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Sample records for hydraulic performance analysis

  1. 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.

  2. 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)

  3. 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)

  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. 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.)

  6. 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.

  7. 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

  8. 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.

  9. Advanced Performance Hydraulic Wind Energy

    Science.gov (United States)

    Jones, Jack A.; Bruce, Allan; Lam, Adrienne S.

    2013-01-01

    The Jet Propulsion Laboratory, California Institute of Technology, has developed a novel advanced hydraulic wind energy design, which has up to 23% performance improvement over conventional wind turbine and conventional hydraulic wind energy systems with 5 m/sec winds. It also has significant cost advantages with levelized costs equal to coal (after carbon tax rebate). The design is equally applicable to tidal energy systems and has passed preliminary laboratory proof-of-performance tests, as funded by the Department of Energy.

  10. 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

  11. Thermal-Hydraulic Analysis of a Once-Through Steam Generator Considering Performance Degradation

    Energy Technology Data Exchange (ETDEWEB)

    Han, Hun Sik; Kang, Han Ok; Yoon, Ju Hyeon; Kim, Young In; Song, Jae Seung; Kim, Keung Koo [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    Several countries have entered into a global race for the commercialization of SMRs, and considerable research and development have been implemented. Among the various reactor designs, many SMRs have adopted an integral type pressurized water reactor (PWR) to enhance the nuclear safety and system reliability. In the integral reactor design, a single reactor pressure vessel contains primary system components such as fuel and core, steam generators, pumps, and a pressurizer. For the component integration into a reactor vessel, it is important to design each component as small as possible. Thus, it is a common practice to employ a once-through steam generator in the integral reactor design due to its advantages in compactness. In general, gradual degradation in thermal-hydraulic performance of the steam generator occurs with time, and it changes slowly the operating point of the steam generator during plant lifetime. Numerical solutions are acquired to evaluate the thermal-hydraulic performance of the steam generator at various AUFs. The design results obtained show that the average tube length of the steam generator is augmented with the increase of design margin to compensate for the design uncertainties and heat transfer area reduction by plugging, fouling, etc. A helically coiled tube once-through steam generator with 30% design margin is considered for comparison of thermal-hydraulic performances according to the degradation rate.

  12. Thermal-Hydraulic Analysis of a Once-Through Steam Generator Considering Performance Degradation

    International Nuclear Information System (INIS)

    Han, Hun Sik; Kang, Han Ok; Yoon, Ju Hyeon; Kim, Young In; Song, Jae Seung; Kim, Keung Koo

    2016-01-01

    Several countries have entered into a global race for the commercialization of SMRs, and considerable research and development have been implemented. Among the various reactor designs, many SMRs have adopted an integral type pressurized water reactor (PWR) to enhance the nuclear safety and system reliability. In the integral reactor design, a single reactor pressure vessel contains primary system components such as fuel and core, steam generators, pumps, and a pressurizer. For the component integration into a reactor vessel, it is important to design each component as small as possible. Thus, it is a common practice to employ a once-through steam generator in the integral reactor design due to its advantages in compactness. In general, gradual degradation in thermal-hydraulic performance of the steam generator occurs with time, and it changes slowly the operating point of the steam generator during plant lifetime. Numerical solutions are acquired to evaluate the thermal-hydraulic performance of the steam generator at various AUFs. The design results obtained show that the average tube length of the steam generator is augmented with the increase of design margin to compensate for the design uncertainties and heat transfer area reduction by plugging, fouling, etc. A helically coiled tube once-through steam generator with 30% design margin is considered for comparison of thermal-hydraulic performances according to the degradation rate

  13. 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.

  14. A flexible tool for hydraulic and water quality performance analysis of green infrastructure

    Science.gov (United States)

    Massoudieh, A.; Alikhani, J.

    2017-12-01

    Models that allow for design considerations of green infrastructure (GI) practices to control stormwater runoff and associated contaminants have received considerable attention in recent years. To be used to evaluate the effect design configurations on the long-term performance of GIs, models should be able to consider processes within GIs with good fidelity. In this presentation, a sophisticated, yet flexible tool for hydraulic and water quality assessment of GIs will be introduced. The tool can be used by design engineers and researchers to capture and explore the effect of design factors and properties of the media employed in the performance of GI systems at a relatively small scale. We deemed it essential to have a flexible GI modeling tool that is capable of simulating GI system components and specific biogeochemical processes affecting contaminants such as evapotranspiration, plant uptake, reactions, and particle-associated transport accurately while maintaining a high degree of flexibility to account for the myriad of GI alternatives. The mathematical framework for a stand-alone GI performance assessment tool has been developed and will be demonstrated. The process-based model framework developed here can be used to model a diverse range of GI practices such as stormwater ponds, green roofs, retention ponds, bioretention systems, infiltration trench, permeable pavement and other custom-designed combinatory systems. An example of the application of the system to evaluate the performance of a rain-garden system will be demonstrated.

  15. Coupled neutronics/thermal-hydraulics analysis of a high-performance light-water reactor fuel assembly

    International Nuclear Information System (INIS)

    Waata, C.L.

    2006-07-01

    The use of water at supercritical pressure as coolant and moderator introduces a challenge in the design of a High-Performance Light-Water Reactor (HPLWR) fuel assembly. At supercritical pressure condition (P=25 MPa), the thermal-hydraulics behaviour of water differs strongly from that at sub-critical pressure due to a rapid variation of the thermal-physical properties across the pseudo-critical line. Due of the strong link between the water (moderation) and the neutron spectrum and subsequently the power distribution, a coupling of neutronics and thermal-hydraulics has become a necessity for reactor concepts operating at supercritical pressure condition. The effect of neutron moderation on the local parameters of thermal-hydraulics and vice-verse in a fuel assembly has to be considered for an accurate design analysis. In this study, the Monte Carlo N-Particle code (MCNP) and the sub-channel code STAFAS (Sub-channel Thermal-hydraulics Analysis of a Fuel Assembly under Supercritical conditions) have been coupled for the design analysis of a fuel assembly with supercritical water as coolant and moderator. Both codes are well known for complex geometry modelling. The MCNP code is used for neutronics analyses and for the prediction of power profiles of individual fuel rods. The sub-channel code STAFAS for the thermal-hydraulics analyses takes into account the coolant properties beyond the critical point as well as separate moderator channels. The coupling procedure is realized automatically. MCNP calculates the power distribution in each fuel rod, which is then transferred into STAFAS to obtain the corresponding thermal-hydraulic conditions in each sub-channel. The new thermal-hydraulic conditions are used to generate a new input deck for the next MCNP calculation. This procedure is repeated until a converged state is achieved. The coupled code system was tested on a proposed fuel assembly design of a HPLWR. An under-relaxation was introduced to achieve convergence

  16. COMMIX analysis of four constant flow thermal upramp experiments performed in a thermal hydraulic model of an advanced LMR

    International Nuclear Information System (INIS)

    Yarlagadda, B.S.

    1989-04-01

    The three-dimensional thermal hydraulics computer code COMMIX-1AR was used to analyze four constant flow thermal upramp experiments performed in the thermal hydraulic model of an advanced LMR. An objective of these analyses was the validation of COMMIX-1AR for buoyancy affected flows. The COMMIX calculated temperature histories of some thermocouples in the model were compared with the corresponding measured data. The conclusions of this work are presented. 3 refs., 5 figs

  17. CFD analysis of sludge accumulation and hydraulic performance of a waste stabilization pond.

    Science.gov (United States)

    Alvarado, Andres; Sanchez, Esteban; Durazno, Galo; Vesvikar, Mehul; Nopens, Ingmar

    2012-01-01

    Sludge management in waste stabilization ponds (WSPs) is essential for safeguarding the system performance. Sludge accumulation patterns in WSPs are strongly influenced by the pond hydrodynamics. CFD modeling was applied to study the relation between velocity profiles and sludge deposition during 10 years of operation of the Ucubamba WSP in Cuenca (Ecuador). One tracer experiment was performed and three sludge accumulation scenarios based on bathymetric surveys were simulated. A residence time distribution (RTD) analysis illustrated the decrease of residence times due to sludge deposition. Sludge accumulation rates were calculated. The influence of flow pattern on the sludge deposition was studied, enabling better planning of future pond operation and desludging.

  18. Transient performance analysis of the master cylinder hydraulic system of a 6.3 MN fineblanking press

    Science.gov (United States)

    Yi, Guodong; Li, Jin

    2018-03-01

    The master cylinder hydraulic system is the core component of the fineblanking press that seriously affects the machine performance. A key issue in the design of the master cylinder hydraulic system is dealing with the heavy shock loads in the fineblanking process. In this paper, an equivalent model of the master cylinder hydraulic system is established based on typical process parameters for practical fineblanking; then, the response characteristics of the master cylinder slider to the step changes in the load and control current are analyzed, and lastly, control strategies for the proportional valve are studied based on the impact of the control parameters on the kinetic stability of the slider. The results show that the kinetic stability of the slider is significantly affected by the step change of the control current, while it is slightly affected by the step change of the system load, which can be improved by adjusting the flow rate and opening time of the proportional valve.

  19. Analyses of hydraulic performance of velocity caps

    DEFF Research Database (Denmark)

    Christensen, Erik Damgaard; Degn Eskesen, Mark Chr.; Buhrkall, Jeppe

    2014-01-01

    The hydraulic performance of a velocity cap has been investigated. Velocity caps are often used in connection with offshore intakes. CFD (computational fluid dynamics) examined the flow through the cap openings and further down into the intake pipes. This was combined with dimension analyses...

  20. 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.

  1. 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.

  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. Analysis of the combined effect of hydrogrinding process and inclination angle on hydraulic performance of diesel injection nozzles

    International Nuclear Information System (INIS)

    Salvador, F.J.; Carreres, M.; Jaramillo, D.; Martínez-López, J.

    2015-01-01

    Highlights: • Effect of inclination angle and rounding radius of diesel nozzle holes is explored. • The study starts with experimental tests and is extended by CFD simulations. • A CFD code with a HEM model for two-phase flow and a RANS approach is used. • Differences in flow parameters, cavitation inception and morphology are analysed. • The flow is generally favoured by low inclination angles and high rounding radius. - Abstract: A computational study to investigate the influence of the orifices inclination and the rounding radius at the orifice inlet (consequence of the hydro-erosive grinding process applied after the orifices machining) over the internal nozzle flow is performed in this paper. The study starts with the analysis of experimental results where the mass flow and momentum flux of two nozzles with very different values of these two variables are compared. This analysis shows relatively small differences in terms of mass flow and momentum flux, since the higher losses associated to the higher deflection of the streamlines with a higher inclination of the orifices are counteracted by the higher rounding radius, which favours the flow entrance to the orifice. To explain this experimental outcome, an extensive computational study involving nine geometries that combine different inclination angles and rounding radius is conducted, in order to quantify the influence of both parameters on the flow separately, as well as to assess the potential of their combination. These geometries are compared in terms of discharge coefficient, critical cavitation conditions and effective injection velocity, among others. Results show differences up to 15% in terms of mass flow rate and 8% for the effective injection velocity among the two extreme cases (lowest inclination and highest hydro-erosion level versus the nozzle with the highest inclination and lowest hydro-erosion level). Given the importance of these phenomena on the subsequent mixing and combustion

  4. 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)

  5. 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

  6. Thermal-hydraulic analysis of heat transfer in subchannels of the European high performance supercritical Water-Cooled Reactor for different CFD turbulence models

    Energy Technology Data Exchange (ETDEWEB)

    Castro, Landy Y.; Rojas, Leorlen Y.; Gamez, Abel; Rosales, Jesus; Gonzalez, Daniel; Garcia, Carlos, E-mail: lcastro@instec.cu, E-mail: leored1984@gmail.com, E-mail: agamezgmf@gmail.com, E-mail: jrosales@instec.cu, E-mail: danielgonro@gmail.com, E-mail: cgh@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas (InSTEC), La Habana (Cuba); Oliveira, Carlos Brayner de, E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil); Dominguez, Dany S., E-mail: dsdominguez@gmail.com [Universidade Estadual de Santa Cruz (UESC), Ilheus, BA (Brazil). Pos-Graduacao em Modelagem Computacional

    2015-07-01

    Chosen as one of six Generation‒IV nuclear-reactor concepts, Supercritical Water-cooled Reactors (SCWRs) are expected to have high thermal efficiencies within the range of 45 - 50% owing to the reactor's high pressures and outlet temperatures. In this reactor, the primary water enters the core under supercritical-pressure condition (25 MPa) at a temperature of 280 deg C and leaves it at a temperature of up to 510 deg C. Due to the significant changes in the physical properties of water at supercritical-pressure, the system is susceptible to local temperature, density and power oscillations. The behavior of supercritical water into the core of the SCWR, need to be sufficiently studied. Most of the methods available to predict the effects of the heat transfer phenomena within the pseudocritical region are based on empirical one-directional correlations, which do not capture the multidimensional effects and do not provide accurate results in regions such as the deteriorated heat transfer regime. In this paper, computational fluid dynamics (CFD) analysis was carried out to study the thermal-hydraulic behavior of supercritical water flows in sub-channels of a typical European High Performance Light Water Reactor (HPLWR) fuel assembly using commercial CFD code CFX-14. It was determined the steady-state equilibrium parameters and calculated the temperature and density distributions. A comparative study for different turbulence models were carried out and the obtained results are discussed. (author)

  7. Thermal-hydraulic analysis of heat transfer in subchannels of the European high performance supercritical Water-Cooled Reactor for different CFD turbulence models

    International Nuclear Information System (INIS)

    Castro, Landy Y.; Rojas, Leorlen Y.; Gamez, Abel; Rosales, Jesus; Gonzalez, Daniel; Garcia, Carlos; Oliveira, Carlos Brayner de; Dominguez, Dany S.

    2015-01-01

    Chosen as one of six Generation‒IV nuclear-reactor concepts, Supercritical Water-cooled Reactors (SCWRs) are expected to have high thermal efficiencies within the range of 45 - 50% owing to the reactor's high pressures and outlet temperatures. In this reactor, the primary water enters the core under supercritical-pressure condition (25 MPa) at a temperature of 280 deg C and leaves it at a temperature of up to 510 deg C. Due to the significant changes in the physical properties of water at supercritical-pressure, the system is susceptible to local temperature, density and power oscillations. The behavior of supercritical water into the core of the SCWR, need to be sufficiently studied. Most of the methods available to predict the effects of the heat transfer phenomena within the pseudocritical region are based on empirical one-directional correlations, which do not capture the multidimensional effects and do not provide accurate results in regions such as the deteriorated heat transfer regime. In this paper, computational fluid dynamics (CFD) analysis was carried out to study the thermal-hydraulic behavior of supercritical water flows in sub-channels of a typical European High Performance Light Water Reactor (HPLWR) fuel assembly using commercial CFD code CFX-14. It was determined the steady-state equilibrium parameters and calculated the temperature and density distributions. A comparative study for different turbulence models were carried out and the obtained results are discussed. (author)

  8. Verification Test of Hydraulic Performance for Reactor Coolant Pump

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Jun; Kim, Jae Shin; Ryu, In Wan; Ko, Bok Seong; Song, Keun Myung [Samjin Ind. Co., Seoul (Korea, Republic of)

    2010-01-15

    According to this project, basic design for prototype pump and model pump of reactor coolant pump and test facilities has been completed. Basic design for prototype pump to establish structure, dimension and hydraulic performance has been completed and through primary flow analysis by computational fluid dynamics(CFD), flow characteristics and hydraulic performance have been established. This pump was designed with mixed flow pump having the following design requirements; specific velocity(Ns); 1080.9(rpm{center_dot}m{sup 3}/m{center_dot}m), capacity; 3115m{sup 3}/h, total head ; 26.3m, pump speed; 1710rpm, pump efficiency; 77.0%, Impeller out-diameter; 349mm, motor output; 360kw, design pressure; 17MPaG. The features of the pump are leakage free due to no mechanical seal on the pump shaft which insures reactor's safety and law noise level and low vibration due to no cooling fan on the motor which makes eco-friendly product. Model pump size was reduced to 44% of prototype pump for the verification test for hydraulic performance of reactor coolant pump and was designed with mixed flow pump and canned motor having the following design requirements; specific speed(NS); 1060.9(rpm{center_dot}m{sup 3}/m{center_dot}m), capacity; 539.4m{sup 3}/h, total head; 21.0m, pump speed; 3476rpm, pump efficiency; 72.9%, Impeller out-diameter; 154mm, motor output; 55kw, design pressure; 1.0MPaG. The test facilities were designed for verification test of hydraulic performance suitable for pump performance test, homologous test, NPSH test(cavitation), cost down test and pressure pulsation test of inlet and outlet ports. Test tank was designed with testing capacity enabling up to 2000m{sup 3}/h and design pressure 1.0MPaG. Auxiliary pump was designed with centrifugal pump having capacity; 1100m{sup 3}/h, total head; 42.0m, motor output; 190kw

  9. Thermal-hydraulic performance analysis of a subchannel with square and triangle fuel rod arrangements using the entropy generation approach

    Institute of Scientific and Technical Information of China (English)

    S.Talebi; M.M.Valoujerdi

    2017-01-01

    The present paper discusses entropy generation in fully developed turbulent flows through a subchannel,arranged in square and triangle arrays.Entropy generation is due to contribution of both heat transfer and pressure drop.Our main objective is to study the effect of key parameters such as spacer grid,fuel rod power distribution,Reynolds number Re,dimensionless heat power ω,lengthto-fuel-diameter ratio λ,and pitch-to-diameter ratio ξ on subchannel entropy generation.The analysis explicitly shows the contribution of heat transfer and pressure drop to the total entropy generation.An analytical formulation is introduced to total entropy generation for situations with uniform and sinusoidal rod power distribution.It is concluded that power distribution affects entropy generation.A smoother power profile leads to less entropy generation.The entropy generation of square rod array bundles is more efficient than that of triangular rod arrays,and spacer grids generate more entropy.

  10. 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.

  11. 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.)

  12. 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.

  13. 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

  14. 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

  15. 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.

  16. 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)

  17. 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.

  18. The Influence of Hydraulic Fracturing on Carbon Storage Performance

    Science.gov (United States)

    Fu, Pengcheng; Settgast, Randolph R.; Hao, Yue; Morris, Joseph P.; Ryerson, Frederick J.

    2017-12-01

    Conventional principles of the design and operation of geologic carbon storage (GCS) require injecting CO2 below the caprock fracturing pressure to ensure the integrity of the storage complex. In nonideal storage reservoirs with relatively low permeability, pressure buildup can lead to hydraulic fracturing of the reservoir and caprock. While the GCS community has generally viewed hydraulic fractures as a key risk to storage integrity, a carefully designed stimulation treatment under appropriate geologic conditions could provide improved injectivity while maintaining overall seal integrity. A vertically contained hydraulic fracture, either in the reservoir rock or extending a limited height into the caprock, provides an effective means to access reservoir volume far from the injection well. Employing a fully coupled numerical model of hydraulic fracturing, solid deformation, and matrix fluid flow, we study the enabling conditions, processes, and mechanisms of hydraulic fracturing during CO2 injection. A hydraulic fracture's pressure-limiting behavior dictates that the near-well fluid pressure is only slightly higher than the fracturing pressure of the rock and is insensitive to injection rate and mechanical properties of the formation. Although a fracture contained solely within the reservoir rock with no caprock penetration, would be an ideal scenario, poroelastic principles dictate that sustaining such a fracture could lead to continuously increasing pressure until the caprock fractures. We also investigate the propagation pattern and injection pressure responses of a hydraulic fracture propagating in a caprock subjected to heterogeneous in situ stress. The results have important implications for the use of hydraulic fracturing as a tool for managing storage performance.

  19. 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.

  20. 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.

  1. 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.

  2. Performance and safety of hydraulic turbines

    International Nuclear Information System (INIS)

    Brekke, H

    2010-01-01

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

  3. Development of fuel performance and thermal hydraulic technology

    International Nuclear Information System (INIS)

    Jung, Youn Ho; Song, K. N.; Kim, H. K. and others

    2000-03-01

    Space grid in LWR fuel assembly is a key structural component to support fuel rods and to enhance heat transfer from fuel rod to the coolant. Therefore, the original spacer grid has been developed. In addition, new phenomena in fuel behavior occurs at the high burnup, so that models to analyze those new phenomena were developed. Results of this project can be summarized as follows. - Seven different spacer grid candidates have been invented and submitted for domestic and US patents. Spacer grid test specimen(3x3 array and 5x5 array) were fabricated for each candidate and the mechanical tests were performed. - Basic technologies in the mechanical and thermal hydraulic behavior in the spacer grid development are studied and relevant test facilities were established - Fuel performance analysis models and programs were developed for the high burnup pellet and cladding, and fuel performance data base were compiled - Procedures of fuel characterization and in-/out of-pile tests were prepared - Conceptual design of fuel rod for integral PWR was carried out. (author)

  4. 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.

  5. Thermal-Hydraulic Performance of Scrubbing Nozzle Used for CFVS

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hyun Chul; Lee, Doo Yong; Jung, Woo Young; Lee, Jong Chan; Kim, Gyu Tae [FNC TECH, Yongin (Korea, Republic of)

    2016-05-15

    A Containment Filtered Venting System (CFVS) is the most interested device to mitigate a threat against containment integrity under the severe accident of nuclear power plant by venting with the filtration of the fission products. FNC technology and partners have been developed the self-priming scrubbing nozzle used for the CFVS which is based on the venturi effect. The thermal-hydraulic performances such as passive scrubbing water suction as well as pressure drop across the nozzle have been tested under various thermal-hydraulic conditions. The two types of test section have been built for testing the thermal-hydraulic performance of the self-priming scrubbing nozzle. Through the visualization loop, the liquid suction performance through the slit, pressure drop across the nozzle are measured. The passive water suction flow through the suction slit at the throat is important parameter to define the scrubbing performance of the self-priming scrubbing nozzle. The water suction flow is increased with the increase of the overhead water level at the same inlet gas flow. It is not so much changed with the change of inlet gas flow at the overhead water level.

  6. Performance and efficiency of a hydraulic hybrid powertrain

    Energy Technology Data Exchange (ETDEWEB)

    Karbaschian, Mohammad Ali [Duisburg-Essen Univ. (Germany). Faculty of Engineering

    2012-11-01

    Hydraulic hybrid powertrains are considered to be a promising technology to save energy and reduce emission in specific automotive fields because of their high power density, components lifetime, and long lasting experience in industries compared to electric hybrid powertrains. Within the first part of the paper, a very brief literature survey on hydraulic hybrid vehicle systems (HHVS) and the related dynamical behaviour is given. No specific activities to improve the efficiency of these systems were detected. Related literature with respect to optimization mainly deals with the management of the system's energy flows trying to control the engine operation point and the high pressure in the system. In the second part, a small simulation study is presented. Therefore, hybrid systems are generally assumed as a Multi-Input-Multi-Output (MIMO) system. The effect of key variables (i.e. accumulator size and pressure, pump/motor displacement and efficiency, valve dynamics) on the system is discussed. The results show that the volume displacement of pump and motor, the performance of the engine, and the state of charge of the accumulator are the most important parameters to specify the efficiency and performance of the hydraulic hybrid powertrain. Additionally, a hybrid hydraulic powertrain with an adjustable state of charge accumulator is compared with one whose state of charge is constant. The result shows the improvement of braking performance and fuel savings. The goal is to optimize the parameters of the system based on the simultaneous consideration of the three (or more) variables for a given load profile with respect to given objectives. (orig.)

  7. 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.

  8. Design and Performance Evaluation of an Electro-Hydraulic Camless Engine Valve Actuator for Future Vehicle Applications.

    Science.gov (United States)

    Nam, Kanghyun; Cho, Kwanghyun; Park, Sang-Shin; Choi, Seibum B

    2017-12-18

    This paper details the new design and dynamic simulation of an electro-hydraulic camless engine valve actuator (EH-CEVA) and experimental verification with lift position sensors. In general, camless engine technologies have been known for improving fuel efficiency, enhancing power output, and reducing emissions of internal combustion engines. Electro-hydraulic valve actuators are used to eliminate the camshaft of an existing internal combustion engines and used to control the valve timing and valve duration independently. This paper presents novel electro-hydraulic actuator design, dynamic simulations, and analysis based on design specifications required to satisfy the operation performances. An EH-CEVA has initially been designed and modeled by means of a powerful hydraulic simulation software, AMESim, which is useful for the dynamic simulations and analysis of hydraulic systems. Fundamental functions and performances of the EH-CEVA have been validated through comparisons with experimental results obtained in a prototype test bench.

  9. 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

  10. 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.

  11. 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.

  12. 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.

  13. 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)

  14. 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.

  15. 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

  16. 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 ...

  17. 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)

  18. 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)

  19. Performance of nano-hydraulic turbine utilizing waterfalls

    Energy Technology Data Exchange (ETDEWEB)

    Ikeda, Toshihiko; Iio, Shouichiro; Tatsuno, Kenji [Faculty of Engineering, Shinshu University, 4-17-1 Wakasato, Nagano, 380-8553 (Japan)

    2010-01-15

    The aim of this investigation was to develop an environmentally friendly nano-hydraulic turbine utilizing waterfalls. A model of an impulse type hydraulic turbine constructed and tested with an indoor type waterfall to arrive at an optimum installation condition. Effects of an installation parameter, namely distance between the rotor and the waterfall on the power performance were studied. The flow field around the rotor was examined visually to clarify influences of installation conditions on the flow field. The flow visualization showed differences of flow pattern around the rotor by the change of flow rate and rotational speed of the rotor. From this study it was found that the power performances of the rotor were changed with the distance between the rotor and the waterfalls. The maximum power coefficient of this turbine is approximately 60%. Also, to respond to changes in the waterfall flow rate, we placed a flat plate on the upper side of the rotor to control the water flow direction. As a result, we found that the coefficient of this turbine is increased with the flow rate and power could be obtained even when the flow rate changed by 3.5 times if the plate was placed on the upper side of the rotor. Although the power coefficient decreased when the plate was installed, the power coefficient still is from 53 to 58%. (author)

  20. Performance analysis

    International Nuclear Information System (INIS)

    2008-05-01

    This book introduces energy and resource technology development business with performance analysis, which has business division and definition, analysis of current situation of support, substance of basic plan of national energy, resource technique development, selection of analysis index, result of performance analysis by index, performance result of investigation, analysis and appraisal of energy and resource technology development business in 2007.

  1. Prediction of Hydraulic Performance of a Scaled-Down Model of SMART Reactor Coolant Pump

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Sun Guk; Park, Jin Seok; Yu, Je Yong; Lee, Won Jae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-08-15

    An analysis was conducted to predict the hydraulic performance of a reactor coolant pump (RCP) of SMART at the off-design as well as design points. In order to reduce the analysis time efficiently, a single passage containing an impeller and a diffuser was considered as the computational domain. A stage scheme was used to perform a circumferential averaging of the flux on the impeller-diffuser interface. The pressure difference between the inlet and outlet of the pump was determined and was used to compute the head, efficiency, and break horse power (BHP) of a scaled-down model under conditions of steady-state incompressible flow. The predicted curves of the hydraulic performance of an RCP were similar to the typical characteristic curves of a conventional mixed-flow pump. The complex internal fluid flow of a pump, including the internal recirculation loss due to reverse flow, was observed at a low flow rate.

  2. Project Startup: Evaluating the Performance of Hydraulic Hybrid Refuse Vehicles

    Energy Technology Data Exchange (ETDEWEB)

    2015-09-01

    The Fleet Test and Evaluation Team at the National Renewable Energy Laboratory (NREL) is evaluating the in-service performance of 10 next-generation hydraulic hybrid refuse vehicles (HHVs), 8 previous-generation HHVs, and 8 comparable conventional diesel vehicles operated by Miami-Dade County's Public Works and Waste Management Department in southern Florida. The HHVs under study - Autocar E3 refuse trucks equipped with Parker Hannifin's RunWise Advanced Series Hybrid Drive systems - can recover as much as 70 percent of the energy typically lost during braking and reuse it to power the vehicle. NREL's evaluation will assess the performance of this technology in commercial operation and help Miami-Dade County determine the ideal routes for maximizing the fuel-saving potential of its HHVs.

  3. 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.

  4. 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)

  5. 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.

  6. 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.

  7. Inferred performance of surface hydraulic barriers from landfill operational data

    International Nuclear Information System (INIS)

    Gross, B.A.; Bonaparte, R.; Othman, M.A.

