WorldWideScience

Sample records for hydraulic fluid safety

  1. CRITICALITY CURVES FOR PLUTONIUM HYDRAULIC FLUID MIXTURES

    International Nuclear Information System (INIS)

    WITTEKIND WD

    2007-01-01

    This Calculation Note performs and documents MCNP criticality calculations for plutonium (100% 239 Pu) hydraulic fluid mixtures. Spherical geometry was used for these generalized criticality safety calculations and three geometries of neutron reflection are: (sm b ullet)bare, (sm b ullet)1 inch of hydraulic fluid, or (sm b ullet)12 inches of hydraulic fluid. This document shows the critical volume and critical mass for various concentrations of plutonium in hydraulic fluid. Between 1 and 2 gallons of hydraulic fluid were discovered in the bottom of HA-23S. This HA-23S hydraulic fluid was reported by engineering to be Fyrquel 220. The hydraulic fluid in GLovebox HA-23S is Fyrquel 220 which contains phosphorus. Critical spherical geometry in air is calculated with 0 in., 1 in., or 12 inches hydraulic fluid reflection

  2. Handbook of hydraulic fluid technology

    CERN Document Server

    Totten, George E

    2011-01-01

    ""The Handbook of Hydraulic Fluid Technology"" serves as the foremost resource for designing hydraulic systems and for selecting hydraulic fluids used in engineering applications. Featuring new illustrations, data tables, as well as practical examples, this second edition is updated with essential information on the latest hydraulic fluids and testing methods. The detailed text facilitates unparalleled understanding of the total hydraulic system, including important hardware, fluid properties, and hydraulic lubricants. Written by worldwide experts, the book also offers a rigorous overview of h

  3. Two-fluid modeling of thermal-hydraulic phenomena for best-estimate LWR safety analysis

    International Nuclear Information System (INIS)

    Yadigaroglu, G.; Andreani, M.

    1989-01-01

    Two-fluid formulation of the conservation equations has allowed modelling of the two-phase flow and heat transfer phenomena and situations involving strong departures in thermal and velocity equilibrium between the phases. The paper reviews the state of the art in modelling critical flows, and certain phase separation phenomena, as well as post-dryout heat transfer situations. Although the two-fluid models and the codes have the potential for correctly modelling such situations, this potential has not always been fully used in practice. (orig.)

  4. Disclosure of hydraulic fracturing fluid chemical additives: analysis of regulations.

    Science.gov (United States)

    Maule, Alexis L; Makey, Colleen M; Benson, Eugene B; Burrows, Isaac J; Scammell, Madeleine K

    2013-01-01

    Hydraulic fracturing is used to extract natural gas from shale formations. The process involves injecting into the ground fracturing fluids that contain thousands of gallons of chemical additives. Companies are not mandated by federal regulations to disclose the identities or quantities of chemicals used during hydraulic fracturing operations on private or public lands. States have begun to regulate hydraulic fracturing fluids by mandating chemical disclosure. These laws have shortcomings including nondisclosure of proprietary or "trade secret" mixtures, insufficient penalties for reporting inaccurate or incomplete information, and timelines that allow for after-the-fact reporting. These limitations leave lawmakers, regulators, public safety officers, and the public uninformed and ill-prepared to anticipate and respond to possible environmental and human health hazards associated with hydraulic fracturing fluids. We explore hydraulic fracturing exemptions from federal regulations, as well as current and future efforts to mandate chemical disclosure at the federal and state level.

  5. Hydraulic fracturing chemicals and fluids technology

    CERN Document Server

    Fink, Johannes

    2013-01-01

    When classifying fracturing fluids and their additives, it is important that production, operation, and completion engineers understand which chemical should be utilized in different well environments. A user's guide to the many chemicals and chemical additives used in hydraulic fracturing operations, Hydraulic Fracturing Chemicals and Fluids Technology provides an easy-to-use manual to create fluid formulations that will meet project-specific needs while protecting the environment and the life of the well. Fink creates a concise and comprehensive reference that enables the engineer to logically select and use the appropriate chemicals on any hydraulic fracturing job. The first book devoted entirely to hydraulic fracturing chemicals, Fink eliminates the guesswork so the engineer can select the best chemicals needed on the job while providing the best protection for the well, workers and environment. Pinpoints the specific compounds used in any given fracturing operation Provides a systematic approach to class...

  6. Understanding, Classifying, and Selecting Environmentally Acceptable Hydraulic Fluids

    Science.gov (United States)

    2016-08-01

    traditional mineral oil; therefore, the life cycle costs over time may be reduced . REPLACEMENT OF EXISTING HYDRAULIC FLUIDS: Hydraulic fluids in existing...properly maintaining the fluid can extend the time interval between fluid changes, thus reducing the overall operating cost of the EA hydraulic fluid. It...Environmentally Acceptable Hydraulic Fluids by Timothy J. Keyser, Robert N. Samuel, and Timothy L. Welp INTRODUCTION: On a daily basis, the United States Army

  7. Fluid Temperature of Aero Hydraulic Systems

    Directory of Open Access Journals (Sweden)

    I. S. Shumilov

    2016-01-01

    Full Text Available In modern supersonic aircrafts due to aerodynamic skin heating a temperature of hydraulics environment significantly exceeds that of permissible for fluids used. The same problem exists for subsonic passenger aircrafts, especially for Airbuses, which have hydraulics of high power where convective heat transfer with the environment is insufficient and there is no required temperature control of fluid. The most significant in terms of heat flow is the flow caused by the loss of power to the pump and when designing the hydraulic system (HS it is necessary to pay very serious attention to it. To use a constant capacity pump is absolutely unacceptable, since HS efficiency in this case is extremely low, and the most appropriate are variable-capacity pumps, cut-off pumps, dual-mode pumps. The HS fluid cooling system should provide high reliability, lightweight, simple design, and a specified heat transfer in all flight modes.A system cooling the fluid by the fuel of feeding lines of the aircraft engines is the most effective, and it is widely used in supersonic aircrafts, where power of cooling system is essential. Subsonic aircrafts widely use convective heat exchangers. In thermal design of the aircraft hydraulics, the focus is generally given to the maximum and minimum temperatures of the HS fluid, the choice of the type of heat exchanger (convective or flow-through, the place of its installation. In calculating the operating temperature of a hydraulic system and its cooling systems it is necessary to determine an increase of the working fluid temperature when throttling it. There are three possible formulas to calculate the fluid temperature in throttling, with the error of a calculated temperature drop from 30% to 4%.The article considers the HS stationary and noon-stationary operating conditions and their calculation, defines temperatures of fluid and methods to control its specified temperature. It also discusses various heat exchanger schemes

  8. Safety valve including a hydraulic brake and hydraulic brake that could be fitted into a valve

    International Nuclear Information System (INIS)

    Chabat-Courrede, Jean.

    1981-01-01

    Making of a safety valve that can be fitted to a containment vessel filled with a non compressible fluid, such as the water system of a nuclear power station. It includes a hydraulic brake located between the valve and the elastic means, close to the valve which completely suppresses the high frequency oscillations of the equipment [fr

  9. Review of fluid and control technology of hydraulic wind turbines

    Science.gov (United States)

    Cai, Maolin; Wang, Yixuan; Jiao, Zongxia; Shi, Yan

    2017-09-01

    This study examines the development of the fluid and control technology of hydraulic wind turbines. The current state of hydraulic wind turbines as a new technology is described, and its basic fluid model and typical control method are expounded by comparing various study results. Finally, the advantages of hydraulic wind turbines are enumerated. Hydraulic wind turbines are expected to become the main development direction of wind turbines.

  10. Review of fluid and control technology of hydraulic wind turbines

    Institute of Scientific and Technical Information of China (English)

    Maolin CAI; Yixuan WANG; Zongxia JIAO; Yan SHI

    2017-01-01

    This study examines the development of the fluid and control technology of hydraulic wind turbines.The current state of hydraulic wind turbines as a new technology is described,and its basic fluid model and typical control method are expounded by comparing various study results.Finally,the advantages of hydraulic wind turbines are enumerated.Hydraulic wind turbines are expected to become the main development direction of wind turbines.

  11. Motion simulation of hydraulic driven safety rod using FSI method

    International Nuclear Information System (INIS)

    Jung, Jaeho; Kim, Sanghaun; Yoo, Yeonsik; Cho, Yeonggarp; Kim, Jong In

    2013-01-01

    Hydraulic driven safety rod which is one of them is being developed by Division for Reactor Mechanical Engineering, KAERI. In this paper the motion of this rod is simulated by fluid structure interaction (FSI) method before manufacturing for design verification and pump sizing. A newly designed hydraulic driven safety rod which is one of reactivity control mechanism is simulated using FSI method for design verification and pump sizing. The simulation is done in CFD domain with UDF. The pressure drop is changed slightly by flow rates. It means that the pressure drop is mainly determined by weight of moving part. The simulated velocity of piston is linearly proportional to flow rates so the pump can be sized easily according to the rising and drop time requirement of the safety rod using the simulation results

  12. Method to Estimate the Dissolved Air Content in Hydraulic Fluid

    Science.gov (United States)

    Hauser, Daniel M.

    2011-01-01

    In order to verify the air content in hydraulic fluid, an instrument was needed to measure the dissolved air content before the fluid was loaded into the system. The instrument also needed to measure the dissolved air content in situ and in real time during the de-aeration process. The current methods used to measure the dissolved air content require the fluid to be drawn from the hydraulic system, and additional offline laboratory processing time is involved. During laboratory processing, there is a potential for contamination to occur, especially when subsaturated fluid is to be analyzed. A new method measures the amount of dissolved air in hydraulic fluid through the use of a dissolved oxygen meter. The device measures the dissolved air content through an in situ, real-time process that requires no additional offline laboratory processing time. The method utilizes an instrument that measures the partial pressure of oxygen in the hydraulic fluid. By using a standardized calculation procedure that relates the oxygen partial pressure to the volume of dissolved air in solution, the dissolved air content is estimated. The technique employs luminescent quenching technology to determine the partial pressure of oxygen in the hydraulic fluid. An estimated Henry s law coefficient for oxygen and nitrogen in hydraulic fluid is calculated using a standard method to estimate the solubility of gases in lubricants. The amount of dissolved oxygen in the hydraulic fluid is estimated using the Henry s solubility coefficient and the measured partial pressure of oxygen in solution. The amount of dissolved nitrogen that is in solution is estimated by assuming that the ratio of dissolved nitrogen to dissolved oxygen is equal to the ratio of the gas solubility of nitrogen to oxygen at atmospheric pressure and temperature. The technique was performed at atmospheric pressure and room temperature. The technique could be theoretically carried out at higher pressures and elevated

  13. Fluid Structure Interaction for Hydraulic Problems

    International Nuclear Information System (INIS)

    Souli, Mhamed; Aquelet, Nicolas

    2011-01-01

    Fluid Structure interaction plays an important role in engineering applications. Physical phenomena such as flow induced vibration in nuclear industry, fuel sloshing tank in automotive industry or rotor stator interaction in turbo machinery, can lead to structure deformation and sometimes to failure. In order to solve fluid structure interaction problems, the majority of numerical tests consists in using two different codes to separately solve pressure of the fluid and structural displacements. In this paper, a unique code with an ALE formulation approach is used to implicitly calculate the pressure of an incompressible fluid applied to the structure. The development of the ALE method as well as the coupling in a computational structural dynamic code, allows to solve more large industrial problems related to fluid structure coupling. (authors)

  14. Application study of magnetic fluid seal in hydraulic turbine

    International Nuclear Information System (INIS)

    Yu, Z Y; Zhang, W

    2012-01-01

    The waterpower resources of our country are abundant, and the hydroelectric power is developed, but at present the main shaft sealing device of hydraulic turbine is easy to wear and tear and the leakage is great. The magnetic fluid seal has the advantages of no contact, no wear, self-healing, long life and so on. In this paper, the magnetic fluid seal would be used in the main shaft of hydraulic turbine, the sealing structure was built the model, meshed the geometry, applied loads and solved by using MULTIPHYSICS in ANSYS software, the influence of the various sealing structural parameters such as tooth width, height, slot width, sealing gap on the sealing property were analyzed, the magnetic fluid sealing device suitable for large-diameter shaft and sealing water was designed, the sealing problem of the hydraulic turbine main shaft was solved effectively which will bring huge economic benefits.

  15. Endurance Pump Test with MIL-PRF-83282 Hydraulic Fluid, Purified with Malabar Purifier

    National Research Council Canada - National Science Library

    Sharma, Shashi

    2004-01-01

    .... Endurance aircraft hydraulic pump tests under carefully controlled conditions were previously conducted using hydraulic fluid purified with a rotating-disk and vacuum type purifier, the portable...

  16. Thermal hydraulic reactor safety analyses and experiments

    International Nuclear Information System (INIS)

    Holmstroem, H.; Eerikaeinen, L.; Kervinen, T.; Kilpi, K.; Mattila, L.; Miettinen, J.; Yrjoelae, V.

    1989-04-01

    The report introduces the results of the thermal hydraulic reactor safety research performed in the Nuclear Engineering Laboratory of the Technical Research Centre of Finland (VTT) during the years 1972-1987. Also practical applications i.e. analyses for the safety authorities and power companies are presented. The emphasis is on description of the state-of-the-art know how. The report describes VTT's most important computer codes, both those of foreign origin and those developed at VTT, and their assessment work, VTT's own experimental research, as well as international experimental projects and other forms of cooperation VTT has participated in. Appendix 8 contains a comprehensive list of the most important publications and technical reports produced. They present the content and results of the research in detail.(orig.)

  17. Theoretical aspects concerning working fluids in hydraulic systems

    Directory of Open Access Journals (Sweden)

    Tița Irina

    2017-01-01

    Full Text Available Among the properties of working fluid, viscosity is the most important as it regards especially to pumps. In order to study the behavior of hydrostatic transmission it is important to create a reliable research instrument for dynamic simulation. Our research expertise being in SimHydraulics consequently this instrument is the suitable block diagram. The purpose of this paper is to present the possible ways to customize the properties of the working fluid in the block diagram.

  18. Savoir Fluide. A newsletter on computational hydraulics and fluid dynamics

    International Nuclear Information System (INIS)

    1997-01-01

    This newsletter reports on computational works performed by the National Laboratory of Hydraulics (LNH) from Electricite de France (EdF). Two papers were selected which concern the simulation of the Paluel nuclear power plant plume and the computation of particles and droplets inside a cooling tower. (J.S.)

  19. Thermal Hydraulic Tests for Reactor Core Safety

    Energy Technology Data Exchange (ETDEWEB)

    Moon, S. K.; Baek, W. P.; Chun, S. Y. (and others)

    2007-06-15

    The main objectives of the present project are to resolve the current issues of reactor core thermal hydraulics, to develop an advanced measurement and analytical techniques, and to perform reactor core safety verification tests. 6x6 reflood experiments, various heat transfer experiments using Freon, and experiments on the spacer grids effects on the post-dryout are carried out using spacer grids developed in Korea in order to resolve the current issues of the reactor core thermal hydraulics. In order to develop a reflood heat transfer model, the detailed reflood phenomena are visualized and measured using round tube and 2x2 rod bundle. A detailed turbulent mixing phenomenon for subchannels is measured using advanced measurement techniques such as LDV and PIV. MARS and MATRA codes developed in Korea are assessed, verified and improved using the obtained experimental data. Finally, a systematic quality assurance program and experimental data generation system has been constructed in order to increase the reliability of the experimental data.

  20. TMI-2 in-vessel hydraulic systems utilize high water and high boron content fluids

    International Nuclear Information System (INIS)

    Baston, V.F.; Hofstetter, K.J.; Hofman, L.A.; Gallagher, R.E.

    1987-01-01

    Choice of a hydraulic fluid for use in the Three Mile Island Unit 2 (TMI-2) reactor vessel defueling equipment required consideration of the following constraints for the hydraulic fluid given an accidental spill into the reactor coolant system (RCS). The TMI-2 RCS hydraulic fluid utilized in the hydraulic operations utilized a solution composition of 95 wt% water and 5 wt% of the above base fluid. The TMI-2 hydraulic system utilizes pressures up to 3500 psi. The selected hydraulic fluid has been in use since December 1986 with minimal operational difficulties

  1. Endurance Pump Tests With Fresh and Purified MIL-PRF-83282 Hydraulic Fluid

    National Research Council Canada - National Science Library

    Sharma, Shashi

    1999-01-01

    .... Two endurance pump tests were conducted with F-16 aircraft hydraulic pumps, using both fresh and purified MIL-PRF-83282 hydraulic fluid, to determine if fluid purification had any adverse effect on pump life...

  2. Loss-of-Fluid Test findings in pressurized water reactor core's thermal-hydraulic behavior

    International Nuclear Information System (INIS)

    Russell, M.

    1983-01-01

    This paper summarizes the pressurized water reactor (PWR) core's thermal-hydraulic behavior findings from experiments performed at the Loss-of-Fluid Test (LOFT) Facility at the Idaho National Engineering Laboratory. The potential impact of these findings on the safety and economics of PWR's generation of electricity is also discussed. Reviews of eight important findings in the core's physical behavior and in experimental methods are presented with supporting evidence

  3. High-water-base hydraulic fluid-irradiation experiments

    International Nuclear Information System (INIS)

    Bradley, E.C.; Meacham, S.A.

    1981-10-01

    A remote system for shearing spent nuclear fuel assemblies is being designed under the direction of the Consolidated Fuel Reprocessing Program (CFRP). The design incorporates a dual hydraulic fluid actuation system in which only one of the fluids, a high-water-base (HWBF), would be exposed to ionizing radiation and radioactive contamination. A commercially available synthetic, solution-type HWBF was selected as the reference. Single-sample irradiation experiments were conducted with three commercial fluids over a range of irradiation exposures. The physical and chemical properties of the irradiated HWBFs were analyzed and compared with unirradiated samples. In general, the results of the analyses showed increasing degradation of fluid properties with increasing irradiation dose. The results also indicated that a synthetic solution-type HWBF would perform satisfactorily in the remote shear system where irradiation doses up to 10 6 Gy (10 8 rad) are expected

  4. High-water-base hydraulic fluid-irradiation experiments

    Energy Technology Data Exchange (ETDEWEB)

    Bradley, E.C.; Meacham, S.A.

    1981-10-01

    A remote system for shearing spent nuclear fuel assemblies is being designed under the direction of the Consolidated Fuel Reprocessing Program (CFRP). The design incorporates a dual hydraulic fluid actuation system in which only one of the fluids, a high-water-base (HWBF), would be exposed to ionizing radiation and radioactive contamination. A commercially available synthetic, solution-type HWBF was selected as the reference. Single-sample irradiation experiments were conducted with three commercial fluids over a range of irradiation exposures. The physical and chemical properties of the irradiated HWBFs were analyzed and compared with unirradiated samples. In general, the results of the analyses showed increasing degradation of fluid properties with increasing irradiation dose. The results also indicated that a synthetic solution-type HWBF would perform satisfactorily in the remote shear system where irradiation doses up to 10/sup 6/ Gy (10/sup 8/ rad) are expected.

  5. Thermal hydraulic tests for reactor safety system -Research on the improvement of nuclear safety-

    International Nuclear Information System (INIS)

    Chung, Moon Ki; Park, Chun Kyeong; Yang, Seon Kyu; Chung, Chang Hwan; Chun, Shee Yeong; Song, Cheol Hwa; Chun, Hyeong Gil; Chang, Seok Kyu; Chung, Heung Joon; Won, Soon Yeon; Cho, Yeong Ro; Kim, Bok Deuk; Min, Kyeong Ho

    1994-07-01

    The present research aims at the development of the thermal hydraulic verification test technology for the reactor safety system of the conventional and advanced nuclear power plant and the development of the advanced thermal hydraulic measuring techniques. (Author)

  6. Organic compounds in hydraulic fracturing fluids and wastewaters: A review.

    Science.gov (United States)

    Luek, Jenna L; Gonsior, Michael

    2017-10-15

    High volume hydraulic fracturing (HVHF) of shale to stimulate the release of natural gas produces a large quantity of wastewater in the form of flowback fluids and produced water. These wastewaters are highly variable in their composition and contain a mixture of fracturing fluid additives, geogenic inorganic and organic substances, and transformation products. The qualitative and quantitative analyses of organic compounds identified in HVHF fluids, flowback fluids, and produced waters are reviewed here to communicate knowledge gaps that exist in the composition of HVHF wastewaters. In general, analyses of organic compounds have focused on those amenable to gas chromatography, focusing on volatile and semi-volatile oil and gas compounds. Studies of more polar and non-volatile organic compounds have been limited by a lack of knowledge of what compounds may be present as well as quantitative methods and standards available for analyzing these complex mixtures. Liquid chromatography paired with high-resolution mass spectrometry has been used to investigate a number of additives and will be a key tool to further research on transformation products that are increasingly solubilized through physical, chemical, and biological processes in situ and during environmental contamination events. Diverse treatments have been tested and applied to HVHF wastewaters but limited information has been published on the quantitative removal of individual organic compounds. This review focuses on recently published information on organic compounds identified in flowback fluids and produced waters from HVHF. Copyright © 2017 Elsevier Ltd. All rights reserved.

  7. Nuclear reactor thermal hydraulics safety analysis and thoughts on FUKUSHIMA

    International Nuclear Information System (INIS)

    Ninokata, Hisashi

    2012-01-01

    The first part of this article is to show my thoughts on the accident at Fukushima Daiichi Nuclear Power Station. It is cited from a summary of my lecture talk in Indonesia, in the beginning of the last December, 2011. This talk was based on my previous lecture and seminar talks including those delivered at MIT, June 16, at the ANS Annual Meeting in Hollywood, Florida, June 28 at NURETH-13 in Toronto, September 27, and others. The content is based on the open and latest information available to date in Japan. It may contain some erroneous or uncertain information. I tried to minimize it to my best capability. Also I tried to eliminate any critical issues or opinions that may jeopardize some people who were involved in. The latter half of this article will be excerpts of my recent R and D activities related to the safety-by-design for sodium cooled fast reactors and light water reactors, thermal hydraulics analysis focusing on the simulation-based technology, in particular subchannel analysis and computational fluid dynamics. (J.P.N.)

  8. Schaum’s outline of fluid mechanics and hydraulics

    CERN Document Server

    Giles, Ranald V; Liu, Cheng

    2014-01-01

    Tough Test Questions? Missed Lectures? Not Enough Time? Fortunately, there's Schaum's. More than 40 million students have trusted Schaum's to help them succeed in the classroom and on exams. Schaum's is the key to faster learning and higher grades in every subject. Each Outline presents all the essential course information in an easy-to-follow, topic-by-topic format. You also get hundreds of examples, solved problems, and practice exercises to test your skills. This Schaum's Outline gives you: 622 fully solved problems; extra practice on topics such as buoyancy and flotation, complex pipeline systems, fluid machinery, flow in open channels, and more; and support for all the major textbooks for fluidmechanics and hydraulics courses. Fully compatible with your classroom text, Schaum's highlights all the important facts you need to know. Use Schaum's to shorten your study time - and get your best test scores! Schaum's Outlines - Problem Solved.

  9. Contamination Control and Monitoring of Tap Water as Fluid in Industrial Tap Water Hydraulic Systems

    DEFF Research Database (Denmark)

    Conrad, Finn; Adelstorp, Anders

    1998-01-01

    Presentation of results and methods addressed to contamination control and monitoring of tap water as fluid in tap water hydraulic systems.......Presentation of results and methods addressed to contamination control and monitoring of tap water as fluid in tap water hydraulic systems....

  10. Operation of a T63 Turbine Engine Using F24 Contaminated Skydrol 5 Hydraulic Fluid

    Science.gov (United States)

    2016-09-01

    hydraulic fluids were originally developed by the Douglas Aircraft Company during the 1940s to reduce fire risk from leaking high pressure mineral oil...thermal load demands in modern hydraulic systems and reduced density to lower weight impact on the aircraft. Eastman Chemical is the current producer of...AFRL-RQ-WP-TM-2016-0155 OPERATION OF A T63 TURBINE ENGINE USING F24 CONTAMINATED SKYDROL 5 HYDRAULIC FLUID Matthew J. Wagner (AFRL/RQTM) James

  11. Overview of Chronic Oral Toxicity Values for Chemicals Present in Hydraulic Fracturing Fluids, Flowback and Produced Waters

    Science.gov (United States)

    as part of EPA's Hydraulic Fracturing Drinking Water Assessment, EPA is summarizing existing toxicity data for chemicals reported to be used in hydraulic fracturing fluids and/or found in flowback or produced waters from hydraulically fractured wells

  12. Development of regulatory technology for thermal-hydraulic safety analysis

    International Nuclear Information System (INIS)

    Bang, Young Seok; Lee, S. H.; Ryu, Y. H.

    2001-02-01

    The present study aims to develop the regulation capability in thermal-hydraulic safety analysis which was required for the reasonable safety regulation in the current NPP, the next generation reactors, and the future-type reactors. The fourth fiscal year of the first phase of the research was focused on the following research topics: Investigation on the current status of the thermal-hydraulic safety analysis technology outside and inside of the country; Review on the improved features of the thermal-hydraulic safety analysis regulatory audit code, RELAP5/MOD3; Assessments of code with LOFT L9-3 ATWS experiment and LSTF SB-SG-10 multiple SGTR experiment; Application of the RELAP5/CANDU code to analyses of SLB and LBLOCA and evaluation of its effect on safety; Application of the code to IAEA PHWR ISP analysis; Assessments of RELAP5 and TRAC with UPTF downcomer injection test and Analysis of LBLOCA with RELAP5 for the performance evaluation of KNGR DVI; Setup of a coupled 3-D kinetics and thermal-hydraulics and application it to a reactivity accident analysis; and Extension of database and improvement of plant input decks. For supporting the resolution of safety issues, loss of RHR event during midloop operation was analyzed for Kori Unit 3, issues on high burnup fuel were reviewed and performance of FRAPCON-3 assessed. Also MSLB was analyzed to figure out the sensitivity of downcomer temperature supporting the PTS risk evaluation of Kori Unit 1. Thermal stratification in pipe was analyzed using the method proposed. And a method predicting the thermal-hydraulic performance of IRWST of KNGR was explored. The PWR ECCS performance criteria was issued as a MOST Article 200-19.and a regulatory guide on evaluation methodology was improved to cover concerns raised from the related licensing review process

  13. Performance and safety of hydraulic turbines

    International Nuclear Information System (INIS)

    Brekke, H

    2010-01-01

    The first part of the paper contains the choice of small turbines for run of the river power plants. Then a discussion is given on the optimization of the performance of different types of large turbines. Finally a discussion on the safety and necessary maintenance of turbines is given with special attention to bolt connections.

  14. Validation of thermal hydraulic codes for fusion reactors safety

    International Nuclear Information System (INIS)

    Sardain, P.; Gulden, W.; Massaut, V.; Takase, K.; Merill, B.; Caruso, G.

    2006-01-01

    A significant effort has been done worldwide on the validation of thermal hydraulic codes, which can be used for the safety assessment of fusion reactors. This work is an item of an implementing agreement under the umbrella of the International Energy Agency. The European part is supported by EFDA. Several programmes related to transient analysis in water-cooled fusion reactors were run in order to assess the capabilities of the codes to treat the main physical phenomena governing the accidental sequences related to water/steam discharge into the vacuum vessel or the cryostat. The typical phenomena are namely the pressurization of a volume at low initial pressure, the critical flow, the flashing, the relief into an expansion volume, the condensation of vapor in a pressure suppression system, the formation of ice on a cryogenic structure, the heat transfer between walls and fluid in various thermodynamic conditions. · A benchmark exercise has been done involving different types of codes, from homogeneous equilibrium to six equations non-equilibrium models. Several cases were defined, each one focusing on a particular phenomenon. · The ICE (Ingress of Coolant Event) facility has been operated in Japan. It has simulated an in-vessel LOCA and the discharge of steam into a pressure suppression system. · The EVITA (European Vacuum Impingement Test Apparatus) facility has been operated in France. It has simulated ingress of coolant into the cryostat, i.e. into a volume at low initial pressure containing surfaces at cryogenic temperature. This paper gives the main lessons gained from these programs, in particular the possibilities for the improvement of the computer codes, extending their capabilities. For example, the water properties have been extended below the triple point. Ice formation models have been implemented. Work has also been done on condensation models. The remaining needs for R-and-D are also highlighted. (author)

  15. Research on the improvement of nuclear safety -Thermal hydraulic tests for reactor safety system-

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Moon Kee; Park, Choon Kyung; Yang, Sun Kyoo; Chun, Se Yung; Song, Chul Hwa; Jun, Hyung Kil; Jung, Heung Joon; Won, Soon Yun; Cho, Yung Roh; Min, Kyung Hoh; Jung, Jang Hwan; Jang, Suk Kyoo; Kim, Bok Deuk; Kim, Wooi Kyung; Huh, Jin; Kim, Sook Kwan; Moon, Sang Kee; Lee, Sang Il [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-06-01

    The present research aims at the development of the thermal hydraulic verification test technology for the safety system of the conventional and advanced nuclear power plant and the development of the advanced thermal hydraulic measuring techniques. In this research, test facilities simulating the primary coolant system and safety system are being constructed for the design verification tests of the existing and advanced nuclear power plant. 97 figs, 14 tabs, 65 refs. (Author).

  16. Thermal-Hydraulic Tests for Reactor Core Safety

    International Nuclear Information System (INIS)

    Chun, Se Young; Chung, Moon Ki; Baek, Won Pil and others

    2005-04-01

    The reflood experiments for single rod annulus geometry have been performed to investigate the effect of spacer grid on thermal-hydraulics under reflood conditions. The reflood experimental loop for 6x6 rod bundle with a spacer grid developed in Korea has been provided. About 8000 data points for Post-CHF heat transfer have been obtained from the experiments About 1400 CHF data points for 3x3 Water and 5x5 Freon rod bundles have been obtained. The existing evaluation methodology for core safety under return-to-power conditions has been investigated using KAERI low flow CHF database. The hydraulic tests for turbulence mixing characteristics in subchannel of 5x5 rod bundle have been carried out using advanced measurement technique, LVD and the database for various spacer grids have been provided. In order to measure the turbulence mixing characteristics in details, the hydraulic loop with a magnified 5x5 rod bundle has been prepared. The database which was constructed through a systematic thermal hydraulic tests for the reflood phenomenon, CHF, Post-CHF is surely to be useful to the industry field, the regulation body and the development of thermal-hydraulic analysis code

  17. ATLAS program for advanced thermal-hydraulic safety research

    International Nuclear Information System (INIS)

    Song, Chul-Hwa; Choi, Ki-Yong; Kang, Kyoung-Ho

    2015-01-01

    Highlights: • Major achievements of the ATLAS program are highlighted in conjunction with both developing advanced light water reactor technologies and enhancing the nuclear safety. • The ATLAS data was shown to be useful for the development and licensing of new reactors and safety analysis codes, and also for nuclear safety enhancement through domestic and international cooperative programs. • A future plan for the ATLAS testing is introduced, covering recently emerging safety issues and some generic thermal-hydraulic concerns. - Abstract: This paper highlights the major achievements of the ATLAS program, which is an integral effect test program for both developing advanced light water reactor technologies and contributing to enhancing nuclear safety. The ATLAS program is closely related with the development of the APR1400 and APR"+ reactors, and the SPACE code, which is a best-estimate system-scale code for a safety analysis of nuclear reactors. The multiple roles of ATLAS testing are emphasized in very close conjunction with the development, licensing, and commercial deployment of these reactors and their safety analysis codes. The role of ATLAS for nuclear safety enhancement is also introduced by taking some examples of its contributions to voluntarily lead to multi-body cooperative programs such as domestic and international standard problems. Finally, a future plan for the utilization of ATLAS testing is introduced, which aims at tackling recently emerging safety issues such as a prolonged station blackout accident and medium-size break LOCA, and some generic thermal-hydraulic concerns as to how to figure out multi-dimensional phenomena and the scaling issue.

  18. Thickened water-based hydraulic fluid with reduced dependence of viscosity on temperature

    Energy Technology Data Exchange (ETDEWEB)

    Deck, C. F.

    1985-01-01

    Improved hydraulic fluids or metalworking lubricants, utilizing mixtures of water, metal lubricants, metal corrosion inhibitors, and an associative polyether thickener, have reduced dependence of the viscosity on temperature achieved by the incorporation therein of an ethoxylated polyether surfactant.

  19. Interstitial hydraulic conductivity and interstitial fluid pressure for avascular or poorly vascularized tumors.

    Science.gov (United States)

    Liu, L J; Schlesinger, M

    2015-09-07

    A correct description of the hydraulic conductivity is essential for determining the actual tumor interstitial fluid pressure (TIFP) distribution. Traditionally, it has been assumed that the hydraulic conductivities both in a tumor and normal tissue are constant, and that a tumor has a much larger interstitial hydraulic conductivity than normal tissue. The abrupt transition of the hydraulic conductivity at the tumor surface leads to non-physical results (the hydraulic conductivity and the slope of the TIFP are not continuous at tumor surface). For the sake of simplicity and the need to represent reality, we focus our analysis on avascular or poorly vascularized tumors, which have a necrosis that is mostly in the center and vascularization that is mostly on the periphery. We suggest that there is an intermediary region between the tumor surface and normal tissue. Through this region, the interstitium (including the structure and composition of solid components and interstitial fluid) transitions from tumor to normal tissue. This process also causes the hydraulic conductivity to do the same. We introduce a continuous variation of the hydraulic conductivity, and show that the interstitial hydraulic conductivity in the intermediary region should be monotonically increasing up to the value of hydraulic conductivity in the normal tissue in order for the model to correspond to the actual TIFP distribution. The value of the hydraulic conductivity at the tumor surface should be the lowest in value. Copyright © 2015 Elsevier Ltd. All rights reserved.

  20. Thermal hydraulic and safety analyses for Pakistan Research Reactor-1

    International Nuclear Information System (INIS)

    Bokhari, I.H.; Israr, M.; Pervez, S.

    1999-01-01

    Thermal hydraulic and safety analysis of Pakistan Research Reactor-1 (PARR-1) utilizing low enriched uranium (LEU) fuel have been performed using computer code PARET. The present core comprises of 29 standard and 5 control fuel elements. Results of the thermal hydraulic analysis show that the core can be operated at a steady-state power level of 10 MW for a flow rate of 950 m 3 /h, with sufficient safety margins against ONB (onset of nucleate boiling) and DNB (departure from nucleate boiling). Safety analysis has been carried out for various modes of reactivity insertions. The events studied include: start-up accident; accidental drop of a fuel element in the core; flooding of a beam tube with water; removal of an in-pile experiment during reactor operation etc. For each of these transients, time histories of reactor power, energy released and clad surface temperature etc. were calculated. The results indicate that the peak clad temperatures remain well below the clad melting temperature during these accidents. It is therefore concluded that the reactor can be safely operated at 10 MW without compromising safety. (author)

  1. Thermal-hydraulic tests for reactor safety system

    International Nuclear Information System (INIS)

    Chun, Se Young; Chung, Moon Ki; Baek, Won Pil

    2002-05-01

    Tests for the safety depressurization system, Sparger adopted for the Korean next generation reactor, APR1400 are carried out for several geometries with the B and C (Blowdown and Condensation) facility in the condition of high temperature and pressure and with a small test facility in the condition of atmospheric temperature and pressure. Tests for the critical heat flux are performed with the RCS(Reactor Coolant System) facility as well as with the Freon CHF Loop in the condition of high temperature and pressure. The atmospheric temperature and pressure facility is utilized for development of the high standard thermal hydraulic measurement technology. The optical method is developed to measure the local thermal-hydraulic behavior for the single and two-phase boiling phenomena

  2. Comparison of EPRI safety valve test data with analytically determined hydraulic results

    International Nuclear Information System (INIS)

    Smith, L.C.; Howe, K.S.

    1983-01-01

    NUREG-0737 (November 1980) and all subsequent U.S. NRC generic follow-up letters require that all operating plant licensees and applicants verify the acceptability of plant specific pressurizer safety valve piping systems for valve operation transients by testing. To aid in this verification process, the Electric Power Research Institute (EPRI) conducted an extensive testing program at the Combustion Engineering Test Facility. Pertinent tests simulating dynamic opening of the safety valves for representative upstream environments were carried out. Different models and sizes of safety valves were tested at the simulated operating conditions. Transducers placed at key points in the system monitored a variety of thermal, hydraulic and structural parameters. From this data, a more complete description of the transient can be made. The EPRI test configuration was analytically modeled using a one-dimensional thermal hydraulic computer program that uses the method of characteristics approach to generate key fluid parameters as a function of space and time. The conservative equations are solved by applying both the implicit and explicit characteristic methods. Unbalanced or wave forces were determined for each straight run of pipe bounded on each side by a turn or elbow. Blowdown forces were included, where appropriate. Several parameters were varied to determine the effects on the pressure, hydraulic forces and timings of events. By comparing these quantities with the experimentally obtained data, an approximate picture of the flow dynamics is arrived at. Two cases in particular are presented. These are the hot and cold loop seal discharge tests made with the Crosby 6M6 spring-loaded safety valve. Included in the paper is a description of the hydraulic code, modeling techniques and assumptions, a comparison of the numerical results with experimental data and a qualitative description of the factors which govern pipe support loading. (orig.)

  3. Mechanical testing of hydraulic fluids II; Mechanische Pruefung von Hydraulikfluessigkeiten II

    Energy Technology Data Exchange (ETDEWEB)

    Kessler, M.; Feldmann, D.G.; Laukart, V.

    2001-09-01

    Since May 1996 the Institute for Mechanical Engineering Design 1 of Technical University of Hamburg-Harburg is working on the topic of ''Mechanical Testing of Hydraulic fluids''. The first project lasting 2 1/2 years was completed in 1999, the results are published as the DGMK report 514. Within these project a testing principle for the ''mechanical testing'' of hydraulic fluids has been derived, a prototype of a test rig was designed and set in operation at the authors' institute. This DGMK-report 514-1 describes the results of the second project, which investigates the operating behaviour of the test-rig more in detail. Several test-runs with a total number of 11 different hydraulic fluids show the dependence of the different lubricating behaviour of the tested fluids and their friction and wear behaviour during the tests in a reproducible way. The aim of the project was to derive a testing principle including the design of a suitable test-rig for the mechanical testing of hydraulic fluids. Based on the described results it can be stated that with the developed test it is possible to test the lubricity of hydraulic fluids reproducible and in correlation to field experiences within a relatively short time, so the target was reached. (orig.)

  4. CONSIDERATIONS ON FLUID DYNAMICS INSIDE A HYDRAULIC SEISMIC ENERGY ABSORBER

    Directory of Open Access Journals (Sweden)

    ȘCHEAUA Fănel

    2013-06-01

    Full Text Available This study presents a method for obtaining a simplified model of a seismic energy dissipation device whose operating principle is based on viscous fluid as a solution for structural isolation against seismic actions. The device operation is based on the resistance force developed by the working fluid when the piston tends to move due to occurrence of a seismic motion. A 3D model achieved is introduced in CFD analysis for emphasize dynamic fluid flow inside the device dissipation cylinder.

  5. Shallow Aquifer Vulnerability From Subsurface Fluid Injection at a Proposed Shale Gas Hydraulic Fracturing Site

    Science.gov (United States)

    Wilson, M. P.; Worrall, F.; Davies, R. J.; Hart, A.

    2017-11-01

    Groundwater flow resulting from a proposed hydraulic fracturing (fracking) operation was numerically modeled using 91 scenarios. Scenarios were chosen to be a combination of hydrogeological factors that a priori would control the long-term migration of fracking fluids to the shallow subsurface. These factors were induced fracture extent, cross-basin groundwater flow, deep low hydraulic conductivity strata, deep high hydraulic conductivity strata, fault hydraulic conductivity, and overpressure. The study considered the Bowland Basin, northwest England, with fracking of the Bowland Shale at ˜2,000 m depth and the shallow aquifer being the Sherwood Sandstone at ˜300-500 m depth. Of the 91 scenarios, 73 scenarios resulted in tracked particles not reaching the shallow aquifer within 10,000 years and 18 resulted in travel times less than 10,000 years. Four factors proved to have a statistically significant impact on reducing travel time to the aquifer: increased induced fracture extent, absence of deep high hydraulic conductivity strata, relatively low fault hydraulic conductivity, and magnitude of overpressure. Modeling suggests that high hydraulic conductivity formations can be more effective barriers to vertical flow than low hydraulic conductivity formations. Furthermore, low hydraulic conductivity faults can result in subsurface pressure compartmentalization, reducing horizontal groundwater flow, and encouraging vertical fluid migration. The modeled worst-case scenario, using unlikely geology and induced fracture lengths, maximum values for strata hydraulic conductivity and with conservative tracer behavior had a particle travel time of 130 years to the base of the shallow aquifer. This study has identified hydrogeological factors which lead to aquifer vulnerability from shale exploitation.

  6. Safety features of subcritical fluid fueled systems

    International Nuclear Information System (INIS)

    Bell, C.R.

    1995-01-01

    Accelerator-driven transmutation technology has been under study at Los Alamos for several years for application to nuclear waste treatment, tritium production, energy generation, and recently, to the disposition of excess weapons plutonium. Studies and evaluations performed to date at Los Alamos have led to a current focus on a fluid-fuel, fission system operating in a neutron source-supported subcritical mode, using molten salt reactor technology and accelerator-driven proton-neutron spallation. In this paper, the safety features and characteristics of such systems are explored from the perspective of the fundamental nuclear safety objectives that any reactor-type system should address. This exploration is qualitative in nature and uses current vintage solid-fueled reactors as a baseline for comparison. Based on the safety perspectives presented, such systems should be capable of meeting the fundamental nuclear safety objectives. In addition, they should be able to provide the safety robustness desired for advanced reactors. However, the manner in which safety objectives and robustness are achieved is very different from that associated with conventional reactors. Also, there are a number of safety design and operational challenges that will have to be addressed for the safety potential of such systems to be credible

  7. Safety features of subcritical fluid fueled systems

    International Nuclear Information System (INIS)

    Bell, C.R.

    1994-01-01

    Accelerator-driven transmutation technology has been under study at Los Alamos for several years for application to nuclear waste treatment, tritium production, energy generation, and recently, to the disposition of excess weapons plutonium. Studies and evaluations performed to date at Los Alamos have led to a current focus on a fluid-fuel, fission system operating in a neutron source-supported subcritical mode, using molten salt reactor technology and accelerator-driven proton-neutron spallation. In this paper, the safety features and characteristics of such systems are explored from the perspective of the fundamental nuclear safety objectives that any reactor-type system should address. This exploration is qualitative in nature and uses current vintage solid-fueled reactors as a baseline for comparison. Based on the safety perspectives presented, such systems should be capable of meeting the fundamental nuclear safety objectives. In addition, they should be able to provide the safety robustness desired for advanced reactors. However, the manner in which safety objectives and robustness are achieved in very different from that associated with conventional reactors. Also, there are a number of safety design and operational challenges that will have to be addressed for the safety potential of such systems to be credible

  8. Safety features of subcritical fluid fueled systems

    Energy Technology Data Exchange (ETDEWEB)

    Bell, C.R. [Los Alamos National Laboratory, NM (United States)

    1995-10-01

    Accelerator-driven transmutation technology has been under study at Los Alamos for several years for application to nuclear waste treatment, tritium production, energy generation, and recently, to the disposition of excess weapons plutonium. Studies and evaluations performed to date at Los Alamos have led to a current focus on a fluid-fuel, fission system operating in a neutron source-supported subcritical mode, using molten salt reactor technology and accelerator-driven proton-neutron spallation. In this paper, the safety features and characteristics of such systems are explored from the perspective of the fundamental nuclear safety objectives that any reactor-type system should address. This exploration is qualitative in nature and uses current vintage solid-fueled reactors as a baseline for comparison. Based on the safety perspectives presented, such systems should be capable of meeting the fundamental nuclear safety objectives. In addition, they should be able to provide the safety robustness desired for advanced reactors. However, the manner in which safety objectives and robustness are achieved is very different from that associated with conventional reactors. Also, there are a number of safety design and operational challenges that will have to be addressed for the safety potential of such systems to be credible.

  9. The study of crosslinked fluid leakoff in hydraulic fracturing physical simulations

    Energy Technology Data Exchange (ETDEWEB)

    Grothe, Vinicius Perrud; Ribeiro, Paulo Roberto [Universidade Estadual de Campinas, SP (Brazil). Faculdade de Engenharia Mecanica. Dept. de Engenharia de Petroleo; Sousa, Jose Luiz Antunes de Oliveira e [Universidade Estadual de Campinas, SP (Brazil). Faculdade de Engenharia. Dept. de Estruturas; Fernandes, Paulo Dore [PETROBRAS S.A., Rio de Janeiro, RJ (Brazil). Centro de Pesquisas

    2000-07-01

    The fluid loss plays an important role in the design and execution of hydraulic fracturing treatments. The main objectives of this work were: the study of the fluid loss associated with the propagation of hydraulic fractures generated at laboratory; and the comparison of two distinct methods for estimating leakoff coefficients - Nolte analysis and the filtrate volume vs. square root of time plot. Synthetic rock samples were used as well as crosslinked hydroxypropyl guar (HPG) fluids in different polymer concentrations. The physical simulations comprised the confinement of (0.1 x 0.1 x 0.1) m{sup 3} rock samples in a load cell for the application of an in situ stress field. Different flow rates were employed in order to investigate shear effects on the overall leakoff coefficient. Horizontal radial fractures were hydraulically induced with approximate diameters, what was accomplished by controlling the injection time. Leakoff coefficients determined by means of the pressure decline analysis were compared to coefficients obtained from static filtration tests, considering similar experimental conditions. The research results indicated that the physical simulation of hydraulic fracturing may be regarded as an useful tool for evaluating the effectiveness of fracturing fluids and that it can supply reliable estimates of fluid loss coefficients. (author)

  10. Research and Development (R&D) on Advanced Nonstructural Materials. Delivery Order 0001: Study of Hydraulic System Component Storage With Operational and Rust-Inhibited Hydraulic Fluids

    National Research Council Canada - National Science Library

    Gschwender, Lois J; Snyder Jr, Carl E; Sharma, Shashi K; Jenney, Tim; Campo, Angela

    2004-01-01

    .... Jars, containing bearings and pistons, as well as hydraulic pumps were stored for up to 3 years in a laboratory environment to determine if operational fluids would protect them from rusting during storage...

  11. Multi-parameter monitoring system for hydraulic fluids; Multi-Parameter Monitoring System fuer Hydraulische Fluessigkeiten

    Energy Technology Data Exchange (ETDEWEB)

    Paul, Sumit; Legner, Wolfgang; Hackner, Angelika; Mueller, Gerhard [EADS Innovation Works, Muenchen (Germany). Bereich Sensors, Electronics and Systems Integration; Baumbach, Volker [Airbus Operations GmbH, Bremen (Germany). Bereich Hydraulic Performance and Integrity

    2011-07-01

    A miniaturised sensor system for aviation hydraulic fluids is presented. The system consists of an optochemical sensor and a particle sensor. The optochemical sensor detects the form of the O-H absorption feature around 3500 cm{sup -1} to reveal the water and acid contamination in the fluid. The particle sensor uses a light barrier principle to derive its particle contamination number. (orig.)

  12. Influence of Concentration and Salinity on the Biodegradability of Organic Additives in Hydraulic Fracturing Fluid

    Science.gov (United States)

    Mouser, P. J.; Kekacs, D.

    2014-12-01

    One of the risks associated with the use of hydraulic fracturing technologies for energy development is the potential release of hydraulic fracturing-related fluids into surface waters or shallow aquifers. Many of the organic additives used in hydraulic fracturing fluids are individually biodegradable, but little is know on how they will attenuate within a complex organic fluid in the natural environment. We developed a synthetic hydraulic fracturing fluid based on disclosed recipes used by Marcellus shale operators to evaluate the biodegradation potential of organic additives across a concentration (25 to 200 mg/L DOC) and salinity gradient (0 to 60 g/L) similar to Marcellus shale injected fluids. In aerobic aqueous solutions, microorganisms removed 91% of bulk DOC from low SFF solutions and 57% DOC in solutions having field-used SFF concentrations within 7 days. Under high SFF concentrations, salinity in excess of 20 g/L inhibited organic compound biodegradation for several weeks, after which time the majority (57% to 75%) of DOC remained in solution. After SFF amendment, the initially biodiverse lake or sludge microbial communities were quickly dominated (>79%) by Pseudomonas spp. Approximately 20% of added carbon was converted to biomass while the remainder was respired to CO2 or other metabolites. Two alcohols, isopropanol and octanol, together accounted for 2-4% of the initial DOC, with both compounds decreasing to below detection limits within 7 days. Alcohol degradation was associated with an increase in acetone at mg/L concentrations. These data help to constrain the biodegradation potential of organic additives in hydraulic fracturing fluids and guide our understanding of the microbial communities that may contribute to attenuation in surface waters.

  13. Dynamic characteristics of Semi-active Hydraulic Engine Mount Based on Fluid-Structure Interaction FEA

    Directory of Open Access Journals (Sweden)

    Tian Jiande

    2015-01-01

    Full Text Available A kind of semi-active hydraulic engine mount is studied in this paper. After careful analysis of its structure and working principle, the FEA simulation of it was divided into two cases. One is the solenoid valve is open, so the air chamber connects to the atmosphere, and Fluid-Structure Interaction was used. Another is the solenoid valve is closed, and the air chamber has pressure, so Fluid-Structure-Gas Interaction was used. The test of this semi-active hydraulic engine mount was carried out to compare with the simulation results, and verify the accuracy of the model. Then the dynamic characteristics-dynamic stiffness and damping angle were analysed by simulation and test. This paper provides theoretical support for the development and optimization of the semi-active hydraulic engine mount.

  14. Effect of rock rheology on fluid leak- off during hydraulic fracturing

    Science.gov (United States)

    Yarushina, V. M.; Bercovici, D.; Oristaglio, M. L.

    2012-04-01

    In this communication, we evaluate the effect of rock rheology on fluid leak­off during hydraulic fracturing of reservoirs. Fluid leak-off in hydraulic fracturing is often nonlinear. The simple linear model developed by Carter (1957) for flow of fracturing fluid into a reservoir has three different regions in the fractured zone: a filter cake on the fracture face, formed by solid additives from the fracturing fluid; a filtrate zone affected by invasion of the fracturing fluid; and a reservoir zone with the original formation fluid. The width of each zone, as well as its permeability and pressure drop, is assumed to remain constant. Physical intuition suggests some straightforward corrections to this classical theory to take into account the pressure dependence of permeability, the compressibility or non-Newtonian rheology of fracturing fluid, and the radial (versus linear) geometry of fluid leak­off from the borehole. All of these refinements, however, still assume that the reservoir rock adjacent to the fracture face is non­deformable. Although the effect of poroelastic stress changes on leak-off is usually thought to be negligible, at the very high fluid pressures used in hydraulic fracturing, where the stresses exceed the rock strength, elastic rheology may not be the best choice. For example, calculations show that perfectly elastic rock formations do not undergo the degree of compaction typically seen in sedimentary basins. Therefore, pseudo-elastic or elastoplastic models are used to fit observed porosity profiles with depth. Starting from balance equations for mass and momentum for fluid and rock, we derive a hydraulic flow equation coupled with a porosity equation describing rock compaction. The result resembles a pressure diffusion equation with the total compressibility being a sum of fluid, rock and pore-space compressibilities. With linear elastic rheology, the bulk formation compressibility is dominated by fluid compressibility. But the possibility

  15. Hydraulic Fracturing and Production Optimization in Eagle Ford Shale Using Coupled Geomechanics and Fluid Flow Model

    Science.gov (United States)

    Suppachoknirun, Theerapat; Tutuncu, Azra N.

    2017-12-01

    With increasing production from shale gas and tight oil reservoirs, horizontal drilling and multistage hydraulic fracturing processes have become a routine procedure in unconventional field development efforts. Natural fractures play a critical role in hydraulic fracture growth, subsequently affecting stimulated reservoir volume and the production efficiency. Moreover, the existing fractures can also contribute to the pressure-dependent fluid leak-off during the operations. Hence, a reliable identification of the discrete fracture network covering the zone of interest prior to the hydraulic fracturing design needs to be incorporated into the hydraulic fracturing and reservoir simulations for realistic representation of the in situ reservoir conditions. In this research study, an integrated 3-D fracture and fluid flow model have been developed using a new approach to simulate the fluid flow and deliver reliable production forecasting in naturally fractured and hydraulically stimulated tight reservoirs. The model was created with three key modules. A complex 3-D discrete fracture network model introduces realistic natural fracture geometry with the associated fractured reservoir characteristics. A hydraulic fracturing model is created utilizing the discrete fracture network for simulation of the hydraulic fracture and flow in the complex discrete fracture network. Finally, a reservoir model with the production grid system is used allowing the user to efficiently perform the fluid flow simulation in tight formations with complex fracture networks. The complex discrete natural fracture model, the integrated discrete fracture model for the hydraulic fracturing, the fluid flow model, and the input dataset have been validated against microseismic fracture mapping and commingled production data obtained from a well pad with three horizontal production wells located in the Eagle Ford oil window in south Texas. Two other fracturing geometries were also evaluated to optimize

  16. Replacement of petroleum based hydraulic fluids with renewable and environmental friendly resource

    International Nuclear Information System (INIS)

    Wan Sani Wan Nik; Noraini Ali

    2000-01-01

    Rational self-interest and good environmental citizenship are forcing the development of renewable and environmentally acceptable hydraulic fluids. Fluids that are at least equivalent in performance plus biodegradable have been formulated in Europe and USA using vegetable oils as base stocks for innovative additive packages. While many of the differences in using vegetable based stocks in place of mineral oils have been adapted to by straightforward formulating changes, the oxidation stability of vegetable-based stock is still a challenging area. This work initiates the investigation in Malaysia in the use of environmentally friendly resource to replace partially the petroleum based hydraulic fluid. The study concentrates more in improving the oxidation stability of the vegetable based stocks. (Author)

  17. Computational Fluid Dynamics Modelling of Hydraulics and Sedimentation in Process Reactors During Aeration Tank Settling

    DEFF Research Database (Denmark)

    Dam Jensen, Mette; Ingildsen, Pernille; Rasmussen, Michael R.

    2005-01-01

    Aeration Tank Settling is a control method alowing settling in the process tank during high hydraulic load. The control method is patented. Aeration Tank Settling has been applied in several waste water treatment plant's using present design of the process tanks. Some process tank designs have...... shown to be more effective than others. To improve the design of less effective plants Computational Fluid Dynamics (CFD) modelling of hydraulics and sedimentation has been applied. The paper discusses the results at one particular plant experiencing problems with partly short-circuiting of the inlet...

  18. Enhancing the safety and efficiency of the driving gear of coal mining machinery by using water as a hydraulic fluid and enhancing the reliability of scraper-chain conveyors; Erhoehung der Sicherheit und Leistungsfaehigkeit der Antriebstechnik von Arbeitsmaschinen durch Verwendung von Wasserhydraulik sowie Erhoehung der Zuverlaessigkeit der Kettenkratzerfoerderer

    Energy Technology Data Exchange (ETDEWEB)

    Reichel, J.; Boeing, R.; Graetz, A.; Loehning, H.D.; Plum, D.

    1997-12-31

    The objective pursued is to increasingly use water or high water-content fluids as a substitute for other hydraulic fluids in driving gear of mining machinery. The state of the art of the technology is represented only by individual solutions achieved for given purposes which are not suitable for other applications, let alone for coal mining machinery. The research project was to identify hydraulic components that will permit the use of water or watery substances as a hydraulic fluid in mining applications. The components have been found and further developed, and finally systems with linear and rotatory drives have been tested at various test facilities in order to derive information on the system behaviour of pressurized fluids and machinery components and their suitability for coal mining applications. (orig./CB) [Deutsch] In der untertaegigen Antriebstechnik sollen vermehrt Wasser und wasserhaltige Fluessigkeiten eingesetzt werden. Der Stand der Technik fuehrt bei der Anwendung von Wasserhydraulik immer wieder nur Einzelloesungen auf, die nicht allgemein und insbesondere im Steinkohlenbergbau angewendet werden koennen. Im Rahmen dieses Forschungsvorhabens wurden fuer die Wasserhydraulik geeignete Komponenten untersucht, weiterentwickelt und schliesslich Systeme mit linearen und rotatorischen Antrieben auf verschiedenen Pruefstaenden erprobt, um Aussagen ueber das Systemverhalten von Druckfluessigkeit und Bauelementen fuer Bergbauanwendungen zu bekommen. (orig./MSK)

  19. Development of thermal-hydraulic models for the safety evaluation of CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Young; Jung, Yun Sik; Hwang, Gi Suk; Kim, Nam Seok [Handong Univ., Pohang (Korea, Republic of); No, Hee Cheon [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2004-02-15

    The objective of the present research is to evaluate the safety analysis for CANDU and to improve the Horizontal Stratification Entrainment Model (HSEM) of RELAP5/MOD3.3. This report includes two items the one is the experimental study of entrainment at horizontal pipe with {+-} 36 .deg. C , {+-} 72 .deg. C branch pies, the other is the model improvement of the moderator heat sink in the Calandria. The off-take experiments on onset of entrainment and branch quality were investigated by using water and air as working fluid, and the experimental data were compared by the previous correlations. The previous correlations could not expect experimental results, thus the weak points of the previous correlations were investigated. The improvement of the previous model continues as the next year research. The thermal hydraulic scaling analysis of SPEL, STERN and ideal linear scaling analysis have been studied. As a result, a new scaling method were needed to design a new experimental facility (HGU). A new scaling method with 1/8 length scale was applied. From these results, the thermal hydraulic model for CFD code simulation was designed and test apparatus has been made. The moderator temperature distribution experiments and CFD code simulation will be continued in next year.

  20. Analysis of molten salt thermal-hydraulics using computational fluid dynamics

    International Nuclear Information System (INIS)

    Yamaji, B.; Csom, G.; Aszodi, A.

    2003-01-01

    To give a good solution for the problem of high level radioactive waste partitioning and transmutation is expected to be a pro missing option. Application of this technology also could extend the possibilities of nuclear energy. Large number of liquid-fuelled reactor concepts or accelerator driven subcritical systems was proposed as transmutors. Several of these consider fluoride based molten salts as the liquid fuel and coolant medium. The thermal-hydraulic behaviour of these systems is expected to be fundamentally different than the behaviour of widely used water-cooled reactors with solid fuel. Considering large flow domains three-dimensional thermal-hydraulic analysis is the method seeming to be applicable. Since the fuel is the coolant medium as well, one can expect a strong coupling between neutronics and thermal-hydraulics too. In the present paper the application of Computational Fluid Dynamics for three-dimensional thermal-hydraulics simulations of molten salt reactor concepts is introduced. In our past and recent works several calculations were carried out to investigate the capabilities of Computational Fluid Dynamics through the analysis of different molten salt reactor concepts. Homogenous single region molten salt reactor concept is studied and optimised. Another single region reactor concept is introduced also. This concept has internal heat exchanges in the flow domain and the molten salt is circulated by natural convection. The analysis of the MSRE experiment is also a part of our work since it may form a good background from the validation point of view. In the paper the results of the Computational Fluid Dynamics calculations with these concepts are presented. In the further work our objective is to investigate the thermal-hydraulics of the multi-region molten salt reactor (Authors)

  1. Hydraulic safety margins and embolism reversal in stems and leaves: Why are conifers and angiosperms so different?

    Science.gov (United States)

    Daniel M. Johnson; Katherine A. McCulloh; David R. Woodruff; Frederick C. Meinzer

    2012-01-01

    Angiosperm and coniferous tree species utilize a continuum of hydraulic strategies. Hydraulic safety margins (defined as differences between naturally occurring xylem pressures and pressures that would cause hydraulic dysfunction, or differences between pressures resulting in loss of hydraulic function in adjacent organs (e.g., stems vs. leaves) tend to be much greater...

  2. Extensive use of computational fluid dynamics in the upgrading of hydraulic turbines

    Energy Technology Data Exchange (ETDEWEB)

    Sabourin, M.; Eremeef, R.; De Henau, V.

    1995-12-31

    Computational fluid dynamics codes, based on turbulent Navier-Stokes equations, allow evaluation of the hydraulic losses of each turbine component with precision. Using those codes with the new generation of computers enables a wide variety of component geometries to be modelled and compared to the original designs under flow conditions obtained from testing, at a reasonable cost and in a relatively short time. This paper reviews the actual method used in the design of a solution to a turbine rehabilitation project involving runner replacement, redesign of upstream components (stay vanes and wicket gates), and downstream components (draft tubes and runner outlets). The paper shows how computational fluid dynamics can help hydraulic engineers to obtain valuable information not only on performance enhancement but also on the phenomena that produce the enhancement, and to reduce the variety of modifications to be tested.

  3. Modeling Studies to Constrain Fluid and Gas Migration Associated with Hydraulic Fracturing Operations

    Science.gov (United States)

    Rajaram, H.; Birdsell, D.; Lackey, G.; Karra, S.; Viswanathan, H. S.; Dempsey, D.

    2015-12-01

    The dramatic increase in the extraction of unconventional oil and gas resources using horizontal wells and hydraulic fracturing (fracking) technologies has raised concerns about potential environmental impacts. Large volumes of hydraulic fracturing fluids are injected during fracking. Incidents of stray gas occurrence in shallow aquifers overlying shale gas reservoirs have been reported; whether these are in any way related to fracking continues to be debated. Computational models serve as useful tools for evaluating potential environmental impacts. We present modeling studies of hydraulic fracturing fluid and gas migration during the various stages of well operation, production, and subsequent plugging. The fluid migration models account for overpressure in the gas reservoir, density contrast between injected fluids and brine, imbibition into partially saturated shale, and well operations. Our results highlight the importance of representing the different stages of well operation consistently. Most importantly, well suction and imbibition both play a significant role in limiting upward migration of injected fluids, even in the presence of permeable connecting pathways. In an overall assessment, our fluid migration simulations suggest very low risk to groundwater aquifers when the vertical separation from a shale gas reservoir is of the order of 1000' or more. Multi-phase models of gas migration were developed to couple flow and transport in compromised wellbores and subsurface formations. These models are useful for evaluating both short-term and long-term scenarios of stray methane release. We present simulation results to evaluate mechanisms controlling stray gas migration, and explore relationships between bradenhead pressures and the likelihood of methane release and transport.

  4. Potential Impacts of Spilled Hydraulic Fracturing Fluid Chemicals on Water Resources: Types, volumes, and physical-chemical properties of chemicals

    Science.gov (United States)

    Hydraulic fracturing (HF) fluid chemicals spilled on-site may impact drinking water resources. While chemicals generally make up <2% of the total injected fluid composition by mass, spills may have undiluted concentrations. HF fluids typically consist of a mixture of base flui...

  5. Vegetable oils as hydraulic fluids for agricultural applications

    Directory of Open Access Journals (Sweden)

    Mendoza, G.

    2011-03-01

    Full Text Available The formulation of environmentally friendly lubricants following the criterion of the European EcoLabel is expensive owing to the lack of technological development in this area. The present work deals with the development of lubricant formulations from vegetable oils, in particular using high oleic sunflower oil as base fluid. These new biolubricants have to perform as good as the reference lubricants used in the real application (an agricultural tractor but with the additional condition and value of their biodegradability without toxicity. Formulation development has been performed by Verkol Lubricantes, involving the selection of the base oil and the design of the additive package. The investigation performed by Tekniker in the laboratory has covered different aspects, characterizing the most important physicochemical properties of the lubricants, including their behavior at low temperatures and their resistance to oxidation. The tribological properties of the new biolubricants have also been studied, analyzing their ability to protect the interacting surface from wear, as well as the level of friction generated during sliding. Moreover, the compatibility of the new formulated oil with all the seals present in the real application has been taken into consideration. The selected lubricant is now being tested in agricultural machinery from AGRIA.

    La formulación de lubricantes amigables con el medioambiente siguiendo los criterios Europeos de la EcoLabel resulta cara debido a la falta de desarrollo tecnológico en esta área. En el presente trabajo se han desarrollado formulaciones de lubricantes a partir de aceites de origen vegetal, en particular empleando como aceite base el GAO (Girasol de Alto Oleico. Estos nuevos lubricantes deben presentar un comportamiento tan bueno como el de los lubricantes de referencia empleados en la aplicación real (un tractor agrícola, pero con la condición y valor añadido de ser biodegradables y no t

  6. Vorticity and turbulence effects in fluid structure interaction an application to hydraulic structure design

    CERN Document Server

    Brocchini, M

    2006-01-01

    This book contains a collection of 11 research and review papers devoted to the topic of fluid-structure interaction.The subject matter is divided into chapters covering a wide spectrum of recognized areas of research, such as: wall bounded turbulence; quasi 2-D turbulence; canopy turbulence; large eddy simulation; lake hydrodynamics; hydraulic hysteresis; liquid impacts; flow induced vibrations; sloshing flows; transient pipe flow and air entrainment in dropshaft.The purpose of each chapter is to summarize the main results obtained by the individual research unit through a year-long activity on a specific issue of the above list. The main feature of the book is to bring state of the art research on fluid structure interaction to the attention of the broad international community.This book is primarily aimed at fluid mechanics scientists, but it will also be of value to postgraduate students and practitioners in the field of fluid structure interaction.

  7. Microbial community changes in hydraulic fracturing fluids and produced water from shale gas extraction.

    Science.gov (United States)

    Murali Mohan, Arvind; Hartsock, Angela; Bibby, Kyle J; Hammack, Richard W; Vidic, Radisav D; Gregory, Kelvin B

    2013-11-19

    Microbial communities associated with produced water from hydraulic fracturing are not well understood, and their deleterious activity can lead to significant increases in production costs and adverse environmental impacts. In this study, we compared the microbial ecology in prefracturing fluids (fracturing source water and fracturing fluid) and produced water at multiple time points from a natural gas well in southwestern Pennsylvania using 16S rRNA gene-based clone libraries, pyrosequencing, and quantitative PCR. The majority of the bacterial community in prefracturing fluids constituted aerobic species affiliated with the class Alphaproteobacteria. However, their relative abundance decreased in produced water with an increase in halotolerant, anaerobic/facultative anaerobic species affiliated with the classes Clostridia, Bacilli, Gammaproteobacteria, Epsilonproteobacteria, Bacteroidia, and Fusobacteria. Produced water collected at the last time point (day 187) consisted almost entirely of sequences similar to Clostridia and showed a decrease in bacterial abundance by 3 orders of magnitude compared to the prefracturing fluids and produced water samplesfrom earlier time points. Geochemical analysis showed that produced water contained higher concentrations of salts and total radioactivity compared to prefracturing fluids. This study provides evidence of long-term subsurface selection of the microbial community introduced through hydraulic fracturing, which may include significant implications for disinfection as well as reuse of produced water in future fracturing operations.

  8. Static Analysis of High-Performance Fixed Fluid Power Drive with a Single Positive-Displacement Hydraulic Motor

    Directory of Open Access Journals (Sweden)

    O. F. Nikitin

    2015-01-01

    Full Text Available The article deals with the static calculations in designing a high-performance fixed fluid power drive with a single positive-displacement hydraulic motor. Designing is aimed at using a drive that is under development and yet unavailable to find and record the minimum of calculations and maximum of existing hydraulic units that enable clear and unambiguous performance, taking into consideration an available assortment of hydraulic units of hydraulic drives, to have the best efficiency.The specified power (power, moment and kinematics (linear velocity or angular velocity of rotation parameters of the output element of hydraulic motor determine the main output parameters of the hydraulic drive and the useful power of the hydraulic drive under development. The value of the overall efficiency of the hydraulic drive enables us to judge the efficiency of high-performance fixed fluid power drive.The energy analysis of a diagram of the high-performance fixed fluid power drive shows that its high efficiency is achieved when the flow rate of fluid flowing into each cylinder and the magnitude of the feed pump unit (pump are as nearly as possible.The paper considers the ways of determining the geometric parameters of working hydromotors (effective working area or working volume, which allow a selection of the pumping unit parameters. It discusses the ways to improve hydraulic drive efficiency. Using the principle of holding constant conductivity allows us to specify the values of the pressure losses in the hydraulic units used in noncatalog modes. In case of no exact matching between the parameters of existing hydraulic power modes and a proposed characteristics of the pump unit, the nearest to the expected characteristics is taken as a working version.All of the steps allow us to create the high-performance fixed fluid power drive capable of operating at the required power and kinematic parameters with high efficiency.

