WorldWideScience

Sample records for natural circulation two-phase

  1. Thermalhydraulic instability analysis of a two phase natural circulation loop

    International Nuclear Information System (INIS)

    Sesini, Paula Aida

    1998-01-01

    This work presents an analysis of a loop operating in natural circulation regime. Experiments were done in a rectangular closed circuit in one and two-phase flows. Numerical analysis were performed initially with the CIRNAT code and afterwards with RELAP5/MOD2. The limitations of CIRNAT were studied and new developments for this code are proposed. (author)

  2. Experimental study of two-phase natural circulation circuit

    Energy Technology Data Exchange (ETDEWEB)

    Lemos, Wanderley Freitas; Su, Jian, E-mail: wlemos@lasme.coppe.ufrj.br, E-mail: sujian@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Faccini, Jose Luiz Horacio, E-mail: faccini@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), RIo de Janeiro, RJ (Brazil). Lab. de Termo-Hidraulica Experimental

    2012-07-01

    This paper reports an experimental study on the behavior of fluid flow in natural circulation under single-and two-phase flow conditions. The natural circulation circuit was designed based on concepts of similarity and scale in proportion to the actual operating conditions of a nuclear reactor. This test equipment has similar performance to the passive system for removal of residual heat presents in Advanced Pressurized Water Reactors (A PWR). The experiment was carried out by supplying water to primary and secondary circuits, as well as electrical power resistors installed inside the heater. Power controller has available to adjust the values for supply of electrical power resistors, in order to simulate conditions of decay of power from the nuclear reactor in steady state. Data acquisition system allows the measurement and control of the temperature at different points by means of thermocouples installed at several points along the circuit. The behavior of the phenomenon of natural circulation was monitored by a software with graphical interface, showing the evolution of temperature measurement points and the results stored in digital format spreadsheets. Besides, the natural circulation flow rate was measured by a flowmeter installed on the hot leg. A flow visualization technique was used the for identifying vertical flow regimes of two-phase natural circulation. Finally, the Reynolds Number was calculated for the establishment of a friction factor correlation dependent on the scale geometrical length, height and diameter of the pipe. (author)

  3. Experimental study of two-phase natural circulation circuit

    International Nuclear Information System (INIS)

    Lemos, Wanderley Freitas; Su, Jian; Faccini, Jose Luiz Horacio

    2012-01-01

    This paper reports an experimental study on the behavior of fluid flow in natural circulation under single-and two-phase flow conditions. The natural circulation circuit was designed based on concepts of similarity and scale in proportion to the actual operating conditions of a nuclear reactor. This test equipment has similar performance to the passive system for removal of residual heat presents in Advanced Pressurized Water Reactors (A PWR). The experiment was carried out by supplying water to primary and secondary circuits, as well as electrical power resistors installed inside the heater. Power controller has available to adjust the values for supply of electrical power resistors, in order to simulate conditions of decay of power from the nuclear reactor in steady state. Data acquisition system allows the measurement and control of the temperature at different points by means of thermocouples installed at several points along the circuit. The behavior of the phenomenon of natural circulation was monitored by a software with graphical interface, showing the evolution of temperature measurement points and the results stored in digital format spreadsheets. Besides, the natural circulation flow rate was measured by a flowmeter installed on the hot leg. A flow visualization technique was used the for identifying vertical flow regimes of two-phase natural circulation. Finally, the Reynolds Number was calculated for the establishment of a friction factor correlation dependent on the scale geometrical length, height and diameter of the pipe. (author)

  4. Transient boiling in two-phase helium natural circulation loops

    Science.gov (United States)

    Furci, H.; Baudouy, B.; Four, A.; Meuris, C.

    2014-01-01

    Two-phase helium natural circulation loops are used for cooling large superconducting magnets, as CMS for LHC. During normal operation or in the case of incidents, transients are exerted on the cooling system. Here a cooling system of this type is studied experimentally. Sudden power changes are operated on a vertical-heated-section natural convection loop, simulating a fast increase of heat deposition on magnet cooling pipes. Mass flow rate, heated section wall temperature and pressure drop variations are measured as a function of time, to assess the time behavior concerning the boiling regime according to the values of power injected on the heated section. The boiling curves and critical heat flux (CHF) values have been obtained in steady state. Temperature evolution has been observed in order to explore the operating ranges where heat transfer is deteriorated. Premature film boiling has been observed during transients on the heated section in some power ranges, even at appreciably lower values than the CHF. A way of attenuating these undesired temperature excursions has been identified through the application of high enough initial heating power.

  5. Simulation of Two-Phase Natural Circulation Loop for Core Cather Cooling Using Air Water

    International Nuclear Information System (INIS)

    Revankar, S. T.; Huang, S. F.; Song, K. W.; Rhee, B. W.; Park, R. J.; Song, J. H.

    2012-01-01

    A closed loop natural circulation system employs thermally induced density gradients in single phase or two-phase liquid form to induce circulation of the working fluid thereby obviating the need for any mechanical moving parts such as pumps and pump controls. This increases the reliability and safety of the cooling system and reduces installation, operation and maintenance costs. That is the reason natural circulation cooling has been considered in advanced reactor core cooling and in engineered safety systems. Natural circulation cooling has been proposed to remove reactor decay heat by external vessel cooling for in-vessel core retention during sever accident scenario. Recently in APR1400 reactor core catcher design natural circulation cooling is proposed to stabilize and cool the corium ejected from the reactor vessel following core melt and breach of reactor vessel. The natural circulation flow is similar to external vessel cooling where water flows through an inclined narrow gap below hot surface and is heated to produce boiling. The two-phase natural circulation enables cooling of the corium pool collected on core catcher. Due to importance of this problem this paper focuses simulation of the two-phase natural circulation through inclined gap using air-water system. Scaling criteria for air-water loop are derived that enable simulation of the flow regimes and natural circulation flow rates in such systems using air-water system

  6. Post Analysis of Two Phase Natural Circulation Mass Flow Rate for CE-PECS

    Energy Technology Data Exchange (ETDEWEB)

    Park, R. J.; Ha, K. S.; Rhee, B. W.; Kim, H. Y. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The coolant in the inclined channel absorbs the decay heat and sensible heat transferred from the corium through the structure of the core catcher body and flows up to the pool as a two phase mixture. On the other hand, some of the pool water will flow into the inlet of the downcomer piping, and will flow into the inclined cooling channel of the core catcher by gravity. The engineered cooling channel is designed to provide effective long-term cooling and stabilization of the corium mixture in the core catcher body while facilitating steam venting. To maintain the integrity of the ex-vessel core catcher, however, it is required that the coolant be circulated at a rate along the inclined cooling channel sufficient to avoid CHF (Critical Heat Flux) on the heating surface of the cooling channel. In this study, post simulations of two phase natural circulation in the CEPECS have been performed to evaluate two phase flow characteristics and the natural circulation mass flow rate in the flow channel using the RELAP5/MOD3 computer code. Post simulations of two phase natural circulation in the CE-PECS have been conducted to evaluate two phase flow characteristics and the natural circulation mass flow rate in the flow channel using the RELAP5/MOD3 computer code. The RELAP5/MOD3 results have shown that the water circulation mass flow rate is approximately 8.7 kg/s in the base case.

  7. Post Analysis of Two Phase Natural Circulation Mass Flow Rate for CE-PECS

    International Nuclear Information System (INIS)

    Park, R. J.; Ha, K. S.; Rhee, B. W.; Kim, H. Y.

    2015-01-01

    The coolant in the inclined channel absorbs the decay heat and sensible heat transferred from the corium through the structure of the core catcher body and flows up to the pool as a two phase mixture. On the other hand, some of the pool water will flow into the inlet of the downcomer piping, and will flow into the inclined cooling channel of the core catcher by gravity. The engineered cooling channel is designed to provide effective long-term cooling and stabilization of the corium mixture in the core catcher body while facilitating steam venting. To maintain the integrity of the ex-vessel core catcher, however, it is required that the coolant be circulated at a rate along the inclined cooling channel sufficient to avoid CHF (Critical Heat Flux) on the heating surface of the cooling channel. In this study, post simulations of two phase natural circulation in the CEPECS have been performed to evaluate two phase flow characteristics and the natural circulation mass flow rate in the flow channel using the RELAP5/MOD3 computer code. Post simulations of two phase natural circulation in the CE-PECS have been conducted to evaluate two phase flow characteristics and the natural circulation mass flow rate in the flow channel using the RELAP5/MOD3 computer code. The RELAP5/MOD3 results have shown that the water circulation mass flow rate is approximately 8.7 kg/s in the base case

  8. Self-organizing maps applied to two-phase flow on natural circulation loop study

    International Nuclear Information System (INIS)

    Castro, Leonardo Ferreira

    2016-01-01

    Two-phase flow of liquid and gas is found in many closed circuits using natural circulation for cooling purposes. Natural circulation phenomenon is important on recent nuclear power plant projects for decay heat removal. The Natural Circulation Facility (Circuito de Circulacao Natural CCN) installed at Instituto de Pesquisas Energeticas e Nucleares, IPEN/CNEN, is an experimental circuit designed to provide thermal hydraulic data related to single and two-phase flow under natural circulation conditions. This periodic flow oscillation behavior can be observed thoroughly in this facility due its glass-made tubes transparency. The heat transfer estimation has been improved based on models that require precise prediction of pattern transitions of flow. This work presents experiments realized at CCN to visualize natural circulation cycles in order to classify two-phase flow patterns associated with phase transients and static instabilities of flow. Images are compared and clustered using Kohonen Self-organizing Maps (SOM's) applied on different digital image features. The Full Frame Discret Cosine Transform (FFDCT) coefficients were used as input for the classification task, enabling good results. FFDCT prototypes obtained can be associated to each flow pattern, enabling a better comprehension of each observed instability. A systematic test methodology was used to verify classifier robustness.

  9. Improving performance of two-phase natural circulation loops by reducing of entropy generation

    International Nuclear Information System (INIS)

    Goudarzi, N.; Talebi, S.

    2015-01-01

    This paper aims to investigate the effects of various parameters on stability behavior and entropy generation through a two-phase natural circulation loop. Two-phase natural circulation systems have low driving head and, consequently, low heat removal capability. To have a higher thermodynamic efficiency, in addition to the stability analysis, minimization of entropy generation by loop should be taken into account in the design of these systems. In the present study, to investigate the stability behavior, the non-linear method (known as the direct solution method or time domain method) which is able to simulate the uniform and non-uniform diameter loops, was applied. To best calculate entropy generation rates, the governing equations of the entropy generation were solved analytically. The effects of various parameters such as operating conditions and geometrical dimensions on the stability behavior and the entropy generation in the two-phase natural circulation loop were then analyzed. - Highlights: • Effects of all important parameters on entropy generation of a loop are studied. • The governing equations of the entropy generation are solved analytically. • Effects of all important parameters on stability of a loop are investigated. • Improvement of two-phase natural circulation loop is investigated.

  10. Classification of natural circulation two-phase flow patterns using fuzzy inference on image analysis

    International Nuclear Information System (INIS)

    Mesquita, R.N. de; Masotti, P.H.F.; Penha, R.M.L.; Andrade, D.A.; Sabundjian, G.; Torres, W.M.

    2012-01-01

    Highlights: ► A fuzzy classification system for two-phase flow instability patterns is developed. ► Flow patterns are classified based on images of natural circulation experiments. ► Fuzzy inference is optimized to use single grayscale profiles as input. - Abstract: Two-phase flow on natural circulation phenomenon has been an important theme on recent studies related to nuclear reactor designs. The accuracy of heat transfer estimation has been improved with new models that require precise prediction of pattern transitions of flow. In this work, visualization of natural circulation cycles is used to study two-phase flow patterns associated with phase transients and static instabilities of flow. A Fuzzy Flow-type Classification System (FFCS) was developed to classify these patterns based only on image extracted features. Image acquisition and temperature measurements were simultaneously done. Experiments in natural circulation facility were adjusted to generate a series of characteristic two-phase flow instability periodic cycles. The facility is composed of a loop of glass tubes, a heat source using electrical heaters, a cold source using a helicoidal heat exchanger, a visualization section and thermocouples positioned over different loop sections. The instability cyclic period is estimated based on temperature measurements associated with the detection of a flow transition image pattern. FFCS shows good results provided that adequate image acquisition parameters and pre-processing adjustments are used.

  11. RELAP5 simulation for one and two-phase natural circulation phenomenon

    International Nuclear Information System (INIS)

    Sabundjian, Gaiane; Andrade, Delvonei Alves de; Umbehaun, Pedro Ernesto; Torres, Walmir Maximo; Castro, Alfredo Jose Alvim de; Braz Filho, Francisco A.; Borges, Eduardo Madeira; Damy. Osvaldo Luiz Almeida; Torres, Eduardo

    2007-01-01

    The objective of this paper is to study the natural circulation phenomenon in one and two-phase regime. There has been a crescent interest in the scientific community in the study of the natural circulation. New generation of compact nuclear reactors uses the natural circulation for residual heat removal in case of accident or shutdown. For this study, the modeling and the simulation of the experimental circuit is performed with the RELAP5 code. The experimental circuit is mounted in the Chemical Engineering Department of the University of Sao Paulo. It is presented in this work the theoretical/experimental comparison for one and two-phase flow. These results will be stored in a database to validate RELAP5 calculations. This work was also used to training some users of RELAP5 from IEAv. (author)

  12. Analysis of Two Phase Natural Circulation Flow in the Cooling Channel of the PECS

    Energy Technology Data Exchange (ETDEWEB)

    Park, R. J; Ha, K. S; Rhee, B. W; Kim, H. Y [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    Decay heat and sensible heat of the relocated and spread corium are removed by the natural circulation flow at the bottom and side wall of the core catcher and the top water cooling of the corium. The coolant in the inclined channel absorbs the decay heat and sensible heat transferred from the corium through the structure of the core catcher body and flows up to the pool as a two phase mixture. On the other hand, some of the pool water will flow into the inlet of the downcomer piping, and will flow into the inclined cooling channel of the core catcher by gravity. As shown in Fig. 1, the engineered cooling channel is designed to provide effective long-term cooling and stabilization of the corium mixture in the core catcher body while facilitating steam venting in the PECS. To maintain the integrity of the ex-vessel core catcher, however, it is necessary that the coolant be sufficiently circulated along the inclined cooling channel to avoid CHF (Critical Heat Flux) on the heating surface of the cooling channel. For this reason, a verification experiment on the cooling capability of the EU-APR1400 core catcher has been performed in the CE (Cooling Experiment)-PECS facility at KAERI. Preliminary simulations of two-phase natural circulation in the CE-PECS were performed to predict two-phase flow characteristics and to determine the natural circulation mass flow rate in the flow channel. In this study, simulations of two-phase natural circulation in a real core catcher of the PECS have been performed to determine the natural circulation mass flow rate in the flow channel using the RELAP5/MOD3 computer code.

  13. Self-organizing maps applied to two-phase flow on natural circulation loop studies

    Energy Technology Data Exchange (ETDEWEB)

    Castro, Leonardo F.; Cunha, Kelly de P.; Andrade, Delvonei A.; Sabundjian, Gaiane; Torres, Walmir M.; Macedo, Luiz A.; Rocha, Marcelo da S.; Masotti, Paulo H.F.; Mesquita, Roberto N. de, E-mail: rnavarro@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    Two-phase flow of liquid and gas is found in many closed circuits using natural circulation for cooling purposes. Natural circulation phenomenon is important on recent nuclear power plant projects for heat removal on 'loss of pump power' or 'plant shutdown' accidents. The accuracy of heat transfer estimation has been improved based on models that require precise prediction of pattern transitions of flow. Self-Organizing Maps are trained to digital images acquired on natural circulation flow instabilities. This technique will allow the selection of the more important characteristics associated with each flow pattern, enabling a better comprehension of each observed instability. This periodic flow oscillation behavior can be observed thoroughly in this facility due its glass-made tubes transparency. The Natural Circulation Facility (Circuito de Circulacao Natural - CCN) installed at Instituto de Pesquisas Energeticas e Nucleares, IPEN/CNEN, is an experimental circuit designed to provide thermal hydraulic data related to one and two phase flow under natural circulation conditions. (author)

  14. Two-phase flow stability structure in a natural circulation system

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, Zhiwei [Nuclear Engineering Laboratory Zurich (Switzerland)

    1995-09-01

    The present study reports a numerical analysis of two-phase flow stability structures in a natural circulation system with two parallel, heated channels. The numerical model is derived, based on the Galerkin moving nodal method. This analysis is related to some design options applicable to integral heating reactors with a slightly-boiling operation mode, and is also of general interest to similar facilities. The options include: (1) Symmetric heating and throttling; (2) Asymmetric heating and symmetric throttling; (3) Asymmetric heating and throttling. The oscillation modes for these variants are discussed. Comparisons with the data from the INET two-phase flow stability experiment have qualitatively validated the present analysis.

  15. NOTICONA--a nonlinear time-domain computer code of two-phase natural circulation instability

    International Nuclear Information System (INIS)

    Su Guanghui; Guo Yujun; Zhang Jinling; Qiu Shuizheng; Jia Dounan; Yu Zhenwan

    1997-10-01

    A microcomputer code, NOTICONA, is developed, which is used for non-linear analysing the two-phase natural circulation systems. The mathematical model of the code includes point source neutron-kinetic model, the feedback of reactivity model, single-phase and two-phase flow model, heat transfer model in different conditions, associated model, etc. NOTICONA is compared with experiments, and its correctness and accuracy are proved. Using NOTICONA, the density wave oscillation (type I) of the 5 MW Test Heating Reactor are calculated, and the marginal stability boundary is obtained

  16. Two-phase flow patterns recognition and parameters estimation through natural circulation test loop image analysis

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, R.N.; Libardi, R.M.P.; Masotti, P.H.F.; Sabundjian, G.; Andrade, D.A.; Umbehaun, P.E.; Torres, W.M.; Conti, T.N.; Macedo, L.A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Nuclear Engineering Center], e-mail: rnavarro@ipen.br

    2009-07-01

    Visualization of natural circulation test loop cycles is used to study two-phase flow patterns associated with phase transients and static instabilities of flow. Experimental studies on natural circulation flow were originally related to accidents and transient simulations relative to nuclear reactor systems with light water refrigeration. In this regime, fluid circulation is mainly caused by a driving force ('thermal head') which arises from density differences due to temperature gradient. Natural circulation phenomenon has been important to provide residual heat removal in cases of 'loss of pump power' or plant shutdown in nuclear power plant accidents. The new generation of compact nuclear reactors includes natural circulation of their refrigerant fluid as a security mechanism in their projects. Two-phase flow patterns have been studied for many decades, and the related instabilities have been object of special attention recently. Experimental facility is an all glass-made cylindrical tubes loop which contains about twelve demineralized water liters, a heat source by an electrical resistor immersion heater controlled by a Variac, and a helicoidal heat exchanger working as cold source. Data is obtained through thermo-pairs distributed over the loop and CCD cameras. Artificial intelligence based algorithms are used to improve (bubble) border detection and patterns recognition, in order to estimate and characterize, phase transitions patterns and correlate them with the periodic static instability (chugging) cycle observed in this circuit. Most of initial results show good agreement with previous numerical studies in this same facility. (author)

  17. Two-phase flow patterns recognition and parameters estimation through natural circulation test loop image analysis

    International Nuclear Information System (INIS)

    Mesquita, R.N.; Libardi, R.M.P.; Masotti, P.H.F.; Sabundjian, G.; Andrade, D.A.; Umbehaun, P.E.; Torres, W.M.; Conti, T.N.; Macedo, L.A.

    2009-01-01

    Visualization of natural circulation test loop cycles is used to study two-phase flow patterns associated with phase transients and static instabilities of flow. Experimental studies on natural circulation flow were originally related to accidents and transient simulations relative to nuclear reactor systems with light water refrigeration. In this regime, fluid circulation is mainly caused by a driving force ('thermal head') which arises from density differences due to temperature gradient. Natural circulation phenomenon has been important to provide residual heat removal in cases of 'loss of pump power' or plant shutdown in nuclear power plant accidents. The new generation of compact nuclear reactors includes natural circulation of their refrigerant fluid as a security mechanism in their projects. Two-phase flow patterns have been studied for many decades, and the related instabilities have been object of special attention recently. Experimental facility is an all glass-made cylindrical tubes loop which contains about twelve demineralized water liters, a heat source by an electrical resistor immersion heater controlled by a Variac, and a helicoidal heat exchanger working as cold source. Data is obtained through thermo-pairs distributed over the loop and CCD cameras. Artificial intelligence based algorithms are used to improve (bubble) border detection and patterns recognition, in order to estimate and characterize, phase transitions patterns and correlate them with the periodic static instability (chugging) cycle observed in this circuit. Most of initial results show good agreement with previous numerical studies in this same facility. (author)

  18. Pressure drop in two-phase He I natural circulation loop at low vapour quality

    International Nuclear Information System (INIS)

    Baudouy, B.