    1997-01-01

    There are few published data on the field performance of surface hydraulic barriers (SHBs) used in waste containment or remediation applications. In contrast, operational data for liner systems used beneath landfills are widely available. These data are frequently collected and reported as a facility permit condition. This paper uses leachate collection system (LCS) and leak detection system (LDS) liquid flow rate and chemical quality data collected from modem landfill double-liner systems to infer the likely hydraulic performance of SHBs. Operational data for over 200 waste management unit liner systems are currently being collected and evaluated by the authors as part of an ongoing research investigation for the United States Environmental Protection Agency (USEPA). The top liner of the double-liner system for the units is either a geomembrane (GMB) alone, geomembrane overlying a geosynthetic clay liner (GMB/GCL), or geomembrane overlying a compacted clay liner (GMB/CCL). In this paper, select data from the USEPA study are used to: (i) infer the likely efficiencies of SHBs incorporating GMBs and overlain by drainage layers; and (ii) evaluate the effectiveness of SHBs in reducing water infiltration into, and drainage from, the underlying waste (i.e., source control). SHB efficiencies are inferred from calculated landfill liner efficiencies and then used to estimate average water percolation rates through SHBs as a function of site average annual rainfall. The effectiveness of SHBs for source control is investigated by comparing LCS liquid flow rates for open and closed landfill cells. The LCS flow rates for closed cells are also compared to the estimated average water percolation rates through SHBs presented in the paper

  8. 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)

  9. 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.

  10. Design, manufacture and performance research of double acting hydraulic press

    OpenAIRE

    Koc, Erdem; Unver, Ertu; Ozturk, Hidayet

    1990-01-01

    This research presents the design and production of a double acting 40 tons capacity hydraulic press. The issues in the design, engineering manufacturing of the hydraulic press are reported specifically on both cylinders generating the same pressure and velocity using a solenoid directional control valve and a flow separating valve. (In Turkish)

  11. 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.

  12. 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...

  13. 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.

  14. 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

  15. 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

  16. 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.

  17. Research of performance prediction to energy on hydraulic turbine

    International Nuclear Information System (INIS)

    Quan, H; Li, R N; Li, Q F; Han, W; Su, Q M

    2012-01-01

    Refer to the low specific speed Francis turbine blade design principle and double-suction pump structure. Then, design a horizontal double-channel hydraulic turbine Francis. Through adding different guide vane airfoil and and no guide vane airfoil on the hydraulic conductivity components to predict hydraulic turbine energy and using Fluent software to numerical simulation that the operating conditions and point. The results show that the blade pressure surface and suction surface pressure is low when the hydraulic turbine installation is added standard positive curvature of the guide vane and modified positive curvature of guide vane. Therefore, the efficiency of energy recovery is low. However, the pressure of negative curvature guide vane and symmetric guide vane added on hydraulic turbine installations is larger than that of the former ones, and it is conducive to working of runner. With the decreasing of guide vane opening, increasing of inlet angle, flow state gets significantly worse. Then, others obvious phenomena are that the reflux and horizontal flow appeared in blade pressure surface. At the same time, the vortex was formed in Leaf Road, leading to the loss of energy. Through analyzing the distribution of pressure, velocity, flow lines of over-current flow in the the back hydraulic conductivity components in above programs we can known that the hydraulic turbine installation added guide vane is more reasonable than without guide vanes, it is conducive to improve efficiency of energy conversion.

  18. 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.

  19. System Design and Performance Test of Hydraulic Intensifier

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyoung Eui; Lee, Gi Chun [Korea Institute of Machinery and Materials, Daejeon (Korea, Republic of); Kim, Jae Hoon [Chungnam National University, Daejeon (Korea, Republic of)

    2010-07-15

    Components such as pressure vessel, hydraulic hose assembly, accumulator, hydraulic cylinder, hydraulic valve, pipe, etc., are tested under the impulse-pressure conditions prescribed in ISO and SAE standards. The impulse pressure test machine needs to have a high pressure, a precise control system and a long life. It should satisfy the requirements for fabrication of the impulse tester to generate ultra high pressure in the hydraulic system. In the impulse tester, a servo-valve control system is adopted; although the control application is convenient, it is expensive owing to the cost of developing the system. The type of the control system determines the pressure wave, which affects the components that are tested. In this study, the manufacturing process and the intensifier system design related to the flow, pressure, and the increasing rate of pressure are investigated. The results indicate the ultra high pressure waves in the system.

  20. Performance of a hydraulic air compressor for use in compressed air energy storage power systems

    Energy Technology Data Exchange (ETDEWEB)

    Berghmans, J. A.; Ahrens, F. W.

    1978-01-01

    A fluid mechanical analysis of a hydraulic air compression system for Compressed Air Energy Storage (CAES) application is presented. With this compression concept, air is charged into an underground reservoir, for later use in power generation, by entraining bubbles into a downward flow of water from a surface reservoir. Upon releasing the air in the underground reservoir, the water is pumped back to the surface. The analytical model delineated is used to predict the hydraulic compressor performance characteristics (pumping power, pump head, compression efficiency) as a function of water flow rate and system geometrical parameters. The results indicate that, although large water pumps are needed, efficiencies as high as 90% (relative to ideal isothermal compression) can be expected. This should result in lower compression power than for conventional compressor systems, while eliminating the need for the usual intercoolers and aftercooler.

  1. Hydraulic performance of compacted clay liners under simulated daily thermal cycles.

    Science.gov (United States)

    Aldaeef, A A; Rayhani, M T

    2015-10-01

    Compacted clay liners (CCLs) are commonly used as hydraulic barriers in several landfill applications to isolate contaminants from the surrounding environment and minimize the escape of leachate from the landfill. Prior to waste placement in landfills, CCLs are often exposed to temperature fluctuations which can affect the hydraulic performance of the liner. Experimental research was carried out to evaluate the effects of daily thermal cycles on the hydraulic performance of CCLs under simulated landfill conditions. Hydraulic conductivity tests were conducted on different soil specimens after being exposed to various thermal and dehydration cycles. An increase in the CCL hydraulic conductivity of up to one order of magnitude was recorded after 30 thermal cycles for soils with low plasticity index (PI = 9.5%). However, medium (PI = 25%) and high (PI = 37.2%) plasticity soils did not show significant hydraulic deviation due to their self-healing potential. Overlaying the CCL with a cover layer minimized the effects of daily thermal cycles, and maintained stable hydraulic performance in the CCLs even after exposure to 60 thermal cycles. Wet-dry cycles had a significant impact on the hydraulic aspect of low plasticity CCLs. However, medium and high plasticity CCLs maintained constant hydraulic performance throughout the test intervals. The study underscores the importance of protecting the CCL from exposure to atmosphere through covering it by a layer of geomembrane or an interim soil layer. Copyright © 2015 Elsevier Ltd. All rights reserved.

  2. Performance optimization of grooved slippers for aero hydraulic pumps

    Directory of Open Access Journals (Sweden)

    Juan Chen

    2016-06-01

    Full Text Available A computational fluid dynamics (CFD simulation method based on 3-D Navier–Stokes equation and Arbitrary Lagrangian–Eulerian (ALE method is presented to analyze the grooved slipper performance of piston pump. The moving domain of grooved slipper is transformed into a fixed reference domain by the ALE method, which makes it convenient to take the effects of rotate speed, body force, temperature, and oil viscosity into account. A geometric model to express the complex structure, which covers the orifice of piston and slipper, vented groove and the oil film, is constructed. Corresponding to different oil film thicknesses calculated in light of hydrostatic equilibrium theory and boundary conditions, a set of simulations is conducted in COMSOL to analyze the pump characteristics and effects of geometry (groove width and radius, orifice size on these characteristics. Furthermore, the mechanics and hydraulics analyses are employed to validate the CFD model, and there is an excellent agreement between simulation and analytical results. The simulation results show that the sealing land radius, orifice size and groove width all dramatically affect the slipper behavior, and an optimum tradeoff among these factors is conducive to optimizing the pump design.

  3. NREL Evaluates Performance of Hydraulic Hybrid Refuse Vehicles

    Energy Technology Data Exchange (ETDEWEB)

    2015-09-01

    This highlight describes NREL's evaluation of the in-service performance of 10 next-generation hydraulic hybrid refuse vehicles (HHVs), 8 previous-generation (model year 2013) HHVs, and 8 comparable conventional diesel vehicles operated by Miami-Dade County's Public Works and Waste Management Department in southern Florida. Launched in March 2015, the on-road portion of this 12-month evaluation focuses on collecting and analyzing vehicle performance data - fuel economy, maintenance costs, and drive cycles - from the HHVs and the conventional diesel vehicles. The fuel economy of heavy-duty vehicles, such as refuse trucks, is largely dependent on the load carried and the drive cycles on which they operate. In the right applications, HHVs offer a potential fuel-cost advantage over their conventional counterparts. This advantage is contingent, however, on driving behavior and drive cycles with high kinetic intensity that take advantage of regenerative braking. NREL's evaluation will assess the performance of this technology in commercial operation and help Miami-Dade County determine the ideal routes for maximizing the fuel-saving potential of its HHVs. Based on the field data, NREL will develop a validated vehicle model using the Future Automotive Systems Technology Simulator, also known as FASTSim, to study the impacts of route selection and other vehicle parameters. NREL is also analyzing fueling and maintenance data to support total-cost-of-ownership estimations and forecasts. The study aims to improve understanding of the overall usage and effectiveness of HHVs in refuse operation compared to similar conventional vehicles and to provide unbiased technical information to interested stakeholders.

  4. 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)

  5. Field investigation on consumer behavior and hydraulic performance of a district heating system in Tianjin, China

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Baoping; Fu, Lin; Di, Hongfa [Department of Building Science, Tsinghua University, Beijing 100084 (China)

    2009-02-15

    With the implementation of heat reforms in China, the application of thermostatic radiator valves (TRVs) has been gaining popularity in the new-style district heating systems (DHSs). The objective of this study was to investigate consumer behavior (including regulation of TRVs and opening of windows) and its influences on the hydraulic performance and energy consumption of individuals and the whole system. The concurrence rate of individual behaviors and hydraulic interactions between individuals were analyzed. This study should be helpful to gain a comprehensive understanding of the new DHSs in China and consider a proper design/control strategy for these systems. Questionnaires and field observations of consumer behavior, tests of hydraulic performance, and surveys of energy consumption were carried out in a DHS in Tianjin, which was one of the heat metering and billing demonstration projects in China. The main results of the tests were as follows: water flow performance in apartment-level heating systems were diverse because consumers' behavior was varied and unpredictable, and the hydraulic interaction between consumers living along the line of a vertical pipe was obvious, and was stronger for terminal consumers with their TRVs set to higher values; however, flow variations in the whole DHS, which included 910 households, were relatively constant. A probability analysis was carried out to explain this phenomenon, and the conclusion was drawn that when there were more than 200 consumers, the stochastic consumer regulation behavior would bring less than 10% of total flow variations. Finally, the power consumption of the circulation pump, heat consumption and energy-saving potential of this type of DHS were discussed and some suggestions for TRV regulation and pump operation were made. (author)

  6. Hydraulic Performance of Lined Permeable Pavement Systems in the Built Environment

    Directory of Open Access Journals (Sweden)

    Jan Støvring

    2018-05-01

    Full Text Available The hydraulic performance of permeable pavement (PP systems has been well demonstrated when based on full or partial on-site infiltration, while there is only limited research on lined PP systems built to provide detention and volume reduction by evaporation only. In this study, we tested the performance of commercially available PP components when constructed as lined PP systems with un-throttled discharge to explore basic hydraulic function in a real-life-setting. Four types of PP surface products and three types of sub-base aggregates were tested in six unique combinations, built as side-by-side parking lots into an existing parking area, each stall having a size of 25 m2 and 0.5 m of depth with individual lining. Based on 12 months of monitoring precipitation and discharge from each stall, total volume reduction ranged from 3% to 37%. Analysis of up to 22 single events, representing return periods of up to two years, revealed marked detention capacities, expressed as median volume reduction of 40%, spanning 27–69% and median lag time of 1:38 h, spanning 0:39–3:16 h, across all stalls. The considerable range in hydraulic properties can be ascribed to both surface and sub-base properties.

  7. 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.

  8. Robust Control Analysis of Hydraulic Turbine Speed

    Science.gov (United States)

    Jekan, P.; Subramani, C.

    2018-04-01

    An effective control strategy for the hydro-turbine governor in time scenario is adjective for this paper. Considering the complex dynamic characteristic and the uncertainty of the hydro-turbine governor model and taking the static and dynamic performance of the governing system as the ultimate goal, the designed logic combined the classical PID control theory with artificial intelligence used to obtain the desired output. The used controller will be a variable control techniques, therefore, its parameters can be adaptively adjusted according to the information about the control error signal.

  9. 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

  10. 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.

  11. 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

  12. 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.

  13. 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

  14. Experimental Analysis of the Thermo-Hydraulic Performance on a Cylindrical Parabolic Concentrating Solar Water Heater with Twisted Tape Inserts in an Absorber Tube

    Science.gov (United States)

    Kumar, Birendra; Nayak, Rajen Kumar; Singh, S. N.

    2018-05-01

    A twisted tape inserted in an absorber tube may be an excellent option to enhance the performance of a cylindrical parabolic concentrating solar collector (CPC). The present work is an experimental study of the flow and heat transfer with and without twisted tape inserts in the absorber tube of a CPC. Results are presented for mass flow rates of water, ṁ=0.0198-0.0525 kg/s, twist ratio, y=5-10 and Reynolds number, Re=2577.46-6785.55. In the present study, we found that the outlet water temperature, collector efficiency and Nusselt number (Nu) are higher in the twisted tapes as compared to those without the twisted tape inserts in the absorber tube of the CPC. For fixed mass flow rate of water ṁ, the To and η increased with the decrease in twist ratio, y, and is higher in lower twist ratio, y=5, of the twisted tapes. The whole experiment was performed at the Indian Institute of Technology (ISM) in Dhanbad, India during the months of March-April 2017. Based on the experimental data, the correlations for the Nu and friction factor were also developed.

  15. 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

  16. 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

  17. Assessing the monitoring performance using a synthetic microseismic catalogue for hydraulic fracturing

    Science.gov (United States)

    Ángel López Comino, José; Kriegerowski, Marius; Cesca, Simone; Dahm, Torsten; Mirek, Janusz; Lasocki, Stanislaw

    2016-04-01

    Hydraulic fracturing is considered among the human operations which could induce or trigger seismicity or microseismic activity. The influence of hydraulic fracturing operations is typically expected in terms of weak magnitude events. However, the sensitivity of the rock mass to trigger seismicity varies significantly for different sites and cannot be easily predicted prior to operations. In order to assess the sensitivity of microseismity to hydraulic fracturing operations, we perform a seismic monitoring at a shale gas exploration/exploitation site in the central-western part of the Peribaltic synclise at Pomerania (Poland). The monitoring will be continued before, during and after the termination of hydraulic fracturing operations. The fracking operations are planned in April 2016 at a depth 4000 m. A specific network setup has been installed since summer 2015, including a distributed network of broadband stations and three small-scale arrays. The network covers a region of 60 km2. The aperture of small scale arrays is between 450 and 950 m. So far no fracturing operations have been performed, but seismic data can already be used to assess the seismic noise and background microseismicity, and to investigate and assess the detection performance of our monitoring setup. Here we adopt a recently developed tool to generate a synthetic catalogue and waveform dataset, which realistically account for the expected microseismicity. Synthetic waveforms are generated for a local crustal model, considering a realistic distribution of hypocenters, magnitudes, moment tensors, and source durations. Noise free synthetic seismograms are superposed to real noise traces, to reproduce true monitoring conditions at the different station locations. We estimate the detection probability for different magnitudes, source-receiver distances, and noise conditions. This information is used to estimate the magnitude of completeness at the depth of the hydraulic fracturing horizontal wells

  18. Staged cost optimization of urban storm drainage systems based on hydraulic performance in a changing environment

    Directory of Open Access Journals (Sweden)

    M. Maharjan

    2009-04-01

    Full Text Available Urban flooding causes large economic losses, property damage and loss of lives. The impact of environmental changes, mainly urbanization and climatic change, leads to increased runoff and peak flows which the drainage system must be able to cope with to reduce potential damage and inconvenience. Allowing for detention storage to compliment the conveyance capacity of the drainage system network is one of the approaches to reduce urban floods. Contemporary practice is to design systems against stationary environmental forcings – including design rainfall, landuse, etc. Due to the rapid change in the climate- and the urban environment, this approach is no longer appropriate, and explicit consideration of gradual changes during the life-time of the drainage system is warranted. In this paper, a staged cost optimization tool based on the hydraulic performance of the drainage system is presented. A one dimensional hydraulic model is used for hydraulic evaluation of the network together with a genetic algorithm based optimization tool to determine optimal intervention timings and responses over the analysis period. The model was applied in a case study area in the city of Porto Alegre, Brazil. It was concluded that considerable financial savings and/or additional level of flood-safety can be achieved by approaching the design problem as a staged plan rather than one-off scheme.

  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. 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

  1. Nuclear power plant thermal-hydraulic performance research program plan

    International Nuclear Information System (INIS)

    1988-07-01

    The purpose of this program plan is to present a more detailed description of the thermal-hydraulic research program than that provided in the NRC Five-Year Plan so that the research plan and objectives can be better understood and evaluated by the offices concerned. The plan is prepared by the Office of Nuclear Regulatory Research (RES) with input from the Office of Nuclear Reactor Regulation (NRR) and updated periodically. The plan covers the research sponsored by the Reactor and Plant Systems Branch and defines the major issues (related to thermal-hydraulic behavior in nuclear power plants) the NRC is seeking to resolve and provides plans for their resolution; relates the proposed research to these issues; defines the products needed to resolve these issues; provides a context that shows both the historical perspective and the relationship of individual projects to the overall objectives; and defines major interfaces with other disciplines (e.g., structural, risk, human factors, accident management, severe accident) needed for total resolution of some issues. This plan addresses the types of thermal-hydraulic transients that are normally considered in the regulatory process of licensing the current generation of light water reactors. This process is influenced by the regulatory requirements imposed by NRC and the consequent need for technical information that is supplied by RES through its contractors. Thus, most contractor programmatic work is administered by RES. Regulatory requirements involve the normal review of industry analyses of design basis accidents, as well as the understanding of abnormal occurrences in operating reactors. Since such transients often involve complex thermal-hydraulic interactions, a well-planned thermal-hydraulic research plan is needed

  2. 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

  3. Hydraulic performance of a multistage array of advanced centrifugal contactors

    International Nuclear Information System (INIS)

    Hodges, M.E.

    1984-01-01

    The hydraulic characteristics of an advanced design centrifugal contactor array have been determined at the Savannah River Laboratory (SRL). The advanced design utilizes couette mixing (Taylor vortices) in the annulus between the rotating and stationary bowls. Excellent phase separation over a wide range of flow conditions was obtained. Interfaces within an entire eight-stage array were controlled with a single weir air pressure. 2 references, 5 figures

  4. Field study comparing the effect of hydraulic mixing on septic tank performance and sludge accumulation.

    Science.gov (United States)

    Almomani, Fares

    2016-01-01

    This study investigates the effect of hydraulic mixing on anaerobic digestion and sludge accumulation in a septic tank. The performance of a septic tank equipped with a hydraulic mixer was compared with that of a similar standard septic tank over a period of 10 months. The study was conducted in two phases: Phase-I--from May to November 2013 (6 months); Phase-II--from January to May 2014 (4 months). Hydraulic mixing effectively reduced the effluent biological oxygen demand (BOD) and total suspended solids, and reduced the sludge accumulation rate in the septic tank. The BOD removal efficiencies during Phase-II were 65% and 75% in the standard septic tank and a septic tank equipped with hydraulic mixer (Smart Digester™), respectively. The effect of hydraulic mixing reduced the rate of sludge accumulation from 0.64 cm/day to 0.27 cm/day, and increased the pump-out interval by a factor of 3.

  5. 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.

  6. 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.

  7. 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

  8. The quest for performance-related specifications for hydraulic cement concrete.

    Science.gov (United States)

    1982-01-01

    This paper reviews some of the problems associated with quality assurance for hydraulic cement concrete and the difficulties of relating the results of quality control and acceptance testing to the performance of the concrete facility. The importance...

  9. 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...

  10. Impact of water temperature and structural parameters on the hydraulic labyrinth-channel emitter performance

    Directory of Open Access Journals (Sweden)

    Ahmed I. Al-Amoud

    2014-06-01

    Full Text Available The effects of water temperature and structural parameters of a labyrinth emitter on drip irrigation hydraulic performance were investigated. The inside structural parameters of the trapezoidal labyrinth emitter include path width (W and length (L, trapezoidal unit numbers (N, height (H, and spacing (S. Laboratory experiments were conducted using five different types of labyrinth-channel emitters (three non-pressure compensating and two pressure-compensating emitters commonly used for subsurface drip irrigation systems. The water temperature effect on the hydraulic characteristics at various operating pressures was recorded and a comparison was made to identify the most effective structural parameter on emitter performance. The pressure compensating emitter flow exponent (x average was 0.014, while non-pressure compensating emitter’s values average was 0.456, indicating that the sensitivity of non-pressure compensating emitters to pressure variation is an obvious characteristic (p<0.001 of this type of emitters. The effects of water temperature on emitter flow rate were insignificant (p>0.05 at various operating pressures, where the flow rate index values for emitters were around one. The effects of water temperature on manufacturer’s coefficient of variation (CV values for all emitters were insignificant (p>0.05. The CV values of the non-pressure compensating emitters were lower than those of pressure compensating emitters. This is typical for most compensating models because they are manufactured with more elements than non-compensating emitters are. The results of regression analysis indicate that N and H are the essential factors (p<0.001 to affect the hydraulic performance.

  11. Comparisons of Hydraulic Performance in Permanent Maglev Pump for Water-Jet Propulsion

    Directory of Open Access Journals (Sweden)

    Puyu Cao

    2014-08-01

    Full Text Available The operation of water-jet propulsion can generate nonuniform inflow that may be detrimental to the performance of the water-jets. To reduce disadvantages of the nonuniform inflow, a rim-driven water-jet propulsion was designed depending on the technology of passive magnetic levitation. Insufficient understanding of large performance deviations between the normal water-jets (shaft and permanent maglev water-jets (shaftless is a major problem in this paper. CFD was directly adopted in the feasibility and superiority of permanent maglev water-jets. Comparison and discussion of the hydraulic performance were carried out. The shaftless duct firstly has a drop in hydraulic losses (K1, since it effectively avoids the formation and evolution of the instability secondary vortex by the normalized helicity analysis. Then, the shaftless intake duct improves the inflow field of the water-jet pump, with consequencing the drop in the backflow and blocking on the blade shroud. So that the shaftless water-jet pump delivers higher flow rate and head to the propulsion than the shaft. Eventually, not only can the shaftless model increase the thrust and efficiency, but it has the ability to extend the working range and broaden the high efficiency region as well.

  12. 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

  13. 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 ...

  14. Neutronic and Thermal-hydraulic Modelling of High Performance Light Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Seppaelae, Malla [VTT Technical Research Centre of Finland, P.O.Box 1000, FI02044 VTT (Finland)

    2008-07-01

    High Performance Light Water Reactor (HPLWR), which is studied in EU project 'HPLWR2', uses water at supercritical pressures as coolant and moderator to achieve higher core outlet temperature and thus higher efficiency compared to present reactors. At VTT Technical Research Centre of Finland, functionality of the thermal-hydraulics in the coupled reactor dynamics code TRAB3D/ SMABRE was extended to supercritical pressures for the analyses of HPLWR. Input models for neutronics and thermal-hydraulics were made for TRAB3D/ SMABRE according to the latest HPLWR design. A preliminary analysis was performed in which the capability of SMABRE in the transition from supercritical pressures to subcritical pressures was demonstrated. Parameterized two-group cross sections for TRAB3D neutronics were received from Hungarian Academy of Sciences KFKI Atomic Energy Research Institute together with a subroutine for handling them. PSG, a new Monte Carlo transport code developed at VTT, was also used to generate two-group constants for HPLWR and comparisons were made with the KFKI cross sections and MCNP calculations. (author)

  15. Neutronic and Thermal-hydraulic Modelling of High Performance Light Water Reactor

    International Nuclear Information System (INIS)

    Seppaelae, Malla

    2008-01-01

    High Performance Light Water Reactor (HPLWR), which is studied in EU project 'HPLWR2', uses water at supercritical pressures as coolant and moderator to achieve higher core outlet temperature and thus higher efficiency compared to present reactors. At VTT Technical Research Centre of Finland, functionality of the thermal-hydraulics in the coupled reactor dynamics code TRAB3D/ SMABRE was extended to supercritical pressures for the analyses of HPLWR. Input models for neutronics and thermal-hydraulics were made for TRAB3D/ SMABRE according to the latest HPLWR design. A preliminary analysis was performed in which the capability of SMABRE in the transition from supercritical pressures to subcritical pressures was demonstrated. Parameterized two-group cross sections for TRAB3D neutronics were received from Hungarian Academy of Sciences KFKI Atomic Energy Research Institute together with a subroutine for handling them. PSG, a new Monte Carlo transport code developed at VTT, was also used to generate two-group constants for HPLWR and comparisons were made with the KFKI cross sections and MCNP calculations. (author)

  16. Hydraulic performance of Compacted Clay Liners (CCLs) under combined temperature and leachate exposures.

    Science.gov (United States)

    Aldaeef, A A; Rayhani, M T

    2014-12-01

    Experimental investigations were carried out to investigate the effect of thermo-chemical exposures on the hydraulic performance of Compacted Clay Liners (CCLs) in landfills. Hydraulic conductivity of most CCL specimens was increased by two to three times their initial values when exposed to 55 °C for 75 days. CCL specimens also experienced increases in their hydraulic conductivities when exposed to leachate at room temperature. This behaviour could be due to the decrease in viscosity when the permeant was changed from tap water to leachate. However, as the leachate exposure time exceeded the first 15 days, hydraulic conductivity readings decreased to as much as one order of magnitude after 75 days of leachate permeation at room temperature. The gradual decrease in the CCLs hydraulic conductivities was most likely due to chemical precipitation and clogging of pore voids within the soils which seemed to reduce the effective pore volume. The rate of hydraulic conductivity reduction due to leachate permeation was slower at higher temperatures, which was attributed to the lower permeant viscosity and lower clogging occurrence. The observed hydraulic behaviours were correlated to the physical, mineral, and chemical properties of the CCLs and described below. Copyright © 2014 Elsevier Ltd. All rights reserved.

  17. 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)

  18. 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.

  19. 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.

  20. 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)

  1. 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.

  2. 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)

  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. 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)

  5. Influence of hummocks and emergent vegetation on hydraulic performance in a surface flow wastewater treatment wetland

    Science.gov (United States)

    Keefe, Steffanie H.; Daniels, Joan S.; Runkel, Robert L.; Wass, Roland D.; Stiles, Eric A.; Barber, Larry B.

    2010-01-01

    A series of tracer experiments were conducted biannually at the start and end of the vegetation growing season in a surface flow wastewater treatment wetland located near Phoenix, AZ. Tracer experiments were conducted prior to and following reconfiguration and replanting of a 1.2 ha treatment wetland from its original design of alternating shallow and deep zones to incorporate hummocks (shallow planting beds situated perpendicular to flow). Tracer test data were analyzed using analysis of moments and the one‐dimensional transport with inflow and storage numerical model to evaluate the effects of the seasonal vegetation growth cycle and hummocks on solute transport. Following reconfiguration, vegetation coverage was relatively small, and minor changes in spatial distribution influenced wetland hydraulics. During start‐up conditions, the wetland underwent an acclimation period characterized by small vegetation coverage and large transport cross‐sectional areas. At the start of the growing season, new growth of emergent vegetation enhanced hydraulic performance. At the end of the growing season, senescing vegetation created short‐circuiting. Wetland hydrodynamics were associated with high volumetric efficiencies and velocity heterogeneities. The hummock design resulted in breakthrough curves characterized by multiple secondary tracer peaks indicative of varied flow paths created by bottom topography.

  6. 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

  7. 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...

  8. 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.

  9. 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

  10. 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.

  11. Hydraulic performance numerical simulation of high specific speed mixed-flow pump based on quasi three-dimensional hydraulic design method

    International Nuclear Information System (INIS)

    Zhang, Y X; Su, M; Hou, H C; Song, P F

    2013-01-01

    This research adopts the quasi three-dimensional hydraulic design method for the impeller of high specific speed mixed-flow pump to achieve the purpose of verifying the hydraulic design method and improving hydraulic performance. Based on the two families of stream surface theory, the direct problem is completed when the meridional flow field of impeller is obtained by employing iterative calculation to settle the continuity and momentum equation of fluid. The inverse problem is completed by using the meridional flow field calculated in the direct problem. After several iterations of the direct and inverse problem, the shape of impeller and flow field information can be obtained finally when the result of iteration satisfies the convergent criteria. Subsequently the internal flow field of the designed pump are simulated by using RANS equations with RNG k-ε two-equation turbulence model. The static pressure and streamline distributions at the symmetrical cross-section, the vector velocity distribution around blades and the reflux phenomenon are analyzed. The numerical results show that the quasi three-dimensional hydraulic design method for high specific speed mixed-flow pump improves the hydraulic performance and reveal main characteristics of the internal flow of mixed-flow pump as well as provide basis for judging the rationality of the hydraulic design, improvement and optimization of hydraulic model

  12. 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.

