical and experimental substantiation of mutual amplification of two kinds of external periodic loads caused by rotation of the reactor coolant pump (RCP) rotor and an earthquake. The data of vibration measurements at an NPP are presented, which confirm the predicted multiple amplification of vibrations in the steam generator and RCP at a certain combination of coolant thermal-hydraulic parameters. It is shown that the vibration frequencies of the main equipment may fall in the frequency band corresponding to the maximal values in the envelope response spectra constructed on the basis of floor accelerograms. The article presents the results from prediction of conditions under which vibroacoustic resonances with external periodic loads take place, which confirm the occurrence of additional earthquake-induced multiple growth of pressure pulsation intensity in the steam generator at the 8.3 Hz frequency and additional multiple growth of vibrations of the RCP and the steam generator cold header at the 16.6 Hz frequency. It is shown that at the elastic wave frequency equal to 8.3 Hz in the coolant, resonance occurs with the frequency of forced vibrations caused by the rotation of the RCP rotor. A conclusion is drawn about the possibility of exceeding the design level of equipment vibrations

  2. Conducting water chemistry of the secondary coolant circuit of VVER-based nuclear power plant units constructed without using copper containing alloys

    Science.gov (United States)

    Tyapkov, V. F.

    2014-07-01

    The secondary coolant circuit water chemistry with metering amines began to be put in use in Russia in 2005, and all nuclear power plant units equipped with VVER-1000 reactors have been shifted to operate with this water chemistry for the past seven years. Owing to the use of water chemistry with metering amines, the amount of products from corrosion of structural materials entering into the volume of steam generators has been reduced, and the flow-accelerated corrosion rate of pipelines and equipment has been slowed down. The article presents data on conducting water chemistry in nuclear power plant units with VVER-1000 reactors for the secondary coolant system equipment made without using copper-containing alloys. Statistical data are presented on conducting ammonia-morpholine and ammonia-ethanolamine water chemistries in new-generation operating power units with VVER-1000 reactors with an increased level of pH. The values of cooling water leaks in turbine condensers the tube system of which is made of stainless steel or titanium alloy are given.

  3. Advanced Neutron Source Reactor (ANSR) phenomena identification and ranking (PIR) for large break loss of coolant accidents (LBLOCA)

    Energy Technology Data Exchange (ETDEWEB)

    Ruggles, A. E. [Oak Ridge National Lab., TN (United States); Tennessee Univ., Knoxville, TN (United States); Cheng, L. Y. [Brookhaven National Lab., Upton, NY (United States); Dimenna, R. A. [Westinghouse Savannah River Co., Aiken, SC (United States); Griffith, P. [Massachusetts Inst. of Tech., Cambridge, MA (United States); Wilson, G. E. [Idaho National Engineering Lab., Idaho Falls, ID (United States)

    1994-06-01

    A team of experts in reactor analysis conducted a phenomena identification and ranking (PIR) exercise for a large break loss-of-coolant accident (LBLOCA) in the Advanced Neutron source Reactor (ANSR). The LBLOCA transient is broken into two separate parts for the PIR exercise. The first part considers the initial depressurization of the system that follows the opening of the break. The second part of the transient includes long-term decay heat removal after the reactor is shut down and the system is depressurized. A PIR is developed for each part of the LBLOCA. The ranking results are reviewed to establish if models in the RELAP5-MOD3 thermalhydraulic code are adequate for use in ANSR LBLOCA simulations. Deficiencies in the RELAP5-MOD3 code are identified and existing data or models are recommended to improve the code for this application. Experiments were also suggested to establish models for situations judged to be beyond current knowledge. The applicability of the ANSR PIR results is reviewed for the entire set of transients important to the ANSR safety analysis.

  4. Project strategy for clean-up of sedimentary radioactive material in Fukushima bay areas using snake-like robotics

    Directory of Open Access Journals (Sweden)

    Cho Hyo Sung

    2015-01-01

    Full Text Available The snake-like robot is used for clean-up project in Fukushima nuclear disaster site. The contaminated water at the Fukushima Daiichi nuclear power plants has been purified by the water treatment system, called Advanced Liquid Processing System, co-developed by Japanese and international technologies. The system is used to remove most remaining radioactive contaminants in water that had to be stored at the facility. In this paper, a snake-like robot, incorporated with Advanced Liquid Processing System is introduced for the severe accident in the nuclear power plants in which human cannot control the cleaning-up in the sea where the radioactive materials have been submerged and some resolved in the sea water. The effective strategy of the cleaning-up is analyzed from the environmental protection aspect with the snake's biomechanics and radioactive hazards.

  5. CORTICAL CLEANUP WITHOUT SIDE PORT IN SMALL INCISION CATARACT SURGERY

    Directory of Open Access Journals (Sweden)

    Udaya Kumar

    2015-11-01

    Full Text Available The aim of study was to achieve complete cortical cleanup and avoid problems related with sideport during Small Incision Cataract Surgery (SICS so as to have a good visual out come with minimal recovery period, and a better quality of life. After nucleus delivery, cortical cleanup is an important step in any cataract surgical procedure. Cortex especially subincisional area (11 to 1 o’clock is difficult to manage intraoperatively. Bimanual irrigation aspiration through two side ports, aspiration by J cannula, iris massage manoeuver, ice cream scoop manoeuver are various techniques of cortical matter aspiration. We acquired the technique of aspiration of subincisional cortex without using side port in all cases by paying attention on type of cataract, status of pupil, use of Adrenalin mixed BSS intraoperatively, Tunnel construction, Capsulorhexis size and capsular rim size at 12 o’clock. MATERIAL AND METHODS In this retrospective study of 1 year from 2013 to 2014, 60 patients (60 eyes aged 40 years or older attending the General Ophthalmic Department were included in the study group with another group of 60 patients (60 eyes as controls. The study was on age related cataracts which are basically. 1 Cortical cataract 2 Nuclear cataract 3 Subcapsular cataract. Proper assessment of cortical cataract based on its maturity such as a Immature b Mature c Hyper mature and d Morgagnian cataract, nucleus for its opalescence and color, size of posterior subcapsular opacity and pupillary status (Dilating well or not with mydriatics were taken into consideration. Eyes with pseudoexfoliation having poor pupillary dilation were also included. Eyes with congenital anomalies, congenital cataract, gross corneal and retinal pathologies, and glaucoma were excluded. RESULTS Among 60 study eyes in the study group 35 presented with cortical, 20 with nuclear cataract and 5 with posterior subcapsular cataracts. In 58(96.6% cases, sideport was not required; 3(5% eyes

  6. North Slope (Wahluke Slope) expedited response action cleanup plan

    Energy Technology Data Exchange (ETDEWEB)

    1994-02-01

    The purpose of this action is to mitigate any threat to public health and the environment from hazards on the North Slope and meet the expedited response action (ERA) objective of cleanup to a degree requiring no further action. The ERA may be the final remediation of the 100-I-3 Operable Unit. A No Action record of decision (ROD) may be issued after remediation completion. The US Department of Energy (DOE) currently owns or administers approximately 140 mi{sup 2} (about 90,000 acres) of land north and east of the Columbia River (referred to as the North Slope) that is part of the Hanford Site. The North Slope, also commonly known as the Wahluke Slope, was not used for plutonium production or support facilities; it was used for military air defense of the Hanford Site and vicinity. The North Slope contained seven antiaircraft gun emplacements and three Nike-Ajax missile positions. These military positions were vacated in 1960--1961 as the defense requirements at Hanford changed. They were demolished in 1974. Prior to government control in 1943, the North Slope was homesteaded. Since the initiation of this ERA in the summer of 1992, DOE signed the modified Hanford Federal Agreement and Consent Order (Tri-Party Agreement) with the Washington Department of Ecology (Ecology) and the US Environmental Protection Agency (EPA), in which a milestone was set to complete remediation activities and a draft closeout report by October 1994. Remediation activities will make the North Slope area available for future non-DOE uses. Thirty-nine sites have undergone limited characterization to determine if significant environmental hazards exist. This plan documents the results of that characterization and evaluates the potential remediation alternatives.

  7. A passively-safe fusion reactor blanket with helium coolant and steel structure

    Energy Technology Data Exchange (ETDEWEB)

    Crosswait, K.M.

    1994-04-01

    Helium is attractive for use as a fusion blanket coolant for a number of reasons. It is neutronically and chemically inert, nonmagnetic, and will not change phase during any off-normal or accident condition. A significant disadvantage of helium, however, is its low density and volumetric heat capacity. This disadvantage manifests itself most clearly during undercooling accident conditions such as a loss of coolant accident (LOCA) or a loss of flow accident (LOFA). This thesis describes a new helium-cooled tritium breeding blanket concept which performs significantly better during such accidents than current designs. The proposed blanket uses reduced-activation ferritic steel as a structural material and is designed for neutron wall loads exceeding 4 MW/m{sup 2}. The proposed geometry is based on the nested-shell concept developed by Wong, but some novel features are used to reduce the severity of the first wall temperature excursion. These features include the following: (1) A ``beryllium-joint`` concept is introduced, which allows solid beryllium slabs to be used as a thermal conduction path from the first wall to the cooler portions of the blanket. The joint concept allows for significant swelling of the beryllium (10 percent or more) without developing large stresses in the blanket structure. (2) Natural circulation of the coolant in the water-cooled shield is used to maintain shield temperatures below 100 degrees C, thus maintaining a heat sink close to the blanket during the accident. This ensures the long-term passive safety of the blanket.

