WorldWideScience

Sample records for hydraulic fluid analysis

  1. 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.

  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. 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)

  4. 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.

  5. 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

  6. 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.

  7. 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...

  8. 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.

  9. 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

  10. 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.)

  11. 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

  12. 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.

  13. 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.

  14. 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

  15. 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

  16. 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)

  17. 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

  18. 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

  19. 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.

  20. 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.

  1. 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)

  2. 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.

  3. 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...

  4. 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.

  5. 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.)

  6. 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

  7. 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...

  8. 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)

  9. 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

  10. 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.

  11. Synovial fluid analysis

    Science.gov (United States)

    Joint fluid analysis; Joint fluid aspiration ... El-Gabalawy HS. Synovial fluid analysis, synovial biopsy, and synovial pathology. In: Firestein GS, Budd RC, Gabriel SE, McInnes IB, O'Dell JR, eds. Kelly's Textbook of ...

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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....

  19. Synovial Fluid Analysis

    Science.gov (United States)

    ... Plasma Free Metanephrines Platelet Count Platelet Function Tests Pleural Fluid Analysis PML-RARA Porphyrin Tests Potassium Prealbumin ... is being tested? Synovial fluid is a thick liquid that acts as a lubricant for the body's ...

  20. 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

  1. 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

  2. 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.

  3. 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.

  4. 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)

  5. 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.)

  6. [Diagnosis: synovial fluid analysis].

    Science.gov (United States)

    Gallo Vallejo, Francisco Javier; Giner Ruiz, Vicente

    2014-01-01

    Synovial fluid analysis in rheumatological diseases allows a more accurate diagnosis in some entities, mainly infectious and microcrystalline arthritis. Examination of synovial fluid in patients with osteoarthritis is useful if a differential diagnosis will be performed with other processes and to distinguish between inflammatory and non-inflammatory forms. Joint aspiration is a diagnostic and sometimes therapeutic procedure that is available to primary care physicians. Copyright © 2014 Elsevier España, S.L. All rights reserved.

  7. Analysis of uncertainties of thermal hydraulic calculations

    International Nuclear Information System (INIS)

    Macek, J.; Vavrin, J.

    2002-12-01

    In 1993-1997 it was proposed, within OECD projects, that a common program should be set up for uncertainty analysis by a probabilistic method based on a non-parametric statistical approach for system computer codes such as RELAP, ATHLET and CATHARE and that a method should be developed for statistical analysis of experimental databases for the preparation of the input deck and statistical analysis of the output calculation results. Software for such statistical analyses would then have to be processed as individual tools independent of the computer codes used for the thermal hydraulic analysis and programs for uncertainty analysis. In this context, a method for estimation of a thermal hydraulic calculation is outlined and selected methods of statistical analysis of uncertainties are described, including methods for prediction accuracy assessment based on the discrete Fourier transformation principle. (author)

  8. 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.

  9. 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.

  10. 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)

  11. 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.

  12. 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

  13. 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...

  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. 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.)

  16. 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.

  17. 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

  18. Comparison for the interfacial and wall friction models in thermal-hydraulic system analysis codes

    International Nuclear Information System (INIS)

    Hwang, Moon Kyu; Park, Jee Won; Chung, Bub Dong; Kim, Soo Hyung; Kim, See Dal

    2007-07-01

    The average equations employed in the current thermal hydraulic analysis codes need to be closed with the appropriate models and correlations to specify the interphase phenomena along with fluid/structure interactions. This includes both thermal and mechanical interactions. Among the closure laws, an interfacial and wall frictions, which are included in the momentum equations, not only affect pressure drops along the fluid flow, but also have great effects for the numerical stability of the codes. In this study, the interfacial and wall frictions are reviewed for the commonly applied thermal-hydraulic system analysis codes, i.e. RELAP5-3D, MARS-3D, TRAC-M, and CATHARE

  19. 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

  20. 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)

  1. 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...

  2. SBWR core thermal hydraulic analysis during startup

    International Nuclear Information System (INIS)

    Lin, J.H.; Huang, R.L.; Sawyer, C.D.

    1993-01-01

    This paper reports on a thermal hydraulic analysis of the SIMPLIFIED BOILING WATER REACTOR (SBWR) during startup. The potential instability during a SBWR startup has drawn the attention of designers, researchers, and engineers. It has not been a concern for a Boiling Water Reactor (BWR) with forced recirculation; however, for SBWR with natural circulation the concern exists. The concern is about the possibility of a geysering mode oscillation during SBWR startup from a cold temperature and a low system pressure with a low natural circulation flow rate. A thermal hydraulic analysis of the SBWR is performed in simulation of the startup using the TRACG computer code. The temperature, pressure, and reactor power profiles of SBWR during the startup are presented. The results are compared with the data of a natural circulation boiling water reactor, the DODEWAARD plant, in which no instabilities have been observed during many startups. It is shown that a SBWR startup which follows proper procedures, geysering and other modes of oscillations can be avoided

  3. Validation of the TEXSAN thermal-hydraulic analysis program

    International Nuclear Information System (INIS)

    Burns, S.P.; Klein, D.E.

    1992-01-01

    The TEXSAN thermal-hydraulic analysis program has been developed by the University of Texas at Austin (UT) to simulate buoyancy driven fluid flow and heat transfer in spent fuel and high level nuclear waste (HLW) shipping applications. As part of the TEXSAN software quality assurance program, the software has been subjected to a series of test cases intended to validate its capabilities. The validation tests include many physical phenomena which arise in spent fuel and HLW shipping applications. This paper describes some of the principal results of the TEXSAN validation tests and compares them to solutions available in the open literature. The TEXSAN validation effort has shown that the TEXSAN program is stable and consistent under a range of operating conditions and provides accuracy comparable with other heat transfer programs and evaluation techniques. The modeling capabilities and the interactive user interface employed by the TEXSAN program should make it a useful tool in HLW transportation analysis

  4. 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.

  5. 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.

  6. 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...

  7. 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

  8. 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

  9. 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.

  10. 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...

  11. Development of realistic thermal-hydraulic system analysis codes ; development of thermal hydraulic test requirements for multidimensional flow modeling

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Kune Yull; Yoon, Sang Hyuk; Noh, Sang Woo; Lee, Il Suk [Seoul National University, Seoul (Korea)

    2002-03-01

    This study is concerned with developing a multidimensional flow model required for the system analysis code MARS to more mechanistically simulate a variety of thermal hydraulic phenomena in the nuclear stem supply system. The capability of the MARS code as a thermal hydraulic analysis tool for optimized system design can be expanded by improving the current calculational methods and adding new models. In this study the relevant literature was surveyed on the multidimensional flow models that may potentially be applied to the multidimensional analysis code. Research items were critically reviewed and suggested to better predict the multidimensional thermal hydraulic behavior and to identify test requirements. A small-scale preliminary test was performed in the downcomer formed by two vertical plates to analyze multidimensional flow pattern in a simple geometry. The experimental result may be applied to the code for analysis of the fluid impingement to the reactor downcomer wall. Also, data were collected to find out the controlling parameters for the one-dimensional and multidimensional flow behavior. 22 refs., 40 figs., 7 tabs. (Author)

  12. Peritoneal Fluid Analysis

    Science.gov (United States)

    ... Get Tested? To help diagnose the cause of peritonitis, an inflammation of the membrane lining the abdomen, ... fever and your healthcare practitioner suspects you have peritonitis or ascites Sample Required? A peritoneal fluid sample ...

  13. 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.

  14. 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.

  15. 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.

  16. Efficiency limit factor analysis for the Francis-99 hydraulic turbine

    Science.gov (United States)

    Zeng, Y.; Zhang, L. X.; Guo, J. P.; Guo, Y. K.; Pan, Q. L.; Qian, J.

    2017-01-01

    The energy loss in hydraulic turbine is the most direct factor that affects the efficiency of the hydraulic turbine. Based on the analysis theory of inner energy loss of hydraulic turbine, combining the measurement data of the Francis-99, this paper calculates characteristic parameters of inner energy loss of the hydraulic turbine, and establishes the calculation model of the hydraulic turbine power. Taken the start-up test conditions given by Francis-99 as case, characteristics of the inner energy of the hydraulic turbine in transient and transformation law are researched. Further, analyzing mechanical friction in hydraulic turbine, we think that main ingredients of mechanical friction loss is the rotation friction loss between rotating runner and water body, and defined as the inner mechanical friction loss. The calculation method of the inner mechanical friction loss is given roughly. Our purpose is that explore and research the method and way increasing transformation efficiency of water flow by means of analysis energy losses in hydraulic turbine.

  17. 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.)

  18. QAPP for Hydraulic Fracturing (HF) Surface Spills Data Analysis

    Science.gov (United States)

    This QAPP provides information concerning the analysis of spills associated with hydraulic fracturing. This project is relevant to both the chemical mixing and flowback and produced water stages of the HF water cycle as found in the HF Study Plan.

  19. CFD thermal-hydraulic analysis of a CANDU fuel channel

    International Nuclear Information System (INIS)

    Catana, A.; Prisecaru, I.; Dupleac, D.; Danila, N.

    2009-01-01

    This paper presents the numerical investigation of a CANDU fuel channel using CFD (Computational fluid dynamics) methodology approach. Limited computer power available at Bucharest University POLITEHNICA forced the authors to analyse only segments of fuel channel namely the significant ones: fuel bundle junctions with adjacent segments, fuel bundle spacer planes with adjacent segments, regular segments of fuel bundles. The computer code used is FLUENT. Fuel bundles contained in pressure tubes forms a complex flow domain. The flow is characterized by high turbulence and in some parts of fuel channel also by multi-phase flow. The flow in the fuel channel has been simulated by solving the equations for conservation of mass and momentum. For turbulence modelling the standard k-e model is employed although other turbulence models can be used as well. In this paper we do not consider heat generation and heat transfer capabilities of CFD methods. Since we consider only some relatively short segments of a CANDU fuel channel we can assume, for this starting stage, that heat transfer is not very important for these short segments of fuel channel. The boundary conditions for CFD analysis are provided by system and sub-channel analysis. In this paper the discussion is focused on some flow parameters behaviour at the bundle junction, spacer's plane configuration, etc. In this paper we present results for Standard CANDU 6 Fuel Bundles as a basis for CFD thermal-hydraulic analysis of INR proposed SEU43 and other new nuclear fuels. (authors)

  20. 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.

  1. 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.

  2. 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)

  3. Analysis of hydraulic bearing effect for vertical-shaft pump

    International Nuclear Information System (INIS)

    Narabayashi, Tadashi; Mawatari, Katsuhiko; Uchida, Ken; Iikura, Takahiko; Hayakawa, Kiyoshi

    1999-01-01

    In inner-rotating non coaxial cylinders, axial flow causes a hydraulic being effect by which the inner cylinder is put at the center of the axis of the outer cylinder, because of the pressure distribution along the surface of the inner cylinder. When the rotating speed becomes higher, whirl force is generated by the pressure distribution in the narrow gap side. Therefore, pocket-type hydraulic being was added between the rotor and the wearing, based on an experiment and flow analysis. The pockets suck a part of discharged water of a pump and pressurize a water along the rotational direction in the pocket. The pressurized water enhance the hydraulic being effect. The analysis results showed good agreement with the experiments, and the analysis method for the hydraulic being for vertical-shaft pump was established. (author)

  4. 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

  5. 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.)

  6. Thermal-hydraulic analysis of nuclear reactors

    CERN Document Server

    Zohuri, Bahman

    2015-01-01

    This text covers the fundamentals of thermodynamics required to understand electrical power generation systems and the application of these principles to nuclear reactor power plant systems. It is not a traditional general thermodynamics text, per se, but a practical thermodynamics volume intended to explain the fundamentals and apply them to the challenges facing actual nuclear power plants systems, where thermal hydraulics comes to play.  Written in a lucid, straight-forward style while retaining scientific rigor, the content is accessible to upper division undergraduate students and aimed at practicing engineers in nuclear power facilities and engineering scientists and technicians in industry, academic research groups, and national laboratories. The book is also a valuable resource for students and faculty in various engineering programs concerned with nuclear reactors. This book also: Provides extensive coverage of thermal hydraulics with thermodynamics in nuclear reactors, beginning with fundamental ...

  7. 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

  8. 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

  9. 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

  10. 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

  11. 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 (

  12. 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.

  13. 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

  14. 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

  15. Analysis of hydraulic characteristics for stream diversion in small stream

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Sang-Jin; Jun, Kye-Won [Chungbuk National University, Cheongju(Korea)

    2001-10-31

    This study is the analysis of hydraulic characteristics for stream diversion reach by numerical model test. Through it we can provide the basis data in flood, and in grasping stream flow characteristics. Analysis of hydraulic characteristics in Seoknam stream were implemented by using computer model HEC-RAS(one-dimensional model) and RMA2(two-dimensional finite element model). As a result we became to know that RMA2 to simulate left, main channel, right in stream is more effective method in analysing flow in channel bends, steep slope, complex bed form effect stream flow characteristics, than HEC-RAS. (author). 13 refs., 3 tabs., 5 figs.

  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.

    2018-01-01

    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.

  17. Horizontal steam generator PGV-1000 thermal-hydraulic analysis

    Energy Technology Data Exchange (ETDEWEB)

    Ubra, O. [Skoda Company, Prague (Switzerland); Doubek, M. [Czech Technical Univ., Prague (Switzerland)

    1995-12-31

    A computer program for the steady state thermal-hydraulic analysis of horizontal steam generator PGV-1000 is presented. The program provides the capability to analyze steam generator PGV-1000 primary side flow and temperature distribution, primary side pressure drops, heat transfer between the primary and secondary sides and multidimensional heat flux distribution. A special attention is paid to the thermal-hydraulics of the secondary side. The code predicts 3-D distribution of the void fraction at the secondary side, mass redistribution under the submerged perforated sheet and the steam generator level profile. By means of developed computer program a detailed thermal-hydraulic study of the PGV-1000 has been carried out. A wide range of calculations has been performed and a set of important steam generator characteristics has been obtained. Some of them are presented in the paper. (orig.). 5 refs.

  18. Dynamic Analysis & Characterization of Conventional Hydraulic Power Supply Units

    DEFF Research Database (Denmark)

    Schmidt, Lasse; Liedhegener, Michael; Bech, Michael Møller

    2016-01-01

    Hydraulic power units operated as constant supply pres-sure systems remain to be widely used in the industry, to supply valve controlled hydraulic drives etc., where the hydraulic power units are constituted by variable pumps with mechanical outlet pressure control, driven by induction motors....... In the analysis of supplied drives, both linear and rotary, emphasis is commonly placed on the drives themselves and the related loads, and the supply system dynamics is often given only little attention, and usually neglected or taken into account in a simplified fashion. The simplified supply system dynamics...... and drives will reduce the flow-to-pressure gain of the supply system, and hence increase the time constant of the sup-ply pressure dynamics. A consequence of this may be large vari-ations in the supply pressure, hence large variations in the pump shaft torque, and thereby the induction motor load torque...

  19. Horizontal steam generator PGV-1000 thermal-hydraulic analysis

    International Nuclear Information System (INIS)

    Ubra, O.; Doubek, M.

    1995-01-01

    A computer program for the steady state thermal-hydraulic analysis of horizontal steam generator PGV-1000 is presented. The program provides the capability to analyze steam generator PGV-1000 primary side flow and temperature distribution, primary side pressure drops, heat transfer between the primary and secondary sides and multidimensional heat flux distribution. A special attention is paid to the thermal-hydraulics of the secondary side. The code predicts 3-D distribution of the void fraction at the secondary side, mass redistribution under the submerged perforated sheet and the steam generator level profile. By means of developed computer program a detailed thermal-hydraulic study of the PGV-1000 has been carried out. A wide range of calculations has been performed and a set of important steam generator characteristics has been obtained. Some of them are presented in the paper. (orig.)

  20. Horizontal steam generator PGV-1000 thermal-hydraulic analysis

    Energy Technology Data Exchange (ETDEWEB)

    Ubra, O [Skoda Company, Prague (Switzerland); Doubek, M [Czech Technical Univ., Prague (Switzerland)

    1996-12-31

    A computer program for the steady state thermal-hydraulic analysis of horizontal steam generator PGV-1000 is presented. The program provides the capability to analyze steam generator PGV-1000 primary side flow and temperature distribution, primary side pressure drops, heat transfer between the primary and secondary sides and multidimensional heat flux distribution. A special attention is paid to the thermal-hydraulics of the secondary side. The code predicts 3-D distribution of the void fraction at the secondary side, mass redistribution under the submerged perforated sheet and the steam generator level profile. By means of developed computer program a detailed thermal-hydraulic study of the PGV-1000 has been carried out. A wide range of calculations has been performed and a set of important steam generator characteristics has been obtained. Some of them are presented in the paper. (orig.). 5 refs.

  1. Probability analysis of dynamical effects of axial piston hydraulic motor

    OpenAIRE

    Sapietova Alzbeta; Dekys Vladimír; Sapieta Milan; Sulka Peter; Gajdos Lukas; Rojek Izabela

    2018-01-01

    The paper presents an analysis of impact force on stopper screw in axial piston hydraulic motor. The solution contains probabilistic description of input variables. If the output parameters of probabilistic solution are compared with arbitrary values and values acquired by analytical solution, the probability of proper operation of the device can be evaluated.

  2. 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

  3. 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)

  4. 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)

  5. Compatibility analysis of DUPIC fuel(4) - thermal hydraulic analysis

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jee Won; Chae, Kyung Myung; Choi, Hang Bok

    2000-07-01

    Thermal-hydraulic compatibility of the DUPIC fuel bundle in the CANDU reactor has been studied. The critical channel power, the critical power ratio, the channel exit quality and the channel flow are calculated for the DUPIC and the standard fuels by using the NUCIRC code. The physical models and associated parametric values for the NUCIRC analysis of the fuels are also presented. Based upon the slave channel analysis, the critical channel power and the critical power ratios have been found to be very similar for the two fuel types. The same dryout model is used in this study for the standard and the DUPIC fuel bundles. To assess the dryout characteristics of the DUPIC fuel bundle, the ASSERT-PV code has been used for the subchannel analysis. Based upon the results of the subchannel analysis, it is found that the dryout location and the power for the two fuel types are indeed very similar. This study shows that thermal performance of the DUPIC fuel is not significantly different from that of the standard fuel.

  6. 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

  7. 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.

  8. 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)

  9. 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.

  10. 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.

  11. 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.

  12. Shock analysis on hydraulic drive control rod during scram

    International Nuclear Information System (INIS)

    Song Wei; Qin Benke; Bo Hanliang

    2013-01-01

    Control rod hydraulic drive mechanism (CRHDM) is a new invention of Institute of Nuclear and New Energy Technology of Tsinghua University. The hydraulic absorber buffers the control rod when it scrams. The control rod fast drop impact experiment was conducted and the key parameters of control rod hydraulic buffering performance were obtained. Based on the test results and according to D'Alembert principle, the maximum inertial impact force on the control rod during the fast drop period was applied as equivalent static load force on the control rod. The deformations and stress distributions on the control rod in this worst case were calculated by using finite element software ABAQUS. Calculation results were compared with the experiment results, and it was verified that nonlinear transient dynamics analysis in this problem can be simplified as static analysis. Damage criterion of the control rod fast drop impact process was also given. And it lays foundation for optimal design of the control rod and hydraulic absorber. (authors)

  13. 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

  14. Thermal-hydraulic analysis for wire-wrapped PWR cores

    Energy Technology Data Exchange (ETDEWEB)

    Diller, P. [General Electric Company, 3901 Castle Hayne Rd., Wilmington, NC 28401 (United States)], E-mail: pdiller@gmail.com; Todreas, N. [Massachusetts Institute of Technology, Cambridge, MA 02139 (United States)], E-mail: todreas@mit.edu; Hejzlar, P. [Massachusetts Institute of Technology, Cambridge, MA 02139 (United States)

    2009-08-15

    This work focuses on the steady-state and transient thermal-hydraulic analyses for PWR cores using wire wraps in a hexagonal array with either U (45% w/o)-ZrH{sub 1.6} (referred to as U-ZrH{sub 1.6}) or UO{sub 2} fuels. Equivalences (thermal-hydraulic and neutronic) were created between grid spacer and wire wrap designs, and were used to apply results calculated for grid spacers to wire wrap designs. Design limits were placed on the pressure drop, critical heat flux (CHF), fuel and cladding temperature and vibrations. The vibrations limits were imposed for flow-induced vibrations (FIV) and thermal-hydraulic vibrations (THV). The transient analysis examined an overpower accident, loss of coolant accident (LOCA) and loss of flow accident (LOFA). The thermal-hydraulic performance of U-ZrH{sub 1.6} and UO{sub 2} were found very similar. Relative to grid spacer designs, wire wrap designs were found to have smaller fretting wear, substantially lower pressure drop and higher CHF. As a result, wire wrap cores were found to offer substantially higher maximum powers than grid spacer cores, allowing for a 25% power increase relative to the grid spacer uprate [Shuffler, C.A., Malen, J.A., Trant, J.M., Todreas, N.E., 2009a. Thermal-hydraulic analysis for grid supported and inverted fueled PWR cores. Nuclear Technology (this special issue devoted to hydride fuel in LWRs)] and a 58% power increase relative to the reference core.

  15. 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

  16. Analysis of slug tests in formations of high hydraulic conductivity.

    Science.gov (United States)

    Butler, James J; Garnett, Elizabeth J; Healey, John M

    2003-01-01

    A new procedure is presented for the analysis of slug tests performed in partially penetrating wells in formations of high hydraulic conductivity. This approach is a simple, spreadsheet-based implementation of existing models that can be used for analysis of tests from confined or unconfined aquifers. Field examples of tests exhibiting oscillatory and nonoscillatory behavior are used to illustrate the procedure and to compare results with estimates obtained using alternative approaches. The procedure is considerably simpler than recently proposed methods for this hydrogeologic setting. Although the simplifications required by the approach can introduce error into hydraulic-conductivity estimates, this additional error becomes negligible when appropriate measures are taken in the field. These measures are summarized in a set of practical field guidelines for slug tests in highly permeable aquifers.

  17. Comparative analysis of hydraulic crane-manipulating installations transport and technological machines and industrial robots hydraulic manipulators

    Directory of Open Access Journals (Sweden)

    Lagerev I.A.

    2016-09-01

    Full Text Available The article presents results of comparative analysis of hydraulic crane-manipulator installations of mobile transport and technological machines and hydraulic manipulators of industrial robots. The comparative analysis is based on consid-eration of a wide range of types and sizes indicated technical devices of both domestic and foreign production: 1580 structures of cranes and more than 450 structures of industrial robots. It was performed in the following areas: func-tional purpose and basic technical characteristics; a design; the loading conditions of the model and failures in operation process; approaches to the design, calculation methods and mathematical modeling. The conclusions about the degree of similarity and the degree of difference hydraulic crane-manipulator installations of transport and technological ma-chines and hydraulic industrial robot manipulators from the standpoint of their design and modeling occurring in them during operation of dynamic and structural processes.

  18. Design and Performance Analysis of a new Rotary Hydraulic Joint

    Science.gov (United States)

    Feng, Yong; Yang, Junhong; Shang, Jianzhong; Wang, Zhuo; Fang, Delei

    2017-07-01

    To improve the driving torque of the robots joint, a wobble plate hydraulic joint is proposed, and the structure and working principle are described. Then mathematical models of kinematics and dynamics was established. On the basis of this, dynamic simulation and characteristic analysis are carried out. Results show that the motion curve of the joint is continuous and the impact is small. Moreover the output torque of the joint characterized by simple structure and easy processing is large and can be rotated continuously.

  19. The analysis of thermal-hydraulic models in MELCOR code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, M H; Hur, C; Kim, D K; Cho, H J [POhang Univ., of Science and TECHnology, Pohang (Korea, Republic of)

    1996-07-15

    The objective of the present work is to verify the prediction and analysis capability of MELCOR code about the progression of severe accidents in light water reactor and also to evaluate appropriateness of thermal-hydraulic models used in MELCOR code. Comparing the results of experiment and calculation with MELCOR code is carried out to achieve the above objective. Specially, the comparison between the CORA-13 experiment and the MELCOR code calculation was performed.

  20. Analysis of hydraulic gradients across the host rock at the proposed Texas Panhandle nuclear-waste repository site

    International Nuclear Information System (INIS)

    Bair, E.S.

    1987-01-01

    Analysis of the direction of ground-water flow across the host rock at the proposed high-level nuclear-waste repository site in Deaf Smith County, Texas, is complicated by vertical and lateral changes in the density of formation fluids in the various hydrogeologic units that overlie and underlie the proposed host rock. Because the concept of hydraulic head is not valid when evaluating vertical hydraulic gradients in a variably-density flow system, other methods were used to determine the direction and magnitude of vertical hydraulic gradients at the proposed site where the specific gravity of formation fluids varies between 1.00 and 1.28. The direction of ground-water flow across the proposed host rock, an 80-foot-thick salt bed in the Lower San Andres Formation, was determined by calculating vertical hydraulic gradients based on formation pressure and fluid density data, and by analysis of pressure-depth diagrams. Based on data from the vicinity of the proposed site, both methods indicate the potential for downflow across the host rock. Downflow or predominantly horizontal flow is considered a favorable prewaste emplacement condition because it prolongs the travel time to the biosphere of any naturally or accidentally released radionuclides

  1. THERMAL HYDRAULIC ANALYSIS OF FIRE DIVERTOR

    International Nuclear Information System (INIS)

    C.B. bAXI; M.A. ULRICKSON; D.E. DRIMEYER; P. HEITZENROEDER

    2000-01-01

    The Fusion Ignition Research Experiment (FIRE) is being designed as a next step in the US magnetic fusion program. The FIRE tokamak has a major radius of 2 m, a minor radius of 0.525 m, and liquid nitrogen cooled copper coils. The aim is to produce a pulse length of 20 s with a plasma current of 6.6 MA and with alpha dominated heating. The outer divertor and baffle of FIRE are water cooled. The worst thermal condition for the outer divertor and baffle is the baseline D-T operating mode (10 T, 6.6 MA, 20 s) with a plasma exhaust power of 67 MW and a peak heat flux of 20 MW/m 2 . A swirl tape (ST) heat transfer enhancement method is used in the outer divertor cooling channels to increase the heat transfer coefficient and the critical heat flux (CHF). The plasma-facing surface consists of tungsten brush. The finite element (FE) analysis shows that for an inlet water temperature of 30 C, inlet pressure of 1.5 MPa and a flow velocity of 10 m/s, the incident critical heat flux is greater than 30 MW/m 2 . The peak copper temperature is 490 C, peak tungsten temperature is 1560 C, and the pressure drop is less than 0.5 MPa. All these results fulfill the design requirements

  2. Thermal hydraulic analysis of BWR containment venting system

    International Nuclear Information System (INIS)

    Baburajan, P.K.; Sharma, Prashant; Paul, U.K.; Gaikwad, Avinash

    2015-01-01

    Installation of additional containment filtered venting system (CFVS) is necessary to depressurize the containment to maintain its mechanical integrity due to over pressurization during severe accident condition. A typical venting system for BWR is modelled using RELAP5 and analysed to investigate the effect of various thermal hydraulic parameters on the operational parameters of the venting system. The venting system consists of piping from the containment to the scrubber tank and exit line from the scrubber tank. The scrubber tank is partially filled with water to enable the scrubbing action to remove the particulate radionuclides from the incoming containment air. The pipe line from the containment is connected to the venturi inlet and the throat of the venturi is open to the scrubber tank water inventory at designed submergence level. The exit of the venturi is open to scrubber tank water. Filters are used in the upper air space of the scrubber tank as mist separator before venting out the air into the atmosphere through the exit vent line. The effect of thermal hydraulic parameters such as inlet fluid temperature, inlet steam content and venturi submergence in the scrubber tank on the venting flow rate, exit steam content, scrubber tank inventory, overflow line and siphon breaker flow rate is analysed. Results show that inlet steam content and the venturi nozzle submergence influence the venting system parameters. (author)

  3. 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.

  4. Hydraulic modeling support for conflict analysis: The Manayunk canal revisited

    International Nuclear Information System (INIS)

    Chadderton, R.A.; Traver, R.G.; Rao, J.N.

    1992-01-01

    This paper presents a study which used a standard, hydraulic computer model to generate detailed design information to support conflict analysis of a water resource use issue. As an extension of previous studies, the conflict analysis in this case included several scenarios for stability analysis - all of which reached the conclusion that compromising, shared access to the water resources available would result in the most benefits to society. This expected equilibrium outcome was found to maximize benefit-cost estimates. 17 refs., 1 fig., 2 tabs

  5. 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.

  6. Thermal-hydraulic analysis of PWR cores in transient condition

    International Nuclear Information System (INIS)

    Silva Galetti, M.R. da.

    1984-01-01

    A calculational methodology for thermal - hydraulic analysis of PWR cores under steady-state and transient condition was selected and made available to users. An evaluation of the COBRA-IIIP/MIT code, used for subchannel analysis, was done through comparison of the code results with experimental data on steady state and transient conditions. As a result, a comparison study allowing spatial and temporal localization of critical heat flux was obtained. A sensitivity study of the simulation model to variations in some empirically determined parameter is also presented. Two transient cases from Angra I FSAR were analysed, showing the evolution of minimum DNBR with time. (Author) [pt

  7. Thermal-hydraulic analysis on the whole module of water cooled ceramic breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, Kecheng; Ma, Xuebin [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); University of Science and Technology of China, Hefei, Anhui, 230027 (China); Cheng, Xiaoman [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Lin, Shuang [University of Science and Technology of China, Hefei, Anhui, 230027 (China); Huang, Kai [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Liu, Songlin, E-mail: slliu@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); University of Science and Technology of China, Hefei, Anhui, 230027 (China)

    2016-11-15

    Highlights: • The 3D thermal hydraulic analysis on the whole module of WCCB is performed by CFD method. • Temperature field and mass flow distribution have been obtained. • The design of WCCB is reasonable from the perspective of thermal-hydraulics. • The scheme for further optimization has been proposed. - Abstract: The Water Cooled Ceramic Breeder blanket (WCCB) is being researched for Chinese Fusion Engineering Test Reactor (CFETR). The thermal-hydraulic analysis is essential because the blanket should remove the high heat flux from the plasma and the volumetric heat generated by neutrons. In this paper, the detailed three dimensional (3D) thermal hydraulic analysis on the whole module of WCCB blanket has been performed by Computational Fluid Dynamics (CFD) method, which is capable of solving conjugate heat transfer between solid structure and fluid. The main results, including temperature field, distribution of mass flow rate and coolant pressure drop, have been calculated simultaneously. These provides beneficial guidance data for the further structural optimization and for the design arrangement of primary and secondary circuit. Under the total heat source of 1.23 MW, the coolant mass flow rate of 5.457 kg/s is required to make coolant water corresponding to the Pressurized Water Reactor (PWR) condition (15.5 MPa, 285 °C–325 °C), generating the total coolant pressure drop (△P) of 0.467 MPa. The results show that the present structural design can make all the materials effectively cooled to the allowable temperature range, except for a few small modifications on the both sides of FW. The main components, including the first wall (FW), cooling plates (CPs), side wall (SWs)&stiffening plates (SPs) and the manifold(1–4), dominate 4.7%/41.7%/13%/40.6% of the total pressure drop, respectively. Additionally, the mass flow rate of each channel has been obtained, showing the peak relative deviation of 3.4% and 2% from the average for the paratactic

  8. Analysis of and H∞ Controller Design For An Electro-Hydraulic Servo Pressure Regulator

    DEFF Research Database (Denmark)

    Stubkier, Søren; Pedersen, Henrik C.; Andersen, Torben Ole

    2011-01-01

    -circuit pumps are still hydraulically controlled, there is however still a need for being able to generate a hydraulic pilot pressure. The focus of the current paper is on the analysis and controller design of an electrohydraulic servo pressure regulator, which generates a hydraulic LS-pressure for a variable...

  9. Hydraulic manipulator design, analysis, and control at Oak Ridge National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Kress, R.L.; Jansen, J.F. [Oak Ridge National Lab., TN (United States). Robotics and Process Systems Div.; Love, L.J. [Oak Ridge Inst. for Science and Education, TN (United States); Basher, A.M.H. [South Carolina State Univ., Orangeburg, SC (United States)

    1996-09-01

    To meet the increased payload capacities demanded by present-day tasks, manipulator designers have turned to hydraulics as a means of actuation. Hydraulics have always been the actuator of choice when designing heavy-life construction and mining equipment such as bulldozers, backhoes, and tunneling devices. In order to successfully design, build, and deploy a new hydraulic manipulator (or subsystem) sophisticated modeling, analysis, and control experiments are usually needed. To support the development and deployment of new hydraulic manipulators Oak Ridge National Laboratory (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 hydraulics laboratory at ORNL has three different manipulators. First is a 6-Degree-of-Freedom (6-DoF), multi-planer, teleoperated, flexible controls test bed used for the development of waste tank clean-up manipulator controls, thermal studies, system characterization, and manipulator tracking. Finally, is a human amplifier test bed used for the development of an entire new class of teleoperated systems. To compliment the hardware in the hydraulics laboratory, ORNL has developed a hydraulics simulation capability including a custom package to model the hydraulic systems and manipulators for performance studies and control development. This paper outlines the history of hydraulic manipulator developments at ORNL, describes the hydraulics laboratory, discusses the use of the equipment within the laboratory, and presents some of the initial results from experiments and modeling associated with these hydraulic manipulators. Included are some of the results from the development of the human amplifier/de-amplifier concepts, the characterization of the thermal sensitivity of hydraulic systems, and end-point tracking accuracy studies. Experimental and analytical

  10. Hydraulic manipulator design, analysis, and control at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Kress, R.L.; Jansen, J.F.; Basher, A.M.H.

    1996-09-01

    To meet the increased payload capacities demanded by present-day tasks, manipulator designers have turned to hydraulics as a means of actuation. Hydraulics have always been the actuator of choice when designing heavy-life construction and mining equipment such as bulldozers, backhoes, and tunneling devices. In order to successfully design, build, and deploy a new hydraulic manipulator (or subsystem) sophisticated modeling, analysis, and control experiments are usually needed. To support the development and deployment of new hydraulic manipulators Oak Ridge National Laboratory (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 hydraulics laboratory at ORNL has three different manipulators. First is a 6-Degree-of-Freedom (6-DoF), multi-planer, teleoperated, flexible controls test bed used for the development of waste tank clean-up manipulator controls, thermal studies, system characterization, and manipulator tracking. Finally, is a human amplifier test bed used for the development of an entire new class of teleoperated systems. To compliment the hardware in the hydraulics laboratory, ORNL has developed a hydraulics simulation capability including a custom package to model the hydraulic systems and manipulators for performance studies and control development. This paper outlines the history of hydraulic manipulator developments at ORNL, describes the hydraulics laboratory, discusses the use of the equipment within the laboratory, and presents some of the initial results from experiments and modeling associated with these hydraulic manipulators. Included are some of the results from the development of the human amplifier/de-amplifier concepts, the characterization of the thermal sensitivity of hydraulic systems, and end-point tracking accuracy studies. Experimental and analytical

  11. 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

  12. Local chemical and thermal-hydraulic analysis of U-tube steam generators

    International Nuclear Information System (INIS)

    Lee, J.Y.; No, H.C.

    1990-01-01

    In order to know how pH distribution affects corrosion in a U-tube steam generator, a study of the combination of water chemistry and thermal-hydraulic conditions is suggested. A two-fluid (unequal velocity and unequal temperature) formulation is proposed to describe the convective transport of volatile species in each phase, and a spherical bubble model is developed on the basis of the penetration theory to describe the interfacial mass transfer. The thermal-hydraulic local conditions are obtained by the U-tube steam generator design analysis code FAUST which is based on the three-dimensional two-fluid model. The results of the present study are compared with dynamic equilibrium model calculations. This study shows that, in contrast with dynamic equilibrium calculations, the pH is lower in the cold-leg side than in the hot-leg side because of liquid recirculation. Just above the tube sheet, however, the lower void fraction in this region than that in the hot-leg region results in higher pH, which agrees with the prediction of the dynamic equilibrium model. (orig.)

  13. Hydraulic separation of plastic wastes: Analysis of liquid-solid interaction.

    Science.gov (United States)

    Moroni, Monica; Lupo, Emanuela; La Marca, Floriana

    2017-08-01

    The separation of plastic wastes in mechanical recycling plants is the process that ensures high-quality secondary raw materials. An innovative device employing a wet technology for particle separation is presented in this work. Due to the combination of the characteristic flow pattern developing within the apparatus and density, shape and size differences among two or more polymers, it allows their separation into two products, one collected within the instrument and the other one expelled through its outlet ducts. The kinematic investigation of the fluid flowing within the apparatus seeded with a passive tracer was conducted via image analysis for different hydraulic configurations. The two-dimensional turbulent kinetic energy results strictly connected to the apparatus separation efficacy. Image analysis was also employed to study the behaviour of mixtures of passive tracer and plastic particles with different physical characteristics in order to understand the coupling regime between fluid and solid phases. The two-dimensional turbulent kinetic energy analysis turned out to be fundamental to this aim. For the tested operating conditions, two-way coupling takes place, i.e., the fluid exerts an influence on the plastic particle and the opposite occurs too. Image analysis confirms the outcomes from the investigation of the two-phase flow via non-dimensional numbers (particle Reynolds number, Stokes number and solid phase volume fraction). Copyright © 2017 Elsevier Ltd. All rights reserved.

  14. 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 ...

  15. Spent nuclear fuel storage pool thermal-hydraulic analysis

    International Nuclear Information System (INIS)

    Gay, R.R.

    1984-01-01

    Storage methods and requirements for spent nuclear fuel at U.S. commercial light water reactors are reviewed in Section 1. Methods of increasing current at-reactor storage capabilities are also outlined. In Section 2 the development of analytical methods for the thermal-hydraulic analysis of spent fuel pools is chronicled, leading up to a discussion of the GFLOW code which is described in Section 3. In Section 4 the verification of GFLOW by comparisons of the code's predictions to experimental data taken inside the fuel storage pool at the Maine Yankee nuclear power plant is presented. The predictions of GFLOW using 72, 224, and 1584 node models of the storage pool are compared to each other and to the experimental data. An example of thermal licensing analysis for Maine Yankee using the GFLOW code is given in Section 5. The GFLOW licensing analysis is compared to previous licensing analysis performed by Yankee Atomic using the RELAP-4 computer code

  16. 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

  17. 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

  18. 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)

  19. 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)

  20. 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...

  1. Development of local TDC model in core thermal hydraulic analysis

    International Nuclear Information System (INIS)

    Kwon, H.S.; Park, J.R.; Hwang, D.H.; Lee, S.K.

    2004-01-01

    The local TDC model consisting of natural mixing and forced mixing part was developed to obtain more realistic local fluid properties in the core subchannel analysis. To evaluate the performance of local TDC model, the CHF prediction capability was tested with the various CHF correlations and local fluid properties at CHF location which are based on the local TDC model. The results show that the standard deviation of measured to predicted CHF ratio (M/P) based on local TDC model can be reduced by about 7% compared to those based on global TDC model when the CHF correlation has no term to account for distance from the spacer grid. (author)

  2. Momentum integral network method for thermal-hydraulic transient analysis

    International Nuclear Information System (INIS)

    Van Tuyle, G.J.

    1983-01-01

    A new momentum integral network method has been developed, and tested in the MINET computer code. The method was developed in order to facilitate the transient analysis of complex fluid flow and heat transfer networks, such as those found in the balance of plant of power generating facilities. The method employed in the MINET code is a major extension of a momentum integral method reported by Meyer. Meyer integrated the momentum equation over several linked nodes, called a segment, and used a segment average pressure, evaluated from the pressures at both ends. Nodal mass and energy conservation determined nodal flows and enthalpies, accounting for fluid compression and thermal expansion

  3. 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.

  4. 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

  5. Analysis of giant electrorheological fluids.

    Science.gov (United States)

    Seo, Youngwook P; Seo, Yongsok

    2013-07-15

    The yield stress dependence on electric field strength for giant electrorheological (GER) fluids over the full range of electric fields was examined using Seo's scaling function which incorporated both the polarization and the conductivity models. If a proper scaling was applied to the yield stress data to collapse them onto a single curve, the Seo's scaling function could correctly fit the yield stress behavior of GER suspensions, even at very high electric field strengths. The model predictions were also compared with recently proposed Choi et al.'s model to allow a consideration of the universal framework of ER fluids. Copyright © 2013 Elsevier Inc. All rights reserved.

  6. Thermal hydraulics analysis of LIBRA-SP target chamber

    International Nuclear Information System (INIS)

    Mogahed, E.A.

    1996-01-01

    LIBRA-SP is a conceptual design study of an inertially confined 1000 MWe fusion power reactor utilizing self-pinched light ion beams. There are 24 ion beams which are arranged around the reactor cavity. The reaction chamber is an upright cylinder with an inverted conical roof resembling a mushroom, and a pool floor. The vertical sides of the cylinder are occupied by a blanket zone consisting of many perforated rigid HT-9 ferritic steel tubes called PERITs (PEr-forated RIgid Tube). The breeding/cooling material, liquid lead-lithium, flows through the PERITs, providing protection to the reflector/vacuum chamber so as to make it a lifetime component. The neutronics analysis and cavity hydrodynamics calculations are performed to account for the neutron heating and also to determine the effects of vaporization/condensation processes on the surface heat flux. The steady state nuclear heating distribution at the midplane is used for thermal hydraulics calculations. The maximum surface temperature of the HT-9 is chosen to not exceed 625 degree C to avoid drastic deterioration of the metal's mechanical properties. This choice restricts the thermal hydraulics performance of the reaction cavity. The inlet first surface coolant bulk temperature is 370 degree C, and the heat exchanger inlet coolant bulk temperature is 502 degree C. 4 refs., 6 figs., 2 tabs

  7. Thermal Hydraulic Analysis on Containment Filtered Venting System

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Young Suk; Park, Tong Kyu; Lee, Doo Yong; Lee, Byung Chul [FNC Technology Co. Ltd., Yongin (Korea, Republic of); Lee, Sang Won; Kim, Hyeong Taek [KHNP-Central Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    In this study, the thermal hydraulic conditions (e. g. pressure and flow rate) at each component have been examined and the sensitivity analysis on CFVS design parameters (e. g. water inventory, volumetric flow rate). The purpose is to know the possible range of flow conditions at each component to determine the optimum size of filtration system. GOTHIC code has been used to simulate the thermal-hydraulic behavior inside of CFVS. The behavior of flows in the CFVS has been investigated. The vessel water level and the flow rates during the CFVS operation are examined. It was observed that the vessel water level would be changed significantly due to steam condensation/thermal expansion and steam evaporation. Therefore, the vessel size and the initial water inventory should be carefully determined to keep the minimum water level required for filtration components and not to flood the components in the upper side of the vessel. It has been also observed that the volumetric flow rate is maintained during the CFVS operation, which is beneficial for pool scrubbing units. However, regarding the significant variations at the orifice downstream, careful design would be necessary.

  8. Development of thermal hydraulic analysis code for IHX of FBR

    International Nuclear Information System (INIS)

    Kumagai, Hiromichi; Naohara, Nobuyuki

    1991-01-01

    In order to obtain flow resistance correlations for thermal-hydrauric analysis code concerned with an intermediate heat exchanger (IHX) of FBR, the hydraulic experiment by air was carried out through a bundle of tubes arranged in an in-line and staggard fashion. The main results are summarized as follows. (1) On pressure loss per unit length of a tube bundle, which is densely a regular triangle arrangement, the in-line fashion is almost the same as the staggard one. (2) In case of 30deg sector model for IHX tube bundle, pressure loss is 1/3 in comparison with the in-line or staggard arrangement. (3) By this experimental data, flow resistance correlations for thermalhydrauric analysis code are obtained. (author)

  9. Thermal hydraulic analysis of the encapsulated nuclear heat source

    Energy Technology Data Exchange (ETDEWEB)

    Sienicki, J.J.; Wade, D.C. [Argonne National Lab., IL (United States)

    2001-07-01

    An analysis has been carried out of the steady state thermal hydraulic performance of the Encapsulated Nuclear Heat Source (ENHS) 125 MWt, heavy liquid metal coolant (HLMC) reactor concept at nominal operating power and shutdown decay heat levels. The analysis includes the development and application of correlation-type analytical solutions based upon first principles modeling of the ENHS concept that encompass both pure as well as gas injection augmented natural circulation conditions, and primary-to-intermediate coolant heat transfer. The results indicate that natural circulation of the primary coolant is effective in removing heat from the core and transferring it to the intermediate coolant without the attainment of excessive coolant temperatures. (authors)

  10. 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.

  11. Analysis of an controller design for an electro-hydraulic servo pressure regulator

    DEFF Research Database (Denmark)

    Pedersen, Henrik C.; Andersen, Torben Ole; Madsen, A. M.

    2009-01-01

    Mobile hydraulics is in a transition phase, where electronic sensors and digital signal processors are starting to become standard on a high number of machines, hereby replacing hydraulic pilot lines and oering new possibilities with regard to both control and feasibility. For controlling some...... of the existing hydraulic components there are, however, still a need for being able to generate a hydraulic pilot pressure, as e.g. almost all open-circuit pumps are hydraulically controlled. The focus of the current paper is therefore on the analysis and controller design an electro-hydraulic servo pressure...... regulator, which generates a hydraulic LS-pressure based on an electrical reference, hereby synergistically integrating knowledge from all parts of the mechatronics area. The servo pressure regulator is used to generate the LS-signal for a variable displacement pump, and the paper rst presents...

  12. Vibration analysis of pipes conveying fluid by transfer matrix method

    International Nuclear Information System (INIS)

    Li, Shuai-jun; Liu, Gong-min; Kong, Wei-tao

    2014-01-01

    Highlights: • A theoretical study on vibration analysis of pipes with FSI is presented. • Pipelines with high fluid pressure and velocity can be solved by developed method. • Several pipeline schemes are discussed to illustrate the application of the method. • The proposed method is easier to apply compared to most existing procedures. • Influence laws of structural and fluid parameters on FSI of pipe are analyzed. -- Abstract: Considering the effects of pipe wall thickness, fluid pressure and velocity, a developed 14-equation model is presented, which describes the fluid–structure interaction behavior of pipelines. The transfer matrix method has been used for numerical modeling of both hydraulic and structural equations. Based on these models and algorithms, several pipeline schemes are presented to illustrate the application of the proposed method. Furthermore, the influence laws of supports, structural properties and fluid parameters on the dynamic response and natural frequencies of pipeline are analyzed, which shows using the optimal supports and structural properties is beneficial to reduce vibration of pipelines

  13. 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...

  14. Development of realistic thermal hydraulic system analysis code

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Jae; Chung, B. D; Kim, K. D. [and others

    2002-05-01

    The realistic safety analysis system is essential for nuclear safety research, advanced reactor development, safety analysis in nuclear industry and 'in-house' plant design capability development. In this project, we have developed a best-estimate multi-dimensional thermal-hydraulic system code, MARS, which is based on the integrated version of the RELAP5 and COBRA-TF codes. To improve the realistic analysis capability, we have improved the models for multi-dimensional two-phase flow phenomena and for advanced two-phase flow modeling. In addition, the GUI (Graphic User Interface) feature were developed to enhance the user's convenience. To develop the coupled analysis capability, the MARS code were linked with the three-dimensional reactor kinetics code (MASTER), the core thermal analysis code (COBRA-III/CP), and the best-estimate containment analysis code (CONTEMPT), resulting in MARS/MASTER/COBRA/CONTEMPT. Currently, the MARS code system has been distributed to 18 domestic organizations, including research, industrial, regulatory organizations and universities. The MARS has been being widely used for the safety research of existing PWRs, advanced PWR, CANDU and research reactor, the pre-test analysis of TH experiments, and others.