  13. 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)

  14. A siphon well model for hydraulic performance optimization and bubble elimination

    Energy Technology Data Exchange (ETDEWEB)

    Fu, Hui, E-mail: fuhui_iwhr@126.com; Ji, Ping; Xia, Qingfu; Guo, Xinlei

    2017-01-15

    Highlights: • A new method was proposed to improve the hydraulic performance and bubble elimination. • The diversion pier and diversion grid were used to stabilize the flow pattern. • Double multi-hole orifices were arranged after the weir. • The new method has a simpler construction and greater bubble elimination. - Abstract: In coastal nuclear power plants, bubble entrainment at the hydraulic jump in the siphon well causes foam pollution and salt fog erosion near the outfall of the siphon well. Thus, bubble elimination in siphon wells has been a topic of considerable interest. This study presents a new hydraulic performance optimization and bubble elimination method based on model experiments. Compared to previous methods, the new method has a simple structure, is effective in eliminating bubbles and is well adapted to different tide levels. The method mainly uses a diversion pier, diversion grid and multi-hole orifices to improve the hydraulic performance, thus reducing bubble entrainment at the hydraulic jump and shortening the bubble movement length in the siphon well. This study provides a valuable reference for the future siphon well design of coastal power plants.

  15. A siphon well model for hydraulic performance optimization and bubble elimination

    International Nuclear Information System (INIS)

    Fu, Hui; Ji, Ping; Xia, Qingfu; Guo, Xinlei

    2017-01-01

    Highlights: • A new method was proposed to improve the hydraulic performance and bubble elimination. • The diversion pier and diversion grid were used to stabilize the flow pattern. • Double multi-hole orifices were arranged after the weir. • The new method has a simpler construction and greater bubble elimination. - Abstract: In coastal nuclear power plants, bubble entrainment at the hydraulic jump in the siphon well causes foam pollution and salt fog erosion near the outfall of the siphon well. Thus, bubble elimination in siphon wells has been a topic of considerable interest. This study presents a new hydraulic performance optimization and bubble elimination method based on model experiments. Compared to previous methods, the new method has a simple structure, is effective in eliminating bubbles and is well adapted to different tide levels. The method mainly uses a diversion pier, diversion grid and multi-hole orifices to improve the hydraulic performance, thus reducing bubble entrainment at the hydraulic jump and shortening the bubble movement length in the siphon well. This study provides a valuable reference for the future siphon well design of coastal power plants.

  16. 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.

  17. 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

  18. 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)

  19. Numerical investigation on thermal-hydraulic performance of new printed circuit heat exchanger model

    International Nuclear Information System (INIS)

    Kim, Dong Eok; Kim, Moo Hwan; Cha, Jae Eun; Kim, Seong O.

    2008-01-01

    Three-dimensional numerical analysis was performed to investigate heat transfer and pressure drop characteristics of supercritical CO 2 flow in new Printed Circuit Heat Exchanger (PCHE) model using commercial CFD code, Fluent 6.3. First, numerical analysis for conventional zigzag channel PCHE model was performed and compared with previous experimental data. Maximum deviation of in-outlet temperature difference and pressure drop from experimental data is about 10%. A new PCHE model has been designed to optimize thermal-hydraulic performance of PCHE. The new PCHE model has several airfoil shape fins (NACA 0020 model), which are designed to streamlined shape. Simulation results showed that in the airfoil shape fin PCHE, total heat transfer rate per unit volume was almost same with zigzag channel PCHE and the pressure drop was reduced to one-twentieth of that in zigzag channel PCHE. In airfoil shape fin PCHE model, the enhancement of heat transfer area and the uniform flow configuration contributed to obtain the same heat transfer performance with zigzag channel PCHE model. And the reduction of pressure drop in airfoil shape fin PCHE model was caused by suppressing generation of separated flow owing to streamlined shape of airfoil fins

  20. 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

  1. Determination of the performance of the Kaplan hydraulic turbines through simplified procedure

    Science.gov (United States)

    Pădureanu, I.; Jurcu, M.; Campian, C. V.; Haţiegan, C.

    2018-01-01

    A simplified procedure has been developed, compared to the complex one recommended by IEC 60041 (i.e. index samples), for measurement of the performance of the hydraulic turbines. The simplified procedure determines the minimum and maximum powers, the efficiency at maximum power, the evolution of powers by head and flow and to determine the correct relationship between runner/impeller blade angle and guide vane opening for most efficient operation of double-regulated machines. The simplified procedure can be used for a rapid and partial estimation of the performance of hydraulic turbines for repair and maintenance work.

  2. 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.

  3. 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.

  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. Characterization of the hydraulic performance of a gully under drainage conditions.

    Science.gov (United States)

    Martins, Ricardo; Leandro, Jorge; de Carvalho, Rita Fernandes

    2014-01-01

    During rainfall events with low return periods (1-20 years) the drainage system can provide some degree of protection to urban areas. The system design is based not only on good hydraulic performance of the surface and the sewer network but also on their linking elements. Although the linking elements are of utmost importance as they allow the exchange of flow between the surface and the sewer network, there is a lack of studies that thoroughly characterize them. One crucial structural part of those elements is the gully. State-of-the-art dual-drainage models often use simplified formulae to replicate the gully hydraulic behaviour that lacks proper validation. This work focuses on simulating, both numerically and experimentally, the hydraulic performance of a 0.6 × 0.3 × 0.3 [m] (L × W × D) gully located inside an 8 × 0.5 × 0.5 [m] rectangular channel. The numerical simulations are conducted with the OpenFOAM toolbox and validated with water level measurements in the Multiple-Linking-Element experimental installation located at the Laboratory of Hydraulics of the University of Coimbra. The results provide a complete three-dimensional insight of the hydraulic behaviour of the flow inside the gully, and discharge coefficient formulae are disclosed that can be directly applied in dual-drainage models as internal boundary conditions.

  6. 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)

  7. 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.

  8. Hydraulic performance of sluice gate with unloaded upstream rotor ...

    African Journals Online (AJOL)

    ... compared to the flow under rotor and weir flow conditions. The video analysis also indicated that significant perturbation exists for the rotor angular speed. The normalized perturbation intensity varied from a minimum of 8% to a maximum of 60%. Keywords: sluice gate, rotor, angular speed, video analysis, hydropower ...

  9. 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)

  10. Thermo-hydraulic performance enhancement of solar air heater ...

    African Journals Online (AJOL)

    DR OKE

    Keywords: Solar air heater; Nusselt number; thermal efficiency; multiple arcs with ... loss; and one or two covers of glass or transparent plastic provide resistance to ..... Methods of testing to determine the thermal performance of solar collectors.

  11. 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

  12. 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.)

  13. Thermal-hydraulic performance of convective boiling jet array impingement

    International Nuclear Information System (INIS)

    Jenkins, R; De Brún, C; Kempers, R; Lupoi, R; Robinson, A J

    2016-01-01

    Jet impingement boiling is investigated with regard to heat transfer and pressure drop performance using a novel laser sintered 3D printed jet impingement manifold design. Water was the working fluid at atmospheric pressure with inlet subcooling of 7 o C. The convective boiling performance of the impinging jet system was investigated for a flat copper target surface for 2700≤Re≤5400. The results indicate that the heat transfer performance of the impinging jet is independent of Reynolds number for fully developed boiling. Also, the investigation of nozzle to plate spacing shows that low spacing delays the onset of nucleate boiling causing a superheat overshoot that is not observed with larger gaps. However, no sensitivity to the gap spacing was measured once boiling was fully developed. The assessment of the pressure drop performance showed that the design effectively transfers heat with low pumping power requirements. In particular, owing to the insensitivity of the heat transfer to flow rate during fully developed boiling, the coefficient of performance of jet impingement boiling in the fully developed boiling regime deteriorates with increased flow rate due to the increase in pumping power flux. (paper)

  14. Simulation of the hydraulic performance of highway filter drains through laboratory models and stormwater management tools.

    Science.gov (United States)

    Sañudo-Fontaneda, Luis A; Jato-Espino, Daniel; Lashford, Craig; Coupe, Stephen J

    2017-05-23

    Road drainage is one of the most relevant assets in transport infrastructure due to its inherent influence on traffic management and road safety. Highway filter drains (HFDs), also known as "French Drains", are the main drainage system currently in use in the UK, throughout 7000 km of its strategic road network. Despite being a widespread technique across the whole country, little research has been completed on their design considerations and their subsequent impact on their hydraulic performance, representing a gap in the field. Laboratory experiments have been proven to be a reliable indicator for the simulation of the hydraulic performance of stormwater best management practices (BMPs). In addition to this, stormwater management tools (SMT) have been preferentially chosen as a design tool for BMPs by practitioners from all over the world. In this context, this research aims to investigate the hydraulic performance of HFDs by comparing the results from laboratory simulation and two widely used SMT such as the US EPA's stormwater management model (SWMM) and MicroDrainage®. Statistical analyses were applied to a series of rainfall scenarios simulated, showing a high level of accuracy between the results obtained in laboratory and using SMT as indicated by the high and low values of the Nash-Sutcliffe and R 2 coefficients and root-mean-square error (RMSE) reached, which validated the usefulness of SMT to determine the hydraulic performance of HFDs.

  15. 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

  16. 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.

  17. 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

  18. 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...

  19. 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

  20. 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 ...

  1. 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.

  2. 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)

  3. The effect of liner hydraulic conductivity on disposal cell performance

    International Nuclear Information System (INIS)

    Yu, C.; Yuan, Y.C.; Chia, Y.P.

    1988-01-01

    Multilayered disposal cells are frequently used for the disposal of radioactive and hazardous wastes. These disposal cells consist of materials with different permeabilities that are placed in various thicknesses at the bottom as well as in the cover of the cell. Typically, a layer of permeable material is placed above a layer with low permeability; the permeable layer functions as a drainage/leachate collection system and the low-permeability layer functions as a migration barrier/liner. This paper analyzes the effects of infiltration through unsaturated soil liners on the long-term performance of the disposal cell. Based on the results of this study, it is concluded that the long-term performance of a disposal cell is dependent on a well-designed cell cover. The design should emphasize a cap with less permeable material to prevent water from infiltrating the disposal cell. An impermeable bottom liner is effective only in the short term; however, it can eventually result in saturation of the wastes and cause the bathtub effect over the long term

  4. Optimal thermal-hydraulic performance for helium-cooled divertors

    International Nuclear Information System (INIS)

    Izenson, M.G.; Martin, J.L.

    1996-01-01

    Normal flow heat exchanger (NFHX) technology offers the potential for cooling divertor panels with reduced pressure drops (<0.5% Δp/p), reduced pumping power (<0.75% pumping/thermal power), and smaller duct sizes than conventional helium heat exchangers. Furthermore, the NFHX can easily be fabricated in the large sizes required for divertors in large tokamaks. Recent experimental and computational results from a program to develop NFHX technology for divertor coolings using porous metal heat transfer media are described. We have tested the thermal and flow characteristics of porous metals and identified the optimal heat transfer material for the divertor heat exchanger. Methods have been developed to create highly conductive thermal bonds between the porous material and a solid substrate. Computational fluid dynamics calculations of flow and heat transfer in the porous metal layer have shown the capability of high thermal effectiveness. An 18-kW NFHX, designed to meet specifications for the international Thermonuclear Experimental Reactor divertor, has been fabricated and tested for thermal and flow performance. Preliminary results confirm design and fabrication methods. 11 refs., 12 figs., 1 tab

  5. Hydraulic numerical analyses of the IFMIF target performance

    International Nuclear Information System (INIS)

    Gordeev, S.; Heinzel, V.; Stieglitz, R.

    2011-01-01

    The target of the International Fusion Material Irradiation Facility (IFMIF) is one of the most crucial components of the planned facility, since due the high power densities only a free surface operated liquid lithium target is potentially capable to match the constraints set by the user community. Therefore in the design process two design options for such a free surface target have been elaborated. Both are analyzed by means of a time dependent large eddy simulation in this paper. The analysis of the numerical data rapidly shows that not only the nozzle but also the flow straighteners installed upstream the two step acceleration nozzle impact the flow especially in the boundary layers of the duct flow domain. Those boundary layers formulate the inlet conditions into the technically relevant target domain. In general both target design options serve an acceptable film thickness in the irradiation area. However, the design option with a straight to curved back wall is less preferable than a design solution with an entirely curved back wall since the surface shape exhibits larger surface fluctuations and a higher fluctuation intensity although the thermal behaviour close to the back wall is marginally better than for the entirely curved option.

  6. Thermal hydraulic performance of naturally aspirated control rod housing assemblies

    International Nuclear Information System (INIS)

    Geiger, G.T.; Randolph, H.W.; Paik, I.K.; Foti, D.J.

    1992-01-01

    Savannah River Site reactors are comprised of heat generating fuel/target assemblies, control rods which regulate reactor power, and heavy water which acts as the coolant and as a moderator. The fuel/target assemblies are cooled by the downflow of heavy water while the control rods are cooled via upflow. Five control rods are grouped with two safety rods in seven-channel assemblies called septifoils. Under normal operating conditions, the reactor power level, radial shape flux and axial power flux are regulated by the positioning of the control rods. The control rods are solid rods of a lithium-aluminum alloy with an thin aluminum outer sheath. Lithium is a good absorber of neutrons and, thus control rod temperatures rise with reactor power. At conditions of sufficiently high reactor power and degraded coolant flow, the control rods could heat sufficiently to cause a metallurigical failure of the sheath leading to molten material coming in contact with water and the possibility of a steam explosion. An accident has been postulated as part of the analysis involving the safety upgrade of Savannah River Site reactors in which the housing is not seated on the pin. Coolant from the upflow pin would not be directed into the housing but, into the moderator space surrounding the housing. Only naturally aspirated cooling due to buoyancy effects would be available to cool the control rods and the coolant mass flow rate would drop significantly from its nominal value. In this study, the mechanisms and limits of cooling heated rods housed in an unseated septifoil are addressed. Experiments were conducted on a shortened, prototypic housing with electrically heated rods to gain an understanding of the phenomena governing the cooling in such a case and develop data which can be used to evaluate predictive models. These experiments are described, their results discussed, and the predictions of current models is presented

  7. 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

  8. 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)

  9. 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.

  10. 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)

  11. Design concept of a pump stage with replaceable hydraulic components and prediction of its performance curves

    International Nuclear Information System (INIS)

    Lugova, S O; Knyazeva, E G; Tverdokhleb, I B; Kochevsky, A N

    2010-01-01

    In many cases, centrifugal pump units are expected to deliver the required performance under varying operating conditions. In particular, the pumps for oil extraction and transportation should deliver a constant head, although their capacity often changes during the life cycle. In order to keep the efficiency at a high level and not to replace a whole pump, the authors suggest to replace in such cases only hydraulic components of the pump (impellers and stationary sections of diffuser channels) that are to be installed in the same casing. The paper describes an approach for designing of radial-flow impellers and sections of diffuser channels to be used as replaceable. It allows for delivering a required head and providing a high efficiency in a wide range of capacities. The components intended for smaller capacities are featured with narrower flow passages. However, the dimensions of replaceable components are the same. The paper describes also a numerical simulation of fluid flow in a pump stage with two sets of replaceable radial-flow impellers and sections of diffuser channels. The CFD software used in this research is ANSYS CFX 11. Good correspondence of results is observed. Difference in flow pattern at various capacities and its influence on the performance curves delivered with replaceable components is demonstrated. Basing on the obtained results, the analysis of energy losses is presented.

  12. Thermal Hydraulic Performance in a Solar Air Heater Channel with Multi V-Type Perforated Baffles

    Directory of Open Access Journals (Sweden)

    Anil Kumar

    2016-07-01

    Full Text Available This article presents heat transfer and fluid flow characteristics in a solar air heater (SAH channel with multi V-type perforated baffles. The flow passage has an aspect ratio of 10. The relative baffle height, relative pitch, relative baffle hole position, flow attack angle, and baffle open area ratio are 0.6, 8.0, 0.42, 60°, and 12%, respectively. The Reynolds numbers considered in the study was in the range of 3000–10,000. The re-normalization group (RNG k-ε turbulence model has been used for numerical analysis, and the optimum relative baffle width has been investigated considering relative baffle widths of 1.0–7.0.The numerical results are in good agreement with the experimental data for the range considered in the study. Multi V-type perforated baffles are shown to have better thermal performance as compared to other baffle shapes in a rectangular passage. The overall thermal hydraulic performance shows the maximum value at the relative baffle width of 5.0.

  13. Drag Reduction and Performance Improvement of Hydraulic Torque Converters with Multiple Biological Characteristics

    Directory of Open Access Journals (Sweden)

    Liu Chunbao

    2016-01-01

    Full Text Available Fish-like, dolphin-like, and bionic nonsmooth surfaces were employed in a hydraulic torque converter to achieve drag reduction and performance improvement, which were aimed at reducing profile loss, impacting loss and friction loss, respectively. YJSW335, a twin turbine torque converter, was bionically designed delicately. The biological characteristics consisted of fish-like blades in all four wheels, dolphin-like structure in the first turbine and the stator, and nonsmooth surfaces in the pump. The prediction performance of bionic YJSW335, obtained by computational fluid dynamics simulation, was improved compared with that of the original model, and then it could be proved that drag reduction had been achieved. The mechanism accounting for drag reduction of three factors was also investigated. After bionic design, the torque ratio and the highest efficiencies of YJSW335 were both advanced, which were very difficult to achieve through traditional design method. Moreover, the highest efficiency of the low speed area and high speed area is 85.65% and 86.32%, respectively. By economic matching analysis of the original and bionic powertrains, the latter can significantly reduce the fuel consumption and improve the operating economy of the loader.

  14. 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.

  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. Effects of draft tube on the hydraulic performance of a Francis turbine

    International Nuclear Information System (INIS)

    Jeon, J H; Byeon, S S; Kim, Y J

    2013-01-01

    The draft tube is an important component of a Francis turbine which influences the hydraulic performance. It is located just under the runner and allowed to decelerate the flow velocity exiting the runner, thereby converting the excess of kinetic energy into static pressure. In this study, we have numerically investigated the hydraulic performance of a Francis turbine on the 15MW hydropower generation with various design parameters (three types of draft tube, thickness of guide vane) through a three-dimensional numerical method with the SST turbulent model. The vortex rope characteristics of the draft tube were confirmed. The results of the vortex flow fields and flow characteristics were graphically depicted with different design parameters and operating conditions

  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. 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)

  19. Numerical investigation of thermal-hydraulic performance of channel with protrusions by turbulent cross flow jet

    Science.gov (United States)

    Sahu, M. K.; Pandey, K. M.; Chatterjee, S.

    2018-05-01

    In this two dimensional numerical investigation, small rectangular channel with right angled triangular protrusions in the bottom wall of test section is considered. A slot nozzle is placed at the middle of top wall of channel which impinges air normal to the protruded surface. A duct flow and nozzle flow combined to form cross flow which is investigated for heat transfer enhancement of protruded channel. The governing equations for continuity, momentum, energy along with SST k-ω turbulence model are solved with finite volume based Computational fluid dynamics code ANSYS FLUENT 14.0. The range of duct Reynolds number considered for this analysis is 8357 to 51760. The ratios of pitch of protrusion to height of duct considered are 0.5, 0.64 and 0.82. The ratios of height of protrusion to height of duct considered are 0.14, 0.23 and 0.29. The effect of duct Reynolds number, pitch and height of protrusion on thermal-hydraulic performance is studied under cross flow condition. It is found that heat transfer rate is more at relatively larger pitch and small pressure drop is found in case of low height of protrusion.

  20. Hydraulic Performance of an Innovative Breakwater for Overtopping Wave Energy Conversion

    Directory of Open Access Journals (Sweden)

    Claudio Iuppa

    2016-11-01

    Full Text Available The Overtopping BReakwaterfor Energy Conversion (OBREC is an overtopping wave energy converter, totally embedded in traditional rubble mound breakwaters. The device consists of a reinforced concrete front reservoir designed with the aim of capturing the wave overtopping in order to produce electricity. The energy is extracted through low head turbines, using the difference between the water levels in the reservoir and the sea water level. This paper analyzes the OBREC hydraulic performances based on physical 2D model tests carried out at Aalborg University (DK. The analysis of the results has led to an improvement in the overall knowledge of the device behavior, completing the main observations from the complementary tests campaign carried out in 2012 in the same wave flume. New prediction formula are presented for wave reflection, the overtopping rate inside the front reservoir and at the rear side of the structure. Such methods have been used to design the first OBREC prototype breakwater in operation since January 2016 at Naples Harbor (Italy.

  1. Evaluation of Environmental and Hydraulic Performance of Bio-Composite Revetment Blocks

    OpenAIRE

    Thamer A.  Ahmeed; Nor A.  Alias; Abdul H.  Ghazali; Mohd. S.  Jaafar

    2006-01-01

    It is necessary to develop a concrete revetment block which can cater for environment and at the same time it will be effective in protecting river banks (stabilize the slope of banks) from scouring during flood. In the present study, the environmental and hydraulic performance of the proposed revetment block was evaluated through laboratory and field tests. The tested revetment block is called bio-composite because it is composed of concrete, plastic mesh and biological material (coconut hus...

  2. 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)

  3. 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.

  4. Experimental study of thermal–hydraulic performance of cam-shaped tube bundle with staggered arrangement

    International Nuclear Information System (INIS)

    Bayat, Hamidreza; Lavasani, Arash Mirabdolah; Maarefdoost, Taher

    2014-01-01

    Highlights: • Thermal–hydraulic performance of a non-circular tube bundle has been investigated experimentally. • Tubes were mounted in staggered arrangement with two longitudinal pitch ratios 1.5 and 2. • Drag coefficient and Nusselt number of tubes in second row was measured. • Friction factor of this tube bundle is lower than circular tube bundle. • Thermal–hydraulic performance of this tube bundle is greater than circular tube bundle. - Abstract: Flow and heat transfer from cam-shaped tube bank in staggered arrangement is studied experimentally. Tubes were located in test section of an open loop wind tunnel with two longitudinal pitch ratios 1.5 and 2. Reynolds number varies in range of 27,000 ⩽ Re D ⩽ 42,500 and tubes surface temperature is between 78 and 85 °C. Results show that both drag coefficient and Nusselt number depends on position of tube in tube bank and Reynolds number. Tubes in the first column have maximum value of drag coefficient, while its Nusselt number is minimum compared to other tubes in tube bank. Moreover, pressure drop from this tube bank is about 92–93% lower than circular tube bank and as a result thermal–hydraulic performance of this tube bank is about 6 times greater than circular tube bank

  5. 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)

  6. 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)

  7. 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

  8. 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.

  9. 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)

  10. Hydraulically amplified self-healing electrostatic actuators with muscle-like performance

    Science.gov (United States)

    Acome, E.; Mitchell, S. K.; Morrissey, T. G.; Emmett, M. B.; Benjamin, C.; King, M.; Radakovitz, M.; Keplinger, C.

    2018-01-01

    Existing soft actuators have persistent challenges that restrain the potential of soft robotics, highlighting a need for soft transducers that are powerful, high-speed, efficient, and robust. We describe a class of soft actuators, termed hydraulically amplified self-healing electrostatic (HASEL) actuators, which harness a mechanism that couples electrostatic and hydraulic forces to achieve a variety of actuation modes. We introduce prototypical designs of HASEL actuators and demonstrate their robust, muscle-like performance as well as their ability to repeatedly self-heal after dielectric breakdown—all using widely available materials and common fabrication techniques. A soft gripper handling delicate objects and a self-sensing artificial muscle powering a robotic arm illustrate the wide potential of HASEL actuators for next-generation soft robotic devices.

  11. Factors affecting the hydraulic performance of infiltration based SUDS in clay

    DEFF Research Database (Denmark)

    Bockhorn, B.; Klint, K.E.S.; Locatelli, Luca

    2017-01-01

    The influence of small scale soil heterogeneity on the hydraulic performance of infiltration based SUDS was studied using field data from a clayey glacial till and groundwater simulations with the integrated surface water and groundwater model HydroGeoSphere. Simulations of homogeneous soil blocks...... with hydraulic properties ranging from sand to clay showed that infiltration capacities vary greatly for the different soil types observed in glacial till. The inclusion of heterogeneities dramatically increased infiltration volume by a factor of 22 for a soil with structural changes above and below the CaC03...... boundary. Infiltration increased further by 8% if tectonic fractures were included and by another 61% if earthworm burrows were added. Comparison of HydroGeoSphere infiltration hydrographs with a simple soakaway model (Roldin et al. 2012) showed similar results for homogenous soils but indicated...

  12. 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

  13. 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.

  14. 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

  15. Towards the optimization of the thermal–hydraulic performance of gyrotron collectors

    Energy Technology Data Exchange (ETDEWEB)

    Savoldi, Laura; Bertani, Cristina [Dipartimento Energia, Politecnico di Torino, 10129 Torino (Italy); Cau, Francesca; Cismondi, Fabio [F4E, Barcelona (Spain); Gantenbein, Gerd; Illy, Stefan [Karlsruhe Institute of Technology (KIT), Institute for Pulsed Power and Microwave Technology (IHM), Kaiserstr. 12, 76131 Karlsruhe (Germany); Monni, Grazia [Dipartimento Energia, Politecnico di Torino, 10129 Torino (Italy); Rozier, Yoann [Thales Electron Devices, 78141 Vélizy-Villacoublay (France); Zanino, Roberto, E-mail: roberto.zanino@polito.it [Dipartimento Energia, Politecnico di Torino, 10129 Torino (Italy)

    2015-11-15

    Different configurations of water-cooled Cu collector for gyrotrons are investigated using the StarCCM + CFD code, aimed at optimizing its thermal–hydraulic (TH) performance. Although the current collectors show a good performance, the collector can be subjected to transient heat loads, due to the spent electron beam, of up to several tens of MW/m{sup 2}, and there is an interest to increase the gyrotron output power in the future. Furthermore, an optimized cooling will lead to improved reliability and lifetime of the collector. Starting from a hypervapotron (HV)-like collector, characterized by 100+ deep rectangular cavities with aspect ratio (AR) = 3, we present in the first part of the paper a single-cavity steady-state parametric analysis of the effect of AR on the heat exhaust capabilities. The investigation is then extended to other collector designs, including circumferential ribs and dimples, in order to assess the options for further improvements of the TH performance. The peak Cu temperature is computed by the code and its minimization is the target of the present optimization exercise. A self-consistent estimate of the heat transfer coefficient between collector and coolant is also obtained, which could be useful for fatigue and lifetime assessments. In the second part of the paper the most promising collector geometries identified in the first part are analyzed in the case of a transient heat load (vertical sweeping), first at the level of a single spatial period of the collector structure, then at the full-collector level. The results of the TH transient analysis are compared with both the results of the first part and with the transient purely thermal analysis of the full collector, showing for all geometries considered in this study a room for cooling efficiency improvement with respect to the HV-like design with AR = 3, at least in the operating conditions considered for this study (V ∼ 4 m/s, almost 100 °C sub-cooling).

  16. Comparative performance of the conjugate gradient and SOR [Successive Over Relaxation] methods for computational thermal hydraulics

    International Nuclear Information System (INIS)

    King, J.B.; Anghaie, S.; Domanus, H.M.

    1987-01-01

    Finite difference approximations to the continuity, momentum, and energy equations in thermal hydraulics codes result in a system of N by N equations for a problem having N field points. In a three dimensional problem, N increases as the problem becomes larger or more complex, and more rapidly as the computational mesh size is reduced. As a consequence, the execution time required to solve the problem increases, which may lead to placing limits on the problem resolution or accuracy. A conventinal method of solution of these systems of equations is the Successive Over Relaxation (SOR) technique. However, for a wide range of problems the execution time may be reduced by using a more efficient linear equation solver. One such method is the conjugate gradient method which was implemented in COMMIX-1B thermal hydraulics code. It was found that the execution time required to solve the resulting system of equations was reduced by a factor of about 2 for some problems. This paper summarizes the characteristics of these iterative solution procedures and compares their performance in modeling of a variety of reactor thermal hydraulic problems, using the COMMIX-1B computer code

  17. Optimisation of hydraulic performance to maximise faecal coliform removal in maturation ponds.

    Science.gov (United States)

    Bracho, Nibis; Lloyd, Barry; Aldana, Gerardo

    2006-05-01

    The present study was conducted with the aim of improving faecal coliform (FC) and faecal streptococcus (FS) removal efficiencies in tertiary maturation stages of a sewage treatment plant in Southern England, where climatic conditions are sub-optimal. The research used intensive field assessments (bacteriological, general quality and hydraulic) to identify the parameters that affect the bacteriological quality of the effluent from three parallel maturation ponds (North, Central and South) of similar geometry and dimensions. An engineering intervention was carried out to convert the South pond to three channels to increase the L/W ratio from 9:1 to 79:1. Hydraulic tracer studies in the South pond with Rhodamine WT showed that the dispersion number 'd' was reduced from 0.37 (dispersed flow) to 0.074 by this intervention under similar flow conditions (4.5l/s). Hydraulic retention time was thus increased by 5h, delay in jet flow short-circuiting was increased from 2.5 to 17.5h thus increasing the exposure times for all elements. As a result of the intervention FC removal increased substantially. Maximum channel-lagoon efficiency of 99.84% was obtained at 4.5l/s and 19 degrees C, when exposure to sunlight was 17 h in summer. It is concluded that the channel configuration produces a higher hydraulic efficiency than conventional maturation ponds. It is therefore recommended as a viable engineering solution which permits a low-cost upgrading of plant performance, requiring no additional land, and with minimal maintenance costs.