  8. Prospects for pyrolysis technologies in managing municipal, industrial, and DOE cleanup wastes

    Energy Technology Data Exchange (ETDEWEB)

    Reaven, S.J. [State Univ. of New York, Stony Brook, NY (United States)

    1994-12-01

    Pyrolysis converts portions of municipal solid wastes, hazardous wastes, and special wastes such as tires, medical wastes, and even old landfills into solid carbon and a liquid or gaseous hydrocarbon stream. Pyrolysis heats a carbonaceous waste stream typically to 290--900 C in the absence of oxygen, and reduces the volume of waste by 90% and its weight by 75%. The solid carbon char has existing markets as an ingredient in many manufactured goods, and as an adsorbent or filter to sequester certain hazardous wastes. Pyrolytic gases may be burned as fuel by utilities, or liquefied for use as chemical feedstocks, or low-pollution motor vehicle fuels and fuel additives. This report analyzes the potential applications of pyrolysis in the Long Island region and evaluates for the four most promising pyrolytic systems their technological and commercial readiness, their applicability to regional waste management needs, and their conformity with DOE requirements for environmental restoration and waste management. This summary characterizes their engineering performance, environmental effects, costs, product applications, and markets. Because it can effectively treat those wastes that are inadequately addressed by current systems, pyrolysis can play an important complementing role in the region`s existing waste management strategy. Its role could be even more significant if the region moves away from existing commitments to incineration and MSW composting. Either way, Long Island could become the center for a pyrolysis-based recovery services industry serving global markets in municipal solid waste treatment and hazardous waste cleanup. 162 refs.

  9. Proceedings of the CSNI specialists meeting on fuel-coolant interactions

    Energy Technology Data Exchange (ETDEWEB)

    None

    1994-03-01

    A specialists meeting on fuel-coolant interactions was held in Santa Barbara, CA from January 5-7, 1993. The meeting was sponsored by the United States Nuclear Regulatory Commission in collaboration with the Committee on the Safety of Nuclear Installation (CSNI) of the OECD Nuclear Energy Agency (NEA) and the University of California at Santa Barbara. The objectives of the meeting are to cross-fertilize on-going work, provide opportunities for mutual check points, seek to focus the technical issues on matters of practical significance and re-evaluate both the objectives as well as path of future research. Individual papers have been cataloged separately.

  10. Physical properties of heavy liquid-metal coolants in a wide temperature range

    Directory of Open Access Journals (Sweden)

    Borisenko A.

    2011-05-01

    Full Text Available The pulse-phase method, the gamma-attenuation method and the method of dumping oscillation of a crucible with a melt were used for measuring the velocity of sound, the density and the kinematic viscosity of a set of liquid-metal coolants for perspective nuclear reactors. There are liquid gallium, indium, tin, lead, bismuth and lead-bismuth eutectic alloy among the melts investigated. The accuracy of the measurements was as high as 0.3%, 0.2 to 0.4% and 1.5% for the ultrasound velocity, the density and the viscosity, correspondingly.

  11. Failures of the thermal barriers of 900 MWe reactor coolant pumps

    Energy Technology Data Exchange (ETDEWEB)

    Peyrouty, P.

    1996-12-01

    This report describes the anomalies encountered in the thermal barriers of the reactor coolant pumps in French 900 MWe PWR power stations. In addition to this specific problem, it demonstrates how the fortuitous discovery of a fault during a sampling test enabled faults of a generic nature to be revealed in components which were not subject to periodic inspection, the failure of which could seriously affect safety. This example demonstrates the risk which can be associated with the deterioration in areas which are not examined periodically and for which there are no preceding signs which would make early detection of deterioration possible.

  12. Analysis of a small break loss-of-coolant accident of pressurized water reactor by APROS

    Energy Technology Data Exchange (ETDEWEB)

    Al-Falahi, A. [Helsinki Univ. of Technology, Espoo (Finland); Haennine, M. [VTT Energy, Espoo (Finland); Porkholm, K. [IVO International, Ltd., Vantaa (Finland)

    1995-09-01

    The purpose of this paper is to study the capability of APROS (Advanced PROcess Simulator) code to simulate the real plant thermal-hydraulic transient of a Small Break Loss-Of-Coolant Accident (SBLOCA) of Loss-Of-Fluid Test (LOFT) facility. The LOFT is a scaled model of a Pressurized Water Reactor (PWR). This work is a part of a larger validation of the APROS thermal-hydraulic models. The results of SBLOCA transient calculated by APROS showed a reasonable agreement with the measured data.

  13. The development of Sn-Li coolant/breeding material for APEX/ALPS applications.

    Energy Technology Data Exchange (ETDEWEB)

    Sze, D.-K.

    1999-07-08

    A Sn-Li alloy has been identified to be a coolant/breeding material for D-T fusion applications. The key feature of this material is its very low vapor pressure, which will be very useful for free surface concepts employed in APEX, ALPS and inertial confinement fission. The vapor is dominated by lithium, which has very low Z. Initial assessment of the material indicates acceptable tritium breeding capability, high thermal conductivity, expected low tritium volubility, and expected low chemical reactivities with water and air. Some key concerns are the high activation and material compatibility issues. The initial assessment of this material, for fission applications, is presented in this paper.

  14. Definition of loss-of-coolant accident radiation source. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    1978-02-01

    Meaningful qualification testing of nuclear reactor components requires a knowledge of the radiation fields expected in a loss-of-coolant accident (LOCA). The overall objective of this program is to define the LOCA source terms and compare these with the output of various simulators employed for radiation qualification testing. The basis for comparison will be the energy deposition in a model reactor component. The results of the calculations are presented and some interpretation of the results given. The energy release rates and spectra were validated by comparison with other calculations using different codes since experimental data appropriate to these calculations do not exist.

  15. Improved solidification influence modelling for Eulerian fuel-coolant interaction codes

    Energy Technology Data Exchange (ETDEWEB)

    Ursic, Mitja, E-mail: mitja.ursic@ijs.s [Jozef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Leskovar, Matjaz; Mavko, Borut [Jozef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia)

    2011-04-15

    Steam explosion experiments revealed important differences in the efficiency between simulant alumina and oxidic corium melts. The experimentally observed differences are importantly attributed to the differences in the melt droplets solidification and void production, which are limiting phenomena in the steam explosion process and have to be adequately modelled in fuel-coolant interaction codes. This article focuses on the modelling of the solidification effect. An improved solidification influence modelling approach for Eulerian fuel-coolant interaction codes was developed and is presented herein. The solidification influence modelling in fuel-coolant interaction codes is strongly related to the modelling of the temperature profile and the mechanical effect of the crust on the fragmentation process. Therefore the first objective was to introduce an improved temperature profile modelling and a fragmentation criterion for partly solidified droplets. The fragmentation criterion was based on the established modified Weber number, which considers the crust stiffness as a stabilizing force acting to retain the crust under presence of the hydrodynamic forces. The modified Weber number was validated on experimental data. The application of the developed improved solidification influence modelling enables an improved determination of the melt droplet mass, which can be efficiently involved in the fine fragmentation during the steam explosion process. Additionally, also the void production modelling is improved, because it is strongly related to the temperature profile modelling in the frame of the solidification influence modelling. Therefore the second objective was to enable an improved solidification influence modelling in codes with an Eulerian formulation of the droplet field. Two additional transported model parameters based on the most important droplets features regarding the fuel-coolant interaction behaviour, were derived. First, the crust stiffness was

  16. Fuel-coolant interaction (FCI) phenomena in reactor safety. Current understanding and future research needs

    Energy Technology Data Exchange (ETDEWEB)

    Speis, T.P. [Maryland Univ., College Park, MD (United States); Basu, S.

    1998-01-01

    This paper gives an account of the current understanding of fuel-coolant interaction (FCI) phenomena in the context of reactor safety. With increased emphasis on accident management and with emerging in-vessel core melt retention strategies for advanced light water reactor (ALWR) designs, recent interest in FCI has broadened to include an evaluation of potential threats to the integrity of reactor vessel lower head and ex-vessel structural support, as well as the role of FCI in debris quenching and coolability. The current understanding of FCI with regard to these issues is discussed, and future research needs to address the issues from a risk perspective are identified. (author)

  17. A novel polymer inclusion membrane based method for continuous clean-up of thiocyanate from gold mine tailings water.

    Science.gov (United States)

    Cho, Youngsoo; Cattrall, Robert W; Kolev, Spas D

    2018-01-05

    Thiocyanate is present in gold mine tailings waters in concentrations up to 1000mgL(-1) and this has a serious environmental impact by not allowing water reuse in the flotation of gold ore. This significantly increases the consumption of fresh water and the amount of wastewater discharged in tailings dams. At the same time thiocyanate in tailings waters often leads to groundwater contamination. A novel continuous membrane-based method for the complete clean-up of thiocyanate in concentrations as high as 1000mgL(-1) from its aqueous solutions has been developed. It employs a flat sheet polymer inclusion membrane (PIM) of composition 70wt% PVC, 20wt% Aliquat 336 and 10wt% 1-tetradecanol which separates counter-current streams of a feed thiocyanate solution and a 1M NaNO3 receiving solution. The PIM-based system has been operated continuously for 45days with 99% separation efficiency. The volume of the receiving solution has been drastically reduced by recirculating it and continuously removing thiocyanate by precipitating it with in-situ generated Cu(I). The newly developed PIM-based thiocyanate clean-up method is environmentally friendly in terms of reagent use and inexpensive with respect to both equipment and running costs. Copyright © 2017 Elsevier B.V. All rights reserved.

  18. Biochemical observations relating to oxidant stress injury in Chernobyl clean-up workers ("liquidators") from Latvia.

