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...

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

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...

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

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

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)

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)

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)

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.

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)

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...

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.

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)

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...

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.

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.

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

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

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.

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.

Development of a 1D thermal-hydraulic analysis code for once-through steam generator in SMRs using straight tubes

Energy Technology Data Exchange (ETDEWEB)

Park, Youngjae; Kim, Iljin; Kim, Hyungdae [Kyung Hee University, Yongin (Korea, Republic of)

2015-10-15

Diverse integral/small-modular reactors (SMRs) have been developed. Once-through steam generator (OTSG) which generates superheated steam without steam separator and dryer was used in the SMRs to reduce volume of steam generator. It would be possible to design a new steam generator with best estimate thermal-hydraulic codes such as RELAP and MARS. However, it is not convenience to use the general purpose thermal-hydraulic analysis code to design a specific component of nuclear power plants. A widely used simulation tool for thermal-hydraulic analysis of drum-type steam generators is ATHOS, which allows 3D analysis. On the other hand, a simple 1D thermal-hydraulic analysis code might be accurate enough for the conceptual design of OTSG. In this study, thermal-hydraulic analysis code for conceptual design of OTSG was developed using 1D homogeneous equilibrium model (HEM). A benchmark calculation was also conducted to verify and validate the prediction accuracy of the developed code by comparing with the analysis results with MARS. Finally, conceptual design of OTSG was conducted by the developed code. A simple 1D thermal-hydraulic analysis code was developed for the purpose of conceptual design OTSG for SMRs. A set of benchmark calculations was conducted to verify and validate the analysis accuracy of the developed code by comparing results obtained with a best-estimated thermal-hydraulic analysis code, MARS. Finally, analysis of two different OTSG design concepts with superheating and recirculation was demonstrated using the developed code.

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.

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

Comparative study of methods to estimate hydraulic parameters in the hydraulically undisturbed Opalinus Clay (Switzerland)

Energy Technology Data Exchange (ETDEWEB)

Yu, C.; Matray, J.-M. [Institut de Radioprotection et de Sûreté Nucléaire, Fontenay-aux-Roses, (France); Yu, C.; Gonçalvès, J. [Aix Marseille Université UMR 6635 CEREGE Technopôle Environnement Arbois-Méditerranée Aix-en-Provence, Cedex 4 (France); and others

2017-04-15

The deep borehole (DB) experiment gave the opportunity to acquire hydraulic parameters in a hydraulically undisturbed zone of the Opalinus Clay at the Mont Terri rock laboratory (Switzerland). Three methods were used to estimate hydraulic conductivity and specific storage values of the Opalinus Clay formation and its bounding formations through the 248 m deep borehole BDB-1: application of a Poiseuille-type law involving petrophysical measurements, spectral analysis of pressure time series and in situ hydraulic tests. The hydraulic conductivity range in the Opalinus Clay given by the first method is 2 × 10{sup -14}-6 × 10{sup -13} m s{sup -1} for a cementation factor ranging between 2 and 3. These results show low vertical variability whereas in situ hydraulic tests suggest higher values up to 7 × 10{sup -12} m s{sup -1}. Core analysis provides economical estimates of the homogeneous matrix hydraulic properties but do not account for heterogeneities at larger scale such as potential tectonic conductive features. Specific storage values obtained by spectral analysis are consistent and in the order of 10{sup -6} m{sup -1}, while formulations using phase shift and gain between pore pressure signals were found to be inappropriate to evaluate hydraulic conductivity in the Opalinus Clay. The values obtained are globally in good agreement with the ones obtained previously at the rock laboratory. (authors)

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)

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.

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

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

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.

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.

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.)

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

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.

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

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

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

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)

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.

Hydraulic manipulator research at ORNL

International Nuclear Information System (INIS)

Kress, R.L.; Jansen, J.F.; Love, L.J.

1997-01-01

Recently, task requirements have dictated that manipulator payload capacity increase to accommodate greater payloads, greater manipulator length, and larger environmental interaction forces. General tasks such as waste storage tank cleanup and facility dismantlement and decommissioning require manipulator life capacities in the range of hundreds of pounds rather than tens of pounds. To meet the increased payload capacities demanded by present-day tasks, manipulator designers have turned once again to hydraulics as a means of actuation. In order to successfully design, build, and deploy a new hydraulic manipulator (or subsystem), sophisticated modeling, analysis, and control experiments are usually needed. Oak Ridge National Laboratory (ORNL) has a history of projects that incorporate hydraulics technology, including mobile robots, teleoperated manipulators, and full-scale construction equipment. In addition, to support the development and deployment of new hydraulic manipulators, ORNL has outfitted a significant experimental laboratory and has developed the software capability for research into hydraulic manipulators, hydraulic actuators, hydraulic systems, modeling of hydraulic systems, and hydraulic controls. The purpose of this article is to describe the past hydraulic manipulator developments and current hydraulic manipulator research capabilities at ORNL. Included are example experimental results from ORNL's flexible/prismatic test stand

Hydraulic manipulator research at ORNL

Energy Technology Data Exchange (ETDEWEB)

Kress, R.L.; Jansen, J.F. [Oak Ridge National Lab., TN (United States); Love, L.J. [Oak Ridge Inst. for Science and Education, TN (United States)

1997-03-01

Recently, task requirements have dictated that manipulator payload capacity increase to accommodate greater payloads, greater manipulator length, and larger environmental interaction forces. General tasks such as waste storage tank cleanup and facility dismantlement and decommissioning require manipulator life capacities in the range of hundreds of pounds rather than tens of pounds. To meet the increased payload capacities demanded by present-day tasks, manipulator designers have turned once again to hydraulics as a means of actuation. In order to successfully design, build, and deploy a new hydraulic manipulator (or subsystem), sophisticated modeling, analysis, and control experiments are usually needed. Oak Ridge National Laboratory (ORNL) has a history of projects that incorporate hydraulics technology, including mobile robots, teleoperated manipulators, and full-scale construction equipment. In addition, to support the development and deployment of new hydraulic manipulators, ORNL has outfitted a significant experimental laboratory and has developed the software capability for research into hydraulic manipulators, hydraulic actuators, hydraulic systems, modeling of hydraulic systems, and hydraulic controls. The purpose of this article is to describe the past hydraulic manipulator developments and current hydraulic manipulator research capabilities at ORNL. Included are example experimental results from ORNL`s flexible/prismatic test stand.

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)

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...

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)

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.

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 ...

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

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)

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.

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

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

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)

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.

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)

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.

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

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

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)

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)

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.

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.

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

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.

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

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

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)

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.

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

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

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

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)

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)

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.

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

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.

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.

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

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.

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.

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.

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

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.

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

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

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)

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)

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

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

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)

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

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)

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

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.

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 ...

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

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)

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

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.

Design of An Energy Efficient Hydraulic Regenerative circuit

Science.gov (United States)

Ramesh, S.; Ashok, S. Denis; Nagaraj, Shanmukha; Adithyakumar, C. R.; Reddy, M. Lohith Kumar; Naulakha, Niranjan Kumar

2018-02-01

Increasing cost and power demand, leads to evaluation of new method to increase through productivity and help to solve the power demands. Many researchers have break through to increase the efficiency of a hydraulic power pack, one of the promising methods is the concept of regenerative. The objective of this research work is to increase the efficiency of a hydraulic circuit by introducing a concept of regenerative circuit. A Regenerative circuit is a system that is used to speed up the extension stroke of the double acting single rod hydraulic cylinder. The output is connected to the input in the directional control value. By this concept, increase in velocity of the piston and decrease the cycle time. For the research, a basic hydraulic circuit and a regenerative circuit are designated and compared both with their results. The analysis was based on their time taken for extension and retraction of the piston. From the detailed analysis of both the hydraulic circuits, it is found that the efficiency by introducing hydraulic regenerative circuit increased by is 5.3%. The obtained results conclude that, implementing hydraulic regenerative circuit in a hydraulic power pack decreases power consumption, reduces cycle time and increases productivity in a longer run.

Development of a multi-dimensional realistic thermal-hydraulic system analysis code, MARS 1.3 and its verification

Energy Technology Data Exchange (ETDEWEB)

Lee, Won Jae; Chung, Bub Dong; Jeong, Jae Jun; Ha, Kwi Seok [Korea Atomic Energy Research Institute, Taejon (Korea)

1998-06-01

A multi-dimensional realistic thermal-hydraulic system analysis code, MARS version 1.3 has been developed. Main purpose of MARS 1.3 development is to have the realistic analysis capability of transient two-phase thermal-hydraulics of Pressurized Water Reactors (PWRs) especially during Large Break Loss of Coolant Accidents (LBLOCAs) where the multi-dimensional phenomena domain the transients. MARS code is a unified version of USNRC developed COBRA-TF, domain the transients. MARS code is a unified version of USNRC developed COBRA-TF, three-dimensional (3D) reactor vessel analysis code, and RELAP5/MOD3.2.1.2, one-dimensional (1D) reactor system analysis code., Developmental requirements for MARS are chosen not only to best utilize the existing capability of the codes but also to have the enhanced capability in code maintenance, user accessibility, user friendliness, code portability, code readability, and code flexibility. For the maintenance of existing codes capability and the enhancement of code maintenance capability, user accessibility and user friendliness, MARS has been unified to be a single code consisting of 1D module (RELAP5) and 3D module (COBRA-TF). This is realized by implicitly integrating the system pressure matrix equations of hydrodynamic models and solving them simultaneously, by modifying the 1D/3D calculation sequence operable under a single Central Processor Unit (CPU) and by unifying the input structure and the light water property routines of both modules. In addition, the code structure of 1D module is completely restructured using the modular data structure of standard FORTRAN 90, which greatly improves the code maintenance capability, readability and portability. For the code flexibility, a dynamic memory management scheme is applied in both modules. MARS 1.3 now runs on PC/Windows and HP/UNIX platforms having a single CPU, and users have the options to select the 3D module to model the 3D thermal-hydraulics in the reactor vessel or other

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.

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.

An overview on rod-bundle thermal-hydraulic analyses

International Nuclear Information System (INIS)

Sha, W.T.

1980-01-01

Three methods used in rod-bundle thermal-hydraulic analysis are summarized. These methods are: (1) subchannel analysis, (2) porous medium formulation with volume porosity, surface permeability, distributed resistance and distributed heat source (sink) and, (3) bench-mark rod-bundle thermal-hydraulic analysis using a boundary-fitted coordinate system. Basic limitations and merits of each method are delineated. (orig.)

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)

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

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)

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.

Coupling analysis of deformation and thermal-hydraulics in a FBR fuel pin bundle using BAMBOO and ASFRE-IV Codes

International Nuclear Information System (INIS)

Ito, Masahiro; Imai, Yasutomo; Uwaba, Tomoyuki; Ohshima, Hiroyuki

2004-03-01

The bundle-duct interaction may occur in sodium cooled wire-wrapped FBR fuel subassemblies in high burn-up conditions. JNC has been developing a bundle deformation analysis code BAMBOO (Behavior Analysis code for Mechanical interaction of fuel Bundle under On-power Operation), a thermal hydraulics analysis code ASFRE-IV (Analysis of Sodium Flow in Reactor Elements - ver. IV) and their coupling method as a simulation system for the evaluation on the integrity of deformed FBR fuel pin bundles. In this study, the simulation system was applied to a coupling analysis of deformation and thermal-hydraulics in the fuel pin-bundle under a steady-state condition just after startup for the purpose of the verification of the simulation system. The iterative calculations of deformation and thermal-hydraulics employed in the coupling analysis provided numerically unstable solutions. From the result, it was found that improvement of the coupling algorithm of BAMBOO and ASFRE-IV is necessary to reduce numerical fluctuations and to obtain better convergence by introducing such computational technique as the optimized under-relaxation method. (author)

Specific storage and hydraulic conductivity tomography through the joint inversion of hydraulic heads and self-potential data

Science.gov (United States)

Ahmed, A. Soueid; Jardani, A.; Revil, A.; Dupont, J. P.

2016-03-01

Transient hydraulic tomography is used to image the heterogeneous hydraulic conductivity and specific storage fields of shallow aquifers using time series of hydraulic head data. Such ill-posed and non-unique inverse problem can be regularized using some spatial geostatistical characteristic of the two fields. In addition to hydraulic heads changes, the flow of water, during pumping tests, generates an electrical field of electrokinetic nature. These electrical field fluctuations can be passively recorded at the ground surface using a network of non-polarizing electrodes connected to a high impedance (> 10 MOhm) and sensitive (0.1 mV) voltmeter, a method known in geophysics as the self-potential method. We perform a joint inversion of the self-potential and hydraulic head data to image the hydraulic conductivity and specific storage fields. We work on a 3D synthetic confined aquifer and we use the adjoint state method to compute the sensitivities of the hydraulic parameters to the hydraulic head and self-potential data in both steady-state and transient conditions. The inverse problem is solved using the geostatistical quasi-linear algorithm framework of Kitanidis. When the number of piezometers is small, the record of the transient self-potential signals provides useful information to characterize the hydraulic conductivity and specific storage fields. These results show that the self-potential method reveals the heterogeneities of some areas of the aquifer, which could not been captured by the tomography based on the hydraulic heads alone. In our analysis, the improvement on the hydraulic conductivity and specific storage estimations were based on perfect knowledge of electrical resistivity field. This implies that electrical resistivity will need to be jointly inverted with the hydraulic parameters in future studies and the impact of its uncertainty assessed with respect to the final tomograms of the hydraulic parameters.

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.

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

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.

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.

Thermal, thermo-hydraulic and thermo-mechanic analysis for fuel elements of IEA-R1 reactor at 5MW

International Nuclear Information System (INIS)

Teixeira e Silva, A.; Silva Macedo, L.V. da

1989-01-01

In connection with the on going conversion of IEA-R1 Research Reactor, operated by IPEN-CNEN/SP, from the use of highly enriched uranium (HEU) fuel to the use of low enriched uranium (LEU) fuel, steady-state thermal and thermo-hydraulic analysis of both existing HEU and proposed LEU cores under 2 MW operating conditions have been carried out. Keeping in mind the possibility of power upgrading, steady-state thermal, thermo-hydraulic and thermomechanical analysis of proposed LEU core under 5 MW operating conditions have also been carried out. The thermal and thermo-hydraulic analysis at 2 MW show that the conversion of the existing HEU core to be proposed LEU core will not change the reactor safety margins. Although the upgrading of the reactor power to 5 MW will result in safety margins lower than in case of 2MW, these will be still sufficient for optimum operation and safe behaviour. The thermomechanical analysis at 5 MW show that the thermal stresses induced in the fuel element will satisfy the design limits for mechanical strenght and elastic stability. (author) [pt

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.

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.

Hydraulic characterization of hydrothermally altered Nopal tuff

Energy Technology Data Exchange (ETDEWEB)

Green, R.T.; Meyer-James, K.A. [Southwest Research Institute, San Antonio, TX (United States); Rice, G. [George Rice and Associates, San Antonio, TX (United States)

1995-07-01

Understanding the mechanics of variably saturated flow in fractured-porous media is of fundamental importance to evaluating the isolation performance of the proposed high-level radioactive waste repository for the Yucca Mountain site. Developing that understanding must be founded on the analysis and interpretation of laboratory and field data. This report presents an analysis of the unsaturated hydraulic properties of tuff cores from the Pena Blanca natural analog site in Mexico. The basic intent of the analysis was to examine possible trends and relationships between the hydraulic properties and the degree of hydrothermal alteration exhibited by the tuff samples. These data were used in flow simulations to evaluate the significance of a particular conceptual (composite) model and of distinct hydraulic properties on the rate and nature of water flow.

Hydraulic characterization of hydrothermally altered Nopal tuff

International Nuclear Information System (INIS)

Green, R.T.; Meyer-James, K.A.; Rice, G.

1995-07-01

Understanding the mechanics of variably saturated flow in fractured-porous media is of fundamental importance to evaluating the isolation performance of the proposed high-level radioactive waste repository for the Yucca Mountain site. Developing that understanding must be founded on the analysis and interpretation of laboratory and field data. This report presents an analysis of the unsaturated hydraulic properties of tuff cores from the Pena Blanca natural analog site in Mexico. The basic intent of the analysis was to examine possible trends and relationships between the hydraulic properties and the degree of hydrothermal alteration exhibited by the tuff samples. These data were used in flow simulations to evaluate the significance of a particular conceptual (composite) model and of distinct hydraulic properties on the rate and nature of water flow

Thermal-hydraulic analysis of the improved TOPAZ-II power system using a heat pipe radiator

Energy Technology Data Exchange (ETDEWEB)

Zhang, Wenwen; Zhang, Dalin, E-mail: dlzhang@mail.xjtu.edu.cn; Tian, Wenxi; Qiu, Suizheng; Su, G.H.

2016-10-15

Highlights: • The system thermal-hydraulic model of the improved space thermionic reactor is developed. • The temperature reactivity feedback effects of the moderator, UO2 fuel, electrodes and reflector are considered. • The alkali metal heat pipe radiator is modeled with the two dimensional heat pipe model. • The steady state and the start-up procedure of the system are analyzed. - Abstract: A system analysis code coupled with the heat pipe model is developed to analyze the thermal-hydraulic characteristics of the improved TOPAZ-II reactor power system with a heat pipe radiator. The core thermal-hydraulic model, neutron physics model, and the coolant loop component models (including pump, volume accumulator, pipes and plenums) are established. The designed heat pipe radiator, which replaces the original pumped loop radiator, is also modeled, including two-dimensional heat pipe analysis model, fin model and coolant transport duct model. The system analysis code and the heat pipe model is coupled in the transport duct model. Steady state condition and start-up procedure of the improved TOPAZ-II system are calculated. The results show that the designed radiator can satisfy the waste heat rejection requirement of the improved power system. Meanwhile, the code can be used to obtained the thermal characteristics of the system transients such as the start-up process.

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)

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

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

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

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.

Proceedings of the third nuclear thermal hydraulics meeting

International Nuclear Information System (INIS)

Anon.

1987-01-01

This book contains the proceedings of the Thermal Hydraulics Division of the American Nuclear Society. The papers presented include: Simulator qualification using engineering codes and Development of thermal hydraulic analysis capabilities for Oyster Creek

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 ...

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)

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)

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)

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

Hydraulic System Design of Hydraulic Actuators for Large Butterfly Valves

Directory of Open Access Journals (Sweden)

Ye HUANG

2014-09-01

Full Text Available Hydraulic control systems of butterfly valves are presently valve-controlled and pump-controlled. Valve-controlled hydraulic systems have serious power loss and generate much heat during throttling. Pump-controlled hydraulic systems have no overflow or throttling losses but are limited in the speed adjustment of the variable-displacement pump, generate much noise, pollute the environment, and have motor power that does not match load requirements, resulting in low efficiency under light loads and wearing of the variable-displacement pump. To overcome these shortcomings, this article designs a closed hydraulic control system in which an AC servo motor drives a quantitative pump that controls a spiral swinging hydraulic cylinder, and analyzes and calculates the structure and parameters of a spiral swinging hydraulic cylinder. The hydraulic system adjusts the servo motor’s speed according to the requirements of the control system, and the motor power matches the power provided to components, thus eliminating the throttling loss of hydraulic circuits. The system is compact, produces a large output force, provides stable transmission, has a quick response, and is suitable as a hydraulic control system of a large butterfly valve.

Slope instability caused by small variations in hydraulic conductivity

Science.gov (United States)

Reid, M.E.

1997-01-01

Variations in hydraulic conductivity can greatly modify hillslope ground-water flow fields, effective-stress fields, and slope stability. In materials with uniform texture, hydraulic conductivities can vary over one to two orders of magnitude, yet small variations can be difficult to determine. The destabilizing effects caused by small (one order of magnitude or less) hydraulic conductivity variations using ground-water flow modeling, finite-element deformation analysis, and limit-equilibrium analysis are examined here. Low hydraulic conductivity materials that impede downslope ground-water flow can create unstable areas with locally elevated pore-water pressures. The destabilizing effects of small hydraulic heterogeneities can be as great as those induced by typical variations in the frictional strength (approximately 4??-8??) of texturally similar materials. Common "worst-case" assumptions about ground-water flow, such as a completely saturated "hydrostatic" pore-pressure distribution, do not account for locally elevated pore-water pressures and may not provide a conservative slope stability analysis. In site characterization, special attention should be paid to any materials that might impede downslope ground-water flow and create unstable regions.

Reactor Thermal Hydraulic Numerical Calculation And Modeling

International Nuclear Information System (INIS)

Duong Ngoc Hai; Dang The Ba

2008-01-01

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

Applied hydraulic transients

CERN Document Server

Chaudhry, M Hanif

2014-01-01

This book covers hydraulic transients in a comprehensive and systematic manner from introduction to advanced level and presents various methods of analysis for computer solution. The field of application of the book is very broad and diverse and covers areas such as hydroelectric projects, pumped storage schemes, water-supply systems, cooling-water systems, oil pipelines and industrial piping systems. Strong emphasis is given to practical applications, including several case studies, problems of applied nature, and design criteria. This will help design engineers and introduce students to real-life projects. This book also: ·         Presents modern methods of analysis suitable for computer analysis, such as the method of characteristics, explicit and implicit finite-difference methods and matrix methods ·         Includes case studies of actual projects ·         Provides extensive and complete treatment of governed hydraulic turbines ·         Presents design charts, desi...

Hydraulic Arm Modeling via Matlab SimHydraulics

Czech Academy of Sciences Publication Activity Database

Věchet, Stanislav; Krejsa, Jiří

2009-01-01

Roč. 16, č. 4 (2009), s. 287-296 ISSN 1802-1484 Institutional research plan: CEZ:AV0Z20760514 Keywords : simulatin modeling * hydraulics * SimHydraulics Subject RIV: JD - Computer Applications, Robotics

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)

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.)

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.

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)

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.

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

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.

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.

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)

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)

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 ...

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

Hydraulic Actuators with Autonomous Hydraulic Supply for the Mainline Aircrafts

Directory of Open Access Journals (Sweden)

I. S. Shumilov

2014-01-01

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

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.

Inherent Limitations of Hydraulic Tomography

Science.gov (United States)

Bohling, Geoffrey C.; Butler, J.J.