  18. 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)

  19. 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)

  20. 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.

  1. 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

  2. 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)

  3. 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...

  4. 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)

  5. 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)

  6. Sludge accumulation and distribution impact the hydraulic performance in waste stabilisation ponds.

    Science.gov (United States)

    Coggins, Liah X; Ghisalberti, Marco; Ghadouani, Anas

    2017-03-01

    Waste stabilisation ponds (WSPs) are used worldwide for wastewater treatment, and throughout their operation require periodic sludge surveys. Sludge accumulation in WSPs can impact performance by reducing the effective volume of the pond, and altering the pond hydraulics and wastewater treatment efficiency. Traditionally, sludge heights, and thus sludge volume, have been measured using low-resolution and labour intensive methods such as 'sludge judge' and the 'white towel test'. A sonar device, a readily available technology, fitted to a remotely operated vehicle (ROV) was shown to improve the spatial resolution and accuracy of sludge height measurements, as well as reduce labour and safety requirements. Coupled with a dedicated software package, the profiling of several WSPs has shown that the ROV with autonomous sonar device is capable of providing sludge bathymetry with greatly increased spatial resolution in a greatly reduced profiling time, leading to a better understanding of the role played by sludge accumulation in hydraulic performance of WSPs. The high-resolution bathymetry collected was used to support a much more detailed hydrodynamic assessment of systems with low, medium and high accumulations of sludge. The results of the modelling show that hydraulic performance is not only influenced by the sludge accumulation, but also that the spatial distribution of sludge plays a critical role in reducing the treatment capacity of these systems. In a range of ponds modelled, the reduction in residence time ranged from 33% in a pond with a uniform sludge distribution to a reduction of up to 60% in a pond with highly channelized flow. The combination of high-resolution measurement of sludge accumulation and hydrodynamic modelling will help in the development of frameworks for wastewater sludge management, including the development of more reliable computer models, and could potentially have wider application in the monitoring of other small to medium water bodies

  7. 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.

  8. 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.

  9. Thermal-Hydraulic Performance of a Corrugated Cooling Fin with Louvered Surfaces

    DEFF Research Database (Denmark)

    Sønderby, Simon Kaltoft; Hosseini, Seyed Mojtaba Mir; Rezaniakolaei, Alireza

    2017-01-01

    The main objective of the article is to investigate thermal-hydraulic performance of a corrugated cooling fin with louvered surfaces. The investigation is carried out using the fin geometry of one most commonly used liquid-to-air heat exchangers. The investigation was carried out by numerically...... simulating the airflow with louvered fin geometry. The simulation model was verified by comparing simulated j- and f-factors with the corresponding values of several experimental correlations. The j-factors deviated less than 10.7 % from two of the experimental correlations, whereas deviations ranging...

  10. Performance of optimised prosthetic ankle designs that are based on a hydraulic variable displacement actuator (VDA)

    OpenAIRE

    Gardiner, JD; Bari, AZ; Kenney, LPJ; Twiste, M; Moser, D; Zahedi, S; Howard, D

    2017-01-01

    Current energy storage and return (ESR) prosthetic\\ud feet only marginally reduce the cost of amputee locomotion\\ud compared to basic solid ankle cushioned heel (SACH) feet,\\ud possibly due to their lack of push-off at the end of stance. To our knowledge, a prosthetic ankle that utilises a hydraulic variable displacement actuator (VDA) to improve push-off performance has not previously been proposed. Therefore, here we report a design optimisation and simulation feasibility study for a VDA ba...

  11. Development of Mitsubishi high thermal performance grid 1 - CFD applicability for thermal hydraulic design

    International Nuclear Information System (INIS)

    Ikeda, K.; Hoshi, M.

    2001-01-01

    Mitsubishi applied the Computational Fluid Dynamics (CFD) evaluation method for designing of the new lower pressure loss and higher DNB performance grid spacer. Reduction of pressure loss of the grid has been estimated by CFD. Also, CFD has been developed as a design tool to predict the coolant mixing ability of vane structures, that is to compare the relative peak spot temperatures around fuel rods at the same heat flux condition. These evaluations have been reflected to the new grid spacer design. The prototype grid was manufactured and some flow tests were performed to examine the thermal hydraulic performance, which were predicted by CFD. The experimental data of pressure loss was in good agreement with CFD prediction. The CFD prediction of flow behaviors at downstream of the mixing vanes was verified by detail cross-flow measurements at rod gaps by the rod LDV system. It is concluded that the applicability of the CFD evaluation method for the thermal hydraulic design of the grid is confirmed. (authors)

  12. 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)

  13. 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

  14. 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)

  15. Modelling and thermal hydraulic analysis of the Angra-2 nuclear reactor using RELAP5-3D code

    International Nuclear Information System (INIS)

    González Mantecón, Javier

    2015-01-01

    The evaluation of Nuclear Power Plants (NPPs) performance during steady-state and accident conditions has been one of the main research subjects in the nuclear field. In order to simulate the behavior of water-cooled reactors, several complex thermal-hydraulic codes systems have been developed. Particularly, the RELAP5 code, developed by the Idaho National Laboratory, is a best-estimate thermal-hydraulic analysis tool and one of the most used in nuclear industry. The RELAP5-3D 3.0.0 code was used to develop a detailed model of Angra 2 nuclear reactor using reference data from the Final Safety Analysis Report. Angra 2 is the second Brazilian NPP, which began commercial operation in 2001. The plant is equipped with a Pressurized Water Reactor (PWR) type with 3771.0 MWt. Simulations of the reactor behavior during normal operation conditions and postulated accident conditions were performed. Results achieved in the reactor steady-state simulation were compared with nominal parameters of the NPP. These results proved to be in good agreement, with relative errors less than 1%. In the transient simulation, the obtained results were coherent and satisfactory. This study demonstrates that the RELAP5-3D model is capable to reproduce the thermal-hydraulic behavior of the Angra-2 PWR during diverse operation conditions and it can contribute for the process of the plant safety analysis. (author)

  16. 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

  17. 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 rational safety design and assessment. This report describes the incorporation of a three-dimensional upper plenum model to ACT and its verification study as part of the program development. To treat the influence of three-dimensional thermal-hydraulic behavior in a upper plenum on the in-core temperature field, the multi-dimensional general purpose thermal-hydraulic analysis program AQUA, which was developed and validated at JNC, was applied as the base of the upper plenum analysis module of ACT. AQUA enables to model the upper plenum configuration including immersed heat exchangers of the direct reactor auxiliary cooling system (DRACS). In coupling core analysis module that consists of the fuel-assembly and the inter-wrapper gap calculation parts with the upper plenum module, different types of computation mesh systems were jointed using the staggered quarter assembly mesh scheme. A coupling algorithm among core, upper plenum and heat transport system modules, which can keep mass, momentum and energy conservation, was developed and optimized in consideration of parallel computing. ACT was applied to analyzing a sodium experiment (PLANDTL-DHX) performed at JNC, which simulated the natural circulation decay heat removal under DRACS operation conditions for the program verification. From the calculation result, the validity of the improved program was confirmed. (author)

  18. Performance-oriented asymptotic tracking control of hydraulic systems with radial basis function network disturbance observer

    Directory of Open Access Journals (Sweden)

    Jian Hu

    2016-05-01

    Full Text Available Uncertainties, including parametric uncertainties and uncertain nonlinearities, always exist in positioning servo systems driven by a hydraulic actuator, which would degrade their tracking accuracy. In this article, an integrated control scheme, which combines adaptive robust control together with radial basis function neural network–based disturbance observer, is proposed for high-accuracy motion control of hydraulic systems. Not only parametric uncertainties but also uncertain nonlinearities (i.e. nonlinear friction, external disturbances, and/or unmodeled dynamics are taken into consideration in the proposed controller. The above uncertainties are compensated, respectively, by adaptive control and radial basis function neural network, which are ultimately integrated together by applying feedforward compensation technique, in which the global stabilization of the controller is ensured via a robust feedback path. A new kind of parameter and weight adaptation law is designed on the basis of Lyapunov stability theory. Furthermore, the proposed controller obtains an expected steady performance even if modeling uncertainties exist, and extensive simulation results in various working conditions have proven the high performance of the proposed control scheme.

  19. 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

  20. Influence of sinusoidal flow on the thermal and hydraulic performance of microchannel heat sink

    International Nuclear Information System (INIS)

    Om, N I; Gunnasegaran, P; Rajasegaran, S

    2013-01-01

    In this paper, the effect of sinusoidal flow on the thermal and hydraulic performance of microchannel heat sink (MCHS) is numerically investigated. This investigation covers Reynolds number in the range of 100 ≤ Re ≤ 1000 and pure water is used as a working fluid. The three-dimensional steady, laminar flow and heat transfer governing equations are solved using finite volume method (FVM). The water flow field and heat transfer performance inside the sinusoidal microchannels is simulated and the results are compared with the straight microchannels. The effect of using sinusoidal microchannels on temperature distribution, Nusselt number, friction factor and thermal resistance is presented in this paper. It is found that with same rectangular cross-section, sinusoidal microchannels have a better heat transfer performance compared to the straight microchannels.

  1. 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)

  2. 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)

  3. 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)

  4. 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)

  5. 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

  6. 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)

  7. Flume Experiments for Optimizing the Hydraulic Performance of a Deep-Water Wetland Utilizing Emergent Vegetation and Obstructions

    Directory of Open Access Journals (Sweden)

    Shang-Shu Shih

    2016-06-01

    Full Text Available Constructed ponds and wetlands are widely used in urban areas for stormwater management, ecological conservation, and pollution treatment. The treatment efficiency of these systems is strongly related to the hydrodynamics and hydraulic residence time. In this study, we developed a physical model and used rhodamine-WT as a tracer to conduct flume experiments. An equivalent Reynolds number was assumed, and the flume was a 1/25-scale model. Emergent obstructions (EOs, submerged obstructions (SOs, and high- and low-density emergent vegetation were placed along the sides of the flume, and 49 tracer tests were performed. We altered the density, spatial extent, aspect ratio, and configurations of the obstructions and emergent vegetation to observe changes in the hydraulic efficiency of a deep-water wetland. In the cases of low-aspect-ratio obstructions, the effects of the EOs on the hydraulic efficiency were significantly stronger than those of the SOs. In contrast, in the cases of high-aspect-ratio obstructions, the improvement effects of the EOs were weaker than those of the SOs. The high-aspect-ratio EOs altered the flow direction and constrained the water conveyance area, which apparently caused a short-circuited flow phenomenon, resulting in a decrease in hydraulic efficiency. Most cases revealed that the emergent vegetation improved the hydraulic efficiency more than the EOs. The high-density emergent vegetation (HEV improved the hydraulic efficiency more than the low-density emergent vegetation (LEV. Three cases involving HEV, two cases involving LEV, and one case involving EOs attained a good hydraulic efficiency (λ > 0.75. To achieve greater water purification, aquatic planting in constructed wetlands should not be overly dense. The HEV configuration in case 3-1 achieved optimum hydraulic performance for compliance with applicable water treatment standards.

  8. Effect of tube bank inclination on the thermal hydraulic performance of air-cooled heat exchangers

    International Nuclear Information System (INIS)

    Monheit, M.; Freim, J.

    1987-01-01

    This paper compares the experimental thermal hydraulic performance of an inclined tube bundle of staggered high finned tubes with that of the same bank in cross flow. The data are shown on a relative basis with the conventional cross-flow arrangement used as the reference layout. The results, presented as plots of heat transfer coefficient and pressure drop per row versus face velocity, are compared with other published data for bare and extended-surface tube banks. Correction factors that account for the deviation from cross-flow performance are presented. Estimated savings in plot area are shown for typical cases. Some possible cost reduction opportunities in specific fields of application are also suggested

  9. 3D Blade Hydraulic Design Method of the Rotodynamic Multiphase Pump Impeller and Performance Research

    Directory of Open Access Journals (Sweden)

    Yongxue Zhang

    2014-02-01

    Full Text Available A hydraulic design method of three-dimensional blade was presented to design the blades of the rotodynamic multiphase pump. Numerical simulations and bench test were conducted to investigate the performance of the example impeller designed by the presented method. The results obtained from the bench test were in good agreement with the simulation results, which indicated the reasonability of the simulation. The distributions of pressure and gas volume fraction were analyzed and the results showed that the designed impeller was good for the transportation of mixture composed of gas and liquid. In addition, the advantage of the impeller designed by the presented method was suitable for using in large volume rate conditions, which were reflected by the comparison of the head performance between this three-dimensional design method and another one.

  10. Study on the performance of the hydraulic and mass-transfer with miniature centrifugal contactor

    International Nuclear Information System (INIS)

    Wang Jianchen; Tang Wencheng; Fan Shilei; Lian Jun

    1997-01-01

    The hydraulic performance and the mass-transfer properties of HNO 3 , Fe 3+ , Nd 3+ are studied in H 2 O-30% TRPO-kerosene system at different conditions with single-stage φ = 10 mm miniature centrifugal contactor. The rotor's speed varies from 4000 r/min to 4500 r/min. The total throughput is less than 600 mL/h. the phase ratio(o/a) changes from 1/10 to 10/1. Under the above experimental conditions, the single contactor operates very well and gives good performance. The stage efficiencies of HNO 3 and Nd 3+ are about 90%. The Fe 3+ extraction is very slow kinetically and the stage efficiency of Fe 3+ is low

  11. Using Pneumatics to Perform Laboratory Hydraulic Conductivity Tests on Gravel with Underdamped Responses

    Science.gov (United States)

    Judge, A. I.

    2011-12-01

    A permeameter has been designed and built to perform laboratory hydraulic conductivity tests on various kinds of gravel samples with hydraulic conductivity values ranging from 0.1 to 1 m/s. The tests are commenced by applying 200 Pa of pneumatic pressure to the free surface of the water column in a riser connected above a cylinder that holds large gravel specimens. This setup forms a permeameter specially designed for these tests which is placed in a barrel filled with water, which acts as a reservoir. The applied pressure depresses the free surface in the riser 2 cm until it is instantly released by opening a ball valve. The water then flows through the base of the cylinder and the specimen like a falling head test, but the water level oscillates about the static value. The water pressure and the applied air pressure in the riser are measured with vented pressure transducers at 100 Hz. The change in diameter lowers the damping frequency of the fluctuations of the water level in the riser, which allows for underdamped responses to be observed for all tests. The results of tests without this diameter change would otherwise be a series of critically damped responses with only one or two oscillations that dampen within seconds and cannot be evaluated with equations for the falling head test. The underdamped responses oscillate about the static value at about 1 Hz and are very sensitive to the hydraulic conductivity of all the soils tested. These fluctuations are also very sensitive to the inertia and friction in the permeameter that are calculated considering the geometry of the permeameter and verified experimentally. Several gravel specimens of various shapes and sizes are tested that show distinct differences in water level fluctuations. The friction of the system is determined by calibrating the model with the results of tests performed where the cylinder had no soil in it. The calculation of the inertia in the response of the water column for the typical testing

  12. Simulation of storage performance on hydropneumatic driveline in dual hybrid hydraulic passenger car

    Directory of Open Access Journals (Sweden)

    Wasbari Faizil

    2017-01-01

    Full Text Available The charging process is one of the critical processes in the hydro-pneumatic driveline storage system. It converts the kinetic energy of the vehicle braking and coasting to the compression energy. This energy is stored in the storage device called the accumulator. The system is planned to be used on the dual hydro-pneumatic hybrid driveline and applied to a hydraulic hybrid passenger car. The aim of this paper is to find the effect of charging parameters on the storage performance through simulation. Through the storage behaviour, the desirable and optimal sizing of the accumulator can be selected. The paper emphasized on the effect of pressure elevation, pre-charge pressure, effective volume, thermal reaction and required time of the accumulator’s charging process. The circuit of charging process has been designed and simulated by using the hydraulic tool in the Automation Studio software. The simulation results were corroborated through the component specification for data rationality. Through the simulation, it was found that pre-charge pressure had a significant effect on the charging process. It determined the efficiency of the effective volume. The higher the pressure elevation, the higher the effective volume. Nevertheless, the more energy required to compress the nitrogen gas in the bladder. Besides, in term of volume displacement, higher volume displacement reduced charging time and lower the fluid temperature. The simulation had been positively highlighted the critical point in charging process which later on, benefited the sizing process in the component selection specification.

  13. Use of geological mapping tools to improve the hydraulic performance of SuDS.

    Science.gov (United States)

    Bockhorn, Britta; Klint, Knud Erik Strøyberg; Jensen, Marina Bergen; Møller, Ingelise

    2015-01-01

    Most cities in Denmark are situated on low permeable clay rich deposits. These sediments are of glacial origin and range among the most heterogeneous, with hydraulic conductivities spanning several orders of magnitude. This heterogeneity has obvious consequences for the sizing of sustainable urban drainage systems (SuDS). We have tested methods to reveal geological heterogeneity at field scale to identify the most suitable sites for the placement of infiltration elements and to minimize their required size. We assessed the geological heterogeneity of a clay till plain in Eastern Jutland, Denmark measuring the shallow subsurface resistivity with a geoelectrical multi-electrode system. To confirm the resistivity data we conducted a spear auger mapping. The exposed sediments ranged from clay tills over sandy clay tills to sandy tills and correspond well to the geoelectrical data. To verify the value of geological information for placement of infiltration elements we carried out a number of infiltration tests on geologically different areas across the field, and we observed infiltration rates two times higher in the sandy till area than in the clay till area, thus demonstrating that the hydraulic performance of SuDS can be increased considerably and oversizing avoided if field geological heterogeneity is revealed before placing SuDS.

  14. Performance of Optimized Prosthetic Ankle Designs That Are Based on a Hydraulic Variable Displacement Actuator (VDA).

    Science.gov (United States)

    Gardiner, James; Bari, Abu Zeeshan; Kenney, Laurence; Twiste, Martin; Moser, David; Zahedi, Saeed; Howard, David

    2017-12-01

    Current energy storage and return prosthetic feet only marginally reduce the cost of amputee locomotion compared with basic solid ankle cushioned heel feet, possibly due to their lack of push-off at the end of stance. To the best of our knowledge, a prosthetic ankle that utilizes a hydraulic variable displacement actuator (VDA) to improve push-off performance has not previously been proposed. Therefore, here we report a design optimization and simulation feasibility study for a VDA-based prosthetic ankle. The proposed device stores the eccentric ankle work done from heel strike to maximum dorsiflexion in a hydraulic accumulator and then returns the stored energy to power push-off. Optimization was used to establish the best spring characteristic and gear ratio between ankle and VDA. The corresponding simulations show that, in level walking, normal push-off is achieved and, per gait cycle, the energy stored in the accumulator increases by 22% of the requirements for normal push-off. Although the results are promising, there are many unanswered questions and, for this approach to be a success, a new miniature, low-losses, and lightweight VDA would be required that is half the size of the smallest commercially available device.

  15. 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.

  16. 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

  17. Thermal Hydraulic Design and Analysis of a Water-Cooled Ceramic Breeder Blanket with Superheated Steam for CFETR

    Science.gov (United States)

    Cheng, Xiaoman; Ma, Xuebin; Jiang, Kecheng; Chen, Lei; Huang, Kai; Liu, Songlin

    2015-09-01

    The water-cooled ceramic breeder blanket (WCCB) is one of the blanket candidates for China fusion engineering test reactor (CFETR). In order to improve power generation efficiency and tritium breeding ratio, WCCB with superheated steam is under development. The thermal-hydraulic design is the key to achieve the purpose of safe heat removal and efficient power generation under normal and partial loading operation conditions. In this paper, the coolant flow scheme was designed and one self-developed analytical program was developed, based on a theoretical heat transfer model and empirical correlations. Employing this program, the design and analysis of related thermal-hydraulic parameters were performed under different fusion power conditions. The results indicated that the superheated steam water-cooled blanket is feasible. supported by the National Special Project for Magnetic Confined Nuclear Fusion Energy of China (Nos. 2013GB108004, 2014GB122000 and 2014GB119000), and National Natural Science Foundation of China (No. 11175207)

  18. Overview of the use of ATHENA for thermal-hydraulic analysis of systems with lead-bismuth coolant

    International Nuclear Information System (INIS)

    Davis, C.B.; Shieh, A. S.

    2000-01-01

    The INEEL and MIT are investigating the suitability of lead-bismuth cooled fast reactor for producing low-cost electricity as well as for actinide burning. This paper is concerned with the general area of thermal-hydraulics of lead-bismuth cooled reactors. The ATHENA code is being used in the thermal-hydraulic design and analysis of lead-bismuth cooled reactors. The ATHENA code was reviewed to determine its applicability for simulating lead-bismuth cooled reactors. Two modifications were made to the code as a result of this review. Specifically, a correlation to represent heat transfer from rod bundles to a liquid metal and a void correlation based on data taken in a mixture of lead-bismuth and steam were added the code. The paper also summarizes the analytical work that is being performed with the code and plans for future analytical work

  19. Overview of the Use of ATHENA for Thermal-Hydraulic Analysis of Systems with Lead-Bismuth Coolant

    Energy Technology Data Exchange (ETDEWEB)

    Davis, Cliff Bybee; Shieh, Arthur Shan Luk

    2000-04-01

    The INEEL and MIT are investigating the suitability of lead-bismuth cooled fast reactor for producing low-cost electricity as well as for actinide burning. This paper is concerned with the general area of thermal-hydraulics of lead-bismuth cooled reactors. The ATHENA code is being used in the thermal-hydraulic design and analysis of lead-bismuth cooled reactors. The ATHENA code was reviewed to determine its applicability for simulating lead-bismuth cooled reactors. Two modifications were made to the code as a result of this review. Specifically, a correlation to represent heat transfer from rod bundles to a liquid metal and a void correlation based on data taken in a mixture of lead-bismuth and steam were added the code. The paper also summarizes the analytical work that is being performed with the code and plans for future analytical work.

  20. Thermal-hydraulic analysis of PWR core including intermediate flow mixers with the THYC code

    International Nuclear Information System (INIS)

    Mur, J.; Meignin, J.C.

    1997-07-01

    Departure from nucleate boiling (DNB) is one of the major limiting factors of pressurized water reactors (PWRs). Safety requires that occurrence of DNB should be precluded under normal or incidental operating conditions. The thermal-hydraulic THYC code developed by EDF is described. The code is devoted to heat and mass transfer in nuclear components. Critical Heat Flux (CHF) is predicted from local thermal-hydraulic parameters such as pressure, mass flow rate, and quality. A three stage methodology to evaluate thermal margins in order to perform standard core design is described. (K.A.)

  1. Thermal-hydraulic analysis of PWR core including intermediate flow mixers with the THYC code

    Energy Technology Data Exchange (ETDEWEB)

    Mur, J. [Electricite de France (EDF), 78 - Chatou (France); Meignin, J.C. [Electricite de France (EDF), 69 - Villeurbanne (France)

    1997-07-01

    Departure from nucleate boiling (DNB) is one of the major limiting factors of pressurized water reactors (PWRs). Safety requires that occurrence of DNB should be precluded under normal or incidental operating conditions. The thermal-hydraulic THYC code developed by EDF is described. The code is devoted to heat and mass transfer in nuclear components. Critical Heat Flux (CHF) is predicted from local thermal-hydraulic parameters such as pressure, mass flow rate, and quality. A three stage methodology to evaluate thermal margins in order to perform standard core design is described. (K.A.) 8 refs.

  2. 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

  3. 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

  4. 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)

  5. 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

  6. 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

  7. 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)

  8. Study of the performance of four repairing material systems for hydraulic structures of concrete dams

    Directory of Open Access Journals (Sweden)

    Kormann A. C. M.

    2003-01-01

    Full Text Available Four types of repairing materials are studied as function of either a conventional concrete or a reference-concrete (RefC, these are: polymer-modified cement mortar (PMor, steel fiber concrete (SFco, epoxy mortar (EMor and silica fume mortar (SFmo, to be applied in hydraulic structures surfaces subjected to a high velocity water flow. Besides the mechanical requests and wearing resistance of hydraulic concrete dam structures, especially the spillway surfaces, the high solar radiation, the environmental temperature and wet and dry cycles, contribute significantly to the reduction of their lifespan. RefC and the SFco were developed based on a usual concrete mixture used in slabs of spillways. The average RefC mixture used was 1: 1.61: 2.99: 0.376, with Pozzolan-modified Portland cement consumption of 425 kg/m³. EMor and PMor mixtures followed the information given by the manufacturers and lab experience. Tests on concrete samples were carried out in laboratory simulating normally found environmental situations in order to control the mechanical resistance and the aging imposed conditions, such as solar radiation and humidity. Also, physicochemical characterizing tests were made for all used materials. From the analyzed results, two of them presented a higher performance: the EMor and SFmo. SFco presented good adherence to the RefC and good mechanical performance. However, it also presented apparent metal corrosion in humidity tests, being indicated for use, with caution, as an intermediate layer in underwater repairs. In a general classification, considering all tests, including their field applications, the better performance material systems were EMor- SFmo> SFco> PMor.

  9. Effect of fin-endwall fillet on thermal hydraulic performance of airfoil printed circuit heat exchanger

    International Nuclear Information System (INIS)

    Ma, Ting; Xin, Fei; Li, Lei; Xu, Xiang-yang; Chen, Yi-tung; Wang, Qiu-wang

    2015-01-01

    Printed circuit heat exchanger (PCHE) is recommended to be used for intermediate heat exchanger in Very High Temperature Reactor (VHTR). One of the key features is that it is manufactured by the photochemical etching in order to maintain the internal structure and metal properties. In this paper, a photochemical etching experiment is conducted to manufacture the airfoil PCHE plate. The result indicates that the airfoil fin is not an ideal airfoil profile, but has a fin-endwall fillet. For the purpose of simplifying the numerical model and saving computational time, a validated model with a single fluid is used to further study the effect of fin-endwall fillet on the thermal hydraulic performance of airfoil PCHE. It is found that the fin-endwall fillet can increase the heat transfer and pressure drop in the cases with the non-dimensional longitudinal pitch being 1.63. The effect of fin-endwall fillet on thermal hydraulic performance decreases with the increase of transverse pitch, but the longitudinal pitch has little effect when the non-dimensional longitudinal pitch is greater than 1.88. In the studied cases, the maximum difference of Nusselt number and friction factor between the two models with and without fin-endwall fillet is up to 6.7% and 6.4%. - Highlights: • Fillets are formed in the endwall of airfoil fins during the photochemical etching. • Two-fluid model can be replaced by single-fluid model to perform simulation. • Fin-endwall fillet can increase heat transfer and pressure drop at ζ_l = 1.63. • Effect of fin-endwall fillet decreases as transverse pitch increases at ζ_l = 1.63. • Longitudinal pitch has little effect at ζ_l ≥ 1.88.

  10. 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

  11. Thermal hydraulic analysis of the IPR-R1 TRIGA research reactor using a RELAP5 model

    International Nuclear Information System (INIS)

    Costa, Antonella L.; Reis, Patricia Amelia L.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Mesquita, Amir Z.; Soares, Humberto V.

    2010-01-01

    The RELAP5 code is widely used for thermal hydraulic studies of commercial nuclear power plants. Current investigations and code adaptations have demonstrated that the RELAP5 code can be also applied for thermal hydraulic analysis of nuclear research reactors with good predictions. Therefore, as a contribution to the assessment of RELAP5/MOD3.3 for research reactors analysis, this work presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor at 50 kilowatts (kW) of power operation. The reactor is located in the Nuclear Technology Development Center (CDTN), Brazil. It is a 250 kW, light water moderated and cooled, graphite-reflected, open pool type research reactor. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of the RELAP5 model validation. The RELAP5 results were also compared with calculated data from the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The results obtained have shown that the RELAP5 model for the IPR-R1 TRIGA reproduces the actual steady-state reactor behavior in good agreement with the available data.

  12. 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)

  13. Hydraulic performance of a low specific speed centrifugal pump with Spanwise-Slotted Blades

    International Nuclear Information System (INIS)

    Ye, D X; Li, H; Wang, Y

    2013-01-01

    The hydraulic efficiency of a low specific speed centrifugal pump is low because of the long and narrow meridian flow passage, and the severe disk friction. Spanwise slotted blade flow control technology has been applied to the low specific speed centrifugal pump. This paper concluded that spanwise slotted blades can improve the pump performance in both experiments and simulations. In order to study the influence to the impeller and volute by spanwise slotted blade, impeller efficiency and volute efficiency were defined. The minimum volute efficiency and the maximum pump efficiency appear at the same time in the design flow condition in the unsteady simulation. The mechanism of spanwise slotted blade flow control technology should be researched furthermore

  14. A Newton-based Jacobian-free approach for neutronic-Monte Carlo/thermal-hydraulic static coupled analysis

    International Nuclear Information System (INIS)

    Mylonakis, Antonios G.; Varvayanni, M.; Catsaros, N.