    2003-01-01

    Steady state pressure drop in a two-phase He I natural circulation loop has been measured at atmospheric pressure. Results are obtained up to 0.2 exit vapor quality for a 14-mm diameter copper tube heated over a length of 1.2 m. Pressure drop assessment, done with the momentum balance equation including subcooling, reveals that the homogeneous model and Friedel's friction multiplier associated with Huq and Loth's void fraction correlations predict data within 15%. (author)

  19. Transient behavior of natural circulation for boiling two-phase flow, 2

    International Nuclear Information System (INIS)

    Aritomi, Masanori; Chiang, Jing-Hsien; Mori, Michitugu.

    1991-01-01

    In this set of experiments, natural circulation in boiling two-phase flow has been investigated for power transients, simulating the start-up process in a natural circulation BWR. This was done in order to understand the underlying mechanism of thermo-hydraulic instability which may appear during a start-up. In this paper, geysering is dealt with especially and the driving mechanism is clarified by investigating the stability related to effects of inlet velocity, subcooling, temperature in an outlet plenum and non-heated length between heated section and the outlet plenum. Furthermore, by considering these results and the operational experience in the Dodewaard reactor, recommendations on how the thermo-hydraulic instabilities can be prevented from occurring are proposed concerning a reactor configuration and start-up procedure for natural circulation BWRs. (author)

  20. Two-phase natural circulation experiments in a pressurized water loop with CANDU geometry

    International Nuclear Information System (INIS)

    Ardron, K.H.; Krishnan, V.S.; McGee, G.R.; Anderson, J.W.D.; Hawley, E.H.

    1984-07-01

    To provide information on two-phase natural circulation in a CANDU-type coolant circuit a series of tests has been performed in the RD-12 loop at the Whiteshell Nuclear Research Establishment. RD-12 is a 10-MPa pressurized-water loop containing two active boilers, two pumps, and two, or four, heated horizontal channels arranged in a symmetrical figure-of-eight configuration characteristic of the CANDU reactor primary heat-transport system. In the tests, single-phase natural circulation was established in the loop and void was introduced by controlled draining, with the surge tank (pressurizer) valved out of the system. The paper reviews the experimental results obtained and describes the evolution of natural circulation flow in particular cases as voidage is progressively increased. The stability behaviour is discussed briefly with reference to a simple stability model

  1. Analytical evaluation of two-phase natural circulation flow characteristics under external reactor vessel cooling

    International Nuclear Information System (INIS)

    Park, Jong Woon

    2009-01-01

    This work proposes an analytical method of evaluating the effects of design and operating parameters on the low-pressure two-phase natural circulation flow through the annular shaped gap at the reactor vessel exterior surface heated by corium (molten core) relocated to the reactor vessel lower plenum after loss of coolant accidents. A natural circulation flow velocity equation derived from steady-state mass, momentum, and energy conservation equations for homogeneous two-phase flow is numerically solved for the core melting conditions of the APR1400 reactor. The solution is compared with existing experiments which measured natural circulation flow through the annular gap slice model. Two kinds of parameters are considered for this analytical method. One is the thermal-hydraulic conditions such as thermal power of corium, pressure and inlet subcooling. The others are those for the thermal insulation system design for the purpose of providing natural circulation flow path outside the reactor vessel: inlet flow area, annular gap clearance and system resistance. A computer program NCIRC is developed for the numerical solution of the implicit flow velocity equation.

  2. Natural circulation in single-phase and two-phase flow

    International Nuclear Information System (INIS)

    Cheung, F.B.; El-Genk, M.S.

    1989-01-01

    Natural circulation usually arises in a closed loop between a heat source and a heat sink were the fluid motion is driven by density difference. It may also occur in enclosures or cavities where the flow is induced primarily by temperature or concentration gradients within the fluid. The subject has recently received special attention by the heat transfer and nuclear reactor safety communities because of it importance to the areas of energy extraction, decay, heat removal in nuclear reactors, solar and geothermal heating, and cooling of electronic equipment. Although many new results and physical insights have been gained of the various natural circulation phenomena, a number of critical issues remain unresolved. These include, for example, transition from laminar to turbulent flow, buoyancy-induced turbulent flow modeling, change of flow regimes, flow field visualization, variable property effects, and flow instability. This symposium volume contains papers presented in the Natural Circulation in Single-Phase and Two-Phase Flow session at the 1989 Winter Annual Meeting of ASME, by authors from different countries including the United States, Japan, Canada, and Brazil. The papers deal with experimental and theoretical studies as well as state-of-the-art reviews, covering a broad spectrum of topics in natural circulation including: variable-conductance thermosyphons, microelectronic chip cooling, natural circulation in anisotropic porous media and in cavities, heat transfer in flat plat solar collectors, shutdown heat removal in fast reactors, cooling of light-water and heavy-water reactors. The breadth of papers contained in this volume clearly reflect the importance of the current interest in natural circulation as a means for passive cooling and heating

  3. Pressure drop in two-phase He I natural circulation loop at low vapour quality

    Energy Technology Data Exchange (ETDEWEB)

    Baudouy, B

    2003-01-01

    Steady state pressure drop in a two-phase He I natural circulation loop has been measured at atmospheric pressure. Results are obtained up to 0.2 exit vapor quality for a 14-mm diameter copper tube heated over a length of 1.2 m. Pressure drop assessment, done with the momentum balance equation including subcooling, reveals that the homogeneous model and Friedel's friction multiplier associated with Huq and Loth's void fraction correlations predict data within 15%. (author)

  4. Low-pressure dynamics of a natural-circulation two-phase flow loop

    International Nuclear Information System (INIS)

    Manera, A.; Kruijf, W.J.M. de; Hartmann, H.; Mudde, R.F.; Hagen, T.H.J.J. van der

    2001-01-01

    Flashing induced oscillations in a natural circulation loop are studied as function of heating power and inlet subcooling in symmetrical and asymmetrical power conditions. To unveil the effects of power/velocity asymmetries on the two-phase flow stability at low power and low pressure conditions different signals at several locations in the loop are recorded. In particular a Laser Doppler Anemometry set-up is used to measure the velocity simultaneously in two parallel channels and a wire-mesh sensor is used to measure the 2D void fraction distribution in a section of the ascendant part of the loop. (orig.)

  5. Results of two-phase natural circulation in hot-leg U-bend simulation experiments

    International Nuclear Information System (INIS)

    Ishii, M.; Lee, S.Y.; Abou El-Seoud, S.

    1987-01-01

    In order to study the two-phase natural circulation and flow termination during a small break loss of coolant accident in LWR, simulation experiments have been performed using two different thermal-hydraulic loops. The main focus of the experiment was the two-phase flow behavior in the hot-leg U-bend typical of BandW LWR systems. The first group of experiments was carried out in the nitrogen gas-water adiabatic simulation loop and the second in the Freon 113 boiling and condensation loop. Both of the loops have been designed as a flow visualization facility and built according to the two-phase flow scaling criteria developed under this program. The nitrogen gas-water system has been used to isolate key hydrodynamic phenomena such as the phase distribution, relative velocity between phases, two-phase flow regimes and flow termination mechanisms, whereas the Freon loop has been used to study the effect of fluid properties, phase changes and coupling between hydrodynamic and heat transfer phenomena. Significantly different behaviors have been observed due to the non-equilibrium phase change phenomena such as the flashing and condensation in the Freon loop. The phenomena created much more unstable hydrodynamic conditions which lead to cyclic or oscillatory flow behaviors

  6. Simulation of the phenomenon of single-phase and two-phase natural circulation

    International Nuclear Information System (INIS)

    Castrillo, Lazara Silveira

    1998-02-01

    Natural convection phenomenon is often used to remove the residual heat from the surfaces of bodies where the heat is generated e.g. during accidents or transients of nuclear power plants. Experimental study of natural circulation can be done in small scale experimental circuits and the results can be extrapolated for larger operational facilities. The numerical analysis of transients can be carried out by using large computational codes that simulate the thermohydraulic behavior in such facilities. The computational code RELAP5/MOD2, (Reactor Excursion and Leak Analysis Program) was developed by U.S. Nuclear Regulatory Commissions's. Division of Reactor Safety Research with the objective of analysis of transients and postulated accidents in the light water reactor (LWR) systems, including small and large ruptures with loss of coolant accidents (LOCA's). The results obtained by the simulation of single-phase and two-phase natural circulation, using the RELAP5/MOD2, are presented in this work. The study was carried out using the experimental circuit built at the 'Departamento de Engenharia Quimica da Escola Politecnica da Universidade de Sao Paulo'. In the circuit, two experiments were carried out with different conditions of power and mass flow, obtaining a single-phase regime with a level of power of 4706 W and flow of 5.10 -5 m 3 /s (3 l/min) and a two-phase regime with a level of power of 6536 W and secondary flow 2,33.10 -5 m 3 /s (1,4 l/min). The study allowed tio evaluate the capacity of the code for representing such phenomena as well as comparing the transients obtained theoretically with the experimental results. The comparative analysis shows that the code represents fairly well the single-phase transient, but the results for two-phase transients, starting from the nodalization and calibration used for the case single-phase transient, did not reproduce faithfully some experimental results. (author)

  7. An experimental study of two-phase natural circulation in an adiabatic flow loop

    International Nuclear Information System (INIS)

    Tan, M.J.; Lambert, G.A.; Ishii, Mamoru.

    1988-01-01

    An experimental investigation was conducted to study the two-phase flow aspect of the phenomena of interruption and resumption of natural circulation, two-phase flow patterns and pattern transitions in the hot legs of B and W light water reactor systems. The test facility was a scaled adiabatic loop designed in accordance with the scaling criteria developed by Kocamustafaogullari and Ishii. The diameter and the height of the hot leg were 10 cm and 5.5 m, respectively; the working fluid pair was nitrogen-water. The effects of the thermal center in the steam generators, friction loss in the cold leg, and configuration of the inlet to the hot leg on the flow conditions in the hot leg were investigated by varying the water level in a gas separator, controlling the size of opening of a friction loss control valve, and using two inlet geometries. Methods for estimating the distribution parameter and the average drift velocity are proposed so that they may be used in the application of one-dimensional drift-flux model to the analysis of the interruption and resumption of natural circulation in a similar geometry. 7 refs., 17 figs., 4 tabs

  8. Comparison between RELAP5 versions for a two-phase natural circulation analysis

    Energy Technology Data Exchange (ETDEWEB)

    Braz Filho, Francisco A.; Ribeiro, Guilherme B.; Sabundjian, Gaianê; Caldeira, Alexandre D., E-mail: fbraz@ieav.cta.br, E-mail: gbribeiro@ieav.cta.br, E-mail: alexdc@ieav.cta.br, E-mail: gdjian@ipen.br [Instituto de Estudos Avançados (IEAv), São José dos Campos, SP (Brazil). Div. de Energia Nuclear; Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2017-11-01

    RELAP5 is one of the most used numerical tools to predict thermal-hydraulic and neutronic phenomena in nuclear reactors. RELAP5-3D is the latest version of this software family, but RELAP5-mod3 is still widely used in Brazilian research institutes and it is also used as benchmark for several nuclear applications. Among these applications, the use of passive heat transfer mechanisms, such as natural circulation, has drawn attention of several studies, especially after the Fukushima-Daiichi accident. Considering this aforementioned aspect, this study proposes a comparison of RELAP5-3D and RELAP5-mod3 versions, focusing on a two-phase natural circulation loop. For comparison purposes, an experimental data set is part of the analysis. Results showed that during the single-phase regime, the temperature difference between versions is negligible. However, when the two-phase flow regime takes place, different wavelengths and amplitudes of flow instabilities were obtained for each version. When compared to the experimental data set, the RELAP5-3D version provided the best prediction results. (author)

  9. Steady state flow analysis of two-phase natural circulation in multiple parallel channel loop

    International Nuclear Information System (INIS)

    Bhusare, V.H.; Bagul, R.K.; Joshi, J.B.; Nayak, A.K.; Kannan, Umasankari; Pilkhwal, D.S.; Vijayan, P.K.

    2016-01-01

    Highlights: • Liquid circulation velocity increases with increasing superficial gas velocity. • Total two-phase pressure drop decreases with increasing superficial gas velocity. • Channels with larger driving force have maximum circulation velocities. • Good agreement between experimental and model predictions. - Abstract: In this work, steady state flow analysis has been carried out experimentally in order to estimate the liquid circulation velocities and two-phase pressure drop in air–water multichannel circulating loop. Experiments were performed in 15 channel circulating loop. Single phase and two-phase pressure drops in the channels have been measured experimentally and have been compared with theoretical model of Joshi et al. (1990). Experimental measurements show good agreement with model.

  10. Two-phase natural circulation experiments in a pressurized water loop with CANDU geometry

    Energy Technology Data Exchange (ETDEWEB)

    Ardron, K.H.; Krishnan, V.S.; McGee, G.R.; Anderson, J.W.D.; Hawley, E.H.

    1984-07-01

    A series of tests has been performed in the RD-12 loop, a 10-MPa pressurized-water loop containing two active boilers, two pumps, and two, or four, heated horizontal channels arranged in a symmetrical figure-of-eight configuration characteristic of the CANDU reactor primary heat-transport system. In the tests, single-phase natural circulation was established in the loop and void was introduced by controlled draining, with the surge tank (pressurizer) valved out of the system. Results indicate that a stable, two-phase, natural circulation flow can usually be established. However, as the void fraction in the loop is increased, large-amplitude flow oscillations can occur. The initial flow oscillations in the two halves of the loop are usually very nearly 180/sup 0/ out-of-phase. However, as the loop inventory is further decreased, an in-phase oscillation component is observed. In tests with two parallel, heated channels in each half-loop, oscillations associated with mass transfer between the channel pairs are also observed. Although flow oscillations can lead to intermittent dryout of the upper elements of the heater-rod assemblies in the horizontal channels, natural circulation cooling appears to be effective until about 50% of the loop inventory is drained; sustained flow stratification then occurs in the heated channels, leading to heater temperature excursions. The paper reviews the experimental results obtained and describes the evolution of natural circulation flow in particular cases as voidage is progressively increased. The stability behavior is discussed briefly with reference to a simple stability model.

  11. An experimental study on two-phase flow pattern in low pressure natural circulation system

    International Nuclear Information System (INIS)

    Wu Shaorong; Han Bing; Zhou Lei; Zhang Youjie; Jiang Shengyao; Wu Xinxin

    1991-10-01

    An experimental study on two-phase flow pattern in the riser of low pressure natural circulation system was performed. The local differential pressure signal was analysed for flow pattern. It is considered that Sr f·d/v can be used to distinguish different flow patterns and it has clear and definite physical meaning. Flow patterns at different inlet temperature with different system pressures (1.5 MPa, 0.24 MPa and 0.1 MPa) are described. It is considered that the flow pattern is only bubble flow without flow pattern change during the period of low quality density-wave instability at 1.5 MPa. There is no density-wave oscillation in the system, when flow pattern is in bubble-intermittent transition area. The effect of flash vaporization on stability at low pressure is discussed

  12. Scaling of the steady state and stability behaviour of single and two-phase natural circulation systems

    International Nuclear Information System (INIS)

    Vijayan, P.K.; Nayak, A.K.; Bade, M.H.; Kumar, N.; Saha, D.; Sinha, R.K.

    2002-01-01

    Scaling methods for both single-phase and two-phase natural circulation systems have been presented. For single-phase systems, simulation of the steady state flow can be achieved by preserving just one nondimensional parameter. For uniform diameter two-phase systems also, it is possible to simulate the steady state behaviour with just one non-dimensional parameter. Simulation of the stability behaviour requires geometric similarity in addition to the similarity of the physical parameters appearing in the governing equations. The scaling laws proposed have been tested with experimental data in case of single-phase natural circulation. (author)

  13. Chaotic oscillations in a low pressure two-phase natural circulation loop under low power and high inlet subcooling conditions

    International Nuclear Information System (INIS)

    Wu, C.Y.; Wang, S.B.; Pan, C.

    1996-01-01

    The oscillation characteristics of a low pressure two-phase natural circulation loop have been investigated experimentally in this study. Experimental results indicate that the characteristics of the thermal hydraulic oscillations can be periodic, with 2-5 fundamental frequencies, or chaotic, depending on the heating power and inlet subcooling. The number of fundamental frequencies of oscillation increases if the inlet subcooling is increased at a given heating power or the heating power is decreased at a given inlet subcooling; chaotic oscillations appear if the inlet subcooling is further increased and/or the heating power is further decreased. A map of the oscillation characteristics is thus established. The change in oscillation characteristics is evident from the time evolution and power spectrum of a thermal hydraulic parameter and the phase portraits of two thermal hydraulic parameters. These results reveal that a strange attractor exists in a low pressure two-phase natural circulation loop with low power and very high inlet subcooling. (orig.)

  14. Experimental study of natural two-phase flow circulation using a visualization technique

    International Nuclear Information System (INIS)

    Vinhas, Pedro A.M.; Su, Jian

    2013-01-01

    This paper presents an experimental study of natural two-phase flow in a circuit that simulates, on a smaller scale, a typical residual heat removal system of passive reactors APWR (Advanced Pressurized Water Reactor). The circuit was formed by a heater, a heat exchanger and piping. The experimental study was the application of a visualization technique, using a high speed camera, for measuring the size and speed of vapor bubbles generated in the heater with different power heating. The camera was positioned in the central region of the pipe connecting the heater to the heat exchanger, where there is a clear passage. The flow of images were processed and analyzed using commercial software that allowed the determination of the length and velocity of the bubbles. The results were then compared with correlations available in literature

  15. Hydrodynamic Instability and Dynamic Burnout in Natural Circulation Two-Phase Flow. An Experimental and Theoretical Study

    Energy Technology Data Exchange (ETDEWEB)

    Becker, Kurt M; Jahnberg, S; Haga, I; Hansson, P T; Mathisen, R P

    1964-09-15

    A theoretical model for predicting the threshold of instability for two phase flow in a natural circulation loop is presented. The model calculates the flow transient caused by a step disturbance of the heat input, and is based upon the conservation laws of mass, momentum and energy in one dimensional form. Empirical correlations are used in the model for estimating the void fractions and the two-phase flow pressure drops. The equations are solved numerically in a finite difference approximation coded for a digital computer. An experimental study of the hydrodynamic instability and dynamic burnout in two-phase flow has been performed in a natural circulation loop in the pressure range from 10 to 70 atg. The test sections were round ducts of 20, 30 and 36 mm inner diameter and 4890 mm heated length. The experimental results showed that within the ranges tested, the stability of the flow increases with increasing pressure and increasing throttling before the test section, but decreases with increasing Inlet subcooling and increasing throttling after the test section. Comparing the natural circulation burnout steam qualities with corresponding forced circulation data shoved that the former data were low by a factor up to 2.5. However, by applying inlet throttling of the flow the burnout values approached and finally coincided with the forced circulation data. The present experimental results as well as data available from other sources have been compared with the stability thresholds obtained with the theoretical model. The comparisons included circular, annular and rod cluster geometries, and the agreement between the experimental and theoretical stability limits was good. Finally the application of the experimental and theoretical results on the assessment of boiling heavy water reactor design is discussed.

  16. Hydrodynamic Instability and Dynamic Burnout in Natural Circulation Two-Phase Flow. An Experimental and Theoretical Study

    International Nuclear Information System (INIS)

    Becker, Kurt M.; Jahnberg, S.; Haga, I.; Hansson, P.T.; Mathisen, R.P.