  9. Purification of Contaminated MIL-PRF-83282 Hydraulic Fluid Using the Pall Purifier and Multiple Process Configurations (Preprint)

    National Research Council Canada - National Science Library

    Snyder, Jr., Carl E; Gschwender, Lois J; Gunderson, Stephen L; Fultz, George W

    2006-01-01

    .... This report describes a project that evaluated the effectiveness of various hydraulic fluid purification process configurations on the removal of water and particulate contaminants from MIL-PRF-83282...

  10. TITAN: an advanced three-dimensional neutronics/thermal-hydraulics code for light water reactor safety analysis

    International Nuclear Information System (INIS)

    Griggs, D.P.; Kazimi, M.S.; Henry, A.F.

    1982-01-01

    The initial development of TITAN, a three-dimensional coupled neutronics/thermal-hydraulics code for LWR safety analysis, has been completed. The transient neutronics code QUANDRY has been joined to the two-fluid thermal-hydraulics code THERMIT with the appropriate feedback mechanisms modeled. A detailed steady-state and transient coupling scheme based on the tandem technique was implemented in accordance with the important structural and operational characteristics of QUANDRY and THERMIT. A two channel sample problem formed the basis for steady-state and transient analyses performed with TITAN. TITAN steady-state results were compared with those obtained with MEKIN and showed good agreement. Null transients, simulated turbine trip transients, and a rod withdrawal transient were analyzed with TITAN and reasonable results were obtained

  11. New tracers identify hydraulic fracturing fluids and accidental releases from oil and gas operations.

    Science.gov (United States)

    Warner, N R; Darrah, T H; Jackson, R B; Millot, R; Kloppmann, W; Vengosh, A

    2014-11-04

    Identifying the geochemical fingerprints of fluids that return to the surface after high volume hydraulic fracturing of unconventional oil and gas reservoirs has important applications for assessing hydrocarbon resource recovery, environmental impacts, and wastewater treatment and disposal. Here, we report for the first time, novel diagnostic elemental and isotopic signatures (B/Cl, Li/Cl, δ11B, and δ7Li) useful for characterizing hydraulic fracturing flowback fluids (HFFF) and distinguishing sources of HFFF in the environment. Data from 39 HFFFs and produced water samples show that B/Cl (>0.001), Li/Cl (>0.002), δ11B (25-31‰) and δ7Li (6-10‰) compositions of HFFF from the Marcellus and Fayetteville black shale formations were distinct in most cases from produced waters sampled from conventional oil and gas wells. We posit that boron isotope geochemistry can be used to quantify small fractions (∼0.1%) of HFFF in contaminated fresh water and likely be applied universally to trace HFFF in other basins. The novel environmental application of this diagnostic isotopic tool is validated by examining the composition of effluent discharge from an oil and gas brine treatment facility in Pennsylvania and an accidental spill site in West Virginia. We hypothesize that the boron and lithium are mobilized from exchangeable sites on clay minerals in the shale formations during the hydraulic fracturing process, resulting in the relative enrichment of boron and lithium in HFFF.

  12. Influences of Hydraulic Fracturing on Fluid Flow and Mineralization at the Vein-Type Tungsten Deposits in Southern China

    Directory of Open Access Journals (Sweden)

    Xiangchong Liu

    2017-01-01

    Full Text Available Wolframite is the main ore mineral at the vein-type tungsten deposits in the Nanling Range, which is a world-class tungsten province. It is disputed how wolframite is precipitated at these deposits and no one has yet studied the links of the mechanical processes to fluid flow and mineralization. Finite element-based numerical experiments are used to investigate the influences of a hydraulic fracturing process on fluid flow and solubility of CO2 and quartz. The fluids are aqueous NaCl solutions and fluid pressure is the only variable controlling solubility of CO2 and quartz in the numerical experiments. Significant fluctuations of fluid pressure and high-velocity hydrothermal pulse are found once rock is fractured by high-pressure fluids. The fluid pressure drop induced by hydraulic fracturing could cause a 9% decrease of quartz solubility. This amount of quartz deposition may not cause a significant decrease in rock permeability. The fluid pressure decrease after hydraulic fracturing also reduces solubility of CO2 by 36% and increases pH. Because an increase in pH would cause a major decrease in solubility of tungsten, the fluid pressure drop accompanying a hydraulic fracturing process facilitates wolframite precipitation. Our numerical experiments provide insight into the mechanisms precipitating wolframite at the tungsten deposits in the Nanling Range as well as other metals whose solubility is strongly dependent on pH.

  13. Characterization of the chemicals used in hydraulic fracturing fluids for wells located in the Marcellus Shale Play.

    Science.gov (United States)

    Chen, Huan; Carter, Kimberly E

    2017-09-15

    Hydraulic fracturing, coupled with the advances in horizontal drilling, has been used for recovering oil and natural gas from shale formations and has aided in increasing the production of these energy resources. The large volumes of hydraulic fracturing fluids used in this technology contain chemical additives, which may be toxic organics or produce toxic degradation byproducts. This paper investigated the chemicals introduced into the hydraulic fracturing fluids for completed wells located in Pennsylvania and West Virginia from data provided by the well operators. The results showed a total of 5071 wells, with average water volumes of 5,383,743 ± 2,789,077 gal (mean ± standard deviation). A total of 517 chemicals was introduced into the formulated hydraulic fracturing fluids. Of the 517 chemicals listed by the operators, 96 were inorganic compounds, 358 chemicals were organic species, and the remaining 63 cannot be identified. Many toxic organics were used in the hydraulic fracturing fluids. Some of them are carcinogenic, including formaldehyde, naphthalene, and acrylamide. The degradation of alkylphenol ethoxylates would produce more toxic, persistent, and estrogenic intermediates. Acrylamide monomer as a primary degradation intermediate of polyacrylamides is carcinogenic. Most of the chemicals appearing in the hydraulic fracturing fluids can be removed when adopting the appropriate treatments. Copyright © 2017 Elsevier Ltd. All rights reserved.

  14. Thermal-hydraulic methods in fast reactor safety

    International Nuclear Information System (INIS)

    Weber, D.P.; Briggs, L.L.

    1985-01-01

    Methods for the solution of thermal-hydraulic problems in liquid metal fast breeder reactors (LMFBRs) arising primarily from transient accident analysis are reviewed. Principal emphasis is given to the important phenomenological issues of sodium boiling and fuel motion. Descriptions of representative phenomenological and mathematical models, computational algorithms, advantages and limitations of the approaches, and current research needs and directions are provided

  15. Divergent Hydraulic Safety Strategies in Three Co-occurring Anacardiaceae Tree Species in a Chinese Savanna.

    Science.gov (United States)

    Zhang, Shu-Bin; Zhang, Jiao-Lin; Cao, Kun-Fang

    2016-01-01

    Vulnerability segmentation, the condition under which plant leaves are more vulnerable to drought-induced cavitation than stems, may act as a "safety valve" to protect stems from hydraulic failure. Evergreen, winter-deciduous, and drought-deciduous tree species co-occur in tropical savannas, but there have been no direct studies on the role of vulnerability segmentation and stomatal regulation in maintaining hydraulic safety in trees with these three leaf phenologies. To this end, we selected three Anacardiaceae tree species co-occurring in a Chinese savanna, evergreen Pistacia weinmanniifolia , drought-deciduous Terminthia paniculata , and winter-deciduous Lannea coromandelica , to study inter-species differentiation in leaf and stem hydraulic safety. We found that the two deciduous species had significantly higher sapwood-specific hydraulic conductivity and leaf-specific hydraulic conductance than the evergreen species. Moreover, two deciduous species were more vulnerable to stem cavitation than the evergreen species, although both drought-deciduous species and evergreen species had drought-resistance leaves. The evergreen species maintained a wide hydraulic safety margin (HSM) in stems and leaves; which was achieved by embolism resistance of both stems and leaves and isohydric stomatal control. Both deciduous species had limited HSMs in stems and leaves, being isohydric in the winter-deciduous species and anisohydric in drought-deciduous species. The difference in water potential at 50% loss of hydraulic conductivity between the leaves and the terminal stems (P50 leaf-stem ) was positive in P. weinmanniifolia and L. coromandelica , whereas, T. paniculata exhibited a lack of vulnerability segmentation. In addition, differences in hydraulic architecture were found to be closely related to other structural traits, i.e., leaf mass per area, wood density, and sapwood anatomy. Overall, the winter-deciduous species exhibits a drought-avoidance strategy that maintains

  16. Polyalkylene glycols, base fluids for special lubricants and hydraulic fluids; Polyalkylenglykole, Basisoele fuer Spezialschmierstoffe und Hydraulikfluessigkeiten

    Energy Technology Data Exchange (ETDEWEB)

    Poellmann, K. [Clariant GmbH (Germany)

    2004-08-01

    For many years polyalkylene glycols have been used as base fluids for special lubricants. In this matter they compete with polyol esters and polyalphaolefines. Synthesis of polyalkylen glycols is founded upon the anionic polymerisation of ethyleneoxid, propyleneoxid and if necessary of other oxigen-containing monomeres. The flexibility of this synthesis is the reason that polyalkylene glycole is a collective term, including a broad group of base fluids with partly extreme different properties. Typical for polyalkylene glycols is a high viscosity-index, watersolubility and adsorbing power for water, low friction numbers, but also the incompatibility with current mineral-oil-soluble additive systems. Because of this quality profile there has been developped specific niche-applications in the lubricant-area for polyalkylene glycols in the last 30 years, where each of the specific benefits has been used. Among them are watercontaining HFC hydraulicfluids, refrigerator oils, and oils for ethylene-compressors. HFC fluids are formulated with high-viscous, water-soluble polyalkylene glycols. For refrigerator oils in motor-car conditioning the R 134A compatibility of water-insoluble polyalkylene glycols is essential. For the use in ethylene-compressors the crucial point is the insolubility of polyalkylene glycol in ethylene. (orig.)

  17. Hydraulically driven control rod concept for integral reactors: fluid dynamic simulation and preliminary test

    International Nuclear Information System (INIS)

    Ricotti, M.E.; Cammi, A.; Lombardi, C.; Passoni, M.; Rizzo, C.; Carelli, M.; Colombo, E.

    2003-01-01

    The paper deals with the preliminary study of the Hydraulically Driven Control Rod concept, tailored for PWR control rods (spider type) with hydraulic drive mechanism completely immersed in the primary water. A specific solution suitable for advanced versions of the IRIS integral reactor is under investigation. The configuration of the Hydraulic Control Rod device, made up by an external movable piston and an internal fixed cylinder, is described. After a brief description of the whole control system, particular attention is devoted to the Control Rod characterization via Computational Fluid Dynamics (CFD) analysis. The investigation of the system behavior, including dynamic equilibrium and stability properties, has been carried out. Finally, preliminary tests were performed in a low pressure, low temperature, reduced length experimental facility. The results are compared with the dynamic control model and CFD simulation model, showing good agreement between simulations and experimental data. During these preliminary tests, the control system performs correctly, allowing stable dynamic equilibrium positions for the Control Rod and stable behavior during withdrawal and insertion steps. (author)

  18. Hydraulic efficiency and safety of vascular and non-vascular components in Pinus pinaster leaves.

    Science.gov (United States)

    Charra-Vaskou, Katline; Badel, Eric; Burlett, Régis; Cochard, Hervé; Delzon, Sylvain; Mayr, Stefan

    2012-09-01

    Leaves, the distal section of the soil-plant-atmosphere continuum, exhibit the lowest water potentials in a plant. In contrast to angiosperm leaves, knowledge of the hydraulic architecture of conifer needles is scant. We investigated the hydraulic efficiency and safety of Pinus pinaster needles, comparing different techniques. The xylem hydraulic conductivity (k(s)) and embolism vulnerability (P(50)) of both needle and stem were measured using the cavitron technique. The conductance and vulnerability of whole needles were measured via rehydration kinetics, and Cryo-SEM and 3D X-ray microtomographic observations were used as reference tools to validate physical measurements. The needle xylem of P. pinaster had lower hydraulic efficiency (k(s) = 2.0 × 10(-4) m(2) MPa(-1) s(-1)) and safety (P(50) = - 1.5 MPa) than stem xylem (k(s) = 7.7 × 10(-4) m(2) MPa(-1) s(-1); P(50) = - 3.6 to - 3.2 MPa). P(50) of whole needles (both extra-vascular and vascular pathways) was - 0.5 MPa, suggesting that non-vascular tissues were more vulnerable than the xylem. During dehydration to - 3.5 MPa, collapse and embolism in xylem tracheids, and gap formation in surrounding tissues were observed. However, a discrepancy in hydraulic and acoustic results appeared compared with visualizations, arguing for greater caution with these techniques when applied to needles. Our results indicate that the most distal parts of the water transport pathway are limiting for hydraulics of P. pinaster. Needle tissues exhibit a low hydraulic efficiency and low hydraulic safety, but may also act to buffer short-term water deficits, thus preventing xylem embolism.

  19. Fluid dynamics of acoustic and hydrodynamic cavitation in hydraulic power systems

    Science.gov (United States)

    Ferrari, A.

    2017-03-01

    Cavitation is the transition from a liquid to a vapour phase, due to a drop in pressure to the level of the vapour tension of the fluid. Two kinds of cavitation have been reviewed here: acoustic cavitation and hydrodynamic cavitation. As acoustic cavitation in engineering systems is related to the propagation of waves through a region subjected to liquid vaporization, the available expressions of the sound speed are discussed. One of the main effects of hydrodynamic cavitation in the nozzles and orifices of hydraulic power systems is a reduction in flow permeability. Different discharge coefficient formulae are analysed in this paper: the Reynolds number and the cavitation number result to be the key fluid dynamical parameters for liquid and cavitating flows, respectively. The latest advances in the characterization of different cavitation regimes in a nozzle, as the cavitation number reduces, are presented. The physical cause of choked flows is explained, and an analogy between cavitation and supersonic aerodynamic flows is proposed. The main approaches to cavitation modelling in hydraulic power systems are also reviewed: these are divided into homogeneous-mixture and two-phase models. The homogeneous-mixture models are further subdivided into barotropic and baroclinic models. The advantages and disadvantages of an implementation of the complete Rayleigh-Plesset equation are examined.

  20. Computational fluid dynamics modelling of hydraulics and sedimentation in process reactors during aeration tank settling.

    Science.gov (United States)

    Jensen, M D; Ingildsen, P; Rasmussen, M R; Laursen, J

    2006-01-01

    Aeration tank settling is a control method allowing settling in the process tank during high hydraulic load. The control method is patented. Aeration tank settling has been applied in several waste water treatment plants using the present design of the process tanks. Some process tank designs have shown to be more effective than others. To improve the design of less effective plants, computational fluid dynamics (CFD) modelling of hydraulics and sedimentation has been applied. This paper discusses the results at one particular plant experiencing problems with partly short-circuiting of the inlet and outlet causing a disruption of the sludge blanket at the outlet and thereby reducing the retention of sludge in the process tank. The model has allowed us to establish a clear picture of the problems arising at the plant during aeration tank settling. Secondly, several process tank design changes have been suggested and tested by means of computational fluid dynamics modelling. The most promising design changes have been found and reported.

  1. The safety feature of hydraulic driving system of control rod for 200 MW nuclear heating reactor

    International Nuclear Information System (INIS)

    Chi Zongbo; Wu Yuanqiang

    1997-01-01

    The hydraulic driving system of control rod is used as control rod drive mechanism in 200 MW nuclear heating reactor. Design of this system is based on passive system, integrating drive and guide of control rod. The author analyzes the inherent safety and the design safety of this system, with mechanism of control rod not ejecting when the pressure of pressure vessel is lost, and calculating result of core not exposing when the amount of coolant is drained by broken pipe. The results indicate that this system has good safety feature, and assures reactor safety under any accident conditions, providing important technology support for 200 MW nuclear heating reactor with inherent safety feature

  2. Effects of hydraulic frac fluids and formation waters on groundwater microbial communities

    Science.gov (United States)

    Krueger, Martin; Jimenez, Nuria

    2017-04-01

    Shale gas is being considered as a complementary energy resource to other fossil fuels. Its exploitation requires using advanced drilling techniques and hydraulic stimulation (fracking). During fracking operations, large amounts of fluids (fresh water, proppants and chemicals) are injected at high pressures into the formations, to create fractures and fissures, and thus to release gas from the source rock into the wellbore. The injected fluid partly remains in the formation, while up to 40% flows back to the surface, together with reservoir waters, sometimes containing dissolved hydrocarbons, high salt concentrations, etc. The aim of our study was to investigate the potential impacts of frac or geogenic chemicals, frac fluid, formation water or flowback on groudnwater microbial communities. Laboratory experiments under in situ conditions (i.e. at in situ temperature, high pressure) were conducted using groundwater samples from three different locations. Series of microcosms containing R2 broth medium or groundwater spiked with either single frac chemicals (including biocides), frac fluids, artificial reservoir water, NaCl, or different mixtures of reservoir water and frac fluid (to simulate flowback) were incubated in the dark. Controls included non-amended and non-inoculated microcosms. Classical microbiological methods and molecular analyses were used to assess changes in the microbial abundance, community structure and function in response to the different treatments. Microbial communities were quite halotolerant and their growth benefited from low concentrations of reservoir waters or salt, but they were negatively affected by higher concentrations of formation waters, salt, biocides or frac fluids. Changes on the microbial community structure could be detected by T-RFLP. Single frac components like guar gum or choline chloride were used as substrates, while others like triethanolamine or light oil distillate hydrogenated prevented microbial growth in

  3. Stimuli Responsive/Rheoreversible Hydraulic Fracturing Fluids for Enhanced Geothermal Energy Production (Part I)

    Science.gov (United States)

    Fernandez, C. A.; Jung, H. B.; Shao, H.; Bonneville, A.; Heldebrant, D.; Hoyt, D.; Zhong, L.; Holladay, J.

    2014-12-01

    Cost-effective yet safe creation of high-permeability reservoirs inside deep crystalline bedrock is the primary challenge for the viability of enhanced geothermal systems and unconventional oil/gas recovery. Current reservoir stimulation processes utilize brute force (hydraulic pressures in the order of hundreds of bar) to create/propagate fractures in the bedrock. Such stimulation processes entail substantial economic costs ($3.3 million per reservoir as of 2011). Furthermore, the environmental impacts of reservoir stimulation are only recently being determined. Widespread concerns about the environmental contamination have resulted in a number of regulations for fracturing fluids advocating for greener fracturing processes. To reduce the costs and environmental impact of reservoir stimulation, we developed an environmentally friendly and recyclable hydraulic fracturing fluid that undergoes a controlled and large volume expansion with a simultaneous increase in viscosity triggered by CO2 at temperatures relevant for reservoir stimulation in Enhanced Geothermal System (EGS). The volume expansion, which will specifically occurs at EGS depths of interest, generates an exceptionally large mechanical stress in fracture networks of highly impermeable rock propagating fractures at effective stress an order of magnitude lower than current technology. This paper will concentrate on the presentation of this CO2-triggered expanding hydrogel formed from diluted aqueous solutions of polyallylamine (PAA). Aqueous PAA-CO2 mixtures also show significantly higher viscosities than conventional rheology modifiers at similar pressures and temperatures due to the cross-linking reaction of PAA with CO2, which was demonstrated by chemical speciation studies using in situ HP-HT 13C MAS-NMR. In addtion, PAA shows shear-thinning behavior, a critical advantage for the use of this fluid system in EGS reservoir stimulation. The high pressure/temperature experiments and their results as well

  4. Can introduction of hydraulic fracturing fluids induce biogenic methanogenesis in the shale reservoirs?

    Science.gov (United States)

    Sharma, S.; Wilson, T.; Wrighton, K. C.; Borton, M.; O'Banion, B.

    2017-12-01

    The hydraulic fracturing fluids (HFF) injected into the shale formation are composed primarily of water, proppant and some chemical additives ( 0.5- 2% by volume). The additives contain a lot of organic and inorganic compounds like ammonium sulfate, guar gum, boric acid, hydrochloric acid, citric acid, potassium carbonate, glutaraldehyde, ethylene glycols which serve as friction reducers, gelling agents, crosslinkers, biocides, corrosion/scale inhibitors, etc. The water and additives introduced into the formation ensue a variety of microbiogechmical reactions in the reservoir. For this study produced, water and gas samples were collected from several old and new Marcellus wells in SE Pennsylvania and NE West Virginia to better understand these microbe-water-rock interactions. The carbon isotopic composition of dissolved inorganic carbon (δ13CDIC) in the produced fluids and CO2 in produced gas (δ13CCO2) are highly enriched with values > +10‰ and +14 ‰ V-PDB respectively. The injected hydraulic fracturing fluid had low δ13CDIC values of detectable carbon in them. The drilling mud and carbonate veins had δ13C values of -1.8 and < 2.0 ‰ V-PDB respectively. Therefore, the high δ13CDIC signatures in produced water are possibly due to the microbial utilization of lighter carbon (12C) by microbes or methanogenic bacteria in the reservoir. It is possible that introduction of C containing nutrients like guar, methanol, methylamines, etc. stimulates certain methanogen species in the reservoir to produce biogenic methane. Genomic analysis reveals that methanogen species like Methanohalophilus or Methanolobus could be the possible sources of biogenic methane in these shale reservoirs. The evidence of microbial methanogenesis raises the possibility of enhanced gas recovery from these shales using biological amendments.

  5. Development of thermal-hydraulic models for the safety evaluation of CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Young; Hwang, Gi Suk; Jung, Yun Sik [Handong Univ., Pohang (Korea, Republic of); No, Hee Cheon; Moon, Young Min [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2003-03-15

    The objective of the present research is to evaluate the safety analysis for CANDU and to improve the Horizontal Stratification Entrainment Model (HSEM) of RELAP5/MOD3.3. This report includes two items : the one is the development of experimental facility for the safety evaluation of CANDU, the other is the results of comparison with the existing correlations and data. The literature reviews are performed and the database for previous off-take experiments are built. By a survey of state-of-the-articles, the deficiencies of previous works and limitations of existing models are examined. The hydraulic behavior branching through the feeder pipes from the header pipe is analyzed and the test facility of off-take experiment is designed and manufactured as the prototype CANDU6, by a proper scaling methodologies. The test facility contains various branch pipes not only for three directions (top, side and bottom), but for arbitrary directions. The experiments about the onset of entrainment and branch quality only for three directions (top, side and bottom) are carried out by using air-water as working fluids. On the whole, the existing correlations predict the present experimental results well branch quality, entrainment, which validates the availability of experimental facility and methodology. Especially, for the branch quality with top and bottom branches, the different results are shown because of the unstable flow regimes in the horizontal pipe and the different branch diameters. The deficiencies of previous works and limitations of existing models are considered. The off-take experiment for arbitrary branch angles continues as the next year research.

  6. Influence of Drought on the Hydraulic Efficiency and the Hydraulic Safety of the Xylem - Case of a Semi-arid Conifer.

    Science.gov (United States)

    Gentine, P.; Guerin, M. F.; von Arx, G.; Martin-Benito, D.; Griffin, K. L.; McDowell, N.; Pockman, W.; Andreu-Hayles, L.

    2017-12-01

    Recent droughts in the Southwest US have resulted in extensive mortality in the pinion pine population (Pinus Edulis). An important factor for resiliency is the ability of a plant to maintain a functional continuum between soil and leaves, allowing water's motion to be sustained or resumed. During droughts, loss of functional tracheids happens through embolism, which can be partially mitigated by increasing the hydraulic safety of the xylem. However, higher hydraulic safety is usually achieved by building narrower tracheids with thicker walls, resulting in a reduction of the hydraulic efficiency of the xylem (conductivity per unit area). Reduced efficiency constrains water transport, limits photosynthesis and might delay recovery after the drought. Supporting existing research on safety-efficiency tradeoff, we test the hypothesis that under dry conditions, isohydric pinions grow xylem that favor efficiency over safety. Using a seven-year experiment with three watering treatments (drought, control, irrigated) in New Mexico, we investigate the effect of drought on the xylem anatomy of pinions' branches. We also compare the treatment effect with interannual variations in xylem structure. We measure anatomical variables - conductivities, cell wall thicknesses, hydraulic diameter, cell reinforcement and density - and preliminarily conclude that treatment has little effect on hydraulic efficiency while hydraulic safety is significantly reduced under dry conditions. Taking advantage of an extremely dry year occurrence during the experiment, we find a sharp increase in vulnerability for xylem tissues built the same year.

  7. Constitutive model development needs for reactor safety thermal-hydraulic codes

    International Nuclear Information System (INIS)

    Kelly, J.M.

    1998-01-01

    This paper discusses the constitutive model development needs for our current and future generation of reactor safety thermal-hydraulic analysis codes. Rather than provide a simple 'shopping list' of models to be improved, a detailed description is given of how a constitutive model works within the computational framework of a current reactor safety code employing the two-fluid model of two-phase flow. The intent is to promote a better understanding of both the types of experiments and the instrumentation needs that will be required in the USNRCs code improvement program. First, a summary is given of the modeling considerations that need to be taken into account when developing constitutive models for use in reactor safety thermal-hydraulic codes. Specifically, the two-phase flow model should be applicable to a control volume formulation employing computational volumes with dimensions on the order of meters but containing embedded structure with a dimension on the order of a centimeter. The closure relations are then required to be suitable when averaged over such large volumes containing millions or even tens of millions of discrete fluid particles (bubbles/drops). This implies a space and time averaging procedure that neglects the intermittency observed in slug and chum turbulent two-phase flows. Furthermore, the geometries encountered in reactor systems are complex, the constitutive relations should therefore be component specific (e.g., interfacial shear in a tube does not represent that in a rod bundle nor in the downcomer). When practicable, future modeling efforts should be directed towards resolving the spatial evolution of two-phase flow patterns through the introduction of interfacial area transport equations and by modeling the individual physical processes responsible for the creation or destruction of interfacial area. Then the example of the implementation and assessment of a subcooled boiling model in a two-fluid code is given. The primary parameter

  8. Characterization of the Oriskany and Berea Sandstones: Evaluating Biogeochemical Reactions of Potential Sandstone–Hydraulic Fracturing Fluid Interaction

    Energy Technology Data Exchange (ETDEWEB)

    Verba, Circe [National Energy Technology Lab. (NETL), Albany, OR (United States); Harris, Aubrey [National Energy Technology Lab. (NETL), Albany, OR (United States)

    2016-07-07

    The Marcellus shale, located in the mid-Atlantic Appalachian Basin, has been identified as a source for natural gas and targeted for hydraulic fracturing recovery methods. Hydraulic fracturing is a technique used by the oil and gas industry to access petroleum reserves in geologic formations that cannot be accessed with conventional drilling techniques (Capo et al., 2014). This unconventional technique fractures rock formations that have low permeability by pumping pressurized hydraulic fracturing fluids into the subsurface. Although the major components of hydraulic fracturing fluid are water and sand, chemicals, such as recalcitrant biocides and polyacrylamide, are also used (Frac Focus, 2015). There is domestic concern that the chemicals could reach groundwater or surface water during transport, storage, or the fracturing process (Chapman et al., 2012). In the event of a surface spill, understanding the natural attenuation of the chemicals in hydraulic fracturing fluid, as well as the physical and chemical properties of the aquifers surrounding the spill site, will help mitigate potential dangers to drinking water. However, reports on the degradation pathways of these chemicals are limited in existing literature. The Appalachian Basin Marcellus shale and its surrounding sandstones host diverse mineralogical suites. During the hydraulic fracturing process, the hydraulic fracturing fluids come into contact with variable mineral compositions. The reactions between the fracturing fluid chemicals and the minerals are very diverse. This report: 1) describes common minerals (e.g. quartz, clay, pyrite, and carbonates) present in the Marcellus shale, as well as the Oriskany and Berea sandstones, which are located stratigraphically below and above the Marcellus shale; 2) summarizes the existing literature of the degradation pathways for common hydraulic fracturing fluid chemicals [polyacrylamide, ethylene glycol, poly(diallyldimethylammonium chloride), glutaraldehyde

  9. Proceedings of the fourth international topical meeting on nuclear thermal hydraulics, operations and safety. Vol. 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

    More than 100 papers were presented. The meeting was divided in 56 sessions and covered the following topics: Plant Operation, Retrofitting and Maintenance Experience; Steam Generator Operation and Maintenance; Artificial Intelligence and Expert Systems; Seismic Technologies for Plant Design and Operations; Aging Management and Life Extension; Two-Phase Flow Modeling and Applications; Severe Accidents and Degraded Core Thermal Hydraulics; Plant Simulators, Analyzers, and Workstations; Advanced Nuclear Fuel Challenges; Recent Nuclear Power Station Decommissioning Experiences in the USA; Application of Probabilistic risk assessment/Probabilistic safety assessment (PRA/PSA) in Design and Modification; Numerical Modeling in Thermal Hydraulics; General Thermal Hydraulics; Severe Accident Management; Licensing and Regulatory Requirements; Advanced Light Water Reactor Designs to Support Reduced Emergency Planning; Best Estimate loss-of-coolant (LOCA) Methodologies; Plant Instrumentation and Control; LWR Fuel Designs for Improved Thermal Hydraulic Performance; Performance Assessment of Radioactive Waste Disposal; Thermal Hydraulics in Passive Reactor Systems; Advances in Man-Machine Interface Design and the Related Human Factors Engineering; Advances in Measurements and Instrumentation; Computer Aided Technology for non-destructive evaluation (NDE) and Plant Maintenance Plant Uprating; Flow-Accelerated Corrosion in Nuclear Power Plants; Advances in Radiological Measurement and Analysis Risk Management and Assessment; Stability in Thermal Hydraulic Systems; Critical heat flux (CHF) and Post Dryout Heat Transfer; Plant Transient and Accident Modeling.

  10. Proceedings of the fourth international topical meeting on nuclear thermal hydraulics, operations and safety. Vol. 1

    International Nuclear Information System (INIS)

    2004-01-01

    More than 100 papers were presented. The meeting was divided in 56 sessions and covered the following topics: Plant Operation, Retrofitting and Maintenance Experience; Steam Generator Operation and Maintenance; Artificial Intelligence and Expert Systems; Seismic Technologies for Plant Design and Operations; Aging Management and Life Extension; Two-Phase Flow Modeling and Applications; Severe Accidents and Degraded Core Thermal Hydraulics; Plant Simulators, Analyzers, and Workstations; Advanced Nuclear Fuel Challenges; Recent Nuclear Power Station Decommissioning Experiences in the USA; Application of Probabilistic risk assessment/Probabilistic safety assessment (PRA/PSA) in Design and Modification; Numerical Modeling in Thermal Hydraulics; General Thermal Hydraulics; Severe Accident Management; Licensing and Regulatory Requirements; Advanced Light Water Reactor Designs to Support Reduced Emergency Planning; Best Estimate loss-of-coolant (LOCA) Methodologies; Plant Instrumentation and Control; LWR Fuel Designs for Improved Thermal Hydraulic Performance; Performance Assessment of Radioactive Waste Disposal; Thermal Hydraulics in Passive Reactor Systems; Advances in Man-Machine Interface Design and the Related Human Factors Engineering; Advances in Measurements and Instrumentation; Computer Aided Technology for non-destructive evaluation (NDE) and Plant Maintenance Plant Uprating; Flow-Accelerated Corrosion in Nuclear Power Plants; Advances in Radiological Measurement and Analysis Risk Management and Assessment; Stability in Thermal Hydraulic Systems; Critical heat flux (CHF) and Post Dryout Heat Transfer; Plant Transient and Accident Modeling

  11. Proceedings of the fourth international topical meeting on nuclear thermal hydraulics, operations and safety. Vol. 2

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

    More than 100 papers presented at the meeting were divided in 56 sessions and covered the following topics: Plant Operation, Retrofitting and Maintenance Experience; Steam Generator Operation and Maintenance; Artificial Intelligence and Expert Systems; Seismic Technologies for Plant Design and Operations; Aging Management and Life Extension; Two-Phase Flow Modeling and Applications; Severe Accidents and Degraded Core Thermal Hydraulics; Plant Simulators, Analyzers, and Workstations; Advanced Nuclear Fuel Challenges; Recent Nuclear Power Station Decommissioning Experiences in the USA; Application of Probabilistic risk assessment/Probabilistic safety assessment (PRA/PSA) in Design and Modification; Numerical Modeling in Thermal Hydraulics; General Thermal Hydraulics; Severe Accident Management; Licensing and Regulatory Requirements; Advanced Light Water Reactor Designs to Support Reduced Emergency Planning; Best Estimate loss-of-coolant (LOCA) Methodologies; Plant Instrumentation and Control; LWR Fuel Designs for Improved Thermal Hydraulic Performance; Performance Assessment of Radioactive Waste Disposal; Thermal Hydraulics in Passive Reactor Systems; Advances in Man-Machine Interface Design and the Related Human Factors Engineering; Advances in Measurements and Instrumentation; Computer Aided Technology for non-destructive evaluation (NDE) and Plant Maintenance Plant Uprating; Flow-Accelerated Corrosion in Nuclear Power Plants; Advances in Radiological Measurement and Analysis Risk Management and Assessment; Stability in Thermal Hydraulic Systems; Critical heat flux (CHF) and Post Dryout Heat Transfer; Plant Transient and Accident Modeling.

  12. Proceedings of the fourth international topical meeting on nuclear thermal hydraulics, operations and safety. Vol. 2

    International Nuclear Information System (INIS)

    2004-01-01

    More than 100 papers presented at the meeting were divided in 56 sessions and covered the following topics: Plant Operation, Retrofitting and Maintenance Experience; Steam Generator Operation and Maintenance; Artificial Intelligence and Expert Systems; Seismic Technologies for Plant Design and Operations; Aging Management and Life Extension; Two-Phase Flow Modeling and Applications; Severe Accidents and Degraded Core Thermal Hydraulics; Plant Simulators, Analyzers, and Workstations; Advanced Nuclear Fuel Challenges; Recent Nuclear Power Station Decommissioning Experiences in the USA; Application of Probabilistic risk assessment/Probabilistic safety assessment (PRA/PSA) in Design and Modification; Numerical Modeling in Thermal Hydraulics; General Thermal Hydraulics; Severe Accident Management; Licensing and Regulatory Requirements; Advanced Light Water Reactor Designs to Support Reduced Emergency Planning; Best Estimate loss-of-coolant (LOCA) Methodologies; Plant Instrumentation and Control; LWR Fuel Designs for Improved Thermal Hydraulic Performance; Performance Assessment of Radioactive Waste Disposal; Thermal Hydraulics in Passive Reactor Systems; Advances in Man-Machine Interface Design and the Related Human Factors Engineering; Advances in Measurements and Instrumentation; Computer Aided Technology for non-destructive evaluation (NDE) and Plant Maintenance Plant Uprating; Flow-Accelerated Corrosion in Nuclear Power Plants; Advances in Radiological Measurement and Analysis Risk Management and Assessment; Stability in Thermal Hydraulic Systems; Critical heat flux (CHF) and Post Dryout Heat Transfer; Plant Transient and Accident Modeling

  13. Radium and barium removal through blending hydraulic fracturing fluids with acid mine drainage.

    Science.gov (United States)

    Kondash, Andrew J; Warner, Nathaniel R; Lahav, Ori; Vengosh, Avner

    2014-01-21

    Wastewaters generated during hydraulic fracturing of the Marcellus Shale typically contain high concentrations of salts, naturally occurring radioactive material (NORM), and metals, such as barium, that pose environmental and public health risks upon inadequate treatment and disposal. In addition, fresh water scarcity in dry regions or during periods of drought could limit shale gas development. This paper explores the possibility of using alternative water sources and their impact on NORM levels through blending acid mine drainage (AMD) effluent with recycled hydraulic fracturing flowback fluids (HFFFs). We conducted a series of laboratory experiments in which the chemistry and NORM of different mix proportions of AMD and HFFF were examined after reacting for 48 h. The experimental data combined with geochemical modeling and X-ray diffraction analysis suggest that several ions, including sulfate, iron, barium, strontium, and a large portion of radium (60-100%), precipitated into newly formed solids composed mainly of Sr barite within the first ∼ 10 h of mixing. The results imply that blending AMD and HFFF could be an effective management practice for both remediation of the high NORM in the Marcellus HFFF wastewater and beneficial utilization of AMD that is currently contaminating waterways in northeastern U.S.A.

  14. Elevational trends in hydraulic efficiency and safety of Pinus cembra roots.

    Science.gov (United States)

    Losso, Adriano; Nardini, Andrea; Nolf, Markus; Mayr, Stefan

    2016-04-01

    In alpine regions, elevational gradients in environmental parameters are reflected by structural and functional changes in plant traits. Elevational changes in plant water relations have also been demonstrated, but comparable information on root hydraulics is generally lacking. We analyzed the hydraulic efficiency (specific hydraulic conductivity k s, entire root system conductance K R) and vulnerability to drought-induced embolism (water potential at 50 % loss of conductivity Ψ 50) of the roots of Pinus cembra trees growing along an elevational transect of 600 m. Hydraulic parameters of the roots were compared with those of the stem and related to anatomical traits {mean conduit diameter (d), wall reinforcement [(t/b)(2)]}. We hypothesized that temperature-related restrictions in root function would cause a progressive limitation of hydraulic efficiency and safety with increasing elevation. We found that both root k s and K R decreased from low (1600 m a.s.l.: k s 5.6 ± 0.7 kg m(-1) s(-1) MPa(-1), K R 0.049 ± 0.005 kg m(-2) s (-1) MPa(-1)) to high elevation (2100 m a.s.l.: k s 4.2 ± 0.6 kg m(-1) s(-1) MPa(-1), K R 0.035 ± 0.006 kg m(-2) s(-1) MPa(-1)), with small trees showing higher K R than large trees. k s was higher in roots than in stems (0.5 ± 0.05 kg m(-1)s(-1)MPa(-1)). Ψ 50 values were similar across elevations and overall less negative in roots (Ψ 50 -3.6 ± 0.1 MPa) than in stems (Ψ 50 -3.9 ± 0.1 MPa). In roots, large-diameter tracheids were lacking at high elevation and (t/b)(2) increased, while d did not change. The elevational decrease in root hydraulic efficiency reflects a limitation in timberline tree hydraulics. In contrast, hydraulic safety was similar across elevations, indicating that avoidance of hydraulic failure is important for timberline trees. As hydraulic patterns can only partly be explained by the anatomical parameters studied, limitations and/or adaptations at the pit level are likely.

  15. The coupled effect of fiber volume fraction and void fraction on hydraulic fluid absorption of quartz/BMI laminates

    Science.gov (United States)

    Hurdelbrink, Keith R.; Anderson, Jacob P.; Siddique, Zahed; Altan, M. Cengiz

    2016-03-01

    Bismaleimide (BMI) resin with quartz (AQ581) fiber reinforcement is a composite material frequently used in aerospace applications, such as engine cowlings and radomes. Various composite components used in aircrafts are exposed to different types of hydraulic fluids, which may lead to anomalous absorption behavior over the service life of the composite. Accurate predictive models for absorption of liquid penetrants are particularly important as the composite components are often exposed to long-term degradation due to absorbed moisture, hydraulic fluids, or similar liquid penetrants. Microstructural features such as fiber volume fraction and void fraction can have a significant effect on the absorption behavior of fiber-reinforced composites. In this paper, hydraulic fluid absorption characteristics of quartz/BMI laminates fabricated from prepregs preconditioned at different relative humidity and subsequently cured at different pressures are presented. The composite samples are immersed into hydraulic fluid at room temperature, and were not subjected to any prior degradation. To generate process-induced microvoids, prepregs were conditioned in an environmental chamber at 2% or 99% relative humidity at room temperature for a period of 24 hours prior to laminate fabrication. To alter the fiber volume fraction, the laminates were fabricated at cure pressures of 68.9 kPa (10 psi) or 482.6 kPa (70 psi) via a hot-press. The laminates are shown to have different levels of microvoids and fiber volume fractions, which were observed to affect the absorption dynamics considerably and exhibited clear non-Fickian behavior. A one-dimensional hindered diffusion model (HDM) was shown to be successful in predicting the hydraulic fluid absorption. Model prediction indicates that as the fabrication pressure increased from 68.9 kPa to 482.6 kPa, the maximum fluid content (M∞) decreased from 8.0% wt. to 1.0% wt. The degree of non-Fickian behavior, measured by hindrance coefficient (

  16. Acoustic emission in a fluid saturated heterogeneous porous layer with application to hydraulic fracture

    Energy Technology Data Exchange (ETDEWEB)

    Nelson, J.T. (California Univ., Berkeley, CA (USA). Dept. of Mechanical Engineering Lawrence Berkeley Lab., CA (USA))

    1988-11-01

    A theoretical model for acoustic emission in a vertically heterogeneous porous layer bounded by semi-infinite solid regions is developed using linearized equations of motion for a fluid/solid mixture and a reflectivity method. Green's functions are derived for both point loads and moments. Numerically integrated propagators represent solutions for intermediate heterogeneous layers in the porous region. These are substituted into a global matrix for solution by Gaussian elimination and back-substitution. Fluid partial stress and seismic responses to dislocations associated with fracturing of a layer of rock with a hydraulically conductive fracture network are computed with the model. A constitutive model is developed for representing the fractured rock layer as a porous material, using commonly accepted relationships for moduli. Derivations of density, tortuosity, and sinuosity are provided. The main results of the model application are the prediction of a substantial fluid partial stress response related to a second mode wave for the porous material. The response is observable for relatively large distances, on the order of several tens of meters. The visco-dynamic transition frequency associated with parabolic versus planar fluid velocity distributions across micro-crack apertures is in the low audio or seismic range, in contrast to materials with small pore size, such as porous rocks, for which the transition frequency is ultrasonic. Seismic responses are predicted for receiver locations both in the layer and in the outlying solid regions. In the porous region, the seismic response includes both shear and dilatational wave arrivals and a second-mode arrival. The second-mode arrival is not observable outside of the layer because of its low velocity relative to the dilatational and shear wave propagation velocities of the solid region.

  17. Fluid driven fracture mechanics in highly anisotropic shale: a laboratory study with application to hydraulic fracturing

    Science.gov (United States)

    Gehne, Stephan; Benson, Philip; Koor, Nick; Enfield, Mark

    2017-04-01

    The finding of considerable volumes of hydrocarbon resources within tight sedimentary rock formations in the UK led to focused attention on the fundamental fracture properties of low permeability rock types and hydraulic fracturing. Despite much research in these fields, there remains a scarcity of available experimental data concerning the fracture mechanics of fluid driven fracturing and the fracture properties of anisotropic, low permeability rock types. In this study, hydraulic fracturing is simulated in a controlled laboratory environment to track fracture nucleation (location) and propagation (velocity) in space and time and assess how environmental factors and rock properties influence the fracture process and the developing fracture network. Here we report data on employing fluid overpressure to generate a permeable network of micro tensile fractures in a highly anisotropic shale ( 50% P-wave velocity anisotropy). Experiments are carried out in a triaxial deformation apparatus using cylindrical samples. The bedding planes are orientated either parallel or normal to the major principal stress direction (σ1). A newly developed technique, using a steel guide arrangement to direct pressurised fluid into a sealed section of an axially drilled conduit, allows the pore fluid to contact the rock directly and to initiate tensile fractures from the pre-defined zone inside the sample. Acoustic Emission location is used to record and map the nucleation and development of the micro-fracture network. Indirect tensile strength measurements at atmospheric pressure show a high tensile strength anisotropy ( 60%) of the shale. Depending on the relative bedding orientation within the stress field, we find that fluid induced fractures in the sample propagate in two of the three principal fracture orientations: Divider and Short-Transverse. The fracture progresses parallel to the bedding plane (Short-Transverse orientation) if the bedding plane is aligned (parallel) with the

  18. Reliability of thermal-hydraulic passive safety systems

    International Nuclear Information System (INIS)

    D'Auria, F.; Araneo, D.; Pierro, F.; Galassi, G.

    2014-01-01

    The scholar will be informed of reliability concepts applied to passive system adopted for nuclear reactors. Namely, for classical components and systems the failure concept is associated with malfunction of breaking of hardware. In the case of passive systems the failure is associated with phenomena. A method for studying the reliability of passive systems is discussed and is applied. The paper deals with the description of the REPAS (Reliability Evaluation of Passive Safety System) methodology developed by University of Pisa (UNIPI) and with results from its application. The general objective of the REPAS methodology is to characterize the performance of a passive system in order to increase the confidence toward its operation and to compare the performances of active and passive systems and the performances of different passive systems

  19. Proceedings of the 10th international topical meeting on nuclear thermal hydraulics, operation and safety (NUTHOS-10)

    International Nuclear Information System (INIS)

    2014-01-01

    The 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operations and Safety (NUTHOS-10) in Okinawa, Japan is sponsored by Atomic Energy Society of Japan, in cooperation with the International Atomic Energy Agency, and co-sponsored by American Nuclear Society Thermal Hydraulics Division among others. Enhanced safety and reducing cost are going together, which can be achieved through continued research and development efforts. NUTHOS keeps you abreast of the most updated information in the advancement of science and technology in nuclear thermal hydraulics, operations and safety, and provides you insights into the future. (J.P.N.)

  20. Common hydraulic fracturing fluid additives alter the structure and function of anaerobic microbial communities

    Science.gov (United States)

    Mumford, Adam C.; Akob, Denise M.; Klinges, J. Grace; Cozzarelli, Isabelle M.

    2018-01-01

    The development of unconventional oil and gas (UOG) resources results in the production of large volumes of wastewater containing a complex mixture of hydraulic fracturing chemical additives and components from the formation. The release of these wastewaters into the environment poses potential risks that are poorly understood. Microbial communities in stream sediments form the base of the food chain and may serve as sentinels for changes in stream health. Iron-reducing organisms have been shown to play a role in the biodegradation of a wide range of organic compounds, and so to evaluate their response to UOG wastewater, we enriched anaerobic microbial communities from sediments collected upstream (background) and downstream (impacted) of an UOG wastewater injection disposal facility in the presence of hydraulic fracturing fluid (HFF) additives: guar gum, ethylene glycol, and two biocides, 2,2-dibromo-3-nitrilopropionamide (DBNPA) and bronopol (C3H6BrNO4). Iron reduction was significantly inhibited early in the incubations with the addition of biocides, whereas amendment with guar gum and ethylene glycol stimulated iron reduction relative to levels in the unamended controls. Changes in the microbial community structure were observed across all treatments, indicating the potential for even small amounts of UOG wastewater components to influence natural microbial processes. The microbial community structure differed between enrichments with background and impacted sediments, suggesting that impacted sediments may have been preconditioned by exposure to wastewater. These experiments demonstrated the potential for biocides to significantly decrease iron reduction rates immediately following a spill and demonstrated how microbial communities previously exposed to UOG wastewater may be more resilient to additional spills.

  1. Early fluid loading for septic patients: Any safety limit needed?

    Science.gov (United States)

    Gong, Yi-Chun; Liu, Jing-Tao; Ma, Peng-Lin

    2018-02-01

    Early adequate fluid loading was the corner stone of hemodynamic optimization for sepsis and septic shock. Meanwhile, recent recommended protocol for fluid resuscitation was increasingly debated on hemodynamic stability vs risk of overloading. In recent publications, it was found that a priority was often given to hemodynamic stability rather than organ function alternation in the early fluid resuscitation of sepsis. However, no safety limits were used at all in most of these reports. In this article, the rationality and safety of early aggressive fluid loading for septic patients were discussed. It was concluded that early aggressive fluid loading improved hemodynamics transitorily, but was probably traded off with a follow-up organ function impairment, such as worsening oxygenation by reduction of lung aeration, in a part of septic patients at least. Thus, a safeguard is needed against unnecessary excessive fluids in early aggressive fluid loading for septic patients. Copyright © 2017 Daping Hospital and the Research Institute of Surgery of the Third Military Medical University. Production and hosting by Elsevier B.V. All rights reserved.

  2. Safety of 5 MW district heating reactor (DHR) and hydraulic dynamic pressure drive control rods

    International Nuclear Information System (INIS)

    Wu Yuanqiang; Wang Dazhong

    1991-11-01

    The principles and movement characteristic of the hydraulic dynamic pressure drive for control rods in 5 MW district heating reactor are described with stress on analysis of its effects on reactor safety features. The drive is different from electric-magnetic drive for PWR or hydraulic drive for BWR. The drive cylinder is driven by dynamic pressure. In the new drive system, the reactor coolant (water) used as actuating medium is pressed by pump, then injected into a step cylinder which is set in the reactor core. The cylinder will move step by step by controlling flow, then the cylinder drives the neutron absorber and controls nuclear reaction. The drive is characterized by simplicity in structure, high reliability, inherent safety, reduction in reactor height, economy, etc

  3. Current status and future prospects for thermal-hydraulics and safety research

    International Nuclear Information System (INIS)

    Park, G.C.

    2000-01-01

    The present paper is to outline the current activities in Korea for the thermal-hydraulics and safety researches, and furthermore illuminate the future aspect of those field under the umbrella of worldwide nuclear prospect. In Korea, a long-term nuclear research plan has been established since 1992, which was recently funded with a fixed monetary rate of Korean won 1.20 per kWh of electricity produced with nuclear power. 11.5% of the fund is assigned for nuclear safety research in 6 areas. Under this program, 3 axes of research body (KAERI, KINS, University) has been operated with close cooperation. Their role, current activities and long-term plan of each body are introduced in the point of thermal-hydraulics' view. (author)

  4. Thermal-hydraulic transient characteristics of ship-propulsion reactor investigated through safety analysis

    International Nuclear Information System (INIS)

    Fujiki, Kazuo; Asaka, Hideaki; Ishida, Toshihisa

    1986-01-01

    Thermal-hydraulic behaviors in the reactor of Nuclear Ship ''Mutsu'' were investigated through safety evaluation of operational transients by using RETRAN and COBRA-IV codes. The results were compared to the transient behaviors of typical commercial PWR and the characteristics of transient thermal-hydraulic behaviors in ship-loaded reactor were figured out. ''Mutsu'' reactor has larger thermal margin than commercial PWR because it is designed to be used as ship-propulsion power source in the load-following operation mode. This margin makes transient behavior in general milder than in commercial PWR but high opening pressure set point of main-steam safety valves leads poor heat-sink condition after reactor trip. The effects of other small-sized components are also investigated. The findings in the paper will be helpful in the design of future advanced reactor for nuclear ship. (author)

  5. Thermal-hydraulics technological strategy roadmap for LWR safety improvement and development

    International Nuclear Information System (INIS)

    Nakamura, Hideo; Arai, Kenji; Oikawa, Hirohide

    2015-01-01

    New version of the Thermal-Hydraulics Safety Evaluation Fundamental Technology Enhancement Strategy Roadmap (TH-RM) was developed by the Atomic Energy Society of Japan (AESJ) for LWR safety improvement and development. The 1st version of TH-RM was prepared in 2009 under collaboration of utilities, vendors, universities, research institutes and technical support organizations (TSO) for regulatory body. The revision was made by three sub-working groups (SWGs) by considering the lessons learned from the Fukushima Daiichi Accident. The 'safety assessment' SWG pursued development of computer codes for safety assessment. The 'fundamental technology' SWG pursued safety improvement and risk reduction via accident management (AM) measures by referring the technical map for severe accident (SA) established by the 'severe accident' SWG. Phenomena and components for counter-measures and/or proper prediction are identified by going through SA progression in both reactor and spent-fuel pool of PWR and BWR. Twelve important technology development subjects were identified, which include melt coolability enhancement to maintain integrity of containment vessel. Fact Sheet was developed to describe each of identified and selected R and D subjects. External hazards are also considered how to cope with from thermal-hydraulic safety point of view. This paper summarizes the revised TH-RM with several examples and future perspectives. (author)

  6. Basic researches on thermo-hydraulic non-equilibrium phenomena related to nuclear reactor safety

    International Nuclear Information System (INIS)

    Sakurai, Akira; Kataoka, Isao; Aritomi, Masanori.

    1989-01-01

    A review was made of recent developments of fundamental researches on thermo-hydraulic non-equilibrium phenomena related to light water reactor safety, in relation to problems to be solved for the improvement of safety analysis codes. As for the problems related to flow con ditions, fundamental researches on basic conservation equations and constitutive equations for transient two-phase flow were reviewed. Regarding to the problems related to thermal non-equilibrium phenomena, fundamental researches on film boiling in pool and forced convection, transient boiling heat transfer and flow behavior caused by pressure transients were reviewed. (author)

  7. Regulatory support activities of JNES by thermal-hydraulic and safety analyses

    International Nuclear Information System (INIS)

    Kasahara, Fumio

    2008-01-01

    Current status and some related topics on regulatory support activities of Japan Nuclear Energy Safety Organization (JNES) by thermal-hydraulic and safety analyses are reported. The safety of nuclear facilities is secured primarily by plant owners and operators. However, the regulatory body NISA (Nuclear and Industrial Safety Agency) has conducted a strict regulation to confirm the adequacy of the site condition as well as the basic and detailed design. The JNES has been conducting independent analyses from applicants (audit analyses, etc.) by direction of NISA and supporting its review. In addition to the audit analysis, thermal-hydraulic and safety analyses are used in such areas as analytical evaluation for investigation of causes of accidents and troubles, level 2 PSA for risk informed regulation, etc. Recent activities of audit analyses are for the application of Tsuruga 3 and 4 (APWR), the spent fuel storage facility for the establishment, and the LMFBR Monju for the core change. For the trouble event evaluation, the criticality accident analysis of Sika1 was carried out and the evaluation of effectiveness of accident management (AM) measure for Tomari 3 (PWR) and Monju was performed. The analytical codes for these evaluations are continuously revised by reflecting the state-of-art technical information and validated using the information provided by the data from JAEA, OECD project, etc. (author)

  8. Near Wellbore Hydraulic Fracture Propagation from Perforations in Tight Rocks: The Roles of Fracturing Fluid Viscosity and Injection Rate

    Directory of Open Access Journals (Sweden)

    Seyed Hassan Fallahzadeh

    2017-03-01

    Full Text Available Hydraulic fracture initiation and near wellbore propagation is governed by complex failure mechanisms, especially in cased perforated wellbores. Various parameters affect such mechanisms, including fracturing fluid viscosity and injection rate. In this study, three different fracturing fluids with viscosities ranging from 20 to 600 Pa.s were used to investigate the effects of varying fracturing fluid viscosities and fluid injection rates on the fracturing mechanisms. Hydraulic fracturing tests were conducted in cased perforated boreholes made in tight 150 mm synthetic cubic samples. A true tri-axial stress cell was used to simulate real far field stress conditions. In addition, dimensional analyses were performed to correspond the results of lab experiments to field-scale operations. The results indicated that by increasing the fracturing fluid viscosity and injection rate, the fracturing energy increased, and consequently, higher fracturing pressures were observed. However, when the fracturing energy was transferred to a borehole at a faster rate, the fracture initiation angle also increased. This resulted in more curved fracture planes. Accordingly, a new parameter, called fracturing power, was introduced to relate fracture geometry to fluid viscosity and injection rate. Furthermore, it was observed that the presence of casing in the wellbore impacted the stress distribution around the casing in such a way that the fracture propagation deviated from the wellbore vicinity.

  9. Hydraulic Darrieus turbines efficiency for free fluid flow conditions versus power farms conditions

    Energy Technology Data Exchange (ETDEWEB)

    Antheaume, Sylvain [Electricite de France, Recherche et Developpement, Laboratoire National d' Hydraulique et Environnement, 6 Quai Watier, 78400 Chatou (France); Maitre, Thierry; Achard, Jean-Luc [Laboratoire des Ecoulements Geophysiques et Industriels, BP 53, 38041 Grenoble (France)

    2008-10-15

    The present study deals with the efficiency of cross flow water current turbine for free stream conditions versus power farm conditions. In the first part, a single turbine for free fluid flow conditions is considered. The simulations are carried out with a new in house code which couples a Navier-Stokes computation of the outer flow field with a description of the inner flow field around the turbine. The latter is based on experimental results of a Darrieus wind turbine in an unbounded domain. This code is applied for the description of a hydraulic turbine. In the second part, the interest of piling up several turbines on the same axis of rotation to make a tower is investigated. Not only is it profitable because only one alternator is needed but the simulations demonstrate the advantage of the tower configuration for the efficiency. The tower is then inserted into a cluster of several lined up towers which makes a barge. Simulations show that the average barge efficiency rises as the distance between towers is decreased and as the number of towers is increased within the row. Thereby, the efficiency of a single isolated turbine is greatly increased when set both into a tower and into a cluster of several towers corresponding to possible power farm arrangements. (author)

  10. Exposure of aircraft maintenance technicians to organophosphates from hydraulic fluids and turbine oils: a pilot study.

    Science.gov (United States)

    Schindler, Birgit Karin; Koslitz, Stephan; Weiss, Tobias; Broding, Horst Christoph; Brüning, Thomas; Bünger, Jürgen

    2014-01-01

    Hydraulic fluids and turbine oils contain organophosphates like tricresyl phosphate isomers, triphenyl phosphate and tributyl phosphate from very small up to high percentages. The aim of this pilot study was to determine if aircraft maintenance technicians are exposed to relevant amounts of organophosphates. Dialkyl and diaryl phosphate metabolites of seven organophosphates were quantified in pre- and post-shift spot urine samples of technicians (N=5) by GC-MS/MS after solid phase extraction and derivatization. Pre- and post shift values of tributyl phosphate metabolites (dibutyl phosphate (DBP): median pre-shift: 12.5 μg/L, post-shift: 23.5 μg/L) and triphenyl phosphate metabolites (diphenyl phosphate (DPP): median pre-shift: 2.9 μg/L, post-shift: 3.5 μg/L) were statistically higher than in a control group from the general population (median DBP: <0.25 μg/L, median DPP: 0.5 μg/L). No tricresyl phosphate metabolites were detected. The aircraft maintenance technicians were occupationally exposed to tributyl and triphenyl phosphate but not to tricresyl phosphate, tri-(2-chloroethyl)- and tri-(2-chloropropyl)-phosphate. Further studies are necessary to collect information on sources, routes of uptake and varying exposures during different work tasks, evaluate possible health effects and to set up appropriate protective measures. Copyright © 2013 Elsevier GmbH. All rights reserved.

  11. Thermal Hydraulic Computational Fluid Dynamics Simulations and Experimental Investigation of Deformed Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Mays, Brian [AREVA Federal Services, Lynchburg, VA (United States); Jackson, R. Brian [TerraPower, Bellevue, WA (United States)

    2017-03-08

    The project, Toward a Longer Life Core: Thermal Hydraulic CFD Simulations and Experimental Investigation of Deformed Fuel Assemblies, DOE Project code DE-NE0008321, was a verification and validation project for flow and heat transfer through wire wrapped simulated liquid metal fuel assemblies that included both experiments and computational fluid dynamics simulations of those experiments. This project was a two year collaboration between AREVA, TerraPower, Argonne National Laboratory and Texas A&M University. Experiments were performed by AREVA and Texas A&M University. Numerical simulations of these experiments were performed by TerraPower and Argonne National Lab. Project management was performed by AREVA Federal Services. The first of a kind project resulted in the production of both local point temperature measurements and local flow mixing experiment data paired with numerical simulation benchmarking of the experiments. The project experiments included the largest wire-wrapped pin assembly Mass Index of Refraction (MIR) experiment in the world, the first known wire-wrapped assembly experiment with deformed duct geometries and the largest numerical simulations ever produced for wire-wrapped bundles.

  12. The blind men and the elephant: the impact of context and scale in evaluating conflicts between plant hydraulic safety and efficiency

    Science.gov (United States)

    Frederick C. Meinzer; Katherine A. McCulloh; Barbara Lachenbruch; David R. Woodruff; Daniel M. Johnson

    2010-01-01

    Given the fundamental importance of xylem safety and efficiency for plant survival and fitness, it is not surprising that these are among the most commonly studied features of hydraulic architecture. However, much remains to be learned about the nature and universality of conflicts between hydraulic safety and efficiency. Although selection for suites of hydraulic...

  13. A HISTORICAL PERSPECTIVE OF NUCLEAR THERMAL HYDRAULICS

    Energy Technology Data Exchange (ETDEWEB)

    D’Auria, F; Rohatgi, Upendra S.

    2017-01-12

    The nuclear thermal-hydraulics discipline was developed following the needs for nuclear power plants (NPPs) and, to a more limited extent, research reactors (RR) design and safety. As in all other fields where analytical methods are involved, nuclear thermal-hydraulics took benefit of the development of computers. Thermodynamics, rather than fluid dynamics, is at the basis of the development of nuclear thermal-hydraulics together with the experiments in complex two-phase situations, namely, geometry, high thermal density, and pressure.

  14. 4D synchrotron X-ray imaging to understand porosity development in shales during exposure to hydraulic fracturing fluid

    Science.gov (United States)

    Kiss, A. M.; Bargar, J.; Kohli, A. H.; Harrison, A. L.; Jew, A. D.; Lim, J. H.; Liu, Y.; Maher, K.; Zoback, M. D.; Brown, G. E.

    2016-12-01

    Unconventional (shale) reservoirs have emerged as the most important source of petroleum resources in the United States and represent a two-fold decrease in greenhouse gas emissions compared to coal. Despite recent progress, hydraulic fracturing operations present substantial technical, economic, and environmental challenges, including inefficient recovery, wastewater production and disposal, contaminant and greenhouse gas pollution, and induced seismicity. A relatively unexplored facet of hydraulic fracturing operations is the fluid-rock interface, where hydraulic fracturing fluid (HFF) contacts shale along faults and fractures. Widely used, water-based fracturing fluids contain oxidants and acid, which react strongly with shale minerals. Consequently, fluid injection and soaking induces a host of fluid-rock interactions, most notably the dissolution of carbonates and sulfides, producing enhanced or "secondary" porosity networks, as well as mineral precipitation. The competition between these mechanisms determines how HFF affects reactive surface area and permeability of the shale matrix. The resultant microstructural and chemical changes may also create capillary barriers that can trap hydrocarbons and water. A mechanistic understanding of the microstructure and chemistry of the shale-HFF interface is needed to design new methodologies and fracturing fluids. Shales were imaged using synchrotron micro-X-ray computed tomography before, during, and after exposure to HFF to characterize changes to the initial 3D structure. CT reconstructions reveal how the secondary porosity networks advance into the shale matrix. Shale samples span a range of lithologies from siliceous to calcareous to organic-rich. By testing shales of different lithologies, we have obtained insights into the mineralogic controls on secondary pore network development and the morphologies at the shale-HFF interface and the ultimate composition of produced water from different facies. These results

  15. Guidelines for Safety Evaluation of a Potential for PWR Steam Generator Tube Failure due to Fluid elastic Instability

    Energy Technology Data Exchange (ETDEWEB)

    Jo, Jong Chull; Do, Kyu Sik; Sheen, Cheol [Nuclear System Evaluation Dept., Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-05-15

    It was found that both SG tube rupture events occurred at North Anna Unit 1 in 1987 and at Mihama Unit 2 in 1991 were caused by a high cycle fatigue due to fluid elastic instability. Therefore, with regard to nuclear safety it is important to design the SG properly in a conservative manner so that the potential for SG U-tube failures due to fluid elastic instability can be minimized. This article provides guidelines for assessing the potential for SG U-tube damage due to fluid elastic instability. This article described guidelines for safety evaluation of a potential for PWR steam generator tube failure due to fluid elastic instability. The guidelines address the requirements for realistically performing the SG thermal-hydraulic analysis and the modal analysis of tubes as well as the criteria for conservatively determining the added mass, the damping ratio and the fluid elastic instability coefficient. The guidelines can be used to predict the potential SG tubes which are susceptible to failure due to fluid elastic instability at operating nuclear power plants and also to evaluate the safety and structural integrity of new SG designs at the licensing review stage. Failure of a pressurized water reactor (PWR) steam generator (SG) tube leads to a leakage of contaminated primary coolant to the secondary system, which has serious safety implications such as the potential for direct release of radioactive fission products to the environment and the loss of coolant. Excessive tube vibration excited by dynamic forces of internal or external fluid flow is called flow-induced vibration (FIV). Among the FIV mechanisms, the so-called fluid elastic instability of SG tubes in cross flow is the most important safety issue in the design of SGs because it may cause severe tube failure in a very short time.

  16. Common Hydraulic Fracturing Fluid Additives Alter the Structure and Function of Anaerobic Microbial Communities.

    Science.gov (United States)

    Mumford, Adam C; Akob, Denise M; Klinges, J Grace; Cozzarelli, Isabelle M

    2018-04-15

    The development of unconventional oil and gas (UOG) resources results in the production of large volumes of wastewater containing a complex mixture of hydraulic fracturing chemical additives and components from the formation. The release of these wastewaters into the environment poses potential risks that are poorly understood. Microbial communities in stream sediments form the base of the food chain and may serve as sentinels for changes in stream health. Iron-reducing organisms have been shown to play a role in the biodegradation of a wide range of organic compounds, and so to evaluate their response to UOG wastewater, we enriched anaerobic microbial communities from sediments collected upstream (background) and downstream (impacted) of an UOG wastewater injection disposal facility in the presence of hydraulic fracturing fluid (HFF) additives: guar gum, ethylene glycol, and two biocides, 2,2-dibromo-3-nitrilopropionamide (DBNPA) and bronopol (C 3 H 6 BrNO 4 ). Iron reduction was significantly inhibited early in the incubations with the addition of biocides, whereas amendment with guar gum and ethylene glycol stimulated iron reduction relative to levels in the unamended controls. Changes in the microbial community structure were observed across all treatments, indicating the potential for even small amounts of UOG wastewater components to influence natural microbial processes. The microbial community structure differed between enrichments with background and impacted sediments, suggesting that impacted sediments may have been preconditioned by exposure to wastewater. These experiments demonstrated the potential for biocides to significantly decrease iron reduction rates immediately following a spill and demonstrated how microbial communities previously exposed to UOG wastewater may be more resilient to additional spills. IMPORTANCE Organic components of UOG wastewater can alter microbial communities and biogeochemical processes, which could alter the rates of

  17. Resolution of thermal-hydraulic safety and licensing issues for the system 80+trademark design

    International Nuclear Information System (INIS)

    Carpentino, S.E.; Ritterbusch, S.E.; Schneider, R.E.

    1995-01-01

    The System 80+ trademark Standard Design is an evolutionary Advanced Light Water Reactor (ALWR) with a generating capacity of 3931 MWt (1350 MWe). The Final Design Approval (FDA) for this design was issued by the Nuclear Regulatory Commission (NRC) in July 1994. The design certification by the NRC is anticipated by the end of 1995 or early 1996. NRC review of the System 80+ design has involved several new safety issues never before addressed in a regulatory atmosphere. In addition, conformance with the Electric Power Research Institute (EPRI) ALWR Utility Requirements Document (URD) required that the System 80+ plant address nuclear industry concerns with regard to design, construction, operation and maintenance of nuclear power plants. A large number of these issues/concerns deals with previously unresolved generic thermal-hydraulic safety issues and severe accident prevention and mitigation. This paper discusses the thermal-hydraulic analyses and evaluations performed for the System 80+ design to resolve safety and licensing issues relevant to both the Nuclear Stream Supply System (NSSS) and containment designs. For the NSSS design, the Safety Depressurization System mitigation capability and resolution of the boron dilution concern are described. Examples of containment design issues dealing with containment shell strength, robustness of the reactor cavity walls and hydrogen mixing under severe accident conditions are also provided. Finally, the overall approach used in the application of NRC's new (NUREG-1465) radiological source term for System 80+ evaluation is described. The robustness of the System 80+ containment design to withstand severe accident consequences was demonstrated through detailed thermal-hydraulic analyses and evaluations. This advanced design to shown to meet NRC severe accident policy goals and ALWR URD requirements without any special design features and unnecessary costs

  18. Resolution of thermal-hydraulic safety and licensing issues for the system 80+{sup {trademark}} design

    Energy Technology Data Exchange (ETDEWEB)

    Carpentino, S.E.; Ritterbusch, S.E.; Schneider, R.E. [ABB-Combustion Engineering, Windsor, CT (United States)] [and others

    1995-09-01

    The System 80+{sup {trademark}} Standard Design is an evolutionary Advanced Light Water Reactor (ALWR) with a generating capacity of 3931 MWt (1350 MWe). The Final Design Approval (FDA) for this design was issued by the Nuclear Regulatory Commission (NRC) in July 1994. The design certification by the NRC is anticipated by the end of 1995 or early 1996. NRC review of the System 80+ design has involved several new safety issues never before addressed in a regulatory atmosphere. In addition, conformance with the Electric Power Research Institute (EPRI) ALWR Utility Requirements Document (URD) required that the System 80+ plant address nuclear industry concerns with regard to design, construction, operation and maintenance of nuclear power plants. A large number of these issues/concerns deals with previously unresolved generic thermal-hydraulic safety issues and severe accident prevention and mitigation. This paper discusses the thermal-hydraulic analyses and evaluations performed for the System 80+ design to resolve safety and licensing issues relevant to both the Nuclear Stream Supply System (NSSS) and containment designs. For the NSSS design, the Safety Depressurization System mitigation capability and resolution of the boron dilution concern are described. Examples of containment design issues dealing with containment shell strength, robustness of the reactor cavity walls and hydrogen mixing under severe accident conditions are also provided. Finally, the overall approach used in the application of NRC`s new (NUREG-1465) radiological source term for System 80+ evaluation is described. The robustness of the System 80+ containment design to withstand severe accident consequences was demonstrated through detailed thermal-hydraulic analyses and evaluations. This advanced design to shown to meet NRC severe accident policy goals and ALWR URD requirements without any special design features and unnecessary costs.