    2017-01-01

    Highlights: •A Newton-based Jacobian-free Monte Carlo/thermal-hydraulic coupling approach is introduced. •OpenMC is coupled with COBRA-EN with a Newton-based approach. •The introduced coupling approach is tested in numerical experiments. •The performance of the new approach is compared with the traditional “serial” coupling approach. -- Abstract: In the field of nuclear reactor analysis, multi-physics calculations that account for the bonded nature of the neutronic and thermal-hydraulic phenomena are of major importance for both reactor safety and design. So far in the context of Monte-Carlo neutronic analysis a kind of “serial” algorithm has been mainly used for coupling with thermal-hydraulics. The main motivation of this work is the interest for an algorithm that could maintain the distinct treatment of the involved fields within a tight coupling context that could be translated into higher convergence rates and more stable behaviour. This work investigates the possibility of replacing the usually used “serial” iteration with an approximate Newton algorithm. The selected algorithm, called Approximate Block Newton, is actually a version of the Jacobian-free Newton Krylov method suitably modified for coupling mono-disciplinary solvers. Within this Newton scheme the linearised system is solved with a Krylov solver in order to avoid the creation of the Jacobian matrix. A coupling algorithm between Monte-Carlo neutronics and thermal-hydraulics based on the above-mentioned methodology is developed and its performance is analysed. More specifically, OpenMC, a Monte-Carlo neutronics code and COBRA-EN, a thermal-hydraulics code for sub-channel and core analysis, are merged in a coupling scheme using the Approximate Block Newton method aiming to examine the performance of this scheme and compare with that of the “traditional” serial iterative scheme. First results show a clear improvement of the convergence especially in problems where significant

  15. Thermal-hydraulic analysis under partial loss of flow accident hypothesis of a plate-type fuel surrounded by two water channels using RELAP5 code

    OpenAIRE

    Itamar Iliuk; José Manoel Balthazar; Ângelo Marcelo Tusset; José Roberto Castilho Piqueira

    2016-01-01

    Thermal-hydraulic analysis of plate-type fuel has great importance to the establishment of safety criteria, also to the licensing of the future nuclear reactor with the objective of propelling the Brazilian nuclear submarine. In this work, an analysis of a single plate-type fuel surrounding by two water channels was performed using the RELAP5 thermal-hydraulic code. To realize the simulations, a plate-type fuel with the meat of uranium dioxide sandwiched between two Zircaloy-4 plates was prop...

  16. Analysis of the steady state hydraulic behaviour of the ITER blanket cooling system

    Energy Technology Data Exchange (ETDEWEB)

    Di Maio, P.A., E-mail: pietroalessandro.dimaio@unipa.it [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Dell’Orco, G.; Furmanek, A. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Garitta, S. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Merola, M.; Mitteau, R.; Raffray, R. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Spagnuolo, G.A.; Vallone, E. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy)

    2015-10-15

    Highlights: • Nominal steady state hydraulic behaviour of ITER blanket standard sector cooling system has been investigated. • Numerical simulations have been run adopting a qualified thermal-hydraulic system code. • Hydraulic characteristic functions and coolant mass flow rates, velocities and pressure drops have been assessed. • Most of the considered circuits are able to effectively cool blanket modules, meeting ITER requirements. - Abstract: The blanket system is the ITER reactor component devoted to providing a physical boundary for plasma transients and contributing to thermal and nuclear shielding of vacuum vessel, magnets and external components. It is expected to be subjected to significant heat loads under nominal conditions and its cooling system has to ensure an adequate cooling, preventing any risk of critical heat flux occurrence while complying with pressure drop limits. At the University of Palermo a study has been performed, in cooperation with the ITER Organization, to investigate the steady state hydraulic behaviour of the ITER blanket standard sector cooling system. A theoretical–computational approach based on the finite volume method has been followed, adopting the RELAP5 system code. Finite volume models of the most critical blanket cooling circuits have been set-up, realistically simulating the coolant flow domain. The steady state hydraulic behaviour of each cooling circuit has been investigated, determining its hydraulic characteristic function and assessing the spatial distribution of coolant mass flow rates, velocities and pressure drops under reference nominal conditions. Results obtained have indicated that the investigated cooling circuits are able to provide an effective cooling to blanket modules, generally meeting ITER requirements in term of pressure drop and velocity distribution, except for a couple of circuits that are being revised.

  17. 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

  18. 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

  19. A Numerical Study on the Improvement of Suction Performance and Hydraulic Efficiency for a Mixed-Flow Pump Impeller

    Directory of Open Access Journals (Sweden)

    Sung Kim

    2014-01-01

    Full Text Available This paper describes a numerical study on the improvement of suction performance and hydraulic efficiency of a mixed-flow pump by impellers. The design of these impellers was optimized using a commercial CFD (computational fluid dynamics code and DOE (design of experiments. The design variables of meridional plane and vane plane development were defined for impeller design. In DOE, variables of inlet part were selected as main design variables in meridional plane, and incidence angle was selected in vane plane development. The verification of the experiment sets that were generated by 2k factorial was done by numerical analysis. The objective functions were defined as the NPSHre (net positive suction head required, total efficiency, and total head of the impellers. The importance of the geometric design variables was analyzed using 2k factorial designs. The interaction between the NPSHre and total efficiency, according to the meridional plane and incidence angle, was discussed by analyzing the 2k factorial design results. The performance of optimally designed model was verified by experiments and numerical analysis and the reliability of the model was retained by comparison of numerical analysis and comparative analysis with the reference model.

  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 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. 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

  5. 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.

  6. 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

  7. High bulk modulus of ionic liquid and effects on performance of hydraulic system.

    Science.gov (United States)

    Kambic, Milan; Kalb, Roland; Tasner, Tadej; Lovrec, Darko

    2014-01-01

    Over recent years ionic liquids have gained in importance, causing a growing number of scientists and engineers to investigate possible applications for these liquids because of their unique physical and chemical properties. Their outstanding advantages such as nonflammable liquid within a broad liquid range, high thermal, mechanical, and chemical stabilities, low solubility for gases, attractive tribological properties (lubrication), and very low compressibility, and so forth, make them more interesting for applications in mechanical engineering, offering great potential for new innovative processes, and also as a novel hydraulic fluid. This paper focuses on the outstanding compressibility properties of ionic liquid EMIM-EtSO4, a very important physical chemically property when IL is used as a hydraulic fluid. This very low compressibility (respectively, very high Bulk modulus), compared to the classical hydraulic mineral oils or the non-flammable HFDU type of hydraulic fluids, opens up new possibilities regarding its usage within hydraulic systems with increased dynamics, respectively, systems' dynamic responses.

  8. Experimental Analysis of Hydraulic Fracture Growth and Acoustic Emission Response in a Layered Formation

    Science.gov (United States)

    Ning, Li; Shicheng, Zhang; Yushi, Zou; Xinfang, Ma; Shan, Wu; Yinuo, Zhang

    2018-04-01

    Microseismic/acoustic emission (AE) monitoring is an essential technology for understanding hydraulic fracture (HF) geometry and stimulated reservoir volume (SRV) during hydraulic fracturing in unconventional reservoirs. To investigate HF growth mechanisms and features of induced microseismic/AE events in a layered formation, laboratory fracturing experiments were performed on shale specimens (30 cm × 30 cm × 30 cm) with multiple bedding planes (BPs) under triaxial stresses. AE monitoring was used to reveal the spatial distribution and hypocenter mechanisms of AE events induced by rock failure. Computerized tomography scanning was used to observe the internal fracture geometry. Experimental results showed that the various HF geometries could be obviously distinguished based on injection pressure curves and AE responses. Fracture complexity was notably increased when vertically growing HFs connected with and opened more BPs. The formation of a complex fracture network was generally indicated by frequent fluctuations in injection pressure curves, intense AE activity, and three-dimensionally distributed AE events. Investigations of the hypocenter mechanisms revealed that shear failure/event dominated in shale specimens. Shear and tensile events were induced in hydraulically connected regions, and shear events also occurred around BPs that were not hydraulically connected. This led to an overestimation of HF height and SRV in layered formations based on the AE location results. The results also showed that variable injection rate and using plugging agent were conducive in promoting HF to penetrate through the weak and high-permeability BPs, thereby increasing the fracture height.

  9. 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.

  10. Hydraulic Hybrid Vehicle Publications | Transportation Research | NREL

    Science.gov (United States)

    Hydraulic Hybrid Vehicle Publications Hydraulic Hybrid Vehicle Publications The following technical papers and fact sheets provide information about NREL's hydraulic hybrid fleet vehicle evaluations . Refuse Trucks Project Startup: Evaluating the Performance of Hydraulic Hybrid Refuse Vehicles. Bob

  11. 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.

  12. 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)

  13. The influence of floc size and hydraulic detention time on the performance of a dissolved air flotation (DAF) pilot unit in the light of a mathematical model.

    Science.gov (United States)

    Moruzzi, R B; Reali, M A P

    2014-12-01

    The influence of floc size and hydraulic detention time on the performance of a dissolved air flotation (DAF) pilot unit was investigated in the light of a known mathematical model. The following design and operational parameters were considered: the hydraulic detention time (tdcz) and hydraulic loading rate in the contact zone, the down-flow loading rate in the clarification zone, the particle size distribution (d F), and the recirculation rate (p). As a reference for DAF performance analysis, the proposed β.td parameter from the above mentioned mathematical model was employed. The results indicated that tdcz is an important factor in DAF performance and that d F and floc size are also determinants of DAF efficiency. Further, β.td was sensitive to both design and operational parameters, which were varied in the DAF pilot plant. The performance of the DAF unit decreases with increasing β.td values because a higher td (considering a fixed β) or a higher β (e.g., higher hydrophobicity of the flocs for a fixed td) would be necessary in the reaction zone to reach desired flotation efficiency.

  14. 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

  15. 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

  16. 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.

  17. 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

  18. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-15

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

  20. Thermal-hydraulic performance of a water-cooled tungsten-rod target for a spallation neutron source

    International Nuclear Information System (INIS)

    Poston, D.I.

    1997-08-01

    A thermal-hydraulic (T-H) analysis is conducted to determine the feasibility and limitations of a water-cooled tungsten-rod target at powers of 1 MW and above. The target evaluated has a 10-cm x 10-cm cross section perpendicular to the beam axis, which is typical of an experimental spallation neutron source - both for a short-pulse spallation source and long-pulse spallation source. This report describes the T-H model and assumptions that are used to evaluate the target. A 1-MW baseline target is examined, and the results indicate that this target should easily handle the T-H requirements. The possibility of operating at powers >1 MW is also examined. The T-H design is limited by the condition that the coolant does not boil (actual limits are on surface subcooling and wall heat flux); material temperature limits are not approached. Three possible methods of enhancing the target power capability are presented: reducing peak power density, altering pin dimensions, and improving coolant conditions (pressure and temperature). Based on simple calculations, it appears that this target concept should have little trouble reaching the 2-MW range (from a purely T-H standpoint), and possibly much higher powers. However, one must keep in mind that these conclusions are based solely on thermal-hydraulics. It is possible, and perhaps likely, that target performance could be limited by structural issues at higher powers, particularly for a short-pulse spallation source because of thermal shock issues

  1. 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.

  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. Numerical simulation of the thermal hydraulic performance of a plate pin fin heat sink

    International Nuclear Information System (INIS)

    Yuan Wuhan; Zhao Jiyun; Tso, C.P.; Wu Tianhua; Liu Wei; Ming Tingzhen

    2012-01-01

    The computational fluid dynamic software FLUENT is used in assessing the electronics cooling potential of a plate pin fin heat sink (PPFHS), including the conjugate effect. The simulation results are validated with reported experimental data. The simulation shows that pin height and air velocity have significant influences on the thermal hydraulic performances of PPFHS while the influences of in-line/staggered array and neighbor pin flow-directional center distance (NPFDCD) of the PPFHS are less notable. In applying the present design to the cooling of a desktop PC CPU at a heat flux of 2.20 W/cm 2 , the temperature can be kept at less than 358 K with an air velocity over 6.5 m/s. - Highlights: ► Pin height and air velocity significantly influence thermal performance of PPFHS. ► Less influence by in-line or staggered array. ► Less influence by neighbor pin flow-directional center distance. ► Design with >6.5 m/s air can cool to 2 flux.

  4. Using automatic calibration method for optimizing the performance of Pedotransfer functions of saturated hydraulic conductivity

    Directory of Open Access Journals (Sweden)

    Ahmed M. Abdelbaki

    2016-06-01

    Full Text Available Pedotransfer functions (PTFs are an easy way to predict saturated hydraulic conductivity (Ksat without measurements. This study aims to auto calibrate 22 PTFs. The PTFs were divided into three groups according to its input requirements and the shuffled complex evolution algorithm was used in calibration. The results showed great modification in the performance of the functions compared to the original published functions. For group 1 PTFs, the geometric mean error ratio (GMER and the geometric standard deviation of error ratio (GSDER values were modified from range (1.27–6.09, (5.2–7.01 to (0.91–1.15, (4.88–5.85 respectively. For group 2 PTFs, the GMER and the GSDER values were modified from (0.3–1.55, (5.9–12.38 to (1.00–1.03, (5.5–5.9 respectively. For group 3 PTFs, the GMER and the GSDER values were modified from (0.11–2.06, (5.55–16.42 to (0.82–1.01, (5.1–6.17 respectively. The result showed that the automatic calibration is an efficient and accurate method to enhance the performance of the PTFs.

  5. Effects of transmembrane hydraulic pressure on performance of forward osmosis membranes.

    Science.gov (United States)

    Coday, Bryan D; Heil, Dean M; Xu, Pei; Cath, Tzahi Y

    2013-03-05

    Forward osmosis (FO) is an emerging membrane separation process that continues to be tested and implemented in various industrial water and wastewater treatment applications. The growing interests in the technology have prompted laboratories and manufacturers to adopt standard testing methods to ensure accurate comparison of membrane performance under laboratory-controlled conditions; however, standardized methods might not capture specific operating conditions unique to industrial applications. Experiments with cellulose triacetate (CTA) and polyamide thin-film composite (TFC) FO membranes demonstrated that hydraulic transmembrane pressure (TMP), common in industrial operation of FO membrane elements, could affect membrane performance. Experiments were conducted with three FO membranes and with increasing TMP up to a maximum of 50 psi (3.45 bar). The feed solution was a mixture of salts and the draw solution was either a NaCl solution or concentrated seawater at similar osmotic pressure. Results revealed that TMP minimally affected water flux, reverse salt flux (RSF), and solute rejection of the CTA membrane. However, water flux through TFC membranes might slightly increase with increasing TMP, and RSF substantially declines with increasing TMP. It was observed that rejection of feed constituents was influenced by TMP and RSF.

  6. 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

  7. 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)

  8. 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.

  9. 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)

  10. 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

  11. 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.

  12. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-10-15

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

  14. Tests Performed on Hydraulic Turbines at Commissioning or after Capital Repairs. Part II. Tests Performed on a 6.5 MW Kaplan Turbine

    Directory of Open Access Journals (Sweden)

    Adrian Cuzmoş

    2015-07-01

    Full Text Available The paper presents the tests performed on a hydraulic turbine on commissioning, the devices, test methods and the results obtained from the respective tests, as well as the conclusions and recommendations resulted from these tests. This kind of tests can be performed for the verification of guarantees.

  15. 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

  16. Thermal-hydraulic analysis of an innovative decay heat removal system for lead-cooled fast reactors

    International Nuclear Information System (INIS)

    Giannetti, Fabio; Vitale Di Maio, Damiano; Naviglio, Antonio; Caruso, Gianfranco

    2016-01-01

    Highlights: • LOOP thermal-hydraulic transient analysis for lead-cooled fast reactors. • Passive decay heat removal system concept to avoid lead freezing. • Solution developed for the diversification of the decay heat removal functions. • RELAP5 vs. RELAP5-3D comparison for lead applications. - Abstract: Improvement of safety requirements in GEN IV reactors needs more reliable safety systems, among which the decay heat removal system (DHR) is one of the most important. Complying with the diversification criteria and based on pure passive and very reliable components, an additional DHR for the ALFRED reactor (Advanced Lead Fast Reactor European Demonstrator) has been proposed and its thermal-hydraulic performances are analyzed. It consists in a coupling of two innovative subsystems: the radiative-based direct heat exchanger (DHX), and the pool heat exchanger (PHX). Preliminary thermal-hydraulic analyses, by using RELAP5 and RELAP5-3D© computer programs, have been carried out showing that the whole system can safely operate, in natural circulation, for a long term. Sensitivity analyses for: the emissivity of the DHX surfaces, the PHX water heat transfer coefficient (HTC) and the lead HTC have been carried out. In addition, the effects of the density variation uncertainty on the results has been analyzed and compared. It allowed to assess the feasibility of the system and to evaluate the acceptable range of the studied parameters. A comparison of the results obtained with RELAP5 and RELAP5-3D© has been carried out and the analysis of the differences of the two codes for lead is presented. The features of the innovative DHR allow to match the decay heat removal performance with the trend of the reactor decay heat power after shutdown, minimizing at the same time the risk of lead freezing. This system, proposed for the diversification of the DHR in the LFRs, could be applicable in the other pool-type liquid metal fast reactors.

  17. Thermal-hydraulic analysis of an innovative decay heat removal system for lead-cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Giannetti, Fabio; Vitale Di Maio, Damiano; Naviglio, Antonio; Caruso, Gianfranco, E-mail: gianfranco.caruso@uniroma1.it

    2016-08-15

    Highlights: • LOOP thermal-hydraulic transient analysis for lead-cooled fast reactors. • Passive decay heat removal system concept to avoid lead freezing. • Solution developed for the diversification of the decay heat removal functions. • RELAP5 vs. RELAP5-3D comparison for lead applications. - Abstract: Improvement of safety requirements in GEN IV reactors needs more reliable safety systems, among which the decay heat removal system (DHR) is one of the most important. Complying with the diversification criteria and based on pure passive and very reliable components, an additional DHR for the ALFRED reactor (Advanced Lead Fast Reactor European Demonstrator) has been proposed and its thermal-hydraulic performances are analyzed. It consists in a coupling of two innovative subsystems: the radiative-based direct heat exchanger (DHX), and the pool heat exchanger (PHX). Preliminary thermal-hydraulic analyses, by using RELAP5 and RELAP5-3D© computer programs, have been carried out showing that the whole system can safely operate, in natural circulation, for a long term. Sensitivity analyses for: the emissivity of the DHX surfaces, the PHX water heat transfer coefficient (HTC) and the lead HTC have been carried out. In addition, the effects of the density variation uncertainty on the results has been analyzed and compared. It allowed to assess the feasibility of the system and to evaluate the acceptable range of the studied parameters. A comparison of the results obtained with RELAP5 and RELAP5-3D© has been carried out and the analysis of the differences of the two codes for lead is presented. The features of the innovative DHR allow to match the decay heat removal performance with the trend of the reactor decay heat power after shutdown, minimizing at the same time the risk of lead freezing. This system, proposed for the diversification of the DHR in the LFRs, could be applicable in the other pool-type liquid metal fast reactors.

  18. Performance of a constructed wetland treating intensive shrimp aquaculture wastewater under high hydraulic loading rate

    International Nuclear Information System (INIS)

    Lin, Y.-F.; Jing, S.-R.; Lee, D.-Y.; Chang, Y.-F.; Chen, Y.-M.; Shih, K.-C.

    2005-01-01

    A water treatment unit, mainly consisting of free water surface (FWS) and subsurface flow (SF) constructed wetland cells, was integrated into a commercial-scale recirculating aquaculture system for intensive shrimp culture. This study investigated performance of the treatment wetlands for controlling water quality. The results showed that the FWS-SF cells effectively removed total suspended solids (55-66%), 5-day biochemical oxygen demand (37-54%), total ammonia (64-66%) and nitrite (83-94%) from the recirculating water under high hydraulic loading rates (1.57-1.95 m/day). This led to a water quality that was suitable for shrimp culture and effluent that always satisfied the discharge standards. The area ratios of wetlands to culture tank being demonstrated (0.43) and calculated (0.096) in this study were both significantly lower than the reported values. Accordingly, a constructed wetland was technically and economically feasible for managing water quality of an intensive aquaculture system. - A constructed wetland was found to be technically and economically feasible for managing water quality of an intensive recirculating aquaculture system

  19. Hydraulic performance of permeable barriers for in situ treatment of contaminated groundwater

    International Nuclear Information System (INIS)

    Smyth, D.J.A.; Shikaze, S.G.; Cherry, J.A.

    1997-01-01

    The passive interception and in situ treatment of dissolved contaminants in groundwater by permeable reactive barriers has recently gained favor at an increasing number of sites as an alternative to conventional approaches to groundwater remediation such as the pump-and-treat method. Permeable reactive barriers have two essential functions. The first is that the barriers must be installed in a position such that all of the plume passes through the reactive system. The second function is to achieve acceptable treatment of the contamination by physical, chemical or biological means within or downgradient of the barrier. In this paper, issues associated with the hydraulic performance of permeable reaction barriers are evaluated using a three-dimensional groundwater flow model. The efficiency of plume capture by permeable wall and funnel-and-gate systems is examined for some generic and for site-specific hydrogeologic systems. The results have important implications to decisions pertaining to the selection, design and installation of permeable reactive barrier systems

  20. Optimization and performance characteristics of servo-piston hydraulic control rod drive mechanism

    International Nuclear Information System (INIS)

    Yu Mingrui; Han Weishi; Wang Ge

    2014-01-01

    This paper introduces the structure and working principles of the servo-piston hydraulic control rod drive mechanism (SHCM), which can be moved continuously and has self-lock capacity. The steady state characteristics of SHCM are simulated using FLUENT codes. Based on comparison with the experimental results, the simulation is proven to be credible as a tool to describe the steady state characteristics. Finally, the influence of structural parameters is analyzed to obtain an optimal design. The experimental results indicate that the traction of the servo-tube is larger in the starting and braking stages. The resistance coefficient of SHCM increases gradually in the starting and lifting stage, and then tends to be stable. This coefficient has a maximum value while the inlet pressure is low. Performance norms of SHCM, such as the anti-disturbance ability and positioning accuracy, are tested, the anti-disturbance ability of the actuator is strong while the inlet pressure is fluctuating. The positioning accuracy is high regardless of the action process (lifting or not). (author)

  1. Experimental study on performance characteristics of servo-piston hydraulic control rod driving mechanism

    International Nuclear Information System (INIS)

    Yu Mingrui; Han Weishi; Zhou Jie; Liu Chunyu; Yang Zhida; Wang Ge

    2014-01-01

    An experimental study on the performance characteristics of the servo-piston hydraulic control rod driving mechanism is carried out, the dynamic processes of the driving mechanism are obtained through the experiments in different working conditions. Combined with the structure characteristics of the driving mechanism, the change rule between the characteristics parameters and the working condition is analyzed. The results indicate that the traction of the servo-tube decreases quickly at first, then slowly and finally trends to be a constant with the working pressure increasing, the tractions are the largest in the startup and deboost phases. The under pressure of the drive cylinder rises slowly and the upper pressure decreases rapidly at the beginning of the rise, the variation trend is opposite in the falling stage. There exists quick and clear flow change processes in the startup and deboost phases, the flow mutation value reduces and the mutation time changes a little with the working pressure increasing. The driving mechanism runs stable and has high sensitivity precision, the load does not vibrate at all when working conditions has small disturbance, a steady transform can be realized among every condition. (authors)

  2. Development of Nanoparticle-Stabilized Foams to Improve Performance of Water-less Hydraulic Fracturing

    Energy Technology Data Exchange (ETDEWEB)

    Prodanovic, Masa [The University of Texas at Austin; Johnston, Keith P. [The University of Texas at Austin

    2017-12-29

    We have successfully created ultra dry carbon-dioxide-in-water and nitrogen-in-water foams (with water content down to 2-5% range), that are remarkably stable at high temperatures (up to 120 deg, C) and pressures (up to 3000psi) and viscous enough (100-200 cP tunable range) to carry proppant. Two generations of these ultra-dry foams have been developed; they are stabilized either with a synergy of surfactants and nanoparticle, or just with viscoelastic surfactants that viscosify the aqueous phase. Not only does this reduce water utilization and disposal, but it minimizes fluid blocking of hydrocarbon production. Further, the most recent development shows successful use of environmentally friendly surfactants at high temperature and pressure. We pay special attention to the role of nanoparticles in stabilization of the foams, specifically for high salinity brines. The preliminary numerical simulation for which shows they open wider fractures with shorter half-length and require less clean-up due to minimal water use. We also tested the stability and sand carrying properties of these foams at high pressure, room temperature conditions in sapphire cell. We performed on a preliminary numerical investigation of applicability for improved oil recovery applications. The applicability was evaluated by running multiphase flow injection simulations in a case-study oil reservoir. The results of this research thus expand the options available to operators for hydraulic fracturing and can simplify the design and field implementation of foamed fracturing fluids.

  3. 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

  4. 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.

  5. 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)

  6. 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

  7. 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)

  8. 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)

  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. 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.

  11. High Bulk Modulus of Ionic Liquid and Effects on Performance of Hydraulic System

    Directory of Open Access Journals (Sweden)

    Milan Kambic

    2014-01-01

    Full Text Available Over recent years ionic liquids have gained in importance, causing a growing number of scientists and engineers to investigate possible applications for these liquids because of their unique physical and chemical properties. Their outstanding advantages such as nonflammable liquid within a broad liquid range, high thermal, mechanical, and chemical stabilities, low solubility for gases, attractive tribological properties (lubrication, and very low compressibility, and so forth, make them more interesting for applications in mechanical engineering, offering great potential for new innovative processes, and also as a novel hydraulic fluid. This paper focuses on the outstanding compressibility properties of ionic liquid EMIM-EtSO4, a very important physical chemically property when IL is used as a hydraulic fluid. This very low compressibility (respectively, very high Bulk modulus, compared to the classical hydraulic mineral oils or the non-flammable HFDU type of hydraulic fluids, opens up new possibilities regarding its usage within hydraulic systems with increased dynamics, respectively, systems’ dynamic responses.

  12. 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

  13. SMART performance analysis methodology

    International Nuclear Information System (INIS)

    Lim, H. S.; Kim, H. C.; Lee, D. J.

    2001-04-01

    To ensure the required and desired operation over the plant lifetime, the performance analysis for the SMART NSSS design is done by means of the specified analysis methodologies for the performance related design basis events(PRDBE). The PRDBE is an occurrence(event) that shall be accommodated in the design of the plant and whose consequence would be no more severe than normal service effects of the plant equipment. The performance analysis methodology which systematizes the methods and procedures to analyze the PRDBEs is as follows. Based on the operation mode suitable to the characteristics of the SMART NSSS, the corresponding PRDBEs and allowable range of process parameters for these events are deduced. With the developed control logic for each operation mode, the system thermalhydraulics are analyzed for the chosen PRDBEs using the system analysis code. Particularly, because of different system characteristics of SMART from the existing commercial nuclear power plants, the operation mode, PRDBEs, control logic, and analysis code should be consistent with the SMART design. This report presents the categories of the PRDBEs chosen based on each operation mode and the transition among these and the acceptance criteria for each PRDBE. It also includes the analysis methods and procedures for each PRDBE and the concept of the control logic for each operation mode. Therefore this report in which the overall details for SMART performance analysis are specified based on the current SMART design, would be utilized as a guide for the detailed performance analysis

  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. 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.

  17. 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)

  18. Thermal-hydraulic analysis for the LBE-cooled natural circulation reactor. Development of the MSG-COPD code and application to the system analysis. Research Document

    International Nuclear Information System (INIS)

    Iwasaki, Takashi; Sakai, Takaaki; Enuma, Yasuhiro; Mizuno, Tomoyasu

    2002-03-01

    Thermal-hydraulic analysis for the Lead-Bismuth eutectic (LBE)-cooled natural circulation reactor has been conducted by using a combined plant dynamics code (MSG-COPD). MSG-COPD has been developed to consider the multi-dimensional thermal-hydraulics effect on the plant dynamics during transients. Plant dynamics analyses for the LBE-cooled STAR-LM reactor, which has been designed by Argonne National Laboratory in U.S.A., have been performed to understand the basic thermal-hydraulic characteristics of the natural circulation reactor. As a result, it has been made clear that cold coolant remains in the lower plenum by the thermal stratification in case of the ULOHS condition with a severe temperature gradient at the stratified surface in the lower plenum. In addition, the flow-redistribution effect in a core channels by the buoyancy force has been evaluated for a candidate LBE-cooled FBR plant concept (LBE-FR), which has been designed by JNC. A linear evaluation method for the flow-redistribution coefficient is proposed for the LBE-FR, and compared with the multi-dimensional results by MSG-COPD. In conclusion, the method shows sufficient performance for the prediction of the flow-redistribution coefficient for typical lateral power distributions in the core. (author)

  19. 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

  20. Analysis of the OECD/NRC BWR Turbine Trip Transient Benchmark with the Coupled Thermal-Hydraulics and Neutronics Code TRAC-M/PARCS

    International Nuclear Information System (INIS)

    Lee, Deokjung; Downar, Thomas J.; Ulses, Anthony; Akdeniz, Bedirhan; Ivanov, Kostadin N.