  15. Development of realistic thermal hydraulic system analysis code

    International Nuclear Information System (INIS)

    Lee, Won Jae; Chung, B. D; Kim, K. D.

    2002-05-01

    The realistic safety analysis system is essential for nuclear safety research, advanced reactor development, safety analysis in nuclear industry and 'in-house' plant design capability development. In this project, we have developed a best-estimate multi-dimensional thermal-hydraulic system code, MARS, which is based on the integrated version of the RELAP5 and COBRA-TF codes. To improve the realistic analysis capability, we have improved the models for multi-dimensional two-phase flow phenomena and for advanced two-phase flow modeling. In addition, the GUI (Graphic User Interface) feature were developed to enhance the user's convenience. To develop the coupled analysis capability, the MARS code were linked with the three-dimensional reactor kinetics code (MASTER), the core thermal analysis code (COBRA-III/CP), and the best-estimate containment analysis code (CONTEMPT), resulting in MARS/MASTER/COBRA/CONTEMPT. Currently, the MARS code system has been distributed to 18 domestic organizations, including research, industrial, regulatory organizations and universities. The MARS has been being widely used for the safety research of existing PWRs, advanced PWR, CANDU and research reactor, the pre-test analysis of TH experiments, and others

  16. 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)

  17. Thermal-Hydraulics analysis of pressurized water reactor core by using single heated channel model

    Directory of Open Access Journals (Sweden)

    Reza Akbari

    2017-08-01

    Full Text Available Thermal hydraulics of nuclear reactor as a basis of reactor safety has a very important role in reactor design and control. The thermal-hydraulic analysis provides input data to the reactor-physics analysis, whereas the latter gives information about the distribution of heat sources, which is needed to perform the thermal-hydraulic analysis. In this study single heated channel model as a very fast model for predicting thermal hydraulics behavior of pressurized water reactor core has been developed. For verifying the results of this model, we used RELAP5 code as US nuclear regulatory approved thermal hydraulics code. The results of developed single heated channel model have been checked with RELAP5 results for WWER-1000. This comparison shows the capability of single heated channel model for predicting thermal hydraulics behavior of reactor core.

  18. Hydraulic analysis and optimization design in Guri rehabilitation project

    Science.gov (United States)

    Cheng, H.; Zhou, L. J.; Gong, L.; Wang, Z. N.; Wen, Q.; Zhao, Y. Z.; Wang, Y. L.

    2016-11-01

    Recently Dongfang was awarded the contract for rehabilitation of 6 units in Guri power plant, the biggest hydro power project in Venezuela. The rehabilitation includes, but not limited to, the extension of output capacity by about 50% and enhancement of efficiency level. To achieve the targets the runner and the guide vanes will be replaced by the newly optimized designs. In addition, the out-of-date stay vanes with straight plate shape will be modified into proper profiles after considering the application feasibility in field. The runner and vane profiles were optimized by using state-of-the-art flow simulation techniques. And the hydraulic performances were confirmed by the following model tests. This paper describes the flow analysis during the optimization procedure and the comparison between various technical concepts.

  19. 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.

  20. Thermal hydraulic analysis of the JMTR improved LEU-core

    Energy Technology Data Exchange (ETDEWEB)

    Tabata, Toshio; Nagao, Yoshiharu; Komukai, Bunsaku; Naka, Michihiro; Fujiki, Kazuo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Takeda, Takashi [Radioactive Waste Management and Nuclear Facility Decommissioning Technology Center, Tokai, Ibaraki (Japan)

    2003-01-01

    After the investigation of the new core arrangement for the JMTR reactor in order to enhance the fuel burn-up and consequently extend the operation period, the ''improved LEU core'' that utilized 2 additional fuel elements instead of formerly installed reflector elements, was adopted. This report describes the results of the thermal-hydraulic analysis of the improved LEU core as a part of safety analysis for the licensing. The analysis covers steady state, abnormal operational transients and accidents, which were described in the annexes of the licensing documents as design bases events. Calculation conditions for the computer codes were conservatively determined based on the neutronic analysis results and others. The results of the analysis, that revealed the safety criteria were satisfied on the fuel temperature, DNBR and primary coolant temperature, were used in the licensing. The operation license of the JMTR with the improved LEU core was granted in March 2001, and the reactor operation with new core started in November 2001 as 142nd operation cycle. (author)

  1. 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...

  2. A LiDAR based analysis of hydraulic hazard mapping

    Science.gov (United States)

    Cazorzi, F.; De Luca, A.; Checchinato, A.; Segna, F.; Dalla Fontana, G.

    2012-04-01

    Mapping hydraulic hazard is a ticklish procedure as it involves technical and socio-economic aspects. On the one hand no dangerous areas should be excluded, on the other hand it is important not to exceed, beyond the necessary, with the surface assigned to some use limitations. The availability of a high resolution topographic survey allows nowadays to face this task with innovative procedures, both in the planning (mapping) and in the map validation phases. The latter is the object of the present work. It should be stressed that the described procedure is proposed purely as a preliminary analysis based on topography only, and therefore does not intend in any way to replace more sophisticated analysis methods requiring based on hydraulic modelling. The reference elevation model is a combination of the digital terrain model and the digital building model (DTM+DBM). The option of using the standard surface model (DSM) is not viable, as the DSM represents the vegetation canopy as a solid volume. This has the consequence of unrealistically considering the vegetation as a geometric obstacle to water flow. In some cases the topographic model construction requires the identification and digitization of the principal breaklines, such as river banks, ditches and similar natural or artificial structures. The geometrical and topological procedure for the validation of the hydraulic hazard maps is made of two steps. In the first step the whole area is subdivided into fluvial segments, with length chosen as a reasonable trade-off between the need to keep the hydrographical unit as complete as possible, and the need to separate sections of the river bed with significantly different morphology. Each of these segments is made of a single elongated polygon, whose shape can be quite complex, especially for meandering river sections, where the flow direction (i.e. the potential energy gradient associated to the talweg) is often inverted. In the second step the segments are analysed

  3. Analysis of the Thermal and Hydraulic Stimulation Program at Raft River, Idaho

    Science.gov (United States)

    Bradford, Jacob; McLennan, John; Moore, Joseph; Podgorney, Robert; Plummer, Mitchell; Nash, Greg

    2017-05-01

    The Raft River geothermal field, located in southern Idaho, roughly 100 miles northwest of Salt Lake City, is the site of a Department of Energy Enhanced Geothermal System project designed to develop new techniques for enhancing the permeability of geothermal wells. RRG-9 ST1, the target stimulation well, was drilled to a measured depth of 5962 ft. and cased to 5551 ft. The open-hole section of the well penetrates Precambrian quartzite and quartz monzonite. The well encountered a temperature of 282 °F at its base. Thermal and hydraulic stimulation was initiated in June 2013. Several injection strategies have been employed. These strategies have included the continuous injection of water at temperatures ranging from 53 to 115 °F at wellhead pressures of approximately 275 psi and three short-term hydraulic stimulations at pressures up to approximately 1150 psi. Flow rates, wellhead and line pressures and fluid temperatures are measured continuously. These data are being utilized to assess the effectiveness of the stimulation program. As of August 2014, nearly 90 million gallons have been injected. A modified Hall plot has been used to characterize the relationships between the bottom-hole flowing pressure and the cumulative injection fluid volume. The data indicate that the skin factor is decreased, and/or the permeability around the wellbore has increased since the stimulation program was initiated. The injectivity index also indicates a positive improvement with values ranging from 0.15 gal/min psi in July 2013 to 1.73 gal/min psi in February 2015. Absolute flow rates have increased from approximately 20 to 475 gpm by February 2 2015. Geologic, downhole temperature and seismic data suggest the injected fluid enters a fracture zone at 5650 ft and then travels upward to a permeable horizon at the contact between the Precambrian rocks and the overlying Tertiary sedimentary and volcanic deposits. The reservoir simulation program FALCON developed at the Idaho National

  4. Analysis of anisotropic shells containing flowing fluid

    International Nuclear Information System (INIS)

    Lakis, A.A.

    1983-01-01

    A general theory for the dynamic analysis of anisotropic thin cylindrical shells containing flowing fluid is presented. The shell may be uniform or non-uniform, provided it is geometrically axially symmetric. This is a finite- element theory, using cylindrical finite elements, but the displacement functions are determined by using classical shell theory. A new solution of the wave equation of the liquid finite element leads to an expression of the fluid pressure, p, as a function of the nodal displacements of the element and three operative forces (inertia, centrifugal and Coriolis) of the moving fluid. (Author) [pt

  5. Discussion on Stochastic Analysis of Hydraulic Vibration in Pressurized Water Diversion and Hydropower Systems

    Directory of Open Access Journals (Sweden)

    Jianxu Zhou

    2018-03-01

    Full Text Available Hydraulic vibration exists in various water conveyance projects and has resulted in different operating problems, but its obvious effects on system’s pressure head and stable operation have not been definitively addressed in the issued codes for engineering design, especially considering the uncertainties of hydraulic vibration. After detailed analysis of the randomness in hydraulic vibration and the commonly used stochastic approaches, in the basic equations for hydraulic vibration analysis, the random parameters and the formed stochastic equations were discussed for further probabilistic characteristic analysis of the random variables. Furthermore, preliminary investigation of the stochastic analysis of hydraulic vibration in pressurized pipelines and possible self-excited vibration in pumped-storage systems was presented for further consideration. The detailed discussion indicates that it is necessary to conduct further and systematic stochastic analysis of hydraulic vibration. Further, with the obtained frequencies and amplitudes in the form of a probability statement, the stochastic characteristics of various hydraulic vibrations can be investigated in detail and these solutions will be more reasonable for practical applications. Eventually, the stochastic analysis of hydraulic vibration will provide a basic premise to introduce its effect into the engineering design of water diversion and hydropower systems.

  6. Eulerian fluid-structure analysis of BWR

    International Nuclear Information System (INIS)

    McMaster, W.H.

    1979-05-01

    A fluid-structure-interaction algorithm is developed for the analysis of the dynamic response of a BWR pressure-suppression pool and containment structure. The method is incorporated into a two-dimensional semi-implicit Eulerian hydrodynamics code, PELE-IC, for the solution of incompressible flow coupled to flexible structures. The fluid, structure, and coupling algorithms have been verified by calculation of solved problems from the literature and by comparison with air and steam blowdown experiments

  7. Analysis of a hydraulic a scaled asymmetric labyrinth weir with Ansys-Fluent

    Science.gov (United States)

    Otálora Carmona, Andrés Humberto; Santos Granados, Germán Ricardo

    2017-04-01

    This document presents the three dimensional computational modeling of a labyrinth weir, using the version 17.0 of the Computational Fluid Dynamics (CFD) software ANSYS - FLUENT. The computational characteristics of the model such as the geometry consideration, the mesh sensitivity, the numerical scheme, and the turbulence modeling parameters. The volume fraction of the water mixture - air, the velocity profile, the jet trajectory, the discharge coefficient and the velocity field are analyzed. With the purpose of evaluating the hydraulic behavior of the labyrinth weir of the Naveta's hydroelectric, in Apulo - Cundinamarca, was development a 1:21 scale model of the original structure, which was tested in the laboratory of the hydraulic studies in the Escuela Colombiana de Ingeniería Julio Garavito. The scale model of the structure was initially developed to determine the variability of the discharge coefficient with respect to the flow rate and their influence on the water level. It was elaborate because the original weir (labyrinth weir with not symmetrical rectangular section), did not have the capacity to work with the design flow of 31 m3/s, because over 15 m3/s, there were overflows in the adduction channel. This variation of efficiency was due to the thickening of the lateral walls by structural requirements. During the physical modeling doing by Rodríguez, H. and Matamoros H. (2015) in the test channel, it was found that, with the increase in the width of the side walls, the discharge coefficient is reduced an average by 34%, generating an increase of the water level by 0.26 m above the structure. This document aims to develop a splicing methodology between the physical models of a labyrinth weir and numerical modeling, using concepts of computational fluid dynamics and finite volume theories. For this, was carried out a detailed analysis of the variations in the different directions of the main hydraulic variables involved in the behavior, such as, the

  8. 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;

  9. Analysis and experimental study on hydraulic balance characteristics in density lock

    International Nuclear Information System (INIS)

    Gu Haifeng; Yan Changqi; Sun Furong

    2009-01-01

    Through the simplified theoretical model, the hydraulic balance condition which should be met in the density lock is obtained, when reactor operates normally and density lock is closed. The main parameters influencing this condition are analyzed, and the results show that the hydraulic balance in the density lock is characterized with self-stability in a certain range. Meantime, a simulating experimental loop is built and experimental verification on the self-stability characteristic is done. Moreover, experimental study is done on the conditions of flow change of work fluids in the primary circuit in the process of stable operations. The experimental results show that the hydraulic balance in the density lock can recovered quickly, depending on the self-stability characteristic without influences on the sealing performance of density lock and normal operation of reactor, after the change of operation parameters breaks the hydraulic balance. (authors)

  10. 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.

  11. 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

  12. 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.

  13. The Phebus FP thermal-hydraulic analysis with Melcor

    International Nuclear Information System (INIS)

    Akgane, Kikuo; Kiso, Yoshihiro; Fukahori, Takanori; Yoshino, Mamoru

    1995-01-01

    The severe accident analysis code MELCOR, version 1.8.2, has been applied for thermal-hydraulic pre-test analysis of the first test of the Phebus FP program (test FPT-0) to study the best test parameters and the applicability of the code. The Phebus FP program is an in-pile test program which has been planned by the French Commissariate a L'Energie Atomique and the Commission of the European Union. The experiments are being conducted by an international collaboration to study the release and transport of fission products (FPs) under conditions assumed to be the most representative of those that would occur in a severe accident. The Phebus FP test apparatus simulates a test bundle of an in-pile section, the circuit including the steam generator U-tubes and the containment. The FPT-0 test was designed to simulate the heat-up and subsequent fuel bundle degradation after a loss of coolant severe accident, using fresh fuel. Two options for fuel degradation models in MELCOR have been applied to fuel degradation behavior. the first model assumes that fuel debris will be formed immediately after the fuel support fails by cladding relocation due to the candling process. The other is the uncollapsed bare fuel pellets option, in which the fuel pellets remain standing in a columnar shape until the fuel reaches its melting point, even if the cladding has been relocated by candling. The thermal-hydraulic behaviors in the circuit and containment of Phebus FP are discussed herein. Flow velocities in the Phebus FP circuit are high in order to produce turbulent flow in a small diameter test pipe. The MELCOR calculation has shown that the length of the hot leg and steam generator are adequate to attain steam temperatures or 700 degrees C and 150 degrees C in the respective outlets. The containment atmosphere temperature and humidity derived by once through integral system calculation show that objective test conditions would be satisfied in the Phebus FP experiment

  14. The Phebus FP thermal-hydraulic analysis with Melcor

    Energy Technology Data Exchange (ETDEWEB)

    Akgane, Kikuo; Kiso, Yoshihiro [Nuclear Power Engineering Corporation, Tokyo (Japan); Fukahori, Takanori [Hitachi Engineering Company, Ltd., Hitachi-shi Ibaraki-ken (Japan); Yoshino, Mamoru [Nuclear Engineering Ltd., Tosabori Nishi-ku (Japan)

    1995-09-01

    The severe accident analysis code MELCOR, version 1.8.2, has been applied for thermal-hydraulic pre-test analysis of the first test of the Phebus FP program (test FPT-0) to study the best test parameters and the applicability of the code. The Phebus FP program is an in-pile test program which has been planned by the French Commissariate a L`Energie Atomique and the Commission of the European Union. The experiments are being conducted by an international collaboration to study the release and transport of fission products (FPs) under conditions assumed to be the most representative of those that would occur in a severe accident. The Phebus FP test apparatus simulates a test bundle of an in-pile section, the circuit including the steam generator U-tubes and the containment. The FPT-0 test was designed to simulate the heat-up and subsequent fuel bundle degradation after a loss of coolant severe accident, using fresh fuel. Two options for fuel degradation models in MELCOR have been applied to fuel degradation behavior. the first model assumes that fuel debris will be formed immediately after the fuel support fails by cladding relocation due to the candling process. The other is the uncollapsed bare fuel pellets option, in which the fuel pellets remain standing in a columnar shape until the fuel reaches its melting point, even if the cladding has been relocated by candling. The thermal-hydraulic behaviors in the circuit and containment of Phebus FP are discussed herein. Flow velocities in the Phebus FP circuit are high in order to produce turbulent flow in a small diameter test pipe. The MELCOR calculation has shown that the length of the hot leg and steam generator are adequate to attain steam temperatures or 700{degrees}C and 150{degrees}C in the respective outlets. The containment atmosphere temperature and humidity derived by once through integral system calculation show that objective test conditions would be satisfied in the Phebus FP experiment.

  15. High fidelity thermal-hydraulic analysis using CFD and massively parallel computers

    International Nuclear Information System (INIS)

    Weber, D.P.; Wei, T.Y.C.; Brewster, R.A.; Rock, Daniel T.; Rizwan-uddin

    2000-01-01

    Thermal-hydraulic analyses play an important role in design and reload analysis of nuclear power plants. These analyses have historically relied on early generation computational fluid dynamics capabilities, originally developed in the 1960s and 1970s. Over the last twenty years, however, dramatic improvements in both computational fluid dynamics codes in the commercial sector and in computing power have taken place. These developments offer the possibility of performing large scale, high fidelity, core thermal hydraulics analysis. Such analyses will allow a determination of the conservatism employed in traditional design approaches and possibly justify the operation of nuclear power systems at higher powers without compromising safety margins. The objective of this work is to demonstrate such a large scale analysis approach using a state of the art CFD code, STAR-CD, and the computing power of massively parallel computers, provided by IBM. A high fidelity representation of a current generation PWR was analyzed with the STAR-CD CFD code and the results were compared to traditional analyses based on the VIPRE code. Current design methodology typically involves a simplified representation of the assemblies, where a single average pin is used in each assembly to determine the hot assembly from a whole core analysis. After determining this assembly, increased refinement is used in the hot assembly, and possibly some of its neighbors, to refine the analysis for purposes of calculating DNBR. This latter calculation is performed with sub-channel codes such as VIPRE. The modeling simplifications that are used involve the approximate treatment of surrounding assemblies and coarse representation of the hot assembly, where the subchannel is the lowest level of discretization. In the high fidelity analysis performed in this study, both restrictions have been removed. Within the hot assembly, several hundred thousand to several million computational zones have been used, to

  16. Hydraulic Geometry Analysis of the Lower Mississippi River

    National Research Council Canada - National Science Library

    Soar, Philip J; Thorne, Colin R; Harmar, Oliver P

    2005-01-01

    The hydraulic geometry of the Lower Mississippi River is primarily the product of the action of natural flows acting on the floodplain materials over centuries and millennia to form an alluvial forming a channel...

  17. Hydraulic ram analysis = Analyse du bélier hydraulique

    NARCIS (Netherlands)

    Verspuy, C.; Tijsseling, A.S.

    1993-01-01

    A simple mathematical model describing the operation of a hydraulic ram is presented. Predictions of the model are compared with measurements done in an earlier stage of the project. The model is used to perform a parameter variation study.

  18. 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

  19. Characterization of fracture networks for fluid flow analysis

    International Nuclear Information System (INIS)

    Long, J.C.S.; Billaux, D.; Hestir, K.; Majer, E.L.; Peterson, J.; Karasaki, K.; Nihei, K.; Gentier, S.; Cox, L.

    1989-06-01

    The analysis of fluid flow through fractured rocks is difficult because the only way to assign hydraulic parameters to fractures is to perform hydraulic tests. However, the interpretation of such tests, or ''inversion'' of the data, requires at least that we know the geometric pattern formed by the fractures. Combining a statistical approach with geophysical data may be extremely helpful in defining the fracture geometry. Cross-hole geophysics, either seismic or radar, can provide tomograms which are pixel maps of the velocity or attenuation anomalies in the rock. These anomalies are often due to fracture zones. Therefore, tomograms can be used to identify fracture zones and provide information about the structure within the fracture zones. This structural information can be used as the basis for simulating the degree of fracturing within the zones. Well tests can then be used to further refine the model. Because the fracture network is only partially connected, the resulting geometry of the flow paths may have fractal properties. We are studying the behavior of well tests under such geometry. Through understanding of this behavior, it may be possible to use inverse techniques to refine the a priori assignment of fractures and their conductances such that we obtain the best fit to a series of well test results simultaneously. The methodology described here is under development and currently being applied to several field sites. 4 refs., 14 figs

  20. 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 ...

  1. Thermal hydraulics analysis of the Advanced High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Dean, E-mail: Dean_Wang@uml.edu [University of Massachusetts Lowell, One University Avenue, Lowell, MA 01854 (United States); Yoder, Graydon L.; Pointer, David W.; Holcomb, David E. [Oak Ridge National Laboratory, 1 Bethel Valley RD #6167, Oak Ridge, TN 37831 (United States)

    2015-12-01

    Highlights: • The TRACE AHTR model was developed and used to define and size the DRACS and the PHX. • A LOFF transient was simulated to evaluate the reactor performance during the transient. • Some recommendations for modifying FHR reactor system component designs are discussed. - Abstract: The Advanced High Temperature Reactor (AHTR) is a liquid salt-cooled nuclear reactor design concept, featuring low-pressure molten fluoride salt coolant, a carbon composite fuel form with embedded coated particle fuel, passively triggered negative reactivity insertion mechanisms, and fully passive decay heat rejection. This paper describes an AHTR system model developed using the Nuclear Regulatory Commission (NRC) thermal hydraulic transient code TRAC/RELAP Advanced Computational Engine (TRACE). The TRACE model includes all of the primary components: the core, downcomer, hot legs, cold legs, pumps, direct reactor auxiliary cooling system (DRACS), the primary heat exchangers (PHXs), etc. The TRACE model was used to help define and size systems such as the DRACS and the PHX. A loss of flow transient was also simulated to evaluate the performance of the reactor during an anticipated transient event. Some initial recommendations for modifying system component designs are also discussed. The TRACE model will be used as the basis for developing more detailed designs and ultimately will be used to perform transient safety analysis for the reactor.

  2. Cold starting characteristics analysis of hydraulic free piston engine

    International Nuclear Information System (INIS)

    Zhang, Shuanlu; Zhao, Zhenfeng; Zhao, Changlu; Zhang, Fujun; Wang, Shan

    2017-01-01

    The cold start characteristic of hydraulic free piston diesel engine may affect its stable operation. Therefore the specific cold start characteristics, such as BDC or TDC positions, pressure in-cylinder, heat release rate, should be investigated in detail. These parameters fluctuate in some regularity in the cod start process. With the development of the free piston engine prototype and the establishment of test bench, the results are obtained. For the dynamic results, the fluctuation range of TDC and BDC positions is 8 mm and decreases with time. The thermodynamic results show that the combustion process is not stable and the pressure in-cylinder fluctuates largely in the cold start process. In addition, the combustion is rapid and knock happens inevitably. In order to investigate the reasons, a CFD model is established for temperature analysis in-cylinder and heat transfer conditions. It is found that higher start wall temperature will lead to more uniform temperature distribution. The delay period may decreases and heat release will move forward. This reason is analyzed by thermodynamic derivation based on the first law of thermodynamics. Finally, the improvement suggestions of cold start strategy are proposed. - Highlights: • The cold start behaviors of HFPE are investigated in detail. • CFD method is used for simulating temperature distribution in start process. • Thermodynamic derivation uncovers the compression temperature distribution. • The improvement suggestions of cold start strategy are proposed.

  3. Thermal Hydraulic Analysis of RPV Support Cooling System for HTGR

    International Nuclear Information System (INIS)

    Min Qi; Wu Xinxin; Li Xiaowei; Zhang Li; He Shuyan

    2014-01-01

    Passive safety is now of great interest for future generation reactors because of its reduction of human interaction and avoidance of failures of active components. reactor pressure vessel (RPV) support cooling system (SCS) for high temperature gas-cooled reactor (HTGR) is a passive safety system and is used to cool the concrete seats for the four RPV supports at its bottom. The SCS should have enough cooling capacity to ensure the temperature of the concrete seats for the supports not exceeding the limit temperature. The SCS system is composed of a natural circulation water loop and an air cooling tower. In the water loop, there is a heat exchanger embedded in the concrete seat, heat is transferred by thermal conduction and convection to the cooling water. Then the water is cooled by the air cooler mounted in the air cooling tower. The driving forces for water and air are offered by the density differences caused by the temperature differences. In this paper, the thermal hydraulic analysis for this system was presented. Methods for decoupling the natural circulation and heat transfer between the water loop and air flow were introduced. The operating parameters for different working conditions and environment temperatures were calculated. (author)

  4. 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.

  5. Transient thermal, hydraulic, and mechanical analysis of a counter flow offset strip fin intermediate heat exchanger using an effective porous media approach

    Science.gov (United States)

    Urquiza, Eugenio

    This work presents a comprehensive thermal hydraulic analysis of a compact heat exchanger using offset strip fins. The thermal hydraulics analysis in this work is followed by a finite element analysis (FEA) to predict the mechanical stresses experienced by an intermediate heat exchanger (IHX) during steady-state operation and selected flow transients. In particular, the scenario analyzed involves a gas-to-liquid IHX operating between high pressure helium and liquid or molten salt. In order to estimate the stresses in compact heat exchangers a comprehensive thermal and hydraulic analysis is needed. Compact heat exchangers require very small flow channels and fins to achieve high heat transfer rates and thermal effectiveness. However, studying such small features computationally contributes little to the understanding of component level phenomena and requires prohibitive computational effort using computational fluid dynamics (CFD). To address this issue, the analysis developed here uses an effective porous media (EPM) approach; this greatly reduces the computation time and produces results with the appropriate resolution [1]. This EPM fluid dynamics and heat transfer computational code has been named the Compact Heat Exchanger Explicit Thermal and Hydraulics (CHEETAH) code. CHEETAH solves for the two-dimensional steady-state and transient temperature and flow distributions in the IHX including the complicating effects of temperature-dependent fluid thermo-physical properties. Temperature- and pressure-dependent fluid properties are evaluated by CHEETAH and the thermal effectiveness of the IHX is also calculated. Furthermore, the temperature distribution can then be imported into a finite element analysis (FEA) code for mechanical stress analysis using the EPM methods developed earlier by the University of California, Berkeley, for global and local stress analysis [2]. These simulation tools will also allow the heat exchanger design to be improved through an

  6. Handbook of mathematical analysis in mechanics of viscous fluids

    CERN Document Server

    Novotný, Antonín

    2018-01-01

    Mathematics has always played a key role for researches in fluid mechanics. The purpose of this handbook is to give an overview of items that are key to handling problems in fluid mechanics. Since the field of fluid mechanics is huge, it is almost impossible to cover many topics. In this handbook, we focus on mathematical analysis on viscous Newtonian fluid. The first part is devoted to mathematical analysis on incompressible fluids while part 2 is devoted to compressible fluids.

  7. Stability analysis for a delay differential equations model of a hydraulic turbine speed governor

    Science.gov (United States)

    Halanay, Andrei; Safta, Carmen A.; Dragoi, Constantin; Piraianu, Vlad F.

    2017-01-01

    The paper aims to study the dynamic behavior of a speed governor for a hydraulic turbine using a mathematical model. The nonlinear mathematical model proposed consists in a system of delay differential equations (DDE) to be compared with already established mathematical models of ordinary differential equations (ODE). A new kind of nonlinearity is introduced as a time delay. The delays can characterize different running conditions of the speed governor. For example, it is considered that spool displacement of hydraulic amplifier might be blocked due to oil impurities in the oil supply system and so the hydraulic amplifier has a time delay in comparison to the time control. Numerical simulations are presented in a comparative manner. A stability analysis of the hydraulic control system is performed, too. Conclusions of the dynamic behavior using the DDE model of a hydraulic turbine speed governor are useful in modeling and controlling hydropower plants.

  8. Hydraulic analysis, Mad River at State Highway 41, Springfield, Ohio

    Science.gov (United States)

    Mayo, Ronald I.

    1977-01-01

    A hydraulic analysis of the lad River in a reach at Springfield, Ohio was made to determine the effects of relocating State Highway 41 in 1S76. The main channel was cleaned by dredging in the vicinity cf the new highway bridge and at the Detroit, Toledo and Ironton Railway bridge upstream. The new highway was placed on a high fill with relief structures for flood plain drainage consisting of a 12-foot corrugated metal pipe culvert and a bridge opening to accommodate the Detroit, Toledo and Ironton Railway and a property access road. The effect of the new highway embankment on drainage from the flood plain was requested. Also requested was the effect that might be expected on the elevation of flood waters above the new highway embankment if the access road through the new highway embankment were raised.The study indicates that the improvement in the capacity of the main channel to carry water was such that, up to a discharge equivalent to a 25-year frequency flood, the water-surface elevation in the reach upstream from the Detroit, Toledo and Ironton Railway bridge would be about 0.6 foot lower than under conditions prior to the construction on State Highway 41. Diversion through the Mad River left bank levee break above the Detroit, Toledo and Ironton Railway bridge to the flood Flain would be decreased about one-half in terms of rate of discharge in cubic feet per second. The maximum difference in elevation cf the flood water between the upstream and downstream side of the new State Highway 41 embankment would be about 0.2 foot, with an additional 0.4 foot to be expected if the access road were raised 1.5 feet.

  9. 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.

  10. Progress of the DUPIC fuel compatibility analysis (II) - thermal-hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Choi, Hang Bok

    2005-03-01

    Thermal-hydraulic compatibility of the DUPIC fuel bundle with a 713 MWe Canada deuterium uranium (CANDU-6) reactor was studied by using both the single channel and sub-channel analysis methods. The single channel analysis provides the fuel channel flow rate, pressure drop, critical channel power, and the channel exit quality, which are assessed against the thermal-hydraulic design requirements of the CANDU-6 reactor. The single channel analysis by the NUCIRC code showed that the thermal-hydraulic performance of the DUPIC fuel is not different from that of the standard CANDU fuel. Regarding the local flow characteristics, the sub-channel analysis also showed that the uncertainty of the critical channel power calculation for the DUPIC fuel channel is very small. As a result, both the single and sub-channel analyses showed that the key thermal-hydraulic parameters of the DUPIC fuel channel do not deteriorate compared to the standard CANDU fuel channel.

  11. Neutronics and thermal-hydraulics analysis of KUHFR

    Energy Technology Data Exchange (ETDEWEB)

    Woodruff, W L [Argonne National Laboratory, Argonne, IL (United States); Mishima, K [KURRI, Osaka (Japan)

    1983-08-01

    control rod worth with reduced enrichment has not yet determined, but only a small decrease in worth is expected. These BOL boron poisoned fuels are also used as the fresh fuel feed for the equilibrium fuel cycle studies contained in this report. The first three cases shown have matching cycle lengths in the equilibrium cycle, while the last case has a considerably longer cycle length. These results are similarly reflected in the 'Maximum Cycle Lengths' shown for unpoisoned BOL cores. Thus, the first three case can be considered comparable. The last case might be considered as an option for an extended cycle length design. The cycle length for this case is increased by about 21%. Obviously, by decreasing the uranium density in the fuel meat (to 2.7 g/cm{sup 3}), the cycle length for this design could be reduced to match that of the other cases. Thermal-hydraulic calculations have been carried out in order to study the safety aspects of the use of reduced enrichment uranium fuel for the KUHFR. The calculations were based on what is outlined in the Safety Analysis Report for the KUHFR and also the IAEA Guidebook for the RERTR program. Only a few combinations of hydraulic parameters have been tested because the reactor safety cannot be discussed without any nuclear physics considerations. For example, any variations in fuel coolant channels may change not only flow velocities but also power peaking factors which may affect the assessment of reactor safety. For this reason, the thermal-hydraulic calculations were carried out only for those specific cases on which neutronics analysis has been already performed. Low enriched uranium (LEU) cases are also included in this study as initial feasibility studies for potential conversion. The computer code PLTEMP has been developed to calculate the flow distribution in the core, fuel plate temperatures and DNB heat fluxes.

  12. 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.

  13. COBRA-SFS [Spent Fuel Storage]: A thermal-hydraulic analysis computer code: Volume 2, User's manual

    International Nuclear Information System (INIS)

    Rector, D.R.; Cuta, J.M.; Lombardo, N.J.; Michener, T.E.; Wheeler, C.L.

    1986-11-01

    COBRA-SFS (Spent Fuel Storage) is a general thermal-hydraulic analysis computer code used to predict temperatures and velocities in a wide variety of systems. The code was refined and specialized for spent fuel storage system analyses for the US Department of Energy's Commercial Spent Fuel Management Program. The finite-volume equations governing mass, momentum, and energy conservation are written for an incompressible, single-phase fluid. The flow equations model a wide range of conditions including natural circulation. The energy equations include the effects of solid and fluid conduction, natural convection, and thermal radiation. The COBRA-SFS code is structured to perform both steady-state and transient calculations; however, the transient capability has not yet been validated. This volume contains the input instructions for COBRA-SFS and an auxiliary radiation exchange factor code, RADX-1. It is intended to aid the user in becoming familiar with the capabilities and modeling conventions of the code

  14. COBRA-SFS [Spent Fuel Storage]: A thermal-hydraulic analysis computer code: Volume 1, Mathematical models and solution method

    International Nuclear Information System (INIS)

    Rector, D.R.; Wheeler, C.L.; Lombardo, N.J.

    1986-11-01

    COBRA-SFS (Spent Fuel Storage) is a general thermal-hydraulic analysis computer code used to predict temperatures and velocities in a wide variety of systems. The code was refined and specialized for spent fuel storage system analyses for the US Department of Energy's Commercial Spent Fuel Management Program. The finite-volume equations governing mass, momentum, and energy conservation are written for an incompressible, single-phase fluid. The flow equations model a wide range of conditions including natural circulation. The energy equations include the effects of solid and fluid conduction, natural convection, and thermal radiation. The COBRA-SFS code is structured to perform both steady-state and transient calculations: however, the transient capability has not yet been validated. This volume describes the finite-volume equations and the method used to solve these equations. It is directed toward the user who is interested in gaining a more complete understanding of these methods

  15. Development of MARS for multi-dimensional and multi-purpose thermal-hydraulic system analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Jae; Chung, Bub Dong; Kim, Kyung Doo; Hwang, Moon Kyu; Jeong, Jae Jun; Ha, Kwi Seok; Joo, Han Gyu [Korea Atomic Energy Research Institute, T/H Safety Research Team, Yusung, Daejeon (Korea)

    2000-10-01

    MARS (Multi-dimensional Analysis of Reactor Safety) code is being developed by KAERI for the realistic thermal-hydraulic simulation of light water reactor system transients. MARS 1.4 has been developed as a final version of basic code frame for the multi-dimensional analysis of system thermal-hydraulics. Since MARS 1.3, MARS 1.4 has been improved to have the enhanced code capability and user friendliness through the unification of input/output features, code models and code functions, and through the code modernization. Further improvements of thermal-hydraulic models, numerical method and user friendliness are being carried out for the enhanced code accuracy. As a multi-purpose safety analysis code system, a coupled analysis system, MARS/MASTER/CONTEMPT, has been developed using multiple DLL (Dynamic Link Library) techniques of Windows system. This code system enables the coupled, that is, more realistic analysis of multi-dimensional thermal-hydraulics (MARS 2.0), three-dimensional core kinetics (MASTER) and containment thermal-hydraulics (CONTEMPT). This paper discusses the MARS development program, and the developmental progress of the MARS 1.4 and the MARS/MASTER/CONTEMPT focusing on major features of the codes and their verification. It also discusses thermal hydraulic models and new code features under development. (author)

  16. Development of MARS for multi-dimensional and multi-purpose thermal-hydraulic system analysis

    International Nuclear Information System (INIS)

    Lee, Won Jae; Chung, Bub Dong; Kim, Kyung Doo; Hwang, Moon Kyu; Jeong, Jae Jun; Ha, Kwi Seok; Joo, Han Gyu

    2000-01-01

    MARS (Multi-dimensional Analysis of Reactor Safety) code is being developed by KAERI for the realistic thermal-hydraulic simulation of light water reactor system transients. MARS 1.4 has been developed as a final version of basic code frame for the multi-dimensional analysis of system thermal-hydraulics. Since MARS 1.3, MARS 1.4 has been improved to have the enhanced code capability and user friendliness through the unification of input/output features, code models and code functions, and through the code modernization. Further improvements of thermal-hydraulic models, numerical method and user friendliness are being carried out for the enhanced code accuracy. As a multi-purpose safety analysis code system, a coupled analysis system, MARS/MASTER/CONTEMPT, has been developed using multiple DLL (Dynamic Link Library) techniques of Windows system. This code system enables the coupled, that is, more realistic analysis of multi-dimensional thermal-hydraulics (MARS 2.0), three-dimensional core kinetics (MASTER) and containment thermal-hydraulics (CONTEMPT). This paper discusses the MARS development program, and the developmental progress of the MARS 1.4 and the MARS/MASTER/CONTEMPT focusing on major features of the codes and their verification. It also discusses thermal hydraulic models and new code features under development. (author)

  17. Thermal hydraulic feasibility analysis of the IBED PHTS for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Carloni, Dario, E-mail: dacarloni@gmail.com [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Pisa University, Via Diotisalvi 2, 56126 Pisa (Italy); Dell’Orco, Giovanni; Babulal, Gopalapillai; Somboli, Fabio; Serio, Luigi [ITER Organisation, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Paci, Sandro [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Pisa University, Via Diotisalvi 2, 56126 Pisa (Italy)

    2013-10-15

    One of the main challenges of the ITER fusion reactor is to effectively remove large amount of heat deposited to the surface of the plasma facing components. The tokamak cooling water system (TCWS) will accomplish the objective of removing about 1 GW of peak heat load from in-vessel components while maintaining pressures and temperatures of the coolant within acceptable and safe limits during different operational scenarios. A study of feasibility has been launched for the IBED PHTS (Integrated Blanket, Edge localized mode coils (ELMs) and Divertor Primary Heat Transfer System; it consists of five independent cooling trains (four operational and one in stand-by), one steam pressurizer, supply and return headers, ring manifolds and connections to the all in-vessel components (i.e. First Wall Blanket, Divertor, ELM, Diagnostics and other Ports clients). The dynamic behaviour of the IBED PHTS has been investigated by means of RELAP5{sup ®} code to simulate the response of the system during plasma pulse and baking operations. Due to the plasma heat deposition on the surfaces of the in-vessel components and subsequent increase in hot leg temperature, a large amount of water volume is transferred from the hot legs of the circuit to the surge-line of the pressurizer during each burn cycle. This causes rapid increase of pressure and temperature of the system and the following actions are proposed to counteract these variations: spray injection in the upper dome of the pressurizer from the Chemical and Volume Control System (CVCS) to reduce the pressure and active control of flow rates through heat exchangers and their bypass loops to regulate the heat transfer from the primary system to the environment via secondary and tertiary loops. This paper focuses on the prediction of the thermal hydraulic behaviour of the IBED PHTS during plasma pulses and baking scenarios, describing the various activity of the analysis, the geometrical assessment of the circuit and the modelling

  18. Status and subjects of thermal-hydraulic analysis for next-generation LWRs

    International Nuclear Information System (INIS)

    2000-03-01

    The status and subjects on thermal-hydraulic analysis for next-generation light water reactors (LWRs) with passive safety systems were surveyed through about 5 years until March 1999 by subcommittee on improvement of reactor thermal-hydraulic analysis codes under the nuclear code committee in Japan Atomic Energy Research Institute. Based on the survey results and discussion, the status and subjects on system analysis for various types of proposed reactor were summarized in 1998 and those on multidimensional two-phase flow analysis were also reviewed, since the multidimensional analysis was recognized as one of the most important subjects through the investigation on system analysis. In this report, the status and subjects for the following were summarized from the survey results and discussion in 1998 and 1999; (1) BWR neutronic/thermal-hydraulic coupled analysis, (2) Evaluation of passive safety system performance and (3) Gas-liquid two-phase flow analysis. The contents in this report are the forefront of thermal-hydraulic analysis for LWRs including test results from several large-scale facilities. We expect that the contents can offer a guideline to improve reactor thermal-hydraulic analysis codes in future. (author)

  19. Comparative study of Thermal Hydraulic Analysis Codes for Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yang Hoon; Jang, Mi Suk; Han, Kee Soo [Nuclear Engineering Service and Solution Co. Ltd., Daejeon (Korea, Republic of)

    2015-05-15

    Various codes are used for the thermal hydraulic analysis of nuclear reactors. The use of some codes among these is limited by user and some codes are not even open to general person. Thus, the use of alternative code is considered for some analysis. In this study, simple thermal hydraulic behaviors are analyzed using three codes to show that alternative codes are possible for the analysis of nuclear reactors. We established three models of the simple u-tube manometer using three different codes. RELAP5 (Reactor Excursion and Leak Analysis Program), SPACE (Safety and Performance Analysis CodE for nuclear power Plants), GOTHIC (Generation of Thermal Hydraulic Information for Containments) are selected for this analysis. RELAP5 is widely used codes for the analysis of system behavior of PWRs. SPACE has been developed based on RELAP5 for the analysis of system behavior of PWRs and licensing of the code is in progress. And GOTHIC code also has been widely used for the analysis of thermal hydraulic behavior in the containment system. The internal behavior of u-tube manometer was analyzed by RELAP5, SPACE and GOTHIC codes. The general transient behavior was similar among 3 codes. However, the stabilized status of the transient period analyzed by REPAP5 was different from the other codes. It would be resulted from the different physical models used in the other codes, which is specialized for the multi-phase thermal hydraulic behavior analysis.

  20. Analytical Thermal Field Theory Applicable to Oil Hydraulic Fluid Film Lubrication

    DEFF Research Database (Denmark)

    Johansen, Per; Roemer, Daniel Beck; Pedersen, Henrik C.

    2014-01-01

    the analytical approach provides an alternative to existing computationally expensive numerical methods. The paper presents the dimensional analysis, which provides the foundation for the derivation of the analytical approximation. Subsequently, the perturbation method is applied in order to find an asymptotic...

  1. Design and analysis of hydraulic ram water pumping system

    Science.gov (United States)

    Hussin, N. S. M.; Gamil, S. A.; Amin, N. A. M.; Safar, M. J. A.; Majid, M. S. A.; Kazim, M. N. F. M.; Nasir, N. F. M.

    2017-10-01

    The current pumping system (DC water pump) for agriculture is powered by household electricity, therefore, the cost of electricity will be increased due to the higher electricity consumption. In addition, the water needs to be supplied at different height of trees and different places that are far from the water source. The existing DC water pump can pump the water to 1.5 m height but it cost money for electrical source. The hydraulic ram is a mechanical water pump that suitable used for agriculture purpose. It can be a good substitute for DC water pump in agriculture use. The hydraulic ram water pumping system has ability to pump water using gravitational energy or the kinetic energy through flowing source of water. This project aims to analyze and develop the water ram pump in order to meet the desired delivery head up to 3 meter height with less operation cost. The hydraulic ram is designed using CATIA software. Simulation work has been done using ANSYS CFX software to validate the working concept. There are three design were tested in the experiment study. The best design reached target head of 3 m with 15% efficiency and flow rate of 11.82l/min. The results from this study show that the less diameter of pressure chamber and higher supply head will create higher pressure.

  2. Applications of the thermit code to 3D thermal hydraulic analysis of LWR cores

    International Nuclear Information System (INIS)

    Reed, W.H.

    1979-01-01

    The THERMIT code calculates the three-dimensional transient thermal hydraulic behavior of light water reactor cores. Its two-fluid dynamics equations for two-phase flow offer improved physical modelling capability needed in the context of calculation coupled to neutron kinetics for feedback. The numerical fluid dynamics method was chosen for reliability over a wider range of transients. An improved heat transfer numerical method is presented which gives better numerical stability and accuracy. A number of example calculations are discussed which give an idea of the power and flexibility of the code

  3. VIPRE-01. a thermal-hydraulic analysis code for reactor cores. Volume 1. Mathematical modeling

    International Nuclear Information System (INIS)

    Stewart, C.W.; Cuta, J.M.; Koontz, A.S.; Kelly, J.M.; Basehore, K.L.; George, T.L.; Rowe, D.S.

    1983-04-01

    VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear reactor core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This volume (Volume 1: Mathematical Modeling) explains the major thermal hydraulic models and supporting correlations in detail

  4. 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.

  5. Transmutation technology development; thermal hydraulic power analysis and structure analysis of the HYPER target beam window

    Energy Technology Data Exchange (ETDEWEB)

    Choi, J. H.; Ju, E. S.; Song, M. K.; Jeon, Y. Z. [Gyeongsang National University, Jinju (Korea)

    2002-03-01

    A thermal hydraulic power analysis, a structure analysis and optimization computation for some design factor for the design of spallation target suitable for HYPER with 1000 MW thermal power in this study was performed. Heat generation formula was used which was evaluated recently based on the LAHET code, mainly to find the maximum beam current under given computation conditions. Thermal hydraulic power of HYPER target system was calculated using FLUENT code, structure conducted by inputting the data into ANSYS. On the temp of beam windows and the pressure distribution calculated using FLUENT. Data transformation program was composed apply the data calculated using FLUENT being commercial CFD code and ANSYS being FEM code for CFX structure analysis. A basic study was conducted on various singular target to obtain fundamental data on the shape for optimum target design. A thermal hydraulic power analysis and structure analysis were conducted on the shapes of parabolic, uniform, scanning beams to choose the optimum shape of beam current analysis was done according to some turbulent model to simulate the real flow. To evaluate the reliability of numerical analysis result, benchmarking of FLUENT code reformed at SNU and Korea Advanced Institute of Science and Technology and it was compared to CFX in the possession of Korea Atomic Energy Research Institute and evaluated. Reliable deviation was observed in the results calculated using FLUENT code, but temperature deviation of about 200 .deg. C was observed in the result from CFX analysis at optimum design condition. Several benchmarking were performed on the basis of numerical analysis concerning conventional HYPER. It was possible to allow a beam arrests of 17.3 mA in the case of the {phi} 350 mm parabolic beam suggested to the optimum in nuclear transmutation when stress equivalent to VON-MISES was calculated to be 140 MPa. 29 refs., 109 figs. (Author)

  6. Ninth Thermal and Fluids Analysis Workshop Proceedings

    Science.gov (United States)

    Sakowski, Barbara (Compiler)

    1999-01-01

    The Ninth Thermal and Fluids Analysis Workshop (TFAWS 98) was held at the Ohio Aerospace Institute in Cleveland, Ohio from August 31 to September 4, 1998. The theme for the hands-on training workshop and conference was "Integrating Computational Fluid Dynamics and Heat Transfer into the Design Process." Highlights of the workshop (in addition to the papers published herein) included an address by the NASA Chief Engineer, Dr. Daniel Mulville; a CFD short course by Dr. John D. Anderson of the University of Maryland; and a short course by Dr. Robert Cochran of Sandia National Laboratories. In addition, lectures and hands-on training were offered in the use of several cutting-edge engineering design and analysis-oriented CFD and Heat Transfer tools. The workshop resulted in international participation of over 125 persons representing aerospace and automotive industries, academia, software providers, government agencies, and private corporations. The papers published herein address issues and solutions related to the integration of computational fluid dynamics and heat transfer into the engineering design process. Although the primary focus is aerospace, the topics and ideas presented are applicable to many other areas where these and other disciplines are interdependent.