    Science.gov (United States)

    Skesters, Andrejs; Zvagule, T; Silova, A; Rusakova, N; Larmane, L; Reste, J; Eglite, M; Rainsford, K D; Callingham, B A; Bake, M-A; Lece, A

    2010-02-01

    To establish if there is further evidence for the long-term oxidant stress injury (as reported previously--Kumerova et al. in Biol Trace Elem Res 77:1-12, 2000) in surviving Chernobyl nuclear power plant (NPP) workers from Latvia. The overall objectives of this study have been to establish if there have been long-term systemic changes in the oxidant/antioxidant status of clean-up workers that might reflect adaptation to the progression of oxidative stress injury. Biochemical analyses of the circulating levels of endogenous oxidants and anti-oxidants were undertaken over two periods (Stage 1 in 1998-1999 and Stage 2 in 2005-2006) at approximately 6-7 years time interval, in order to establish if there have been time-dependent changes in the parameters that may be important for the health of the clean-up workers. The biochemical analyses included (a) plasma levels of the anti-oxidant, selenium, (b) blood and plasma levels of glutathione peroxidase, (c) red blood cell catalase, (d) plasma total oxidant status as lipid peroxides and hydroperoxides, (e) plasma ceruloplasmin, and (f) total blood levels of zinc and copper. The circulating content of lipid peroxides, plasma oxidisability, lipid peroxides, catalase, Zn, and Cu were elevated above normal values at both the stages of this study. Glutathione peroxidase was increased above normal values at Stage 1 but not at Stage 2. The most pronounced changes between Stage 1 and Stage 2 were (a) a reduction by about (1/2) in the content of lipid peroxides and lipid peroxidation, but not in the blood oxidisability and (b) increased plasma selenium. The data show that there may be a partial improvement in the anti-oxidant/oxidant status of the Chernobyl NPP workers over the 7-year period of investigation. The NPP patients may be undergoing progressive reduction in blood oxidants accompanied by adaptation to oxidant stress injury due to the increased anti-oxidant activity measured in their plasma and blood.

  19. Chemical gel barriers as low-cost alternative to containment and in situ cleanup of hazardous wastes to protect groundwater

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-01-01

    Chemical gel barriers are being considered as a low-cost alternative for containment and in situ cleanup of hazardous wastes to protect groundwater. Most of the available gels in petroleum application are non-reactive and relative impermeable, providing a physical barriers for all fluids and contaminants. However, other potential systems can be envisioned. These systems could include gels that are chemically reactive and impermeable such that most phase are captured by the barriers but the contaminants could diffuse through the barriers. Another system that is chemically reactive and permeable could have potential applications in selectivity capturing contaminants while allowing water to pass through the barriers. This study focused on chemically reactive and permeable gel barriers. The gels used in experiment are DuPont LUDOX SM colloidal silica gel and Pfizer FLOPAAM 1330S hydrolyzed polyacrylamide (HPAM) gel.

  20. The premixing and propagation phases of fuel-coolant interactions: a review of recent experimental studies and code developments

    Energy Technology Data Exchange (ETDEWEB)

    Antariksawan, A.R. [Reactor Safety Technology Research Center of BATAN (Indonesia); Moriyama, Kiyofumi; Park, Hyun-sun; Maruyama, Yu; Yang, Yanhua; Sugimoto, Jun

    1998-09-01

    A vapor explosion (or an energetic fuel-coolant interactions, FCIs) is a process in which hot liquid (fuel) transfers its internal energy to colder, more volatile liquid (coolant); thus the coolant vaporizes at high pressure and expands and does works on its surroundings. Traditionally, the energetic fuel-coolant interactions could be distinguished in subsequent stages: premixing (or coarse mixing), triggering, propagation and expansion. Realizing that better and realistic prediction of fuel-coolant interaction consequences will be available understanding the phenomenology in the premixing and propagation stages, many experimental and analytical studies have been performed during more than two decades. A lot of important achievements are obtained during the time. However, some fundamental aspects are still not clear enough; thus the works are directed to that direction. In conjunction, the model/code development is pursuit. This is aimed to provide a scaling tool to bridge the experimental results to the real geometries, e.g. reactor pressure vessel, reactor containment. The present review intends to collect the available information on the recent works performed to study the premixing and propagation phases. (author). 97 refs.

  1. Sensitivity Analysis of Core Damage from Reactor Coolant Pump Seal Leakage during Extended Loss of All AC Power

    Energy Technology Data Exchange (ETDEWEB)

    Park, Da Hee; Kim, Min Gi; Lee, Kyung Jin; Hwang, Su hyun; Lee, Byung Chul [FNC Technology Co. Ltd., Yongin (Korea, Republic of); Yoon, Duk Joo; Lee, Seung Chan [Korea Hydro and Nuclear Power Co. Ltd., Daejeon (Korea, Republic of)

    2015-10-15

    In this study, in order to comprehend the Fukushima accident, the sensitivity analysis was performed to analyze the behavior of Reactor Coolant System (RCS) during ELAP using the RELAP5/MOD3.3 code. The Fukushima accident was caused by tsunami resulted in Station Black Out (SBO) followed by the reactor core melt-down and release of radioactive materials. After the accident, the equipment and strategies for the Extended Loss of All AC Power (ELAP) were recommended strongly. In this analysis, sensitivity studies for the RCP seal failure of the OPR1000 type NPP were performed by using RELAP5/MOD3.3 code. Six cases with different leakage rate of RCP seal were studied for ELAP with operator action or not. The main findings are summarized as follows: (1) Without the operator action, the core uncovery time is determined by the leakage rate of RCP seal. When the leakage rate per RCP seal are 5 gpm, 50 gpm, and 300 gpm respectively, the core uncovery time are 1.62 hr, 1.58 hr, and 1.29 hr respectively. Namely, If the leakage rate of RCP seal was much bigger, the uncover time of core would be shorter. (2) In case that the cooling by SG secondary side was performed using the TDAFP and SG ADV, the core uncovery time was significantly extended.

  2. Development of advanced techniques for life management and inspection of advanced heavy water reactor (AWHR) coolant channel components

    Energy Technology Data Exchange (ETDEWEB)

    Madhusoodanan, K.; Sinha, S.K.; Kumar, K.; Shyam, T.V.; Panwar, S.; Sharma, B.S.V.G. [Bhabha Atomic Research Centre, Reactor Engineering Div., Trombay, Mumbai (India); Sinha, R. K. [Bhabha Atomic Research Centre, Reactor Design and Development Group., Trombay, Mumbai (India)

    2011-07-01

    Operating life of pressure tubes of Pressurized Heavy Water Reactor (PHWR) is limited due to the presence of various issues associated with the material like hydrogen pick up, delayed hydride cracking, axial elongation and increase in diameter due to irradiation creep and growth. Periodic monitoring of the health of the pressure tube under in-situ conditions is essential to ensure the safe operation of the reactor. New designs of reactor call for innovative design philosophy, modification in fabrication route of pressure tube, development of reactor specific tools, both analytical and hardware for assessing the fitness for service of the pressure tube. Feedback from existing reactors has enhanced the understanding about life limiting parameters. This paper gives an insight into the life limiting issues associated with pressure tube and the efforts pursued for development of life management techniques for coolant channel of Advanced Heavy Water Reactor (AHWR) designed in India. The tools and techniques for in-situ property/hydrogen measurement, pulsed eddy current technique for zirconium alloy in-homogeneity characterization, horizontal shear wave EMAT system for dissimilar metal weld inspection, sliver sampling of vertical channel etc. are elaborated in the paper. (author)

  3. Behavior of an improved Zr fuel cladding with oxidation resistant coating under loss-of-coolant accident conditions

    Science.gov (United States)

    Park, Dong Jun; Kim, Hyun Gil; Jung, Yang Il; Park, Jung Hwan; Yang, Jae Ho; Koo, Yang Hyun

    2016-12-01

    This study investigates protective coatings for improving the high temperature oxidation resistance of Zr fuel claddings for light water nuclear reactors. FeCrAl alloy and Cr layers were deposited onto Zr plates and tubes using cold spraying. For the FeCrAl/Zr system, a Mo layer was introduced between the FeCrAl coating and the Zr matrix to prevent inter-diffusion at high temperatures. Both the FeCrAl and Cr coatings improved the oxidation resistance compared to that of the uncoated Zr alloy when exposed to a steam environment at 1200 °C. The ballooning behavior and mechanical properties of the coated cladding samples were studied under simulated loss-of-coolant accident conditions. The coated samples showed higher burst temperatures, lower circumferential strain, and smaller rupture openings compared to the uncoated Zr. Although 4-point bend tests of the coated samples showed a small increase in the maximum load, ring compression tests of a sectioned sample showed increased ductility.

  4. 基于相依竞争失效模型的主泵多状态可靠性分析%Multi-state Reliability for Coolant Pump Based on Dependent Competitive Failure Model

    Institute of Scientific and Technical Information of China (English)

    尚彦龙; 蔡琦; 赵新文; 陈玲

    2013-01-01

    基于核动力主泵运行环境和性能退化机理,考虑自身振动和外部冲击对其性能退化的影响,建立了主泵冲击与退化相依竞争失效过程的可靠度模型。采用该模型计算了考虑性能退化的主泵在振动和外部冲击条件下的退化状态概率和可靠度,为基于使用环境的核动力主泵的多状态可靠性分析提供了一种有效的分析途径。分析结果可为设计变更和维修优化提供决策依据。%By taking into account the effect of degradation due to internal vibration and external shocks , and based on service environment and degradation mechanism of nuclear power plant coolant pump ,a multi-state reliability model of coolant pump was proposed for the system that involves competitive failure process between shocks and degradation .Using this model ,degradation state probability and system reliability were obtained under the consideration of internal vibration and external shocks for the degraded coolant pump .It provided an effective method to reliability analysis for coolant pump in nuclear power plant based on operating environment .The results can provide a decision making basis for design changing and maintenance optimization .