    2004-01-01

    An analysis of the Peach Bottom Unit 2 Turbine Trip 2 (TT2) experiment has been performed using the U.S. Nuclear Regulatory Commission coupled thermal-hydraulics and neutronics code TRAC-M/PARCS. The objective of the analysis was to assess the performance of TRAC-M/PARCS on a BWR transient with significance in two-phase flow and spatial variations of the neutron flux. TRAC-M/PARCS results are found to be in good agreement with measured plant data for both steady-state and transient phases of the benchmark. Additional analyses of four fictitious extreme scenarios are performed to provide a basis for code-to-code comparisons and comprehensive testing of the thermal-hydraulics/neutronics coupling. The obtained results of sensitivity studies on the effect of direct moderator heating on transient simulation indicate the importance of this modeling aspect

  1. Effects of Hydraulic Retention Time on the Performance of a Membrane Bioreactor Treating Municipal Wastewater

    Directory of Open Access Journals (Sweden)

    Hadi Falahati

    2017-09-01

    Full Text Available There is an increasing demand for effluents of higher quality from wastewater treatment plants due to the more stringent quality standards as well as the increasing pressure on water resources worldwide, which calls for effluent recycle and reuse. Membrane bioreactors (MBRs have been recently gaining rapid popularity as a promising technology for wastewater treatment. In order to improve the quality of the effluent from Shiraz wastewater treatment plant, an on-site pilot-scale membrane bioreactor was operated for 9 months. The pilot plant built at Shiraz wastewater treatment plant consisted of an aerobic reactor and a membrane compartment containing one submerged hollow fiber membrane module. In this study, eleven different aerobic hydraulic retention times (HRT ranging from 2 to 12 hours were tested to determine the membrane capacity and to investigate the performance of the system in removing total ammonia nitrogen, organic matter, total suspended solids, and turbidity.The system recorded a perfectly stable removal efficiency over the whole experimental period, except for the 2-hour aerobic HRT, so that its COD and BOD reductions exceeded 95% and 99%, respectively. Moreover, the system achieved complete nitrification in a stable manner during the whole study period, except for the 2-hour aerobic operation period. TSS concentration was almost zero and turbidity was less than 1 NTU. Membrane capacity measurements showed an average flux of 5.5 Lm-2h-1 with a mean trans-membrane pressure difference of 30 kPa. Results showed that the MBR outperformed the conventional sewage treatment processes. Additionally, it was not affected by aerobic HRT changes (12, 10, 8, 6, 4, and 3h. Based on the effluent qulity, teh system may be recommended for application toward water reuse in industrial and agricultural settings

  2. Effects of Volute Design and Number of Impeller Blades on Lateral Impeller Forces and Hydraulic Performance

    Directory of Open Access Journals (Sweden)

    Daniel O. Baun

    2003-01-01

    Full Text Available A comparison is made between the characteristics of the measured lateral impeller forces and the hydraulic performances of a four- and a five-vane impeller, each operating in a spiral volute, a concentric volute, and a double volute. The pump's rotor was supported in magnetic bearings. In addition to supporting and controlling the rotor motion, the magnetic bearings also served as active load cells and were used to measure the impeller forces acting on the pump's rotor. The lateral impeller force characteristics, as a function of a normalized flow coefficient, were virtually identical in the four- and five-vane impellers in each respective volute type. The measured impeller forces for each volute type were compared with correlations in the literature. The measured forces from the double volute configurations agreed with the forces from a correlation model over the full flow range. Single volute configurations compared well with the predictions of a published correlation at high flow rates, ϕ/ϕn>0.5. Concentric volute configurations compared well with a published correlation at low flow rates, ϕ/ϕn<0.4. The head-versus-flow characteristics of the four-vane impeller in each volute type were stable over a greater flow range than the corresponding characteristics of the five-vane impeller. At higher flow rates in the stable region of the head's characteristic curves near the best efficiency point, the five-vane impeller produced higher head than did the four-vane impeller in each volute type.

  3. Thermal-hydraulic analysis of research reactor core with different LEU fuel types using RELAP5

    Energy Technology Data Exchange (ETDEWEB)

    El-Sahlamy, Neama M. [Nuclear and Radiological Regulatory Authority, Cairo (Egypt)

    2017-11-15

    In the current work, comparisons between the core performances when using different LEU fuels are done. The fuels tested are UA1{sub X}-A1, U{sub 3}O{sub 8}-Al, and U{sub 3}Si{sub 2}-Al fuels with 19.7 % enrichment. Calculations are done using RELAP5 code to evaluate the thermal-hydraulic performance of the IAEA benchmark 10 MW reactor. First, a reassessment of the slow reactivity insertion transient with UA1{sub X}-A1 LEU fuel to compare the results with those reported in the IAEA TECDOC [1]. Then, comparisons between the thermal-hydraulic core performances when using the three LEU fuels are done. The assessment is performed at initial power of 1.0 W. The reactor power is calculated using the RELAP5 point kinetic model. The reactivity feedback, from changes in water density and fuel temperature, is considered for all cases. From the results it is noticed that U{sub 3}Si{sub 2}-Al fuel gives the best fuel performance since it has the minimum value of peak fuel temperature and the minimum peak clad surface temperature, as operating parameters. Also, it gives the maximum value of the Critical Heat Flux Ratio and the lowest tendency to flow instability occurrence.

  4. Performance Analysis of MYSEA

    Science.gov (United States)

    2012-09-01

    Services FSD Federated Services Daemon I&A Identification and Authentication IKE Internet Key Exchange KPI Key Performance Indicator LAN Local Area...spection takes place in different processes in the server architecture. Key Performance Indica- tor ( KPI )s associated with the system need to be...application and risk analysis of security controls. Thus, measurement of the KPIs is needed before an informed tradeoff between the performance penalties

  5. A numerical study on improving the thermal hydraulic performance of printed circuit heat exchanger using the supercritical carbon dioxide

    Energy Technology Data Exchange (ETDEWEB)

    Park, Bo Guen; Kim, Dae Hyun; Chung, Jin Taek [Dept. of Mechanical Engineering, Korea University, Seoul (Korea, Republic of)

    2015-10-15

    The objective of this study is to propose a new channel shape that improves thermal-hydraulic performance. The existing Zigzag channel has high pressure loss due to flow separation and reverse flow. To improve this disadvantage, partial straight channel is inserted into bended points. Also, the effects of straight channel's length change on heat transfer and pressure loss are analyzed. Thermal-hydraulic performance of the new shape and existing Zigzag channel are quantitatively compared in terms of Goodness Factor. Mass flow rate was changed from 1.41 x 10{sup -4} kg/s to 2.48 x 10{sup -4} kg/s . The average volume goodness factor of 1mm straight channel shape was increased by 25% compared to the Zigzag channel.

  6. IAEA Coordinated Research Project on HTGR Reactor Physics, Thermal-hydraulics and Depletion Uncertainty Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bostelmann, F. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    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 (SA) and uncertainty analysis (UA) methods. Uncertainty originates from errors in physical data, manufacturing uncertainties, modelling and computational algorithms. (The interested reader is referred to the large body of published SA and UA literature for a more complete overview of the various types of uncertainties, methodologies and results obtained). SA is helpful for ranking the various sources of uncertainty and error in the results of core analyses. SA and UA are required to address cost, safety, and licensing needs and should be applied to all aspects of reactor multi-physics simulation. SA and UA can guide experimental, modelling, and algorithm research and development. Current SA and UA rely either on derivative-based methods such as stochastic sampling methods or on generalized perturbation theory to obtain sensitivity coefficients. Neither approach addresses all needs. In order to benefit from recent advances in modelling 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. Only a parallel effort in advanced simulation and in nuclear data improvement will be able to provide designers with more robust and well validated calculation tools to meet design target accuracies. In February 2009, the Technical Working Group on Gas-Cooled Reactors (TWG-GCR) of the International Atomic Energy Agency (IAEA) recommended that the proposed Coordinated Research Program (CRP) on

  7. ANTEO+: A subchannel code for thermal-hydraulic analysis of liquid metal cooled systems

    Energy Technology Data Exchange (ETDEWEB)

    Lodi, F., E-mail: francesco.lodi5@unibo.it [DIN – Laboratory of Montecuccolino, University of Bologna, Via dei Colli 16, 40136 Bologna (Italy); Grasso, G., E-mail: giacomo.grasso@enea.it [ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development), via Martiri di Monte Sole, 4, 40129 Bologna (Italy); Mattioli, D., E-mail: davide.mattioli@enea.it [ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development), via Martiri di Monte Sole, 4, 40129 Bologna (Italy); Sumini, M., E-mail: marco.sumini@unibo.it [DIN – Laboratory of Montecuccolino, University of Bologna, Via dei Colli 16, 40136 Bologna (Italy)

    2016-05-15

    Highlights: • The code structure is presented in detail. • The performed validation is outlined. • Results are critically discussed assessing code accuracy. • Conclusions are drawn and ground for future work identified. - Abstract: Liquid metal cooled fast reactors are promising options for achieving the high degrees of safety and sustainability demanded by the Generation IV paradigm. Among the critical aspects to be addressed in the design process, thermal-hydraulics is one of the most challenging; in order to embed safety in the core conceptualization, these aspects are to be considered at the very beginning of the design process, and translated in a design perspective. For achieving these objectives the subchannel code ANTEO+ has been conceived, able to simulate pin bundle arrangements cooled by liquid metals. The main purposes of ANTEO+ are simplifying the problem description maintaining the required accuracy, enabling a more transparent interface with the user, and having a clear and identifiable application domain, in order to help the user interpreting the results and, mostly, defining their confidence. Since ANTEO+ relies on empirical correlations, the validation phase is of paramount importance along with a clear discussion on the simplifications adopted in modeling the conservation equations. In the present work a detailed description of ANTEO+ structure is given along with a thorough validation of the main models implemented for flow split, pressure drops and subchannel temperatures. The analysis confirmed the ability of ANTEO+ in reproducing experimental data in its anticipated validity domain, with a relatively high degree of accuracy when compared to other classical subchannel tools like ENERGY-II, COBRA-IV-I-MIT and BRS-TVS.

  8. 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)

  9. Responses of Woody Plant Functional Traits to Nitrogen Addition: A Meta-Analysis of Leaf Economics, Gas Exchange, and Hydraulic Traits.

    Science.gov (United States)

    Zhang, Hongxia; Li, Weibin; Adams, Henry D; Wang, Anzhi; Wu, Jiabing; Jin, Changjie; Guan, Dexin; Yuan, Fenghui

    2018-01-01

    Atmospheric nitrogen (N) deposition has been found to significantly affect plant growth and physiological performance in terrestrial ecosystems. Many individual studies have investigated how N addition influences plant functional traits, however these investigations have usually been limited to a single species, and thereby do not allow derivation of general patterns or underlying mechanisms. We synthesized data from 56 papers and conducted a meta-analysis to assess the general responses of 15 variables related to leaf economics, gas exchange, and hydraulic traits to N addition among 61 woody plant species, primarily from temperate and subtropical regions. Results showed that under N addition, leaf area index (+10.3%), foliar N content (+7.3%), intrinsic water-use efficiency (+3.1%) and net photosynthetic rate (+16.1%) significantly increased, while specific leaf area, stomatal conductance, and transpiration rate did not change. For plant hydraulics, N addition significantly increased vessel diameter (+7.0%), hydraulic conductance in stems/shoots (+6.7%), and water potential corresponding to 50% loss of hydraulic conductivity ( P 50 , +21.5%; i.e., P 50 became less negative), while water potential in leaves (-6.7%) decreased (became more negative). N addition had little effect on vessel density, hydraulic conductance in leaves and roots, or water potential in stems/shoots. N addition had greater effects on gymnosperms than angiosperms and ammonium nitrate fertilization had larger effects than fertilization with urea, and high levels of N addition affected more traits than low levels. Our results demonstrate that N addition has coupled effects on both carbon and water dynamics of woody plants. Increased leaf N, likely fixed in photosynthetic enzymes and pigments leads to higher photosynthesis and water use efficiency, which may increase leaf growth, as reflected in LAI results. These changes appear to have downstream effects on hydraulic function through increases

  10. Thermal-hydraulic simulation and analysis of Research Reactor Cooling Systems

    International Nuclear Information System (INIS)

    EL Khatib, H.H.A.

    2013-01-01

    The objective of the present study is to formulate a model to simulate the thermal hydraulic behavior of integrated cooling system in a typical material testing reactor (MTR) under loss of ultimate heat sink, the model involves three interactively coupled sub-models for reactor core, heat exchanger and cooling tower. The developed model predicts the temperature profiles in addition it predicts inlet and outlet temperatures of the hot and cold stream as well as the heat exchangers and cooling tower. The model is validated against PARET code for steady-state operation and also verified by the reactor operational records, and then the model is used to simulate the thermal-hydraulic behavior of the reactor under a loss of ultimate heat sink. The simulation is performed for two operational regimes named regime I of (11 MW) thermal power and three operated cooling tower cells and regime II of (22 MW) thermal power and six operated cooling tower cells. In regime I, the simulation is performed for 1, 2 and 3 cooling tower failed cells while in regime II, it is performed for 1, 2, 3, 4, 5 and 6 cooling tower failed cells. The safety action is conducted by the reactor protection system (RPS) named power reduction safety action, it is triggered to decrease the reactor power by amount of 20% of the present power when the water inlet temperature to the core reaches 43 degree C and a scram (emergency shutdown) is triggered in case of the inlet temperature reaches 44 degree C. The model results are analyzed and discussed. The temperature profiles of fuel, clad and coolant are predicted during transient where its maximum values are far from thermal hydraulic limits.

  11. 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)

  12. Comparison of thermo-hydraulic analysis with measurements for HELIOS. The scaled integral test loop for PEACER

    International Nuclear Information System (INIS)

    Cho, Jae Hyun; Lim, Jun; Kim, Ji Hak; Hwang, Il Soon

    2009-01-01

    A scaled-down Lead-Bismuth Eutectic circulating integral test loop named as HELIOS (Heavy Eutectic liquid metal Loop for Integral test of Operability and Safety of PEACER) has been employed to characterize steady-state isothermal forced circulation behavior and non-isothermal natural circulation capability of the lead and lead-alloy cooled advanced nuclear energy systems (LACANES). In this time, thermal-hydraulic experiments have been carried out using HELIOS following rigorous calibration campaigns on sensors for temperature and pressure, especially isothermal steady-state forced convection using by the pump. The isothermal steady-state forced convection test was performed to obtain the pressure loss information including friction loss coefficients and form loss coefficients. Then its data were compared with multi-approaching analysis including hand calculation results and computer simulation code results. (MARS-LBE, CFX). We report the results of comparisons between the analysis and measurements together. (author)

  13. Application of sensitivity analysis to a simplified coupled neutronic thermal-hydraulics transient in a fast reactor using Adjoint techniques

    International Nuclear Information System (INIS)

    Gilli, L.; Lathouwers, D.; Kloosterman, J.L.; Van der Hagen, T.H.J.J.

    2011-01-01

    In this paper a method to perform sensitivity analysis for a simplified multi-physics problem is presented. The method is based on the Adjoint Sensitivity Analysis Procedure which is used to apply first order perturbation theory to linear and nonlinear problems using adjoint techniques. The multi-physics problem considered includes a neutronic, a thermo-kinetics, and a thermal-hydraulics part and it is used to model the time dependent behavior of a sodium cooled fast reactor. The adjoint procedure is applied to calculate the sensitivity coefficients with respect to the kinetic parameters of the problem for two reference transients using two different model responses, the results obtained are then compared with the values given by a direct sampling of the forward nonlinear problem. Our first results show that, thanks to modern numerical techniques, the procedure is relatively easy to implement and provides good estimation for most perturbations, making the method appealing for more detailed problems. (author)

  14. 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

  15. Hydraulic/mechanical modeling of smectitic materials for HMC analytical evaluation of the long term performance of TRU geological repository - 59090

    International Nuclear Information System (INIS)

    Kobayashi, Ichizo; Owada, Hitoshi; Ishii, Tomoko

    2012-01-01

    Aiming at evaluation of the long term performance of transuranic (TRU) geological repositories, the hydraulic/ mechanical/chemical (HMC) analysis method has been studied. In this phase of research (four years) the hydraulic/mechanical modeling of smectitic materials for HMC analyses has been studied. In this paper, new experimental methods for investigation of the hydraulic/mechanical behavior of smectitic materials were developed. For hydraulic modeling, the measurement method of the specific surface area of compacted smectitic materials was developed using X-ray diffraction (XRD). The results of the method were applied to the Kozeny-Carman law. Since the specific surface area represents the microstructure of smectitic materials such as the degree of swelling, it was found that the Kozeny-Carman law using measured specific surface area of compacted smectitic materials was useful in evaluating the hydraulic performance of smectitic materials. Moreover, since the Kozeny-Carman law can take the alteration of content of pore water into consideration by not only a coefficient of viscosity but also by changes in specific surface area, the Kozeny-Carman law will be more suitable to chemical and mechanical couple analyses than the ordinary Darcy's law. For the mechanical modeling, the procedure of one dimensional exhausting compression test was developed. The tests gave the dry density and compression stress relation in the state of full-saturation of smectitic materials with varying water content. The relations between the dry density and compression stress in the state of fully saturation were termed fully saturation lines. The group of iso-grams of degree of saturation and water content were also given with this test. It was found that the fully-saturation line is consistent with swelling deformation-pressure relation in the equilibrium state. The results indicated that the swelling deformation-pressure relation does not depend on the saturation manner, such as the

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

    International Nuclear Information System (INIS)

    Haapalehto, T.

    1995-01-01

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

  17. 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...

  18. 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.

  19. 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

  20. 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

  1. Thermal-hydraulics analysis of a PWR reactor using zircaloy and carbide silicon reinforced with type S fibers as fuel claddings: Simulation of a channel blockage transient

    Energy Technology Data Exchange (ETDEWEB)

    Matuck, Vinicius; Ramos, Mario C.; Faria, Rochkhudson B.; Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia, E-mail: rochkdefaria@yahoo.com.br, E-mail: matuck747@gmail.com, E-mail: patricialire@yahoo.com.br, E-mail: marc5663@gmail.com, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte (Brazil). Departamento de Engenharia Nuclear

    2017-11-01

    A detailed thermal-hydraulic reactor model using as reference data from the Angra 2 Final Safety Analysis Report (FSAR) has been developed and SiC reinforced with Hi-Nicalon type S fibers (SiC HNS) was used as fuel cladding. The goal is to compare its behavior from the thermal viewpoint with the Zircaloy, at the steady- state and transient conditions. The RELAP-3D was used to perform the thermal-hydraulic analysis and a blockage transient has been investigated at full power operation. The transient considered is related to total obstruction of a core cooling channel of one fuel assembly. The calculations were performed using a point kinetic model. The reactor behavior after this transient was analyzed and the time evolution of cladding and coolant temperatures mass flow and void fraction are presented. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-12-15

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

  3. Feasibility study on thermal-hydraulic performance in tight-lattice rod bundles for reduced-moderation water reactors

    International Nuclear Information System (INIS)

    Ohnuki, A.; Kureta, M.; Liu, W.; Tamai, H.; Akimoto, H.

    2004-01-01

    Research and development project for investigating thermal-hydraulic performance in tight-lattice rod bundles for Reduced-Moderation Water Reactor (RMWR) started at Japan Atomic Energy Research Institute (JAERI) in 2002. The RMWR can attain the favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured light-water reactor technologies. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio by reducing the moderation of neutron. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. The confirmation of thermal-hydraulic feasibility is one of the most important issues for the RMWR because of the tight-lattice configuration. The project has mainly consisted of a large-scale thermal-hydraulic test and development of analytical methods named modeling engineering. In the large-scale test, 37-rod bundle experiments can be performed. Steady-state critical power experiments have been achieved in the test facility and the experimental data reveal the feasibility of RMWR

  4. 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

  5. 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.

  6. 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

  7. 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)

  8. 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)

  9. 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.

  10. 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)

  11. 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.

  12. 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)

  13. Uncertainty and Sensitivity Analysis Applied to the Validation of BWR Bundle Thermal-Hydraulic Calculations

    International Nuclear Information System (INIS)

    Hernandez-Solis, Augusto

    2010-04-01

    This work has two main objectives. The first one is to enhance the validation process of the thermal-hydraulic features of the Westinghouse code POLCA-T. This is achieved by computing a quantitative validation limit based on statistical uncertainty analysis. This validation theory is applied to some of the benchmark cases of the following macroscopic BFBT exercises: 1) Single and two phase bundle pressure drops, 2) Steady-state cross-sectional averaged void fraction, 3) Transient cross-sectional averaged void fraction and 4) Steady-state critical power tests. Sensitivity analysis is also performed to identify the most important uncertain parameters for each exercise. The second objective consists in showing the clear advantages of using the quasi-random Latin Hypercube Sampling (LHS) strategy over simple random sampling (SRS). LHS allows a much better coverage of the input uncertainties than SRS because it densely stratifies across the range of each input probability distribution. The aim here is to compare both uncertainty analyses on the BWR assembly void axial profile prediction in steady-state, and on the transient void fraction prediction at a certain axial level coming from a simulated re-circulation pump trip scenario. It is shown that the replicated void fraction mean (either in steady-state or transient conditions) has less variability when using LHS than SRS for the same number of calculations (i.e. same input space sample size) even if the resulting void fraction axial profiles are non-monotonic. It is also shown that the void fraction uncertainty limits achieved with SRS by running 458 calculations (sample size required to cover 95% of 8 uncertain input parameters with a 95% confidence), result in the same uncertainty limits achieved by LHS with only 100 calculations. These are thus clear indications on the advantages of using LHS. Finally, the present study contributes to a realistic analysis of nuclear reactors, in the sense that the uncertainties of

  14. 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

  15. An Improved Rate-Transient Analysis Model of Multi-Fractured Horizontal Wells with Non-Uniform Hydraulic Fracture Properties

    Directory of Open Access Journals (Sweden)

    Youwei He

    2018-02-01

    Full Text Available Although technical advances in hydraulically fracturing and drilling enable commercial production from tight reservoirs, oil/gas recovery remains at a low level. Due to the technical and economic limitations of well-testing operations in tight reservoirs, rate-transient analysis (RTA has become a more attractive option. However, current RTA models hardly consider the effect of the non-uniform production on rate decline behaviors. In fact, PLT results demonstrate that production profile is non-uniform. To fill this gap, this paper presents an improved RTA model of multi-fractured horizontal wells (MFHWs to investigate the effects of non-uniform properties of hydraulic fractures (production of fractures, fracture half-length, number of fractures, fracture conductivity, and vertical permeability on rate transient behaviors through the diagnostic type curves. Results indicate obvious differences on the rate decline curves among the type curves of uniform properties of fractures (UPF and non-uniform properties of fractures (NPF. The use of dimensionless production integral derivative curve magnifies the differences so that we can diagnose the phenomenon of non-uniform production. Therefore, it’s significant to incorporate the effects of NPF into the RDA models of MFHWs, and the model proposed in this paper enables us to better evaluate well performance based on long-term production data.

  16. Feeding kinematics, suction, and hydraulic jetting performance of harbor seals (Phoca vitulina.

    Directory of Open Access Journals (Sweden)

    Christopher D Marshall

    Full Text Available The feeding kinematics, suction and hydraulic jetting capabilities of captive harbor seals (Phoca vitulina were characterized during controlled feeding trials. Feeding trials were conducted using a feeding apparatus that allowed a choice between biting and suction, but also presented food that could be ingested only by suction. Subambient pressure exerted during suction feeding behaviors was directly measured using pressure transducers. The mean feeding cycle duration for suction-feeding events was significantly shorter (0.15±0.09 s; P<0.01 than biting feeding events (0.18±0.08 s. Subjects feeding in-water used both a suction and a biting feeding mode. Suction was the favored feeding mode (84% of all feeding events compared to biting, but biting comprised 16% of feeding events. In addition, seals occasionally alternated suction with hydraulic jetting, or used hydraulic jetting independently, to remove fish from the apparatus. Suction and biting feeding modes were kinematically distinct regardless of feeding location (in-water vs. on-land. Suction was characterized by a significantly smaller gape (1.3±0.23 cm; P<0.001 and gape angle (12.9±2.02°, pursing of the rostral lips to form a circular aperture, and pursing of the lateral lips to occlude lateral gape. Biting was characterized by a large gape (3.63±0.21 cm and gape angle (28.8±1.80°; P<0.001 and lip curling to expose teeth. The maximum subambient pressure recorded was 48.8 kPa. In addition, harbor seals were able to jet water at food items using suprambient pressure, also known as hydraulic jetting. The maximum hydraulic jetting force recorded was 53.9 kPa. Suction and hydraulic jetting where employed 90.5% and 9.5%, respectively, during underwater feeding events. Harbor seals displayed a wide repertoire of behaviorally flexible feeding strategies to ingest fish from the feeding apparatus. Such flexibility of feeding strategies and biomechanics likely forms the basis of their

  17. The development and calibration of a physical model to assist in optimising the hydraulic performance and design of maturation ponds.

    Science.gov (United States)

    Aldana, G J; Lloyd, B J; Guganesharajah, K; Bracho, N

    2005-01-01

    A physical and a computational fluid dynamic (CFD) model (HYDRO-3D) were developed to simulate the effects of novel maturation pond configurations, and critical environmental factors (wind speed and direction) on the hydraulic efficiency (HE) of full-scale maturation ponds. The aims of the study were to assess the reliability of the physical model and convergence with HYDRO-3D, as tools for assessing and predicting best hydraulic performance of ponds. The physical model of the open ponds was scaled to provide a similar nominal retention time (NRT) of 52 hours. Under natural conditions, with a variable prevailing westerly wind opposite to the inlet, a rhodamine tracer study on the full-scale prototype pond produced a mean hydraulic retention time (MHRT) of 18.5 hours (HE = 35.5%). Simulations of these wind conditions, but with constant wind speed and direction in both the physical model and HYDRO-3D, produced a higher MHRT of 21 hours in both models and an HE of 40.4%. In the absence of wind tracer studies in the open pond physical model revealed incomplete mixing with peak concentrations leaving the model in several hours, but an increase in MHRT to 24.5-28 hours (HE = 50.2-57.1%). Although wind blowing opposite to the inlet flow increases dispersion (mixing), it reduced hydraulic performance by 18-25%. Much higher HE values were achieved by baffles (67-74%) and three channel configurations (69-92%), compared with the original open pond configuration. Good agreement was achieved between the two models where key environmental and flow parameters can be controlled and set, but it is difficult to accurately simulate full-scale works conditions due to the unpredictability of natural hourly and daily fluctuation in these parameters.