  19. Evaluation on thermal-hydraulic characteristics for passive safety device of APR1400

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Seong Yeon; Lee, S. H.; Son, M. K. [Korea Association for Nuclear Technology, Taejon (Korea, Republic of); Jee, M. S.; Chung, M. H. [Chungnam National Univ., Taejon (Korea, Republic of)

    2001-07-15

    To establish evaluation and verification guideline for the APR1400, thermal-hydraulic characteristics for fuel rod bundle, reactor vessel and fluidic device is analyzed using FLUENT. Scope and major results of research are as follows : Thermal-hydraulic characteristics for nuclear fuel rod bundle: design data for nuclear fuel rod bundle and structure are surveyed, and 3 x 3 sub-channel model is adopted to investigate the fluid flow and heat transfer characteristics in fuel rod bundle. Computational results are compared with the heat transfer data measured by naphthalene sublimation method, and numerical analysis and evaluation are performed at various design conditions and flow conditions. Thermal-hydraulic characteristics for reactor vessel: reactor vessel design data are surveyed to develop numerical model. Porous media model is applied for fuel rod bundle, and full-scale, three dimensional simulation is performed at actual operating conditions. Distributions of velocity, pressure and temperature are discussed. Flow characteristics for fluidic device: three dimensional numerical model for fluidic device is developed, and numerical results are compared with experimental data obtained at KAERI in order to verify numerical simulation. In addition, variation of flow rate is investigated at various elapsed times after valve operating, and flow characteristics is analyzed at low and high flow rate conditions, respectively.

  20. Thermal-Hydraulic System Codes in Nulcear Reactor Safety and Qualification Procedures

    Directory of Open Access Journals (Sweden)

    Alessandro Petruzzi

    2008-01-01

    Full Text Available In the last four decades, large efforts have been undertaken to provide reliable thermal-hydraulic system codes for the analyses of transients and accidents in nuclear power plants. Whereas the first system codes, developed at the beginning of the 1970s, utilized the homogenous equilibrium model with three balance equations to describe the two-phase flow, nowadays the more advanced system codes are based on the so-called “two-fluid model” with separation of the water and vapor phases, resulting in systems with at least six balance equations. The wide experimental campaign, constituted by the integral and separate effect tests, conducted under the umbrella of the OECD/CSNI was at the basis of the development and validation of the thermal-hydraulic system codes by which they have reached the present high degree of maturity. However, notwithstanding the huge amounts of financial and human resources invested, the results predicted by the code are still affected by errors whose origins can be attributed to several reasons as model deficiencies, approximations in the numerical solution, nodalization effects, and imperfect knowledge of boundary and initial conditions. In this context, the existence of qualified procedures for a consistent application of qualified thermal-hydraulic system code is necessary and implies the drawing up of specific criteria through which the code-user, the nodalization, and finally the transient results are qualified.

  1. Coupled neutronics/thermal-hydraulics and safety characteristics of liquid-fueled molten salt reactors

    Energy Technology Data Exchange (ETDEWEB)

    Qiu, Suizheng; Zhang, Dalin; Liu, Minghao; Liu, Limin; Xu, Rongshuan; Gong, Cheng; Su, Guanghui [Xi' an Jiaotong Univ. (China). State Key Laboratory of Multiphase Flow in Power Engineering

    2016-05-15

    Molten salt reactor (MSR) as one candidate of the Generation IV advanced nuclear power systems is attracted more attention in China due to its top ranked fuel cycle and thorium utilization. The MSRs are characterized by using liquid-fuel, which offers complicated coupling problem of neutronics and thermal hydraulics. In this paper, the fundamental model and numerical method are established to calculate and analyze the safety characteristics for liquid-fuel MSRs. The theories and methodologies are applied to the MOSART concept. The liquid-fuel flow effects on neutronics, reactivity coefficients and three operation parameters' influences at steady state are obtained, which provide the basic information for safety analysis. The unprotected loss of flow transient is calculated, the results of which shows the inherent safety characteristics of MOSART due to its strong negative reactivity feedbacks.

  2. Coupled neutronics/thermal-hydraulics and safety characteristics of liquid-fueled molten salt reactors

    International Nuclear Information System (INIS)

    Qiu, Suizheng; Zhang, Dalin; Liu, Minghao; Liu, Limin; Xu, Rongshuan; Gong, Cheng; Su, Guanghui

    2016-01-01

    Molten salt reactor (MSR) as one candidate of the Generation IV advanced nuclear power systems is attracted more attention in China due to its top ranked fuel cycle and thorium utilization. The MSRs are characterized by using liquid-fuel, which offers complicated coupling problem of neutronics and thermal hydraulics. In this paper, the fundamental model and numerical method are established to calculate and analyze the safety characteristics for liquid-fuel MSRs. The theories and methodologies are applied to the MOSART concept. The liquid-fuel flow effects on neutronics, reactivity coefficients and three operation parameters' influences at steady state are obtained, which provide the basic information for safety analysis. The unprotected loss of flow transient is calculated, the results of which shows the inherent safety characteristics of MOSART due to its strong negative reactivity feedbacks.

  3. Fast reactor safety and computational thermo-fluid dynamics approaches

    International Nuclear Information System (INIS)

    Ninokata, Hisashi; Shimizu, Takeshi

    1993-01-01

    This article provides a brief description of the safety principle on which liquid metal cooled fast breeder reactors (LMFBRs) is based and the roles of computations in the safety practices. A number of thermohydraulics models have been developed to date that successfully describe several of the important types of fluids and materials motion encountered in the analysis of postulated accidents in LMFBRs. Most of these models use a mixture of implicit and explicit numerical solution techniques in solving a set of conservation equations formulated in Eulerian coordinates, with special techniques included to specific situations. Typical computational thermo-fluid dynamics approaches are discussed in particular areas of analyses of the physical phenomena relevant to the fuel subassembly thermohydraulics design and that involve describing the motion of molten materials in the core over a large scale. (orig.)

  4. Essentials of fluid dynamics with applications to hydraulics, aeronautics, meteorology and other subjets

    CERN Document Server

    Prandtl, Ludwig

    1953-01-01

    Equilibrium of liquids and gases ; kinematics : dynamics of frictionless fluids ; motion of viscous fluids : turbulence : fluid resistance : practical applications ; flow with appreciable volume changes (dynamics of gases) ; miscellaneous topics.

  5. Fluid dispersal from safety cannulas: an in vitro comparative test.

    Science.gov (United States)

    Rosenthal, Victor D; Hughes, Gavin

    2015-03-01

    We report a comparative laboratory study between 2 peripheral intravenous catheters equipped with a passive fully automatic safety mechanism to assess generation of blood droplets during withdrawal. One presented no fluid droplets, whereas the other presented droplets in 48% and 60% for the best and worst case, with analysis of variance showing positive effects on the number of droplets generated (P blood splatter. Copyright © 2015 Association for Professionals in Infection Control and Epidemiology, Inc. Published by Elsevier Inc. All rights reserved.

  6. A reactive transport modelling approach to assess the leaching potential of hydraulic fracturing fluids associated with coal seam gas extraction

    Science.gov (United States)

    Mallants, Dirk; Simunek, Jirka; Gerke, Kirill

    2015-04-01

    Coal Seam Gas production generates large volumes of "produced" water that may contain compounds originating from the use of hydraulic fracturing fluids. Such produced water also contains elevated concentrations of naturally occurring inorganic and organic compounds, and usually has a high salinity. Leaching of produced water from storage ponds may occur as a result of flooding or containment failure. Some produced water is used for irrigation of specific crops tolerant to elevated salt levels. These chemicals may potentially contaminate soil, shallow groundwater, and groundwater, as well as receiving surface waters. This paper presents an application of scenario modelling using the reactive transport model for variably-saturated media HP1 (coupled HYDRUS-1D and PHREEQC). We evaluate the fate of hydraulic fracturing chemicals and naturally occurring chemicals in soil as a result of unintentional release from storage ponds or when produced water from Coal Seam Gas operations is used in irrigation practices. We present a review of exposure pathways and relevant hydro-bio-geo-chemical processes, a collation of physico-chemical properties of organic/inorganic contaminants as input to a set of generic simulations of transport and attenuation in variably saturated soil profiles. We demonstrate the ability to model the coupled processes of flow and transport in soil of contaminants associated with hydraulic fracturing fluids and naturally occurring contaminants.

  7. Preliminary thermal-hydraulic and safety analysis of China DFLL-TBM system

    Energy Technology Data Exchange (ETDEWEB)

    Li, Wei [School of Nuclear Science and Technology, Xi’an Jiaotong University, No. 28, Xianning West Road, Xi’an, Shanxi 710049 (China); Tian, Wenxi, E-mail: wxtian@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, No. 28, Xianning West Road, Xi’an, Shanxi 710049 (China); Qiu, Suizheng; Su, Guanghui; Jiao, Hong [School of Nuclear Science and Technology, Xi’an Jiaotong University, No. 28, Xianning West Road, Xi’an, Shanxi 710049 (China); Bai, Yunqing; Chen, Hongli [Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Wu, Yican, E-mail: yican.Wu@Fds.Org.Cn [Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui 230031 (China)

    2013-06-15

    Highlights: • Thermal-hydraulic and safety analysis on DFLL-TBM system is performed. • The TBM FW maximum temperature is 541 °C under steady state condition. • The TBM FW maximum temperature does not exceed the melt point of CLAM steel 1500 °C. • Neither the VV pressurization nor vault pressure build-up goes beyond 0.2 MPa. -- Abstract: China has proposed the dual-functional lithium-lead (DFLL) tritium breeding blanket concept for testing in ITER as a test blanket module (TBM), to demonstrate the technologies of tritium self-sufficiency, high-grade heat extraction and efficient electricity production which are needed for DEMO and fusion power plant. Safety assessment of the TBM and its auxiliary system should be conducted to deal with ITER safety issues directly caused by the TBM system failure during the design process. In this work, three potential initial events (PIEs) – in-vessel loss of helium (He) coolant and ex-vessel loss of He coolant and loss of flow without scram (LOFWS) – were analyzed for the TBM system with a modified version of the RELAP5/MOD3 code containing liquid lithium-lead eutectic (LiPb). The code also comprised an empirical expression for MHD pressure drop relevant to three-dimensional (3D) effect, the Lubarsky–Kaufman convective heat transfer correlation for LiPb flow and the Gnielinski convective heat transfer correlation for He flow. Since both LiPb and He serve as TBM coolants, the LiPb and He ancillary cooling systems were modeled to investigate the thermal-hydraulic characteristic of the TBM system and its influence on ITER safety under those accident conditions. The TBM components and the coolants flow within the TBM were simulated with one-dimensional heat structures and their associated hydrodynamic components. ITER enclosures including vacuum vessel (VV), port cell and TCWS vault were also covered in the model for accident analyses. Through this best estimate approach, the calculation indicated that the current

  8. Xylem hydraulic safety margins in woody plants: coordination of stomatal control of xylem tension with hydraulic capacitance

    Science.gov (United States)

    Frederick C. Meinzer; Daniel M. Johnson; Barbara Lachenbruch; Katherine A. McCulloh; David R. Woodruff

    2009-01-01

    The xylem pressure inducing 50% loss of hydraulic conductivity due to embolism (P50) is widely used for comparisons of xylem vulnerability among species and across aridity gradients. However, despite its utility as an index of resistance to catastrophic xylem failure under extreme drought, P50 may have no special...

  9. Fracture Propagation, Fluid Flow, and Geomechanics of Water-Based Hydraulic Fracturing in Shale Gas Systems and Electromagnetic Geophysical Monitoring of Fluid Migration

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jihoon; Um, Evan; Moridis, George

    2014-12-01

    We investigate fracture propagation induced by hydraulic fracturing with water injection, using numerical simulation. For rigorous, full 3D modeling, we employ a numerical method that can model failure resulting from tensile and shear stresses, dynamic nonlinear permeability, leak-off in all directions, and thermo-poro-mechanical effects with the double porosity approach. Our numerical results indicate that fracture propagation is not the same as propagation of the water front, because fracturing is governed by geomechanics, whereas water saturation is determined by fluid flow. At early times, the water saturation front is almost identical to the fracture tip, suggesting that the fracture is mostly filled with injected water. However, at late times, advance of the water front is retarded compared to fracture propagation, yielding a significant gap between the water front and the fracture top, which is filled with reservoir gas. We also find considerable leak-off of water to the reservoir. The inconsistency between the fracture volume and the volume of injected water cannot properly calculate the fracture length, when it is estimated based on the simple assumption that the fracture is fully saturated with injected water. As an example of flow-geomechanical responses, we identify pressure fluctuation under constant water injection, because hydraulic fracturing is itself a set of many failure processes, in which pressure consistently drops when failure occurs, but fluctuation decreases as the fracture length grows. We also study application of electromagnetic (EM) geophysical methods, because these methods are highly sensitive to changes in porosity and pore-fluid properties due to water injection into gas reservoirs. Employing a 3D finite-element EM geophysical simulator, we evaluate the sensitivity of the crosswell EM method for monitoring fluid movements in shaly reservoirs. For this sensitivity evaluation, reservoir models are generated through the coupled flow

  10. Thermal-hydraulic code for estimating safety limits of nuclear reactors with plate type fuels

    Energy Technology Data Exchange (ETDEWEB)

    Castellanos, Duvan A.; Moreira, João L.; Maiorino, Jose R.; Rossi, Pedro R.; Carajilescov, Pedro, E-mail: duvan.castellanos@ufabc.edu.br, E-mail: joao.moreira@ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br, E-mail: pedro.rossi@ufabc.edu.br, E-mail: pedro.carajilescov10@gmail.com [Universidade Federal do ABC (UFABC), Santo André, SP (Brazil). Centro de Engenharias, Modelagem e Ciências Sociais Aplicadas

    2017-07-01

    To ensure the normal and safe operation of PWR type nuclear reactors is necessary the knowledge of nuclear and heat transfer properties of the fuel, coolant and structural materials. The thermal-hydraulic analysis of nuclear reactors yields parameters such as the distribution of fuel and coolant temperatures, and the departure from nucleated boiling ratio. Usually computational codes are used to analyze the safety performance of the core. This research work presents a computer code for performing thermal-hydraulic analyses of nuclear reactors with plate-type fuel elements operating at low pressure and temperature (research reactors) or high temperature and pressure (naval propulsion or small power reactors). The code uses the sub-channel method based on geometric and thermal-hydraulic conditions. In order to solve the conservation equations for mass, momentum and energy, each sub-channel is divided into control volumes in the axial direction. The mass flow distribution for each fuel element of core is obtained. Analysis of critical heat flux is performed in the hottest channel. The code considers the radial symmetry and the chain or cascade method for two steps in order to facilitate the whole analysis. In the first step, we divide the core into channels with size equivalent to a fuel assembly. >From this analysis, the channel with the largest enthalpy is identified as the hot assembly. In the second step, we divide the hottest fuel assembly into sub-channels with size equivalent to one actual coolant channel. As in the previous step, the sub-channel with largest final enthalpy is identified as the hottest sub-channel. For the code validation, we considered results from the chinese CARR research reactor. The code reproduced well the CARR reactor results, yielding detailed information such as static pressure in the channel, mass flow rate distribution among the fuel channels, coolant, clad and centerline fuel temperatures, quality and local heat and critical heat

  11. Thermal-hydraulic code for estimating safety limits of nuclear reactors with plate type fuels

    International Nuclear Information System (INIS)

    Castellanos, Duvan A.; Moreira, João L.; Maiorino, Jose R.; Rossi, Pedro R.; Carajilescov, Pedro

    2017-01-01

    To ensure the normal and safe operation of PWR type nuclear reactors is necessary the knowledge of nuclear and heat transfer properties of the fuel, coolant and structural materials. The thermal-hydraulic analysis of nuclear reactors yields parameters such as the distribution of fuel and coolant temperatures, and the departure from nucleated boiling ratio. Usually computational codes are used to analyze the safety performance of the core. This research work presents a computer code for performing thermal-hydraulic analyses of nuclear reactors with plate-type fuel elements operating at low pressure and temperature (research reactors) or high temperature and pressure (naval propulsion or small power reactors). The code uses the sub-channel method based on geometric and thermal-hydraulic conditions. In order to solve the conservation equations for mass, momentum and energy, each sub-channel is divided into control volumes in the axial direction. The mass flow distribution for each fuel element of core is obtained. Analysis of critical heat flux is performed in the hottest channel. The code considers the radial symmetry and the chain or cascade method for two steps in order to facilitate the whole analysis. In the first step, we divide the core into channels with size equivalent to a fuel assembly. >From this analysis, the channel with the largest enthalpy is identified as the hot assembly. In the second step, we divide the hottest fuel assembly into sub-channels with size equivalent to one actual coolant channel. As in the previous step, the sub-channel with largest final enthalpy is identified as the hottest sub-channel. For the code validation, we considered results from the chinese CARR research reactor. The code reproduced well the CARR reactor results, yielding detailed information such as static pressure in the channel, mass flow rate distribution among the fuel channels, coolant, clad and centerline fuel temperatures, quality and local heat and critical heat

  12. Lateral hydraulic forces calculation on PWR fuel assemblies with computational fluid dynamics codes

    International Nuclear Information System (INIS)

    Corpa Masa, R.; Jimenez Varas, G.; Moreno Garcia, B.

    2016-01-01

    To be able to simulate the behavior of nuclear fuel under operating conditions, it is required to include all the representative loads, including the lateral hydraulic forces which were not included traditionally because of the difficulty of calculating them in a reliable way. Thanks to the advance in CFD codes, now it is possible to assess them. This study calculates the local lateral hydraulic forces, caused by the contraction and expansion of the flow due to the bow of the surrounding fuel assemblies, on of fuel assembly under typical operating conditions from a three loop Westinghouse PWR reactor. (Author)

  13. Estimation of changes in dynamic hydraulic force in a magnetically suspended centrifugal blood pump with transient computational fluid dynamics analysis.

    Science.gov (United States)

    Masuzawa, Toru; Ohta, Akiko; Tanaka, Nobuatu; Qian, Yi; Tsukiya, Tomonori

    2009-01-01

    The effect of the hydraulic force on magnetically levitated (maglev) pumps should be studied carefully to improve the suspension performance and the reliability of the pumps. A maglev centrifugal pump, developed at Ibaraki University, was modeled with 926 376 hexahedral elements for computational fluid dynamics (CFD) analyses. The pump has a fully open six-vane impeller with a diameter of 72.5 mm. A self-bearing motor suspends the impeller in the radial direction. The maximum pressure head and flow rate were 250 mmHg and 14 l/min, respectively. First, a steady-state analysis was performed using commercial code STAR-CD to confirm the model's suitability by comparing the results with the real pump performance. Second, transient analysis was performed to estimate the hydraulic force on the levitated impeller. The impeller was rotated in steps of 1 degrees using a sliding mesh. The force around the impeller was integrated at every step. The transient analysis revealed that the direction of the radial force changed dynamically as the vane's position changed relative to the outlet port during one circulation, and the magnitude of this force was about 1 N. The current maglev pump has sufficient performance to counteract this hydraulic force. Transient CFD analysis is not only useful for observing dynamic flow conditions in a centrifugal pump but is also effective for obtaining information about the levitation dynamics of a maglev pump.

  14. Research and development program for PWR safety at the CEA reactor thermal hydraulics laboratories

    International Nuclear Information System (INIS)

    Bernard, M.

    1995-01-01

    Since the start of the French electronuclear program, the three partners Fermate, EDF and Cea (DRN and IPSN) have devoted considerable effort to research and development for safety issues. In particular an important program on thermal hydraulics was initiated at the beginning of the seventies. It is illustrated by the development of the CATHARE thermalhydraulic safety code which includes physical models derived from a large experimental support program and the construction of the BETHSY integral facility which is aimed to assess both the CATHARE code and the physical relevance of the accident management procedures to be applied on reactors. The state of the art on this program is described with particular emphasis on the capabilities and the assessment of the last version of CATHARE and the lessons drawn from 50 BETHSY tests performed so far. The future plans for safety research cover the following strategy: - to solve the few problems identified on present computing tools and extend the assessment - to solve the few problems identified on present computing tools and extend the assessment - to perform safety studies on the basis of plant operation feedback - to contribute to treating the safety issues related to the future reactors and in particular the case of severe accidents which have to be taken into account from the design stage. The program on severe accidents is aimed to support the design studies performed by the industrial partners and to provide computing tools which model the various phases of severe accidents and will be validated on experiments performed with real and simulating materials. The main lines of the program are: - the development of the TOLBIAC 3D code for the thermal hydraulics of core melt pools, which will be validated against the Bali experiment presently under construction - the Sultan experiment, to study the capability of cooling by external flooding of the reactor vessel - the development of the MC-3D code for core melt

  15. Experimental Validation of Modelled Fluid Forces in Fast Switching Hydraulic On/Off Valves

    DEFF Research Database (Denmark)

    Nørgård, Christian; Bech, Michael Møller; Roemer, Daniel Beck

    2015-01-01

    A prototype of a fast switching valve for a digital hydraulic machine has been designed and manufactured. The valve is composed of an annular seat plunger connected with a moving coil actuator as the force producing element. The valve prototype is designed for flow rates of 600 l/min with less th...

  16. Related research with thermo hydraulics safety by means of Trace code

    International Nuclear Information System (INIS)

    Chaparro V, F. J.; Del Valle G, E.; Rodriguez H, A.; Gomez T, A. M.; Sanchez E, V. H.; Jager, W.

    2014-10-01

    In this article the results of the design of a pressure vessel of a BWR/5 similar to the type of Laguna Verde NPP are presented, using the Trace code. A thermo hydraulics Vessel component capable of simulating the behavior of fluids and heat transfer that occurs within the reactor vessel was created. The Vessel component consists of a three-dimensional cylinder divided into 19 axial sections, 4 azimuthal sections and two concentric radial rings. The inner ring is used to contain the core and the central part of the reactor, while the outer ring is used as a down comer. Axial an azimuthal divisions were made with the intention that the dimensions of the internal components, heights and orientation of the external connections match the reference values of a reactor BWR/5 type. In the model internal components as, fuel assemblies, steam separators, jet pumps, guide tubes, etc. are included and main external connections as, steam lines, feed-water or penetrations of the recirculation system. The model presents significant simplifications because the object is to keep symmetry between each azimuthal section of the vessel. In most internal components lack a detailed description of the geometry and initial values of temperature, pressure, fluid velocity, etc. given that it only considered the most representative data, however with these simulations are obtained acceptable results in important parameters such as the total flow through the core, the pressure in the vessel, percentage of vacuums fraction, pressure drop in the core and the steam separators. (Author)

  17. Degradation of phosphate ester hydraulic fluid in power station turbines investigated by a three-magnet unilateral magnet array.

    Science.gov (United States)

    Guo, Pan; He, Wei; García-Naranjo, Juan C

    2014-04-14

    A three-magnet array unilateral NMR sensor with a homogeneous sensitive spot was employed for assessing aging of the turbine oils used in two different power stations. The Carr-Purcell-Meiboom-Gill (CPMG) sequence and Inversion Recovery-prepared CPMG were employed for measuring the ¹H-NMR transverse and longitudinal relaxation times of turbine oils with different service status. Two signal components with different lifetimes were obtained by processing the transverse relaxation curves with a numeric program based on the Inverse Laplace Transformation. The long lifetime components of the transverse relaxation time T₂eff and longitudinal relaxation time T₁ were chosen to monitor the hydraulic fluid aging. The results demonstrate that an increase of the service time of the turbine oils clearly results in a decrease of T₂eff,long and T₁,long. This indicates that the T₂eff,long and T₁,long relaxation times, obtained from the unilateral magnetic resonance measurements, can be applied as indices for degradation of the hydraulic fluid in power station turbines.

  18. Measurement of fluid film thickness on the valve plate in oil hydraulic axial piston pumps (I): bearing pad effects

    International Nuclear Information System (INIS)

    Kim, Jong Ki; Jung, Jae Youn

    2003-01-01

    The tribological mechanism between the valve plate and the cylinder block in oil hydraulic axial piston pumps plays an important role on high power density. In this study, the fluid film thickness between the valve plate and the cylinder block was measured with discharge pressure and rotational speed by use of a gap sensor, and a slip ring system in the operating period. To investigate the effect of the valve plate shapes, we designed two valve plates with different shapes: the first valve plate was without a bearing pad, while the second valve plate had a bearing pad. It was found that both valve plates behaved differently with respect to the fluid film thickness characteristics. The leakage flow rates and the shaft torque were also experimented in order to clarify the performance difference between the valve plate without a bearing pad and the valve plate with a bearing pad. From the results of this study, we found out that in the oil hydraulic axial piston pumps, the valve plate with a bearing pad showed better film thickness contours than the valve plate without a bearing pad

  19. Finite mixture models for sensitivity analysis of thermal hydraulic codes for passive safety systems analysis

    Energy Technology Data Exchange (ETDEWEB)

    Di Maio, Francesco, E-mail: francesco.dimaio@polimi.it [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Nicola, Giancarlo [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Zio, Enrico [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Chair on System Science and Energetic Challenge Fondation EDF, Ecole Centrale Paris and Supelec, Paris (France); Yu, Yu [School of Nuclear Science and Engineering, North China Electric Power University, 102206 Beijing (China)

    2015-08-15

    Highlights: • Uncertainties of TH codes affect the system failure probability quantification. • We present Finite Mixture Models (FMMs) for sensitivity analysis of TH codes. • FMMs approximate the pdf of the output of a TH code with a limited number of simulations. • The approach is tested on a Passive Containment Cooling System of an AP1000 reactor. • The novel approach overcomes the results of a standard variance decomposition method. - Abstract: For safety analysis of Nuclear Power Plants (NPPs), Best Estimate (BE) Thermal Hydraulic (TH) codes are used to predict system response in normal and accidental conditions. The assessment of the uncertainties of TH codes is a critical issue for system failure probability quantification. In this paper, we consider passive safety systems of advanced NPPs and present a novel approach of Sensitivity Analysis (SA). The approach is based on Finite Mixture Models (FMMs) to approximate the probability density function (i.e., the uncertainty) of the output of the passive safety system TH code with a limited number of simulations. We propose a novel Sensitivity Analysis (SA) method for keeping the computational cost low: an Expectation Maximization (EM) algorithm is used to calculate the saliency of the TH code input variables for identifying those that most affect the system functional failure. The novel approach is compared with a standard variance decomposition method on a case study considering a Passive Containment Cooling System (PCCS) of an Advanced Pressurized reactor AP1000.

  20. Thermo-hydraulic characteristics of ship propulsion reactor in the conditions of ship motions and safety assessment

    International Nuclear Information System (INIS)

    Kobayashi, Michiyuki; Aya, Izuo; Inasaka, Fujio; Murata, Hiroyuki; Odano, Naoteru; Shiozaki, Koki

    1998-01-01

    A research project from 1995-1999 had a plan to make experimental studies on (1) safety of nuclear ship loaded with an integral ship propulsion reactor (2) effects of pulsating flow on the thermo-hydraulic characteristics of ship propulsion reactor and (3) thermo-hydraulic behaviors of the reactor container at the time of accident in a passively safe ship propulsion reactor. Development of a data base for ship propulsion reactor was attempted using previous experimental data on the thermo-hydraulic characteristics of the reactor in the institute in addition to the present results aiming to make general analytical evaluation for the safety of the engineering-simulation system for nuclear ship. A general data base was obtained by integrating the data list and the analytical program for static characteristics. A test equipment which allows to visualize the pulsating flow was produced and visualization experiments have started. (M.N.)

  1. Application study of fluid pressure energy recycling of decarbonisation process by C4H6O3 in ammonia synthesis systems by hydraulic turbochargers

    Science.gov (United States)

    Ji, Yunguang; Xu, Yangyang; Li, Hongtao; Oklejas, Michael; Xue, Shuqi

    2018-01-01

    A new type of hydraulic turbocharger energy recovery system was designed and applied in the decarbonisation process by propylene carbonate of a 100k tons ammonia synthesis system firstly in China. Compared with existing energy recovery devices, hydraulic turbocharger energy recovery system runs more smoothly, has lower failure rate, longer service life and greater comprehensive benefits due to its unique structure, simpler adjustment process and better adaptability to fluid fluctuation.

  2. Screening of Gas-Cooled Reactor Thermal-Hydraulic and Safety Analysis Tools and Experimental Database

    International Nuclear Information System (INIS)

    Lee, Won Jae; Kim, Min Hwan; Lee, Seung Wook

    2007-08-01

    This report is a final report of I-NERI Project, 'Screening of Gas-cooled Reactor Thermal Hydraulic and Safety Analysis Tools and Experimental Database 'jointly carried out by KAERI, ANL and INL. In this study, we developed the basic technologies required to develop and validate the VHTR TH/safety analysis tools and evaluated the TH/safety database information. The research tasks consist of; 1) code qualification methodology (INL), 2) high-level PIRTs for major nucleus set of events (KAERI, ANL, INL), 3) initial scaling and scoping analysis (ANL, KAERI, INL), 4) filtering of TH/safety tools (KAERI, INL), 5) evaluation of TH/safety database information (KAERI, INL, ANL) and 6) key scoping analysis (KAERI). The code qualification methodology identifies the role of PIRTs in the R and D process and the bottom-up and top-down code validation methods. Since the design of VHTR is still evolving, we generated the high-level PIRTs referencing 600MWth block-type GT-MHR and 400MWth pebble-type PBMR. Nucleus set of events that represents the VHTR safety and operational transients consists of the enveloping scenarios of HPCC (high pressure conduction cooling: loss of primary flow), LPCC/Air-Ingress (low pressure conduction cooling: loss of coolant), LC (load changes: power maneuvering), ATWS (anticipated transients without scram: reactivity insertion), WS (water ingress: water-interfacing system break) and HU (hydrogen-side upset: loss of heat sink). The initial scaling analysis defines dimensionless parameters that need to be reflected in mixed convection modeling and the initial scoping analysis provided the reference system transients used in the PIRTs generation. For the PIRTs phenomena, we evaluated the modeling capability of the candidate TH/safety tools and derived a model improvement need. By surveying and evaluating the TH/safety database information, a tools V and V matrix has been developed. Through the key scoping analysis using available database, the modeling

  3. Screening of Gas-Cooled Reactor Thermal-Hydraulic and Safety Analysis Tools and Experimental Database

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Jae; Kim, Min Hwan; Lee, Seung Wook (and others)

    2007-08-15

    This report is a final report of I-NERI Project, 'Screening of Gas-cooled Reactor Thermal Hydraulic and Safety Analysis Tools and Experimental Database 'jointly carried out by KAERI, ANL and INL. In this study, we developed the basic technologies required to develop and validate the VHTR TH/safety analysis tools and evaluated the TH/safety database information. The research tasks consist of; 1) code qualification methodology (INL), 2) high-level PIRTs for major nucleus set of events (KAERI, ANL, INL), 3) initial scaling and scoping analysis (ANL, KAERI, INL), 4) filtering of TH/safety tools (KAERI, INL), 5) evaluation of TH/safety database information (KAERI, INL, ANL) and 6) key scoping analysis (KAERI). The code qualification methodology identifies the role of PIRTs in the R and D process and the bottom-up and top-down code validation methods. Since the design of VHTR is still evolving, we generated the high-level PIRTs referencing 600MWth block-type GT-MHR and 400MWth pebble-type PBMR. Nucleus set of events that represents the VHTR safety and operational transients consists of the enveloping scenarios of HPCC (high pressure conduction cooling: loss of primary flow), LPCC/Air-Ingress (low pressure conduction cooling: loss of coolant), LC (load changes: power maneuvering), ATWS (anticipated transients without scram: reactivity insertion), WS (water ingress: water-interfacing system break) and HU (hydrogen-side upset: loss of heat sink). The initial scaling analysis defines dimensionless parameters that need to be reflected in mixed convection modeling and the initial scoping analysis provided the reference system transients used in the PIRTs generation. For the PIRTs phenomena, we evaluated the modeling capability of the candidate TH/safety tools and derived a model improvement need. By surveying and evaluating the TH/safety database information, a tools V and V matrix has been developed. Through the key scoping analysis using available database, the

  4. Bootstrap and Order Statistics for Quantifying Thermal-Hydraulic Code Uncertainties in the Estimation of Safety Margins

    Directory of Open Access Journals (Sweden)

    Enrico Zio

    2008-01-01

    Full Text Available In the present work, the uncertainties affecting the safety margins estimated from thermal-hydraulic code calculations are captured quantitatively by resorting to the order statistics and the bootstrap technique. The proposed framework of analysis is applied to the estimation of the safety margin, with its confidence interval, of the maximum fuel cladding temperature reached during a complete group distribution blockage scenario in a RBMK-1500 nuclear reactor.

  5. Investigations into the use of water glycol as the hydraulic fluid in a servo system

    International Nuclear Information System (INIS)

    Cole, G.V.

    1984-07-01

    The effects of water glycol on the performance of a hydraulic system and on the life of the system components have been investigated and a guide to the design of systems using water glycol is given. The dynamic performance of the system using water-glycol was compared with that using mineral oil, then the system was endurance tested to determine its service life. (author)

  6. Status and subjects of thermal-hydraulic analysis for next-generation LWRs with passive safety systems

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-03-01

    The present status and subjects on thermal-hydraulic analysis for next-generation light water reactors (LWRs) with passive safety systems were summarized based on survey results and discussion by subcommittee on improvement of reactor thermal-hydraulic analysis codes under nuclear code committee in Japan Atomic Energy Research Institute. This survey was performed to promote the research of improvement of reactor thermal-hydraulic analysis codes in future. In the first part of this report, the status and subjects on system analysis and those on evaluation of passive safety system performance are summarized for various types of reactor proposed before. In the second part, the status and subjects on multidimensional two-phase flow analysis are reviewed, since the multidimensional analysis was recognized as one of most important subjects through the investigation in the first part. Besides, databases for bubbly flow and annular dispersed flow were explored, those are needed to assess and verify each multidimensional analytical method. The contents in this report are the forefront of thermal-hydraulic analysis for LWRs and those include current findings for the development of multidimensional two-phase flow analytical method. Thus, we expect that the contents can offer various useful information against the improvement of reactor thermal-hydraulic analysis codes in future. (author)

  7. Automated System of Diagnostic Monitoring at Bureya HPP Hydraulic Engineering Installations: a New Level of Safety

    International Nuclear Information System (INIS)

    Musyurka, A. V.

    2016-01-01

    This article presents the design, hardware, and software solutions developed and placed in service for the automated system of diagnostic monitoring (ASDM) for hydraulic engineering installations at the Bureya HPP, and assuring a reliable process for monitoring hydraulic engineering installations. Project implementation represents a timely solution of problems addressed by the hydraulic engineering installation diagnostics section.

  8. Automated System of Diagnostic Monitoring at Bureya HPP Hydraulic Engineering Installations: a New Level of Safety

    Energy Technology Data Exchange (ETDEWEB)

    Musyurka, A. V., E-mail: musyurkaav@burges.rushydro.ru [Bureya HPP (a JSC RusGidro affiliate) (Russian Federation)

    2016-09-15

    This article presents the design, hardware, and software solutions developed and placed in service for the automated system of diagnostic monitoring (ASDM) for hydraulic engineering installations at the Bureya HPP, and assuring a reliable process for monitoring hydraulic engineering installations. Project implementation represents a timely solution of problems addressed by the hydraulic engineering installation diagnostics section.

  9. Final report of the 'Nordic thermal-hydraulic and safety network (NOTNET)' - Project

    International Nuclear Information System (INIS)

    Tuunanen, J.; Tuomainen, M.

    2005-04-01

    A Nordic network for thermal-hydraulics and nuclear safety research was started. The idea of the network is to combine the resources of different research teams in order to carry out more ambitious and extensive research programs than would be possible for the individual teams. From the very beginning, the end users of the research results have been integrated to the network. Aim of the network is to benefit the partners involved in nuclear energy in the Nordic Countries (power companies, reactor vendors, safety regulators, research units). First task within the project was to describe the resources (personnel, know-how, simulation tools, test facilities) of the various teams. Next step was to discuss with the end users about their research needs. Based on these steps, few most important research topics with defined goals were selected, and coarse road maps were prepared for reaching the targets. These road maps will be used as a starting point for planning the actual research projects in the future. The organisation and work plan for the network were established. National coordinators were appointed, as well as contact persons in each participating organisation, whether research unit or end user. This organisation scheme is valid for the short-term operation of NOTNET when only Nordic organisations take part in the work. Later on, it is possible to enlarge the network e.g. within EC framework programme. The network can now start preparing project proposals and searching funding for the first common research projects. (au)

  10. Safety System for Controlling Fluid Flow into a Suction Line

    Science.gov (United States)

    England, John Dwight (Inventor); Kelley, Anthony R. (Inventor); Cronise, Raymond J. (Inventor)

    2018-01-01

    A safety system includes a sleeve fitted within a pool's suction line at its inlet. The sleeve terminates with a plate that resides within the suction line. The plate has holes formed therethrough. A housing defining distinct channels is fitted in the sleeve so that the distinct channels lie within the sleeve. Each of the distinct channels has a first opening on one end thereof and a second opening on another end thereof. The second openings reside in the sleeve. The first openings are in fluid communication with the water in the pool, and are distributed around a periphery of an area of the housing that prevents coverage of all the first openings when a human interacts therewith. A first sensor is coupled to the sleeve to sense pressure therein, and a second pressure sensor is coupled to the plate to sense pressure in one of the plates' holes.

  11. Breakdown of Preservative Fluid MIL-PRF-46170 in Aircraft Hydraulic Systems

    National Research Council Canada - National Science Library

    Moorman, Jeffrey

    2001-01-01

    .... Additional information obtained from outside sources is also summarized for background. Laboratory pump testing showed rapid filter dogging with small amounts of preservative fluid (MU-PRF-46l70) in the system...

  12. Challenges in coupled thermal-hydraulics and neutronics simulations for LWR safety analysis

    International Nuclear Information System (INIS)

    Ivanov, Kostadin; Avramova, Maria

    2007-01-01

    The simulation of nuclear power plant accident conditions requires three-dimensional (3D) modeling of the reactor core to ensure a realistic description of physical phenomena. The operational flexibility of Light Water Reactor (LWR) plants can be improved by utilizing accurate 3D coupled neutronics/thermal-hydraulics calculations for safety margins evaluations. There are certain requirements to the coupling of thermal-hydraulic system codes and neutron-kinetics codes that ought to be considered. The objective of these requirements is to provide accurate solutions in a reasonable amount of CPU time in coupled simulations of detailed operational transient and accident scenarios. These requirements are met by the development and implementation of six basic components of the coupling methodologies: ways of coupling (internal or external coupling); coupling approach (integration algorithm or parallel processing); spatial mesh overlays; coupled time-step algorithms; coupling numerics (explicit, semi-implicit and implicit schemes); and coupled convergence schemes. These principles of the coupled simulations are discussed in details along with the scientific issues associated with the development of appropriate neutron cross-section libraries for coupled code transient modeling. The current trends in LWR nuclear power generation and regulation as well as the design of next generation LWR reactor concepts along with the continuing computer technology progress stimulate further development of these coupled code systems. These efforts have been focused towards extending the analysis capabilities as well as refining the scale and level of detail of the coupling. This article analyses the coupled phenomena and modeling challenges on both global (assembly-wise) and local (pin-wise) levels. The issues related to the consistent qualification of coupled code systems as well as their application to different types of LWR transients are presented. Finally, the advances in numerical

  13. Experimental studies of thermo-hydraulic processes during passive safety systems operation in new WWER NPP projects

    International Nuclear Information System (INIS)

    Morozov, A.V.; Remizov, O.V.; Kalyakin, D.S.

    2014-01-01

    The results of experimental study of thermal-hydraulic processes during operation of the passive safety systems of WWER reactors of new generation are given. The interaction processes of counter flows of saturated steam and cold water in vertical steam-line of the auxiliary passive core reflood system from secondary hydraulic accumulator are studied. The peculiarities of undeveloped boiling on single horizontal tube heating by steam and steam-gas mixture, which is character for WWER steam generator condensing mode, are investigated [ru

  14. Assessment and Application of the ROSE Code for Reactor Outage Thermal-Hydraulic and Safety Analysis

    International Nuclear Information System (INIS)

    Liang, Thomas K.S.; Ko, F.-K.; Dai, L.-C.

    2001-01-01

    The currently available tools, such as RELAP5, RETRAN, and others, cannot easily and correctly perform the task of analyzing the system behavior during plant outages. Therefore, a medium-sized program aiming at reactor outage simulation and evaluation, such as midloop operation (MLO) with loss of residual heat removal (RHR), has been developed. Important thermal-hydraulic processes involved during MLO with loss of RHR can be properly simulated by the newly developed reactor outage simulation and evaluation (ROSE) code. The two-region approach with a modified two-fluid model has been adopted to be the theoretical basis of the ROSE code.To verify the analytical model in the first step, posttest calculations against the integral midloop experiments with loss of RHR have been performed. The excellent simulation capacity of the ROSE code against the Institute of Nuclear Energy Research Integral System Test Facility test data is demonstrated. To further mature the ROSE code in simulating a full-sized pressurized water reactor, assessment against the WGOTHIC code and the Maanshan momentary-loss-of-RHR event has been undertaken. The successfully assessed ROSE code is then applied to evaluate the abnormal operation procedure (AOP) with loss of RHR during MLO (AOP 537.4) for the Maanshan plant. The ROSE code also has been successfully transplanted into the Maanshan training simulator to support operator training. How the simulator was upgraded by the ROSE code for MLO will be presented in the future

  15. Extensive use of computational fluid dynamics in the upgrading of hydraulic turbines

    Energy Technology Data Exchange (ETDEWEB)

    Sabourin, M.; De Henau, V. [GEC Alsthom Electromechanical Inc., Tracy, PQ (Canada); Eremeef, R. [GEC Alsthom Neyrpic, Grenoble (France)

    1995-12-31

    The use of computational fluid flow dynamics (CFD) and the Navier Stokes equations by GEC Alsthom for turbine rehabilitation were discussed. The process of runner rehabilitation was discussed from a fluid flow perspective, which accounts for the spiral case-distributor set and draft tube. The Kootenay turbine rehabilitation was described with regard to it spiral case and stay vane. The numerical analysis used to model upstream components was explained. The influence of draft tube effects was emphasized as an important efficiency factor. The differences between draft tubes at Sir Adam Beck 2 and La Grande 2 were discussed. Computational fluid flow modelling was claimed to have produced global performance enhancements in a reasonably short time, and at a reasonable cost. 6 refs., 6 figs., 4 tabs.

  16. Basic hydraulics

    CERN Document Server

    Smith, P D

    1982-01-01

    BASIC Hydraulics aims to help students both to become proficient in the BASIC programming language by actually using the language in an important field of engineering and to use computing as a means of mastering the subject of hydraulics. The book begins with a summary of the technique of computing in BASIC together with comments and listing of the main commands and statements. Subsequent chapters introduce the fundamental concepts and appropriate governing equations. Topics covered include principles of fluid mechanics; flow in pipes, pipe networks and open channels; hydraulic machinery;

  17. The trade-off between safety and efficiency in hydraulic architecture in 31 woody species in a karst area.

    Science.gov (United States)

    Fan, Da-Yong; Jie, Sheng-Lin; Liu, Chang-Cheng; Zhang, Xiang-Ying; Xu, Xin-Wu; Zhang, Shou-Ren; Xie, Zong-Qiang

    2011-08-01

    plants. Whole-plant hydraulic adjustment may decouple the trade-off relationship between safety and efficiency at the branch level.

  18. Thermo-hydraulic characteristics of ship propulsion reactor in the conditions of ship motions and safety assessment

    International Nuclear Information System (INIS)

    Kobayashi, Michiyuki; Murata, Hiroyuki; Sawada, Kenichi; Inasaka, Fujio; Aya, Izuo; Shiozaki, Koki

    1999-01-01

    By inputting the experimental data, information and others on thermo-hydraulic characteristics of integrated ship propulsion reactor accumulated hitherto by the Ship Research Institute and some recent cooperation results into the nuclear ship engineering simulation system, it was conducted not only to contribute an improvement study on next ship reactor by executing general analysis and evaluation on motion characteristics under ship body motion conditions, safety at accidents, and others of the integrated ship reactor but also to investigate and prepare some measures to apply fundamental experiment results based on obtained here information to safety countermeasure of the nuclear ships. In 1997 fiscal year, on safety of the integrated ship propulsion reactor loading nuclear ship, by adding experimental data on unstable flow analysis and information on all around of the analysis to general data base fundamental program, development to intellectual data base program was intended; on effect of pulsation flow on thermo-hydraulic characteristics of ship propulsion reactor; after pulsation flow visualization experiment, experimental equipment was reconstructed into heat transfer type to conduct numerical analysis of pulsation flow by confirming validity of numerical analysis code under comparison with the visualization experiment results; and on thermo-hydraulic behavior in storage container at accident of active safety type ship propulsion reactor; a flashing vibration test using new apparatus finished on its higher pressurization at last fiscal year to examine effects of each parameter such as radius and length of exhausting nozzle and pool water temperature. (G.K.)

  19. Experimental and modeling hydraulic studies of foam drilling fluid flowing through vertical smooth pipes

    Directory of Open Access Journals (Sweden)

    Amit Saxena

    2017-06-01

    Full Text Available Foam has emerged as an efficient drilling fluid for the drilling of low pressure, fractured and matured reservoirs because of its the ability to reduce formation damage, fluid loss, differential sticking etc. However the compressible nature along with its complicated rheology has made its implementation a multifaceted task. Knowledge of the hydrodynamic behavior of drilling fluid within the borehole is the key behind successful implementation of drilling job. However, little effort has been made to develop the hydrodynamic models for the foam flowing with cuttings through pipes of variable diameter. In the present study, hydrodynamics of the foam fluid was investigated through the vertical smooth pipes of different pipe diameters, with variable foam properties in a flow loop system. Effect of cutting loading on pressure drop was also studied. Thus, the present investigation estimates the differential pressure loss across the pipe. The flow loop permits foam flow through 25.4 mm, 38.1 mm and 50.8 mm diameter pipes. The smaller diameter pipes are used to replicate the annular spaces between the drill string and wellbore. The developed model determines the pressure loss along the pipe and the results are compared with a number of existing models. The developed model is able to predict the experimental results more accurately.

  20. Control system for the feed of pressurized fluid in a hydraulic circuit as a function of the state of the locking or unlocking of two mechanical organs

    International Nuclear Information System (INIS)

    Huet, Y.; Perichon, C.

    1985-01-01

    The control system comprises two hydraulic cylinders of which rods are integral with the mechanical organs. The piston of the first cylinder separates the chamber of this one in two parts. The piston of the second cylinder separates its chamber in three parts. The inlet chamber of the two cylinders are connected to pressurized fluid feed pipes, and the outlet chambers to a depressurization pipe. According to the position of the piston depending itself on the state of locking or unlocking of the rods, an interconnection pipe and a feed pipe of the pressurized fluid hydraulic circuit communicate with a chamber or another one. The feed of the hydraulic circuit is possible only the two rods are unlocked. The invention applies more particularly to the feed of the control circuit of an emergency seal of the primary pump of a pressurized water nuclear reactor [fr

  1. Integrated Experimental and Computational Study of Hydraulic Fracturing and the Use of Alternative Fracking Fluids

    Science.gov (United States)

    Viswanathan, H.; Carey, J. W.; Karra, S.; Porter, M. L.; Rougier, E.; Zhang, D.; Makedonska, N.; Middleton, R. S.; Currier, R.; Gupta, R.; Lei, Z.; Kang, Q.; O'Malley, D.; Hyman, J.

    2014-12-01

    Shale gas is an unconventional fossil energy resource that is already having a profound impact on US energy independence and is projected to last for at least 100 years. Production of methane and other hydrocarbons from low permeability shale involves hydrofracturing of rock, establishing fracture connectivity, and multiphase fluid-flow and reaction processes all of which are poorly understood. The result is inefficient extraction with many environmental concerns. A science-based capability is required to quantify the governing mesoscale fluid-solid interactions, including microstructural control of fracture patterns and the interaction of engineered fluids with hydrocarbon flow. These interactions depend on coupled thermo-hydro-mechanical-chemical (THMC) processes over scales from microns to tens of meters. Determining the key mechanisms in subsurface THMC systems has been impeded due to the lack of sophisticated experimental methods to measure fracture aperture and connectivity, multiphase permeability, and chemical exchange capacities at the high temperature, pressure, and stresses present in the subsurface. This project uses innovative high-pressure microfluidic and triaxial core flood experiments on shale to explore fracture-permeability relations and the extraction of hydrocarbon. These data are integrated with simulations including lattice Boltzmann modeling of pore-scale processes, finite-element/discrete element models of fracture development in the near-well environment, discrete-fracture modeling of the reservoir, and system-scale models to assess the economics of alternative fracturing fluids. The ultimate goal is to make the necessary measurements to develop models that can be used to determine the reservoir operating conditions necessary to gain a degree of control over fracture generation, fluid flow, and interfacial processes over a range of subsurface conditions.

  2. Thermal-hydraulic codes validation for safety analysis of NPPs with RBMK

    International Nuclear Information System (INIS)

    Brus, N.A.; Ioussoupov, O.E.

    2000-01-01

    This work is devoted to validation of western thermal-hydraulic codes (RELAP5/MOD3 .2 and ATHLET 1.1 Cycle C) in application to Russian designed light water reactors. Such validation is needed due to features of RBMK reactor design and thermal-hydraulics in comparison with PWR and BWR reactors, for which these codes were developed and validated. These validation studies are concluded with a comparison of calculation results of modeling with the thermal-hydraulics codes with the experiments performed earlier using the thermal-hydraulics test facilities with the experimental data. (authors)

  3. Fluid dynamics of acoustic and hydrodynamic cavitation in hydraulic power systems

    OpenAIRE

    Ferrari, A.

    2017-01-01

    Cavitation is the transition from a liquid to a vapour phase, due to a drop in pressure to the level of the vapour tension of the fluid. Two kinds of cavitation have been reviewed here: acoustic cavitation and hydrodynamic cavitation. As acoustic cavitation in engineering systems is related to the propagation of waves through a region subjected to liquid vaporization, the available expressions of the sound speed are discussed. One of the main effects of hydrodynamic cavitation in the nozzles ...

  4. Are separate-phase thermal-hydraulic models better than mixture-fluid approaches? It depends. Rather not

    International Nuclear Information System (INIS)

    Hoeld, A.

    2004-01-01

    The thermal-hydraulic theory of single- and especially two-phase flow systems used for plant transient analysis is dominated by separate-phase models. The corresponding mostly very comprehensive codes (TRAC, RELAP, CATHARE, ATHLET etc.) are looked as to be by far more efficient than a 3 eq. mixture-fluid approach and code also if they show deficiencies in describing flow situations within inner loops as for example the distribution into parallel channels (and thus the simulation of 3D thermal-hydraulic phenomena). This may be justified if comparing them to the very simple 'homogeneous equilibrium models (HEM)', but not if looking to the more refined non-homogeneous 'separate-region' mixture-fluid approaches based on appropriate drift-flux correlation packages which can have, on the contrary, enormous advantages with respect to such separate-phase models. Especially if comparing the basic (and starting) eqs. of such theoretical models of both types the differences are remarkable. Single-phase and mixture-fluid models start from genuine conservation eqs. for mass, energy and momentum, demanding (in case of two-phase flow) additionally an adequate drift flux package (in order to get a relation for a fourth independent variable), a heat transfer coefficients package (over the whole range of the possible fields of application) and correlations for single- and two-phase friction. The other types of models are looking at each phase separately with corresponding 'field' eqs. for each phase, connected by exchange (=closure) terms which substitute the classical constitutive packages for drift, heat transfer and friction. That the drift-flux, heat transfer into a coolant channel and friction along a wall and between the phases is described better by a separate-phase approach is at least doubtful. The corresponding mixture-fluid correlations are based over a wide range on a treasure of experience and measurements, their pseudo-stationary treatment can (due to their small time

  5. Hydraulic patterns and safety margins, from stem to stomata, in three eastern US tree species

    Science.gov (United States)

    D.M. Johnson; K.A. McCulloh; F.C. Meinzer; D.R. Woodruff; D.M. Eissenstat

    2011-01-01

    Adequate water transport is necessary to prevent stomatal closure and allow for photosynthesis. Dysfunction in the water transport pathway can result in stomatal closure, and can be deleterious to overall plant health and survival. Although much is known about small branch hydraulics, little is known about the coordination of leaf and stem hydraulic function....

  6. Design and verification of additional filtration for the application of ecological transmission and hydraulic fluids in tractorc

    Directory of Open Access Journals (Sweden)

    Pavel Máchal

    2013-01-01

    Full Text Available This contribution presents the design and function verification of additional filtration. It is intended for the common transmission and hydraulic oil filling of tractors. The main role of this filtration concept is to ensure a high level of oil cleanness as a condition for the application of ecologic fluids in tractors. The next one is to decrease the wear of lubricated tractor components, the degradation of oil and eventually to extend the interval of oil change. The designed additional filtering is characterized by ease installation through the use of quick couplings and hoses to the external hydraulic circuit. Therefore, the filtration is suitable for various tractor types. Filter element has been designed with the filter ability 1micron and the ability to separate to 0.5 dm3 of water from oil. Function of additional filtration was verified during the 150 engine hours of tractor operation. During this time period the oil contamination was evaluated on the basis of chemical elements content such as Fe, Cu, Si, Al, Ni, Mo and Cr. The additive concentration was evaluated on the basis of chemical elements content such as Ca, P and Zn. During the test operation of tractor the concentration decrease of chemical elements reached the values 25.53 % (Fe, 23.53 % (Si, 25 % (Al and 5.5 % (Cu. The decrease of additive concentration reached only medium level (6.6 %. Therefore, the designed additional filtration doesn’t remove additives from oil. Based on the evaluation of the content of chemical elements (that representing contamination and additives, we can say that the designed filtering method is suitable for use in agricultural tractors.

  7. Twenty-third water reactor safety information meeting: Volume 1, plenary session, high burnup fuel behavior, thermal hydraulic research. Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1996-03-01

    This three-volume report contains papers presented at the Twenty- Third Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 23-25, 1995. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Italy, Japan, Norway, Russia, Sweden, and Switzerland. This document, Volume 1, present topics on High Burnup Fuel Behavior, Thermal Hydraulic Research, and Plenary Session topics. Individual papers have been cataloged separately.

  8. Twenty-third water reactor safety information meeting: Volume 1, plenary session, high burnup fuel behavior, thermal hydraulic research. Proceedings

    International Nuclear Information System (INIS)

    Monteleone, S.

    1996-03-01

    This three-volume report contains papers presented at the Twenty- Third Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 23-25, 1995. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Italy, Japan, Norway, Russia, Sweden, and Switzerland. This document, Volume 1, present topics on High Burnup Fuel Behavior, Thermal Hydraulic Research, and Plenary Session topics. Individual papers have been cataloged separately

  9. Literature survey of heat transfer and hydraulic resistance of water, carbon dioxide, helium and other fluids at supercritical and near-critical pressures

    Energy Technology Data Exchange (ETDEWEB)

    Pioro, I.L.; Duffey, R.B

    2003-04-01

    This survey consists of 430 references, including 269 Russian publications and 161 Western publications devoted to the problems of heat transfer and hydraulic resistance of a fluid at near-critical and supercritical pressures. The objective of the literature survey is to compile and summarize findings in the area of heat transfer and hydraulic resistance at supercritical pressures for various fluids for the last fifty years published in the open Russian and Western literature. The analysis of the publications showed that the majority of the papers were devoted to the heat transfer of fluids at near-critical and supercritical pressures flowing inside a circular tube. Three major working fluids are involved: water, carbon dioxide, and helium. The main objective of these studies was the development and design of supercritical steam generators for power stations (utilizing water as a working fluid) in the 1950s, 1960s, and 1970s. Carbon dioxide was usually used as the modeling fluid due to lower values of the critical parameters. Helium, and sometimes carbon dioxide, were considered as possible working fluids in some special designs of nuclear reactors. (author)

  10. Preliminary fluid channel design and thermal-hydraulic analysis of glow discharge cleaning permanent electrode

    Energy Technology Data Exchange (ETDEWEB)

    Cai, Lijun, E-mail: cailj@swip.ac.cn [Southwestern Institute of Physics, Chengdu (China); Lin, Tao; Wang, Yingqiao; Wang, Mingxu [Southwestern Institute of Physics, Chengdu (China); Maruyama, So; Yang, Yu; Kiss, Gabor [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France)

    2016-11-01

    Highlights: • The plasma facing closure cap has to survive after 30,000 thermal heat load cycles. • 0.35 MW/m2 radiation heat load plus nuclear heat load are very challenging for stainless steel. • Multilayer structure has been designed by using advanced welding and drilling technology to solve the neutron heating problem. • Accurate volumetric load application in analysis model by CFX has been mastered. - Abstract: Glow discharge cleaning (GDC) shall be used on ITER device to reduce and control impurity and hydrogenic fuel out-gassing from in-vessel plasma facing components. After first plasma, permanent electrode (PE) will be used to replace Temporary Electrode (TE) for subsequent operation. Two fundamental scenarios i.e., GDC and Plasma Operation State (POS) should be considered for electrode design, which requires the heat load caused by plasma radiation and neutron heating must be taken away by cooling water flowing inside the electrode. In this paper, multilayer cooling channels inside PE are preliminarily designed, and snakelike route in each layer is adopted to improve the heat exchange. Detailed thermal-hydraulic analyses have been done to validate the design feasibility or rationality. The analysis results show that during GDC the cooling water inlet and outlet temperature difference is far less than the allowable temperature rise under water flow rate 0.15 kg/s compromised by many factors. For POS, the temperature rise and pressure drop are within the design goals, but high thermal stress occurs on the front surface of closure cap of electrode. After several iterations of optimization of the closure cap, the equivalent strain range after 30,000 loading cycles for POS is well below 0.3% design goals.

  11. Information and dialogue process on safety and environmental effects of the hydraulic fracturing technology; Der Informations- und Dialogprozess zur Sicherheit und Umweltvertraeglichkeit der Fracking-Technologie

    Energy Technology Data Exchange (ETDEWEB)

    Borchardt, Dietrich; Richter, Sandra [Helmholtz-Zentrum fuer Umweltforschung - UFZ, Magdeburg (Germany); Ewen, Christoph [team ewen, Darmstadt (Germany); Hammerbacher, Ruth [hammerbacher gmbh - beratung und projekte, Osnabrueck (Germany)

    2012-10-15

    After the big success of hydraulic fracturing in the USA, natural gas utilities are now planning natural gas production from nonconventional deposits (shale gas, coal seam gas) by hydraulic fracturing also in Germany. In order to calm public fears, an 'information and dialogue process on safety and environmental effects of the hydraulic fracturing technology' was initiated. A risk study carried out by a team of neutral experts gives recommendations for a well-founded, careful and realistic discussion of the environmental compatibility of hydraulic fracturing.

  12. Information and dialogue process on safety and environmental effects of the hydraulic fracturing technology; Der Informations- und Dialogprozess zur Sicherheit und Umweltvertraeglichkeit der Fracking-Technologie

    Energy Technology Data Exchange (ETDEWEB)

    Borchardt, Dietrich; Richter, Sandra [Helmholtz-Zentrum fuer Umweltforschung - UFZ, Magdeburg (Germany); Ewen, Christoph [team ewen, Darmstadt (Germany); Hammerbacher, Ruth [hammerbacher gmbh - beratung und projekte, Osnabrueck (Germany)

    2012-10-15

    After the big success of hydraulic fracturing in the USA, natural gas utilities are now planning natural gas production from nonconventional deposits (shale gas, coal seam gas) by hydraulic fracturing also in Germany. In order to calm public fears, an 'information and dialogue process on safety and environmental effects of the hydraulic fracturing technology' was initiated. A risk study carried out by a team of neutral experts gives recommendations for a well-founded, careful and realistic discussion of the environmental compatibility of hydraulic fracturing.

  13. Selection and use of fire-resistant hydraulic fluids for underground mining equipment. [Oil-in-water emulsions; water-in-oil emulsions; phosphate esters; chlorinated hydrocarbons

    Energy Technology Data Exchange (ETDEWEB)

    Harrison, A J

    1981-02-01

    During the initial introduction of fire-resistant fluids to the Canadian underground mining industry, all hydraulic systems for which they were being considered were originally designed for operation with mineral oil. This meant that each system had to be individually examined and assessed with regard to its suitability in terms of acceptable component life and operation, at the same time as the selection of a fluid was being undertaken. Fluid selection by cost differential, toxicity content and fire resistancy was narrowed to types HFB and HFC, with HFB water-in-oil emulsion being the preferred fluid based on performance characteristics. By incorporating British mining industry experience and superior fluid types with practical trials, it was found that by modifing the design of some systems and slightly derating the operational parameters of individual components, it was possible to obtain a system performance comparable to that obtained when mineral oil was being used.

  14. Streaming Potential Modeling to Understand the Identification of Hydraulically Active Fractures and Fracture-Matrix Fluid Interactions Using the Self-Potential Method

    Science.gov (United States)

    Jougnot, D.; Roubinet, D.; Linde, N.; Irving, J.

    2016-12-01

    Quantifying fluid flow in fractured media is a critical challenge in a wide variety of research fields and applications. To this end, geophysics offers a variety of tools that can provide important information on subsurface physical properties in a noninvasive manner. Most geophysical techniques infer fluid flow by data or model differencing in time or space (i.e., they are not directly sensitive to flow occurring at the time of the measurements). An exception is the self-potential (SP) method. When water flows in the subsurface, an excess of charge in the pore water that counterbalances electric charges at the mineral-pore water interface gives rise to a streaming current and an associated streaming potential. The latter can be measured with the SP technique, meaning that the method is directly sensitive to fluid flow. Whereas numerous field experiments suggest that the SP method may allow for the detection of hydraulically active fractures, suitable tools for numerically modeling streaming potentials in fractured media do not exist. Here, we present a highly efficient two-dimensional discrete-dual-porosity approach for solving the fluid-flow and associated self-potential problems in fractured domains. Our approach is specifically designed for complex fracture networks that cannot be investigated using standard numerical methods due to computational limitations. We then simulate SP signals associated with pumping conditions for a number of examples to show that (i) accounting for matrix fluid flow is essential for accurate SP modeling and (ii) the sensitivity of SP to hydraulically active fractures is intimately linked with fracture-matrix fluid interactions. This implies that fractures associated with strong SP amplitudes are likely to be hydraulically conductive, attracting fluid flow from the surrounding matrix.

  15. Life Cycle Assessment of age-related environmental impact of biogenic hydraulic fluids; Life Cycle Assessment der alterungsbedingten Umweltvertraeglichkeit biogener Hydraulik-Schmierstoffe

    Energy Technology Data Exchange (ETDEWEB)

    Bressling, Jana

    2012-07-01

    Biogenic hydraulic fluids, based on synthetic esters (category: HEES), have an excellent environmental profile in the unused state, so that they are typically classified into water hazard class 1 or as ''not hazardous to water''. During storage at room temperature and tribological application, occurring chemical and toxicological changes take no account in the classification of lubricants until now. However, the ageing and oxidation stability gets increasing importance, since it determines the service life of lubricants in tribological systems in addition to the storage time. Since it always comes to direct and uncontrolled entries into the environment in case of accidents or hydraulic leaks, it is essential to assess whether there is an environmental hazard by waste oils. With an increased use of biogenic hydraulic fluids in environmentally sensitive areas, thus the need for an appropriate monitoring and assessment approach as part of a Life Cycle Assessment (LCA). The aquatic and miniaturised test procedures applied in this work with the Water Soluble Fraction (WSF) concept, allows a simple and quick screening of age-related ecotoxic potential of lubricants by oxidative processes and tribological application. For detection of genotoxic potential the umu-test is a suitable indicator test to detect geno- and cytotoxic effects by oxidative reactions. The determination of biodegradability is essential for the assessment of the environmental impact of hydraulic fluids. The optimised biodegradability test system ''O2/CO2-Headspace Test'' has proved itself as a suitable procedure for the investigation of biogenic lubricants within the scope of a LCA and shows therefore a comparable method of the required test procedures for the assignment of ecolabels. In addition, the combination of biological test procedures and chemical analysis allows a comprehensive investigation of effects and causes of age-related changes of hydraulic

  16. Improvement of auditing technology of safety analysis through thermal-hydraulic separate effect tests

    Energy Technology Data Exchange (ETDEWEB)

    No, Hee Cheon; Moon, Young Min; Lee, Dong Won; Lee, Sang Ik; Kim, Eung Soo; Yeom, Keum Soo [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2002-03-15

    The objective of the present research is to perform the separate effect tests and to assess the RELAP5/MOD3.2 code for the analysis of thermal-hydraulic behavior in the reactor coolant system and the improvement of the auditing technology of safety analysis. Three Separate Effect Tests (SETs) are the reflux condensation in the U-tube, the direct contact condensation in the hot-leg and the mixture level buildup in the pressurizer. The experimental data and the empirical correlations are obtained through SETs. On the ases of the three SET works, models in RELAP5 are modified and improved, which are compared with the data. The Korea Standard Nuclear Power Plant (KSNP) are assessed using the modified RELAP5. In the reflux condensation test, the data of heat transfer coefficients and flooding are obtained and the condensation models are modified using the non-iterative model, as results, modified code better predicts the data. In the direct contact condensation test, the data of heat transfer coefficients are obtained for the cocurrent and countercurrent flow between the mixture gas and the water in condition of horizontal stratified flow. Several condensation and friction models are modified, which well predict the present data. In the mixture level test, the data for the mixture level and the onset of water draining into the surge line are obtained. The standard RELAP5 over-predicts the mixture level and the void fraction in the pressurizer. Simple modification of model related to the pool void fraction is suggested. The KSNP is assessed using the standard and the modified RELAP5 resulting from the experimental and code works for the SETs. In case of the pressurizer manway opening with available secondary side of the steam generators, the modified code predicts that the collapsed level in the pressurizer is little accumulated. The presence and location of the opening and the secondary condition of the steam generators have an effect on the coolant inventory. The

  17. Status and topics of thermal-hydraulic analysis for next-generation LWRs with passive safety systems

    International Nuclear Information System (INIS)

    Aritomi, Masanori; Ohnuki, Akira; Arai, Kenji; Kikuta, Michitaka; Yonomoto, Taisuke; Araya, Fumimasa; Akimoto, Hajime

    1999-01-01

    For increasing of electric power demand and reducing of carbon dioxide exhaust in the 21st century, studies of the next-generation light water reactor (LWR) with passive safety systems are developing in the world: AP-600 (by Westing House Co.); SBWR (by General Electric Co.); SWR1000 (by Siemens Co.); NP21 (by Mitsubishi Heavy Industry Co., et al.); JPSR (by JAERI). The passive equipment using natural circulation and natural convection are installed in the passive safety system, instead of active safety equipment, such as pumps, etc. It remains still as a important issue, however, to verify the reliability on the functions of the passive equipment, since that the driving forces of the passive equipment are small at comparison with the active safety equipment. The various subjects of thermal-hydraulic analysis for the next-generation light water reactors, such as temperature stratification in the passive safety systems, vapor condensation in the mixture of non-condensable gases and the interactions of the passive safety system with the primary cooling system, are illustrated and discussed in the paper. (M. Suetake)

  18. Advances in thermal hydraulic and neutronic simulation for reactor analysis and safety

    International Nuclear Information System (INIS)

    Tentner, A.M.; Blomquist, R.N.; Canfield, T.R.; Ewing, T.F.; Garner, P.L.; Gelbard, E.M.; Gross, K.C.; Minkoff, M.; Valentin, R.A.