  7. Seismic analysis of hydraulic control rod driving system

    International Nuclear Information System (INIS)

    Zheng, Yanhua; Bo, Hanliang; Dong, Duo

    2002-01-01

    A simplified mathematical model was developed for the Hydraulic Control Rod Driving System (HCRDS) of a 200 MW nuclear heating reactor, which incorporated the design of its chamfer-hole step cylinder, to analyze its seismic response characteristics. The control rod motion was analyzed for different sine-wave vibration loadings on platform vibrator. The vibration frequency domain and the minimum acceleration amplitude of the control rod needed to cause the control rod to step to its next setting were compared with the design acceleration amplitude spectrum. The system design was found to be safety within the calculated limits. The safety margin increased with increasing frequency. (author)

  8. 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

  9. Hydrodynamic analysis of clastic injection and hydraulic fracturing structures in the Jinding Zn-Pb deposit, Yunnan, China

    Directory of Open Access Journals (Sweden)

    Guoxiang Chi

    2012-01-01

    Full Text Available The Jinding Zn-Pb deposit has been generally considered to have formed from circulating basinal fluids in a relatively passive way, with fluid flow being controlled by structures and sedimentary facies, similar to many other sediments-hosted base metal deposits. However, several recent studies have revealed the presence of sand injection structures, intrusive breccias, and hydraulic fractures in the open pit of the Jinding deposit and suggested that the deposit was formed from explosive release of overpressured fluids. This study reports new observations of fluid overpressure-related structures from underground workings (Paomaping and Fengzishan, which show clearer crosscutting relationships than in the open pit. The observed structures include: 1 sand (±rock fragment dikes injecting into fractures in solidified rocks; 2 sand (±rock fragment bodies intruding into unconsolidated or semi-consolidated sediments; 3 disintegrated semi-consolidated sand bodies; and 4 veins and breccias formed from hydraulic fracturing of solidified rocks followed by cementation of hydrothermal minerals. The development of ore minerals (sphalerite in the cement of the various clastic injection and hydraulic fractures indicate that these structures were formed at the same time as mineralization. The development of hydraulic fractures and breccias with random orientation indicates small differential stress during mineralization, which is different from the stress field with strong horizontal shortening prior to mineralization. Fluid flow velocity may have been up to more than 11 m/s based on calculations from the size of the fragments in the clastic dikes. The clastic injection and hydraulic fracturing structures are interpreted to have formed from explosive release of overpressured fluids, which may have been related to either magmatic intrusions at depth or seismic activities that episodically tapped an overpressured fluid reservoir. Because the clastic injection

  10. Thermal-Hydraulic Analysis Tasks for ANAV NPPs in Support of Plant Operation and Control

    Directory of Open Access Journals (Sweden)

    L. Batet

    2007-11-01

    Full Text Available Thermal-hydraulic analysis tasks aimed at supporting plant operation and control of nuclear power plants are an important issue for the Asociación Nuclear Ascó-Vandellòs (ANAV. ANAV is the consortium that runs the Ascó power plants (2 units and the Vandellòs-II power plant. The reactors are Westinghouse-design, 3-loop PWRs with an approximate electrical power of 1000 MW. The Technical University of Catalonia (UPC thermal-hydraulic analysis team has jointly worked together with ANAV engineers at different levels in the analysis and improvement of these reactors. This article is an illustration of the usefulness of computational analysis for operational support. The contents presented were operational between 1985 and 2001 and subsequently changed slightly following various organizational adjustments. The paper has two different parts. In the first part, it describes the specific aspects of thermal-hydraulic analysis tasks related to operation and control and, in the second part, it briefly presents the results of three examples of analyses that were performed. All the presented examples are related to actual situations in which the scenarios were studied by analysts using thermal-hydraulic codes and prepared nodalizations. The paper also includes a qualitative evaluation of the benefits obtained by ANAV through thermal-hydraulic analyses aimed at supporting operation and plant control.

  11. Thermal-hydraulic analysis for changing feedwater check valve leakage rate testing methodology

    Energy Technology Data Exchange (ETDEWEB)

    Fuller, R.; Harrell, J.

    1996-12-01

    The current design and testing requirements for the feedwater check valves (FWCVs) at the Grand Gulf Nuclear Station are established from original licensing requirements that necessitate extremely restrictive air testing with tight allowable leakage limits. As a direct result of these requirements, the original high endurance hard seats in the FWCVs were modified with elastomeric seals to provide a sealing surface capable of meeting the stringent air leakage limits. However, due to the relatively short functional life of the elastomeric seals compared to the hard seats, the overall reliability of the sealing function actually decreased. This degraded performance was exhibited by frequent seal failures and subsequent valve repairs. The original requirements were based on limited analysis and the belief that all of the high energy feedwater vaporized during the LOCA blowdown. These phenomena would have resulted in completely voided feedwater lines and thus a steam environment within the feedwater leak pathway. To challenge these criteria, a comprehensive design basis accident analysis was developed using the RELAP5/MOD3.1 thermal-hydraulic code. Realistic assumptions were used to more accurately model the post-accident fluid conditions within the feedwater system. The results of this analysis demonstrated that no leak path exists through the feedwater lines during the reactor blowdown phase and that sufficient subcooled water remains in various portions of the feedwater piping to form liquid water loop seals that effectively isolate this leak path. These results provided the bases for changing the leak testing requirements of the FWCVs from air to water. The analysis results also established more accurate allowable leakage limits, determined the real effective margins associated with the FWCV safety functions, and led to design changes that improved the overall functional performance of the valves.

  12. Thermal-hydraulic analysis for changing feedwater check valve leakage rate testing methodology

    International Nuclear Information System (INIS)

    Fuller, R.; Harrell, J.

    1996-01-01

    The current design and testing requirements for the feedwater check valves (FWCVs) at the Grand Gulf Nuclear Station are established from original licensing requirements that necessitate extremely restrictive air testing with tight allowable leakage limits. As a direct result of these requirements, the original high endurance hard seats in the FWCVs were modified with elastomeric seals to provide a sealing surface capable of meeting the stringent air leakage limits. However, due to the relatively short functional life of the elastomeric seals compared to the hard seats, the overall reliability of the sealing function actually decreased. This degraded performance was exhibited by frequent seal failures and subsequent valve repairs. The original requirements were based on limited analysis and the belief that all of the high energy feedwater vaporized during the LOCA blowdown. These phenomena would have resulted in completely voided feedwater lines and thus a steam environment within the feedwater leak pathway. To challenge these criteria, a comprehensive design basis accident analysis was developed using the RELAP5/MOD3.1 thermal-hydraulic code. Realistic assumptions were used to more accurately model the post-accident fluid conditions within the feedwater system. The results of this analysis demonstrated that no leak path exists through the feedwater lines during the reactor blowdown phase and that sufficient subcooled water remains in various portions of the feedwater piping to form liquid water loop seals that effectively isolate this leak path. These results provided the bases for changing the leak testing requirements of the FWCVs from air to water. The analysis results also established more accurate allowable leakage limits, determined the real effective margins associated with the FWCV safety functions, and led to design changes that improved the overall functional performance of the valves

  13. Thermal-hydraulic analysis of LTS cables for the DEMO TF coil using simplified models

    Directory of Open Access Journals (Sweden)

    Lewandowska Monika

    2017-03-01

    Full Text Available The conceptual design activities for the DEMOnstration reactor (DEMO – the prototype fusion power plant – are conducted in Europe by the EUROfusion Consortium. In 2015, three design concepts of the DEMO toroidal field (TF coil were proposed by Swiss Plasma Center (EPFL-SPC, PSI Villigen, Italian National Agency for New Technologies (ENEA Frascati, and Atomic Energy and Alternative Energies Commission (CEA Cadarache. The proposed conductor designs were subjected to complete mechanical, electromagnetic, and thermal-hydraulic analyses. The present study is focused on the thermal-hydraulic analysis of the candidate conductor designs using simplified models. It includes (a hydraulic analysis, (b heat removal analysis, and (c assessment of the maximum temperature and the maximum pressure in each conductor during quench. The performed analysis, aimed at verification whether the proposed design concepts fulfil the established acceptance criteria, provides the information for further improvements of the coil and conductors design.

  14. An Analysis of Reconstituted Fluid Milk Pricing Policy

    OpenAIRE

    Glen D. Whipple

    1983-01-01

    This analysis suggests that alteration of the reconstituted fluid milk pricing provisions of federal and state milk market orders would have a substantial impact on market equilibrium. A reactive programming model of the U.S. milk market was used to simulate the effects of altered reconstituted fluid milk pricing policy. The solutions indicate that reconstituted fluid milk, as a lower cost alternative to fresh fluid milk, would make up a substantial portion of the fluid milk consumption in so...

  15. A cyclostationary multi-domain analysis of fluid instability in Kaplan turbines

    Science.gov (United States)

    Pennacchi, P.; Borghesani, P.; Chatterton, S.

    2015-08-01

    Hydraulic instabilities represent a critical problem for Francis and Kaplan turbines, reducing their useful life due to increase of fatigue on the components and cavitation phenomena. Whereas an exhaustive list of publications on computational fluid-dynamic models of hydraulic instability is available, the possibility of applying diagnostic techniques based on vibration measurements has not been investigated sufficiently, also because the appropriate sensors seldom equip hydro turbine units. The aim of this study is to fill this knowledge gap and to exploit fully, for this purpose, the potentiality of combining cyclostationary analysis tools, able to describe complex dynamics such as those of fluid-structure interactions, with order tracking procedures, allowing domain transformations and consequently the separation of synchronous and non-synchronous components. This paper will focus on experimental data obtained on a full-scale Kaplan turbine unit, operating in a real power plant, tackling the issues of adapting such diagnostic tools for the analysis of hydraulic instabilities and proposing techniques and methodologies for a highly automated condition monitoring system.

  16. Analysis of fluid structural instability in water

    International Nuclear Information System (INIS)

    Piccirillo, N.

    1997-02-01

    Recent flow testing of stainless steel hardware in a high pressure/high temperature water environment produced an apparent fluid-structural instability. The source of instability was investigated by studying textbook theory and by performing NASTRAN finite element analyses. The modal analyses identified the mode that was being excited, but the flutter instability analysis showed that the design is stable if minimal structural damping is present. Therefore, it was suspected that the test hardware was the root cause of the instability. Further testing confirmed this suspicion

  17. 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.

  18. In-depth analysis of bicycle hydraulic disc brakes

    Science.gov (United States)

    Maier, Oliver; Györfi, Benedikt; Wrede, Jürgen; Arnold, Timo; Moia, Alessandro

    2017-10-01

    Hydraulic Disc Brakes (HDBs) represent the most recent and innovative bicycle braking system. Especially Electric Bicycles (EBs), which are becoming more and more popular, are equipped with this powerful, unaffected by environmental influences, and low-wear type of brakes. As a consequence of the high braking performance, typical bicycle braking errors lead to more serious accidents. This is the starting point for the development of a Braking Dynamics Assistance system (BDA) to prevent front wheel lockup and nose-over (falling over the handlebars). One of the essential prerequisites for the system design is a better understanding of bicycle HDBs' characteristics. A physical simulation model and a test bench have been built for this purpose. The results of the virtual and real experiments conducted show a high correlation and allow valuable insights into HDBs on bicycles, which have not been studied scientifically in any depth so far.

  19. Validation of a multidimensional computational fluid dynamics model for subcooled flow boiling analysis

    Energy Technology Data Exchange (ETDEWEB)

    Braz Filho, Francisco A.; Caldeira, Alexandre D.; Borges, Eduardo M., E-mail: fbraz@ieav.cta.b, E-mail: alexdc@ieav.cta.b, E-mail: eduardo@ieav.cta.b [Instituto de Estudos Avancados (IEAv/CTA), Sao Jose dos Campos, SP (Brazil). Div. de Energia Nuclear

    2011-07-01

    In a heated vertical channel, the subcooled flow boiling regime occurs when the bulk fluid temperature is lower than the saturation temperature, but the fluid temperature reaches the saturation point near the channel wall. This phenomenon produces a significant increase in heat flux, limited by the critical heat flux. This study is particularly important to the thermal-hydraulics analysis of pressurized water reactors. The purpose of this work is the validation of a multidimensional model to analyze the subcooled flow boiling comparing the results with experimental data found in literature. The computational fluid dynamics code FLUENT was used with Eulerian multiphase model option. The calculated values of wall temperature in the liquid-solid interface presented an excellent agreement when compared to the experimental data. Void fraction calculations presented satisfactory results in relation to the experimental data in pressures of 15, 30 and 45 bars. (author)

  20. Validation of a multidimensional computational fluid dynamics model for subcooled flow boiling analysis

    International Nuclear Information System (INIS)

    Braz Filho, Francisco A.; Caldeira, Alexandre D.; Borges, Eduardo M.

    2011-01-01

    In a heated vertical channel, the subcooled flow boiling regime occurs when the bulk fluid temperature is lower than the saturation temperature, but the fluid temperature reaches the saturation point near the channel wall. This phenomenon produces a significant increase in heat flux, limited by the critical heat flux. This study is particularly important to the thermal-hydraulics analysis of pressurized water reactors. The purpose of this work is the validation of a multidimensional model to analyze the subcooled flow boiling comparing the results with experimental data found in literature. The computational fluid dynamics code FLUENT was used with Eulerian multiphase model option. The calculated values of wall temperature in the liquid-solid interface presented an excellent agreement when compared to the experimental data. Void fraction calculations presented satisfactory results in relation to the experimental data in pressures of 15, 30 and 45 bars. (author)

  1. Strain measurement in and analysis for hydraulic test of CPR1000 reactor pressure vessel

    International Nuclear Information System (INIS)

    Zhou Dan; Zhuang Dongzhen

    2013-01-01

    The strain measurement in hydraulic test of CPR1000 reactor pressure vessel performed in Dongfang Heavy Machinery Co., Ltd. is introduced. The detail test scheme and method was introduced and the measurement results of strain and stress was given. Meanwhile the finite element analysis was performed for the pressure vessel, which was generally matched with the measurement results. The reliability of strain measurement was verified and the high strength margin of vessel was shown, which would give a good reference value for the follow-up hydraulic tests and strength analysis of reactor pressure vessel. (authors)

  2. Neutronic and thermal hydraulic analysis for production of fission molybdenum-99 at Pakistan Research Reactor-1

    Energy Technology Data Exchange (ETDEWEB)

    Mushtaq, A. [Isotope Production Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad (Pakistan)], E-mail: mushtaqa@pinstech.org.pk; Iqbal, Massod; Bokhari, Ishtiaq Hussain; Mahmood, Tariq; Mahmood, Tayyab; Ahmad, Zahoor; Zaman, Qamar [Nuclear Engineering Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad (Pakistan)

    2008-02-15

    Neutronic and thermal hydraulic analysis for the fission molybdenum-99 production at PARR-1 has been performed. Low enriched uranium foil (<20% {sup 235}U) will be used as target material. Annular target designed by ANL (USA) will be irradiated in PARR-1 for the production of 100 Ci of molybdenum-99 at the end of irradiation, which will be sufficient to prepare required {sup 99}Mo/{sup 99m}Tc generators at PINSTECH and its supply in the country. Neutronic and thermal hydraulic analysis were performed using various codes. Data shows that annular targets can be safely irradiated in PARR-1 for production of required amount of fission molybdenum-99.

  3. Development and analysis of hydraulic-material transfer analysis code taking density current into account

    International Nuclear Information System (INIS)

    Saito, Hiroaki; Iriya, Yoshikazu

    1999-03-01

    It is an important issue to select site for the underground disposal of high level radioactive waste in a stable environment. Modelling of hydraulics in the freshwater/seawater boundaries is required. In this study, the analyzer code has been modified, in order to enable the analysis under more various conditions. Input/output functions were modified, after the functions of each module and major parameters were reconsidered. The modification included the change of input mode, from dialogue mode to file mode. Specifications of input/output and parameters are described. (A. Yamamoto)

  4. Numerical Analysis of Thermo Hydraulic Conditions in Car Fog Lamp

    Science.gov (United States)

    Ramšak, M.; Žunič, Z.; Škerget, L.; Jurejevčič, T.

    2009-08-01

    In the article a coupled heat transfer in the solid and fluid inside of a car fog lamp is presented using CFD software CFX [1]. All three basic principles of heat transfer are dealt with: conduction, convection and radiation. Two different approaches to radiation modeling are compared. Laminar and turbulent flow modeling are compared since computed Rayleight number indicates transitional flow regime. Results are in good agreement with the measurements.

  5. Approximation generation for correlations in thermal-hydraulic analysis codes

    International Nuclear Information System (INIS)

    Pereira, Luiz C.M.; Carmo, Eduardo G.D. do

    1997-01-01

    A fast and precise evaluation of fluid thermodynamic and transport properties is needed for the efficient mass, energy and momentum transport phenomena simulation related to nuclear plant power generation. A fully automatic code capable to generate suitable approximation for correlations with one or two independent variables is presented. Comparison in terms of access speed and precision with original correlations currently used shows the adequacy of the approximation obtained. (author). 4 refs., 8 figs., 1 tab

  6. 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.

  7. Fluid-Induced Vibration Analysis for Reactor Internals Using Computational FSI Method

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Jong Sung; Yi, Kun Woo; Sung, Ki Kwang; Im, In Young; Choi, Taek Sang [KEPCO E and C, Daejeon (Korea, Republic of)

    2013-10-15

    This paper introduces a fluid-induced vibration analysis method which calculates the response of the RVI to both deterministic and random loads at once and utilizes more realistic pressure distribution using the computational Fluid Structure Interaction (FSI) method. As addressed above, the FIV analysis for the RVI was carried out using the computational FSI method. This method calculates the response to deterministic and random turbulence loads at once. This method is also a simple and integrative method to get structural dynamic responses of reactor internals to various flow-induced loads. Because the analysis of this paper omitted the bypass flow region and Inner Barrel Assembly (IBA) due to the limitation of computer resources, it is necessary to find an effective way to consider all regions in the RV for the FIV analysis in the future. Reactor coolant flow makes Reactor Vessel Internals (RVI) vibrate and may affect the structural integrity of them. U. S. NRC Regulatory Guide 1.20 requires the Comprehensive Vibration Assessment Program (CVAP) to verify the structural integrity of the RVI for Fluid-Induced Vibration (FIV). The hydraulic forces on the RVI of OPR1000 and APR1400 were computed from the hydraulic formulas and the CVAP measurements in Palo Verde Unit 1 and Yonggwang Unit 4 for the structural vibration analyses. In this method, the hydraulic forces were divided into deterministic and random turbulence loads and were used for the excitation forces of the separate structural analyses. These forces are applied to the finite element model and the responses to them were combined into the resultant stresses.

  8. From the direct numerical simulation to system codes-perspective for the multi-scale analysis of LWR thermal hydraulics

    International Nuclear Information System (INIS)

    Bestion, D.

    2010-01-01

    A multi-scale analysis of water-cooled reactor thermal hydraulics can be used to take advantage of increased computer power and improved simulation tools, including Direct Numerical Simulation (DNS), Computational Fluid Dynamics (CFD) (in both open and porous mediums), and system thermalhydraulic codes. This paper presents a general strategy for this procedure for various thermalhydraulic scales. A short state of the art is given for each scale, and the role of the scale in the overall multi-scale analysis process is defined. System thermalhydraulic codes will remain a privileged tool for many investigations related to safety. CFD in porous medium is already being frequently used for core thermal hydraulics, either in 3D modules of system codes or in component codes. CFD in open medium allows zooming on some reactor components in specific situations, and may be coupled to the system and component scales. Various modeling approaches exist in the domain from DNS to CFD which may be used to improve the understanding of flow processes, and as a basis for developing more physically based models for macroscopic tools. A few examples are given to illustrate the multi-scale approach. Perspectives for the future are drawn from the present state of the art and directions for future research and development are given

  9. Thermal Hydraulic Analysis Of Thorium-Based Annular Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Han, Kyu Hyun [Korea Institute of Nuclear Safety, 19, Guseong-dong, Yuseong-gu, Daejeon, 305-338 (Korea, Republic of)

    2008-07-01

    Thermal hydraulic characteristics of thorium-based fuel assemblies loaded with annular seed pins have been analyzed using AMAP combined with MATRA, and compared with those of the existing thorium-based assemblies. MATRA and AMAP showed good agreements for the pressure drops at the internal sub-channels. The pressure drop generally increased in the cases of the assemblies loaded with annular seed pins due to the larger wetted perimeter, but an exception existed. In the inner sub-channels of the seed pins, mass fluxes were high due to the grid form losses in the outer sub-channels. About 43% of the heat generated from the seed pin flowed into the inner sub-channel and the rest into the outer sub-channel, which implies the inner to outer wall heat flux ratio was approximately 1.2. The maximum temperatures of the annular seed pins were slightly above 500 deg. C. The MDNBRs of the assemblies loaded with annular seed pins were higher than those of the existing assemblies. Due to the fact that inter-channel mixing cannot occur in the inner sub-channels, temperatures and enthalpies were higher in the inner sub-channels. (author)

  10. Thermal Hydraulic Analysis of K-DEMO Single Blanket Module for Preliminary Accident Analysis using MELCOR

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

    To develop the Korean fusion commercial reactor, preliminary design concept for K-DEMO (Korean fusion demonstration reactor) has been announced by NFRI (National Fusion Research Institute). This pre-conceptual study of K-DEMO has been introduced to identify technical details of a fusion power plant for the future commercialization of fusion reactor in Korea. Before this consideration, to build the K-DEMO, accident analysis is essential. Since the Fukushima accident, which is severe accident from unexpected disaster, safety analysis of nuclear power plant has become important. The safety analysis of both fission and fusion reactors is deemed crucial in demonstrating the low radiological effect of these reactors on the environment, during severe accidents. A risk analysis of K-DEMO should be performed, as a prerequisite for the construction of a fusion reactor. In this research, thermal-hydraulic analysis of single blanket module of K-DEMO is conducted for preliminary accident analysis for K-DEMO. Further study about effect of flow distributer is conducted. The normal K-DEMO operation condition is applied to the boundary condition and simulated to verify the material temperature limit using MELCOR. MELCOR is fully integrated, relatively fast-running code developed by Sandia National Laboratories. MELCOR had been used for Light Water Reactors and fusion reactor version of MELCOR was developed for ITER accident analysis. This study shows the result of thermal-hydraulic simulation of single blanket module with MELCOR which is severe accident code for nuclear fusion safety analysis. The difference of mass flow rate for each coolant channel with or without flow distributer is presented. With flow distributer, advantage of broadening temperature gradient in the K-DEMO blanket module and increase mass flow toward first wall is obtained. This can enhance the safety of K-DEMO blanket module. Most 13 .deg. C temperature difference in blanket module is obtained.

  11. Hydraulic characteristics of a radioactive waste repository groundwater analysis

    International Nuclear Information System (INIS)

    1986-09-01

    This report deals with the deep drilling program executed in northern Switzerland by the National Cooperative for the Storage of Radioactive Wastes (NAGRA). Investigations were aimed at describing geologic conditions with respect to waste disposal. One of the main effort was directed at identifying properties and behaviour of groundwater. Among the activities involved was the collecting of groundwater samples for laboratory investigations. The methods used and experience gained during drilling fluid tracing, water sampling and quality control of extracted groundwater are described. The technical constraints (depth, temperature, borehole diameter) led to the deployment of specialized equipment, parts of which were still at the experimental stage [fr

  12. Analysis of core physics and thermal-hydraulics results of control rod withdrawal experiments in the LOFT facility

    International Nuclear Information System (INIS)

    Varacalle, D.J. Jr.; Chen, T.H.; Harvego, E.A.; Ollikkala, H.

    1983-01-01

    Two anticipated transient experiments simulating an uncontrolled control rod withdrawal event in a pressurized water reactor (PWR) were conducted in the Loss-of-Fluid Test (LOFT) Facility at the Idaho National Engineering Laboratory. The scaled LOFT 50-MW(t) PWR includes most of the principal features of larger commercial PWRs. The experiments tested the ability of reactor analysis codes to accurately calculate core reactor physics and thermal-hydraulic phenomena in an integral reactor system. The initial conditions and scaled operating parameters for the experiments were representative of those expected in a commercial PWR. In both experiments, all four LOFT control rod assemblies were withdrawn at a reactor power of 37.5 MW and a system pressure of 14.8 MPa

  13. Thermal-hydraulic analysis of the semiscale Mod-1 blowdown heat transfer test series

    International Nuclear Information System (INIS)

    Cozzuol, J.M.

    1976-06-01

    Selected experimental thermal-hydraulic data from the recent Semiscale Mod-1 blowdown heat transfer test series are analyzed from an experimental viewpoint with emphasis on explaining those phenomena which influence core fluid behavior. Comparisons are made between the trends measured by the system instrumentation and the trends predicted by the RELAP4 computer code to aid in obtaining an understanding of the interactions between phenomena occurring in different parts of the system. The analyses presented in this report are valuable for evaluating the adequacy and improving the predictive capability of analytical models developed to predict the system response of a pressurized water reactor during a postulated loss-of-coolant accident

  14. Sensitivity Analysis for Hydraulic Behavior of Shiraz Plain Aquifer Using PMWIN

    Directory of Open Access Journals (Sweden)

    Ahmad Reza karimipour

    2011-07-01

    Full Text Available In this study, hydraulic behavior of Shirazplain aquifer, with an area of ~300 km2, was simulated using PMWIN model. The performance of recently constructed drainage system in the plain was modeled and parameters affecting hydraulic behavior of the aquifer were analyzed. Measured rainfall and evaporation rates in the plain, recharge and discharge rates through the aqueducts, Khoshk and Chenar Rahdar rivers, as well as amount of water discharged from production wells and recharge due to returned wastewater were considered in the model. Plain hydrodynamic coefficients were estimated via calibration and sensitivity analysis of the model was performed for four important parameters. Results showed that the model is most sensitive to recharge rate and hydraulic conductivity, respectively, such that a small variation in these two parameters causes a dramatic change in hydraulic head distribution in the plain. Furthermore, specific yield coefficient influences the seasonal water level fluctuations, but the aqueducts conductance coefficient only affects the aqueduct radius of influence with little effect on the overall hydraulic behavior of the plain.

  15. TRAC-B thermal-hydraulic analysis of the Black Fox boiling water reactor

    International Nuclear Information System (INIS)

    Martin, R.P.

    1993-05-01

    Thermal-hydraulic analyses of six hypothetical accident scenarios for the General Electric Black Fox Nuclear Project boiling water reactor were performed using the TRAC-BF1 computer code. This work is sponsored by the US Nuclear Regulatory Commission and is being done in conjunction with future analysis work at the US Nuclear Regulatory Commission Technical Training Center in Chattanooga, Tennessee. These accident scenarios were chosen to assess and benchmark the thermal-hydraulic capabilities of the Black Fox Nuclear Project simulator at the Technical Training Center to model abnormal transient conditions

  16. SUPERENERGY-2: a multiassembly, steady-state computer code for LMFBR core thermal-hydraulic analysis

    Energy Technology Data Exchange (ETDEWEB)

    Basehore, K.L.; Todreas, N.E.

    1980-08-01

    Core thermal-hydraulic design and performance analyses for Liquid Metal Fast Breeder Reactors (LMFBRs) require repeated detailed multiassembly calculations to determine radial temperature profiles and subchannel outlet temperatures for various core configurations and subassembly structural analyses. At steady-state, detailed core-wide temperature profiles are required for core restraint calculations and subassembly structural analysis. In addition, sodium outlet temperatures are routinely needed for each reactor operating cycle. The SUPERENERGY-2 thermal-hydraulic code was designed specifically to meet these designer needs. It is applicable only to steady-state, forced-convection flow in LMFBR core geometries.

  17. Thermal - hydraulic analysis of pressurizer water reactors using the model of open lateral boundary

    International Nuclear Information System (INIS)

    Borges, R.C.

    1980-10-01

    A computational method is developed for thermal-hydraulic analysis, where the channel may be analysed by more than one independent steps of calculation. This is made possible by the incorporation of the model of open lateral boundary in the code COBRA-IIIP, which permits the determination of the subchannel of an open lattice PWR core in a multi-step calculation. The thermal-hydraulic code COBRA-IIIP, developed at the Massachusetts Institute of Technology, is used as the basic model for this study. (Author) [pt

  18. SUPERENERGY-2: a multiassembly, steady-state computer code for LMFBR core thermal-hydraulic analysis

    International Nuclear Information System (INIS)

    Basehore, K.L.; Todreas, N.E.

    1980-08-01

    Core thermal-hydraulic design and performance analyses for Liquid Metal Fast Breeder Reactors (LMFBRs) require repeated detailed multiassembly calculations to determine radial temperature profiles and subchannel outlet temperatures for various core configurations and subassembly structural analyses. At steady-state, detailed core-wide temperature profiles are required for core restraint calculations and subassembly structural analysis. In addition, sodium outlet temperatures are routinely needed for each reactor operating cycle. The SUPERENERGY-2 thermal-hydraulic code was designed specifically to meet these designer needs. It is applicable only to steady-state, forced-convection flow in LMFBR core geometries

  19. Kinematic and Dynamic Simulation Analysis of Hydraulic Excavator’s Working Equipment based on ADAMS

    Directory of Open Access Journals (Sweden)

    Yu Hong Yan

    2016-01-01

    Full Text Available This paper establishes the 3D excavator model according to the actual size in UG firstly. Then based on the virtual simulation software ADAMS, the virtual prototype of the working device is built by adding interrelated constraints(kinematic pair and hydraulic cylinder driving function and load secondly. This paper gets the main parameters of the excavator working scope and the pressure situation change curves of point of each hydraulic cylinder by making kinematic and dynamic simulation analysis of hydraulic excavator’s working equipment at last. The conclusion providing design theory and improvement for the excavator’s working device, which also play an important role in improving the level of China’s excavator design, enhancing excavator’s performance and promoting the rapid development of excavator industry.

  20. 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.

  1. Multi-scale analysis of nuclear reactor thermal-hydraulics-first applications using the NEPTUNE platform

    International Nuclear Information System (INIS)

    Guelfi, A.; Boucker, M.; Mimouni, S.; Bestion, D.; Boudier, P.

    2005-01-01

    The NEPTUNE project aims at building a new two-phase flow thermal-hydraulics platform for nuclear reactor simulation. EDF (Electricite de France) and CEA (Commissariat a l'Energie Atomique) with the co-sponsorship of IRSN (Institut de Radioprotection et Surete Nucleaire) and FRAMATOME-ANP, are jointly developing the NEPTUNE multi-scale platform that includes new physical models and numerical methods for each of the computing scales. One usually distinguishes three different scales for industrial simulations: the 'system' scale, the 'component' scale (subchannel analysis) and CFD (Computational Fluid Dynamics). In addition DNS (Direct Numerical Simulation) can provide information at a smaller scale that can be useful for the development of the averaged scales. The NEPTUNE project also includes work on software architecture and research on new numerical methods for coupling codes since both are required to improve industrial calculations. All these R and D challenges have been defined in order to meet industrial needs and the underlying stakes (mainly the competitiveness and the safety of Nuclear Power Plants). This paper focuses on three high priority needs: DNB (Departure from Nucleate Boiling) prediction, directly linked to fuel performance; PTS (Pressurized Thermal Shock), a key issue when studying the lifespan of critical components and LBLOCA (Large Break Loss of Coolant Accident), a reference accident for safety studies. For each of these industrial applications, we provide a review of the last developments within the NEPTUNE platform and we present the first results. A particular attention is also given to physical validation and the needs for further experimental data. (authors)

  2. Sensitivity analysis of hydraulic fracturing Using an extended finite element method for the PKN model

    NARCIS (Netherlands)

    Garikapati, Hasini; Verhoosel, Clemens V.; van Brummelen, Harald; Diez, Pedro; Papadrakakis, M.; Papadopoulos, V.; Stefanou, G.; Plevris, V.

    2016-01-01

    Hydraulic fracturing is a process that is surrounded by uncertainty, as available data on e.g. rock formations is scant and available models are still rudimentary. In this contribution sensitivity analysis is carried out as first step in studying the uncertainties in the model. This is done to

  3. 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

  4. Uncertainty Evaluation of the SFR Subchannel Thermal-Hydraulic Modeling Using a Hot Channel Factors Analysis

    International Nuclear Information System (INIS)

    Choi, Sun Rock; Cho, Chung Ho; Kim, Sang Ji

    2011-01-01

    In an SFR core analysis, a hot channel factors (HCF) method is most commonly used to evaluate uncertainty. It was employed to the early design such as the CRBRP and IFR. In other ways, the improved thermal design procedure (ITDP) is able to calculate the overall uncertainty based on the Root Sum Square technique and sensitivity analyses of each design parameters. The Monte Carlo method (MCM) is also employed to estimate the uncertainties. In this method, all the input uncertainties are randomly sampled according to their probability density functions and the resulting distribution for the output quantity is analyzed. Since an uncertainty analysis is basically calculated from the temperature distribution in a subassembly, the core thermal-hydraulic modeling greatly affects the resulting uncertainty. At KAERI, the SLTHEN and MATRA-LMR codes have been utilized to analyze the SFR core thermal-hydraulics. The SLTHEN (steady-state LMR core thermal hydraulics analysis code based on the ENERGY model) code is a modified version of the SUPERENERGY2 code, which conducts a multi-assembly, steady state calculation based on a simplified ENERGY model. The detailed subchannel analysis code MATRA-LMR (Multichannel Analyzer for Steady-State and Transients in Rod Arrays for Liquid Metal Reactors), an LMR version of MATRA, was also developed specifically for the SFR core thermal-hydraulic analysis. This paper describes comparative studies for core thermal-hydraulic models. The subchannel analysis and a hot channel factors based uncertainty evaluation system is established to estimate the core thermofluidic uncertainties using the MATRA-LMR code and the results are compared to those of the SLTHEN code

  5. Thermal hydraulics of accelerator driven system: validation and analysis

    International Nuclear Information System (INIS)

    Kumari, I.; Khanna, A.

    2014-01-01

    This paper presents validation of RELAP5/Mod4.0 code modified to incorporate Lead Bismuth Eutectic (LBE)fluid properties for simulation of Accelerator Driven System (ADS) against Barone's NACIE facility.Results of mass flow rates (MFR), Reynolds number, heat transfer coefficients, temperatures and temperature difference for three powers (10.8, 21.7 and 32.5 kW) under natural circulation of LBE match with Barone's values within 7%,18%,37%, 5% and 8% of relative error respectively. After this validation Indian ADS for thermal power of 15 kW has been simulated. Simulated profiles of temperature, MFR and pressure drop LBE and air are reported. Air and LBE temperatures of present work match with literature design values within 5% of relative error. (author)

  6. Internal fluid flow management analysis for Clinch River Breeder Reactor Plant sodium pumps

    International Nuclear Information System (INIS)

    Cho, S.M.; Zury, H.L.; Cook, M.E.; Fair, C.E.

    1978-12-01

    The Clinch River Breeder Reactor Plant (CRBRP) sodium pumps are currently being designed and the prototype unit is being fabricated. In the design of these large-scale pumps for elevated temperature Liquid Metal Fast Breeder Reactor (LMFBR) service, one major design consideration is the response of the critical parts to severe thermal transients. A detailed internal fluid flow distribution analysis has been performed using a computer code HAFMAT, which solves a network of fluid flow paths. The results of the analytical approach are then compared to the test data obtained on a half-scale pump model which was tested in water. The details are presented of pump internal hydraulic analysis, and test and evaluation of the half-scale model test results

  7. A Dynamic Behavior of the Nuclear Test Rig with Coolant using the Fluid-Structural interaction Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Tae-Ho; Hong, Jintae; Ahn, Sung-Ho; Joung, Chang-Young; Jang, Seo-Yun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Yeon, Kon-Whi [Chungnam National University, Daejeon (Korea, Republic of)

    2016-10-15

    In this paper, the dynamic behavior of the test rig in the coolant flow simulator is evaluated by using the 2-way fluid-structural interaction analysis. The maximum value and location of the deformation and equivalent stress in the test rig is confirmed. The fluid-structural interaction analysis is applied to perform the fluid and structural analysis A fluid-structure interaction analysis is used to simulate the relationship between the deformation and hydraulic pressure. There are two types of fluid-structural interaction analysis. One is a 1-way direction analysis in which the hydraulic pressure is calculated by a CFD and transmitted to the surface of the structure, and a structural analysis is then performed. The other is a 2-way direction analysis that is performed by changing the data between the deformation of the structural and pressure of the coolant water for every time step. The location of the maximum deformation of the test rig is the bottom parts of the test rig. It is expected that the equivalent stress of the test rig is occurred. The maximum equivalent stress in the test rig under the circulation of the coolant is 90.1 MPa. The location of the maximum stress in the test rig is the connect part between the fuel rod and flow divider. A safety factor on the test rig is 3, approximately. The deformation motion of the test rig at the bottom part of the test rig is caused about the fluid-induced vibration. A test on the fluid-induced vibration of the test rig will be performed and compared with results of the analysis in further paper.

  8. Thermal-hydraulic analysis of SMART steam generator tube rupture using TASS/SMR-S code

    International Nuclear Information System (INIS)

    Kim, Hee-Kyung; Kim, Soo Hyoung; Chung, Young-Jong; Kim, Hyeon-Soo

    2013-01-01

    Highlights: ► The analysis was performed from the viewpoint of primary coolant leakage. ► The thermal hydraulic responses and the maximum leakage have been identified. ► There is no direct release into the atmosphere caused by an SGTR accident. ► SMART safety system works well against an SGTR accident. - Abstract: A steam generator tube rupture (SGTR) accident analysis for SMART was performed using the TASS/SMR-S code. SMART with a rated thermal power of 330 MWt has been developed at the Korea Atomic Energy Research Institute. The TASS/SMR-S code can analyze the thermal hydraulic phenomena of SMART in a full range of reactor operating conditions. An SGTR is one of the most important accidents from a thermal hydraulic and radiological viewpoint. A conservative analysis against a SMART SGTR was performed. The major concern of this analysis is to find the thermal hydraulic responses and maximum leakage amount from a primary to a secondary side caused by an SGTR accident. A sensitivity study searching for the conservative thermal hydraulic conditions, break locations, reactivity and other conditions was performed. The dominant parameters related with the integral leak are the high RCS pressure, low core inlet coolant temperature and low break location of the SG cassette. The largest integral leak comes to 28 tons in the most conservative case during 1 h. But there is no direct release into the atmosphere because the secondary system pressure is maintained with a sufficient margin for the design pressure. All leaks go to the condenser. The analysis results show that the primary and secondary system pressures are maintained below the design pressure and the SMART safety system is working well against an SGTR accident

  9. Application of hydraulic network analysis to motor operated butterfly valves in nuclear power plants

    International Nuclear Information System (INIS)

    Eldiwany, B.H.; Kalsi, M.S.

    1992-01-01

    This paper presents the application of hydraulic network analysis to evaluate the performance of butterfly valves in nuclear power plant applications. Required actuation torque for butterfly valves in high-flow applications is often dictated by peak dynamic torque. The peak dynamic torque, which occurs at some intermediate disc position, requires accurate evaluation of valve flow rate and pressure drop throughout the valve stroke. Valve flow rate and pressure drop are significantly affected by the valve flow characteristics and the hydraulic system characteristics, such as pumping capability, piping resistances, single and parallel flow paths, system hydrostatic pressure, and the location of the motor-operated valve (MOV) within the system. A hydraulic network analysis methodology that addresses the effect of these parameters on the MOV performance is presented. The methodology is based on well-established engineering principles. The application of this methodology requires detailed characteristics of both the MOV and the hydraulic system in which it is installed. The valve characteristics for this analysis can be obtained by flow testing or from the valve manufacturer. Even though many valve users, valve manufacturers, and engineering standards have recognized the importance of performing these analyses, none has provided a detailed procedure for doing so

  10. Cerebrospinal Fluid (CSF) Analysis: MedlinePlus Lab Test Information

    Science.gov (United States)

    ... K. Brunner & Suddarth's Handbook of Laboratory and Diagnostic Tests. 2nd Ed, Kindle. Philadelphia: Wolters Kluwer Health, Lippincott Williams & Wilkins; c2014. Cerebrospinal Fluid Analysis; 144 p. Johns ...

  11. Quantification of LOCA core damage frequency based on thermal-hydraulics analysis

    International Nuclear Information System (INIS)

    Cho, Jaehyun; Park, Jin Hee; Kim, Dong-San; Lim, Ho-Gon

    2017-01-01

    Highlights: • We quantified the LOCA core damage frequency based on the best-estimated success criteria analysis. • The thermal-hydraulic analysis using MARS code has been applied to Korea Standard Nuclear Power Plants. • Five new event trees with new break size boundaries and new success criteria were developed. • The core damage frequency is 5.80E−07 (/y), which is 12% less than the conventional PSA event trees. - Abstract: A loss-of-coolant accident (LOCA) has always been significantly considered one of the most important initiating events. However, most probabilistic safety assessment models, up to now, have undoubtedly adopted the three groups of LOCA, and even an exact break size boundary that used in WASH-1400 reports was published in 1975. With an awareness of the importance of a realistic PSA for a risk-informed application, several studies have tried to find the realistic thermal-hydraulic behavior of a LOCA, and improve the PSA model. The purpose of this research is to obtain realistic results of the LOCA core damage frequency based on a success criteria analysis using the best-estimate thermal-hydraulics code. To do so, the Korea Standard Nuclear Power Plant (KSNP) was selected for this study. The MARS code was used for a thermal hydraulics analysis and the AIMS code was used for the core damage quantification. One of the major findings in the thermal hydraulics analysis was that the decay power is well removed by only a normal secondary cooling in LOCAs of below 1.4 in and by only a high pressure safety injection in LOCAs of 0.8–9.4 in. Based on the thermal hydraulics results regarding new break size boundaries and new success criteria, five new event trees (ETs) were developed. The core damage frequency of new LOCA ETs is 5.80E−07 (/y), which is 12% less than the conventional PSA ETs. In this research, we obtained not only thermal-hydraulics characteristics for the entire break size of a LOCA in view of the deterministic safety

  12. Quantification of LOCA core damage frequency based on thermal-hydraulics analysis

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jaehyun, E-mail: chojh@kaeri.re.kr; Park, Jin Hee; Kim, Dong-San; Lim, Ho-Gon

    2017-04-15

    Highlights: • We quantified the LOCA core damage frequency based on the best-estimated success criteria analysis. • The thermal-hydraulic analysis using MARS code has been applied to Korea Standard Nuclear Power Plants. • Five new event trees with new break size boundaries and new success criteria were developed. • The core damage frequency is 5.80E−07 (/y), which is 12% less than the conventional PSA event trees. - Abstract: A loss-of-coolant accident (LOCA) has always been significantly considered one of the most important initiating events. However, most probabilistic safety assessment models, up to now, have undoubtedly adopted the three groups of LOCA, and even an exact break size boundary that used in WASH-1400 reports was published in 1975. With an awareness of the importance of a realistic PSA for a risk-informed application, several studies have tried to find the realistic thermal-hydraulic behavior of a LOCA, and improve the PSA model. The purpose of this research is to obtain realistic results of the LOCA core damage frequency based on a success criteria analysis using the best-estimate thermal-hydraulics code. To do so, the Korea Standard Nuclear Power Plant (KSNP) was selected for this study. The MARS code was used for a thermal hydraulics analysis and the AIMS code was used for the core damage quantification. One of the major findings in the thermal hydraulics analysis was that the decay power is well removed by only a normal secondary cooling in LOCAs of below 1.4 in and by only a high pressure safety injection in LOCAs of 0.8–9.4 in. Based on the thermal hydraulics results regarding new break size boundaries and new success criteria, five new event trees (ETs) were developed. The core damage frequency of new LOCA ETs is 5.80E−07 (/y), which is 12% less than the conventional PSA ETs. In this research, we obtained not only thermal-hydraulics characteristics for the entire break size of a LOCA in view of the deterministic safety

  13. Transient flow analysis of the single cylinder for the control rod hydraulic driving system

    International Nuclear Information System (INIS)

    Sun, Xinming; Qin, Benke; Bo, Hanliang

    2017-01-01

    Highlights: • The control rod hydraulic driving system(CRHDS) is a new type of built-in control rod drive technology. The hydraulic cylinder is the main component of the CRHDS. • Transient flow phenomenon in the CRHDS is studied by experiments under different working conditions. • The working mechanism of the hydraulic cylinder step motion and the key characteristic parameters are analyzed based on the experimental results. - Abstract: The control rod hydraulic driving system (CRHDS) is a new type of built-in control rod drive technology. In the CRHDS the pulse flow from the pump into the hydraulic cylinder of the control rod hydraulic drive mechanism (CRHDM) is regulated by the integrated valve to perform the step motion of the reactor control rod. Transient flow occurs in the CRHDS during control rod step motion process which is studied by experiments. The time-history curves of flow rate, pressure and inner cylinder displacement were analyzed, and the results show that the water hammer pressure peak during the step-up motion is high, while there are no obvious pressure fluctuations in the corresponding step-down motion. In the step-up process, the pressure fluctuation amplitude increases with the increase of CRHDS driving pressure. The step-up time and the pressure increasing time before step-up decreases with the driving pressure. The step-up pressure increases with the driving pressure. In the step-down process, the step-down time, the step-down pressure and the pressure decreasing time before step-down do not change with the increase of the driving pressure. The experimental results lay the base for the working principle and vibration reduction analysis of the CRHDS and it’s also helpful for improvement of the working performance of the key facilities and instruments of the CRHDS loop.

  14. Calculation and analysis of thermal–hydraulics fluctuations in pressurized water reactors

    International Nuclear Information System (INIS)

    Malmir, Hessam; Vosoughi, Naser

    2015-01-01

    Highlights: • Single-phase thermal–hydraulics noise equations are originally derived in the frequency domain. • The fluctuations of all the coolant parameters are calculated, without any simplifying assumptions. • The radial distribution of the temperature fluctuations in the fuel, gap and cladding are taken into account. • The closed-loop calculations are performed by means of the point kinetics noise theory. • Both the space- and frequency-dependence of the thermal–hydraulics fluctuations are analyzed. - Abstract: Analysis of thermal–hydraulics fluctuations in pressurized water reactors (e.g., local and global temperature or density fluctuations, as well as primary and charging pumps fluctuations) has various applications in calculation or measurement of the core dynamical parameters (temperature or density reactivity coefficients) in addition to thermal–hydraulics surveillance and diagnostics. In this paper, the thermal–hydraulics fluctuations in PWRs are investigated. At first, the single-phase thermal–hydraulics noise equations (in the frequency domain) are originally derived, without any simplifying assumptions. The fluctuations of all the coolant parameters, as well as the radial distribution of the temperature fluctuations in the fuel, gap and cladding are taken into account. Then, the derived governing equations are discretized using the finite volume method (FVM). Based on the discretized equations and the proposed algorithm of solving, a single heated channel noise calculation code (SHC-Noise) is developed, by which the steady-state and fluctuating parameters of PWR fuel assemblies can be calculated. The noise sources include the inlet coolant temperature and velocity fluctuations, in addition to the power density noises. The developed SHC-Noise code is benchmarked in different cases and scenarios. Furthermore, to show the effects of the power feedbacks, the closed-loop calculations are performed by means of the point kinetics noise

  15. Methodology for thermal hydraulic conceptual design and performance analysis of KALIMER core

    International Nuclear Information System (INIS)

    Young-Gyun Kim; Won-Seok Kim; Young-Jin Kim; Chang-Kue Park

    2000-01-01

    This paper summarizes the methodology for thermal hydraulic conceptual design and performance analysis which is used for KALIMER core, especially the preliminary methodology for flow grouping and peak pin temperature calculation in detail. And the major technical results of the conceptual design for the KALIMER 98.03 core was shown and compared with those of KALIMER 97.07 design core. The KALIMER 98.03 design core is proved to be more optimized compared to the 97.07 design core. The number of flow groups are reduced from 16 to 11, and the equalized peak cladding midwall temperature from 654 deg. C to 628 deg. C. It was achieved from the nuclear and thermal hydraulic design optimization study, i.e. core power flattening and increase of radial blanket power fraction. Coolant flow distribution to the assemblies and core coolant/component temperatures should be determined in core thermal hydraulic analysis. Sodium flow is distributed to core assemblies with the overall goal of equalizing the peak cladding midwall temperatures for the peak temperature pin of each bundle, thus pin cladding damage accumulation and pin reliability. The flow grouping and the peak pin temperature calculation for the preliminary conceptual design is performed with the modules ORFCE-F60 and ORFCE-T60 respectively. The basic subchannel analysis will be performed with the SLTHEN code, and the detailed subchannel analysis will be done with the MATRA-LMR code which is under development for the K-Core system. This methodology was proved practical to KALIMER core thermal hydraulic design from the related benchmark calculation studies, and it is used to KALIMER core thermal hydraulic conceptual design. (author)

  16. Thermal-hydraulic analysis of spent fuel storage systems

    International Nuclear Information System (INIS)

    Rector, D.R.; Wheeler, C.L.; Lombardo, N.J.