    1964-09-01

    A theoretical model for predicting the threshold of instability for two phase flow in a natural circulation loop is presented. The model calculates the flow transient caused by a step disturbance of the heat input, and is based upon the conservation laws of mass, momentum and energy in one dimensional form. Empirical correlations are used in the model for estimating the void fractions and the two-phase flow pressure drops. The equations are solved numerically in a finite difference approximation coded for a digital computer. An experimental study of the hydrodynamic instability and dynamic burnout in two-phase flow has been performed in a natural circulation loop in the pressure range from 10 to 70 atg. The test sections were round ducts of 20, 30 and 36 mm inner diameter and 4890 mm heated length. The experimental results showed that within the ranges tested, the stability of the flow increases with increasing pressure and increasing throttling before the test section, but decreases with increasing Inlet subcooling and increasing throttling after the test section. Comparing the natural circulation burnout steam qualities with corresponding forced circulation data shoved that the former data were low by a factor up to 2.5. However, by applying inlet throttling of the flow the burnout values approached and finally coincided with the forced circulation data. The present experimental results as well as data available from other sources have been compared with the stability thresholds obtained with the theoretical model. The comparisons included circular, annular and rod cluster geometries, and the agreement between the experimental and theoretical stability limits was good. Finally the application of the experimental and theoretical results on the assessment of boiling heavy water reactor design is discussed

  17. Theoretical model of two-phase drift flow on natural circulation

    International Nuclear Information System (INIS)

    Yang Xingtuan; Jiang Shengyao; Zhang Youjie

    2002-01-01

    Some expressions, such as sub-cooled boiling in the heating section, condensation near the riser inlet, flashing in the riser, and pressure balance in the steam-space, have been theoretically deduced from the physical model of 5 MW heating reactor test loop. The thermodynamics un-equilibrium etc have been considered too. A entire drift model with four equations has been formed, which can be applied to natural circulation system with low pressure and low steam quality. By means of introducing the concept of condensation layer, condensing of bubbles in the sub-cooled liquid has been formulated for the first time. The restrictive equations of the steam space pressure and liquid level have been offered. The equations can be solved by means of integral method, then by using Rung-Kutta-Verner method the final results is obtained

  18. Similarity analysis and scaling criteria for LWRs under single-phase and two-phase natural circulation

    International Nuclear Information System (INIS)

    Ishii, M.; Kataoka, I.

    1983-03-01

    Scaling criteria for a natural circulation loop under single phase and two-phase flow conditions have been derived. For a single phase case the continuity, integral momentum, and energy equations in one-dimensional area average forms have been used. From this, the geometrical similarity groups, friction number, Richardson number, characteristic time constant ratio, Biot number, and heat source number are obtained. The Biot number involves the heat transfer coefficient which may cause some difficulties in simulating the turbulent flow regime. For a two-phase flow case, the similarity groups obtained from a perturbation analysis based on the one-dimensional drift-flux model have been used. The physical significance of the phase change number, subcooling number, drift-flux number, friction number are discussed and conditions imposed by these groups are evaluated. In the two-phase flow case, the critical heat flux is one of the most important transients which should be simulated in a scale model. The above results are applied to the LOFT facility in case of a natural circulation simulation. Some preliminary conclusions on the feasibility of the facility have been obtained

  19. Similarity analysis and scaling criteria for LWRs under single-phase and two-phase natural circulation

    Energy Technology Data Exchange (ETDEWEB)

    Ishii, M.; Kataoka, I.

    1983-03-01

    Scaling criteria for a natural circulation loop under single phase and two-phase flow conditions have been derived. For a single phase case the continuity, integral momentum, and energy equations in one-dimensional area average forms have been used. From this, the geometrical similarity groups, friction number, Richardson number, characteristic time constant ratio, Biot number, and heat source number are obtained. The Biot number involves the heat transfer coefficient which may cause some difficulties in simulating the turbulent flow regime. For a two-phase flow case, the similarity groups obtained from a perturbation analysis based on the one-dimensional drift-flux model have been used. The physical significance of the phase change number, subcooling number, drift-flux number, friction number are discussed and conditions imposed by these groups are evaluated. In the two-phase flow case, the critical heat flux is one of the most important transients which should be simulated in a scale model. The above results are applied to the LOFT facility in case of a natural circulation simulation. Some preliminary conclusions on the feasibility of the facility have been obtained.

  20. Assessment of the MARS Code Using the Two-Phase Natural Circulation Experiments at a Core Catcher Test Facility

    Directory of Open Access Journals (Sweden)

    Dong Hun Lee

    2017-01-01

    Full Text Available A core catcher has been developed to maintain the integrity of nuclear reactor containment from molten corium during a severe accident. It uses a two-phase natural circulation for cooling molten corium. Flow in a typical core catcher is unique because (i it has an inclined cooling channel with downwards-facing heating surface, of which flow processes are not fully exploited, (ii it is usually exposed to a low-pressure condition, where phase change causes dramatic changes in the flow, and (iii the effects of a multidimensional flow are very large in the upper part of the core catcher. These features make computational analysis more difficult. In this study, the MARS code is assessed using the two-phase natural circulation experiments that had been conducted at the CE-PECS facility to verify the cooling performance of a core catcher. The code is a system-scale thermal-hydraulic (TH code and has a multidimensional TH component. The facility was modeled by using both one- and three-dimensional components. Six experiments at the facility were selected to investigate the parametric effects of heat flux, pressure, and form loss. The results show that MARS can predict the two-phase flow at the facility reasonably well. However, some limitations are obviously revealed.

  1. Development of TPNCIRC code for Evaluation of Two-Phase Natural Circulation Flow Performance under External Reactor Vessel Cooling Conditions

    International Nuclear Information System (INIS)

    Choi, A-Reum; Song, Hyuk-Jin; Park, Jong-Woon

    2015-01-01

    During a severe accident, corium is relocated to the lower head of the nuclear reactor pressure vessel (RPV). Design concept of retaining the corium inside a nuclear reactor pressure vessel (RPV) through external cooling under hypothetical core melting accidents is called external reactor vessel cooling (ERVC). In this respect, validated two-phase natural circulation flow (TPNC) model is necessary to determine the adequacy of the ERVC design and operating conditions such as inlet area, form losses, gap distance, riser length and coolant conditions. The most important model generally characterizing the TPNC are void fraction and two-phase friction factors. Typical experimental and analytical studies to be referred to on two-phase circulation flow characteristics are those by Reyes, Gartia et al. based on Vijayan et al., Nayak et al. and Dubey et al. In the present paper, two-phase natural circulation (TPNC) flow characteristics under external reactor vessel cooling (ERVC) conditions are studied using two existing TPNC flow models of Reyes and Gartia et al. incorporating more improved void fraction and two-phase friction models. These models and correlations are integrated into a computer program, TPNCIRC, which can handle candidate ERVC design parameters, such as inlet, riser and downcomer flow lengths and areas, gap size between reactor vessel and surrounding insulations, minor loss factors and operating parameters of decay power, pressure and subcooling. Accuracy of the TPNCIRC program is investigated with respect to the flow rate and void fractions for existing measured data from a general experiment and ULPU specifically designed for the AP1000 in-vessel retention. Also, the effect of some important design parameters are examined for the experimental and plant conditions. Using the flow models and correlations are integrated into a computer program, TPNCIRC, a number of correlations have been examined. This seems coming from the differences of void fractions

  2. Evaluations of two-phase natural circulation flow induced in the reactor vessel annular gap under ERVC conditions

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Kwang Soon, E-mail: tomo@kaeri.re.kr [Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Cheung, Fan-Bill [The Pennsylvania State University, University Park, PA 16802 (United States); Park, Rae Joon; Kim, Sang Baik [Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)

    2012-12-15

    Highlights: Black-Right-Pointing-Pointer Two-phase natural circulation flow induced in insulation gap was investigated. Black-Right-Pointing-Pointer Half-scaled non-heating experiments were performed to evaluate flow behavior. Black-Right-Pointing-Pointer The loop-integrated momentum equation was formulated and solved asymptotically. Black-Right-Pointing-Pointer First-order approximate solution was obtained and agreed with experimental data. - Abstract: The process of two-phase natural circulation flow induced in the annular gap between the reactor vessel and the insulation under external reactor vessel cooling conditions was investigated experimentally and analytically in this study. HERMES-HALF experiments were performed to observe and quantify the induced two-phase natural circulation flow in the annular gap. A half-scaled non-heating experimental facility was designed by utilizing the results of a scaling analysis to simulate the APR1400 reactor and its insulation system. The behavior of the boiling-induced two-phase natural circulation flow in the annular gap was observed, and the liquid mass flow rates driven by the natural circulation loop and the void fraction distribution were measured. Direct flow visualization revealed that choking would occur under certain flow conditions in the minimum gap region near the shear keys. Specifically, large recirculation flows were observed in the minimum gap region for large air injection rates and small outlet areas. Under such conditions, the injected air could not pass through the minimum gap region, resulting in the occurrence of choking near the minimum gap with a periodical air back flow being generated. Therefore, a design modification of the minimum gap region needs to be done to facilitate steam venting and to prevent choking from occurring. To complement the HERMES-HALF experimental effort, an analytical study of the dependence of the induced natural circulation mass flow rate on the inlet area and the

  3. Experimental study on saturated boiling of two phase natural circulation under low pressure in narrow rectangular channels

    Energy Technology Data Exchange (ETDEWEB)

    Li, Zi-chao; Qi, Shi; Zhou, Tao; Li, Bing; Shahzad, Muhammad Ali [North China Electric Power Univ., Beijing (China). School of Nuclear Science and Engineering; Beijing Key Laboratory of Passive Safety Technology for Nuclear Energy, Beijing (China); Huang, Yan-ping [Nuclear Reactor Thermal Hydraulics Technology, Chengdu (China). CNNC Key Lab.

    2017-12-15

    Saturated boiling of two-phase natural circulation has been experimentally investigated based on a natural circulation device with narrow rectangular channels. When heating power reaches a certain range, it is possible to observe the phenomenon of saturated boiling and flow pattern transition in the system. The results show the heat transfer coefficient of saturated boiling decreases with the increasing of pressure, heating power and size of narrow rectangle channels. The buoyancy force causing mixed convection decreases the heat transfer coefficient. Finally, a dimensionless number is introduced, which reflects length to width ratio of rectangular narrow section and Rayleigh number, in order to revise the presented correlation. All errors fall within the range of ±15%.

  4. Flow visualization of bubble behavior under two-phase natural circulation flow conditions using high speed digital camera

    Energy Technology Data Exchange (ETDEWEB)

    Lemos, Wanderley F.; Su, Jian, E-mail: wlemos@con.ufrj.br, E-mail: sujian@lasme.coppe.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Faccini, Jose L.H., E-mail: faccini@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Lab. de Termo-Hidraulica Experimental

    2013-07-01

    The The present work aims at identifying flow patterns and measuring interfacial parameters in two-phase natural circulation by using visualization technique with high-speed digital camera. The experiments were conducted in the Natural Circulation Circuit (CCN), installed at Nuclear Engineering Institute/CNEN. The thermo-hydraulic circuit comprises heater, heat exchanger, expansion tank, the pressure relief valve and pipes to interconnect the components. A glass tube is installed at the midpoint of the riser connected to the heater outlet. The natural circulation circuit is complemented by acquisition system of values of temperatures, flow and graphic interface. The instrumentation has thermocouples, volumetric flow meter, rotameter and high-speed digital camera. The experimental study is performed through analysis of information from measurements of temperatures at strategic points along the hydraulic circuit, besides natural circulation flow rates. The comparisons between analytical and experimental values are validated by viewing, recording and processing of the images for the flows patterns. Variables involved in the process of identification of flow regimes, dimensionless parameters, the phase velocity of the flow, initial boiling point, the phenomenon of 'flashing' pre-slug flow type were obtained experimentally. (author)

  5. Flow visualization of bubble behavior under two-phase natural circulation flow conditions using high speed digital camera

    International Nuclear Information System (INIS)

    Lemos, Wanderley F.; Su, Jian; Faccini, Jose L.H.

    2013-01-01

    The The present work aims at identifying flow patterns and measuring interfacial parameters in two-phase natural circulation by using visualization technique with high-speed digital camera. The experiments were conducted in the Natural Circulation Circuit (CCN), installed at Nuclear Engineering Institute/CNEN. The thermo-hydraulic circuit comprises heater, heat exchanger, expansion tank, the pressure relief valve and pipes to interconnect the components. A glass tube is installed at the midpoint of the riser connected to the heater outlet. The natural circulation circuit is complemented by acquisition system of values of temperatures, flow and graphic interface. The instrumentation has thermocouples, volumetric flow meter, rotameter and high-speed digital camera. The experimental study is performed through analysis of information from measurements of temperatures at strategic points along the hydraulic circuit, besides natural circulation flow rates. The comparisons between analytical and experimental values are validated by viewing, recording and processing of the images for the flows patterns. Variables involved in the process of identification of flow regimes, dimensionless parameters, the phase velocity of the flow, initial boiling point, the phenomenon of 'flashing' pre-slug flow type were obtained experimentally. (author)

  6. Study on liquid-metal MHD power generation system with two-phase natural circulation. Applicability to fast reactor conditions

    International Nuclear Information System (INIS)

    Saito, Masaki

    2001-03-01

    Feasibility study of the liquid-metal MHD power generation system combined with the high-density two-phase natural circulation has been performed for the applicability to the simple, autonomic energy conversion system of the liquid-metal cooled fast reactor. The present system has many promising aspects not only in the energy conversion process, but also in safety and economical improvements of the liquid-metal cooled fast reactor. In the previous report, as the first step of the feasibility study, the cycle analyses were performed to examine the effects of the main system parameters on the fundamental characteristics of the system. It was found that the cycle efficiency of the present system is enough competitive with that of the conventional steam turbine system. It was also found that the cycle efficiency depends strongly on the gas-liquid slip ratio in the two-phase flow channel. However, it is very difficult to estimate the gas-liquid slip ratio theoretically, especially in the heavy liquid metal two-phase natural circulation. For example, the effects of MHD load on the two-phase flow characteristics, such as the void fraction and gas-liquid slip ratio are not known well. In the present study, therefore, as the second step of the feasibility study, a series of the experiments were performed to investigate, especially, the effect of MHD load at the single-phase shown-comer flow channel on the characteristics of the two-phase natural circulation. In the first series of the experiments, Woods-metal (Density: 9517 Kg/m 3 ) and nitrogen gas were chosen as the two-phase working fluids. The MHD pressure drop was simulated by the ball valve. The experiments with water and nitrogen gas were also performed to check the effects of the physical properties. From the present experiments, it is found that the average void fraction in the two-phase flow channel is determined by the force balance between the MHD pressure drop, frictional and pressure losses in the tube, and

  7. Experimental study of gas–liquid two-phase flow through packed bed under natural circulation conditions

    International Nuclear Information System (INIS)

    Chen, Shao-Wen; Miwa, Shuichiro; Griffiths, Matt

    2016-01-01

    Dry-out phenomena in packed beds or porous media may cause a significant digression of cooling/reaction performance in heat transfer/chemical reactor systems. One of the phenomena responsible for the dry-out in packed beds is known as the counter-current flow limitation (CCFL). In order to investigate the CCFL phenomena induced by gas–liquid two-phase flow in packed beds inside a pool, a natural circulation packed bed test facility was designed and constructed. A total of 27 experimental conditions covering various packing media sizes (sphere diameters: 3.0, 6.4 and 9.5 mm), packed bed heights (15, 35 and 50 cm) and water level heights (1.0, 1.5 and 2.0 m) were tested to examine the CCFL criteria with adiabatic air–water two-phase flow under natural circulation conditions. Both CCFL and flow reversal phenomena were observed, and the experimental data including instantaneous and time-averaged void fraction, differential pressure and superficial gas–liquid velocities were collected. The CCFL criteria were determined when periodical oscillations of void fraction and differential pressure appear. In addition, the Wallis correlation for CCFL was utilized for data analysis, and the Wallis coefficient, C, was determined experimentally from the packed bed CCFL tests. Compared to the existing data-sets in literature, the higher C values obtained in the present experiment suggest a possibly higher dry-out heat flux for natural circulation debris systems, which may be due to the water supply from both top and bottom surfaces of the packed beds. Considering the effects of bed height and hydraulic diameter of the packing media, a newly developed model for the Wallis coefficient, C, under natural circulation CCFL is presented. The present model can predict the experimental data with an averaged absolute error of ±7.9%. (author)

  8. Study on liquid-metal MHD power generation system with two-phase natural circulation. Applicability to fast reactor conditions

    International Nuclear Information System (INIS)

    Saito, Masaki

    2000-03-01

    Feasibility study of the liquid-metal MHD power generation system combined with the high-density two-phase natural circulation has been performed for the applicability to the simple, autonomic energy conversion system of the liquid-metal cooled fast reactor. The present system has many promising aspects not only in the energy conversion process, but also in safety and economical improvements of the liquid-metal cooled fast reactor. For example, the high cycle efficiency can be expected because of the similarity of the present cycle to the Ericsson cycle. Sodium-Water Interaction problem can be excluded by proper combination of the working fluids. As the economical feature, the present system is so simple that the liquid-metal main circular pump, the steam turbine generator, and even the steam generator can be excluded if the thermodynamic working fluid is injected directly into the high temperature liquid metal MHD working fluid. In addition, the present system has the potential to be applied to various heat sources including solar energy because of the high flexibility of the operation temperature. In the present paper, as the first step of the feasibility study, the cycle analyses were performed to examine the effects of the main system parameters on the fundamental characteristics of the system. It is found that the cycle efficiency of the present system is enough competitive with that of the conventional steam turbine system. It is, however, found that the cycle efficiency depends strongly on the gas-liquid slip ratio in the two-phase flow channel. As the conclusions, it is recommended to perform experimental study to obtain the fundamental data, such as the gas-liquid slip ratio in the high-density liquid-metal two-phase natural circulation. (author)

  9. The research on static bifurcation characteristics and parametric effect of two-phase natural circulation and passive system

    International Nuclear Information System (INIS)

    Xu Jijun; Yang Yanhua; Kuang Bo; Yao Wei; Zhang Ronghua; Tong Lili

    2001-01-01

    The formation of dissipative structures has long been known to occur in hydrodynamics. The two-phase natural circulation and passive system (TPNCPS) instability is a dissipative structure problem in multiphase hydrodynamics. The spectrum of the static bifurcation solutions (SBS) of TPNCPS through the variation of a parameter (one or more) has been derived in terms of Bifurcation Theory and DERPAR Numerical Method. Based on the appearance of Thermal-Siphon Hysteresis, the transport heat capability, static excursion criterion, stationary margin, transport heat capability of specific mass flow-rate and the disappear of bifurcation-the transition of single-valued region with the change of parameter have been defined. Such phenomena are the problems of describing self-organization, i.e. detailed study of stationary and/or time dependent status evolving with changes of characteristic parameter. A comparison between computational curves and low-pressure experimental data shows the tendency of evolutionary processes compatibly. The further tests are needed

  10. Analytical study of flow instability behaviour in a boiling two-phase natural circulation loop under low quality conditions

    International Nuclear Information System (INIS)

    Nayak, A.K.; Kumar, N.; Vijayan, P.K.; Saha, D.; Sinha, R.K.

    2002-01-01

    Analytical investigations have been carried out to study the flow instability behaviour in a boiling two-phase natural circulation loop under low quality conditions. For this purpose, the computer code TINFLO-S has been developed. The code solves the conservation equations of mass, momentum and energy and equation of state for homogeneous equilibrium twophase flow using linear analytical technique. The results of the code have been validated with the experimental data of the loop for both the steady state and stability. The study reveals that the stability behaviour of low quality flow oscillations is different from that of the high quality flow oscillations. The instability reduces with increase in power and throttling at the inlet of the heater. The instability first increases and then reduces with increase in pressure at any subcooling. The effects of diameter of riser pipe, heater and the height of the riser on this instability are also investigated. (orig.) [de

  11. Examination of transient characteristics of two-phase natural circulation within a Freon-113 boiling/condensation loop

    International Nuclear Information System (INIS)

    Tanimoto, K.; Ishii, M.