    1993-01-01

    This paper describes several large-scale computational models developed at Argonne National Laboratory for the simulation and analysis of thermal-hydraulic and neutronic events in nuclear reactors and nuclear power plants. The impact of advanced parallel computing technologies on these computational models is emphasized

  19. The potential for spills and leaks of hydraulic fracturing related fluids on well sites and from road incidents.

    Science.gov (United States)

    Clancy, Sarah; Worrall, Fred; Davies, Richard; Gluyas, Jon

    2017-04-01

    recovered. The most common cause of leakage each year is equipment failure; these results highlight the need for good regulation and maintenance onsite. The UK's Institute of Directors suggests several shale gas production scenarios for the UK and how this would influence truck movement. One of their scenarios suggests the development of well pads with 10-wells and 40 laterals (one well pad with 10 well each with 4 laterals). This type of well pad would be projected to use 544,000 m3 of water, which would generate between 11155-31288 truck movements over 20 years, or 6.1-17.1 per day if averaged over 5 years. Dairy farmers in the UK produce 11 million m3 of milk a year, which if the tanker has a capacity of 30 m3, equates to approximately 366667 milk tanker journeys a year. This study assesses the number of road incidents and milk tanker spills and predicts the likelihood of such events for fluids involved in hydraulic fracturing.

  20. Development of best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B. D.; Lee, W. J.; Lim, H. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-04-15

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool, i.e. RELAP5/MOD3. This scope of project is first step of the whole project, thus focus to the establishment of improvement area. The study was performed by reconsideration of the previous code assessment works and investigation of AECL design analysis tools. In order to identify the thermal hydraulic phenomena for events, the whole system of CANDU plant was divided into main functional systems and subcomponents. Each phenomena was addressed to the each subcomponent. FinaIly improvement areas of model development for auditing tool were established based on the identified phenomena.

  1. Evaluation of operational safety at Babcock and Wilcox Plants: Volume 2, Thermal-hydraulic results

    International Nuclear Information System (INIS)

    Wheatley, P.D.; Davis, C.B.; Callow, R.A.; Fletcher, C.D.; Dobbe, C.A.; Beelman, R.J.

    1987-11-01

    The Nuclear Regulatory Commission has initiated a research program to develop a methodology to assess the operational performance of Babcock and Wilcox plants and to apply this methodology on a trial basis. The methodology developed for analyzing Babcock and Wilcox plants integrated methods used in both thermal-hydraulics and human factors and compared results with information used in the assessment of risk. The integrated methodology involved an evaluation of a selected plant for each pressurized water reactor vendor during a limited number of transients. A plant was selected to represent each vendor, and three transients were identified for analysis. The plants were Oconee Unit 1 for Babcock and Wilcox, H.B. Robinson Unit 2 for Westinghouse, and Calvert Cliffs Unit 1 for Combustion Engineering. The three transients were a complete loss of all feedwater, a small-break loss-of-coolant accident, and a steam-generator overfill with auxiliary feedwater. Included in the integrated methodology was an assessment of the thermal-hydraulic behavior, including event timing, of the plants during the three transients. Thermal-hydraulic results are presented in this volume (Volume 2) of the report. 26 refs., 30 figs., 7 tabs

  2. International benchmark study of advanced thermal hydraulic safety analysis codes against measurements on IEA-R1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hainoun, A., E-mail: pscientific2@aec.org.sy [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Doval, A. [Nuclear Engineering Department, Av. Cmdt. Luis Piedrabuena 4950, C.P. 8400 S.C de Bariloche, Rio Negro (Argentina); Umbehaun, P. [Centro de Engenharia Nuclear – CEN, IPEN-CNEN/SP, Av. Lineu Prestes 2242-Cidade Universitaria, CEP-05508-000 São Paulo, SP (Brazil); Chatzidakis, S. [School of Nuclear Engineering, Purdue University, West Lafayette, IN 47907 (United States); Ghazi, N. [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Park, S. [Research Reactor Design and Engineering Division, Basic Science Project Operation Dept., Korea Atomic Energy Research Institute (Korea, Republic of); Mladin, M. [Institute for Nuclear Research, Campului Street No. 1, P.O. Box 78, 115400 Mioveni, Arges (Romania); Shokr, A. [Division of Nuclear Installation Safety, Research Reactor Safety Section, International Atomic Energy Agency, A-1400 Vienna (Austria)

    2014-12-15

    Highlights: • A set of advanced system thermal hydraulic codes are benchmarked against IFA of IEA-R1. • Comparative safety analysis of IEA-R1 reactor during LOFA by 7 working teams. • This work covers both experimental and calculation effort and presents new out findings on TH of RR that have not been reported before. • LOFA results discrepancies from 7% to 20% for coolant and peak clad temperatures are predicted conservatively. - Abstract: In the framework of the IAEA Coordination Research Project on “Innovative methods in research reactor analysis: Benchmark against experimental data on neutronics and thermal hydraulic computational methods and tools for operation and safety analysis of research reactors” the Brazilian research reactor IEA-R1 has been selected as reference facility to perform benchmark calculations for a set of thermal hydraulic codes being widely used by international teams in the field of research reactor (RR) deterministic safety analysis. The goal of the conducted benchmark is to demonstrate the application of innovative reactor analysis tools in the research reactor community, validation of the applied codes and application of the validated codes to perform comprehensive safety analysis of RR. The IEA-R1 is equipped with an Instrumented Fuel Assembly (IFA) which provided measurements for normal operation and loss of flow transient. The measurements comprised coolant and cladding temperatures, reactor power and flow rate. Temperatures are measured at three different radial and axial positions of IFA summing up to 12 measuring points in addition to the coolant inlet and outlet temperatures. The considered benchmark deals with the loss of reactor flow and the subsequent flow reversal from downward forced to upward natural circulation and presents therefore relevant phenomena for the RR safety analysis. The benchmark calculations were performed independently by the participating teams using different thermal hydraulic and safety

  3. Implicit Treatment of Technical Specification and Thermal Hydraulic Parameter Uncertainties in Gaussian Process Model to Estimate Safety Margin

    Directory of Open Access Journals (Sweden)

    Douglas A. Fynan

    2016-06-01

    Full Text Available The Gaussian process model (GPM is a flexible surrogate model that can be used for nonparametric regression for multivariate problems. A unique feature of the GPM is that a prediction variance is automatically provided with the regression function. In this paper, we estimate the safety margin of a nuclear power plant by performing regression on the output of best-estimate simulations of a large-break loss-of-coolant accident with sampling of safety system configuration, sequence timing, technical specifications, and thermal hydraulic parameter uncertainties. The key aspect of our approach is that the GPM regression is only performed on the dominant input variables, the safety injection flow rate and the delay time for AC powered pumps to start representing sequence timing uncertainty, providing a predictive model for the peak clad temperature during a reflood phase. Other uncertainties are interpreted as contributors to the measurement noise of the code output and are implicitly treated in the GPM in the noise variance term, providing local uncertainty bounds for the peak clad temperature. We discuss the applicability of the foregoing method to reduce the use of conservative assumptions in best estimate plus uncertainty (BEPU and Level 1 probabilistic safety assessment (PSA success criteria definitions while dealing with a large number of uncertainties.

  4. Numerical investigation on hydraulic fracture cleanup and its impact on the productivity of a gas well with a non-Newtonian fluid model

    Energy Technology Data Exchange (ETDEWEB)

    Friedel, T. [Schlumberger Data and Consulting Services, Sugar Land, TX (United States)

    2006-07-01

    There are many damage mechanisms associated with hydraulically fractured gas wells. These include hydraulic damage caused by invading fluids during the treatment and damage due to the stresses exerted on the fracture face. Damage to the proppant pack can also reduce conductivity and non-Darcy flow. However, these are not the only impacts of impaired productivity in tight-gas reservoirs, which do not respond to hydraulic fracturing as expected. Some sustain a flat production profile or show only a slow increase in production rate for several weeks or months. This is due to poor rock quality, strong stress dependency in permeability, hydraulic and mechanical damage. Another reason for the poor performance is related to the cleanup of the cross-linked fracturing fluid with its non-Newtonian characteristics. This paper presented an improved 3-phase cleanup model for the investigation of polymer gel cleanup. Yield stress was considered according to the Herschel-Bulkley rheology model. The viscosity model is based on the exact analytical solution, including the plug flow zone. According to data in the published literature, half of the gel phase can be recovered. The gel saturation gradually increases towards the fracture tips, thereby lowering the fracture conductivities. The residing gel damages the permeability and porosity of the proppant pack or causes damage to the fracture face, thereby reducing production potential. These results are in agreement with field observations where fracture half-lengths, conductivities and productivity are also lower than expected. Preliminary results suggest that capillary forces and load-water recovery have little influence on gel cleanup. 16 refs., 2 tabs., 17 figs.

  5. Validation studies of thermal-hydraulic code for safety analysis of nuclear power plants

    International Nuclear Information System (INIS)

    Haapalehto, T.

    1995-01-01

    The thesis gives an overview of the validation process for thermal-hydraulic system codes and it presents in more detail the assessment and validation of the French code CATHARE for VVER calculations. Three assessment cases are presented: loop seal clearing, core reflooding and flow in a horizontal steam generator. The experience gained during these assessment and validation calculations has been used to analyze the behavior of the horizontal steam generator and the natural circulation in the geometry of the Loviisa nuclear power plant. Large part of the work has been performed in cooperation with the CATHARE-team in Grenoble, France. (41 refs., 11 figs., 8 tabs.)

  6. Coupled neutronic and thermal-hydraulic code benchmark activities at the International Nuclear Safety Center

    International Nuclear Information System (INIS)

    Podlazov, L. N.

    1998-01-01

    Two realistic benchmark problems are defined and used to assess the performance of coupled thermal-hydraulic and neutronic codes used in simulating dynamic processes in VVER-1000 and RBMK reactor systems. One of the problems simulates a design basis accident involving the ejection of three control and protection system rods from a VVER-1000 reactor. The other is based on a postulated rod withdrawal from an operating RBMK reactor. Preliminary results calculated by various codes are compared. While these results show significant differences, the intercomparisons performed so far provide a basis for further evaluation of code limitations and modeling assumptions

  7. PANDA a multi-purpose thermal-hydraulics facility devoted to nuclear reactor containment safety analysis

    International Nuclear Information System (INIS)

    Paladino, Domenico

    2014-01-01

    This paper presents the multi purpose facility PANDA devised for the safety analysis of nuclear reactor containment. The passive safety systems for LWRs have been explained with details about the PAssive Nachzerfallswärmeabfuhr und Druck-Abbau Testanlage (PANDA)

  8. Development of best estimate auditing code for CANDU thermal hydraulic safety analysis

    International Nuclear Information System (INIS)

    Hwnag, M.

    2001-04-01

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool , i.e. RELAP5/MOD3. This scope of project is a fourth step of the whole project, applying the RELAP5/MOD3/CANDU+ version for the real CANDU plant LOCA Analysis and D2O leakage incident. There are three main models under investigation, i.e. Moody critical flow model, flow regime model of horizontal CANDU bundle, and fuel element heatup model when the stratification occurs, especially when CANDU LOCA is tested. Also, for Wolsung unit 1 D2O leakage incident analysis, the plant behavior is predicited with the newly developed version for the first 1000 seconds after onset of the incident, with the main interest aiming for system pressure, level control system, and thermal hydraulic transient behavior of the secondary system. The model applided for this particular application includes heat transfer model of nuclear fuel assembly, decay heat model, and MOV (Motor Operated Valve) model. Finally, the code maintenance work, mainly correcting the known errors, is presented

  9. Related research with thermo hydraulics safety by means of Trace code; Investigaciones relacionadas con seguridad termohidraulica con el codigo TRACE

    Energy Technology Data Exchange (ETDEWEB)

    Chaparro V, F. J.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, UP - Adolfo Lopez Mateos, Edif. 9, 07738 Mexico D. F. (Mexico); Rodriguez H, A.; Gomez T, A. M. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Sanchez E, V. H.; Jager, W., E-mail: evalle@esfm.ipn.mx [Karlsruhe Institute of Technology, Hermann-von-Helmholtz Platz I, D-76344 Eggenstein - Leopoldshafen (Germany)

    2014-10-15

    In this article the results of the design of a pressure vessel of a BWR/5 similar to the type of Laguna Verde NPP are presented, using the Trace code. A thermo hydraulics Vessel component capable of simulating the behavior of fluids and heat transfer that occurs within the reactor vessel was created. The Vessel component consists of a three-dimensional cylinder divided into 19 axial sections, 4 azimuthal sections and two concentric radial rings. The inner ring is used to contain the core and the central part of the reactor, while the outer ring is used as a down comer. Axial an azimuthal divisions were made with the intention that the dimensions of the internal components, heights and orientation of the external connections match the reference values of a reactor BWR/5 type. In the model internal components as, fuel assemblies, steam separators, jet pumps, guide tubes, etc. are included and main external connections as, steam lines, feed-water or penetrations of the recirculation system. The model presents significant simplifications because the object is to keep symmetry between each azimuthal section of the vessel. In most internal components lack a detailed description of the geometry and initial values of temperature, pressure, fluid velocity, etc. given that it only considered the most representative data, however with these simulations are obtained acceptable results in important parameters such as the total flow through the core, the pressure in the vessel, percentage of vacuums fraction, pressure drop in the core and the steam separators. (Author)

  10. The hydraulics of the pressurized water reactors

    International Nuclear Information System (INIS)

    Bouchter, J.C.; Barbier, D.; Caruso, A.

    1999-01-01

    The SFEN organized, the 10 june 1999 at Paris, a meeting in the domain of the PWR hydraulics and in particular the hydraulic phenomena concerning the vessel and the vapor generators. The papers presented showed the importance of the industrial stakes with their associated phenomena: cores performance and safety with the more homogenous cooling system, the rods and the control rods wear, the temperature control, the fluid-structure interactions. A great part was also devoted to the progresses in the domain of the numerical simulation and the models and algorithms qualification. (A.L.B.)

  11. A Comprehensive Prediction Model of Hydraulic Extended-Reach Limit Considering the Allowable Range of Drilling Fluid Flow Rate in Horizontal Drilling.

    Science.gov (United States)

    Li, Xin; Gao, Deli; Chen, Xuyue

    2017-06-08

    Hydraulic extended-reach limit (HERL) model of horizontal extended-reach well (ERW) can predict the maximum measured depth (MMD) of the horizontal ERW. The HERL refers to the well's MMD when drilling fluid cannot be normally circulated by drilling pump. Previous model analyzed the following two constraint conditions, drilling pump rated pressure and rated power. However, effects of the allowable range of drilling fluid flow rate (Q min  ≤ Q ≤ Q max ) were not considered. In this study, three cases of HERL model are proposed according to the relationship between allowable range of drilling fluid flow rate and rated flow rate of drilling pump (Q r ). A horizontal ERW is analyzed to predict its HERL, especially its horizontal-section limit (L h ). Results show that when Q min  ≤ Q r  ≤ Q max (Case I), L h depends both on horizontal-section limit based on rated pump pressure (L h1 ) and horizontal-section limit based on rated pump power (L h2 ); when Q min  drilling fluid flow rate, while L h2 keeps decreasing as the drilling fluid flow rate increases. The comprehensive model provides a more accurate prediction on HERL.

  12. Proceedings of fifth international topical meeting on nuclear thermal hydraulics, operations and safety

    International Nuclear Information System (INIS)

    1997-01-01

    The fifth international topical meeting on nuclear thermohydraulics, operations and safety was convened in Beijing in April 14-18, 1997. The topical meeting was sponsored by the Chinese Nuclear Society and cosponsored by American Nuclear Society, Atomic Energy Society of Japan, American Society of Mechanical Engineers, Canada Nuclear Society, Korean Nuclear Society, Mexican Nuclear Society, Nuclear Society of Slovenia and Spanish Nuclear Society. There were 262 articles were published in the meeting. They are related nuclear power thermohydraulics, operations and safety

  13. Development of best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B. D.; Lee, W. J.; Lim, H. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2000-03-15

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model if existing PWR auditing tool, i.e. RELAP5/MOD3. This scope of project is a third step of the whole project, and expand the RELAP5/MOD3/CANDU version for implementation of LOCA analysis. There are three main area of model development, i.e. moody critical flow model, flow regime model of horizontal CANDU bundle, and fuel element heatup model when the stratification occurs. Newly developed version, namely RELAP5/MOD3/CANDU+ is applicable to CANDU plant analysis with keeping the function of light water reactor analysis. The limited validations of model installation were performed. Assessment of CHF model using AECL separated effect test and calculation for Wolsong 2 plant were performed also for the applicability test of the developed version.

  14. Thermal hydraulic codes for LWR safety analysis - present status and future perspective

    Energy Technology Data Exchange (ETDEWEB)

    Staedtke, H. [Commission of the European Union, Ispra (Italy)

    1997-07-01

    The aim of the present paper is to give a review on the current status and future perspective of present best-estimate Thermal Hydraulic codes. Reference is made to internationally well-established codes which have reached a certain state of maturity. The first part of the paper deals with the common basic code features with respect to the physical modelling and their numerical methods used to describe complex two-phase flow and heat transfer processes. The general predictive capabilities are summarized identifying some remaining code deficiencies and their underlying limitations. The second part discusses various areas including physical modelling, numerical techniques and informatic structure where the codes could be substantially improved.

  15. Development of a best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B.D.; Lee, W.J.; Lim, H.S. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-03-01

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool , i.e. RELAP5/MOD3. This scope of project is a third step of the whole project, and expand the RELAP5/MOD3/CANDU version for implementation of LOCA Analysis. There are three main area of model development, i.e. Moody critical flow model, flow regime model of horizontal CANDU bundle, and fuel element heatup model when the stratification occurs. Newly developed version, namely RELAP5/MOD3/CANDU+ is applicable to CANDU plant analysis with keeping the function of light water reactor analysis. The limited validations of model installation were performed. Assessment of CHF model using AECL separated effect test and calculation for Wolsong 2 plant were performed also for the applicability test of the developed version. 15 refs., 37 figs., 8 tabs. (Author)

  16. Thermal hydraulic codes for LWR safety analysis - present status and future perspective

    International Nuclear Information System (INIS)

    Staedtke, H.

    1997-01-01

    The aim of the present paper is to give a review on the current status and future perspective of present best-estimate Thermal Hydraulic codes. Reference is made to internationally well-established codes which have reached a certain state of maturity. The first part of the paper deals with the common basic code features with respect to the physical modelling and their numerical methods used to describe complex two-phase flow and heat transfer processes. The general predictive capabilities are summarized identifying some remaining code deficiencies and their underlying limitations. The second part discusses various areas including physical modelling, numerical techniques and informatic structure where the codes could be substantially improved

  17. Development of a best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B.D.; Lee, W.J.; Lim, H.S. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-04-01

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool, i.e. RELAP5/MOD3. This scope of project is a second step of the whole project, and focus to the implementation of CANDU models based on the previous study. FORTRAN 90 language have been used for the development of RELAP5.MOD3/CANDU PC version. For the convenience of the previous Workstation users, the FOTRAN 77 version has been coded also and implanted into the original RELAP5 source file. The verification of model implementation has been performed through the simple verification calculations using the CANDU version. 6 refs., 15 figs., 7 tabs. (Author)

  18. Thermal-hydraulic analysis code development and application to passive safety reactor at JAERI

    International Nuclear Information System (INIS)

    Araya, F.

    1995-01-01

    After a brief overview of safety assessment process, the author describes the LOCA analysis code system developed in JAERI. It comprises audit calculation code (WREM, WREM-J2, Japanese own code and BE codes (2D/3D, ICAP, ROSA). The codes are applied to development of Japanese passive safety reactor concept JPSR. Special attention is paid to the passive heat removal system and phenomena considered to occur under loss of heat sink event. Examples of LOCA analysis based on operation of JPSR for the cases of heat removal by upper RHR and heat removal from core to atmosphere are given. Experiments for multi-dimensional flow field in RPV and steam condensation in water pool are used for understanding the phenomena in passive safety reactors. The report is in a poster form only. 1 tab., 13 figs

  19. Some Findings from Thermal-Hydraulic Validation Tests for SMART Passive Safety System

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyun Sik; Bae, Hwang; Ryu, Sung-Uk; Ryu, Hyobong; Shin, Yong-Cheol; Min, Kyoung-Ho; Yi, Sung-Jae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    To satisfy the domestic and international needs for nuclear safety improvement after the Fukushima accident, an effort to improve its safety has been studied, and a Passive Safety System (PSS) for SMART has been designed. In addition, an Integral Test Loop for the SMART design (SMART-ITL, or FESTA) has been constructed and it finished its commissioning tests in 2012. Consequently, a set of Design Base Accident (DBA) scenarios have been simulated using SMARTITL. Recently, a test program to validate the performance of the SMART PSS was launched and its scaled-down test facility was additionally installed at the existing SMART-ITL facility. In this paper, some findings from the validation tests for the SMART PSS will be summarized. The acquired data will be used to validate the safety analysis code and its related models, to evaluate the performance of SMART PSS, and to provide base data during the application phase of SDA revision and construction licensing. A test program to validate the performance of SMARS PSS was launched with an additional scaleddown test facility of SMART PSS, which will be installed at the existing SMART-ITL facility. In this paper, some findings from the validation tests of the SMART passive safety system during 2013-2014 were summarized. They include a couple of SMART PSS tests using active pumps and several 1-train SMART PSS tests. From the test results it was estimated that the SMART PSS has sufficient cooling capability to deal with the SBLOCA scenario of SMART. During the SBLOCA scenario, in the CMT the water layer inventory was well stratified thermally and the safety injection water was injected efficiently into the RPV from the initial period and cools down the RCS properly.

  20. Thermal effects on fluid flow and hydraulic fracturing from wellbores and cavities in low-permeability formations

    Energy Technology Data Exchange (ETDEWEB)

    Yarlong Wang [Petro-Geotech Inc., Calgary, AB (Canada); Papamichos, Euripides [IKU Petroleum Research, Trondheim (Norway)

    1999-07-01

    The coupled heat-fluid-stress problem of circular wellbore or spherical cavity subjected to a constant temperature change and a constant fluid flow rate is considered. Transient analytical solutions for temperature, pore pressure and stress are developed by coupling conductive heat transfer with Darcy fluid flow in a poroelastic medium. They are applicable to lower permeability porous media suitable for liquid-waste disposal and also simulating reservoir for enhanced oil recovery, where conduction dominates the heat transfer process. A full range of solutions is presented showing separately the effects of temperature and fluid flow on pore pressure and stress development. It is shown that injection of warm fluid can be used to restrict fracture development around wellbores and cavities and generally to optimise a fluid injection operation. Both the limitations of the solutions and the convective flow effect are addressed. (Author)

  1. Final report of the 'Nordic thermal-hydraulic and safety network (NOTNET)' - Project

    Energy Technology Data Exchange (ETDEWEB)

    Tuunanen, J.; Tuomainen, M. [VTT Processes (Finland)

    2005-04-01

    A Nordic network for thermal-hydraulics and nuclear safety research was started. The idea of the network is to combine the resources of different research teams in order to carry out more ambitious and extensive research programs than would be possible for the individual teams. From the very beginning, the end users of the research results have been integrated to the network. Aim of the network is to benefit the partners involved in nuclear energy in the Nordic Countries (power companies, reactor vendors, safety regulators, research units). First task within the project was to describe the resources (personnel, know-how, simulation tools, test facilities) of the various teams. Next step was to discuss with the end users about their research needs. Based on these steps, few most important research topics with defined goals were selected, and coarse road maps were prepared for reaching the targets. These road maps will be used as a starting point for planning the actual research projects in the future. The organisation and work plan for the network were established. National coordinators were appointed, as well as contact persons in each participating organisation, whether research unit or end user. This organisation scheme is valid for the short-term operation of NOTNET when only Nordic organisations take part in the work. Later on, it is possible to enlarge the network e.g. within EC framework programme. The network can now start preparing project proposals and searching funding for the first common research projects. (au)

  2. Development of best estimate auditing code for CANDU thermal-hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Bub Dong; Lee, Won Jae; Hwang, Moon Kyu; Lim, Hong Sik [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-04-01

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool, i.e. RELAP5/MOD3.The study was performed by reconsideration of the previous code assessment works and phenomena identification for essential accident scenario. Improvement areas of model development for auditing tool were identified based on the code comparison and PIRT results. Nine models have been improved significantly for the analysis of LOCA and Mon LOCA event. Conceptual problem or separate effect assessment have been performed to verify the model improvement. The linking calculation with CONTAIN 2.0 has been also enabled to establish the unified auditing code system. Analysis for the CANDU plant real transient and hypothetical LOCA bas been performed using the improved version. It has been concluded that the developed version can be utilized for the auditing analysis of LOCA and non-LOCA event for the CANDU reactor. 25 refs., 84 figs., 36 tabs. (Author)

  3. Inconsistency in the average hydraulic models used in nuclear reactor design and safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jee Won; Roh, Gyu Hong; Choi, Hang Bok [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    One of important inconsistencies in the six-equation model predictions has been found to be the force experienced by a single bubble placed in a convergent stream of liquid. Various sets of governing equations yield different amount of forces to hold the bubble stationary in a convergent nozzle. By using the first order potential flow theory, it is found that the six-equation model can not be used to estimate the force experienced by a deformed bubble. The theoretical value of the particle stress of a bubble in a convergent nozzle flow has been found to be a function of the Weber number when bubble distortion is allowed. This force has been calculated by using different sets of governing equations and compared with the theoretical value. It is suggested in this study that the bubble size distribution function can be used to remove the presented inconsistency by relating the interfacial variables with different moments of the bubble size distribution function. This study also shows that the inconsistencies in the thermal-hydraulic governing equation can be removed by mechanistic modeling of the phasic interface. 11 refs., 3 figs. (Author)

  4. Thermal-hydraulic studies on the safety of VVER-440 type nuclear power plants

    International Nuclear Information System (INIS)

    Tuunanen, J.

    1994-01-01

    The thesis includes several thermal-hydraulic analyses related to the Loviisa VVER-440 nuclear power plant units. The work consists of experimental studies, analysis of the experiments, analysis of some plant transients and development of a calculational model for calculation of boric concentrations in the reactor. In the first part of thesis, in the case of simulation of boric acid solution behaviour during long-term cooling period of LOCAs, experiments were performed in scaled-down test facilities. The experimental data together with the results of RELAP5/MOD3 simulations were used to develop a model for calculations of boric acid concentrations in the reactor during LOCAs. In the second part, in the case of simulation of horizontal generators, experiments were performed with PACTEL integral test loop to simulate loss of feedwater transients. The PACTEL experiments as well as earlier REWETT-III natural circulation tests, were analyzed with RELAP5/MOD3 Version 5m5 code. The third part of the work consists of simulations of Loviisa VVER reactor pump trip transients with RELAP5/MOD1-Eur, RELAP5/MOD3 and CATHARE codes. (56 refs., 9 figs.)

  5. Inconsistency in the average hydraulic models used in nuclear reactor design and safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jee Won; Roh, Gyu Hong; Choi, Hang Bok [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    One of important inconsistencies in the six-equation model predictions has been found to be the force experienced by a single bubble placed in a convergent stream of liquid. Various sets of governing equations yield different amount of forces to hold the bubble stationary in a convergent nozzle. By using the first order potential flow theory, it is found that the six-equation model can not be used to estimate the force experienced by a deformed bubble. The theoretical value of the particle stress of a bubble in a convergent nozzle flow has been found to be a function of the Weber number when bubble distortion is allowed. This force has been calculated by using different sets of governing equations and compared with the theoretical value. It is suggested in this study that the bubble size distribution function can be used to remove the presented inconsistency by relating the interfacial variables with different moments of the bubble size distribution function. This study also shows that the inconsistencies in the thermal-hydraulic governing equation can be removed by mechanistic modeling of the phasic interface. 11 refs., 3 figs. (Author)

  6. Feedback from uncertainties propagation research projects conducted in different hydraulic fields: outcomes for engineering projects and nuclear safety assessment.

    Science.gov (United States)

    Bacchi, Vito; Duluc, Claire-Marie; Bertrand, Nathalie; Bardet, Lise

    2017-04-01

    In recent years, in the context of hydraulic risk assessment, much effort has been put into the development of sophisticated numerical model systems able reproducing surface flow field. These numerical models are based on a deterministic approach and the results are presented in terms of measurable quantities (water depths, flow velocities, etc…). However, the modelling of surface flows involves numerous uncertainties associated both to the numerical structure of the model, to the knowledge of the physical parameters which force the system and to the randomness inherent to natural phenomena. As a consequence, dealing with uncertainties can be a difficult task for both modelers and decision-makers [Ioss, 2011]. In the context of nuclear safety, IRSN assesses studies conducted by operators for different reference flood situations (local rain, small or large watershed flooding, sea levels, etc…), that are defined in the guide ASN N°13 [ASN, 2013]. The guide provides some recommendations to deal with uncertainties, by proposing a specific conservative approach to cover hydraulic modelling uncertainties. Depending of the situation, the influencing parameter might be the Strickler coefficient, levee behavior, simplified topographic assumptions, etc. Obviously, identifying the most influencing parameter and giving it a penalizing value is challenging and usually questionable. In this context, IRSN conducted cooperative (Compagnie Nationale du Rhone, I-CiTy laboratory of Polytech'Nice, Atomic Energy Commission, Bureau de Recherches Géologiques et Minières) research activities since 2011 in order to investigate feasibility and benefits of Uncertainties Analysis (UA) and Global Sensitivity Analysis (GSA) when applied to hydraulic modelling. A specific methodology was tested by using the computational environment Promethee, developed by IRSN, which allows carrying out uncertainties propagation study. This methodology was applied with various numerical models and in

  7. Coupled high fidelity thermal hydraulics and neutronics for reactor safety simulations

    International Nuclear Information System (INIS)

    Vincent A. Mousseau; Hongbin Zhang; Haihua Zhao

    2008-01-01

    This work is a continuation of previous work on the importance of accuracy in the simulation of nuclear reactor safety transients. This work is qualitative in nature and future work will be more quantitative. The focus of this work will be on a simplified single phase nuclear reactor primary. The transient of interest investigates the importance of accuracy related to passive (inherent) safety systems. The transient run here will be an Unprotected Loss of Flow (ULOF) transient. Here the coolant pump is turned off and the un-SCRAM-ed reactor transitions from forced to free convection (Natural circulation). Results will be presented that show the difference that the first order in time truncation physics makes on the transient. The purpose of this document is to illuminate a possible problem in traditional reactor simulation approaches. Detailed studies need to be done on each simulation code for each transient analyzed to determine if the first order truncation physics plays an important role

  8. Thermal-hydraulic safety aspects related to irradiation capabilities in MTR reactors

    International Nuclear Information System (INIS)

    Khedr, A.

    2009-01-01

    MTR research reactor such as ETRR-2 is an open pool type reactor that has a capability for irradiation into a number of irradiation boxes (IBs) installed at different positions on a separate grid called irradiation grid (I G). The I B has a lower removable plug to open or close its lower nozzle according to the I B is used or not.Increasing the used No. of I Bs in irradiation means that a valuable change in the flow distribution on the I G will occur. This paper is focused on the optimum number of I Bs that could be used without deterioration the cooling of I G components and avoiding the formation of hot spots. RELAP5 system code is used for thermal hydraulic analysis of the I G cooling system. Mathematical models and fortran program is developed to calculate the heat distribution in the I G components and the equivalent nozzle diameter that compensate the I B pressure drop due to the irradiated material (I M). This equivalent diameter simulates the used I B nozzle in the RELAP5 input deck. The results show that, the internal flow into the I Bs has significant effect on the coolability of the I G components. The number of I Bs that can be used is inversely proportional with the reactor power, the IM's void fraction and directly proportional with the PCS flow rate. Different cases of operating power and void fraction at two values for PCS flow are studied. In all of the cases considered limited number of the I Bs is permissible to use in order to avoid the excessive heating of the I G components

  9. An investigation of fluid mixing with safety injection in advanced reactors

    International Nuclear Information System (INIS)

    Cha, Jong Hee; Won, Soon Yean; Chung, Moon Ki; Jun, Hyung Gil

    1994-01-01

    The objective of this work is to investigate the fluid mixing phenomena in aspect of pressurized thermal shock(PTS) in an advanced PWR vessel downcomer during transient cooldown with safety injection. It provides comparison of fluid mixing characteristics between AP 600 DVI, designed by Westinghouse, and ABB CE System 80+ DVI, and the effects of deflector at the reactor downcomer. In order to investigate the fluid mixing phenomena in the downcomer of an advanced PWR, the flow visualization tests and the salt concentration tests were conducted in a 1/7-scale acrylic transparent model, which was designed and built based on AP 600 reactor geometry. The behaviour of the safety injection flow in downcomer associated with mixing phenomenon can be observed during visualization test, and time-dependent mixing rate between safety injection fluid and existing coolant can be determined with concentration test. Visualization tests were performed by the dye injection method. The results of concentration measurements were compared with the calculation using the REMIX code. During the tests, difference between AP 600 DVI flow and ABB CE System 80+ DVI flow and the effect of the deflector were observed

  10. Thermal-hydraulics associated with nuclear education and research

    International Nuclear Information System (INIS)

    Yokobori, Seiichi

    2011-01-01

    This article was the rerecording of the author's lecture at the fourth 'Future Energy Forum' (aiming at improving nuclear safety and economics) held in December 2010. The lecture focused on (1) importance of thermal hydraulics associated with nuclear education and research (critical heat flux, two-phase flow and multiphase flow), (2) emerging trend of maintenance engineering (fluid induced vibration, flow accelerated corrosion and stress corrosion cracks), (3) fostering sensible nuclear engineer with common engineering sense, (4) balanced curriculum of basics and advanced research, (5) computerized simulation and fluid mechanics, (6) crucial point of thermo hydraulics education (viscosity, flux, steam and power generation), (7) safety education and human resources development (indispensable technologies such as defence in depth) and (8) topics of thermo hydraulics research (vortices of curbed pipes and visualization of two-phase flow). (T. Tanaka)

  11. Progress of thermal hydraulic evaluation methods and experimental studies on a sodium-cooled fast reactor and its safety in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kamide, Hideki, E-mail: kamide.hideki@jaea.go.jp; Ohshima, Hiroyuki, E-mail: ohshima.hiroyuki@jaea.go.jp; Sakai, Takaaki, E-mail: sakai.takaaki@jaea.go.jp; Tanaka, Masaaki, E-mail: tanaka.masaaki@jaea.go.jp

    2017-02-15

    Highlights: • Thermal hydraulic issues for safety design criteria of sodium cooled fast reactors. • Measurement of velocity data in a subchannel surrounded by wire wrapped fuel-pins. • Statistical evaluation of core hot spot temperature during natural circulation. • Simulation of dynamics of molten fuel pool in a core disruptive accident. • V&V procedure of a multi-dimensional thermal hydraulic code on thermal striping. - Abstract: In the framework of the Generation-IV International Forum, the safety design criteria (SDC) incorporating safety-related R&D results on innovative technologies and lessons learned from Fukushima Dai-ichi nuclear power plants accident has been established to provide the set of general criteria for the safety designs of structures, systems and components of Generation-IV Sodium-cooled Fast Reactors (Gen-IV SFRs). A number of thermal-hydraulic evaluations are necessary to meet the concept of the criteria in the design studies of Gen-IV SFRs. This paper focuses on four kinds of thermal-hydraulic issues associated with the SDC, i.e., fuel subassembly thermal-hydraulics, natural circulation decay heat removal, core disruptive accidents, and thermal striping. Progress of evaluation methods on these issues is shown with activities on verification and validation (V&V) and experimental studies towards commercialization of SFR in Japan. These evaluation methods are planned to be eventually integrated into a comprehensive numerical simulation system that can be applied to all possible phenomena in SFR systems and that can be expected to become an effective tool for the development of human resource and the handing our knowledge and technologies down.

  12. Uncertainties assessment for safety margins evaluation in MTR reactors core thermal-hydraulic design

    International Nuclear Information System (INIS)

    Gimenez, M.; Schlamp, M.; Vertullo, A.

    2002-01-01

    This report contains a bibliographic review and a critical analysis of different methodologies used for uncertainty evaluation in research reactors core safety related parameters. Different parameters where uncertainties are considered are also presented and discussed, as well as their intrinsic nature regarding the way their uncertainty combination must be done. Finally a combined statistical method with direct propagation of uncertainties and a set of basic parameters as wall and DNB temperatures, CHF, PRD and their respective ratios where uncertainties should be considered is proposed. (author)

  13. Improvement of auditing technology of safety analysis through thermal-hydraulic separate effect tests

    Energy Technology Data Exchange (ETDEWEB)

    No, Hee Cheon; Park, Hyun Sik; Kim, Hyoung Tae; Moon, Young Min; Choi, Sung Won; Hwang, Do Hyun [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2000-03-15

    The direct-contact condensation hear transfer coefficients are experimentally obtained in the following conditions : pure steam/steam in the presence of noncondensible gas, horizontal/slightly inclined pipe, cocurrent/countercurrent stratified flow with water. The empirical correlation for liquid Nusselt number is developed in conditions of the slightly inclined pipe and the cocurrent stratified flow. The several models - the wall friction coefficient, the interfacial friction coefficient, the correlation of direct-contact condensation with noncondensible gases, and the correlation of wall film condensation - in the RELAP5/MOD3.2 code are modified, As results, RELAP5/MOD3.2 is improved. The present experimental data is used for evaluating the improved code. The standard RELAP5/MOD3.2 code is modified using the non-iterative modeling, which is a mechanistic model and does not require any interfacial information such as the interfacial temperature, The modified RELAP5/MOD3.2 code os used to simulate the horizontally stratified in-tube condensation experiment which represents the direct-contact condensation phenomena in a hot leg of a nuclear reactor. The modeling capabilities of the modified code as well as the standard code are assessed using several hot-leg condensation experiments. The modified code gives better prediction over local experimental data of liquid void fraction and interfacial heat transfer coefficient than the standard code. For the separate effect test of the thermal-hydraulic phenomena in the pressurizer, the scaling analysis is performed to obtain a similarity of the phenomena between the Korea Standard Nuclear Power Plant(KSNPP) and the present experimental facility. The diameters and lengths of the hot-leg, the surge line and the pressurizer are scaled down with the similitude of CCFL and velocity. The ratio of gas flow rate is 1/25. The experimental facility is composed of the air-water supply tank, the horizontal pipe, the surge line and the

  14. Trends in hydraulic fracturing distributions and treatment fluids, additives, proppants, and water volumes applied to wells drilled in the United States from 1947 through 2010: data analysis and comparison to the literature

    Science.gov (United States)

    Gallegos, Tanya J.; Varela, Brian A.

    2015-01-01

    Hydraulic fracturing is presently the primary stimulation technique for oil and gas production in low-permeability, unconventional reservoirs. Comprehensive, published, and publicly available information regarding the extent, location, and character of hydraulic fracturing in the United States is scarce. This national spatial and temporal analysis of data on nearly 1 million hydraulically fractured wells and 1.8 million fracturing treatment records from 1947 through 2010 (aggregated in Data Series 868) is used to identify hydraulic fracturing trends in drilling methods and use of proppants, treatment fluids, additives, and water in the United States. These trends are compared to the literature in an effort to establish a common understanding of the differences in drilling methods, treatment fluids, and chemical additives and of how the newer technology has affected the water use volumes and areal distribution of hydraulic fracturing. Historically, Texas has had the highest number of records of hydraulic fracturing treatments and associated wells in the United States documented in the datasets described herein. Water-intensive horizontal/directional drilling has also increased from 6 percent of new hydraulically fractured wells drilled in the United States in 2000 to 42 percent of new wells drilled in 2010. Increases in horizontal drilling also coincided with the emergence of water-based “slick water” fracturing fluids. As such, the most current hydraulic fracturing materials and methods are notably different from those used in previous decades and have contributed to the development of previously inaccessible unconventional oil and gas production target areas, namely in shale and tight-sand reservoirs. Publicly available derivative datasets and locations developed from these analyses are described.

  15. Hydraulic Properties of Closely Spaced Dipping Open Fractures Intersecting a Fluid-Filled Borehole Derived From Tube Wave Generation and Scattering

    Science.gov (United States)

    Minato, Shohei; Ghose, Ranajit; Tsuji, Takeshi; Ikeda, Michiharu; Onishi, Kozo

    2017-10-01

    Fluid-filled fractures and fissures often determine the pathways and volume of fluid movement. They are critically important in crustal seismology and in the exploration of geothermal and hydrocarbon reservoirs. We introduce a model for tube wave scattering and generation at dipping, parallel-wall fractures intersecting a fluid-filled borehole. A new equation reveals the interaction of tube wavefield with multiple, closely spaced fractures, showing that the fracture dip significantly affects the tube waves. Numerical modeling demonstrates the possibility of imaging these fractures using a focusing analysis. The focused traces correspond well with the known fracture density, aperture, and dip angles. Testing the method on a VSP data set obtained at a fault-damaged zone in the Median Tectonic Line, Japan, presents evidences of tube waves being generated and scattered at open fractures and thin cataclasite layers. This finding leads to a new possibility for imaging, characterizing, and monitoring in situ hydraulic properties of dipping fractures using the tube wavefield.

  16. Thermal hydraulic test for reactor safety system; a visualization study on flow boiling and bubble behavior

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Soon Heung; Baek, Won Pil; Ban, In Cheol [Korea Advanced Institute of Science and Technology, Taejeon (Korea)

    2002-03-01

    The project contribute to understand and to clarify the physical mechanism of flow nucleate boiling and CHF phenomena through the visualization experiments. the results are useful in the development of the enhancement device of heat transfer and to enhance nuclear fuel safety 1. Visual experimental facility 2. Application method of visualization Technique 3. Visualization results of flow nucleate boiling regime - Overall Bubble Behavior on the Heated Surface - Bubble Behavior near CHF Condition - Identification of Flow Structure - Three-layer flow structure 4. Quantifying of bubble parameter through a digital image processing - Image Processing Techniques - Classification of objects and measurements of the size - Three dimensional surface plot with using the luminance 5. Development and estimation of a correlation between bubble diameter and flow parameter - The effect of system parameter on bubble diameter - The development of a bubble diameter correlation . 49 refs., 42 figs., 7 tabs. (Author)

  17. Thermally Actuated Hydraulic Pumps

    Science.gov (United States)

    Jones, Jack; Ross, Ronald; Chao, Yi

    2008-01-01

    Thermally actuated hydraulic pumps have been proposed for diverse applications in which direct electrical or mechanical actuation is undesirable and the relative slowness of thermal actuation can be tolerated. The proposed pumps would not contain any sliding (wearing) parts in their compressors and, hence, could have long operational lifetimes. The basic principle of a pump according to the proposal is to utilize the thermal expansion and contraction of a wax or other phase-change material in contact with a hydraulic fluid in a rigid chamber. Heating the chamber and its contents from below to above the melting temperature of the phase-change material would cause the material to expand significantly, thus causing a substantial increase in hydraulic pressure and/or a substantial displacement of hydraulic fluid out of the chamber. Similarly, cooling the chamber and its contents from above to below the melting temperature of the phase-change material would cause the material to contract significantly, thus causing a substantial decrease in hydraulic pressure and/or a substantial displacement of hydraulic fluid into the chamber. The displacement of the hydraulic fluid could be used to drive a piston. The figure illustrates a simple example of a hydraulic jack driven by a thermally actuated hydraulic pump. The pump chamber would be a cylinder containing encapsulated wax pellets and containing radial fins to facilitate transfer of heat to and from the wax. The plastic encapsulation would serve as an oil/wax barrier and the remaining interior space could be filled with hydraulic oil. A filter would retain the encapsulated wax particles in the pump chamber while allowing the hydraulic oil to flow into and out of the chamber. In one important class of potential applications, thermally actuated hydraulic pumps, exploiting vertical ocean temperature gradients for heating and cooling as needed, would be used to vary hydraulic pressures to control buoyancy in undersea research

  18. Neutronic, thermal-hydraulics and safety calculations of a Miniplate Irradiation Device (MID) of dispersion type fuel elements

    International Nuclear Information System (INIS)

    Domingos, Douglas Borges

    2010-01-01

    Neutronic, thermal-hydraulics and accident analysis calculations were developed to estimate the safety of a Miniplate Irradiation Device (MID) to be placed in the IEA-R1 reactor core. The irradiation device is used to receive miniplates of U 3 O 8 -Al and U 3 Si 2 - Al dispersion fuels, LEU type (19.75 % 235 U) with uranium densities of, respectively, 3.2 gU/cm 3 and 4.8 gU/cm 3 . The fuel miniplates will be irradiated to nominal 235 U burnup levels of 50% and 80%, in order to qualify the above high-density dispersion fuels to be used in the Brazilian Multipurpose Reactor (RMB), now in the conception phase. For the neutronic calculation, the computer codes CITATION and 2DB were utilized. The computer code FLOW was used to calculate the coolant flow rate in the irradiation device, allowing the determination of the fuel miniplate temperatures with the computer model MTRCR-IEA-R1. A postulated Loss of Coolant Accident (LOCA) was analyzed with the computer codes LOSS and TEMPLOCA, allowing the calculation of the fuel miniplate temperatures after the reactor pool draining. The calculations showed that the irradiation should occur without adverse consequences in the IEA-R1 reactor. (author)

  19. RDS; A systematic approach towards system thermal hydraulics input code development for a comprehensive deterministic safety analysis

    International Nuclear Information System (INIS)

    Mohd Faiz Salim; Ridha Roslan; Mohd Rizal Mamat

    2013-01-01

    Full-text: Deterministic Safety Analysis (DSA) is one of the mandatory requirements conducted for Nuclear Power Plant licensing process, with the aim of ensuring safety compliance with relevant regulatory acceptance criteria. DSA is a technique whereby a set of conservative deterministic rules and requirements are applied for the design and operation of facilities or activities. Computer codes are normally used to assist in performing all required analysis under DSA. To ensure a comprehensive analysis, the conduct of DSA should follow a systematic approach. One of the methodologies proposed is the Standardized and Consolidated Reference Experimental (and Calculated) Database (SCRED) developed by University of Pisa. Based on this methodology, the use of Reference Data Set (RDS) as a pre-requisite reference document for developing input nodalization was proposed. This paper shall describe the application of RDS with the purpose of assessing its effectiveness. Two RDS documents were developed for an Integral Test Facility of LOBIMOD2 and associated Test A1-83. Data and information from various reports and drawings were referred in preparing the RDS. The results showed that by developing RDS, it has made possible to consolidate all relevant information in one single document. This is beneficial as it enables preservation of information, promotes quality assurance, allows traceability, facilitates continuous improvement, promotes solving of contradictions and finally assisting in developing thermal hydraulic input regardless of whichever code selected. However, some disadvantages were also recognized such as the need for experience in making engineering judgments, language barrier in accessing foreign information and limitation of resources. Some possible improvements are suggested to overcome these challenges. (author)

  20. RDS - A systematic approach towards system thermal hydraulics input code development for a comprehensive deterministic safety analysis

    International Nuclear Information System (INIS)

    Salim, Mohd Faiz; Roslan, Ridha; Ibrahim, Mohd Rizal Mamat

    2014-01-01

    Deterministic Safety Analysis (DSA) is one of the mandatory requirements conducted for Nuclear Power Plant licensing process, with the aim of ensuring safety compliance with relevant regulatory acceptance criteria. DSA is a technique whereby a set of conservative deterministic rules and requirements are applied for the design and operation of facilities or activities. Computer codes are normally used to assist in performing all required analysis under DSA. To ensure a comprehensive analysis, the conduct of DSA should follow a systematic approach. One of the methodologies proposed is the Standardized and Consolidated Reference Experimental (and Calculated) Database (SCRED) developed by University of Pisa. Based on this methodology, the use of Reference Data Set (RDS) as a pre-requisite reference document for developing input nodalization was proposed. This paper shall describe the application of RDS with the purpose of assessing its effectiveness. Two RDS documents were developed for an Integral Test Facility of LOBI-MOD2 and associated Test A1-83. Data and information from various reports and drawings were referred in preparing the RDS. The results showed that by developing RDS, it has made possible to consolidate all relevant information in one single document. This is beneficial as it enables preservation of information, promotes quality assurance, allows traceability, facilitates continuous improvement, promotes solving of contradictions and finally assisting in developing thermal hydraulic input regardless of whichever code selected. However, some disadvantages were also recognized such as the need for experience in making engineering judgments, language barrier in accessing foreign information and limitation of resources. Some possible improvements are suggested to overcome these challenges

  1. RDS - A systematic approach towards system thermal hydraulics input code development for a comprehensive deterministic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Salim, Mohd Faiz, E-mail: mohdfaizs@tnb.com.my [Nuclear Energy Department, Tenaga Nasional Berhad, Level 32, Dua Sentral, 50470 Kuala Lumpur (Malaysia); Roslan, Ridha [Nuclear Installation Division, Atomic Energy Licensing Board, Batu 24, Jalan Dengkil, 43800 Dengkil, Selangor (Malaysia); Ibrahim, Mohd Rizal Mamat [Technical Support Division, Malaysian Nuclear Agency, Bangi, 43000 Kajang, Selangor (Malaysia)

    2014-02-12

    Deterministic Safety Analysis (DSA) is one of the mandatory requirements conducted for Nuclear Power Plant licensing process, with the aim of ensuring safety compliance with relevant regulatory acceptance criteria. DSA is a technique whereby a set of conservative deterministic rules and requirements are applied for the design and operation of facilities or activities. Computer codes are normally used to assist in performing all required analysis under DSA. To ensure a comprehensive analysis, the conduct of DSA should follow a systematic approach. One of the methodologies proposed is the Standardized and Consolidated Reference Experimental (and Calculated) Database (SCRED) developed by University of Pisa. Based on this methodology, the use of Reference Data Set (RDS) as a pre-requisite reference document for developing input nodalization was proposed. This paper shall describe the application of RDS with the purpose of assessing its effectiveness. Two RDS documents were developed for an Integral Test Facility of LOBI-MOD2 and associated Test A1-83. Data and information from various reports and drawings were referred in preparing the RDS. The results showed that by developing RDS, it has made possible to consolidate all relevant information in one single document. This is beneficial as it enables preservation of information, promotes quality assurance, allows traceability, facilitates continuous improvement, promotes solving of contradictions and finally assisting in developing thermal hydraulic input regardless of whichever code selected. However, some disadvantages were also recognized such as the need for experience in making engineering judgments, language barrier in accessing foreign information and limitation of resources. Some possible improvements are suggested to overcome these challenges.

  2. Adsorption of hydraulic fracturing fluid components 2-butoxyethanol and furfural onto granular activated carbon and shale rock.

    Science.gov (United States)

    Manz, Katherine E; Haerr, Gregory; Lucchesi, Jessica; Carter, Kimberly E

    2016-12-01

    The objective of this study was to understand the adsorption ability of a surfactant and a non-surfactant chemical additive used in hydraulic fracturing onto shale and GAC. Experiments were performed at varying temperatures and sodium chloride concentrations to establish these impacts on the adsorption of the furfural (a non-surfactant) and 2-Butoxyethanol (2-BE) (a surfactant). Experiments were carried out in continuously mixed batch experiments with Langmuir and Freundlich isotherm modeling. The results of the experiments showed that adsorption of these compounds onto shale does not occur, which may allow these compounds to return to the surface in flowback and produced waters. The adsorption potential for these chemicals onto GAC follows the assumptions of the Langmuir model more strongly than those of the Freundlich model. The results show uptake of furfural and 2-BE occurs within 23 h in the presence of DI water, 0.1 mol L -1 sodium chloride, and in lab synthesized hydraulic fracturing brine. Based on the data, 83% of the furfural and 62% of the 2-BE was adsorbed using GAC. Copyright © 2016 Elsevier Ltd. All rights reserved.

  3. Pemetrexed safety and pharmacokinetics in patients with third-space fluid

    DEFF Research Database (Denmark)

    Dickgreber, Nicolas J; Sorensen, Jens Benn; Paz-Ares, Luis G

    2010-01-01

    Pemetrexed is established as first-line treatment with cisplatin for malignant pleural mesothelioma and advanced nonsquamous non-small-cell lung cancer (NSCLC) and as single-agent second-line treatment for nonsquamous NSCLC. Because the structure and pharmacokinetics of pemetrexed are similar to ...... to those of methotrexate, and methotrexate is associated with severe toxicity in patients with third-space fluid (TSF), the safety of pemetrexed in patients with TSF was evaluated....

  4. Improvement of auditing technology of safety analysis through thermal-hydraulic separate effect tests

    Energy Technology Data Exchange (ETDEWEB)

    No, Hee Cheon; Park, Hyun Sik; Kim, Hyougn Tae; Moon, Young Min; Choi, Sung Won; Heo, Sun [Korea Advanced Institute Science and Technology, Taejon (Korea, Republic of)

    1999-04-15

    The loss-of-RHR accident during midloop operation has been important as results of the probabilistic safety analysis. The condensation models In RELAP5/MOD3 are not proper to analyze the midloop operation. To audit and improve the model in RELAP5/MOD3.2, several items of separate effect tests have been performed. The 29 sets of reflux condensation data is obtained and the correlation is developed with these heat transfer coefficient's data. In the experiment of the direct contact condensation in hot leg, the apparatus setting is finished and a few experimental data is obtained. Non-iterative model is used to predict the model in RELAP5/MOD3.2 with the results of reflux condensation and evaluates better than the present model. The results of the direct contact condensation in a hot leg represent to be similar with the present model. The study of the CCF and liquid entrainment in a surge line and pressurizer is selected as the third separate experiment and is on performance.

  5. Small hydraulic turbine drives

    Science.gov (United States)

    Rostafinski, W. A.

    1970-01-01

    Turbine, driven by the fluid being pumped, requires no external controls, is completely integrated into the flow system, and has bearings which utilize the main fluid for lubrication and cooling. Torque capabilities compare favorably with those developed by positive displacement hydraulic motors.

  6. Comprehensive thermal-hydraulic and thermal-mechanical analysis of core and fuel rods for the safety validation of real refueling at the Kozloduy WWER-440

    Energy Technology Data Exchange (ETDEWEB)

    Stefanova, S; Panajotov, D; Ilieva, B; Vitkova, M; Simeonova, V; Passage, G; Manolova, M [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. za Yadrena Izsledvaniya i Yadrena Energetika

    1996-12-31

    Safety analysis aimed at determination of thermal-hydraulic and thermal-mechanical margins of core and fuel rods has been carried out using computer codes COBSOFM and PIN-micro. Thermal-hydraulic calculations for the part of the core with maximum heat flux during steady-state regime show that the coolant, cladding and fuel temperatures are within the design limits. A severe accident with reactor blackout has been simulated. It is found that at 95% probability level there is no boiling crisis anywhere in the core. The thermal-mechanical parameters of working assembly fuel rod with maximum load have been calculated. The assembly linear power reached a maximum of 25 kW/m during the second fuel cycle, the fuel temperature remaining well below 1000{sup o} C. As the fuel assembly with typical power history has enough safety margins, it was proposed to use it for one more cycle. 4 refs., 12 figs.

  7. Application of RELAP5/MOD3.3 to Calculate Thermal Hydraulic Behavior of the Pressurizer Safety Valve Performance Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Hyun; Kim, Young Ae; Oh, Seung Jong; Park, Jong Woon [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of)

    2007-10-15

    The increase of the acceptance tolerance of Pressurizer Safety Valve (PSV) test is vital for the safe operation of nuclear power plants because the frequent tests may make the valves decrepit and become a cause of leak. Recently, Korea Hydro and Nuclear Power Company (KHNP) is building a PSV performance test facility to provide the technical background data for the relaxation of the acceptance tolerance of PSV including the valve pop-up characteristics and the loop seal dynamics (if the plant has the loop seal in the upstream of PSV). The discharge piping and supports must be designed to withstand severe transient hydrodynamic loads when the safety valve actuates. The evaluation of hydrodynamic loads is a two-step process: first the thermal hydraulic behavior in the piping must be defined, and then the hydrodynamic loads are calculated from the thermal hydraulic parameters such as pressure and mass flow. The hydrodynamic loads are used as input to the structural analysis.

  8. Vertical hydraulic conductivity of a clayey-silt aquitard: accelerated fluid flow in a centrifuge permeameter compared with in situ conditions

    Science.gov (United States)

    Timms, W. A.; Crane, R.; Anderson, D. J.; Bouzalakos, S.; Whelan, M.; McGeeney, D.; Rahman, P. F.; Guinea, A.; Acworth, R. I.

    2014-03-01

    Evaluating the possibility of leakage through low permeability geological strata is critically important for sustainable water supplies, extraction of fuels from strata such as coal beds, and confinement of waste within the earth. Characterizing low or negligible flow rates and transport of solutes can require impractically long periods of field or laboratory testing, but is necessary for evaluations over regional areas and over multi-decadal timescales. The current work reports a custom designed centrifuge permeameter (CP) system, which can provide relatively rapid and reliable hydraulic conductivity (K) measurement compared to column permeameter tests at standard gravity (1g). Linear fluid velocity through a low K porous sample is linearly related to g-level during a CP flight unless consolidation or geochemical reactions occur. The CP module is designed to fit within a standard 2 m diameter, geotechnical centrifuge with a capacity for sample dimensions of 30 to 100 mm diameter and 30 to 200 mm in length. At maximum RPM the resultant centrifugal force is equivalent to 550g at base of sample or a total stress of ~2 MPa. K is calculated by measuring influent and effluent volumes. A custom designed mounting system allows minimal disturbance of drill core samples and a centrifugal force that represents realistic in situ stress conditions is applied. Formation fluids were used as influent to limit any shrink-swell phenomena which may alter the resultant K value. Vertical hydraulic conductivity (Kv) results from CP testing of core from the sites in the same clayey silt formation varied (10-7 to 10-9 m s-1, n = 14) but higher than 1g column permeameter tests of adjacent core using deionized water (10-9 to 10-11 m s-1, n = 7). Results at one site were similar to in situ Kv values (3 × 10-9 m s-1) from pore pressure responses within a 30 m clayey sequence in a homogenous area of the formation. Kv sensitivity to sample heterogeneity was observed, and anomalous flow via

  9. RETRAN-02: a program for transient thermal-hydraulic analysis of complex fluid-flow systems. Volume 4. Applications

    International Nuclear Information System (INIS)

    Peterson, C.E.; Gose, G.C.; McFadden, J.H.

    1983-01-01

    RETRAN-02 represents a significant achievement in the development of a versatile and reliable computer program for use in best estimate transient thermal-hydraulic analysis of light water reactor systems. The RETRAN-02 computer program is an extension of the RETRAN-01 program designed to provide analysis capabilities for 1) BWR and PWR transients, 2) small break loss of coolant accidents, 3) balance of plant modeling, and 4) anticipated transients without scram, while maintaining the analysis capabilities of the predecessor code. The RETRAN-02 computer code is constructed in a semimodular and dynamic dimensioned form where additions to the code can be easily carried out as new and improved models are developed. This report (the fourth of a five volume computer code manual) describes the verification and validation of RETRAN-02

  10. Seismic monitoring of hydraulic fracturing: techniques for determining fluid flow paths and state of stress away from a wellbore

    Energy Technology Data Exchange (ETDEWEB)

    Fehler, M.; House, L.; Kaieda, H.

    1986-01-01

    Hydraulic fracturing has gained in popularity in recent years as a way to determine the orientations and magnitudes of tectonic stresses. By augmenting conventional hydraulic fracturing measurements with detection and mapping of the microearthquakes induced by fracturing, we can supplement and idependently confirm information obtained from conventional analysis. Important information obtained from seismic monitoring includes: the state of stress of the rock, orientation and spacing of the major joint sets, and measurements of rock elastic parameters at locations distant from the wellbore. While conventional well logging operations can provide information about several of these parameters, the zone of interrogation is usually limited to the immediate proximity of the borehole. The seismic waveforms of the microearthquakes contain a wealth of information about the rock in regions that are otherwise inaccessible for study. By reliably locating the hypocenters of many microearthquakes, we have inferred the joint patterns in the rock. We observed that microearthquake locations do not define a simple, thin, planar distribution, that the fault plane solutions are consistent with shear slippage, and that spectral analysis indicates that the source dimensions and slip along the faults are small. Hence we believe that the microearthquakes result from slip along preexisting joints, and not from tensile extension at the tip of the fracture. Orientations of the principal stresses can be estimated by using fault plane solutions of the larger microearthquakes. By using a joint earthquake location scheme, and/or calibrations with downhole detonators, rock velocities and heterogeneities thereof can be investigated in rock volumes that are far enough from the borehole to be representative of intrincis rock properties.

  11. Statistical safety evaluation of BWR turbine trip scenario using coupled neutron kinetics and thermal hydraulics analysis code SKETCH-INS/TRACE5.0

    International Nuclear Information System (INIS)

    Ichikawa, Ryoko; Masuhara, Yasuhiro; Kasahara, Fumio

    2012-01-01

    The Best Estimate Plus Uncertainty (BEPU) method has been prepared for the regulatory cross-check analysis at Japan Nuclear Energy Safety Organization (JNES) on base of the three-dimensional neutron-kinetics/thermal-hydraulics coupled code SKETCH-INS/TRACE5.0. In the preparation, TRACE5.0 is verified against the large-scale thermal-hydraulic tests carried out with NUPEC facility. These tests were focused on the pressure drop of steam-liquid two phase flow and void fraction distribution. From the comparison of the experimental data with other codes (RELAP5/MOD3.3 and TRAC-BF1), TRACE5.0 was judged better than other codes. It was confirmed that TRACE5.0 has high reliability for thermal hydraulics behavior and are used as a best-estimate code for the statistical safety evaluation. Next, the coupled code SKETCH-INS/TRACE5.0 was applied to turbine trip tests performed at the Peach Bottom-2 BWR4 Plant. The turbine trip event shows the rapid power peak due to the voids collapse with the pressure increase. The analyzed peak value of core power is better simulated than the previous version SKETCH-INS/TRAC-BF1. And the statistical safety evaluation using SKETCH-INS/TRACE5.0 was applied to the loss of load transient for examining the influence of the choice of sampling method. (author)

  12. Determination of MIL-H-6083 Hydraulic Fluid In-Service Use Limits for Self-Propelled Artillery

    Science.gov (United States)

    1991-09-01

    determined using the American Society for Testing and Materials (ASTM) D1744 Karl Fischer Reagent method . The specification limit is 0.05% (500 pans per...cazefully controlled. TOTAL ACID NUMBER The acid number was determined by the ASTM D664 potentiometric titration test method . Unfortunately, data were...fluid condition t results with AOAP tent date was found. The Navy Patch Kit method for particle contamination meamrement was evaluated as a possible

  13. Safety, tolerability, and cerebrospinal fluid penetration of ursodeoxycholic Acid in patients with amyotrophic lateral sclerosis.

    Science.gov (United States)

    Parry, Gareth J; Rodrigues, Cecilia M P; Aranha, Marcia M; Hilbert, Sarah J; Davey, Cynthia; Kelkar, Praful; Low, Walter C; Steer, Clifford J

    2010-01-01

    Amyotrophic lateral sclerosis is a progressive degenerative disease, which typically leads to death in 3 to 5 years. Neuronal cell death offers a potential target for therapeutic intervention. Ursodeoxycholic acid is a cytoprotective, endogenous bile acid that has been shown to be neuroprotective in experimental Huntington and Alzheimer diseases, retinal degeneration, and ischemic and hemorrhagic stroke. The objective of this research was to study the safety and the tolerability of ursodeoxycholic acid in amyotrophic lateral sclerosis and document effective and dose-dependent cerebrospinal fluid penetration. Eighteen patients were randomly assigned to receive ursodeoxycholic acid at doses of 15, 30, and 50 mg/kg of body weight per day. Serum and cerebrospinal fluid were obtained for analysis after 4 weeks of treatment. Treatment-emergent clinical and laboratory events were monitored weekly. Our data indicated that ursodeoxycholic acid is well tolerated by all subjects at all doses. We also showed that ursodeoxycholic acid is well absorbed after oral administration and crosses the blood-brain barrier in a dose-dependent manner. These results show excellent safety and tolerability of ursodeoxycholic acid. The drug penetrates the cerebrospinal fluid in a dose-dependent manner. A large, placebo-controlled clinical trial is needed to assess the efficacy of ursodeoxycholic acid in treating amyotrophic lateral sclerosis.

  14. A two-fluid two-phase model for thermal-hydraulic analysis of a U-tube steam generator

    International Nuclear Information System (INIS)

    Hung, Huanjen; Chieng, Chingchang; Pei, Baushei; Wang, Songfeng

    1993-01-01

    The Advanced Thermal-Hydraulic Analysis Code for Nuclear Steam Generators (ATHANS) was developed on the basis of the THERMIT-UTSG computer code for U-tube steam generators. The main features of the ATHANS model are as follows: (a) the equations are solved in cylindrical coordinates, (b) the number and the arrangement of the control volumes inside the steam generator can be chosen by the user, (c) the virtual mass effect is incorporated, and (d) the conjugate gradient squared method is employed to accelerate and improve the numerical convergence. The performance of the model is successfully validated by comparison with the test data from a Westinghouse model F steam generator at the Maanshan nuclear power plant. Better agreement with the test data can be obtained by a finer grid system using a cylindrical coordinate system and the virtual mass effect. With these advanced features, ATHANS provides the basic framework for further studies on the problems of steam generators, such as analyses of secondary-side corrosion and tube ruptures

  15. Implementation of wall film condensation model to two-fluid model in component thermal hydraulic analysis code CUPID - 15237

    International Nuclear Information System (INIS)

    Lee, J.H.; Park, G.C.; Cho, H.K.