    1987-01-01

    This paper describes the COBRA-SFS (Spent Fuel Storage) computer code, which is designed to predict flow and temperature distributions in spent nuclear fuel storage and transportation systems. The decay heat generated by spent fuel in a dry storage cask is removed through a combination of conduction, natural convection, and thermal radiation. One major advantage of COBRA-SFS is that fluid recirculation within the cask is computed directly by solving the mass and momentum conservation equations. In addition, thermal radiation heat transfer is modeled using detailed radiation exchange factors based on quarter-rod segments. The equations governing mass, momentum, and energy conservation for incompressible flows are presented, and the semi-implicit solution method is described. COBRA-SFS predictions are compared to temperature data from a spent fuel storage cask test and the effect of different fill media on the cladding temperature distribution is discussed. The effect of spent fuel consolidation on cask thermal performance is also investigated. 16 refs., 6 figs., 2 tabs

  17. Thermal-hydraulic experiment and analysis for interim dry storage of spent nuclear fuel

    International Nuclear Information System (INIS)

    Yoo, Seung Hun

    2011-02-01

    The experimental and numerical studies of interim storages for nuclear spent fuels have been performed to investigate thermal-hydraulic characteristics of the dry storage systems and to propose new methodologies for the analysis and the design. Three separate researches have been performed in the present study: (a) Development of a scaling methodology and thermal-hydraulic experiment of a single spent fuel assembly simulating a dry storage cask: (b) Full-scope simulation of a dry storage cask by the use of Computational Fluid Dynamics (CFD) code: (c) Thermal-hydraulic design of a tunnel-type interim storage facility. In the first study, a scaling methodology has been developed to design a scaled-down canister. The scaling was performed in two steps. For the first step, the height of a spent fuel assembly was reduced from full height to half height. In order to consider the effect of height reduction on the natural convection, the scaling law of Ishii and Kataoka (1984) was employed. For the second step, the quantity of spent fuel assemblies was reduced from multiple assemblies to a single assembly. The scaling methodology was validated through the comparison with the experiment of the TN24P cask. The Peak Cladding Temperature (PCT), temperature gradients, and the axial and radial temperature distribution in the nondimensional forms were in good agreement with the experimental data. Based on the developed methodology, we have performed a single assembly experiment which was designed to simulate the full scale of the TN24P cask. The experimental data was compared with the CFD calculations. It turns out that their PCTs were less than the maximum allowable temperature for the fuel cladding and that the differences of their PCTs were agreed within 3 .deg. C, which was less than measurement uncertainty. In the second study, the full-scope simulations of the TN24P cask were performed by FLUENT. In order to investigate the sensitivity of the numerical and physical

  18. 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)

  19. Thermal-hydraulic analysis of a 600 MW supercritical CFB boiler with low mass flux

    International Nuclear Information System (INIS)

    Pan Jie; Yang Dong; Chen Gongming; Zhou Xu; Bi Qincheng

    2012-01-01

    Supercritical Circulating Fluidized Bed (CFB) boiler becomes an important development trend for coal-fired power plant and thermal-hydraulic analysis is a key factor for the design and operation of water wall. According to the boiler structure and furnace-sided heat flux, the water wall system of a 600 MW supercritical CFB boiler is treated in this paper as a flow network consisting of series-parallel loops, pressure grids and connecting tubes. A mathematical model for predicting the thermal-hydraulic characteristics in boiler heating surface is based on the mass, momentum and energy conservation equations of these components, which introduces numerous empirical correlations available for heat transfer and hydraulic resistance calculation. Mass flux distribution and pressure drop data in the water wall at 30%, 75% and 100% of the boiler maximum continuous rating (BMCR) are obtained by iteratively solving the model. Simultaneity, outlet vapor temperatures and metal temperatures in water wall tubes are estimated. The results show good heat transfer performance and low flow resistance, which implies that the water wall design of supercritical CFB boiler is applicable. - Highlights: → We proposed a model for thermal-hydraulic analysis of boiler heating surface. → The model is applied in a 600 MW supercritical CFB boiler. → We explore the pressure drop, mass flux and temperature distribution in water wall. → The operating safety of boiler is estimated. → The results show good heat transfer performance and low flow resistance.

  20. Thermal modeling of a hydraulic hybrid vehicle transmission based on thermodynamic analysis

    International Nuclear Information System (INIS)

    Kwon, Hyukjoon; Sprengel, Michael; Ivantysynova, Monika

    2016-01-01

    Hybrid vehicles have become a popular alternative to conventional powertrain architectures by offering improved fuel efficiency along with a range of environmental benefits. Hydraulic Hybrid Vehicles (HHV) offer one approach to hybridization with many benefits over competing technologies. Among these benefits are lower component costs, more environmentally friendly construction materials, and the ability to recover a greater quantity of energy during regenerative braking which make HHVs partially well suited to urban environments. In order to further the knowledge base regarding HHVs, this paper explores the thermodynamic characteristics of such a system. A system model is detailed for both the hydraulic and thermal components of a closed circuit hydraulic hybrid transmission following the FTP-72 driving cycle. Among the new techniques proposed in this paper is a novel method for capturing rapid thermal transients. This paper concludes by comparing the results of this model with experimental data gathered on a Hardware-in-the-Loop (HIL) transmission dynamometer possessing the same architecture, components, and driving cycle used within the simulation model. This approach can be used for several applications such as thermal stability analysis of HHVs, optimal thermal management, and analysis of the system's thermodynamic efficiency. - Highlights: • Thermal modeling for HHVs is introduced. • A model for the hydraulic and thermal system is developed for HHVs. • A novel method for capturing rapid thermal transients is proposed. • The thermodynamic system diagram of a series HHV is predicted.

  1. Transient thermal-hydraulic/neutronic analysis in a VVER-1000 reactor core

    International Nuclear Information System (INIS)

    Seyed khalil Mousavian; Mohammad Mohsen Ertejaei; Majid Shahabfar

    2005-01-01

    Full text of publication follows: Nowadays, coupled thermal-hydraulic and three-dimensional neutronic codes in order to consider different feedback effects is state of the art subject in nuclear engineering researches. In this study, RELAP5/COBRA and WIMS/CITATION codes are implemented to investigate the VVER-1000 reactor core parameters during Large Break Loss of Coolant Accident (LB-LOCA). In a LB-LOCA, the primary side pressure, coolant density and fuel temperature strongly decrease but the cladding temperature experiences a strong peak. For this purpose, the RELAP5 Best Estimate (BE) system code is used to simulate the LB-LOCA analysis in VVER-1000 nuclear thermal-hydraulic loops. Also, the modified COBRA-IIIc software as a sub-channel analysis code is applied for modeling of VVER-1000 reactor core. Moreover, WIMS and CITATION as a cross section and 3-D neutron flux codes are coupled with thermal-hydraulic codes with the aim of consider the spatial effects through the reactor core. For this reason, suitable software is developed to link and speed up the coupled thermalhydraulic and three-dimensional neutronic calculations. This software utilizes of external coupling concept in order to integrate thermal-hydraulic and neutronic calculations. (authors)

  2. Single-channel model for steady thermal-hydraulic analysis in nuclear reactor

    International Nuclear Information System (INIS)

    Zhang Xiaoying; Huang Yuanyuan

    2010-01-01

    This article established a single-channel model for steady analysis in the reactor and an example of thermal-hydraulic analysis was made by using this model, including the Maximum heat flux density of fuel element, enthalpy, Coolant flow, various kinds of pressure drop, enthalpy increase in average tube and thermal tube. I also got the Coolant temperature distribution and the fuel element temperature distribution and analysis of the final result. The results show that some relevant parameters which we got in this paper are well coincide with the actual operating parameters. It is also show that the single-channel model can be used to the steady thermal-hydraulic analysis. (authors)

  3. Analysis and evaluation of the Dual Fluid Reactor concept

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Xiang

    2017-06-27

    The Dual Fluid Reactor is a molten salt fast reactor developed by IFK in Berlin based on the Gen-IV Molten-Salt Reactor concept and the Liquid-Metal Cooled Reactor. The design aims to combine these two concepts to improve these two concepts. The Dissertation focuses on the concept and performs diverse calculations and estimations on the subjects of neutron physics, depletion and thermal-hydraulic behaviors to validate the new features of the concept. Based on the results it is concluded that this concept is feasible to its desired purpose and with great potential.

  4. Adaptive Finite Element-Discrete Element Analysis for Microseismic Modelling of Hydraulic Fracture Propagation of Perforation in Horizontal Well considering Pre-Existing Fractures

    Directory of Open Access Journals (Sweden)

    Yongliang Wang

    2018-01-01

    Full Text Available Hydrofracturing technology of perforated horizontal well has been widely used to stimulate the tight hydrocarbon reservoirs for gas production. To predict the hydraulic fracture propagation, the microseismicity can be used to infer hydraulic fractures state; by the effective numerical methods, microseismic events can be addressed from changes of the computed stresses. In numerical models, due to the challenges in accurately representing the complex structure of naturally fractured reservoir, the interaction between hydraulic and pre-existing fractures has not yet been considered and handled satisfactorily. To overcome these challenges, the adaptive finite element-discrete element method is used to refine mesh, effectively identify the fractures propagation, and investigate microseismic modelling. Numerical models are composed of hydraulic fractures, pre-existing fractures, and microscale pores, and the seepage analysis based on the Darcy’s law is used to determine fluid flow; then moment tensors in microseismicity are computed based on the computed stresses. Unfractured and naturally fractured models are compared to assess the influences of pre-existing fractures on hydrofracturing. The damaged and contact slip events were detected by the magnitudes, B-values, Hudson source type plots, and focal spheres.

  5. Thermo-hydraulic and structural analysis for finger-based concept of ITER blanket first wall

    International Nuclear Information System (INIS)

    Kim, Byoung-Yoon; Ahn, Hee-Jae

    2011-01-01

    The blanket first wall is one of the main plasma facing components in ITER tokamak. The finger-typed first wall was proposed through the current design progress by ITER organization. In this concept, each first wall module is composed of a beam and twenty fingers. The main function of the first wall is to remove efficiently the high heat flux loading from the fusion plasma during its operation. Therefore, the thermal and structural performance should be investigated for the proposed finger-based design concept of first wall. The various case studies were performed for a unit finger model considering different loading conditions. The finite element model was made for a half of a module using symmetric boundary conditions to reduce the computational effort. The thermo-hydraulic analysis was performed to obtain the pressure drop and temperature profiles. Then the structural analysis was carried out using the maximum temperature distribution obtained in thermo-hydraulic analysis. Finally, the transient thermo-hydraulic analysis was performed for the generic first wall module to obtain the temperature evolution history considering cyclic heat flux loading with nuclear heating. After that, the thermo-mechanical analysis was performed at the time step when the maximum temperature gradient was occurred. Also, the stress analysis was performed for the component with a finger and a beam to check the residual stress of the component after thermal shrinkage assembly.

  6. Thermo-hydraulic Analysis of a Water-cooled Printed Circuit Heat Exchanger in a Small-scale Nitrogen Loop

    International Nuclear Information System (INIS)

    Kim, Chan Soo; Hong, Sung Deok; Kim, Min Hwan; Shim, Jaesool; Lee, Gyung Dong

    2013-01-01

    The development of high-temperature heat exchangers is very important because of its higher operation temperature and pressure than those of common light water reactors and industrial process plants. In particular, the intermediate heat exchanger is a key-challenged high temperature component in a Very High Temperature gas-cooled Reactor (VHTR). A printed circuit heat exchanger is one of the candidates for an intermediate heat exchanger in a VHTR. The printed circuit heat exchanger (PCHE) was developed and commercialized by HEATRIC. The compactness is better than any other heat exchanger types, because its core matrices are fabricated by diffusion bonding with photo-chemically etched micro-channels. Various tests and analysis have been performed to verify the performance of PCHE. The thermal stress analysis of the high temperature PCHE is necessary to endure the extremely operation condition of IHX. In this study, the thermo-hydraulic analysis for the laboratory-scale PCHE is performed to provide the input data for the boundary conditions of a structural analysis. The results from the first-principal calculation are compared with those from computational fluid dynamics code analysis. COMSOL 4.3a analysis is successfully performed at the uniform pressure drop condition in a set of flow channel stacks. The heat-exchanged region concentrated to the nitrogen inlet cause the uniform mass velocity distribution in the channels, therefore there is little difference between two analytical results

  7. Analysis of the jet pipe electro-hydraulic servo valve with finite element methods

    Directory of Open Access Journals (Sweden)

    Kaiyu Zhao

    2018-01-01

    Full Text Available The dynamic characteristics analysis about the jet pipe electro-hydraulic servo valve based on experience and mathematical derivation was difficult and not so precise. So we have analysed the armature feedback components, torque motor and jet pipe receiver in electrohydraulic servo valve by sophisticated finite element analysis tools respectively and have got physical meaning data on these parts. Then the data were fitted by Matlab and the mathematical relationships among them were calculated. We have done the dynamic multi-physical fields’ Simulink co-simulation using above mathematical relationship, and have got the input-output relationship of the overall valve, the frequency response and step response. This work can show the actual working condition accurately. At the same time, we have considered the materials and the impact of the critical design dimensions in the finite element analysis process. It provides some new ideas to the overall design of jet pipe electro-hydraulic servo valve.

  8. Sensitivity Analysis of Hydraulic Methods Regarding Hydromorphologic Data Derivation Methods to Determine Environmental Water Requirements

    Directory of Open Access Journals (Sweden)

    Alireza Shokoohi

    2015-07-01

    Full Text Available This paper studies the accuracy of hydraulic methods in determining environmental flow requirements. Despite the vital importance of deriving river cross sectional data for hydraulic methods, few studies have focused on the criteria for deriving this data. The present study shows that the depth of cross section has a meaningful effect on the results obtained from hydraulic methods and that, considering fish as the index species for river habitat analysis, an optimum depth of 1 m should be assumed for deriving information from cross sections. The second important parameter required for extracting the geometric and hydraulic properties of rivers is the selection of an appropriate depth increment; ∆y. In the present research, this parameter was found to be equal to 1 cm. The uncertainty of the environmental discharge evaluation, when allocating water in areas with water scarcity, should be kept as low as possible. The Manning friction coefficient (n is an important factor in river discharge calculation. Using a range of "n" equal to 3 times the standard deviation for the study area, it is shown that the influence of friction coefficient on the estimation of environmental flow is much less than that on the calculation of river discharge.

  9. Trend analysis of troubles caused by thermal-hydraulic phenomena at nuclear power plants

    International Nuclear Information System (INIS)

    Komatsu, Teruo

    2010-01-01

    The Institute of Nuclear Safety System (INSS) is promoting researches to improve the safety and reliability of nuclear power plants. In the present study, our attention was focused on troubles attributed to thermal-hydraulic phenomena in particular, trend analysis were carried out to learn lessons from these troubles and to prevent their recurrence. Through our survey, we found the following two points. First, many thermal-hydraulics related troubles can be attributed to design faults, since we found some events in foreign countries took place after inadequate facility renovation. To ensure appropriate design verification, it is important to take account of state-of-the-art science and technology and at the same time to pay attention to the compatibility with the initial design concept. Second point, thermal-hydraulic related troubles are common and recurrent to nuclear power plants worldwide. Japanese utilities are planning to introduce some of overseas experiences to their plants, such as power uprate and renovations of aged facilities. It is important to learn lessons from experiences paying close attention continuously to overseas trouble events, including thermal-hydraulics related events, and to use them to improve safety and reliability of nuclear power plants. (author)

  10. Discussion on sealing performance required in disposal system. Hydraulic analysis of tunnel intersections

    International Nuclear Information System (INIS)

    Sugita, Yutaka; Takahashi, Yoshiaki; Uragami, Manabu; Kitayama, Kazumi; Fujita, Tomoo; Kawakami, Susumu; Yui, Mikazu; Umeki, Hiroyuki; Miyamoto, Yoichi

    2005-09-01

    The sealing performance of a repository must be considered in the safety assessment of the geological disposal system of the high-level radioactive waste. NUMO and JNC established 'Technical Commission on Sealing Technology of Repository' based on the cooperation agreement. The objectives of this commission are to present the concept on the sealing performance required in the disposal system and to develop the direction for future R and D programme for design requirements of closure components (backfilling material, clay plug, etc.) in the presented concept. In the first phase of this commission, the current status of domestic and international sealing technologies were reviewed; and repository components and repository environments were summarized subsequently, the hydraulic analysis of tunnel intersections, where a main tunnel and a disposal tunnel in a disposal panel meet, were performed, considering components in and around the engineered barrier system (EBS). Since all tunnels are connected in the underground facility, understanding the hydraulic behaviour of tunnel intersections is an important issue to estimate migration of radionuclides from the EBS and to evaluate the required sealing performance in the disposal system. In the analytical results, it was found that the direction of hydraulic gradient, hydraulic conductivities of concrete and backfilling materials and the position of clay plug had impact on flow condition around the EBS. (author)

  11. Techniques for the thermal/hydraulic analysis of LMFBR check valves

    International Nuclear Information System (INIS)

    Cho, S.M.; Kane, R.S.

    1979-01-01

    A thermal/hydraulic analysis of the check valves in liquid sodium service for LMFBR plants is required to provide temperature data for thermal stress analysis of the valves for specified transient conditions. Because of the complex three-dimensional flow pattern within the valve, the heat transfer analysis techniques for less complicated shapes could not be used. This paper discusses the thermal analysis techniques used to assure that the valve stress analysis is conservative. These techniques include a method for evaluating the recirculating flow patterns and for selecting appropriately conservative heat transfer correlations in various regions of the valve

  12. Thermal hydraulic analysis of Pb-Bi cooled HYPER fuel assemblies using SLTHEN code

    International Nuclear Information System (INIS)

    Tak, Nam Il; Song, Tae Y.; Park, Won S.; Kim, Chang Hyun

    2002-12-01

    In the present work, the existing SLTHEN code, which had been originally developed for subchannel analysis of sodium cooled fast reactors, was modified and applied to the Pb-Bi cooled HYPER core which consists of 237 fuel assemblies (TRU assemblies). In the analysis of single fuel assembly having chopped cosine power profile, the validation and the assessment of usefulness of the modified SLTHEN were focused. In the quantitative comparison, the results of the modified SLTHEN agreed well with those of analytical calculations and of MATRA. For the qualitative approaches, the sensitivity calculations for intra-assembly gap flow and turbulent mixing parameter were used. The sensitivity analysis results showed that the modified SLTHEN can provide reasonable simulations of subchannel thermal hydraulics. In particular, turbulent mixing parameter which is known as the most uncertain parameter in subchannel analyses did not affect largely the maximum cladding temperature. Therefore, it can be said that the results of single assembly show the usefulness of the modified SLTHEN code for thermal hydraulic analysis and design of HYPER under the conceptual design stage. In order to assess intra-assembly heat transfer, subchannel analyses were implemented for two types of 7 assemblies; 1) artificial 7 fuel assemblies to maximize intra-assembly heat transfer, 2) central 7 fuel assemblies in the HYPER reference core. The results showed that the modified SLTHEN can reasonably simulate intra-heat transfer and the amount of intra-assembly heat transfer is not so large in HYPER conditions. Particularly, intra-heat transfer did not affect the maximum coolant and the maximum cladding temperatures which are major parameters in conceptual core designs. The capability of full core thermal hydraulic analysis was confirmed by the analysis of 45 fuel assemblies in 1/6 HYPER core at the first cycle. The SLTHEN predicted that the reference design parameters are acceptable in terms of thermal

  13. Development of Regulatory Thermal-Hydraulic Analysis System (RETAS)

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Seung-Hoon; Kim, In-Goo; Kim, Hho-Jung; Cho, Yong Jin [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2007-10-15

    A review is provided of the reasons why the Korea Institute of Nuclear Safety needs improvement of the existing codes employed for a regulatory audit. The proposed new organization of the codes, developed or to be developed, is presented together with illustrative applications. Inspection of the quality assurance activities is planned to ensure the robustness of MARS (Multi-dimensional Analysis for Reactor Safety) code, served as a pivot of the organization.

  14. Development of Regulatory Thermal-Hydraulic Analysis System (RETAS)

    International Nuclear Information System (INIS)

    Ahn, Seung-Hoon; Kim, In-Goo; Kim, Hho-Jung; Cho, Yong Jin

    2007-01-01

    A review is provided of the reasons why the Korea Institute of Nuclear Safety needs improvement of the existing codes employed for a regulatory audit. The proposed new organization of the codes, developed or to be developed, is presented together with illustrative applications. Inspection of the quality assurance activities is planned to ensure the robustness of MARS (Multi-dimensional Analysis for Reactor Safety) code, served as a pivot of the organization

  15. 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.

  16. An improved analysis of gravity drainage experiments for estimating the unsaturated soil hydraulic functions

    Science.gov (United States)

    Sisson, James B.; van Genuchten, Martinus Th.

    1991-04-01

    The unsaturated hydraulic properties are important parameters in any quantitative description of water and solute transport in partially saturated soils. Currently, most in situ methods for estimating the unsaturated hydraulic conductivity (K) are based on analyses that require estimates of the soil water flux and the pressure head gradient. These analyses typically involve differencing of field-measured pressure head (h) and volumetric water content (θ) data, a process that can significantly amplify instrumental and measurement errors. More reliable methods result when differencing of field data can be avoided. One such method is based on estimates of the gravity drainage curve K'(θ) = dK/dθ which may be computed from observations of θ and/or h during the drainage phase of infiltration drainage experiments assuming unit gradient hydraulic conditions. The purpose of this study was to compare estimates of the unsaturated soil hydraulic functions on the basis of different combinations of field data θ, h, K, and K'. Five different data sets were used for the analysis: (1) θ-h, (2) K-θ, (3) K'-θ (4) K-θ-h, and (5) K'-θ-h. The analysis was applied to previously published data for the Norfolk, Troup, and Bethany soils. The K-θ-h and K'-θ-h data sets consistently produced nearly identical estimates of the hydraulic functions. The K-θ and K'-θ data also resulted in similar curves, although results in this case were less consistent than those produced by the K-θ-h and K'-θ-h data sets. We conclude from this study that differencing of field data can be avoided and hence that there is no need to calculate soil water fluxes and pressure head gradients from inherently noisy field-measured θ and h data. The gravity drainage analysis also provides results over a much broader range of hydraulic conductivity values than is possible with the more standard instantaneous profile analysis, especially when augmented with independently measured soil water retention data.

  17. Slip analysis of squeezing flow using doubly stratified fluid

    Science.gov (United States)

    Ahmad, S.; Farooq, M.; Javed, M.; Anjum, Aisha

    2018-06-01

    The non-isothermal flow is modeled and explored for squeezed fluid. The influence of velocity, thermal and solutal slip effects on transport features of squeezed fluid are analyzed through Darcy porous channel when fluid is moving due to squeezing of upper plate towards the stretchable lower plate. Dual stratification effects are illustrated in transport equations. A similarity analysis is performed and reduced governing flow equations are solved using moderated and an efficient convergent approach i.e. Homotopic technique. The significant effects of physical emerging parameters on flow velocity, temperature and fluid concentration are reporting through various plots. Graphical explanations for drag force, Nusselt and Sherwood numbers are stated and examined. The results reveal that minimum velocity field occurs near the plate, whereas it increases far away from the plate for strong velocity slip parameter. Furthermore, temperature and fluid concentration significantly decreases with increased slip effects. The current analysis is applicable in some advanced technological processes and industrial fluid mechanics.

  18. Determination of minimum sample size for fault diagnosis of automobile hydraulic brake system using power analysis

    Directory of Open Access Journals (Sweden)

    V. Indira

    2015-03-01

    Full Text Available Hydraulic brake in automobile engineering is considered to be one of the important components. Condition monitoring and fault diagnosis of such a component is very essential for safety of passengers, vehicles and to minimize the unexpected maintenance time. Vibration based machine learning approach for condition monitoring of hydraulic brake system is gaining momentum. Training and testing the classifier are two important activities in the process of feature classification. This study proposes a systematic statistical method called power analysis to find the minimum number of samples required to train the classifier with statistical stability so as to get good classification accuracy. Descriptive statistical features have been used and the more contributing features have been selected by using C4.5 decision tree algorithm. The results of power analysis have also been verified using a decision tree algorithm namely, C4.5.

  19. Development of steady thermal-hydraulic analysis code for China advanced research reactor

    International Nuclear Information System (INIS)

    Tian Wenxi; Qiu Suizheng; Guo Yun; Su Guanghui; Jia Dounan; Liu Tiancai; Zhang Jianwei

    2006-01-01

    A multi-channel model steady-state thermal-hydraulic analysis code was developed for China Advanced Research Reactor (CARR). By simulating the whole reactor core, the detailed flow distribution in the core was obtained. The result shows that the structure size plays the most important role in flow distribution and the influence of core power could be neglected under single-phase flow. The temperature field of fuel element under unsymmetrical cooling condition was also obtained, which is necessary for the further study such as stress analysis etc. of the fuel element. At the same time, considering the hot channel effect including engineering factor and nuclear factor, calculation of hot channel was carried out and it is proved that all thermal-hydraulic parameters accord with the Safety Regulation of CARR. (authors)

  20. Thermal-hydraulic software development for nuclear waste transportation cask design and analysis

    International Nuclear Information System (INIS)

    Brown, N.N.; Burns, S.P.; Gianoulakis, S.E.; Klein, D.E.

    1991-01-01

    This paper describes the development of a state-of-the-art thermal-hydraulic software package intended for spent fuel and high-level nuclear waste transportation cask design and analysis. The objectives of this software development effort are threefold: (1) to take advantage of advancements in computer hardware and software to provide a more efficient user interface, (2) to provide a tool for reducing inefficient conservatism in spent fuel and high-level waste shipping cask design by including convection as well as conduction and radiation heat transfer modeling capabilities, and (3) to provide a thermal-hydraulic analysis package which is developed under a rigorous quality assurance program established at Sandia National Laboratories. 20 refs., 5 figs., 2 tabs

  1. Biochemical Analysis of Synovial Fluid, Cerebrospinal Fluid and Vitreous Humor at Early Postmortem Intervals in Donkeys

    Directory of Open Access Journals (Sweden)

    Doha Yahia

    2014-01-01

    Full Text Available Biochemical analysis of body fluids after death is a helpful tool in veterinary forensic medicine. Synovial fluid, cerebrospinal fluid (CSF and vitreous humor are easily accessible and well preserved from contamination. Five donkeys (Equus africanus asinus aged 1 - 2 years old were subjected to the study. Samples (Synovial fluid, CSF and vitreous humor were collected before death (antimortem and then at 2, 4, 6, 8, 10 and 12 hours postmortem. Samples were analyzed for glucose, chloride, sodium, magnesium, potassium, enzymes and total protein. Synovial fluid analysis showed that glucose concentration started to decrease at 6 hours postmortem, while magnesium level increased with time. Other parameters were more stable. CSF analysis showed several changes related to time after death as the decrease in glucose and sodium levels, and the increased levels of potassium, magnesium, calcium and total protein. Vitreous analysis revealed a reduction in glucose level and increased potassium and magnesium concentrations. The present study concluded that biochemical analysis of synovial fluid, vitreous humor and CSF can help in determination of time since death in donkeys. This study recommend using CSF for determination of early post-mortem intervals.

  2. Quantifying Water-Rock Interactions during Hydraulic Fracturing from the Analysis of Flowback Water

    Science.gov (United States)

    Osselin, F.; Nightingale, M.; Kloppmann, W.; Gaucher, E.; Clarkson, C.; Mayer, B.

    2017-12-01

    Hydraulic fracturing technologies have facilitated the rapid development of shale gas and other unconventional resources throughout the world. In order to get sufficient access to the trapped hydrocarbon, it is necessary to fracture the bedrock and increase its permeability. Fracturing fluids are usually composed of tens of thousand of cubic meters of low salinity water with numerous additives, such as viscosity agent or breakers. The objective of this study was to investigate and quantify the water-rock interactions during hydraulic fracturing. This study was based on repeated sampling of flowback water from a hydraulically fractured well in Alberta, Canada. The flowback water was sampled 24 times during the first week and one last time after one, and analyzed for major ions and trace elements, as well as stable isotopes of sulfate and water among others. Results showed that salinity rapidly increases up to 100 000 mg/L at the end of the first week. We demonstrate that conservative species such as Na and Cl follow a clear two end-members mixing line, while some species including sulfate had much higher concentrations (8 times higher than the expected value from the mixing line). This indicates that the rapid increase of salinity in flowback water is caused by both mixing with formation water initially present in the shale formation, and from water-rock interactions triggered by the fracturing fluid and in some cases by the additives. Stable isotope data suggest that additional sulfate is mobilized as a consequence of pyrite oxidation, releasing sulfate, iron and potentially other heavy metals into the flowback water. This release of excess sulfate can be detrimental because it has the potential to promote scaling of sulfate minerals. Moreover, pyrite oxidation is a highly acidifying reaction and this may decrease the effectiveness of other additives, and promote carbonate minerals dissolution enhancing further scaling. We propose that a better control of the

  3. Isogeometric Analysis and Shape Optimization in Fluid Mechanics

    DEFF Research Database (Denmark)

    Nielsen, Peter Nørtoft

    This thesis brings together the fields of fluid mechanics, as the study of fluids and flows, isogeometric analysis, as a numerical method to solve engineering problems using computers, and shape optimization, as the art of finding "best" shapes of objects based on some notion of goodness. The flow...... approximations, and for shape optimization purposes also due to its tight connection between the analysis and geometry models. The thesis is initiated by short introductions to fluid mechanics, and to the building blocks of isogeometric analysis. As the first contribution of the thesis, a detailed description...... isogeometric analysis may serve as a natural framework for shape optimization within fluid mechanics. We construct an efficient regularization measure for avoiding inappropriate parametrizations during optimization, and various numerical examples of shape optimization for fluids are considered, serving...

  4. Thermal-hydraulics Analysis of a Radioisotope-powered Mars Hopper Propulsion System

    International Nuclear Information System (INIS)

    O'Brien, Robert C.; Klein, Andrew C.; Taitano, William T.; Gibson, Justice; Myers, Brian; Howe, Steven D.

    2011-01-01

    Thermal-hydraulics analyses results produced using a combined suite of computational design and analysis codes are presented for the preliminary design of a concept Radioisotope Thermal Rocket (RTR) propulsion system. Modeling of the transient heating and steady state temperatures of the system is presented. Simulation results for propellant blow down during impulsive operation are also presented. The results from this study validate the feasibility of a practical thermally capacitive RTR propulsion system.

  5. A Study on the Dynamic Analysis of the Nuclear Fuel Test Rig Using 1-Way Fluid-Structure Coupled Analysis

    International Nuclear Information System (INIS)

    Yang, Tae-Ho; Hong, Jin-Tae; Ahn, Sung-Ho; Joung, Chang-Young; Heo, Sung-Ho; Jang, Seo-Yun

    2015-01-01

    1-way fluid-structure coupled analysis is used to estimate the dynamic characteristic of the fuel test rig. the motion at the bottom of the test rig is confirmed. The maximum deformation of the test rig is 0.11 mm. The structural integrity of the test rig is performed by using the comparison with the Von-mises stress of the analysis and yield stress of the material. It is evaluated that the motion at the bottom of the test rig is able to cause other structural problem. Using the 2-way fluid-structural coupled analysis, the structural integrity of the test rig will be performed in further paper. The cooling water with specific flow rate was flowed in the nuclear fuel test rig. The structural integrity of the test rig was affected by the vibration. The fluid-induced vibration test had to be performed to obtain the amplitude of the vibration on the structure. Various test systems was developed. Flow-induced vibration and pressure drop experimental tester was developed in Korea Atomic Energy Research Institute. The vibration test with high fluid flow rate was difficult by the tester. To generate the nuclear fuel test environment, coolant flow simulation system was developed. The scaled nuclear fuel test was able to be performed by the simulation system. The mock-up model of the test rig was used in the simulation system. The mock-up model in the simulation system was manufactured with scaled down full model. In this paper, the fluid induced vibration characteristic of the full model in the nuclear fuel test is studied. The hydraulic pressure on the velocity of the fluid was calculated. The static structure analysis was performed by using the pressure. The structural integrity was assessed using the results of the analysis

  6. Challenges in thermal and hydraulic analysis of ADS target systems

    International Nuclear Information System (INIS)

    Groetzbach, G.; Batta, A.; Lefhalm, C.-H.; Otic, I.

    2004-01-01

    The liquid metal cooled spallation targets of Accelerator Driven nuclear reactor Systems obey high thermal loads; in addition some flow and cooling conditions are of a prototypical character; in contrast the operating conditions for the engaged materials are narrow; thus, the target development requires a very careful analysis by experimental and numerical means. Especially the cooling of the steel window, which is heated by the proton beam, needs special care. Some of the main goals of the experimental and numerical analyses of the thermal dynamics of those systems are discusses. The prediction of locally detached flows and of flows with larger recirculation areas suffers from insufficient turbulence modeling; this has to be compensated by using prototypical model experiments, e.g. with water, to select the adequate models and numerical schemes. The well known problems with the Reynolds analogy in predicting the heat transfer in liquid metals requires always prototypic liquid metal experiments to select and adapt the turbulent heat flux models. The uncertainties in liquid metal experiments cannot be neglected; so it is necessary to perform CFD calculations and experiments always hand in hand and to develop improve turbulent heat flux models. One contribution to an improved 3 or 4-equation model is deduced from recent Direct Numerical Simulation (DNS) data. (author)

  7. Preliminary LOCA analysis of the westinghouse small modular reactor using the WCOBRA/TRAC-TF2 thermal-hydraulics code

    Energy Technology Data Exchange (ETDEWEB)

    Liao, J.; Kucukboyaci, V. N.; Nguyen, L.; Frepoli, C. [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2012-07-01

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (> 225 MWe) integral pressurized water reactor (iPWR) with all primary components, including the steam generator and the pressurizer located inside the reactor vessel. The reactor core is based on a partial-height 17x17 fuel assembly design used in the AP1000{sup R} reactor core. The Westinghouse SMR utilizes passive safety systems and proven components from the AP1000 plant design with a compact containment that houses the integral reactor vessel and the passive safety systems. A preliminary loss of coolant accident (LOCA) analysis of the Westinghouse SMR has been performed using the WCOBRA/TRAC-TF2 code, simulating a transient caused by a double ended guillotine (DEG) break in the direct vessel injection (DVI) line. WCOBRA/TRAC-TF2 is a new generation Westinghouse LOCA thermal-hydraulics code evolving from the US NRC licensed WCOBRA/TRAC code. It is designed to simulate PWR LOCA events from the smallest break size to the largest break size (DEG cold leg). A significant number of fluid dynamics models and heat transfer models were developed or improved in WCOBRA/TRAC-TF2. A large number of separate effects and integral effects tests were performed for a rigorous code assessment and validation. WCOBRA/TRAC-TF2 was introduced into the Westinghouse SMR design phase to assist a quick and robust passive cooling system design and to identify thermal-hydraulic phenomena for the development of the SMR Phenomena Identification Ranking Table (PIRT). The LOCA analysis of the Westinghouse SMR demonstrates that the DEG DVI break LOCA is mitigated by the injection and venting from the Westinghouse SMR passive safety systems without core heat up, achieving long term core cooling. (authors)

  8. 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

  9. Computer code for the thermal-hydraulic analysis of ITU TRIGA Mark-II reactor

    International Nuclear Information System (INIS)

    Ustun, G.; Durmayaz, A.

    2002-01-01

    Istanbul Technical University (ITU) TRIGA Mark-II reactor core consists of ninety vertical cylindrical elements located in five rings. Sixty-nine of them are fuel elements. The reactor is operated and cooled with natural convection by pool water, which is also cooled and purified in external coolant circuits by forced convection. This characteristic leads to consider both the natural and forced convection heat transfer in a 'porous-medium analysis'. The safety analysis of the reactor requires a thermal-hydraulic model of the reactor to determine the thermal-hydraulic parameters in each mode of operation. In this study, a computer code cooled TRIGA-PM (TRIGA - Porous Medium) for the thermal-hydraulic analysis of ITU is considered. TRIGA Mark-II reactor code has been developed to obtain velocity, pressure and temperature distributions in the reactor pool as a function of core design parameters and pool configuration. The code is a transient, thermal-hydraulic code and requires geometric and physical modelling parameters. In the model, although the reactor is considered as only porous medium, the other part of the reactor pool is considered partly as continuum and partly as porous medium. COMMIX-1C code is used for the benchmark purpose of TRIGA-PM code. For the normal operating conditions of the reactor, estimations of TRIGA-PM are in good agreement with those of COMMIX-1C. After some more improvements, this code will be employed for the estimation of LOCA scenario, which can not be analyses by COMMIX-1C and the other multi-purpose codes, considering a break at one of the beam tubes of the reactor

  10. Thermal-hydraulic analysis techniques for axisymmetric pebble bed nuclear reactor cores

    International Nuclear Information System (INIS)

    Stroh, K.R.

    1979-03-01

    The pebble bed reactor's cylindrical core volume contains a random bed of small, spherical fuel-moderator elements. These graphite spheres, containing a central region of dispersed coated-particle fissile and fertile material, are cooled by high pressure helium flowing through the connected interstitial voids. A mathematical model and numerical solution technique have been developed which allow calculation of macroscopic values of thermal-hydraulic variables in an axisymmetric pebble bed nuclear reactor core. The computer program PEBBLE is based on a mathematical model which treats the bed macroscopically as a generating, conducting porous medium. The steady-state model uses a nonlinear Forchheimer-type relation between the coolant pressure gradient and mass flux, with newly derived coefficients for the linear and quadratic resistance terms. The remaining equations in the model make use of mass continuity, and thermal energy balances for the solid and fluid phases

  11. 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.

  12. Numerical Methods for an Analysis of Hydrogen Behaviors Coupled with Thermal Hydraulics in a NPP Containment

    International Nuclear Information System (INIS)

    Kim, Jongtae; Park, Rae-Joon; Hong, Seong-Wan; Kim, Gun-Hong

    2016-01-01

    In a containment safety analysis, multi-dimensional characteristics in thermal hydraulics are very important because the flow paths are not confined in a large free volume of the containment. The analysis is difficult because of a difference in length scales between a characteristic length of the flow and representative length of the containment. In order to simulate hydrogen and steam behaviors in a containment during postulated severe accidents, the GASFLOW code as a multi-dimensional analysis tool for NPP containment has been used for years because of its computational efficiency. Though GASFLOW is well developed for a real NPP containment analysis, there exist shortcomings in nodalization, two-phase and turbulence models. It is based on a Cartesian or cylindrical coordinate mesh, so it is impractical to refine a mesh locally in a region with a physical or geometrical complication. In this paper, the importance of the hydrogen safety in an NPP containment and requirements of the analysis tool was described. And physical models necessary for the hydrogen safety analysis code were listed. As a member of international collaborative project HYMERES for containment thermal hydraulics, KAERI is actively participating in an analytic working group. As an analysis tool for blind benchmarkes, the analysis code described in this paper was used. From the blind benchmark analyses, it was found that the code is very promising for hydrogen safety analysis. Currently, it is proposed to develop the code collaboratively in a hydrogen safety community based on an open-source strategy

  13. Numerical Methods for an Analysis of Hydrogen Behaviors Coupled with Thermal Hydraulics in a NPP Containment

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jongtae; Park, Rae-Joon; Hong, Seong-Wan; Kim, Gun-Hong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    In a containment safety analysis, multi-dimensional characteristics in thermal hydraulics are very important because the flow paths are not confined in a large free volume of the containment. The analysis is difficult because of a difference in length scales between a characteristic length of the flow and representative length of the containment. In order to simulate hydrogen and steam behaviors in a containment during postulated severe accidents, the GASFLOW code as a multi-dimensional analysis tool for NPP containment has been used for years because of its computational efficiency. Though GASFLOW is well developed for a real NPP containment analysis, there exist shortcomings in nodalization, two-phase and turbulence models. It is based on a Cartesian or cylindrical coordinate mesh, so it is impractical to refine a mesh locally in a region with a physical or geometrical complication. In this paper, the importance of the hydrogen safety in an NPP containment and requirements of the analysis tool was described. And physical models necessary for the hydrogen safety analysis code were listed. As a member of international collaborative project HYMERES for containment thermal hydraulics, KAERI is actively participating in an analytic working group. As an analysis tool for blind benchmarkes, the analysis code described in this paper was used. From the blind benchmark analyses, it was found that the code is very promising for hydrogen safety analysis. Currently, it is proposed to develop the code collaboratively in a hydrogen safety community based on an open-source strategy.

  14. 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.

  15. State of Art of the CAREM-25 Hydraulic Control Rod Drives Feasibility Analysis

    International Nuclear Information System (INIS)

    Mazufri, C.M; Mazzi, R.O

    2000-01-01

    The proposed design adopted for the control rod drives for the CAREM reactor is based on a hydraulic system.As any innovative device, the design process requires to obtain experimental evidence to identify the most important control parameters and to set their relationship with other design parameters, in order to guarantee its feasibility as a previous step to the design qualification tests at the working conditions at the reactor.This paper features a global evaluation of the analysis performed and experimental results obtained in a low pressure loop, design improvements, limiting phenomena identified and corrective actions analyzed or proposed.The evaluation is based on a repetitivity, sensitivity and scalability study of the control parameters and test conditions, as well as the dynamic response between rod drive and the hydraulic system and features related with the mechanical design.Obtained results show that present system has an adequate response compatible with functional and manufacturing requirements

  16. ATWS thermal-hydraulic analysis for Krsko Full Scope Simulator validation

    International Nuclear Information System (INIS)

    Parzer, I.; Kljenak, I.

    2005-01-01

    The purpose of this analysis was to simulate Anticipated Transient without Scram transient for Krsko NPP. The results of these calculations were used for annual ANSI/ANS validation of reactor coolant system thermal-hydraulic response predicted by Krsko Full Scope Simulator. For the thermal-hydraulic analyses the RELAP5/MOD3.3 code and the input model for NPP Krsko, delivered by NPP Krsko, was used. In the presented paper the most severe ATWS scenario has been analyzed, starting with the loss of Main Feedwater at both steam generators. Thus, gradual loss of secondary heat sink occurred. On top of that, control rods were not supposed to scram, leaving the chain reaction to be controlled only by inherent physical properties of the fuel and moderator and eventual actions of the BOP system. The primary system response has been studied assuming AMSAC availability. (author)

  17. Analysis of birth-death fluid queues

    NARCIS (Netherlands)

    van Doorn, Erik A.; Scheinhardt, Willem R.W.

    1996-01-01

    We present a survey of techniques for analysing the performance of a reservoir which receives and releases fluid at rates which are determined by the state of a background birth-death process. The reservoir is assumed to be infinitely large, but the state space of the modulating birth-death process

  18. Analysis of birth-death fluid queues

    OpenAIRE

    van Doorn, Erik A.; Scheinhardt, Willem R.W.

    1996-01-01

    We present a survey of techniques for analysing the performance of a reservoir which receives and releases fluid at rates which are determined by the state of a background birth-death process. The reservoir is assumed to be infinitely large, but the state space of the modulating birth-death process may be finite or infinite.

  19. Yield stress fluids slowly yield to analysis

    NARCIS (Netherlands)

    Bonn, D.; Denn, M.M.

    2009-01-01

    We are surrounded in everyday life by yield stress fluids: materials that behave as solids under small stresses but flow like liquids beyond a critical stress. For example, paint must flow under the brush, but remain fixed in a vertical film despite the force of gravity. Food products (such as

  20. Steady-state thermal-hydraulic design analysis of the Advanced Neutron Source reactor

    International Nuclear Information System (INIS)

    Yoder, G.L. Jr.; Dixon, J.R.; Elkassabgi, Y.; Felde, D.K.; Giles, G.E.; Harrington, R.M.; Morris, D.G.; Nelson, W.R.; Ruggles, A.E.; Siman-Tov, M.; Stovall, T.K.

    1994-05-01

    The Advanced Neutron Source (ANS) is a research reactor that is planned for construction at Oak Ridge National Laboratory. This reactor will be a user facility with the major objective of providing the highest continuous neutron beam intensities of any reactor in the world. Additional objectives for the facility include providing materials irradiation facilities and isotope production facilities as good as, or better than, those in the High Flux Isotope Reactor. To achieve these objectives, the reactor design uses highly subcooled heavy water as both coolant and moderator. Two separate core halves of 67.6-L total volume operate at an average power density of 4.5 MW(t)/L, and the coolant flows upward through the core at 25 m/s. Operating pressure is 3.1 MPa at the core inlet with a 1.4-MPa pressure drop through the core region. Finally, in order to make the resources available for experimentation, the fuel is designed to provide a 17-d fuel cycle with an additional 4 d planned in each cycle for the refueling process. This report examines the codes and models used to develop the thermal-hydraulic design for ANS, as well as the correlations and physical data; evaluates thermal-hydraulic uncertainties; reports on thermal-hydraulic design and safety analysis; describes experimentation in support of the ANS reactor design and safety analysis; and provides an overview of the experimental plan

  1. Methodology for thermal-hydraulics analysis of pool type MTR fuel research reactors

    International Nuclear Information System (INIS)

    Umbehaun, Pedro Ernesto

    2000-01-01

    This work presents a methodology developed for thermal-hydraulic analysis of pool type MTR fuel research reactors. For this methodology a computational program, FLOW, and a model, MTRCR-IEAR1 were developed. FLOW calculates the cooling flow distribution in the fuel elements, control elements, irradiators, and through the channels formed among the fuel elements and among the irradiators and reflectors. This computer program was validated against experimental data for the IEA-R1 research reactor core at IPEN-CNEN/SP. MTRCR-IEAR1 is a model based on the commercial program Engineering Equation Solver (EES). Besides the thermal-hydraulic analyses of the core in steady state accomplished by traditional computational programs like COBRA-3C/RERTR and PARET, this model allows to analyze parallel channels with different cooling flow and/or geometry. Uncertainty factors of the variables from neutronic and thermalhydraulic calculation and also from the fabrication of the fuel element are introduced in the model. For steady state analyses MTRCR-IEAR1 showed good agreement with the results of COBRA-3C/RERTR and PARET. The developed methodology was used for the calculation of the cooling flow distribution and the thermal-hydraulic analysis of a typical configuration of the IEA-R1 research reactor core. (author)

  2. Thermo-hydraulic analysis of the cool-down of the EDIPO test facility

    Science.gov (United States)

    Lewandowska, Monika; Bagnasco, Maurizio

    2011-09-01

    The first cool-down of the EDIPO (European DIPOle) test facility is foreseen to take place in 2011 by means of the existing 1.2 kW cryoplant at EPFL-CRPP Villigen. In this work, the thermo-hydraulic analysis of the EDIPO cool-down is performed in order both to assess the its duration and to optimize the procedure. The cool-down is driven by the helium flowing in both the outer cooling channel and in the windings connected hydraulically in parallel. We take into account limitations due to the pressure drop in the cooling circuit and the refrigerator capacity as well as heat conduction in the iron yoke. Two schemes of the hydraulic cooling circuit in the EDIPO windings are studied (coils connected in series and coils connected in parallel). The analysis is performed by means of an analytical model complemented by and numerical model. The results indicate that the cool-down to 5 K can be achieved in about 12 days.