    1998-01-01

    Transient characteristics of two-phase natural circulation within a Freon-113 loop with a large condenser have been examined mainly focused on the flashing phenomenon. General behavior was described and parametric studies were performed. The items observed were the period and duration of flashing, peak flow rate, amount of flow carryover per flashing, lowest-peak liquid level within the condenser, and the peak void distribution in the riser section. The parameters considered were the heater power input, valve friction at the heater inlet (simulating the loopwise friction), condenser cooling, degree of subcooling at the heater inlet, and the heat loss to the surroundings. As a whole, the heater power input, valve friction, and the rate of condenser cooling played important roles in flashing while the other effects being marginal. In general, the flow appeared to be more unstable with the larger condensing surface which causes the condensation-induced flashing. (orig.)

  12. An experimental study of the heat transfer performance of a rectangular two-phase natural circulation loop

    International Nuclear Information System (INIS)

    Chen, K.S.; Chen, Y.Y.; Tsai, S.T.

    1990-01-01

    An experimental study is presented for the heat transfer performance of a rectangular, two-phase, natural-circulation loop with water-steam as the working fluid. Local temperature measurements of the core fluid and the wall were made, and the overall heat transfer coefficients of the evaporator, the condenser, and the loop system were obtained and correlated in terms of the fluid properties, heat flux conditions, and the liquid charge level. An overheat phenomenon at very low charge level was also observed. Result of a preliminary analysis shows that if the liquid charge level is below the fractional volume of the connecting tube between the condenser and the evaporator, an overheat phenomenon will occur

  13. Single and two-phase natural circulation in Westinghouse pressurized water reactor simulators: Phenomena, analysis and scaling

    International Nuclear Information System (INIS)

    Schultz, R.R.; Chapman, J.C.; Kukita, Y.; Motley, F.E.; Stumpf, H.; Chen, Y.S.; Tasaka, K.

    1987-01-01

    Natural circulation data obtained in the 1/48 scale W four loop PWR simulator - the Large Scale Test Facility (LSTF) are discussed and summarized. Core cooling modes, the primary fluid state, the primary loop mass flow and localized natural circulation phenomena occurring in the steam generator are presented. TRAC-PF1 LSTF model (using both a 1 U-tube and a 3 U-tube steam generator model) analyses of the LSTF natural circulation data including the SG recirculation patterns are presented and compared to the data. The LSTF data are then compared to similar natural circulation data obtained in the Primarkreislaufe (PKL) and the Semiscale facilities. Based on the 1/48 to 1/1705 scaling range which exists between the facilities, the implications of these data towrard natural circulation behavior in commercial plants are briefly discussed

  14. Self-organizing maps applied to two-phase flow on natural circulation loop study; Aplicacao de mapas auto-organizaveis na classificacao de padroes de escoamento bifasico

    Energy Technology Data Exchange (ETDEWEB)

    Castro, Leonardo Ferreira

    2016-11-01

    Two-phase flow of liquid and gas is found in many closed circuits using natural circulation for cooling purposes. Natural circulation phenomenon is important on recent nuclear power plant projects for decay heat removal. The Natural Circulation Facility (Circuito de Circulacao Natural CCN) installed at Instituto de Pesquisas Energeticas e Nucleares, IPEN/CNEN, is an experimental circuit designed to provide thermal hydraulic data related to single and two-phase flow under natural circulation conditions. This periodic flow oscillation behavior can be observed thoroughly in this facility due its glass-made tubes transparency. The heat transfer estimation has been improved based on models that require precise prediction of pattern transitions of flow. This work presents experiments realized at CCN to visualize natural circulation cycles in order to classify two-phase flow patterns associated with phase transients and static instabilities of flow. Images are compared and clustered using Kohonen Self-organizing Maps (SOM's) applied on different digital image features. The Full Frame Discret Cosine Transform (FFDCT) coefficients were used as input for the classification task, enabling good results. FFDCT prototypes obtained can be associated to each flow pattern, enabling a better comprehension of each observed instability. A systematic test methodology was used to verify classifier robustness.

  15. An experimental study on the two-phase natural circulation flow through the gap between reactor vessel and insulation under ERVC

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Kwang-Soon; Park, Rae-Joon; Cho, Young-Ro; Kim, Sang-Baik; Kim, Hwan-Yeol; Kim, Hee-dong

    2005-04-01

    As part of a study on a two-phase natural circulation flow between the outer reactor vessel and the insulation material in the reactor cavity under an external reactor vessel cooling of APR1400, T-HERMES-SMALL and HERMES-HALF experiments have been performed. For the T-HERMES-SMALL experiments, an 1/21.6 scaled experimental facility was prepared utilizing the results of a scaling analysis to simulate the APR1400 reactor and insulation system. The liquid mass flow rates driven by natural circulation loop were measured by varying the wall heat flux, upper outlet area and configuration, and water head condition. The experimental data were also compared with numerical ones given by simple loop analysis. And non-heating small-scaled experiments have also been performed to certify the hydraulic similarity of the heating experiments by injecting air equivalent to the steam generated in the heating experimental condition. The HERMES-HALF experiment is a half-scaled / non-heating experimental study on the two-phase natural circulation through the annular gap between the reactor vessel and the insulation. The behaviors of the two-phase natural circulation flow in the insulation gap were observed, and the liquid mass flow rates driven by natural circulation loop were measured by varying the air injection rate, the coolant inlet area and configuration, and the outlet area and also the water head condition of coolant reservoir. From the experimental flow observation, the recirculation flows in the near region of the shear key were identified. At a higher air injection rate condition, higher recirculation flows and choking phenomenon in the near region of the shear key were observed. As the water inlet areas increased, the natural circulation mass flow rates asymptotically increased, that is, they converged at a specific value. And the experimental correlations for the natural circulation mass flow rates along with a variation of the inlet / outlet area and wall heat flux were

  16. Effect of Loop Diameter on the Steady State and Stability Behaviour of Single-Phase and Two-Phase Natural Circulation Loops

    Directory of Open Access Journals (Sweden)

    P. K. Vijayan

    2008-01-01

    Full Text Available In natural circulation loops, the driving force is usually low as it depends on the riser height which is generally of the order of a few meters. The heat transport capability of natural circulation loops (NCLs is directly proportional to the flow rate it can generate. With low driving force, the straightforward way to enhance the flow is to reduce the frictional losses. A simple way to do this is to increase the loop diameter which can be easily adopted in pressure tube designs such as the AHWR and the natural circulation boilers employed in fossil-fuelled power plants. Further, the loop diameter also plays an important role on the stability behavior. An extensive experimental and theoretical investigation of the effect of loop diameter on the steady state and stability behavior of single- and two-phase natural circulation loops have been carried out and the results of this study are presented in this paper.

  17. Inlet throttling effect on the boiling two-phase flow stability in a natural circulation loop with a chimney

    International Nuclear Information System (INIS)

    Furuya, M.; Inada, F.; Yasuo, A.

    2001-01-01

    Experiments have been conducted to investigate an effect of inlet restriction on the thermal-hydraulic stability. A Test facility used in this study was designed and constructed to have non-dimensional values that are nearly equal to those of natural circulation BWR. Experimental results showed that driving force of the natural circulation at the stability boundary was described as a function of heat flux and inlet subcooling independent of inlet restriction. In order to extend experimental database regarding thermal-hydraulic stability to different inlet restriction, numerical analysis was carried out based on the homogeneous flow model. Stability maps in reference to the core inlet subcooling and heat flux were presented for various inlet restrictions using the above-mentioned function. Instability region during the inlet subcooling shifted to the higher inlet subcooling with increasing inlet restriction and became larger with increasing heat flux. (orig.)

  18. Experimental study on two-phase flow natural circulation in a core catcher cooling channel for EU-APR1400 using air-water system

    Energy Technology Data Exchange (ETDEWEB)

    Song, Ki Won [Division of Advanced Nuclear Engineering, POSTECH, Pohang 790-784 (Korea, Republic of); Korea Atomic Energy Research Institute, Daejeon 34057 (Korea, Republic of); Nguyen, Thanh Hung [School of Nuclear Engineering, Purdue University, West Lafayette, IN 47906 (United States); Ha, Kwang Soon; Kim, Hwan Yeol; Song, Jinho [Korea Atomic Energy Research Institute, Daejeon 34057 (Korea, Republic of); Park, Hyun Sun [Division of Advanced Nuclear Engineering, POSTECH, Pohang 790-784 (Korea, Republic of); Revankar, Shripad T., E-mail: shripad@postech.ac.kr [Division of Advanced Nuclear Engineering, POSTECH, Pohang 790-784 (Korea, Republic of); School of Nuclear Engineering, Purdue University, West Lafayette, IN 47906 (United States); Kim, Moo Hwan [Division of Advanced Nuclear Engineering, POSTECH, Pohang 790-784 (Korea, Republic of); Korea Institute of Nuclear Safety, Daejeon 305-338 (Korea, Republic of)

    2017-05-15

    Highlights: • Two-phase flow regimes and transition behavior were observed in the coolant channel. • Test were conducted for natural circulation with air-water. • Data were obtained on flow regime, void fraction, flow rates and re-wetting time. • The data were related to a cooling capability of core catcher system. - Abstract: Ex-vessel core catcher cooling system driven by natural circulation is designed using a full scaled air-water system. A transparent half symmetric section of a core catcher coolant channel of a pressurized water reactor was designed with instrumentations for local void fraction measurement and flow visualization. Two designs of air-water top separator water tanks are studied including one with modified ‘super-step’ design which prevents gas entrainment into down-comer. In the experiment air flow rates are set corresponding to steam generation rate for given corium decay power. Measurements of natural circulation flow rate, spatial local void fraction distribution and re-wetting time near the top wall are carried out for various air flow rates which simulate boiling-induced vapor generation. Since heat transfer and critical heat flux are strongly dependent on the water mass flow rate and development of two-phase flow on the heated wall, knowledge of two-phase flow characteristics in the coolant channel is essential. Results on flow visualization showing two phase flow structure specifically near the high void accumulation regions, local void profiles, rewetting time, and natural circulation flow rate are presented for various air flow rates that simulate corium power levels. The data are useful in assessing the cooling capability of and safety of the core catcher system.

  19. Simulation of the phenomenon of single-phase and two-phase natural circulation; Simulacao do fenomeno de circulacao natural mono e bifasica

    Energy Technology Data Exchange (ETDEWEB)

    Castrillo, Lazara Silveira

    1998-02-01

    Natural convection phenomenon is often used to remove the residual heat from the surfaces of bodies where the heat is generated e.g. during accidents or transients of nuclear power plants. Experimental study of natural circulation can be done in small scale experimental circuits and the results can be extrapolated for larger operational facilities. The numerical analysis of transients can be carried out by using large computational codes that simulate the thermohydraulic behavior in such facilities. The computational code RELAP5/MOD2, (Reactor Excursion and Leak Analysis Program) was developed by U.S. Nuclear Regulatory Commissions's. Division of Reactor Safety Research with the objective of analysis of transients and postulated accidents in the light water reactor (LWR) systems, including small and large ruptures with loss of coolant accidents (LOCA's). The results obtained by the simulation of single-phase and two-phase natural circulation, using the RELAP5/MOD2, are presented in this work. The study was carried out using the experimental circuit built at the 'Departamento de Engenharia Quimica da Escola Politecnica da Universidade de Sao Paulo'. In the circuit, two experiments were carried out with different conditions of power and mass flow, obtaining a single-phase regime with a level of power of 4706 W and flow of 5.10{sup -5} m{sup 3}/s (3 l/min) and a two-phase regime with a level of power of 6536 W and secondary flow 2,33.10{sup -5} m{sup 3}/s (1,4 l/min). The study allowed tio evaluate the capacity of the code for representing such phenomena as well as comparing the transients obtained theoretically with the experimental results. The comparative analysis shows that the code represents fairly well the single-phase transient, but the results for two-phase transients, starting from the nodalization and calibration used for the case single-phase transient, did not reproduce faithfully some experimental results. (author)

  20. Detailed evaluation of two phase natural circulation flow in the cooling channel of the ex-vessel core catcher for EU-APR1400

    Energy Technology Data Exchange (ETDEWEB)

    Park, Rae-Joon, E-mail: rjpark@kaeri.re.kr; Ha, Kwang-Soon; Rhee, Bo-Wook; Kim, Hwan Yeol

    2016-03-15

    Highlights: • Ex-vessel core catcher of PECS is installed in EU-APR1400. • CE-PECS has been conducted to test a cooling capability of the PECS. • Two phase flow in CE-PECS and PECS was analyzed using RELAP5/MOD3. • RELAP5 results are very similar to the CE-PECS data. • The super-step design is suitable for steam injection into the downcomer in PECS. - Abstract: The ex-vessel core catcher of the PECS (Passive Ex-vessel corium retaining and Cooling System) is installed to retain and cool down the corium in the reactor cavity of the EU (European Union)-APR (Advanced Power Reactor) 1400. A verification experiment on the cooling capability of the PECS has been conducted in the CE (Cooling Experiment)-PECS. Simulations of a two-phase natural circulation flow using the RELAP5/MOD3 computer code in the CE-PECS and PECS have been conducted to predict the two-phase flow characteristics, to determine the natural circulation mass flow rate in the cooling channel, and to evaluate the scaling in the experimental design of the CE-PECS. Particularly from a comparative study of the prototype PECS and the scaled test facility of the CE-PECS, the orifice loss coefficient in the CE-PECS was found to be 6 to maintain the coolant circulation mass flux, which is approximately 273.1 kg/m{sup 2} s. The RELAP5 results on the coolant circulation mass flow rate are very similar to the CE-PECS experimental results. An increase in the coolant injection temperature and the heat flux lead to an increase in the coolant circulation mass flow rate. In the base case simulation, a lot of vapor was injected into the downcomer, which leads to an instability of the two-phase natural circulation flow. A super-step design at a downcomer inlet is suitable to prevent vapor injection into the downcomer piping.

  1. Detailed evaluation of two phase natural circulation flow in the cooling channel of the ex-vessel core catcher for EU-APR1400

    International Nuclear Information System (INIS)

    Park, Rae-Joon; Ha, Kwang-Soon; Rhee, Bo-Wook; Kim, Hwan Yeol

    2016-01-01

    Highlights: • Ex-vessel core catcher of PECS is installed in EU-APR1400. • CE-PECS has been conducted to test a cooling capability of the PECS. • Two phase flow in CE-PECS and PECS was analyzed using RELAP5/MOD3. • RELAP5 results are very similar to the CE-PECS data. • The super-step design is suitable for steam injection into the downcomer in PECS. - Abstract: The ex-vessel core catcher of the PECS (Passive Ex-vessel corium retaining and Cooling System) is installed to retain and cool down the corium in the reactor cavity of the EU (European Union)-APR (Advanced Power Reactor) 1400. A verification experiment on the cooling capability of the PECS has been conducted in the CE (Cooling Experiment)-PECS. Simulations of a two-phase natural circulation flow using the RELAP5/MOD3 computer code in the CE-PECS and PECS have been conducted to predict the two-phase flow characteristics, to determine the natural circulation mass flow rate in the cooling channel, and to evaluate the scaling in the experimental design of the CE-PECS. Particularly from a comparative study of the prototype PECS and the scaled test facility of the CE-PECS, the orifice loss coefficient in the CE-PECS was found to be 6 to maintain the coolant circulation mass flux, which is approximately 273.1 kg/m"2 s. The RELAP5 results on the coolant circulation mass flow rate are very similar to the CE-PECS experimental results. An increase in the coolant injection temperature and the heat flux lead to an increase in the coolant circulation mass flow rate. In the base case simulation, a lot of vapor was injected into the downcomer, which leads to an instability of the two-phase natural circulation flow. A super-step design at a downcomer inlet is suitable to prevent vapor injection into the downcomer piping.

  2. Modeling of flashing-induced instabilities in the start-up phase of natural-circulation BWRs using the two-phase flow code FLOCAL

    Energy Technology Data Exchange (ETDEWEB)

    Manera, A. [Forschungszentrum Rossendorf e.V. (FZR), Institute of Safety Research, P.O.B. 510119, D-01324 Dresden (Germany) and Interfaculty Reactor Institute, Delft University of Technology, Mekelweg 15, 2629 JB Delft (Netherlands)]. E-mail: a.manera@fz-rossendorf.de; Rohde, U. [Forschungszentrum Rossendorf e.V. (FZR), Institute of Safety Research, P.O.B. 510119, D-01324 Dresden (Germany); Prasser, H.-M. [Forschungszentrum Rossendorf e.V. (FZR), Institute of Safety Research, P.O.B. 510119, D-01324 Dresden (Germany); Hagen, T.H.J.J. van der [Interfaculty Reactor Institute, Delft University of Technology, Mekelweg 15, 2629 JB Delft (Netherlands)

    2005-06-01

    This paper reports on the modeling and simulation of flashing-induced instabilities in natural-circulation systems, with special emphasis on natural-circulation boiling water reactors (BWRs). For the modeling the 4-equation two-phase model FLOCAL [Rohde, U., 1986. Ein teoretisches Modell fur Zweiphasen-stromungen in wassergekulthen Kernreaktoren und seine Anwendung zur Analyse des Naturumlaufs im Heizreaktor AST-500. Ph.D. dissertation, Akademie der Wissenschaften der DDR, Dresden], developed at the Forschungszentrum Rossendorf (FZR, Germany), has been used. The model allows for the liquid and vapor to be in thermal non-equilibrium and, via drift-flux models, to have different velocities. The phenomenology of the instability has been studied and the dominating physical effects have been determined. The results of the simulations have been compared qualitatively and quantitatively with experiments [Manera, A., van der Hagen, T.H.J.J., 2003. Stability of natural-circulation-cooled boiling water reactor during start up: experimental results. Nuc. Technol., 143] that have been carried out within the framework of a European project (NACUSP) on the CIRCUS facility. The facility, built at the Delft University of Technology in The Netherlands, is a water/steam 1:1 height-scaled loop of a typical natural-circulation-cooled BWR.

  3. Input Calibration and Validation of RELAP5 Against CIRCUS-IV Single Channel Tests on Natural Circulation Two-Phase Flow Instability

    Directory of Open Access Journals (Sweden)

    Viet-Anh Phung

    2015-01-01

    Full Text Available RELAP5 is a system thermal-hydraulic code that is used to perform safety analysis on nuclear reactors. Since the code is based on steady state, two-phase flow regime maps, there is a concern that RELAP5 may provide significant errors for rapid transient conditions. In this work, the capability of RELAP5 code to predict the oscillatory behavior of a natural circulation driven, two-phase flow at low pressure is investigated. The simulations are compared with a series of experiments that were performed in the CIRCUS-IV facility at the Delft University of Technology. For this purpose, we developed a procedure for calibration of the input and code validation. The procedure employs (i multiple parameters measured in different regimes, (ii independent consideration of the subsections of the loop, and (iii assessment of importance of the uncertain input parameters. We found that predicted system parameters are less sensitive to variations of the uncertain input and boundary conditions in high frequency oscillations regime. It is shown that calculation results overlap experimental values, except for the high frequency oscillations regime where the maximum inlet flow rate was overestimated. This finding agrees with the idea that steady state, two-phase flow regime maps might be one of the possible reasons for the discrepancy in case of rapid transients in two-phase systems.

  4. Thermal-hydraulic oscillations in a low pressure two-phase natural circulation loop at low powers and high inlet subcoolings

    International Nuclear Information System (INIS)

    Wang, S.B.; Wu, J.Y.; Chin Pan; Lin, W.K.

    2004-01-01

    The stability of a natural circulation boiling loop is of great importance and interests for both academic researches and many industrial applications, such as next generation boiling water reactors. The present study investigated the thermal-hydraulic oscillation behavior in a low pressure two-phase natural circulation loop at low powers and high inlet subcoolings. The experiments were conducted at atmospheric pressure with heating power ranging from 4 to 8 kW and inlet subcooling ranging from 27 to 75 deg. C. Significant oscillations in loop mass flow rate, pressure drop in each section, and heated wall and fluid temperatures are present for all the cases studied here. The oscillation is typically quasi-periodic and with flow reversal with magnitudes smaller than forward flows. The magnitude of wall temperature oscillation could be as high as 60 deg. C, which will be of serious concern for practical applications. It is found that the first fundamental oscillation (large magnitude oscillation) frequency increases with increase in heated power and with decrease in inlet subcooling. (author)

  5. Experimental study of natural two-phase flow circulation using a visualization technique; Estudo experimental da circulacao natural bifasica utilizando uma tecnica de visualizacao

    Energy Technology Data Exchange (ETDEWEB)

    Vinhas, Pedro A.M., E-mail: Pedro_mvinhas@poli.ufrj.br [Universidade Federal do Rio de Janeiro (UFRJ), Rio de Janeiro, RJ (Brazil). Dept. de Engenharia Nuclear; Faccini, Jose L.H., E-mail: faccini@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil); Su, Jian, E-mail: sujian@lasme.coppe.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

    2013-07-01

    This paper presents an experimental study of natural two-phase flow in a circuit that simulates, on a smaller scale, a typical residual heat removal system of passive reactors APWR (Advanced Pressurized Water Reactor). The circuit was formed by a heater, a heat exchanger and piping. The experimental study was the application of a visualization technique, using a high speed camera, for measuring the size and speed of vapor bubbles generated in the heater with different power heating. The camera was positioned in the central region of the pipe connecting the heater to the heat exchanger, where there is a clear passage. The flow of images were processed and analyzed using commercial software that allowed the determination of the length and velocity of the bubbles. The results were then compared with correlations available in literature.