    2015-01-01

    In the containment of a nuclear reactor, the wall condensation occurs when containment cooling system and structures remove the mass and energy release and this phenomenon is of great importance to ensure containment integrity. If the phenomenon occurs in the presence of non-condensable gases, their accumulation near the condensate film leads to significant reduction in heat transfer during the condensation. This study aims at simulating the wall film condensation in the presence of non-condensable gas using CUPID, a computational multi-fluid dynamics code, which is developed by the Korea Atomic Energy Research Institute (KAERI) for the analysis of transient two-phase flows in nuclear reactor components. In order to simulate the wall film condensation in containment, the code requires a proper wall condensation model and liquid film model applicable to the analysis of the large scale system. In the present study, the liquid film model and wall film condensation model were implemented in the two-fluid model of CUPID. For the condensation simulation, a wall function approach with heat and mass transfer analogy was applied in order to save computational time without considerable refinement for the boundary layer. This paper presents the implemented wall film condensation model and then, introduces the simulation result using CUPID with the model for a conceptual condensation problem in a large system. (authors)

  16. Computational Fluid Dynamics for Nuclear Reactor Safety-5 (CFD4NRS-5). Workshop Proceedings, 9-11 September 2014, Zurich, Switzerland

    International Nuclear Information System (INIS)

    Smith, Brian L.; Andreani, Michele; Badillo, Arnoldo; Dehbi, Abdel; Sato, Yohei; Smith, Brian L.; Dreier, Joerg; Kapulla, Ralf; Niceno, Bojan; Sharabi, Medhat; Bestion, Dominique; Bieder, Ulrich; Coste, Pierre; Martinez, Jean Marc; Zigh, Ghani; Boyd, Chris; Prasser, Horst-Michael; Kerenyi, Nora; Adams, Robert; Bolesch, Christian; D'Aleo, Paolo; Eismann, Ralph; Kickhofel, John; Lafferty, Nathan; Saxena, Abhishek; Kissane, Martin; ); Ulses, Anthony; ); Bartosiewicz, Yann; Seynhaeve, Jean-Marie; Caraghiaur, Diana; Munoz Cobo, Jose Luis; Glaeser, Horst; Buchholz, Sebastian; Scheuerer, Martina; Hassan, Yassin; In, Wang-Kee; Song, Chul-Hwa; Yoon, Han-Young; Kim, J.W.; Koncar, Bostjan; Tiselj, Iztoc; Lakehal, Djamel; Yadigaroglu, George; Lo, Simon; Manera, Annalisa; Petrov, Victor; Mimouni, Stephane; Benhamadouche, Sofiane; Morii, Tadashi; Suikkanen, Heikki; Toppila, Timo; Angele, Kristian; Baglietto, Emilio; Cheng, Xu; Graffard, Estelle; Ko, Jordan; Hoehne, Thomas; Lucas, Dirk; Krepper, Eckhard; Laurien, Eckart; Moretti, Fabio; Piro, Markus; Roelofs, Ferry; Veber, Pascal; Watanabe, Tadashi; Yan, Jin; Yeoh, Guan

    2016-01-01

    This present workshop, the 5. Computational Fluid Dynamics for Nuclear-Reactor Safety (CFD4NRS-5), in the biennial series of such Nuclear Energy Agency (NEA) and International Atomic Energy Agency (IAEA) sponsored events, a tradition which began in Garching in 2006, follows the format and objectives of its predecessors in creating a forum whereby numerical analysts and experimentalists can exchange information in the application of computational fluid dynamics (CFD) to nuclear power plant (NPP) safety and future design issues. The emphasis, as always, was, in a congenial atmosphere, to offer exposure to state-of-the-art (single-phase and multi-phase) CFD applications reflecting topical issues arising in NPP design and safety, but in particular to promote the release of high-resolution experimental data to continue the CFD validation process in this application area. The reason for the increased use of multi-dimensional CFD methods is that a number of important thermal-hydraulic phenomena occurring in NPPs cannot be adequately predicted using traditional one-dimensional system hydraulics codes with the required accuracy and spatial resolution when strong three-dimensional motions prevail. Established CFD codes already contain empirical models for simulating turbulence, heat transfer, multi-phase interaction and chemical reactions. Nonetheless, such models must be validated against test data before they can be used with confidence. The necessary validation procedure is performed by comparing model predictions against trustworthy experimental data. However, reliable model assessment requires CFD simulations to be undertaken with full control over numerical errors and input uncertainties. The writing groups originally set up by the NEA have been consistently promoting the use of best practice guidelines (BPGs) in the application of CFD for just this purpose, and BPGs remain a central pillar of the simulation material accepted at this current workshop, as it was at its

  17. Study on safety of a nuclear ship having an integral marine water reactor. Intelligent information database program concerned with thermal-hydraulic characteristics

    International Nuclear Information System (INIS)

    Inasaka, Fujio; Nariai, Hideki; Kobayashi, Michiyuki; Murata, Hiroyuki; Aya, Izuo

    2001-01-01

    As a high economical marine reactor with sufficient safety functions, an integrated type marine water reactor has been considered most promising. At the National Maritime Research Institute, a series of the experimental studies on the thermal-hydraulic characteristics of an integrated/passive-safety type marine water reactor such as the flow boiling of a helical-coil type steam generator, natural circulation of primary water under a ship rolling motion and flashing-condensation oscillation phenomena in pool water has been conducted. This current study aims at making use of the safety analysis or evaluation of a future marine water reactor by developing an intelligent information database program concerned with the thermal-hydraulic characteristics of an integral/passive-safety reactor on the basis of the above-mentioned valuable experimental knowledge. Since the program was created as a Windows application using the Visual Basic, it is available to the public and can be easily installed in the operating system. Main functions of the program are as follows: (1) steady state flow boiling analysis and determination of stability limit for any helical-coil type once-through steam generator design. (2) analysis and comparison with the flow boiling data, (3) reference and graphic display of the experimental data, (4) indication of the knowledge information such as analysis method and results of the study. The program will be useful for the design of not only the future integrated type marine water reactor but also the small sized water reactor. (author)

  18. Adjoint sensitivity analysis of the RELAPS/MOD3.2 two-fluid thermal-hydraulic code system

    International Nuclear Information System (INIS)

    Ionescu-Bujor, M.

    2000-10-01

    This work presents the implementation of the Adjoint Sensitivity Analysis Procedure (ASAP) for the non-equilibrium, non-homogeneous two-fluid model, including boron concentration and non-condensable gases, of the RELAP5/MOD3.2 code. The end-product of this implementation is the Adjoint Sensitivity Model (ASM-REL/TF), which is derived for both the differential and discretized equations underlying the two-fluid model with non-condensable(s). The consistency requirements between these two representations are also highlighted. The validation of the ASM-REL/TF has been carried out by using sample problems involving: (i) liquid-phase only, (ii) gas-phase only, and (iii) two-phase mixture (of water and steam). Thus the 'Two-Loops with Pumps' sample problem supplied with RELAP5/MOD3.2 has been used to verify the accuracy and stability of the numerical solution of the ASM-REL/TF when only the liquid-phase is present. Furthermore, the 'Edwards Pipe' sample problem, also supplied with RELAP5/MOD3.2, has been used to verify the accuracy and stability of the numerical solution of the ASM-REL/TF when both (i.e., liquid and gas) phases are present. In addition, the accuracy and stability of the numerical solution of the ASM-REL/TF have been verified when only the gas-phase is present by using modified 'Two-Loops with Pumps' and the 'Edwards Pipe' sample problems in which the liquid and two-phase fluids, respectively, were replaced by pure steam. The results obtained for these sample problems depict typical sensitivities of junction velocities and volume-averaged pressures to perturbations in initial conditions, and indicate that the numerical solution of the ASM-REL/TF is as robust, stable, and accurate as the original RELAP5/MOD3.2 calculations. In addition, the solution of the ASM-REL/TF has been used to calculate sample sensitivities of volume-averaged pressures to variations in the pump head. (orig.) [de

  19. Development of thermal hydraulic evaluation code for CANDU reactors

    International Nuclear Information System (INIS)

    Kim, Man Woong; Yu, Seon Oh; Choi, Yong Seog; Shin, Chull; Hwang, Soo Hyun

    2004-02-01

    To enhance the safety of operating CANDU reactors, the establishment of the safety analysis codes system for CANDU reactors is in progress. As for the development of thermal-hydraulic analysis code for CANDU system, the studies for improvement of evaluation model inside RELAP/CANDU code and the development of safety assessment methodology for GSI (Generic Safety Issues) are in progress as a part of establishment of CANDU safety assessment system. To develop the 3-D thermal-hydraulic analysis code for moderator system, the CFD models for analyzing the CANDU-6 moderator circulation are developed. One model uses a structured grid system with the porous media approach for the 380 Calandria tubes in the core region. The other uses a unstructured grid system on the real geometry of 380 Calandria tubes, so that the detailed fluid flow between the Calandria tubes can be observed. As to the development of thermal-hydraulic analysis code for containment, the study on the applicability of CONTAIN 2.0 code to a CANDU containment was conducted and a simulation of the thermal-hydraulic phenomena during the accident was performed. Besides, the model comparison of ESFs (Engineered Safety Features) inside CONTAIN 2.0 code and PRESCON code has also conducted

  20. HYDRAULIC SERVO CONTROL MECHANISM

    Science.gov (United States)

    Hussey, R.B.; Gottsche, M.J. Jr.

    1963-09-17

    A hydraulic servo control mechanism of compact construction and low fluid requirements is described. The mechanism consists of a main hydraulic piston, comprising the drive output, which is connected mechanically for feedback purposes to a servo control piston. A control sleeve having control slots for the system encloses the servo piston, which acts to cover or uncover the slots as a means of controlling the operation of the system. This operation permits only a small amount of fluid to regulate the operation of the mechanism, which, as a result, is compact and relatively light. This mechanism is particuiarly adaptable to the drive and control of control rods in nuclear reactors. (auth)

  1. Some safety studies of the MEGAPIE spallation source target performed using computational fluid dynamics

    International Nuclear Information System (INIS)

    Smith, B.L.

    2011-01-01

    Such a target forms part of the evolutionary Accelerator-Driven System (ADS) concept in which neutrons are generated in an otherwise sub-critical core by spallation reactions resulting from bombardment by a proton beam. The international project MEGAPIE had the objective of demonstrating the feasibility of the spallation process for a particular target design under strict test conditions. The test was carried over a period of four months at the end of 2006 at the SINQ facility of the Paul Scherrer Institute in Switzerland. The design studies carried out for the MEGAPIE target prior to irradiation using Computational Fluid Dynamics (CFD) resulted in an optimum flow configuration being defined for the coolant circulation. Simultaneously, stresses in the structural components were examined using Finite Element Method (FEM) techniques. To this purpose, an interface program was written which enabled different specialist groups to carry out the thermal hydraulics and structural mechanics analyses within the project with fully consistent model data. Results for steady-state operation of the target show that the critical lower target components are adequately cooled, and that stresses and displacements are well within tolerances. Transient analyses were also performed to demonstrate the robustness of the design in the event of abnormal operation, including pump failure and burn-through of the target casing by the proton beam. In the latter case, the CFD analyses complemented and extended full-scale tests. (author)

  2. Some safety studies of the MEGAPIE spallation source target performed using computational fluid dynamics

    Energy Technology Data Exchange (ETDEWEB)

    Smith, B.L., E-mail: brian.smith@psi.ch [Paul Scherrer Institute, OHSA/C08, 5232 Villigen PSI (Switzerland)

    2011-07-01

    Such a target forms part of the evolutionary Accelerator-Driven System (ADS) concept in which neutrons are generated in an otherwise sub-critical core by spallation reactions resulting from bombardment by a proton beam. The international project MEGAPIE had the objective of demonstrating the feasibility of the spallation process for a particular target design under strict test conditions. The test was carried over a period of four months at the end of 2006 at the SINQ facility of the Paul Scherrer Institute in Switzerland. The design studies carried out for the MEGAPIE target prior to irradiation using Computational Fluid Dynamics (CFD) resulted in an optimum flow configuration being defined for the coolant circulation. Simultaneously, stresses in the structural components were examined using Finite Element Method (FEM) techniques. To this purpose, an interface program was written which enabled different specialist groups to carry out the thermal hydraulics and structural mechanics analyses within the project with fully consistent model data. Results for steady-state operation of the target show that the critical lower target components are adequately cooled, and that stresses and displacements are well within tolerances. Transient analyses were also performed to demonstrate the robustness of the design in the event of abnormal operation, including pump failure and burn-through of the target casing by the proton beam. In the latter case, the CFD analyses complemented and extended full-scale tests. (author)

  3. Hydraulic limits on maximum plant transpiration and the emergence of the safety-efficiency trade-off.

    Science.gov (United States)

    Manzoni, Stefano; Vico, Giulia; Katul, Gabriel; Palmroth, Sari; Jackson, Robert B; Porporato, Amilcare

    2013-04-01

    Soil and plant hydraulics constrain ecosystem productivity by setting physical limits to water transport and hence carbon uptake by leaves. While more negative xylem water potentials provide a larger driving force for water transport, they also cause cavitation that limits hydraulic conductivity. An optimum balance between driving force and cavitation occurs at intermediate water potentials, thus defining the maximum transpiration rate the xylem can sustain (denoted as E(max)). The presence of this maximum raises the question as to whether plants regulate transpiration through stomata to function near E(max). To address this question, we calculated E(max) across plant functional types and climates using a hydraulic model and a global database of plant hydraulic traits. The predicted E(max) compared well with measured peak transpiration across plant sizes and growth conditions (R = 0.86, P efficiency trade-off in plant xylem. Stomatal conductance allows maximum transpiration rates despite partial cavitation in the xylem thereby suggesting coordination between stomatal regulation and xylem hydraulic characteristics. © 2013 The Authors. New Phytologist © 2013 New Phytologist Trust.

  4. Vibration of hydraulic machinery

    CERN Document Server

    Wu, Yulin; Liu, Shuhong; Dou, Hua-Shu; Qian, Zhongdong

    2013-01-01

    Vibration of Hydraulic Machinery deals with the vibration problem which has significant influence on the safety and reliable operation of hydraulic machinery. It provides new achievements and the latest developments in these areas, even in the basic areas of this subject. The present book covers the fundamentals of mechanical vibration and rotordynamics as well as their main numerical models and analysis methods for the vibration prediction. The mechanical and hydraulic excitations to the vibration are analyzed, and the pressure fluctuations induced by the unsteady turbulent flow is predicted in order to obtain the unsteady loads. This book also discusses the loads, constraint conditions and the elastic and damping characters of the mechanical system, the structure dynamic analysis, the rotor dynamic analysis and the system instability of hydraulic machines, including the illustration of monitoring system for the instability and the vibration in hydraulic units. All the problems are necessary for vibration pr...

  5. Thermal hydraulic analysis of aggressive secondary cooldown in a small break loss of coolant accident with a total loss of high pressure safety injection

    International Nuclear Information System (INIS)

    Han, Seok Jung; Lim, Ho Gon; Yang, Joon Eon

    2003-01-01

    To support the development of a Probabilistic Safety Assessment (PSA) model usable in Riskinformed Applications (RIA) for Korea Standard Nuclear power Plants (KSNP), we have performed a thermal hydraulic analysis of Aggressive Secondary Cooldown (ASC) in a 2-inch Small Break Loss Of Coolant Accident (SBLOCA) with a total loss of High Pressure Safety Injection (HPSI). The present study focuses on the estimation of the success criteria of ASC, and the enhanced understanding of the detailed thermal hydraulic behavior and phenomena. The results have shown that the Reactor Coolant System (RCS) pressure can be reduced to the Low Pressure Safety Injection (LPSI) operation conditions without core damage. It was also shown that more relaxed success criteria compared to those in the previous PSA models of KSNP could be used in the new PSA model. However, it was found that the results could be affected by various parameters related with ASC operation, i.e., reference temperature for the calculation of the cooldown rate and its control method

  6. Thermal hydraulic analysis of aggressive secondary cooldown in small break loss of coolant accident with total loss of high pressure safety injection

    International Nuclear Information System (INIS)

    Han, S. J.; Im, H. K.; Yang, J. U.

    2003-01-01

    Recently, Probabilistic Safety Assessment (PSA) has being applied to various fields as a basic technique of Risk-Informed Applications (RIA). To use RIA, the present study focuses on the detailed thermal hydraulic analyses for major accident sequences and success criteria to support a development of PSA model for Korea Standard Nuclear Power plant (KSNP). The primary purpose of the present study is to evaluate the success criteria of Aggressive Secondary Cooldown (ASC) in Small Break Loss Of Coolant Accident (SBLOCA) with total loss of High Pressure Safety Injection (HPSI) and to enhance the understanding of related thermal hydraulic behavior and phenomena. The accident scenario was 2 inch coldleg break LOCA without HPSI, with 1/2 Low Pressure Safety Injection (LPSI), and performing ASC limited by 55.6 .deg. C /hr (100 .deg. F/hr) cooldown rate at 15 minute after reactor trip, which successively reaches the LPSI condition for about 1.5hr after starting ASC operation with the Peak Cladding Temperature (PCT) of the hottest rod below the core damage criteria 1204.4 .deg. C (2200 .deg. F). In the present study, more relaxed success criteria than the previous PSA for KSNP could be generated under an assumption that operator should maintain the adequate ASC operation. However, it is necessary to evaluate uncertainties arisen from the related parameters of the ASC operation

  7. Twenty-fifth water reactor safety information meeting: Proceedings. Volume 3: Thermal hydraulic research and codes; Digital instrumentation and control; Structural performance

    International Nuclear Information System (INIS)

    Monteleone, S.

    1998-04-01

    This three-volume report contains papers presented at the conference. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Japan, Norway, and Russia. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. This volume contains the following: (1) thermal hydraulic research and codes; (2) digital instrumentation and control; (3) structural performance

  8. The Process of Hydraulic Fracturing

    Science.gov (United States)

    Hydraulic fracturing, know as fracking or hydrofracking, produces fractures in a rock formation by pumping fluids (water, proppant, and chemical additives) at high pressure down a wellbore. These fractures stimulate the flow of natural gas or oil.

  9. Water Hydraulic Systems

    DEFF Research Database (Denmark)

    Conrad, Finn

    2005-01-01

    The paper presents research results using IT-Tools for CAD and dynamic modelling, simulation, analysis, and design of water hydraulic actuators for motion control of machines, lifts, cranes and robots. Matlab/Simulink and CATIA are used as IT-Tools. The contributions include results from on......-going research projects on fluid power and mechatronics based on tap water hydraulic servovalves and linear servo actuators and rotary vane actuators for motion control and power transmission. Development and design a novel water hydraulic rotary vane actuator for robot manipulators. Proposed mathematical...... modelling, control and simulation of a water hydraulic rotary vane actuator applied to power and control a two-links manipulator and evaluate performance. The results include engineering design and test of the proposed simulation models compared with IHA Tampere University’s presentation of research...

  10. Preparation of a thermal-hydraulic design method for driver core fuel pins of a new in-pile experimental reactor for FBR safety research

    International Nuclear Information System (INIS)

    Mizuno, Masahiro; Yamaguchi, Katsuhisa; Uto, Nariaki

    1999-07-01

    A design study of a new in-pile experimental reactor, SERAPH (Safety Engineering Reactor for Accident PHenomenology), for FBR safety research has progressed at JNC (Japan Nuclear Cycle Development Institute). SERAPH is intended for various in-pile experiments to be performed under quasi-steady state and various transient operation modes. In order to evaluate the driver core performance in conducting such experiments, clarify the relating design issues to be resolved and refine the experimental needs, it is indispensable to comprehend the allowable margin for the thermal-hydraulic fuel pin design since it largely affects the strategy for the driver core design. This report presents a thermal-hydraulic design method for the driver core fuel pins, which is a combination of a two-dimensional time-dependent heat transfer analysis code TAC-2D and a general non-linear finite-element structural analysis code FINAS. In TAC-2D, the allowable spatial mesh and the time step sizes are evaluated. The code is modified so as to treat time-dependent thermal properties, include an improved gap heat-transfer model and treat the change of intra-pin gap width under transient modes, for the purpose of improving the accuracy of evaluating heat transfer characteristics which gives a significant impact on the thermal-hydraulic design. As for FINAS, the number of element nodes and spatial meshes required to obtain adequate accuracy for the thermal stress characteristics of a fuel pellet during transient modes are investigated. In addition, post-processing tools are newly developed to process the calculation results obtained from these codes. The results of this work contribute to advancing the fuel pin design study for SERAPH as well with the investigation on the technique of manufacturing fuel pins. (author)

  11. Computational fluid dynamic simulation of pressurizer safety valve loop seal purge phenomena in nuclear power plants

    International Nuclear Information System (INIS)

    Park, Jong Woon

    2012-01-01

    In Korean 3 Loop plants a water loop seal pipe is installed containing condensed water upstream of a pressurizer safety valve to protect the valve disk from the hot steam environment. The loop seal water purge time is a key parameter in safety analyses for overpressure transients, because it delays valve opening. The loop seal purge time is uncertain to measure by test and thus 3-dimensional realistic computational fluid dynamics (CFD) model is developed in this paper to predict the seal water purge time before full opening of the valve which is driven by steam after water purge. The CFD model for a typical pressurizer safety valve with a loop seal pipe is developed using the computer code of ANSYS CFX 11. Steady-state simulations are performed for full discharge of steam at the valve full opening. Transient simulations are performed for the loop seal dynamics and to estimate the loop seal purge time. A sudden pressure drop higher than 2,000 psia at the tip of the upper nozzle ring is expected from the steady-state calculation. Through the transient simulation, almost loop seal water is discharged within 1.2 second through the narrow opening between the disk and the nozzle of the valve. It can be expected that the valve fully opens at least before 1.2 second because constant valve opening is assumed in this CFX simulation, which is conservative because the valve opens fully before the loop seal water is completely discharged. The predicted loop seal purge time is compared with previous correlation. (orig.)

  12. Computational fluid dynamic simulation of pressurizer safety valve loop seal purge phenomena in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Woon [Dongguk Univ., Gyeongju (Korea, Republic of). Nuclear and Energy Engineering Dept.

    2012-11-15

    In Korean 3 Loop plants a water loop seal pipe is installed containing condensed water upstream of a pressurizer safety valve to protect the valve disk from the hot steam environment. The loop seal water purge time is a key parameter in safety analyses for overpressure transients, because it delays valve opening. The loop seal purge time is uncertain to measure by test and thus 3-dimensional realistic computational fluid dynamics (CFD) model is developed in this paper to predict the seal water purge time before full opening of the valve which is driven by steam after water purge. The CFD model for a typical pressurizer safety valve with a loop seal pipe is developed using the computer code of ANSYS CFX 11. Steady-state simulations are performed for full discharge of steam at the valve full opening. Transient simulations are performed for the loop seal dynamics and to estimate the loop seal purge time. A sudden pressure drop higher than 2,000 psia at the tip of the upper nozzle ring is expected from the steady-state calculation. Through the transient simulation, almost loop seal water is discharged within 1.2 second through the narrow opening between the disk and the nozzle of the valve. It can be expected that the valve fully opens at least before 1.2 second because constant valve opening is assumed in this CFX simulation, which is conservative because the valve opens fully before the loop seal water is completely discharged. The predicted loop seal purge time is compared with previous correlation. (orig.)

  13. Mechanics of Hydraulic Fractures

    Science.gov (United States)

    Detournay, Emmanuel

    2016-01-01

    Hydraulic fractures represent a particular class of tensile fractures that propagate in solid media under pre-existing compressive stresses as a result of internal pressurization by an injected viscous fluid. The main application of engineered hydraulic fractures is the stimulation of oil and gas wells to increase production. Several physical processes affect the propagation of these fractures, including the flow of viscous fluid, creation of solid surfaces, and leak-off of fracturing fluid. The interplay and the competition between these processes lead to multiple length scales and timescales in the system, which reveal the shifting influence of the far-field stress, viscous dissipation, fracture energy, and leak-off as the fracture propagates.

  14. Hydraulic hoisting and backfilling

    Science.gov (United States)

    Sauermann, H. B.

    In a country such as South Africa, with its large deep level mining industry, improvements in mining and hoisting techniques could result in substantial savings. Hoisting techniques, for example, may be improved by the introduction of hydraulic hoisting. The following are some of the advantages of hydraulic hoisting as against conventional skip hoisting: (1) smaller shafts are required because the pipes to hoist the same quantity of ore hydraulically require less space in the shaft than does skip hoisting equipment; (2) the hoisting capacity of a mine can easily be increased without the necessity of sinking new shafts. Large savings in capital costs can thus be made; (3) fully automatic control is possible with hydraulic hoisting and therefore less manpower is required; and (4) health and safety conditions will be improved.

  15. 75 FR 5553 - Federal Motor Vehicle Safety Standards; Motor Vehicle Brake Fluids

    Science.gov (United States)

    2010-02-03

    ... NHTSA comments S6.2 Wet Equilibrium Boiling Point... Appendix E of SAE J1703 Appendix E of SAE J1703 No.... Definition of ``Brake Fluid'' To apply FMVSS No. 116 to brake fluid that contacts EPDM rubber, we propose to expand the definition of ``brake fluid'' at S4 of the standard to expressly state that ``brake fluid...

  16. Thermal hydraulic analysis of aggressive secondary cooldown in a small break loss of coolant accident with a total loss of high pressure safety injection

    International Nuclear Information System (INIS)

    Han, Seok Jung; Lim, Ho Gon; Yang, Joon Eon

    2003-03-01

    Recently, Probabilistic Safety Assessment (PSA) has being applied to various fields as a basic technique of Risk-Informed Applications (RIA). The present study focuses on detailed thermal hydraulic analyses for major accident sequences and success criteria to support a development of PSA model using RIA for Korea Standard Nuclear Power plant (KSNP). The primary purpose of the present study in this year is to evaluate the success cri-teria of Aggressive Secondary Cooldown (ASC) in a Small Size Loss Of Coolant Accident (SBLOCA) without HPSI and to enhance the understanding of related thermal hydraulic behavior and phenomena. An effort was made to evaluate the system success criteria and a mission time for the recovery action by an operator to prevent the core damage for that accident scenario. The accident scenario for KSNP was a 2 inch coldleg break LOCA with a total loss of High Pressure Safety Injection (HPSI) and 1/2 Low Pressure Safety Injection (LPSI) available and perform-ing ASC limited by 55.6 .deg. C/hr (100 .deg. F/hr) cooldown rate at 15 minute after reactor trip. It successively reached the LPSI condition for about 1.5hr after starting the ASC operation with the Peak Cladding Temperature (PCT) of the hottest rod below the core damage criteria of 1204.4 .deg. C (2200 .deg. F). Sensitivity studies were performed for (1) cool-ant average temperature parameters, (2) ASC operation control method, (3) operation start time, (4) 1 inch break size. The present analysis identified thermal hydraulic phenomena and parameters affecting on the behavior, which consist of coolant break flow and inventory, parameters governing secondary heat removal, ASC operation control method, and its reference temperature parameters. In the present study, more relaxed success criteria than the previous PSA for KSNP could be generated under an assumption that an operator should maintain the ade-quate ASC operation. However, it is necessary to evaluate the uncertainties arisen from the

  17. Hydraulic structures

    CERN Document Server

    Chen, Sheng-Hong

    2015-01-01

    This book discusses in detail the planning, design, construction and management of hydraulic structures, covering dams, spillways, tunnels, cut slopes, sluices, water intake and measuring works, ship locks and lifts, as well as fish ways. Particular attention is paid to considerations concerning the environment, hydrology, geology and materials etc. in the planning and design of hydraulic projects. It also considers the type selection, profile configuration, stress/stability calibration and engineering countermeasures, flood releasing arrangements and scouring protection, operation and maintenance etc. for a variety of specific hydraulic structures. The book is primarily intended for engineers, undergraduate and graduate students in the field of civil and hydraulic engineering who are faced with the challenges of extending our understanding of hydraulic structures ranging from traditional to groundbreaking, as well as designing, constructing and managing safe, durable hydraulic structures that are economical ...

  18. Subsea Hydraulic Leakage Detection and Diagnosis

    OpenAIRE

    Stavenes, Thomas

    2010-01-01

    The motivation for this thesis is reduction of hydraulic emissions, minimizing of process emergency shutdowns, exploitation of intervention capacity, and reduction of costs. Today, monitoring of hydraulic leakages is scarce and the main way to detect leakage is the constant need for filling of hydraulic fluid to the Hydraulic Power Unit (HPU). Leakage detection and diagnosis has potential, which would be adressed in this thesis. A strategy towards leakage detection and diagnosis is given....

  19. Analysis for thermal fluid dynamics in downcomer of JAERI passive safety reactor (JPSR)

    International Nuclear Information System (INIS)

    Kunii, Katsuhiko; Iwamura, Takamichi; Murao, Yoshio

    1995-01-01

    The driving-force of the natural circulation in the residual heat removal system for the JPSR (JAERI Passive Safety Reactor) under a steady condition is given as a gravity force based on the density (temperature) difference between hotter coolant in core and upper plenum and cooler coolant in downcomer. The downcomer is a very important flow pass in the system to obtain the enough driving-force because the flow pass has a three-dimensional annulus geometry long in vertical and circumference directions respectively and narrow in radius direction so that the thermal fluid flow pattern in downcomer directly relates to generation of the density difference. The density difference could naturally become smaller unless the coolant flowing into downcomer spreads widely in the whole region of it. The numerical analysis has been performed taking account of the downcomer being a three-dimensional annulus flow pass with the purposes to investigate the possibilities of the followings: (1) promotion of making the flow pattern and temperature distribution uniform in downcomer by applying a mechanical device at the inlet part of downcomer (installing a baffle) to increase the driving-force of the natural circulation, (2) achievement of an enough driving-force of the natural circulation to remove the residual heat, (3) approximation of three-dimensional thermal fluid flow in downcomer to simple one-dimensional one assumed on the preliminary design of the passive residual heat removal system. The following conclusions were obtained: (1) The effect of the baffle on the driving-force of natural circulation is little being considered due to the enhancing of mixing on thermal fluid flow in case with baffle, (2) Though the flow pattern becomes three-dimensional in some case such as large vortex flow not to be able to approximate simply to one-dimensional, the required driving-force can be obtained, (3) The driving-force can be estimated as the almost same functional value for time

  20. Co-ordination of growth, gas exchange and hydraulics define the carbon safety margin in tree species with contrasting drought strategies.

    Science.gov (United States)

    Mitchell, P J; O'Grady, A P; Tissue, D T; Worledge, D; Pinkard, E A

    2014-05-01

    Gas exchange, growth, water transport and carbon (C) metabolism diminish during drought according to their respective sensitivities to declining water status. The timing of this sequence of declining physiological functions may determine how water and C relations compromise plant survival. In this paper, we test the hypothesis that the degree of asynchrony between declining C supply (photosynthesis) and C demand (growth and respiration) determines the rate and magnitude of changes in whole-plant non-structural carbohydrates (NSC) during drought. Two complementary experiments using two tree species (Eucalyptus globulus Labill. and Pinus radiata D. Don) with contrasting drought response strategies were performed to (i) assess changes in radial stem growth, transpiration, leaf water potential and gas exchange in response to chronic drought, and (ii) evaluate the concomitant impacts of these drought responses on the temporal patterns of NSC during terminal drought. The three distinct phases of water stress were delineated by thresholds of growth cessation and stomatal closure that defined the 'carbon safety margin' (i.e., the difference between leaf water potential when growth is zero and leaf water potential when net photosynthesis is zero). A wider C safety margin in E. globulus was defined by an earlier cessation of growth relative to photosynthesis that reduced the demand for NSC while maintaining C acquisition. By contrast, the narrower C safety margin in P. radiata was characterized by a synchronous decline in growth and photosynthesis, whereby growth continued under a declining supply of NSC from photosynthesis. The narrower C safety margin in P. radiata was associated with declines in starch concentrations after ∼ 90 days of chronic drought and significant depletion of starch in all organs at mortality. The observed divergence in the sensitivity of drought responses is indicative of a potential trade-off between maintaining hydraulic safety and adequate C

  1. Statistically based uncertainty analysis for ranking of component importance in the thermal-hydraulic safety analysis of the Advanced Neutron Source Reactor

    International Nuclear Information System (INIS)

    Wilson, G.E.

    1992-01-01

    The Analytic Hierarchy Process (AHP) has been used to help determine the importance of components and phenomena in thermal-hydraulic safety analyses of nuclear reactors. The AHP results are based, in part on expert opinion. Therefore, it is prudent to evaluate the uncertainty of the AHP ranks of importance. Prior applications have addressed uncertainty with experimental data comparisons and bounding sensitivity calculations. These methods work well when a sufficient experimental data base exists to justify the comparisons. However, in the case of limited or no experimental data the size of the uncertainty is normally made conservatively large. Accordingly, the author has taken another approach, that of performing a statistically based uncertainty analysis. The new work is based on prior evaluations of the importance of components and phenomena in the thermal-hydraulic safety analysis of the Advanced Neutron Source Reactor (ANSR), a new facility now in the design phase. The uncertainty during large break loss of coolant, and decay heat removal scenarios is estimated by assigning a probability distribution function (pdf) to the potential error in the initial expert estimates of pair-wise importance between the components. Using a Monte Carlo sampling technique, the error pdfs are propagated through the AHP software solutions to determine a pdf of uncertainty in the system wide importance of each component. To enhance the generality of the results, study of one other problem having different number of elements is reported, as are the effects of a larger assumed pdf error in the expert ranks. Validation of the Monte Carlo sample size and repeatability are also documented

  2. Calculation of dynamic hydraulic forces in nuclear plant piping systems

    International Nuclear Information System (INIS)

    Choi, D.K.

    1982-01-01

    A computer code was developed as one of the tools needed for analysis of piping dynamic loading on nuclear power plant high energy piping systems, including reactor safety and relief value upstream and discharge piping systems. The code calculates the transient hydraulic data and dynamic forces within the one-dimensional system, caused by a pipe rupture or sudden value motion, using a fixed space and varying time grid-method of characteristics. Subcooled, superheated, homogeneous two-phase and transition flow regimes are considered. A non-equilibrium effect is also considered in computing the fluid specific volume and fluid local sonic velocity in the two-phase mixture. Various hydraulic components such as a spring loaded or power operated value, enlarger, orifice, pressurized tank, multiple pipe junction (tee), etc. are considered as boundary conditions. Comparisons of calculated results with available experimental data shows a good agreement. (Author)

  3. Thermal fluid flow analysis in downcomer of JAERI passive safety light water reactor (JPSR)

    International Nuclear Information System (INIS)

    Kunii, K.; Iwamura, T.; Murao, Y.

    1995-01-01

    The residual heat for the JPSR (JAERI Passive Safety Light Water Reactor) is removed by a natural-circulation of coolant flowing through downcomer. The numerical analysis has been performed taking account of the downcomer being a three-dimensional annulus flow pass with the purposes to confirm the abilities of (1) approximation of three-dimensional thermal fluid flow in downcomer to simple one-dimensional one assumed on the preliminary design of the passive residual heat removal system and (2) achievement of an enough driving-force of the natural circulation to remove the residual heat. The following results were obtained : (1) Flow pattern in downcomer shows remarkable three-dimensionality (multi-dimensionality) at lower inlet flow rate not to be able to approximate to one-dimensional flow field. However, the temperature distribution does not deviate from uniform one so much even if the multi-dimensional flow such as large vortex arises. (2) It can be expected to obtain the required enough driving-force at a steady state in any case of inlet flow rate where multi-dimensional flow pattern appears. (3) The increase ratio of the driving-force with the time-integrated coolant amount can be estimated as two functional curves in case of higher and other lower inlet flow rates not dependent only on the respective inlet flow rate. (Author)

  4. Hydraulic Hybrid Fleet Vehicle Testing | Transportation Research | NREL

    Science.gov (United States)

    Hydraulic Hybrid Fleet Vehicle Evaluations Hydraulic Hybrid Fleet Vehicle Evaluations How Hydraulic Hybrid Vehicles Work Hydraulic hybrid systems can capture up to 70% of the kinetic energy that would -pressure reservoir to a high-pressure accumulator. When the vehicle accelerates, fluid in the high-pressure

  5. A two-step approach for the preliminary evaluation of the thermal-hydraulics and safety of the ELSY open square core design

    International Nuclear Information System (INIS)

    Meloni, Paride; Bandini, Giacomino; Polidori, Massimiliano; Cervone, Antonio; Manservisi, Sandro

    2009-01-01

    Several innovative solutions for a liquid metal fast reactor design have been investigated in the EURATOM Sixth Framework Programme and an open-assembly core design for the ELSY (European Lead-cooled System) reactor has been proposed by ENEA. The development of this new reactor, based on innovative neutronic and safety considerations, requires a new approach to the thermal-hydraulic (T/H) core design. In this paper a new two-step approach of the T/H analysis for this open-assembly core is presented and, in particular is used for the evaluation of the preliminary core design of a 1500 MW lead fast reactor with open square lattice and three fuel radial zones with different levels of enrichment. In the first step a preliminary thermal-hydraulic and safety evaluation of the core neutronic design is investigated by using a one-dimensional RELAP5 model for independent channel analysis. Then two and three-dimensional effects are taken into account by using a dedicated tool for the evaluation of assembly mixing effects. The RELAP5 model, based on pressure loss and heat transfer correlations available for heavy liquid metal flows in rod bundle, consists of completely independent assemblies and therefore it can be used for a conservative evaluation of the thermal-hydraulics of the core reactor. Due to the open-lattice configuration, the two and three-dimensional effects are important and they are taken into account by using a simplified three-dimensional numerical model of an open square lattice reactor core, developed with the purpose of analyzing the whole core behavior. The numerical simulation is performed at assembly length level taking into account the local fluctuations of turbulent viscosity and energy exchange coefficients at sub-channel level through transfer operators based on parametric coefficients. A preliminary evaluation of the mixing effects between assembly flows on the temperature field has been performed by using an average assembly turbulent viscosity

  6. Hydraulic turbines

    International Nuclear Information System (INIS)

    Meluk O, G.

    1998-01-01

    The hydraulic turbines are defined according to the specific speed, in impulse turbines and in reaction turbines. Currently, the Pelton turbines (of impulse) and the Francis and Kaplan turbines (of reaction), they are the most important machines in the hydroelectric generation. The hydraulic turbines are capable of generating in short times, large powers, from its loads zero until the total load and reject the load instantly without producing damages in the operation. When the hydraulic resources are important, the hydraulic turbines are converted in the axle of the electric system. Its combination with thermoelectric generation systems, it allow the continuing supply of the variations in demand of energy system. The available hydraulic resource in Colombia is of 93085 MW, of which solely 9% is exploited, become 79% of all the electrical country generation, 21% remaining is provided by means of the thermoelectric generation

  7. Thermal-Hydraulic Effects of Stud Shape and Size on the Safety Margin of Core Catcher System

    Energy Technology Data Exchange (ETDEWEB)

    Song, Kyusang; Son, Hong Hyun; Jeong, Uiju; Kim, Sung Joong [Hanyang University, Seoul (Korea, Republic of)

    2015-10-15

    With the ERVC strategy, an additional system (core catcher system) to catch molten core penetrating the reactor pressure vessel (RPV) was proposed for advanced light water reactor. The newly engineered corium cooling system, that is, an ex-vessel core catcher system has been designed and adapted in some nuclear power plants such as VVER-1000, EPR, ESBWR, EU-APR1400 to mention a few. For example, Russia adopted a crucible-type core catcher for VVER-1000. On the other hand, a way to catch melt spreading is adopted by several countries, such as EPR in France, ESBWR in USA, ABWR in japan, and EU-APR1400 in Korea In Korea, the core catcher system has been designed and implemented for the European Advanced Power Reactor 1400 (EU-APR1400) to acquire a European license certificate. It is to confine molten materials in the reactor cavity while maintaining a coolable geometry in case that RPV failure occurs. The core catcher system consists of a carbon steel body, sacrificial material, protection material and engineered cooling channel. While installation of the studs is unavoidable, the studs tend to interfere in the smooth streamline of the core catcher channel. The distorted streamline could affect the overall thermal-hydraulic performance including two-phase heat transfer coefficient and critical heat flux (CHF) of the system. Thus, it is of importance to investigate the thermal-hydraulic effects of studs on the coolability, especially the CHF of the core catcher system. With aforementioned importance, pool boiling experiments were carried out with stud shape of, rectangular, cylinder, and elliptic and for stud sizes of 10, 15, 20, and 25 mm under the condition of atmospheric saturated water. A particular attention was focused on observing local vapor behavior around the studs and finding any hot spots, where the vapors are accumulated. The occurrence of the CHF is anticipated at the back side of the studs. The visual observation and CHF measurements indicate that the

  8. Thermal-Hydraulic Effects of Stud Shape and Size on the Safety Margin of Core Catcher System

    International Nuclear Information System (INIS)

    Song, Kyusang; Son, Hong Hyun; Jeong, Uiju; Kim, Sung Joong

    2015-01-01

    With the ERVC strategy, an additional system (core catcher system) to catch molten core penetrating the reactor pressure vessel (RPV) was proposed for advanced light water reactor. The newly engineered corium cooling system, that is, an ex-vessel core catcher system has been designed and adapted in some nuclear power plants such as VVER-1000, EPR, ESBWR, EU-APR1400 to mention a few. For example, Russia adopted a crucible-type core catcher for VVER-1000. On the other hand, a way to catch melt spreading is adopted by several countries, such as EPR in France, ESBWR in USA, ABWR in japan, and EU-APR1400 in Korea In Korea, the core catcher system has been designed and implemented for the European Advanced Power Reactor 1400 (EU-APR1400) to acquire a European license certificate. It is to confine molten materials in the reactor cavity while maintaining a coolable geometry in case that RPV failure occurs. The core catcher system consists of a carbon steel body, sacrificial material, protection material and engineered cooling channel. While installation of the studs is unavoidable, the studs tend to interfere in the smooth streamline of the core catcher channel. The distorted streamline could affect the overall thermal-hydraulic performance including two-phase heat transfer coefficient and critical heat flux (CHF) of the system. Thus, it is of importance to investigate the thermal-hydraulic effects of studs on the coolability, especially the CHF of the core catcher system. With aforementioned importance, pool boiling experiments were carried out with stud shape of, rectangular, cylinder, and elliptic and for stud sizes of 10, 15, 20, and 25 mm under the condition of atmospheric saturated water. A particular attention was focused on observing local vapor behavior around the studs and finding any hot spots, where the vapors are accumulated. The occurrence of the CHF is anticipated at the back side of the studs. The visual observation and CHF measurements indicate that the

  9. Proceedings of the US Nuclear Regulatory Commission twentieth water reactor safety information meeting; Volume 2, Severe accident research, Thermal hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Weiss, A.J. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1993-03-01

    This three-volume report contains papers presented at the Twentieth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 21--23, 1992. The papers describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included 10 different papers presented by researchersfrom CEC, China, Finland, France, Germany, Japan, Spain and Taiwan. Selected papers have been processed separately for inclusion in the Energy Science and Technology Database.

  10. Experimental Study of Hydraulic Control Rod Drive Mechanism for Passive IN-core Cooling System of Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Kim, In Guk; Kim, Kyung Mo; Jeong, Yeong Shin; Bang, In Cheol [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    CAREM 25 (27 MWe safety systems using hydraulic control rod drives (CRD) studied critical issues that were rod drops with interrupted flow [3]. Hydraulic control rod drive suggested fast shutdown condition using a large gap between piston and cylinder in order to fast drop of neutron absorbing rods. A Passive IN-core Cooling system (PINCs) was suggested for safety enhancement of pressurized water reactors (PWR), small modular reactor (SMR), sodium fast reactor (SFR) in UNIST. PINCs consist of hydraulic control rod drive mechanism (Hydraulic CRDM) and hybrid control rod assembly with heat pipe combined with control rod. The schematic diagram of the hydraulic CRDM for PINCs is shown in Fig. 1. The experimental results show the steady state and transient behavior of the upper cylinder at a low pressure and low temperature. The influence of the working fluid temperature and cylinder mass are investigated. Finally, the heat removal between evaporator section and condenser section is compared with or without the hybrid control rod. Heat removal test of the hybrid heat pipe with hydraulic CRDM system showed the heat transfer coefficient of the bundle hybrid control rod and its effect on evaporator pool. The preliminary test both hydraulic CRDM and heat removal system was conducted, which showed the possibility of the in-core hydraulic drive system for application of PINCs.

  11. Hydraulic Structures

    Data.gov (United States)

    Department of Homeland Security — This table is required whenever hydraulic structures are shown in the flood profile. It is also required if levees are shown on the FIRM, channels containing the...

  12. Development of safety analysis methodology for moderator system failure of CANDU-6 reactor by thermal-hydraulics/physics coupling

    International Nuclear Information System (INIS)

    Kim, Jong Hyun; Jin, Dong Sik; Chang, Soon Heung

    2013-01-01

    Highlights: • Developed new safety analysis methodology of moderator system failures for CANDU-6. • The new methodology used the TH-physics coupling concept. • Thermalhydraulic code is CATHENA, physics code is RFSP-IST. • Moderator system failure ends to the subcriticality through self-shutdown. -- Abstract: The new safety analysis methodology for the CANDU-6 nuclear power plant (NPP) moderator system failure has been developed by using the coupling technology with the thermalhydraulic code, CATHENA and reactor core physics code, RFSP-IST. This sophisticated methodology can replace the legacy methodology using the MODSTBOIL and SMOKIN-G2 in the field of the thermalhydraulics and reactor physics, respectively. The CATHENA thermalhydraulic model of the moderator system can simulate the thermalhydraulic behaviors of all the moderator systems such as the calandria tank, head tank, moderator circulating circuit and cover gas circulating circuit and can also predict the thermalhydraulic property of the moderator such as moderator density, temperature and water level in the calandria tank as the moderator system failures go on. And these calculated moderator thermalhydraulic properties are provided to the 3-dimensional neutron kinetics solution module – CERBRRS of RFSP-IST as inputs, which can predict the change of the reactor power and provide the calculated reactor power to the CATHENA. These coupling calculations are performed at every 2 s time steps, which are equivalent to the slow control of CANDU-6 reactor regulating systems (RRS). The safety analysis results using this coupling methodology reveal that the reactor operation enters into the self-shutdown mode without any engineering safety system and/or human interventions for the postulated moderator system failures of the loss of heat sink and moderator inventory, respectively

  13. Hydraulic fracture monitoring in hard rock at 410 m depth with an advanced fluid-injection protocol and extensive sensor array

    Science.gov (United States)

    Zang, Arno; Stephansson, Ove; Stenberg, Leif; Plenkers, Katrin; Specht, Sebastian; Milkereit, Claus; Schill, Eva; Kwiatek, Grzegorz; Dresen, Georg; Zimmermann, Günter; Dahm, Torsten; Weber, Michael

    2017-02-01

    In this paper, an underground experiment at the Äspö Hard Rock Laboratory (HRL) is described. Main goal is optimizing geothermal heat exchange in crystalline rock mass at depth by multistage hydraulic fracturing with minimal impact on the environment, that is, seismic events. For this, three arrays with acoustic emission, microseismicity and electromagnetic sensors are installed mapping hydraulic fracture initiation and growth. Fractures are driven by three different water injection schemes (continuous, progressive and pulse pressurization). After a brief review of hydraulic fracture operations in crystalline rock mass at mine scale, the site geology and the stress conditions at Äspö HRL are described. Then, the continuous, single-flow rate and alternative, multiple-flow rate fracture breakdown tests in a horizontal borehole at depth level 410 m are described together with the monitoring networks and sensitivity. Monitoring results include the primary catalogue of acoustic emission hypocentres obtained from four hydraulic fractures with the in situ trigger and localizing network. The continuous versus alternative water injection schemes are discussed in terms of the fracture breakdown pressure, the fracture pattern from impression packer result and the monitoring at the arrays. An example of multistage hydraulic fracturing with several phases of opening and closing of fracture walls is evaluated using data from acoustic emissions, seismic broad-band recordings and electromagnetic signal response. Based on our limited amount of in situ tests (six) and evaluation of three tests in Ävrö granodiorite, in the multiple-flow rate test with progressively increasing target pressure, the acoustic emission activity starts at a later stage in the fracturing process compared to the conventional fracturing case with continuous water injection. In tendency, also the total number and magnitude of acoustic events are found to be smaller in the progressive treatment with

  14. Hydraulic design development of Xiluodu Francis turbine

    International Nuclear Information System (INIS)

    Wang, Y L; Li, G Y; Shi, Q H; Wang, Z N

    2012-01-01

    Hydraulic optimization design with CFD (Computational Fluid Dynamics) method, hydraulic optimization measures and model test results in the hydraulic development of Xiluodu hydropower station by DFEM (Dongfang Electric Machinery) of DEC (Dongfang Electric Corporation) of China were analyzed in this paper. The hydraulic development conditions of turbine, selection of design parameter, comparison of geometric parameters and optimization measure of turbine flow components were expatiated. And the measures of improving turbine hydraulic performance and the results of model turbine acceptance experiment were discussed in details.

  15. An experimental study on the thermal-hydraulic phenomena in the Hybrid Safety Injection Tank using a separate effect test facility

    International Nuclear Information System (INIS)

    Ryu, Sung Uk; Ryu, Hyobong; Park, Hyun-Sik; Yi, Sung-Jae

    2016-01-01

    Highlights: • The experimental study on the pressure balancing between the Hybrid SIT and PZR. • The effects of different variables affecting the pressure balancing are investigated. • A sensitivity analysis on the pressure variations of the Hybrid SIT. - Abstract: This paper reports an experimental research for investigating thermal hydraulic phenomena of Hybrid Safety Injection Tank (Hybrid SIT) using a separate effect test facility in Korea Atomic Energy Research Institute (KAERI). The Hybrid SIT is a passive safety injection system that enables the safety injection water to be injected into the reactor pressure vessel throughout all operating pressures by connecting the top of the SIT and the pressurizer (PZR). The separate effect test (SET) facility of Hybrid SIT, which is designed based on the APR+ power plant, comprises a PZR, Hybrid SIT, pressure balancing line (PBL), injection line (IL), nitrogen gas line, and refueling water tank (RWT). Furthermore, the pressure loss range of the SET facility was analyzed and compared with that of the reference nuclear power plant. In this research, a condition for balancing the pressure between the Hybrid SIT and PZR is examined and the effects of different variables affecting the pressure balancing, which are flow rate, injection velocity of steam and initial water level, are also investigated. The condition for balancing the pressure between the Hybrid SIT and PZR was derived theoretically from a pressure network for the Hybrid SIT, pressurizer, and reactor pressure vessel. Additionally, a sensitivity analysis as a theoretical approach was conducted on the pressure variations in relation to the rate of steam condensation inside the Hybrid SIT. The results showed that pressure of the Hybrid SIT was predominantly determined by the rate of steam condensation. The results showed that if the rate of condensation increased or decreased by 10%, the Hybrid SIT pressure at the pressure balancing point decreased or

  16. The hydraulics of the pressurized water reactors; L'hydraulique des reacteurs a eau pressurisee

    Energy Technology Data Exchange (ETDEWEB)

    Bouchter, J.C. [CEA Cadarache, SMET, 13 - Saint-Paul-lez-Durance (France); Barbier, D. [CEA/Grenoble, Dept. de Thermohydraulique et de Physique, DTP/SH2C, 38 (France); Caruso, A. [Electricite de France, Service Etudes et Projets Thermiques et Nucleaires, 75 - Paris (France)] [and others

    1999-07-02

    The SFEN organized, the 10 june 1999 at Paris, a meeting in the domain of the PWR hydraulics and in particular the hydraulic phenomena concerning the vessel and the vapor generators. The papers presented showed the importance of the industrial stakes with their associated phenomena: cores performance and safety with the more homogenous cooling system, the rods and the control rods wear, the temperature control, the fluid-structure interactions. A great part was also devoted to the progresses in the domain of the numerical simulation and the models and algorithms qualification. (A.L.B.)

  17. PMK-2, the First Integral Thermal-Hydraulics Tests for the Safety Evaluation of VVER-440/213 Nuclear Power Plants

    Directory of Open Access Journals (Sweden)

    Gy. Ézsöl

    2012-01-01

    Full Text Available The PMK-2 facility is a full-pressure thermal-hydraulic model of the primary and partly the secondary circuit of the VVER-type units of Paks NPP. The facility was the first integral-type facility for VVERs. The PMK-2 was followed later by the PACTEL (for VVER-440, the ISB, and PSB for VVER-1000. Since the startup of the facility in 1985, 55 experiments have been performed primarily in international frameworks with the participation of experts from 29 European and overseas countries forming a scientific school to better understand VVER system behaviour and reach a high level of modelling of accident sequences. The ATHLET, CATHARE, and RELAP5 codes have been validated including both qualitative and quantitative assessments. The former was almost exclusively applied to the early phase of validation by integral experiments, while the quantitative assessments have been performed by the Fast Fourier Transform Based Method. Paper gives comprehensive information on the design features of PMK-2 facility with a special respect to the representativeness of phenomena, the experiments performed, and the results of the validation of ATHLET, CATHARE, and RELAP5 codes. Safety significance of the PMK-2 projects is also discussed.

  18. PMK-2, the First Integral Thermal-Hydraulics Tests for the Safety Evaluation of VVER-440/213 Nuclear Power Plants

    International Nuclear Information System (INIS)

    Ezsol, G.; Perneczky, L.; Szabados, L.; Toth, I.

    2012-01-01

    The PMK-2 facility is a full-pressure thermal-hydraulic model of the primary and partly the secondary circuit of the VVER-type units of Paks NPP. The facility was the first integral-type facility for VVERs. The PMK-2 was followed later by the PACTEL (for VVER-440), the ISB, and PSB for VVER-1000. Since the startup of the facility in 1985, 55 experiments have been performed primarily in international frameworks with the participation of experts from 29 European and overseas countries forming a scientific school to better understand VVER system behaviour and reach a high level of modelling of accident sequences. The ATHLET, CATHARE, and RELAP5 codes have been validated including both qualitative and quantitative assessments. The former was almost exclusively applied to the early phase of validation by integral experiments, while the quantitative assessments have been performed by the Fast Fourier Transform Based Method. Paper gives comprehensive information on the design features of PMK-2 facility with a special respect to the representativeness of phenomena, the experiments performed, and the results of the validation of ATHLET, CATHARE, and RELAP5 codes. Safety significance of the PMK-2 projects is also discussed

  19. Horizon Expansion of Thermal-Hydraulic Activities into HTGR Safety Analysis Including Gas-Turbine Cycle and Hydrogen Plant

    International Nuclear Information System (INIS)

    No, Hee Cheon; Yoon, Ho Joon; Kim, Seung Jun; Lee, Byeng Jin; Kim, Ji Hwan; Kim, Hyeun Min; Lim, Hong Sik

    2009-01-01

    We present three nuclear/hydrogen-related R and D activities being performed at KAIST: air-ingressed LOCA analysis code development, gas turbine analysis tool development, and hydrogen-production system analysis model development. The ICE numerical technique widely used for the safety analysis of water-reactors is successfully implemented into GAMMA, with which we solve the basic equations for continuity, momentum conservation, energy conservation of the gas mixture, and mass conservation of 6 species (He, N2, O2, CO, CO2, and H2O). GAMMA has been extensively validated using data from 14 test facilities. We developed a tool to predict the characteristics of HTGR helium turbines based on the through flow calculation with a Newton- Raphson method that overcomes the weakness of the conventional method based on the successive iteration scheme. It is found that the current method reaches stable and quick convergence even under the off-normal condition with the same degree of accuracy. The dynamic equations for the distillation column of HI process are described with 4 material components involved in the HI process: H2O, HI, I2, H2. For the HI process we improved the Neumann model based on the NRTL (Non-Random Two-Liquid) model. The improved Neumann model predicted a total pressure with 8.6% maximum relative deviation from the data and 2.5% mean relative deviation, and liquid-liquid-separation with 9.52% maximum relative deviation from the data

  20. A study to investigate viscous coupling effects on the hydraulic conductance of fluid layers in two-phase flow at the pore level.

    Science.gov (United States)

    Shams, Mosayeb; Raeini, Ali Q; Blunt, Martin J; Bijeljic, Branko

    2018-07-15

    This paper examines the role of momentum transfer across fluid-fluid interfaces in two-phase flow. A volume-of-fluid finite-volume numerical method is used to solve the Navier-Stokes equations for two-phase flow at the micro-scale. The model is applied to investigate viscous coupling effects as a function of the viscosity ratio, the wetting phase saturation and the wettability, for different fluid configurations in simple pore geometries. It is shown that viscous coupling effects can be significant for certain pore geometries such as oil layers sandwiched between water in the corner of mixed wettability capillaries. A simple parametric model is then presented to estimate general mobility terms as a function of geometric properties and viscosity ratio. Finally, the model is validated by comparison with the mobilities computed using direct numerical simulation. Copyright © 2018 The Authors. Published by Elsevier Inc. All rights reserved.

  1. Modeling and stability of electro-hydraulic servo of hydraulic excavator

    Science.gov (United States)

    Jia, Wenhua; Yin, Chenbo; Li, Guo; Sun, Menghui

    2017-11-01

    The condition of the hydraulic excavator is complicated and the working environment is bad. The safety and stability of the control system is influenced by the external factors. This paper selects hydraulic excavator electro-hydraulic servo system as the research object. A mathematical model and simulation model using AMESIM of servo system is established. Then the pressure and flow characteristics are analyzed. The design and optimization of electro-hydraulic servo system and its application in engineering machinery is provided.

  2. On the Versatility of Rheoreversible, Stimuli-responsive Hydraulic-Fracturing Fluids for Enhanced Geothermal Systems: Effect of Reservoir pH

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez, Carlos A.; Shao, Hongbo; Bonneville, Alain; Varga, Tamas; Zhong, Lirong

    2016-04-25

    Abstract The primary challenge for the feasibility of enhanced geothermal systems (EGS) is to cost-effectively create high-permeability reservoirs inside deep crystalline bedrock. Although fracturing fluids are commonly used for oil/gas, standard fracturing methods are not developed or proven for EGS temperatures and pressures. Furthermore, the environmental impacts of currently used fracturing methods are only recently being determined. These authors recently reported an environmentally benign, CO2-activated, rheoreversible fracturing fluid that enhances permeability through fracturing due to in situ volume expansion and gel formation. The potential of this novel fracturing fluid is evaluated in this work towards its application at geothermal sites under different pH conditions. Laboratory-scale fracturing experiments using Coso Geothermal rock cores under different pH environments were performed followed by X-ray microtomography characterization. The results demonstrate that CO2-reactive aqueous solutions of environmentally amenable polyallylamine (PAA) consistently and reproducibly creates/propagates fracture networks through highly impermeable crystalline rock from Coso EGS sites at considerably lower effective stress as compared to conventional fracturing fluids. In addition, permeability was significantly enhanced in a wide range of formation-water pH values. This effective, and environmentally-friendly fracturing fluid technology represents a potential alternative to conventional fracturing fluids.

  3. Design of Pumps for Water Hydraulic Systems

    DEFF Research Database (Denmark)

    Klit, Peder; Olsen, Stefan; Bech, Thomas Nørgaard

    1999-01-01

    This paper considers the development of two pumps for water hydraulic applications. The pumps are based on two different working principles: The Vane-type pump and the Gear-type pump. Emphasis is put on the considerations that should be made to account for water as the hydraulic fluid.......KEYWORDS: water, pump, design, vane, gear....

  4. Development of thermal-hydraulic safety codes for HTGRs with gas-turbine and hydrogen process cycles

    International Nuclear Information System (INIS)

    No, Hee Cheon; Yoon, Ho Joon; Lee, Byung Jin; Kim, Yong Soo; Jin, Hyeng Gon; Kim, Ji Hwan; Kim, Hyeun Min; Lim, Hong Sik

    2008-01-01

    We present three nuclear/hydrogen-related R and D activities being performed at KAIST: air-ingressed LOCA analysis code development, gas turbine analysis tool development, and hydrogen-production system analysis model development. The ICE numerical technique widely used for the safety analysis of water-reactors is successfully implemented into GAMMA in which we solve the basic equations for continuity, momentum conservation, energy conservation of the gas mixture, and mass conservation of 6 species (He, N2, O2, CO, CO2, and H2O). GAMMA has been extensively validated using data from 14 test facilities. We developed SANA code to predict the characteristics of HTGR helium turbines based on the throughflow calculation with a Newton-Raphson method that overcomes the weakness of the conventional method based on the successive iteration scheme. It is found out that the current method reaches stable and quick convergence even under the off-normal condition with the same degree of accuracy. The GAMMA-SANA coupled code was assessed by comparing its results with the steady-state of the GTHTR300, and the load reduction transient was simulated for the 100% to 70% power operation. The calculation results confirm that two-dimensional throughflow modeling can be successfully used to describe the gas turbine behavior. The dynamic equations for the distillation column of the HI process in the I-S cycle are described with 4 material components involved in the HI process: H2O, HI, I2, and H2. For the VLE prediction in the HI process we improved the Neumann model based on the NRTL (Non-Random Two-Liquid) model. Relative to the experimental data, the improved Neumann model shows deviations of 8.6% in maximum and 2.5% in average for the total pressure, and 9.5% in maximum for the liquid-liquid separation composition. Through a parametric analysis using the published experimental data related to the Bunsen reaction and liquid-liquid separation, an optimized operating condition for the

  5. Tree Hydraulics: How Sap Rises

    Science.gov (United States)

    Denny, Mark

    2012-01-01

    Trees transport water from roots to crown--a height that can exceed 100 m. The physics of tree hydraulics can be conveyed with simple fluid dynamics based upon the Hagen-Poiseuille equation and Murray's law. Here the conduit structure is modelled as conical pipes and as branching pipes. The force required to lift sap is generated mostly by…

  6. Analysis of INDOT current hydraulic policies : [technical summary].

    Science.gov (United States)

    2011-01-01

    Hydraulic design often tends to be on a conservative side for safety reasons. Hydraulic structures are typically oversized with the goal being reduced future maintenance costs, and to reduce the risk of property owner complaints. This approach leads ...

  7. Multiphase Flow Dynamics 5 Nuclear Thermal Hydraulics

    CERN Document Server

    Kolev, Nikolay Ivanov

    2012-01-01

    The present Volume 5 of the successful book package "Multiphase Flow Dynamics" is devoted to nuclear thermal hydraulics which is a substantial part of nuclear reactor safety. It provides knowledge and mathematical tools for adequate description of the process of transferring the fission heat released in materials due to nuclear reactions into its environment. It step by step introduces into the heat release inside the fuel, temperature fields in the fuels, the "simple" boiling flow in a pipe described using ideas of different complexity like equilibrium, non equilibrium, homogeneity, non homogeneity. Then the "simple" three-fluid boiling flow in a pipe is described by gradually involving the mechanisms like entrainment and deposition, dynamic fragmentation, collisions, coalescence, turbulence. All heat transfer mechanisms are introduced gradually discussing their uncertainty. Different techniques are introduced like boundary layer treatments or integral methods. Comparisons with experimental data at each step...

  8. Multiphase flow dynamics 5 nuclear thermal hydraulics

    CERN Document Server

    Kolev, Nikolay Ivanov

    2015-01-01

    This Volume 5 of the successful book package "Multiphase Flow Dynamics" is devoted to nuclear thermal hydraulics which is a substantial part of nuclear reactor safety. It provides knowledge and mathematical tools for adequate description of the process of transferring the fission heat released in materials due to nuclear reactions into its environment. It step by step introduces into the heat release inside the fuel, temperature fields in the fuels, the "simple" boiling flow in a pipe described using ideas of different complexity like equilibrium, non equilibrium, homogeneity, non homogeneity. Then the "simple" three-fluid boiling flow in a pipe is described by gradually involving the mechanisms like entrainment and deposition, dynamic fragmentation, collisions, coalescence, turbulence. All heat transfer mechanisms are introduced gradually discussing their uncertainty. Different techniques are introduced like boundary layer treatments or integral methods. Comparisons with experimental data at each step demons...

  9. Cerebrospinal fluid protein and glucose examinations and tuberculosis:
Will laboratory safety regulations force a change of practice?

    Science.gov (United States)

    Tormey, William P; O'Hagan, Christopher

    2015-01-01

    Cerebrospinal fluid (CSF) protein and glucose examinations are usually performed in chemical pathology departments on autoanalysers. Tuberculosis (TB) is a group 3 biological agent under Directive 2000/54/EC of the European Parliament but in the biochemistry laboratory, no extra precautions are taken in its analysis in possible TB cases. The issue of laboratory practice and safety in the biochemical analyses of CSF specimens, when tuberculosis infection is in question is addressed in the context of ambiguity in the implementation of current national and international health and safety regulations. Additional protective measures for laboratory staff during the analysis of CSF TB samples should force a change in current laboratory practice and become a regulatory issue under ISO 15189. Annual Mantoux skin test or an interferon-γ release assay for TB should be mandatory for relevant staff. This manuscript addresses the issue of biochemistry laboratory practice and safety in the biochemical analyses of CSF specimens when tuberculosis infection is in question in the context of the ambiguity of statutory health and safety regulations.

  10. 30 CFR 250.459 - What are the safety requirements for drilling fluid-handling areas?

    Science.gov (United States)

    2010-07-01

    ... addition: (1) If natural means provide adequate ventilation, then a mechanical ventilation system is not... areas where adequate ventilation is provided by natural means. You must test and recalibrate gas... install and maintain a ventilation system and gas monitors. Drilling fluid-handling areas must have the...

  11. 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)

  12. Reactor Thermal Hydraulic Numerical Calculation And Modeling

    International Nuclear Information System (INIS)

    Duong Ngoc Hai; Dang The Ba

    2008-01-01

    In the paper the results of analysis of thermal hydraulic state models using the numerical codes such as COOLOD, EUREKA and RELAP5 for simulation of the reactor thermal hydraulic states are presented. The calculations, analyses of reactor thermal hydraulic state and safety were implemented using different codes. The received numerical results, which were compared each to other, to experiment measurement of Dalat (Vietnam) research reactor and published results, show their appropriateness and capacity for analyses of different appropriate cases. (author)

  13. A fluid-solid finite element method for the analysis of reactor safety problems

    International Nuclear Information System (INIS)

    Mitra, Santanu; Kumar, Ashutosh; Sinhamahapatra, K.P.

    2006-01-01

    The work presented herein can broadly be categorized as a fluid-structure interaction problem. The response of a circular cylindrical structure subjected to cross flow is examined using the finite element method for both the liquid and the structure domains. The cylindrical tube is mounted elastically at the ends and is free to move under the action of the unsteady flow-induced forces. The fluid is considered to be acoustic compressible and viscous. A Galerkin finite element method implemented on a triangular mesh is used to solve the time-dependent Navier-Stokes equations. The cylinder motion is modeled using a five-degrees of freedom generalized shell element structural dynamics model. The numerical simulations of the response of the calandria tubes/pressure tubes, adjustor rod and shut-off rod of a nuclear reactor are presented. A few typical results are presented to assess the accuracy and applicability of the developed modules

  14. Failure analysis of fire resistant fluid (FRF piping used in hydraulic control system at oil-fired thermal power generation plant

    Directory of Open Access Journals (Sweden)

    Muhammad Akram

    2017-04-01

    Full Text Available This is a case study regarding frequent forced outages in an oil-fired power generating station due to failure of fire resistant fluid (FRF piping of material ASTM A-304. This analysis was done to find out the most probable cause of failure and to rectify the problem. Methods for finding and analyzing the cracks include nondestructive testing techniques such as visual testing (VT and dye penetrant testing (PT along with that periodic monitoring after rectification of problem. The study revealed that pitting and pit to crack transitions were formed in stainless steel piping containing high pressure (system pressure 115 bars fire resistant fluid. However, after replacement of piping the pitting and cracking reoccurred. It was observed that due to possible exposure to chlorinated moisture in surrounding environment pitting was formed which then transformed into cracks. The research work discussed in this paper illustrates the procedure used in detection of the problem and measures taken to solve the problem.

  15. Hydraulic Actuators with Autonomous Hydraulic Supply for the Mainline Aircrafts

    Directory of Open Access Journals (Sweden)

    I. S. Shumilov

    2014-01-01

    Full Text Available Applied in the aircraft control systems, hydraulic servo actuators with autonomous hydraulic supply, so-called, hydraulic actuators of integrated configuration, i.e. combination of a source of hydraulic power and its load in the single unit, are aimed at increasing control system reliability both owing to elimination of the pipelines connecting the actuator to the hydraulic supply source, and owing to avoidance of influence of other loads failure on the actuator operability. Their purpose is also to raise control system survivability by eliminating the long pipeline communications and their replacing for the electro-conductive power supply system, thus reducing the vulnerability of systems. The main reason for a delayed application of the hydraulic actuators in the cutting-edge aircrafts was that such aircrafts require hydraulic actuators of considerably higher power with considerable heat releases, which caused an unacceptable overheat of the hydraulic actuators. Positive and negative sides of the hydraulic actuators, their alternative options of increased reliability and survivability, local hydraulic systems as an advanced alternative to independent hydraulic actuators are considered.Now to use hydraulic actuators in mainline aircrafts is inexpedient since there are the unfairly large number of the problems reducing, first and last, safety of flights, with no essential weight and operational advantages. Still works to create competitive hydraulic actuators ought to be continued.Application of local hydraulic systems (LHS will allow us to reduce length of pressure head and drain pipelines and mass of pipelines, as well as to raise their general fail-safety and survivability. Application of the LHS principle will allow us to use a majority of steering drive advantages. It is necessary to allocate especially the following:- ease of meeting requirements for the non-local spread of the engine weight;- essentially reducing length and weight of

  16. National Laboratory of Hydraulics. 1996 progress report

    International Nuclear Information System (INIS)

    1996-01-01

    This progress report of the National Laboratory of Hydraulics (LNH) of Electricite de France (EdF) summarizes, first, the research and development studies carried out in 1996 for the development of research tools for industrial fluid mechanics and environmental hydraulics and for the development of computer tools (computer codes and softwares for fluid mechanics modeling, modeling of reactive, compressible, two-phase and turbulent flows and of complex chemical kinetics using finite elements and finite volume methods). A second parts describes the research studies performed for other services of EdF, concerning: the functioning of nuclear reactors (thermohydraulic studies of the reactor vessel and of the primary coolant circuit, gas flows following severe accidents, fluid-structure thermal coupling etc...), fossil fuel power plants, the equipment and operation of thermal power plants and hydraulic power plants, the use of electric power. A third part summarizes the river and marine hydraulic studies carried out for other companies. (J.S.)