  3. Thermohydrodynamic analysis of cryogenic liquid turbulent flow fluid film bearings

    Science.gov (United States)

    Andres, Luis San

    1993-01-01

    A thermohydrodynamic analysis is presented and a computer code developed for prediction of the static and dynamic force response of hydrostatic journal bearings (HJB's), annular seals or damper bearing seals, and fixed arc pad bearings for cryogenic liquid applications. The study includes the most important flow characteristics found in cryogenic fluid film bearings such as flow turbulence, fluid inertia, liquid compressibility and thermal effects. The analysis and computational model devised allow the determination of the flow field in cryogenic fluid film bearings along with the dynamic force coefficients for rotor-bearing stability analysis.

  4. 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.

  5. Hydraulic head interpolation using ANFIS—model selection and sensitivity analysis

    Science.gov (United States)

    Kurtulus, Bedri; Flipo, Nicolas

    2012-01-01

    The aim of this study is to investigate the efficiency of ANFIS (adaptive neuro fuzzy inference system) for interpolating hydraulic head in a 40-km 2 agricultural watershed of the Seine basin (France). Inputs of ANFIS are Cartesian coordinates and the elevation of the ground. Hydraulic head was measured at 73 locations during a snapshot campaign on September 2009, which characterizes low-water-flow regime in the aquifer unit. The dataset was then split into three subsets using a square-based selection method: a calibration one (55%), a training one (27%), and a test one (18%). First, a method is proposed to select the best ANFIS model, which corresponds to a sensitivity analysis of ANFIS to the type and number of membership functions (MF). Triangular, Gaussian, general bell, and spline-based MF are used with 2, 3, 4, and 5 MF per input node. Performance criteria on the test subset are used to select the 5 best ANFIS models among 16. Then each is used to interpolate the hydraulic head distribution on a (50×50)-m grid, which is compared to the soil elevation. The cells where the hydraulic head is higher than the soil elevation are counted as "error cells." The ANFIS model that exhibits the less "error cells" is selected as the best ANFIS model. The best model selection reveals that ANFIS models are very sensitive to the type and number of MF. Finally, a sensibility analysis of the best ANFIS model with four triangular MF is performed on the interpolation grid, which shows that ANFIS remains stable to error propagation with a higher sensitivity to soil elevation.

  6. Uncertainty analysis for results of thermal hydraulic codes of best-estimate-type

    International Nuclear Information System (INIS)

    Alva N, J.

    2010-01-01

    In this thesis, some fundamental knowledge is presented about uncertainty analysis and about diverse methodologies applied in the study of nuclear power plant transient event analysis, particularly related to thermal hydraulics phenomena. These concepts and methodologies mentioned in this work come from a wide bibliographical research in the nuclear power subject. Methodologies for uncertainty analysis have been developed by quite diverse institutions, and they have been widely used worldwide for application to results from best-estimate-type computer codes in nuclear reactor thermal hydraulics and safety analysis. Also, the main uncertainty sources, types of uncertainties, and aspects related to best estimate modeling and methods are introduced. Once the main bases of uncertainty analysis have been set, and some of the known methodologies have been introduced, it is presented in detail the CSAU methodology, which will be applied in the analyses. The main objective of this thesis is to compare the results of an uncertainty and sensibility analysis by using the Response Surface Technique to the application of W ilks formula, apply through a loss coolant experiment and an event of rise in a BWR. Both techniques are options in the part of uncertainty and sensibility analysis of the CSAU methodology, which was developed for the analysis of transients and accidents at nuclear power plants, and it is the base of most of the methodologies used in licensing of nuclear power plants practically everywhere. Finally, the results of applying both techniques are compared and discussed. (Author)

  7. Generalized Fluid System Simulation Program (GFSSP) Version 6 - General Purpose Thermo-Fluid Network Analysis Software

    Science.gov (United States)

    Majumdar, Alok; Leclair, Andre; Moore, Ric; Schallhorn, Paul

    2011-01-01

    GFSSP stands for Generalized Fluid System Simulation Program. It is a general-purpose computer program to compute pressure, temperature and flow distribution in a flow network. GFSSP calculates pressure, temperature, and concentrations at nodes and calculates flow rates through branches. It was primarily developed to analyze Internal Flow Analysis of a Turbopump Transient Flow Analysis of a Propulsion System. GFSSP development started in 1994 with an objective to provide a generalized and easy to use flow analysis tool for thermo-fluid systems.

  8. 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...

  9. Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ohshima, Hiroyuki; Uwaba, Tomoyuki [Japan Atomic Energy Agency (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan); Hashimoto, Akihiko; Imai, Yasutomo [NDD Corporation (1-1-6 Jounan, Mito, Ibaraki 310-0803, Japan) (Japan); Ito, Masahiro [NESI Inc. (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan)

    2015-12-31

    A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.

  10. 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

  11. TITAN: an advanced three-dimensional coupled neutronic/thermal-hydraulics code for light water nuclear reactor core analysis

    International Nuclear Information System (INIS)

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

    1984-06-01

    The three-dimensional nodal neutronics code QUANDRY and the three-dimensional two-fluid thermal-hydraulics code THERMIT are combined into TITAN. Steady-state and transient coupling methodologies based upon a tandem structure were devised and implemented. Additional models for nuclear feedback, equilibrium xenon and direct moderator heating were added. TITAN was tested using a boiling water two channel problem and the coupling methodologies were shown to be effective. Simulated turbine trip transients and several control rod withdrawal transients were analyzed with good results. Sensitivity studies indicated that the time-step size can affect transient results significantly. TITAN was also applied to a quarter core PWR problem based on a real reactor geometry. The steady-state results were compared to a solution produced by MEKIN-B and poor agreement between the horizontal power shapes was found. Calculations with various mesh spacings showed that the mesh spacings in the MEKIN-B analysis were too large to produce accurate results with a finite difference method. The TITAN results were shown to be reasonable. A pair of control rod ejection accidents were also analyzed with TITAN. A comparison of the TITAN PWR control rod ejection results with results from coupled point kinetics/thermal-hydraulics analyses showed that the point kinetics method used (adiabatic method for control rod reactivities, steady-state flux shape for core-averaged reactivity feedback) underpredicted the power excursion in one case and overpredicted it in the other. It was therefore concluded that point kinetics methods should be used with caution and that three-dimensional codes like TITAN are superior for analyzing PWR control rod ejection transients

  12. Introduction of thermal-hydraulic analysis code and system analysis code for HTGR

    International Nuclear Information System (INIS)

    Tanaka, Mitsuhiro; Izaki, Makoto; Koike, Hiroyuki; Tokumitsu, Masashi

    1984-01-01

    Kawasaki Heavy Industries Ltd. has advanced the development and systematization of analysis codes, aiming at lining up the analysis codes for heat transferring flow and control characteristics, taking up HTGR plants as the main object. In order to make the model of flow when shock waves propagate to heating tubes, SALE-3D which can analyze a complex system was developed, therefore, it is reported in this paper. Concerning the analysis code for control characteristics, the method of sensitivity analysis in a topological space including an example of application is reported. The flow analysis code SALE-3D is that for analyzing the flow of compressible viscous fluid in a three-dimensional system over the velocity range from incompressibility limit to supersonic velocity. The fundamental equations and fundamental algorithm of the SALE-3D, the calculation of cell volume, the plotting of perspective drawings and the analysis of the three-dimensional behavior of shock waves propagating in heating tubes after their rupture accident are described. The method of sensitivity analysis was added to the analysis code for control characteristics in a topological space, and blow-down phenomena was analyzed by its application. (Kako, I.)

  13. Application of ADINA fluid element for transient response analysis of fluid-structure system

    International Nuclear Information System (INIS)

    Sakurai, Y.; Kodama, T.; Shiraishi, T.

    1985-01-01

    Pressure propagation and Fluid-Structure Interaction (FSI) in 3D space were simulated by general purpose finite element program ADINA using the displacement-based fluid element which presumes inviscid and compressible fluid with no net flow. Numerical transient solution was compared with the measured data of an FSI experiment and was found to fairly agree with the measured. In the next step, post analysis was conducted for a blowdown experiment performed with a 1/7 scaled reactor pressure vessel and a flexible core barrel and the code performance was found to be satisfactory. It is concluded that the transient response of the core internal structure of a PWR during the initial stage of LOCA can be analyzed by the displacement-based finite fluid element and the structural element. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-01-15

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

  15. Estimating hydraulic parameters of the Açu-Brazil aquifer using the computer analysis of micrographs

    Science.gov (United States)

    de Lucena, Leandson R. F.; da Silva, Luis R. D.; Vieira, Marcela M.; Carvalho, Bruno M.; Xavier Júnior, Milton M.

    2016-04-01

    The conventional way of obtaining hydraulic parameters of aquifers is through the interpretation of aquifer tests that requires a fairly complex logistics in terms of equipment and personnel. On the other way, the processing and analysis of digital images of two-dimensional rock sample micrographs presents itself as a promising (simpler and cheaper) alternative procedure for obtaining estimates for hydraulics parameters. This methodology involves the sampling of rocks, followed by the making and imaging of thin rock samples, image segmentation, three-dimensional reconstruction and flow simulation. This methodology was applied to the outcropping portion of the Açu aquifer in the northeast of Brazil, and the computational analyses of the thin rock sections of the acquired samples produced effective porosities between 11.2% and 18.5%, and permeabilities between 52.4 mD and 1140.7 mD. Considering that the aquifer is unconfined, these effective porosity values can be used effectively as storage coefficients. The hydraulic conductivities produced by adopting different water dynamic viscosities at the temperature of 28 °C in the conversion of the permeabilities result in values in the range of [ 6.03 ×10-7, 1.43 ×10-5 ] m/s, compatible with the local hydrogeology.

  16. The analysis of thermal-hydraulic performances of nuclear ship reactor

    International Nuclear Information System (INIS)

    Wakabayashi, Shinshichi; Hamada, Masao

    1975-01-01

    Thermal-hydraulic performances in the core of nuclear ship reactor was analysed by thermal-hydraulic analyser codes, AMRTC and COBRA-11+DNBCAL. This reactor is of a pressurized water type and incorporates the steam generator within the reactor vessel with the rated power of 330 MWt, which is developed by Nuclear Ship Research Panel Seven (NSR-7) in The Shipbuilding Research Association of Japan. Fuel temperature distributions, coolant temperature distributions, void fractions in coolant and minimum burn out ratio etc. were calculated. Results are as follows; a) The maximum temperature of fuel center is 1,472 0 C that corresponds to 53% as small as the melting point (2,800 0 C). b) Subcooled boiling exists in the core and the maximum void fraction is less than 4%. c) The minimum burn out ratio is not less than the minimum allowable limit of 1.25. It was found from the results of analysis that this reactor was able to be operated wide margin with respect to thermal-hydraulic design limits at the rated power. (auth.)

  17. Quench characterization and thermo hydraulic analysis of SST-1 TF magnet busbar

    Energy Technology Data Exchange (ETDEWEB)

    Sharma, A.N., E-mail: ansharma@ipr.res.in [Institute for Plasma Research, Gandhinagar (India); Pradhan, S. [Institute for Plasma Research, Gandhinagar (India); Duchateau, J.L. [CEA Cadarache, 13108 St Paul lez Durance Cedex (France); Khristi, Y.; Prasad, U.; Doshi, K.; Varmora, P.; Tanna, V.L.; Patel, D.; Panchal, A. [Institute for Plasma Research, Gandhinagar (India)

    2015-01-15

    Highlights: • Details of SST-1 TF busbar quench detection. • Simulation of slow propagating normal zone. • Thermo hydraulic analyses of TF busbar in current feeder system. - Abstract: Toroidal field (TF) magnet system of steady-state superconducting tokamak-1 (SST-1) has 16 superconducting coils. TF coils are cooled with forced flow supercritical helium at 0.4 MPa, at 4.5 K and operate at nominal current of 10,000 A. Prior to TF magnet system assembly in SST-1 tokamak, each TF coil was tested individually in a test cryostat. During these tests, TF coil was connected to a pair of conventional helium vapor cooled current leads. The connecting busbar was made from the same base cable-in-conduit-conductor (CICC) of SST-1 superconducting magnet system. Quenches experimentally observed in the busbar sections of the single coil test setups have been analyzed in this paper. A steady state thermo hydraulic analysis of TF magnet busbar in actual SST-1 tokamak assembly has been done. The experimental observations of quench and results of relevant thermo hydraulic analyses have been used to predict the safe operation regime of TF magnet system busbar during actual SST-1 tokamak operational scenarios.

  18. Thermal hydraulic analysis of the IPR-R1 TRIGA reactor

    International Nuclear Information System (INIS)

    Veloso, Marcelo Antonio; Fortini, Maria Auxiliadora

    2002-01-01

    The subchannel approach, normally employed for the analysis of power reactor cores that work under forced convection, have been used for the thermal hydraulic evaluation of a TRIGA Mark I reactor, named IPR-R1, at 250 kW power level. This was accomplished by using the PANTERA-1P subchannel code, which has been conveniently adapted to the characteristics of natural convection of TRIGA reactors. The analysis of results indicates that the steady state operation of IPR-R1 at 250 kW do not imply risks to installations, workers and public. (author)

  19. Computational fluid dynamics analysis of a mixed flow pump impeller

    African Journals Online (AJOL)

    ATHARVA

    International Journal of Engineering, Science and Technology ... From the CFD analysis software and advanced post processing tools the complex flow inside the ... The numerical simulation can provide quite accurate information on the fluid ...

  20. An Integrated Solution for Performing Thermo-fluid Conjugate Analysis

    Science.gov (United States)

    Kornberg, Oren

    2009-01-01

    A method has been developed which integrates a fluid flow analyzer and a thermal analyzer to produce both steady state and transient results of 1-D, 2-D, and 3-D analysis models. The Generalized Fluid System Simulation Program (GFSSP) is a one dimensional, general purpose fluid analysis code which computes pressures and flow distributions in complex fluid networks. The MSC Systems Improved Numerical Differencing Analyzer (MSC.SINDA) is a one dimensional general purpose thermal analyzer that solves network representations of thermal systems. Both GFSSP and MSC.SINDA have graphical user interfaces which are used to build the respective model and prepare it for analysis. The SINDA/GFSSP Conjugate Integrator (SGCI) is a formbase graphical integration program used to set input parameters for the conjugate analyses and run the models. The contents of this paper describes SGCI and its thermo-fluids conjugate analysis techniques and capabilities by presenting results from some example models including the cryogenic chill down of a copper pipe, a bar between two walls in a fluid stream, and a solid plate creating a phase change in a flowing fluid.

  1. 3D casing-distributor analysis with a novel block coupled OpenFOAM solver for hydraulic design application

    International Nuclear Information System (INIS)

    Devals, C; Zhang, Y; Dompierre, J; Guibault, F; Vu, T C; Mangani, L

    2014-01-01

    Nowadays, computational fluid dynamics is commonly used by design engineers to evaluate and compare losses in hydraulic components as it is less expensive and less time consuming than model tests. For that purpose, an automatic tool for casing and distributor analysis will be presented in this paper. An in-house mesh generator and a Reynolds Averaged Navier-Stokes equation solver using the standard k-ω SST turbulence model will be used to perform all computations. Two solvers based on the C++ OpenFOAM library will be used and compared to a commercial solver. The performance of the new fully coupled block solver developed by the University of Lucerne and Andritz will be compared to the standard 1.6ext segregated simpleFoam solver and to a commercial solver. In this study, relative comparisons of different geometries of casing and distributor will be performed. The present study is thus aimed at validating the block solver and the tool chain and providing design engineers with a faster and more reliable analysis tool that can be integrated into their design process

  2. 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

  3. Coupled neutronics/thermal-hydraulics for analysis of molten salt reactor

    International Nuclear Information System (INIS)

    Guo, Zhangpeng; Zhou, Jianjun; Zhang, Dalin; Chaudri, Khurrum Saleem; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng

    2013-01-01

    Highlights: ► A multiple-channel analysis code (MAC) is developed to be coupled with MCNP. ► 1/8 of core is simulated in MCNP and thermal-hydraulic code. ► The coupling calculation can achieve stable state after a few iterations. ► The coupling calculation results are in reasonable agreement with the analytic solutions of the ORNL. ► Parametric studies of MSR are performed to provide valuable information for future design MSR. -- Abstract: The Generation IV International Forum (GIF) selected molten salt reactor (MSR) among six advanced reactor types. It is characterized by a liquid circulating fuel that also serves as coolant. In this study, a multiple-channel analysis code (MAC) is developed and it is coupled with MCNP4c to analyze the neutronics/thermal-hydraulics behavior of molten salt reactor experiment (MSRE). The MAC calculates thermal-hydraulic parameters, such as temperature distribution, flow distribution and pressure drop. MCNP4c performs the analysis of effective multiplication factor, neutron flux and power distribution. A linkage code is developed to exchange data between MAC and MCNP to implement coupling iteration process until the power convergence is achieved. The coupling calculation can achieve converged solution after a few iterations. The results are in reasonable agreement with the analytic solutions from the ORNL. For further design analysis, parametric studies are performed to provide valuable information for new design of MSR. The effect of inlet temperature, graphite to molten salt volume ratio (G/Ms) from varying channel diameter and different power levels on the effective multiplication factor, neutron flux, graphite lifetime and temperature distribution are discussed in detail

  4. Interaction between thermal/hydraulics, human factors and system analysis for assessing feed and bleed risk benefits

    International Nuclear Information System (INIS)

    Lanore, J.M.; Caron, J.L.

    1987-11-01

    For probabilistic analysis of accident sequences, thermal/hydraulics, human factors and systems operation problems are frequently closely interrelated. This presentation will discuss a typical example which illustrates this interrelation: total loss of feedwater flow. It will present thermal/hydraulic analysises performed, how the T/H analysises are related to human factors and systems operation, and how, based on this, the failure probability of the feed and bleed cooling mode was evaluated

  5. Nuclear fuel element design and thermal-hydraulic analysis of Wolsung-1, 600 MWe CANDU-PHWR (Part II)

    International Nuclear Information System (INIS)

    Suk, H.C; Lee, J.C.; Suh, K.S.; Yuk, K.E.; Whang, W.; Park, J.S.; Eim, J.S.; Bang, K.H.; Eim, M.S.; Rim, C.S.

    1982-01-01

    The main objective of the present thermal hydraulic analysis is to determine the thermal hydraulic characteristics of Wolsung-1 600 MWe CANDU-PHW reactor under normal operation. This is to verify and expedite the development of the nuclear fuel design and fabrication as well as the management. The computer program package developed for the stated objective are DOD81, CANREPP, PLOC81 and COBRA-CANDU. (Author)

  6. 3D thermal-hydraulic analysis on core of PWR nuclear power station

    International Nuclear Information System (INIS)

    Yao Zhaohui; Wang Xuefang; Shen Mengyu

    1997-01-01

    Thermal hydraulic analysis of core is of great importance in reactor safety analysis. A computer code, thermal hydraulic analysis porous medium analysis (THAPMA), has been developed to simulate the flow and heat transfer characteristics of reactor components. It has been proved reliable by several numerical tests. In the THAPMA code, a new difference scheme and solution method have been studied in developing the computer software. For the difference scheme, a second order accurate, high resolution scheme, called WSUC scheme, has been proposed. This scheme is total variation bounded and unconditionally stable in convective numeral stability. Numerical tests show that the WSUC is better in accuracy and resolution than the 1-st order upwind, 2-nd order upwind, SOUCUP by Zhu and Rodi. In solution method, a modified PISO algorithm is used, which is not only simpler but also more accurate and more rapid in convergence than the original PISO algorithm. Moreover, the modified PISO algorithm can effectively solve steady and transient state problem. Besides, with the THAPMA code, the flow and heat transfer phenomena in reactor core have been numerically simulated in the light of the design condition of Qinshan PWR nuclear power station (the second-term project). The simulation results supply a theoretical basis for the core design

  7. Thermal-hydraulic analysis of the Three Mile Island Unit 2 reactor accident with THALES code

    International Nuclear Information System (INIS)

    Hashimoto, Kazuichiro; Soda, Kunihisa

    1991-10-01

    The OECD Nuclear Energy Agency (NEA) has established a Task Group in the Committee on the Safety of Nuclear Installations (CSNI) to perform an analysis of Three Mile Island Unit 2 (TMI-2) accident as a standard problem to benchmark severe accident computer codes and to assess the capability of the codes. The TMI-2 Analysis Exercise was performed at the Japan Atomic Energy Research Institute (JAERI) using the THALES (Thermal-Hydraulic Analysis of Loss-of-Coolant, Emergency Core Cooling and Severe Core Damage) - PM1/TMI code. The purpose of the analysis is to verify the capability of THALES-PM1/TMI code to describe accident progression in the actual plant. The present paper describes the final result of the TMI-2 Analysis Exercise performed at JAERI. (author)

  8. 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)

  9. Software Tool for Automated Failure Modes and Effects Analysis (FMEA) of Hydraulic Systems

    DEFF Research Database (Denmark)

    Stecki, J. S.; Conrad, Finn; Oh, B.

    2002-01-01

    Offshore, marine,aircraft and other complex engineering systems operate in harsh environmental and operational conditions and must meet stringent requirements of reliability, safety and maintability. To reduce the hight costs of development of new systems in these fields improved the design...... management techniques and a vast array of computer aided techniques are applied during design and testing stages. The paper present and discusses the research and development of a software tool for automated failure mode and effects analysis - FMEA - of hydraulic systems. The paper explains the underlying...

  10. A General Model for Thermal, Hydraulic and Electric Analysis of Superconducting Cables

    CERN Document Server

    Bottura, L; Rosso, C

    2000-01-01

    In this paper we describe a generic, multi-component and multi-channel model for the analysis of superconducting cables. The aim of the model is to treat in a general and consistent manner simultaneous thermal, electric and hydraulic transients in cables. The model is devised for most general situations, but reduces in limiting cases to most common approximations without loss of efficiency. We discuss here the governing equations, and we write them in a matrix form that is well adapted to numerical treatment. We finally demonstrate the model capability by comparison with published experimental data on current distribution in a two-strand cable.

  11. VIPRE-01: a thermal-hydraulic analysis code for reactor cores. Volume 2. User's manual

    International Nuclear Information System (INIS)

    Cuta, J.M.; Koontz, A.S.; Stewart, C.W.; Montgomery, S.D.

    1983-04-01

    VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear energy reactor core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This volume (Volume 2: User's Manual) describes the input requirements of VIPRE and its auxiliary programs, SPECSET, ASP and DECCON, and lists the input instructions for each code

  12. ANTEO: An optimised PC computer code for the steady state thermal hydraulic analysis of rod bundles

    International Nuclear Information System (INIS)

    Cevolani, S.

    1996-07-01

    The paper deals with the description of a Personal Computer oriented subchannel code, devoted to the steady state thermal hydraulic analysis of nuclear reactor fuel bundles. The development of a such code was made possible by two facts: first, the increase the computing power of the desk machines; secondly, the fact several years of experience into operate subchannels codes have shown how to simplify many of the physical models without a sensible loss of accuracy. For sake of validation, the developed code was compared with a traditional subchannel code, the COBRA one. The results of the comparison show a very good agreement between the two codes

  13. Comparative analysis between Hec-RAS models and IBER in the hydraulic assessment of bridges

    OpenAIRE

    Rincón, Jean; Pérez, María; Delfín, Guillermo; Freitez, Carlos; Martínez, Fabiana

    2017-01-01

    This work aims to perform a comparative analysis between the Hec-RAS and IBER models, in the hydraulic evaluation of rivers with structures such as bridges. The case of application was the La Guardia creek, located in the road that communicates the cities of Barquisimeto-Quíbor, Venezuela. The first phase of the study consisted in the comparison of the models from the conceptual point of view and the management of both. The second phase focused on the case study, and the comparison of ...

  14. PIXE analysis of cerebrospinal fluid before and after brain transplantation

    International Nuclear Information System (INIS)

    Ma Xinpei; Wang Junke.

    1992-01-01

    Considering methodology of PIXE quantitative analysis based on Inner-standard, we provide a simple and convenient method to measure the elemental relative sensitivity curve. The concentrations of 16 various elements in cerebrospinal fluid samples before and after brain transplantation have been investigated and compared with those of normal person's and transplanted tissues. The experimental results show that the brain transplantation results in apparently curative effects in compensating and regulating the element concentrations in cerebrospinal fluid and improvement of elemental physiological metabolism. It illustrates that the appropriate concentrations of trace elements in cerebrospinal fluid play an undoubtedly important role in keeping the normal physiological function of brain and central nervous system. (author)

  15. Wavelet analysis of polarization maps of polycrystalline biological fluids networks

    Science.gov (United States)

    Ushenko, Y. A.

    2011-12-01

    The optical model of human joints synovial fluid is proposed. The statistic (statistic moments), correlation (autocorrelation function) and self-similar (Log-Log dependencies of power spectrum) structure of polarization two-dimensional distributions (polarization maps) of synovial fluid has been analyzed. It has been shown that differentiation of polarization maps of joint synovial fluid with different physiological state samples is expected of scale-discriminative analysis. To mark out of small-scale domain structure of synovial fluid polarization maps, the wavelet analysis has been used. The set of parameters, which characterize statistic, correlation and self-similar structure of wavelet coefficients' distributions of different scales of polarization domains for diagnostics and differentiation of polycrystalline network transformation connected with the pathological processes, has been determined.

  16. AIR INGRESS ANALYSIS: COMPUTATIONAL FLUID DYNAMIC MODELS

    Energy Technology Data Exchange (ETDEWEB)

    Chang H. Oh; Eung S. Kim; Richard Schultz; Hans Gougar; David Petti; Hyung S. Kang

    2010-08-01

    The Idaho National Laboratory (INL), under the auspices of the U.S. Department of Energy, is performing research and development that focuses on key phenomena important during potential scenarios that may occur in very high temperature reactors (VHTRs). Phenomena Identification and Ranking Studies to date have ranked an air ingress event, following on the heels of a VHTR depressurization, as important with regard to core safety. Consequently, the development of advanced air ingress-related models and verification and validation data are a very high priority. Following a loss of coolant and system depressurization incident, air will enter the core of the High Temperature Gas Cooled Reactor through the break, possibly causing oxidation of the in-the core and reflector graphite structure. Simple core and plant models indicate that, under certain circumstances, the oxidation may proceed at an elevated rate with additional heat generated from the oxidation reaction itself. Under postulated conditions of fluid flow and temperature, excessive degradation of the lower plenum graphite can lead to a loss of structural support. Excessive oxidation of core graphite can also lead to the release of fission products into the confinement, which could be detrimental to a reactor safety. Computational fluid dynamic model developed in this study will improve our understanding of this phenomenon. This paper presents two-dimensional and three-dimensional CFD results for the quantitative assessment of the air ingress phenomena. A portion of results of the density-driven stratified flow in the inlet pipe will be compared with results of the experimental results.

  17. Modeling and analysis of a meso-hydraulic climbing robot with artificial muscle actuation.

    Science.gov (United States)

    Chapman, Edward M; Jenkins, Tyler E; Bryant, Matthew

    2017-07-10

    This paper presents a fully coupled electro-hydraulic model of a bio-inspired climbing robot actuated by fluidic artificial muscles (FAMs). This analysis expands upon previous FAM literature by considering not only the force and contraction characteristics of the actuator, but the complete hydraulic and electromechanical circuits as well as the dynamics of the climbing robot. This analysis allows modeling of the time-varying applied pressure, electrical current, and actuator contraction for accurate prediction of the robot motion, energy consumption, and mechanical work output. The developed model is first validated against mechanical and electrical data collected from a proof-of-concept prototype robot. The model is then employed to study the system-level sensitivities of the robot locomotion efficiency and average climbing speed to several design and operating parameters. The results of this analysis demonstrate that considering only the transduction efficiency of the FAM actuators is insufficient to maximize the efficiency of the complete robot, and that a holistic approach can lead to significant improvements in performance. © 2017 IOP Publishing Ltd.

  18. Development of numerical simulation technology for high resolution thermal hydraulic analysis

    International Nuclear Information System (INIS)

    Yoon, Han Young; Kim, K. D.; Kim, B. J.; Kim, J. T.; Park, I. K.; Bae, S. W.; Song, C. H.; Lee, S. W.; Lee, S. J.; Lee, J. R.; Chung, S. K.; Chung, B. D.; Cho, H. K.; Choi, S. K.; Ha, K. S.; Hwang, M. K.; Yun, B. J.; Jeong, J. J.; Sul, A. S.; Lee, H. D.; Kim, J. W.

    2012-04-01

    A realistic simulation of two phase flows is essential for the advanced design and safe operation of a nuclear reactor system. The need for a multi dimensional analysis of thermal hydraulics in nuclear reactor components is further increasing with advanced design features, such as a direct vessel injection system, a gravity driven safety injection system, and a passive secondary cooling system. These features require more detailed analysis with enhanced accuracy. In this regard, KAERI has developed a three dimensional thermal hydraulics code, CUPID, for the analysis of transient, multi dimensional, two phase flows in nuclear reactor components. The code was designed for use as a component scale code, and/or a three dimensional component, which can be coupled with a system code. This report presents an overview of the CUPID code development and preliminary assessment, mainly focusing on the numerical solution method and its verification and validation. It was shown that the CUPID code was successfully verified. The results of the validation calculations show that the CUPID code is very promising, but a systematic approach for the validation and improvement of the physical models is still needed

  19. Sensitivity analysis of thermal hydraulic response in containment at core meltdown accident

    International Nuclear Information System (INIS)

    Kobayashi, Kensuke; Ishigami, Tsutomu; Horii, Hideo; Chiba, Takemi.

    1985-01-01

    A sensitivity analysis of thermal hydraulic response in a containment during a 'station blackout' (the loss of all AC power) accident at Browns Ferry unit one plant was performed with the computer code MARCH 1.0. In the analysis, the plant station batteries were assumed to be available for 4h after the initiation of the accident. The thermal hydraulic response in the containment was calculated by varying several input data for MARCH 1.0 independently and the deviation among calculated results were investigated. The sensitivity analysis showed that (a) the containment would fail due to the overtemperature without any operator actions for plant recovery, which would be strongly dependent on the model of the debris-concrete interaction and the input parameters for specifying the containment failure modes in MARCH 1.0, (b) a core melting temperature and an amount of water left in a primary system at the end of the meltdown were identified as important parameters which influenced the time of the containment failure, and (c) experimental works regarding the parameters mentioned above could be recommended. (author)

  20. ANALYSIS OF HYDRAULIC LOAD OF SELECTED WASTEWATER TREATMENT PLANT IN JASŁO COUNTY

    Directory of Open Access Journals (Sweden)

    Dariusz Piotr Młyński

    2016-12-01

    Full Text Available The paper presents an analysis of hydraulic load in selected a wastewater treatment plant (WTP in Jasło County: in Przysieki, Kołaczyce and Szebnie. The study was based on the records of daily sewage volume entering the treatment plants within a multi-year period of 2011-2014. The analysis took into account the average daily amount of incoming sewage, the maximum daily peaking factor for the incoming sewage, changes in the sewage volume depending on specific month and the intervals with the greatest frequency of occurrence were designated. The analysis revealed that investigated wastewater treatment plants were hydraulically underloaded. Moreover it was conclude a significant variables of inflowing sewage amount. The sewage admission was the largest in spring and summer periods. Sewage volume interval most often occurring at the WTP in Przysieki was the one between 320 and 480 m3•d-1, for Kołaczyce between 290 and 320 m3•d-1 and Szebnie between 120 and 240 m3•d-1.

  1. Numerical analysis of the performance of rock weirs: Effects of structure configuration on local hydraulics

    Science.gov (United States)

    Holmquist-Johnson, C. L.

    2009-01-01

    River spanning rock structures are being constructed for water delivery as well as to enable fish passage at barriers and provide or improve the aquatic habitat for endangered fish species. Current design methods are based upon anecdotal information applicable to a narrow range of channel conditions. The complex flow patterns and performance of rock weirs is not well understood. Without accurate understanding of their hydraulics, designers cannot address the failure mechanisms of these structures. Flow characteristics such as jets, near bed velocities, recirculation, eddies, and plunging flow govern scour pool development. These detailed flow patterns can be replicated using a 3D numerical model. Numerical studies inexpensively simulate a large number of cases resulting in an increased range of applicability in order to develop design tools and predictive capability for analysis and design. The analysis and results of the numerical modeling, laboratory modeling, and field data provide a process-based method for understanding how structure geometry affects flow characteristics, scour development, fish passage, water delivery, and overall structure stability. Results of the numerical modeling allow designers to utilize results of the analysis to determine the appropriate geometry for generating desirable flow parameters. The end product of this research will develop tools and guidelines for more robust structure design or retrofits based upon predictable engineering and hydraulic performance criteria. ?? 2009 ASCE.

  2. Selected hydraulic test analysis techniques for constant-rate discharge tests

    International Nuclear Information System (INIS)

    Spane, F.A. Jr.

    1993-03-01

    The constant-rate discharge test is the principal field method used in hydrogeologic investigations for characterizing the hydraulic properties of aquifers. To implement this test, the aquifer is stressed by withdrawing ground water from a well, by using a downhole pump. Discharge during the withdrawal period is regulated and maintained at a constant rate. Water-level response within the well is monitored during the active pumping phase (i.e., drawdown) and during the subsequent recovery phase following termination of pumping. The analysis of drawdown and recovery response within the stress well (and any monitored, nearby observation wells) provides a means for estimating the hydraulic properties of the tested aquifer, as well as discerning formational and nonformational flow conditions (e.g., wellbore storage, wellbore damage, presence of boundaries, etc.). Standard analytical methods that are used for constant-rate pumping tests include both log-log type-curve matching and semi-log straight-line methods. This report presents a current ''state of the art'' review of selected transient analysis procedures for constant-rate discharge tests. Specific topics examined include: analytical methods for constant-rate discharge tests conducted within confined and unconfined aquifers; effects of various nonideal formation factors (e.g., anisotropy, hydrologic boundaries) and well construction conditions (e.g., partial penetration, wellbore storage) on constant-rate test response; and the use of pressure derivatives in diagnostic analysis for the identification of specific formation, well construction, and boundary conditions

  3. Development of a thermal-hydraulic code for reflood analysis in a PWR experimental loop

    International Nuclear Information System (INIS)

    Alves, Sabrina P.; Mesquita, Amir Z.; Rezende, Hugo C.; Palma, Daniel A.P.

    2017-01-01

    A process of fundamental importance in the event of Loss of Coolant Accident (LOCA) in Pressurized Water nuclear Reactors (PWR) is the reflood of the core or rewetting of nuclear fuels. The Nuclear Technology Development Center (CDTN) has been developing since the 70’s programs to allow Brazil to become independent in the field of reactor safety analysis. To that end, in the 80’s was designed, assembled and commissioned one Rewetting Test Facility (ITR in Portuguese). This facility aims to investigate the phenomena involved in the thermal hydraulic reflood phase of a Loss of Coolant Accident in a PWR nuclear reactor. This work aim is the analysis of physical and mathematical models governing the rewetting phenomenon, and the development a thermo-hydraulic simulation code of a representative experimental circuit of the PWR reactors core cooling channels. It was possible to elaborate and develop a code called REWET. The results obtained with REWET were compared with the experimental results of the ITR, and with the results of the Hydroflut code, that was the old program previously used. An analysis was made of the evolution of the wall temperature of the test section as well as the evolution of the front for two typical tests using the two codes calculation, and experimental results. The result simulated by REWET code for the rewetting time also came closer to the experimental results more than those calculated by Hydroflut code. (author)

  4. Verification of combined thermal-hydraulic and heat conduction analysis code FLOWNET/TRUMP

    International Nuclear Information System (INIS)

    Maruyama, Soh; Fujimoto, Nozomu; Sudo, Yukio; Kiso, Yoshihiro; Murakami, Tomoyuki.

    1988-09-01

    This report presents the verification results of the combined thermal-hydraulic and heat conduction analysis code, FLOWNET/TRUMP which has been utilized for the core thermal hydraulic design, especially for the analysis of flow distribution among fuel block coolant channels, the determination of thermal boundary conditions for fuel block stress analysis and the estimation of fuel temperature in the case of fuel block coolant channel blockage accident in the design of the High Temperature Engineering Test Reactor(HTTR), which the Japan Atomic Energy Research Institute has been planning to construct in order to establish basic technologies for future advanced very high temperature gas-cooled reactors and to be served as an irradiation test reactor for promotion of innovative high temperature new frontier technologies. The verification of the code was done through the comparison between the analytical results and experimental results of the Helium Engineering Demonstration Loop Multi-channel Test Section(HENDEL T 1-M ) with simulated fuel rods and fuel blocks. (author)

  5. Verification of combined thermal-hydraulic and heat conduction analysis code FLOWNET/TRUMP

    Science.gov (United States)

    Maruyama, Soh; Fujimoto, Nozomu; Kiso, Yoshihiro; Murakami, Tomoyuki; Sudo, Yukio

    1988-09-01

    This report presents the verification results of the combined thermal-hydraulic and heat conduction analysis code, FLOWNET/TRUMP which has been utilized for the core thermal hydraulic design, especially for the analysis of flow distribution among fuel block coolant channels, the determination of thermal boundary conditions for fuel block stress analysis and the estimation of fuel temperature in the case of fuel block coolant channel blockage accident in the design of the High Temperature Engineering Test Reactor(HTTR), which the Japan Atomic Energy Research Institute has been planning to construct in order to establish basic technologies for future advanced very high temperature gas-cooled reactors and to be served as an irradiation test reactor for promotion of innovative high temperature new frontier technologies. The verification of the code was done through the comparison between the analytical results and experimental results of the Helium Engineering Demonstration Loop Multi-channel Test Section(HENDEL T(sub 1-M)) with simulated fuel rods and fuel blocks.

  6. Development of a thermal-hydraulic code for reflood analysis in a PWR experimental loop

    Energy Technology Data Exchange (ETDEWEB)

    Alves, Sabrina P.; Mesquita, Amir Z.; Rezende, Hugo C., E-mail: sabrinapral@gmail.com, E-mail: amir@cdtn.brm, E-mail: hcr@cdtn.br, E-mail: hcr@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Palma, Daniel A.P., E-mail: dapalma@cnen.gov.br [Comissão Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    A process of fundamental importance in the event of Loss of Coolant Accident (LOCA) in Pressurized Water nuclear Reactors (PWR) is the reflood of the core or rewetting of nuclear fuels. The Nuclear Technology Development Center (CDTN) has been developing since the 70’s programs to allow Brazil to become independent in the field of reactor safety analysis. To that end, in the 80’s was designed, assembled and commissioned one Rewetting Test Facility (ITR in Portuguese). This facility aims to investigate the phenomena involved in the thermal hydraulic reflood phase of a Loss of Coolant Accident in a PWR nuclear reactor. This work aim is the analysis of physical and mathematical models governing the rewetting phenomenon, and the development a thermo-hydraulic simulation code of a representative experimental circuit of the PWR reactors core cooling channels. It was possible to elaborate and develop a code called REWET. The results obtained with REWET were compared with the experimental results of the ITR, and with the results of the Hydroflut code, that was the old program previously used. An analysis was made of the evolution of the wall temperature of the test section as well as the evolution of the front for two typical tests using the two codes calculation, and experimental results. The result simulated by REWET code for the rewetting time also came closer to the experimental results more than those calculated by Hydroflut code. (author)

  7. Hydraulic analysis of river training cross-vanes as part of post-restoration monitoring

    Directory of Open Access Journals (Sweden)

    T. A. Endreny

    2011-07-01

    Full Text Available River restoration design methods are incrementally improved by studying and learning from monitoring data in previous projects. In this paper we report post-restoration monitoring data and simulation analysis for a Natural Channel Design (NCD restoration project along 1600 m of the Batavia Kill (14 km2 watershed in the Catskill Mountains, NY. The restoration project was completed in 2002 with goals to reduce bank erosion and determine the efficacy of NCD approaches for restoring headwater streams in the Catskill Mountains, NY. The NCD approach used a reference-reach to determine channel form, empirical relations between the project site and reference site bankfull dimensions to size channel geometry, and hydraulic and sediment computations based on a bankfull (1.3 yr return interval discharge to test channel capacity and sediment stability. The NCD project included 12 cross-vanes and 48 j-hook vanes as river training structures along 19 meander bends to protect against bank erosion and maintain scour pools for fish habitat. Monitoring data collected from 2002 to 2004 were used to identify aggradation of pools in meander bends and below some structures. Aggradation in pools was attributed to the meandering riffle-pool channel trending toward step-pool morphology and cross-vane arms not concentrating flow in the center of the channel. The aggradation subsequently caused flow splitting and 4 partial point bar avulsions during a spring 2005 flood with a 25-yr return interval. Processing the pre-flood monitoring data with hydraulic analysis software provided clues the reach was unstable and preventative maintenance was needed. River restoration and monitoring teams should be trained in robust hydraulic analytical methods that help them extend project restoration goals and structure stability.

  8. 3-D thermal hydraulic analysis of transient heat removal from fast reactor core using immersion coolers

    International Nuclear Information System (INIS)

    Chvetsov, I.; Volkov, A.

    2000-01-01

    For advanced fast reactors (EFR, BN-600M, BN-1600, CEFR) the special complementary loop is envisaged in order to ensure the decay heat removal from the core in the case of LOF accidents. This complementary loop includes immersion coolers that are located in the hot reactor plenum. To analyze the transient process in the reactor when immersion coolers come into operation one needs to involve 3-D thermal hydraulics code. Furthermore sometimes the problem becomes more complicated due to necessity of simulation of the thermal hydraulics processes into the core interwrapper space. For example on BN-600M and CEFR reactors it is supposed to ensure the effective removal of decay heat from core subassemblies by specially arranged internal circulation circuit: 'inter-wrapper space'. For thermal hydraulics analysis of the transients in the core and in the whole reactor including hot plenum with immersion coolers and considering heat and mass exchange between the main sodium flow and sodium that moves in the inter-wrapper space the code GRIFIC (the version of GRIF code family) was developed in IPPE. GRIFIC code was tested on experimental data obtained on RAMONA rig under conditions simulating decay heat removal of a reactor with the use of immersion coolers. Comparison has been made of calculated and experimental result, such as integral characteristics (flow rate through the core and water temperature at the core inlet and outlet) and the local temperatures (at thermocouple location) as well. In order to show the capabilities of the code some results of the transient analysis of heat removal from the core of BN-600M - type reactor under loss-of-flow accident are presented. (author)

  9. Analysis of BTEX groundwater concentrations from surface spills associated with hydraulic fracturing operations.

    Science.gov (United States)

    Gross, Sherilyn A; Avens, Heather J; Banducci, Amber M; Sahmel, Jennifer; Panko, Julie M; Tvermoes, Brooke E

    2013-04-01

    Concerns have arisen among the public regarding the potentialfor drinking-water contamination from the migration of methane gas and hazardous chemicals associated with hydraulic fracturing and horizontal drilling. However, little attention has been paid to the potentialfor groundwater contamination resulting from surface spills from storage and production facilities at active well sites. We performed a search for publically available data regarding groundwater contamination from spills at ULS. drilling sites. The Colorado Oil and Gas Conservation Commission (COGCC) database was selected for further analysis because it was the most detailed. The majority ofspills were in Weld County, Colorado, which has the highest density of wells that used hydraulic fracturing for completion, many producing both methane gas and crude oil. We analyzed publically available data reported by operators to the COGCC regarding surface spills that impacted groundwater From July 2010 to July 2011, we noted 77 reported surface spills impacting the groundwater in Weld County, which resulted in surface spills associated with less than 0.5% of the active wells. The reported data included groundwater samples that were analyzed for benzene, toluene, ethylbenzene, andxylene (BTEX) components of crude oil. For groundwater samples taken both within the spill excavation area and on the first reported date of sampling, the BTEX measurements exceeded National Drinking Water maximum contaminant levels (MCLs) in 90, 30, 12, and 8% of the samples, respectively. However, actions taken to remediate the spills were effective at reducing BJTEX levels, with at least 84% of the spills reportedly achieving remediation as of May 2012. Our analysis demonstrates that surface spills are an important route of potential groundwater contamination from hydraulic fracturing activities and should be a focus of programs to protect groundwater While benzene can occur naturally in groundwater sources, spills and migration

  10. An analysis of chemicals and other constituents found in produced water from hydraulically fractured wells in California and the challenges for wastewater management.

    Science.gov (United States)

    Chittick, Emily A; Srebotnjak, Tanja

    2017-12-15

    As high-volume hydraulic fracturing (HF) has grown substantially in the United States over the past decade, so has the volume of produced water (PW), i.e., briny water brought to the surface as a byproduct of oil and gas production. According to a recent study (Groundwater Protection Council, 2015), more than 21 billion barrels of PW were generated in 2012. In addition to being high in TDS, PW may contain hydrocarbons, PAH, alkylphenols, naturally occurring radioactive material (NORM), metals, and other organic and inorganic substances. PW from hydraulically fractured wells includes flowback water, i.e., injection fluids containing chemicals and additives used in the fracturing process such as friction reducers, scale inhibitors, and biocides - many of which are known to cause serious health effects. It is hence important to gain a better understanding of the chemical composition of PW and how it is managed. This case study of PW from hydraulically fractured wells in California provides a first aggregate chemical analysis since data collection began in accordance with California's 2013 oil and gas well stimulation law (SB4, Pavley). The results of analyzing one-time wastewater analyses of 630 wells hydraulically stimulated between April 1, 2014 and June 30, 2015 show that 95% of wells contained measurable and in some cases elevated concentrations of BTEX and PAH compounds. PW from nearly 500 wells contained lead, uranium, and/or other metals. The majority of hazardous chemicals known to be used in HF operations, including formaldehyde and acetone, are not reported in the published reports. The prevalent methods for dealing with PW in California - underground injection and open evaporation ponds - are inadequate for this waste stream due to risks from induced seismicity, well integrity failure, well upsets, accidents and spills. Beneficial reuse of PW, such as for crop irrigation, is as of yet insufficiently safety tested for consumers and agricultural workers as

  11. The preliminary thermal–hydraulic analysis of a water cooled blanket concept design based on RELAP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Guanghuai; Peng, Changhong; Guo, Yun, E-mail: guoyun79@ustc.edu.cn

    2016-11-01

    Highlights: • The superheated steam and PWR schemes are analyzed by RELAP5 code. • The influence of non-uniform heating sources is include. • A supposed slow flow decrease case is discussed and the PWR scheme is better. - Abstract: Water cooled blanket (WCB) is very important in the conceptual design and energy transfer in future fusion power plant. One conceptual design of WCB is under computational testing. RELAP5 code, which is mature and often used in transient analysis in Pressurizer water reactor (PWR), is selected as the simulation tool. The complex inner flow channels and heat sources are simplified according to its thermal–hydraulic characteristics. Then the nodal model for REALP5 is built for approximating the conceptual design. Two typical operating plans, superheated steam scheme and PWR scheme, are analyzed. After some adjustments of the inlet flow resistance coefficients of some flow channels, the reasonable stable conditions of both operation plans can be obtained. The stable fluid and wall temperature distributions and pressure drops are studied. At last, a supposed slow flow decreasing is discussed under two operating conditions separately. According to present results, the superheated steam scheme still needs to be further optimized. The PWR scheme shows a very good safety feature.