2010-01-01

We offer a cautionary note in response to an increasing level of enthusiasm regarding high-resolution aquifer characterization with hydraulic tomography. We use synthetic examples based on two recent field experiments to demonstrate that a high degree of nonuniqueness remains in estimates of hydraulic parameter fields even when those estimates are based on simultaneous analysis of a number of carefully controlled hydraulic tests. We must, therefore, be careful not to oversell the technique to the community of practicing hydrogeologists, promising a degree of accuracy and resolution that, in many settings, will remain unattainable, regardless of the amount of effort invested in the field investigation. No practically feasible amount of hydraulic tomography data will ever remove the need to regularize or bias the inverse problem in some fashion in order to obtain a unique solution. Thus, along with improving the resolution of hydraulic tomography techniques, we must also strive to couple those techniques with procedures for experimental design and uncertainty assessment and with other more cost-effective field methods, such as geophysical surveying and, in unconsolidated formations, direct-push profiling, in order to develop methods for subsurface characterization with the resolution and accuracy needed for practical field applications. Copyright ?? 2010 The Author(s). Journal compilation ?? 2010 National Ground Water Association.

Groundwater potentiality mapping using geoelectrical-based aquifer hydraulic parameters: A GIS-based multi-criteria decision analysis modeling approach

Directory of Open Access Journals (Sweden)

Kehinde Anthony Mogaji Hwee San Lim

2017-01-01

Full Text Available This study conducted a robust analysis on acquired 2D resistivity imaging data and borehole pumping test records to optimize groundwater potentiality mapping in Perak province, Malaysia using derived aquifer hydraulic properties. The transverse resistance (TR parameter was determined from the interpreted 2D resistivity imaging data by applying the Dar-Zarrouk parameter equation. Linear regression and GIS techniques were used to regress the estimated values for TR parameters with the aquifer transmissivity values extracted from the geospatially produced BPT records-based aquifer transmissivity map to develop the aquifer transmissivity parameter predictive (ATPP model. The reliability evaluated ATPP model using the Theil inequality coefficient measurement approach was used to establish geoelectrical-based hydraulic parameters (GHP modeling equations for the modeling of transmissivity (Tr, hydraulic conductivity (K, storativity (St, and hydraulic diffusivity (D properties. The applied GHP modeling equation results to the delineated aquifer media was used to produce aquifer potential conditioning factor maps for Tr, K, St, and D. The maps were modeled to develop an aquifer potential mapping index (APMI model via applying the multi-criteria decision analysis-analytic hierarchy process principle. The area groundwater reservoir productivity potential model map produced based on the processed APMI model estimates in the GIS environment was found to be 71% accurate. This study establishes a good alternative approach to determine aquifer hydraulic parameters even in areas where pumping test information is unavailable using a cost effective geophysical data. The produced map can be explored for hydrological decision making.

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

Hydraulic turbines

International Nuclear Information System (INIS)

Meluk O, G.

1998-01-01

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

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.

Controlling a negative loaded hydraulic cylinder using pressure feedback

DEFF Research Database (Denmark)

Hansen, M.R.; Andersen, T.O.

2010-01-01

This paper is concerned with the inherent oscillatory nature of pressure compensated velocity control of a hydraulic cylinder subjected to a negative load and suspended by means of an over-center valve. Initially, a linearized stability analysis of such a hydraulic circuit is carried out clearly ...... in a nonlinear time domain simulation model validating the linear stability analysis....

Physico-empirical approach for mapping soil hydraulic behaviour

Directory of Open Access Journals (Sweden)

G. D'Urso

1997-01-01

Full Text Available Abstract: Pedo-transfer functions are largely used in soil hydraulic characterisation of large areas. The use of physico-empirical approaches for the derivation of soil hydraulic parameters from disturbed samples data can be greatly enhanced if a characterisation performed on undisturbed cores of the same type of soil is available. In this study, an experimental procedure for deriving maps of soil hydraulic behaviour is discussed with reference to its application in an irrigation district (30 km2 in southern Italy. The main steps of the proposed procedure are: i the precise identification of soil hydraulic functions from undisturbed sampling of main horizons in representative profiles for each soil map unit; ii the determination of pore-size distribution curves from larger disturbed sampling data sets within the same soil map unit. iii the calibration of physical-empirical methods for retrieving soil hydraulic parameters from particle-size data and undisturbed soil sample analysis; iv the definition of functional hydraulic properties from water balance output; and v the delimitation of soil hydraulic map units based on functional properties.

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

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.

Modeling and stability of electro-hydraulic servo of hydraulic excavator

Science.gov (United States)

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

2017-11-01

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

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

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)

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.

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...

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)

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

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.

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.

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.

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

Thermo-Hydraulic Modelling of Buffer and Backfill

International Nuclear Information System (INIS)

Pintado, X.; Rautioaho, E.

2013-09-01

The temporal evolution of saturation, liquid pressure and temperature in the components of the engineered barrier system was studied using numerical methods. A set of laboratory tests was conducted to calibrate the parameters employed in the models. The modelling consisted of thermal, hydraulic and thermo-hydraulic analysis in which the significant thermo-hydraulic processes, parameters and features were identified. CODE B RIGHT was used for the finite element modelling and supplementary calculations were conducted with analytical methods. The main objective in this report is to improve understanding of the thermo-hydraulic processes and material properties that affect buffer behaviour in the Olkiluoto repository and to determine the parametric requirements of models for the accurate prediction of this behaviour. The analyses consisted of evaluating the influence of initial canister temperature and gaps in the buffer, and the role played by fractures and the rock mass located between fractures in supplying water for buffer and backfill saturation. In the thermo-hydraulic analysis, the primary processes examined were the effects of buffer drying near the canister on temperature evolution and the manner in which heat flow affects the buffer saturation process. Uncertainties in parameters and variations in the boundary conditions, modelling geometry and thermo-hydraulic phenomena were assessed with a sensitivity analysis. The material parameters, constitutive models, and assumptions made were carefully selected for all the modelling cases. The reference parameters selected for the simulations were compared and evaluated against laboratory measurements. The modelling results highlight the importance of understanding groundwater flow through the rock mass and from fractures in the rock in order to achieve reliable predictions regarding buffer saturation, since saturation times could range from a few years to tens of thousands of years depending on the hydrogeological

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)

Hydraulic design of Three Gorges right bank powerhouse turbine for improvement of hydraulic stability

International Nuclear Information System (INIS)

Shi, Q

2010-01-01

This paper presents the hydraulic design of Three Gorges Right Bank Powerhouse turbine for improvement of hydraulic stability. The technical challenges faced in the hydraulic design of the turbine are given. The method of hydraulic design for improving the hydraulic stability and particularly for eliminating the upper part load pressure pulsations is clarified. The final hydraulic design results of Three Gorges Right Bank Powerhouse turbine based on modern hydraulic design techniques are presented.

Hydraulic design of Three Gorges right bank powerhouse turbine for improvement of hydraulic stability

Energy Technology Data Exchange (ETDEWEB)

Shi, Q, E-mail: qhshi@dfem.com.c [Dong Fang Electrical Machinery Co., Ltd., DEC 188, Huanghe West Road, Deyang, 618000 (China)

2010-08-15

This paper presents the hydraulic design of Three Gorges Right Bank Powerhouse turbine for improvement of hydraulic stability. The technical challenges faced in the hydraulic design of the turbine are given. The method of hydraulic design for improving the hydraulic stability and particularly for eliminating the upper part load pressure pulsations is clarified. The final hydraulic design results of Three Gorges Right Bank Powerhouse turbine based on modern hydraulic design techniques are presented.

A Study on thermal-hydraulic characteristics of the coolant materials for the transmutation reactor

Energy Technology Data Exchange (ETDEWEB)

Chung, Chang Hyun; You, Young Woo; Cho, Jae Seon; Kim, Ju Youl; Kim, Do Hyoung; Kim, Yoon Ik; Yang, Hui Chang [Seoul National University, Taejon (Korea)

1998-03-01

The objective of this study is to provide the direction of transmutation reactor design in terms of thermal hydraulics especially through the analysis of thermal hydraulic characteristics of various candidate materials for the transmutation reactor coolant. In this study, the characteristics of coolant materials used in current nuclear power plants and candidate materials for transmutation reactor are analyzed and compared. To evaluate the thermal hydraulic characteristics, the preliminary thermal-hydraulic calculation is performed for the candidate coolant materials of transmutation reactor. An analysis of thermal-hydraulic characteristics of transmutation reactor. An analysis of thermal-hydraulic characteristics of Sodium, Lead, Lead-Bismuth, and Lead-Lithium among the liquid metals considered as the coolant of transmutation reactor is performed by using computational fluid dynamics code FLUENT, and SIMPLER algorithm. (author). 50 refs., 40 figs., 30 tabs.

Cross-cutting european thermal-hydraulics research for innovative nuclear systems

International Nuclear Information System (INIS)

Roelofs, F.; Class, A.; Cheng, X.; Meloni, P.; Van Tichelen, K.; Boudier, P.; Prasser, M.

2010-01-01

Thermal-hydraulics is recognized as a key scientific subject in the development of different innovative nuclear reactor systems. From the thermal-hydraulic point of view, different innovative reactors are mainly characterized by their coolants (gas, water, liquid metals and molten salt). This results in different micro- and macroscopic behavior of flow and heat transfer and requires specific models and advanced analysis tools. However, many common thermal-hydraulic issues are identified among various innovative nuclear systems. In Europe, such cross-cutting thermal-hydraulic issues are the subject of the 7. framework programme THINS (Thermal-Hydraulics of Innovative Nuclear Systems) project which runs from 2010 until 2014. This paper will describe the activities in this project which address the main identified thermal hydraulics issues for innovative nuclear systems. (authors)

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

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.

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...

Advantages of Oscillatory Hydraulic Tomography

Science.gov (United States)

Kitanidis, P. K.; Bakhos, T.; Cardiff, M. A.; Barrash, W.

2012-12-01

Characterizing the subsurface is significant for most hydrogeologic studies, such as those involving site remediation and groundwater resource explo¬ration. A variety of hydraulic and geophysical methods have been developed to estimate hydraulic conductivity and specific storage. Hydraulic methods based on the analysis of conventional pumping tests allow the estimation of conductivity and storage without need for approximate petrophysical relations, which is an advantage over most geophysical methods that first estimate other properties and then infer values of hydraulic parameters. However, hydraulic methods have the disadvantage that the head-change signal decays with distance from the pumping well and thus becomes difficult to separate from noise except in close proximity to the source. Oscillatory hydraulic tomography (OHT) is an emerging technology to im¬age the subsurface. This method utilizes the idea of imposing sinusoidally varying pressure or discharge signals at several points, collecting head observations at several other points, and then processing these data in a tomographic fashion to estimate conductivity and storage coefficients. After an overview of the methodology, including a description of the most important potential advantages and challenges associated with this approach, two key promising features of the approach will be discussed. First, the signal at an observation point is orthogonal to and thus can be separated from nuisance inputs like head fluctuation from production wells, evapotranspiration, irrigation, and changes in the level of adjacent streams. Second, although the signal amplitude may be weak, one can extract the phase and amplitude of the os¬cillatory signal by collecting measurements over a longer time, thus compensating for the effect of large distance through longer sampling period.

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

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

Plant hydraulic diversity buffers forest ecosystem responses to drought

Science.gov (United States)

Anderegg, W.; Konings, A. G.; Trugman, A. T.; Pacala, S. W.; Yu, K.; Sulman, B. N.; Sperry, J.; Bowling, D. R.

2017-12-01

Drought impacts carbon, water, and energy cycles in forests and may pose a fundamental threat to forests in future climates. Plant hydraulic transport of water is central to tree drought responses, including curtailing of water loss and the risk of mortality during drought. The effect of biodiversity on ecosystem function has typically been examined in grasslands, yet the diversity of plant hydraulic strategies may influence forests' response to drought. In a combined analysis of eddy covariance measurements, remote-sensing data of plant water content variation, model simulations, and plant hydraulic trait data, we test the degree to which plant water stress schemes influence the carbon cycle and how hydraulic diversity within and across ecosystems affects large-scale drought responses. We find that current plant functional types are not well-suited to capture hydraulic variation and that higher hydraulic diversity buffers ecosystem variation during drought. Our results demonstrate that tree functional diversity, particularly hydraulic diversity, may be critical to simulate in plant functional types in current land surface model projections of future vegetation's response to climate extremes.

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)

BEPU-FSAR: establishing a background for extension of nuclear thermal hydraulic principles to non thermal-hydraulic codes

Energy Technology Data Exchange (ETDEWEB)

Menzel, Francine; Sabundjian, Gaianê, E-mail: franmenzel@gmail.com, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil); D’Auria, Francesco, E-mail: f.dauria@ing.unipi.it [University of Pisa, San Piero a Grado Nuclear Research Group (Italy)

2017-07-01

Nuclear thermal hydraulic and accident analysis are based in three pillar activities, which consists in: Scaling, Coupling and V and V. Each of them are established technology, with key documents to describe and widely used. The final goal of this work is to apply the BEPU methodology in all parts of FSAR where analytical techniques are needed (BEPU-FSAR) and for that the crucial step is the transfer of the BEPU concepts into the other areas. In this sense, the issue is how to adapt to other disciplines the pillar activities presented in the thermal hydraulic area. For that we need to identify which elements can be applied in the other areas, to show that the proposed methodology is feasible. This work aims to discuss the first steps towards a BEPU-FSAR methodology and to show that the Scaling, Coupling and V and V elements, currently done for thermal-hydraulic codes, can be also done for different codes, which are used to perform different analysis included on a FSAR of a generic plant. (author)

Thermally Actuated Hydraulic Pumps

Science.gov (United States)

Jones, Jack; Ross, Ronald; Chao, Yi

2008-01-01

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

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.

Development of thermal hydraulic evaluation code for CANDU reactors

International Nuclear Information System (INIS)

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

2004-02-01

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

Continuous hydrologic simulation and flood-frequency, hydraulic, and flood-hazard analysis of the Blackberry Creek watershed, Kane County, Illinois

Science.gov (United States)

Soong, David T.; Straub, Timothy D.; Murphy, Elizabeth A.

2006-01-01

Results of hydrologic model, flood-frequency, hydraulic model, and flood-hazard analysis of the Blackberry Creek watershed in Kane County, Illinois, indicate that the 100-year and 500-year flood plains range from approximately 25 acres in the tributary F watershed (a headwater subbasin at the northeastern corner of the watershed) to almost 1,800 acres in Blackberry Creek main stem. Based on 1996 land-cover data, most of the land in the 100-year and 500-year flood plains was cropland, forested and wooded land, and grassland. A relatively small percentage of urban land was in the flood plains. The Blackberry Creek watershed has undergone rapid urbanization in recent decades. The population and urbanized lands in the watershed are projected to double from the 1990 condition by 2020. Recently, flood-induced damage has occurred more frequently in urbanized areas of the watershed. There are concerns about the effect of urbanization on flood peaks and volumes, future flood-mitigation plans, and potential effects on the water quality and stream habitats. This report describes the procedures used in developing the hydrologic models, estimating the flood-peak discharge magnitudes and recurrence intervals for flood-hazard analysis, developing the hydraulic model, and the results of the analysis in graphical and tabular form. The hydrologic model, Hydrological Simulation Program-FORTRAN (HSPF), was used to perform the simulation of continuous water movements through various patterns of land uses in the watershed. Flood-frequency analysis was applied to an annual maximum series to determine flood quantiles in subbasins for flood-hazard analysis. The Hydrologic Engineering Center-River Analysis System (HEC-RAS) hydraulic model was used to determine the 100-year and 500-year flood elevations, and to determine the 100-year floodway. The hydraulic model was calibrated and verified using high water marks and observed inundation maps for the July 17-18, 1996, flood event. Digital

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

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)

Mine drivage in hydraulic mines

Energy Technology Data Exchange (ETDEWEB)

Ehkber, B Ya

1983-09-01

From 20 to 25% of labor cost in hydraulic coal mines falls on mine drivage. Range of mine drivage is high due to the large number of shortwalls mined by hydraulic monitors. Reducing mining cost in hydraulic mines depends on lowering drivage cost by use of new drivage systems or by increasing efficiency of drivage systems used at present. The following drivage methods used in hydraulic mines are compared: heading machines with hydraulic haulage of cut rocks and coal, hydraulic monitors with hydraulic haulage, drilling and blasting with hydraulic haulage of blasted rocks. Mining and geologic conditions which influence selection of the optimum mine drivage system are analyzed. Standardized cross sections of mine roadways driven by the 3 methods are shown in schemes. Support systems used in mine roadways are compared: timber supports, roof bolts, roof bolts with steel elements, and roadways driven in rocks without a support system. Heading machines (K-56MG, GPKG, 4PU, PK-3M) and hydraulic monitors (GMDTs-3M, 12GD-2) used for mine drivage are described. Data on mine drivage in hydraulic coal mines in the Kuzbass are discussed. From 40 to 46% of roadways are driven by heading machines with hydraulic haulage and from 12 to 15% by hydraulic monitors with hydraulic haulage.

Toward a Stakeholder Perspective on Social Stability Risk of Large Hydraulic Engineering Projects in China: A Social Network Analysis

Directory of Open Access Journals (Sweden)

Zhengqi He

2018-04-01

Full Text Available In China, large hydraulic engineering projects have made a great contribution to social economic development; at the same time, they also lead to social risks that affect social stability. The pluralism of stakeholders in large hydraulic engineering projects and the complex interrelationship among stakeholders are the important factors affecting social stability risk. Previous studies of social stability risk have mainly focused on risk identification and risk assessment, without considering the relationships among stakeholders and their linkages of risks. For large hydraulic engineering projects, this paper investigated the relevant risk factors and their interrelationships through a literature review and interviews that represented stakeholder perspectives. The key social stability risk factors were identified based on social network analysis. A multi-channel project financial system, a perfect interest compensation mechanism, an efficient prevention mechanism of group events, and a complete project schedule control system were proposed to mitigate the social stability risks. This study combined stakeholder management with risk management by using social network analysis, providing reference for the social stability risk management of large engineering projects in China.

Effect of hydraulic hysteresis on the stability of infinite slopes under steady infiltration

Science.gov (United States)

Chen, Pan; Mirus, Benjamin B.; Lu, Ning; Godt, Jonathan W.

2017-01-01

Hydraulic hysteresis, including capillary soil water retention (SWR), air entrapment SWR, and hydraulic conductivity, is a common phenomenon in unsaturated soils. However, the influence of hydraulic hysteresis on suction stress, and subsequently slope stability, is generally ignored. This paper examines the influence of each of these three types of hysteresis on slope stability using an infinite slope stability analysis under steady infiltration conditions. First, hypothetical slopes for representative silty and sandy soils are examined. Then a monitored hillslope in the San Francisco Bay Area, California is assessed, using observed rainfall conditions and measured hydraulic and geotechnical properties of the colluvial soil. Results show that profiles of suction stress and the corresponding factor of safety are generally strongly affected by hydraulic hysteresis. Results suggest that each of the three types of hydraulic hysteresis may play a major role in the occurrence of slope failure, indicating that ignoring hydraulic hysteresis will likely lead to underestimates of failure potential and hence to inaccurate slope stability analysis.

Horizontal steam generator thermal-hydraulics

Energy Technology Data Exchange (ETDEWEB)

Ubra, O. [SKODA Praha Company, Prague (Czechoslovakia); Doubek, M. [Czech Technical Univ., Prague (Czechoslovakia)

1995-09-01

Horizontal steam generators are typical components of nuclear power plants with pressure water reactor type VVER. Thermal-hydraulic behavior of horizontal steam generators is very different from the vertical U-tube steam generator, which has been extensively studied for several years. To contribute to the understanding of the horizontal steam generator thermal-hydraulics a computer program for 3-D steady state analysis of the PGV-1000 steam generator has been developed. By means of this computer program, a detailed thermal-hydraulic and thermodynamic study of the horizontal steam generator PGV-1000 has been carried out and a set of important steam generator characteristics has been obtained. The 3-D distribution of the void fraction and 3-D level profile as functions of load and secondary side pressure have been investigated and secondary side volumes and masses as functions of load and pressure have been evaluated. Some of the interesting results of calculations are presented in the paper.

Basic hydraulics

CERN Document Server

Smith, P D

1982-01-01

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

Development of neutronics and thermal hydraulics coupled code – SAC-RIT for plate type fuel and its application to reactivity initiated transient analysis

International Nuclear Information System (INIS)

Singh, Tej; Kumar, Jainendra; Mazumdar, Tanay; Raina, V.K.

2013-01-01

Highlights: • A point reactor kinetics code coupled with thermal hydraulics of plate type fuel is developed. • This code is applicable for two phase flow of coolant. • Safety analysis of IAEA benchmark reactor core is carried out. • Results agree well with the results available in literature. - Abstract: A point reactor kinetics code SAC-RIT, acronym of Safety Analysis Code for Reactivity Initiated Transient, coupled with thermal hydraulics of two phase coolant flow for plate type fuel, is developed to calculate reactivity initiated transient analysis of nuclear research and test reactors. Point kinetics equations are solved by fourth order Runge Kutta method. Reactivity feedback effect is included into the code. Solution of kinetics equations gives neutronic power and it is then fed into a thermal hydraulic code where mass, momentum and thermal energy conservation equations are solved by explicit finite difference method to find out fuel, clad and coolant temperatures during transients. In this code, all possible flow regimes including laminar flow, transient flow and turbulent flow have been covered. Various heat transfer coefficients suitable for single liquid, sub-cooled boiling, saturation boiling, film boiling and single vapor phases are incorporated in the thermal hydraulic code

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)

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

International Nuclear Information System (INIS)

Portezan Filho, Otavio

1997-01-01

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

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)

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

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)

REVIEW OF ENERGY-SAVING TECHNOLOGIES IN MODERN HYDRAULIC DRIVES

Directory of Open Access Journals (Sweden)

Mykola Karpenko

2017-12-01

Full Text Available This paper focuses on review of modern energy­saving technologies in hydraulic drives. Described main areas of energy conservation in hydraulic drive (which in turn are divided into many under the directions and was established the popularity of them. Reviewed the comparative analysis of efficiency application of various strategies for energy saving in a hydraulic drive. Based on the review for further research a combined method of real­time control systems with energy­saving algorithms and regeneration unit – selected for maxing efficiency in hydraulic drive. Scientific papers (40 papers, what introduced in review, is not older than 15 years in the databases “Sciencedirect” and “Scopus”.

Thermal-Hydraulic Tests for Reactor Core Safety

International Nuclear Information System (INIS)

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

2005-04-01

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

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)

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

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.

The influence of thermodynamic state of mineral hydraulic oil on flow rate through radial clearance at zero overlap inside the hydraulic components

Directory of Open Access Journals (Sweden)

Knežević Darko M.