  5. Method for calculating coolant resonance frequencies under normal and accident conditions in nuclear power plants with WWER-type pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Proskuryakov, K.N. (Moskovskij Ehnergeticheskij Inst. (USSR))

    1983-03-01

    Mathematical models are proposed for calculating acoustic oscillation resonance frequencies in the coolant in various components of the WWER type primary circuit (core, steam generator, pressurizer, piping). Due to the correspondence between model calculations and experimental results obtained in operating nuclear power plants, the developed models can be used for practical calculations. The possibility of calculating the eigenfrequencies of the coolant oscillation under different operating conditions leads to the interpretation of operational data, to the analysis of operational conditions, to the detection of coolant boiling in the reactor, and to design changes in order to prevent resonance oscillations within the coolant.

  6. Heavily Oiled Salt Marsh following the Deepwater Horizon Oil Spill, Ecological Comparisons of Shoreline Cleanup Treatments and Recovery

    National Research Council Canada - National Science Library

    Zengel, Scott; Bernik, Brittany M; Rutherford, Nicolle; Nixon, Zachary; Michel, Jacqueline

    2015-01-01

    The Deepwater Horizon oil spill affected hundreds of kilometers of coastal wetland shorelines, including salt marshes with persistent heavy oiling that required intensive shoreline "cleanup" treatment...

  7. Validation of extraction, clean-up and DR CALUX {sup registered} bioanalysis. Pt. I. Feedingstuff

    Energy Technology Data Exchange (ETDEWEB)

    Besselink, H.; Jonas, A.; Brouwer, B. [BioDetection Systems BV (BDS), Amsterdam (Netherlands); Pijnappels, M.; Swinkels, A. [Masterlab BV, Boxmeer (Netherlands)

    2004-09-15

    Feed safety is a high priority issue for the feed sector as it directly impacts at the beginning of the food-chain. Currently stringent EU limit values are in force for dioxins in feedingstuffs for animal and public health protection. Biodetection Systems BV's (BDS) DR CALUX {sup registered} bioassay is a cost-effective and rapid method to measure low levels of dioxins and dioxin-like compounds in various matrices. The use of bioassays for monitoring dioxins in feed allows the (pre)-selection of samples suspected of being contaminated above limit values with dioxins. To permit bioassays to be used for screening feedingstuffs for the presence of dioxins and related compounds, the EU has laid down general requirements for the determination of dioxins and dioxin-like PCBs in feedingstuffs and specific requirements for cell-based bioassays. To ensure reliability the DR CALUX {sup registered} bioassay, validated methods for extraction and DR CALUX {sup registered} analysis are necessary. Within the framework of the development of DR CALUX {sup registered} analysis methods, extraction and clean-up methods for 13 feedingstuffs were evaluated, selected and validated. In this paper we present the results of this substantive multi-matrix feedingstuffs validation study.

  8. Development of a Calicum-Based Sorbent for Hot Gas Cleanup.

    Energy Technology Data Exchange (ETDEWEB)

    Wheelock, T.W.; Constant, K.; Doraiswamy, L.K.; Akiti, T.; Zhu, J.; Amanda, A.; Roe, R.

    1997-09-01

    Further review of the technical literature has provided additional information which will support the development of a superior calcium-based sorbent for hot gas cleanup in IGCC systems. Two general methods of sorbent preparation are being investigated. One method involves impregnating a porous refractory substrate with calcium while another method involves pelletizing lime or other calcium containing materials with a suitable binder. Several potential substrates, which are made of alumina and are commercially available, have been characterized by various methods. The surface area and apparent density of the materials have been measured, and it has been shown that some of the high surface area materials (i.e., 200-400 m{sub 2}/g) undergo a large decrease in surface area when heated to higher temperatures. Some of the lower surface area materials (i.e., 1-30 m{sub 2}/g) have been successfully impregnated with calcium by soaking them in a calcium nitrate solution and then heat treating them to decompose the nitrate. Potentially useful sorbents have also been prepared by pelletizing type I Portland cement and mixtures of cement and lime.

  9. Cleanup of metals and hydrocarbons contaminated soils using the ChemTech process

    Energy Technology Data Exchange (ETDEWEB)

    Stephenson, R.; Yan, V.; Lim, S. [Klohn-Crippen Consultants Ltd., Richmond, BC (Canada)

    1997-10-01

    The ChemTech soil treatment process, an on-site ex-situ system, comprised of a three-phase fluidized bed to scour, emulsify and chemically leach soil contaminants into a process water, was described. The cleaned soils are then removed from the process circuit by means of a hydrodynamic classifier. At this point they are suitable for return to the excavation site. The process was demonstrated on a pilot scale in January 1997 by Klohn-Crippen Consultants at a demonstration program of emerging and innovative technologies sponsored by the Bay Area Defence Conversion Action Team (BADCAT), to assist with the remediation of twelve closing military bases in the San Francisco area. The ChemTest demonstration involved the removal of copper, chromium, lead and zinc from the Hunter Point Naval Reserve, plus treatability tests on a number of other contaminated soil samples. The ChemTech process was selected by federal and state regulatory agencies from 21 proposed technologies on the basis of performance, effectiveness, low cost, and absence of secondary environmental impacts. This paper provides details of the demonstration program, addresses the applicability of the technology to other sites, and provides cost estimates of unit cleanup costs. 3 refs., 4 tabs., 4 figs.

  10. TRANSPORTATION MODAL CHOICE IN COOLANT IMPORTATION THROUGH TOTAL COSTS MINIMIZATION: A CASE STUDY

    Directory of Open Access Journals (Sweden)

    Marcela de Souza Leite

    2016-07-01

    Full Text Available Transportation plays a very significant role when it comes to the costs of a company representing on average 60% of logistics costs, so its management is very important for any company. The transportation modal choice is one of the most important transportation decisions. The purpose of this article is to select the transportation mode which is able to minimize total costs, and consistent with the objectives of customer service on the coolant import, which is used in plasma cutting machines. With the installation of a distribution center in Brazil and the professionalization of the logistics department of the company, it was decided to re-evaluate the transportation mode previously chosen to import some items. To determine the best mode of transportation was used basic compensation costs, in other words the cost compensation of using the shuttle service to the indirect cost of inventory related to the modal performance. Through the study, it was possible to observe it may be possible to save up to 73% on the coolant international transportation by changing the transportation mode used by the company.

  11. Effect of internal coolant crossflow orientation on the discharge coefficient of shaped film-cooling holes

    Energy Technology Data Exchange (ETDEWEB)

    Gritsch, M.; Saumweber, C.; Schulz, A.; Wittig, S.; Sharp, E.

    2000-01-01

    Discharge coefficients of three film-cooling hole geometries are presented over a wide range of engine like conditions. The hole geometries comprise a cylindrical hole and two holes with a diffuser-shaped exit portion (a fanshaped and a laidback fanshaped hole). For all three hole geometries the hole axis was inclined 30 deg with respect to the direction of the external (hot gas) flow. The flow conditions considered were the hot gas crossflow Mach number (up to 0.6), the coolant crossflow Mach number (up to 0.6) and the pressure ratio across the hole (up to 2). The effect of internal crossflow approach direction, perpendicular or parallel to the main flow direction, is particularly addressed in the present study. Comparison is made of the results for a parallel and perpendicular orientation, showing that the coolant crossflow orientation has a strong impact on the discharge behavior of the different hole geometries. The discharge coefficients were found to strongly depend on both hole geometry and crossflow conditions. Furthermore, the effects of internal and external crossflow on the discharge coefficients were described by means of correlations used to derive a predicting scheme for discharge coefficients. A comparison between predictions and measurements reveals the capability of the method proposed.

  12. In-vessel ITER tubing failure rates for selected materials and coolants

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, T.D. [Rensselaer Polytechnic Institute, Troy, NY (United States); Cadwallader, L.C. [EG& G Idaho Inc., Idaho Falls, ID (United States)

    1994-03-01

    Several materials have been suggested for fabrication of ITER in-vessel coolant tubing: beryllium, copper, Inconel, niobium, stainless steel, titanium, and vanadium. This report generates failure rates for the materials to identify the best performer from an operational safety and availability perspective. Coolant types considered in this report are helium gas, liquid lithium, liquid sodium, and water. Failure rates for the materials are generated by including the influence of ITER`s operating environment and anticipated tubing failure mechanisms with industrial operating experience failure rates. The analyses define tubing failure mechanisms for ITER as: intergranular attack, flow erosion, helium induced swelling, hydrogen damage, neutron irradiation embrittlement, cyclic fatigue, and thermal cycling. K-factors, multipliers, are developed to model each failure mechanism and are applied to industrial operating experience failure rates to generate tubing failure rates for ITER. The generated failure rates identify the best performer by its expected reliability. With an average leakage failure rate of 3.1e-10(m-hr){sup {minus}1}and an average rupture failure rate of 3.1e-11(m-hr){sup {minus}1}, titanium proved to be the best performer of the tubing materials. The failure rates generated in this report are intended to serve as comparison references for design safety and optimization studies. Actual material testing and analyses are required to validate the failure rates.

  13. LOSS-OF-COOLANT ACIDENT SIMULATIONS IN THE NATIONAL RESEARCH UNIVERSAL REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    Bennett, W D; Goodman, R L; Heaberlin, S W; Hesson, G M; Nealley, C; Kirg, L L; Marshall, R K; McNair, G W; Meitzler, W D; Neally, G W; Parchen, L J; Pilger, J P; Rausch, W N; Russcher, G E; Schreiber, R E; Wildung, N J

    1981-02-01

    Pressurized water reactor loss-of-coolant accident (LOCA) phenomena are being simulated with a series of experiments in the U-2 loop of the National Research Universal Reactor at Chalk River, Ontario, Canada. The first of these experiments includes up to 45 parametric thermal-hydraulic tests to establish the relationship among the reflood delay time of emergency coolant, the reflooding rate, and the resultant fuel rod cladding peak temperature. Subsequent experiments establish the fuel rod failure characteristics at selected peak cladding temperatures. Fuel rod cladding pressurization simulates high burnup fission gas pressure levels of modern PWRs. This document contains both an experiment overview of the LOCA simulation program and a review of the safety analyses performed by Pacific Northwest Laboratory (PNL) to define the expected operating conditions as well as to evaluate the worst case operating conditions. The primary intent of this document is to supply safety information required by the Chalk River Nuclear Laboratories (CRNL), to establish readiness to proceed from one test phase to the next and to establish the overall safety of the experiment. A hazards review summarizes safety issues, normal operation and three worst case accidents that have been addressed during the development of the experiment plan.