  12. The preliminary thermal–hydraulic analysis of a water cooled blanket concept design based on RELAP5 code

    International Nuclear Information System (INIS)

    Wang, Guanghuai; Peng, Changhong; Guo, Yun

    2016-01-01

    Highlights: • The superheated steam and PWR schemes are analyzed by RELAP5 code. • The influence of non-uniform heating sources is include. • A supposed slow flow decrease case is discussed and the PWR scheme is better. - Abstract: Water cooled blanket (WCB) is very important in the conceptual design and energy transfer in future fusion power plant. One conceptual design of WCB is under computational testing. RELAP5 code, which is mature and often used in transient analysis in Pressurizer water reactor (PWR), is selected as the simulation tool. The complex inner flow channels and heat sources are simplified according to its thermal–hydraulic characteristics. Then the nodal model for REALP5 is built for approximating the conceptual design. Two typical operating plans, superheated steam scheme and PWR scheme, are analyzed. After some adjustments of the inlet flow resistance coefficients of some flow channels, the reasonable stable conditions of both operation plans can be obtained. The stable fluid and wall temperature distributions and pressure drops are studied. At last, a supposed slow flow decreasing is discussed under two operating conditions separately. According to present results, the superheated steam scheme still needs to be further optimized. The PWR scheme shows a very good safety feature.

  13. Thermal-Hydraulic Analysis for SBLOCA in OPR1000 and Evaluation of Uncertainty for PSA

    International Nuclear Information System (INIS)

    Kim, Tae Jin; Park, Goon Cherl

    2012-01-01

    Probabilistic Safety assessment (PSA) is a mathematical tool to evaluate numerical estimates of risk for nuclear power plants (NPPs). But PSA has the problems about quality and reliability since the quantification of uncertainties from thermal hydraulic (TH) analysis has not been included in the quantification of overall uncertainties in PSA. From the former research, it is proved that the quantification of uncertainties from best-estimate LBLOCA analysis can improve the PSA quality by modifying the core damage frequency (CDF) from the existing PSA report. Basing on the similar concept, this study considers the quantification of SBLOCA analysis results. In this study, however, operator error parameters are also included in addition to the phenomenon parameters which are considered in LBLOCA analysis

  14. Difference analysis for fluid-structure interaction

    International Nuclear Information System (INIS)

    Giencke, E.; Forkel, M.

    1979-01-01

    For solving fluid structure interaction problems it is possible to organize the compter programs for the difference method in the same way as for the finite element method by establishing the difference equations with the principial of virtual work. In the finite element method the individual localized functions for the approximation of the potential function PHI will be chosen also as virtual functions delta PHI. Deriving difference equations the virtual states are simple as possible and the approximation of the potential function may be linear or parabolic. The equations become symmetric both for points in the interiour and the boundaries and for grids with rectangular and triangular elements. The boundary and edge-conditions shall established for elastic walls and for the free surface. For regular rectangular and triangular grids it is possible to derive on the same way multipoint difference equations, which for the same numbers of unknowns are two orders better in accuracy as the usual difference or the finite element equations. Some examples for the pressure distribution in a BWR-steel-containment due to steam bubble collaps at the condenser pipes will be shown. (orig.)

  15. 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

  16. Development of subchannel analysis code MATRA-LMR for KALIMER subassembly thermal-hydraulics

    International Nuclear Information System (INIS)

    Won-Seok Kim; Young-Gyun Kim

    2000-01-01

    In the sodium cooled liquid metal reactors, the design limit are imposed on the maximum temperatures of claddings and fuel pins. Thus an accurate prediction of core coolant/fuel temperature distribution is essential to the LMR core thermal-hydraulic design. The detailed subchannel thermal-hydraulic analysis code MATRA-LMR (Multichannel Analyzer for Steady States and Transients in Rod Arrays for Liquid Metal Reactors) is being developed for KALIMER core design and analysis, based on COBRA-IV-i and MATRA. The major modifications and improvements implemented into MATRA-LMR are as follows: a) nonuniform axial noding capability, b) sodium properties calculation subprogram, c) sodium coolant heat transfer correlations, and d) most recent pressure drop correlations, such as Novendstern, Chiu-Rohsenow-Todreas and Cheng-Todreas. To assess the development status of this code, the benchmark calculations were performed with the ORNL 19 pin tests and EBR-II seven-assembly SLTHEN calculation results. The calculation results of MATRA-LMR for ORNL 19-pin assembly tests and EBR-II 91-pin experiments were compared to the measurements, and to SABRE4 and SLTHEN code calculation results, respectively. In this comparison, the differences are found among the three codes because of the pressure drop and the thermal mixing modellings. Finally, the major technical results of the conceptual design for the KALIMER 98.03 core have been compared with the calculations of MATRA-LMR, SABRE4 and SLTHEN codes. (author)

  17. 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.

  18. 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)

  19. Analysis of selenium in body fluids: A review

    International Nuclear Information System (INIS)

    Alaejos, M.S.; Romero, C.D.

    1995-01-01

    This article reviews numerous analytical techniques for determining trace amounts of selenium in body fluids. In addition, sampling storage and treatment procedures are evaluated. The analytical techniques reviewed include the following: spectrofluorometry and spectrophotometry; atomic absorption spectrometry; fluorescence and atomic emission spectroscopy; mass spectroscopy; X-ray spectrometric analysis; neutron activation analysis; chromatographic methods; and electrochemical methods. 469 refs

  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. Analysis and selection of a system for hydraulic transport of slags in the Mironovskii power plant

    Energy Technology Data Exchange (ETDEWEB)

    Mirgorodskii, V.G.; Mova, M.E.; Korenev, V.E.; Grechikhin, Yu.A. (Donetskii Politekhnicheskii Institut (USSR))

    1991-01-01

    Discusses systems for hydraulic transport of ashes and slags from combustion of black coal (with an ash content of 40.5%) in the Mironovskii power plant. Three systems are comparatively evaluated: hydraulic transport under influence of gravity, hydraulic transport with a system of dredging pumps, or an airlift pump system. Design of each system, its operation and types of pumps or airlift systems are discussed. The evaluation concentrates on the hydraulic transport system with 1 to 3 airlift pumps each with a capacity ranging from 110 to 890 m{sup 3}/h. Optimum design of the airlift hydraulic system for slag and ash transport is described.

  2. Gastric fluid versus amniotic fluid analysis for the identification of intra-amniotic infection due to Ureaplasma species.

    Science.gov (United States)

    Kim, Sun Min; Romero, Roberto; Lee, JoonHo; Chaemsaithong, Piya; Docheva, Nikolina; Yoon, Bo Hyun

    2016-01-01

    Early neonatal sepsis is often due to intra-amniotic infection. The stomach of the neonate contains fluid swallowed before and during delivery. The presence of bacteria as well as neutrophils detected by culture or Gram stain of the gastric fluid during the first day of life is suggestive of exposure to bacteria or inflammation. We undertook this study to determine the relationship between gastric fluid analysis and amniotic fluid obtained by transabdominal amniocentesis in the detection of Ureaplasma species, the most frequent microorganisms responsible for intra-amniotic infection. The study population consisted of 100 singleton pregnant women who delivered preterm neonates (Ureaplasma species was performed. Intra-amniotic inflammation was defined as an elevated amniotic fluid matrix metalloproteinase-8 concentration (>23 ng/mL). (1) Ureaplasma species were detected by culture or PCR in 18% (18/100) of amniotic fluid samples and in 5% (5/100) of gastric fluid samples; (2) among the amniotic fluid cases positive for Ureaplasma species, these microorganisms were identified in 27.8% (5/18) of gastric fluid samples; (3) none of the cases negative for Ureaplasma species in the amniotic fluid were found to be positive for these microorganisms in the gastric fluid; (4) patients with amniotic fluid positive for Ureaplasma species but with gastric fluid negative for these microorganisms had a significantly higher rate of intra-amniotic inflammation, acute histologic chorioamnionitis, and neonatal death than those with both amniotic fluid and gastric fluid negative for Ureaplasma species; and (5) no significant differences were observed in the rate of intra-amniotic inflammation, acute histologic chorioamnionitis, and neonatal death between patients with amniotic fluid positive for Ureaplasma species but with gastric fluid negative for these microorganisms and those with both amniotic fluid and gastric fluid positive for Ureaplasma species. Gastric fluid analysis has 100

  3. Linear and nonlinear analysis of fluid slosh dampers

    Science.gov (United States)

    Sayar, B. A.; Baumgarten, J. R.

    1982-11-01

    A vibrating structure and a container partially filled with fluid are considered coupled in a free vibration mode. To simplify the mathematical analysis, a pendulum model to duplicate the fluid motion and a mass-spring dashpot representing the vibrating structure are used. The equations of motion are derived by Lagrange's energy approach and expressed in parametric form. For a wide range of parametric values the logarithmic decrements of the main system are calculated from theoretical and experimental response curves in the linear analysis. However, for the nonlinear analysis the theoretical and experimental response curves of the main system are compared. Theoretical predictions are justified by experimental observations with excellent agreement. It is concluded finally that for a proper selection of design parameters, containers partially filled with viscous fluids serve as good vibration dampers.

  4. Computation and analysis of cavitating flow in Francis-class hydraulic turbines

    Science.gov (United States)

    Leonard, Daniel J.

    Hydropower is the most proven renewable energy technology, supplying the world with 16% of its electricity. Conventional hydropower generates a vast majority of that percentage. Although a mature technology, hydroelectric generation shows great promise for expansion through new dams and plants in developing hydro countries. Moreover, in developed hydro countries, such as the United States, installing generating units in existing dams and the modern refurbishment of existing plants can greatly expand generating capabilities with little to no further impact on the environment. In addition, modern computational technology and fluid dynamics expertise has led to substantial improvements in modern turbine design and performance. Cavitation has always presented a problem in hydroturbines, causing performance breakdown, erosion, damage, vibration, and noise. While modern turbines are usually designed to be cavitation-free at their best efficiency point, due to the variable demand of the energy market it is fairly common to operate at off-design conditions. Here, cavitation and its deleterious effects are unavoidable, and hence, cavitation is a limiting factor on the design and operation of these turbines. Multiphase Computational Fluid Dynamics (CFD) has been used in recent years to model cavitating flow for a large range of problems, including turbomachinery. However, CFD of cavitating flow in hydroturbines is still in its infancy. This dissertation presents steady-periodic Reynolds-averaged Navier-Stokes simulations of a cavitating Francis-class hydroturbine at model and prototype scales. Computational results of the reduced-scale model and full-scale prototype, undergoing performance breakdown, are compared with empirical model data and prototype performance estimations based on standard industry scalings from the model data. Mesh convergence of the simulations is also displayed. Comparisons are made between the scales to display that cavitation performance breakdown

  5. Thermal-hydraulic and neutronic analysis of pressurized water reactor cores

    International Nuclear Information System (INIS)

    Alves, C.H.

    1982-01-01

    A computational code, named CANAL2, was developed for the simulation of the steady-state and transient behaviour of a Pressurized Water Reactor core. The conservation equations for the control volumes are obtained by area-averaging of the two-fluid model conservation equations and reducing them to the drift-flux model formulation. The resulting equations are aproximated by finite differences and solved by a marching-type numerical scheme. The model takes into account the exchange of mass, momentum and energy between adjacent subchannels of a fuel bundle. Turbulent mixing and diversion crossflow are considered. Correlations are provided for several heat trans and flow regimes and selected according to the local conditons. During transients core power can be evaluated by a point-Kinetics model. Fuel and coolant temperatures are feedback to the neutronics. The heat conduction equation is solved in the fuel using the Crank-Nicolson scheme. Temperature-dependent correlations are provided for the fuel and cladding thermal conductivities. Several runs were made with the code CANAL2 using the available experimental and calculated data in the open literature. Results indicate that CANAL2 is a good calculational tool for the thermal-hydraulics of PWR cores. A few refinements will make the code useful for design. (Author) [pt

  6. 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

  7. Coupled Viscous Fluid Flow and Joint Deformation Analysis for Grout Injection in a Rock Joint

    Science.gov (United States)

    Kim, Hyung-Mok; Lee, Jong-Won; Yazdani, Mahmoud; Tohidi, Elham; Nejati, Hamid Reza; Park, Eui-Seob

    2018-02-01

    Fluid flow modeling is a major area of interest within the field of rock mechanics. The main objective of this study is to gain insight into the performance of grout injection inside jointed rock masses by numerical modeling of grout flow through a single rock joint. Grout flow has been widely simulated using non-Newtonian Bingham fluid characterized by two main parameters of dynamic viscosity and shear yield strength both of which are time dependent. The increasing value of these properties with injection time will apparently affect the parameters representing the grouting performance including grout penetration length and volumetric injection rate. In addition, through hydromechanical coupling a mutual influence between the injection pressure from the one side and the joint opening/closing behavior and the aperture profile variation on the other side is anticipated. This is capable of producing a considerable impact on grout spread within the rock joints. In this study based on the Bingham fluid model, a series of numerical analysis has been conducted using UDEC to simulate the flow of viscous grout in a single rock joint with smooth parallel surfaces. In these analyses, the time-dependent evolution of the grout fluid properties and the hydromechanical coupling have been considered to investigate their impact on grouting performance. In order to verify the validity of these simulations, the results of analyses including the grout penetration length and the injection flow rate were compared with a well-known analytical solution which is available for the simple case of constant grout properties and non-coupled hydraulic analysis. The comparison demonstrated that the grout penetration length can be overestimated when the time-dependent hardening of grout material is not considered. Moreover, due to the HM coupling, it was shown that the joint opening induced by injection pressure may have a considerable increasing impression on the values of penetration length and

  8. Fundamental approaches for analysis thermal hydraulic parameter for Puspati Research Reactor

    International Nuclear Information System (INIS)

    Hashim, Zaredah; Lanyau, Tonny Anak; Farid, Mohamad Fairus Abdul; Kassim, Mohammad Suhaimi; Azhar, Noraishah Syahirah

    2016-01-01

    The 1-MW PUSPATI Research Reactor (RTP) is the one and only nuclear pool type research reactor developed by General Atomic (GA) in Malaysia. It was installed at Malaysian Nuclear Agency and has reached the first criticality on 8 June 1982. Based on the initial core which comprised of 80 standard TRIGA fuel elements, the very fundamental thermal hydraulic model was investigated during steady state operation using the PARET-code. The main objective of this paper is to determine the variation of temperature profiles and Departure of Nucleate Boiling Ratio (DNBR) of RTP at full power operation. The second objective is to confirm that the values obtained from PARET-code are in agreement with Safety Analysis Report (SAR) for RTP. The code was employed for the hot and average channels in the core in order to calculate of fuel’s center and surface, cladding, coolant temperatures as well as DNBR’s values. In this study, it was found that the results obtained from the PARET-code showed that the thermal hydraulic parameters related to safety for initial core which was cooled by natural convection was in agreement with the designed values and safety limit in SAR

  9. Fundamental approaches for analysis thermal hydraulic parameter for Puspati Research Reactor

    Science.gov (United States)

    Hashim, Zaredah; Lanyau, Tonny Anak; Farid, Mohamad Fairus Abdul; Kassim, Mohammad Suhaimi; Azhar, Noraishah Syahirah

    2016-01-01

    The 1-MW PUSPATI Research Reactor (RTP) is the one and only nuclear pool type research reactor developed by General Atomic (GA) in Malaysia. It was installed at Malaysian Nuclear Agency and has reached the first criticality on 8 June 1982. Based on the initial core which comprised of 80 standard TRIGA fuel elements, the very fundamental thermal hydraulic model was investigated during steady state operation using the PARET-code. The main objective of this paper is to determine the variation of temperature profiles and Departure of Nucleate Boiling Ratio (DNBR) of RTP at full power operation. The second objective is to confirm that the values obtained from PARET-code are in agreement with Safety Analysis Report (SAR) for RTP. The code was employed for the hot and average channels in the core in order to calculate of fuel's center and surface, cladding, coolant temperatures as well as DNBR's values. In this study, it was found that the results obtained from the PARET-code showed that the thermal hydraulic parameters related to safety for initial core which was cooled by natural convection was in agreement with the designed values and safety limit in SAR.

  10. Design and thermal-hydraulic analysis of PFC baking for SST-1 Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Chaudhuri, Paritosh E-mail: paritosh@ipr.res.in; Reddy, D. Chenna; Khirwadkar, S.; Prakash, N. Ravi; Santra, P.; Saxena, Y.C

    2001-09-01

    The Steady-State Superconducting Tokamak (SST-1) is a medium-size tokamak with super-conducting magnetic field coils. Plasma facing components (PFC) of the SST-1, consisting of divertors, passive stabilisers, baffles, and poloidal limiters, are designed to be compatible for steady-state operation. Except for the poloidal limiters, all other PFC are structurally continuous in the toroidal direction. As SST-1 is designed to run double-null divertor plasmas, these components also have up-down symmetry. A closed divertor configuration is chosen to produce high recycling and high pumping speed in the divertor region. The passive stabilisers are located close to the plasma to provide stability against the vertical instability of the elongated plasma. The main consideration in the design of the PFC is the steady-state heat removal of up to 1 MW/m{sup 2}. In addition to removing high heat fluxes, the PFC are also designed to be compatible for baking at 350 deg. C. Different flow parameters and various tube layouts have been examined to select the optimum thermal-hydraulic parameters and tube layout for different PFC of SST-1. Thermal response of the PFC during baking has been performed analytically (using a Fortran code) and two-dimensional finite element analysis using ANSYS. The detailed thermal hydraulics and thermal responses of PFC baking is presented in this paper.

  11. Design and thermal-hydraulic analysis of PFC baking for SST-1 Tokamak

    International Nuclear Information System (INIS)

    Chaudhuri, Paritosh; Reddy, D. Chenna; Khirwadkar, S.; Prakash, N. Ravi; Santra, P.; Saxena, Y.C.

    2001-01-01

    The Steady-State Superconducting Tokamak (SST-1) is a medium-size tokamak with super-conducting magnetic field coils. Plasma facing components (PFC) of the SST-1, consisting of divertors, passive stabilisers, baffles, and poloidal limiters, are designed to be compatible for steady-state operation. Except for the poloidal limiters, all other PFC are structurally continuous in the toroidal direction. As SST-1 is designed to run double-null divertor plasmas, these components also have up-down symmetry. A closed divertor configuration is chosen to produce high recycling and high pumping speed in the divertor region. The passive stabilisers are located close to the plasma to provide stability against the vertical instability of the elongated plasma. The main consideration in the design of the PFC is the steady-state heat removal of up to 1 MW/m 2 . In addition to removing high heat fluxes, the PFC are also designed to be compatible for baking at 350 deg. C. Different flow parameters and various tube layouts have been examined to select the optimum thermal-hydraulic parameters and tube layout for different PFC of SST-1. Thermal response of the PFC during baking has been performed analytically (using a Fortran code) and two-dimensional finite element analysis using ANSYS. The detailed thermal hydraulics and thermal responses of PFC baking is presented in this paper

  12. Thermo-hydraulic analysis of the generic equatorial port plug design

    International Nuclear Information System (INIS)

    Rodríguez, E.; Guirao, J.; Ordieres, J.; Cortizo, J.L.; Iglesias, S.

    2012-01-01

    Highlights: ► Thermo-hydraulic transient performance evaluation and optimization of the GEPP structure cooling/heating system under neutronic heating and baking conditions. ► The optimization of the GEPP box structure's cooling system includes positioning and minimization of number and size of gun drilled channels, complying with the flow and functional requirements during operating and baking conditions. - Abstract: The port-based ITER diagnostic systems are housed primarily in two locations, the equatorial and upper port plugs. The port plug structure provides confinement function, maintains ultra-high vacuum quality and the first confinement barrier for radioactive materials at the ports. The port plug structure design, from the ITER International Organisation (IO), is cooled and heated by pressurized water which flows through a series of gun-drilled water channels and water pipes. The cooling function is required to remove nuclear heating due to radiation during operation of ITER, while the heating function is intended to heat up uniformly the machine during baking condition. The work presented provides coupled thermo-hydraulic analysis and optimization of a Generic Equatorial Port Plug (GEPP) structure cooling and heating system. The optimization performed includes positioning, minimization of number and size of gun drilled channels, complying with the flow and functional requirements during operating and baking conditions.

  13. Shock Mechanism Analysis and Simulation of High-Power Hydraulic Shock Wave Simulator

    Directory of Open Access Journals (Sweden)

    Xiaoqiu Xu

    2017-01-01

    Full Text Available The simulation of regular shock wave (e.g., half-sine can be achieved by the traditional rubber shock simulator, but the practical high-power shock wave characterized by steep prepeak and gentle postpeak is hard to be realized by the same. To tackle this disadvantage, a novel high-power hydraulic shock wave simulator based on the live firing muzzle shock principle was proposed in the current work. The influence of the typical shock characteristic parameters on the shock force wave was investigated via both theoretical deduction and software simulation. According to the obtained data compared with the results, in fact, it can be concluded that the developed hydraulic shock wave simulator can be applied to simulate the real condition of the shocking system. Further, the similarity evaluation of shock wave simulation was achieved based on the curvature distance, and the results stated that the simulation method was reasonable and the structural optimization based on software simulation is also beneficial to the increase of efficiency. Finally, the combination of theoretical analysis and simulation for the development of artillery recoil tester is a comprehensive approach in the design and structure optimization of the recoil system.

  14. Coupled neutronic/thermal-hydraulic analysis of the HPLWR three pass core

    International Nuclear Information System (INIS)

    Monti, Lanfranco; Starflinger, Joerg; Schulenberg, Thomas

    2008-01-01

    The High Performance Light Water Reactor is an innovative Gen-IV reactor cooled and moderated with water at supercritical pressure. The three pass core concept has been proposed to reduce peaking factors, i.e. hot-channel effects, and it further increases the core heterogeneity, which is mainly due to pronounced water density reduction. For this kind of nuclear reactor, the significant feedbacks - which exist between the properties of the components and the power generation rate - can not be neglected and require a coupled Neutronic/Thermal-Hydraulic analysis even for steady state conditions. The main goal of this paper is to present the developed tool for coupled analyses of the HPLWR. Two state-of-the-art codes have been chosen for Thermal-Hydraulic and Neutronic core analyses, namely TRACE and ERANOS, and they have been coupled with in an iterative procedure in which they are run in series until a steady state condition has been reached. In the simplifying assumptions of uniform enrichment distribution, zero burn-up and ignoring the effect of the control rods, the obtained steady state condition will be discussed and a core power map, flow rate redistribution as well as water and fuel temperature variations will be presented. (author)

  15. Fundamental approaches for analysis thermal hydraulic parameter for Puspati Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hashim, Zaredah, E-mail: zaredah@nm.gov.my; Lanyau, Tonny Anak, E-mail: tonny@nm.gov.my; Farid, Mohamad Fairus Abdul; Kassim, Mohammad Suhaimi [Reactor Technology Centre, Technical Support Division, Malaysia Nuclear Agency, Ministry of Science, Technology and Innovation, Bangi, 43000, Kajang, Selangor Darul Ehsan (Malaysia); Azhar, Noraishah Syahirah [Universiti Teknologi Malaysia, 80350, Johor Bahru, Johor Darul Takzim (Malaysia)

    2016-01-22

    The 1-MW PUSPATI Research Reactor (RTP) is the one and only nuclear pool type research reactor developed by General Atomic (GA) in Malaysia. It was installed at Malaysian Nuclear Agency and has reached the first criticality on 8 June 1982. Based on the initial core which comprised of 80 standard TRIGA fuel elements, the very fundamental thermal hydraulic model was investigated during steady state operation using the PARET-code. The main objective of this paper is to determine the variation of temperature profiles and Departure of Nucleate Boiling Ratio (DNBR) of RTP at full power operation. The second objective is to confirm that the values obtained from PARET-code are in agreement with Safety Analysis Report (SAR) for RTP. The code was employed for the hot and average channels in the core in order to calculate of fuel’s center and surface, cladding, coolant temperatures as well as DNBR’s values. In this study, it was found that the results obtained from the PARET-code showed that the thermal hydraulic parameters related to safety for initial core which was cooled by natural convection was in agreement with the designed values and safety limit in SAR.

  16. Transient thermal-hydraulic characteristics analysis software for PWR nuclear power systems

    International Nuclear Information System (INIS)

    Wu Yingwei; Zhuang Chengjun; Su Guanghui; Qiu Suizheng

    2010-01-01

    A point reactor neutron kinetics model, a two-phase drift-flow U-tube steam generator model, an advanced non-equilibrium three regions pressurizer model, and a passive emergency core decay heat-removed system model are adopted in the paper to develop the computerized analysis code for PWR transient thermal-hydraulic characteristics, by Compaq Visual Fortran 6.0 language. Visual input, real-time processing and dynamic visualization output are achieved by Microsoft Visual Studio. NET language. The reliability verification of the soft has been conducted by RELAP 5, and the verification results show that the software is with high calculation precision, high calculation speed, modern interface, luxuriant functions and strong operability. The software was applied to calculate the transient accident conditions for QSNP, and the analysis results are significant to the practical engineering applications. (authors)

  17. Development of a steady thermal-hydraulic analysis code for the China Advanced Research Reactor

    Institute of Scientific and Technical Information of China (English)

    TIAN Wenxi; QIU Suizheng; GUO Yun; SU Guanghui; JIA Dounan; LIU Tiancai; ZHANG Jianwei

    2007-01-01

    A multi-channel model steady-state thermalhydraulic analysis code was developed for the China Advanced Research Reactor (CARR). By simulating the whole reactor core, the detailed mass flow distribution in the core was obtained. The result shows that structure size plays the most important role in mass flow distribution, and the influence of core power could be neglected under singlephase flow. The temperature field of the fuel element under unsymmetrical cooling condition was also obtained, which is necessary for further study such as stress analysis, etc. Of the fuel element. At the same time, considering the hot channel effect including engineering factor and nuclear factor, calculation of the mean and hot channel was carried out and it is proved that all thermal-hydraulic parameters satisfy the "Safety design regulation of CARR".

  18. Analysis of the thermal hydraulics and core degradation behavior in the PHEBUS-FPT1 test train with impact/SAMPSON code

    International Nuclear Information System (INIS)

    Terada, Masafumi; Ikeda, Takashi; Nakahara, Katsuhiko; Shirakawa, Noriyuki; Horie, Hideki; Katsuragi, Kazuyuki; Yamagishi, Makoto; Ito, Takahiro

    2003-01-01

    As one of the verification studies of SAMPSON code, PHEBUS-FPT1, which is authorized as the International Standard Problem-46, was analyzed about the in-core phenomena with four modules, the molten core relocation analysis (MCRA) module, the fuel rod heat up analysis (FRHA) module, the fission product release analysis (FPRA) module, and the analysis control module (ACM) of SAMPSON. This paper describes the analysis of thermal hydraulics and core degradation behavior in the test train. Two-dimensional version of MCRA models the whole structure of the test train in the cylindrical system, including the fuel bundle and the shroud. FRHA models eighteen irradiated fuel rods, two fresh fuel rods, and one control rod in the center of the bundle. FRHA evaluates the transient behavior of fuel rods and releases failed fuel components to MCRA. MCRA evaluates the fluid dynamics of steam and debris considering the thermal and fluid mechanical interaction between them, and at the same time the thermal interaction between gas/debris and shroud material. By the phase change model of MCRA, molten debris forms debris pool and a part of them possibly freezes on fuel rods or shroud surface, then forms crust. This combination of modules of SAMPSON was proved to be capable for modeling the PHEBUS-FPT1 in-core phenomena sufficiently. The analysis has shown sufficient agreement with test results regarding to steam flow rates at the outlet, reproducing its reduction due to hydrogen generation, steam and shroud temperature, and debris relocation behavior. (author)

  19. Steady-state thermal hydraulic analysis and flow channel blockage accident analysis of JRR-3 silicide core

    International Nuclear Information System (INIS)

    Kaminaga, Masanori

    1997-03-01

    JRR-3 is a light water moderated and cooled, beryllium and heavy water reflected pool type research reactor using low enriched uranium (LEU) plate-type fuels. Its thermal power is 20 MW. The core conversion program from uranium-aluminum (UAl x -Al) dispersion type fuel (aluminide fuel) to uranium-silicon-aluminum (U 3 Si 2 -Al) dispersion type fuel (silicide fuel) is currently conducted at the JRR-3. This report describes about the steady-state thermal hydraulic analysis results and the flow channel blockage accident analysis result. In JRR-3, there are two operation mode. One is high power operation mode up to 20 MW, under forced convection cooling using the primary and the secondary cooling systems. The other is low power operation mode up to 200 kW, under natural circulation cooling between the reactor core and the reactor pool without the primary and the secondary cooling systems. For the analysis of the flow channel blockage accident, COOLOD code was used. On the other hand, steady-state thermal hydraulic analysis for both of the high power operation mode under forced convection cooling and low power operation under natural convection cooling, COOLOD-N2 code was used. From steady-state thermal hydraulic analysis results of both forced and natural convection cooling, fuel temperature, minimum DNBR etc. meet the design criteria and JRR-3 LEU silicide core has enough safety margin under normal operation conditions. Furthermore, flow channel blockage accident analysis results show that one channel flow blockage accident meet the safety criteria for accident conditions which have been established for JRR-3 LEU silicide core. (author)

  20. Evaluation of the flow forces on an open centre directional control valve by means of a computational fluid dynamic analysis

    International Nuclear Information System (INIS)

    Amirante, R.; Del Vescovo, G.; Lippolis, A.

    2006-01-01

    The aim of the present paper is the evaluation of the driving forces acting on a 4/3 hydraulic open center directional control valve spool by means of a complete numerical analysis. In a previous paper by the same authors, the valve was inserted in a closed hydraulic circuit and was tested with different pump flow rate values to obtain experimental results about the driving forces. The experimental results are used in this paper to evaluate and validate the numerical analysis of the valve. The obtained numerical results show important differences between an open center valve and a closed center one, the latter being extensively analyzed in the literature. The numerical analysis is performed by using the commercial code 'Fluent', and the numerical results show the complete flow field inside the valve. The aim of this analysis is to evaluate the valve fluid dynamic performance, exploiting computational fluid dynamics (CFD) techniques, in order to give the reliable indications needed to define the valve design criteria and avoid expensive experimental tests

  1. 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.

  2. Mathematical theory of compressible viscous fluids analysis and numerics

    CERN Document Server

    Feireisl, Eduard; Pokorný, Milan

    2016-01-01

    This book offers an essential introduction to the mathematical theory of compressible viscous fluids. The main goal is to present analytical methods from the perspective of their numerical applications. Accordingly, we introduce the principal theoretical tools needed to handle well-posedness of the underlying Navier-Stokes system, study the problems of sequential stability, and, lastly, construct solutions by means of an implicit numerical scheme. Offering a unique contribution – by exploring in detail the “synergy” of analytical and numerical methods – the book offers a valuable resource for graduate students in mathematics and researchers working in mathematical fluid mechanics. Mathematical fluid mechanics concerns problems that are closely connected to real-world applications and is also an important part of the theory of partial differential equations and numerical analysis in general. This book highlights the fact that numerical and mathematical analysis are not two separate fields of mathematic...

  3. The role of thermal-hydraulic computation system in LTMP for simulation in order to support the design and analysis

    International Nuclear Information System (INIS)

    Bambang Teguh, P.; Turyana, I.

    1997-01-01

    In order to support the activities of LTMP and other Indonesia research institutions in the field of thermal-hydraulic, LTMP is equipped with several software, one of which is thermalhydraulic code TRIO-VF developed by CEA (commissariat a Energie Atomique), France. TRIO-VF is a computer code to solve general equations of thermal-hydraulic in 3D. The code can be used for numerical simulation of laminar or turbulent flow, with or without the presence of heat or mass transfer. these simulations or predictions are important step in the conception of thermalhydraulic equipment (vessels, heat and components of nuclear reactors). The fluid flow can be in the domain where internal obstacles (plate, tube bundel...etc.) are present

  4. Optimised design and thermal-hydraulic analysis of the IFMIF/HFTM test section

    Energy Technology Data Exchange (ETDEWEB)

    Gordeev, S.; Heinzel, V.; Lang, K.H.; Moeslang, A.; Schleisiek, K.; Slobodtchouk, V.; Stratmanns, E.

    2003-10-01

    On the basis of previous concepts, analyses and experiments, the high flux test module (HFTM) for the International Fusion Materials Irradiation Facility (IFMIF) was further optimised. The work focused on the design and the thermal hydraulic analysis of the HFTM section containing the material specimens to be irradiated, the ''test section'', with the main objective to improve the concept with respect to the optimum use of the available irradiation volume and to the temperature of the specimens. Particular emphasis was laid on the application of design principles which assure stable and reproducible thermal conditions. The present work has confirmed the feasibility and suitability of the optimised design of the HFTM test section with chocolate plate like shaped rigs. In particular it has been shown that the envisaged irradiation temperatures can be reached with acceptable temperature differences inside the specimen stack. The latter can be achieved only by additional electrical heating of the axial ends of the capsules. Division of the heater in three sections with separate power supply and control units is necessary. Maintaining of the temperatures during beam-off periods likewise requires electrical heating. The required electrical heaters - mineral isolated wires - are commercially available. The potential of the CFD code STAR-CD for the thermal hydraulic analysis of complex systems like the HFTM was confirmed. Nevertheless, experimental confirmation is desirable. Suitable experiments are under preparation. To verify the assumptions made on the thermal conductivity of the contact faces and layers between the two shells of the rig, dedicated experiments are suggested. The present work must be complemented by a thermal mechanical analysis of the module. Most critical component in this respect seems to be the rig wall. Furthermore, it will be necessary to investigate the response of the HFTM to power transients, and to determine the requirements

  5. Preliminary analysis of K-DEMO thermal hydraulic system using MELCOR; Parametric study of hydrogen explosion

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

    K-DEMO (Korean fusion demonstration reactor) is future reactor for the commercializing the fusion power generation. The Design of K-DEMO is similar to that of ITER but the fusion energy generation is much bigger because ITER is experimental reactor. For this reason, K-DEMO uses more fusion reaction with bigger amount of tritium. Higher fusion power means more neutron generation that can irradiate the structure around fusion plasma. Fusion reactor can produce many kinds of radioactive material in the accident. Because of this hazard, preliminary safety analysis is mandatory before its construction. Concern for safety problem of accident of fusion/fission reactor has been growing after Fukushima accident which is severe accident from unexpected disaster. To model the primary heat transfer system, in this study, MARS-KS thermal hydraulic analysis is referred. Lee et al. and Kim et al. conducted thermal hydraulic analysis using MARS-KS and multiple module simulation to deal with the phenomena of first wall corrosion for each plasma pulse. This study shows the relationship between vacuum vessel rupture area and source term leakage after hydrogen explosion. For the conservative study, first wall heating is not terminated because the heating inside the vacuum vessel increase the pressure inside VV. Pressurizer, steam generator and turbine is not damaged. 6.69 kg of tritiated water (HTO) and 1 ton of dust is modeled which is ITER guideline. The entire system of K-DEMO is smaller than that of ITER. For this reason, lots of aerosol is release into environment although the safety system like DS is maintained. This result shows that the safety system of K-DEMO should use much more safety system.

  6. Optimised design and thermal-hydraulic analysis of the IFMIF/HFTM test section

    International Nuclear Information System (INIS)

    Gordeev, S.; Heinzel, V.; Lang, K.H.; Moeslang, A.; Schleisiek, K.; Slobodtchouk, V.; Stratmanns, E.

    2003-10-01

    On the basis of previous concepts, analyses and experiments, the high flux test module (HFTM) for the International Fusion Materials Irradiation Facility (IFMIF) was further optimised. The work focused on the design and the thermal hydraulic analysis of the HFTM section containing the material specimens to be irradiated, the ''test section'', with the main objective to improve the concept with respect to the optimum use of the available irradiation volume and to the temperature of the specimens. Particular emphasis was laid on the application of design principles which assure stable and reproducible thermal conditions. The present work has confirmed the feasibility and suitability of the optimised design of the HFTM test section with chocolate plate like shaped rigs. In particular it has been shown that the envisaged irradiation temperatures can be reached with acceptable temperature differences inside the specimen stack. The latter can be achieved only by additional electrical heating of the axial ends of the capsules. Division of the heater in three sections with separate power supply and control units is necessary. Maintaining of the temperatures during beam-off periods likewise requires electrical heating. The required electrical heaters - mineral isolated wires - are commercially available. The potential of the CFD code STAR-CD for the thermal hydraulic analysis of complex systems like the HFTM was confirmed. Nevertheless, experimental confirmation is desirable. Suitable experiments are under preparation. To verify the assumptions made on the thermal conductivity of the contact faces and layers between the two shells of the rig, dedicated experiments are suggested. The present work must be complemented by a thermal mechanical analysis of the module. Most critical component in this respect seems to be the rig wall. Furthermore, it will be necessary to investigate the response of the HFTM to power transients, and to determine the requirements on the electrical

  7. A practical sensitivity analysis method for ranking sources of uncertainty in thermal–hydraulics applications

    Energy Technology Data Exchange (ETDEWEB)

    Pourgol-Mohammad, Mohammad, E-mail: pourgolmohammad@sut.ac.ir [Department of Mechanical Engineering, Sahand University of Technology, Tabriz (Iran, Islamic Republic of); Hoseyni, Seyed Mohsen [Department of Basic Sciences, East Tehran Branch, Islamic Azad University, Tehran (Iran, Islamic Republic of); Hoseyni, Seyed Mojtaba [Building & Housing Research Center, Tehran (Iran, Islamic Republic of); Sepanloo, Kamran [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of)

    2016-08-15

    Highlights: • Existing uncertainty ranking methods prove inconsistent for TH applications. • Introduction of a new method for ranking sources of uncertainty in TH codes. • Modified PIRT qualitatively identifies and ranks uncertainty sources more precisely. • The importance of parameters is calculated by a limited number of TH code executions. • Methodology is applied successfully on LOFT-LB1 test facility. - Abstract: In application to thermal–hydraulic calculations by system codes, sensitivity analysis plays an important role for managing the uncertainties of code output and risk analysis. Sensitivity analysis is also used to confirm the results of qualitative Phenomena Identification and Ranking Table (PIRT). Several methodologies have been developed to address uncertainty importance assessment. Generally, uncertainty importance measures, mainly devised for the Probabilistic Risk Assessment (PRA) applications, are not affordable for computationally demanding calculations of the complex thermal–hydraulics (TH) system codes. In other words, for effective quantification of the degree of the contribution of each phenomenon to the total uncertainty of the output, a practical approach is needed by considering high computational burden of TH calculations. This study aims primarily to show the inefficiency of the existing approaches and then introduces a solution to cope with the challenges in this area by modification of variance-based uncertainty importance method. Important parameters are identified by the modified PIRT approach qualitatively then their uncertainty importance is quantified by a local derivative index. The proposed index is attractive from its practicality point of view on TH applications. It is capable of calculating the importance of parameters by a limited number of TH code executions. Application of the proposed methodology is demonstrated on LOFT-LB1 test facility.

  8. A practical sensitivity analysis method for ranking sources of uncertainty in thermal–hydraulics applications

    International Nuclear Information System (INIS)

    Pourgol-Mohammad, Mohammad; Hoseyni, Seyed Mohsen; Hoseyni, Seyed Mojtaba; Sepanloo, Kamran

    2016-01-01

    Highlights: • Existing uncertainty ranking methods prove inconsistent for TH applications. • Introduction of a new method for ranking sources of uncertainty in TH codes. • Modified PIRT qualitatively identifies and ranks uncertainty sources more precisely. • The importance of parameters is calculated by a limited number of TH code executions. • Methodology is applied successfully on LOFT-LB1 test facility. - Abstract: In application to thermal–hydraulic calculations by system codes, sensitivity analysis plays an important role for managing the uncertainties of code output and risk analysis. Sensitivity analysis is also used to confirm the results of qualitative Phenomena Identification and Ranking Table (PIRT). Several methodologies have been developed to address uncertainty importance assessment. Generally, uncertainty importance measures, mainly devised for the Probabilistic Risk Assessment (PRA) applications, are not affordable for computationally demanding calculations of the complex thermal–hydraulics (TH) system codes. In other words, for effective quantification of the degree of the contribution of each phenomenon to the total uncertainty of the output, a practical approach is needed by considering high computational burden of TH calculations. This study aims primarily to show the inefficiency of the existing approaches and then introduces a solution to cope with the challenges in this area by modification of variance-based uncertainty importance method. Important parameters are identified by the modified PIRT approach qualitatively then their uncertainty importance is quantified by a local derivative index. The proposed index is attractive from its practicality point of view on TH applications. It is capable of calculating the importance of parameters by a limited number of TH code executions. Application of the proposed methodology is demonstrated on LOFT-LB1 test facility.

  9. Dynamic analysis of multibody system immersed in a fluid medium

    International Nuclear Information System (INIS)

    Wu, R.W.; Liu, L.K.; Levy, S.

    1977-01-01

    This paper is concerned primarily with the development and evaluation of an analysis method for the reponse prediction of immersed systems to seismic and other dynamic excitations. For immersed multibody systems, the hydrodynamic interaction causes coupled motion among the solid bodies. Also, under intense external excitations, impact between bodies may occur. The complex character of such systems inhibit the use of conventional analytical solutions in closed form. Therefore, approximate numerical schemes have been devised. For an incompressible, inviscid fluid, the hydrodynamic forces exerted by the fluid on solid bodies are determined to be linearly proportional to the acceleration of the vibrating solid bodies; i.e., the presence of the fluid only affects the inertia of the solid body system. A finite element computer program has been developed for computing this hydrodynamic (or added) mass effect. This program can be used to determine the hydrodynamic mass of a two-dimensional fluid field with solid bodies of arbitrary geometry. Triangular elements and linear pressure interpolation function are used to discretize the fluid region. The component element method is used to determine the dynamic response of the multibody system to externally applied mechanical loading or support excitation. The present analysis method for predicting the dynamic response of submerged multibody system is quite general and pertains to any number of solid bodies. However in this paper, its application is demonstrated only for 4 and 25 body systems. (Auth.)

  10. Fractal analysis of the hydraulic conductivity on a sandy porous media reproduced in a laboratory facility.

    Science.gov (United States)

    de Bartolo, S.; Fallico, C.; Straface, S.; Troisi, S.; Veltri, M.

    2009-04-01

    The complexity characterization of the porous media structure, in terms of the "pore" phase and the "solid" phase, can be carried out by means of the fractal geometry which is able to put in relationship the soil structural properties and the water content. It is particularly complicated to describe analytically the hydraulic conductivity for the irregularity of the porous media structure. However these can be described by many fractal models considering the soil structure as the distribution of particles dimensions, the distribution of the solid aggregates, the surface of the pore-solid interface and the fractal mass of the "pore" and "solid" phases. In this paper the fractal model of Yu and Cheng (2002) and Yu and Liu (2004), for a saturated bidispersed porous media, was considered. This model, using the Sierpinsky-type gasket scheme, doesn't contain empiric constants and furnishes a well accord with the experimental data. For this study an unconfined aquifer was reproduced by means of a tank with a volume of 10 Ã- 7 Ã- 3 m3, filled with a homogeneous sand (95% of SiO2), with a high percentage (86.4%) of grains between 0.063mm and 0.125mm and a medium-high permeability. From the hydraulic point of view, 17 boreholes, a pumping well and a drainage ring around its edge were placed. The permeability was measured utilizing three different methods, consisting respectively in pumping test, slug test and laboratory analysis of an undisturbed soil cores, each of that involving in the measurement a different support volume. The temporal series of the drawdown obtained by the pumping test were analyzed by the Neuman-type Curve method (1972), because the saturated part above the bottom of the facility represents an unconfined aquifer. The data analysis of the slug test were performed by the Bouwer & Rice (1976) method and the laboratory analysis were performed on undisturbed saturated soil samples utilizing a falling head permeameter. The obtained values either of the

  11. 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.

  12. Seminal Fluid Analysis And Biophysical Profile: Findings And ...

    African Journals Online (AJOL)

    Seminal Fluid Analysis And Biophysical Profile: Findings And Relevance In Infertile Males In Ilorin, Nigeria. EK Oghagbon, AAG Jimoh, SA Adebisi. Abstract. To determine if there was a bearing of body mass index (BMI) on male infertility, a cross-sectional study of males of infertile couples, attending our infertility clinic was ...

  13. HAMOC: a computer program for fluid hammer analysis

    International Nuclear Information System (INIS)

    Johnson, H.G.

    1975-12-01

    A computer program has been developed for fluid hammer analysis of piping systems attached to a vessel which has undergone a known rapid pressure transient. The program is based on the characteristics method for solution of the partial differential equations of motion and continuity. Column separation logic is included for situations in which pressures fall to saturation values

  14. Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1

    Directory of Open Access Journals (Sweden)

    Muhammad Atta

    2011-01-01

    Full Text Available The burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1, reference operating core, has been carried out utilizing standard computer codes WIMS/D4, CITATION, and RELAP5/MOD3.4. Reactor codes WIMS/D4 and CITATION have been used for the calculations of neutronic parameters including peaking factors and power profiles at different burn-up considering a xenon free core and also the equilibrium xenon values. RELAP5/MOD3.4 code was utilized for the determination of peak fuel centerline, clad and coolant temperatures to ensure the safety of the reactor throughout the cycle. The calculations reveal that the reactor is safe and no nucleate boiling will commence at any part of the core throughout the cycle and that the safety margin increases with burnup as peaking factors decrease.

  15. Hydraulic and structural co-simulation analysis of turbine runner during operation

    International Nuclear Information System (INIS)

    Markov, Zoran; Popovski, Predrag; Lipej, Andrej; Djelic, Vesko

    2006-01-01

    Modern concept of HPP refurbishment procedure consists of many aspects of the turbine re-design. One of the most useful data is the previous operational data during the lifetime of the unit. In many cases, high stressed areas are damaged. Lack of the measurements makes the solution of the problems and verification of the numerical results very difficult. This work represents an integrated approach in solving hydraulic and structural problems in design stage or optimization of an aial hydro turbine. CFD approach is implemented in solving the flow through a complete aial turbine, taking into account all the necessary factors influencing the real flow. Frozen rotor condition is taken as an input in the computations. The results from the CFD calculations are used as an input for the performed FEA modeling and structural analysis.