2016-01-01

Full Text Available In control hydraulic components (servo valves, LS regulators, etc. there is a need for precise mathematical description of fluid flow through radial clearances between the control piston and body of component at zero overlap, small valve opening and small lengths of overlap. Such a mathematical description would allow for a better dynamic analysis and stability analysis of hydraulic systems. The existing formulas in the literature do not take into account the change of the physical properties of the fluid with a change of thermodynamic state of the fluid to determine the flow rate through radial clearances in hydraulic components at zero overlap, a small opening, and a small overlap lengths, which leads to the formation of insufficiently precise mathematical models. In this paper model description of fluid flow through radial clearances at zero overlap is developed, taking into account the changes of physical properties of hydraulic fluid as a function of pressure and temperature. In addition, the experimental verification of the mathematical model is performed.

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.

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.

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...

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.

TISKTH-3: a couple neutronics/thermal-hydraulics code for the transient analysis of light water reactors

International Nuclear Information System (INIS)

Peng Muzhang; Zhang Quan; Wang Guoli; Zhang Yuman

1988-01-01

TISKTH-3 is a coupled neutronics/thermal-hydraulics code for the transient analysis. A 3-dimensional neutron kinetics equation solved by the Nodal Green's Function Method is used for the neutronics model of the code. A homogeneous equilibrium model with a complete boiling curve and two numerical solutions of the implicit and explicit scheme is used for the thermal-hydraulics model of the code. A 2-dimensional heat conduction equation with variable conductivity solved by the method of weighted residuals is used for the fuel rod heat transfer model of the code. TISKTH-3 is able to analyze the fast transient process and complicate accident situations in the core. The initative applications have shown that the stability and convergency in the calculations with the code are satisfactory

TISKTH-3: a couple neutronics/thermal-hydraulics code for the transient analysis of light water reactors

Energy Technology Data Exchange (ETDEWEB)

Muzhang, Peng; Quan, Zhang; Guoli, Wang; Yuman, Zhang

1988-03-01

TISKTH-3 is a coupled neutronics/thermal-hydraulics code for the transient analysis. A 3-dimensional neutron kinetics equation solved by the Nodal Green's Function Method is used for the neutronics model of the code. A homogeneous equilibrium model with a complete boiling curve and two numerical solutions of the implicit and explicit scheme is used for the thermal-hydraulics model of the code. A 2-dimensional heat conduction equation with variable conductivity solved by the method of weighted residuals is used for the fuel rod heat transfer model of the code. TISKTH-3 is able to analyze the fast transient process and complicate accident situations in the core. The initative applications have shown that the stability and convergency in the calculations with the code are satisfactory.

Handbook of hydraulic fluid technology

CERN Document Server

Totten, George E

2011-01-01

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

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.

Thermal-Hydraulic Experiments and Modelling for Advanced Nuclear Reactor Systems

International Nuclear Information System (INIS)

Song, C. H.; Chung, M. K.; Park, C. K. and others

2005-04-01

The objectives of the project are to study thermal hydraulic characteristics of reactor primary system for the verification of the reactor safety and to evaluate new safety concepts of new safety design features. To meet the research goal, several thermal hydraulic experiments were performed and related thermal hydraulic models were developed with the experimental data which were produced through the thermal hydraulic experiments. Followings are main research topics; - Multi-dimensional Phenomena in a Reactor Vessel Downcomer - Condensation Load and Thermal Mixing in the IRWST - Development of Thermal-Hydraulic Models for Two-Phase Flow - Development of Measurement Techniques for Two-Phase Flow - Supercritical Reactor T/H Characteristics Analysis From the above experimental and analytical studies, new safety design features of the advanced power reactors were verified and lots of the safety issues were also resolved

Thermal-Hydraulic Experiments and Modelling for Advanced Nuclear Reactor Systems

Energy Technology Data Exchange (ETDEWEB)

Song, C. H.; Chung, M. K.; Park, C. K. and others

2005-04-15

The objectives of the project are to study thermal hydraulic characteristics of reactor primary system for the verification of the reactor safety and to evaluate new safety concepts of new safety design features. To meet the research goal, several thermal hydraulic experiments were performed and related thermal hydraulic models were developed with the experimental data which were produced through the thermal hydraulic experiments. Followings are main research topics; - Multi-dimensional Phenomena in a Reactor Vessel Downcomer - Condensation Load and Thermal Mixing in the IRWST - Development of Thermal-Hydraulic Models for Two-Phase Flow - Development of Measurement Techniques for Two-Phase Flow - Supercritical Reactor T/H Characteristics Analysis From the above experimental and analytical studies, new safety design features of the advanced power reactors were verified and lots of the safety issues were also resolved.

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

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

The hydraulic conductivity of sediments: A pore size perspective

KAUST Repository

Ren, X.W.

2017-12-06

This article presents an analysis of previously published hydraulic conductivity data for a wide range of sediments. All soils exhibit a prevalent power trend between the hydraulic conductivity and void ratio. Data trends span 12 orders of magnitude in hydraulic conductivity and collapse onto a single narrow trend when the hydraulic conductivity data are plotted versus the mean pore size, estimated using void ratio and specific surface area measurements. The sensitivity of hydraulic conductivity to changes in the void ratio is higher than the theoretical value due to two concurrent phenomena: 1) percolating large pores are responsible for most of the flow, and 2) the larger pores close first during compaction. The prediction of hydraulic conductivity based on macroscale index parameters in this and similar previous studies has reached an asymptote in the range of kmeas/5≤kpredict≤5kmeas. The remaining uncertainty underscores the important role of underlying sediment characteristics such as pore size distribution, shape, and connectivity that are not measured with index properties. Furthermore, the anisotropy in hydraulic conductivity cannot be recovered from scalar parameters such as index properties. Overall, results highlight the robustness of the physics inspired data scrutiny based Hagen–Poiseuille and Kozeny-Carman analyses.

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.

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

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

Experimental Determination of Hydraulic Properties of Unsaturated Calcarenites

Science.gov (United States)

Turturro, Antonietta Celeste; Andriani, Gioacchino Francesco; Clementina Caputo, Maria; Maggi, Sabino

2013-04-01

Understanding hydraulic properties is essential in the modeling of flow and solute transport through the vadose zone, to which problems of soil and groundwater pollution are related. The vadose zone, in fact, is of great importance in controlling groundwater recharge and transport of contaminants into and through the subsoil. The aim of this work is to determine experimentally in laboratory the hydraulic properties of unsaturated calcarenites using an approach including petrophysical determinations and methods for measuring water retention. For this purpose, samples of calcarenites belonging to the Calcarenite di Gravina Fm.(Pliocene-early Pleistocene), came from two different quarry districts located in Southern Italy (Canosa di Puglia and Massafra), were utilized. The water retention function, θ(h), which binds the water content, θ, to water potential, h, was determined in the laboratory by means two different experimental methods: the WP4-T psychrometer and the suction table. At last, a simple mathematical equation represented by van Genuchten's model is fitted to the experimental data and the unknown empirical parameters of this model are determined. Textural analysis on thin sections using optical petrographic microscopy and evaluation of total and effective porosity by means of standard geotechnical laboratory tests, mercury intrusion porosimetry and image analysis were also performed. In particular, a comparison between mercury porosimetry data and results of photomicrograph computer analysis through the methods of quantitative stereology was employed for providing pore size distributions. The results of this study identify the relationship between the hydraulic behavior, described by the water retention function, and pore size distribution for the calcarenites that are not easy to hydraulically characterize. This relationship could represent a useful tool to infer the unsaturated hydraulic properties of calcarenites and in general this approach could be

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.

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)

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)

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

Control rod drive hydraulic device

International Nuclear Information System (INIS)

Takekawa, Toru.

1994-01-01

The device of the present invention can reliably prevent a possible erroneous withdrawal of control rod driving mechanism when the pressure of a coolant line is increased by isolation operation of hydraulic control units upon periodical inspection for a BWR type reactor. That is, a coolant line is connected to the downstream of a hydraulic supply device. The coolant line is connected to a hydraulic control unit. A coolant hydraulic detection device and a pressure setting device are disposed to the coolant line. A closing signal line and a returning signal line are disposed, which connect the hydraulic supply device and a flow rate control valve for the hydraulic setting device. In the device of the present invention, even if pressure of supplied coolants is elevated due to isolation of hydraulic control units, the elevation of the hydraulic pressure can be prevented. Accordingly, reliability upon periodical reactor inspection can be improved. Further, the facility is simplified and the installation to an existent facility is easy. (I.S.)

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.

Study on Characteristics of Hydraulic Servo System for Force Control of Hydraulic Robots

International Nuclear Information System (INIS)

Kim, Hyo-gon; Han, Changsoo; Lee, Jong-won; Park, Sangdeok

2015-01-01

Because a hydraulic actuator has high power and force densities, this allows the weight of the robot's limbs to be reduced. This allows for good dynamic characteristics and high energy efficiency. Thus, hydraulic actuators are used in some exoskeleton robots and quadrupedal robots that require high torque. Force control is useful for robot compliance with a user or environment. However, force control of a hydraulic robot is difficult because a hydraulic servo system is highly nonlinear from a control perspective. In this study, a nonlinear model was used to develop a simulation program for a hydraulic servo system consisting of a servo valve, transmission lines, and a cylinder. The problems and considerations with regard to the force control performance for a hydraulic servo system were investigated. A force control method using the nonlinear model was proposed, and its effect was evaluated with the simulation program

Study on Characteristics of Hydraulic Servo System for Force Control of Hydraulic Robots

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyo-gon; Han, Changsoo [Hanyang University, Seoul (Korea, Republic of); Lee, Jong-won [Korea University of Science and Technology, Seoul (Korea, Republic of); Park, Sangdeok [Korea Institute of Industrial Technology, Seoul (Korea, Republic of)

2015-02-15

Because a hydraulic actuator has high power and force densities, this allows the weight of the robot's limbs to be reduced. This allows for good dynamic characteristics and high energy efficiency. Thus, hydraulic actuators are used in some exoskeleton robots and quadrupedal robots that require high torque. Force control is useful for robot compliance with a user or environment. However, force control of a hydraulic robot is difficult because a hydraulic servo system is highly nonlinear from a control perspective. In this study, a nonlinear model was used to develop a simulation program for a hydraulic servo system consisting of a servo valve, transmission lines, and a cylinder. The problems and considerations with regard to the force control performance for a hydraulic servo system were investigated. A force control method using the nonlinear model was proposed, and its effect was evaluated with the simulation program.

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.

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.

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...

Hydraulic Hybrid Vehicles

Science.gov (United States)

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

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.

The hydraulic wheel

International Nuclear Information System (INIS)

Alvarez Cardona, A.

1985-01-01

The present article this dedicated to recover a technology that key in disuse for the appearance of other techniques. It is the hydraulic wheel with their multiple possibilities to use their energy mechanical rotational in direct form or to generate electricity directly in the fields in the place and to avoid the high cost of transport and transformation. The basic theory is described that consists in: the power of the currents of water and the hydraulic receivers. The power of the currents is determined knowing the flow and east knowing the section of the flow and its speed; they are given you formulate to know these and direct mensuration methods by means of floodgates, drains and jumps of water. The hydraulic receivers or properly this hydraulic wheels that are the machines in those that the water acts like main force and they are designed to transmit the biggest proportion possible of absolute work of the water, the hydraulic wheels of horizontal axis are the common and they are divided in: you rotate with water for under, you rotate with side water and wheels with water for above. It is analyzed each one of them, their components are described; the conditions that should complete to produce a certain power and formulate them to calculate it. There are 25 descriptive figures of the different hydraulic wheels

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.

Determination of thermal-hydraulic loads on reactor internals in a DBA-situation

International Nuclear Information System (INIS)

Ville Lestinen; Timo Toppila

2005-01-01

Full text of publication follows: According to Finnish regulatory requirements, reactor internals have to stay intact in a design basis accident (DBA) situation, so that control rods can still penetrate into the core. To fulfill this demand some criteria must be followed in periodical in-service inspections. This is the motivation for studying and developing more detailed methods for analysis of thermal-hydraulic loads on reactor internals during the DBA-situation for the Loviisa NPP in Finland. The objective of this research program is to connect thermal-hydraulic and mechanical analysis methods with the goal to produce a reliable method for determination of thermal-hydraulic and mechanical loads on reactor internals in the accident situation. The tools studied are thermal-hydraulic system codes, computational fluid dynamics (CFD) codes and finite element analysis (FEA) codes. This paper concentrates mainly on thermal-hydraulic part of the research, but also the mechanical aspects are discussed. Firstly, the paper includes a short literary review of the available methods to analyse the described problem including both thermal-hydraulic and structural analysis parts. Secondly, different possibilities to carry out thermal-hydraulic analyses have been studied. The DBA-case includes complex physical phenomena and therefore modelling is difficult. The accident situation can be for example LLOCA. When the pipe has broken, the pressure decreases and water starts to evaporate, which consumes energy and that way limits the pressure decrease. After some period of time, the system reaches a new equilibrium state. To perform exact thermal-hydraulic analysis also two phase phenomena must be included. Therefore CFD codes are not capable of modelling the DBA situation very well, but the use of CFD codes requires that the effect of two phase flow must be added somehow. One method to calculate two phase phenomena with CFD codes is to use thermal-hydraulic system codes to calculate

Inducement of Design Parameters for Reliability Improvement of Servo Actuator for Hydraulic Valve Operation

Energy Technology Data Exchange (ETDEWEB)

Sung, Baek Ju; Kim, Do Sik [Korea Institute of Machinery and Materials, Daejeon (Korea, Republic of)

2014-05-15

The precision hydraulic valve is widely used in various industrial field like aircraft, automobile, and general machinery. Servo actuator is the most important device for driving the precise hydraulic valve. The reliable operation of servo actuator effects on the overall hydraulic system. The performance of servo actuator relies on frequency response and step response according to arbitrary input signal. In this paper, we performed the analysis for the components of servo actuator to satisfy the reliable operation and response characteristics through the reliability analysis, and also induced the design parameters to realize the reliable operation and fast response characteristics of servo actuator for hydraulic valve operation through the empirical knowledge of experts and electromagnetic theories. We suggested the design equations to determine the values of design parameters of servo actuator as like bobbin size, length of yoke and plunger and turn number of coil, and verified the achieved design values through FEM analysis and performance tests using some prototypes of servo actuators adapted in hydraulic valve.

Light-water-reactor coupled neutronic and thermal-hydraulic codes

International Nuclear Information System (INIS)

Diamond, D.J.

1982-01-01

An overview is presented of computer codes that model light water reactor cores with coupled neutronics and thermal-hydraulics. This includes codes for transient analysis and codes for steady state analysis which include fuel depletion and fission product buildup. Applications in nuclear design, reactor operations and safety analysis are given and the major codes in use in the USA are identified. The neutronic and thermal-hydraulic methodologies and other code features are outlined for three steady state codes (PDQ7, NODE-P/B and SIMULATE) and four dynamic codes (BNL-TWIGL, MEKIN, RAMONA-3B, RETRAN-02). Speculation as to future trends with such codes is also presented

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.

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...

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

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

Investigation and Development of the Thermal Preparation System of the Trailbuilder Machinery Hydraulic Actuator

Science.gov (United States)

Konev, V.; Polovnikov, E.; Krut, O.; Merdanov, Sh; Zakirzakov, G.

2017-07-01

It’s determined that the main part of trailbuilders operated in the North is the technology equipped by the hydraulic actuator. Further development of the northern territories will demand using of various means and ways machinery thermal preparation, and also the machinery of the northern fulfillment. On this basis problems in equipment operation are defined. One of the main is efficiency supplying of a hydraulic actuator. On the basis of the operating conditions’ analysis of trailbuilder hydraulic actuator operation it is determined, that under low negative temperatures the means of thermal preparation are necessary. The existing systems warm up only a hydraulic tank or warming up of the hydro equipment before the machinery operation is carried out under loading with intensive wears. Thus, with the purpose to raise the efficiency of thermal hydraulic actuator, operated far from stationary bases autonomous, energy saving, not expensive in creation and operation systems are necessary. In accordance with the analysis of means and ways of the thermal preparation of the hydraulic actuator and the thermal balance calculations of the (internal) combustion engine the system of the hydraulic actuator heating is offered and is being investigated. It contains a local hydraulic actuator warming up and the system of internal combustion engine heat utilization. Within research operation conditions of the local hydraulic actuator heating are viewed and determined, taking into account constructive changes to the local hydraulic actuator heating. Mathematical modelling of the heat technical process in the modernized hydraulic actuator is considered. As a result temperature changes of the heat-transfer and the hydraulic cylinder in time are determined. To check the theoretical researches and to define dependences on hydraulic actuator warming up, the experimental installation is made. It contains the measuring equipment, a small tank with the heat exchanger of the burnt gases

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.)

Characteristic Length Scales in Fracture Networks: Hydraulic Connectivity through Periodic Hydraulic Tests

Science.gov (United States)

Becker, M.; Bour, O.; Le Borgne, T.; Longuevergne, L.; Lavenant, N.; Cole, M. C.; Guiheneuf, N.

2017-12-01

Determining hydraulic and transport connectivity in fractured bedrock has long been an important objective in contaminant hydrogeology, petroleum engineering, and geothermal operations. A persistent obstacle to making this determination is that the characteristic length scale is nearly impossible to determine in sparsely fractured networks. Both flow and transport occur through an unknown structure of interconnected fracture and/or fracture zones making the actual length that water or solutes travel undetermined. This poses difficulties for flow and transport models. For, example, hydraulic equations require a separation distance between pumping and observation well to determine hydraulic parameters. When wells pairs are close, the structure of the network can influence the interpretation of well separation and the flow dimension of the tested system. This issue is explored using hydraulic tests conducted in a shallow fractured crystalline rock. Periodic (oscillatory) slug tests were performed at the Ploemeur fractured rock test site located in Brittany, France. Hydraulic connectivity was examined between three zones in one well and four zones in another, located 6 m apart in map view. The wells are sufficiently close, however, that the tangential distance between the tested zones ranges between 6 and 30 m. Using standard periodic formulations of radial flow, estimates of storativity scale inversely with the square of the separation distance and hydraulic diffusivity directly with the square of the separation distance. Uncertainty in the connection paths between the two wells leads to an order of magnitude uncertainty in estimates of storativity and hydraulic diffusivity, although estimates of transmissivity are unaffected. The assumed flow dimension results in alternative estimates of hydraulic parameters. In general, one is faced with the prospect of assuming the hydraulic parameter and inverting the separation distance, or vice versa. Similar uncertainties exist

Investigation of coupling scheme for neutronic and thermal-hydraulic codes

International Nuclear Information System (INIS)

Wang Guoli; Yu Jianfeng; Pen Muzhang; Zhang Yuman.

1988-01-01

Recently, a number of coupled neutronics/thermal-hydraulics codes have been used in reaction design and safty analysis, which have been obtained by coupling previous neutronic and thermal-hydraulic codes. The different coupling schemes affect computer time and accuracy of calculation results. Numberical experiments of several different coupling schemes and some heuristic results are described

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.

BWR 9 X 9 Fuel Assembly Thermal-Hydraulic Tests (2): Hydraulic Vibration Test

International Nuclear Information System (INIS)

Yoshiaki Tsukuda; Katsuichiro Kamimura; Toshiitsu Hattori; Akira Tanabe; Noboru Saito; Masahiko Warashina; Yuji Nishino

2002-01-01

Nuclear Power Engineering Corporation (NUPEC) conducted thermal-hydraulic projects for verification of thermal-hydraulic design reliability for BWR high-burnup 8 x 8 and 9 x 9 fuel assemblies, entrusted by the Ministry of Economy, Trade and Industry (METI). As a part of the NUPEC thermal-hydraulic projects, hydraulic vibration tests using full-scale test assemblies simulating 9 x 9 fuel assemblies were carried out to evaluate BWR fuel integrity. The test data were applied to development of a new correlation for the estimation of fuel rod vibration amplitude. (authors)

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

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.

Stochastic Spectral Analysis for Characterizing Hydraulic Diffusivity in an Alluvial Fan Aquifer with River Stimulus

Science.gov (United States)

Wang, Y. L.; Zha, Y.; Yeh, T. C. J.; Wen, J. C.

2015-12-01

Estimation of subsurface hydraulic diffusivity was carried out to understand the characteristics of Zhuoshui River alluvial fan, Taiwan. The fan, an important agricultural and industrial region with high water demand, is located at middle Taiwan with an area of 1800 km2. The prior geo-investigations suggest that the main recharge region of the fan is at an apex along the river. The distribution of soil hydraulic diffusivity was estimated by fusing naturally recurring stimulus provided by river and groundwater head. Specifically, the variance and power spectrum provided by temporal and spatial change of groundwater head in response to river stage variations are analyzed to estimate hydraulic diffusivity distribution. It is found that the hydraulic diffusivity of the fan is at the range from 0.08 to 16 m2/s. The average hydraulic diffusivity at the apex, middle, and tail of the fan along the river is about 0.4, 0.6, and 1.0 m2/s, respectively.

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

International Nuclear Information System (INIS)

Lemonnier, H.

2005-01-01

The main topics covered by the NURETH 11 meeting are the thermal-hydraulics of existing and future nuclear power plants as foreseen by the Generation IV worldwide initiative. Normal operation and accidental situations are also relevant topics of the Conference. The topics cover modeling, experiments, instrumentation and numerical simulations related to flow and heat transfer in nuclear reactors with a special emphasis on the advances of multiphase CFD methods. The first part of this Book of Abstracts enumerates the Organizing Scientific Societies, the Sponsors of the Conference, the Conference Chairs, and the members of the Steering Committee and of the Technical Program Committee. The second part of this Book of Abstracts contains the list of the titles of the contributed papers. Each item includes the log number of the paper, the abstract of which can therefore be easily located in the next section of this book. The titles of the papers have been sorted out by topics to provide a synthetic view of the contributions in a selected domain. The last section of this Book includes an index of authors and co-authors with a reference to the log number(s) of their contributed paper(s). Finally, the CD-Rom of the Conference Proceedings containing the full-length papers is inserted at the inside back cover. Sessions content: A - two-phase flow and heat transfer fundamentals: computational and mathematical techniques (numerical schemes, LBM, BEM, mesh-less, etc.); contact angle and wettability phenomena; experiments and data bases for the assessment and the verification of 3D models; flow regime identification and modelling; heat transfer near critical pressure and supercritical water reactors; interfacial area (data base, modeling, measurement techniques); instrumentation techniques; micro-scale basic phenomena, fluid flow and heat transfer; scaling methods; counter current flow; B - code developments: containment analysis; core thermal-hydraulics and subchannel analysis

Thermal-hydraulic analysis of PWR small assembly for irradiation test of CARR

International Nuclear Information System (INIS)

Yin Hao; Zou Yao; Liu Xingmin

2015-01-01

The thermal-hydraulic behaviors of the PWR 4 × 4 small assembly tested in the high temperature and high pressure loop of China Advanced Research Reactor were analyzed. The CFD method was used to carry out 3D simulation of the model, thus detailed thermal-hydraulic parameters were obtained. Firstly, the simplified model was simulated to give the 3D temperature and velocity distributions and analyze the heat transfer process. Then the whole scale small assembly model was simulated and the simulation results were compared with those of simplified rod bundle model. Its flow behavior was studied and flow mixing characteristics of the grids were analyzed, and the mixing factor of the grid was calculated and can be used for further thermal-hydraulic study. It is shown that the highest temperature of the fuel rod meets the design limit and the mixing effect of the grid is obvious. (authors)

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)

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.