  14. Determination of the {sup 129}I in primary coolant of PWR

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Ke Chon; Park, Yong Joon; Song, Kyu Seok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-02-15

    Among the radioactive wastes generated from the nuclear power plant, a radioactive nuclide such as {sup 129}I is classified as a difficult-to-measure (DTM) nuclide, owing to its low specific activity. Therefore, the establishment of an analytical procedure, including a chemical separation for {sup 129}I as a representative DTM, becomes essential. In this report, the adsorption and recovery rate were measured by adding {sup 125}I as a radio-isotopic tracer (t1/2 = 60.14 d) to the simulation sample, in order to measure the activity concentration of {sup 129}I in a pressurized-water reactor primary coolant. The optimum condition for the maximum recovery yield of iodine on the anion exchange resins (AG1 x2, 50-100 mesh, Clform) was found to be at pH 7. In this report, the effect of the boron content in a pressurized-water reactor primary coolant on the separation process of {sup 129}I was examined, as was the effect of {sup 3}H on the measurement of the activity of iodine. As a result, no influence of the boron content and of the simultaneous {sup 3}H presence was found with activity concentrations of {sup 3}H lower than 50 Bq/mL, and with a boron concentration of less than 2,000 {mu}g/mL.

  15. Experimental and analytical studies of melt jet-coolant interactions: a synthesis

    Energy Technology Data Exchange (ETDEWEB)

    Dinh, T.N.; Bui, V.A.; Nourgaliev, R.R.; Green, J.A.; Sehgal, B.R. [Royal Inst. of Tech., Stockholm (Sweden). Div. of Nuclear Power Safety

    1998-01-01

    Instability and fragmentation of a core melt jet in water have been actively studied during the past ten years. Several models, and a few computer codes, have been developed. However, there are, still, large uncertainties, both, in interpreting experimental results and in predicting reactor-scale processes. Steam explosion and debris coolability, as reactor safety issues, are related to the jet fragmentation process. A better understanding of the physics of jet instability and fragmentation is crucial for assessments of fuel-coolant interactions (FCIs). This paper presents research, conducted at the Division of Nuclear Power Safety, Royal Institute of Technology (RIT/NPS), Stockholm, concerning molten jet-coolant interactions, as a precursor for premixing. First, observations were obtained from scoping experiments with simulant fluids. Second, the linear perturbation method was extended and applied to analyze the interfacial-instability characteristics. Third, two innovative approachs to CFD modeling of jet fragmentation were developed and employed for analysis. The focus of the studies was placed on (a) identifying potential factors, which may affect the jet instability, (b) determining the scaling laws, and (c) predicting the jet behavior for severe accidents conditions. In particular, the effects of melt physical properties, and the thermal hydraulics of the mixing zone, on jet fragmentation were investigated. Finally, with the insights gained from a synthesis of the experimental results and analysis results, a new phenomenological concept, named `macrointeractions concept of jet fragmentation` is proposed. (author)

  16. Heat Exchanger Can Assembly for Provision of Helium Coolant Streams for Cryomodule Testing below 2K

    Science.gov (United States)

    Smith, E. N.; Eichhorn, R.; Quigley, P.; Sabol, D.; Shore, C.; Widger, D.

    2017-02-01

    A series of heat exchanger can (HXC) assemblies have been designed, constructed and built to utilize existing 4.2 K liquefaction and compressor capabilities to provide helium gas coolant streams of 80 K, 4.5 K, and liquid from 1.6 to 2.0 K for operating cryomodules containing from one to six superconducting RF cavities built for an energy recovery linear accelerator. Designs for the largest assemblies required up to 100 W of cooling at 1.8 K with precise temperature control, especially during cool-down, and up to 2000 W at 80 K (with a 40 K temperature rise). A novel feature of these assemblies was the use of relatively inexpensive brazed stainless steel plate heat exchangers intended for room-temperature operation with water or oil, but which in practice worked well at cryogenic temperatures. The choice of operating temperatures/pressures were to provide single-phase helium flow for better control of coolant distribution in the 80 K and 4.5 K streams, to take advantage of locally elevated heat capacity near the critical point for the 4.5 K stream, and in the region below 2 K to get the best possible Q from the niobium cavities under test.

  17. Surface and subsurface cleanup protocol for radionuclides, Gunnison, Colorado, UMTRA project processing site. Final [report

    Energy Technology Data Exchange (ETDEWEB)

    1993-09-01

    Surface and subsurface soil cleanup protocols for the Gunnison, Colorado, processing sits are summarized as follows: In accordance with EPA-promulgated land cleanup standards (40 CFR 192), in situ Ra-226 is to be cleaned up based on bulk concentrations not exceeding 5 and 15 pCi/g in 15-cm surface and subsurface depth increments, averaged over 100-m{sup 2} grid blocks, where the parent Ra-226 concentrations are greater than, or in secular equilibrium with, the Th-230 parent. A bulk interpretation of these EPA standards has been accepted by the Nuclear Regulatory Commission (NRC), and while the concentration of the finer-sized soil fraction less than a No. 4 mesh sieve contains the higher concentration of radioactivity, the bulk approach in effect integrates the total sample radioactivity over the entire sample mass. In locations where Th-230 has differentially migrated in subsoil relative to Ra-226, a Th-230 cleanup protocol has been developed in accordance with Supplemental Standard provisions of 40 CFR 192 for NRC/Colorado Department of Health (CDH) approval for timely implementation. Detailed elements of the protocol are contained in Appendix A, Generic Protocol from Thorium-230 Cleanup/Verification at UMTRA Project Processing Sites. The cleanup of other radionuclides or nonradiological hazards that pose a significant threat to the public and the environment will be determined and implemented in accordance with pathway analysis to assess impacts and the implications of ALARA specified in 40 CFR 192 relative to supplemental standards.

  18. Surface and subsurface cleanup protocol for radionuclides, Gunnison, Colorado, UMTRA project processing site. Final [report

    Energy Technology Data Exchange (ETDEWEB)

    1993-09-01

    Surface and subsurface soil cleanup protocols for the Gunnison, Colorado, processing sits are summarized as follows: In accordance with EPA-promulgated land cleanup standards (40 CFR 192), in situ Ra-226 is to be cleaned up based on bulk concentrations not exceeding 5 and 15 pCi/g in 15-cm surface and subsurface depth increments, averaged over 100-m{sup 2} grid blocks, where the parent Ra-226 concentrations are greater than, or in secular equilibrium with, the Th-230 parent. A bulk interpretation of these EPA standards has been accepted by the Nuclear Regulatory Commission (NRC), and while the concentration of the finer-sized soil fraction less than a No. 4 mesh sieve contains the higher concentration of radioactivity, the bulk approach in effect integrates the total sample radioactivity over the entire sample mass. In locations where Th-230 has differentially migrated in subsoil relative to Ra-226, a Th-230 cleanup protocol has been developed in accordance with Supplemental Standard provisions of 40 CFR 192 for NRC/Colorado Department of Health (CDH) approval for timely implementation. Detailed elements of the protocol are contained in Appendix A, Generic Protocol from Thorium-230 Cleanup/Verification at UMTRA Project Processing Sites. The cleanup of other radionuclides or nonradiological hazards that pose a significant threat to the public and the environment will be determined and implemented in accordance with pathway analysis to assess impacts and the implications of ALARA specified in 40 CFR 192 relative to supplemental standards.

  19. Assessment of Candidate Molten Salt Coolants for the Advanced High Temperature Reactor (AHTR)

    Energy Technology Data Exchange (ETDEWEB)

    Williams, D.F.

    2006-03-24

    The Advanced High-Temperature Reactor (AHTR) is a novel reactor design that utilizes the graphite-matrix high-temperature fuel of helium-cooled reactors, but provides cooling with a high-temperature fluoride salt. For applications at temperatures greater than 900 C the AHTR is also referred to as a Liquid-Salt-Cooled Very High-Temperature Reactor (LS-VHTR). This report provides an assessment of candidate salts proposed as the primary coolant for the AHTR based upon a review of physical properties, nuclear properties, and chemical factors. The physical properties most relevant for coolant service were reviewed. Key chemical factors that influence material compatibility were also analyzed for the purpose of screening salt candidates. Some simple screening factors related to the nuclear properties of salts were also developed. The moderating ratio and neutron-absorption cross-section were compiled for each salt. The short-lived activation products, long-lived transmutation activity, and reactivity coefficients associated with various salt candidates were estimated using a computational model. Table A presents a summary of the properties of the candidate coolant salts. Certain factors in this table, such as melting point, vapor pressure, and nuclear properties, can be viewed as stand-alone parameters for screening candidates. Heat-transfer properties are considered as a group in Sect. 3 in order to evaluate the combined effects of various factors. In the course of this review, it became apparent that the state of the properties database was strong in some areas and weak in others. A qualitative map of the state of the database and predictive capabilities is given in Table B. It is apparent that the property of thermal conductivity has the greatest uncertainty and is the most difficult to measure. The database, with respect to heat capacity, can be improved with modern instruments and modest effort. In general, ''lighter'' (low-Z) salts tend to

  20. Development of an annular linear induction electromagnetic pump for the na-coolant circulation of LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hee Reyoung; Lee, Yong Bum; Kim, Yong Kyun; Nam, Ho Yun [KAERI, Taejon (Korea, Republic of)