  18. Protective coating as a factor to ensure the strength and hydraulic performance of recoverable pipelines

    Directory of Open Access Journals (Sweden)

    Orlov Vladimir Aleksandrovich

    2015-01-01

    Full Text Available The authors present an analysis of various types of internal protective pipeline coatings to ensure the strength and hydraulic characteristics of a remodeled pipeline and related coating methods for effective trenchless renovation of engineering systems, water supply systems and sanitation. As protective coating the authors considered a round profile tube of a smaller diameter than of the old pipe, close to the old pipe, sprayed lining on the basis of inorganic and inorganic materials. The article analyzes the methods of trenchless renovation for applying protective coatings: routing in the old pipeline of new pipes made of polymeric materials or polymeric sleeves, centrifugal spraying on the inner surface of pipelines’ inorganic and organic protective coatings. Special attention was paid to bag technology, providing the required strength properties at specific values of the modulus of elasticity and a number of external factors such as the depth of the existing pipe, the existence and magnitude of the horizon groundwater over it. Also attention is paid to the application technology of tape coatings ribbed profile on the inner surface of pipelines. This technology has a unique feature, which is the ability of recoverable pipeline functioning during its renovation by winding an endless belt and the formation of a new pipe. The tape coating winding is carried out by different types of spiral winding machines. The thickness of the protective coating layer forming the tube remains minimal. Inorganic cement-sand and organic coatings were considered as alternative options for repair of pipelines, which allow to localize the defects in the form of a fistula, minor cracks and other damages. However it is noted that a cement-sandy covering is inferior to organic, because it does not provide the strength characteristics of the pipeline system. The main advantage of the organic coating is mudding fistula of a large diameter, making a high wear

  19. 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

  20. Numerical comparisons of the performance of a hydraulic coupling with different pump rotational speeds

    International Nuclear Information System (INIS)

    Luo, Y; Feng, L H; Liu, S H; Chen, T J; Fan, H G

    2013-01-01

    A hydraulic coupling is a hydrodynamic device for transmitting rotating mechanical power. It is widely used in the machinery industry because of its advantages of high energy transmission efficiency, shock absorption and good adaptability, etc. In this paper, SIMPLEC algorithm and SST k-ω turbulence model were employed to simulate the steady state flows at operating conditions of two different rotational speeds (3000r/min and 7500 r/min) of the pump of a specified hydraulic coupling model. The results indicate the existence of similarity in the distributions of the flow fields between the two speeds, but the efficiency at the optimum condition is larger with higher rotational speed. It is concluded that the similarity principle of the efficiency of the hydraulic couplings does not apply in this case due to the relatively high rotating speed and small geometric specifications. It is also shown that the radially stratified pressure distribution on the torus section becomes more obvious with larger speed ratios, since the centrifugal movement plays more dominant roles over the circulating movement in these situations. When the speed ratio is small, with the completion of the circulating flow, the pressure distribution presents in a more circular pattern around the neutral zone of the torus section

  1. Development of a coupled neutronic/thermal-hydraulic tool with multi-scale capabilities and applications to HPLWR core analysis

    International Nuclear Information System (INIS)

    Monti, Lanfranco; Starflinger, Joerg; Schulenberg, Thomas

    2011-01-01

    Highlights: → Advanced analysis and design techniques for innovative reactors are addressed. → Detailed investigation of a 3 pass core design with a multi-physics-scales tool. → Coupled 40-group neutron transport/equivalent channels TH core analyses methods. → Multi-scale capabilities: from equivalent channels to sub-channel pin-by-pin study. → High fidelity approach: reduction of conservatism involved in core simulations. - Abstract: The High Performance Light Water Reactor (HPLWR) is a thermal spectrum nuclear reactor cooled and moderated with light water operated at supercritical pressure. It is an innovative reactor concept, which requires developing and applying advanced analysis tools as described in the paper. The relevant water density reduction associated with the heat-up, together with the multi-pass core design, results in a pronounced coupling between neutronic and thermal-hydraulic analyses, which takes into account the strong natural influence of the in-core distribution of power generation and water properties. The neutron flux gradients within the multi-pass core, together with the pronounced dependence of water properties on the temperature, require to consider a fine spatial resolution in which the individual fuel pins are resolved to provide precise evaluation of the clad temperature, currently considered as one of the crucial design criteria. These goals have been achieved considering an advanced analysis method based on the usage of existing codes which have been coupled with developed interfaces. Initially neutronic and thermal-hydraulic full core calculations have been iterated until a consistent solution is found to determine the steady state full power condition of the HPLWR core. Results of few group neutronic analyses might be less reliable in case of HPLWR 3-pass core than for conventional LWRs because of considerable changes of the neutron spectrum within the core, hence 40 groups transport theory has been preferred to the

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-03-15

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

  3. Development of a kinetics analysis code for fuel solution combined with thermal-hydraulics analysis code PHOENICS and analysis of natural-cooling characteristic test of TRACY. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Watanabe, Shouichi; Yamane, Yuichi; Miyoshi, Yoshinori [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    Since exact information is not always acquired in the criticality accident of fuel solution, parametric survey calculations are required for grasping behaviors of the thermal-hydraulics. On the other hand, the practical methods of the calculation with can reduce the computation time with allowable accuracy will be also required, since the conventional method takes a long calculation time. In order to fulfill the requirement, a two-dimensional (R-Z) nuclear-kinetics analysis code considering thermal-hydraulic based on the multi-region kinetic equations with one-group neutron energy was created by incorporating with the thermal-hydraulics analysis code PHOENICS for all-purpose use the computation time of the code was shortened by separating time mesh intervals of the nuclear- and heat-calculations from that of the hydraulics calculation, and by regulating automatically the time mesh intervals in proportion to power change rate. A series of analysis were performed for the natural-cooling characteristic test using TRACY in which the power changed slowly for 5 hours after the transient power resulting from the reactivity insertion of a 0.5 dollar. It was found that the code system was able to calculate within the limit of practical time, and acquired the prospect of reproducing the experimental values considerably for the power and temperature change. (author)

  4. 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

  5. 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.

  6. 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

  7. 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

  8. 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)

  9. Network performance analysis

    CERN Document Server

    Bonald, Thomas

    2013-01-01

    The book presents some key mathematical tools for the performance analysis of communication networks and computer systems.Communication networks and computer systems have become extremely complex. The statistical resource sharing induced by the random behavior of users and the underlying protocols and algorithms may affect Quality of Service.This book introduces the main results of queuing theory that are useful for analyzing the performance of these systems. These mathematical tools are key to the development of robust dimensioning rules and engineering methods. A number of examples i

  10. Analysis of Thermal-Hydraulic Behavior of CMT in the SMART-ITL Facility

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, Byong Guk; Bae, Hwang; Ryu, Sung-Uk; Ryu, Hyobong; Byun, Sun-Joon; Yi, Sung-Jae; Park, Hyun-Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    SMART, an integral small modular reactor, received a standard design approval in 2012 and now extends its safety features through replacing active safety injection pumps by passive safety injection systems: core makeup tanks (CMT) and safety injection tanks (SIT). SMART-ITL has been built in a full height scale and 1/49 area and power scale. One train of CMT and SIT has been installed and their thermal-hydraulic behaviors have been identified through a series of tests. In this paper, initial condensation characteristics as well as force balance around the CMT will be discussed for a representative test. PSIS are added into SMART for better treatment of accidents with prolonged station blackout. In the SMART-ITL, the CMT and SIT are installed to evaluate their performance and a series of tests have been conducted. In this paper, the thermal-hydraulic behavior of CMT is addressed based on the experimental data, especially focusing on the issues of fierce condensation after opening of the isolation valve and driving force balance around the CMT.

  11. Analysis of Thermal-Hydraulic Behavior of CMT in the SMART-ITL Facility

    International Nuclear Information System (INIS)

    Jeon, Byong Guk; Bae, Hwang; Ryu, Sung-Uk; Ryu, Hyobong; Byun, Sun-Joon; Yi, Sung-Jae; Park, Hyun-Sik

    2015-01-01

    SMART, an integral small modular reactor, received a standard design approval in 2012 and now extends its safety features through replacing active safety injection pumps by passive safety injection systems: core makeup tanks (CMT) and safety injection tanks (SIT). SMART-ITL has been built in a full height scale and 1/49 area and power scale. One train of CMT and SIT has been installed and their thermal-hydraulic behaviors have been identified through a series of tests. In this paper, initial condensation characteristics as well as force balance around the CMT will be discussed for a representative test. PSIS are added into SMART for better treatment of accidents with prolonged station blackout. In the SMART-ITL, the CMT and SIT are installed to evaluate their performance and a series of tests have been conducted. In this paper, the thermal-hydraulic behavior of CMT is addressed based on the experimental data, especially focusing on the issues of fierce condensation after opening of the isolation valve and driving force balance around the CMT

  12. High performance hydraulic design techniques of mixed-flow pump impeller and diffuser

    International Nuclear Information System (INIS)

    Kim, Sung; Lee, Kyoung Yong; Kim, Joon Hyung; Kim, Jin Hyuk; Jung, Uk Hee; Choi, Young Seok

    2015-01-01

    In this paper, we describe a numerical study about the performance improvement of a mixed-flow pump by optimizing the design of the impeller and diffuser using a commercial computational fluid dynamics (CFD) code and design-of-experiments (DOE). The design variables of impeller and diffuser in the vane plane development were defined with a fixed meridional plane. The design variables were defined by the vane plane development, which indicates the blade-angle distributions and length of the impeller and diffuser. The vane plane development was controlled using the blade-angle in a fixed meridional plane. The blade shape of the impeller and diffuser were designed using a traditional method in which the inlet and exit angles are connected smoothly. First, the impeller optimum design was performed with impeller design variables. The diffuser optimum design was performed with diffuser design variables while the optimally designed impeller shape was fixed. The importance of the impeller and diffuser design variables was analyzed using 2 k factorial designs, and the design optimization of the impeller and diffuser design variables was determined using the response surface method (RSM). The objective functions were defined as the total head (Ht) and the total efficiency (ηt) at the design flow rate. The optimally designed model was verified using numerical analysis, and the numerical analysis results for both the optimum model and the reference model were compared to determine the reasons for the improved pump performance. A pump performance test was carried out for the optimum model, and its reliability was proved by a comparative analysis of the results of the numerical analysis and an experiment using the optimum model.

  13. 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

  14. 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.

  15. 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

  16. 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.)

  17. 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.

  18. 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.

  19. 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)

  20. 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".

  1. Thermal-hydraulic analysis under partial loss of flow accident hypothesis of a plate-type fuel surrounded by two water channels using RELAP5 code

    Directory of Open Access Journals (Sweden)

    Itamar Iliuk

    2016-01-01

    Full Text Available Thermal-hydraulic analysis of plate-type fuel has great importance to the establishment of safety criteria, also to the licensing of the future nuclear reactor with the objective of propelling the Brazilian nuclear submarine. In this work, an analysis of a single plate-type fuel surrounding by two water channels was performed using the RELAP5 thermal-hydraulic code. To realize the simulations, a plate-type fuel with the meat of uranium dioxide sandwiched between two Zircaloy-4 plates was proposed. A partial loss of flow accident was simulated to show the behavior of the model under this type of accident. The results show that the critical heat flux was detected in the central region along the axial direction of the plate when the right water channel was blocked.

  2. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-03-15

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

  4. Evaluation of the hydraulic and biological performance of the portable floating fish collector at Cougar Reservoir and Dam, Oregon, 2014

    Science.gov (United States)

    Beeman, John W.; Evans, Scott D.; Haner, Philip V.; Hansel, Hal C.; Hansen, Amy C.; Hansen, Gabriel S.; Hatton, Tyson W.; Sprando, Jamie M.; Smith, Collin D.; Adams, Noah S.

    2016-01-12

    The biological and hydraulic performance of a new portable floating fish collector (PFFC) located in a cul-de-sac within the forebay of Cougar Dam, Oregon, was evaluated during 2014. The purpose of the PFFC was to explore surface collection as a means to capture juvenile salmonids at one or more sites using a small, cost-effective, pilot-scale device. The PFFC used internal pumps to draw attraction flow over an inclined plane about 3 meters (m) deep, through a flume at a design velocity of as much as 6 feet per second (ft/s), and to empty a small amount of water and any entrained fish into a collection box. Performance of the PFFC was evaluated at 64 cubic feet per second (ft3/s) (Low) and 109 ft3/s (High) inflow rates alternated using a randomized-block schedule from May 27 to December 16, 2014. The evaluation of the biological performance was based on trap catch; behaviors, locations, and collection of juvenile Chinook salmon (Oncorhynchus tshawytscha) tagged with acoustic transmitters plus passive integrated transponder (PIT) tags; collection of juvenile Chinook salmon implanted with only PIT tags; and untagged fish monitored near and within the PFFC using acoustic cameras. The evaluation of hydraulic performance was based on measurements of water velocity and direction of flow in the PFFC.

  5. Controls of Hydraulic Wind Turbine

    Directory of Open Access Journals (Sweden)

    Zhang Yin

    2016-01-01

    Full Text Available In this paper a hydraulic wind turbine generator system was proposed based on analysis the current wind turbines technologies. The construction and principles were introduced. The mathematical model was verified using MATLAB and AMsim. A displacement closed loop of swash plate of motor and a speed closed loop of generator were setup, a PID control is introduced to maintain a constant speed and fixed frequency at wind turbine generator. Simulation and experiment demonstrated that the system can connect grid to generate electric and enhance reliability. The control system demonstrates a high performance speed regulation and effectiveness. The results are great significant to design a new type hydraulic wind turbine system.

  6. Use of Hydraulic Model for Water Loss Reduction

    OpenAIRE

    Mindaugas Rimeika; Anželika Jurkienė

    2016-01-01

    Hydraulic modeling is the modern way to apply world water engineering experience in every day practice. Hydraulic model is an effective tool in order to perform analysis of water supply system, optimization of its operation, assessment of system efficiency potential, evaluation of water network development, fire flow capabilities, energy saving opportunities and water loss reduction and ect. Hydraulic model shall include all possible engineering elements and devices allocated in a real water ...

  7. 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

  8. 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.

  9. 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

  10. 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.

  11. 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.

  12. Assessment of the impact of neutronic/thermal-hydraulic coupling on the design and performance of nuclear reactors for space propulsion

    International Nuclear Information System (INIS)

    Aithal, S.M.; Aldemir, T.; Vafai, K.

    1994-01-01

    A series of studies has been performed to investigate the potential impact of the coupling between neutronics and thermal hydraulics on the design and performance assessment of solid core reactors for nuclear thermal space propulsion, using the particle bed reactor (PBR) concept as an example system. For a given temperature distribution in the reactor, the k eff and steady-state core power distribution are obtained from three-dimensional, continuous energy Monte Carlo simulations using the MCNP code. For a given core power distribution, determination of the temperature distribution in the core and hydrogen-filled annulus between the reflector and pressure vessel is based on a nonthermal equilibrium analysis. The results show that a realistic estimation of fuel, core size, and control requirements for PBRs using hydrogenous moderators, as well as optimization of the overall engine design, may require coupled neutronic/thermal-hydraulic studies. However, it may be possible to estimate the thermal safety margins and propellant exit temperatures based on power distributions obtained from neutronic calculations at room temperature. The results also show that, while variation of the hydrogen flow rate in the annulus has been proposed as a partial control mechanism for PBRs, such control mechanism may not be feasible for PBRs with high moderator-to-fuel ratios and hence soft core neutron spectra

  13. Data management implementation plan for the Bear Creek Valley treatability study phase 2 hydraulic performance testing, Oak Ridge Y-12 Plant, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    1997-12-01

    The overall objective of the Bear Creek Valley treatability study is to provide site-specific data defining potential treatment technologies applicable to contaminated groundwater and surface water. The ultimate goal of this effort is to install a treatment system that will remove uranium, technetium, nitrate, and several metals from groundwater before it reaches Bear Creek. This project, the Bear Creek Valley treatability study Phase 2 hydraulic performance testing, directly supports the Bear Creek Valley Feasibility Study. Specific project objectives include (1) installing monitoring and extraction wells, (2) installing a groundwater extraction trench, (3) performing pumping tests of the extraction wells and trench, (4) determining hydraulic gradients, and (5) collecting water quality parameters. The primary purpose of environmental data management is to provide a system for generating and maintaining technically defensible data. To meet current regulatory requirements for the Environmental Restoration Program, complete documentation of the information flow must be established. To do so, each step in the data management process (collection, management, storage, and analysis) must be adequately planned and documented. This document will serve to identify data management procedures, expected data types and flow, and roles and responsibilities for all data management activities associated with this project

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-04-01

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

  15. 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.

  16. 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.

  17. 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.)

  18. 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

  19. 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

  20. Thermal-Hydraulic Performance of Cross-Shaped Spiral Fuel in High-Power-Density BWRs

    International Nuclear Information System (INIS)

    Conboy, Thomas; Hejzlar, Pavel

    2006-01-01

    Power up-rating of existing nuclear reactors promises to be an area of great study for years to come. One of the major approaches to efficiently increasing power density is by way of advanced fuel design, and cross-shaped spiral-fuel has shown such potential in previous studies. Our work aims to model the thermal-hydraulic consequences of filling a BWR core with these spiral-shaped pins. The helically-wound pins have a cross-section resembling a 4-petaled flower. They fill an assembly in a tight bundle, their dimensions chosen carefully such that the petals of neighboring pins contact each other at their outer-most extent in a self-supporting lattice, absent of grid spacers. Potential advantages of this design raise much optimism from a thermal-hydraulic perspective. These spiral rods possess about 40% larger surface area than traditional rods, resulting in increased cooling and a proportional reduction in average surface heat flux. The thin petal-like extensions help by lowering thermal resistance between the hot central region of the pin and the bulk coolant flow, decreasing the maximum fuel temperature by 200 deg. C according to Finite Element (COSMOS) models. However, COSMOS models also predict a potential problem area at the 'elbow' region of two adjoining petals, where heat flux peaking is twice that along the extensions. Preliminary VIPRE models, which account only for the surface area increase, predict a 22% increase in critical power. It is also anticipated that the spiral twist would provide the flowing coolant with an additional radial velocity component, and likely promote turbulence and mixing within an assembly. These factors are expected to provide further margin for increased power density, and are currently being incorporated into the VIPRE model. The reduction in pressure drop inherent in any core without grid-spacers is also expected to be significant in aiding core stability, though this has not yet been quantified. Spiral-fuel seems to be a

  1. Influence of armour porosity on the hydraulic stability of cube armour layers

    OpenAIRE

    Medina Folgado, Josep Ramón; Molines Llodra, Jorge; GÓMEZ MARTÍN, MARÍA ESTHER

    2014-01-01

    Armour placement and packing density directly affect construction costs and hydraulic performance of mound breakwaters. In this paper, the literature concerning the influence of armour porosity on the hydraulic stability of single- and double-layer armours is discussed. Qualitative and quantitative estimations for the influence of armour porosity and packing density on the hydraulic stability are given for the most common concrete armour units. The analysis focuses on specific 2D hydraulic st...

  2. Digital switched hydraulics

    Science.gov (United States)

    Pan, Min; Plummer, Andrew

    2018-06-01

    This paper reviews recent developments in digital switched hydraulics particularly the switched inertance hydraulic systems (SIHSs). The performance of SIHSs is presented in brief with a discussion of several possible configurations and control strategies. The soft switching technology and high-speed switching valve design techniques are discussed. Challenges and recommendations are given based on the current research achievements.

  3. Derivation of site-specific relationships between hydraulic parameters and p-wave velocities based on hydraulic and seismic tomography

    Energy Technology Data Exchange (ETDEWEB)

    Brauchler, R.; Doetsch, J.; Dietrich, P.; Sauter, M.

    2012-01-10

    In this study, hydraulic and seismic tomographic measurements were used to derive a site-specific relationship between the geophysical parameter p-wave velocity and the hydraulic parameters, diffusivity and specific storage. Our field study includes diffusivity tomograms derived from hydraulic travel time tomography, specific storage tomograms, derived from hydraulic attenuation tomography, and p-wave velocity tomograms, derived from seismic tomography. The tomographic inversion was performed in all three cases with the SIRT (Simultaneous Iterative Reconstruction Technique) algorithm, using a ray tracing technique with curved trajectories. The experimental set-up was designed such that the p-wave velocity tomogram overlaps the hydraulic tomograms by half. The experiments were performed at a wellcharacterized sand and gravel aquifer, located in the Leine River valley near Göttingen, Germany. Access to the shallow subsurface was provided by direct-push technology. The high spatial resolution of hydraulic and seismic tomography was exploited to derive representative site-specific relationships between the hydraulic and geophysical parameters, based on the area where geophysical and hydraulic tests were performed. The transformation of the p-wave velocities into hydraulic properties was undertaken using a k-means cluster analysis. Results demonstrate that the combination of hydraulic and geophysical tomographic data is a promising approach to improve hydrogeophysical site characterization.

  4. Improving the Hydraulic Performance of Stormwater Infiltration Systems in Clay Tills

    DEFF Research Database (Denmark)

    Bockhorn, Britta

    investigations on two typical Danish clay till sites, and one modeling study with the integrated surface water and groundwater model HydroGeoSphere. The saturated hydraulic conductivity (Ksat) is the most critical soil physical parameter when it comes to sizing stormwater infiltration systems. In the first study......, different field methods for Ksat estimation, the double ring infiltrometer, the Guelph permeameter and falling head infiltration tests in a small excavation, were compared and evaluated for their capability to return realistic Ksat values in tills. The double ring infiltrometer and the Guelph permeameter...... represent suitable methods for sizing stormwater infiltration systems if measurements are combined with geological knowledge from maps of near-surface deposits and borehole descriptions. If space allows, the more invasive infiltration tests in a small excavation are recommended, because measurements...

  5. Thermal-hydraulic performance of the finned surface of a compact heat exchanger

    International Nuclear Information System (INIS)

    Errasti Cabrera, Michel

    2015-01-01

    In this work the thermal-hydraulic behavior of the finned surface of a compact heat exchanger is obtained in tube-fin configuration corrugated (wavy). Through numerical simulation are determined average values ​​of intensification of heat transfer and pressure loss in the inter-channel finned. The objective is to characterize the surface to use as a reference, to make comparisons with other heat exchange surfaces enhanced using traditional techniques combined with more current, such as vortex generators. The study is conducted in laminar flow, with Reynolds numbers below 1000. In the working model compact exchanger tubes and corrugated fins (wavy) heat is described, and the results of the coefficient of overall heat transfer and the pressure drop are explained from the local characteristics of the velocity field and temperature inside the heat exchanger. (Full text)

  6. Performance analysis in saber.

    Science.gov (United States)

    Aquili, Andrea; Tancredi, Virginia; Triossi, Tamara; De Sanctis, Desiree; Padua, Elvira; DʼArcangelo, Giovanna; Melchiorri, Giovanni

    2013-03-01

    Fencing is a sport practiced by both men and women, which uses 3 weapons: foil, épée, and saber. In general, there are few scientific studies available in international literature; they are limited to the performance analysis of fencing bouts, yet there is nothing about saber. There are 2 kinds of competitions in the World Cup for both men and women: the "FIE GP" and "A." The aim of this study was to carry out a saber performance analysis to gain useful indicators for the definition of a performance model. In addition, it is expected to verify if it could be influenced by the type of competition and if there are differences between men and women. Sixty bouts: 33 FIE GP and 27 "A" competitions (35 men's and 25 women's saber bouts) were analyzed. The results indicated that most actions are offensive (55% for men and 49% for women); the central area of the piste is mostly used (72% for men and 67% for women); the effective fighting time is 13.6% for men and 17.1% for women, and the ratio between the action and break times is 1:6.5 for men and 1:5.1 for women. A lunge is carried out every 23.9 seconds by men and every 20 seconds by women, and a direction change is carried out every 65.3 seconds by men and every 59.7 seconds by women. The data confirm the differences between the saber and the other 2 weapons. There is no significant difference between the data of the 2 different kinds of competitions.

  7. Advances of study on thermal-hydraulic performance in tight-lattice rod bundles for reduced-moderation water reactors

    International Nuclear Information System (INIS)

    Akira Ohnuki; Kazuyuki Takase; Masatoshi Kureta; Hiroyuki Yoshida; Hidesada Tamai; Wei Liu; Toru Nakatsuka; Hajime Akimoto

    2005-01-01

    R and D project to investigate thermal-hydraulic performance in tight-lattice rod bundles for Reduced-Moderation Water Reactor (RMWR) is started at Japan Atomic Energy Research Institute in collaboration with power company, reactor vendors, universities since 2002. The RMWR can attain the favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured LWR technologies. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio by reducing the moderation of neutron. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. The confirmation of thermal-hydraulic feasibility is one of the most important R and D items for the RMWR because of the tight-lattice configuration. In this paper, we will show the R and D plan and describe some advances on experimental and analytical studies. The experimental study is performed mainly using large-scale (37-rod bundle) test facility and the analytical one aims to develop a predictable technology for geometry effects such as gap between rods, grid spacer configuration etc. using advanced 3-D two-phase flow simulation methods. Steady-state and transient critical power experiments are conducted with the test facility (Gap width between rods: 1.0 mm) and the experimental data reveal the feasibility of RMWR. (authors)

  8. Plant performance on Mediterranean green roofs: interaction of species-specific hydraulic strategies and substrate water relations.

    Science.gov (United States)

    Raimondo, Fabio; Trifilò, Patrizia; Lo Gullo, Maria A; Andri, Sergio; Savi, Tadeja; Nardini, Andrea

    2015-01-20

    Recent studies have highlighted the ecological, economic and social benefits assured by green roof technology to urban areas. However, green roofs are very hostile environments for plant growth because of shallow substrate depths, high temperatures and irradiance and wind exposure. This study provides experimental evidence for the importance of accurate selection of plant species and substrates for implementing green roofs in hot and arid regions, like the Mediterranean area. Experiments were performed on two shrub species (Arbutus unedo L. and Salvia officinalis L.) grown in green roof experimental modules with two substrates slightly differing in their water retention properties, as derived from moisture release curves. Physiological measurements were performed on both well-watered and drought-stressed plants. Gas exchange, leaf and xylem water potential and also plant hydraulic conductance were measured at different time intervals following the last irrigation. The substrate type significantly affected water status. Arbutus unedo and S. officinalis showed different hydraulic responses to drought stress, with the former species being substantially isohydric and the latter one anisohydric. Both A. unedo and S. officinalis were found to be suitable species for green roofs in the Mediterranean area. However, our data suggest that appropriate choice of substrate is key to the success of green roof installations in arid environments, especially if anisohydric species are employed. Published by Oxford University Press on behalf of the Annals of Botany Company.

  9. 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

    Laboratory is being used to simulate and visualize the effects of the injection. The simulation model uses a discrete fracture network generated for RRG-9 using acoustic borehole imaging and analysis of microseismic activity. By adjusting the permeability of the fractures, a pressure history match for the first part of the stimulation program was obtained. The results of this model indicate that hydraulic fracturing is the dominant mechanism for permeability improvement for this part of the stimulation program.

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

    Science.gov (United States)

    Leonard, Daniel J.

    can occur more abruptly in the model than the prototype, due to lack of Froude similitude between the two. When severe cavitation occurs, clear differences are observed in vapor content between the scales. A stage-by-stage performance decomposition is conducted to analyze the losses within individual components of each scale of the machine. As cavitation becomes more severe, the losses in the draft tube account for an increasing amount of the total losses in the machine. More losses occur in the model draft tube as cavitation formation in the prototype draft tube is prevented by the larger hydrostatic pressure gradient across the machine. Additionally, unsteady Detached Eddy Simulations of the fully-coupled cavitating hydroturbine are performed for both scales. Both mesh and temporal convergence studies are provided. The temporal and spectral content of fluctuations in torque and pressure are monitored and compared between single-phase, cavitating, model, and prototype cases. A shallow draft tube induced runner imbalance results in an asymmetric vapor distribution about the runner, leading to more extensive growth and collapse of vapor on any individual blade as it undergoes a revolution. Unique frequency components manifest and persist through the entire machine only when cavitation is present in the hub vortex. Large maximum pressure spikes, which result from vapor collapse, are observed on the blade surfaces in the multiphase simulations, and these may be a potential source of cavitation damage and erosion. Multiphase CFD is shown to be an accurate and effective technique for simulating and analyzing cavitating flow in Francis-class hydraulic turbines. It is recommended that it be used as an industrial tool to supplement model cavitation experiments for all types of hydraulic turbines. Moreover, multiphase CFD can be equally effective as a research tool, to investigate mechanisms of cavitating hydraulic turbines that are not understood, and to uncover unique new

  12. Development of RETRAN-03/MOV code for thermal-hydraulic analysis of nuclear reactor under moving conditions

    International Nuclear Information System (INIS)

    Kim, Hak Jae; Park, Goon Cherl

    1996-01-01

    Nuclear ship reactors have several; features different from land-based PWR's. Especially, effects of ship motions on reactor thermal-hydraulics and good load following capability for abrupt load changes are essential characteristics of nuclear ship reactors. This study modified the RETRAN-03 to analyze the thermal-hydraulic transients under three-dimensional ship motions, named RETRAN-03/MOV in order to apply to future marine reactors. First Japanese nuclear ship MUTSU reactor have been analyzed under various ship motions to verify this code. Calculations have been performed under rolling,heaving and stationary inclination conditions during normal operation. Also, the natural circulation has been analyzed, which can provide the decay heat removed to ensure the passive safety of marine reactors. As results, typical thermal-hydraulic characteristics of marine reactors such as flow rate oscillations and S/G water level oscillations have been successfully simulated at various conditions. 7 refs., 11 figs. (author)

  13. Effect of wet-dry cycles on polymer treated bentonite in seawater : swelling ability, hydraulic conductivity and crack analysis

    NARCIS (Netherlands)

    De Camillis, Michela; Di Emidio, Gemmina; Bezuijen, Adam; Verastegui Flores, Daniel; Van Stappen, Jeroen; Cnudde, Veerle

    2017-01-01

    Waste disposal facilities are often isolated by clay liners in order to prevent pollutant migration into the ground. Bentonite is used as barrier material thanks to the low conductivity to water. However, the hydraulic performance may be impaired by contact with aggressive liquids due to cation

  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. Thermal-hydraulic Analysis of High-temperature Cover Gas Region in STELLA-2

    Energy Technology Data Exchange (ETDEWEB)

    Jo, Youngchul; Son, Seok-Kwon; Yoon, Jung; Eoh, Jaehyuk; Jeong, Ji-Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The first phase of the program was focused on the key sodium component tests, and the second one has been concentrated on the sodium thermal-hydraulic integral effect test (STELLA-2). Based on its platform, simulation of the PGSFR transient will be made to evaluate plant dynamic behaviors as well as to demonstrate decay heat removal performance. Therefore, most design features of PGSFR have been modeled in STELLA-2 as closely as possible. The similarities of temperature and pressure between the model (STELLA-2) and the prototype (PGSFR) have been well preserved to reflect thermal-hydraulic behavior with natural convection as well as heat transfer between structure and sodium coolant inside the model reactor vessel (RV). For this reason, structural integrity of the entire test section should be confirmed as in the prototype. In particular, since the model reactor head in STELLA-2 supports key components and internal structures, its structural integrity exposed to high-temperature cover gas region should be confirmed. In order to reduce thermal radiation heat transfer from the hot sodium pool during normal operation, a dedicated insulation layer has been installed at the downward surface of the model reactor head to prevent direct heat flux from the sodium free surface at 545 .deg. C. Three-dimensional conjugate heat transfer analyses for the full-shape geometry of the upper part of the model reactor vessel in STELLA-2 have been carried out. Based on the results, steady-state temperature distributions in the cover gas region and the model reactor head itself have been obtained and the design requirement in temperature of the model reactor head has been newly proposed to be 350 .deg. C. For any elevated temperature conditions in STELLA-2, it was confirmed that the model reactor head generally satisfied the requirement. The CFD database constructed from this study will be used to optimize geometric parameters such as thicknesses and/or types of the insulator.