ENERGY EFFICIENCY OF DIESEL LOCOMOTIVE HYDRAULIC TRANSMISSION TESTS AT LOCOMOTIVE REPAIR PLANT

Directory of Open Access Journals (Sweden)

B. E. Bodnar

2015-10-01

Full Text Available Purpose. In difficult economic conditions, cost reduction of electricity consumption for the needs of production is an urgent task for the country’s industrial enterprises. Technical specifications of enterprises, which repair diesel locomotive hydraulic transmission, recommend conducting a certain amount of evaluation and regulatory tests to monitor their condition after repair. Experience shows that a significant portion of hydraulic transmission defects is revealed by bench tests. The advantages of bench tests include the ability to detect defects after repair, ease of maintenance of the hydraulic transmission and relatively low labour intensity for eliminating defects. The quality of these tests results in the transmission resource and its efficiency. Improvement of the technology of plant post-repairs hydraulic tests in order to reduce electricity consumption while testing. Methodology. The possible options for hydraulic transmission test bench improvement were analysed. There was proposed an energy efficiency method for diesel locomotive hydraulic transmission testing in locomotive repair plant environment. This is achieved by installing additional drive motor which receives power from the load generator. Findings. Based on the conducted analysis the necessity of improving the plant stand testing of hydraulic transmission was proved. The variants of the stand modernization were examined. The test stand modernization analysis was conducted. Originality. The possibility of using electric power load generator to power the stand electric drive motor or the additional drive motor was theoretically substantiated. Practical value. A variant of hydraulic transmission test stand based on the mutual load method was proposed. Using this method increases the hydraulic transmission load range and power consumption by stand remains unchanged. The additional drive motor will increase the speed of the input shaft that in its turn wil allow testing in

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)

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.

Power Management in Mobile Hydraulic Applications - An Approach for Designing Hydraulic Power Supply Systems

DEFF Research Database (Denmark)

Pedersen, Henrik Clemmensen

2004-01-01

Throughout the last three decades energy consumption has become one of the primary design aspects in hydraulic systems, especially for mobile hydraulic systems, as power and cooling capacity here is at limited disposal. Considering the energy usage, this is dependent on component efficiency, but ...... the hydraulic power supply in the most energy efficient way, when considering a number of load situations. Finally an example of the approach is shown to prove its validity.}......Throughout the last three decades energy consumption has become one of the primary design aspects in hydraulic systems, especially for mobile hydraulic systems, as power and cooling capacity here is at limited disposal. Considering the energy usage, this is dependent on component efficiency...

Views on the future of thermal hydraulic modeling

Energy Technology Data Exchange (ETDEWEB)

Ishii, M. [Purdue Univ., West Lafayette, IN (United States)

1997-07-01

It is essential for the U.S. NRC to sustain the highest level of the thermal-hydraulics and reactor safety research expertise and continuously improve their accident analysis capability. Such expertise should span over four different areas which are strongly related to each other. These are: (1) Reactor Safety Code Development, (2) Two-phase Flow Modeling, (3) Instrumentation and Fundamental Experimental Research, and (4) Separate Effect and Integral Test. The NRC is already considering a new effort in the area of advanced thermal-hydraulics effort. Its success largely depends on the availability of a significantly improved two-phase flow formulation and constitutive relations supported by detailed experimental data. Therefore, it is recommended that the NRC start significant research efforts in the areas of two-phase flow modeling, instrumentation, basic and separate effect experiments which should be pursued systematically and with clearly defined objectives. It is desirable that some international program is developed in this area. This paper is concentrated on those items in the thermal-hydraulic area which eventually determine the quality of future accident analysis codes.

Views on the future of thermal hydraulic modeling

International Nuclear Information System (INIS)

Ishii, M.

1997-01-01

It is essential for the U.S. NRC to sustain the highest level of the thermal-hydraulics and reactor safety research expertise and continuously improve their accident analysis capability. Such expertise should span over four different areas which are strongly related to each other. These are: (1) Reactor Safety Code Development, (2) Two-phase Flow Modeling, (3) Instrumentation and Fundamental Experimental Research, and (4) Separate Effect and Integral Test. The NRC is already considering a new effort in the area of advanced thermal-hydraulics effort. Its success largely depends on the availability of a significantly improved two-phase flow formulation and constitutive relations supported by detailed experimental data. Therefore, it is recommended that the NRC start significant research efforts in the areas of two-phase flow modeling, instrumentation, basic and separate effect experiments which should be pursued systematically and with clearly defined objectives. It is desirable that some international program is developed in this area. This paper is concentrated on those items in the thermal-hydraulic area which eventually determine the quality of future accident analysis codes

Effects of turbine's selection on hydraulic transients in the long pressurized water conveyance system

International Nuclear Information System (INIS)

Zhou, J X; Hu, M; Cai, F L; Huang, X T

2014-01-01

For a hydropower station with longer water conveyance system, an optimum turbine's selection will be beneficial to its reliable and stable operation. Different optional turbines will result in possible differences of the hydraulic characteristics in the hydromechanical system, and have different effects on the hydraulic transients' analysis and control. Therefore, the premise for turbine's selection is to fully understand the properties of the optional turbines and their effects on the hydraulic transients. After a brief introduction of the simulation models for hydraulic transients' computation and stability analysis, the effects of hydraulic turbine's characteristics at different operating points on the hydro-mechanical system's free vibration analysis were theoretically investigated with the hydraulic impedance analysis of the hydraulic turbine. For a hydropower station with long water conveyance system, based on the detailed hydraulic transients' computation respectively for two different optional turbines, the effects of the turbine's selection on hydraulic transients were analyzed. Furthermore, considering different operating conditions for each turbine and the similar operating conditions for these two turbines, free vibration analysis was comprehensively carried out to reveal the effects of turbine's impedance on system's vibration characteristics. The results indicate that, respectively with two different turbines, most of the controlling parameters under the worst cases have marginal difference, and few shows obvious differences; the turbine's impedances under different operating conditions have less effect on the natural angular frequencies; different turbine's characteristics and different operating points have obvious effects on system's vibration stability; for the similar operating conditions of these two turbines, system's vibration characteristics are basically consistent with

Hydraulic Yaw System

DEFF Research Database (Denmark)

Stubkier, Søren; Pedersen, Henrik C.; Mørkholt, M.

a hydraulic soft yaw system, which is able to reduce the loads on the wind turbine significantly. A full scale hydraulic yaw test rig is available for experiments and tests. The test rig is presented as well as the system schematics of the hydraulic yaw system....... the HAWC2 aeroelastic code and an extended model of the NREL 5MW turbine combined with a simplified linear model of the turbine, the parameters of the soft yaw system are optimized to reduce loading in critical components. Results shows that a significant reduction in fatigue and extreme loads to the yaw...... system and rotor shaft when utilizing the soft yaw drive concept compared to the original stiff yaw system. The physical demands of the hydraulic yaw system are furthermore examined for a life time of 20 years. Based on the extrapolated loads, the duty cycles show that it is possible to construct...

Hydraulic lifter of a drilling unit

Energy Technology Data Exchange (ETDEWEB)

Velikovskiy, L S; Demin, A V; Shadchinov, L M

1979-01-08

The invention refers to drilling equipment, in particular, devices for lowering and lifting operations during drilling. A hydraulic lifter of the drilling unit is suggested which contains a hydraulic cylinder, pressure line and hollow plunger whose cavities are hydraulically connected. In order to improve the reliability of the hydraulic lifter by balancing the forces of compression in the plunger of the hydraulic cylinder, a closed vessel is installed inside the plunger and rigidly connected to its ends. Its cavity is hydraulically connected to the pressure line.

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

Process management using component thermal-hydraulic function classes

Science.gov (United States)

Morman, J.A.; Wei, T.Y.C.; Reifman, J.

1999-07-27

A process management expert system where following malfunctioning of a component, such as a pump, for determining system realignment procedures such as for by-passing the malfunctioning component with on-line speeds to maintain operation of the process at full or partial capacity or to provide safe shut down of the system while isolating the malfunctioning component. The expert system uses thermal-hydraulic function classes at the component level for analyzing unanticipated as well as anticipated component malfunctions to provide recommended sequences of operator actions. Each component is classified according to its thermal-hydraulic function, and the generic and component-specific characteristics for that function. Using the diagnosis of the malfunctioning component and its thermal hydraulic class, the expert system analysis is carried out using generic thermal-hydraulic first principles. One aspect of the invention employs a qualitative physics-based forward search directed primarily downstream from the malfunctioning component in combination with a subsequent backward search directed primarily upstream from the serviced component. Generic classes of components are defined in the knowledge base according to the three thermal-hydraulic functions of mass, momentum and energy transfer and are used to determine possible realignment of component configurations in response to thermal-hydraulic function imbalance caused by the malfunctioning component. Each realignment to a new configuration produces the accompanying sequence of recommended operator actions. All possible new configurations are examined and a prioritized list of acceptable solutions is produced. 5 figs.

Process management using component thermal-hydraulic function classes

Science.gov (United States)

Morman, James A.; Wei, Thomas Y. C.; Reifman, Jaques

1999-01-01

A process management expert system where following malfunctioning of a component, such as a pump, for determining system realignment procedures such as for by-passing the malfunctioning component with on-line speeds to maintain operation of the process at full or partial capacity or to provide safe shut down of the system while isolating the malfunctioning component. The expert system uses thermal-hydraulic function classes at the component level for analyzing unanticipated as well as anticipated component malfunctions to provide recommended sequences of operator actions. Each component is classified according to its thermal-hydraulic function, and the generic and component-specific characteristics for that function. Using the diagnosis of the malfunctioning component and its thermal hydraulic class, the expert system analysis is carried out using generic thermal-hydraulic first principles. One aspect of the invention employs a qualitative physics-based forward search directed primarily downstream from the malfunctioning component in combination with a subsequent backward search directed primarily upstream from the serviced component. Generic classes of components are defined in the knowledge base according to the three thermal-hydraulic functions of mass, momentum and energy transfer and are used to determine possible realignment of component configurations in response to thermal-hydraulic function imbalance caused by the malfunctioning component. Each realignment to a new configuration produces the accompanying sequence of recommended operator actions. All possible new configurations are examined and a prioritized list of acceptable solutions is produced.

Hydraulic Soft Yaw System for Multi MW Wind Turbines

DEFF Research Database (Denmark)

Stubkier, Søren

energy and an increase in the loading of the wind turbine structure and components. This dissertation examines the hypothesis that there are advantages of basing a yaw system on hydraulic components instead of normal electrical components. This is done through a state of the art analysis followed...... in the wind turbine yaw system along with minor reductions in the blades and main shaft. Optimization of the damping and stiffness of the hydraulic soft yaw system have been conducted and an optimum found for load reduction. Linear control algorithms for control of damping pressure peaks have been developed...... the full turbine code in FAST, and the mathematical model of the hydraulic yaw system in Matlab/Simulink and Amesim is developed in order to analyze a full scale model of the hydraulic yaw system in combination with the implemented friction model for the yaw system. These results are also promising...

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

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.

Numerical Simulation of Hydraulic Fracture Propagation Guided by Single Radial Boreholes

Directory of Open Access Journals (Sweden)

Tiankui Guo

2017-10-01

Full Text Available Conventional hydraulic fracturing is not effective in target oil development zones with available wellbores located in the azimuth of the non-maximum horizontal in-situ stress. To some extent, we think that the radial hydraulic jet drilling has the function of guiding hydraulic fracture propagation direction and promoting deep penetration, but this notion currently lacks an effective theoretical support for fracture propagation. In order to verify the technology, a 3D extended finite element numerical model of hydraulic fracturing promoted by the single radial borehole was established, and the influences of nine factors on propagation of hydraulic fracture guided by the single radial borehole were comprehensively analyzed. Moreover, the term ‘Guidance factor (Gf’ was introduced for the first time to effectively quantify the radial borehole guidance. The guidance of nine factors was evaluated through gray correlation analysis. The experimental results were consistent with the numerical simulation results to a certain extent. The study provides theoretical evidence for the artificial control technology of directional propagation of hydraulic fracture promoted by the single radial borehole, and it predicts the guidance effect of a single radial borehole on hydraulic fracture to a certain extent, which is helpful for planning well-completion and fracturing operation parameters in radial borehole-promoted hydraulic fracturing technology.

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)

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)

INFORMATION-MEASURING TEST SYSTEM OF DIESEL LOCOMOTIVE HYDRAULIC TRANSMISSIONS

Directory of Open Access Journals (Sweden)

I. V. Zhukovytskyy

2015-08-01

Full Text Available Purpose. The article describes the process of developing the information-measuring test system of diesel locomotives hydraulic transmission, which gives the possibility to obtain baseline data to conduct further studies for the determination of the technical condition of diesel locomotives hydraulic transmission. The improvement of factory technology of post-repair tests of hydraulic transmissions by automating the existing hydraulic transmission test stands according to the specifications of the diesel locomotive repair enterprises was analyzed. It is achieved based on a detailed review of existing foreign information-measuring test systems for hydraulic transmission of diesel locomotives, BelAZ earthmover, aircraft tug, slag car, truck, BelAZ wheel dozer, some brands of tractors, etc. The problem for creation the information-measuring test systems for diesel locomotive hydraulic transmission is being solved, starting in the first place from the possibility of automation of the existing test stand of diesel locomotives hydraulic transmission at Dnipropetrovsk Diesel Locomotive Repair Plant "Promteplovoz". Methodology. In the work the researchers proposed the method to create a microprocessor automated system of diesel locomotives hydraulic transmission stand testing in the locomotive plant conditions. It acts by justifying the selection of the necessary sensors, as well as the application of the necessary hardware and software for information-measuring systems. Findings. Based on the conducted analysis there was grounded the necessity of improvement the plant hydraulic transmission stand testing by creating a microprocessor testing system, supported by the experience of developing such systems abroad. Further research should be aimed to improve the accuracy and frequency of data collection by adopting the more modern and reliable sensors in tandem with the use of filtering software for electromagnetic and other interference. Originality. The

Thermal-hydraulic analysis of an annular fuel element: The Achilles' heel of the particle bed reactor

International Nuclear Information System (INIS)

Dibben, M.J.; Tuttle, R.F.

1993-01-01

The low pressure nuclear thermal propulsion (LPNTP) concept offers significant improvements in rocket engine specific impulse over rockets employment chemical propulsion. This study investigated a parametric thermal-hydraulic analysis of an annular fueld element, also referred to as a fuel pipe, using the computer code ATHENA (Advanced Thermal Hydraulic Energy Network Analyzer). The fuelpipe is an annular particle bed fuel element of the reactor with radially inward flow of hydrogen through the element. In this study, the outlet temperature of the hydrogen is parametrically related to key effects, including the reactor power at two different pressure drops, the effect of power coupling for in-core testing, and the effect of hydrogen flow rates. Results show that the temperature is linearly related to the reactor power, but not to pressure drop, and that cross flow inside the fuelpipe occurs at approximately 0.3 percent of the radial flow rates

Hydraulic Structures : Caissons

NARCIS (Netherlands)

Voorendt, M.Z.; Molenaar, W.F.; Bezuyen, K.G.

These lecture notes on caissons are part of the study material belonging to the course 'Hydraulic Structures 1' (code CTB3355), part of the Bachelor of Science education and the Hydraulic Engineering track of the Master of Science education for civil engineering students at Delft University of

Virginia Power thermal-hydraulics methods

International Nuclear Information System (INIS)

Anderson, R.C.; Basehore, K.L.; Harrell, J.R.

1987-01-01

Virginia Power's nuclear safety analysis group is responsible for the safety analysis of reload cores for the Surry and North Anna power stations, including the area of core thermal-hydraulics. Postulated accidents are evaluated for potential departure from nucleate boiling violations. In support of these tasks, Virginia Power has employed the COBRA code and the W-3 and WRB-1 DNB correlations. A statistical DNBR methodology has also been developed. The code, correlations and statistical methodology are discussed

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)

Slip flow coefficient analysis in water hydraulics gear pump for environmental friendly application

International Nuclear Information System (INIS)

Yusof, A A; Wasbari, F; Zakaria, M S; Ibrahim, M Q

2013-01-01

Water hydraulics is the sustainable option in developing fluid power systems with environmental friendly approach. Therefore, an investigation on water-based external gear pump application is being conducted, as a low cost solution in the shifting effort of using water, instead of traditional oil hydraulics in fluid power application. As the gear pump is affected by fluid viscosity, an evaluation has been conducted on the slip flow coefficient, in order to understand to what extent the spur gear pump can be used with water-based hydraulic fluid. In this paper, the results of a simulated study of variable-speed fixed displacement gear pump are presented. The slip flow coefficient varies from rotational speed of 250 RPM to 3500 RPM, and provides volumetric efficiency ranges from 9 % to 97% accordingly

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

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 ...

Determining the Conditions for the Hydraulic Impacts Emergence at Hydraulic Systems

Directory of Open Access Journals (Sweden)

Mazurenko A.S.

2017-08-01

Full Text Available This research aim is to develop a method for modeling the conditions for the critical hydrau-lic impacts emergence on thermal and nuclear power plants’ pipeline systems pressure pumps depart-ing from the general provisions of the heat and hydrodynamic instability theory. On the developed method basis, the conditions giving rise to the reliability-critical hydraulic impacts emergence on pumps for the thermal and nuclear power plants’ typical pipeline system have been determined. With the flow characteristic minimum allowable (critical sensitivity, the flow velocity fluctuations ampli-tude reaches critical values at which the pumps working elements’ failure occurs. The critical hydrau-lic impacts emergence corresponds to the transition of the vibrational heat-hydrodynamic instability into an aperiodic one. As research revealed, a highly promising approach as to the preventing the criti-cal hydraulic impacts related to the foreground use of pumps having the most sensitive consumption (at supply network performance (while other technical characteristics corresponding to that parame-ter. The research novelty refers to the suggested method elaborated by the authors’ team, which, in contrast to traditional approaches, is efficient in determining the pump hydraulic impact occurrence conditions when the vibrational heat-hydrodynamic instability transition to the aperiodic instability.

Water hydraulic applications in hazardous environments

International Nuclear Information System (INIS)

Siuko, M.; Koskinen, K.T.; Vilenius, M.J.

1996-01-01

Water hydraulic technology provides several advantages for devices operating in critical environment. Though water hydraulics has traditionally been used in very rough applications, gives recent strong development of components possibility to build more sophisticated applications and devices with similar capacity and control properties than those of oil hydraulics without the disadvantages of oil hydraulic systems. In this paper, the basic principles, possibilities and advantages of water hydraulics are highlighted, some of the most important design considerations are presented and recent developments of water hydraulic technology are presented. Also one interesting application area, ITER fusion reactor remote handling devices, are discussed. (Author)

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

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)

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)

Automated software for hydraulic simulation of pipeline operation

Directory of Open Access Journals (Sweden)

Hurgin Roman

2018-01-01

Full Text Available Design of modern water supply systems of large cities as well as their management via renovation of hydraulic models poses time-consuming tasks to researchers, and coping with this task requires specific approaches. When tackling these tasks, water services companies come across a lot of information about various objects of water infrastructure, the majority of which is located underground. In those cases, modern computer-aided design systems containing various components come to help. These systems help to solve a wide array of problems using existing information regarding pipelines, analysis and optimization of their basic parameters. CAD software is becoming an integral part of water supply systems management in large cities, and its capabilities allow engineering and operating companies to not only collect all the necessary data concerning water supply systems in any given city, but also to conduct research aimed at improving various parameters of these systems, including optimization of their hydraulic properties which directly determine the quality of water. This paper contains the analysis of automated CAD software for hydraulic design and management of city water supply systems in order to provide safe and efficient operation of these water supply systems. Authors select the most suitable software that might be used to provide hydraulic compatibility of old and new sections of water supply ring mains after selective or continuous draw-in renovation and decrease in diameter of distribution networks against the background of water consumption decrease in the cities.

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.

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.

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.

Thermal-hydraulic characteristic of the PGV-1000 steam generator

International Nuclear Information System (INIS)

Ubra, O.; Doubek, M.

1995-01-01

Horizontal steam generators are typical parts of nuclear power plants with pressure water reactor type VVER. By means of this computer program, a detailed thermal-hydraulic study of the horizontal steam generator PGV-1000 has been carried out and a special attention has been paid to the thermal-hydraulics of the secondary side. A set of important steam generator characteristics has been obtained and analyzed. Some of the interesting results of the analysis are presented in the paper. (author)

Current and anticipated uses of thermal hydraulic codes in Korea

Energy Technology Data Exchange (ETDEWEB)

Kim, Kyung-Doo; Chang, Won-Pyo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1997-07-01

In Korea, the current uses of thermal hydraulic codes are categorized into 3 areas. The first application is in designing both nuclear fuel and NSSS. The codes have usually been introduced based on the technology transfer programs agreed between KAERI and the foreign vendors. Another area is in the supporting of the plant operations and licensing by the utility. The third category is research purposes. In this area assessments and some applications to the safety issue resolutions are major activities using the best estimate thermal hydraulic codes such as RELAP5/MOD3 and CATHARE2. Recently KEPCO plans to couple thermal hydraulic codes with a neutronics code for the design of the evolutionary type reactor by 2004. KAERI also plans to develop its own best estimate thermal hydraulic code, however, application range is different from KEPCO developing code. Considering these activities, it is anticipated that use of the best estimate hydraulic analysis code developed in Korea may be possible in the area of safety evaluation within 10 years.

3D Hydraulic tomography from joint inversion of the hydraulic heads and self-potential data. (Invited)

Science.gov (United States)

Jardani, A.; Soueid Ahmed, A.; Revil, A.; Dupont, J.