  16. Energy-saving analysis of hydraulic hybrid excavator based on common pressure rail.

    Science.gov (United States)

    Shen, Wei; Jiang, Jihai; Su, Xiaoyu; Karimi, Hamid Reza

    2013-01-01

    Energy-saving research of excavators is becoming one hot topic due to the increasing energy crisis and environmental deterioration recently. Hydraulic hybrid excavator based on common pressure rail (HHEC) provides an alternative with electric hybrid excavator because it has high power density and environment friendly and easy to modify based on the existing manufacture process. This paper is focused on the fuel consumption of HHEC and the actuator dynamic response to assure that the new system can save energy without sacrificing performance. Firstly, we introduce the basic principle of HHEC; then, the sizing process is presented; furthermore, the modeling period which combined mathematical analysis and experiment identification is listed. Finally, simulation results show that HHEC has a fast dynamic response which can be accepted in engineering and the fuel consumption can be reduced 21% to compare the original LS excavator and even 32% after adopting another smaller engine.

  17. Performance Degradation Analysis of Aviation Hydraulic Piston Pump Based on Mixed Wear Theory

    Directory of Open Access Journals (Sweden)

    C. Zhang

    2017-06-01

    Full Text Available This paper focuses on the mathematical modeling of axial piston pump through dividing the failure development of friction pair into lubrication, mixed lubrication and abrasion. Directing to the wedge-shaped oil film between cylinder block and valve plate, the support force distribution under the temperature variance was obtained. Considering the rough peak of valve plate, the contact load model is built under plastic deformation and elastic deformation and the corresponding wear volume is calculated. Computing the wear and tear along the counter-clockwise, the total amount of friction and wear can be calculated. Simulation and preliminary wear particle monitoring test indicates that proposed modeling and analysis can effectively reflect the real abrasion process of hydraulic piston pump.

  18. Nonlinear Dynamical Analysis of Hydraulic Turbine Governing Systems with Nonelastic Water Hammer Effect

    Directory of Open Access Journals (Sweden)

    Junyi Li

    2014-01-01

    Full Text Available A nonlinear mathematical model for hydroturbine governing system (HTGS has been proposed. All essential components of HTGS, that is, conduit system, turbine, generator, and hydraulic servo system, are considered in the model. Using the proposed model, the existence and stability of Hopf bifurcation of an example HTGS are investigated. In addition, chaotic characteristics of the system with different system parameters are studied extensively and presented in the form of bifurcation diagrams, time waveforms, phase space trajectories, Lyapunov exponent, chaotic attractors, and Poincare maps. Good correlation can be found between the model predictions and theoretical analysis. The simulation results provide a reasonable explanation for the sustained oscillation phenomenon commonly seen in operation of hydroelectric generating set.

  19. High heat flux thermal-hydraulic analysis of ITER divertor and blanket systems

    International Nuclear Information System (INIS)

    Raffray, A.R.; Chiocchio, S.; Ioki, K.; Tivey, R.; Krassovski, D.; Kubik, D.

    1998-01-01

    Three separate cooling systems are used for the divertor and blanket components, based mainly on flow routing access and on grouping together components with the highest heat load levels and uncertainties: divertor, limiter/outboard baffle, and primary first wall/inboard baffle. The coolant parameters for these systems are set to accommodate peak heat load conditions with a reasonable critical heat flux (CHF) margin. Material temperature constraints and heat transport system space and cost requirements are also taken into consideration. This paper summarises the three cooling system designs and highlights the high heat flux thermal-hydraulic analysis carried out in converging on the design values for the coolant operating parameters. Application of results from on-going high heat flux R and D and a brief description of future R and D effort to address remaining issues are also included. (orig.)

  20. FLICA III. A digital computer program for thermal-hydraulic analysis of reactors and experimental loops

    International Nuclear Information System (INIS)

    Plas, Roger.

    1975-05-01

    This computer program describes the flow and heat transfer in steady and transient state in two-phase flows. It is the present stage of the evolution about FLICA, FLICA II and FLICA II B codes which have been used and developed at CEA for the thermal-hydraulic analysis of reactors and experimental loops with heating rod bundles. In the mathematical model all the significant terms of the fundamental hydrodynamic equations are taken into account with the approximations of turbulent viscosity and conductivity. The two-phase flow is calculated by the homogeneous model with slip. In the flow direction an implicit resolution scheme is available, which make possible to study partial or total flow blockage, with upstream and downstream effects. A special model represents the helical wire effects in out-of pile experimental rod bundles [fr

  1. 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)

  2. Development of system analysis code for thermal-hydraulic simulation of integral reactor, Rex-10

    International Nuclear Information System (INIS)

    Lee, Y. G.; Kim, J. W.; Yoon, S. J.; Park, G. C.

    2010-10-01

    Rex-10 is an environment-friendly and economical small-scale nuclear reactor to provide the energy for district heating as well as the electric power in micro-grid. This integral reactor comprises several innovative concepts supported by advanced primary circuit components, low coolant parameters and natural circulation cooling. To evaluate the system performance and thermal-hydraulic behavior of the reactor, a system analysis code is being developed so that the new designs and technologies adopted in Rex-10 can be reflected. The research efforts are absorbed in programming the simple and fast-running thermal-hydraulic analysis software. The details of hydrodynamic governing equations component models and numerical solution scheme used in this code are presented in this paper. On the basis of one-dimensional momentum integral model, the models of point reactor neutron kinetics for thorium-fueled core, physical processes in the steam-gas pressurizer, and heat transfers in helically coiled steam generator are implemented to the system code. Implicit numerical scheme is employed to momentum and energy equations to assure the numerical stability. The accuracy of simulation is validated by applying the solution method to the Rex-10 test facility. Calculated natural circulation flow rate and coolant temperature at steady-state are compared to the experimental data. The validation is also carried out for the transients in which the sudden reduction in the core power or the feedwater flow takes place. The code's capability to predict the steady-state flow by natural convection and the qualitative behaviour of the primary system in the transients is confirmed. (Author)

  3. Development of system analysis code for thermal-hydraulic simulation of integral reactor, Rex-10

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-10-15

    Rex-10 is an environment-friendly and economical small-scale nuclear reactor to provide the energy for district heating as well as the electric power in micro-grid. This integral reactor comprises several innovative concepts supported by advanced primary circuit components, low coolant parameters and natural circulation cooling. To evaluate the system performance and thermal-hydraulic behavior of the reactor, a system analysis code is being developed so that the new designs and technologies adopted in Rex-10 can be reflected. The research efforts are absorbed in programming the simple and fast-running thermal-hydraulic analysis software. The details of hydrodynamic governing equations component models and numerical solution scheme used in this code are presented in this paper. On the basis of one-dimensional momentum integral model, the models of point reactor neutron kinetics for thorium-fueled core, physical processes in the steam-gas pressurizer, and heat transfers in helically coiled steam generator are implemented to the system code. Implicit numerical scheme is employed to momentum and energy equations to assure the numerical stability. The accuracy of simulation is validated by applying the solution method to the Rex-10 test facility. Calculated natural circulation flow rate and coolant temperature at steady-state are compared to the experimental data. The validation is also carried out for the transients in which the sudden reduction in the core power or the feedwater flow takes place. The code's capability to predict the steady-state flow by natural convection and the qualitative behaviour of the primary system in the transients is confirmed. (Author)

  4. Application of symplectic integrator to numerical fluid analysis

    International Nuclear Information System (INIS)

    Tanaka, Nobuatsu

    2000-01-01

    This paper focuses on application of the symplectic integrator to numerical fluid analysis. For the purpose, we introduce Hamiltonian particle dynamics to simulate fluid behavior. The method is based on both the Hamiltonian formulation of a system and the particle methods, and is therefore called Hamiltonian Particle Dynamics (HPD). In this paper, an example of HPD applications, namely the behavior of incompressible inviscid fluid, is solved. In order to improve accuracy of HPD with respect to space, CIVA, which is a highly accurate interpolation method, is combined, but the combined method is subject to problems in that the invariants of the system are not conserved in a long-time computation. For solving the problems, symplectic time integrators are introduced and the effectiveness is confirmed by numerical analyses. (author)

  5. Supercritical fluid chromatography for lipid analysis in foodstuffs.

    Science.gov (United States)

    Donato, Paola; Inferrera, Veronica; Sciarrone, Danilo; Mondello, Luigi

    2017-01-01

    The task of lipid analysis has always challenged separation scientists, and new techniques in chromatography were often developed for the separation of lipids; however, no single technique or methodology is yet capable of affording a comprehensive screening of all lipid species and classes. This review acquaints the role of supercritical fluid chromatography within the field of lipid analysis, from the early developed capillary separations based on pure CO 2 , to the most recent techniques employing packed columns under subcritical conditions, including the niche multidimensional techniques using supercritical fluids in at least one of the separation dimensions. A short history of supercritical fluid chromatography will be introduced first, from its early popularity in the late 1980s, to the sudden fall and oblivion until the last decade, experiencing a regain of interest within the chromatographic community. Afterwards, the subject of lipid nomenclature and classification will be briefly dealt with, before discussing the main applications of supercritical fluid chromatography for food analysis, according to the specific class of lipids. © 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  6. Effective hydraulic resistance of a nozzle generating hybrid-synthetic jet in an electrodynamic actuator - Part II: Analysis and correlations

    Czech Academy of Sciences Publication Activity Database

    Tesař, Václav; Kordík, Jozef

    Roč. 199, September (2013), s. 391-400 ISSN 0924-4247 R&D Projects: GA TA ČR TA02020795; GA ČR(CZ) GCP101/11/J019; GA ČR GPP101/12/P556 Institutional research plan: CEZ:AV0Z20760514 Institutional support: RVO:61388998 Keywords : nozzle * periodic jet * hydraulic resistance Subject RIV: BK - Fluid Dynamics Impact factor: 1.943, year: 2013 http://dx.doi.org/10.1016/j.sna.2013.01.003

  7. Vibration analysis of partially cracked plate submerged in fluid

    Science.gov (United States)

    Soni, Shashank; Jain, N. K.; Joshi, P. V.

    2018-01-01

    The present work proposes an analytical model for vibration analysis of partially cracked rectangular plates coupled with fluid medium. The governing equation of motion for the isotropic plate based on the classical plate theory is modified to accommodate a part through continuous line crack according to simplified line spring model. The influence of surrounding fluid medium is incorporated in the governing equation in the form of inertia effects based on velocity potential function and Bernoulli's equations. Both partially and totally submerged plate configurations are considered. The governing equation also considers the in-plane stretching due to lateral deflection in the form of in-plane forces which introduces geometric non-linearity into the system. The fundamental frequencies are evaluated by expressing the lateral deflection in terms of modal functions. The assessment of the present results is carried out for intact submerged plate as to the best of the author's knowledge the literature lacks in analytical results for submerged cracked plates. New results for fundamental frequencies are presented as affected by crack length, fluid level, fluid density and immersed depth of plate. By employing the method of multiple scales, the frequency response and peak amplitude of the cracked structure is analyzed. The non-linear frequency response curves show the phenomenon of bending hardening or softening and the effect of fluid dynamic pressure on the response of the cracked plate.

  8. 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

  9. Amino acid analysis in biological fluids by GC-MS

    OpenAIRE

    Kaspar, Hannelore

    2009-01-01

    Amino acids are intermediates in cellular metabolism and their quantitative analysis plays an important role in disease diagnostics. A gas chromatography-mass spectrometry (GC-MS) based method was developed for the quantitative analysis of free amino acids as their propyl chloroformate derivatives in biological fluids. Derivatization with propyl chloroformate could be carried out directly in the biological samples without prior protein precipitation or solid-phase extraction of the amino acid...

  10. 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...

  11. A review of the current thermal-hydraulic modeling of the Jules Horowitz Reactor: A loss of flow accident analysis

    International Nuclear Information System (INIS)

    Pegonen, R.; Bourdon, S.; Gonnier, C.; Anglart, H.

    2014-01-01

    Highlights: • CEA methodology for thermal-hydraulic calculations in the JHR reactor is described. • Thermal-hydraulics of the JHR is analyzed during LOFA using CATHARE and FLICA4. • Safety criteria, important modeling parameters and correlations are presented. • Possible improvements of the current methodology are discussed and proposed. - Abstract: The newest European high performance material testing reactor, the Jules Horowitz Reactor, will support existing and future nuclear reactor designs. The reactor is under construction at CEA Cadarache research center in France and is expected to start operation at the end of this decade. R and D and analytical works have already been performed to set-up the methodology for thermal-hydraulic calculations of the reactor. This paper presents the off-line coupled thermal-hydraulic modeling of the reactor using the CATHARE system code and the FLICA4 core analysis code. The main objective of the present work is to analyze the thermal-hydraulic calculations of the reactor during the loss of flow accident using CEA methodology. Possible improvements of the current methodology are shortly discussed and suggested

  12. Mathematical simulation of fluid flow and analysis of flow pattern in the flow path of low-head Kaplan turbine

    Directory of Open Access Journals (Sweden)

    A. V. Rusanov

    2016-12-01

    Full Text Available The results of numerical investigation of spatial flow of viscous incompressible fluid in flow part of Kaplan turbine PL20 Kremenchug HPP at optimum setting angle of runner blade φb = 15° and at maximum setting angle φb = 35° are shown. The flow simulation has been carried out on basis of numerical integration of the Reynolds equations with an additional term containing artificial compressibility. The differential two-parameter model of Menter (SST has been applied to take into account turbulent effects. Numerical integration of the equations is carried out using an implicit quasi-monotone Godunov type scheme of second - order accuracy in space and time. The calculations have been conducted with the help of the software system IPMFlow. The analysis of fluid flow in the flow part elements is shown and the values of hydraulic losses and local cavitation coefficient have been obtained. Comparison of calculated and experimental results has been carried out.

  13. 3D Volumetric Analysis of Fluid Inclusions Using Confocal Microscopy

    Science.gov (United States)

    Proussevitch, A.; Mulukutla, G.; Sahagian, D.; Bodnar, B.

    2009-05-01

    Fluid inclusions preserve valuable information regarding hydrothermal, metamorphic, and magmatic processes. The molar quantities of liquid and gaseous components in the inclusions can be estimated from their volumetric measurements at room temperatures combined with knowledge of the PVTX properties of the fluid and homogenization temperatures. Thus, accurate measurements of inclusion volumes and their two phase components are critical. One of the greatest advantages of the Laser Scanning Confocal Microscopy (LSCM) in application to fluid inclsion analsyis is that it is affordable for large numbers of samples, given the appropriate software analysis tools and methodology. Our present work is directed toward developing those tools and methods. For the last decade LSCM has been considered as a potential method for inclusion volume measurements. Nevertheless, the adequate and accurate measurement by LSCM has not yet been successful for fluid inclusions containing non-fluorescing fluids due to many technical challenges in image analysis despite the fact that the cost of collecting raw LSCM imagery has dramatically decreased in recent years. These problems mostly relate to image analysis methodology and software tools that are needed for pre-processing and image segmentation, which enable solid, liquid and gaseous components to be delineated. Other challenges involve image quality and contrast, which is controlled by fluorescence of the material (most aqueous fluid inclusions do not fluoresce at the appropriate laser wavelengths), material optical properties, and application of transmitted and/or reflected confocal illumination. In this work we have identified the key problems of image analysis and propose some potential solutions. For instance, we found that better contrast of pseudo-confocal transmitted light images could be overlayed with poor-contrast true-confocal reflected light images within the same stack of z-ordered slices. This approach allows one to narrow

  14. Development of heat transfer package for core thermal-hydraulic design and analysis of upgraded JRR-3

    International Nuclear Information System (INIS)

    Sudo, Yukio; Ikawa, Hiromasa; Kaminaga, Masanori

    1985-01-01

    A heat transfer package was developed for the core thermal-hydraulic design and analysis of the Japan Research Reactor-3 (JRR-3) which is to be remodeled to a 20 MWt pool-type, light water-cooled reactor with 20 % low enriched uranium (LEU) plate-type fuel. This paper presents the constitution of the developed heat transfer package and the applicability of the heat transfer correlations adopted in it, based on the heat transfer experiments in which thermal-hydraulic features of the new JRR-3 core were properly reflected. (author)

  15. Analysis of the Phebus FPT0 containment thermal hydraulics with the Jericho and Trio-VF codes

    International Nuclear Information System (INIS)

    Layly, V.D.; Spitz, P.; Mailliat, A.

    1994-01-01

    This paper presents the analysis of the thermal hydraulic behavior of the containment, during the Phebus FPT0 test performed on December 2, 1993, with the Jericho code which deals with the thermal hydraulics of containment in the severe accident field. This code is part of Escadre which is the French system of codes in charge of predicting PWR severe accidents. After summarizing the relevant Jericho code characteristics and the preliminary assessment work for the Phebus conditions, we briefly describe the REPF 502 test facility and report the thermal hydraulic FPT0 experimental protocol. Then, the experiment / Jericho calculation comparisons are analysed. Because the Jericho code assumes a well-mixed atmosphere, some additional 3-D calculations have been carried out in order to get further insight on the convection flow patterns and qualify the well-mixed atmosphere assumption in the Phebus containment. (author). 9 refs., 12 figs

  16. KUGEL: a thermal, hydraulic, fuel performance, and gaseous fission product release code for pebble bed reactor core analysis

    International Nuclear Information System (INIS)

    Shamasundar, B.I.; Fehrenbach, M.E.

    1981-05-01

    The KUGEL computer code is designed to perform thermal/hydraulic analysis and coated-fuel particle performance calculations for axisymmetric pebble bed reactor (PBR) cores. This computer code was developed as part of a Department of Energy (DOE)-funded study designed to verify the published core performance data on PBRs. The KUGEL code is designed to interface directly with the 2DB code, a two-dimensional neutron diffusion code, to obtain distributions of thermal power, fission rate, fuel burnup, and fast neutron fluence, which are needed for thermal/hydraulic and fuel performance calculations. The code is variably dimensioned so that problem size can be easily varied. An interpolation routine allows variable mesh size to be used between the 2DB output and the two-dimensional thermal/hydraulic calculations

  17. 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

  18. BWR transient analysis using neutronic / thermal hydraulic coupled codes including uncertainty quantification

    International Nuclear Information System (INIS)

    Hartmann, C.; Sanchez, V.; Tietsch, W.; Stieglitz, R.

    2012-01-01

    The KIT is involved in the development and qualification of best estimate methodologies for BWR transient analysis in cooperation with industrial partners. The goal is to establish the most advanced thermal hydraulic system codes coupled with 3D reactor dynamic codes to be able to perform a more realistic evaluation of the BWR behavior under accidental conditions. For this purpose a computational chain based on the lattice code (SCALE6/GenPMAXS), the coupled neutronic/thermal hydraulic code (TRACE/PARCS) as well as a Monte Carlo based uncertainty and sensitivity package (SUSA) has been established and applied to different kind of transients of a Boiling Water Reactor (BWR). This paper will describe the multidimensional models of the plant elaborated for TRACE and PARCS to perform the investigations mentioned before. For the uncertainty quantification of the coupled code TRACE/PARCS and specifically to take into account the influence of the kinetics parameters in such studies, the PARCS code has been extended to facilitate the change of model parameters in such a way that the SUSA package can be used in connection with TRACE/PARCS for the U and S studies. This approach will be presented in detail. The results obtained for a rod drop transient with TRACE/PARCS using the SUSA-methodology showed clearly the importance of some kinetic parameters on the transient progression demonstrating that the coupling of a best-estimate coupled codes with uncertainty and sensitivity tools is very promising and of great importance for the safety assessment of nuclear reactors. (authors)

  19. Transient analysis and thermal hydraulic margins of GHARR-1 using the PARET/NAL code

    International Nuclear Information System (INIS)

    Adoo, N.A.

    2009-06-01

    The PARET code has been adapted by the IAEA for testing transient behaviour in research reactors. The PARET code provides a coupled thermal hydrodynamic and point kinetics capability with a continuous reactivity feedback and an optional voiding model that estimates the voiding produced by the subcooled boiling. The present version of the PARET/ANL 73 code provides a convenient means of assessing the various models and correlations proposed for the use in the analysis of research reactor behaviour. The Monte Carlo N-Particle code (MCNP) has been used to obtain power peaking profile for a two channel PARET/ANL model. A PARET model with the corresponding neutronics and thermal hydraulic characteristics for the miniature neutron source reactor (MNSR) has been used to simulate reactivity accidents for the Ghana Research Reactor - 1(GHARR-1) under the MNSR operation conditions of natural circulation, normal operation and reactivity insertion accidents. The simulation results via the insertion of large reactivity demonstrated the high inherent safety features of the MNSR for which the high negative reactivity feedback of moderator temperature limits power excursion and avoids consequently the escalation of clad temperature to the level of onset of sub-cooled void formation. The hot channel peaking factors for both radial and axial were found to be 1.17 and 1.44 respectively. Thermal hydraulic performance characteristics were investigated and the safety margins determined. The peak clad and coolant temperatures ranged from 59.18 0 C to 106.75 0 C and 42.95 0 C to 178.44 0 C respectively at which nucleate boiling will occur within the flow channels of the core. (au)

  20. ERP-IV-A program for transient thermal-hydraulic analysis of PWR plant

    International Nuclear Information System (INIS)

    Dai Anguo; Tang Jiahuan; Qian Huifu; Gao Zhikang

    1987-12-01

    The author deal with the descriptions of physical model of transient process in PWR plant and the function of ERP-IV (ERR-IV Transient Thermo-Hydraulic Analysis Code). The code has been developed for safety analysis and design transient. The code is characterized by the multi-loop long-term, short term, wide-range plant simulation with the capability to analyze natural circulation condition. The description of ERP-IV includes following parts: reactor, primary coolant loops, pressurizer, steam generators, main steam system, turbine, feedwater system, steam dump, relive valves, and safety valves in secondary side, etc.. The code can use for accident analysis, such as loss of all A.C. power to power plant auxiliaries (a station blackout), loss of normal feedwater, loss of load, loss of condenser vacuum and other events causing a turbine trip, complete loss of forced reactor coolant flow, uncontrolled rod cluster control assembly bank withdrawal. It can also be used for accident analysis of the emergency and limiting conditions, such as feedwater line break and main steam line rupture. It can also be utilized as a tool for system design studies, component design, setpoint studies and design transition studies, etc

  1. HTGR reactor physics, thermal-hydraulics and depletion uncertainty analysis: a proposed IAEA coordinated research project

    International Nuclear Information System (INIS)

    Tyobeka, Bismark; Reitsma, Frederik; Ivanov, Kostadin

    2011-01-01

    The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of HTGR design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The predictive capability of coupled neutronics/thermal hydraulics and depletion simulations for reactor design and safety analysis can be assessed with sensitivity analysis and uncertainty analysis methods. In order to benefit from recent advances in modeling and simulation and the availability of new covariance data (nuclear data uncertainties) extensive sensitivity and uncertainty studies are needed for quantification of the impact of different sources of uncertainties on the design and safety parameters of HTGRs. Uncertainty and sensitivity studies are an essential component of any significant effort in data and simulation improvement. In February 2009, the Technical Working Group on Gas-Cooled Reactors recommended that the proposed IAEA Coordinated Research Project (CRP) on the HTGR Uncertainty Analysis in Modeling be implemented. In the paper the current status and plan are presented. The CRP will also benefit from interactions with the currently ongoing OECD/NEA Light Water Reactor (LWR) UAM benchmark activity by taking into consideration the peculiarities of HTGR designs and simulation requirements. (author)

  2. Transient three-dimensional thermal-hydraulic analysis of nuclear reactor fuel rod arrays: general equations and numerical scheme

    International Nuclear Information System (INIS)

    Wnek, W.J.; Ramshaw, J.D.; Trapp, J.A.; Hughes, E.D.; Solbrig, C.W.

    1975-11-01

    A mathematical model and a numerical solution scheme for thermal-hydraulic analysis of fuel rod arrays are given. The model alleviates the two major deficiencies associated with existing rod array analysis models, that of a correct transverse momentum equation and the capability of handling reversing and circulatory flows. Possible applications of the model include steady state and transient subchannel calculations as well as analysis of flows in heat exchangers, other engineering equipment, and porous media

  3. Baseflow recession analysis in a large shale play: Climate variability and anthropogenic alterations mask effects of hydraulic fracturing

    Science.gov (United States)

    Arciniega-Esparza, Saúl; Breña-Naranjo, Jose Agustín; Hernández-Espriú, Antonio; Pedrozo-Acuña, Adrián; Scanlon, Bridget R.; Nicot, Jean Philippe; Young, Michael H.; Wolaver, Brad D.; Alcocer-Yamanaka, Victor Hugo

    2017-10-01

    Water resources development and landscape alteration exert marked impacts on water-cycle dynamics, including areas subjected to hydraulic fracturing (HF) for exploitation of unconventional oil and gas resources found in shale or tight sandstones. Here we apply a conceptual framework for linking baseflow analysis to changes in water demands from different sectors (e.g. oil/gas extraction, irrigation, and municipal consumption) and climatic variability in the semiarid Eagle Ford play in Texas, USA. We hypothesize that, in water-limited regions, baseflow (Qb) changes are partly due (along with climate variability) to groundwater abstraction. For a more realistic assessment, the analysis was conducted in two different sets of unregulated catchments, located outside and inside the Eagle Ford play. Three periods were considered in the analysis related to HF activities: pre-development (1980-2000), moderate (2001-2008) and intensive (2009-2015) periods. Results indicate that in the Eagle Ford play region, temporal changes in baseflow cannot be directly related to the increase in hydraulic fracturing. Instead, substantial baseflow declines during the intensive period of hydraulic fracturing represent the aggregated effects from the combination of: (1) a historical exceptional drought during 2011-2012; (2) increased groundwater-based irrigation; and (3) an intensive hydraulic fracturing activity.

  4. Analysis of the Coupled Influence of Hydraulic Conductivity and Porosity Heterogeneity on Probabilistic Risk Analysis

    Science.gov (United States)

    Libera, A.; Henri, C.; de Barros, F.

    2017-12-01

    Heterogeneities in natural porous formations, mainly manifested through the hydraulic conductivity (K) and, to a lesser degree, the porosity (Φ), largely control subsurface flow and solute transport. The influence of the heterogeneous structure of K on flow and solute transport processes has been widely studied, whereas less attention is dedicated to the joint heterogeneity of conductivity and porosity fields. Our study employs computational tools to investigate the joint effect of the spatial variabilities of K and Φ on the transport behavior of a solute plume. We explore multiple scenarios, characterized by different levels of heterogeneity of the geological system, and compare the computational results from the joint K and Φ heterogeneous system with the results originating from the generally adopted constant porosity case. In our work, we assume that the heterogeneous porosity is positively correlated to hydraulic conductivity. We perform numerical Monte Carlo simulations of conservative and reactive contaminant transport in a 3D aquifer. Contaminant mass and plume arrival times at multiple control planes and/or pumping wells operating under different extraction rates are analyzed. We employ different probabilistic metrics to quantify the risk at the monitoring locations, e.g., increased lifetime cancer risk and exceedance of Maximum Contaminant Levels (MCLs), under multiple transport scenarios (i.e., different levels of heterogeneity, conservative or reactive solutes and different contaminant species). Results show that early and late arrival times of the solute mass at the selected sensitive locations (i.e. control planes/pumping wells) as well as risk metrics are strongly influenced by the spatial variability of the Φ field.

  5. 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)

  6. 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)

  7. Thermal-hydraulic analysis of loss-of-coolant accident in the JMTR

    International Nuclear Information System (INIS)

    Sakurai, Fumio; Oyamada, Rokuro

    1985-02-01

    The reevaluation of the Loss-of-Coolant Accident (LOCA) was required through the process of a safety review for the Japan Materials Testing Reactor (JMTR) core conversion from the high-enriched uranium fuel (Enrichment : 93%) to the medium-enriched uranium fuel (Enrichment : 45%). The following were concluded by thermal-hydraulic analysis of a LOCA caused by a double-ended pipe break in the JMTR primary cooling system. (1) The fuel in the core does not burn-out as long as it is covered with water. (2) A larger siphon break valve (larger than phi60mm) should be installed instead of the present one (phi25mm) on the primary cooling system in order to prevent the core from being uncovered with water in case of a LOCA caused by a double-ended pipe break. The present siphon break valve was installed to keep the core covered with water in case of a LOCA caused by a small pipe rupture. In this analysis, the Siphon Breaker Analysis Code (SBAC) was written in order to analyse the size of the siphon break valve and its accuracy was confirmed to be within 5% through a verification experiment. (author)

  8. FE Analysis of Rock with Hydraulic-Mechanical Coupling Based on Continuum Damage Evolution

    Directory of Open Access Journals (Sweden)

    Yongliang Wang

    2016-01-01

    Full Text Available A numerical finite element (FE analysis technology is presented for efficient and reliable solutions of rock with hydraulic-mechanical (HM coupling, researching the seepage characteristics and simulating the damage evolution of rock. To be in accord with the actual situation, the rock is naturally viewed as heterogeneous material, in which Young’s modulus, permeability, and strength property obey the typical Weibull distribution function. The classic Biot constitutive relation for rock as porous medium is introduced to establish a set of equations coupling with elastic solid deformation and seepage flow. The rock is subsequently developed into a novel conceptual and practical model considering the damage evolution of Young’s modulus and permeability, in which comprehensive utilization of several other auxiliary technologies, for example, the Drucker-Prager strength criterion, the statistical strength theory, and the continuum damage evolution, yields the damage variable calculating technology. To this end, an effective and reliable numerical FE analysis strategy is established. Numerical examples are given to show that the proposed method can establish heterogeneous rock model and be suitable for different load conditions and furthermore to demonstrate the effectiveness and reliability in the seepage and damage characteristics analysis for rock.

  9. Hydraulic analysis of emergency core cooling system of reactor RP-10

    International Nuclear Information System (INIS)

    Gallardo Padilla, Alberto; Moreyra, Geraldo Lazaro; Nieto Malpartida, Manuel

    2002-01-01

    For design of the Emergency Core Cooling System (ECCS) of reactor RP-10 from Peru is very important the hydraulic analysis of this system. In this paper, based on a basic design of the ECCS are showed the conservation equations, the parabolic movement, being deduced from them the equations to evaluate regarding the time the variables to consider in the design: level of the emergency water in the reserve tank, flow, reaches of sprinkle, etc. In this analysis is considered a quasi-stationary flow for simplify the calculation. The developed model was implemented in a computer program denominated ECCSRP10, in language Fortran 77, whose results are shown in form graph. From analysis of results we can conclude that for the system of pipe of the ECCS the appropriate diameter is of 2 , and that the maximum flow possible to give is of 5 m 3 /h for to assure a minimum time of refrigeration of 150000 seconds. Experimental tests were made in a prototype of the pipe system being demonstrated that the obtained results of the simplified calculation agree with the values registered with a global approach of 10%. (author)

  10. Thermal-hydraulic analysis of total loss of steam generator feed water in WWER-440

    International Nuclear Information System (INIS)

    Sabotinov, L.; Cadet-Mercier, S.

    2001-01-01

    The analysis is carried out for a WWER-440/V270 with upgraded primary safety valves (replacement of the existing PRZ safety valves with Pilot Operated Relief Valves (PORV) of the type SEBIM (France)) The current analysis is focused on the scenario 'Total Loss of SGs Feed Water' with application of the operator action of primary system 'Feed and Bleed' in order to check the effectiveness of the installed pressurizer SEBIM valves and to verify that the operator can cool down the reactor system and cope with this accident. The calculations have been performed at the Institute of Protection and Nuclear Safety (IPSN) in Fontenay-aux-Roses with the computer code CATHARE 2 Version 1.3L1. CATHARE is a French best estimate thermal-hydraulic program for accident analysis in the light water nuclear reactors, developed with the participation of the IPSN (Institut de Protection et Surete Nucleaire), CEA (Commissariat a l'Energie Atomique), Framatome and EdF (Electricite de France). (author)

  11. Steady state thermal hydraulic analysis of LMR core using COBRA-K code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Eui Kwang; Kim, Young Gyun; Kim Young In; Kim Young Cheol

    1997-02-01

    A thermal hydraulics analysis code COBRA-K is being developed by the KAERI LMR core design technology development team. COBRA-K is a part of the integrated computation system for LMR core design and analysis, the K-CORE system. COBRA-K is supposed to predict the flow and temperature distributions in LMR core. COBRA-K is an extension of the previously published COBRA-IV-I code with several functional improvements. Specially COBRA-K has been improved to analyze single and multi-assembly, and whole-core in the transient condition. This report describes the overall features of COBRA-K and gives general input descriptions. The 19 pin assembly experimental data of ORNL were used to verify the accuracy of this code for the steady state analysis. The comparative results show good agreements between the calculated and the measured data. And COBRA-K can be used to predict flow and temperature distributions for the LMR core design. (author). 7 refs., 6 tabs., 13 figs.

  12. 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.

  13. Analysis of Hydraulic Fluids and Lubricating Oils for the Formation of Trimethylolpropane Phosphate (TMP-P)

    Science.gov (United States)

    1989-08-09

    as to whether TMP-P was actually present, or wether the result represented only a campound which co-chromatographed at the same retention time as TMP...X E 841 76. Milbrath, Dean S., Engel, Judith L., Verkade, john G., and Casida, John E. Structure-Toxicity relationsips of 1-substituted-4-alkyl-2,6,7

  14. Post-excavation analysis of a revised hydraulic model of the Room 209 fracture, URL, Manitoba, Canada

    International Nuclear Information System (INIS)

    Winberg, A.; Tin Chan; Griffiths, P.; Nakka, B.

    1989-10-01

    An excavation response test was conducted in the Room 209 on the 240 m level of the AECL Underground Research Laboratory. Model predictions prior to excavation were made of the geomechanical response of the rock mass and the hydraulic response of an intercepted fracture. The model results were compared with excavation response data collected in a comprehensive instrument array. The work performed has addressed discrepancies between calculated and in-situ measured hydraulic response as part of a post-test analysis. Already existing hydraulic conceptual models of the fracture were revised and any available information was included in the new model. The model reproduced the pre-excavation hydraulic head distribution and hydraulic test results in terms of normalized flow rate within 5% and 75%, respectively. It was also found that the model reproduced the results of cross-hole hydraulic interference tests at least from a qualitative standpoint. The next stage of the modelling addressed the response of the model to a simulation of the excavated pilot tunnel. The preliminary results suggested the presence of a skin of different permeability in a thin zone around the periphery of the tunnel. By altering the permeability in the floor and along the walls and roof of the periphery, a better correspondence between calculated and measured drawdown was obtained. The same also applied for measured groundwater inflow in quantity, though not for the actual distribution on inflow. As probable causes for the interpreted positive skin in the crown and wall, temporary partial unsaturation and propulsion of debris into the fracture were suggested. The negative skin in the floor was interpreted as an effect of the dense and high energy charges used in the excavation process. (authors)

  15. 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

  16. Supercritical fluid chromatography in drug analysis: a literature survey.

    Science.gov (United States)

    Salvador, A; Jaime, M A; Becerra, G; Guardia, M de L

    1996-08-01

    The applications of supercritical fluid chromatography to the analysis of drugs have been carefully revised from the literature compiled in the Analytical Abstracts until March 1994. Easy-to-read tables provide useful information about the state-of-the-art and possibilities offered by SFC in pharmaceutical analysis. The tables comprise extensive data about samples analyzed, pharmaceutical principles determined, solvents used and sample quantity injected, supercritical fluids and modifiers employed, injection system, instrumentation, experimental conditions for chromatographic separations (density, pressure, flow, temperature), characteristics of columns employed (type, support, length, diameter, particle film thickness, stationary phase), detectors, type of restrictors, and also some analytical features of the methods developed (such as retention time, resolution, sensitivity, limit of detection and relative standard deviation).

  17. Neutronic analysis of two-fluid thorium molten salt reactor

    International Nuclear Information System (INIS)

    Frybort, Jan; Vocka, Radim

    2009-01-01

    The aim of this paper is to evaluate features of the two-fluid MSBR through a parametric study and compare its properties to one-fluid MSBR concepts. The starting point of the analysis is the original ORNL 1000 MWe reactor design, although simplified to some extent. We studied the influence of dimensions of distinct reactor parts - fuel and fertile channels radius, plenum height, design etc. - on fundamental reactor properties: breeding ratio and doubling time, reactor inventory, graphite lifetime, and temperature feedback coefficients. The calculations were carried out using MCNP5 code. Based on obtained results we proposed an improved reactor design. Our results show clear advantages of the concept with two separate fluoride salts if compared to the one fluid concept in breading, doubling time, and temperature feedback coefficients. Limitations of the two-fluid concept - particularly the graphite lifetime - are also pointed out. The reactor design can be a subject of further optimizations, namely from the viewpoint of reactor safety. (author)

  18. 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)

  19. Development of whole core thermal-hydraulic analysis program ACT. 3. Coupling core module with primary heat transport system module

    International Nuclear Information System (INIS)

    Ohtaka, Masahiko; Ohshima, Hiroyuki

    1998-10-01

    A whole core thermal-hydraulic analysis program ACT is being developed for the purpose of evaluating detailed in-core thermal hydraulic phenomena of fast reactors including inter-wrapper flow under various reactor operation conditions. In this work, the core module as a main part of the ACT developed last year, which simulates thermal-hydraulics in the subassemblies and the inter-subassembly gaps, was coupled with an one dimensional plant system thermal-hydraulic analysis code LEDHER to simulate transients in the primary heat transport system and to give appropriate boundary conditions to the core model. The effective algorithm to couple these two calculation modules was developed, which required minimum modification of them. In order to couple these two calculation modules on the computing system, parallel computing technique using PVM (Parallel Virtual Machine) programming environment was applied. The code system was applied to analyze an out-of-pile sodium experiment simulating core with 7 subassemblies under transient condition for code verification. It was confirmed that the analytical results show a similar tendency of experimental results. (author)

  20. Improvement of multi-dimensional realistic thermal-hydraulic system analysis code, MARS 1.3

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Jae; Chung, Bub Dong; Jeong, Jae Jun; Ha, Kwi Seok

    1998-09-01

    The MARS (Multi-dimensional Analysis of Reactor Safety) code is a multi-dimensional, best-estimate thermal-hydraulic system analysis code. This report describes the new features that have been improved in the MARS 1.3 code since the release of MARS 1.3 in July 1998. The new features include: - implementation of point kinetics model into the 3D module - unification of the heat structure model - extension of the control function to the 3D module variables - improvement of the 3D module input check function. Each of the items has been implemented in the developmental version of the MARS 1.3.1 code and, then, independently verified and assessed. The effectiveness of the new features is well verified and it is shown that these improvements greatly extend the code capability and enhance the user friendliness. Relevant input data changes are also described. In addition to the improvements, this report briefly summarizes the future code developmental activities that are being carried out or planned, such as coupling of MARS 1.3 with the containment code CONTEMPT and the three-dimensional reactor kinetics code MASTER 2.0. (author). 8 refs.

  1. Improvement of multi-dimensional realistic thermal-hydraulic system analysis code, MARS 1.3

    International Nuclear Information System (INIS)

    Lee, Won Jae; Chung, Bub Dong; Jeong, Jae Jun; Ha, Kwi Seok

    1998-09-01

    The MARS (Multi-dimensional Analysis of Reactor Safety) code is a multi-dimensional, best-estimate thermal-hydraulic system analysis code. This report describes the new features that have been improved in the MARS 1.3 code since the release of MARS 1.3 in July 1998. The new features include: - implementation of point kinetics model into the 3D module - unification of the heat structure model - extension of the control function to the 3D module variables - improvement of the 3D module input check function. Each of the items has been implemented in the developmental version of the MARS 1.3.1 code and, then, independently verified and assessed. The effectiveness of the new features is well verified and it is shown that these improvements greatly extend the code capability and enhance the user friendliness. Relevant input data changes are also described. In addition to the improvements, this report briefly summarizes the future code developmental activities that are being carried out or planned, such as coupling of MARS 1.3 with the containment code CONTEMPT and the three-dimensional reactor kinetics code MASTER 2.0. (author). 8 refs

  2. Numerical experiment on variance biases and Monte Carlo neutronics analysis with thermal hydraulic feedback

    International Nuclear Information System (INIS)

    Hyung, Jin Shim; Beom, Seok Han; Chang, Hyo Kim

    2003-01-01

    Monte Carlo (MC) power method based on the fixed number of fission sites at the beginning of each cycle is known to cause biases in the variances of the k-eigenvalue (keff) and the fission reaction rate estimates. Because of the biases, the apparent variances of keff and the fission reaction rate estimates from a single MC run tend to be smaller or larger than the real variances of the corresponding quantities, depending on the degree of the inter-generational correlation of the sample. We demonstrate this through a numerical experiment involving 100 independent MC runs for the neutronics analysis of a 17 x 17 fuel assembly of a pressurized water reactor (PWR). We also demonstrate through the numerical experiment that Gelbard and Prael's batch method and Ueki et al's covariance estimation method enable one to estimate the approximate real variances of keff and the fission reaction rate estimates from a single MC run. We then show that the use of the approximate real variances from the two-bias predicting methods instead of the apparent variances provides an efficient MC power iteration scheme that is required in the MC neutronics analysis of a real system to determine the pin power distribution consistent with the thermal hydraulic (TH) conditions of individual pins of the system. (authors)

  3. Simplified thermal-hydraulic analysis of single phase natural circulation circuit with two heat exchangers

    Energy Technology Data Exchange (ETDEWEB)

    Pinheiro, Larissa Cunha; Su, Jian, E-mail: larissa@lasme.coppe.ufrj.br, E-mail: sujian@lasme.coppe.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenhraria Nuclear; Cotta, Renato Machado, E-mail: cotta@mecanica.coppe.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (POLI/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Dept. de Engenharia Mecanica

    2015-07-01

    Single phase natural circulation circuits composed of two convective heat exchangers and connecting tubes are important for the passive heat removal from spent fuel pools (SFP). To keep the structural integrity of the stored spent fuel assemblies, continuously cooling has to be provided in order to avoid increase at the pool temperature and subsequent uncovering of the fuel and enhanced reaction between water and metal releasing hydrogen. Decay heat can achieve considerably high amounts of energy e.g. in the AP1000, considering the emergency fuel assemblies, the maximum heat decay will reach 13 MW in the 15th day (Westinghouse Electric Company, 2010). A highly efficient alternative to do so is by means of natural circulation, which is cost-effective compared to active cooling systems and is inherently safer since presents less associated devices and no external work is required. Many researchers have investigated safety and stability aspects of natural circulation loops (NCL). However, there is a lack of literature concerning the improvement of NCL through a standard unified methodology, especially for natural circulation circuits with two heat exchangers. In the present study, a simplified thermal-hydraulic analysis of single phase natural circulation circuit with two heat exchanges is presented. Relevant dimensionless key groups were proposed to for the design and safety analysis of a scaled NCL for the cooling of spent fuel storage pool with convective cooling and heating. (author)

  4. Multi-Function Waste Tank Facility thermal hydraulic analysis for Title II design

    International Nuclear Information System (INIS)

    Cramer, E.R.

    1994-01-01

    The purpose of this work was to provide the thermal hydraulic analysis for the Multi-Function Waste Tank Facility (MWTF) Title II design. Temperature distributions throughout the tank structure were calculated for subsequent use in the structural analysis and in the safety evaluation. Calculated temperatures of critical areas were compared to design allowables. Expected operating parameters were calculated for use in the ventilation system design and in the environmental impact documentation. The design requirements were obtained from the MWTF Functional Design Criteria (FDC). The most restrictive temperature limit given in the FDC is the 200 limit for the haunch and dome steel and concrete. The temperature limit for the rest of the primary and secondary tanks and concrete base mat and supporting pad is 250 F. Also, the waste should not be allowed to boil. The tank geometry was taken from ICF Kaiser Engineers Hanford drawing ES-W236A-Z1, Revision 1, included here in Appendix B. Heat removal rates by evaporation from the waste surface were obtained from experimental data. It is concluded that the MWTF tank cooling system will meet the design temperature limits for the design heat load of 700,000 Btu/h, even if cooling flow is lost to the annulus region, and temperatures change very slowly during transients due to the high heat capacity of the tank structure and the waste. Accordingly, transients will not be a significant operational problem from the viewpoint of meeting the specified temperature limits

  5. Peptidome analysis of amniotic fluid from pregnancies with preeclampsia.

    Science.gov (United States)

    Qian, Yating; Zhang, Lei; Rui, Can; Ding, Hongjuan; Mao, Pengyuan; Ruan, Hongjie; Jia, Ruizhe

    2017-11-01

    Preeclampsia (PE), a life‑threatening, complicated pregnancy‑associated disease, has recently become a research focus in obstetrics. However, the peptidome of the amniotic fluid in PE patients has rarely been investigated. The present study used peptidomic profiling to perform a comparative analysis of human amniotic fluid between normal and PE pregnancies. Centrifugal ultrafiltration and liquid chromatography‑tandem mass spectrometry (LC‑MS/MS) was combined with isotopomeric dimethyl labels to gain a deeper understanding of the role of proteins and the peptidome in the onset of PE. Following ultrafiltration and LC‑MS/MS, 352 peptides were identified. Of these, 23 peptides were observed to be significantly differentially expressed (6 downregulated and 17 upregulated; POntology and Blastp analyses, the functions and biological activities of these 23 peptides were identified and revealed to include autophagy, signal transduction, receptor activity, enzymatic activity and nucleic acid binding. In addition, a bibliographic search revealed that some of the identified peptides, including Titin, are crucial to the pathogenesis underlying PE. The present study identified 23 peptides expressed at significantly different levels in the amniotic fluid of PE and normal pregnancies. A comprehensive peptidome analysis is more efficient than a simple biomarker analysis at revealing deficiencies and improving the detection rate in diseases. These analyses therefore provide a substantial advantage in applications aimed at the discovery of disease‑specific biomarkers.

  6. 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 ...

  7. 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....

  8. CFD Analysis on the Thermal Hydraulic Performance of an SAH Duct with Multi V-Shape Roughened Ribs

    Directory of Open Access Journals (Sweden)

    Anil Kumar

    2016-05-01

    Full Text Available This study presents the heat transfer and fluid flow characteristics in a rib-roughened SAH (solar air heater channel. The artificial roughness of the rectangular channel was in the form of a thin circular wire in discrete multi V-pattern rib geometries. The effect of this geometry on heat transfer, fluid flow, and performance augmentation was investigated using the CFD (computational fluid dynamics. The roughness parameters were a relative discrete distance of 0.69, a relative rib height of 0.043, a relative rib pitch of 10, a relative rib width of 6.0, and a flow-attack-angle of 60°. The discrete width ratios and Reynolds numbers ranged from 0.5 to 2.0 and from 2000 to 20,000, respectively. The CFD results using the renormalization k-epsilon model were in good agreement with the empirical relationship. This model was used to investigate the heat transfer and fluid flow characteristics in the multi V-pattern rib roughened SAH channel. The thermo-hydraulic performance was found to be the best for the discrete width ratio of 1.0. A discrete multi V-pattern rib combined with dimple staggered ribs also had better overall thermal performance compared to other rib shapes.

  9. Use of computer programs STLK1 and STWT1 for analysis of stream-aquifer hydraulic interaction

    Science.gov (United States)

    Desimone, Leslie A.; Barlow, Paul M.