  16. 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

  17. 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

  18. 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.

  19. 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)

  20. Uncertainty in the determination of soil hydraulic parameters and its influence on the performance of two hydrological models of different complexity

    Directory of Open Access Journals (Sweden)

    G. Baroni

    2010-02-01

    Full Text Available Data of soil hydraulic properties forms often a limiting factor in unsaturated zone modelling, especially at the larger scales. Investigations for the hydraulic characterization of soils are time-consuming and costly, and the accuracy of the results obtained by the different methodologies is still debated. However, we may wonder how the uncertainty in soil hydraulic parameters relates to the uncertainty of the selected modelling approach. We performed an intensive monitoring study during the cropping season of a 10 ha maize field in Northern Italy. The data were used to: i compare different methods for determining soil hydraulic parameters and ii evaluate the effect of the uncertainty in these parameters on different variables (i.e. evapotranspiration, average water content in the root zone, flux at the bottom boundary of the root zone simulated by two hydrological models of different complexity: SWAP, a widely used model of soil moisture dynamics in unsaturated soils based on Richards equation, and ALHyMUS, a conceptual model of the same dynamics based on a reservoir cascade scheme. We employed five direct and indirect methods to determine soil hydraulic parameters for each horizon of the experimental profile. Two methods were based on a parameter optimization of: a laboratory measured retention and hydraulic conductivity data and b field measured retention and hydraulic conductivity data. The remaining three methods were based on the application of widely used Pedo-Transfer Functions: c Rawls and Brakensiek, d HYPRES, and e ROSETTA. Simulations were performed using meteorological, irrigation and crop data measured at the experimental site during the period June – October 2006. Results showed a wide range of soil hydraulic parameter values generated with the different methods, especially for the saturated hydraulic conductivity Ksat and the shape parameter α of the van Genuchten curve. This is reflected in a variability of

  1. Thermal-hydraulic unreliability of passive systems

    International Nuclear Information System (INIS)

    Tzanos, C.P.; Saltos, N.T.

    1995-01-01

    Advanced light water reactor designs like AP600 and the simplified boiling water reactor (SBWR) use passive safety systems for accident prevention and mitigation. Because these systems rely on natural forces for their operation, their unavailability due to hardware failures and human error is significantly smaller than that of active systems. However, the coolant flows predicted to be delivered by these systems can be subject to significant uncertainties, which in turn can lead to a significant uncertainty in the predicted thermal-hydraulic performance of the plant under accident conditions. Because of these uncertainties, there is a probability that an accident sequence for which a best estimate thermal-hydraulic analysis predicts no core damage (success sequence) may actually lead to core damage. For brevity, this probability will be called thermal-hydraulic unreliability. The assessment of this unreliability for all the success sequences requires very expensive computations. Moreover, the computational cost increases drastically as the required thermal-hydraulic reliability increases. The required computational effort can be greatly reduced if a bounding approach can be used that either eliminates the need to compute thermal-hydraulic unreliabilities, or it leads to the analysis of a few bounding sequences for which the required thermal-hydraulic reliability is relatively small. The objective of this paper is to present such an approach and determine the order of magnitude of the thermal-hydraulic unreliabilities that may have to be computed

  2. ANN Model for Predicting the Impact of Submerged Aquatic Weeds Existence on the Hydraulic Performance of Branched Open Channel System Accompanied by Water Structures

    International Nuclear Information System (INIS)

    Abdeen, Mostafa A. M.; Abdin, Alla E.

    2007-01-01

    The existence of hydraulic structures in a branched open channel system urges the need for considering the gradually varied flow criterion in evaluating the different hydraulic characteristics in this type of open channel system. Computations of hydraulic characteristics such as flow rates and water surface profiles in branched open channel system with hydraulic structures require tremendous numerical effort especially when the flow cannot be assumed uniform. In addition, the existence of submerged aquatic weeds in this branched open channel system adds to the complexity of the evaluation of the different hydraulic characteristics for this system. However, this existence of aquatic weeds can not be neglected since it is very common in Egyptian open channel systems. Artificial Neural Network (ANN) has been widely utilized in the past decade in civil engineering applications for the simulation and prediction of the different physical phenomena and has proven its capabilities in the different fields. The present study aims towards introducing the use of ANN technique to model and predict the impact of submerged aquatic weeds existence on the hydraulic performance of branched open channel system. Specifically the current paper investigates a branched open channel system that consists of main channel supplies water to two branch channels that are infested by submerged aquatic weeds and have water structures such as clear over fall weirs and sluice gates. The results of this study showed that ANN technique was capable, with small computational effort and high accuracy, of predicting the impact of different infestation percentage for submerged aquatic weeds on the hydraulic performance of branched open channel system with two different hydraulic structures

  3. 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

  4. Development of hydraulic analysis code for optimizing thermo-chemical is process reactors

    International Nuclear Information System (INIS)

    Terada, Atsuhiko; Hino, Ryutaro; Hirayama, Toshio; Nakajima, Norihiro; Sugiyama, Hitoshi

    2007-01-01

    The Japan Atomic Energy Agency has been conducting study on thermochemical IS process for water splitting hydrogen production. Based on the test results and know-how obtained through the bench-scale test, a pilot test plant, which has a hydrogen production performance of 30 Nm 3 /h, is being designed conceptually as the next step of the IS process development. In design of the IS pilot plant, it is important to make chemical reactors compact with high performance from the viewpoint of plant cost reduction. A new hydraulic analytical code has been developed for optimizing mixing performance of multi-phase flow involving chemical reactions especially in the Bunsen reactor. Complex flow pattern with gas-liquid chemical interaction involving flow instability will be characterized in the Bunsen reactor. Preliminary analytical results obtained with above mentioned code, especially flow patterns induced by swirling flow agreed well with that measured by water experiments, which showed vortex breakdown pattern in a simplified Bunsen reactor. (author)

  5. Coupled fuel performance and thermal-hydraulics simulation with BISON and RELAP-7 at accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Martineau, R.C., E-mail: Richard.Martineau@inl.gov [Idaho National Laboratory, Idaho Falls, ID (United States)

    2015-07-01

    'Full text:' RELAP-7 is expected to be the next in the RELAP nuclear reactor safety/systems analysis application series developed at the Idaho National Laboratory (INL). The development of RELAP-7 began in 2011 to support the Risk Informed Safety Margins Characterization (RISMC) Pathway of Department of Energy's (DOE) Light Water Reactor Sustainability (LWRS) Program. The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in software design, numerical methods, and physical models in order to provide capabilities needed for the RISMC methodology and to support modern nuclear power safety analysis. RELAP-7 is built using the INL's modern scientific software development framework, MOOSE (Multi-physics Object Oriented Simulation Environment). MOOSE provides improved implicit numerical schemes, including higher-order integration in both space and time, and yielding converged second-order accuracy for RELAP-7. The code structure is based on multiple physical component models such as pipes, junctions, pumps, etc. This component-based software architecture allows RELAP-7 to quickly adopt different physical models for different applications. One of the main advantages of building RELAP-7 on the MOOSE framework is that tight coupling with other MOOSE-based applications solving physics not present in RELAP-7 requires little to no additional lines of code. For example, the RELAP-7 core channel component is based upon a one-dimensional flow channel and a three-zone two-dimensional heat structure designed to represent fuel, gap, and cladding conjugate heat transfer with the coolant. However, the RELAP-7 application does not carry the fuels performance physics to analyze irradiated fuel, especially for accident scenarios. Here, we demonstrate the tightly coupled capability of the BISON nuclear fuels performance application with RELAP-7 for the station black out (SBO) accident scenario (Fukushima type event) and

  6. 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.

  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. 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.

  11. Design of a Novel Electro-hydraulic Drive Downhole Tractor

    Science.gov (United States)

    Fang, Delei; Shang, Jianzhong; Yang, Junhong; Wang, Zhuo; Wu, Wei

    2018-02-01

    In order to improve the traction ability and the work efficiency of downhole tractor in oil field, a novel electro-hydraulic drive downhole tractor was designed. The tractor’s supporting mechanism and moving mechanism were analyzed based on the tractor mechanical structure. Through the introduction of hydraulic system, the hydraulic drive mechanism and the implementation process were researched. Based on software, analysis of tractor hydraulic drive characteristic and movement performance were simulated, which provide theoretical basis for the development of tractor prototype.

  12. 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

  13. 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)

  14. 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

  15. 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)

  16. Preliminary analysis on incore performance of nuclear fuel: pt. 4

    International Nuclear Information System (INIS)

    Noh, S.K.; Chang, M.H.; Lee, C.C.; Chung, Y.H.; Kuk, K.Y.; Park, C.Y.; Lee, S.K.

    1981-01-01

    An analysis has been performed for thermal hydraulic design parameters of Wolsung-1 reactor core in steady state with the help of a computer code COBRA-IV-I. The design parameters are coolant enthalpy, flow velocity, coolant quality, pressure and fuel temperature distribution. The maximum power channel has been taken into account in this work. The results appear to be reasonably agreeable with data from PSR'S, with the maximum difference between this work and PSR'S being 4.3%

  17. 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.

  18. 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

  19. 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)

  20. 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)

  1. 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

  2. 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 ...

  3. Thermal hydraulic model validation for HOR mixed core fuel management

    International Nuclear Information System (INIS)

    Gibcus, H.P.M.; Vries, J.W. de; Leege, P.F.A. de

    1997-01-01

    A thermal-hydraulic core management model has been developed for the Hoger Onderwijsreactor (HOR), a 2 MW pool-type university research reactor. The model was adopted for safety analysis purposes in the framework of HEU/LEU core conversion studies. It is applied in the thermal-hydraulic computer code SHORT (Steady-state HOR Thermal-hydraulics) which is presently in use in designing core configurations and for in-core fuel management. An elaborate measurement program was performed for establishing the core hydraulic characteristics for a variety of conditions. The hydraulic data were obtained with a dummy fuel element with special equipment allowing a.o. direct measurement of the true core flow rate. Using these data the thermal-hydraulic model was validated experimentally. The model, experimental tests, and model validation are discussed. (author)

  4. Monitoring performance for hydraulic fracturing using synthetic microseismic catalogue at the Wysin site (Poland)

    Science.gov (United States)

    Ángel López Comino, José; Cesca, Simone; Kriegerowski, Marius; Heimann, Sebastian; Dahm, Torsten; Mirek, Janusz; Lasocky, Stanislaw

    2017-04-01

    Previous analysis to assess the monitoring performance of a dedicated seismic network are always useful to determine its capability of detecting, locating and characterizing target seismicity. This work focuses on a hydrofracking experiment in Poland, which is monitored in the framework of the SHEER (SHale gas Exploration and Exploitation induced Risks) EU project. The seismic installation is located near Wysin (Poland), in the central-western part of the Peribaltic synclise at Pomerania. The network setup includes a distributed network of six broadband stations, three shallow borehole stations and three small-scale arrays. We assess the monitoring performance prior operations, using synthetic seismograms. Realistic full waveform are generated and combined with real noise before fracking operations, to produce either event based or continuous synthetic waveforms. Background seismicity is modelled by double couple (DC) focal mechanisms. Non-DC sources resemble induced tensile fractures opening in the direction of the minimal compressive stress and closing in the same direction after the injection. Microseismic sources are combined with a realistic crustal model, distribution of hypocenters, magnitudes and source durations. The network detection performance is then assessed in terms of Magnitude of Completeness (Mc) through two different techniques: i) using an amplitude threshold approach, taking into account a station dependent noise level and different values of signal-to-noise ratio (SNR) and ii) through the application of an automatic detection algorithm to the continuous synthetic dataset. In the first case, we compare the maximal amplitude of noise free synthetic waveforms with the different noise levels. Imposing the simultaneous detection at e.g. 4 stations for a robust detection, the Mc is assessed and can be adjusted by empirical relationships for different SNR values. We find that different source mechanisms have different detection threshold. The

  5. 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.

  6. 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.

  7. Neutronics and Thermal Hydraulics Analysis of a Conceptual Ultra-High Temperature MHD Cermet Fuel Core for Nuclear Electric Propulsion

    Directory of Open Access Journals (Sweden)

    Jian Song

    2018-04-01

    Full Text Available Nuclear electric propulsion (NEP offers unique advantages for the interplanetary exploration. The extremely high conversion efficiency of magnetohydrodynamics (MHD conversion nuclear reactor makes it a highly potential space power source in the future, especially for NEP systems. Research on ultra-high temperature reactor suitable for MHD power conversion is performed in this paper. Cermet is chosen as the reactor fuel after a detailed comparison with the (U,ZrC graphite-based fuel and mixed carbide fuel. A reactor design is carried out as well as the analysis of the reactor physics and thermal-hydraulics. The specific design involves fuel element, reactor core, and radiation shield. Two coolant channel configurations of fuel elements are considered and both of them can meet the demands. The 91 channel configuration is chosen due to its greater heat transfer performance. Besides, preliminary calculation of nuclear criticality safety during launch crash accident is also presented. The calculation results show that the current design can meet the safety requirements well.

  8. Thermal hydraulic-Mechanic Integrated Simulation for Advanced Cladding Thermal Shock Fracture Analysis during Reflood Phase in LBLOCA

    Energy Technology Data Exchange (ETDEWEB)

    Son, Seong Min; Lee, You Ho; Cho, Jae Wan; Lee, Jeong Ik [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    This study suggested thermal hydraulic-mechanical integrated stress based methodology for analyzing the behavior of ATF type claddings by SiC-Duplex cladding LBLOCA simulation. Also, this paper showed that this methodology could predict real experimental result well. That concept for enhanced safety of LWR called Advanced Accident-Tolerance Fuel Cladding (ATF cladding, ATF) is researched actively. However, current nuclear fuel cladding design criteria for zircaloy cannot be apply to ATF directly because those criteria are mainly based on limiting their oxidation. So, the new methodology for ATF design criteria is necessary. In this study, stress based analysis methodology for ATF cladding design criteria is suggested. By simulating LBLOCA scenario of SiC cladding which is the one of the most promising candidate of ATF. Also we'll confirm our result briefly through comparing some facts from other experiments. This result is validating now. Some of results show good performance with 1-D failure analysis code for SiC fuel cladding that already developed and validated by Lee et al,. It will present in meeting. Furthermore, this simulation presented the possibility of understanding the behavior of cladding deeper. If designer can predict the dangerous region and the time precisely, it may be helpful for designing nuclear fuel cladding geometry and set safety criteria.

  9. The examination of the effects of the flow baffle on the thermal hydraulic performance of IHX shell side

    International Nuclear Information System (INIS)

    Kim, W. K.; Sim, Y. S.; Kim, S. O.; Baek, B. J.

    2002-01-01

    The effects of the flow baffle on the thermal hydraulic performance of IHX shell side has been examined using ASTEEPL, 2DHX code for the variation of baffle distance, baffle cut and baffle flow hole design data. When KALIMER design data were selected as a base for the study, a shell baffle structures does not influence on a total heat transfer rate and it is better to reduce baffle distance rather than baffle flow hole size from the view of pressure loss in the IHX. Radial thermal imbalance between tubes was reduced to 6 .deg. C when the number of baffle is beyond 13. And thermal imbalance decreased also when flow hole size decreases. If flow hole clogs perfectly, thermal imbalance can be reduced up to 2 .deg. C

  10. 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)

  11. 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 ...

  12. Meta-analysis Reveals that Hydraulic Traits Explain Cross-Species Patterns of Drought-Induced Tree Mortality across the Globe

    Science.gov (United States)

    Anderegg, W.

    2016-12-01

    Drought-induced tree mortality has been observed globally and is expected to increase under climate change scenarios, with large potential consequences for the terrestrial carbon sink. Predicting mortality across species is crucial for assessing the effects of climate extremes on forest community biodiversity, composition, and carbon sequestration. However, the physiological traits associated with elevated risk of mortality in diverse ecosystems remain unknown, though these could greatly improve understanding and prediction of tree mortality in forests. We performed a meta-analysis on species' mortality rates across 475 species from 33 studies around the globe to assess which traits determine a species' mortality risk. We found that species-specific mortality anomalies from community mortality rate in a given drought were associated with plant hydraulic traits. Across all species, mortality was best predicted by a low hydraulic safety margin - the difference between typical minimum xylem water potential and that causing xylem dysfunction - and xylem vulnerability to embolism. Angiosperms and gymnosperms experienced roughly equal mortality risk. Our results provide broad support that hydraulic traits capture key mechanisms determining tree death and highlight that physiological traits can improve vegetation models' prediction of tree mortality during climate extremes. We conclude with thoughts about a revised framework for future tree mortality research.

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

    International Nuclear Information System (INIS)

    Ichikawa, Ryoko; Masuhara, Yasuhiro; Kasahara, Fumio

    2012-01-01

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

  14. 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)

  15. Study on development of virtual reactor core laboratory (1). Development of prototype coupled neutronic, thermal-hydraulic and structural analysis system

    International Nuclear Information System (INIS)

    Uto, Nariaki; Sugaya, Toshio; Tsukimori, Kazuyuki; Negishi, Hitoshi; Enuma, Yasuhiro; Sakai, Takaaki

    1999-09-01

    A study on development of virtual reactor core laboratory, which is to conduct numerical experiments representative of complicated physical phenomena in practical reactor core systems on a computational environment, has progressed at Japan Nuclear Cycle Development Institute (JNC). The study aims at systematic evaluation of these phenomena into which nuclear reactions, thermal-hydraulic characteristics, structural responses and fuel behaviors combine, and effective utilization of the obtained comprehension for core design. This report presents a production of a prototype computational system which is required to construct the virtual reactor core laboratory. This system is to evaluate reactor core performance under the coupled neutronic, thermal-hydraulic and structural phenomena, and is composed of two analysis tools connected by a newly developed interface program; 1) an existing space-dependent coupled neutronic and thermal-hydraulic analysis system arranged at JNC and 2) a core deformation analysis code. It acts on a cluster of several DEC/Alpha workstations. A specific library called MPI1 (Message Passing Interface 1) is incorporated as a tool for communicating among the analysis modules consisting of the system. A series of calculations for simulating a sequence of Unprotected Loss Of Heat Sink (ULOHS) coupled with rapid drop of some neutron absorber devices in a prototype fast reactor is tried to investigate how the system works. The obtained results show the core deformation behavior followed by the reactivity change that can be properly evaluated. The results of this report show that the system is expected to be useful for analyzing sensitivity of reactor core performance with respect to uncertainties of various design parameters and establishing a concept of passive safety reactor system, taking into account space distortion of neutron flux distribution during abnormal events as well as reactivity feedback from core deformation. (author)

  16. Analysis of soil hydraulic and thermal properties for land surface modeling over the Tibetan Plateau

    Science.gov (United States)

    Zhao, Hong; Zeng, Yijian; Lv, Shaoning; Su, Zhongbo

    2018-06-01

    Soil information (e.g., soil texture and porosity) from existing soil datasets over the Tibetan Plateau (TP) is claimed to be inadequate and even inaccurate for determining soil hydraulic properties (SHP) and soil thermal properties (STP), hampering the understanding of the land surface process over TP. As the soil varies across three dominant climate zones (i.e., arid, semi-arid and subhumid) over the TP, the associated SHP and STP are expected to vary correspondingly. To obtain an explicit insight into the soil hydrothermal properties over the TP, in situ and laboratory measurements of over 30 soil property profiles were obtained across the climate zones. Results show that porosity and SHP and STP differ across the climate zones and strongly depend on soil texture. In particular, it is proposed that gravel impact on porosity and SHP and STP are both considered in the arid zone and in deep layers of the semi-arid zone. Parameterization schemes for porosity, SHP and STP are investigated and compared with measurements taken. To determine the SHP, including soil water retention curves (SWRCs) and hydraulic conductivities, the pedotransfer functions (PTFs) developed by Cosby et al. (1984) (for the Clapp-Hornberger model) and the continuous PTFs given by Wösten et al. (1999) (for the Van Genuchten-Mualem model) are recommended. The STP parameterization scheme proposed by Farouki (1981) based on the model of De Vries (1963) performed better across the TP than other schemes. Using the parameterization schemes mentioned above, the uncertainties of five existing regional and global soil datasets and their derived SHP and STP over the TP are quantified through comparison with in situ and laboratory measurements. The measured soil physical properties dataset is available at https://data.4tu.nl/repository/uuid:c712717c-6ac0-47ff-9d58-97f88082ddc0" target="_blank">https://data.4tu.nl/repository/uuid:c712717c-6ac0-47ff-9d58-97f88082ddc0.

  17. Hydraulic pressure pulses with elastic and plastic structural flexibility: test and analysis (LWBR Development Program)

    International Nuclear Information System (INIS)

    Schwirian, R.E.

    1978-03-01

    Pressure pulse tests were conducted with a flexible test section in a test vessel filled with room temperature water. The pressure pulses were generated with a drop hammer and piston pulse generator and were of a sufficient magnitude to cause plastic deformation of the test section. Because of the strong pressure relief effect of the deforming test section, pressure pulse magnitudes were below 265 psig in magnitude and had durations of 50 to 55 msecs. Calculations performed with the FLASH-35 bi-linear hysteresis model of structural deformation show good agreement with experiment. In particular, FLASH 35 adequately predicts the decrease in peak pressure and the increase in pulse duration due to elastic and plastic deformation of the test section. Predictions of flexible member motion are good, but are less satisfactory than the pressure pulse results due to uncertainties in the values of yield point and beyond yield stiffness used to model the various flexible members. Coupled with this is a strong sensitivity of the FLASH 35 predictions to the values of yield point and beyond yield stiffness chosen for the various flexible members. The test data versus calculation comparisons presented here provide preliminary qualification for FLASH 35 calculations of transient hydraulic pressures and pressure differentials in the presence of flexible structural members which deform both elastically and plastically

  18. Parametric thermo-hydraulic analysis of the TF system of JT-60SA during fast discharge

    International Nuclear Information System (INIS)

    Polli, Gian Mario; Lacroix, Benoit; Zani, Louis; Besi Vetrella, Ugo; Cucchiaro, Antonio

    2013-01-01

    Highlights: • We modeled the central clock-wise pancake of JT-60SA TF magnet at the EOB. • We simulated a quench followed by a fast discharge. • We evaluated the temperature and pressure rises in the nominal configuration. • We evaluated the effect of several parameter changes on the thermal-hydraulic response of the system. -- Abstract: The evolution of the conductor temperature and of the helium pressure of the central pancake of the TF superconducting magnet of the JT-60SA tokamak in a quench scenario are here discussed. The quench is triggered by a heat disturbance applied at the end of burning and followed by a fast safety discharge. A parametric study aimed at assessing the robustness of the calculation is also addressed with special regard to the voltage threshold, used to define the occurrence of the quench, and to the time delay, that cover all the possible delays in the fast discharge after quench detection. Finally, due to sensitivity analyses the influences of different parameters were assessed: the material properties of the strands (RRR, copper fraction), the magnitude and the spatial length of the triggering disturbance and the magnetic field distribution. The numerical evaluations were performed in the framework of the Broader Approach Agreement in collaboration with CEA, ENEA and the JT-60SA European Home Team using the 1D code Gandalf [1

  19. Transient thermal hydraulic modeling and analysis of ITER divertor plate system

    International Nuclear Information System (INIS)

    El-Morshedy, Salah El-Din; Hassanein, Ahmed

    2009-01-01

    A mathematical model has been developed/updated to simulate the steady state and transient thermal-hydraulics of the International Thermonuclear Experimental Reactor (ITER) divertor module. The model predicts the thermal response of the armour coating, divertor plate structural materials and coolant channels. The selected heat transfer correlations cover all operating conditions of ITER under both normal and off-normal situations. The model also accounts for the melting, vaporization, and solidification of the armour material. The developed model is to provide a quick benchmark of the HEIGHTS multidimensional comprehensive simulation package. The present model divides the coolant channels into a specified axial regions and the divertor plate into a specified radial zones, then a two-dimensional heat conduction calculation is created to predict the temperature distribution for both steady and transient states. The model is benchmarked against experimental data performed at Sandia National Laboratory for both bare and swirl tape coolant channel mockups. The results show very good agreements with the data for steady and transient states. The model is then used to predict the thermal behavior of the ITER plasma facing and structural materials due to plasma instability event where 60 MJ/m 2 plasma energy is deposited over 500 ms. The results for ITER divertor response is analyzed and compared with HEIGHTS results.

  20. Transient thermal hydraulic modeling and analysis of ITER divertor plate system

    Energy Technology Data Exchange (ETDEWEB)

    El-Morshedy, Salah El-Din [Argonne National Laboratory, Argonne, IL (United States); Atomic Energy Authority, Cairo (Egypt)], E-mail: selmorshedy@etrr2-aea.org.eg; Hassanein, Ahmed [Purdue University, West Lafayette, IN (United States)], E-mail: hassanein@purdue.edu

    2009-12-15

    A mathematical model has been developed/updated to simulate the steady state and transient thermal-hydraulics of the International Thermonuclear Experimental Reactor (ITER) divertor module. The model predicts the thermal response of the armour coating, divertor plate structural materials and coolant channels. The selected heat transfer correlations cover all operating conditions of ITER under both normal and off-normal situations. The model also accounts for the melting, vaporization, and solidification of the armour material. The developed model is to provide a quick benchmark of the HEIGHTS multidimensional comprehensive simulation package. The present model divides the coolant channels into a specified axial regions and the divertor plate into a specified radial zones, then a two-dimensional heat conduction calculation is created to predict the temperature distribution for both steady and transient states. The model is benchmarked against experimental data performed at Sandia National Laboratory for both bare and swirl tape coolant channel mockups. The results show very good agreements with the data for steady and transient states. The model is then used to predict the thermal behavior of the ITER plasma facing and structural materials due to plasma instability event where 60 MJ/m{sup 2} plasma energy is deposited over 500 ms. The results for ITER divertor response is analyzed and compared with HEIGHTS results.

  1. Thermal hydraulic and power cycle analysis of liquid lithium blanket designs

    International Nuclear Information System (INIS)

    Misra, B.; Stevens, H.C.; Maroni, V.A.

    1977-01-01

    Thermal hydraulic and power cycle analyses were performed for the first-wall and blanket systems of tokamak-type fusion reactors under a typical set of design and operating conditions. The analytical results for lithium-cooled blanket cells show that with stainless steel as construction material and with no divertor present, the maximum allowable neutron wall loading is approximately 2 MW/m 2 and is limited by thermal stress criteria. With vanadium alloy as construction material and no divertor present, the maximum allowable neutron wall loading is approximately 8 MW/m 2 and is limited by an interplay of constraints imposed on the maximum allowable structural temperature and the minimum allowable coolant inlet temperature. With a divertor these wall loadings can be increased by from 40 to 90 percent. The cost of the vanadium system is found to be competitive with the stainless steel system because of the higher allowable structural temperatures and concomitant higher thermal efficiencies afforded by the vanadium alloys

  2. 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

  3. 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...

  4. 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)

  5. 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)

  6. 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

  7. 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

  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. Preliminary Thermo-hydraulic Core Design Analysis of Korea Advanced Nuclear Thermal Engine Rocket for Space Application

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Seung Hyun; Lee, Jeong Ik; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-05-15

    {sub th} power and electricity generation with 100 kW{sub th} idle power. Consequently, KANUTER has the characteristics of a compact and lightweight system, excellent propellant efficiency, bimodal capability, and mission versatility as indicated in the reference design parameters. This thermo-hydraulic design analysis was carried out to estimate the optimum FWT of the unique SLHC fuel design in the core and thereby the maximum rocket performance. The FWT affects the mechanical strength of the SLHC fuel assembly as well as the thermo-hydraulic capability mainly depending on the heat transfer area of fuel. The thicker fuel wafer is mechanically strong with low pressure drop, while the thinner fuel wafer is thermally robust with less mechanical strength and higher shear stress in the core.

  11. 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

  12. 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

  13. 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 the