2013-12-01

Pumping tests are usually employed to predict the hydraulic conductivity filed from the inversion of the head measurements. Nevertheless, the inverse problem is strongly underdetermined and a reliable imaging requires a considerable number of wells. We propose to add more information to the inversion of the heads by adding (non-intrusive) streaming potentials (SP) data. The SP corresponds to perturbations in the local electrical field caused directly by the fow of the ground water. These SP are obtained with a set of the non-polarising electrodes installed at the ground surface. We developed a geostatistical method for the estimation of the hydraulic conductivity field from measurements of hydraulic heads and SP during pumping and injection experiments. We use the adjoint state method and a recent petrophysical formulation of the streaming potential problem in which the streaming coupling coefficient is derived from the hydraulic conductivity allowed reducing of the unknown parameters. The geostatistical inverse framework is applied to three synthetic case studies with different number of the wells and electrodes used to measure the hydraulic heads and the streaming potentials. To evaluate the benefits of the incorporating of the streaming potential to the hydraulic data, we compared the cases in which the data are coupled or not to map the hydraulic conductivity. The results of the inversion revealed that a dense distribution of electrodes can be used to infer the heterogeneities in the hydraulic conductivity field. Incorporating the streaming potential information to the hydraulic head data improves the estimate of hydraulic conductivity field especially when the number of piezometers is limited.

Hydraulic Shearing and Hydraulic Jacking Observed during Hydraulic Stimulations in Fractured Geothermal Reservoir in Pohang, Korea

Science.gov (United States)

Min, K. B.; Park, S.; Xie, L.; Kim, K. I.; Yoo, H.; Kim, K. Y.; Choi, J.; Yoon, K. S.; Yoon, W. S.; Lee, T. J.; Song, Y.

2017-12-01

Enhanced Geothermal System (EGS) relies on sufficient and irreversible enhancement of reservoir permeability through hydraulic stimulation and possibility of such desirable change of permeability is an open question that can undermine the universality of EGS concept. We report results of first hydraulic stimulation campaign conducted in two deep boreholes in fractured granodiorite geothermal reservoir in Pohang, Korea. Borehole PX-1, located at 4.22 km, was subjected to the injection of 3,907 m3 with flow rate of up to 18 kg/s followed by bleeding off of 1,207 m3. The borehole PX-2, located at 4.35 km, was subjected to the injection of 1,970 m3 with flow rate of up to 46 kg/sIn PX-1, a sharp distinct decline of wellhead pressure was observed at around 16 MPa of wellhead pressure which was similar to the predicted injection pressure to induce hydraulic shearing. Injectivity interpretation before and after the hydraulic shearing indicates that permanent increase of permeability was achieved by a factor of a few. In PX-2, however, injectivity was very small and hydraulic shearing was not observed due possibly to the near wellbore damage made by the remedying process of lost circulation such as using lost circulation material during drilling. Flow rate of larger than 40 kg/s was achieved at very high well head pressure of nearly 90 MPa. Hydraulic jacking, that is reversible opening and closure of fracture with change of injection pressure, was clearly observed. Although sharp increase of permeability due to fracture opening was achieved with elevated injection pressure, the increased permeability was reversed with decreased injection pressure.Two contrasting response observed in the same reservoir at two different boreholes which is apart only 600 m apart provide important implication that can be used for the stimulation strategy for EGS.This work was supported by the New and Renewable Energy Technology Development Program of the Korea Institute of Energy Technology

Improvement of computer programs 'BAMBOO' and 'ASFRE-IV' for coupling analysis of deformation and thermal-hydraulics in a high burn-up fuel subassembly of fast reactor

International Nuclear Information System (INIS)

Uwaba, Tomoyuki; Ohshima, Hiroyuki; Imai, Yasutomo

2003-04-01

A simulation system of a deformed fuel subassembly is being developed for the structure integrity of high burn-up wire-spacer-type fuel subassemblies of sodium-cooled fast breeder reactors. This report describes a computer program improvement work for coupling analyses of deformation and thermal-hydraulics in a fuel subassembly as part of the simulation system development. In this work, a function of data conversion as an interface between a bundle deformation analysis program BAMBOO and a thermal hydraulic analysis program ASFRE-IV was incorporated to each program. BAMBOO was improved to accept the coolant temperature data from ASFRE-IV and to offer bundle deformation data to ASFRE-IV. ASFRE-IV was also improved to offer the coolant temperature data to BAMBOO and to obtain the bundle deformation data from BAMBOO. Improved BAMBOO and ASFRE-IV were applied to an analysis of 169-pin bundle for the program verification. It was confirmed that the coupling analysis gave the physically reasonable results on both deformation and thermal hydraulic behaviors in the fuel subassembly. (author)

Network hydraulics inclusion in water quality event detection using multiple sensor stations data.

Science.gov (United States)

Oliker, Nurit; Ostfeld, Avi

2015-09-01

Event detection is one of the current most challenging topics in water distribution systems analysis: how regular on-line hydraulic (e.g., pressure, flow) and water quality (e.g., pH, residual chlorine, turbidity) measurements at different network locations can be efficiently utilized to detect water quality contamination events. This study describes an integrated event detection model which combines multiple sensor stations data with network hydraulics. To date event detection modelling is likely limited to single sensor station location and dataset. Single sensor station models are detached from network hydraulics insights and as a result might be significantly exposed to false positive alarms. This work is aimed at decreasing this limitation through integrating local and spatial hydraulic data understanding into an event detection model. The spatial analysis complements the local event detection effort through discovering events with lower signatures by exploring the sensors mutual hydraulic influences. The unique contribution of this study is in incorporating hydraulic simulation information into the overall event detection process of spatially distributed sensors. The methodology is demonstrated on two example applications using base runs and sensitivity analyses. Results show a clear advantage of the suggested model over single-sensor event detection schemes. Copyright © 2015 Elsevier Ltd. All rights reserved.

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.

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.

Hydraulic Analysis of the Contribution of Emergency Water to C. N. Almaraz Systems Affected as a Result of the Complementary Technical Instructions issued by the CSN after Fukushima

International Nuclear Information System (INIS)

Vilar Carmona, G.; Puertas Munoz, S.; Arguello Tara, A.; Sanz Roman, F. J.

2013-01-01

This paper presents the study and hydraulic analysis of the capacity required contribution of emergency water to the Almaraz NPP to power systems deal with the accidental events outside the bases of design defined in the Complementary technical instructions generated by the CSN after Fukushima. Through the program of balanced hydraulic SBAL, developed by entrepreneurs Grouped (EE.AA) and used in multiple security systems analysis, and based on designs and requirements to be fulfilled by the Almaraz NPP of the different strategies are set, have developed a series of hydraulic models that they have allowed the definition and dimensioning of the portable media and the new connections required in the central systems.

Hydraulic Stability of Accropode Armour

DEFF Research Database (Denmark)

Jensen, T.; Burcharth, H. F.; Frigaard, Peter

The present report describes the hydraulic model tests of Accropode armour layers carried out at the Hydraulics Laboratory at Aalborg University from November 1995 through March 1996. The objective of the model tests was to investigate the hydraulic stability of Accropode armour layers...... with permeable core (crushed granite with a gradation of 5-8 mm). The outcome of this study is described in "Hydraulic Stability of Single-Layer Dolos and Accropode Armour Layers" by Christensen & Burcharth (1995). In January/February 1996, Research Assistant Thomas Jensen carried out a similar study...

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

Assimilation of temperature and hydraulic gradients for quantifying the spatial variability of streambed hydraulics

Science.gov (United States)

Huang, Xiang; Andrews, Charles B.; Liu, Jie; Yao, Yingying; Liu, Chuankun; Tyler, Scott W.; Selker, John S.; Zheng, Chunmiao

2016-08-01

Understanding the spatial and temporal characteristics of water flux into or out of shallow aquifers is imperative for water resources management and eco-environmental conservation. In this study, the spatial variability in the vertical specific fluxes and hydraulic conductivities in a streambed were evaluated by integrating distributed temperature sensing (DTS) data and vertical hydraulic gradients into an ensemble Kalman filter (EnKF) and smoother (EnKS) and an empirical thermal-mixing model. The formulation of the EnKF/EnKS assimilation scheme is based on a discretized 1D advection-conduction equation of heat transfer in the streambed. We first systematically tested a synthetic case and performed quantitative and statistical analyses to evaluate the performance of the assimilation schemes. Then a real-world case was evaluated to calculate assimilated specific flux. An initial estimate of the spatial distributions of the vertical hydraulic gradients was obtained from an empirical thermal-mixing model under steady-state conditions using a constant vertical hydraulic conductivity. Then, this initial estimate was updated by repeatedly dividing the assimilated specific flux by estimates of the vertical hydraulic gradients to obtain a refined spatial distribution of vertical hydraulic gradients and vertical hydraulic conductivities. Our results indicate that optimal parameters can be derived with fewer iterations but greater simulation effort using the EnKS compared with the EnKF. For the field application in a stream segment of the Heihe River Basin in northwest China, the average vertical hydraulic conductivities in the streambed varied over three orders of magnitude (5 × 10-1 to 5 × 102 m/d). The specific fluxes ranged from near zero (qz < ±0.05 m/d) to ±1.0 m/d, while the vertical hydraulic gradients were within the range of -0.2 to 0.15 m/m. The highest and most variable fluxes occurred adjacent to a debris-dam and bridge pier. This phenomenon is very likely

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)

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.

Analysis of Grain Size Distribution and Hydraulic Conductivity for a Variety of Sediment Types with Application to Wadi Sediments

KAUST Repository

Rosas Aguilar, Jorge

2013-05-01

Grain size distribution, porosity, and hydraulic conductivity from over 400 unlithified sediment samples were analized. The measured hydraulic conductivity values were then compared to values calculated using 20 different empirical equations commonly used to estimate hydraulic conductivity from grain size analyses. It was found that most of the hydraulic conductivity values estimated from the empirical equations correlated very poorly to the measured hydraulic conductivity values. Modifications of the empirical equations, including changes to special coefficients and statistical off sets, were made to produce modified equations that considerably improve the hydraulic conductivity estimates from grain size data for beach, dune, off shore marine, and wadi sediments. Expected hydraulic conductivity estimation errors were reduced. Correction factors were proposed for wadi sediments, taking mud percentage and the standard deviation (in phi units) into account.

Hydraulic hoisting and backfilling

Science.gov (United States)

Sauermann, H. B.

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

A field assessment of the value of steady shape hydraulic tomography for characterization of aquifer heterogeneities

Science.gov (United States)

Bohling, Geoffrey C.; Butler, James J.; Zhan, Xiaoyong; Knoll, Michael D.

2007-01-01

Hydraulic tomography is a promising approach for obtaining information on variations in hydraulic conductivity on the scale of relevance for contaminant transport investigations. This approach involves performing a series of pumping tests in a format similar to tomography. We present a field‐scale assessment of hydraulic tomography in a porous aquifer, with an emphasis on the steady shape analysis methodology. The hydraulic conductivity (K) estimates from steady shape and transient analyses of the tomographic data compare well with those from a tracer test and direct‐push permeameter tests, providing a field validation of the method. Zonations based on equal‐thickness layers and cross‐hole radar surveys are used to regularize the inverse problem. The results indicate that the radar surveys provide some useful information regarding the geometry of the K field. The steady shape analysis provides results similar to the transient analysis at a fraction of the computational burden. This study clearly demonstrates the advantages of hydraulic tomography over conventional pumping tests, which provide only large‐scale averages, and small‐scale hydraulic tests (e.g., slug tests), which cannot assess strata connectivity and may fail to sample the most important pathways or barriers to flow.

CFD studies on thermal hydraulics of spallation targets

International Nuclear Information System (INIS)

Tak, N.I.; Batta, A.; Cheng, X.

2005-01-01

Full text of publication follows: Due to the fast advances in computer hardware as well as software in recent years, more and more interests have been aroused to use computational fluid dynamics (CFD) technology in nuclear engineering and designs. During recent many years, Forschungszentrum Karlsruhe (FZK) has been actively involved in the thermal hydraulic analysis and design of spallation targets. To understand the thermal hydraulic behaviors of spallation targets very detailed simulations are necessary because of their complex geometries, complicated boundary conditions such as spallation heat distributions, and very strict design limits. A CFD simulation is believed to be the best for this purpose even though the validation of CFD codes are not perfectly completed yet in specific topics like liquid metal heat transfer. The research activities on three spallation targets (i.e., MEGAPIE, TRADE, and XADS targets) are currently very active in Europe in order to consolidate the European ADS road-map. In the thermal hydraulics point of view, two kinds of the research activities, i.e., (1) numerical design and (2) experimental work, are required to achieve the objectives of these targets. It should be noted that CFD studies play important role on both kinds of two activities. A preliminary design of a target can be achieved by sophisticated CFD analysis and pre-and-post analyses of an experimental work using a CFD code help the design of the test section of the experiment as well as the analysis of the experimental results. The present paper gives an overview about the recent CFD studies relating to thermal hydraulics of the spallation targets recently involved in FZK. It covers numerical design studies as well as CFD studies to support experimental works. The CFX code has been adopted for the studies. Main recent results for the selected examples performed by FZK are presented and discussed with their specific lessons learned. (authors)

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)

Proceedings of transient thermal-hydraulics and coupled vessel and piping system responses 1991

International Nuclear Information System (INIS)

Wang, G.Y.; Shin, Y.W.; Moody, F.J.

1991-01-01

This book reports on transient thermal-hydraulics and coupled vessel and piping system responses. Topics covered include: nuclear power plant containment designs; analysis of control rods; gate closure of hydraulic turbines; and shock wave solutions for steam water mixtures in piping systems

Evaluation of Fish Passage at Whitewater Parks Using 2D and 3D Hydraulic Modeling

Science.gov (United States)

Hardee, T.; Nelson, P. A.; Kondratieff, M.; Bledsoe, B. P.

2016-12-01

In-stream whitewater parks (WWPs) are increasingly popular recreational amenities that typically create waves by constricting flow through a chute to increase velocities and form a hydraulic jump. However, the hydraulic conditions these structures create can limit longitudinal habitat connectivity and potentially inhibit upstream fish migration, especially of native fishes. An improved understanding of the fundamental hydraulic processes and potential environmental effects of whitewater parks is needed to inform management decisions about Recreational In-Channel Diversions (RICDs). Here, we use hydraulic models to compute a continuous and spatially explicit description of velocity and depth along potential fish swimming paths in the flow field, and the ensemble of potential paths are compared to fish swimming performance data to predict fish passage via logistic regression analysis. While 3d models have been shown to accurately predict trout movement through WWP structures, 2d methods can provide a more cost-effective and manager-friendly approach to assessing the effects of similar hydraulic structures on fish passage when 3d analysis in not feasible. Here, we use 2d models to examine the hydraulics in several WWP structures on the North Fork of the St. Vrain River at Lyons, Colorado, and we compare these model results to fish passage predictions from a 3d model. Our analysis establishes a foundation for a practical, transferable and physically-rigorous 2d modeling approach for mechanistically evaluating the effects of hydraulic structures on fish passage.

A global data set of soil hydraulic properties and sub-grid variability of soil water retention and hydraulic conductivity curves

Science.gov (United States)

Montzka, Carsten; Herbst, Michael; Weihermüller, Lutz; Verhoef, Anne; Vereecken, Harry

2017-07-01

Agroecosystem models, regional and global climate models, and numerical weather prediction models require adequate parameterization of soil hydraulic properties. These properties are fundamental for describing and predicting water and energy exchange processes at the transition zone between solid earth and atmosphere, and regulate evapotranspiration, infiltration and runoff generation. Hydraulic parameters describing the soil water retention (WRC) and hydraulic conductivity (HCC) curves are typically derived from soil texture via pedotransfer functions (PTFs). Resampling of those parameters for specific model grids is typically performed by different aggregation approaches such a spatial averaging and the use of dominant textural properties or soil classes. These aggregation approaches introduce uncertainty, bias and parameter inconsistencies throughout spatial scales due to nonlinear relationships between hydraulic parameters and soil texture. Therefore, we present a method to scale hydraulic parameters to individual model grids and provide a global data set that overcomes the mentioned problems. The approach is based on Miller-Miller scaling in the relaxed form by Warrick, that fits the parameters of the WRC through all sub-grid WRCs to provide an effective parameterization for the grid cell at model resolution; at the same time it preserves the information of sub-grid variability of the water retention curve by deriving local scaling parameters. Based on the Mualem-van Genuchten approach we also derive the unsaturated hydraulic conductivity from the water retention functions, thereby assuming that the local parameters are also valid for this function. In addition, via the Warrick scaling parameter λ, information on global sub-grid scaling variance is given that enables modellers to improve dynamical downscaling of (regional) climate models or to perturb hydraulic parameters for model ensemble output generation. The present analysis is based on the ROSETTA PTF

Social costs from proximity to hydraulic fracturing in New York State

International Nuclear Information System (INIS)

Popkin, Jennifer H.; Duke, Joshua M.; Borchers, Allison M.; Ilvento, Thomas

2013-01-01

The study reports data from an economic choice experiment to determine the likely welfare impacts of hydraulic fracturing, in this case using natural gas extracted by hydraulic fracturing for household electricity. Data were collected from an Internet survey of 515 residents of New York State. The welfare analysis indicated that on average households incur a welfare loss from in-state hydraulic fracturing as the source of their electricity. The evidence suggests that households in shale counties bear more costs from HF electricity than households out of shale counties. The average welfare loss is substantive, estimated at 40–46% of average household electric bills in shale counties and 16–20% of bills in counties without shale. The evidence also suggests that relative proximity to HF well sites also increases cost borne by households. -- Highlights: •New York households were surveyed to determine impacts of hydraulic fracturing. •Households on average lose welfare if hydraulic fracturing gas provides their electricity. •The average welfare loss is estimated to be 16–46% of respondents’ electricity bill. •The welfare impacts were heterogeneous, with some predicted to have welfare gain. •Proximity to hydraulic fracturing wells decreases welfare, on average

CRITICALITY CURVES FOR PLUTONIUM HYDRAULIC FLUID MIXTURES

International Nuclear Information System (INIS)

WITTEKIND WD

2007-01-01

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

VIPRE-01: a thermal-hydraulic analysis code for reactor cores. Volume 3. Programmer's manual. Final report

International Nuclear Information System (INIS)

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

1983-05-01

VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear-reactor-core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This is Volume 3, the Programmer's Manual. It explains the codes' structures and the computer interfaces

Co-occurring woody species have diverse hydraulic strategies and mortality rates during an extreme drought: Belowground hydraulic failure during drought

Energy Technology Data Exchange (ETDEWEB)

Johnson, Daniel M. [College of Natural Resources, University of Idaho, Moscow ID 83844 USA; Domec, Jean-Christophe [Bordeaux Sciences Agro, UMR INRA-ISPA 1391, Gradignan 33195 France; Nicholas School of the Environment, Duke University, Durham NC 27708 USA; Carter Berry, Z. [College of Natural Resources, University of Idaho, Moscow ID 83844 USA; Department of Natural Resources and the Environment, University of New Hampshire, Durham NH 03824 USA; Schwantes, Amanda M. [Nicholas School of the Environment, Duke University, Durham NC 27708 USA; McCulloh, Katherine A. [Department of Botany, University of Wisconsin-Madison, Madison WI 53705 USA; Woodruff, David R. [US Forest Service, Pacific Northwest Research Station, Corvallis OR 97331 USA; Wayne Polley, H. [Grassland, Soil & Water Research Laboratory USDA-Agricultural Research Service, Temple TX 76502 USA; Wortemann, Remí [INRA Nancy, UMR INRA-UL 1137 Ecologie et Ecophysiologie Forestières, Champenoux 54280 France; Swenson, Jennifer J. [Nicholas School of the Environment, Duke University, Durham NC 27708 USA; Scott Mackay, D. [Department of Geography, State University of New York, Buffalo NY 14261 USA; McDowell, Nate G. [Pacific Northwest National Laboratory, Richland WA 99352 USA; Jackson, Robert B. [Department of Earth System Science, Woods Institute for the Environment, and Precourt Institute for Energy, Stanford University, Stanford CA 94305 USA

2018-01-29

From 2011 to 2013, Texas experienced its worst drought in recorded history. This event provided a unique natural experiment to assess species-specific responses to extreme drought and mortality of four co-occurring woody species: Quercus fusiformis, Diospyros texana, Prosopis glandulosa and Juniperus ashei. We examined hypothesized mechanisms that could promote these species’ diverse mortality patterns using post-drought measurements on surviving trees coupled to retrospective process modeling. The species exhibited a wide range of gas exchange responses, hydraulic strategies, and mortality rates. Multiple proposed indices of mortality mechanisms were not consistent with the observed mortality patterns across species, including measures of iso/anisohydry, photosynthesis, carbohydrate depletion, and hydraulic safety margins. Large losses of growing season whole-tree conductance (driven by belowground losses of conductance), and shallower rooting depths, were associated with species that exhibited greater mortality. Based on this retrospective analysis, we suggest that species more vulnerable to drought were more likely to have succumbed to hydraulic failure belowground.

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

Design and Optimization of Fast Switching Valves for Large Scale Digital Hydraulic Motors

DEFF Research Database (Denmark)

Roemer, Daniel Beck

The present thesis is on the design, analysis and optimization of fast switching valves for digital hydraulic motors with high power ratings. The need for such high power motors origins in the potential use of hydrostatic transmissions in wind turbine drive trains, as digital hydraulic machines...... have been shown to improve the overall efficiency and efficient operation range compared to traditional hydraulic machines. Digital hydraulic motors uses electronically controlled independent seat valves connected to the pressure chambers, which must be fast acting and exhibit low pressure losses...... to enable efficient operation. These valves are complex components to design, as multiple design aspects are present in these integrated valve units, with conflicting objectives and interdependencies. A preliminary study on a small scale single-cylinder digital hydraulic pump has initially been conducted...

Evaluation of thermal-hydraulic parameter uncertainties in a TRIGA research reactor

International Nuclear Information System (INIS)

Mesquita, Amir Z.; Costa, Antonio C.L.; Ladeira, Luiz C.D.; Rezende, Hugo C.; Palma, Daniel A.P.

2015-01-01

Experimental studies had been performed in the TRIGA Research Nuclear Reactor of CDTN/CNEN to find out the its thermal hydraulic parameters. Fuel to coolant heat transfer patterns must be evaluated as function of the reactor power in order to assess the thermal hydraulic performance of the core. The heat generated by nuclear fission in the reactor core is transferred from fuel elements to the cooling system through the fuel-cladding (gap) and the cladding to coolant interfaces. As the reactor core power increases the heat transfer regime from the fuel cladding to the coolant changes from single-phase natural convection to subcooled nucleate boiling. This paper presents the uncertainty analysis in the results of the thermal hydraulics experiments performed. The methodology used to evaluate the propagation of uncertainty in the results was done based on the pioneering article of Kline and McClintock, with the propagation of uncertainties based on the specification of uncertainties in various primary measurements. The uncertainty analysis on thermal hydraulics parameters of the CDTN TRIGA fuel element is determined, basically, by the uncertainty of the reactor's thermal power. (author)

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)

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)

Hydraulic Yaw System for Wind Turbines with New Compact Hydraulic Motor Principle

DEFF Research Database (Denmark)

Sørensen, Rasmus Mørk; Hansen, Michael Rygaard; Mouritsen, Ole Ø.