    1998-07-01

    The EM (ElectroMagnetic) pump operated by Lorentz force (J x B) is developed for the sodium coolant circulation of LMFBR (Liquid Metal Fast Breeder Reactors). Design and experimental characterization are carried out on the linear induction EM pump of the narrow annular channel type. The pump which obtains propulsion force resultantly by the three phase symmetric alternating input currents is analyzed by the electrical equivalent circuit method used in the analyses of the induction machines. Then, the equivalent circuit for the pump consists of equivalent variables of primary and secondary resistances and magnetizing and leakage reactances given as functions of pump geometrical and electrical variables by Laithwaithe's standard formulae. Developing pressure-flowrate relation given by pump variables is sought from the balance equation on the circuit. Developing pressure and efficiency of the pump according to the pump variables are analyzed for the pump with a flowrate of 200 l/min. It is shown that pump is mainly characterized by length of the core, diameter of the inner core and channel gap geometrically and by input frequency electrically. Optimum values of pump geometrical and operational variables are determined to maximize the developing force and overall efficiency. The pump has geometrical size of 60 cm in length, 4.27 cm in inner core diameter and electrical input of 6,428 VA and 17 Hz. Optimally designed pump is manufactured by the consideration of material and operational requirements in the chemically-active sodium environment with high temperature of 600 .deg. C. Silicon-iron steel plates with high magnetic permeability in the high temperature are stacked for generation of the high magnetic flux and alumina-dispersion-strengthened-copper bands are used as exciting coils. Each turn of coil is insulated by asbestos band to protect electrical short in the high temperature. Stainless steel which can be compatible with sodium is selected as structural

  1. Validation of computational fluid dynamics calculation using Rossendorf coolant mixing model flow measurements in primary loop of coolant in a pressurized water reactor model

    Energy Technology Data Exchange (ETDEWEB)

    Farkas, Istvan; Hutli, Ezddin; Faekas, Tatiana; Takacs, Antal; Guba, Attila; Toth, Ivan [Dept. of Thermohydraulics, Centre for Energy Research, Hungarian Academy of Sciences, Budapest (Hungary)

    2016-08-15

    The aim of this work is to simulate the thermohydraulic consequences of a main steam line break and to compare the obtained results with Rossendorf Coolant Mixing Model (ROCOM) 1.1 experimental results. The objective is to utilize data from steady-state mixing experiments and computational fluid dynamics (CFD) calculations to determine the flow distribution and the effect of thermal mixing phenomena in the primary loops for the improvement of normal operation conditions and structural integrity assessment of pressurized water reactors. The numerical model of ROCOM was developed using the FLUENT code. The positions of the inlet and outlet boundary conditions and the distribution of detailed velocity/turbulence parameters were determined by preliminary calculations. The temperature fields of transient calculation were averaged in time and compared with time-averaged experimental data. The perforated barrel under the core inlet homogenizes the flow, and therefore, a uniform temperature distribution is formed in the pressure vessel bottom. The calculated and measured values of lowest temperature were equal. The inlet temperature is an essential parameter for safety assessment. The calculation predicts precisely the experimental results at the core inlet central region. CFD results showed a good agreement (both qualitatively and quantitatively) with experimental results.

  2. Support of the launching of motor car air conditioning systems with the coolant CO{sub 2} (R744). Test bench measurements and practical trials; Unterstuetzung der Markteinfuehrung von Pkw-Klimaanlagen mit dem Kaeltemittel CO{sub 2} (R744). Pruefstandsmessungen und Praxistest

    Energy Technology Data Exchange (ETDEWEB)

    Lemke, Nicholas; Mildenberger, Julia [Technische Univ. Braunschweig (Germany); Graz, Martin [Obrist Engineering GmbH, Lustenau (Austria)

    2011-10-15

    In the research project two passenger car air-conditioning systems were analyzed with regard to cooling capacity and efficiency. The results were compared with one another. The first system was a standard air-conditioning unit using R134a as a refrigerant. As a second system a CO{sub 2} (R744) prototype HVAC unit was used. Both units were investigated on one hand installed in a car on a dynamometer by Obrist Engineering GmbH and on the other hand installed in a calorimetric test rig by Technische Universitaet Braunschweig, Institut fuer Thermodynamik. While the tests in the calorimetric test rig showed comparable efficiencies and cooling capacities for both setups, consumption advantages were determined for the R744- air-conditioning unit installed in the vehicle by the company Obrist. With CO{sub 2} (R744) as a refrigerant for mobile air-conditioning systems an environmental friendly solution is available. (orig.)

  3. System and method for conditioning intake air to an internal combustion engine

    Energy Technology Data Exchange (ETDEWEB)

    Sellnau, Mark C.

    2015-08-04

    A system for conditioning the intake air to an internal combustion engine includes a means to boost the pressure of the intake air to the engine and a liquid cooled charge air cooler disposed between the output of the boost means and the charge air intake of the engine. Valves in the coolant system can be actuated so as to define a first configuration in which engine cooling is performed by coolant circulating in a first coolant loop at one temperature, and charge air cooling is performed by coolant flowing in a second coolant loop at a lower temperature. The valves can be actuated so as to define a second configuration in which coolant that has flowed through the engine can be routed through the charge air cooler. The temperature of intake air to the engine can be controlled over a wide range of engine operation.

  4. EXPERIMENTAL STUDY OF LOCAL HYDRODYNAMICS AND MASS EXCHANGE PROCESSES OF COOLANT IN FUEL ASSEMBLIES OF PRESSURIZED WATER REACTORS

    Directory of Open Access Journals (Sweden)

    S. M. Dmitriev

    2016-01-01

    Full Text Available The results of experimental studies of local hydrodynamics and mass exchange of coolant flow behind spacer and mixing grids of different structural versions that were developed for fuel assemblies of domestic and foreign nuclear reactors are presented in the article. In order to carry out the study the models of the following fuel assemblies have been fabricated: FA for VVER and VBER, FA-KVADRAT for PWR-reactor and FA for KLT-40C reactor. All the models have been fabricated with a full geometrical similarity with full-scale fuel assemblies. The study was carried out by simulating the flow of coolant in a core by air on an aerodynamic test rig. In order to measure local hydrodynamic characteristics of coolant flow five-channel Pitot probes were used that enable to measure the velocity vector in a point by its three components. The tracerpropane method was used for studying mass transfer processes. Flow hydrodynamics was studied by measuring cross-section velocities of coolant flow and coolant rates according to the model cells. The investigation of mass exchange processes consisted of a study of concentration distribution for tracer in experimental model, in determination of attenuation lengths of mass transfer processes behind mixing grids, in calculating of inter-cellar mass exchange coefficient. The database on coolant flow in fuel assemblies for different types of reactors had been accumulated that formed the basis of the engineering substantiation of reactor cores designs. The recommendations on choice of optimal versions of mixing grids have been taken into consideration by implementers of the JSC “OKBM Afrikantov” when creating commissioned fuel assemblies. The results of the study are used for verification of CFD-codes and CFD programs of detailed cell-by-cell calculation of reactor cores in order to decrease conservatism for substantiation of thermal-mechanical reliability.

  5. Contribution to the optimization of the chemical and radiochemical purification of pressurized water nuclear power plants primary coolant; Contribution a l'optimisation de la purification chimique et radiochimique du fluide primaire des centrales nucleaires a eau sous pression

    Energy Technology Data Exchange (ETDEWEB)

    Elain, L

    2004-12-15

    The primary coolant of pressurised water reactors is permanently purified thanks to a device, composed of filters and the demineralizers furnished with ion exchange resins (IER), located in the chemical and volume control system (CVCS). The study of the retention mechanisms of the radio-contaminants by the IER implies, initially, to know the speciation of the primary coolant percolant through the demineralizers. Calculations of theoretical speciation of the primary coolant were carried out on the basis of known composition of the primary coolant and thanks to the use of an adapted chemical speciation code. A complementary study, dedicated to silver behaviour, considered badly extracted, suggests metallic aggregates existence generated by the radiolytic reduction of the Ag{sup +} ions. An analysis of the purification curves of the elements Ni, Fe, Co, Cr, Mn, Sb and their principal radionuclides, relating to the cold shutdown of Fessenheim 1-cycle 20 and Tricastin 2-cycle 21, was carried out, in the light of a model based on the concept of a coupling well term - source term. Then, a thermodynamic modelling of ion exchange phenomena in column was established. The formation of the permutation front and the enrichment zones planned was validated by frontal analysis experiments of synthetic fluids (mixtures of Ni(B(OH){sub 4}){sub 2}, LiB(OH){sub 4} and AgB(OH){sub 4} in medium B(OH){sub 3})), and of real fluid during the putting into service of the device mini-CVCS at the time of Tricastin 2 cold shutdown. New tools are thus proposed, opening the way with an optimised management of demineralizers and a more complete interpretation of the available experience feedback. (author)

  6. Application of Diagnostic/Prognostic Methods to Critical Equipment for the Spent Nuclear Fuel Cleanup Program

    Energy Technology Data Exchange (ETDEWEB)

    Casazza, Lawrence O.; Jarrell, Donald B.; Koehler, Theresa M.; Meador, Richard J.; Wallace, Dale E.