    1999-01-01

    Quantifying the hydraulic interaction of aquifers and streams is important in the analysis of stream base fow, flood-wave effects, and contaminant transport between surface- and ground-water systems. This report describes the use of two computer programs, STLK1 and STWT1, to analyze the hydraulic interaction of streams with confined, leaky, and water-table aquifers during periods of stream-stage fuctuations and uniform, areal recharge. The computer programs are based on analytical solutions to the ground-water-flow equation in stream-aquifer settings and calculate ground-water levels, seepage rates across the stream-aquifer boundary, and bank storage that result from arbitrarily varying stream stage or recharge. Analysis of idealized, hypothetical stream-aquifer systems is used to show how aquifer type, aquifer boundaries, and aquifer and streambank hydraulic properties affect aquifer response to stresses. Published data from alluvial and stratifed-drift aquifers in Kentucky, Massachusetts, and Iowa are used to demonstrate application of the programs to field settings. Analytical models of these three stream-aquifer systems are developed on the basis of available hydrogeologic information. Stream-stage fluctuations and recharge are applied to the systems as hydraulic stresses. The models are calibrated by matching ground-water levels calculated with computer program STLK1 or STWT1 to measured ground-water levels. The analytical models are used to estimate hydraulic properties of the aquifer, aquitard, and streambank; to evaluate hydrologic conditions in the aquifer; and to estimate seepage rates and bank-storage volumes resulting from flood waves and recharge. Analysis of field examples demonstrates the accuracy and limitations of the analytical solutions and programs when applied to actual ground-water systems and the potential uses of the analytical methods as alternatives to numerical modeling for quantifying stream-aquifer interactions.

  10. Probabilistic Risk Assessment of Hydraulic Fracturing in Unconventional Reservoirs by Means of Fault Tree Analysis: An Initial Discussion

    Science.gov (United States)

    Rodak, C. M.; McHugh, R.; Wei, X.

    2016-12-01

    The development and combination of horizontal drilling and hydraulic fracturing has unlocked unconventional hydrocarbon reserves around the globe. These advances have triggered a number of concerns regarding aquifer contamination and over-exploitation, leading to scientific studies investigating potential risks posed by directional hydraulic fracturing activities. These studies, balanced with potential economic benefits of energy production, are a crucial source of information for communities considering the development of unconventional reservoirs. However, probabilistic quantification of the overall risk posed by hydraulic fracturing at the system level are rare. Here we present the concept of fault tree analysis to determine the overall probability of groundwater contamination or over-exploitation, broadly referred to as the probability of failure. The potential utility of fault tree analysis for the quantification and communication of risks is approached with a general application. However, the fault tree design is robust and can handle various combinations of regional-specific data pertaining to relevant spatial scales, geological conditions, and industry practices where available. All available data are grouped into quantity and quality-based impacts and sub-divided based on the stage of the hydraulic fracturing process in which the data is relevant as described by the USEPA. Each stage is broken down into the unique basic events required for failure; for example, to quantify the risk of an on-site spill we must consider the likelihood, magnitude, composition, and subsurface transport of the spill. The structure of the fault tree described above can be used to render a highly complex system of variables into a straightforward equation for risk calculation based on Boolean logic. This project shows the utility of fault tree analysis for the visual communication of the potential risks of hydraulic fracturing activities on groundwater resources.

  11. Thermal-hydraulic analysis of Ignalina NPP compartments response to group distribution header rupture using RALOC4 code

    International Nuclear Information System (INIS)

    Urbonavicius, E.

    2000-01-01

    The Accident Localisation System (ALS) of Ignalina NPP is a containment of pressure suppression type designed to protect the environment from the dangerous impact of the radioactivity. The failure of ALS could lead to contamination of the environment and prescribed public radiation doses could be exceeded. The purpose of the presented analysis is to perform long term thermal-hydraulic analysis of compartments response to Group Distribution Header rupture and verify if design pressure values are not exceeded. (authors)

  12. Packaged integrated opto-fluidic solution for harmful fluid analysis

    Science.gov (United States)

    Allenet, T.; Bucci, D.; Geoffray, F.; Canto, F.; Couston, L.; Jardinier, E.; Broquin, J.-E.

    2016-02-01

    Advances in nuclear fuel reprocessing have led to a surging need for novel chemical analysis tools. In this paper, we present a packaged lab-on-chip approach with co-integration of optical and micro-fluidic functions on a glass substrate as a solution. A chip was built and packaged to obtain light/fluid interaction in order for the entire device to make spectral measurements using the photo spectroscopy absorption principle. The interaction between the analyte solution and light takes place at the boundary between a waveguide and a fluid micro-channel thanks to the evanescent part of the waveguide's guided mode that propagates into the fluid. The waveguide was obtained via ion exchange on a glass wafer. The input and the output of the waveguides were pigtailed with standard single mode optical fibers. The micro-scale fluid channel was elaborated with a lithography procedure and hydrofluoric acid wet etching resulting in a 150+/-8 μm deep channel. The channel was designed with fluidic accesses, in order for the chip to be compatible with commercial fluidic interfaces/chip mounts. This allows for analyte fluid in external capillaries to be pumped into the device through micro-pipes, hence resulting in a fully packaged chip. In order to produce this co-integrated structure, two substrates were bonded. A study of direct glass wafer-to-wafer molecular bonding was carried-out to improve detector sturdiness and durability and put forward a bonding protocol with a bonding surface energy of γ>2.0 J.m-2. Detector viability was shown by obtaining optical mode measurements and detecting traces of 1.2 M neodymium (Nd) solute in 12+/-1 μL of 0.01 M and pH 2 nitric acid (HNO3) solvent by obtaining an absorption peak specific to neodymium at 795 nm.

  13. Full vessel CFD analysis on thermal-hydraulic characteristics of CPR1000 PWR

    International Nuclear Information System (INIS)

    Chao Yanmeng; Yang Lixin; Zhang Mingqian

    2014-01-01

    To obtain flow distributions and thermal-hydraulic properties in a full vessel PWR under limited computation ability and time, a full vessel simulation model of CPR1000 was built based on two simplification methods. One simplified the inner geometry of the control rod guide tubes using equivalent flow area. Another substituted the core by a porous domain to maintain the pressure drop and temperature rise. After the computation, global and localized flow distributions, hydraulic loads of some main assemblies were obtained, as well as other thermal-hydraulic properties. The results indicate the flow distribution in the full vessel is asymmetrical. Therefore it is essential to use the full vessel model to simulate. The calculated thermal-hydraulic characteristics agree well with the operation statistics, providing the reference data for the reactor safety operation. (authors)

  14. Case Study Analysis of the Impacts of Water Acquisition for Hydraulic Fracturing on Local Water Availability

    Science.gov (United States)

    Hydraulic fracturing (HF) is used to develop unconventional gas reserves, but the technology requires large volumes of water, placing demands on local water resources and potentially creating conflict with other users and ecosystems. This study examines the balance between water ...

  15. Design and performance characteristic analysis of servo valve-type water hydraulic poppet valve

    International Nuclear Information System (INIS)

    Park, Sung Hwan

    2009-01-01

    For water hydraulic system control, the flow or pressure control using high-speed solenoid valve controlled by PWM control method could be a good solution for prevention of internal leakage. However, since the PWM control of on-off valves cause extensive flow and pressure fluctuation, it is difficult to control the water hydraulic actuators precisely. In this study, the servo valve-type water hydraulic valve using proportional poppet as the main valve is designed and the performance characteristics of the servo valve-type water hydraulic valve are analyzed. Furthermore, it is demonstrated through experiments that a decline in control chamber pressure that follows the change of pilot flow is caused by the occurrence of cavitation around the proportional poppet, and that fundamental characteristics of the developed valve remain unaffected by the occurrence of cavitation

  16. Development of whole core thermal-hydraulic analysis program ACT. 4. Simplified fuel assembly model and parallelization by MPI

    International Nuclear Information System (INIS)

    Ohshima, Hiroyuki

    2001-10-01

    A whole core thermal-hydraulic analysis program ACT is being developed for the purpose of evaluating detailed in-core thermal hydraulic phenomena of fast reactors including the effect of the flow between wrapper-tube walls (inter-wrapper flow) under various reactor operation conditions. As appropriate boundary conditions in addition to a detailed modeling of the core are essential for accurate simulations of in-core thermal hydraulics, ACT consists of not only fuel assembly and inter-wrapper flow analysis modules but also a heat transport system analysis module that gives response of the plant dynamics to the core model. This report describes incorporation of a simplified model to the fuel assembly analysis module and program parallelization by a message passing method toward large-scale simulations. ACT has a fuel assembly analysis module which can simulate a whole fuel pin bundle in each fuel assembly of the core and, however, it may take much CPU time for a large-scale core simulation. Therefore, a simplified fuel assembly model that is thermal-hydraulically equivalent to the detailed one has been incorporated in order to save the simulation time and resources. This simplified model is applied to several parts of fuel assemblies in a core where the detailed simulation results are not required. With regard to the program parallelization, the calculation load and the data flow of ACT were analyzed and the optimum parallelization has been done including the improvement of the numerical simulation algorithm of ACT. Message Passing Interface (MPI) is applied to data communication between processes and synchronization in parallel calculations. Parallelized ACT was verified through a comparison simulation with the original one. In addition to the above works, input manuals of the core analysis module and the heat transport system analysis module have been prepared. (author)

  17. Fluid Dynamic Models for Bhattacharyya-Based Discriminant Analysis.

    Science.gov (United States)

    Noh, Yung-Kyun; Hamm, Jihun; Park, Frank Chongwoo; Zhang, Byoung-Tak; Lee, Daniel D

    2018-01-01

    Classical discriminant analysis attempts to discover a low-dimensional subspace where class label information is maximally preserved under projection. Canonical methods for estimating the subspace optimize an information-theoretic criterion that measures the separation between the class-conditional distributions. Unfortunately, direct optimization of the information-theoretic criteria is generally non-convex and intractable in high-dimensional spaces. In this work, we propose a novel, tractable algorithm for discriminant analysis that considers the class-conditional densities as interacting fluids in the high-dimensional embedding space. We use the Bhattacharyya criterion as a potential function that generates forces between the interacting fluids, and derive a computationally tractable method for finding the low-dimensional subspace that optimally constrains the resulting fluid flow. We show that this model properly reduces to the optimal solution for homoscedastic data as well as for heteroscedastic Gaussian distributions with equal means. We also extend this model to discover optimal filters for discriminating Gaussian processes and provide experimental results and comparisons on a number of datasets.

  18. Application of Lie group analysis in geophysical fluid dynamics

    CERN Document Server

    Ibragimov, Ranis

    2011-01-01

    This is the first monograph dealing with the applications of the Lie group analysis to the modeling equations governing internal wave propagation in the deep ocean. A new approach to describe the nonlinear interactions of internal waves in the ocean is presented. While the central idea of the book is to investigate oceanic internal waves through the prism of Lie group analysis, it is also shown for the first time that internal wave beams, representing exact solutions to the equation of motion of stratified fluid, can be found by solving the given model as invariant solutions of nonlinear equat

  19. CFD analysis of thermal-hydraulic behavior in SCWR typical flow channels

    International Nuclear Information System (INIS)

    Gu, H.Y.; Cheng, X.; Yang, Y.H.

    2008-01-01

    Investigations on thermal-hydraulic behavior in SCWR fuel assembly have obtained a significant attention in the international SCWR community. However, there is still a lack of understanding and ability to predict the heat transfer behavior of supercritical water. In this paper, CFD analysis is carried out to study the flow and heat transfer behavior of supercritical water in sub-channels of both square and triangular rod bundles. Effect of various parameters, e.g. thermal boundary conditions and pitch-to-diameter ratio on the thermal-hydraulic behavior is investigated. Two boundary conditions, i.e., constant heat flux at the outer surface of cladding and constant heat density in the fuel pin are applied. The results show that the structure of the secondary flow mainly depends on the rod bundle configuration as well as the pitch-to-diameter ratio, whereas, the amplitude of the secondary flow is affected by the thermal boundary conditions, as well. The secondary flow is much stronger in a square lattice than that in a triangular lattice. The turbulence behavior is similar in both square and triangular lattices. The dependence of the amplitude of the turbulent velocity fluctuation across the gap on Reynolds number becomes prominent in both lattices as the pitch-to-diameter ratio increases. The effect of thermal boundary conditions on turbulent velocity fluctuation is negligibly small. For both lattices with small pitch-to-diameter ratios (P/D < 1.3), the mixing coefficient is about 0.022. Both secondary flow and turbulent mixing show unusual behavior in the vicinity of the pseudo-critical point. Further investigation is needed. A strong circumferential non-uniformity of wall temperature and heat transfer is observed in tight lattices at constant heat flux boundary conditions, especially in square lattices. In the case with constant heat density of fuel pin, the circumferential conductive heat transfer significantly reduces the non-uniformity of circumferential

  20. Spectral analysis of viscous static compressible fluid equilibria

    Energy Technology Data Exchange (ETDEWEB)

    Nunez, Manuel [Departamento de Analisis Matematico, Universidad de Valladolid, Valladolid (Spain)

    2001-05-25

    It is generally assumed that the study of the spectrum of the linearized Navier-Stokes equations around a static state will provide information about the stability of the equilibrium. This is obvious for inviscid barotropic compressible fluids by the self-adjoint character of the relevant operator, and rather easy for viscous incompressible fluids by the compact character of the resolvent. The viscous compressible linearized system, both for periodic and homogeneous Dirichlet boundary problems, satisfies neither condition, but it does turn out to be the generator of an immediately continuous, almost stable semigroup, which justifies the analysis of the spectrum as predictive of the initial behaviour of the flow. As for the spectrum itself, except for a unique negative finite accumulation point, it is formed by eigenvalues with negative real part, and nonreal eigenvalues are confined to a certain bounded subset of complex numbers. (author)

  1. Thermal Hydraulics Analysis for the 3MW TRIGA MARK-II Research Reactor Under Transient Condition

    International Nuclear Information System (INIS)

    Huda, M.Q.; Bhuiyan, S.I.; Mondal, M.A.W.

    1996-12-01

    Some important thermal hydraulic parameters of the 3 MW TRIGA MARK-II research reactor operating under transient condition were investigated using two computer codes PULTRI and TEMPUL. Major transient parameters, such as, peak power and prompt energy released after pulse, maximum fuel and coolant temperature, surface heat flux, time and radial distribution of temperature within fuel element after pulse, fuel, fuel-cladding gap width variation, etc. were computer and compared with the experimental and operational values as reported in the safety Analysis Report (SAR). It was observed that pulsing of the reactor inserting an excess reactivity of $2.00 shoots the reactor power level to 854.353 MW compared to an experimental value of 852 MW; the maximum fuel temperature corresponding to this peak power was found to be 846.76 o C which is much less than the limiting maximum value of fuel temperature of 1150 0 C as reported in SAR. During a pulse if the film boiling occurs for a peak adiabatic fuel temperature of 1000 o C, the calculated outer cladding wall temperature was observed to be 702.39 0 C compared to a value of 760 o C reported in SAR under the same condition. The investigated other results were also found to be in good agreement with the values reported in the SAR. 16 refs., 22 figs. (author)

  2. COBRA/TRAC analysis of two-dimensional thermal-hydraulic behavior in SCTF reflood tests

    International Nuclear Information System (INIS)

    Iwamura, Takamichi; Ohnuki, Akira; Sobajima, Makoto; Adachi, Hiromichi

    1987-01-01

    The effects of radial power distribution and non-uniform upper plenum water accumulation on thermal-hydraulic behavior in the core were observed in the reflood tests with Slab Core Test Facility (SCTF). In order to examine the predictability of these two effects by a multi-dimensional analysis code, the COBRA/TRAC calculations were made. The calculated results indicated that the heat transfer enhancement in high power bundles above quench front was caused by high vapor flow rate in those bundles due to the radial power distribution. On the other hand, the heat transfer degradation in the peripheral bundles under the condition of non-uniform upper plenum water accumulation was caused by the lower flow rates of vapor and entrained liquid above the quench front in those bundles by the reason that vapor concentrated in the center bundles due to the cross flow induced by the horizontal pressure gradient in the core. The above-mentioned two-dimensional heat transfer behaviors calculated with the COBRA/TRAC code is similar to those observed in SCTF tests and therefore those calculations are useful to investigate the mechanism of the two-dimensional effects in SCTF reflood tests. (author)

  3. Independent assessment of MELCOR as a severe accident thermal-hydraulic/source term analysis tool

    International Nuclear Information System (INIS)

    Madni, I.K.; Eltawila, F.

    1994-01-01

    MELCOR is a fully integrated computer code that models all phases of the progression of severe accidents in light water reactor nuclear power plants, and is being developed for the US Nuclear Regulatory Commission (NRC) by Sandia National Laboratories (SNL). Brookhaven National Laboratory (BNL) has a program with the NRC called ''MELCOR Verification, Benchmarking, and Applications,'' whose aim is to provide independent assessment of MELCOR as a severe accident thermal-hydraulic/source term analysis tool. The scope of this program is to perform quality control verification on all released versions of MELCOR, to benchmark MELCOR against more mechanistic codes and experimental data from severe fuel damage tests, and to evaluate the ability of MELCOR to simulate long-term severe accident transients in commercial LWRs, by applying the code to model both BWRs and PWRs. Under this program, BNL provided input to the NRC-sponsored MELCOR Peer Review, and is currently contributing to the MELCOR Cooperative Assessment Program (MCAP). This paper presents a summary of MELCOR assessment efforts at BNL and their contribution to NRC goals with respect to MELCOR

  4. Lead-Bismuth Eutectic cooled experimental Accelerator Driven System. Windowless target unit thermal-hydraulic analysis

    International Nuclear Information System (INIS)

    Bianchi, F.; Ferri, R.; Moreau, V.

    2004-01-01

    A main concern related to the peaceful use of nuclear energy is the safe management of nuclear wastes, with particular attention to long-lived fission products. An increasing attention has recently been addressed to transmutation systems (Accelerator Driven System: ADS) able to 'burn' the actinides and some of the long-lived fission products (High-Level Waste: HLW), transforming them in short or medium-lived wastes that may be easier managed and stored in the geological disposal, with the consequent easier acceptability by population. An ADS consists of a subcritical-core coupled with an accelerator by means of a target. This paper deals with the thermal-hydraulic analysis, performed with STAR-CD and RELAP5 codes for the windowless target unit of Lead-Bismuth Eutectic (LBE) cooled experimental ADS (XADS), both to assess its behaviour during operational and accident sequences and to provide input data for the thermal-mechanical analyses. It also reports a description of modifications properly implemented in the codes used for the assessment of this kind of plants. (author)

  5. The detection of cavitation in hydraulic machines by use of ultrasonic signal analysis

    International Nuclear Information System (INIS)

    Gruber, P; Odermatt, P; Etterlin, M; Lerch, T; Frei, M; Farhat, M

    2014-01-01

    This presentation describes an experimental approach for the detection of cavitation in hydraulic machines by use of ultrasonic signal analysis. Instead of using the high frequency pulses (typically 1MHz) only for transit time measurement different other signal characteristics are extracted from the individual signals and its correlation function with reference signals in order to gain knowledge of the water conditions. As the pulse repetition rate is high (typically 100Hz), statistical parameters can be extracted of the signals. The idea is to find patterns in the parameters by a classifier that can distinguish between the different water states. This classification scheme has been applied to different cavitation sections: a sphere in a water flow in circular tube at the HSLU in Lucerne, a NACA profile in a cavitation tunnel and a Francis model test turbine both at LMH in Lausanne. From the signal raw data several statistical parameters in the time and frequency domain as well as from the correlation function with reference signals have been determined. As classifiers two methods were used: neural feed forward networks and decision trees. For both classification methods realizations with lowest complexity as possible are of special interest. It is shown that three signal characteristics, two from the signal itself and one from the correlation function are in many cases sufficient for the detection capability. The final goal is to combine these results with operating point, vibration, acoustic emission and dynamic pressure information such that a distinction between dangerous and not dangerous cavitation is possible

  6. Application of perturbation methods and sensitivity analysis to water hammer problems in hydraulic networks

    International Nuclear Information System (INIS)

    Balino, Jorge L.; Larreteguy, Axel E.; Andrade Lima, Fernando R.

    1995-01-01

    The differential method was applied to the sensitivity analysis for water hammer problems in hydraulic networks. Starting from the classical water hammer equations in a single-phase liquid with friction, the state vector comprising the piezometric head and the velocity was defined. Applying the differential method the adjoint operator, the adjoint equations with the general form of their boundary conditions, and the general form of the bilinear concomitant were calculated. The discretized adjoint equations and the corresponding boundary conditions were programmed and solved by using the so called method of characteristics. As an example, a constant-level tank connected through a pipe to a valve discharging to atmosphere was considered. The bilinear concomitant was calculated for this particular case. The corresponding sensitivity coefficients due to the variation of different parameters by using both the differential method and the response surface generated by the computer code WHAT were also calculated. The results obtained with these methods show excellent agreement. (author). 11 refs, 2 figs, 2 tabs

  7. CFD thermal-hydraulic analysis of a CANDU fuel channel with SEU43 type fuel bundle

    International Nuclear Information System (INIS)

    Catana, A.; Prisecaru, Ilie; Dupleac, D.; Danila, Nicolae

    2009-01-01

    This paper presents the numerical investigation of a CANDU fuel channel using CFD (Computational Fluid Dynamics) methodology approach, when SEU43 fuel bundles are used. Comparisons with STD37 fuel bundles are done in order to evaluate the influence of geometrical differences of the fuel bundle types on fluid flow properties. We adopted a strategy to analyze only the significant segments of fuel channel, namely : - the fuel bundle junctions with adjacent segments; - the fuel bundle spacer planes with adjacent segments; - the fuel bundle segments with turbulence enhancement buttons; - and the regular segments of fuel bundles. The computer code used is an academic version of FLUENT code, available from UPB. The complex flow domain of fuel bundles contained in pressure tube and operating conditions determine a high turbulence flow and in some parts of fuel channel also a multi-phase flow. Numerical simulation of the flow in the fuel channel has been achieved by solving the equations for conservation of mass, momentum and energy. For turbulence model the standard k-model is employed although other turbulence models can be used. In this paper we do not consider heat generation and heat transfer capabilities of CFD methods. Boundary conditions for CFD analysis are provided by system and sub-channel analysis. In this paper the discussion is focused on some flow parameters behaviour at the bundle junction, spacer's plane configuration, etc. of a SEU43 fuel bundle in conditions of a typical CANDU 6 fuel channel starting from some experience gained in a previous work. (authors)

  8. Thermo hydraulic analysis of narrow channel effect in supercritical-pressure light water reactor

    International Nuclear Information System (INIS)

    Zhou Tao; Chen Juan; Cheng Wanxu

    2012-01-01

    Highlights: ► Detailed thermal analysis with different narrow gaps between fuel rods is given. ► Special characteristics of narrow channels effect on heat transfer in supercritical pressure are shown. ► Reasonable size selection of gaps between fuel rods is proposed for SCWR. - Abstract: The size of the gap between fuel rods has important effects on flow and heat transfer in a supercritical-pressure light water reactor. Based on thermal analysis at different coolant flow rates, the reasonable value range of gap size between fuel rods is obtained, for which the maximum cladding temperature safety limits and installation technology are comprehensively considered. Firstly, for a given design flow rate of coolant, thermal hydraulic analysis of supercritical pressure light water reactor with different gap sizes is provided by changing the fuel rod pitch only. The results show that, by means of reducing the gap size between fuel rods, the heat transfer coefficients between coolant and fuel rod, as well as the heat transfer coefficient between coolant and water rod, would both increase noticeably. Furthermore, the maximum cladding temperature will significantly decrease when the moderator temperature is decreased but coolant temperature remains essentially constant. Meanwhile, the reduction in the maximum cladding temperature in the inner assemblies is much larger than that in the outer assemblies. In addition, the maximum cladding temperature could be further reduced by means of increasing coolant flow rate for each gap size. Finally, the characteristics of narrow channels effect are proposed, and the maximum allowable gap between fuel rods is obtained by making full use of the enhancing narrow channels effect on heat transfer, and concurrently considering installation. This could provide a theoretical reference for supercritical-pressure light water reactor design optimization, in which the effects of gap size and flow rate on heat transfer are both considered.

  9. Installation of aerosol behavior model into multi-dimensional thermal hydraulic analysis code AQUA

    International Nuclear Information System (INIS)

    Kisohara, Naoyuki; Yamaguchi, Akira

    1997-12-01

    The safety analysis of FBR plant system for sodium leak phenomena needs to evaluate the deposition of the aerosol particle to the components in the plant, the chemical reaction of aerosol to humidity in the air and the effect of the combustion heat through aerosol to the structural component. For this purpose, ABC-INTG (Aerosol Behavior in Containment-INTeGrated Version) code has been developed and used until now. This code calculates aerosol behavior in the gas area of uniform temperature and pressure by 1 cell-model. Later, however, more detailed calculation of aerosol behavior requires the installation of aerosol model into multi-cell thermal hydraulic analysis code AQUA. AQUA can calculate the carrier gas flow, temperature and the distribution of the aerosol spatial concentration. On the other hand, ABC-INTG can calculate the generation, deposition to the wall and flower, agglomeration of aerosol particle and figure out the distribution of the aerosol particle size. Thus, the combination of these two codes enables to deal with aerosol model coupling the distribution of the aerosol spatial concentration and that of the aerosol particle size. This report describes aerosol behavior model, how to install the aerosol model to AQUA and new subroutine equipped to the code. Furthermore, the test calculations of the simple structural model were executed by this code, appropriate results were obtained. Thus, this code has prospect to predict aerosol behavior by the introduction of coupling analysis with multi-dimensional gas thermo-dynamics for sodium combustion evaluation. (J.P.N.)

  10. TRAC-PF1/MOD1: an advanced best-estimate computer program for pressurized water reactor thermal-hydraulic analysis

    International Nuclear Information System (INIS)

    Liles, D.R.; Mahaffy, J.H.

    1986-07-01

    The Los Alamos National Laboratory is developing the Transient Reactor Analysis Code (TRAC) to provide advanced best-estimate predictions of postulated accidents in light-water reactors. The TRAC-PF1/MOD1 program provides this capability for pressurized water reactors and for many thermal-hydraulic test facilities. The code features either a one- or a three-dimensional treatment of the pressure vessel and its associated internals, a two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field and solute tracking, flow-regime-dependent constitutive equation treatment, optional reflood tracking capability for bottom-flood and falling-film quench fronts, and consistent treatment of entire accident sequences including the generation of consistent initial conditions. The stability-enhancing two-step (SETS) numerical algorithm is used in the one-dimensional hydrodynamics and permits this portion of the fluid dynamics to violate the material Courant condition. This technique permits large time steps and, hence, reduced running time for slow transients

  11. TRAC-PF1/MOD1: an advanced best-estimate computer program for pressurized water reactor thermal-hydraulic analysis

    Energy Technology Data Exchange (ETDEWEB)

    Liles, D.R.; Mahaffy, J.H.

    1986-07-01

    The Los Alamos National Laboratory is developing the Transient Reactor Analysis Code (TRAC) to provide advanced best-estimate predictions of postulated accidents in light-water reactors. The TRAC-PF1/MOD1 program provides this capability for pressurized water reactors and for many thermal-hydraulic test facilities. The code features either a one- or a three-dimensional treatment of the pressure vessel and its associated internals, a two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field and solute tracking, flow-regime-dependent constitutive equation treatment, optional reflood tracking capability for bottom-flood and falling-film quench fronts, and consistent treatment of entire accident sequences including the generation of consistent initial conditions. The stability-enhancing two-step (SETS) numerical algorithm is used in the one-dimensional hydrodynamics and permits this portion of the fluid dynamics to violate the material Courant condition. This technique permits large time steps and, hence, reduced running time for slow transients.

  12. 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

  13. The fluid–solid coupling analysis of screw conveyor in drilling fluid centrifuge based on ANSYS

    Directory of Open Access Journals (Sweden)

    Hongbin Liu

    2015-09-01

    Full Text Available In the centrifugal separations of drilling fluid, screw conveyor is a critical component to push and separate the sediment. The work performance and structural parameters of conveyor are immediately related to the production capability, the working life and the separating effect of the centrifuge. The existing researches always use the theoretical calculation of the approximate loads to analyze the strength of conveyor, and it cannot reflect the stress situations accurately. In order to ensure the precise mastery of the working performance, this article obtained pressure distribution under working conditions from CFX evaluation and gained equivalent stress and deformation under several load conditions by using the ANSYS Workbench platform to check the strength of conveyor. The results showed that the influence of centrifugal hydraulic pressure was less than that of centrifugal force on the strength and deformation of conveyor. Besides, the maximum equivalent stress occurred at the inside of the feed opening, while the maximum deformation occurred at the conveyor blade edge of taper extremity. Furthermore, whether considered the feed opening or not, the computing model had a great influence on the analysis results, and the simplified loads had a great influence on the deformation analysis results. The methods and results from this article can provide reference for the design and the improvement of screw conveyor.

  14. 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

  15. 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

  16. Analysis of thermo-hydraulic behavior of coolant during discharge of pressurized high-temperature water

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Sobajima, Makoto; Sasaki, Shinobu; Onishi, Nobuaki; Shiba, Masayoshi

    1978-01-01

    The present report describes results of the analysis of the LOFT semiscale experiment No. 1011 using remodeled RELAP-3 code, performed at the Idaho National Engineering Laboratory to simulate a postulated loss-of-coolant accident in a pressurized water reactor. It was clarified through the analysis that coolant behavior during blowdown was influenced variously by the system components in the primary loop, comparing with coolant discharge from a pressure vessel. Good agreement was obtained between experimental and analytical results when phase separation was assumed in upper plenum and downcomer, since experimental data indicated existence of liquid level in those parts. It was also found that the use of the Wilson's equation to calculate bubble rise velocity and the use of discharge coefficient as the function of fluid quality at break location to calculate discharge flow rate resulted in good agreement with experimental data. (auth.)

  17. Proceedings of the 11th Thermal and Fluids Analysis Workshop

    Science.gov (United States)

    Sakowski, Barbara

    2002-07-01

    The Eleventh Thermal & Fluids Analysis WorkShop (TFAWS 2000) was held the week of August 21-25 at The Forum in downtown Cleveland. This year's annual event focused on building stronger links between research community and the engineering design/application world and celebrated the theme "Bridging the Gap Between Research and Design". Dr. Simon Ostrach delivered the keynote address "Research for Design (R4D)" and encouraged a more deliberate approach to performing research with near-term engineering design applications in mind. Over 100 persons attended TFAWS 2000, including participants from five different countries. This year's conference devoted a full-day seminar to the discussion of analysis and design tools associated with aeropropulsion research at the Glenn Research Center. As in previous years, the workshop also included hands-on instruction in state-of-the-art analysis tools, paper sessions on selected topics, short courses and application software demonstrations. TFAWS 2000 was co-hosted by the Thermal/Fluids Systems Design and Analysis Branch of NASA GRC and by the Ohio Aerospace Institute and was co-chaired by Barbara A. Sakowski and James R. Yuko. The annual NASA Delegates meeting is a standard component of TFAWS where the civil servants of the various centers represented discuss current and future events which affect the Community of Applied Thermal and Fluid ANalystS (CATFANS). At this year's delegates meeting the following goals (among others) were set by the collective body of delegates participation of all Centers in the NASA material properties database (TPSX) update: (1) developing and collaboratively supporting multi-center proposals; (2) expanding the scope of TFAWS to include other federal laboratories; (3) initiation of a white papers on thermal tools and standards; and (4) formation of an Agency-wide TFAWS steering committee.

  18. 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

  19. Program ELM: A tool for rapid thermal-hydraulic analysis of solid-core nuclear rocket fuel elements

    International Nuclear Information System (INIS)

    Walton, J.T.

    1992-11-01

    This report reviews the state of the art of thermal-hydraulic analysis codes and presents a new code, Program ELM, for analysis of fuel elements. ELM is a concise computational tool for modeling the steady-state thermal-hydraulics of propellant flow through fuel element coolant channels in a nuclear thermal rocket reactor with axial coolant passages. The program was developed as a tool to swiftly evaluate various heat transfer coefficient and friction factor correlations generated for turbulent pipe flow with heat addition which have been used in previous programs. Thus, a consistent comparison of these correlations was performed, as well as a comparison with data from the NRX reactor experiments from the Nuclear Engine for Rocket Vehicle Applications (NERVA) project. This report describes the ELM Program algorithm, input/output, and validation efforts and provides a listing of the code

  20. Coupled fluid-structure method for pressure suppression analysis

    International Nuclear Information System (INIS)

    McMaster, W.H.; Norris, D.M. Jr.; Goudreau, G.L.

    1979-01-01

    We have coupled an incompressible Eulerian hydrodynamic algorithm to a Lagrangian finite-element shell algorithm for the analysis of pressure suppression in boiling water reactors. The computer program calculates loads and structural response from air and steam blowdown and the oscillating condensation of steam bubbles in a water pool. The fluid, structure, and coupling algorithms have been verified by the calculation of solved problems from the literature and from air and steam blowdown experiments. The foundation of the program is the semi-implicit, two-dimensional SOLA algorithm. The shell structure algorithm uses conventional thin-shell theory with transverse shear. The finite-element spatial discretization employs piecewise-linear interpolation functions and one-point quadrature applied to conical frustra. We use the Newmark implicit time-integration method implemented as a one-step module. The algorithms are strongly coupled in the iteration loop using the iterated pressure in the fluid to drive the structure. The coupling algorithm requires normal velocity compatibility at the fluid-structure interface and incompressibility of the computational Eulerian zone overlaid by the structure. This is accomplished by iterating on the pressure field which is applied to the structure during each iteration until both conditions are satisfied

  1. Quantitative Analysis of Karst Conduit Structure Parameters and Hydraulic Parameters Based on Tracer Test

    Science.gov (United States)

    Qiang, Z.; Zhiqiang, Z.; Xu, M.; Jinyu, S.; Jihong, Q.

    2017-12-01

    The Old Town of Lijiang is famous as the world cultural heritage since 1997, while characterized by its ancient buildings and natural scenery, water is the soul of the town. Around Heilongtan Springs, there are a large quantity of springs at the Old Town of Lijiang , which is an important part of the World Cultural Heritage. Heilongtan Springs is 2420m above the sea level, the annual variation of the flow rate varies greatly (0 8042 x 104 m3 / year). Recharge area Jiuzihai depressions is 6km long, 3km wide and 2800m above sea level, as main karst water recharge area karst funnel and the sink hole are developing on this planation surface, in the research area medium to thick layers of limestone made up Beiya formation (T2b) of Triassic system distributed widely, karst is strongly developed and the fissure caves water occurrence. In order to exploring the application of tracer test in karst hydrogeology, a tracer test was conducted from Jiuzihai depressions to Ganze Spring. Based on the hydrogeological conditions in the study area, tracer test was used for analysis of groundwater connectivity and flow field characteristics, quantitative analysis of Tracer Breakthrough Curves (BTC) with code Qtracer2. The results demonstated that there are hydraulic connection between Jiuzihai depressions with Ganze Spring, and there are other karst conduits in this area. The longitudinal dispersivity coefficient is 0.24 m2/s, longitudinal dispersivity is 12.06m, flow-channel volume is 3.08×104 m3, flow-channel surface area is 3.27×107m2, mean diameter is 1.42m, Reynolds number is 25187, Froude number is 0.0061, respectively. The groundwater in this area is in a slow turbulent state. The results are of great significance to understand the law of groundwater migration, establish groundwater quality prediction model and exploit karst water resources effectively.

  2. Thermal-Hydraulic Analysis of SWAMUP Facility Using ATHLET-SC Code

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Zidi; Cao, Zhen; Liu, Xiaojing, E-mail: xiaojingliu@sjtu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, Shanghai (China)

    2015-03-16

    During the loss of coolant accident (LOCA) of supercritical water-cooled reactor (SCWR), the pressure in the reactor system will undergo a rapid decrease from the supercritical pressure to the subcritical condition. This process is called trans-critical transients, which is of crucial importance for the LOCA analysis of SCWR. In order to simulate the trans-critical transient, a number of system codes for SCWR have been developed up to date. However, the validation work for the trans-critical models in these codes is still missing. The test facility Supercritical WAter MUltiPurpose loop (SWAMUP) with 2 × 2 rod bundle in Shanghai Jiao Tong University (SJTU) will be applied to provide test data for code validation. Some pre-test calculations are important and necessary to show the feasibility of the experiment. In this study, trans-critical transient analysis is performed for the SWAMUP facility with the system code ATHLET-SC, which is modified in SJTU, for supercritical water system. This paper presents the system behavior, e.g., system pressure, coolant mass flow, cladding temperature during the depressurization. The effects of some important parameters such as heating power, depressurization rate on the system characteristics are also investigated in this paper. Additionally, some sensitivities study of the code models, e.g., heat transfer coefficient, critical heat flux correlation are analyzed and discussed. The results indicate that the revised system code ATHLET-SC is capable of simulating thermal-hydraulic behavior during the trans-critical transient. According to the results, the cladding temperature during the transient is kept at a low value. However, the pressure difference of the heat exchanger after depressurization could reach 6 MPa, which should be considered in the experiment.

  3. Thermal-hydraulics analysis for advanced fuel to be used in Candu 600 nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Catana, Alexandru [RAAN, Institute for Nuclear Research, Str. Campului Nr. 1, Pitesti, Arges (Romania); Danila, Nicolae; Prisecaru, Ilie; Dupleac, Daniel [University POLITEHNICA of Bucharest (Romania)

    2008-07-01

    Two Candu 600 pressure tube nuclear reactors cover about 17% of Romania's electricity demand. These nuclear reactors are moderated/cooled with D{sub 2}O, fuelled on-power with Natural Uranium (NU) dioxide encapsulated in a standard (STD37) fuel bundle. High neutron economy is achieved using D{sub 2}O as moderator and coolant in separated systems. To reduce fuel cycle costs, programs were initiated in Canada, S.Korea, Argentina and Romania for the design and build new fuel bundles able to accommodate different fuel compositions. Candu core structure and modular fuel bundles, permits flexible fuel cycles. The main expected achievements are: reduced fuel cycle costs, increased discharge burn-up, plutonium and minor actinides management, thorium cycle, use of recycled PWR and in the same time waste minimization and operating cost reduction. These new fuel bundles are to be used in already operated Candu reactors. Advanced fuel bundle were proposed: CANFLEX bundle (Canada, S-Korea); the Romanian 'SEU43' bundle (Fig 1). In this paper thermal-hydraulic analysis in sub-channel approach is presented for SEU43. Comparisons with standard (STD37) fuel bundles are made using SEU-NU for NU fuel composition and SEU-0.96, for recycled uranium (RU) fuel with 0.96% U-235. Extended and comprehensive analysis must be made in order to assess the TH behaviour of SEU43. In this paper, considering STD37, SEU43-NU and SEU43-0.96 fuel bundles, main TH parameters were analysed: pressure drop, fuel highest temperatures, coolant density, critical heat flux. Differences between these fuel types are outlined. Benefits are: fuel costs reduction, spent fuel waste minimization, increase in competitiveness of nuclear power. Safety margins must be, at least, conserved. (authors)

  4. Thermal-hydraulics analysis for advanced fuel to be used in Candu 600 nuclear reactors

    International Nuclear Information System (INIS)

    Catana, Alexandru; Danila, Nicolae; Prisecaru, Ilie; Dupleac, Daniel

    2008-01-01

    Two Candu 600 pressure tube nuclear reactors cover about 17% of Romania's electricity demand. These nuclear reactors are moderated/cooled with D 2 O, fuelled on-power with Natural Uranium (NU) dioxide encapsulated in a standard (STD37) fuel bundle. High neutron economy is achieved using D 2 O as moderator and coolant in separated systems. To reduce fuel cycle costs, programs were initiated in Canada, S.Korea, Argentina and Romania for the design and build new fuel bundles able to accommodate different fuel compositions. Candu core structure and modular fuel bundles, permits flexible fuel cycles. The main expected achievements are: reduced fuel cycle costs, increased discharge burn-up, plutonium and minor actinides management, thorium cycle, use of recycled PWR and in the same time waste minimization and operating cost reduction. These new fuel bundles are to be used in already operated Candu reactors. Advanced fuel bundle were proposed: CANFLEX bundle (Canada, S-Korea); the Romanian 'SEU43' bundle (Fig 1). In this paper thermal-hydraulic analysis in sub-channel approach is presented for SEU43. Comparisons with standard (STD37) fuel bundles are made using SEU-NU for NU fuel composition and SEU-0.96, for recycled uranium (RU) fuel with 0.96% U-235. Extended and comprehensive analysis must be made in order to assess the TH behaviour of SEU43. In this paper, considering STD37, SEU43-NU and SEU43-0.96 fuel bundles, main TH parameters were analysed: pressure drop, fuel highest temperatures, coolant density, critical heat flux. Differences between these fuel types are outlined. Benefits are: fuel costs reduction, spent fuel waste minimization, increase in competitiveness of nuclear power. Safety margins must be, at least, conserved. (authors)

  5. Experimental-based Modelling and Simulation of Water Hydraulic Mechatronics Test Facilities for Motion Control and Operation in Environmental Sensitive Applications` Areas

    DEFF Research Database (Denmark)

    Conrad, Finn; Pobedza, J.; Sobczyk, A.

    2003-01-01

    The paper presents experimental-based modelling, simulation, analysis and design of water hydraulic actuators for motion control of machines, lifts, cranes and robots. The contributions includes results from on-going research projects on fluid power and mechatronics based on tap water hydraulic...

  6. Critical analysis of soil hydraulic conductivity determination using monoenergetic gamma radiation attenuation

    International Nuclear Information System (INIS)

    Portezan Filho, Otavio

    1997-01-01

    Three soil samples of different textures: LVA (red yellow latosol), LVE (dark red latosol) and LRd (dystrophic dark red latosol) were utilized for unsaturated hydraulic conductivity K(θ) measurements. Soil bulk densities and water contents during internal water drainage were measured by monoenergetic gamma radiation attenuation, using homogeneous soil columns assembled in the laboratory. The measurements were made with a collimated gamma beam of 0.003 m in diameter using a Nal(Tl) (3'' x 3 '') detector and a 137 Cs gamma source of 74 X 10 8 Bq and 661.6 KeV. Soil columns were scanned with the gamma beam from 0.01 to 0.20 m depth, in 0.01m steps, for several soil water redistribution times. The results show a great variability of the unsaturated hydraulic conductivity relation K(θ), even though homogeneous soils were used. The variability among methods is significantly smaller in relation to variability in space. The assumption of unit hydraulic gradient during redistribution of soil water utilized in the methods of Hillel, Libardi and Sisson leads to hydraulic conductivity values that increase in depth. The exponential character of the K(θ) relationship, is responsible for the difficulty of estimating soil hydraulic conductivity, which is a consequence of small variations in the porous arrangement, even in samples supposed to be homogeneous. (author)

  7. Parameters Analysis of Hydraulic-Electrical Energy Regenerative Absorber on Suspension Performance

    Directory of Open Access Journals (Sweden)

    Han Zhang

    2014-05-01

    Full Text Available To recycle the vibration energy of vehicles over rough roads, a hydraulic-electricity energy regenerative suspension (HEERS was designed in the present work, and simulations were performed with focus on its performance. On the basis of the system principle, the mathematical model of hydraulic-electrical energy regenerative absorber (HEERA and two degrees of freedom (DOF suspension dynamic model were constructed. Using the model of HEERA, simulations on force-displacement and force-velocity characteristics were performed with a 1.67 Hz frequency and a sinusoidal input adopted. And then in combination with HEERA model and two DOF suspension models, simulations on the performance of HEERS also were carried out. Finally, the influences of charging pressure and volume of the accumulator, hydraulic motor displacement, orifice area of check valve, and inner diameter of hydraulic pipelines on the performance of HEERA and HEERS were investigated in depth. The simulation results indicated that (i the damping characteristic of HEERA was coincident with the damping characteristics of traditional absorber; (ii the most remarkable influencing factor on the performance of HEERS was the hydraulic motor displacement, followed by orifice area of check valve, inner diameter of pipelines, and charging pressure of accumulator, while the effects of charging volume of accumulator were quite limited.

  8. CCP Sensitivity Analysis by Variation of Thermal-Hydraulic Parameters of Wolsong-3, 4

    Energy Technology Data Exchange (ETDEWEB)

    You, Sung Chang [KHNP, Daejeon (Korea, Republic of)

    2016-10-15

    The PHWRs are tendency that ROPT(Regional Overpower Protection Trip) setpoint is decreased with reduction of CCP(Critical Channel Power) due to aging effects. For this reason, Wolsong unit 3 and 4 has been operated less than 100% power due to the result of ROPT setpoint evaluation. Typically CCP for ROPT evaluation is derived at 100% PHTS(Primary Heat Transport System) boundary conditions - inlet header temperature, header to header different pressure and outlet header pressure. Therefore boundary conditions at 100% power were estimated to calculate the thermal-hydraulic model at 100% power condition. Actually thermal-hydraulic boundary condition data for Wolsong-3 and 4 cannot be taken at 100% power condition at aged reactor condition. Therefore, to create a single-phase thermal-hydraulic model with 80% data, the validity of the model was confirmed at 93.8%(W3), 94.2%(W4, in the two-phase). And thermal-hydraulic boundary conditions at 100% power were calculated to use this model. For this reason, the sensitivities by varying thermal-hydraulic parameters for CCP calculation were evaluated for Wolsong unit 3 and 4. For confirming the uncertainties by variation PHTS model, sensitivity calculations were performed by varying of pressure tube roughness, orifice degradation factor and SG fouling factor, etc. In conclusion, sensitivity calculation results were very similar and the linearity was constant.

  9. Stochastic Spectral Analysis for Characterizing Hydraulic Diffusivity in an Alluvial Fan Aquifer with River Stimulus

    Science.gov (United States)

    Wang, Y. L.; Zha, Y.; Yeh, T. C. J.; Wen, J. C.

    2015-12-01

    Estimation of subsurface hydraulic diffusivity was carried out to understand the characteristics of Zhuoshui River alluvial fan, Taiwan. The fan, an important agricultural and industrial region with high water demand, is located at middle Taiwan with an area of 1800 km2. The prior geo-investigations suggest that the main recharge region of the fan is at an apex along the river. The distribution of soil hydraulic diffusivity was estimated by fusing naturally recurring stimulus provided by river and groundwater head. Specifically, the variance and power spectrum provided by temporal and spatial change of groundwater head in response to river stage variations are analyzed to estimate hydraulic diffusivity distribution. It is found that the hydraulic diffusivity of the fan is at the range from 0.08 to 16 m2/s. The average hydraulic diffusivity at the apex, middle, and tail of the fan along the river is about 0.4, 0.6, and 1.0 m2/s, respectively.

  10. VIPRE-01: a thermal-hydraulic analysis code for reactor cores. Volume 3. Programmer's manual. Final report

    International Nuclear Information System (INIS)

    Stewart, C.W.; Koontz, A.S.; Cuta, J.M.; Montgomery, S.D.

    1983-05-01

    VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear-reactor-core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This is Volume 3, the Programmer's Manual. It explains the codes' structures and the computer interfaces

  11. Development of whole core thermal-hydraulic analysis program ACT. 4. Incorporation of three-dimensional upper plenum model

    International Nuclear Information System (INIS)

    Ohshima, Hiroyuki

    2003-03-01

    The thermal-hydraulic analysis computer program ACT is under development for the evaluation of detailed flow and temperature fields in a core region of fast breeder reactors under various operation conditions. The purpose of this program development is to contribute not only to clarifying thermal hydraulic characteristics that cannot be revealed by experiments due to measurement difficulty but also to performing