2011-01-01

This paper presents a new hydraulic yaw system for wind turbines. The basic component is a new type of hydraulic motor characterized by an extraordinary high specific displacement yielding high output torque in a compact form. The focus in the paper is the volumetric efficiency of the motor, which...

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

Directory of Open Access Journals (Sweden)

Yongliang Wang

2018-01-01

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

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

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

Residual stresses associated with the hydraulic expansion of steam generator tubing into tubesheets

International Nuclear Information System (INIS)

Middlebrooks, W.B.; Harrod, D.L.; Gold, R.E.

1991-01-01

Westinghouse has used three different processes for the full depth expansion of tubes into the tube sheets of recirculating nuclear steam generators: mechanical rolling, explosive expansion and hydraulic expansion. Each process aims at expanding tubes tightly to tube sheets, leaving the smallest possible secondary side crevice depth, and minimizing the residual stress in the expanded tubes, all for the purpose of mitigating the effect of corrosion phenomena. The hydraulic expansion process was qualified and has been implemented since 1978, and more than 1.1 million tube ends have been hydraulically expanded into production units. In this paper, the results of the recent analytical studies related to the residual stress in the expanded tubes are summarized. The method of hydraulic expansion is explained, and some important parameters are given. Finite element method, theoretical incremental analysis, tube sheet yielding and residual stress, contact pressure, sensitivity analysis and temperature effect in the central region of tube sheets, and the residual stress in the transition zone are described. (K.I.)

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.

Estimation of ground water hydraulic parameters

Energy Technology Data Exchange (ETDEWEB)

Hvilshoej, Soeren

1998-11-01

The main objective was to assess field methods to determine ground water hydraulic parameters and to develop and apply new analysis methods to selected field techniques. A field site in Vejen, Denmark, which previously has been intensively investigated on the basis of a large amount of mini slug tests and tracer tests, was chosen for experimental application and evaluation. Particular interest was in analysing partially penetrating pumping tests and a recently proposed single-well dipole test. Three wells were constructed in which partially penetrating pumping tests and multi-level single-well dipole tests were performed. In addition, multi-level slug tests, flow meter tests, gamma-logs, and geologic characterisation of soil samples were carried out. In addition to the three Vejen analyses, data from previously published partially penetrating pumping tests were analysed assuming homogeneous anisotropic aquifer conditions. In the present study methods were developed to analyse partially penetrating pumping tests and multi-level single-well dipole tests based on an inverse numerical model. The obtained horizontal hydraulic conductivities from the partially penetrating pumping tests were in accordance with measurements obtained from multi-level slug tests and mini slug tests. Accordance was also achieved between the anisotropy ratios determined from partially penetrating pumping tests and multi-level single-well dipole tests. It was demonstrated that the partially penetrating pumping test analysed by and inverse numerical model is a very valuable technique that may provide hydraulic information on the storage terms and the vertical distribution of the horizontal and vertical hydraulic conductivity under both confined and unconfined aquifer conditions. (EG) 138 refs.

Hydraulic turbines and auxiliary equipment

Energy Technology Data Exchange (ETDEWEB)

Luo Gaorong [Organization of the United Nations, Beijing (China). International Centre of Small Hydroelectric Power Plants

1995-07-01

This document presents a general overview on hydraulic turbines and auxiliary equipment, emphasizing the turbine classification, in accordance with the different types of turbines, standard turbine series in China, turbine selection based on the basic data required for the preliminary design, general hill model curves, chart of turbine series and the arrangement of application for hydraulic turbines, hydraulic turbine testing, and speed regulating device.

Subsea Hydraulic Leakage Detection and Diagnosis

OpenAIRE

Stavenes, Thomas

2010-01-01

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

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)

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)

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.

Hydraulic fracturing of rock-fill dam

Directory of Open Access Journals (Sweden)

Jun-Jie WANG

2016-02-01

Full Text Available The condition in which hydraulic fracturing in core of earth-rock fill dam maybe induced, the mechanism by which the reason of hydraulic fracturing canbe explained, and the failure criterion by which the occurrence of hydraulicfracturing can be determined, were investigated. The condition dependson material properties such as, cracks in the core and low permeability ofcore soil, and “water wedging” action in cracks. An unsaturated core soiland fast impounding are the prerequisites for the formation of “waterwedging” action. The mechanism of hydraulic fracturing can be explainedby fracture mechanics. The crack propagation induced by water pressuremay follow any of mode I, mode II and mixed mode I-II. Based on testingresults of a core soil, a new criterion for hydraulic fracturing was suggested,from which mechanisms of hydraulic fracturing in the core of rock-fill damwere discussed. The results indicated that factors such as angle betweencrack surface and direction of principal stress, local stress state at thecrack, and fracture toughness KIC of core soil may largely affect theinduction of hydraulic fracturing and the mode of the propagation of thecrack.The condition in which hydraulic fracturing in core of earth-rock fill dam maybe induced, the mechanism by which the reason of hydraulic fracturing canbe explained, and the failure criterion by which the occurrence of hydraulicfracturing can be determined, were investigated. The condition dependson material properties such as, cracks in the core and low permeability ofcore soil, and “water wedging” action in cracks. An unsaturated core soiland fast impounding are the prerequisites for the formation of “waterwedging” action. The mechanism of hydraulic fracturing can be explainedby fracture mechanics. The crack propagation induced by water pressuremay follow any of mode I, mode II and mixed mode I-II. Based on testingresults of a core soil, a new criterion for hydraulic fracturing

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)

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

Science.gov (United States)

Liu, L J; Schlesinger, M

2015-09-07

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

Insight into the hydraulics and resilience of Ponderosa pine seedlings using a mechanistic ecohydrologic model

Science.gov (United States)

Maneta, M. P.; Simeone, C.; Dobrowski, S.; Holden, Z.; Sapes, G.; Sala, A.; Begueria, S.

2017-12-01

In semiarid regions, drought-induced seedling mortality is considered to be caused by failure in the tree hydraulic column. Understanding the mechanisms that cause hydraulic failure and death in seedlings is important, among other things, to diagnose where some tree species may fail to regenerate, triggering demographic imbalances in the forest that could result in climate-driven shifts of tree species. Ponderosa pine is a common lower tree line species in the western US. Seedlings of ponderosa pine are often subject to low soil water potentials, which require lower water potentials in the xylem and leaves to maintain the negative pressure gradient that drives water upward. The resilience of the hydraulic column to hydraulic tension is species dependent, but from greenhouse experiments, we have identified general tension thresholds beyond which loss of xylem conductivity becomes critical, and mortality in Ponderosa pine seedlings start to occur. We describe this hydraulic behavior of plants using a mechanistic soil-vegetation-atmosphere transfer model. Before we use this models to understand water-stress induced seedling mortality at the landscape scale, we perform a modeling analysis of the dynamics of soil moisture, transpiration, leaf water potential and loss of plant water conductivity using detailed data from our green house experiments. The analysis is done using a spatially distributed model that simulates water fluxes, energy exchanges and water potentials in the soil-vegetation-atmosphere continuum. Plant hydraulic and physiological parameters of this model were calibrated using Monte Carlo methods against information on soil moisture, soil hydraulic potential, transpiration, leaf water potential and percent loss of conductivity in the xylem. This analysis permits us to construct a full portrait of the parameter space for Ponderosa pine seedling and generate posterior predictive distributions of tree response to understand the sensitivity of transpiration

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".

Combined hydraulic and regenerative braking system

Science.gov (United States)

Venkataperumal, R.R.; Mericle, G.E.

1979-08-09

A combined hydraulic and regenerative braking system and method for an electric vehicle is disclosed. The braking system is responsive to the applied hydraulic pressure in a brake line to control the braking of the vehicle to be completely hydraulic up to a first level of brake line pressure, to be partially hydraulic at a constant braking force and partially regenerative at a linearly increasing braking force from the first level of applied brake line pressure to a higher second level of brake line pressure, to be partially hydraulic at a linearly increasing braking force and partially regenerative at a linearly decreasing braking force from the second level of applied line pressure to a third and higher level of applied line pressure, and to be completely hydraulic at a linearly increasing braking force from the third level to all higher applied levels of line pressure.

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)

Unsaturated hydraulic behaviour of a permeable pavement: Laboratory investigation and numerical analysis by using the HYDRUS-2D model

Science.gov (United States)

Turco, Michele; Kodešová, Radka; Brunetti, Giuseppe; Nikodem, Antonín; Fér, Miroslav; Piro, Patrizia

2017-11-01

An adequate hydrological description of water flow in permeable pavement systems relies heavily on the knowledge of the unsaturated hydraulic properties of the construction materials. Although several modeling tools and many laboratory methods already exist in the literature to determine the hydraulic properties of soils, the importance of an accurate materials hydraulic description of the permeable pavement system, is increasingly recognized in the fields of urban hydrology. Thus, the aim of this study is to propose techniques/procedures on how to interpret water flow through the construction system using the HYDRUS model. The overall analysis includes experimental and mathematical procedures for model calibration and validation to assess the suitability of the HYDRUS-2D model to interpret the hydraulic behaviour of a lab-scale permeable pavement system. The system consists of three porous materials: a wear layer of porous concrete blocks, a bedding layers of fine gravel, and a sub-base layer of coarse gravel. The water regime in this system, i.e. outflow at the bottom and water contents in the middle of the bedding layer, was monitored during ten irrigation events of various durations and intensities. The hydraulic properties of porous concrete blocks and fine gravel described by the van Genuchten functions were measured using the clay tank and the multistep outflow experiments, respectively. Coarse gravel properties were set at literature values. In addition, some of the parameters (Ks of the concrete blocks layer, and α, n and Ks of the bedding layer) were optimized with the HYDRUS-2D model from water fluxes and soil water contents measured during irrigation events. The measured and modeled hydrographs were compared using the Nash-Sutcliffe efficiency (NSE) index (varied between 0.95 and 0.99) while the coefficient of determination R2 was used to assess the measured water content versus the modelled water content in the bedding layer (R2 = 0.81 ÷ 0.87) . The

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.

Determination of hydraulic properties of unsaturated soil via inverse modeling

International Nuclear Information System (INIS)

Kodesova, R.

2004-01-01

The method for determining the hydraulic properties of unsaturated soil with inverse modeling is presented. A modified cone penetrometer has been designed to inject water into the soil through a screen, and measure the progress of the wetting front with two tensiometer rings positioned above the screen. Cumulative inflow and pressure head readings are analyzed to obtain estimates of the hydraulic parameters describing K(h) and θ(h). Optimization results for tests at one side are used to demonstrate the possibility to evaluate either the wetting branches of the soil hydraulic properties, or the wetting and drying curves simultaneously, via analysis of different parts of the experiment. The optimization results are compared to the results of standard laboratory and field methods. (author)

Charging valve of the full hydraulic braking system

Directory of Open Access Journals (Sweden)

Jinshi Chen

2016-03-01

Full Text Available It is known that the full hydraulic braking system has excellent braking performance. As the key component of the full hydraulic braking system, the parameters of the accumulator charging valve have a significant effect on the braking performance. In this article, the key parameters of the charging valve are analyzed through the static theoretical and an Advanced Modeling Environment for performing Simulation of engineering systems (AMESim simulation model of the dual-circuit accumulator charging valve is established based on the real structure parameters first. Second, according to the results of the dynamic simulation, the dynamic characteristics of the charging pressure, the flow rate, and the frequency of the charging valve are studied. The key parameters affecting the serial production are proposed and some technical advices for improving the performance of the full hydraulic system are provided. Finally, the theoretical analysis is validated by the simulation results. The comparison between the simulation results and the experimental results indicates that the simulated AMESim model of the charging valve is accurate and credible with the error rate inside 0.5% compared with the experimental result. Hence, the performance of the charging valve meets the request of the full hydraulic braking system exactly.

A flexible coupling scheme for Monte Carlo and thermal-hydraulics codes

Energy Technology Data Exchange (ETDEWEB)

Hoogenboom, J. Eduard, E-mail: J.E.Hoogenboom@tudelft.nl [Delft University of Technology (Netherlands); Ivanov, Aleksandar; Sanchez, Victor, E-mail: Aleksandar.Ivanov@kit.edu, E-mail: Victor.Sanchez@kit.edu [Karlsruhe Institute of Technology, Institute of Neutron Physics and Reactor Technology, Eggenstein-Leopoldshafen (Germany); Diop, Cheikh, E-mail: Cheikh.Diop@cea.fr [CEA/DEN/DANS/DM2S/SERMA, Commissariat a l' Energie Atomique, Gif-sur-Yvette (France)

2011-07-01

A coupling scheme between a Monte Carlo code and a thermal-hydraulics code is being developed within the European NURISP project for comprehensive and validated reactor analysis. The scheme is flexible as it allows different Monte Carlo codes and different thermal-hydraulics codes to be used. At present the MCNP and TRIPOLI4 Monte Carlo codes can be used and the FLICA4 and SubChanFlow thermal-hydraulics codes. For all these codes only an original executable is necessary. A Python script drives the iterations between Monte Carlo and thermal-hydraulics calculations. It also calls a conversion program to merge a master input file for the Monte Carlo code with the appropriate temperature and coolant density data from the thermal-hydraulics calculation. Likewise it calls another conversion program to merge a master input file for the thermal-hydraulics code with the power distribution data from the Monte Carlo calculation. Special attention is given to the neutron cross section data for the various required temperatures in the Monte Carlo calculation. Results are shown for an infinite lattice of PWR fuel pin cells and a 3 x 3 fuel BWR pin cell cluster. Various possibilities for further improvement and optimization of the coupling system are discussed. (author)

A flexible coupling scheme for Monte Carlo and thermal-hydraulics codes

International Nuclear Information System (INIS)

Hoogenboom, J. Eduard; Ivanov, Aleksandar; Sanchez, Victor; Diop, Cheikh

2011-01-01

A coupling scheme between a Monte Carlo code and a thermal-hydraulics code is being developed within the European NURISP project for comprehensive and validated reactor analysis. The scheme is flexible as it allows different Monte Carlo codes and different thermal-hydraulics codes to be used. At present the MCNP and TRIPOLI4 Monte Carlo codes can be used and the FLICA4 and SubChanFlow thermal-hydraulics codes. For all these codes only an original executable is necessary. A Python script drives the iterations between Monte Carlo and thermal-hydraulics calculations. It also calls a conversion program to merge a master input file for the Monte Carlo code with the appropriate temperature and coolant density data from the thermal-hydraulics calculation. Likewise it calls another conversion program to merge a master input file for the thermal-hydraulics code with the power distribution data from the Monte Carlo calculation. Special attention is given to the neutron cross section data for the various required temperatures in the Monte Carlo calculation. Results are shown for an infinite lattice of PWR fuel pin cells and a 3 x 3 fuel BWR pin cell cluster. Various possibilities for further improvement and optimization of the coupling system are discussed. (author)

Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

Energy Technology Data Exchange (ETDEWEB)

Baek, W. P.; Song, C. H.; Kim, Y. S. and others

2005-02-15

The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform various integral effect tests for design, operation, and safety regulation of pressurized water reactors. During the first phase of this project (1997.8{approx}2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished: a full-height, 1/300-volume-scaled full pressure facility for APR1400, an evolutionary pressurized water reactor that was developed by Korean industry. Main objectives of the present phase (2002.4{approx}2005.2), was to optimize the facility design and to construct the experimental facility. We have performed following researches: 1) Optimization of the basic design of the thermal-hydraulic integral effect test facility for PWRs - ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) - Reduced height design for APR1400 (+ specific design features of KSNP safety injection systems) - Thermal-hydraulic scaling based on three-level scaling methodology by Ishii et al. 2) Construction of the ATLAS facility - Detailed design of the test facility - Manufacturing and procurement of components - Installation of the facility 3) Development of supporting technology for integral effect tests - Development and application of advanced instrumentation technology - Preliminary analysis of test scenarios - Development of experimental procedures - Establishment and implementation of QA system/procedure.

A Computational Model of Hydraulic Volume Displacement Drive

Directory of Open Access Journals (Sweden)

V. N. Pil'gunov

2014-01-01

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

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.

Hydraulic lifter for an underwater drilling rig

Energy Technology Data Exchange (ETDEWEB)

Garan' ko, Yu L

1981-01-15

A hydraulic lifter is suggested for an underwater drilling rig. It includes a base, hydraulic cylinders for lifting the drilling pipes connected to the clamp holder and hydraulic distributor. In order to simplify the design of the device, the base is made with a hollow chamber connected to the rod cavities and through the hydraulic distributor to the cavities of the hydraulic cylinders for lifting the drilling pipes. The hydraulic distributor is connected to the hydrosphere through the supply valve with control in time or by remote control. The base is equipped with reverse valves whose outlets are on the support surface of the base.

Verifying the prevalence, properties, and congruent hydraulics of at-many-stations hydraulic geometry (AMHG) for rivers in the continental United States

Science.gov (United States)

Barber, Caitline A.; Gleason, Colin J.

2018-01-01

Hydraulic geometry (HG) has long enabled daily discharge estimates, flood risk monitoring, and water resource and habitat assessments, among other applications. At-many-stations HG (AMHG) is a newly discovered form of HG with an evolving understanding. AMHG holds that there are temporally and spatially invariant ('congruent') depth, width, velocity, and discharge values that are shared by all stations of a river. Furthermore, these river-wide congruent hydraulics have been shown to link at-a-station HG (AHG) in space, contrary to previous expectation of AHG as spatially unpredictable. To date, AMHG has only been thoroughly examined on six rivers, and its congruent hydraulics are not well understood. To address the limited understanding of AMHG, we calculated AMHG for 191 rivers in the United States using USGS field-measured data from over 1900 gauging stations. These rivers represent nearly all geologic and climatic settings found in the continental U.S. and allow for a robust assessment of AMHG across scales. Over 60% of rivers were found to have AMHG with strong explanatory power to predict AHG across space (defined as r2 > 0.6, 118/191 rivers). We also found that derived congruent hydraulics bear little relation to their observed time-varying counterparts, and the strength of AMHG did not correlate with any available observed or congruent hydraulic parameters. We also found that AMHG is expressed at all fluvial scales in this study. Some statistically significant spatial clusters of rivers with strong and weak AMHG were identified, but further research is needed to identify why these clusters exist. Thus, this first widespread empirical investigation of AMHG leads us to conclude that AMHG is indeed a widely prevalent natural fluvial phenomenon, and we have identified linkages between known fluvial parameters and AMHG. Our work should give confidence to future researchers seeking to perform the necessary detailed hydraulic analysis of AMHG.

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.

EFFECTIVE APPLICATIO N OF LIDAR DATA IN T WO - DIMENSIONAL HYDRAULIC MODELLING

Directory of Open Access Journals (Sweden)

Bakuła Krzysztof

2014-12-01

Full Text Available This paper presents aspects of ALS data usage in two - dimensional hydraulic modelling including generation of high - precision digital terrain models, t heir effective processing which is a compromise between the resolution and the accuracy of the processed data, as well as information about the roughness of the land cover providing information that could compete with information from topographic databases and orthophotomaps. Still evolving ALS technology makes it possible to collect the data with constantly increasing spatial resolution that guarantees correct representation of the terrain shape and height. It also provides a reliable description of the la nd cover. However, the size of generated files may cause roblems in their effective usage in the 2D hydraulic modeling where Saint - Venant’s equations are implemented. High - resolution elevation models make it impossible or prolong the duration of the calcu lations for large areas in complex algorithms defining a model of the water movement, which is directly related to the cost of the hydraulic analysis. As far as an effective usage of voluminous datasets is concerned, the data reduction is recommended. Suc h a process should reduce the size of the data files, maintain their accuracy and keep the appropriate structure to allow their further application in the hydraulic modelling. An application of only a few percent of unprocessed datasets, selected with the use of specified filtering algorithms and spatial analysis tools, can give the same result of the hydraulic modeling obtained in a significantly shorter time than the result of the comparable operation on unprocessed datasets. Such an approach, however, is not commonly used, which means the most reliable hydraulic models are applied only in small areas in the largest cities. Another application of ALS data is its potential usage in digital roughness model creation for 2D hydraulic models. There are many po ssibilities of roughness

46 CFR 112.50-3 - Hydraulic starting.

Science.gov (United States)

2010-10-01

... POWER SYSTEMS Emergency Diesel and Gas Turbine Engine Driven Generator Sets § 112.50-3 Hydraulic starting. A hydraulic starting system must meet the following: (a) The hydraulic starting system must be a... 46 Shipping 4 2010-10-01 2010-10-01 false Hydraulic starting. 112.50-3 Section 112.50-3 Shipping...

Hydraulic testing in crystalline rock

International Nuclear Information System (INIS)

Almen, K.E.; Andersson, J.E.; Carlsson, L.; Hansson, K.; Larsson, N.A.

1986-12-01

Swedish Geolocical Company (SGAB) conducted and carried out single-hole hydraulic testing in borehole Fi 6 in the Finnsjoen area of central Sweden. The purpose was to make a comprehensive evaluation of different methods applicable in crystalline rocks and to recommend methods for use in current and scheduled investigations in a range of low hydraulic conductivity rocks. A total of eight different methods of testing were compared using the same equipment. This equipment was thoroughly tested as regards the elasticity of the packers and change in volume of the test section. The use of a hydraulically operated down-hole valve enabled all the tests to be conducted. Twelve different 3-m long sections were tested. The hydraulic conductivity calculated ranged from about 5x10 -14 m/s to 1x10 -6 m/s. The methods used were water injection under constant head and then at a constant rate-of-flow, each of which was followed by a pressure fall-off period. Water loss, pressure pulse, slug and drill stem tests were also performed. Interpretation was carried out using standard transient evaluation methods for flow in porous media. The methods used showed themselves to be best suited to specific conductivity ranges. Among the less time-consuming methods, water loss, slug and drill stem tests usually gave somewhat higher hydraulic conductivity values but still comparable to those obtained using the more time-consuming tests. These latter tests, however, provided supplementary information on hydraulic and physical properties and flow conditions, together with hydraulic conductivity values representing a larger volume of rock. (orig./HP)

LOOP-3, Hydraulic Stability in Heated Parallel Channels

Energy Technology Data Exchange (ETDEWEB)

Davies, A L [AEEW, Dorset (United Kingdom)

1968-02-01

1 - Nature of physical problem solved: Hydraulic stability in parallel channels. 2 - Method of solution: Calculation of transfer functions developed in reference (10 below). 3 - Restrictions on the complexity of the problem: Only due to assumptions in analysis (see ref.)