    2002-02-28

    The management of the Spent Nuclear Fuel (SNF) project at the Hanford K-Basin in the 100 N Area has successfully restructured the preventive maintenance, spare parts inventory requirements, and the operator rounds data requirements. In this investigation, they continue to examine the different facets of the operations and maintenance (O&M) of the K-Basin cleanup project in search of additional reliability and cost savings. This report focuses on the initial findings of a team of PNNL engineers engaged to identify potential opportunities for reducing the cost of O&M through the application of advanced diagnostics (fault determination) and prognostics (residual life/reliability determination). The objective is to introduce predictive technologies to eliminate or reduce high impact equipment failures. The PNNL team in conjunction with the SNF engineers found the following major opportunities for cost reduction and/or enhancing reliability: (1) Provide data routing and automated analysis from existing detection systems to a display center that will engage the operations and engineering team. This display will be operator intuitive with system alarms and integrated diagnostic capability. (2) Change operating methods to reduce major transients induced in critical equipment. This would reduce stress levels on critical equipment. (3) Install a limited sensor set on failure prone critical equipment to allow degradation or stressor levels to be monitored and alarmed. This would provide operators and engineers with advance guidance and warning of failure events. Specific methods for implementation of the above improvement opportunities are provided in the recommendations. They include an Integrated Water Treatment System (IWTS) decision support system, introduction of variable frequency drives on certain pump motors, and the addition of limited diagnostic instrumentation on specified critical equipment.

  7. Flow tests of a single fuel element coolant channel for a compact fast reactor for space power

    Science.gov (United States)

    Springborn, R. H.

    1971-01-01

    Water flow tests were conducted on a single-fuel-element cooling channel for a nuclear concept to be used for space power. The tests established a method for measuring coolant flow rate which is applicable to water flow testing of a complete mockup of the reference reactor. The inlet plenum-to-outlet plenum pressure drop, which approximates the overall core pressure drop, was measured and correlated with flow rate. This information can be used for reactor coolant flow and heat transfer calculations. An analytical study of the flow characteristics was also conducted.

  8. Clean-up and disposal process of polluted sediments from urban rivers

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    In this paper, the discussion is concentrated on the properties of the polluted sediments and the combination of clean-up and disposal process for the upper layer heavily polluted sediments with good flowability. Based on the systematic analyses of various clean-up processes, a suitable engineering process has been evaluated and recommended. The process has been applied to the river reclamation in Yangpu District of Shanghai metropolis. An improved centrifuge is used for dewatering the dredged sludge,which plays an important role in the combination of clean-up and disposal process. The assessment of the engineering process shows its environmental and technical economy feasibility, which is much better than that of traditional dredging-disposal processes.

  9. Mechanism Involved in Trichloroethylene-Induced Liver Cancer: Importance to Environmental Cleanup

    Energy Technology Data Exchange (ETDEWEB)

    Bull, Richard J.; Thrall, Brian D.

    1999-06-01

    The objective of this project is to develop critical data for improving risk-based cleanup standards for trichloroethylene (TCE). Importance to DOE. Cleanup costs for chlorinated solvents found on DOE sites are most frequently driven by TCE because it is the most widespread contaminant and is generally present at the highest concentrations. Data that would permit increases in risk-based standards for TCE would reduce complex wide cleanup costs by hundreds of millions of dollars. Current Regulatory Actions that Research will Impact. EPA is currently reviewing its risk assessment for TCE. Richard J. Bull has worked with EPA on this review by writing the mode of action section of their determination. A presentation by James Cogliano of EPA at the 1999 Annual Society of Toxicology Meeting indicates that they have accepted the concept of nonlinear extrapolation for liver tumor induction by TCE. This project will end in FY 1999 with its major technical and policy objectives satisfied.

  10. Summary of Model Toxics Control Act (MTCA) Potential Impacts Related to Hanford Cleanup and the Tri-Party Agreement (TPA)

    Energy Technology Data Exchange (ETDEWEB)

    IWATATE, D.F.

    2000-07-14

    This white paper provides an initial assessment of the potential impacts of the Model Toxics Control Act (MTCA) regulations (and proposed revisions) on the Hanford site cleanup and addresses concerns that MTCA might impose inappropriate or unachievable clean-up levels and drive clean-up costs higher. The white paper and supporting documentation (Appendices A and B) provide DOE with a concise and up-to-date review of potential MTCA impacts to cost and schedule for the Hanford site activities. MTCA, Chapter 70.105D RCW, is the State of Washington's risk based law governing clean-up of contaminated sites and is implemented by The Washington Department of Ecology (Ecology) under the MTCA Clean-up Regulations, Chapter 173-340 WAC. Hanford cleanup is subject to the MTCA requirements as Applicable, Relevant and Appropriate Requirements (ARARs) for those areas of Hanford being managed under the authority of the Federal Resource Conservation and Recovery Act (RCRA), Comprehensive Environmental Response, Compensation and Liability Act (CERCLA), and the state Dangerous Waste Regulations. MTCA provides Ecology with authority to implement site clean-up actions under both the federal RCRA and CERCLA regulations as well as the state regulations. Most of the Hanford clean-up actions are being implemented under the CERCLA program, however, there is a trend is toward increased use of MTCA procedures and standards. The application of MTCA to the Hanford clean-up has been an evolving process with some of the Hanford clean-up actions considering MTCA standards as an ARAR and using MTCA procedures for remedy selection. The increased use and application of MTCA standards and procedures could potentially impact both cost and schedule for the Hanford cleanup.

  11. A study on removal of cobalt from the primary coolant by continuous electrode-ionization with various conducting spacers

    Energy Technology Data Exchange (ETDEWEB)

    Yeon, K.H.; Song, J.H.; Moon, S.H. [Department of Environmental Science and Engineering, Kwangju Inst. of Science and Technology (K-JIST) (Korea, Republic of)

    2002-07-01

    CEDI is a hybrid separation system of electrodialysis and ion exchange processes. This system does not require chemicals to regenerate the ion exchange resin and to concentrate the wastewater. In a CEDI system, the ion exchange resin bed plays a major role in the reduction of the high electrical resistance in the dilute compartment, while the ion exchange membranes lead to depletion and concentration of the solutions in the dilute compartment and concentrate compartment, respectively. The production of high purity water in the primary coolant of a nuclear power plant was investigated using a CEDI process along with various ion-conducting spacers, such as an ion exchange resin (IX), polyurethane-coated ion exchange beads (IEPU), and an ion exchange textile (IET). The ion exchange resin was introduced into the ion-depleting compartments of an electrodialysis (ED) stack, and has been used to reduce the electrical resistance of the stack since ED cannot be applied economically to the treatment of dilute solutions due to their high electrical resistances and the development of the polarization phenomena. However, packing the resin beads in the compartment and assembling the stack is laborious work, while attaining a free flowing solution is difficult because the resin beads are driven downward by gravity in the diluted compartment. Nevertheless, a resin-packed ED stack has recently been developed by Millipore, and is now commercially available from U.S. Filter as industrial units. We set out to prepare improved ion-conducting materials suitable for use in CEDI stacks. To this end, IEPU was prepared using a blending method that produces mixtures of resin beads and powder by allophanate/biuret cross-linking. IET was prepared by the radiation grafting of styrene-fulfonic acid or trimethyl-ammonium chloride onto polypropylene non-woven fabric. (authors)

  12. SGN's experience in the field of decommissioning and site cleanup

    Energy Technology Data Exchange (ETDEWEB)

    Fouques, F. [SGN, Montigny-le-Bretonneux, 78182 Saint Quentin-en-Yvelines Cedex (France); Destrait, L. [SGN, 30204 Bagnols sur Ceze Cedex (France)

    2003-07-01

    As early as the 1980's, SGN participated in dismantling projects at CEA and COGEMA plants in France. The experience gained has since been applied to many projects in France and abroad. In close collaboration with the customer, SGN is prime contractor on a cleanup and dismantling project, from preliminary studies and tool and process development to release of the site. SGN's areas of expertise include waste retrieval, decontamination processes, intervention robotics, cutting tools and waste management and treatment. SGN's proposal is based on proven methods and feedback from earlier projects. SGN is currently participating in many cleanup and dismantling projects, including the three (LRTP project at Chernobyl, Marcoule Plant UP1 and Hanford in the U.S.A) presented below. The contents is as follows: 1. Introduction; 2. LRTP Project at Chernobyl; 2.1. Description of interim waste storage; 2.2. Organization; 2.3. Plant characteristics; 2.4. Process Implemented (waste retrieval from the storage tanks; waste sampling and pretreatment; Volume reduction; Cementing); 2.5. Schedule; 3. Marcoule plant UP1; 3.1 Description of plant UP1; 3.2 Organization; 3.3. Cleanup/Dismantling program; 3.3.1. Purpose of the decommissioning and cleanup operations; 3.3.2. D and D techniques Implemented (Cleanup techniques used; Examples of remote handling equipment): 3.3.3. Purpose of the waste retrieval and packaging operations; 3.3.4. Purpose of the dismantling operations; 4. Hanford in the U.S.A.; 4.1. Decontamination and cleanup of hot cells; 4.2. Liquid and sludge retrieval; 4.3. Retrieval and packaging of spent nuclear fuel.

  13. The mental health of clean-up workers 18 years after the Chernobyl accident.

    Science.gov (United States)

    Loganovsky, K; Havenaar, J M; Tintle, N L; Guey, L T; Kotov, R; Bromet, E J

    2008-04-01

    The psychological aftermath of the Chernobyl accident is regarded as the largest public health problem unleashed by the accident to date. Yet the mental health of the clean-up workers, who faced the greatest radiation exposure and threat to life, has not been systematically evaluated. This study describes the long-term psychological effects of Chernobyl in a sample of clean-up workers in Ukraine. The cohorts were 295 male clean-up workers sent to Chernobyl between 1986 and 1990 interviewed 18 years after the accident (71% participation rate) and 397 geographically matched controls interviewed as part of the Ukraine World Mental Health (WMS) Survey 16 years after the accident. The World Health Organization (WHO) Composite International Diagnostic Interview (CIDI) was administered. We examined group differences in common psychiatric disorders, suicide ideation and severe headaches, differential effects of disorder on days lost from work, and in the clean-up workers, the relationship of exposure severity to disorder and current trauma and somatic symptoms. Analyses were adjusted for age in 1986 and mental health prior to the accident. Relatively more clean-up workers than controls experienced depression (18.0% v. 13.1%) and suicide ideation (9.2% v. 4.1%) after the accident. In the year preceding interview, the rates of depression (14.9% v. 7.1%), post-traumatic stress disorder (PTSD) (4.1% v. 1.0%) and headaches (69.2% v. 12.4%) were elevated. Affected workers lost more work days than affected controls. Exposure level was associated with current somatic and PTSD symptom severity. Long-term mental health consequences of Chernobyl were observed in clean-up workers.