  6. The influence of the friction coefficients, the power and the flowrate on the two-phase flow instabilities in natural circulation

    International Nuclear Information System (INIS)

    Kerris, A.M.

    1987-04-01

    A study of non-linear effects of a two phases flow instabilities has been done using theoretical and experimental models. The experimental model is based on a boiling channel placed in a water loop of natural convection. Studies of stability, with introduction of fluctuations over different parameters, have been achieved using the two models. The results of the experimental model agree with those of the theoretical model. It was found that the head loss coefficients K I and K E of the inlet and outlet channel respectively, had an influence over the stability of the system. An increase in the former produces an increase in the stability while an increase in the latter has the effect of increasing the instability. In the experiment, oscillations of the flow rate were observed. Two types of oscillations were noticed: (1) small oscillations called pressure drop oscillations, (2) large oscillations called density wave oscillations. A study of the variation of these two types of oscillations with power and the coefficient K I and K E had been achieved. It was found that pressure drop oscillations disappeared with the increase of power and the density waves oscillations increased with the increase of power

  7. Contribution to the study of the thermal and hydrodynamical properties of a two-phase natural circulation flow of normal helium (He I) for the cooling of superconducting magnets

    International Nuclear Information System (INIS)

    Benkheira, L.

    2007-06-01

    The method of cooling based on the thermosyphon principle is of great interest because of its simplicity, its passivity and its low cost. It is adopted to cool down to 4,5 K the superconducting magnet of the CMS particles detector of the Large Hadron Collider (LHC) experiment under construction at CERN, Geneva. This work studies heat and mass transfer characteristics of two phase He I in a natural circulation loop. The experimental set-up consists of a thermosyphon single branch loop mainly composed of a phase separator, a downward tube, and a test section. The experiments were conducted with varying several parameters such as the diameter of the test section (10 mm or 14 mm) and the applied heat flux up to the appearance of the boiling crisis. These experiments have permitted to determine the laws of evolution of the various parameters characterizing the flow (circulation mass flow rate, vapour mass flow rate, vapour quality, friction coefficient, two phase heat transfer coefficient and the critical heat flux) as a function of the applied heat flux. On the base of the obtained results, we discuss the validity of the various existing models in the literature. We show that the homogeneous model is the best model to predict the hydrodynamical properties of this type of flow in the vapour quality range 0≤x≤30%. Moreover, we propose two models for the prediction of the two phase heat transfer coefficient and the density of the critical heat flux. The first one considers that the effects of the forced convection and nucleate boiling act simultaneously and contribute to heat transfer. The second one correlates the measured critical heat flux density with the ratio altitude to diameter. (author)

  8. Natural Circulation with Boiling

    Energy Technology Data Exchange (ETDEWEB)

    Mathisen, R P

    1967-09-15

    A number of parameters with dominant influence on the power level at hydrodynamic instability in natural circulation, two-phase flow, have been studied experimentally. The geometrical dependent quantities were: the system driving head, the boiling channel and riser dimensions, the single-phase as well as the two phase flow restrictions. The parameters influencing the liquid properties were the system pressure and the test section inlet subcooling. The threshold of instability was determined by plotting the noise characteristics in the mass flow records against power. The flow responses to artificially obtained power disturbances at instability conditions were also measured in order to study the nature of hydrodynamic instability. The results presented give a review over relatively wide ranges of the main parameters, mainly concerning the coolant performance in both single and parallel boiling channel flow. With regard to the power limits the experimental results verified that the single boiling channel performance was intimately related to that of the parallel channels. In the latter case the additional inter-channel factors with attenuating effects were studied. Some optimum values of the parameters were observed.

  9. Contribution to the study of the thermal and hydrodynamical properties of a two-phase natural circulation flow of normal helium (He I) for the cooling of superconducting magnets; Contribution a l'etude des proprietes thermiques et hydrodynamiques d'un ecoulement d'helium normal (He I) diphasique en circulation naturelle pour le refroidissement des aimants supraconducteurs

    Energy Technology Data Exchange (ETDEWEB)

    Benkheira, L

    2007-06-15

    The method of cooling based on the thermosyphon principle is of great interest because of its simplicity, its passivity and its low cost. It is adopted to cool down to 4,5 K the superconducting magnet of the CMS particles detector of the Large Hadron Collider (LHC) experiment under construction at CERN, Geneva. This work studies heat and mass transfer characteristics of two phase He I in a natural circulation loop. The experimental set-up consists of a thermosyphon single branch loop mainly composed of a phase separator, a downward tube, and a test section. The experiments were conducted with varying several parameters such as the diameter of the test section (10 mm or 14 mm) and the applied heat flux up to the appearance of the boiling crisis. These experiments have permitted to determine the laws of evolution of the various parameters characterizing the flow (circulation mass flow rate, vapour mass flow rate, vapour quality, friction coefficient, two phase heat transfer coefficient and the critical heat flux) as a function of the applied heat flux. On the base of the obtained results, we discuss the validity of the various existing models in the literature. We show that the homogeneous model is the best model to predict the hydrodynamical properties of this type of flow in the vapour quality range 0{<=}x{<=}30%. Moreover, we propose two models for the prediction of the two phase heat transfer coefficient and the density of the critical heat flux. The first one considers that the effects of the forced convection and nucleate boiling act simultaneously and contribute to heat transfer. The second one correlates the measured critical heat flux density with the ratio altitude to diameter. (author)

  10. Natural circulation in a scaled PWR integral test facility

    International Nuclear Information System (INIS)

    Kiang, R.L.; Jeuck, P.R. III

    1987-01-01

    Natural circulation is an important mechanism for cooling a nuclear power plant under abnormal operating conditions. To study natural circulation, we modeled a type of pressurized water reactor (PWR) that incorporates once-through steam generators. We conducted tests of single-phase natural circulations, two-phase natural circulations, and a boiler condenser mode. Because of complex geometry, the natural circulations observed in this facility exhibit some phenomena not commonly seen in a simple thermosyphon loop

  11. Experimental study of the natural circulation phenomena

    International Nuclear Information System (INIS)

    Sabundjian, Gaiane; Andrade, Delvonei Alves de; Umbehaun, Pedro E.; Torres, Walmir M.; Castro, Alfredo Jose Alvim de; Belchior Junior, Antonio; Rocha, Ricardo Takeshi Vieira da; Damy, Osvaldo Luiz de Almeida; Torres, Eduardo

    2006-01-01

    The objective of this paper is to study the natural circulation in experimental loops and extend the results to nuclear facilities. New generation of compact nuclear power plants use the natural circulation as cooling and residual heat removal systems in case of accidents or shutdown. Lately the interest in this phenomenon, by scientific community, has increased. The experimental loop, described in this paper, was assembled at Escola Politecnica - USP at the Chemical Engineering Department. It is the goal to generate information to help with the understanding of the one and two phase natural circulation phenomena. Some experiments were performed with different levels of heat power and different flow of the cooling water at the secondary circuit. The data generated from these experiments are going to be used to validate some computational thermal hydraulic codes. Experimental results for one and two phase regimes are presented as well as the proposed model to simulate the flow regimes with the RELAP5 code. (author)

  12. Natural circulation in simulated LMFBR fuel assemblies

    International Nuclear Information System (INIS)

    Levin, A.E.; Carbajo, J.J.; Lloyd, D.B.; Montgomery, B.H.; Rose, S.D.; Wantland, J.L.

    1985-01-01

    Natural circulation experiments have been performed using simulated liquid metal fast breeder reactor fuel assemblies in the Thermal-Hydraulic Out-of-Reactor Safety (THORS) facility, an engineering-scale sodium loop. Objective of these tests has been to provide experimental data under conditions that might be encountered during a partial or total loss of the shutdown heat removal system (SHRS) in a reactor. The experiments have included single- and two-phase tests under quasi-steady and transient conditions, at both nominal and non-nominal system conditions. Results from these test indicate that the potential for reactor damage during degraded SHRS operation is extremely slight, and that natural circulation can be a major contributor to safe operation of the system in both single- and two-phase flow during such operation

  13. Quenching phenomena in natural circulation loop

    International Nuclear Information System (INIS)

    Umekawa, Hisashi; Ozawa, Mamoru; Ishida, Naoki

    1995-01-01

    Quenching phenomena has been investigated experimentally using circulation loop of liquid nitrogen. During the quenching under natural circulation, the heat transfer mode changes from film boiling to nucleate boiling, and at the same time flux changes with time depending on the vapor generation rate and related two-phase flow characteristics. Moreover, density wave oscillations occur under a certain operating condition, which is closely related to the dynamic behavior of the cooling curve. The experimental results indicates that the occurrence of the density wave oscillation induces the deterioration of effective cooling of the heat surface in the film and the transition boiling regions, which results in the decrease in the quenching velocity

  14. Quenching phenomena in natural circulation loop

    Energy Technology Data Exchange (ETDEWEB)

    Umekawa, Hisashi; Ozawa, Mamoru [Kansai Univ., Osaka (Japan); Ishida, Naoki [Daihatsu Motor Company, Osaka (Japan)

    1995-09-01

    Quenching phenomena has been investigated experimentally using circulation loop of liquid nitrogen. During the quenching under natural circulation, the heat transfer mode changes from film boiling to nucleate boiling, and at the same time flux changes with time depending on the vapor generation rate and related two-phase flow characteristics. Moreover, density wave oscillations occur under a certain operating condition, which is closely related to the dynamic behavior of the cooling curve. The experimental results indicates that the occurrence of the density wave oscillation induces the deterioration of effective cooling of the heat surface in the film and the transition boiling regions, which results in the decrease in the quenching velocity.

  15. Isolation of circulating tumor cells using photoacoustic flowmetry and two phase flow

    Science.gov (United States)

    O'Brien, Christine M.; Rood, Kyle D.; Gupta, Sagar K.; Mosley, Jeffrey D.; Goldschmidt, Benjamin S.; Sharma, Nikhilesh; Sengupta, Shramik; Viator, John A.

    2011-03-01

    Melanoma is the deadliest form of skin cancer, yet current diagnostic methods are inadequately sensitive. Patients must wait until secondary tumors form before malignancy can be diagnosed and treatment prescribed. Detection of cells that have broken off the original tumor and flow through the blood or lymph system can provide data for diagnosing and monitoring cancer. Our group utilizes the photoacoustic effect to detect metastatic melanoma cells, which contain the pigmented granule melanin. As a rapid laser pulse irradiates melanoma, the melanin undergoes thermo-elastic expansion and ultimately creates a photoacoustic wave. Thus, melanoma patient's blood samples can be enriched, leaving the melanoma in a white blood cell (WBC) suspension. Irradiated melanoma cells produce photoacoustic waves, which are detected with a piezoelectric transducer, while the optically transparent WBCs create no signals. Here we report an isolation scheme utilizing two-phase flow to separate detected melanoma from the suspension. By introducing two immiscible fluids through a t-junction into one flow path, the analytes are compartmentalized. Therefore, the slug in which the melanoma cell is located can be identified and extracted from the system. Two-phase immiscible flow is a label free technique, and could be used for other types of pathological analytes.

  16. 两相自然循环非能动系统静态分岔特性和参数效应研究%The Research on Static Bifurcation Characteristics and ParametricEffect of Two-Phase Natural Circulation and Passive System

    Institute of Scientific and Technical Information of China (English)

    徐济鋆; 杨燕华; 匡波; 姚伟; 张荣华; 佟立丽

    2001-01-01

    耗散结构在流体动力学中存在,两相自然循环非能动系统稳定性是耗散结构在多相流体动力学中的问题,运用分岔理论及其DERPAR数值方法给出系统随参数变化的演化过程的静态分岔解图图谱。以静态分岔解图呈现的迟滞现象(多平衡解)定义了输热能力限、静态漂移稳定性准则、稳定性裕度、比质量流量输热能力、分岔现象消失—呈单值性的临界参数区的系统演化的自组织特性的物理本质—系统的稳态和时间相关态随着特征参数的演化。用低压下的实验数据与理论计算分析进行比较,变化趋势吻合,尚需进一步验证。%The formation of dissipative structures has long been known tooccur in hydrodynamics. The two-phase natural circulation and passive system (TPNCPS) instability is a disspative structure problem in multiphase hydrodynamics. The spectrum of the static bifurcation solutions (SBS) of TPNCPS through the variation of a parameter (one or more) has been derived in terms of Bifurcation Theory and DERPAR Numerical Method. Based on the appearance of Thermal-Siphon Hysteresis, the transprot heat capability, static excursion criterion, stationary margin, transport heat capability of specific mass flow-rate and the disappear of bifurcation-the transition of single-valued region with the change of parameter have been defined. Such phenomena are the problems of describing self-organization, i. e. detailed study of stationary and/or time dependent status evolving with changes of characteristic parameters. A comparison between computational curves and low-pressure experimental data shows the tendency of evolutionary processes compatibly. The further tests is needed.

  17. Natural circulation in an integral CANDU test facility

    International Nuclear Information System (INIS)

    Ingham, P.J.; Sanderson, T.V.; Luxat, J.C.; Melnyk, A.J.

    2000-01-01

    Over 70 single- and two-phase natural circulation experiments have been completed in the RD-14M facility, an integral CANDU thermalhydraulic test loop. This paper describes the RD-14M facility and provides an overview of the impact of key parameters on the results of natural circulation experiments. Particular emphasis will be on phenomena which led to heat up at high system inventories in a small subset of experiments. Clarification of misunderstandings in a recently published comparison of the effectiveness of natural circulation flows in RD-14M to integral facilities simulating other reactor geometries will also be provided. (author)

  18. Natural-circulation-cooling characteristics during PWR accident simulations

    International Nuclear Information System (INIS)

    Adams, J.P.; McCreery, G.E.; Berta, V.T.

    1983-01-01

    A description of natural circulation cooling characteristics is presented. Data were obtained from several pressurized water reactor accident simulations in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR). The reliability of natural circulation cooling, its cooling effectiveness, and the effect of changing system conditions are described. Quantitative comparison of flow rates and time constants with theory for both single- and two-phase fluid conditions were made. It is concluded that natural circulation cooling can be relied on in plant recovery procedures in the absence of forced convection whenever the steam generator heat sink is available

  19. Natural Circulation Characteristics of an Integral Pressurized Water Reactor

    International Nuclear Information System (INIS)

    Junli Gou; Suizheng Qiu; Guanghui Su; Dounan Jia

    2006-01-01

    Natural circulation potential is of great importance to the inherent safety of a nuclear reactor. This paper presents a theoretical investigation on the natural circulation characteristics of an integrated pressurized water reactor. Through numerically solved the one-dimensional model, the steady-state single phase conservative equations for the primary circuit and the steady-state two-phase drift-flux conservative equations for the secondary side of the once-through steam generator, the natural circulation characteristics are studied. Based on the preliminary calculation analysis, it is found that natural circulation mass flow rate is proportional to the exponential function of the power, and the value of the exponent is related to working conditions of the steam generator secondary side. The higher height difference between the core center and the steam generator center is favorable to the heat removal capacity of the natural circulation. (authors)

  20. Gas-liquid two-phase flows in double inlet cyclones for natural gas separation

    DEFF Research Database (Denmark)

    Yang, Yan; Wang, Shuli; Wen, Chuang

    2017-01-01

    The gas-liquid two-phase flow within a double inlet cyclone for natural gasseparation was numerically simulated using the discrete phase model. The numericalapproach was validated with the experimental data, and the comparison resultsagreed well with each other. The simulation results showed...... that the strong swirlingflow produced a high centrifugal force to remove the particles from the gas mixture.The larger particles moved downward on the internal surface and were removeddue to the outer vortex near the wall. Most of the tiny particles went into the innervortex zones and escaped from the up...

  1. Analysis of the hydrodynamic stability of natural circulation

    International Nuclear Information System (INIS)

    Olive, J.; Baby, J.P.

    1980-01-01

    A mathematical model (EOLE) for the analysis of the stability of boilers with natural circulation is discussed. The method employed consists in linearizing one-dimensional flow equations and in integrating them while employing the Laplace transformation. The properties of a two-phase fluid are schematized by a homogeneous model with slip. The computation results in the circulation loop transfer functions and its natural modes of oscillation (frequency and damping). A discussion follows which compares results obtained with this method to those of other existing models in the case of a straight pipe with forced circulation. Agreement proved to be satisfactory. The results are then given of a parametric study involving the stability of a PWR natural circulation steam generator. These results show that the model can satisfy, at least qualitatively, trends observed empirically or obtained with other more complex theoretical models. (author)

  2. Single and two-phase similarity analysis of a reduced-scale natural convection loop relative to a full-scale prototype

    International Nuclear Information System (INIS)

    Botelho, David A.; Faccini, Jose L.H.

    2002-01-01

    The main topic in this paper is a new device being considered to improve nuclear reactor safety employing the natural circulation. A scaled experiment used to demonstrate the performance of the device is also described. We also applied a similarity analysis method for single and two-phase natural convection loop flow to the IEN CCN experiment and to an APEX like experiment to verify the degree of similarity relative to a full-scale prototype like the AP600. Most of the CCN similarity numbers that represent important single and two-phase similarity conditions are comparable to the APEX like loop non-dimensional numbers calculated employing the same methodology. Despite the much smaller geometric, pressure, and power scales, we conclude that the IEN CCN has single and two-phase natural circulation similarity numbers that represent fairly well the full-scale prototype. even lacking most complementary primary and safety systems, this IEN circuit provided a much valid experience to develop human, experimental, and analytical resources, besides its utilization as a training tool. (author)

  3. Two-phase modelling of thermal dissipation in a natural basin

    International Nuclear Information System (INIS)

    Baltrenas, P.; Vaitiekunas, P.; Katinas, V.; Markevicius, A.

    2004-01-01

    The state of two-phase flow 'liquid-gas' has been modeled numerically by the three-dimensional method of complex research of heat and mass transfer. This allows examining the interaction of some transfer processes in a natural cooling basin (the Druksiai lake): the wind power and direction, variable water density, the coefficient of heat conduction and heat transfer of the water-air interface. Combined effect of these natural actions determines the heat amount that the basin is able to dissipate to the surrounding atmospheric media in thermal equilibrium (without changes in the mean water temperature). This paper presents a number of the most widely used expressions for the coefficients of vertical and horizontal heat transfer. On the basis of stream velocity and mean temperature profiles measured in the cooling pond as well as on that of their time variations suggestions are made that the mixing rate at the water surface is caused by natural space - time variation of the wind, and can be described by the value of eddy viscosity coefficient - 1 m2/s (numerical modeling with 0,9-1,3 m2/s). The wind influences the surface of the lake according to the experimental data, i e 1-3 % of the mean wind velocity. The model applies to the weakly wind, approximately 1-5 m/s of the mean wind velocity. Comparison of experimental and numerical results showed a qualitative agreement. For a better quantitative approximation, it is necessary to have more boundary conditions variable with time and to solve unsteady set equations for transfer processes. (author)

  4. NPP Krsko natural circulation performance evaluation

    International Nuclear Information System (INIS)

    Segon, Velimir; Bajs, Tomislav; Frogheri, Monica

    1999-01-01

    The present document deals with an evaluation of the natural circulation performance of the Krsko nuclear power plant. Two calculation have been performed using the NPP Krsko nodalization (both similar to the LOBI A2-77 natural circulation experiment) - the first with the present steam generators at NPP Krsko (Westinghouse, 18% plugged), the second with the future steam generators (Siemens, 0% plugged). The results were evaluated using the natural circulation flow map derived in /1/, and were compared to evaluate the influence of the new steam generators on the natural circulation performance. (author)

  5. Basic natural circulation characteristics of SBWR

    International Nuclear Information System (INIS)

    Kuran, S.; Soekmen, C. N.