Thermal hydraulic issues and challenges for current and new generation FBRs

Energy Technology Data Exchange (ETDEWEB)

Chellapandi, P.; Velusamy, K., E-mail: kvelu@igcar.gov.in

2015-12-01

Highlights: • We present challenges in thermal hydraulic design of sodium cooled fast reactors. • We present roadmap of Indian fast reactor program and innovative design concepts. • Analysis methodology for thermal striping and thermal stratification are highlighted. • Design solutions for gas entrainment are presented. • Experimental approaches for normal and post accident decay heat removal are highlighted. - Abstract: Pool type sodium cooled fast reactors pose several design challenges and among them, certain thermal hydraulics and structural mechanics issues are special. High frequency temperature fluctuations due to thermal striping, thermal stratifications and sodium free level fluctuations at the liquid–cover gas interfaces are to be investigated carefully to eliminate high cycle thermal fatigue of structures. Solutions to address the core thermal hydraulics call for high power computing. Innovative concepts and methods are developed to carry out plant dynamics and safety studies. Particularly, extensive numerical and experimental simulation techniques are needed for understanding and solving the gas entrainment mechanisms and its effects on core safety. Though decay heat removal through natural convection is achievable in a pool type SFR, demonstration of design solutions conceived in the reactor and performance of diverse systems under all operating conditions, especially over prolonged station blackout situations needs advanced CFD computations and should be validated by relatively large scale simulated experiments. These issues are addressed in this paper under five broad topics: special thermal hydraulic issues to be addressed in SFR, thermal hydraulic design and analysis, plant dynamics studies, safety studies and evolving thermal hydraulic studies for the future FBRs. The 500 MWe Prototype Fast Breeder Reactor (PFBR) is taken as the reference design for addressing the issues. Indian fast reactor programme is highlighted in the introduction

Extended Analytic Linear Model of Hydraulic Cylinder With Respect Different Piston Areas and Volumes

Directory of Open Access Journals (Sweden)

Petr KOŇAŘÍK

2009-06-01

Full Text Available Standard analytic linear model of hydraulic cylinder usually comes from assumptions of identical action piston areas on both sides of hydraulic cylinder (double piston rod and suitable operation point, which is usually chosen in the middle of piston. By reason of that volumes inside of cylinder are than same. Moreover for control of that arrangement of hydraulic cylinder, usually controlled by 4/3 servovalve, the same mount of flows comes in and comes out to each of chambers of hydraulic cylinder. Presented paper deal with development of extended form of analytic linear model of single piston rod hydraulic cylinder which respects different action piston areas and volumes inside of chambers of hydraulic cylinder and also two different input flows of hydraulic cylinder. In extended model are also considered possibilities of different dead volumes in hoses and intake parts of hydraulic cylinder. Dead volume has impact on damping of hydraulic cylinder. Because the system of hydraulic cylinder is generally presented as a integrative system with inertia of second order: eq , we can than obtain time constants and damping of hydraulic cylinder for each of analytic form model. The model has arisen for needs of model fractionation on two parts. Part of behaviour of chamber A and part of behaviour of chamber B of cylinder. It was created for the reason of analysis and synthesis of control parameters of regulation circuit of multivalve control concept of hydraulic drive with separately controlled chamber A and B which could be then used for.

Thermal-hydraulic interfacing code modules for CANDU reactors

Energy Technology Data Exchange (ETDEWEB)

Liu, W.S.; Gold, M.; Sills, H. [Ontario Hydro Nuclear, Toronto (Canada)] [and others

1997-07-01

The approach for CANDU reactor safety analysis in Ontario Hydro Nuclear (OHN) and Atomic Energy of Canada Limited (AECL) is presented. Reflecting the unique characteristics of CANDU reactors, the procedure of coupling the thermal-hydraulics, reactor physics and fuel channel/element codes in the safety analysis is described. The experience generated in the Canadian nuclear industry may be useful to other types of reactors in the areas of reactor safety analysis.

Thermal-hydraulic interfacing code modules for CANDU reactors

International Nuclear Information System (INIS)

Liu, W.S.; Gold, M.; Sills, H.

1997-01-01

The approach for CANDU reactor safety analysis in Ontario Hydro Nuclear (OHN) and Atomic Energy of Canada Limited (AECL) is presented. Reflecting the unique characteristics of CANDU reactors, the procedure of coupling the thermal-hydraulics, reactor physics and fuel channel/element codes in the safety analysis is described. The experience generated in the Canadian nuclear industry may be useful to other types of reactors in the areas of reactor safety analysis

Hydraulic design development of Xiluodu Francis turbine

International Nuclear Information System (INIS)

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

2012-01-01

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

The study of crosslinked fluid leakoff in hydraulic fracturing physical simulations

Energy Technology Data Exchange (ETDEWEB)

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

2000-07-01

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

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)

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.

Evaluating temporal changes in hydraulic conductivities near karst-terrain dams: Dokan Dam (Kurdistan-Iraq)

Science.gov (United States)

Dafny, Elad; Tawfeeq, Kochar Jamal; Ghabraie, Kazem

2015-10-01

Dam sites provide an outstanding opportunity to explore dynamic changes in the groundwater flow regime because of the high hydraulic gradient rapidly induced in their surroundings. This paper investigates the temporal changes of the hydraulic conductivities of the rocks and engineered structures via a thorough analysis of hydrological data collected at the Dokam Dam, Iraq, and a numerical model that simulates the Darcian component of the seepage. Analysis of the data indicates increased seepage with time and suggests that the hydraulic conductivity of the rocks increased as the conductivity of the grout curtain decreased. Conductivity changes on the order of 10-8 m/s, in a 20-yr period were quantified using the numerical analysis. It is postulated that the changes in hydraulic properties in the vicinity of Dokan Dam are due to suspension of fine materials, interbedded in small fissures in the rocks, and re-settlement of these materials along the curtain. Consequently, the importance of the grout curtain to minimize the downstream seepage, not only as a result of the conductivity contrast with the rocks, but also as a barrier to suspended clay sediments, is demonstrated. The numerical analysis also helped us to estimate the proportion of the disconnected karstic conduit flow to the overall flow.

Hydraulic characterization of volcanic rocks in Pahute Mesa using an integrated analysis of 16 multiple-well aquifer tests, Nevada National Security Site, 2009–14

Science.gov (United States)

Garcia, C. Amanda; Jackson, Tracie R.; Halford, Keith J.; Sweetkind, Donald S.; Damar, Nancy A.; Fenelon, Joseph M.; Reiner, Steven R.

2017-01-20

An improved understanding of groundwater flow and radionuclide migration downgradient from underground nuclear-testing areas at Pahute Mesa, Nevada National Security Site, requires accurate subsurface hydraulic characterization. To improve conceptual models of flow and transport in the complex hydrogeologic system beneath Pahute Mesa, the U.S. Geological Survey characterized bulk hydraulic properties of volcanic rocks using an integrated analysis of 16 multiple-well aquifer tests. Single-well aquifer-test analyses provided transmissivity estimates at pumped wells. Transmissivity estimates ranged from less than 1 to about 100,000 square feet per day in Pahute Mesa and the vicinity. Drawdown from multiple-well aquifer testing was estimated and distinguished from natural fluctuations in more than 200 pumping and observation wells using analytical water-level models. Drawdown was detected at distances greater than 3 miles from pumping wells and propagated across hydrostratigraphic units and major structures, indicating that neither faults nor structural blocks noticeably impede or divert groundwater flow in the study area.Consistent hydraulic properties were estimated by simultaneously interpreting drawdown from the 16 multiple-well aquifer tests with an integrated groundwater-flow model composed of 11 well-site models—1 for each aquifer test site. Hydraulic properties were distributed across volcanic rocks with the Phase II Pahute Mesa-Oasis Valley Hydrostratigraphic Framework Model. Estimated hydraulic-conductivity distributions spanned more than two orders of magnitude in hydrostratigraphic units. Overlapping hydraulic conductivity ranges among units indicated that most Phase II Hydrostratigraphic Framework Model units were not hydraulically distinct. Simulated total transmissivity ranged from 1,600 to 68,000 square feet per day for all pumping wells analyzed. High-transmissivity zones exceeding 10,000 square feet per day exist near caldera margins and extend

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.

Development of Thermal-hydraulic Analysis Methodology for Multi-module Breeding Blankets in K-DEMO

Energy Technology Data Exchange (ETDEWEB)

Kim, Geon-Woo; Lee, Jeong-Hun; Park, Goon-Cherl; Cho, Hyoung-Kyu [Seoul National University, Seoul (Korea, Republic of); Im, Kihak [National Fusion Research Institute, Daejeon (Korea, Republic of)

2015-05-15

In this paper, the purpose of the analyses is to extend the capability of MARS-KS to the entire blanket system which includes a few hundreds of single blanket modules. Afterwards, the plan for the whole blanket system analysis using MARS-KS is introduced and the result of the multiple blanket module analysis is summarized. A thermal-hydraulic analysis code for a nuclear reactor safety, MARS-KS, was applied for the conceptual design of the K-DEMO breeding blanket thermal analysis. Then, a methodology to simulate multiple blanket modules was proposed, which uses a supervisor program to handle each blanket module individually at first and then distribute the flow rate considering pressure drops arises in each module. For a feasibility test of the proposed methodology, 10 outboard blankets in a toroidal field sector were simulated, which are connected with each other through the inlet and outlet common headers. The calculation results of flow rates, pressure drops, and temperatures showed the validity of the calculation and thanks to the parallelization using MPI, almost linear speed-up could be obtained.

Virtual prototype simulation of hydraulic shovel kinematics for spatial characterization in surface mining operations

Energy Technology Data Exchange (ETDEWEB)

S. Frimpong; Y. Li [University of Missouri-Rolla, Rolla, MO (United States). Department of Mining and Nuclear Engineering

2005-12-15

Hydraulic shovels are large-capacity equipment for excavating and loading dump trucks in constrained surface mining environments. Kinematics simulation of such equipment allows mine planning engineers to plan, design and control their spatial environments to achieve operating safety and efficiency. In this study, a hydraulic shovel was modelled as a mechanical manipulator with five degrees of freedom comprising the crawler, upper, boom, stick, bucket and bucket door components. The model was captured in a schematic diagram consisting of a six-bar linkage using the symbolic notation of Denavit and Hartenberg (Ho and Sriwattanathmma 1989). Homogeneous transformation matrices were used to capture the spatial configuration between adjacent links. The forward kinematics method was used to formulate the kinematics equations by attaching Cartesian coordinates to the schematic shovel diagram. Based on the kinematics model, a 3D virtual prototype of the hydraulic shovel was built in the Automatic Dynamic Analysis of Mechanical Systems (ADAMS) environment to simulate the motions of the hydraulic shovel with selected time steps. The simulator was validated using real-world data with animation and numerical analysis of the digging, swinging and dumping motions of the shovel machinery. The superimposed display of the deployment of the hydraulic shovel in three phases allows a detailed motion examination of the system. The numerical results of linear and angular displacements of the bucket tip and bucket door can be used to analyse the kinematics motion of the hydraulic shovel for its optimization. This simulator provides a solid foundation for further dynamics modelling and dynamic hydraulic shovel performance studies.

Measurement of in-situ hydraulic conductivity in the Cretaceous Pierre Shale

International Nuclear Information System (INIS)

Neuzil, C.E.; Bredehoeft, J.D.

1981-01-01

A recent study of the hydrology of the Cretaceous Pierre Shale utilized three techniques for measuring the hydraulic conductivity of tight materials. Regional hydraulic conductivity was obtained from a hydrodynamic model analysis of the aquifer-aquitard system which includes the Pierre Shale. Laboratory values were obtained from consolidation tests on core samples. In-situ values of hydraulic conductivity were obtained by using a borehole slug test designed specifically for tight formations. The test is conducted by isolating a portion of the borehole with one or two packers, abruptly pressurizing the shut-in portion, and recording the pressure decay with time. The test utilizes the analytical solution for pressure decay as water flows into the surrounding formation. Consistent results were obtained using the test on three successively smaller portions of a borehole in the Pierre Shale. The in-situ tests and laboratory tests yielded comparable values; the regional hydraulic conductivity was two to three orders of magnitude larger. This suggests that the lower values represent intergranular hydraulic conductivity of the intact shale and the regional values represent secondary permeability due to fractures. Calculations based on fracture flow theory demonstrate that small fractures could account for the observed differences

Uncertainty in hydraulic tests in fractured rock

International Nuclear Information System (INIS)

Ji, Sung-Hoon; Koh, Yong-Kwon

2014-01-01

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

Analysis and interpretation of borehole hydraulic tests in deep boreholes: principles, model development, and applications

International Nuclear Information System (INIS)

Pickens, J.F.; Grisak, G.E.; Avis, J.D.; Belanger, D.W.

1987-01-01

A review of the literature on hydraulic testing and interpretive methods, particularly in low-permeability media, indicates a need for a comprehensive hydraulic testing interpretive capability. Physical limitations on boreholes, such as caving and erosion during continued drilling, as well as the high costs associated with deep-hole rigs and testing equipment, often necessitate testing under nonideal conditions with respect to antecedent pressures and temperatures. In these situations, which are common in the high-level nuclear waste programs throughout the world, the interpretive requirements include the ability to quantitatively account for thermally induced pressure responses and borehole pressure history (resulting in a time-dependent pressure profile around the borehole) as well as equipment compliance effects in low-permeability intervals. A numerical model was developed to provide the capability to handle these antecedent conditions. Sensitivity studies and practical applications are provided to illustrate the importance of thermal effects and antecedent pressure history. It is demonstrated theoretically and with examples from the Swiss (National Genossenschaft fuer die Lagerung radioaktiver Abfaelle) regional hydrogeologic characterization program that pressure changes (expressed as hydraulic head) of the order of tens to hundreds of meters can results from 1 0 to 2 0 C temperature variations during shut-in (packer isolated) tests in low-permeability formations. Misinterpreted formation pressures and hydraulic conductivity can also result from inaccurate antecedent pressure history. Interpretation of representative formation properties and pressures requires that antecedent pressure information and test period temperature data be included as an integral part of the hydraulic test analyses

Thermal-hydraulic effects of transition to improved System 80TM fuel

International Nuclear Information System (INIS)

Rodack, T.; Joffre, P.F.; Kapoor, R.K.

2004-01-01

ABB CE's improved System 80 TM PWR fuel design includes GUARDIAN debris-resistant features and laser-welded Zircaloy grids. The GUARDIAN features include an Inconel grid with debris-filtering features located just above the Lower End Fitting, and a solid fuel rod bottom end cap that extends above the filtering features. Tests and analyses were done to establish the impact of these design improvements on fuel assembly hydraulic performance. Further analysis was done to determine the mixed core thermal-hydraulic performance as the transition is made over two fuel cycles to a full core of the improved System 80 TM fuel. Results confirm that the Thermal-Hydraulic (T-H) effects of the reduction in hydraulic resistance between the improved and resident fuel due to the laser-welded Zircaloy grids offsets the effects of the increased resistance GUARDIAN grid. Therefore, the mechanically improved System 80 TM fuel can be implemented with no net impact on Departure from Nucleate Boiling (DNB) margin in transition cores. (author)

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

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.

The 25 kWe solar thermal Stirling hydraulic engine system: Conceptual design

Science.gov (United States)

White, Maurice; Emigh, Grant; Noble, Jack; Riggle, Peter; Sorenson, Torvald

1988-01-01

The conceptual design and analysis of a solar thermal free-piston Stirling hydraulic engine system designed to deliver 25 kWe when coupled to a 11 meter test bed concentrator is documented. A manufacturing cost assessment for 10,000 units per year was made. The design meets all program objectives including a 60,000 hr design life, dynamic balancing, fully automated control, more than 33.3 percent overall system efficiency, properly conditioned power, maximum utilization of annualized insolation, and projected production costs. The system incorporates a simple, rugged, reliable pool boiler reflux heat pipe to transfer heat from the solar receiver to the Stirling engine. The free-piston engine produces high pressure hydraulic flow which powers a commercial hydraulic motor that, in turn, drives a commercial rotary induction generator. The Stirling hydraulic engine uses hermetic bellows seals to separate helium working gas from hydraulic fluid which provides hydrodynamic lubrication to all moving parts. Maximum utilization of highly refined, field proven commercial components for electric power generation minimizes development cost and risk.

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.

Proceedings of the 1991 national conference on hydraulic engineering

International Nuclear Information System (INIS)

Shane, R.M.

1991-01-01

This book contains the proceedings of the 1991 National Conference of Hydraulic Engineering. The conference was held in conjunction with the International Symposium on Ground Water and a Software Exchange that facilitated exchange of information on recent software developments of interest to hydraulic engineers. Also included in the program were three mini-symposia on the Exclusive Economic Zone, Data Acquisition, and Appropriate Technology. Topics include sedimentation; appropriate technology; exclusive economic zone hydraulics; hydraulic data acquisition and display; innovative hydraulic structures and water quality applications of hydraulic research, including the hydraulics of aerating turbines; wetlands; hydraulic and hydrologic extremes; highway drainage; overtopping protection of dams; spillway design; coastal and estuarine hydraulics; scale models; computation hydraulics; GIS and expert system applications; watershed response to rainfall; probabilistic approaches; and flood control investigations

Undular Hydraulic Jump

Directory of Open Access Journals (Sweden)

Oscar Castro-Orgaz

2015-04-01

Full Text Available The transition from subcritical to supercritical flow when the inflow Froude number Fo is close to unity appears in the form of steady state waves called undular hydraulic jump. The characterization of the undular hydraulic jump is complex due to the existence of a non-hydrostatic pressure distribution that invalidates the gradually-varied flow theory, and supercritical shock waves. The objective of this work is to present a mathematical model for the undular hydraulic jump obtained from an approximate integration of the Reynolds equations for turbulent flow assuming that the Reynolds number R is high. Simple analytical solutions are presented to reveal the physics of the theory, and a numerical model is used to integrate the complete equations. The limit of application of the theory is discussed using a wave breaking condition for the inception of a surface roller. The validity of the mathematical predictions is critically assessed using physical data, thereby revealing aspects on which more research is needed

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.)

Magnetic Field and Torque Output of Packaged Hydraulic Torque Motor

Directory of Open Access Journals (Sweden)

Liang Yan

2018-01-01

Full Text Available Hydraulic torque motors are one key component in electro-hydraulic servo valves that convert the electrical signal into mechanical motions. The systematic characteristics analysis of the hydraulic torque motor has not been found in the previous research, including the distribution of the electromagnetic field and torque output, and particularly the relationship between them. In addition, conventional studies of hydraulic torque motors generally assume an evenly distributed magnetic flux field and ignore the influence of special mechanical geometry in the air gaps, which may compromise the accuracy of analyzing the result and the high-precision motion control performance. Therefore, the objective of this study is to conduct a detailed analysis of the distribution of the magnetic field and torque output; the influence of limiting holes in the air gaps is considered to improve the accuracy of both numerical computation and analytical modeling. The structure and working principle of the torque motor are presented first. The magnetic field distribution in the air gaps and the magnetic saturation in the iron blocks are analyzed by using a numerical approach. Subsequently, the torque generation with respect to the current input and assembly errors is analyzed in detail. This shows that the influence of limiting holes on the magnetic field is consistent with that on torque generation. Following this, a novel modified equivalent magnetic circuit is proposed to formulate the torque output of the hydraulic torque motor analytically. The comparison among the modified equivalent magnetic circuit, the conventional modeling approach and the numerical computation is conducted, and it is found that the proposed method helps to improve the modeling accuracy by taking into account the effect of special geometry inside the air gaps.

Structural Integrity Assessment for SSDM Hydraulic Cylinder of JRTR

International Nuclear Information System (INIS)

Kim, Sanghaun; Lee, Jin Haeng; Cho, Yeonggarp; Yoo, Yeonsik

2014-01-01

In HANARO, there are four Control Rod Drive Mechanisms (CRDMs) with an individual step motor and four Shutoff (SO) Units with an individual hydraulic system located at the top of reactor pool. The absorber rods in SO units are poised at the top of the core by the hydraulic force during normal operation. The rods of SO units drop by gravity as the first reactor showdown mechanism when a trip is commended by the reactor protection system (RPS). The CRDMs act as the first reactor shutdown mechanism and reactor regulating as well. The top-mounted SSDM driven by the hydraulic system for the JRTR is under design in KAERI. The SSDM provides an alternate and independent means of reactor shutdown. The second shutdown rods (SSRs) of the SSDM are poised at the top of the core by the hydraulic system during the normal operation and drop by gravity for the reactor trip. Based on the proven technology of the design, operation and maintenance for HANARO, the SSDM for the JRTR has been optimized by the design improvement from the experience and test. This paper aims for the structural integrity assessment for SSDM hydraulic cylinder which is designed on the basis of the SO unit of HANARO but optimized with the new core environment (i. e., geometrical, physical, etc.) of JRTR. A stress analysis of the hydraulic cylinder for the SSDM used in JRTR has been performed through the conservative approach with the uncertainties in the system design step. The crank's pinch load with no slip between the bearing (stiffener) plate of hydraulic cylinder and base plate of mount bracket during SSE has been calculated by considering the design and seismic load combination. The stress by the load combination satisfies the Class 3 criteria given Table NG-3325 of Section III of the ASME Code. The maximum stresses are at the clamp contact region in the cylinder

Selective perceptions of hydraulic fracturing.

Science.gov (United States)

Sarge, Melanie A; VanDyke, Matthew S; King, Andy J; White, Shawna R

2015-01-01

Hydraulic fracturing (HF) is a focal topic in discussions about domestic energy production, yet the American public is largely unfamiliar and undecided about the practice. This study sheds light on how individuals may come to understand hydraulic fracturing as this unconventional production technology becomes more prominent in the United States. For the study, a thorough search of HF photographs was performed, and a systematic evaluation of 40 images using an online experimental design involving N = 250 participants was conducted. Key indicators of hydraulic fracturing support and beliefs were identified. Participants showed diversity in their support for the practice, with 47 percent expressing low support, 22 percent high support, and 31 percent undecided. Support for HF was positively associated with beliefs that hydraulic fracturing is primarily an economic issue and negatively associated with beliefs that it is an environmental issue. Level of support was also investigated as a perceptual filter that facilitates biased issue perceptions and affective evaluations of economic benefit and environmental cost frames presented in visual content of hydraulic fracturing. Results suggested an interactive relationship between visual framing and level of support, pointing to a substantial barrier to common understanding about the issue that strategic communicators should consider.