  17. Safety evaluation of the loss of fluid test facility project No. 394

    International Nuclear Information System (INIS)

    1975-05-01

    Assessment of the safety of the LOFT facility and subsequent recommendations have been based on a comparison of the LOFT facility to requirements for commercial power reactors. In this comparison, the many unique features of the LOFT facility were considered including the low power level, the limited operational use as a test reactor, and the remoteness of the site. Based on this assessment, it is concluded, that while the likelihood of an accidental release of fission products may be greater than for a commercial power reactor, the consequences of such a release are reduced by the lower fission product inventory, the remoteness of the site and the capability of evacuating the Idaho National Engineering Laboratory (INEL) and adjacent areas. There is reasonable assurance that the public health and safety will not be endangered due to operation of this facility, specifically: The INEL site is acceptable with respect to location, land use, population distribution, controlled access, hydrology, meteorology, geology and seismology. Sufficient engineered safety features have been included to assure that the potential offsite doses are well within 10 CFR Part 100 guidelines. The LOFT facility has been designed in general accordance with standards, guides and codes which are comparable to those applied to commercial power reactors and any exceptions to these have been based on the unique features of the LOFT facility. Certain matters including the final safety analyses based on detailed component designs, Technical Specifications, LOCE controls and detailed program plan have not been reviewed but we assume will properly be resolved by ERDA, which has the ultimate responsibility for the safety of this facility. Changes to the facility design or program plan such as removal of the fueled Mobile Test Assembly or blowdowns to the containment vessel also will require additional analyses and review. (U.S.)

  18. Calculation of fluid-structure interaction for reactor safety with the Cassiopee code

    International Nuclear Information System (INIS)

    Graveleau, J.L.; Louvet, P.D.

    1979-01-01

    The cassiopee code is an eulerian-lagrangian coupled code for computations where the hydrodynamic is coupled with structural domains. It is completely explicit. The fluid zones may be computed either in lagrangian or in eulerian coordinates; thin shells can be computed wih their flexural behaviour; elastic plastic zones must be calculated in a lagrangian way. This code is under development in Cadarache. Its purpose is to compute the hypothetical core disruptive accident of a LMFBR when lagrangian codes are not sufficient. This paper contains a description of the code and two examples of computations, one of which has been compared with experimental results

  19. Hydraulics national laboratory; Laboratoire national d`hydraulique

    Energy Technology Data Exchange (ETDEWEB)

    Chabard, J P

    1996-12-31

    The hydraulics national laboratory is a department of the service of applications of electric power and environment from the direction of studies and researches of Electricite de France. It has to solve the EDF problems concerning the fluids mechanics and hydraulics. Problems in PWR type reactors, fossil fuel power plants, circulating fluidized bed power plants, hydroelectric power plants relative to fluid mechanics and hydraulics studied and solved in 1995 are explained in this report. (N.C.)

  20. Hydraulics national laboratory; Laboratoire national d`hydraulique

    Energy Technology Data Exchange (ETDEWEB)

    Chabard, J.P.

    1995-12-31

    The hydraulics national laboratory is a department of the service of applications of electric power and environment from the direction of studies and researches of Electricite de France. It has to solve the EDF problems concerning the fluids mechanics and hydraulics. Problems in PWR type reactors, fossil fuel power plants, circulating fluidized bed power plants, hydroelectric power plants relative to fluid mechanics and hydraulics studied and solved in 1995 are explained in this report. (N.C.)

  1. Effect of Poroelasticity on Hydraulic Fracture Interactions

    DEFF Research Database (Denmark)

    Usui, Tomoya; Salimzadeh, Saeed; Paluszny, Adriana

    2017-01-01

    This study investigates, by performing finite element-based simulations, the influence of fluid leak-off and poroelasticity on growth of multiple hydraulic fractures that initiate from a single horizontal well. In this research, poroelastic deformation of the matrix is coupled with fluid flow in ...

  2. Spiral groove seal. [for hydraulic rotating shaft

    Science.gov (United States)

    Ludwig, L. P. (Inventor)

    1973-01-01

    Mating flat surfaces inhibit leakage of a fluid around a stationary shaft. A spiral groove pattern produces a pumping action toward the fluid when the shaft rotates which prevents leakage while a generated hydraulic lifting force separates the mating surfaces to minimize wear.

  3. Hydraulic manipulator

    International Nuclear Information System (INIS)

    Sinha, A.K.; Srikrishnamurty, G.

    1990-01-01

    Successful operation of nuclear plant is largely dependent on safe handling of radio-active material. In order to reduce this handling problem and minimise the exposure of radiation, various handling equipment and manipulators have been developed according to the requirements. Manufacture of nuclear fuel, which is the most important part of the nuclear industry, involves handling of uranium ingots weighing approximately 250 kg. This paper describes a specially designed hydraulic manipulator for handling of the ingots in a limited space. It was designed to grab and handle the ingots in any position. This has following drive motions: (1)gripping and releasing, (2)lifting and lowering (z-motion), (3)rotation about the horizontal axis (azimuth drive), (4)rotation about the job axis, and (5)rotation about the vertical axis. For horizontal motion (X and Y axis motion) this equipment is mounted on a motorised trolley, so that it can move inside the workshop. For all drives except the rotation about the job axis, hydraulic cylinders have been used with a battery operated power pack. Trolley drive is also given power from same battery. This paper describes the design aspects of this manipulator. (author). 4 figs

  4. Development of design technology on thermal-hydraulic performance in tight-lattice rod bundle. 4. Large paralleled simulation by the advanced two-fluid model code

    International Nuclear Information System (INIS)

    Misawa, Takeharu; Yoshida, Hiroyuki; Akimoto, Hajime

    2008-01-01

    In Japan Atomic Energy Agency (JAEA), the Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been developed. For thermal design of FLWR, it is necessary to develop analytical method to predict boiling transition of FLWR. Japan Atomic Energy Agency (JAEA) has been developing three-dimensional two-fluid model analysis code ACE-3D, which adopts boundary fitted coordinate system to simulate complex shape channel flow. In this paper, as a part of development of ACE-3D to apply to rod bundle analysis, introduction of parallelization to ACE-3D and assessments of ACE-3D are shown. In analysis of large-scale domain such as a rod bundle, even two-fluid model requires large number of computational cost, which exceeds upper limit of memory amount of 1 CPU. Therefore, parallelization was introduced to ACE-3D to divide data amount for analysis of large-scale domain among large number of CPUs, and it is confirmed that analysis of large-scale domain such as a rod bundle can be performed by parallel computation with keeping parallel computation performance even using large number of CPUs. ACE-3D adopts two-phase flow models, some of which are dependent upon channel geometry. Therefore, analyses in the domains, which simulate individual subchannel and 37 rod bundle, are performed, and compared with experiments. It is confirmed that the results obtained by both analyses using ACE-3D show agreement with past experimental result qualitatively. (author)

  5. Percutaneous catheter drainage of thoracic fluid: the usefulness and safety of bedside trocar placement under ultrasound guidance

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Heon [Seoul Medical Center, Seoul (Korea, Republic of)

    2006-07-15

    The author wanted to evaluate the usefulness and safety of the trocar technique for US-guided bedside catheter placement into thoracic fluid collections, and this technique has generally been reserved for the larger or superficial fluid collections. 42 drainage procedures were performed in 38 patients at the bedside. The patients were positioned supine or semi-upright. A drainage catheter system with a stylet and cannula assembly was used and all of the catheters were inserted using the trocar technique. The procedures consisted of drainage of empyema (n=14), malignant effusion (n=13), lung abscess (n=3), massive transudate (n=8), hemothorax (n=2) and chest wall hematoma (n=2). The clinical results were classified as successful (complete and partially successful), failure or undetermined. The medical records and images were retrospectively reviewed to evaluate the success rate, the complications and the procedure time. Technical success was achieved in all of the 42 procedures. With using the trocar technique, all the catheters were placed into even the small collections without significant complications. Drainage was successful in 36 (85.7%) of the 42 procedures. The average volume of thoracic fluid that was aspirated manually at the time of catheter placement was 420 mL (range: 35 to 1470 mL). The procedure time was less than 10 minutes from US-localization to complete catheter placement in all of the procedures. The trocar technique under US guidance can be an efficient and safe alternative to the Seldinger or guide-wire exchange technique for bedside catheter placement in the critically ill or hemodynamically unstable patients.

  6. Computational Fluid Dynamics (CFD) in Nuclear Reactor Safety (NRS) - Proceedings of the workshop on Experiments and CFD Code Application to Nuclear Reactor Safety (XCFD4NRS)

    International Nuclear Information System (INIS)

    2008-01-01

    Computational Fluid Dynamics (CFD) is to an increasing extent being adopted in nuclear reactor safety analyses as a tool that enables specific safety relevant phenomena occurring in the reactor coolant system to be better described. The Committee on the Safety of Nuclear Installations (CSNI), which is responsible for the activities of the OECD Nuclear Energy Agency that support advancing the technical base of the safety of nuclear installations, has in recent years conducted an important activity in the CFD area. This activity has been carried out within the scope of the CSNI working group on the analysis and management of accidents (GAMA), and has mainly focused on the formulation of user guidelines and on the assessment and verification of CFD codes. It is in this GAMA framework that a first workshop CFD4NRS was organized and held in Garching, Germany in 2006. Following the CFD4NRS workshop, this XCFD4NRS Workshop was intended to extend the forum created for numerical analysts and experimentalists to exchange information in the field of Nuclear Reactor Safety (NRS) related activities relevant to Computational Fluid Dynamics (CFD) validation, but this time with more emphasis placed on new experimental techniques and two-phase CFD applications. The purpose of the workshop was to provide a forum for numerical analysts and experimentalists to exchange information in the field of NRS-related activities relevant to CFD validation, with the objective of providing input to GAMA CFD experts to create a practical, state-of-the-art, web-based assessment matrix on the use of CFD for NRS applications. The scope of XCFD4NRS includes: - Single-phase and two-phase CFD simulations with an emphasis on validation in areas such as: boiling flows, free-surface flows, direct contact condensation and turbulent mixing. These applications should relate to NRS-relevant issues such as: pressurized thermal shocks, critical heat flux, pool heat exchangers, boron dilution, hydrogen

  7. Technical-evaluation report on the proposed technical-specification changes for the inservice surveillance of safety-related hydraulic and mechanical snubbers at the Millstone Nuclear Power Station, Unit 2 (Docket No. 50-336)

    International Nuclear Information System (INIS)

    Selan, J.C.

    1983-01-01

    This report documents the technical evaluation of the proposed Technical Specification changes to Limiting Conditions for Operation, Surveillance Requirements and Bases for safety-related hydraulic and mechanical snubbers at the Millstone Nuclear Power Station, Unit 2. The evaluation is to determine whether the proposed Technical Specifications are in conformance with the model Standard Technical Specification set forth by the NRC. A check list, Appendix A of this report, compares the licensee's submittal with the NRC requirements and includes Proposed Resolution of the Deviations

  8. Technical-evaluation report on the proposed technical-specification changes for the inservice surveillance of safety-related hydraulic and mechanical snubbers at the Indian Point Nuclear Power Plant, Unit 3 (Docket No. 50-286)

    International Nuclear Information System (INIS)

    Selan, J.C.

    1983-01-01

    This report documents the technical evaluation of the proposed Technical Specification changes to Limiting Conditions for Operation, Surveillance Requirements and Bases for safety-related hydraulic and mechanical snubbers at the Indian Point Nuclear Power Plant, Unit 3. The evaluation is to determine whether the proposed Technical Specifications are in conformance with the model Standard Technical Specification set forth by the NRC. A check list, Appendix A of this report, compares the licensee's submittal with the NRC requirements and includes Proposed Resolution of the Deviations

  9. Technical-evaluation report on the proposed technical-specification changes for the inservice surveillance of safety-related hydraulic and mechanical snubbers at the James A. Fitzpatrick Nuclear Power Plant (Docket No. 50-333)

    International Nuclear Information System (INIS)

    Selan, J.C.

    1983-01-01

    This report documents the technical evaluation of the proposed Technical Specification changes to Limiting Conditions for Operation, Surveillance Requirements and Bases for safety-related hydraulic and mechanical snubbers at the James A. FitzPatrick Nuclear Power Plant. The evaluation is to determine whether the proposed Technical Specifications are in conformance with the model Standard Technical Specification set forth by the NRC. A check list, Appendix A of this report, compares the licensee's submittal with the NRC requirements and includes Proposed Resolution of the Deviations

  10. Improved Flow Modeling in Transient Reactor Safety Analysis Computer Codes

    International Nuclear Information System (INIS)

    Holowach, M.J.; Hochreiter, L.E.; Cheung, F.B.

    2002-01-01

    A method of accounting for fluid-to-fluid shear in between calculational cells over a wide range of flow conditions envisioned in reactor safety studies has been developed such that it may be easily implemented into a computer code such as COBRA-TF for more detailed subchannel analysis. At a given nodal height in the calculational model, equivalent hydraulic diameters are determined for each specific calculational cell using either laminar or turbulent velocity profiles. The velocity profile may be determined from a separate CFD (Computational Fluid Dynamics) analysis, experimental data, or existing semi-empirical relationships. The equivalent hydraulic diameter is then applied to the wall drag force calculation so as to determine the appropriate equivalent fluid-to-fluid shear caused by the wall for each cell based on the input velocity profile. This means of assigning the shear to a specific cell is independent of the actual wetted perimeter and flow area for the calculational cell. The use of this equivalent hydraulic diameter for each cell within a calculational subchannel results in a representative velocity profile which can further increase the accuracy and detail of heat transfer and fluid flow modeling within the subchannel when utilizing a thermal hydraulics systems analysis computer code such as COBRA-TF. Utilizing COBRA-TF with the flow modeling enhancement results in increased accuracy for a coarse-mesh model without the significantly greater computational and time requirements of a full-scale 3D (three-dimensional) transient CFD calculation. (authors)

  11. A strategy for the qualification of multi-fluid approaches for nuclear reactor safety

    Energy Technology Data Exchange (ETDEWEB)

    Lucas, D., E-mail: D.Lucas@hzdr.de; Rzehak, R.; Krepper, E.; Ziegenhein, Th.; Liao, Y.; Kriebitzsch, S.; Apanasevich, P.

    2016-04-01

    CFD-simulations for two-phase flows applying the multi-fluid approach are not yet qualified to provide reliable predictions for unknown flows. Among others, one important reason is the missing agreement within the community on closure models to be used. Considering specific phenomena or not, using different models and adjustable constants, most papers presenting a model validation end up with a good agreement with experimental data. However a case by case selection of models and constants does not help to improve the predictive capabilities of such models. For this reason the definition of baseline models considering all known phenomena that could be important is proposed. In such baseline models all parameter have to be defined, i.e., there are no tuning parameters by definition. Therefore these baseline models have to be applied to many experiments with different complexity. Shortcomings of the models and our physical understanding of the complex flow phenomena have to be identified by detailed analyses on the deviations between experimental data and simulation results. A modification of the baseline model will only be done if it bases on physical considerations and improves the overall performance of the model. This requires a huge effort, but seems to be the only way to improve the situation. In particular more complete experimental data are required. Joint activities on the development of such baseline models are desirable. The paper illustrates this strategy by a baseline model for polydisperse bubbly flows which is presently developed at HZDR.

  12. Hydraulics calculation in drilling simulator

    Science.gov (United States)

    Malyugin, Aleksey A.; Kazunin, Dmitry V.

    2018-05-01

    The modeling of drilling hydraulics in the simulator system is discussed. This model is based on the previously developed quasi-steady model of an incompressible fluid flow. The model simulates the operation of all parts of the hydraulic drilling system. Based on the principles of creating a common hydraulic model, a set of new elements for well hydraulics was developed. It includes elements that correspond to the in-drillstring and annular space. There are elements controlling the inflow from the reservoir into the well and simulating the lift of gas along the annulus. New elements of the hydrosystem take into account the changing geometry of the well, loss in the bit, characteristics of the fluids including viscoplasticity. There is an opportunity specify the complications, the main one of which is gas, oil and water inflow. Correct work of models in cases of complications makes it possible to work out various methods for their elimination. The coefficients of the model are adjusted on the basis of incomplete experimental data provided by operators of drilling platforms. At the end of the article the results of modeling the elimination of gas inflow by a continuous method are presented. The values displayed in the simulator (drill pipe pressure, annulus pressure, input and output flow rates) are in good agreement with the experimental data. This exercise took one hour, which is less than the time on a real rig with the same configuration of equipment and well.

  13. Model for polygonal hydraulic jumps

    DEFF Research Database (Denmark)

    Martens, Erik Andreas; Watanabe, Shinya; Bohr, Tomas

    2012-01-01

    We propose a phenomenological model for the polygonal hydraulic jumps discovered by Ellegaard and co-workers [Nature (London) 392, 767 (1998); Nonlinearity 12, 1 (1999); Physica B 228, 1 (1996)], based on the known flow structure for the type-II hydraulic jumps with a "roller" (separation eddy...... nonhydrostatic pressure contributions from surface tension in light of recent observations by Bush and co-workers [J. Fluid Mech. 558, 33 (2006); Phys. Fluids 16, S4 (2004)]. The model can be analyzed by linearization around the circular state, resulting in a parameter relationship for nearly circular polygonal...... states. A truncated but fully nonlinear version of the model can be solved analytically. This simpler model gives rise to polygonal shapes that are very similar to those observed in experiments, even though surface tension is neglected, and the condition for the existence of a polygon with N corners...

  14. Hydraulic Characterization Activities in Support of the Shaft-Seals Fluid-Flow Modeling Integration into the WIPP EPA Compliance Certification Application

    International Nuclear Information System (INIS)

    Knowles, M.K.; Hurtado, L.D.; Dale, Tim

    1997-12-01

    The Waste Isolation Pilot Plant (WIPP) is a planned geologic repository for permanent disposal of transuranic waste generated by the U.S. Department of Energy. Disposal regions consist of panels and drifts mined from the bedded salt of the Salado Formation at a depth of approximately 650 m below the surface. This lithology is part of the 225 million year old Delaware Basin, and is geographically located in southeastern New Mexico. Four shafts service the facility needs for air intake, exhaust, waste handling, and salt handling. As the science advisor for the project, Sandia National Laboratories developed the WIPP shaft sealing system design. This design is a fundamental component of the application process for facility licensing, and has been found acceptable by stakeholders and regulatory agencies. The seal system design is founded on results obtained from laboratory and field experiments, numerical modeling, and engineering judgment. This paper describes a field test program to characterize the fluid flow properties in the WIPP shafts at representative seal locations. This work was conducted by Duke Engineering and Services under contract to Sandia National Laboratories in support of the seal system design

  15. Proceedings of the Twenty-First Water Reactor Safety Information Meeting: Volume 1, Plenary session; Advanced reactor research; advanced control system technology; advanced instrumentation and control hardware; human factors research; probabilistic risk assessment topics; thermal hydraulics; thermal hydraulic research for advanced passive LWRs

    International Nuclear Information System (INIS)

    Monteleone, S.

    1994-04-01

    This three-volume report contains 90 papers out of the 102 that were presented at the Twenty-First Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 25--27, 1993. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Germany, Japan, Russia, Switzerland, Taiwan, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. Individual papers have been cataloged separately. This document, Volume 1 covers the following topics: Advanced Reactor Research; Advanced Instrumentation and Control Hardware; Advanced Control System Technology; Human Factors Research; Probabilistic Risk Assessment Topics; Thermal Hydraulics; and Thermal Hydraulic Research for Advanced Passive Light Water Reactors

  16. Proceedings of the Twenty-First Water Reactor Safety Information Meeting: Volume 1, Plenary session; Advanced reactor research; advanced control system technology; advanced instrumentation and control hardware; human factors research; probabilistic risk assessment topics; thermal hydraulics; thermal hydraulic research for advanced passive LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [Brookhaven National Lab., Upton, NY (United States)] [comp.

    1994-04-01

    This three-volume report contains 90 papers out of the 102 that were presented at the Twenty-First Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 25--27, 1993. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Germany, Japan, Russia, Switzerland, Taiwan, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. Individual papers have been cataloged separately. This document, Volume 1 covers the following topics: Advanced Reactor Research; Advanced Instrumentation and Control Hardware; Advanced Control System Technology; Human Factors Research; Probabilistic Risk Assessment Topics; Thermal Hydraulics; and Thermal Hydraulic Research for Advanced Passive Light Water Reactors.

  17. Scaling Analysis of the Single-Phase Natural Circulation: the Hydraulic Similarity

    International Nuclear Information System (INIS)

    Yu, Xin-Guo; Choi, Ki-Yong

    2015-01-01

    These passive safety systems all rely on the natural circulation to cool down the reactor cores during an accident. Thus, a robust and accurate scaling methodology must be developed and employed to both assist in the design of a scaled-down test facility and guide the tests in order to mimic the natural circulation flow of its prototype. The natural circulation system generally consists of a heat source, the connecting pipes and several heat sinks. Although many applauding scaling methodologies have been proposed during last several decades, few works have been dedicated to systematically analyze and exactly preserve the hydraulic similarity. In the present study, the hydraulic similarity analyses are performed at both system and local level. By this mean, the scaling criteria for the exact hydraulic similarity in a full-pressure model have been sought. In other words, not only the system-level but also the local-level hydraulic similarities are pursued. As the hydraulic characteristics of a fluid system is governed by the momentum equation, the scaling analysis starts with it. A dimensionless integral loop momentum equation is derived to obtain the dimensionless numbers. In the dimensionless momentum equation, two dimensionless numbers, the dimensionless flow resistance number and the dimensionless gravitational force number, are identified along with a unique hydraulic time scale, characterizing the system hydraulic response. A full-height full-pressure model is also made to see which model among the full-height model and reduced-height model can preserve the hydraulic behavior of the prototype. From the dimensionless integral momentum equation, a unique hydraulic time scale, which characterizes the hydraulic response of a single-phase natural circulation system, is identified along with two dimensionless parameters: the dimensionless flow resistance number and the dimensionless gravitational force number. By satisfying the equality of both dimensionless numbers

  18. Scaling Analysis of the Single-Phase Natural Circulation: the Hydraulic Similarity

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Xin-Guo; Choi, Ki-Yong [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    These passive safety systems all rely on the natural circulation to cool down the reactor cores during an accident. Thus, a robust and accurate scaling methodology must be developed and employed to both assist in the design of a scaled-down test facility and guide the tests in order to mimic the natural circulation flow of its prototype. The natural circulation system generally consists of a heat source, the connecting pipes and several heat sinks. Although many applauding scaling methodologies have been proposed during last several decades, few works have been dedicated to systematically analyze and exactly preserve the hydraulic similarity. In the present study, the hydraulic similarity analyses are performed at both system and local level. By this mean, the scaling criteria for the exact hydraulic similarity in a full-pressure model have been sought. In other words, not only the system-level but also the local-level hydraulic similarities are pursued. As the hydraulic characteristics of a fluid system is governed by the momentum equation, the scaling analysis starts with it. A dimensionless integral loop momentum equation is derived to obtain the dimensionless numbers. In the dimensionless momentum equation, two dimensionless numbers, the dimensionless flow resistance number and the dimensionless gravitational force number, are identified along with a unique hydraulic time scale, characterizing the system hydraulic response. A full-height full-pressure model is also made to see which model among the full-height model and reduced-height model can preserve the hydraulic behavior of the prototype. From the dimensionless integral momentum equation, a unique hydraulic time scale, which characterizes the hydraulic response of a single-phase natural circulation system, is identified along with two dimensionless parameters: the dimensionless flow resistance number and the dimensionless gravitational force number. By satisfying the equality of both dimensionless numbers

  19. Assessment of computational fluid dynamics (CFD) for nuclear reactor safety problems

    International Nuclear Information System (INIS)

    Smith, B. L.; Andreani, M.; Bieder, U.; Bestion, D.; Ducros, F.; Graffard, E.; Heitsch, M.; Scheuerer, M.; Henriksson, M.; Hoehne, T.; Rohde, U.; Lucas, D.; Komen, E.; Houkema, M.; Mahaffy, J.; Moretti, F.; Morii, T.; Muehlbauer, P.; Song, C.H.; Zigh, G.; Menter, F.; Watanabe, T.

    2008-01-01

    The basic objective of the present work was to provide documented evidence of the need to perform CFD simulations in Nuclear Reactor Safety (NRS), concentrating on single-phase applications, and to assess the competence of the present generation of CFD codes to perform these simulations reliably. The fulfilling of this objective involves multiple tasks, summarized as: to provide a classification of NRS problems requiring CFD analysis, to identify and catalogue existing CFD assessment bases, to identify shortcomings in CFD approaches, to put into place a means for extending the CFD assessment database, with an emphasis on NRS applications. The resulting document is presented here. After some introductory remarks, chapter 3 lists twenty-two NRS issues for which it is considered that the application of CFD would bring real benefits in terms of better predictive capability. This classification is followed by a short description of the safety issue, a state-of-the-art summary of what has been attempted, and what is still needed to be done to improve reliability. Chapter 4 details the assessment bases that have already been established in both the nuclear and non-nuclear domains, and discusses the usefulness and relevance of the work to NRS applications, where appropriate. This information is augmented in Chapter 5 by descriptions of the existing CFD assessment bases that have been established around specific, NRS problems. Typical examples are experiments devoted to the boron dilution issue, pressurised thermal shock, and thermal fatigue in pipes. Chapter 6 is devoted to identifying the technology gaps which need to be closed to make CFD a more trustworthy analytical tool. Some deficiencies identified are lack of a Phenomenon Identification and Ranking Table (PIRT), limitations in the range of application of turbulence models, coupling of CFD with neutronics and system codes, and computer power limitations. Most CFD codes currently being used have their own, custom

  20. Hydraulic engine valve actuation system including independent feedback control

    Science.gov (United States)

    Marriott, Craig D

    2013-06-04

    A hydraulic valve actuation assembly may include a housing, a piston, a supply control valve, a closing control valve, and an opening control valve. The housing may define a first fluid chamber, a second fluid chamber, and a third fluid chamber. The piston may be axially secured to an engine valve and located within the first, second and third fluid chambers. The supply control valve may control a hydraulic fluid supply to the piston. The closing control valve may be located between the supply control valve and the second fluid chamber and may control fluid flow from the second fluid chamber to the supply control valve. The opening control valve may be located between the supply control valve and the second fluid chamber and may control fluid flow from the supply control valve to the second fluid chamber.

  1. FONESYS: The FOrum and NEtwork of SYStem Thermal-Hydraulic Codes in Nuclear Reactor Thermal-Hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, S.H., E-mail: k175ash@kins.re.kr [Korea Institute of Nuclear Safety (KINS) (Korea, Republic of); Aksan, N., E-mail: nusr.aksan@gmail.com [University of Pisa San Piero a Grado Nuclear Research Group (GRNSPG) (Italy); Austregesilo, H., E-mail: henrique.austregesilo@grs.de [Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) (Germany); Bestion, D., E-mail: dominique.bestion@cea.fr [Commissariat à l’énergie atomique et aux énergies alternatives (CEA) (France); Chung, B.D., E-mail: bdchung@kaeri.re.kr [Korea Atomic Energy Research Institute (KAERI) (Korea, Republic of); D’Auria, F., E-mail: f.dauria@ing.unipi.it [University of Pisa San Piero a Grado Nuclear Research Group (GRNSPG) (Italy); Emonot, P., E-mail: philippe.emonot@cea.fr [Commissariat à l’énergie atomique et aux énergies alternatives (CEA) (France); Gandrille, J.L., E-mail: jeanluc.gandrille@areva.com [AREVA NP (France); Hanninen, M., E-mail: markku.hanninen@vtt.fi [VTT Technical Research Centre of Finland (VTT) (Finland); Horvatović, I., E-mail: i.horvatovic@ing.unipi.it [University of Pisa San Piero a Grado Nuclear Research Group (GRNSPG) (Italy); Kim, K.D., E-mail: kdkim@kaeri.re.kr [Korea Atomic Energy Research Institute (KAERI) (Korea, Republic of); Kovtonyuk, A., E-mail: a.kovtonyuk@ing.unipi.it [University of Pisa San Piero a Grado Nuclear Research Group (GRNSPG) (Italy); Petruzzi, A., E-mail: a.petruzzi@ing.unipi.it [University of Pisa San Piero a Grado Nuclear Research Group (GRNSPG) (Italy)

    2015-01-15

    Highlights: • We briefly presented the project called Forum and Network of System Thermal-Hydraulics Codes in Nuclear Reactor Thermal-Hydraulics (FONESYS). • We presented FONESYS participants and their codes. • We explained FONESYS projects motivation, its main targets and working modalities. • We presented FONESYS position about projects topics and subtopics. - Abstract: The purpose of this article is to present briefly the project called Forum and Network of System Thermal-Hydraulics Codes in Nuclear Reactor Thermal-Hydraulics (FONESYS), its participants, the motivation for the project, its main targets and working modalities. System Thermal-Hydraulics (SYS-TH) codes, also as part of the Best Estimate Plus Uncertainty (BEPU) approaches, are expected to achieve a more-and-more relevant role in nuclear reactor technology, safety and design. Namely, the number of code-users can easily be predicted to increase in the countries where nuclear technology is exploited. Thus, the idea of establishing a forum and a network among the code developers and with possible extension to code users has started to have major importance and value. In this framework the FONESYS initiative has been created. The main targets of FONESYS are: • To promote the use of SYS-TH Codes and the application of the BEPU approaches. • To establish acceptable and recognized procedures and thresholds for Verification and Validation (V and V). • To create a common ground for discussing envisaged improvements in various areas, including user-interface, and the connection with other numerical tools, including Computational Fluid Dynamics (CFD) Codes.

  2. Animal study assessing safety of an acoustic coupling fluid that holds the potential to avoid surgically induced artifacts in 3D ultrasound guided operations

    International Nuclear Information System (INIS)

    Jakola, Asgeir S; Jørgensen, Arve; Selbekk, Tormod; Michler, Ralf-Peter; Solheim, Ole; Torp, Sverre H; Sagberg, Lisa M; Aadahl, Petter; Unsgård, Geirmund

    2014-01-01

    Use of ultrasound in brain tumor surgery is common. The difference in attenuation between brain and isotonic saline may cause artifacts that degrade the ultrasound images, potentially affecting resection grades and safety. Our research group has developed an acoustic coupling fluid that attenuates ultrasound energy like the normal brain. We aimed to test in animals if the newly developed acoustic coupling fluid may have harmful effects. Eight rats were included for intraparenchymal injection into the brain, and if no adverse reactions were detected, 6 pigs were to be included with injection of the coupling fluid into the subarachnoid space. Animal behavior, EEG registrations, histopathology and immunohistochemistry were used in assessment. In total, 14 animals were included, 8 rats and 6 pigs. We did not detect any clinical adverse effects, seizure activity on EEG or histopathological signs of tissue damage. The novel acoustic coupling fluid intended for brain tumor surgery appears safe in rats and pigs under the tested circumstances

  3. Safety

    International Nuclear Information System (INIS)

    1998-01-01

    A brief account of activities carried out by the Nuclear power plants Jaslovske Bohunice in 1997 is presented. These activities are reported under the headings: (1) Nuclear safety; (2) Industrial and health safety; (3) Radiation safety; and Fire protection

  4. DESIGN AND CONSTRUCTION OF A HYDRAULIC PISTON

    OpenAIRE

    Santos De la Cruz, Eulogio; Rojas Lazo, Oswaldo; Yenque Dedios, Julio; Lavado Soto, Aurelio

    2014-01-01

    A hydraulic system project includes the design, materials selection and construction of the hydraulic piston, hydraulic circuit and the joint with the pump and its accesories. This equiment will be driven by the force of moving fluid, whose application is in the devices of machines, tools, printing, perforation, packing and others. El proyecto de un sistema hidráulico, comprende el diseño, selección de materiales y construcción del pistón hidráulico, circuito hidráulico y el ensamble con l...

  5. Safety and effectiveness evaluation of a domestic peritoneal dialysis fluid packed in non-PVC bags: study protocol for a randomized controlled trial.

    Science.gov (United States)

    Zhou, Jianhui; Cao, Xueying; Lin, Hongli; Ni, Zhaohui; He, Yani; Chen, Menghua; Zheng, Hongguang; Chen, Xiangmei

    2015-12-29

    Peritoneal dialysis is an important type of renal replacement therapy for uremic patients. In peritoneal dialysis, fluids fill in and flow out of the abdominal cavity three to five times per day. Usually, the fluid is packed in a polyvinyl chloride (PVC) bag. Safety concerns have arisen over di-(2-ethylhexyl) phthalate, which is essential in the formation of PVC materials. In 2011, the National Development and Reform Commission of China released a catalog of industrial structural adjustments, mandating the elimination of PVC bags for intravenous infusion and food containers. Although bags for peritoneal dialysis fluid were not included in the elimination list, several manufacturers began to develop new materials for fluid bags. HUAREN peritoneal dialysis fluid consists of the same electrolytes and buffer agent as in Baxter fluid, but is packed in bags that do not contain PVC. This multicenter randomized controlled trial was designed to compare peritoneal dialysis fluid packed in non-PVC-containing and PVC-containing bags. Further, the study sought to determine the proper dose of peritoneal dialysis fluid and the actual survival rates of Chinese patients undergoing peritoneal dialysis. The study participants are adults undergoing continuous ambulatory peritoneal dialysis for 30 days to 6 months. All eligible patients are randomized (1:1) to peritoneal dialysis with Baxter and HUAREN dialysis fluids (initial dose, 6 l/day), with dosages adjusted according to a unified protocol. The primary outcomes are the 1-, 2-, 3-, 4-, and 5-year overall survival rates. Secondary outcome measures include technique survival rates, reductions in estimated glomerular filtration rate, nutritional status, quality of life, cardiovascular events, medical costs and drop-out rates. Safety outcome measures include adverse events, changes in vital signs and laboratory parameters, peritonitis, allergies, and quality of products. This study is the first to evaluate the long-term safety and

  6. Hydraulic conductivity of rock fractures

    International Nuclear Information System (INIS)

    Zimmerman, R.W.; Bodvarsson, G.S.

    1994-10-01

    Yucca Mountain, Nevada contains numerous geological units that are highly fractured. A clear understanding of the hydraulic conductivity of fractures has been identified as an important scientific problem that must be addressed during the site characterization process. The problem of the flow of a single-phase fluid through a rough-walled rock fracture is discussed within the context of rigorous fluid mechanics. The derivation of the cubic law is given as the solution to the Navier-Stokes equations for flow between smooth, parallel plates, the only fracture geometry that is amenable to exact treatment. The various geometric and kinetic conditions that are necessary in order for the Navier-Stokes equations to be replaced by the more tractable lubrication or Hele-Shaw equations are studied and quantified. Various analytical and numerical results are reviewed pertaining to the problem of relating the effective hydraulic aperture to the statistics of the aperture distribution. These studies all lead to the conclusion that the effective hydraulic aperture is always less than the mean aperture, by a factor that depends on the ratio of the mean value of the aperture to its standard deviation. The tortuosity effect caused by regions where the rock walls are in contact with each other is studied using the Hele-Shaw equations, leading to a simple correction factor that depends on the area fraction occupied by the contact regions. Finally, the predicted hydraulic apertures are compared to measured values for eight data sets from the literature for which aperture and conductivity data were available on the same fracture. It is found that reasonably accurate predictions of hydraulic conductivity can be made based solely on the first two moments of the aperture distribution function, and the proportion of contact area. 68 refs

  7. A Computational Model of Hydraulic Volume Displacement Drive

    Directory of Open Access Journals (Sweden)

    V. N. Pil'gunov

    2014-01-01

    Full Text Available The paper offers a computational model of industrial-purpose hydraulic drive with two hydraulic volume adjustable working chamber machines (pump and motor. Adjustable pump equipped with the pressure control unit can be run together with several adjustable hydraulic motors on the principle of three-phase hydraulic socket-outlet with high-pressure lines, drain, and drainage system. The paper considers the pressure-controlled hydrostatic transmission with hydraulic motor as an output link. It shows a possibility to create a saving hydraulic drive using a functional tie between the adjusting parameters of the pump and hydraulic motor through the pressure difference, torque, and angular rate of the hydraulic motor shaft rotation. The programmable logic controller can implement such tie. The Coulomb and viscous frictions are taken into consideration when developing a computational model of the hydraulic volume displacement drive. Discharge balance considers external and internal leakages in equivalent clearances of hydraulic machines, as well as compression loss volume caused by hydraulic fluid compressibility and deformation of pipe walls. To correct dynamic properties of hydraulic drive, the paper offers that in discharge balance are included the additional regulated external leakages in the open circuit of hydraulic drive and regulated internal leakages in the closed-loop circuit. Generalized differential equations having functional multipliers and multilinked nature have been obtained to describe the operation of hydraulic positioning and speed drive with two hydraulic volume adjustable working chamber machines. It is shown that a proposed computational model of hydraulic drive can be taken into consideration in development of LS («Load-Sensing» drives, in which the pumping pressure is tuned to the value required for the most loaded slave motor to overcome the load. Results attained can be used both in designing the industrial-purpose heavy

  8. Fluid and particle mechanics

    CERN Document Server

    Michell, S J

    2013-01-01

    Fluid and Particle Mechanics provides information pertinent to hydraulics or fluid mechanics. This book discusses the properties and behavior of liquids and gases in motion and at rest. Organized into nine chapters, this book begins with an overview of the science of fluid mechanics that is subdivided accordingly into two main branches, namely, fluid statics and fluid dynamics. This text then examines the flowmeter devices used for the measurement of flow of liquids and gases. Other chapters consider the principle of resistance in open channel flow, which is based on improper application of th

  9. 11. international topical meeting on nuclear reactor thermal-hydraulics (NURETH-11)

    International Nuclear Information System (INIS)

    Lemonnier, H.

    2005-01-01

    ; aerosol transport, deposition and re-entrainment; steam generators thermal-hydraulics; system codes development and assessment; uncertainties analysis; diffuse interface methods and interface tracking methods; C - severe accidents and fires: molten core natural convection and physico-chemical phenomena, modeling and experiments; fuel coolant interaction, modeling and experiments; debris bed cooling; combustion and fires, modeling and experiments; molten corium concrete interaction; D - advanced code developments: fast transient modelling and experiments; multidimensional single-phase or two-phase flow and heat transfer modeling; neutronics and thermal-hydraulics coupling; fluid and structures mechanical interactions; coupled thermal-hydraulics of fluids and structures; thermal-hydraulic dependent corrosion and ablation; E - operation and safety of existing reactors: instabilities and nonlinear dynamics; NPP transients and accidents analysis; RBMK and VVER safety analysis, including the OECD benchmark; F - experimental thermal-hydraulics: boiling heat transfer; CHF and post-CHF heat transfer; condensation heat transfer; integral testing; vibrations, wear and thermal fatigue phenomena; fuel design and performance; G - advanced reactors thermal-hydraulics (gen IV, INPRO, fusion, hydrogen production): accelerator driven reactors; advanced pressurized water reactors thermal-hydraulics; gas cooled fast reactors; gas cooled high temperature reactors; lead and lead-bismuth cooled reactors; future and existing sodium cooled reactors; molten salt reactors; H - waste management thermal-hydraulics: thermal-hydraulics problems related to waste processing and storage; I - thermal-hydraulics of non electricity generating nuclear equipment: sono-fusion (cavitation induced bubble fusion; hydrogen producing nuclear reactors

  10. Advanced modelling and numerical strategies in nuclear thermal-hydraulics

    International Nuclear Information System (INIS)

    Staedtke, H.

    2001-01-01

    The first part of the lecture gives a brief review of the current status of nuclear thermal hydraulics as it forms the basis of established system codes like TRAC, RELAP5, CATHARE or ATHLET. Specific emphasis is given to the capabilities and limitations of the underlying physical modelling and numerical solution strategies with regard to the description of complex transient two-phase flow and heat transfer conditions as expected to occur in PWR reactors during off-normal and accident conditions. The second part of the lecture focuses on new challenges and future needs in nuclear thermal-hydraulics which might arise with regard to re-licensing of old plants using bestestimate methodologies or the design and safety analysis of Advanced Light Water Reactors relying largely on passive safety systems. In order to meet these new requirements various advanced modelling and numerical techniques will be discussed including extended wellposed (hyperbolic) two-fluid models, explicit modelling of interfacial area transport or higher order numerical schemes allowing a high resolution of local multi-dimensional flow processes.(author)

  11. Directly acting spring loaded safety valves as shock reducing measure

    International Nuclear Information System (INIS)

    Ismaier, A.; Schluecker, E.

    2010-01-01

    Hydraulic shocks as induced by fast closure of armatures or by sudden pump failures are massive impacts in piping systems and require extensive measures to absorb the generated load. Basically the avoidance of water hammers are preferable but in case of emergency shutdowns unavoidable hydraulic shocks have to be reduced by appropriate measures. The authors describe experiments with spring loaded safety valves as shock reducing measures. It was shown that the vale dimensions is essential for the efficacy. A realistic modeling is possible using the one-dimensional fluid mechanics code ROLAST.

  12. Uncertainty in hydraulic tests in fractured rock

    International Nuclear Information System (INIS)

    Ji, Sung-Hoon; Koh, Yong-Kwon

    2014-01-01

    Interpretation of hydraulic tests in fractured rock has uncertainty because of the different hydraulic properties of a fractured rock to a porous medium. In this study, we reviewed several interesting phenomena which show uncertainty in a hydraulic test at a fractured rock and discussed their origins and the how they should be considered during site characterisation. Our results show that the estimated hydraulic parameters of a fractured rock from a hydraulic test are associated with uncertainty due to the changed aperture and non-linear groundwater flow during the test. Although the magnitude of these two uncertainties is site-dependent, the results suggest that it is recommended to conduct a hydraulic test with a little disturbance from the natural groundwater flow to consider their uncertainty. Other effects reported from laboratory and numerical experiments such as the trapping zone effect (Boutt, 2006) and the slip condition effect (Lee, 2014) can also introduce uncertainty to a hydraulic test, which should be evaluated in a field test. It is necessary to consider the way how to evaluate the uncertainty in the hydraulic property during the site characterisation and how to apply it to the safety assessment of a subsurface repository. (authors)

  13. Hydraulic fluid used for power transmission

    Energy Technology Data Exchange (ETDEWEB)

    Heikkilae, M.; Pikulinsky, K.; Leisio, C.

    1996-11-01

    In early October another 50-kilowatt wind turbine was provided with new power transmission technology at the Kopparnaes Energy Park in Inkoo, Finland, west of Helsinki. The new technology is thought to make this wind turbine located on the south coast of Finland more efficient, lighter, and cheaper. Certain aspects of this new technology can be applied to older wind turbines. (orig.)

  14. Liquid Chromatographic Analysis of Hydraulic Fluids.

    Science.gov (United States)

    1979-11-01

    chemical mixtures of a petroleum- or nonpetroleum-base stock component formulated with various additives which may be present in trace amounts or...absorb UV radiation near the monitoring wavelength may swamp the detector signal and therefore should be avoided in 1JV detection. The recorder trace of...Also, organic phosphites , thiophosphates, and sulfides are used to inhibit oxidative catalysis by metal ions. The oxidation inhibitor in 6083D-0 is BPC

  15. Analysis of buffering process of control rod hydraulic absorber

    International Nuclear Information System (INIS)

    Bao Jishi; Qin Benke; Bo Hanliang

    2011-01-01

    Control Rod Hydraulic Drive Mechanism(CRHDM) is a newly invented build-in control rod drive mechanism. Hydraulic absorber is the key part of this mechanism, and is used to cushion the control rod when the rod scrams. Thus, it prevents the control rod from being deformed and damaged. In this paper dynamics program ANSYS CFX is used to calculate all kinds of flow conditions in hydraulic absorber to obtain its hydraulic characteristics. Based on the flow resistance coefficients obtained from the simulation results, fluid mass and momentum equations were developed to get the trend of pressure change in the hydraulic cylinder and the displacement of the piston rod during the buffering process of the control rod. The results obtained in this paper indicate that the hydraulic absorber meets the design requirement. The work in this paper will be helpful for the design and optimization of the control rod hydraulic absorber. (author)

  16. Use of computational fluid dynamics codes for safety analysis of nuclear reactor systems, including containment. Summary report of a technical meeting

    International Nuclear Information System (INIS)

    2003-11-01

    Safety analysis is an important tool for justifying the safety of nuclear power plants. Typically, this type of analysis is performed by means of system computer codes with one dimensional approximation for modelling real plant systems. However, in the nuclear area there are issues for which traditional treatment using one dimensional system codes is considered inadequate for modelling local flow and heat transfer phenomena. There is therefore increasing interest in the application of three dimensional computational fluid dynamics (CFD) codes as a supplement to or in combination with system codes. There are a number of both commercial (general purpose) CFD codes as well as special codes for nuclear safety applications available. With further progress in safety analysis techniques, the increasing use of CFD codes for nuclear applications is expected. At present, the main objective with respect to CFD codes is generally to improve confidence in the available analysis tools and to achieve a more reliable approach to safety relevant issues. An exchange of views and experience can facilitate and speed up progress in the implementation of this objective. Both the International Atomic Energy Agency (IAEA) and the Nuclear Energy Agency of the Organisation for Economic Co-operation and Development (OECD/NEA) believed that it would be advantageous to provide a forum for such an exchange. Therefore, within the framework of the Working Group on the Analysis and Management of Accidents of the NEA's Committee on the Safety of Nuclear Installations, the IAEA and the NEA agreed to jointly organize the Technical Meeting on the Use of Computational Fluid Dynamics Codes for Safety Analysis of Reactor Systems, including Containment. The meeting was held in Pisa, Italy, from 11 to 14 November 2002. The publication constitutes the report of the Technical Meeting. It includes short summaries of the presentations that were made and of the discussions as well as conclusions and

  17. Hydraulic Hybrid Vehicles

    Science.gov (United States)

    EPA and the United Parcel Service (UPS) have developed a hydraulic hybrid delivery vehicle to explore and demonstrate the environmental benefits of the hydraulic hybrid for urban pick-up and delivery fleets.

  18. SAFETY

    CERN Multimedia

    Niels Dupont

    2013-01-01

    CERN Safety rules and Radiation Protection at CMS The CERN Safety rules are defined by the Occupational Health & Safety and Environmental Protection Unit (HSE Unit), CERN’s institutional authority and central Safety organ attached to the Director General. In particular the Radiation Protection group (DGS-RP1) ensures that personnel on the CERN sites and the public are protected from potentially harmful effects of ionising radiation linked to CERN activities. The RP Group fulfils its mandate in collaboration with the CERN departments owning or operating sources of ionising radiation and having the responsibility for Radiation Safety of these sources. The specific responsibilities concerning "Radiation Safety" and "Radiation Protection" are delegated as follows: Radiation Safety is the responsibility of every CERN Department owning radiation sources or using radiation sources put at its disposition. These Departments are in charge of implementing the requi...

  19. Advanced Hydraulic Studies on Enhancing Particle Removal

    DEFF Research Database (Denmark)

    He, Cheng

    clarifier. The inlet zone of an existing rectangular storm water clarifier was redesigned to improve the fluid flow conditions and reduce the hydraulic head loss in order to remove the lamellar plates and adapt the clarifier to the needs of high-rate clarification of storm water with flocculant addition...... excessive local head losses and helped to select structural changes to reduce such losses. The analysis of the facility showed that with respect to hydraulic operation, the facility is a complex, highly non-linear hydraulic system. Within the existing constraints, a few structural changes examined......The removal of suspended solids and attached pollutants is one of the main treatment processes in wastewater treatment. This thesis presents studies on the hydraulic conditions of various particle removal facilities for possible ways to increase their treatment capacity and performance by utilizing...

  20. TH3D, a three-dimensional thermal hydraulic tool, for design and safety analysis of HTRS - HTR2008-58178

    International Nuclear Information System (INIS)

    Hossain, K.; Buck, M.; Bernnat, W.; Lohnert, G.

    2008-01-01

    The institute of nuclear engineering and energy systems (IKE), Univ. of Stuttgart (Germany)) has developed a new thermal hydraulic tool which can be used for three-dimensional thermal hydraulic analysis of pebble bed as well as block type HTRs. During nominal operation, the flow inside the gas-cooled High Temperature Reactor is essentially single-phase, impressible, and non-isothermal. So, at least one gas phase has to be considered beside the solid phase for thermal hydraulic analysis of HTRs. Each phase (e.g. solid, gas) is considered as a continuum which occupies only its respective fraction of. the control volume. Thermal non-equilibrium is considered between phases and time dependent energy conservation equations for solid and gas phases are solved. Simplified momentum conservation equation for gas obtained from porous media approximation is solved along with the time dependent mass conservation equation. Pro visions for simulating more than one gas component are available in this newly developed code TH3D which could be required for simulating some accident situations (e.g air / water ingress by pipes break). The interaction between phases is made by a set of constitutive equations which re/v on semi-empirical correlations obtained from different experiments. Finite volume method with a staggered grid approach is used for spatial discretization and a fully implicit, time adaptive, multi step method is used for time-dependent discretization. A benchmark calculation which is oriented to the pebble i fuel reactor PBMR-400 and a 3D calculation were presented in HTR -2006 conference and will also be published in Nuclear Engineering and Design (NED) journal. In order to demonstrate the capabilities of TH3D for simulating all block type HTRs. A benchmark calculation which is proposed by IAEA CRP-3 and oriented to the Gas Turbine Modular Helium Reactor (GT-MHR) is performed. calculations are performed for the steady state case (nominal operation) as well as for Loss

  1. Calculation Sheet for the Basic Design of the ATLAS Fluid System

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyun Sik; Moon, S. K.; Yun, B. J.; Kwon, T. S.; Choi, K. Y.; Cho, S.; Park, C. K.; Lee, S. J.; Kim, Y. S.; Song, C. H.; Baek, W. P.; Hong, S. D

    2007-03-15

    The basic design of an integral effect test loop for pressurized water reactors (PWRs), the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), has been carried out by Thermal-Hydraulics Safety Research Team in Korea Atomic Energy Research Institute (KAERI). The ATLAS facility has been designed to have the length scale of 1/2 and area scale of 1/144 compared with the reference plant, APR1400, and is scaled for full pressure and temperature conditions. This report includes calculation sheets for the basic design of ATLAS fluid systems, which are consisted of a reactor pressure vessel with core simulator, the primary loop piping, a pressurizer, reactor coolant pumps, steam generators, the secondary system, the safety system, the auxiliary system, and the heat loss compensation system. The present calculation sheets will be used to help understanding the basic design of the ATLAS fluid system and its based scaling methodology.

  2. Calculation Sheet for the Basic Design of the ATLAS Fluid System

    International Nuclear Information System (INIS)

    Park, Hyun Sik; Moon, S. K.; Yun, B. J.; Kwon, T. S.; Choi, K. Y.; Cho, S.; Park, C. K.; Lee, S. J.; Kim, Y. S.; Song, C. H.; Baek, W. P.; Hong, S. D.

    2007-03-01

    The basic design of an integral effect test loop for pressurized water reactors (PWRs), the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), has been carried out by Thermal-Hydraulics Safety Research Team in Korea Atomic Energy Research Institute (KAERI). The ATLAS facility has been designed to have the length scale of 1/2 and area scale of 1/144 compared with the reference plant, APR1400, and is scaled for full pressure and temperature conditions. This report includes calculation sheets for the basic design of ATLAS fluid systems, which are consisted of a reactor pressure vessel with core simulator, the primary loop piping, a pressurizer, reactor coolant pumps, steam generators, the secondary system, the safety system, the auxiliary system, and the heat loss compensation system. The present calculation sheets will be used to help understanding the basic design of the ATLAS fluid system and its based scaling methodology

  3. Analysis of hydraulic instability of ANS involute fuel plates

    International Nuclear Information System (INIS)

    Sartory, W.K.

    1991-11-01

    Curved shell equations for the involute Advanced Neutron Source (ANS) fuel plates are coupled to two-dimensional hydraulic channel flow equations that include fluid friction. A complete set of fluid and plate boundary conditions is applied at the entrance and exit and along the sides of the plate and the channel. The coupled system is linearized and solved to assess the hydraulic instability of the plates

  4. Hydraulic brake-system for a bicycle

    NARCIS (Netherlands)

    Van Frankenhuyzen, J.

    2007-01-01

    The invention relates to a hydraulic brake system for a bicycle which may or may not be provided with an auxiliary motor, comprising a brake disc and brake claws cooperating with the brake disc, as well as fluid-containing channels (4,6) that extend between an operating organ (1) and the brake

  5. Analyses of hydraulic performance of velocity caps

    DEFF Research Database (Denmark)

    Christensen, Erik Damgaard; Degn Eskesen, Mark Chr.; Buhrkall, Jeppe

    2014-01-01

    The hydraulic performance of a velocity cap has been investigated. Velocity caps are often used in connection with offshore intakes. CFD (computational fluid dynamics) examined the flow through the cap openings and further down into the intake pipes. This was combined with dimension analyses...

  6. Separation and pattern formation in hydraulic jumps

    DEFF Research Database (Denmark)

    Bohr, Tomas; Ellegaard, C.; Hansen, A. Espe

    1998-01-01

    We present theory and experiments on the circular hydraulic jump in the stationary regime. The theory can handle the situation in which the fluid flows over an edge far away from the jump. In the experiments the external height is controlled, and a series of transitions in the flow structure appe...

  7. Thermal-hydraulic analysis and design improvement for coolant channel of ITER shield block

    International Nuclear Information System (INIS)

    Zhao Ling; Li Huaqi; Zheng Jiantao; Yi Jingwei; Kang Weishan; Chen Jiming

    2013-01-01

    As an important part for ITER, shield block is used to shield the neutron heat. The structure design of shield block, especially the inner coolant channel design will influence its cooling effect and safety significantly. In this study, the thermal-hydraulic analysis for shield block has been performed by the computational fluid dynamics software, some optimization suggestions have been proposed and thermal-hydraulic characteristics of the improved model has been analyzed again. The analysis results for improved model show that pressure drop through flow path near the inlet and outlet region of the shield block has been reduced, and the total pressure drop in cooling path has been reduced too; the uniformity of the mass flowrate distribution and the velocity distribution have been improved in main cooling branches; the local highest temperature of solid domain reduced considerably, which could avoid thermal stress becoming too large because of coolant effect unevenly. (authors)

  8. Numerical modeling of local scour around hydraulic structure in sandy beds by dynamic mesh method

    Science.gov (United States)

    Fan, Fei; Liang, Bingchen; Bai, Yuchuan; Zhu, Zhixia; Zhu, Yanjun

    2017-10-01

    Local scour, a non-negligible factor in hydraulic engineering, endangers the safety of hydraulic structures. In this work, a numerical model for simulating local scour was constructed, based on the open source code computational fluid dynamics model OpenFOAM. We consider both the bedload and suspended load sediment transport in the scour model and adopt the dynamic mesh method to simulate the evolution of the bed elevation. We use the finite area method to project data between the three-dimensional flow model and the two-dimensional (2D) scour model. We also improved the 2D sand slide method and added it to the scour model to correct the bed bathymetry when the bed slope angle exceeds the angle of repose. Moreover, to validate our scour model, we conducted and compared the results of three experiments with those of the developed model. The validation results show that our developed model can reliably simulate local scour.

  9. Requirements to be taken into account when designing safety-related mechanical components conveying or containing pressurized fluid and classified as level 2 or 3

    International Nuclear Information System (INIS)

    1984-12-01

    RFS or Regles Fondamentales de Surete (Basic Safety Rules) applicable to certain types of nuclear facilities lay down requirements with which compliance, for the type of facilities and within the scope of application covered by the RFS, is considered to be equivalent to compliance with technical French regulatory practice. The object of the RFS is to take advantage of standardization in the field of safety, while allowing for technical progress in that field. They are designed to enable the operating utility and contractors to know the rules pertaining to various subjects which are considered to be acceptable by the Service Central de Surete des Installations Nucleaires, or the SCSIN (Central Department for the Safety of Nuclear Facilities). These RFS should make safety analysis easier and lead to better understanding between experts and individuals concerned with the problems of nuclear safety. The SCSIN reserves the right to modify, when considered necessary, any RFS and specify, if need be, the terms under which a modification is deemed retroactive. The purpose of this RFS is to specify the requirements to be taken into account when designing mechanical components conveying or containing pressurized fluid and which are in safety class 2 or 3

  10. Safety

    International Nuclear Information System (INIS)

    2001-01-01

    This annual report of the Senior Inspector for the Nuclear Safety, analyses the nuclear safety at EDF for the year 1999 and proposes twelve subjects of consideration to progress. Five technical documents are also provided and discussed concerning the nuclear power plants maintenance and safety (thermal fatigue, vibration fatigue, assisted control and instrumentation of the N4 bearing, 1300 MW reactors containment and time of life of power plants). (A.L.B.)

  11. Twenty-second water reactor safety information meeting. Volume 2: Severe accident research, thermal hydraulic research for advanced passive LWRs, high-burnup fuel behavior

    International Nuclear Information System (INIS)

    Monteleone, S.

    1995-04-01

    This three-volume report contains papers presented at the Twenty-Second Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 24-26, 1994. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Finland, France, Italy, Japan, Russia, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting

  12. Twenty-second water reactor safety information meeting. Volume 2: Severe accident research, thermal hydraulic research for advanced passive LWRs, high-burnup fuel behavior

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.

    1995-04-01

    This three-volume report contains papers presented at the Twenty-Second Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 24-26, 1994. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Finland, France, Italy, Japan, Russia, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting.

  13. Coupled neutronic-thermal-hydraulics analysis in a coolant subchannel of a PWR using CFD techniques

    Energy Technology Data Exchange (ETDEWEB)

    Ribeiro, Felipe P.; Su, Jian, E-mail: sujian@nuclear.ufrj.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

    2017-07-01

    The high capacity of Computational Fluid Dynamics code to predict multi-dimensional thermal-hydraulics behaviour and the increased availability of capable computer systems are making that method a good tool to simulate phenomena of thermal-hydraulics nature in nuclear reactors. However, since there are no neutron kinetics models available in commercial CFD codes to the present day, the application of CFD in the nuclear reactor safety analyses is still limited. The present work proposes the implementation of the point kinetics model (PKM) in ANSYS - Fluent to predict the neutronic behaviour in a Westinghouse Sequoyah nuclear reactor, coupling with the phenomena of heat conduction in the rod and thermal-hydraulics in the cooling fluid, via the reactivity feedback. Firstly, a mesh convergence and turbulence model study was performed, using the Reynolds-Average Navier-Stokes method, with square arrayed rod bundle featuring pitch to diameter ratio of 1:32. Secondly, simulations using the k-! SST turbulence model were performed with an axial distribution of the power generation in the fuel to analyse the heat transfer through the gap and cladding, and its in fluence on the thermal-hydraulics behaviour of the cooling fluid. The wall shear stress distribution for the centre-line rods and the dimensionless velocity were evaluated to validate the model, as well as the in fluence of the mass flow rate variation on the friction factor. The coupled model enabled to perform a dynamic analysis of the nuclear reactor during events of insertion of reactivity and shutdown of primary coolant pumps. (author)

  14. An evaluation of calculation procedures affecting the constituent factors of equivalent circulating density for drilling hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Bailey, William J

    1997-12-31

    This Dr. ing. thesis covers a study of drilling hydraulics offshore. The purpose of drilling hydraulics is to provide information about downhole pressure, suitable surface pump rates, the quality of hole cleaning and optimum tripping speeds during drilling operations. Main fields covered are drilling hydraulics, fluid characterisation, pressure losses, and equivalent circulating density. 197 refs., 23 figs., 22 tabs.

  15. An evaluation of calculation procedures affecting the constituent factors of equivalent circulating density for drilling hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Bailey, William J.

    1996-12-31

    This Dr. ing. thesis covers a study of drilling hydraulics offshore. The purpose of drilling hydraulics is to provide information about downhole pressure, suitable surface pump rates, the quality of hole cleaning and optimum tripping speeds during drilling operations. Main fields covered are drilling hydraulics, fluid characterisation, pressure losses, and equivalent circulating density. 197 refs., 23 figs., 22 tabs.

  16. Hydraulic braking system for loads subjected to impacts and vibrations

    International Nuclear Information System (INIS)

    1980-01-01

    This invention concerns a hydraulic braking system for loads subjected to impacts and vibrations. These double acting telescopic type hydraulic braking systems possess significant drawbacks linked to possibly important hydraulic leaks due to (a) the use of many dynamic seals in such appliances and (b) the effects of the environment of the system on these seals, particularly when employed in nuclear power stations where the seals reach significant temperatures and are subjected to radiation. Under this invention a remedy is suggested to such drawbacks by integrating means to offset automatically the leaks and the accumulation of hydraulic fluid expansions, as well as facilities to show if such leaks have occurred [fr

  17. Numerical simulation of temperature's sensitivity of chamfer hole's resistance on hydraulic step cylinder

    International Nuclear Information System (INIS)

    Jinhua, Wang; Hanliang, Bo; Wenxiang, Zheng; Jinnong, Yang

    2003-01-01

    The control rod drive is a very important device for controlling nuclear reactor startup, operation, shut down, and power change. The ability of the control rod drive to move safely and reliably directly relates to reactor safety. The Hydraulic Control Rod Drive System (HCRDS) is a new type of control rod drive system developed by the Institute of Nuclear Energy Technology (INET) of Tsinghua University for Nuclear Heating Reactors. The HCRDS, designed using the hydrodynamic principle, has many advantages, including having the structure complete in the vessel, no possible ejection accident, short drive line, simple movable parts structure and safe shutdown during accidents. The hydraulic step cylinder is the key part for the HCRDS. In the process of reactor startup, the variation of temperature could make the water's density and viscosity change, and the force from the water flow would change accordingly. These factors could influence the performance of the hydraulic step cylinder. In this paper, the temperature sensitivity of the chamfer hole's resistance in the hydraulic step cylinder was studied with the Computational Fluid Dynamics (CFD) program CFX5.5. The results were satisfactory: the discipline of variation of the chamfer hole's resistance with the outer tube's position was the same at different temperatures, the discrepancy of the chamfer hole's resistance was small for the same position at different temperatures, the chamfer hole's resistance decreased gradually with the increase of temperature, and the decrease extent was relatively small

  18. Aging and service wear of hydraulic and mechanical snubbers used on safety-related piping and components of nuclear power plants. Phase I study

    Energy Technology Data Exchange (ETDEWEB)

    Bush, S H; Heasler, P G; Dodge, R E

    1986-02-01

    This report presents an overview of hydraulic and mechanical snubbers used on nuclear piping systems and components, based on information from the literature and other sources. The functions and functional requirements of snubbers are discussed. The real versus perceived need for snubbers is reviewed, based primarily on studies conducted by a Pressure Vessel Research Committee. Tests conducted to qualify snubbers, to accept them on a case-by-case basis, and to establish their fitness for continued operation are reviewed. This report had two primary purposes. The first was to assess the effects of various aging mechanisms on snubber operation. The second was to determine the efficacy of existing tests in determining the effects of aging and degradation mechanisms. These tests include breakaway force, drag force, velocity/ acceleration range for activation in tension or compression, release rates within specified tension/compression limits, and restricted thermal movement. The snubber operating experience was reviewed using licensee event reports and other historical data for a period of more than 10 years. Data were statistically analyzed using arbitrary snubber populations. Value-impact was considered in terms of exposure to a radioactive environment for examination/ testing and the influence of lost snubber function and subsequent testing program expansion on the costs and operation of a nuclear power plant. The implications of the observed trends were assessed; recommendations include modifying or improving examination and testing procedures to enhance snubber reliability. Optimization of snubber populations by selective removal of unnecessary snubbers was also considered. (author)

  19. Thermal-hydraulic unreliability of passive systems

    International Nuclear Information System (INIS)

    Tzanos, C.P.; Saltos, N.T.

    1995-01-01

    Advanced light water reactor designs like AP600 and the simplified boiling water reactor (SBWR) use passive safety systems for accident prevention and mitigation. Because these systems rely on natural forces for their operation, their unavailability due to hardware failures and human error is significantly smaller than that of active systems. However, the coolant flows predicted to be delivered by these systems can be subject to significant uncertainties, which in turn can lead to a significant uncertainty in the predicted thermal-hydraulic performance of the plant under accident conditions. Because of these uncertainties, there is a probability that an accident sequence for which a best estimate thermal-hydraulic analysis predicts no core damage (success sequence) may actually lead to core damage. For brevity, this probability will be called thermal-hydraulic unreliability. The assessment of this unreliability for all the success sequences requires very expensive computations. Moreover, the computational cost increases drastically as the required thermal-hydraulic reliability increases. The required computational effort can be greatly reduced if a bounding approach can be used that either eliminates the need to compute thermal-hydraulic unreliabilities, or it leads to the analysis of a few bounding sequences for which the required thermal-hydraulic reliability is relatively small. The objective of this paper is to present such an approach and determine the order of magnitude of the thermal-hydraulic unreliabilities that may have to be computed

  20. Hydraulic regenerative system for a light vehicle

    OpenAIRE

    Orpella Aceret, Jordi; Guinart Trayter, Xavier

    2009-01-01

    The thesis is based in a constructed light vehicle that must be improved by adding a hydraulic energy recovery system. This vehicle named as TrecoLiTH, participated in the Formula Electric and Hybrid competition (Formula EHI) 2009 in Italy -Rome- and won several awards. This system consists in two hydraulic motors hub mounted which are used to store fluid at high pressure in an accumulator when braking. Through a valve the pressure will flow from the high pressure accumulator to the low press...

  1. Lateral hydraulic forces calculation on PWR fuel assemblies with computational fluid dynamics codes; Calculo de fuerzas laterales hidraulicas en elementos combustibles tipo PWR con codigos de dinamica de fluidos coputacional

    Energy Technology Data Exchange (ETDEWEB)

    Corpa Masa, R.; Jimenez Varas, G.; Moreno Garcia, B.

    2016-08-01

    To be able to simulate the behavior of nuclear fuel under operating conditions, it is required to include all the representative loads, including the lateral hydraulic forces which were not included traditionally because of the difficulty of calculating them in a reliable way. Thanks to the advance in CFD codes, now it is possible to assess them. This study calculates the local lateral hydraulic forces, caused by the contraction and expansion of the flow due to the bow of the surrounding fuel assemblies, on of fuel assembly under typical operating conditions from a three loop Westinghouse PWR reactor. (Author)

  2. Contribution to the study of thermal-hydraulic problems in nuclear reactors

    International Nuclear Information System (INIS)

    Cognet, G.

    1998-01-01

    In nuclear reactors, whatever the type considered, Pressurized Water Water Reactors (PWRs), Fast Breeder reactors (FBRs)..., thermal-hydraulics, the science of fluid mechanics and thermal behaviour, plays an essential role, both in nominal operating and accidental conditions. Fluid can either be the primary fluid (liquid or gas) or a very specific fluid called corium, which, in case of severe accident, could result from core and environning structure melting. The work reported here represents a 20-year contribution to thermal-hydraulic issues which could occur in FBRs and PWRs. Working on these two types of reactors, both in nominal and severe accident situations, has allowed me to compare the problems and to realize the importance of communication between research teams. The evolution in the complexity of studied problems, unavoidable in order to reduce costs and significantly improve safety, has led me from numerical modelling of single-phase flow turbulence to high temperature real melt experiments. The difficulties encountered in understanding the observed phenomena and in increasing experimental databases for computer code qualification have often entailed my participation in specific measurement device developments or adaptations, in particular non-intrusive devices generally based on optical techniques. Being concerned about the end-use of this research work, I actively participated in 'in-situ' thermalhydraulic experiments in the FBRs: Phenix and Super-Phenix, of which I appreciated their undeniable scientific contribution. In my opinion, the thermal-hydraulic questions related to severe accidents are the most complex as they are at the cross-roads of several scientific specialities. Consequently, they require a multi-disciplinary approach and a continuous see-saw motion between experimentalists and modelling teams. After a brief description of the various problems encountered, the main ones are reported. Finally, the importance for research teams to

  3. Overview of NSSS Fluid System Design of PGSFR

    Energy Technology Data Exchange (ETDEWEB)

    Han, Ji-Woong; Choi, Seok-Ki; Kim, Seong-O; Kim, Eui-Kwang; Kim, Dehee; Hong, Jonggan; Ye, Huee-Youl; Yeom, Sujin; Ryu, Seungho; Yoon, Jung; Choi, Sun Rock; Park, Jin-Seok; Lee, Tae-Ho Lee [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In this paper an overview on the NSSS fluid system design of PGSFR is described based on the issued design documents. System concepts and major components design concepts for PHTS, IHTS, DHRS and SWRPRS were developed. Thermal-hydraulic characteristics were analyzed based on CFD simulation. The design bases and concepts for auxiliary systems were also developed. The upstream design requirements of fluid system such as system design requirements, component design requirements, I and C design requirements, BOP interface design requirements, design guides and P and IDs were produced. The control logic and computer code for the analysis for operational characteristics is under progress. The protection system consists of a safety grade PPS and a non-safety grade DPS (Diverse Protection System). The DPS provides a diverse method to trip the reactor to satisfy the requirements relative to ATWS (Anticipated Transients Without Scram) as well as Defense-In-Depth and Diversity.