  14. Study on the effect of the impeller and diffuser blade number on reactor coolant pump performances

    Science.gov (United States)

    Long, Y.; Yin, J. L.; Wang, D. Z.; Li, T. B.

    2016-05-01

    In this paper, CFD approach was employed to study how the blade number of impeller and diffuser influences reactor coolant pump performances. The three-dimensional pump internal flow channel was modelled by pro/E software, Reynolds-averaged Naiver-Stokes equations with the k-ε turbulence model were solved by the computational fluid dynamics software CFX. By post-processing on the numerical results, the performance curves of reactor coolant pump were obtained. The results are as follows, with the blade number of the impeller increasing, the head of the pump with different diffuser universally increases in the 8Q n∼1.2Q n conditions, and at different blade number of the diffuser, the head increases with the blade number of the impeller increasing. In 1.0Q n condition, when the blades number combination of impeller and diffuser chooses 4+16, 7+14 and 6+18, the head curves exist singular points. In 1.2Q n condition, the head curve still exists singular point in 6+18. With the blade number of the impeller increasing, the efficiency of the pump with different diffuser universally decreases in the 0.8Q n and 1.0Q n conditions, but in 1.2Q n condition, the efficiency of the pump with different diffuser universally increases. In 1.0Q n condition, the impellers of 4 and 5 blades are better. When the blade number combination of impeller and diffuser choose 4+11, 4+17, 4+18, 5+12, 5+17 and 5+18, the efficiencies relatively have higher values. With the blade number of the impeller increasing, the hydraulic shaft power of the pump with different diffuser universally increases in the 0.8Q n∼1.2Q n conditions, and with the blade number of the diffuser increasing, the power of different impeller overall has small fluctuation, but tends to be uniform. This means the increase of the diffuser blade number has less influence on shaft power.The influence on the head and flow by the matching relationship of the blades number between impeller and diffuser is very complicated, which

  15. Simulation of Heat Transfer to the Gas Coolant with Low Prandtl Number Value

    Directory of Open Access Journals (Sweden)

    T. N. Kulikova

    2015-01-01

    Full Text Available The work concerns the simulating peculiarities of heat transfer to the gas coolants with low values of the Prandtl number, in particular, to the binary mixtures of inert gases.The paper presents simulation results of heat transfer to the fully established flow of a helium-xenon mixture in the round tube of 6 mm in diameter with the boundary condition of the second kind. It considers a flow of three helium-xenon mixtures with different helium content and molecular Prandtl numbers within the range 0.239–0.322 and with Reynolds numbers ranged from 10000 to 50000. During numerical simulation a temperature factor changed from 1.034 to 1.061. CFD-code STAR-CCM+ that is designed for solving a wide range of problems of hydrodynamics, heat transfer and stress was used as the primary software.The applicability of the five models for the turbulent Prandtl number is examined. It is shown that the choice of the model has a significant influence on the heat transfer coefficient. The paper presents structural characteristics of the flow in the wall region. It estimates a thermal stabilization section to be approximately as long as 30 diameters of tube.Simulation results are compared with the known data on heat transfer to gas coolants with low values of the Prandtl number. It is shown that V2F low-Reynolds number -ε turbulence model with an approximation for the turbulent Prandtl number used according Kays-CrawfordWeigand gives the best compliance with the results predicted by relationships of Kays W.M. and Petukhov B.S. The approximating correlation summarizes a set of simulation results.Application of the work results is reasonable when conducting the numerical simulation of heat transfer to binary gas mixtures in channels of different forms. The presented approximating correlation allows rapid estimate of heat transfer coefficients to the gas coolants with a low value of the molecular Prandl number within the investigated range with a flow through the

  16. Mechanism involved in trichloroethylene-induced liver cancer: Importance to environmental cleanup. 1997 annual progress report

    Energy Technology Data Exchange (ETDEWEB)

    Bull, R.J.

    1997-06-01

    'The Pacific Northwest National Lab. was awarded ten (10) Environmental Management Science Program (EMSP) research grants in Fiscal Year 1996. This section gives a summary of how each grant is addressing significant DOE cleanup issues, including those at the Hanford Site. The technical progress made to date in each of these research projects is addressed in more detail in the individual progress reports contained in this document. This research is primarily focused in three areas-Tank Waste Remediation, Soil and Groundwater Cleanup, and Health Effects.'

  17. Dyscirculatory encephalopathy in Chernobyl disaster clean-up workers (a 20-year study).

    Science.gov (United States)

    Podsonnaya, I V; Shumakher, G I; Golovin, V A

    2010-05-01

    Results obtained over 20-years of following 536 Chernobyl clean-up workers and 436 control subjects are presented. Dyscirculatory encephalopathy developed more frequently in persons exposed to radiation at age 30 years. As compared with the control group, workers were characterized by early onset of disease, faster progression, stable symptomatology for 5-6 years, and further progression of disease in the form of autonomic dysfunction, psycho-organic syndrome, and epilepsy. Major strokes were also more common in clean-up workers.

  18. DECOMMISSIONING AND ENVRIONMENTAL CLEANUP OF SMALL ARMS TRAINING FACILITY

    Energy Technology Data Exchange (ETDEWEB)

    Kmetz, T.

    2012-12-04

    constituents posed a migration risk to groundwater. The NTCR action involved removal of approximately 12,092 m3 (15,816 yd3) of spent bullets and lead-impacted soil and off-site disposal. The removal action included soils from the berm area, a fill area that received scraped soils from the berm, and soil from a drainage ditch located on the edge of the berm area. Also included in the removal action was a mixture of soil, concrete, and asphalt from the other three range areas. Under this action, 11,796 m3 (15,429 yd3) of hazardous waste and impacted soil were removed from the SATA and transported to a permitted hazardous waste disposal facility (Lone Mountain Facility in Oklahoma) and 296 m3 (387 yd3) of nonhazardous waste (primarily concrete debris) were removed and transported to a local solid waste landfill for disposal. During the excavation process, the extent was continuously assessed through the use of a hand-held, field-portable X-ray fluorescence unit with results verified using confirmation sampling with certified laboratory analysis. Following the completion of the excavation and confirmation sampling, final contouring, grading, and establishment of vegetative cover was performed to stabilize the affected areas. The NTCR action began on August 17, 2010, and mechanical completion was achieved on April 27, 2011. The selected removal action met the removal action objectives (RAOs), is protective of human health and the environment both in the short- and long-term, was successful in removing potential ecological risks, and is protective of surface water and groundwater. Furthermore, the selected NTCR action met residential cleanup goals and resulted in the release of the SEA from restricted use contributing to the overall footprint reduction at SRS.

  19. Volume overload cleanup: An approach for on-line SPE-GC, GPC-GC, and GPC-SPE-GC

    NARCIS (Netherlands)

    Kerkdijk, H.; Mol, H.G.J.; Nagel, B. van der

    2007-01-01

    A new concept for cleanup, based on volume overloading of the cleanup column, has been developed for on-line coupling of gel permeation chromatography (GPC), solid-phase extraction (SPE), or both, to gas chromatography (GC). The principle is outlined and the applicability demonstrated by the determi

  20. Commissioning of the STAR test section for experimental simulation of loss of coolant accident using the EC-208 instrumented fuel assembly of the IEA-R1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Maprelian, Eduardo; Torres, Walmir M.; Prado, Adelk C.; Umbehaun, Pedro E.; Franca, Renato L.; Santos, Samuel C.; Macedo, Luiz A.; Sabundjian, Gaiane, E-mail: emaprel@ipen.br, E-mail: wmtorres@ipen.br, E-mail: acprado@ipen.br, E-mail: umbehaun@ipen.br, E-mail: rlfranca@ipen.br, E-mail: samuelcs@ipen.br, E-mail: lamacedo@ipen.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SO (Brazil)

    2015-07-01

    The three basic safety functions of Research Reactors (RR) are the safe shutdown of the reactor, the proper cooling of the decay heat of the fuel elements and the confinement of radioactive materials. Compared to Nuclear Power Reactors, RR power release is small, yet its three safety functions must be met to ensure the integrity of the reactor. During a loss of coolant accident (LOCA) in pool type RR, partial or complete loss of pool water may occur, with consequent partial or complete uncovering of the fuel assemblies. In such an accident, the decay heat removal safety function must not be compromised. The Test Section for Experimental Simulation of Loss of Coolant Accident (STAR) is in commissioning phase. This test section will provide experimental data on partial and total uncovering of the EC-208 instrumented fuel assembly (IFA) irradiated in the IEA-R1. Experimental results will be useful in validation of computer codes for RR safety analysis, particularly on heat removal efficiency aspects (safety function) in accident conditions. STAR comprises a base on which is installed the IFA, the cylindrical stainless steel hull, the compressed air system for the test section emptying and refilling, and the instrumentation for temperature and level measurements. The commissioning tests or pre-operational check, consist of several preliminary tests to verify experimental procedures, the difficulties during assembling of STAR in the pool, the difficulties in control the emptying and refilling velocities, as well as, the repeatability capacity, tests of equipment, valves and systems and tests of instrumentation and data acquisition system. Safety, accuracy and easiness of operation will be checked. (author)