Birth of a hydraulic jump

Science.gov (United States)

Duchesne, Alexis; Bohr, Tomas; Andersen, Anders

2017-11-01

The hydraulic jump, i.e., the sharp transition between a supercritical and a subcritical free-surface flow, has been extensively studied in the past centuries. However, ever since Leonardo da Vinci asked it for the first time, an important question has been left unanswered: How does a hydraulic jump form? We present an experimental and theoretical study of the formation of stationary hydraulic jumps in centimeter wide channels. Two starting situations are considered: The channel is, respectively, empty or filled with liquid, the liquid level being fixed by the wetting properties and the boundary conditions. We then change the flow-rate abruptly from zero to a constant value. In an empty channel, we observe the formation of a stationary hydraulic jump in a two-stage process: First, the channel fills by the advancing liquid front, which undergoes a transition from supercritical to subcritical at some position in the channel. Later the influence of the downstream boundary conditions makes the jump move slowly upstream to its final position. In the pre-filled channel, the hydraulic jump forms at the injector edge and then moves downstream to its final position.

A 6-DOF vibration isolation system for hydraulic hybrid vehicles

Science.gov (United States)

Nguyen, The; Elahinia, Mohammad; Olson, Walter W.; Fontaine, Paul

2006-03-01

This paper presents the results of vibration isolation analysis for the pump/motor component of hydraulic hybrid vehicles (HHVs). The HHVs are designed to combine gasoline/diesel engine and hydraulic power in order to improve the fuel efficiency and reduce the pollution. Electric hybrid technology is being applied to passenger cars with small and medium engines to improve the fuel economy. However, for heavy duty vehicles such as large SUVs, trucks, and buses, which require more power, the hydraulic hybridization is a more efficient choice. In function, the hydraulic hybrid subsystem improves the fuel efficiency of the vehicle by recovering some of the energy that is otherwise wasted in friction brakes. Since the operation of the main component of HHVs involves with rotating parts and moving fluid, noise and vibration are an issue that affects both passengers (ride comfort) as well as surrounding people (drive-by noise). This study looks into the possibility of reducing the transmitted noise and vibration from the hydraulic subsystem to the vehicle's chassis by using magnetorheological (MR) fluid mounts. To this end, the hydraulic subsystem is modeled as a six degree of freedom (6-DOF) rigid body. A 6-DOF isolation system, consisting of five mounts connected to the pump/motor at five different locations, is modeled and simulated. The mounts are designed by combining regular elastomer components with MR fluids. In the simulation, the real loading and working conditions of the hydraulic subsystem are considered and the effects of both shock and vibration are analyzed. The transmissibility of the isolation system is monitored in a wide range of frequencies. The geometry of the isolation system is considered in order to sustain the weight of the hydraulic system without affecting the design of the chassis and the effectiveness of the vibration isolating ability. The simulation results shows reduction in the transmitted vibration force for different working cycles of

Model-based nonlinear control of hydraulic servo systems: Challenges, developments and perspectives

Science.gov (United States)

Yao, Jianyong

2018-06-01

Hydraulic servo system plays a significant role in industries, and usually acts as a core point in control and power transmission. Although linear theory-based control methods have been well established, advanced controller design methods for hydraulic servo system to achieve high performance is still an unending pursuit along with the development of modern industry. Essential nonlinearity is a unique feature and makes model-based nonlinear control more attractive, due to benefit from prior knowledge of the servo valve controlled hydraulic system. In this paper, a discussion for challenges in model-based nonlinear control, latest developments and brief perspectives of hydraulic servo systems are presented: Modelling uncertainty in hydraulic system is a major challenge, which includes parametric uncertainty and time-varying disturbance; some specific requirements also arise ad hoc difficulties such as nonlinear friction during low velocity tracking, severe disturbance, periodic disturbance, etc.; to handle various challenges, nonlinear solutions including parameter adaptation, nonlinear robust control, state and disturbance observation, backstepping design and so on, are proposed and integrated, theoretical analysis and lots of applications reveal their powerful capability to solve pertinent problems; and at the end, some perspectives and associated research topics (measurement noise, constraints, inner valve dynamics, input nonlinearity, etc.) in nonlinear hydraulic servo control are briefly explored and discussed.

Informational Entropy and Bridge Scour Estimation under Complex Hydraulic Scenarios

Science.gov (United States)

Pizarro, Alonso; Link, Oscar; Fiorentino, Mauro; Samela, Caterina; Manfreda, Salvatore

2017-04-01

Bridges are important for society because they allow social, cultural and economic connectivity. Flood events can compromise the safety of bridge piers up to the complete collapse. The Bridge Scour phenomena has been described by empirical formulae deduced from hydraulic laboratory experiments. The range of applicability of such models is restricted by the specific hydraulic conditions or flume geometry used for their derivation (e.g., water depth, mean flow velocity, pier diameter and sediment properties). We seek to identify a general formulation able to capture the main dynamic of the process in order to cover a wide range of hydraulic and geometric configuration, allowing to extend our analysis in different contexts. Therefore, exploiting the Principle of Maximum Entropy (POME) and applying it on the recently proposed dimensionless Effective flow work, W*, we derived a simple model characterized by only one parameter. The proposed Bridge Scour Entropic (BRISENT) model shows good performances under complex hydraulic conditions as well as under steady-state flow. Moreover, the model was able to capture the evolution of scour in several hydraulic configurations even if the model contains only one parameter. Furthermore, results show that the model parameter is controlled by the geometric configurations of the experiment. This offers a possible strategy to obtain a priori model parameter calibration. The BRISENT model represents a good candidate for estimating the time-dependent scour depth under complex hydraulic scenarios. The authors are keen to apply this idea for describing the scour behavior during a real flood event. Keywords: Informational entropy, Sediment transport, Bridge pier scour, Effective flow work.

Thermal-hydraulic modeling of porous bed reactors

International Nuclear Information System (INIS)

Araj, K.J.; Nourbakhsh, H.P.

1987-01-01

Optimum design of nuclear reactor core requires an iterative approach between the thermal-hydraulic, neutronic and operational analysis. This paper concentrates on the thermal-hydraulic behavior of a hydrogen cooled, small particle bed reactor (PBR). The PBR core, modeled here, consists of a hexagonal array of fuel elements embedded in a moderator matrix. The fuel elements are annular packed beds of fuel particles held between two porous cylindrical frits. These particles, 500 to 600 μm in diameter, have a uranium carbide core, which is coated by two layers of graphite and an outer coating of zirconium carbide. Coolant flow, radially inward, from the cold frit through the packed bed and hot frit and axially out the channel, formed by the hot frit, to a common plenum. 5 refs., 1 fig., 2 tabs

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.

Hydraulic Hybrid Fleet Vehicle Testing | Transportation Research | NREL

Science.gov (United States)

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

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)

Analysis of in-R12 CHF data: influence of hydraulic diameter and heating length; test of Weisman boiling crisis model

International Nuclear Information System (INIS)

Czop, V.; Herer, C.; Souyri, A.; Garnier, J.

1993-09-01

In order to progress on the comprehensive modelling of the boiling crisis phenomenon, Electricite de France (EDF), Commissariat a l'Energie Atomique (CEA) and FRAMATOME have set up experimental programs involving in-R12 tests: the EDF APHRODITE program and the CEA-EDF-FRAMATOME DEBORA program. The first phase in these programs aims to acquire critical heat flux (CHF) data banks, within large thermal-hydraulic parameter ranges, both in cylindrical and annular configurations, and with different hydraulic diameters and heating lengths. Actually, three data banks have been considered in the analysis, all of them concerning in-R12 round tube tests: - the APHRODITE data bank, obtained at EDF with a 13 mn inside diameter, - the DEBORA data bank, obtained at CEA with a 19.2 mm inside diameter, - the KRISTA data bank, obtained at KfK with a 8 mm inside diameter. The analysis was conducted using CHF correlations and with the help of an advanced mathematical tool using pseudo-cubic thin plate type Spline functions. Two conclusions were drawn: -no influence of the heating length on our CHF results, - the influence of the diameter on the CHF cannot be simply expressed by an exponential function of this parameter, as thermal-hydraulic parameters also have an influence. Some calculations with Weisman and Pei theoretical boiling crisis model have been compared to experimental values: fairly good agreement was obtained, but further study must focus on improving the modelling of the influence of pressure and mass velocity. (authors). 12 figs., 4 tabs., 21 refs

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.)

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.

Conceptual assessment and thermal hydraulic analysis of MVDS system for the dry storage of reduced metal fuel

International Nuclear Information System (INIS)

Lee, J. C.; Bang, K. S.; Shin, H. S.; Joo, J. S.; Su, K. S.; Kim, H. D.

2003-01-01

Conceptual assessment and thermal hydraulic analysis of MVDS storage system have been carried out for application of reduced metal fuel. The storage concept was established considering the optimum weight, storage volume and thermal efficiency. The capacity of MVDS system for loading the reduced metal fuel has four times as compared with existing PWR fuel storage system. In the results of thermal analysis, the maximum temperature of metal fuel was estimated to be 110 .deg. C which is lower than the allowable value under normal operation condition. Therefore, it is shown that the MVDS system can feasibly accomodate the reduced metal fuel in aspect of thermal safety

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

International Nuclear Information System (INIS)

Bestion, D.

2010-01-01

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

Development of thermal-hydraulic analysis methodology for multiple modules of water-cooled breeder blanket in fusion DEMO reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Geon-Woo; Lee, Jeong-Hun [Department of Nuclear Engineering, Seoul National University 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Cho, Hyoung-Kyu, E-mail: chohk@snu.ac.kr [Department of Nuclear Engineering, Seoul National University 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Park, Goon-Cherl [Department of Nuclear Engineering, Seoul National University 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Im, Kihak [National Fusion Research Institute, 169-148, Yuseong-gu, Daejeon 305-806 (Korea, Republic of)

2016-02-15

Highlights: • A methodology to simulate the K-DEMO blanket system was proposed. • The results were compared with the CFD, to verify the prediction capability of MARS. • 46 Blankets in a single sector in K-DEMO were simulated using MARS-KS. • Supervisor program was devised to handle each blanket module individually. • The calculation results showed the flow rates, pressure drops, and temperatures. - Abstract: According to the conceptual design of the fusion DEMO reactor proposed by the National Fusion Research Institute of Korea, the water-cooled breeding blanket system incorporates a total of 736 blanket modules. The heat flux and neutron wall loading to each blanket module vary along their poloidal direction, and hence, thermal analysis for at least one blanket sector is required to confirm that the temperature limitations of the materials are satisfied in all the blanket modules. The present paper proposes a methodology of thermal analysis for multiple modules of the blanket system using a nuclear reactor thermal-hydraulic analysis code, MARS-KS. In order to overcome the limitations of the code, caused by the restriction on the number of computational nodes, a supervisor program was devised, which handles each blanket module separately at first, and then corrects the flow rate, considering pressure drops that occur in each module. For a feasibility test of the proposed methodology, 46 blankets in a single sector were simulated; the calculation results of the parameters, such as mass flow, pressure drops, and temperature distribution in the multiple blanket modules showed that the multi-module analysis method can be used for efficient thermal-hydraulic analysis of the fusion DEMO reactor.

Hydraulics and pneumatics

CERN Document Server

Parr, Andrew

2006-01-01

Nearly all industrial processes require objects to be moved, manipulated or subjected to some sort of force. This is frequently accomplished by means of electrical equipment (such as motors or solenoids), or via devices driven by air (pneumatics) or liquids (hydraulics).This book has been written by a process control engineer as a guide to the operation of hydraulic and pneumatic systems for all engineers and technicians who wish to have an insight into the components and operation of such a system.This second edition has been fully updated to include all recent developments su

Monitoring hydraulic stimulation using telluric sounding

Science.gov (United States)

Rees, Nigel; Heinson, Graham; Conway, Dennis

2018-01-01

The telluric sounding (TS) method is introduced as a potential tool for monitoring hydraulic fracturing at depth. The advantage of this technique is that it requires only the measurement of electric fields, which are cheap and easy when compared with magnetotelluric measurements. Additionally, the transfer function between electric fields from two locations is essentially the identity matrix for a 1D Earth no matter what the vertical structure. Therefore, changes in the earth resulting from the introduction of conductive bodies underneath one of these sites can be associated with deviations away from the identity matrix, with static shift appearing as a galvanic multiplier at all periods. Singular value decomposition and eigenvalue analysis can reduce the complexity of the resulting telluric distortion matrix to simpler parameters that can be visualised in the form of Mohr circles. This technique would be useful in constraining the lateral extent of resistivity changes. We test the viability of utilising the TS method for monitoring on both a synthetic dataset and for a hydraulic stimulation of an enhanced geothermal system case study conducted in Paralana, South Australia. The synthetic data example shows small but consistent changes in the transfer functions associated with hydraulic stimulation, with grids of Mohr circles introduced as a useful diagnostic tool for visualising the extent of fluid movement. The Paralana electric field data were relatively noisy and affected by the dead band making the analysis of transfer functions difficult. However, changes in the order of 5% were observed from 5 s to longer periods. We conclude that deep monitoring using the TS method is marginal at depths in the order of 4 km and that in order to have meaningful interpretations, electric field data need to be of a high quality with low levels of site noise.[Figure not available: see fulltext.

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.

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.

Thermal hydraulic considerations in liquid-metal-cooled components of tokamak fusion reactors

International Nuclear Information System (INIS)

Picologlou, B.F.; Reed, C.B.; Hua, T.Q.

1989-01-01

The basic considerations of MHD thermal hydraulics for liquid-metal-cooled blankets and first walls of tokamak fusion reactors are discussed. The liquid-metal MHD program of Argonne National Laboratory (ANL) dedicated to analytical and experimental investigations of reactor relevant MHD flows and development of relevant thermal hydraulic design tools is presented. The status of the experimental program and examples of local velocity measurements are given. An account of the MHD codes developed to date at ANL is also presented as is an example of a 3-D thermal hydraulic analysis carried out with such codes. Finally, near term plans for experimental investigations and code development are outlined. 20 refs., 8 figs., 1 tab

Characteristics of Air Entrainment in Hydraulic Jump

Science.gov (United States)

Albarkani, M. S. S.; Tan, L. W.; Al-Gheethi, A.

2018-04-01

The characteristics of hydraulic jump, especially the air entrainment within jump is still not properly understood. Therefore, the current work aimed to determine the size and number of air entrainment formed in hydraulic jump at three different Froude numbers and to obtain the relationship between Froude number with the size and number of air entrainment in hydraulic jump. Experiments of hydraulic jump were conducted in a 10 m long and 0.3 m wide Armfield S6MKII glass-sided tilting flume. Hydraulic jumps were produced by flow under sluice gate with varying Froude number. The air entrainment of the hydraulic jump was captured with a Canon Power Shot SX40 HS digital camera in video format at 24 frames per second. Three discharges have been considered, i.e. 0.010 m3/s, 0.011 m3/s, and 0.013 m3/s. For hydraulic jump formed in each discharge, 32 frames were selected for the purpose of analysing the size and number of air entrainment in hydraulic jump. The results revealed that that there is a tendency to have greater range in sizes of air bubbles as Fr1 increases. Experiments with Fr1 = 7.547. 7.707, and 7.924 shown that the number of air bubbles increases exponentially with Fr1 at a relationship of N = 1.3814 e 0.9795Fr1.

Rapid hydraulic recovery in Eucalyptus pauciflora after drought: linkages between stem hydraulics and leaf gas exchange.

Science.gov (United States)

Martorell, Sebastià; Diaz-Espejo, Antonio; Medrano, Hipólito; Ball, Marilyn C; Choat, Brendan

2014-03-01

In woody plants, photosynthetic capacity is closely linked to rates at which the plant hydraulic system can supply water to the leaf surface. Drought-induced embolism can cause sharp declines in xylem hydraulic conductivity that coincide with stomatal closure and reduced photosynthesis. Recovery of photosynthetic capacity after drought is dependent on restored xylem function, although few data exist to elucidate this coordination. We examined the dynamics of leaf gas exchange and xylem function in Eucalyptus pauciflora seedlings exposed to a cycle of severe water stress and recovery after re-watering. Stomatal closure and leaf turgor loss occurred at water potentials that delayed the extensive spread of embolism through the stem xylem. Stem hydraulic conductance recovered to control levels within 6 h after re-watering despite a severe drought treatment, suggesting an active mechanism embolism repair. However, stomatal conductance did not recover after 10 d of re-watering, effecting tighter control of transpiration post drought. The dynamics of recovery suggest that a combination of hydraulic and non-hydraulic factors influenced stomatal behaviour post drought. © 2013 John Wiley & Sons Ltd.

Estimating Hydraulic Resistance for Floodplain Mapping and Hydraulic Studies from High-Resolution Topography: Physical and Numerical Simulations

Science.gov (United States)

Minear, J. T.

2017-12-01

One of the primary unknown variables in hydraulic analyses is hydraulic resistance, values for which are typically set using broad assumptions or calibration, with very few methods available for independent and robust determination. A better understanding of hydraulic resistance would be highly useful for understanding floodplain processes, forecasting floods, advancing sediment transport and hydraulic coupling, and improving higher dimensional flood modeling (2D+), as well as correctly calculating flood discharges for floods that are not directly measured. The relationship of observed features to hydraulic resistance is difficult to objectively quantify in the field, partially because resistance occurs at a variety of scales (i.e. grain, unit and reach) and because individual resistance elements, such as trees, grass and sediment grains, are inherently difficult to measure. Similar to photogrammetric techniques, Terrestrial Laser Scanning (TLS, also known as Ground-based LiDAR) has shown great ability to rapidly collect high-resolution topographic datasets for geomorphic and hydrodynamic studies and could be used to objectively quantify the features that collectively create hydraulic resistance in the field. Because of its speed in data collection and remote sensing ability, TLS can be used both for pre-flood and post-flood studies that require relatively quick response in relatively dangerous settings. Using datasets collected from experimental flume runs and numerical simulations, as well as field studies of several rivers in California and post-flood rivers in Colorado, this study evaluates the use of high-resolution topography to estimate hydraulic resistance, particularly from grain-scale elements. Contrary to conventional practice, experimental laboratory runs with bed grain size held constant but with varying grain-scale protusion create a nearly twenty-fold variation in measured hydraulic resistance. The ideal application of this high-resolution topography

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

International Nuclear Information System (INIS)

Fujiki, Kazuo; Asaka, Hideaki; Ishida, Toshihisa

1986-01-01

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

Hydraulic testing in granite using the sinusoidal variation of pressure

International Nuclear Information System (INIS)

Black, J.H.; Holmes, D.C.; Noy, D.J.

1982-09-01

Access to two boreholes at the Carwynnen test site in Cornwall enabled the trial of a number of innovative approaches to the hydrogeology of fractured crystalline rock. These methods ranged from the use of seisviewer data to measure the orientation of fractures to the use of the sinusoidal pressure technique to measure directional hydraulic diffusivity. The testing began with a short programme of site investigation consisting of borehole caliper and seisviewer logging followed by some single borehole hydraulic tests. The single borehole hydraulic testing was designed to assess whether the available boreholes and adjacent rock were suitable for testing using the sinusoidal method. The main testing methods were slug and pulse tests and were analysed using the fissured porous medium analysis proposed in Barker and Black (1983). Derived hydraulic conductivity (K) ranged from 2 x 10 -12 m/sec to 5 x 10 -7 m/sec with one near-surface zone of high K being perceived in both boreholes. The results were of the form which is typical of fractured rock and indicated a combination of high fracture frequency and permeable granite matrix. The results are described and discussed. (author)

Compressed air piping, 241-SY-101 hydraulic pump retrieval trailer

International Nuclear Information System (INIS)

Wilson, T.R.

1994-01-01

The following Design Analysis was prepared by the Westinghouse Hanford Company to determine pressure losses in the compressed air piping installed on the hydraulic trailer for the 241-SY-101 pump retrieval mission

Hydraulic gradients in rock aquifers

International Nuclear Information System (INIS)

Dahlblom, P.

1992-05-01

This report deals with fractured rock as a host for deposits of hazardous waste. In this context the rock, with its fractures containing moving groundwater, is called the geological barrier. The desired properties of the geological barrier are low permeability to water, low hydraulic gradients and ability to retain matter dissolved in the water. The hydraulic gradient together with the permeability and the porosity determines the migration velocity. Mathematical modelling of the migration involves calculation of the water flow and the hydrodynamic dispersion of the contaminant. The porous medium approach can be used to calculate mean flow velocities and hydrodynamic dispersion of a large number of fractures are connected, which means that a large volume have to be considered. It is assumed that the porous medium approach can be applied, and a number of idealized examples are shown. It is assumed that the groundwater table is replenished by percolation at a constant rate. One-dimensional analytical calculations show that zero hydraulic gradients may exist at relatively large distance from the coast. Two-dimensional numerical calculations show that it may be possible to find areas with low hydraulic gradients and flow velocities within blocks surrounded by areas with high hydraulic conductivity. (au)

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

Effect of physical property of supporting media and variable hydraulic loading on hydraulic characteristics of advanced onsite wastewater treatment system.

Science.gov (United States)

Sharma, Meena Kumari; Kazmi, Absar Ahmad

2015-01-01

A laboratory-scale study was carried out to investigate the effects of physical properties of the supporting media and variable hydraulic shock loads on the hydraulic characteristics of an advanced onsite wastewater treatment system. The system consisted of two upflow anaerobic reactors (a septic tank and an anaerobic filter) accommodated within a single unit. The study was divided into three phases on the basis of three different supporting media (Aqwise carriers, corrugated ring and baked clay) used in the anaerobic filter. Hydraulic loadings were based on peak flow factor (PFF), varying from one to six, to simulate the actual conditions during onsite wastewater treatment. Hydraulic characteristics of the system were identified on the basis of residence time distribution analyses. The system showed a very good hydraulic efficiency, between 0.86 and 0.93, with the media of highest porosity at the hydraulic loading of PFF≤4. At the higher hydraulic loading of PFF 6 also, an appreciable hydraulic efficiency of 0.74 was observed. The system also showed good chemical oxygen demand and total suspended solids removal efficiency of 80.5% and 82.3%, respectively at the higher hydraulic loading of PFF 6. Plug-flow dispersion model was found to be the most appropriate one to describe the mixing pattern of the system, with different supporting media at variable loading, during the tracer study.