  4. International training program in support of safety analysis. 3D S.UN.COP-scaling uncertainty and 3D thermal-hydraulics/neutron-kinetics coupled codes seminars

    International Nuclear Information System (INIS)

    Petruzzi, Alessandro; D'Auria, Francesco; Bajs, Tomislav; Reventos, Francesc; Hassan, Yassin

    2007-01-01

    Thermal-hydraulic system computer codes are extensively used worldwide for analysis of nuclear facilities by utilities, regulatory bodies, nuclear power plant designers and vendors, nuclear fuel companies, research organizations, consulting companies, and technical support organizations. The computer code user represents a source of uncertainty that can influence the results of system code calculations. This influence is commonly known as the user effect' and stems from the limitations embedded in the codes as well as from the limited capability of the analysis to use the codes. Code user training and qualification is an effective means for reducing the variation of results caused by the application of the codes by different users. This paper describes a systematic approach to training code users who, upon completion of the training, should be able to perform calculations making the best possible use of the capabilities of best estimate codes. In other words, the program aims at contributing towards solving the problem of user effect. The 3D S.UN.COP (Scaling, Uncertainty and 3D COuPled code calculations) seminars have been organized as follow-up of the proposal to IAEA for the Permanent Training Course for System Code Users. Six seminars have been held at University of Pisa (2003, 2004), at The Pennsylvania State University (2004), at University of Zagreb (2005), at the School of Industrial Engineering of Barcelona (January-February 2006) and in Buenos Aires, Argentina (October 2006), being this last one requested by ARN (Autoridad Regulatoria Nuclear), NA-SA (Nucleoelectrica Argentina S.A) and CNEA (Comision Nacional de Energia Atomica). It was recognized that such courses represented both a source of continuing education for current code users and a mean for current code users to enter the formal training structure of a proposed 'permanent' stepwise approach to user training. The 3D S.UN.COP 2006 in Barcelona was successfully held with the attendance of 33

  5. International Training Program in Support of Safety Analysis: 3D S.UN.COP-Scaling, Uncertainty and 3D Thermal-Hydraulics/Neutron-Kinetics Coupled Codes Seminars

    International Nuclear Information System (INIS)

    Petruzzi, Alessandro; D'Auria, Francesco; Bajs, Tomislav; Reventos, Francesc

    2006-01-01

    Thermal-hydraulic system computer codes are extensively used worldwide for analysis of nuclear facilities by utilities, regulatory bodies, nuclear power plant designers and vendors, nuclear fuel companies, research organizations, consulting companies, and technical support organizations. The computer code user represents a source of uncertainty that can influence the results of system code calculations. This influence is commonly known as the 'user effect' and stems from the limitations embedded in the codes as well as from the limited capability of the analysts to use the codes. Code user training and qualification is an effective means for reducing the variation of results caused by the application of the codes by different users. This paper describes a systematic approach to training code users who, upon completion of the training, should be able to perform calculations making the best possible use of the capabilities of best estimate codes. In other words, the program aims at contributing towards solving the problem of user effect. The 3D S.UN.COP (Scaling, Uncertainty and 3D COuPled code calculations) seminars have been organized as follow-up of the proposal to IAEA for the Permanent Training Course for System Code Users [1]. Five seminars have been held at University of Pisa (2003, 2004), at The Pennsylvania State University (2004), at University of Zagreb (2005) and at the School of Industrial Engineering of Barcelona (2006). It was recognized that such courses represented both a source of continuing education for current code users and a mean for current code users to enter the formal training structure of a proposed 'permanent' stepwise approach to user training. The 3D S.UN.COP 2006 was successfully held with the attendance of 33 participants coming from 18 countries and 28 different institutions (universities, vendors, national laboratories and regulatory bodies). More than 30 scientists (coming from 13 countries and 23 different institutions) were

  6. Monitoring hydraulic fractures: state estimation using an extended Kalman filter

    International Nuclear Information System (INIS)

    Rochinha, Fernando Alves; Peirce, Anthony

    2010-01-01

    There is considerable interest in using remote elastostatic deformations to identify the evolving geometry of underground fractures that are forced to propagate by the injection of high pressure viscous fluids. These so-called hydraulic fractures are used to increase the permeability in oil and gas reservoirs as well as to pre-fracture ore-bodies for enhanced mineral extraction. The undesirable intrusion of these hydraulic fractures into environmentally sensitive areas or into regions in mines which might pose safety hazards has stimulated the search for techniques to enable the evolving hydraulic fracture geometries to be monitored. Previous approaches to this problem have involved the inversion of the elastostatic data at isolated time steps in the time series provided by tiltmeter measurements of the displacement gradient field at selected points in the elastic medium. At each time step, parameters in simple static models of the fracture (e.g. a single displacement discontinuity) are identified. The approach adopted in this paper is not to regard the sequence of sampled elastostatic data as independent, but rather to treat the data as linked by the coupled elastic-lubrication equations that govern the propagation of the evolving hydraulic fracture. We combine the Extended Kalman Filter (EKF) with features of a recently developed implicit numerical scheme to solve the coupled free boundary problem in order to form a novel algorithm to identify the evolving fracture geometry. Numerical experiments demonstrate that, despite excluding significant physical processes in the forward numerical model, the EKF-numerical algorithm is able to compensate for the un-modeled dynamics by using the information fed back from tiltmeter data. Indeed the proposed algorithm is able to provide reasonably faithful estimates of the fracture geometry, which are shown to converge to the actual hydraulic fracture geometry as the number of tiltmeters is increased. Since the location of

  7. Virginia Power thermal-hydraulics methods

    International Nuclear Information System (INIS)

    Anderson, R.C.; Basehore, K.L.; Harrell, J.R.

    1987-01-01

    Virginia Power's nuclear safety analysis group is responsible for the safety analysis of reload cores for the Surry and North Anna power stations, including the area of core thermal-hydraulics. Postulated accidents are evaluated for potential departure from nucleate boiling violations. In support of these tasks, Virginia Power has employed the COBRA code and the W-3 and WRB-1 DNB correlations. A statistical DNBR methodology has also been developed. The code, correlations and statistical methodology are discussed

  8. A study on the effect of fluidic device installed in a safety injection tank on thermal-hydraulic phenomena of large break loss of coolant accident

    International Nuclear Information System (INIS)

    Chung, Young Jong; Bae, Kyoo Hwan; Song, Jin Ho; Sim, Suk Ku; Park, Jong Kyun

    1999-03-01

    The performance of the Safety Injection Tank (SIT) with fluidic device (advanced SIT) is analyzed for the large break loss of coolant accident (LBLOCA) using RELAP5/MOD3.1-KREM. First the case is analyzed using the conventional SIT. Among various cases the case with 4-split downcomer, discharge coefficient Cd=0.6, MCP trip with reactor trip and break location of cold leg discharge side with the pressurizer is found to be the most limiting case. For the same condition, the advanced SIT results the similar PCT, however it can maintain adequately the liquid level in the downcomer. By changing the ECCS location from the current injection to the cold leg elevations, PCT is improved by 75 K. (Author). 6 refs., 4 tabs., 54 figs

  9. Risk-based Comparative Study of Fluid Power Pitch Concepts

    DEFF Research Database (Denmark)

    Liniger, Jesper; Pedersen, Henrik Clemmensen; N. Soltani, Mohsen

    2017-01-01

    Proper functioning of the pitch system is essential to both normal operation and safety critical shut down of modern multi megawatt wind turbines. Several studies on field failure rates for such turbines show that pitch systems are a major contributor to failures which entails an increased risk....... Thus, more reliable and safe concepts are needed. A review of patents and patent applications covering fluid power pitch concepts, reveals that many propose closed-type hydraulic systems. This paper proposes a closed-type concept with a bootstrap reservoir. In contrary to a conventional system where...

  10. Small scale thermal-hydraulic experiment for stable operation of a pius-type reactor

    International Nuclear Information System (INIS)

    Tasaka, K.; Tamaki, M.; Imai, S.; Irianto, I.D.; Tsuji, Y.; Kukita, Y.

    1994-01-01

    Thermal-hydraulic experiments using a small-scale atmospheric pressure test loop have been performed for the Process Inherent Ultimate Safety (PIUS)-type reactor to develop the new pump speed feedback control system. Three feedback control systems based on the measurement of flow rate, differential pressure, and fluid temperature distribution in the lower density lock have been proposed and confirmed by a series of experiments. Each of the feedback control systems had been verified in the simulation experiment such as a start-up simulation test. The automatic pump speed control based on the fluid temperature at the lower density lock was quite effective to maintain the stratified interface between primary water and borated pool water for stable operation of the reactor. (author)

  11. Thermal Fluid Engineering

    International Nuclear Information System (INIS)

    Jang, Byeong Ju

    1984-01-01

    This book is made up of 5 chapters. They are fluid mechanics, fluid machines, Industrial thermodynamics, steam boiler and steam turbine. It introduces hydrostatics, basic theory of fluid movement and law of momentum. It also deals with centrifugal pump, axial flow pump, general hydraulic turbine, and all phenomena happening in the pump. It covers the law of thermodynamics, perfect gas, properties of steam, and flow of gas and steam and water tube boiler. Lastly it explains basic format, theory, loss and performance as well as principle part of steam turbine.

  12. Digital switched hydraulics

    Science.gov (United States)

    Pan, Min; Plummer, Andrew

    2018-06-01

    This paper reviews recent developments in digital switched hydraulics particularly the switched inertance hydraulic systems (SIHSs). The performance of SIHSs is presented in brief with a discussion of several possible configurations and control strategies. The soft switching technology and high-speed switching valve design techniques are discussed. Challenges and recommendations are given based on the current research achievements.

  13. Hydraulic Structures : Caissons

    NARCIS (Netherlands)

    Voorendt, M.Z.; Molenaar, W.F.; Bezuyen, K.G.

    These lecture notes on caissons are part of the study material belonging to the course 'Hydraulic Structures 1' (code CTB3355), part of the Bachelor of Science education and the Hydraulic Engineering track of the Master of Science education for civil engineering students at Delft University of

  14. Neutronic, thermal-hydraulics and safety calculations of a Miniplate Irradiation Device (MID) of dispersion type fuel elements; Calculos neutronicos, termo-hidraulicos e de seguranca de um dispositivo para Irradiacao de miniplacas (DIM) de elementos combustiveis tipo dispersao

    Energy Technology Data Exchange (ETDEWEB)

    Domingos, Douglas Borges

    2010-07-01

    Neutronic, thermal-hydraulics and accident analysis calculations were developed to estimate the safety of a Miniplate Irradiation Device (MID) to be placed in the IEA-R1 reactor core. The irradiation device is used to receive miniplates of U{sub 3}O{sub 8}-Al and U{sub 3}Si{sub 2}- Al dispersion fuels, LEU type (19.75 % {sup 235}U) with uranium densities of, respectively, 3.2 gU/cm{sup 3} and 4.8 gU/cm{sup 3}. The fuel miniplates will be irradiated to nominal {sup 235}U burnup levels of 50% and 80%, in order to qualify the above high-density dispersion fuels to be used in the Brazilian Multipurpose Reactor (RMB), now in the conception phase. For the neutronic calculation, the computer codes CITATION and 2DB were utilized. The computer code FLOW was used to calculate the coolant flow rate in the irradiation device, allowing the determination of the fuel miniplate temperatures with the computer model MTRCR-IEA-R1. A postulated Loss of Coolant Accident (LOCA) was analyzed with the computer codes LOSS and TEMPLOCA, allowing the calculation of the fuel miniplate temperatures after the reactor pool draining. The calculations showed that the irradiation should occur without adverse consequences in the IEA-R1 reactor. (author)

  15. Thermal hydraulic model validation for HOR mixed core fuel management

    International Nuclear Information System (INIS)

    Gibcus, H.P.M.; Vries, J.W. de; Leege, P.F.A. de

    1997-01-01

    A thermal-hydraulic core management model has been developed for the Hoger Onderwijsreactor (HOR), a 2 MW pool-type university research reactor. The model was adopted for safety analysis purposes in the framework of HEU/LEU core conversion studies. It is applied in the thermal-hydraulic computer code SHORT (Steady-state HOR Thermal-hydraulics) which is presently in use in designing core configurations and for in-core fuel management. An elaborate measurement program was performed for establishing the core hydraulic characteristics for a variety of conditions. The hydraulic data were obtained with a dummy fuel element with special equipment allowing a.o. direct measurement of the true core flow rate. Using these data the thermal-hydraulic model was validated experimentally. The model, experimental tests, and model validation are discussed. (author)

  16. FRACTURING FLUID CHARACTERIZATION FACILITY

    Energy Technology Data Exchange (ETDEWEB)

    Subhash Shah

    2000-08-01

    Hydraulic fracturing technology has been successfully applied for well stimulation of low and high permeability reservoirs for numerous years. Treatment optimization and improved economics have always been the key to the success and it is more so when the reservoirs under consideration are marginal. Fluids are widely used for the stimulation of wells. The Fracturing Fluid Characterization Facility (FFCF) has been established to provide the accurate prediction of the behavior of complex fracturing fluids under downhole conditions. The primary focus of the facility is to provide valuable insight into the various mechanisms that govern the flow of fracturing fluids and slurries through hydraulically created fractures. During the time between September 30, 1992, and March 31, 2000, the research efforts were devoted to the areas of fluid rheology, proppant transport, proppant flowback, dynamic fluid loss, perforation pressure losses, and frictional pressure losses. In this regard, a unique above-the-ground fracture simulator was designed and constructed at the FFCF, labeled ''The High Pressure Simulator'' (HPS). The FFCF is now available to industry for characterizing and understanding the behavior of complex fluid systems. To better reflect and encompass the broad spectrum of the petroleum industry, the FFCF now operates under a new name of ''The Well Construction Technology Center'' (WCTC). This report documents the summary of the activities performed during 1992-2000 at the FFCF.

  17. Thermal-Hydraulic Experiments and Modelling for Advanced Nuclear Reactor Systems

    International Nuclear Information System (INIS)

    Song, C. H.; Baek, W. P.; Chung, M. K.

    2007-06-01

    The objectives of the project are to study thermal hydraulic characteristics of advanced nuclear reactor system for evaluating key thermal-hydraulic phenomena relevant to new safety concepts. To meet the research goal, several thermal hydraulic experiments were performed and related thermal hydraulic models were developed with the experimental data which were produced through the thermal hydraulic experiments. The Followings are main research topics: - Multi-dimensional Phenomena in a Reactor Vessel Downcomer - Condensation-induced Thermal Mixing in a Pool - Development of Thermal-Hydraulic Models for Two-Phase Flow - Construction of T-H Data Base

  18. Trends in Design of Water Hydraulics

    DEFF Research Database (Denmark)

    Conrad, Finn

    2005-01-01

    ordinary tap water and the range of application areas are illustrated with examples, in particular within the food processing industry, humidification operations, water mist systems for fire fighting, high water pressure cleaners, water moisturising systems for wood processing, lumber drying process...... operate with pure water from the tap without additives of any kind. Hence water hydraulics takes the benefit of pure water as fluid being environmentally friendly, easy to clean sanitary design, non-toxic, non-flammable, inexpensive, readily available and easily disposable. The low-pressure tap water...... and accessories running with sea-water as fluid are available. A unique solution is to use reverse osmosis to generate drinking water from sea-water, and furthermore for several off-shore applications. Furthermore, tap water hydraulic components of the Nessie® family and examples of measured performance...

  19. Feasibility study for objective oriented design of system thermal hydraulic analysis program

    International Nuclear Information System (INIS)

    Chung, Bub Dong; Jeong, Jae Jun; Hwang, Moon Kyu

    2008-01-01

    The system safety analysis code, such as RELAP5, TRAC, CATHARE etc. have been developed based on Fortran language during the past few decades. Refactoring of conventional codes has been also performed to improve code readability and maintenance. However the programming paradigm in software technology has been changed to use objects oriented programming (OOP), which is based on several techniques, including encapsulation, modularity, polymorphism, and inheritance. In this work, objective oriented program for system safety analysis code has been tried utilizing modernized C language. The analysis, design, implementation and verification steps for OOP system code development are described with some implementation examples. The system code SYSTF based on three-fluid thermal hydraulic solver has been developed by OOP design. The verifications of feasibility are performed with simple fundamental problems and plant models. (author)

  20. Hydraulic fracture considerations in oil sand overburden dams

    Energy Technology Data Exchange (ETDEWEB)

    Cameron, R.; Madden, B.; Danku, M. [Syncrude Canada Ltd., Fort McMurray, AB (Canada)

    2008-07-01

    This paper discussed hydraulic fracture potential in the dry-filled temporary dams used in the oil sands industry. Hydraulic fractures can occur when reservoir fluid pressures are greater than the minimum stresses in a dam. Stress and strain conditions are influenced by pore pressures, levels of compaction in adjacent fills as well as by underlying pit floor and abutment conditions. Propagation pressure and crack initiation pressures must also be considered in order to provide improved hydraulic fracture protection to dams. Hydraulic fractures typically result in piping failures. Three cases of hydraulic fracture at oil sands operations in Alberta were presented. The study showed that hydraulic fracture failure modes must be considered in dam designs, particularly when thin compacted lift of dry fill are used to replace wetted clay cores. The risk of hydraulic fractures can be reduced by eliminating in situ bedrock irregularities and abutments. Overpressure heights, abutment sloping, and the sloping of fills above abutments, as well as the dam's width and base conditions must also be considered in relation to potential hydraulic fractures. It was concluded that upstream sand beaches and internal filters can help to prevent hydraulic fractures in dams in compacted control zones. 5 refs., 16 figs.

  1. Two phase flow arising in hydraulics

    Czech Academy of Sciences Publication Activity Database

    Straškraba, Ivan

    2015-01-01

    Roč. 60, č. 1 (2015), s. 21-33 ISSN 0862-7940 R&D Projects: GA ČR GA201/08/0012 Institutional support: RVO:67985840 Keywords : compressible fluid * Navier-Stokes equations * hydraulic systems Subject RIV: BA - General Mathematics Impact factor: 0.507, year: 2015 http://link.springer.com/article/10.1007/s10492-015-0083-9

  2. Hydraulic upright of a mine support

    Energy Technology Data Exchange (ETDEWEB)

    Solomakhin, A N; Il' in, V A; Ponomarenko, Yu F; Shakhmeyster, Yu L

    1979-04-30

    The hydraulic upright of a mine support, which includes a housing, piston with compacting element and dirt collector, rod and guide sleeve, is described. In order to improve protection of the piston element from abrasive particles and to reduce the pressure differential the piston of the upright is also equipped with a compaction ring, whose lateral surface has a groove beneath the compacting element. The surface on the side of the working fluid supply is made conical in order to remove dirt.

  3. Safety actuator of the Cabri reactor as a function of its power and cooling fluid flow rate

    International Nuclear Information System (INIS)

    Bertrand, Jean; Da Costa Vieira, David; Tattegrain, Alain

    1969-04-01

    This report present a device which is to provide a stop command to the Cabri reactor when the rate of its power to the cooling fluid rate reaches a value determined with respect to water temperature in the circuit. The stop command is delivered by an actuator which opens a relay contact when the power reaches a specific value. The authors present the device, its characteristics, and principle. They also present the different amplifier circuits, the input and output circuits (flow rate input, temperature input, and output circuit), the energy supply, and the various adjustments

  4. Scaling the viscous circular hydraulic jump

    Science.gov (United States)

    Argentina, Mederic; Cerda, Enrique; Duchesne, Alexis; Limat, Laurent

    2017-11-01

    The formation mechanism of hydraulic jumps has been proposed by Belanger in 1828 and rationalised by Lord Rayleigh in 1914. As the Froude number becomes higher than one, the flow super criticality induces an instability which yields the emergence of a steep structure at the fluid surface. Strongly deformed liquid-air interface can be observed as a jet of viscous fluid impinges a flat boundary at high enough velocity. In this experimental setup, the location of the jump depends on the viscosity of the liquid, as shown by T. Bohr et al. in 1997. In 2014, A. Duchesne et al. have established the constancy of the Froude number at jump. Hence, it remains a contradiction, in which the radial hydraulic jump location might be explained through inviscid theory, but is also viscosity dependent. We present a model based on the 2011 Rojas et al. PRL, which solves this paradox. The agreement with experimental measurements is excellent not only for the prediction of the position of the hydraulic jump, but also for the determination of the fluid thickness profile. We predict theoretically the critical value of the Froude number, which matches perfectly to that measured by Duchesne et al. We acknowledge the support of the CNRS and the Universit Cte d'Azur, through the IDEX funding.

  5. SAFETY

    CERN Multimedia

    M. Plagge, C. Schaefer and N. Dupont

    2013-01-01

    Fire Safety – Essential for a particle detector The CMS detector is a marvel of high technology, one of the most precise particle measurement devices we have built until now. Of course it has to be protected from external and internal incidents like the ones that can occur from fires. Due to the fire load, the permanent availability of oxygen and the presence of various ignition sources mostly based on electricity this has to be addressed. Starting from the beam pipe towards the magnet coil, the detector is protected by flooding it with pure gaseous nitrogen during operation. The outer shell of CMS, namely the yoke and the muon chambers are then covered by an emergency inertion system also based on nitrogen. To ensure maximum fire safety, all materials used comply with the CERN regulations IS 23 and IS 41 with only a few exceptions. Every piece of the 30-tonne polyethylene shielding is high-density material, borated, boxed within steel and coated with intumescent (a paint that creates a thick co...

  6. SAFETY

    CERN Multimedia

    C. Schaefer and N. Dupont

    2013-01-01

      “Safety is the highest priority”: this statement from CERN is endorsed by the CMS management. An interpretation of this statement may bring you to the conclusion that you should stop working in order to avoid risks. If the safety is the priority, work is not! This would be a misunderstanding and misinterpretation. One should understand that “working safely” or “operating safely” is the priority at CERN. CERN personnel are exposed to different hazards on many levels on a daily basis. However, risk analyses and assessments are done in order to limit the number and the gravity of accidents. For example, this process takes place each time you cross the road. The hazard is the moving vehicle, the stake is you and the risk might be the risk of collision between both. The same principle has to be applied during our daily work. In particular, keeping in mind the general principles of prevention defined in the late 1980s. These principles wer...

  7. Engineering applications of pneumatics and hydraulics

    CERN Document Server

    Turner, Ian C

    2014-01-01

    Assuming only the most basic knowledge of the physics of fluids, this book aims to equip the reader with a sound understanding of fluid power systems and their uses in practical engineering. In line with the strongly practical bias of the book, maintenance and trouble-shooting are covered, with particular emphasis on safety systems and regulations.

  8. Hydraulic jumps in a channel

    DEFF Research Database (Denmark)

    Bonn, D.; Andersen, Anders Peter; Bohr, Tomas

    2009-01-01

    We present a study of hydraulic jumps with flow predominantly in one direction, created either by confining the flow to a narrow channel with parallel walls or by providing an inflow in the form of a narrow sheet. In the channel flow, we find a linear height profile upstream of the jump as expected......'s mixing-length theory with a mixing length that is proportional to the height of the fluid layer. Using averaged boundary-layer equations, taking into account the friction with the channel walls and the eddy viscosity, the flow both upstream and downstream of the jump can be understood. For the downstream...... subcritical flow, we assume that the critical height is attained close to the channel outlet. We use mass and momentum conservation to determine the position of the jump and obtain an estimate which is in rough agreement with our experiment. We show that the averaging method with a varying velocity profile...

  9. Hydraulically amplified PZT mems actuator

    Science.gov (United States)

    Miles, Robin R.

    2004-11-02

    A hydraulically amplified microelectromechanical systems actuator. A piece of piezoelectric material or stacked piezo bimorph is bonded or deposited as a thin film. The piece is operatively connected to a primary membrane. A reservoir is operatively connected to the primary membrane. The reservoir contains a fluid. A membrane is operatively connected to the reservoir. In operation, energizing the piezoelectric material causing the piezoelectric material to bow. Bowing of the piezoelectric material causes movement of the primary membrane. Movement of the primary membrane results in a force in being transmitted to the liquid in the reservoir. The force in the liquid causes movement of the membrane. Movement of the membrane results in an operating actuator.

  10. Hydraulic Yaw System

    DEFF Research Database (Denmark)

    Stubkier, Søren; Pedersen, Henrik C.; Mørkholt, M.

    a hydraulic soft yaw system, which is able to reduce the loads on the wind turbine significantly. A full scale hydraulic yaw test rig is available for experiments and tests. The test rig is presented as well as the system schematics of the hydraulic yaw system....... the HAWC2 aeroelastic code and an extended model of the NREL 5MW turbine combined with a simplified linear model of the turbine, the parameters of the soft yaw system are optimized to reduce loading in critical components. Results shows that a significant reduction in fatigue and extreme loads to the yaw...... system and rotor shaft when utilizing the soft yaw drive concept compared to the original stiff yaw system. The physical demands of the hydraulic yaw system are furthermore examined for a life time of 20 years. Based on the extrapolated loads, the duty cycles show that it is possible to construct...

  11. An assessment of surface mud system design options for minimizing the health, safety, and environmental impact concerns associated with drilling fluids

    International Nuclear Information System (INIS)

    Minton, R.C.; Bailey, M.G.

    1991-01-01

    In this paper a drilling fluid surface system design concept is proposed that resolves the Environmental, occupational hygiene and safety issues associated with conventional designs. Automation of the chemical handling and dosing system is the central element of the concept which, when fully integrated into the system, permits a significant reduction in the surface volume requirements. This, in turn, results in weight and capital cost savings, offsetting the cost of the processing and treatment plant, and a smaller overall footprint for the system. Adoption of the design philosophy results in a safe, healthy working environment in which all of the waste streams are managed so as to minimize the overall environmental impact of the drilling process

  12. Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Sessions 17-24

    Energy Technology Data Exchange (ETDEWEB)

    Block, R.C.; Feiner, F. [American Nuclear Society, La Grange Park, IL (United States)

    1995-09-01

    Technical papers accepted for presentation at the Seventh International Topical Meeting on Nuclear Reactor Thermal-Hydraulics are included in the present Proceedings. Except for the invited papers in the plenary session, all other papers are contributed papers. The topics of the meeting encompass all major areas of nuclear thermal-hydraulics, including analytical and experimental works on the fundamental mechanisms of fluid flow and heat transfer, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Because of the complex nature of nuclear reactors and power plants, several papers deal with the combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. The participation in the conference by the authors from several countries and four continents makes the Proceedings a comprehensive review of the recent progress in the field of nuclear reactor thermal-hydraulics worldwide. Individual papers have been cataloged separately.

  13. Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Sessions 17-24

    International Nuclear Information System (INIS)

    Block, R.C.; Feiner, F.

    1995-09-01

    Technical papers accepted for presentation at the Seventh International Topical Meeting on Nuclear Reactor Thermal-Hydraulics are included in the present Proceedings. Except for the invited papers in the plenary session, all other papers are contributed papers. The topics of the meeting encompass all major areas of nuclear thermal-hydraulics, including analytical and experimental works on the fundamental mechanisms of fluid flow and heat transfer, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Because of the complex nature of nuclear reactors and power plants, several papers deal with the combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. The participation in the conference by the authors from several countries and four continents makes the Proceedings a comprehensive review of the recent progress in the field of nuclear reactor thermal-hydraulics worldwide. Individual papers have been cataloged separately

  14. Risky consumption habits and safety of fluid milk available in retail sales outlets in Viçosa, Minas Gerais State, Brazil.

    Science.gov (United States)

    Pieri, Fabio Alessandro; Colombo, Monique; Merhi, Carolina Milner; Juliati, Vinícius Augusto; Ferreira, Marcello Sebe; Nero, Marcelo Antônio; Nero, Luis Augusto

    2014-06-01

    This study aimed to assess raw milk consumption habits in the urban population of Viçosa, Minas Gerais, Brazil, and the microbiological safety and quality of the fluid milk available in retail sales outlets in the same region. A simplified questionnaire regarding raw milk consumption was applied to the persons responsible for food acquisition in 411 residences. The regular consumption of raw milk was observed by 18.5% of the interviewers, and lack of knowledge of possible risks related to this food product. Microbiological safety and quality were assessed for raw (n=69), pasteurized (n=80), and ultra-high-temperature (UHT)-treated milk (n=80) by analyzing the counts of mesophilic aerobes, coliforms, and Escherichia coli, and detection of Listeria monocytogenes and Salmonella spp.; raw milk samples were also subjected to enumeration of coagulase-positive Staphylococcus. Concerning raw milk, 59.4% of the samples were considered as produced in inadequate hygienic conditions, 5.8% of the samples presented counts of coagulase-positive Staphylococcus lower than 100 colony-forming units (CFU)/mL, and no samples presented with positive results for L. monocytogenes or Salmonella spp. All pasteurized and UHT milk samples presented with low counts of mesophilic aerobes and coliforms, while L. monocytogenes and Salmonella spp. were absent. The data demonstrated that raw milk was consumed by the population studied. Despite the absence of potential hazards, raw milk was of poor hygienic quality, in contrast with the processed fluid milk available in retail sales outlets that was safe and of good hygienic quality, highlighting the suitability of pasteurized and UHT milk for human consumption.

  15. Coordination of stem and leaf hydraulic conductance in southern California shrubs: a test of the hydraulic segmentation hypothesis.

    Science.gov (United States)

    Pivovaroff, Alexandria L; Sack, Lawren; Santiago, Louis S

    2014-08-01

    Coordination of water movement among plant organs is important for understanding plant water use strategies. The hydraulic segmentation hypothesis (HSH) proposes that hydraulic conductance in shorter lived, 'expendable' organs such as leaves and longer lived, more 'expensive' organs such as stems may be decoupled, with resistance in leaves acting as a bottleneck or 'safety valve'. We tested the HSH in woody species from a Mediterranean-type ecosystem by measuring leaf hydraulic conductance (Kleaf) and stem hydraulic conductivity (KS). We also investigated whether leaves function as safety valves by relating Kleaf and the hydraulic safety margin (stem water potential minus the water potential at which 50% of conductivity is lost (Ψstem-Ψ50)). We also examined related plant traits including the operating range of water potentials, wood density, leaf mass per area, and leaf area to sapwood area ratio to provide insight into whole-plant water use strategies. For hydrated shoots, Kleaf was negatively correlated with KS , supporting the HSH. Additionally, Kleaf was positively correlated with the hydraulic safety margin and negatively correlated with the leaf area to sapwood area ratio. Consistent with the HSH, our data indicate that leaves may act as control valves for species with high KS , or a low safety margin. This critical role of leaves appears to contribute importantly to plant ecological specialization in a drought-prone environment. © 2014 The Authors. New Phytologist © 2014 New Phytologist Trust.

  16. Safety

    International Nuclear Information System (INIS)

    Jones, P.M.S.

    1987-01-01

    Aspects of fission reactors are considered - control, heat removal and containment. Brief descriptions of the reactor accidents at the SL-1 reactor (1961), Windscale (1957), Browns Ferry (1975), Three Mile Island (1979) and Chernobyl (1986) are given. The idea of inherently safe reactor designs is discussed. Safety assessment is considered under the headings of preliminary hazard analysis, failure mode analysis, event trees, fault trees, common mode failure and probabalistic risk assessments. These latter can result in a series of risk distributions linked to specific groups of fault sequences and specific consequences. A frequency-consequence diagram is shown. Fatal accident incidence rates in different countries including the United Kingdom for various industries are quoted. The incidence of fatal cancers from occupational exposure to chemicals is tabulated. Human factors and the acceptability of risk are considered. (U.K.)

  17. A New Method Solving Contact/Detach Problem in Fluid and Structure Interaction Simulation with Application in Modeling of a Safety Valve

    Directory of Open Access Journals (Sweden)

    Zheng Guo

    2010-01-01

    Full Text Available A new virtual baffle methodology is implemented to solve contact/detach problem which is often encountered in fluid and structure interaction simulations while using dynamic grids technique. The algorithm is based on tetrahedral unstructured grid, and a zero thickness baffle face is generated between actually contacted two objects. In computation process, this baffle face is divided into two parts representing convective and blocked area, respectively; the area of each part is calculated according to the actual displacement between the two objects. Convective part in a baffle face is treated as inner interface between cells, and on blocked part wall boundary condition is applied; so convective and blocking effect can be achieved on a single baffle face. This methodology can simulate real detaching process starting from contact, that is, zero displacement, while it has no restriction to minimum grid cell size. The methodology is then applied in modeling of a complicated safety valve opening process, involving multidisciplinary fluid and structure interaction and dynamic grids. The results agree well with experimental data, which proves that the virtual baffle method is successful.

  18. The hydraulic wheel

    International Nuclear Information System (INIS)

    Alvarez Cardona, A.

    1985-01-01

    The present article this dedicated to recover a technology that key in disuse for the appearance of other techniques. It is the hydraulic wheel with their multiple possibilities to use their energy mechanical rotational in direct form or to generate electricity directly in the fields in the place and to avoid the high cost of transport and transformation. The basic theory is described that consists in: the power of the currents of water and the hydraulic receivers. The power of the currents is determined knowing the flow and east knowing the section of the flow and its speed; they are given you formulate to know these and direct mensuration methods by means of floodgates, drains and jumps of water. The hydraulic receivers or properly this hydraulic wheels that are the machines in those that the water acts like main force and they are designed to transmit the biggest proportion possible of absolute work of the water, the hydraulic wheels of horizontal axis are the common and they are divided in: you rotate with water for under, you rotate with side water and wheels with water for above. It is analyzed each one of them, their components are described; the conditions that should complete to produce a certain power and formulate them to calculate it. There are 25 descriptive figures of the different hydraulic wheels

  19. Verification Test of Hydraulic Performance for Reactor Coolant Pump

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Jun; Kim, Jae Shin; Ryu, In Wan; Ko, Bok Seong; Song, Keun Myung [Samjin Ind. Co., Seoul (Korea, Republic of)

    2010-01-15

    According to this project, basic design for prototype pump and model pump of reactor coolant pump and test facilities has been completed. Basic design for prototype pump to establish structure, dimension and hydraulic performance has been completed and through primary flow analysis by computational fluid dynamics(CFD), flow characteristics and hydraulic performance have been established. This pump was designed with mixed flow pump having the following design requirements; specific velocity(Ns); 1080.9(rpm{center_dot}m{sup 3}/m{center_dot}m), capacity; 3115m{sup 3}/h, total head ; 26.3m, pump speed; 1710rpm, pump efficiency; 77.0%, Impeller out-diameter; 349mm, motor output; 360kw, design pressure; 17MPaG. The features of the pump are leakage free due to no mechanical seal on the pump shaft which insures reactor's safety and law noise level and low vibration due to no cooling fan on the motor which makes eco-friendly product. Model pump size was reduced to 44% of prototype pump for the verification test for hydraulic performance of reactor coolant pump and was designed with mixed flow pump and canned motor having the following design requirements; specific speed(NS); 1060.9(rpm{center_dot}m{sup 3}/m{center_dot}m), capacity; 539.4m{sup 3}/h, total head; 21.0m, pump speed; 3476rpm, pump efficiency; 72.9%, Impeller out-diameter; 154mm, motor output; 55kw, design pressure; 1.0MPaG. The test facilities were designed for verification test of hydraulic performance suitable for pump performance test, homologous test, NPSH test(cavitation), cost down test and pressure pulsation test of inlet and outlet ports. Test tank was designed with testing capacity enabling up to 2000m{sup 3}/h and design pressure 1.0MPaG. Auxiliary pump was designed with centrifugal pump having capacity; 1100m{sup 3}/h, total head; 42.0m, motor output; 190kw

  20. Status of SPACE Safety Analysis Code Development

    International Nuclear Information System (INIS)

    Lee, Dong Hyuk; Yang, Chang Keun; Kim, Se Yun; Ha, Sang Jun

    2009-01-01

    In 2006, the Korean the Korean nuclear industry started developing a thermal-hydraulic analysis code for safety analysis of PWR(Pressurized Water Reactor). The new code is named as SPACE(Safety and Performance Analysis Code for Nuclear Power Plant). The SPACE code can solve two-fluid, three-field governing equations in one dimensional or three dimensional geometry. The SPACE code has many component models required for modeling a PWR, such as reactor coolant pump, safety injection tank, etc. The programming language used in the new code is C++, for new generation of engineers who are more comfortable with C/C++ than old FORTRAN language. This paper describes general characteristics of SPACE code and current status of SPACE code development

  1. An XFEM Model for Hydraulic Fracturing in Partially Saturated Rocks

    Directory of Open Access Journals (Sweden)

    Salimzadeh Saeed

    2016-01-01

    Full Text Available Hydraulic fracturing is a complex multi-physics phenomenon. Numerous analytical and numerical models of hydraulic fracturing processes have been proposed. Analytical solutions commonly are able to model the growth of a single hydraulic fracture into an initially intact, homogeneous rock mass. Numerical models are able to analyse complex problems such as multiple hydraulic fractures and fracturing in heterogeneous media. However, majority of available models are restricted to single-phase flow through fracture and permeable porous rock. This is not compatible with actual field conditions where the injected fluid does not have similar properties as the host fluid. In this study we present a fully coupled hydro-poroelastic model which incorporates two fluids i.e. fracturing fluid and host fluid. Flow through fracture is defined based on lubrication assumption, while flow through matrix is defined as Darcy flow. The fracture discontinuity in the mechanical model is captured using eXtended Finite Element Method (XFEM while the fracture propagation criterion is defined through cohesive fracture model. The discontinuous matrix fluid velocity across fracture is modelled using leak-off loading which couples fracture flow and matrix flow. The proposed model has been discretised using standard Galerkin method, implemented in Matlab and verified against several published solutions. Multiple hydraulic fracturing simulations are performed to show the model robustness and to illustrate how problem parameters such as injection rate and rock permeability affect the hydraulic fracturing variables i.e. injection pressure, fracture aperture and fracture length. The results show the impact of partial saturation on leak-off and the fact that single-phase models may underestimate the leak-off.

  2. Critical discharge of fluids and gases

    International Nuclear Information System (INIS)

    Seewald, Michael

    2012-01-01

    The thermal hydraulic relations during discharge of fluids and gases are complex and a closed solution does not seem to be available. For the modeling of leakage accidents in nuclear power plants basic considerations are suitable for statements on the maximum mass flow, and thus the leak rate. The maximum mass flow is reached when the critical velocity is reached in the smallest cross section. This allows the appropriate design of safety systems for one-phase and two-phase flows. For German NPP simulators the hydrodynamics simulation program RELAP5-3D is used. The simulator center operates a 1:10 scale gas model of a two-loop PWR type reactor. The observable phenomena have occurred in nuclear power plants. The characteristics for a visualization of two-phase flows are not available in the simulation software and have to be added by correlations with experimental results. The realization of expectations on digital visualization techniques is discussed.

  3. 2-D CFD time-dependent thermal-hydraulic simulations of CANDU-6 moderator flows

    Energy Technology Data Exchange (ETDEWEB)

    Mehdi Zadeh, Foad [Department of Engineering Physics/Polytechnique Montréal, Montréal, QC (Canada); Étienne, Stéphane [Department of Mechanical Engineering/Polytechnique Montréal, Montréal, QC (Canada); Teyssedou, Alberto, E-mail: alberto.teyssedou@polymtl.ca [Department of Engineering Physics/Polytechnique Montréal, Montréal, QC (Canada)

    2016-12-01

    Highlights: • 2-D time-dependent CFD simulations of CANDU-6 moderator flows are presented. • A thermal-hydraulic code using thermal physical fluid properties is used. • The numerical approach and convergence is validated against available data. • Flow configurations are correlated using Richardson’s number. • Frequency components indicate moderator flow oscillations vs. Richardson numbers. - Abstract: The distribution of the fluid temperature and mass density of the moderator flow in CANDU-6 nuclear power reactors may affect the reactivity coefficient. For this reason, any possible moderator flow configuration and consequently the corresponding temperature distributions must be studied. In particular, the variations of the reactivity may result in major safety issues. For instance, excessive temperature excursions in the vicinity of the calandria tubes nearby local flow stagnation zones, may bring about partial boiling. Moreover, steady-state simulations have shown that for operating condition, intense buoyancy forces may be dominant, which can trigger a thermal stratification. Therefore, the numerical study of the time-dependent flow transition to such a condition, is of fundamental safety concern. Within this framework, this paper presents detailed time-dependent numerical simulations of CANDU-6 moderator flow for a wide range of flow conditions. To get a better insight of the thermal-hydraulic phenomena, the simulations were performed by covering long physical-time periods using an open-source code (Code-Saturne V3) developed by Électricité de France. The results show not only a region where the flow is characterized by coherent structures of flow fluctuations but also the existence of two limit cases where fluid oscillations disappear almost completely.

  4. Thermal-hydraulic calculation and water hammer analysis on CEFR loop system

    International Nuclear Information System (INIS)

    Hao Pengfei; Zhang Xiwen; Cai Weidong; Wang Xuefang

    1997-01-01

    China Experimental Fast Reactor (CEFR) is one of the '863' High-technical Projects. It is necessary to study the hydraulic and thermal Characteristic of CEFR loop system in order to guarantee the safety of operation. The results of the thermal-hydraulic calculation have been given. The main points are as follows: 1. The simplified model is built according to the loop system of CEFR, and the calculation method which is called 'NODE'-'BRANCH' is applied. This method includes two aspects, one is the theoretical analysis that is based on fluid mechanics and heat transfer theory. The other is the engineering calculation. These two aspects are connected in the computation. On the basis of the work mentioned above, the stable state computation is presented. In order to prevent serious damage caused by power failure accident, the courses of surplus reactor heat removing through two different systems have been simulated in the computation. 2. By using the fluid dynamics theory, the simplified model and the equipment boundary conditions of loop system are given. The water hammer computation is processed during the valve closing and pump stopping accidents. Some pictures of water hammer wave are presented, and the most dangerous state in the accident is also given

  5. Directly acting spring loaded safety valves as shock reducing measure; Direkt wirkende, federbelastete Sicherheitsventile als Druckstossreduzierende Massnahme

    Energy Technology Data Exchange (ETDEWEB)

    Ismaier, A.; Schluecker, E. [Erlangen-Nuernberg Univ. (DE). Lehrstuhl fuer Prozessmaschinen und Anlagentechnik (IPAT)

    2010-05-15

    Hydraulic shocks as induced by fast closure of armatures or by sudden pump failures are massive impacts in piping systems and require extensive measures to absorb the generated load. Basically the avoidance of water hammers are preferable but in case of emergency shutdowns unavoidable hydraulic shocks have to be reduced by appropriate measures. The authors describe experiments with spring loaded safety valves as shock reducing measures. It was shown that the vale dimensions is essential for the efficacy. A realistic modeling is possible using the one-dimensional fluid mechanics code ROLAST.

  6. Fluid circulation control device

    International Nuclear Information System (INIS)

    Benard, Henri; Henocque, Jean.

    1982-01-01

    Horizontal fluid circulation control device, of the type having a pivoting flap. This device is intended for being fitted in the pipes of hydraulic installation, particularly in a bleed and venting system of a nuclear power station shifting radioactive or contaminated liquids. The characteristic of this device is the cut-out at the top of the flap to allow the air contained in the pipes to flow freely [fr

  7. Thermal-Hydraulic Experiments and Modelling for Advanced Nuclear Reactor Systems

    International Nuclear Information System (INIS)

    Song, C. H.; Chung, M. K.; Park, C. K. and others

    2005-04-01

    The objectives of the project are to study thermal hydraulic characteristics of reactor primary system for the verification of the reactor safety and to evaluate new safety concepts of new safety design features. To meet the research goal, several thermal hydraulic experiments were performed and related thermal hydraulic models were developed with the experimental data which were produced through the thermal hydraulic experiments. Followings are main research topics; - Multi-dimensional Phenomena in a Reactor Vessel Downcomer - Condensation Load and Thermal Mixing in the IRWST - Development of Thermal-Hydraulic Models for Two-Phase Flow - Development of Measurement Techniques for Two-Phase Flow - Supercritical Reactor T/H Characteristics Analysis From the above experimental and analytical studies, new safety design features of the advanced power reactors were verified and lots of the safety issues were also resolved

  8. Thermal-Hydraulic Experiments and Modelling for Advanced Nuclear Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Song, C. H.; Chung, M. K.; Park, C. K. and others

    2005-04-15

    The objectives of the project are to study thermal hydraulic characteristics of reactor primary system for the verification of the reactor safety and to evaluate new safety concepts of new safety design features. To meet the research goal, several thermal hydraulic experiments were performed and related thermal hydraulic models were developed with the experimental data which were produced through the thermal hydraulic experiments. Followings are main research topics; - Multi-dimensional Phenomena in a Reactor Vessel Downcomer - Condensation Load and Thermal Mixing in the IRWST - Development of Thermal-Hydraulic Models for Two-Phase Flow - Development of Measurement Techniques for Two-Phase Flow - Supercritical Reactor T/H Characteristics Analysis From the above experimental and analytical studies, new safety design features of the advanced power reactors were verified and lots of the safety issues were also resolved.

  9. Factory Acceptance Test Procedure Westinghouse 100 ton Hydraulic Trailer

    International Nuclear Information System (INIS)

    Aftanas, B.L.

    1994-01-01

    This Factory Acceptance Test Procedure (FAT) is for the Westinghouse 100 Ton Hydraulic Trailer. The trailer will be used for the removal of the 101-SY pump. This procedure includes: safety check and safety procedures; pre-operation check out; startup; leveling trailer; functional/proofload test; proofload testing; and rolling load test

  10. Multiphase flow models for hydraulic fracturing technology

    Science.gov (United States)

    Osiptsov, Andrei A.

    2017-10-01

    The technology of hydraulic fracturing of a hydrocarbon-bearing formation is based on pumping a fluid with particles into a well to create fractures in porous medium. After the end of pumping, the fractures filled with closely packed proppant particles create highly conductive channels for hydrocarbon flow from far-field reservoir to the well to surface. The design of the hydraulic fracturing treatment is carried out with a simulator. Those simulators are based on mathematical models, which need to be accurate and close to physical reality. The entire process of fracture placement and flowback/cleanup can be conventionally split into the following four stages: (i) quasi-steady state effectively single-phase suspension flow down the wellbore, (ii) particle transport in an open vertical fracture, (iii) displacement of fracturing fluid by hydrocarbons from the closed fracture filled with a random close pack of proppant particles, and, finally, (iv) highly transient gas-liquid flow in a well during cleanup. The stage (i) is relatively well described by the existing hydralics models, while the models for the other three stages of the process need revisiting and considerable improvement, which was the focus of the author’s research presented in this review paper. For stage (ii), we consider the derivation of a multi-fluid model for suspension flow in a narrow vertical hydraulic fracture at moderate Re on the scale of fracture height and length and also the migration of particles across the flow on the scale of fracture width. At the stage of fracture cleanaup (iii), a novel multi-continua model for suspension filtration is developed. To provide closure relationships for permeability of proppant packings to be used in this model, a 3D direct numerical simulation of single phase flow is carried out using the lattice-Boltzmann method. For wellbore cleanup (iv), we present a combined 1D model for highly-transient gas-liquid flow based on the combination of multi-fluid and

  11. A 6-DOF vibration isolation system for hydraulic hybrid vehicles

    Science.gov (United States)

    Nguyen, The; Elahinia, Mohammad; Olson, Walter W.; Fontaine, Paul

    2006-03-01

    This paper presents the results of vibration isolation analysis for the pump/motor component of hydraulic hybrid vehicles (HHVs). The HHVs are designed to combine gasoline/diesel engine and hydraulic power in order to improve the fuel efficiency and reduce the pollution. Electric hybrid technology is being applied to passenger cars with small and medium engines to improve the fuel economy. However, for heavy duty vehicles such as large SUVs, trucks, and buses, which require more power, the hydraulic hybridization is a more efficient choice. In function, the hydraulic hybrid subsystem improves the fuel efficiency of the vehicle by recovering some of the energy that is otherwise wasted in friction brakes. Since the operation of the main component of HHVs involves with rotating parts and moving fluid, noise and vibration are an issue that affects both passengers (ride comfort) as well as surrounding people (drive-by noise). This study looks into the possibility of reducing the transmitted noise and vibration from the hydraulic subsystem to the vehicle's chassis by using magnetorheological (MR) fluid mounts. To this end, the hydraulic subsystem is modeled as a six degree of freedom (6-DOF) rigid body. A 6-DOF isolation system, consisting of five mounts connected to the pump/motor at five different locations, is modeled and simulated. The mounts are designed by combining regular elastomer components with MR fluids. In the simulation, the real loading and working conditions of the hydraulic subsystem are considered and the effects of both shock and vibration are analyzed. The transmissibility of the isolation system is monitored in a wide range of frequencies. The geometry of the isolation system is considered in order to sustain the weight of the hydraulic system without affecting the design of the chassis and the effectiveness of the vibration isolating ability. The simulation results shows reduction in the transmitted vibration force for different working cycles of

  12. HANARO thermal hydraulic accident analysis

    Energy Technology Data Exchange (ETDEWEB)

    Park, Chul; Kim, Heon Il; Lee, Bo Yook; Lee, Sang Yong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-06-01

    For the safety assessment of HANARO, accident analyses for the anticipated operational transients, accident scenarios and limiting accident scenarios were conducted. To do this, the commercial nuclear reactor system code. RELAP5/MOD2 was modified to RELAP5/KMRR; the thermal hydraulic correlations and the heat exchanger model was changed to incorporate HANARO characteristics. This report summarizes the RELAP/KMRR calculation results and the subchannel analyses results based on the RELAP/KMRR results. During the calculation, major concern was placed on the integrity of the fuel. For all the scenarios, the important accident analysis parameters, i.e., fuel centerline temperatures and the minimum critical heat flux ratio(MCHFR), satisfied safe design limits. It was verified, therefore, that the HANARO was safely designed. 21 tabs., 89 figs., 39 refs. (Author) .new.

  13. Hydraulic Fracturing and the Environment

    Science.gov (United States)

    Ayatollahy Tafti, T.; Aminzadeh, F.; Jafarpour, B.; de Barros, F.

    2013-12-01

    In this presentation, we highlight two key environmental concerns of hydraulic fracturing (HF), namely induced seismicity and groundwater contamination (GC). We examine the induced seismicity (IS) associated with different subsurface fluid injection and production (SFIP) operations and the key operational parameters of SFIP impacting it. In addition we review the key potential sources for possible water contamination. Both in the case of IS and GC we propose modeling and data analysis methods to quantify the risk factors to be used for monitoring and risk reduction. SFIP include presents a risk in hydraulic fracturing, waste water injection, enhanced oil recovery as well as geothermal energy operations. Although a recent report (NRC 2012) documents that HF is not responsible for most of the induced seismicities, we primarily focus on HF here. We look into vaious operational parameters such as volume and rate of water injection, the direction of the well versus the natural fracture network, the depth of the target and the local stress field and fault system, as well as other geological features. The latter would determine the potential for triggering tectonic related events by small induced seismicity events. We provide the building blocks for IS risk assessment and monitoring. The system we propose will involve adequate layers of complexity based on mapped seismic attributes as well as results from ANN and probabilistic predictive modeling workflows. This leads to a set of guidelines which further defines 'safe operating conditions' and 'safe operating zones' which will be a valuable reference for future SFIP operations. We also illustrate how HF can lead to groundwater aquifer contamination. The source of aquifer contamination can be the hydrocarbon gas or the chemicals used in the injected liquid in the formation. We explore possible pathways of contamination within and discuss the likelihood of contamination from each source. Many of the chemical compounds used

  14. Analysis of the RBMK-1500 type reactor emergency core cooling system behavior, taking into account the specified hydraulic characteristics of fast acting motor valves

    International Nuclear Information System (INIS)

    Kaliatka, A.; Ognerubov, V.; Adomavicius, A.; Ziedelis, S.

    2005-01-01

    During the accident analysis of nuclear power plants, reliability and uncertainty of results depends on adequateness of mathematical models of main elements and phenomena in systems important to safety. The best way for qualification of these models is collation with relevant experimental data. However, at the case of lack of such data modern computational fluid dynamics codes can be used for this purpose. This paper presents the results of an attempt to specify the hydraulic characteristics of the fast acting motor valves as well as to demonstrate the impact of these characteristics to transient processes in emergency core cooling system of the RBMK-1500 type reactor. For these purposes the finite element model of fast acting motor valve was developed and analyzed, using two separate computational fluid dynamics codes in parallel: CFX5 and COSMOS/FLOWORKS. Both all main design particularities and changes of flow structure during valve opening (closure) process were taken into account. It was demonstrated, that the obtained dependencies of changes of hydraulic loss coefficient in respect of relative valve opening (closure) rate substantially differ from those commonly used in thermal-hydraulic calculations of nuclear reactors. This difference is extremely big at the square one of the valve opening process, when the value of the valve hydraulic resistance is most important to flow of coolant channelized to the group distribution header. The series of thermal-hydraulic calculations of the maximum design-basis accident initiated by full break of main circulation pump pressure header were performed. The obtained dependencies of changes of hydraulic loss coefficient in respect of relative valve opening (closure) rate as well as those commonly used in thermal-hydraulic code RELAP5 were used. The results of calculations show, that in the initial stage of accident flow of coolant going from emergency core cooling system via fast acting motor valves to group distribution

  15. Energy harvesting from hydraulic pressure fluctuations

    International Nuclear Information System (INIS)

    Cunefare, K A; Skow, E A; Erturk, A; Savor, J; Verma, N; Cacan, M R

    2013-01-01

    State-of-the-art hydraulic hose and piping systems employ integral sensor nodes for structural health monitoring to avoid catastrophic failures. Energy harvesting in hydraulic systems could enable self-powered wireless sensor nodes for applications such as energy-autonomous structural health monitoring and prognosis. Hydraulic systems inherently have a high energy intensity associated with the mean pressure and flow. Accompanying the mean pressure is the dynamic pressure ripple, which is caused by the action of pumps and actuators. Pressure ripple is a deterministic source with a periodic time-domain behavior conducive to energy harvesting. An energy harvester prototype was designed for generating low-power electricity from pressure ripples. The prototype employed an axially-poled off-the-shelf piezoelectric stack. A housing isolated the stack from the hydraulic fluid while maintaining a mechanical coupling allowing for dynamic-pressure-induced deflection of the stack. The prototype exhibited an off-resonance energy harvesting problem since the fundamental resonance of the piezoelectric stack was much higher than the frequency content of the pressure ripple. The prototype was designed to provide a suitable power output for powering sensors with a maximum output of 1.2 mW. This work also presents electromechanical model simulations and experimental characterization of the piezoelectric power output from the pressure ripple in terms of the force transmitted into the harvester. (paper)

  16. High bulk modulus of ionic liquid and effects on performance of hydraulic system.

    Science.gov (United States)

    Kambic, Milan; Kalb, Roland; Tasner, Tadej; Lovrec, Darko

    2014-01-01

    Over recent years ionic liquids have gained in importance, causing a growing number of scientists and engineers to investigate possible applications for these liquids because of their unique physical and chemical properties. Their outstanding advantages such as nonflammable liquid within a broad liquid range, high thermal, mechanical, and chemical stabilities, low solubility for gases, attractive tribological properties (lubrication), and very low compressibility, and so forth, make them more interesting for applications in mechanical engineering, offering great potential for new innovative processes, and also as a novel hydraulic fluid. This paper focuses on the outstanding compressibility properties of ionic liquid EMIM-EtSO4, a very important physical chemically property when IL is used as a hydraulic fluid. This very low compressibility (respectively, very high Bulk modulus), compared to the classical hydraulic mineral oils or the non-flammable HFDU type of hydraulic fluids, opens up new possibilities regarding its usage within hydraulic systems with increased dynamics, respectively, systems' dynamic responses.

  17. Cavitation in Hydraulic Machinery

    Energy Technology Data Exchange (ETDEWEB)

    Kjeldsen, M.

    1996-11-01

    The main purpose of this doctoral thesis on cavitation in hydraulic machinery is to change focus towards the coupling of non-stationary flow phenomena and cavitation. It is argued that, in addition to turbulence, superimposed sound pressure fluctuations can have a major impact on cavitation and lead to particularly severe erosion. For the design of hydraulic devices this finding may indicate how to further limit the cavitation problems. Chapter 1 reviews cavitation in general in the context of hydraulic machinery, emphasizing the initial cavitation event and the role of the water quality. Chapter 2 discusses the existence of pressure fluctuations for situations common in such machinery. Chapter 3 on cavitation dynamics presents an algorithm for calculating the nucleation of a cavity cluster. Chapter 4 describes the equipment used in this work. 53 refs., 55 figs.,10 tabs.

  18. Trade-offs between xylem hydraulic properties, wood anatomy and yield in Populus.

    Science.gov (United States)

    Hajek, Peter; Leuschner, Christoph; Hertel, Dietrich; Delzon, Sylvain; Schuldt, Bernhard

    2014-07-01

    Trees face the dilemma that achieving high plant productivity is accompanied by a risk of drought-induced hydraulic failure due to a trade-off in the trees' vascular system between hydraulic efficiency and safety. By investigating the xylem anatomy of branches and coarse roots, and measuring branch axial hydraulic conductivity and vulnerability to cavitation in 4-year-old field-grown aspen plants of five demes (Populus tremula L. and Populus tremuloides Michx.) differing in growth rate, we tested the hypotheses that (i) demes differ in wood anatomical and hydraulic properties, (ii) hydraulic efficiency and safety are related to xylem anatomical traits, and (iii) aboveground productivity and hydraulic efficiency are negatively correlated to cavitation resistance. Significant deme differences existed in seven of the nine investigated branch-related anatomical and hydraulic traits but only in one of the four coarse-root-related anatomical traits; this likely is a consequence of high intra-plant variation in root morphology and the occurrence of a few 'high-conductivity roots'. Growth rate was positively related to branch hydraulic efficiency (xylem-specific conductivity) but not to cavitation resistance; this indicates that no marked trade-off exists between cavitation resistance and growth. Both branch hydraulic safety and hydraulic efficiency significantly depended on vessel size and were related to the genetic distance between the demes, while the xylem pressure causing 88% loss of hydraulic conductivity (P88 value) was more closely related to hydraulic efficiency than the commonly used P50 value. Deme-specific variation in the pit membrane structure may explain why vessel size was not directly linked to growth rate. We conclude that branch hydraulic efficiency is an important growth-influencing trait in aspen, while the assumed trade-off between productivity and hydraulic safety is weak. © The Author 2014. Published by Oxford University Press. All rights reserved

  19. Cryogenics safety

    International Nuclear Information System (INIS)

    Reider, R.

    1977-01-01

    The safety hazards associated with handling cryogenic fluids are discussed in detail. These hazards include pressure buildup when a cryogenic fluid is heated and becomes a gas, potential damage to body tissues due to surface contact, toxic risk from breathing air altered by cryogenic fluids, dangers of air solidification, and hazards of combustible cryogens such as liquified oxygen, hydrogen, or natural gas or of combustible mixtures. Safe operating procedures and emergency planning are described

  20. Hydraulics and pneumatics

    CERN Document Server

    Parr, Andrew

    2006-01-01

    Nearly all industrial processes require objects to be moved, manipulated or subjected to some sort of force. This is frequently accomplished by means of electrical equipment (such as motors or solenoids), or via devices driven by air (pneumatics) or liquids (hydraulics).This book has been written by a process control engineer as a guide to the operation of hydraulic and pneumatic systems for all engineers and technicians who wish to have an insight into the components and operation of such a system.This second edition has been fully updated to include all recent developments su

  1. Hydraulic Arm Modeling via Matlab SimHydraulics

    Czech Academy of Sciences Publication Activity Database

    Věchet, Stanislav; Krejsa, Jiří

    2009-01-01

    Roč. 16, č. 4 (2009), s. 287-296 ISSN 1802-1484 Institutional research plan: CEZ:AV0Z20760514 Keywords : simulatin modeling * hydraulics * SimHydraulics Subject RIV: JD - Computer Applications, Robotics

  2. Thermal hydraulic feasibility assessment of the spent nuclear fuel project

    International Nuclear Information System (INIS)

    Heard, F.J.

    1996-01-01

    A series of analyses have been completed investigating the thermal-hydraulic performance and feasibility of the Spent Nuclear Fuel Project (SNFP) Integrated Process Strategy (IPS). The goal was to develop a series of thermal-hydraulic models that could respond to all process and safety related issues that may arise pertaining to the SNFP, as well as provide a basis for validation of the results. Results show that there is a reasonable envelope for process conditions and requirements that are thermally and hydraulically acceptable

  3. Hydraulic method of working large super-drift pillars

    Energy Technology Data Exchange (ETDEWEB)

    Rad' ko, B.V.; Syroezhkin, P.V.; Durov, V.S.

    1987-03-01

    Describes the method of hydraulic coal extraction introduced in the Pioneer mine belonging to the Dobropol'eugol' coal association. This method was found to reduce the number of collection and ventilation roadways needed significantly, increase their stability, reduce coal loss and increase safety, particularly when mining pillars up to 80 m high. Large scale diagram of hydraulic mining layout shows: ventilation gate, hydraulic monitors, mine roadway, cross-cut, and collection roadways. A table shows pillar dimensions and depth and economic savings for different seams in the mine.

  4. Estimating Hydraulic Conductivities in a Fractured Shale Formation from Pressure Pulse Testing and 3d Modeling

    Science.gov (United States)

    Courbet, C.; DICK, P.; Lefevre, M.; Wittebroodt, C.; Matray, J.; Barnichon, J.

    2013-12-01

    In the framework of its research on the deep disposal of radioactive waste in shale formations, the French Institute for Radiological Protection and Nuclear Safety (IRSN) has developed a large array of in situ programs concerning the confining properties of shales in their underground research laboratory at Tournemire (SW France). One of its aims is to evaluate the occurrence and processes controlling radionuclide migration through the host rock, from the disposal system to the biosphere. Past research programs carried out at Tournemire covered mechanical, hydro-mechanical and physico-chemical properties of the Tournemire shale as well as water chemistry and long-term behaviour of the host rock. Studies show that fluid circulations in the undisturbed matrix are very slow (hydraulic conductivity of 10-14 to 10-15 m.s-1). However, recent work related to the occurrence of small scale fractures and clay-rich fault gouges indicate that fluid circulations may have been significantly modified in the vicinity of such features. To assess the transport properties associated with such faults, IRSN designed a series of in situ and laboratory experiments to evaluate the contribution of both diffusive and advective process on water and solute flux through a clay-rich fault zone (fault core and damaged zone) and in an undisturbed shale formation. As part of these studies, Modular Mini-Packer System (MMPS) hydraulic testing was conducted in multiple boreholes to characterize hydraulic conductivities within the formation. Pressure data collected during the hydraulic tests were analyzed using the nSIGHTS (n-dimensional Statistical Inverse Graphical Hydraulic Test Simulator) code to estimate hydraulic conductivity and formation pressures of the tested intervals. Preliminary results indicate hydraulic conductivities of 5.10-12 m.s-1 in the fault core and damaged zone and 10-14 m.s-1 in the adjacent undisturbed shale. Furthermore, when compared with neutron porosity data from borehole

  5. Comparison of in-plant performance test data with analytic prediction of reactor safety system injection transient (U)

    International Nuclear Information System (INIS)

    Roy, B.N.; Neill, C.H. Jr.

    1993-01-01

    This paper compares the performance test data from injection transients for both of the subsystems of the Supplementary Safety System of the Savannah River Site production reactor with analytical predictions from an in-house thermal hydraulic computer code. The code was initially developed for design validation of the new Supplementary Safety System subsystem, but is shown to be equally capable of predicting the performance of the Supplementary Safety System existing subsystem even though the two subsystem transient injections have marked differences. The code itself was discussed and its validation using prototypic tests with simulated fluids was reported in an earlier paper (Roy and Nomm 1991)

  6. Hydraulic Fracturing: Paving the Way for a Sustainable Future?

    Directory of Open Access Journals (Sweden)

    Jiangang Chen

    2014-01-01

    Full Text Available With the introduction of hydraulic fracturing technology, the United States has become the largest natural gas producer in the world with a substantial portion of the production coming from shale plays. In this review, we examined current hydraulic fracturing literature including associated wastewater management on quantity and quality of groundwater. We conclude that proper documentation/reporting systems for wastewater discharge and spills need to be enforced at the federal, state, and industrial level. Furthermore, Underground Injection Control (UIC requirements under SDWA should be extended to hydraulic fracturing operations regardless if diesel fuel is used as a fracturing fluid or not. One of the biggest barriers that hinder the advancement of our knowledge on the hydraulic fracturing process is the lack of transparency of chemicals used in the practice. Federal laws mandating hydraulic companies to disclose fracturing fluid composition and concentration not only to federal and state regulatory agencies but also to health care professionals would encourage this practice. The full disclosure of fracturing chemicals will allow future research to fill knowledge gaps for a better understanding of the impacts of hydraulic fracturing on human health and the environment.

  7. Hydraulic fracturing: paving the way for a sustainable future?

    Science.gov (United States)

    Chen, Jiangang; Al-Wadei, Mohammed H; Kennedy, Rebekah C M; Terry, Paul D

    2014-01-01

    With the introduction of hydraulic fracturing technology, the United States has become the largest natural gas producer in the world with a substantial portion of the production coming from shale plays. In this review, we examined current hydraulic fracturing literature including associated wastewater management on quantity and quality of groundwater. We conclude that proper documentation/reporting systems for wastewater discharge and spills need to be enforced at the federal, state, and industrial level. Furthermore, Underground Injection Control (UIC) requirements under SDWA should be extended to hydraulic fracturing operations regardless if diesel fuel is used as a fracturing fluid or not. One of the biggest barriers that hinder the advancement of our knowledge on the hydraulic fracturing process is the lack of transparency of chemicals used in the practice. Federal laws mandating hydraulic companies to disclose fracturing fluid composition and concentration not only to federal and state regulatory agencies but also to health care professionals would encourage this practice. The full disclosure of fracturing chemicals will allow future research to fill knowledge gaps for a better understanding of the impacts of hydraulic fracturing on human health and the environment.

  8. Mine drivage in hydraulic mines

    Energy Technology Data Exchange (ETDEWEB)

    Ehkber, B Ya

    1983-09-01

    From 20 to 25% of labor cost in hydraulic coal mines falls on mine drivage. Range of mine drivage is high due to the large number of shortwalls mined by hydraulic monitors. Reducing mining cost in hydraulic mines depends on lowering drivage cost by use of new drivage systems or by increasing efficiency of drivage systems used at present. The following drivage methods used in hydraulic mines are compared: heading machines with hydraulic haulage of cut rocks and coal, hydraulic monitors with hydraulic haulage, drilling and blasting with hydraulic haulage of blasted rocks. Mining and geologic conditions which influence selection of the optimum mine drivage system are analyzed. Standardized cross sections of mine roadways driven by the 3 methods are shown in schemes. Support systems used in mine roadways are compared: timber supports, roof bolts, roof bolts with steel elements, and roadways driven in rocks without a support system. Heading machines (K-56MG, GPKG, 4PU, PK-3M) and hydraulic monitors (GMDTs-3M, 12GD-2) used for mine drivage are described. Data on mine drivage in hydraulic coal mines in the Kuzbass are discussed. From 40 to 46% of roadways are driven by heading machines with hydraulic haulage and from 12 to 15% by hydraulic monitors with hydraulic haulage.