    2001-01-01

    Natural circulation is an important passive heat removal mechanism for both existing and next generation light water reactors. Simplified Boiling Water Reactor (SBWR) is one of the advanced light water reactors that rely on natural circulation for normal as well as emergency core cooling. In this study, basic natural circulation characteristics of this reactor are examined on a flow loop that simulates the operation of SBWR. On this model, effect of system operating parameters on the steady state natural circulation characteristics inside the loop is studied via solving the transcendental equation for loop flow rate

  6. Survey of natural-circulation cooling in U.S. pressurized water reactors

    International Nuclear Information System (INIS)

    Boyack, B.E.

    1985-01-01

    Literature describing natural circulation analyses, experiments, and plant operation have been obtained from the Nuclear Regulatory Commission, reactor vendors, utility-sponsored research groups, utilities, national laboratories, and foreign sources. These have been reviewed and significant results and conclusions identified. Three modes of natural-circulation cooling are covered: single phase, two-phase, and reflux condensation. Single-phase natural circulation is amply verified by plant operational data, test data from scaled experimental facilities, and analysis with assessed computer codes. Ample evidence also exists that two-phase natural circulation can successfully cool pressurized water reactors. This mode occurs during certain events such as small-break loss-of-coolant accidents. The data base for reflux condensation is primarily from tests in scaled experimental facilities. There are no plant operational data and only limited assessment of thermal-hydraulic systems codes has been performed. Further work is needed before this mode of natural circulation can be confidently used

  7. Two-Phase Flow Simulations In a Natural Rock Fracture using the VOF Method

    International Nuclear Information System (INIS)

    Crandall, Dustin; Ahmadi, Goodarz; Smith, Duane H.; Bromhal, Grant

    2010-01-01

    Standard models of two-phase flow in porous media have been shown to exhibit several shortcomings that might be partially overcome with a recently developed model based on thermodynamic principles (Hassanizadeh and Gray, 1990). This alternative two-phase flow model contains a set of new and non-standard parameters, including specific interfacial area. By incorporating interfacial area production, destruction, and propagation into functional relationships that describe the capillary pressure and saturation, a more physical model has been developed. Niessner and Hassanizadeh (2008) have examined this model numerically and have shown that the model captures saturation hysteresis with drainage/imbibition cycles. Several static experimental studies have been performed to examine the validity of this new thermodynamically based approach; these allow the determination of static parameters of the model. To date, no experimental studies have obtained information about the dynamic parameters required for the model. A new experimental porous flow cell has been constructed using stereolithography to study two-phase flow phenomena (Crandall et al. 2008). A novel image analysis tool was developed for an examination of the evolution of flow patterns during displacement experiments (Crandall et al. 2009). This analysis tool enables the direct quantification of interfacial area between fluids by matching known geometrical properties of the constructed flow cell with locations identified as interfaces from images of flowing fluids. Numerous images were obtained from two-phase experiments within the flow cell. The dynamic evolution of the fluid distribution and the fluid-fluid interface locations were determined by analyzing these images. In this paper, we give a brief introduction to the thermodynamically based two-phase flow model, review the properties of the stereolithography flow cell, and show how the image analysis procedure has been used to obtain dynamic parameters for the

  8. Natural circulation in reactor coolant system

    International Nuclear Information System (INIS)

    Han, J.T.

    1987-01-01

    Reactor coolant system (RCS) natural circulation in a PWR is the buoyancy-driven coolant circulation between the core and the upper-plenum region (in-vessel circulation) with or without a countercurrent flow in the hot leg piping between the vessel and steam generators (ex-vessel circulation). This kind of multidimensional bouyancy-driven flow circulation serves as a means of transferring the heat from the core to the structures in the upper plenum, hot legs, and possibly steam generators. As a result, the RCS piping and other pressure boundaries may be heated to high temperatures at which the structural integrity is challenged. RCS natural circulation is likely to occur during the core uncovery period of the TMLB' accident in a PWR when the vessel upper plenum and hot leg are already drained and filled with steam and possibly other gaseous species. RCS natural circulation is being studied for the Surry plant during the TMLB' accident in which station blackout coincides with the loss of auxiliary feedwater and no operator actions. The effects of the multidimensional RCS natural circulation during the TMLB' accident are discussed

  9. Reliability evaluation of a natural circulation system

    International Nuclear Information System (INIS)

    Jafari, Jalil; D'Auria, Francesco; Kazeminejad, Hossein; Davilu, Hadi

    2003-01-01

    This paper discusses a reliability study performed with reference to a passive thermohydraulic natural circulation (NC) system, named TTL-1. A methodology based on probabilistic techniques has been applied with the main purpose to optimize the system design. The obtained results have been adopted to estimate the thermal-hydraulic reliability (TH-R) of the same system. A total of 29 relevant parameters (including nominal values and plausible ranges of variations) affecting the design and the NC performance of the TTL-1 loop are identified and a probability of occurrence is assigned for each value based on expert judgment. Following procedures established for the uncertainty evaluation of thermal-hydraulic system codes results, 137 system configurations have been selected and each configuration has been analyzed via the Relap5 best-estimate code. The reference system configuration and the failure criteria derived from the 'mission' of the passive system are adopted for the evaluation of the system TH-R. Four different definitions of a less-than-unity 'reliability-values' (where unity represents the maximum achievable reliability) are proposed for the performance of the selected passive system. This is normally considered fully reliable, i.e. reliability-value equal one, in typical Probabilistic Safety Assessment (PSA) applications in nuclear reactor safety. The two 'point' TH-R values for the considered NC system were found equal to 0.70 and 0.85, i.e. values comparable with the reliability of a pump installed in an 'equivalent' forced circulation (active) system having the same 'mission'. The design optimization study was completed by a regression analysis addressing the output of the 137 calculations: heat losses, undetected leakage, loop length, riser diameter, and equivalent diameter of the test section have been found as the most important parameters bringing to the optimal system design and affecting the TH-R. As added values for this work, the comparison has

  10. Visualization of Two Phase Natural Convection Flow in a Vertical Pipe using the Sulfuric Acid - Copper Sulfate Electroplating System

    Energy Technology Data Exchange (ETDEWEB)

    Ohk, Seung-Min; Chae, Myeong-Seon; Chung, Bum-Jin [Kyung Hee Univ., Yongin (Korea, Republic of)

    2014-10-15

    The passive containment cooling system (PCCS) driven by natural forces convection gain draws research interests after Fukushima NPP accident. The PCCS was classified into three categories: Containment pressure suppression, Containment passive heat removal/pressure suppression systems and Passive containment spray. Among the types of containment passive heat removal/pressure suppression systems, the system composed of an internal heat exchanger and an external coolant tank is considered. In a severe accident condition, the heat from the containment atmosphere is transferred to the outer surface of the heat exchanger by the convection and condensation of the mixture of steam and gases. On the other hand, the heat is transferred to external pool by single phase or two phase natural convection inside of heat exchanger pipes. The study aimed at investigating the influence of the diameter (D) and height (H) of the heat exchanger pipes on the single phase and two phase natural convection heat transfer. As the initial stage of the study, the two phase natural convection flow inside a vertical pipe is visualized. In order to achieve the aim with ample test rig, a sulfuric acid - cooper sulfate electroplating system was employed based on the analogy between heat and mass transfer. The reduction of hydrogen ion at the cathode surface at high potential was used to simulate the boiling phenomena. This study tried to visualize the boiling heat transfer inside a vertical pipe using a cupric acid-copper sulfate (H{sub 2}SO{sub 4}-CuSO{sub 4}) electroplating system. This seems to be successful so far. However further study has to be done to compare the result with real two phase flow situation. The surface tension and surface characteristics are to be tuned to simulate the real situation.

  11. Assessment of CATHARE2 V1.5qR6 using the experimental data of BETHSY natural circulation tests

    International Nuclear Information System (INIS)

    Huang Yanping; Jia Dounan

    2003-01-01

    The assessment of CATHARE2 V1.5qR6 is carried out against the experimental data of BETHSY natural circulation test-4. 1a-TC. Results show that the experimental process under single phase natural circulation can be predicted very well by CATHARE2 V1.5qR6, the primary mass inventory at the transition points from single phase natural circulation to two-phase natural circulation and from two-phase natural circulation to reflux condensation mode are also predicted correctly. The predicted results for thermohydraulic parameters of two-phase natural circulation and reflux condensation mode are not so good. Generally speaking, the prediction capability of CATHARE2 V1.5 for strong and two-phase flow process should be improved further in future

  12. Mathematical Model of Two Phase Flow in Natural Draft Wet-Cooling Tower Including Flue Gas Injection

    Directory of Open Access Journals (Sweden)

    Hyhlík Tomáš

    2016-01-01

    Full Text Available The previously developed model of natural draft wet-cooling tower flow, heat and mass transfer is extended to be able to take into account the flow of supersaturated moist air. The two phase flow model is based on void fraction of gas phase which is included in the governing equations. Homogeneous equilibrium model, where the two phases are well mixed and have the same velocity, is used. The effect of flue gas injection is included into the developed mathematical model by using source terms in governing equations and by using momentum flux coefficient and kinetic energy flux coefficient. Heat and mass transfer in the fill zone is described by the system of ordinary differential equations, where the mass transfer is represented by measured fill Merkel number and heat transfer is calculated using prescribed Lewis factor.

  13. Power Disturbances Close to Hydrodynamic Instability in Natural Circulation Two-Phase Flow

    International Nuclear Information System (INIS)

    Mathisen, R.P.; Eklind, O.

    1967-07-01

    In certain boiling reactor designs high positive void coefficients could exist and under certain circumstances cause instability. Control systems may therefore be desired. In such a controlled reactor there could remain superimposed low frequency power oscillations of some magnitude. The object of the current experiments in SKALVAN was to examine whether or not such slow oscillations could influence the hydrodynamic stability limit of the individual boiling channels. While operating the loop close to the threshold of hydrodynamic instability, the power was pulsed in the boiling channel. The pulse widths had a lower limit of 0.65 sec due to the contactor time constant. The square wave power oscillation amplitude ΔQ/Q was 12.2 %, and the interval T between the pulses was varied in the range 0 0 /T 0 was the mass flow oscillation period. The corresponding mass flow oscillations remained damped for all disturbance periods which were examined. With minimum test section inlet restrictions the power level at instability was much lower than that at burnout conditions. At higher restrictions these phenomena occurred at approximately equivalent power levels. The experiments with minimum inlet restrictions were also performed beyond the instability threshold. In this case it was possible to exceed the nominal burnout point temporarily by 5 per cent or more for periods of the order of magnitude 1 second. Even now the boiling channel conditions were not so severely affected that the burnout detectors tripped, and the power disturbances caused low frequency modulated wave trains

  14. Power Disturbances Close to Hydrodynamic Instability in Natural Circulation Two-Phase Flow

    Energy Technology Data Exchange (ETDEWEB)

    Mathisen, R P; Eklind, O

    1967-07-15

    In certain boiling reactor designs high positive void coefficients could exist and under certain circumstances cause instability. Control systems may therefore be desired. In such a controlled reactor there could remain superimposed low frequency power oscillations of some magnitude. The object of the current experiments in SKALVAN was to examine whether or not such slow oscillations could influence the hydrodynamic stability limit of the individual boiling channels. While operating the loop close to the threshold of hydrodynamic instability, the power was pulsed in the boiling channel. The pulse widths had a lower limit of 0.65 sec due to the contactor time constant. The square wave power oscillation amplitude {delta}Q/Q was 12.2 %, and the interval T between the pulses was varied in the range 0 < T{sub 0}/T < 0. 5 where T{sub 0} was the mass flow oscillation period. The corresponding mass flow oscillations remained damped for all disturbance periods which were examined. With minimum test section inlet restrictions the power level at instability was much lower than that at burnout conditions. At higher restrictions these phenomena occurred at approximately equivalent power levels. The experiments with minimum inlet restrictions were also performed beyond the instability threshold. In this case it was possible to exceed the nominal burnout point temporarily by 5 per cent or more for periods of the order of magnitude 1 second. Even now the boiling channel conditions were not so severely affected that the burnout detectors tripped, and the power disturbances caused low frequency modulated wave trains.

  15. Investigations on hydrodynamic stability of two phase flow in a low pressure natural circulation system

    Energy Technology Data Exchange (ETDEWEB)

    Shaorong, Wu; Dazhong, Wang; Meisheng, Yao; Jinhai, Bo; Yunxian, Tong; Shengyao, Jiang; Bing, Han [Institute of Nuclear Energy and Technology, Tsingua Univ., Beijing (China)

    1997-09-01

    Appropriately scaled ``Loop Stability`` tests and ``Channel Stability`` tests were performed with single heated channel system and two parallel channel system separately at the Institute of Nuclear Energy Technology (INET) of the Tsinghua University in China. A broad range of several operational parameters such as heating power, system pressure, test inlet subcooling and resistance coefficient were investigated. It was found that under certain geometric conditions and operating parameters a self-sustaining, low frequency, even amplitude mass flow oscillation may be excited at very low steam qualities and subcooling conditions. Stability maps under different conditions have been provided to assist the design of the NHR. (author). 6 refs, 15 figs.

  16. Experimental study of natural circulation circuit

    Energy Technology Data Exchange (ETDEWEB)

    Lemos, Wanderley F.; Su, Jian, E-mail: wlemos@lasme.coppe.ufrj.br, E-mail: sujian@lasme.coppe.ufrj.br [Coordenacao dos Programas de Pos-Graduacao de Engenharia (LASME/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Lab. de Simulacao e Metodos Numericos; Faccini, Jose L.H., E-mail: faccini@ien.gov.br [Instituto de Engenharia Nuclear (LTE/IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Lab. de Termo-Hidraulica Experimental

    2011-07-01

    This work presents an experimental study about fluid flows behavior in natural circulation, under conditions of single-phase flow. The experiment was performed through experimental thermal-hydraulic circuit built at IEN. This test equipment has performance similar to passive system of residual heat removal present in Advanced Pressurized Water Reactors (APWR). This experimental study aims to observing and analyzing the natural circulation phenomenon, using this experimental circuit that was dimensioned and built based on concepts of similarity and scale. This philosophy allows the analysis of natural circulation behavior in single-phase flow conditions proportionally to the functioning real conditions of a nuclear reactor. The experiment was performed through procedures to initialization of hydraulic feeding of primary and secondary circuits and electrical energizing of resistors installed inside heater. Power controller has availability to adjust values of electrical power to feeding resistors, in order to portray several conditions of energy decay of nuclear reactor in a steady state. Data acquisition system allows the measurement and monitoring of the evolution of the temperature in various points through thermocouples installed in strategic points along hydraulic circuit. The behavior of the natural circulation phenomenon was monitored by graphical interface on computer screen, showing the temperature evolutions of measuring points and results stored in digital spreadsheets. The results stored in digital spreadsheets allowed the getting of data to graphic construction and discussion about natural circulation phenomenon. Finally, the calculus of Reynolds number allowed the establishment for a correlation of friction in function of geometric scales of length, heights and cross section of tubing, considering a natural circulation flow throughout in the region of hot leg. (author)

  17. Experimental study of natural circulation circuit

    International Nuclear Information System (INIS)

    Lemos, Wanderley F.; Su, Jian; Faccini, Jose L.H.

    2011-01-01

    This work presents an experimental study about fluid flows behavior in natural circulation, under conditions of single-phase flow. The experiment was performed through experimental thermal-hydraulic circuit built at IEN. This test equipment has performance similar to passive system of residual heat removal present in Advanced Pressurized Water Reactors (APWR). This experimental study aims to observing and analyzing the natural circulation phenomenon, using this experimental circuit that was dimensioned and built based on concepts of similarity and scale. This philosophy allows the analysis of natural circulation behavior in single-phase flow conditions proportionally to the functioning real conditions of a nuclear reactor. The experiment was performed through procedures to initialization of hydraulic feeding of primary and secondary circuits and electrical energizing of resistors installed inside heater. Power controller has availability to adjust values of electrical power to feeding resistors, in order to portray several conditions of energy decay of nuclear reactor in a steady state. Data acquisition system allows the measurement and monitoring of the evolution of the temperature in various points through thermocouples installed in strategic points along hydraulic circuit. The behavior of the natural circulation phenomenon was monitored by graphical interface on computer screen, showing the temperature evolutions of measuring points and results stored in digital spreadsheets. The results stored in digital spreadsheets allowed the getting of data to graphic construction and discussion about natural circulation phenomenon. Finally, the calculus of Reynolds number allowed the establishment for a correlation of friction in function of geometric scales of length, heights and cross section of tubing, considering a natural circulation flow throughout in the region of hot leg. (author)

  18. Engineering Judgment and Natural Circulation Calculations

    International Nuclear Information System (INIS)

    Ferreri, J.C.; Ferreri, J.C.

    2011-01-01

    The analysis performed to establish the validity of computer code results in the particular field of natural circulation flow stability calculations is presented in the light of usual engineering practice. The effects of discretization and closure correlations are discussed and some hints to avoid undesired mistakes in the evaluations performed are given. Additionally, the results are presented for an experiment relevant to the way in which a (small) number of skilled, nuclear safety analysts and researchers react when facing the solution of a natural circulation problem. These results may be also framed in the concept of Engineering Judgment and are potentially useful for Knowledge Management activities.

  19. Natural circulation under severe accident conditions

    International Nuclear Information System (INIS)

    Pafford, D.J.; Hanson, D.J.; Tung, V.X.; Chmielewski, S.V.

    1992-01-01

    Research is being conducted to better understand natural circulation phenomena in mixtures of steam and noncondensibles and its influence on the temperature of the vessel internals and the hot leg, pressurizer surge line, and steam generator tubes. The temperature of these structures is important because their failure prior to reactor vessel lower head failure could reduce the likelihood of containment failure as a result of direct containment heating. Computer code calculations (MELPROG, SCDAP/RELAP5/MOD3) predict high fluid temperatures in the upper plenum resulting from in-vessel natural circulation. Using a simple model for the guide tube phenomena, high upper plenum temperatures are shown to be consistent with the relatively low temperatures that were deduced metallurgically from leadscrews removed from the TMI-2 upper plenum. Evaluation of the capabilities of the RELAP5/MOD3 computer code to predict natural circulation behavior was also performed. The code was used to model the Westinghouse natural circulation experimental facility. Comparisons between code calculations and results from experiments show good agreement

  20. Development of an ex-vessel corium debris bed with two-phase natural convection in a flooded cavity

    International Nuclear Information System (INIS)

    Kim, Eunho; Lee, Mooneon; Park, Hyun Sun; Moriyama, Kiyofumi; Park, Jin Ho

    2016-01-01

    Highlights: • For ex-vessel severe accidents in LWRs with wet-cavity strategy, development of debris bed with two-phase natural convection flow due to thermal characteristics of prototypic corium particles was investigated experimentally by using simulant particles and local air bubble control system. • Based on the experimental results of this study, an analytical model was established to describe the spreading of the debris bed in terms of two-phase flow and the debris injection parameters. • This model was then used to analyze the formation of debris beds at the reactor scale, and a sensitivity analysis was carried out based on key accident parameters. - Abstract: During severe accidents of light water reactors (LWRs), the coolability of relocated corium from the reactor vessel is a significant safety issue and a threat to the integrity of containment. With a flooded cavity, a porous debris bed is expected to develop on the bottom of the pool due to breakup and fragmentation of the melt jet. As part of the coolability assessment under accident conditions, the geometrical configuration of the debris bed is important. The Debris Bed Research Apparatus for Validation of the Bubble-Induced Natural Convection Effect Issue (DAVINCI) experimental apparatus facility was constructed to investigate the formation of debris beds under the influence of a two-phase flow induced by steam generation due to the decay heat of the debris bed. Using this system, five kilograms of stainless steel simulant debris were injected from the top of the water level, while air bubbles simulating the vapor flow were injected from the bottom of the particle catcher plate. The airflow rate was determined based on the quantity of settled debris, which will form a heat source due to the decay of corium. The radial distribution of the settled debris was examined using a ‘gap–tooth’ approach. Based on the experimental results of this study, an analytical model was established to

  1. Development of an ex-vessel corium debris bed with two-phase natural convection in a flooded cavity

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Eunho; Lee, Mooneon; Park, Hyun Sun, E-mail: hejsunny@postech.ac.kr; Moriyama, Kiyofumi; Park, Jin Ho

    2016-03-15

    Highlights: • For ex-vessel severe accidents in LWRs with wet-cavity strategy, development of debris bed with two-phase natural convection flow due to thermal characteristics of prototypic corium particles was investigated experimentally by using simulant particles and local air bubble control system. • Based on the experimental results of this study, an analytical model was established to describe the spreading of the debris bed in terms of two-phase flow and the debris injection parameters. • This model was then used to analyze the formation of debris beds at the reactor scale, and a sensitivity analysis was carried out based on key accident parameters. - Abstract: During severe accidents of light water reactors (LWRs), the coolability of relocated corium from the reactor vessel is a significant safety issue and a threat to the integrity of containment. With a flooded cavity, a porous debris bed is expected to develop on the bottom of the pool due to breakup and fragmentation of the melt jet. As part of the coolability assessment under accident conditions, the geometrical configuration of the debris bed is important. The Debris Bed Research Apparatus for Validation of the Bubble-Induced Natural Convection Effect Issue (DAVINCI) experimental apparatus facility was constructed to investigate the formation of debris beds under the influence of a two-phase flow induced by steam generation due to the decay heat of the debris bed. Using this system, five kilograms of stainless steel simulant debris were injected from the top of the water level, while air bubbles simulating the vapor flow were injected from the bottom of the particle catcher plate. The airflow rate was determined based on the quantity of settled debris, which will form a heat source due to the decay of corium. The radial distribution of the settled debris was examined using a ‘gap–tooth’ approach. Based on the experimental results of this study, an analytical model was established to

  2. Two phase sampling

    CERN Document Server

    Ahmad, Zahoor; Hanif, Muhammad

    2013-01-01

    The development of estimators of population parameters based on two-phase sampling schemes has seen a dramatic increase in the past decade. Various authors have developed estimators of population using either one or two auxiliary variables. The present volume is a comprehensive collection of estimators available in single and two phase sampling. The book covers estimators which utilize information on single, two and multiple auxiliary variables of both quantitative and qualitative nature. Th...