  9. Advanced thermal-hydraulic and neutronic codes: current and future applications. Summary and conclusions

    International Nuclear Information System (INIS)

    2001-05-01

    An OECD Workshop on Advanced Thermal-Hydraulic and Neutronic Codes Applications was held from 10 to 13 April 2000, in Barcelona, Spain, sponsored by the Committee on the Safety of Nuclear Installations (CSNI) of the OECD Nuclear Energy Agency (NEA). It was organised in collaboration with the Spanish Nuclear Safety Council (CSN) and hosted by CSN and the Polytechnic University of Catalonia (UPC) in collaboration with the Spanish Electricity Association (UNESA). The objectives of the Workshop were to review the developments since the previous CSNI Workshop held in Annapolis [NEA/CSNI/ R(97)4; NUREG/CP-0159], to analyse the present status of maturity and remnant needs of thermal-hydraulic (TH) and neutronic system codes and methods, and finally to evaluate the role of these tools in the evolving regulatory environment. The Technical Sessions and Discussion Sessions covered the following topics: - Regulatory requirements for Best-Estimate (BE) code assessment; - Application of TH and neutronic codes for current safety issues; - Uncertainty analysis; - Needs for integral plant transient and accident analysis; - Simulators and fast running codes; - Advances in next generation TH and neutronic codes; - Future trends in physical modeling; - Long term plans for development of advanced codes. The focus of the Workshop was on system codes. An incursion was made, however, in the new field of applying Computational Fluid Dynamic (CFD) codes to nuclear safety analysis. As a general conclusion, the Barcelona Workshop can be considered representative of the progress towards the targets marked at Annapolis almost four years ago. The Annapolis Workshop had identified areas where further development and specific improvements were needed, among them: multi-field models, transport of interfacial area, 2D and 3D thermal-hydraulics, 3-D neutronics consistent with level of details of thermal-hydraulics. Recommendations issued at Annapolis included: developing small pilot/test codes for

  10. Proceedings of the workshop on advanced thermal-hydraulic and neutronic codes: current and future applications

    International Nuclear Information System (INIS)

    2001-01-01

    An OECD Workshop on Advanced Thermal-Hydraulic and Neutronic Codes Applications was held from 10 to 13 April 2000, in Barcelona, Spain, sponsored by the Committee on the Safety of Nuclear Installations (CSNI) of the OECD Nuclear Energy Agency (NEA). It was organised in collaboration with the Spanish Nuclear Safety Council (CSN) and hosted by CSN and the Polytechnic University of Catalonia (UPC) in collaboration with the Spanish Electricity Association (UNESA). The objectives of the Workshop were to review the developments since the previous CSNI Workshop held in Annapolis [NEA/CSNI/ R(97)4; NUREG/CP-0159], to analyse the present status of maturity and remnant needs of thermal-hydraulic (TH) and neutronic system codes and methods, and finally to evaluate the role of these tools in the evolving regulatory environment. The Technical Sessions and Discussion Sessions covered the following topics: - Regulatory requirements for Best-Estimate (BE) code assessment; - Application of TH and neutronic codes for current safety issues; - Uncertainty analysis; - Needs for integral plant transient and accident analysis; - Simulators and fast running codes; - Advances in next generation TH and neutronic codes; - Future trends in physical modeling; - Long term plans for development of advanced codes. The focus of the Workshop was on system codes. An incursion was made, however, in the new field of applying Computational Fluid Dynamic (CFD) codes to nuclear safety analysis. As a general conclusion, the Barcelona Workshop can be considered representative of the progress towards the targets marked at Annapolis almost four years ago. The Annapolis Workshop had identified areas where further development and specific improvements were needed, among them: multi-field models, transport of interfacial area, 2D and 3D thermal-hydraulics, 3-D neutronics consistent with level of details of thermal-hydraulics. Recommendations issued at Annapolis included: developing small pilot/test codes for

  11. Optimisation of Working Areas in Discrete Hydraulic Power Take off-system for Wave Energy Converters

    DEFF Research Database (Denmark)

    Hansen, Anders Hedegaard; Hansen, Rico Hjerm; Pedersen, Henrik C.

    2012-01-01

    Fluid power is the leading technology in Power Take Off(PTO) systems in Wave Energy Converters(WEC’s), due to the capability of generating high force at low velocity. However, as hydraulic force controlling system may suffer from large energy losses the efficiency of the hydraulic PTO systems may...

  12. Hydraulic simulation of the systems of a nuclear power plant for charges calculation in piping

    International Nuclear Information System (INIS)

    Masriera, N.

    1990-01-01

    This work presents a general description of the methodology used by the ENACE S.A. Fluids Working Group for hydraulics simulation of a nuclear power plant system for the calculation charges in piping. (Author) [es

  13. Hydraulic shock absorbers

    International Nuclear Information System (INIS)

    Thatcher, G.; Davidson, D. F.

    1984-01-01

    A hydraulic shock absorber of the dash pot kind for use with electrically conducting liquid such as sodium, has magnet means for electro magnetically braking a stream of liquid discharged from the cylinder. The shock absorber finds use in a liquid metal cooled nuclear reactor for arresting control rods

  14. Preparation of hydraulic cement

    Energy Technology Data Exchange (ETDEWEB)

    1921-08-28

    A process for the preparation of hydraulic cement by the use of oil-shale residues is characterized in that the oil-shale refuse is mixed with granular basic blast-furnace slag and a small amount of portland cement and ground together.

  15. Modelling of Hydraulic Robot

    DEFF Research Database (Denmark)

    Madsen, Henrik; Zhou, Jianjun; Hansen, Lars Henrik

    1997-01-01

    This paper describes a case study of identifying the physical model (or the grey box model) of a hydraulic test robot. The obtained model is intended to provide a basis for model-based control of the robot. The physical model is formulated in continuous time and is derived by application...

  16. Manual Hydraulic Structures

    NARCIS (Netherlands)

    Molenaar, W.F.; Voorendt, M.Z.

    This manual is the result of group work and origins in Dutch lecture notes that have been used since long time. Amongst the employees of the Hydraulic Engineering Department that contributed to this work are dr.ir. S. van Baars, ir.K.G.Bezuijen, ir.G.P.Bourguignon, prof.ir.A.Glerum,

  17. Water Treatment Technology - Hydraulics.

    Science.gov (United States)

    Ross-Harrington, Melinda; Kincaid, G. David

    One of twelve water treatment technology units, this student manual on hydraulics provides instructional materials for three competencies. (The twelve units are designed for a continuing education training course for public water supply operators.) The competencies focus on the following areas: head loss in pipes in series, function loss in…

  18. Fluid Power, Rate Training Manual.

    Science.gov (United States)

    Bureau of Naval Personnel, Washington, DC.

    Fundamentals of hydraulics and pneumatics are presented in this manual, prepared for regular navy and naval reserve personnel who are seeking advancement to Petty Officer Third Class. The history of applications of compressed fluids is described in connection with physical principles. Selection of types of liquids and gases is discussed with a…

  19. Thermal-hydraulic interfacing code modules for CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Liu, W.S.; Gold, M.; Sills, H. [Ontario Hydro Nuclear, Toronto (Canada)] [and others

    1997-07-01

    The approach for CANDU reactor safety analysis in Ontario Hydro Nuclear (OHN) and Atomic Energy of Canada Limited (AECL) is presented. Reflecting the unique characteristics of CANDU reactors, the procedure of coupling the thermal-hydraulics, reactor physics and fuel channel/element codes in the safety analysis is described. The experience generated in the Canadian nuclear industry may be useful to other types of reactors in the areas of reactor safety analysis.

  20. Thermal-hydraulic interfacing code modules for CANDU reactors

    International Nuclear Information System (INIS)

    Liu, W.S.; Gold, M.; Sills, H.

    1997-01-01

    The approach for CANDU reactor safety analysis in Ontario Hydro Nuclear (OHN) and Atomic Energy of Canada Limited (AECL) is presented. Reflecting the unique characteristics of CANDU reactors, the procedure of coupling the thermal-hydraulics, reactor physics and fuel channel/element codes in the safety analysis is described. The experience generated in the Canadian nuclear industry may be useful to other types of reactors in the areas of reactor safety analysis

  1. Cradle modification for hydraulic ram

    International Nuclear Information System (INIS)

    Koons, B.M.

    1995-01-01

    The analysis of the cradle hydraulic system considers stress, weld strength, and hydraulic forces required to lift and support the cradle/pump assembly. The stress and weld strength of the cradle modifications is evaluated to ensure that they meet the requirements of the American Institute for Steel Construction (AISC 1989). The hydraulic forces are evaluated to ensure that the hydraulic system is capable of rotating the cradle and pump assembly to the vertical position (between 70 degrees and 90 degrees)

  2. HYDRAULIC AND PHYSICAL PROPERTIES OF MCU SALTSTONE

    International Nuclear Information System (INIS)

    Dixon, K; Mark Phifer, M

    2008-01-01

    The Saltstone Disposal Facility (SDF), located in the Z-Area of the Savannah River Site (SRS), is used for the disposal of low-level radioactive salt solution. The SDF currently contains two vaults: Vault 1 (6 cells) and Vault 4 (12 cells). Additional disposal cells are currently in the design phase. The individual cells of the saltstone facility are filled with saltstone., Saltstone is produced by mixing the low-level radioactive salt solution, with blast furnace slag, fly ash, and cement or lime to form a dense, micro-porous, monolithic, low-level radioactive waste form. The saltstone is pumped into the disposal cells where it subsequently solidifies. Significant effort has been undertaken to accurately model the movement of water and contaminants through the facility. Key to this effort is an accurate understanding of the hydraulic and physical properties of the solidified saltstone. To date, limited testing has been conducted to characterize the saltstone. The primary focus of this task was to estimate the hydraulic and physical properties of MCU (Modular Caustic Side Solvent Extraction Unit) saltstone relative to two permeating fluids. These fluids included simulated groundwater equilibrated with vault concrete and simulated saltstone pore fluid. Samples of the MCU saltstone were prepared by the Savannah River National Laboratory (SRNL) and allowed to cure for twenty eight days prior to testing. These samples included two three-inch diameter by six inch long mold samples and three one-inch diameter by twelve inch long mold samples

  3. HYDRAULIC AND PHYSICAL PROPERTIES OF MCU SALTSTONE

    Energy Technology Data Exchange (ETDEWEB)

    Dixon, K; Mark Phifer, M

    2008-03-19

    The Saltstone Disposal Facility (SDF), located in the Z-Area of the Savannah River Site (SRS), is used for the disposal of low-level radioactive salt solution. The SDF currently contains two vaults: Vault 1 (6 cells) and Vault 4 (12 cells). Additional disposal cells are currently in the design phase. The individual cells of the saltstone facility are filled with saltstone., Saltstone is produced by mixing the low-level radioactive salt solution, with blast furnace slag, fly ash, and cement or lime to form a dense, micro-porous, monolithic, low-level radioactive waste form. The saltstone is pumped into the disposal cells where it subsequently solidifies. Significant effort has been undertaken to accurately model the movement of water and contaminants through the facility. Key to this effort is an accurate understanding of the hydraulic and physical properties of the solidified saltstone. To date, limited testing has been conducted to characterize the saltstone. The primary focus of this task was to estimate the hydraulic and physical properties of MCU (Modular Caustic Side Solvent Extraction Unit) saltstone relative to two permeating fluids. These fluids included simulated groundwater equilibrated with vault concrete and simulated saltstone pore fluid. Samples of the MCU saltstone were prepared by the Savannah River National Laboratory (SRNL) and allowed to cure for twenty eight days prior to testing. These samples included two three-inch diameter by six inch long mold samples and three one-inch diameter by twelve inch long mold samples.

  4. Elevator and hydraulics; Elevator to yuatsu

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, I. [Hitachi, Ltd., Tokyo (Japan)

    1994-07-15

    A hydraulic type elevator is installed in relatively lower buildings as compared with a rope type elevator, but the ratio in the number of installation of the former elevator is increasing. This paper explains from its construction and features to especially various control systems for the riding comfort and safety. A direct push-up system with hydraulic jacks arranged beneath a car, and an indirect push-up system that has hydraulic jacks arranged on flank of a car and transmits the movement of a plunger via a rope are available. The latter system eliminates the need of large holes to embed hydraulic jacks. While the speed is controlled by controlling flow rates of high-pressure oil, the speed, position, acceleration and even time differential calculus of the acceleration must be controlled severely. The system uses two-step control for the through-speed and the landing speed. Different systems that have been realized may include compensation for temperatures in flow rate control valves, load pressures, and oil viscosity, from learning control to fuzzy control for psychological effects, or control of inverters in motors. 13 refs., 12 figs., 1 tab.

  5. Order of 11 October 1977 on general safety measures applicable to fluids, radioactive waste and irradiated and non-irradiated fuels in large nuclear installations

    International Nuclear Information System (INIS)

    1978-01-01

    This Order by the Minister of Industry, Commerce and Crafts and the Minister of Labour was made in implementation of Section 40 of Decree No. 75-306 of 28 april 1975 on the protection of workers against the hazards of ionizing radiation in large nuclear installations. It lays down the safety measures to be taken as regards construction of the installation to limit radioactive dispersal and exposure of workers. The Order specifies the characteristics of the piping and vessels as well as the materials to be used for construction of such vessels and piping. The radioactive fluids must be contained in especially designed pipes and vessels and transport of radioactive substances within installations must be carried out with the approval of the person responsible for radiation protection as defined in Decree No 75-306. Finally all possible measures must be taken to eliminate risks of criticality, in particular when the quantity of fissile materials likely to be assembled is likely to exceed the limits fixed by Order of 25 January 1867. (NEA) [fr

  6. Hydraulic turbines and auxiliary equipment

    Energy Technology Data Exchange (ETDEWEB)

    Luo Gaorong [Organization of the United Nations, Beijing (China). International Centre of Small Hydroelectric Power Plants

    1995-07-01

    This document presents a general overview on hydraulic turbines and auxiliary equipment, emphasizing the turbine classification, in accordance with the different types of turbines, standard turbine series in China, turbine selection based on the basic data required for the preliminary design, general hill model curves, chart of turbine series and the arrangement of application for hydraulic turbines, hydraulic turbine testing, and speed regulating device.

  7. Hydraulic Hybrid Vehicle Publications | Transportation Research | NREL

    Science.gov (United States)

    Hydraulic Hybrid Vehicle Publications Hydraulic Hybrid Vehicle Publications The following technical papers and fact sheets provide information about NREL's hydraulic hybrid fleet vehicle evaluations . Refuse Trucks Project Startup: Evaluating the Performance of Hydraulic Hybrid Refuse Vehicles. Bob

  8. FEASIBILITY OF HYDRAULIC FRACTURING OF SOILS TO IMPROVE REMEDIAL ACTIONS

    Science.gov (United States)

    Hydraulic fracturing, a method of increasing fluid flow within the subsurface, should improve the effectiveness of several remedial techniques, including pump and treat, vapor extraction, bio-remediation, and soil-flushing. he technique is widely used to increase the yields of oi...

  9. Experimental study of hydraulic transport of coarse basalt

    Czech Academy of Sciences Publication Activity Database

    Matoušek, Václav; Vlasák, Pavel; Chára, Zdeněk; Konfršt, Jiří

    2015-01-01

    Roč. 148, č. 2 (2015), s. 93-100 ISSN 1741-7597 R&D Projects: GA ČR GAP105/10/1574 Institutional support: RVO:67985874 Keywords : hydraulics * hydrodynamics * dredging * pipes * pipelines Subject RIV: BK - Fluid Dynamics Impact factor: 0.281, year: 2015

  10. Determination of unsaturated hydraulic conductivity of alfisol soil in ...

    African Journals Online (AJOL)

    The hydrolic conductivity of soil measures the ease at which water moves through the soil by determining the flux density of water passing through the soil. The estimation of hydraulic conductivity indicates how fluids flow throuhg a substance and thus determine the water balance in the soil profile. The trend lines of ...

  11. The Influence of Hydraulic Fracturing on Carbon Storage Performance

    Science.gov (United States)

    Fu, Pengcheng; Settgast, Randolph R.; Hao, Yue; Morris, Joseph P.; Ryerson, Frederick J.

    2017-12-01

    Conventional principles of the design and operation of geologic carbon storage (GCS) require injecting CO2 below the caprock fracturing pressure to ensure the integrity of the storage complex. In nonideal storage reservoirs with relatively low permeability, pressure buildup can lead to hydraulic fracturing of the reservoir and caprock. While the GCS community has generally viewed hydraulic fractures as a key risk to storage integrity, a carefully designed stimulation treatment under appropriate geologic conditions could provide improved injectivity while maintaining overall seal integrity. A vertically contained hydraulic fracture, either in the reservoir rock or extending a limited height into the caprock, provides an effective means to access reservoir volume far from the injection well. Employing a fully coupled numerical model of hydraulic fracturing, solid deformation, and matrix fluid flow, we study the enabling conditions, processes, and mechanisms of hydraulic fracturing during CO2 injection. A hydraulic fracture's pressure-limiting behavior dictates that the near-well fluid pressure is only slightly higher than the fracturing pressure of the rock and is insensitive to injection rate and mechanical properties of the formation. Although a fracture contained solely within the reservoir rock with no caprock penetration, would be an ideal scenario, poroelastic principles dictate that sustaining such a fracture could lead to continuously increasing pressure until the caprock fractures. We also investigate the propagation pattern and injection pressure responses of a hydraulic fracture propagating in a caprock subjected to heterogeneous in situ stress. The results have important implications for the use of hydraulic fracturing as a tool for managing storage performance.

  12. Implementation of CFD module in the KORSAR thermal-hydraulic system code

    Energy Technology Data Exchange (ETDEWEB)

    Yudov, Yury V.; Danilov, Ilia G.; Chepilko, Stepan S. [Alexandrov Research Inst. of Technology (NITI), Sosnovy Bor (Russian Federation)

    2015-09-15

    The Russian KORSAR/GP (hereinafter KORSAR) computer code was developed by a joint team from Alexandrov NITI and OKB ''Gidropress'' for VVER safety analysis and certified by the Rostechnadzor of Russia in 2009. The code functionality is based on a 1D two-fluid model for calculation of two-phase flows. A 3D CFD module in the KORSAR computer code is being developed by Alexandrov NITI for representing 3D effects in the downcomer and lower plenum during asymmetrical loop operation. The CFD module uses Cartesian grid method with cut cell approach. The paper presents a numerical algorithm for coupling 1D and 3D thermal- hydraulic modules in the KORSAR code. The combined pressure field is calculated by the multigrid method. The performance efficiency of the algorithm for coupling 1D and 3D modules was demonstrated by solving the benchmark problem of mixing cold and hot flows in a T-junction.

  13. Current and anticipated uses of thermal hydraulic codes at the Japan Atomic Energy Research Institute

    International Nuclear Information System (INIS)

    Akimoto, Hajime; Kukita; Ohnuki, Akira

    1997-01-01

    The Japan Atomic Energy Research Institute (JAERI) is conducting several research programs related to thermal-hydraulic and neutronic behavior of light water reactors (LWRs). These include LWR safety research projects, which are conducted in accordance with the Nuclear Safety Commission's research plan, and reactor engineering projects for the development of innovative reactor designs or core/fuel designs. Thermal-hydraulic and neutronic codes are used for various purposes including experimental analysis, nuclear power plant (NPP) safety analysis, and design assessment

  14. Applied mathematical methods in nuclear thermal hydraulics

    International Nuclear Information System (INIS)

    Ransom, V.H.; Trapp, J.A.

    1983-01-01

    Applied mathematical methods are used extensively in modeling of nuclear reactor thermal-hydraulic behavior. This application has required significant extension to the state-of-the-art. The problems encountered in modeling of two-phase fluid transients and the development of associated numerical solution methods are reviewed and quantified using results from a numerical study of an analogous linear system of differential equations. In particular, some possible approaches for formulating a well-posed numerical problem for an ill-posed differential model are investigated and discussed. The need for closer attention to numerical fidelity is indicated

  15. Hydraulic system for driving control rods

    International Nuclear Information System (INIS)

    Okuzumi, Naoaki.

    1982-01-01

    Purpose: To enable safety reactor shut down upon occurrence of an abnormal excess pressure in a hydraulic control unit. Constitution: The actuation pressure for a pressure switch that generates a scram signal is set lower than the release pressure set to a pressure release valve. Thus, if the pressure of nitrogen gas in a nitrogen container increases such as upon exposure of the hydraulic control unit to a high temperature, the pressure switch is actuated at first to generate the scram signal and a scram valve is opened to supply water at high pressure to control rod drives under the driving force of the nitrogen gas at high pressure to rapidly insert the control element into the reactor and shut down it. If the pressure of the nitrogen gas still increases after the scram, the pressure release valve is opened to release the nitrogen gas at high temperature to the atmosphere. Since the scram is attained before the actuation of the pressure release valve, safety reactor shut down can be attained and the hydraulic control unit can be protected. (Sekiya, K.)

  16. Forschungszentrum Rossendorf, Institute of Safety Research. Annual report 2003

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

    The work of the institute is directed to the assessment and enhancement of the safety of technical plants and to the increase of the efficiency and environmental sustainability of those facilities. Subjects of investigation are equally nuclear plants and installations of process industries. To achieve the goals mentioned, the institute is mainly engaged in the scientific fields of thermal fluid dynamics including magneto-hydrodynamics (MHD) and materials sciences. In 2003, the ISR worked on the following main scientific projects. Sub-programme: Plant and Rector Safety. Project: accident analysis of nuclear reactors, safety of materials and components, particle and radiation transport, safety and efficiency of chemical processes. Sub-programme: Thermal Fluid Dynamics. Project: magneto-hydrodynamics, thermal fluid dynamics of multiphase systems. Considerable progress could also be achieved in the CFD simulation of two-phase flows. New approaches for the forces acting on steam bubbles in a water flow could be developed and implemented into the CFX code in close cooperation with the CFX developer ANSYS/CFX. The qualified models allow to simulate the evolution of bubble size specific radial void distribution profiles along the flow path. These theoretical studies and the related experiments at the Rossendorf TOPFLOW test facility represent an important part of the German CFD network that aims at the improvement of thermal hydraulic calculation methods in reactor safety. (orig.)

  17. Forschungszentrum Rossendorf, Institute of Safety Research. Annual report 2003

    International Nuclear Information System (INIS)

    2004-01-01

    The work of the institute is directed to the assessment and enhancement of the safety of technical plants and to the increase of the efficiency and environmental sustainability of those facilities. Subjects of investigation are equally nuclear plants and installations of process industries. To achieve the goals mentioned, the institute is mainly engaged in the scientific fields of thermal fluid dynamics including magneto-hydrodynamics (MHD) and materials sciences. In 2003, the ISR worked on the following main scientific projects. Sub-programme: Plant and Rector Safety. Project: accident analysis of nuclear reactors, safety of materials and components, particle and radiation transport, safety and efficiency of chemical processes. Sub-programme: Thermal Fluid Dynamics. Project: magneto-hydrodynamics, thermal fluid dynamics of multiphase systems. Considerable progress could also be achieved in the CFD simulation of two-phase flows. New approaches for the forces acting on steam bubbles in a water flow could be developed and implemented into the CFX code in close cooperation with the CFX developer ANSYS/CFX. The qualified models allow to simulate the evolution of bubble size specific radial void distribution profiles along the flow path. These theoretical studies and the related experiments at the Rossendorf TOPFLOW test facility represent an important part of the German CFD network that aims at the improvement of thermal hydraulic calculation methods in reactor safety. (orig.)

  18. Hydraulic manipulator research at ORNL

    International Nuclear Information System (INIS)

    Kress, R.L.; Jansen, J.F.; Love, L.J.

    1997-01-01

    Recently, task requirements have dictated that manipulator payload capacity increase to accommodate greater payloads, greater manipulator length, and larger environmental interaction forces. General tasks such as waste storage tank cleanup and facility dismantlement and decommissioning require manipulator life capacities in the range of hundreds of pounds rather than tens of pounds. To meet the increased payload capacities demanded by present-day tasks, manipulator designers have turned once again to hydraulics as a means of actuation. In order to successfully design, build, and deploy a new hydraulic manipulator (or subsystem), sophisticated modeling, analysis, and control experiments are usually needed. Oak Ridge National Laboratory (ORNL) has a history of projects that incorporate hydraulics technology, including mobile robots, teleoperated manipulators, and full-scale construction equipment. In addition, to support the development and deployment of new hydraulic manipulators, ORNL has outfitted a significant experimental laboratory and has developed the software capability for research into hydraulic manipulators, hydraulic actuators, hydraulic systems, modeling of hydraulic systems, and hydraulic controls. The purpose of this article is to describe the past hydraulic manipulator developments and current hydraulic manipulator research capabilities at ORNL. Included are example experimental results from ORNL's flexible/prismatic test stand

  19. Hydraulic manipulator research at ORNL

    Energy Technology Data Exchange (ETDEWEB)

    Kress, R.L.; Jansen, J.F. [Oak Ridge National Lab., TN (United States); Love, L.J. [Oak Ridge Inst. for Science and Education, TN (United States)

    1997-03-01

    Recently, task requirements have dictated that manipulator payload capacity increase to accommodate greater payloads, greater manipulator length, and larger environmental interaction forces. General tasks such as waste storage tank cleanup and facility dismantlement and decommissioning require manipulator life capacities in the range of hundreds of pounds rather than tens of pounds. To meet the increased payload capacities demanded by present-day tasks, manipulator designers have turned once again to hydraulics as a means of actuation. In order to successfully design, build, and deploy a new hydraulic manipulator (or subsystem), sophisticated modeling, analysis, and control experiments are usually needed. Oak Ridge National Laboratory (ORNL) has a history of projects that incorporate hydraulics technology, including mobile robots, teleoperated manipulators, and full-scale construction equipment. In addition, to support the development and deployment of new hydraulic manipulators, ORNL has outfitted a significant experimental laboratory and has developed the software capability for research into hydraulic manipulators, hydraulic actuators, hydraulic systems, modeling of hydraulic systems, and hydraulic controls. The purpose of this article is to describe the past hydraulic manipulator developments and current hydraulic manipulator research capabilities at ORNL. Included are example experimental results from ORNL`s flexible/prismatic test stand.

  20. Hydraulically actuated artificial muscles

    Science.gov (United States)

    Meller, M. A.; Tiwari, R.; Wajcs, K. B.; Moses, C.; Reveles, I.; Garcia, E.

    2012-04-01

    Hydraulic Artificial Muscles (HAMs) consisting of a polymer tube constrained by a nylon mesh are presented in this paper. Despite the actuation mechanism being similar to its popular counterpart, which are pneumatically actuated (PAM), HAMs have not been studied in depth. HAMs offer the advantage of compliance, large force to weight ratio, low maintenance, and low cost over traditional hydraulic cylinders. Muscle characterization for isometric and isobaric tests are discussed and compared to PAMs. A model incorporating the effect of mesh angle and friction have also been developed. In addition, differential swelling of the muscle on actuation has also been included in the model. An application of lab fabricated HAMs for a meso-scale robotic system is also presented.

  1. Transitioning from interpretive to predictive in thermal hydraulic codes

    International Nuclear Information System (INIS)

    Mousseau, V.A.

    2004-01-01

    The current thermal hydraulic codes in use in the US, RELAP and TRAC, where originally written in the mid to late 1970's. At that time computers were slow, expensive, and had small memories. Because of these constraints, sacrifices had to be made, both in physics and numerical methods, which resulted in limitations on the accuracy of the solutions. Significant changes have occurred that induce very different requirements for the thermal hydraulic codes to be used for the future GEN-IV nuclear reactors. First, computers speed and memory grow at an exponential rate while the costs hold constant or decrease. Second, passive safety systems in modern designs stretch the length of relevant transients to many days. Finally, costs of experiments have grown very rapidly. Because of these new constraints, modern thermal hydraulic codes will be relied on for a significantly larger portion of bringing a nuclear reactor on line. Simulation codes will have to define in which part of state space experiments will be run. They will then have to be able to extend the small number of experiments to cover the large state space in which the reactors will operate. This data extrapolation mode will be referred to as 'predictive'. One of the keys to analyzing the accuracy of a simulation is to consider the entire domain being simulated. For example, in a reactor design where the containment is coupled to the reactor cooling system through radiative heat transfer, the accuracy of a transient includes the containment, the radiation heat transfer, the fluid flow in the cooling system, the thermal conduction in the solid, and the neutron transport in the reactor. All of this physics is coupled together in one nonlinear system through material properties, cross sections, heat transfer coefficients, and other mechanisms that exchange mass, momentum, and energy. Traditionally, these different physical domains, (containment, cooling system, nuclear fuel, etc.) have been solved in different

  2. Undular Hydraulic Jump

    Directory of Open Access Journals (Sweden)

    Oscar Castro-Orgaz

    2015-04-01

    Full Text Available The transition from subcritical to supercritical flow when the inflow Froude number Fo is close to unity appears in the form of steady state waves called undular hydraulic jump. The characterization of the undular hydraulic jump is complex due to the existence of a non-hydrostatic pressure distribution that invalidates the gradually-varied flow theory, and supercritical shock waves. The objective of this work is to present a mathematical model for the undular hydraulic jump obtained from an approximate integration of the Reynolds equations for turbulent flow assuming that the Reynolds number R is high. Simple analytical solutions are presented to reveal the physics of the theory, and a numerical model is used to integrate the complete equations. The limit of application of the theory is discussed using a wave breaking condition for the inception of a surface roller. The validity of the mathematical predictions is critically assessed using physical data, thereby revealing aspects on which more research is needed

  3. Thermal hydraulics in undergraduate nuclear engineering education

    International Nuclear Information System (INIS)

    Theofanous, T.G.

    1986-01-01

    The intense safety-related research efforts of the seventies in reactor thermal hydraulics have brought about the recognition of the subject as one of the cornerstones of nuclear engineering. Many nuclear engineering departments responded by building up research programs in this area, and mostly as a consequence, educational programs, too. Whether thermal hydraulics has fully permeated the conscience of nuclear engineering, however, remains yet to be seen. The lean years that lie immediately ahead will provide the test. The purpose of this presentation is to discuss the author's own educational activity in undergraduate nuclear engineering education over the past 10 yr or so. All this activity took place at Purdue's School of Nuclear Engineering. He was well satisfied with the results and expects to implement something similar at the University of California in Santa Barbara in the near future

  4. Effect of hydraulic hysteresis on the stability of infinite slopes under steady infiltration

    Science.gov (United States)

    Chen, Pan; Mirus, Benjamin B.; Lu, Ning; Godt, Jonathan W.

    2017-01-01

    Hydraulic hysteresis, including capillary soil water retention (SWR), air entrapment SWR, and hydraulic conductivity, is a common phenomenon in unsaturated soils. However, the influence of hydraulic hysteresis on suction stress, and subsequently slope stability, is generally ignored. This paper examines the influence of each of these three types of hysteresis on slope stability using an infinite slope stability analysis under steady infiltration conditions. First, hypothetical slopes for representative silty and sandy soils are examined. Then a monitored hillslope in the San Francisco Bay Area, California is assessed, using observed rainfall conditions and measured hydraulic and geotechnical properties of the colluvial soil. Results show that profiles of suction stress and the corresponding factor of safety are generally strongly affected by hydraulic hysteresis. Results suggest that each of the three types of hydraulic hysteresis may play a major role in the occurrence of slope failure, indicating that ignoring hydraulic hysteresis will likely lead to underestimates of failure potential and hence to inaccurate slope stability analysis.

  5. Thermal-Hydraulic Experiment To Test The Stable Operation Of A PIUS Type Reactor

    International Nuclear Information System (INIS)

    Irianto, Djoko; Kanji, T.; Kukita, Y.

    1996-01-01

    An advanced type of reaktor concept as the Process Inherent Ultimate Safety (PIUS) reactor was based on intrinsically passive safety considerations. The stable operation of a PIUS type reactor is based on the automation of circulation pump speed. An automatic circulation pump speed control system by using a measurement of the temperature distribution in the lower density lock is proposed the PIUS-type reactor. In principle this control system maintains the fluid temperature at the axial center of the lower density lock at average of the fluid temperatures below and above the lower density lock. This control system will prevent the poison water from penetrating into the core during normal operation. The effectiveness of this control system was successfully confirmed by a series of experiments using atmospheric pressure thermal-hydraulic test loop which simulated the PIUS principle. The experiments such as: start-up and power ramping tests for normal operation simulation and loss of feedwater test for an accident condition simulation, carried out in JAERI

  6. Quasi-open loop hydraulic ram incremental actuator with power conserving properties

    International Nuclear Information System (INIS)

    Raymond, E.T.; Robinson, C.W.

    1982-01-01

    An electric stepping motor, operated by command signals from a computer or a microprocessor, rotates a rotary control member of a distributor valve, for sequencing hydraulic pressure and hence flow to the cylinders of an axial piston hydraulic machine. A group of the cylinders are subjected to pressure and flow and the remaining cylinders are vented to a return line. Rotation of the rotary control valve member sequences pressurization by progressively adding a cylinder to the forward edge to the pressurized group and removing a cylinder from the trailing edge of the pressurized group. The double ended pistons of each new pressurized group function to drive a wobble plate into a new position of equilibrium and then hold it in such position until another change in the makeup of the pressurized group. These pistons also displace hydraulic fluid from the opposite cylinder head which serves as the output of a pumping element. An increment of displacement of the wobble plate occurs in direct response to each command pulse that is received by the stepping motor. Wobble plate displacement drives the rotary valve of the hydraulic power transfer unit, causing it to transfer hydraulic fluid from a first expansible chamber on one side of a piston in a hydraulic ram to a second expansible chamber on the opposite side of the piston. Reverse drive of the hydraulic power transfer unit reverses the direction of transfer of hydraulic fluid between the two expansible chambers

  7. 4-H NFPA Fluid Power Challenge

    OpenAIRE

    Bonnett, Erika D

    2016-01-01

    The 4-H NFPA Fluid Power Challenge partnered Purdue Polytechnic Institute and Indiana 4-H with the National Fluid Power Association and Center for Compact and Efficient Fluid Power to provide teams of Indiana youth in 6-8th grades with opportunity to learn about hydraulics, engineering design, and other STEM skills. This created an opportunity to give youth a learning experience with STEM through hands-on, experiential learning activities. Youth experienced a one day workshop in which they wo...

  8. CFD [computational fluid dynamics] And Safety Factors. Computer modeling of complex processes needs old-fashioned experiments to stay in touch with reality.

    Energy Technology Data Exchange (ETDEWEB)

    Leishear, Robert A.; Lee, Si Y.; Poirier, Michael R.; Steeper, Timothy J.; Ervin, Robert C.; Giddings, Billy J.; Stefanko, David B.; Harp, Keith D.; Fowley, Mark D.; Van Pelt, William B.

    2012-10-07

    Computational fluid dynamics (CFD) is recognized as a powerful engineering tool. That is, CFD has advanced over the years to the point where it can now give us deep insight into the analysis of very complex processes. There is a danger, though, that an engineer can place too much confidence in a simulation. If a user is not careful, it is easy to believe that if you plug in the numbers, the answer comes out, and you are done. This assumption can lead to significant errors. As we discovered in the course of a study on behalf of the Department of Energy's Savannah River Site in South Carolina, CFD models fail to capture some of the large variations inherent in complex processes. These variations, or scatter, in experimental data emerge from physical tests and are inadequately captured or expressed by calculated mean values for a process. This anomaly between experiment and theory can lead to serious errors in engineering analysis and design unless a correction factor, or safety factor, is experimentally validated. For this study, blending times for the mixing of salt solutions in large storage tanks were the process of concern under investigation. This study focused on the blending processes needed to mix salt solutions to ensure homogeneity within waste tanks, where homogeneity is required to control radioactivity levels during subsequent processing. Two of the requirements for this task were to determine the minimum number of submerged, centrifugal pumps required to blend the salt mixtures in a full-scale tank in half a day or less, and to recommend reasonable blending times to achieve nearly homogeneous salt mixtures. A full-scale, low-flow pump with a total discharge flow rate of 500 to 800 gpm was recommended with two opposing 2.27-inch diameter nozzles. To make this recommendation, both experimental and CFD modeling were performed. Lab researchers found that, although CFD provided good estimates of an average blending time, experimental blending times varied

  9. Hydraulic System Design of Hydraulic Actuators for Large Butterfly Valves

    Directory of Open Access Journals (Sweden)

    Ye HUANG

    2014-09-01

    Full Text Available Hydraulic control systems of butterfly valves are presently valve-controlled and pump-controlled. Valve-controlled hydraulic systems have serious power loss and generate much heat during throttling. Pump-controlled hydraulic systems have no overflow or throttling losses but are limited in the speed adjustment of the variable-displacement pump, generate much noise, pollute the environment, and have motor power that does not match load requirements, resulting in low efficiency under light loads and wearing of the variable-displacement pump. To overcome these shortcomings, this article designs a closed hydraulic control system in which an AC servo motor drives a quantitative pump that controls a spiral swinging hydraulic cylinder, and analyzes and calculates the structure and parameters of a spiral swinging hydraulic cylinder. The hydraulic system adjusts the servo motor’s speed according to the requirements of the control system, and the motor power matches the power provided to components, thus eliminating the throttling loss of hydraulic circuits. The system is compact, produces a large output force, provides stable transmission, has a quick response, and is suitable as a hydraulic control system of a large butterfly valve.

  10. Modernization of the turbine control technique and the turbine hydraulics aimed to improved maneuverability in the load range, system safety and plant availability, plant transparency for diagnosis and long-term performance

    International Nuclear Information System (INIS)

    Baran, Detlef

    2012-01-01

    In the contribution H.Mauell GmbH presents modernization projects for the nuclear power plants Tihange-3 and Doel-4. The project volume included control technique and the turbine hydraulics for the steam turbo generating set including turbine auxiliary devices and two turbine feeding pumps. The modernizations were successfully completed in 2010 and 2011, respectively. The nuclear power plants are trouble-free operated.

  11. A Study on thermal-hydraulic characteristics of the coolant materials for the transmutation reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Chang Hyun; You, Young Woo; Cho, Jae Seon; Kim, Ju Youl; Kim, Do Hyoung; Kim, Yoon Ik; Yang, Hui Chang [Seoul National University, Taejon (Korea)

    1998-03-01

    The objective of this study is to provide the direction of transmutation reactor design in terms of thermal hydraulics especially through the analysis of thermal hydraulic characteristics of various candidate materials for the transmutation reactor coolant. In this study, the characteristics of coolant materials used in current nuclear power plants and candidate materials for transmutation reactor are analyzed and compared. To evaluate the thermal hydraulic characteristics, the preliminary thermal-hydraulic calculation is performed for the candidate coolant materials of transmutation reactor. An analysis of thermal-hydraulic characteristics of transmutation reactor. An analysis of thermal-hydraulic characteristics of Sodium, Lead, Lead-Bismuth, and Lead-Lithium among the liquid metals considered as the coolant of transmutation reactor is performed by using computational fluid dynamics code FLUENT, and SIMPLER algorithm. (author). 50 refs., 40 figs., 30 tabs.

  12. In-place testing of hydraulic snubbers

    International Nuclear Information System (INIS)

    Raymont, J.M. Jr.

    1986-01-01

    Over the last few years, an increasing number of utilities have implemented periodic in-service inspection (ISI) programs of their hydraulic snubbers. This thrust has caused the nuclear power industry to seek cost-effective means of testing hydraulic snubbers. This paper reviews the following aspects of in-place testing and develops a technical justification for its use as a viable alternative to test bench testing. (1) A detailed examination of how in-place testing works is provided. Discussed are the hydraulic principles, fluid flow paths, and snubber test setup. (2) A comparison of the test bench and in-place test machines is provided. The discussion reviews the similarities and differences between the two test methods as well as the test results. (3) The need for correlation of in-place test results back to test bench data with a snubber footprint is discussed. (4) The issue of partial load testing with extrapolation to full load testing is discussed and compared with full load testing. The hydraulic principles as well as the costs and benefits of partial load versus full load testing are compared. (5) In-place test machine technology is reviewed. The operating principles, accuracies, and limitations are presented. (6) Actual test data are provided and reviewed on a test-by-test basis. (7) Lessons learned from actual in-place test jobs are reviewed. (8) In-place test procedures and calibration practices are outlined to illustrate the nature of the required planning on the part of the utility

  13. Historical perspective of thermal reactor safety in light water reactors

    International Nuclear Information System (INIS)

    Levy, S.

    1986-01-01

    A brief history of thermal reactor safety in U.S. light water reactors is provided in this paper. Important shortcomings in safety philosophy evolution versus time are identified and potential corrective actions are suggested. It should be recognized, that this analysis represents only one person's opinion and that most historical accountings reflect the author's biases and specific areas of knowledge. In that sense, many of the examples used in this paper are related to heat transfer and fluid flow safety issues, which explains why it has been included in a Thermal Hydraulics session. One additional note of caution: the value of hindsight and the selective nature of human memory when looking at the past cannot be overemphasized in any historical perspective

  14. Hydraulic shock damper for fuel assemblies of nuclear reactors

    International Nuclear Information System (INIS)

    Jabson, F.S.

    1978-01-01

    A typical embodiment of this invention provides a hydraulic mechanism for alleviating the effect of seismic forces and other stresses that are applied to a fuel assembly in a nuclear reactor. Illustratively, hollow guide posts potrude into a fuel assembly end fitting grid from biased spring pads. Plungers that move with the spring pads plug one end of each of the respective guide posts. Plates on the end fitting grid that have individual holes for fluid discharge partially plug the other ends of the respective guide posts, thereby providing a hydraulic means for absorbing the longitudinal component of seismic shocks and other anticipated forces. (Auth.)

  15. Fracture Evolution Following a Hydraulic Stimulation within an EGS Reservoir

    Energy Technology Data Exchange (ETDEWEB)

    Mella, Michael [Univ. of Utah, Salt Lake City, UT (United States). Energy and Geoscience Inst.

    2016-08-31

    The objective of this project was to develop and demonstrate an approach for tracking the evolution of circulation immediately following a hydraulic stimulation in an EGS reservoir. Series of high-resolution tracer tests using conservative and thermally reactive tracers were designed at recently created EGS reservoirs in order to track changes in fluid flow parameters such as reservoir pore volume, flow capacity, and effective reservoir temperature over time. Data obtained from the project would be available for the calibration of reservoir models that could serve to predict EGS performance following a hydraulic stimulation.

  16. Contribution to the study of thermal-hydraulic problems in nuclear reactors; Contribution a l`etude de problemes de thermohydraulique dans les reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Cognet, G

    1998-07-07

    In nuclear reactors, whatever the type considered, Pressurized Water Water Reactors (PWRs), Fast Breeder reactors (FBRs)..., thermal-hydraulics, the science of fluid mechanics and thermal behaviour, plays an essential role, both in nominal operating and accidental conditions. Fluid can either be the primary fluid (liquid or gas) or a very specific fluid called corium, which, in case of severe accident, could result from core and environning structure melting. The work reported here represents a 20-year contribution to thermal-hydraulic issues which could occur in FBRs and PWRs. Working on these two types of reactors, both in nominal and severe accident situations, has allowed me to compare the problems and to realize the importance of communication between research teams. The evolution in the complexity of studied problems, unavoidable in order to reduce costs and significantly improve safety, has led me from numerical modelling of single-phase flow turbulence to high temperature real melt experiments. The difficulties encountered in understanding the observed phenomena and in increasing experimental databases for computer code qualification have often entailed my participation in specific measurement device developments or adaptations, in particular non-intrusive devices generally based on optical techniques. Being concerned about the end-use of this research work, I actively participated in `in-situ` thermalhydraulic experiments in the FBRs: Phenix and Super-Phenix, of which I appreciated their undeniable scientific contribution. In my opinion, the thermal-hydraulic questions related to severe accidents are the most complex as they are at the cross-roads of several scientific specialities. Consequently, they require a multi-disciplinary approach and a continuous see-saw motion between experimentalists and modelling teams. After a brief description of the various problems encountered, the main ones are reported. Finally, the importance for research teams to

  17. Observation of ground deformation associated with hydraulic fracturing and seismicity in the Western Canadian Sedimentary Basin

    Science.gov (United States)

    Kubanek, J.; Liu, Y.; Harrington, R. M.; Samsonov, S.

    2017-12-01

    In North America, the number of induced earthquakes related to fluid injection due to the unconventional recovery of oil and gas resources has increased significantly within the last five years. Recent studies demonstrate that InSAR is an effective tool to study surface deformation due to large-scale wastewater injection, and highlight the value of surface deformation monitoring with respect to understanding evolution of pore pressure and stress at depth - vital parameters to forecast fault reactivation, and thus, induced earthquakes. In contrast to earthquakes related to the injection of large amounts of wastewater, seismic activity related to the hydraulic fracturing procedure itself was, until recently, considered to play a minor role without significant hazard. In the Western Canadian Sedimentary Basin (WCSB), however, Mw>4 earthquakes have recently led to temporary shutdown of industrial injection activity, causing multi-million dollar losses to operators and raising safety concerns with the local population. Recent studies successfully utilize seismic data and modeling to link seismic activity with hydraulic fracturing in the WCSB. Although the study of surface deformation is likely the most promising tool for monitoring integrity of a well and to derive potential signatures prior to moderate or large induced events, InSAR has, to date, not been utilized to detect surface deformation related to hydraulic fracturing and seismicity. We therefore plan to analyze time-series of SAR data acquired between 1991 to present over two target sites in the WCSB that will enable the study of long- and short-term deformation. Since the conditions for InSAR are expected to be challenging due to spatial and temporal decorrelation, we have designed corner reflectors that will be installed at one target site to improve interferometric performance. The corner reflectors will be collocated with broadband seismometers and Trimble SeismoGeodetic Systems that simultaneously measure

  18. Hydraulic Stability of Accropode Armour

    DEFF Research Database (Denmark)

    Jensen, T.; Burcharth, H. F.; Frigaard, Peter

    The present report describes the hydraulic model tests of Accropode armour layers carried out at the Hydraulics Laboratory at Aalborg University from November 1995 through March 1996. The objective of the model tests was to investigate the hydraulic stability of Accropode armour layers...... with permeable core (crushed granite with a gradation of 5-8 mm). The outcome of this study is described in "Hydraulic Stability of Single-Layer Dolos and Accropode Armour Layers" by Christensen & Burcharth (1995). In January/February 1996, Research Assistant Thomas Jensen carried out a similar study...

  19. Hydraulic fracturing proppants

    Directory of Open Access Journals (Sweden)

    V. P. P. de Campos

    Full Text Available Abstract Hydrocarbon reservoirs can be classified as unconventional or conventional depending on the oil and gas extraction difficulty, such as the need for high-cost technology and techniques. The hydrocarbon extraction from bituminous shale, commonly known as shale gas/oil, is performed by using the hydraulic fracturing technique in unconventional reservoirs where 95% water, 0.5% of additives and 4.5% of proppants are used. Environmental problems related to hydraulic fracturing technique and better performance/development of proppants are the current challenge faced by companies, researchers, regulatory agencies, environmentalists, governments and society. Shale gas is expected to increase USA fuel production, which triggers the development of new proppants and technologies of exploration. This paper presents a review of the definition of proppants, their types, characteristics and situation in the world market and information about manufacturers. The production of nanoscale materials such as anticorrosive and intelligent proppants besides proppants with carbon nanotubes is already carried out on a scale of tonnes per year in Belgium, Germany and Asia countries.

  20. Hydraulic jett mixing

    International Nuclear Information System (INIS)

    Ackerman, J.R.

    1989-01-01

    Efficient mixing of reactants into a waste stream has always been a problem in that there has been no mixer capable of combining all the elements of enhanced mixing into a single piece of equipment. Through the development of a mixing system for the mining industry to treat acid mine water containing heavy metals, a versatile new hydraulic jetting static mixer has been developed that has no moving parts and a clean bore with no internal components. This paper reports that the main goal of the development of the hydraulic jett mixer was to reduce the size of the tankage required for an acid mine drainage (AMD) treatment plant through development of a static mixing device that could coincidentally aerate the treatment flow. This process equipment being developed would simultaneously adjust the pH and oxidize the metals allowing formation of the hydroxide sludges required for sedimentation and removal of the metals from the treatment stream. In effect, the device eliminates two reaction tanks, the neutralization/mixing tank and the aeration tank

  1. Applied hydraulic transients

    CERN Document Server

    Chaudhry, M Hanif

    2014-01-01

    This book covers hydraulic transients in a comprehensive and systematic manner from introduction to advanced level and presents various methods of analysis for computer solution. The field of application of the book is very broad and diverse and covers areas such as hydroelectric projects, pumped storage schemes, water-supply systems, cooling-water systems, oil pipelines and industrial piping systems. Strong emphasis is given to practical applications, including several case studies, problems of applied nature, and design criteria. This will help design engineers and introduce students to real-life projects. This book also: ·         Presents modern methods of analysis suitable for computer analysis, such as the method of characteristics, explicit and implicit finite-difference methods and matrix methods ·         Includes case studies of actual projects ·         Provides extensive and complete treatment of governed hydraulic turbines ·         Presents design charts, desi...

  2. Thermo-hydraulic simulations of the experimental fast reactor core

    International Nuclear Information System (INIS)

    Silveira Luz, M. da; Braz Filho, F.A.; Borges, E.M.

    1985-01-01

    A study of the core and performance of metallic fuel of the experimental fast reactor, from the thermal-hydraulic point of view, was carried out employing the COBRA IV-I code. The good safety characteristics of this reactor and the feasibility of using metallic fuel in experimental fast reactor were demonstrated. (Author) [pt

  3. Technology and control for hydraulic manipulators

    International Nuclear Information System (INIS)

    Measson, Y.; David, O.; Louveau, F.; Friconneau, J.P.

    2003-01-01

    Hydraulic manipulators are candidate for fusion reactor maintenance. Their main advantages are their large payload with respect to volume and mass, their reliability and their robustness. However, due to their force control limitations, they are disqualified for precise manipulation and are dangerous for the environment and themselves in case of unexpected collision. CEA, in collaboration with CYBERNETIX and IFREMER has developed the advanced hydraulic robot MAESTRO. Force and hybrid control has been developed in order to avoid the previous problems. Using 'pressure' control servo-valve instead of the standard 'flow' control servo-valve (standard configuration of the MAESTRO) makes a real simplification of the control loop. No more pressure sensors are needed for monitoring the hydraulic joint in force control mode and using this kind of valves makes big safety improvements. The French company IN-LHC, designed and manufactured a prototype of servo-valve that fits the performances and space constraints of the Maestro arm. A characterisation of this new product was made on a mock-up and a set of these prototypes integrated in the Maestro slave-arm. A comparison between the two actuating technologies was made and showed that the performances of the pressure servo-valves make it applicable to general application

  4. Strategic need for a multi-purpose thermal hydraulic loop for support of advanced reactor technologies

    Energy Technology Data Exchange (ETDEWEB)

    O' Brien, James E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sabharwall, Piyush [Idaho National Lab. (INL), Idaho Falls, ID (United States); Yoon, Su -Jong [Idaho National Lab. (INL), Idaho Falls, ID (United States); Housley, Gregory K. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-09-01

    This report presents a conceptual design for a new high-temperature multi fluid, multi loop test facility for the INL to support thermal hydraulic, materials, and thermal energy storage research for nuclear and nuclear-hybrid applications. In its initial configuration, the facility will include a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX) and a secondary heat exchanger (SHX). Research topics to be addressed with this facility include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs) at prototypical operating conditions, flow and heat transfer issues related to core thermal hydraulics in advanced helium-cooled and salt-cooled reactors, and evaluation of corrosion behavior of new cladding materials and accident-tolerant fuels for LWRs at prototypical conditions. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST) facility. Research performed in this facility will advance the state of the art and technology readiness level of high temperature intermediate heat exchangers (IHXs) for nuclear applications while establishing the INL as a center of excellence for the development and certification of this technology. The thermal energy storage capability will support research and demonstration activities related to process heat delivery for a variety of hybrid energy systems and grid stabilization strategies. Experimental results obtained from this research will assist in development of reliable predictive models for thermal hydraulic design and safety codes over the range of expected advanced reactor operating conditions. Proposed/existing IHX heat transfer and friction correlations and criteria will be assessed with information on materials compatibility and instrumentation

  5. Proceedings of the 10. Meeting on Reactor Physics and Thermal Hydraulics

    International Nuclear Information System (INIS)

    Santos Bastos, W. dos

    1995-01-01

    These proceedings presents all the Meeting papers emphasizing specific aspects on reactor physics method, criticality, fuel management, nuclear data, safety analysis, simulation and shielding, neutronics, thermal hydraulics, reactor operation and computational methods

  6. Design, test and model of a hybrid magnetostrictive hydraulic actuator

    International Nuclear Information System (INIS)

    Chaudhuri, Anirban; Yoo, Jin-Hyeong; Wereley, Norman M

    2009-01-01

    The basic operation of hybrid hydraulic actuators involves high frequency bi-directional operation of an active material that is converted to uni-directional motion of hydraulic fluid using valves. A hybrid actuator was developed using magnetostrictive material Terfenol-D as the driving element and hydraulic oil as the working fluid. Two different lengths of Terfenol-D rod, 51 and 102 mm, with the same diameter, 12.7 mm, were used. Tests with no load and with load were carried out to measure the performance for uni-directional motion of the output piston at different pumping frequencies. The maximum no-load flow rates were 24.8 cm 3 s −1 and 22.7 cm 3 s −1 with the 51 mm and 102 mm long rods respectively, and the peaks were noted around 325 Hz pumping frequency. The blocked force of the actuator was close to 89 N in both cases. A key observation was that, at these high pumping frequencies, the inertial effects of the fluid mass dominate over the viscous effects and the problem becomes unsteady in nature. In this study, we also develop a mathematical model of the hydraulic hybrid actuator in the time domain to show the basic operational principle under varying conditions and to capture phenomena affecting system performance. Governing equations for the pumping piston and output shaft were obtained from force equilibrium considerations, while compressibility of the working fluid was taken into account by incorporating the bulk modulus. Fluid inertia was represented by a lumped parameter approach to the transmission line model, giving rise to strongly coupled ordinary differential equations. The model was then used to calculate the no-load velocities of the actuator at different pumping frequencies and simulation results were compared with experimental data for model validation

  7. Hydraulic Limits on Maximum Plant Transpiration

    Science.gov (United States)

    Manzoni, S.; Vico, G.; Katul, G. G.; Palmroth, S.; Jackson, R. B.; Porporato, A. M.

    2011-12-01

    Photosynthesis occurs at the expense of water losses through transpiration. As a consequence of this basic carbon-water interaction at the leaf level, plant growth and ecosystem carbon exchanges are tightly coupled to transpiration. In this contribution, the hydraulic constraints that limit transpiration rates under well-watered conditions are examined across plant functional types and climates. The potential water flow through plants is proportional to both xylem hydraulic conductivity (which depends on plant carbon economy) and the difference in water potential between the soil and the atmosphere (the driving force that pulls water from the soil). Differently from previous works, we study how this potential flux changes with the amplitude of the driving force (i.e., we focus on xylem properties and not on stomatal regulation). Xylem hydraulic conductivity decreases as the driving force increases due to cavitation of the tissues. As a result of this negative feedback, more negative leaf (and xylem) water potentials would provide a stronger driving force for water transport, while at the same time limiting xylem hydraulic conductivity due to cavitation. Here, the leaf water potential value that allows an optimum balance between driving force and xylem conductivity is quantified, thus defining the maximum transpiration rate that can be sustained by the soil-to-leaf hydraulic system. To apply the proposed framework at the global scale, a novel database of xylem conductivity and cavitation vulnerability across plant types and biomes is developed. Conductivity and water potential at 50% cavitation are shown to be complementary (in particular between angiosperms and conifers), suggesting a tradeoff between transport efficiency and hydraulic safety. Plants from warmer and drier biomes tend to achieve larger maximum transpiration than plants growing in environments with lower atmospheric water demand. The predicted maximum transpiration and the corresponding leaf water

  8. Preliminary assessment of water-based nano-fluids for use as coolants in PWRs

    International Nuclear Information System (INIS)

    Jacopo Buongiorno

    2005-01-01

    Full text of publication follows: The impact of using water-based fluids with small additions (<2% vol.) of nano-sized (10-100 nm) particle populations as coolants for current and advanced PWRs is evaluated. Such 'engineered' fluids (known as nano-fluids) are attractive because the presence of the nano-particles enhances energy transport considerably. As a result, nano-fluids are known to have (i) higher thermal conductivity than water (up to 20% depending on nano-particle material, size and volumetric fraction), (ii) higher heat transfer coefficients (up to 40%), (iii) higher CHF (up to 300% in pool boiling), and (iv) comparable pressure drop. Furthermore, nano-fluids appear to be very stable suspensions with little or no sedimentation, because of the small size of the dispersed particles and their typically low volumetric fractions. The ultimate objective of this work is to assess whether existing PWRs could be retro-fitted with a water-based nano-fluid coolant, to increase safety margins, reduce stored energy, and/or allow for power up-rates. Also, advanced PWRs could be designed with nano-fluids. The linear heat generation rate in PWRs is limited by a) fuel centerline melting, b) cladding overheating (CHF), and c) stored energy release following a large-break LOCA. Mechanisms b) and c) are usually the most limiting. For given geometry and linear power, it is obvious that the core with the nano-fluid coolant will have higher margins to CHF and LOCA limits. Conversely, for given margins, a higher linear power can be accommodated by the nano-fluid-cooled core. Standard thermal-hydraulic models for the PWR hot fuel pin (including a RELAP model for the LOCA) have been used to quantify the benefit of using nano-fluid coolants on the performance of a PWR. (author)

  9. Slip flow coefficient analysis in water hydraulics gear pump for environmental friendly application

    International Nuclear Information System (INIS)

    Yusof, A A; Wasbari, F; Zakaria, M S; Ibrahim, M Q

    2013-01-01

    Water hydraulics is the sustainable option in developing fluid power systems with environmental friendly approach. Therefore, an investigation on water-based external gear pump application is being conducted, as a low cost solution in the shifting effort of using water, instead of traditional oil hydraulics in fluid power application. As the gear pump is affected by fluid viscosity, an evaluation has been conducted on the slip flow coefficient, in order to understand to what extent the spur gear pump can be used with water-based hydraulic fluid. In this paper, the results of a simulated study of variable-speed fixed displacement gear pump are presented. The slip flow coefficient varies from rotational speed of 250 RPM to 3500 RPM, and provides volumetric efficiency ranges from 9 % to 97% accordingly

  10. Thermal reactor safety

    International Nuclear Information System (INIS)

    1980-06-01

    Information is presented concerning new trends in licensing; seismic considerations and system structural behavior; TMI-2 risk assessment and thermal hydraulics; statistical assessment of potential accidents and verification of computational methods; issues with respect to improved safety; human factors in nuclear power plant operation; diagnostics and activities in support of recovery; LOCA transient analysis; unresolved safety issues and other safety considerations; and fission product transport

  11. Thermal reactor safety

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-01

    Information is presented concerning new trends in licensing; seismic considerations and system structural behavior; TMI-2 risk assessment and thermal hydraulics; statistical assessment of potential accidents and verification of computational methods; issues with respect to improved safety; human factors in nuclear power plant operation; diagnostics and activities in support of recovery; LOCA transient analysis; unresolved safety issues and other safety considerations; and fission product transport.

  12. Cortical gluing and Ringer lactate solution inflation to avoid cortical mantle collapse and subdural fluid collections in pediatric neurosurgery: safety and feasibility.

    Science.gov (United States)

    Mirone, Giuseppe; Ruggiero, Claudio; Spennato, Pietro; Aliberti, Ferdinando; Trischitta, Vincenzo; Cinalli, Giuseppe

    2015-06-01

    Subdural fluid collections following intraventricular and/or paraventricular procedures in pediatric neurosurgery are common and can be hard to treat. We describe our technique to close cortical defects by the aid of a fibrin adhesive and subsequent Ringer inflation with the aim to avoid cortical mantle collapse and to prevent the development of subdural fluid collections. We report the preliminary results of a prospective study on a consecutive series of 29 children who underwent 37 transcortical or transcallosal surgical procedures since 2008 in our department. In 17 procedures, we performed a transcortical approach on lesions, and in other 19 operations, we operated by a transcallosal. In 5/17 transcortical approaches (29%) and in 3/20 transcallosal approaches (15%), we observed a 5-mm-thick subdural fluid collection of the 5 patients with subdural fluid collections in the transcortical group, 3 patients (17%) underwent surgery for symptomatic or progressive subdural fluid collections. Of the 3 patients in the transcallosal group, a subduro-peritoneal shunt was necessary only for 1 patient (5%). At the very end of the treatment (including chemotherapy and radiotherapy), it was possible to remove the subduro-peritoneal shunt in all these patients because of disappearance of the subdural fluid collections. In pediatric patients after transcortical or transcallosal procedures, the use of a fibrin adhesive to seal surgical opening and subsequent inflation of the residual cavity with Ringer lactate solution to avoid cortical mantle collapse seems safe and appears to prevent the development of subdural fluid collections.

  13. Economic and hydraulic divergences underpin ecological differentiation in the Bromeliaceae.

    Science.gov (United States)

    Males, Jamie; Griffiths, Howard

    2018-01-01

    Leaf economic and hydraulic theories have rarely been applied to the ecological differentiation of speciose herbaceous plant radiations. The role of character trait divergences and network reorganization in the differentiation of the functional types in the megadiverse Neotropical Bromeliaceae was explored by quantifying a range of leaf economic and hydraulic traits in 50 diverse species. Functional types, which are defined by combinations of C 3 or Crassulacean acid metabolism (CAM) photosynthesis, terrestrial or epiphytic habits, and non-specialized, tank-forming or atmospheric morphologies, segregated clearly in trait space. Most classical leaf economic relationships were supported, but they were weakened by the presence of succulence. Functional types differed in trait-network architecture, suggesting that rewiring of trait-networks caused by innovations in habit and photosynthetic pathway is an important aspect of ecological differentiation. The hydraulic data supported the coupling of leaf hydraulics and gas exchange, but not the hydraulic safety versus efficiency hypothesis, and hinted at an important role for the extra-xylary compartment in the control of bromeliad leaf hydraulics. Overall, our findings highlight the fundamental importance of structure-function relationships in the generation and maintenance of ecological diversity. © 2017 The Authors Plant, Cell & Environment Published by John Wiley & Sons Ltd.

  14. Process of preparing hydraulic cement

    Energy Technology Data Exchange (ETDEWEB)

    1919-12-11

    A process of preparing hydraulic cement from oil shale or shale coke is characterized in that the oil shale or shale coke after the distillation is burned long and hot to liberate the usual amount of carbonic acid and then is fine ground to obtain a slow hardening hydraulic cement.

  15. Control rod drive hydraulic device

    International Nuclear Information System (INIS)

    Takekawa, Toru.

    1994-01-01

    The device of the present invention can reliably prevent a possible erroneous withdrawal of control rod driving mechanism when the pressure of a coolant line is increased by isolation operation of hydraulic control units upon periodical inspection for a BWR type reactor. That is, a coolant line is connected to the downstream of a hydraulic supply device. The coolant line is connected to a hydraulic control unit. A coolant hydraulic detection device and a pressure setting device are disposed to the coolant line. A closing signal line and a returning signal line are disposed, which connect the hydraulic supply device and a flow rate control valve for the hydraulic setting device. In the device of the present invention, even if pressure of supplied coolants is elevated due to isolation of hydraulic control units, the elevation of the hydraulic pressure can be prevented. Accordingly, reliability upon periodical reactor inspection can be improved. Further, the facility is simplified and the installation to an existent facility is easy. (I.S.)

  16. Dynamic Characteristics of Communication Lines with Distributed Parameters to Control the Throttle-controlled Hydraulic Actuators

    Directory of Open Access Journals (Sweden)

    D. N. Popov

    2015-01-01

    Full Text Available The article considers a mathematical model of the hydraulic line for remote control of electro-hydraulic servo drive (EHSD with throttle control. This type of hydraulic lines is designed as a backup to replace the electrical connections, which are used to control EHSD being remote from the site with devices located to form the control signals of any object. A disadvantage of electric connections is that they are sensitive to magnetic fields and thereby do not provide the required reliability of the remote control. Hydraulic lines have no this disadvantage and therefore are used in aircraft and other industrial systems. However, dynamic characteristics of hydraulic systems still have been investigated insufficiently in the case of transmitting control signals at a distance at which the signal may be distorted when emerging the wave processes.The article results of mathematical simulation, which are verified through physical experimentation, largely eliminate the shortcomings of said information.The mathematical model described in the paper is based on the theory of unsteady pressure compressible fluids. In the model there are formulas that provide calculation of frequency characteristics of the hydraulic lines under hydraulic oscillations of the laminar flow parameters of viscous fluid.A real mock-up of the system under consideration and an experimental ad hoc unit are used to verify the results of mathematically simulated hydraulic systems.Calculated logarithmic amplitude and phase frequency characteristics compared with those obtained experimentally prove, under certain conditions, the proposed theoretical method of calculation. These conditions have to ensure compliance with initial parameters of fluid defined under stationary conditions. The applied theory takes into consideration a non-stationary hydraulic resistance of the line when calculating frequency characteristics.The scientific novelty in the article material is presented in

  17. Current and anticipated uses of thermal hydraulic codes in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyung-Doo; Chang, Won-Pyo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-07-01

    In Korea, the current uses of thermal hydraulic codes are categorized into 3 areas. The first application is in designing both nuclear fuel and NSSS. The codes have usually been introduced based on the technology transfer programs agreed between KAERI and the foreign vendors. Another area is in the supporting of the plant operations and licensing by the utility. The third category is research purposes. In this area assessments and some applications to the safety issue resolutions are major activities using the best estimate thermal hydraulic codes such as RELAP5/MOD3 and CATHARE2. Recently KEPCO plans to couple thermal hydraulic codes with a neutronics code for the design of the evolutionary type reactor by 2004. KAERI also plans to develop its own best estimate thermal hydraulic code, however, application range is different from KEPCO developing code. Considering these activities, it is anticipated that use of the best estimate hydraulic analysis code developed in Korea may be possible in the area of safety evaluation within 10 years.

  18. Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Baek, W. P.; Song, C. H.; Kim, Y. S. and others

    2005-02-15

    The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform various integral effect tests for design, operation, and safety regulation of pressurized water reactors. During the first phase of this project (1997.8{approx}2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished: a full-height, 1/300-volume-scaled full pressure facility for APR1400, an evolutionary pressurized water reactor that was developed by Korean industry. Main objectives of the present phase (2002.4{approx}2005.2), was to optimize the facility design and to construct the experimental facility. We have performed following researches: 1) Optimization of the basic design of the thermal-hydraulic integral effect test facility for PWRs - ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) - Reduced height design for APR1400 (+ specific design features of KSNP safety injection systems) - Thermal-hydraulic scaling based on three-level scaling methodology by Ishii et al. 2) Construction of the ATLAS facility - Detailed design of the test facility - Manufacturing and procurement of components - Installation of the facility 3) Development of supporting technology for integral effect tests - Development and application of advanced instrumentation technology - Preliminary analysis of test scenarios - Development of experimental procedures - Establishment and implementation of QA system/procedure.

  19. Study on the application of energy storage system in offshore wind turbine with hydraulic transmission

    International Nuclear Information System (INIS)

    Fan, Yajun; Mu, Anle; Ma, Tao

    2016-01-01

    Highlights: • Hydraulic offshore wind turbine is capable of outputting near constant power. • Open loop hydraulic transmission uses seawater as the working fluid. • Linear control strategy distributes total flow according to demand and supply. • Constant pressure hydraulic accumulator stores/releases the surplus energy. • Simulations show the dynamic performance of the hybrid system. - Abstract: A novel offshore wind turbine comprising fluid power transmission and energy storage system is proposed. In this wind turbine, the conventional mechanical transmission is replaced by an open-loop hydraulic system, in which seawater is sucked through a variable displacement pump in nacelle connected directly with the rotor and utilized to drive a Pelton turbine installed on the floating platform. Aiming to smooth and stabilize the output power, an energy storage system with the capability of flexible charging and discharging is applied. The related mathematical model is developed, which contains some sub-models that are categorized as the wind turbine rotor, hydraulic pump, transmission pipeline, proportional valve, accumulator and hydraulic turbine. A linear control strategy is adopted to distribute the flow out of the proportional valve through comparing the demand power with captured wind energy by hydraulic pump. Ultimately, two time domain simulations demonstrate the operation of the hybrid system when the hydraulic accumulator is utilized and show how this system can be used for load leveling and stabilizing the output power.

  20. Wind tunnel experiments to prove a hydraulic passive rotor speed control concept for variable speed wind turbines (poster)

    NARCIS (Netherlands)

    Diepeveen, N.F.B.; Jarquin Laguna, A.

    2012-01-01

    As alternative to geared and direct drive solutions, fluid power drive trains are being developed by several institutions around the world. The common configuration is where the wind turbine rotor is coupled to a hydraulic pump. The pump is connected through a high pressure line to a hydraulic motor