  3. Natural circulation in pressurized water reactors

    International Nuclear Information System (INIS)

    Bastos, J.L.F.; Loureiro, L.V.; Rocha, R.T.V. da; Umbehaun, P.E.

    1992-01-01

    Several analytical modelling have been done for steady-state and slow transients conditions, besides more sophisticated studies considering two and three dimensional effects in a very simple geometry. Under severe accident conditions for PWR a code to analyse natural circulation has been developed by Westinghouse. This paper discusses the problem of natural circulation in a complex geometry similar to that of nuclear power plants. A first experiment has been done at the integral test facility of 'Co-ordination of Special Projects-Ministry of Naval Affairs' (Coordenadoria para Projetos Especiais -Ministerio da Marinha, COPESP) for several flux conditions. The results obtained were compared with numerical simulations for the steady-state regime. 09 refs, 05 figs, 01 tab. (B.C.A.)

  4. Engineering Judgment and Natural Circulation Calculations

    OpenAIRE

    Ferreri, J. C.

    2011-01-01

    The analysis performed to establish the validity of computer code results in the particular field of natural circulation flow stability calculations is presented in the light of usual engineering practice. The effects of discretization and closure correlations are discussed and some hints to avoid undesired mistakes in the evaluations performed are given. Additionally, the results are presented for an experiment relevant to the way in which a (small) number of skilled, nuclear safety analysts...

  5. Startup transient simulation for natural circulation boiling water reactors in PUMA facility

    International Nuclear Information System (INIS)

    Kuran, S.; Xu, Y.; Sun, X.; Cheng, L.; Yoon, H.J.; Revankar, S.T.; Ishii, M.; Wang, W.

    2006-01-01

    In view of the importance of instabilities that may occur at low-pressure and -flow conditions during the startup of natural circulation boiling water reactors, startup simulation experiments were performed in the Purdue University Multi-Dimensional Integral Test Assembly (PUMA) facility. The simulations used pressure scaling and followed the startup procedure of a typical natural circulation boiling water reactor. Two simulation experiments were performed for the reactor dome pressures ranging from 55 kPa to 1 MPa, where the instabilities may occur. The experimental results show the signature of condensation-induced oscillations during the single-phase-to-two-phase natural circulation transition. The results also suggest that a rational startup procedure is needed to overcome the startup instabilities in natural circulation boiling water reactor designs

  6. Capability of the RELAP5 code to simulate natural circulation behaviour in test facilities

    International Nuclear Information System (INIS)

    Mangal, Amit; Jain, Vikas; Nayak, A.K.

    2011-01-01

    In the present study, one of the extensively used best estimate code RELAP5 has been used for simulation of steady state, transient and stability behavior of natural circulation based experimental facilities, such as the High-Pressure Natural Circulation Loop (HPNCL) and the Parallel Channel Loop (PCL) installed and operating at BARC. The test data have been generated for a range of pressure, power and subcooling conditions. The computer code RELAP5/MOD3.2 was applied to predict the transient natural circulation characteristics under single-phase and two-phase conditions, thresholds of flow instability, amplitude and frequency of flow oscillations for different operating conditions of the loops. This paper presents the effect of nodalisation in prediction of natural circulation behavior in test facilities and a comparison of experimental data in with that of code predictions. The errors associated with the predictions are also characterized

  7. Prediction to natural circulation in semiscale SBLOCA test, S-NC-8B

    International Nuclear Information System (INIS)

    Bang, Young Seok; Seul, Kwang Won; Lee, Sukho; Kim, Hho Jung

    1995-01-01

    Natural circulation and the associated thermal-hydraulic behavior are predicted by RELAP5/MOD3.1 code against the test S-NC-8B, which simulated 0.1% equivalent SBLOCA in PWR. The Semiscale Mod-2A facility and the test-specific initial/boundary condition are modeled. The calculation result is compared with the experiment data in terms of natural circulation characteristic and the code predictability is evaluated on natural circulation. As a result, flow rate during single-and two-phase natural circulation modes is well predicted and slightly overpredicted with oscillation in transition and reflux regimes. Additional sensitivity calculations are attempted with different discharge coefficient and break modeling to investigate the break flow effect

  8. Non linear stability analysis of parallel channels with natural circulation

    Energy Technology Data Exchange (ETDEWEB)

    Mishra, Ashish Mani; Singh, Suneet, E-mail: suneet.singh@iitb.ac.in

    2016-12-01

    Highlights: • Nonlinear instabilities in natural circulation loop are studied. • Generalized Hopf points, Sub and Supercritical Hopf bifurcations are identified. • Bogdanov–Taken Point (BT Point) is observed by nonlinear stability analysis. • Effect of parameters on stability of system is studied. - Abstract: Linear stability analysis of two-phase flow in natural circulation loop is quite extensively studied by many researchers in past few years. It can be noted that linear stability analysis is limited to the small perturbations only. It is pointed out that such systems typically undergo Hopf bifurcation. If the Hopf bifurcation is subcritical, then for relatively large perturbation, the system has unstable limit cycles in the (linearly) stable region in the parameter space. Hence, linear stability analysis capturing only infinitesimally small perturbations is not sufficient. In this paper, bifurcation analysis is carried out to capture the non-linear instability of the dynamical system and both subcritical and supercritical bifurcations are observed. The regions in the parameter space for which subcritical and supercritical bifurcations exist are identified. These regions are verified by numerical simulation of the time-dependent, nonlinear ODEs for the selected points in the operating parameter space using MATLAB ODE solver.

  9. Passive decay heat removal by natural circulation

    International Nuclear Information System (INIS)

    Vijayan, P.K.; Venkat Raj, V.; Kakodkar, A.; Mehta, S.K.

    1990-01-01

    The standardised 235 MWe PHWRs being built in India are the pressure tube type, heavy water moderated, heavy water cooled and natural uranium fuelled reactors. Several passive safety features are incorporated in these reactors. These include: (1) Containment pressure reduction and fission product trapping with the help of suppression pool following LOCA. (2) Emergency coolant injection by means of accumulators. (3) Large heat sink provided by the low temperature moderator under accident conditions. (4) Low excess reactivity, through the use of natural uranium fuel and on power fuelling. (5) Residual heat removal by means of natural circulation, etc. of which the last item is the subject matter of this report. (author). 8 refs, 10 figs

  10. Density wave oscillations of a boiling natural circulation loop induced by flashing

    Energy Technology Data Exchange (ETDEWEB)

    Furuya, Masahiro; Inada, Fumio; Yasuo, Akira [Central Research Institute of Electric Power Industry, Tokyo (Japan)

    1995-09-01

    Experiments are conducted to investigate two-phase flow instabilities in a boiling natural circulation loop with a chimney due to flashing in the chimney at lower pressure. The test facility used in this experiment is designed to have non-dimensional values which are nearly equal to those of natural circulation BWR. Stability maps in reference to the heat flux, the inlet subcooling, the system pressure are presented. This instability is suggested to be density wave oscillations due to flashing in the chimney, and the differences from other phenomena such as flow pattern oscillations and geysering phenomena are discussed by investigating the dynamic characteristics, the oscillation period, and the transient flow pattern.

  11. Experimental investigations in high-pressure natural circulation loop: progress report for the period January-June, 1999

    International Nuclear Information System (INIS)

    Naveen Kumar; Rajalakshmi, R.; Kulkarni, R.D.; Sagar, T.V.; Vijayan, P.K.; Saha, D.

    2000-02-01

    The Advanced Heavy Water Reactor employs natural circulation as the normal mode of coolant circulation. This is expected to enhance safety and reliability as it eliminates all safety issues associated with the pump failure. Two-phase natural circulation, however, is susceptible to several types of instabilities. In addition, the flow rate in a natural circulation loop is a dependent quantity and is not known a priori. Reliable calculations of the flow rate and stability behaviour are essential to ensure the success of AHWR design. Hence computer codes developed to predict the steady state flow rate and stability behaviour require validation against test data under natural circulation. For this purpose a high-pressure natural circulation loop has been designed, constructed and commissioned. Steady state experiments have been carried out in this loop to study the effect of pressure on natural circulation flow rate. The experimental results for this case are presented in this report. More experiments are planned in future to study the various aspects of two-phase natural circulation. (author)

  12. Instability of single-phase natural circulation

    International Nuclear Information System (INIS)

    Xie Heng; Zhang Jinling; Jia Dounan

    1997-01-01

    The author has investigated the instability of single-phase flows in natural circulation loops. The momentum equation and energy equation are made dimensionless according to some definitions, and some important dimensionless parameters are gotten. The authors decomposed the mean mass flowrate and temperature into a steady solution and a small disturbance equations. Through solving the disturbance equations, the authors get the neutral stability curves. The authors have studied the effect of the two parameters which represent the ratio of buoyancy force to the friction loss in the loop on the stability of loops. The authors also have studied the effect of the difference of height between the center of heat source and the heat sink on the stability

  13. Experiments on the basic behavior of supercritical CO{sub 2} natural circulation

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Guangxu [CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology, Nuclear Power Institute of China (China); Huang, Yanping, E-mail: hyanping007@163.com [CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology, Nuclear Power Institute of China (China); Wang, Junfeng; Lv, Fa [CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology, Nuclear Power Institute of China (China); Leung, Laurence K.H. [Canadian Nuclear Laboratories, 286 Plant Road, Chalk River, Ontario (Canada)

    2016-04-15

    Highlights: • Steady-state behavior of supercritical CO{sub 2} natural circulation was studied. • Effects of pressure and inlet temperature were carefully investigated. • No instabilities were found in present study. • The maximum of mass flow was obtained at outlet temperature much higher than T{sub pc}. • Inlet temperature has vital effect on mass flow rate. - Abstract: To study the steady-state characteristics of supercritical carbon dioxide natural circulation, experiments were carried out in a simple rectangular loop with vertically placed heating section. The effects of system pressure and inlet temperature on the system behavior were also investigated. No instabilities were found in the present experiments. The maximum of mass flow rate was obtained at a heating section outlet temperature much higher than the pseudo-critical temperature. The maximum value of mass flow rate increased with system pressure just as in two-phase natural circulation systems. Inlet temperature significantly affected the steady-state characteristics of supercritical carbon dioxide natural circulation system. A small temperature difference of 14 °C in the natural circulation system could induce a mass flow rate with considerably high Re up to 9.1 × 10{sup 4}, which indicates the potential for supercritical carbon dioxide to be used as a high efficient natural circulation working fluid.

  14. Power limit and quality limit of natural circulation reactor

    International Nuclear Information System (INIS)

    Zhao Guochang; Ma Changwen

    1997-01-01

    The circulation characteristics of natural circulation reactor in boiling regime are researched. It is found that, the circulation mass flow rate and the power have a peak value at a mass quality respectively. Therefore, the natural circulation reactor has a power limit under certain technological condition. It can not be increased steadily by continually increasing the mass quality. Corresponding to this, the mass quality of natural circulation reactor has a reasonable limit. The relations between the maximum power and the reactor parameters, such as the resistance coefficient, the working pressure and so on, are analyzed. It is pointed out that the power limit of natural circulation reactor is about 1000 MW at present technological condition. Taking the above result and low quality stability experimental result into account, the authors recommend that the reasonable mass quality of natural circulation reactor working in boiling regime is from 2% to 3% under the researched working pressure

  15. Capacitance sensor for void fraction measurement in a natural circulation refrigeration circuit

    International Nuclear Information System (INIS)

    Rocha, Marcelo S.; Cabral, Eduardo L.L.; Simoes-Moreira, Jose R.

    2009-01-01

    Natural circulation is widely used in nuclear reactors for residual heat refrigeration. In this work, a conductance probe is designed and constructed to measure the instantaneous bulk void fraction in a vertical tube section. This probe is installed in a natural circulation refrigeration loop designed to simulate a nuclear reactor primary refrigeration circuit. During the operation of the natural circulation loop several gas-liquid flow patterns are observed, including oscillatory flow. The instantaneous signal generated by the capacitance probe allows the calculation of the two-phase flow void fraction. The void fraction obtained by the probe will be compared with the theoretical void fraction calculated by the computational program RELAP5/MOD3.2.2 gamma. The probe design and electronics, as well as the previous results obtained are presented and discussed. (author)

  16. Reverse primary-side flow in steam generators during natural circulation cooling

    International Nuclear Information System (INIS)

    Stumpf, H.; Motley, F.; Schultz, R.; Chapman, J.; Kukita, Y.

    1987-01-01

    A TRAC model of the Large Scale Test Facility with a 3-tube steam-generator model was used to analyze natural-circulation test ST-NC-02. For the steady state at 100% primary mass inventory, TRAC was in excellent agreement with the natural-circulation flow rate, the temperature distribution in the steam-generator tubes, and the temperature drop from the hot leg to the steam-generator inlet plenum. TRAC also predicted reverse flow in the long tubes. At reduced primary mass inventories, TRAC predicted the three natural-circulation flow regimes: single phase, two phase, and reflux condensation. TRAC did not predict the cyclic fill-and-dump phenomenon seen briefly in the test. TRAC overpredicted the two-phase natural-circulation flow rate. Since the core is well cooled at this time, the result is conservative. An important result of the analysis is that TRAC was able to predict the core dryout and heatup at approximately the same primary mass inventory as in the test. 4 refs., 8 figs., 2 tabs

  17. Theoretical research for natural circulation operational characteristic of ship nuclear machinery under ocean conditions

    Energy Technology Data Exchange (ETDEWEB)

    Yan Binghuo [Department of Nuclear Science and Engineering, Naval University of Engineering, Wuhan 430033 (China)], E-mail: yanbh1986@163.com; Yu Lei [Department of Nuclear Science and Engineering, Naval University of Engineering, Wuhan 430033 (China)], E-mail: yulei301@163.com

    2009-06-15

    Based on the two-phase drift flux model and the multi-pressure nodes matrix solving method, natural circulation thermal hydraulic analysis models for the Nuclear Machinery (NM) under ocean conditions are developed. The neutron physical activities and the responses of the reactivity control systems are described by the two-group, 3-dimensional space and time dependent neutron kinetics model. Reactivity feedback is calculated by coupling the neutron physics and thermal hydraulic codes, and is tested by comparison with experiments. Using the models developed, the natural circulation operating characteristics of NM in rolling and pitching motions and the transitions between forced circulation (FC) to natural circulation (NC) are analyzed. The results show that the influence of the rolling motion increases as the rolling amplitude is increased, and as the rolling period becomes shorter. The results also show that for this NM, with the same rolling period and rolling angle, the influence of pitching motion on natural circulation is greater than that of rolling motion. Furthermore, the oscillation period for pitching motion is the same as the pitching period, while the oscillation period for rolling is one half of the rolling period. In the ocean environment, excessive flow oscillation of the natural circulation may cause the control rods to respond so frequently that the NM would not be able to realize the transition from the FC to NC steadily. However, the influence of ocean environment on the transition from NC to FC is limited.

  18. Theoretical investigation on the steady-state natural circulation characteristics of a new type of pressurized water reactor

    International Nuclear Information System (INIS)

    Gou Junli; Qiu Suizheng; Su Guanghui; Jia Dounan

    2006-01-01

    This article presents a theoretical investigation on the steady-state natural circulation characteristics of a new type of pressurized water reactor. Through numerically solving the one-dimensional steady-state single-phase conservative equations for the primary circuit and the steady-state two-phase drift-flux conservative equations for the secondary side of the steam generator, the natural circulation characteristics were studied. On the basis of the preliminary calculation analysis, it was found that natural circulation mass flow rate was proportional to the exponential function of the power and that the value of the exponent is related to the operating conditions of the secondary side of the steam generator. The higher the outlet pressure of the secondary side of the steam generator, the higher the primary natural circulation mass flow rate. The larger height difference between the core center and the steam generator center is favorable for the heat removal capacity of the natural circulation. (authors)

  19. Analysis and research on natural circulation capacity of HFETR

    International Nuclear Information System (INIS)

    Xu Taozhong; Duan Tianyuan

    2010-01-01

    For the operating characteristics of HFETR, the numerical model of HFETR was established by RELAP5/MOD3 to analysis the maximal natural circulation capacity. Combining with the reactor running condition, the influence of the system pressure was analyzed by ascending power in step method and the pool water temperature on natural circulation characteristics was analyzed by integral power method. The results show that the natural circulation capacity are 0.9 and 2.0 MW separately under low pressure and high pressure, the natural circulation capacity increases as the running pressure increases, however the natural circulation capacity decreases as the coolant temperature increases in the pressure vessel. Based on the computational result and the theoretical deduction, a correlation was proposed to predicate the relationship between the natural circulation mass flow and the core power under different coolant temperatures. (authors)

  20. Transient modelling of a natural circulation loop under variable pressure

    International Nuclear Information System (INIS)

    Vianna, Andre L.B.; Faccini, Jose L.H.; Su, Jian; Instituto de Engenharia Nuclear

    2017-01-01

    The objective of the present work is to model the transient operation of a natural circulation loop, which is one-tenth scale in height to a typical Passive Residual Heat Removal system (PRHR) of an Advanced Pressurized Water Nuclear Reactor and was designed to meet the single and two-phase flow similarity criteria to it. The loop consists of a core barrel with electrically heated rods, upper and lower plena interconnected by hot and cold pipe legs to a seven-tube shell heat exchanger of countercurrent design, and an expansion tank with a descending tube. A long transient characterized the loop operation, during which a phenomenon of self-pressurization, without self-regulation of the pressure, was experimentally observed. This represented a unique situation, named natural circulation under variable pressure (NCVP). The self-pressurization was originated in the air trapped in the expansion tank and compressed by the loop water dilatation, as it heated up during each experiment. The mathematical model, initially oriented to the single-phase flow, included the heat capacity of the structure and employed a cubic polynomial approximation for the density, in the buoyancy term calculation. The heater was modelled taking into account the different heat capacities of the heating elements and the heater walls. The heat exchanger was modelled considering the coolant heating, during the heat exchanging process. The self-pressurization was modelled as an isentropic compression of a perfect gas. The whole model was computationally implemented via a set of finite difference equations. The corresponding computational algorithm of solution was of the explicit, marching type, as for the time discretization, in an upwind scheme, regarding the space discretization. The computational program was implemented in MATLAB. Several experiments were carried out in the natural circulation loop, having the coolant flow rate and the heating power as control parameters. The variables used in the

  1. Transient modelling of a natural circulation loop under variable pressure

    Energy Technology Data Exchange (ETDEWEB)

    Vianna, Andre L.B.; Faccini, Jose L.H.; Su, Jian, E-mail: avianna@nuclear.ufrj.br, E-mail: sujian@nuclear.ufrj.br, E-mail: faccini@ien.gov.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Lab. de Termo-Hidraulica Experimental

    2017-07-01

    The objective of the present work is to model the transient operation of a natural circulation loop, which is one-tenth scale in height to a typical Passive Residual Heat Removal system (PRHR) of an Advanced Pressurized Water Nuclear Reactor and was designed to meet the single and two-phase flow similarity criteria to it. The loop consists of a core barrel with electrically heated rods, upper and lower plena interconnected by hot and cold pipe legs to a seven-tube shell heat exchanger of countercurrent design, and an expansion tank with a descending tube. A long transient characterized the loop operation, during which a phenomenon of self-pressurization, without self-regulation of the pressure, was experimentally observed. This represented a unique situation, named natural circulation under variable pressure (NCVP). The self-pressurization was originated in the air trapped in the expansion tank and compressed by the loop water dilatation, as it heated up during each experiment. The mathematical model, initially oriented to the single-phase flow, included the heat capacity of the structure and employed a cubic polynomial approximation for the density, in the buoyancy term calculation. The heater was modelled taking into account the different heat capacities of the heating elements and the heater walls. The heat exchanger was modelled considering the coolant heating, during the heat exchanging process. The self-pressurization was modelled as an isentropic compression of a perfect gas. The whole model was computationally implemented via a set of finite difference equations. The corresponding computational algorithm of solution was of the explicit, marching type, as for the time discretization, in an upwind scheme, regarding the space discretization. The computational program was implemented in MATLAB. Several experiments were carried out in the natural circulation loop, having the coolant flow rate and the heating power as control parameters. The variables used in the

  2. System identification on two-phase flow stability

    International Nuclear Information System (INIS)

    Wu Shaorong; Zhang Youjie; Wang Dazhong; Bo Jinghai; Wang Fei

    1996-01-01

    The theoretical principle, experimental method and results of interrelation analysis identification for the instability of two-phase flow are described. A completely new concept of test technology and method on two-phase flow stability was developed by using he theory of information science on system stability and system identification for two-phase flow stability in thermo-physics field. Application of this method would make it possible to identify instability boundary of two-phase flow under stable operation conditions of two-phase flow system. The experiment was carried out on the thermohydraulic test system HRTL-5. Using reverse repeated pseudo-random sequences of heating power as input signal sources and flow rate as response function in the test, the two-phase flow stability and stability margin of the natural circulation system are investigated. The effectiveness and feasibility of identifying two-phase flow stability by using this system identification method were experimentally demonstrated. Basic data required for mathematics modeling of two-phase flow and analysis of two-phase flow stability were obtained, which are useful for analyzing, monitoring of the system operation condition, and forecasting of two-phase flow stability in engineering system

  3. Study of the hydrodynamic stability of natural-circulation steam generators

    International Nuclear Information System (INIS)

    Olive, J.

    1981-01-01

    This report presents a mathematical model of a study of the stability of natural-circulation steam generators. The method used consists in linearizing the equations for the single-dimensional flow and integrating them by using Laplace's transformation. The properties of the two-phase fluids are described by a homegeneous model with slip. The results of the calculation are the transfer functions of the circulation loop and its own oscillation modes (period and damping). Comparison of the results obtained by this method with those from other existing methods in the case of a straight tube with forced flow have proved satisfactory. Lastly, the results of a parametric study on the stability of a natural-circulation steam generator for a PWR unit are presented. The results show that the model is capable of reproducing at least qualitatively the trends observed experimentally or obtained by other more complex theoretical models [fr

  4. The development and study on passive natural circulation

    International Nuclear Information System (INIS)

    Zhou Tao; Li Jingjing; Ju Zhongyun; Huang Yanping; Xiao Zejun

    2013-01-01

    Passive natural circulation is getting more and more important in the field of nuclear power engineering. This article cited a passive natural circulation in the nuclear power system application, analyzed the potential problems during operation, described current mathematical research methods of the reliability of passive natural cycle analysis, briefly summarized the advantages and disadvantages of these methods, and finally got an outlook of the direction of passive natural circulation. Since the presence of passive natural circulation may get failure, sufficient attention and active research should be paid in response to the physical process failure of the running passive natural circulation system and its reliability. To ensure system security during the operation, the operation process should combine active with non-dynamic; while selecting an accurate model, perfect passive reliability analysis methods to achieve accurate theoretical calculations and experimental verification. (authors)

  5. An experimental study on the flow instabilities and critical heat flux under natural circulation

    International Nuclear Information System (INIS)

    Kim, Yun Il

    1993-02-01

    This study has been carried out to investigate the hydrodynamic stabilities of natural circulation and to analyze Critical Heat Flux (CHF) characteristics for the natural and forced circulation. A low pressure experimental loop was constructed, and experiments under various conditions have been performed. In the experiments of the natural circulation, flow oscillations and the average mass flux have been observed. Several parameters such as heat flux, the inlet temperature of test section, friction valve opening and riser length have been varied in order to investigate their effects on the flow stability of the natural circulation system. The results show that the flow instability has strongly dependent on geometric conditions and operating parameters, the inlet temperature and the heat flux of test section. It was found that unstable region for the heat flux and the inlet temperature exists between the single-phase stable region of low heat and low inlet temperature and the two-phase stable region of very high heat flux and high inlet temperature. The CHF data from the natural and forced circulation experiments have been compared each other to identify the effects of the flow instabilities on the CHF for the natural circulation mode. The test conditions were low flow less than 70 kg/m 2 s of water in vertical round tube with diameter of 0.008m at near atmospheric pressure. In this study, no difference in CHF values is observed between natural and fored circulation. Since low flow usually has the oscillation characteristic of relatively low amplitude and high frequency, the effect of the flow instabilities on the CHF seems to be negligible

  6. Natural Circulation Phenomena and Modelling for Advanced Water Cooled Reactors

    International Nuclear Information System (INIS)

    2012-03-01

    The role of natural circulation in advanced water cooled reactor design has been extended with the adoption of passive safety systems. Some designs utilize natural circulation to remove core heat during normal operation. Most passive safety systems used in evolutionary and innovative water cooled reactor designs are driven by natural circulation. The use of passive systems based on natural circulation can eliminate the costs associated with the installation, maintenance and operation of active systems that require multiple pumps with independent and redundant electric power supplies. However, considering the weak driving forces of passive systems based on natural circulation, careful design and analysis methods must be employed to ensure that the systems perform their intended functions. Several IAEA Member States with advanced reactor development programmes are actively conducting investigations of natural circulation to support the development of advanced water cooled reactor designs with passive safety systems. To foster international collaboration on the enabling technology of passive systems that utilize natural circulation, in 2004 the IAEA initiated a coordinated research project (CRP) on Natural Circulation Phenomena, Modelling and Reliability of Passive Systems that Utilize Natural Circulation. Three reports were published within the framework of this CRP. The first report (IAEA-TECDOC-1474) contains the material developed for the first IAEA training course on natural circulation in water cooled nuclear power plants. The second report (IAEA-TECDOC-1624) describes passive safety systems in a wide range of advanced water cooled nuclear power plant designs, with the goal of gaining insights into system design, operation and reliability. This third, and last, report summarizes the research studies completed by participating institutes during the CRP period.

  7. Linear stability analysis of the gas injection augmented natural circulation of STAR-LM

    International Nuclear Information System (INIS)

    Yeon-Jong Yoo; Qiao Wu; James J Sienicki

    2005-01-01

    Full text of publication follows: A linear stability analysis has been performed for the gas injection augmented natural circulation of the Secure Transportable Autonomous Reactor - Liquid Metal (STAR-LM). Natural circulation is of great interest for the development of Generation-IV nuclear energy systems due to its vital role in the area of passive safety and reliability. One of such systems is STAR-LM under development by Argonne National Laboratory. STAR-LM is a 400 MWt class modular, proliferation-resistant, and passively safe liquid metal-cooled fast reactor system that uses inert lead (Pb) coolant and the advanced power conversion system that consists of a gas turbine Brayton cycle utilizing supercritical carbon dioxide (CO 2 ) to obtain higher plant efficiency. The primary loop of STAR-LM relies only on the natural circulation to eliminate the use of circulation pumps for passive safety consideration. To enhance the natural circulation of the primary coolant, STAR-LM optionally incorporates the additional driving force provided by the injection of noncondensable gas into the primary coolant above the reactor core, which is effective in removing heat from the core and transferring it to the secondary working fluid without the attainment of excessive coolant temperature at nominal operating power. Therefore, it naturally raises the concern about the natural circulation instability due to the relatively high temperature change in the core and the two-phase flow condition in the hot leg above the core. For the ease of analysis, the flow path of the loop was partitioned into five thermal-hydraulically distinct sections, i.e., heated core, unheated core, hot leg, heat exchanger, and cold leg. The one-dimensional single-phase flow field equations governing the natural circulation, i.e., continuity, momentum, and energy equations, were used for each section except the hot leg. For the hot leg, the one-dimensional homogeneous equilibrium two-phase flow field

  8. Natural circulation and stability performance of BWRs (NACUSP)

    International Nuclear Information System (INIS)

    Aguirre, C.; Caruge, D.; Castrillo, F.; Dominicus, G.; Geutjes, A.J.; Saldo, V.; Hagen, T.H.J.J. van der; Hennig, D.; Huggenberger, M.; Ketelaar, K.C.J.; Manera, A.; Munoz-Cobo, J.L.; Prasser, H.-M.; Rohde, U.; Royer, E.; Yadigaroglu, G.

    2005-01-01

    From the beginning of BWR technology it was realized that a BWR can become unstable under particular circumstances caused by a feedback between the thermal-hydraulics and the neutronics. This instability can result in oscillations of the power and the flow rate, which is an unwanted phenomenon. The NACUSP project addresses the stability issues in current and future BWRs by expanding the basic understanding through well structured testing and analyses of experimental data, by analyses of existing operational stability data from three different European reactors (Forsmark, Leibstadt, Cofrentes), by applying this knowledge via efficient models and validated computer codes to operating reactors and reactor designs, and by developing general guidelines for reactor operation and design on how to avoid BWR instabilities. In order to cover the parameter range as efficiently as possible, four existing, sophisticated thermohydraulic test facilities (CLOTAIRE [Gouirand, J.M., 1988. CLOTAIRE Program, description and manufacturing of the mock-up, CEA Cadarache, DRE/STRE/LGV 88-876.] DESIRE [van de Graaf, R., van der Hagen, T.H.J.J., Mudde, R.F., 1994. Two-phase flow scaling laws for a simulated BWR assembly. Nucl. Eng. Des. 148, 455-462.] CIRCUS [de Kruijf, W.J.M., van der Hagen, T.H.J.J., Mudde, R.F., 2000. CIRCUS; a natural circulation two-phase flow facility, Eurotherm Seminar No. 63, 6-8 September 1999 Genoa, Italy, 391-395] and PANDA [Dreier, J., Huggenberger, M., Aubert, C., Bandurski, T., Fischer, O., Healzer, J., Lomperski, S., Strassberger, H.-J., Varadi, G., Yadigaroglu, G., 1996. The PANDA facility and first test results, Kerntechnik 61, 214-222]) have been selected. To extrapolate from small-scale separate-effect testing conditions to full-scale integral reactor conditions one needs to rely on the performance of computer codes (MONA [Hoyer, N., 1994. MONA, a 7-Equation Transient two-phase flow model for LWR dynamics, Proceedings of the International Conference on

  9. A novel start-up procedure for natural-circulation boiling water reactors

    International Nuclear Information System (INIS)

    Annalisa Manera; Frank Schaefer

    2005-01-01

    Full text of publication follows: The elimination of recirculation pumps and associated systems, as proposed for natural-circulation Boiling Water Reactors (BWRs), allow a great simplification in the design of BWRs. On the other hand, it has been shown both experimentally and analytically that such a new reactor configuration makes the system susceptible to thermal-hydraulic instabilities during the start-up phase (so-called flashing-induced instabilities). Therefore, appropriate start-up procedures have to be planned to avoid instabilities in natural-circulation BWRs. Not many proposals of start-up procedures for natural-circulation BWRs are reported in literature, but all authors agree on the fact that the system should be pressurized before the transition to two-phase circulation is allowed. Nayak [1] and Jiang and coauthors [2] proposed to externally pressurize the system by injecting in the pressure vessel respectively steam produced in a separate boiler or nitrogen. Once the pressure in the reactor vessel is high enough, the reactor power can be increased to achieve two-phase natural circulation. Unfortunately, the procedure suggested by Nayak requires an external boiler of adequate volume and power and the related connecting piping to the reactor vessel, while the procedure suggested by Jiang and coauthors requires an additional system for the nitrogen storage and the related connecting piping to the reactor vessel. The external pressurization does not accomplish to the requirements of simplicity that are at the very base of natural circulation BWRs design and it is thus not recommendable. Cheung and Rao [3] suggested a start-up procedure in which the reactor is first filled with water at 80 deg. C at a pressure of 0.55 bar. The reactor is made critical and is pressurized in conditions of single-phase circulation up to a pressure of 63 bar. At this pressure a sudden transition to two-phase operation is achieved by opening the MSIVs (Main Steam Isolation

  10. SCDAP/RELAP5 applications to RCS natural circulation

    International Nuclear Information System (INIS)

    Bayless, P.D.

    1988-01-01

    The effects of natural circulation flows in the reactor coolant system during a TMLB' sequence were investigated. Both in-vessel circulation and hot leg countercurrent flow were modeled in the Surry nuclear power plant using the SCDAP/RELAP5 computer code. The transient was analyzed until after fuel rod relocation had begun. The delays in the onset of relocation resulting from the natural circulation flows were not significant compared to SCDAP/RELAP5 calculations without natural circulation modeled, but were large compared to the analyses presented in NUREG-1150. The most significant aspect of the natural circulations flows was the heating of ex-vessel structures. Surge line failure is likely to occur before the vessel is breached by the molten core, while steam generator tube failure is not expected

  11. An analytical and experimental investigation of natural circulation transients in a model pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Massoud, M

    1987-01-01

    Natural Circulation phenomena in a simulated PWR was investigated experimentally and analytically. The experimental investigation included determination of system characteristics as well as system response to the imposed transient under symmetric and asymmetric operations. System characteristics were used to obtain correlation for heat transfer coefficient in heat exchangers, system flow resistance, and system buoyancy heat. Asymmetric transients were imposed to study flow oscillation and possible instability. The analytical investigation encompassed development of mathematical model for single-phase, steady-state and transient natural circulation as well as modification of existing model for two-phase flow analysis of phenomena such as small break LOCA, high pressure coolant injection and pump coast down. The developed mathematical model for single-phase analysis was computer coded to simulate the imposed transients. The computer program, entitled ''Symmetric and Asymmetric Analysis of Single-Phase Flow (SAS),'' were employed to simulate the imposed transients. It closely emulated the system behavior throughout the transient and subsequent steady-state. Modifications for two-phase flow analysis included addition of models for once-through steam generator and electric heater rods. Both programs are faster than real time. Off-line, they can be used for prediction and training applications while on-line they serve for simulation and signal validation. The programs can also be used to determine the sensitivity of natural circulation behavior to variation of inputs such as secondary distribution and power transients.

  12. An analytical and experimental investigation of natural circulation transients in a model pressurized water reactor

    International Nuclear Information System (INIS)

    Massoud, M.

    1987-01-01

    Natural Circulation phenomena in a simulated PWR was investigated experimentally and analytically. The experimental investigation included determination of system characteristics as well as system response to the imposed transient under symmetric and asymmetric operations. System characteristics were used to obtain correlation for heat transfer coefficient in heat exchangers, system flow resistance, and system buoyancy heat. Asymmetric transients were imposed to study flow oscillation and possible instability. The analytical investigation encompassed development of mathematical model for single-phase, steady-state and transient natural circulation as well as modification of existing model for two-phase flow analysis of phenomena such as small break LOCA, high pressure coolant injection and pump coast down. The developed mathematical model for single-phase analysis was computer coded to simulate the imposed transients. The computer program, entitled ''Symmetric and Asymmetric Analysis of Single-Phase Flow (SAS),'' were employed to simulate the imposed transients. It closely emulated the system behavior throughout the transient and subsequent steady-state. Modifications for two-phase flow analysis included addition of models for once-through steam generator and electric heater rods. Both programs are faster than real time. Off-line, they can be used for prediction and training applications while on-line they serve for simulation and signal validation. The programs can also be used to determine the sensitivity of natural circulation behavior to variation of inputs such as secondary distribution and power transients

  13. Characteristic of onset of nucleate boiling in natural circulation

    International Nuclear Information System (INIS)

    Zhou Tao; Yang Ruichang; Liu Ruolei

    2006-01-01

    Two kinds of thermodynamics quality at onset of nucleate boiling with sub-cooled boiling were calculated for force circulation by using Bergles and Rohesenow method or Davis and Anderson method, and natural circulation by using Tsinghua University project group's empirical equations suggested in our natural circulation experiment at same condition. The characteristic of onset of nucleate boiling with subcooled boiling in natural circulation were pointed out. The research result indicates that the thermodynamics quality at onset of nucleate boiling with subcooled boiling in natural circulation is more sensitive for heat and inlet temperature and system pressure. Producing of onset of nucleate boiling with subcooled boiling is early at same condition. The research result also indicates more from microcosmic angle of statistical physics that the phenomena are caused by the effects of characteristic of dissipative structure of natural circulation in self organization, fluctuation force and momentum force of dynamics on thermodynamics equilibrium. these can lay good basis for study and application on sub-cooled boiling in natural circulation in future. (authors)

  14. Dynamics and developing of natural circulation cooling from vertical upflow and downflow conditions

    International Nuclear Information System (INIS)

    Yang, B.W.; Ouyang, W.

    2004-01-01

    Several research programs have been conducted to evaluate the capability of natural circulation cooling of reactors following a loss of cooling accident. Both experimental and RELAP5 simulation results were obtained for these studies in a facility with vertical heated tube(s) and a unheated bypass channel. The analytical results showed that, under a certain power level, a natural circulation pattern can be developed from both initial upflow and downflow conditions, and maintained for a significant cooling period. This power level, for the discussion of this paper, is defined as the natural circulation cooling (NCC) power limit. Two import factors, namely the pump coastdown rate and the initial flow direction, are examined in this paper. In the benchmark case, as compared to the experimental results, the RELAP5 simulation program accurately predicted the transient phenomena from forced convection through flow reversal, then, into natural circulation cooling. Generally, the two-phase NCC power limit is higher and also more stable for the cases with initial upflow forced convection than for the cases with initial downflow. The transient phenomena (dynamics) of the natural circulation cooling was examined by varying the pump coast down rate in approaching the flow reversal natural circulation. A significant pump coastdown effect on the NCC power limit was observed for the analytical tests with initial downflow forced convection. For the tests with initial downflow condition, the higher the coastdown rate (or the shorter the coastdown period), the higher the NCC power limit. For the case with initial upflow forced convection, there may be an optimal coastdown rate for a given subcooled condition. However, for the subcooled condition used in this study, the effect of pump coast down rate is not as significant as in the downward forced convection. (author)

  15. Development and verification of a space-dependent dynamic model of a natural circulation steam generator

    International Nuclear Information System (INIS)

    Mewdell, C.G.; Harrison, W.C.; Hawley, E.H.

    1980-01-01

    This paper describes the development and verification of a Non-Linear Space-Dependent Dynamic Model of a Natural Circulation Steam Generator typical of boilers used in CANDU nuclear power stations. The model contains a detailed one-dimensional dynamic description of both the primary and secondary sides of an integral pre-heater natural circulation boiler. Two-phase flow effects on the primary side are included. The secondary side uses a drift-flux model in the boiling sections and a detailed non-equilibrium point model for the steam drum. The paper presents the essential features of the final model called BOILER-2, its solution scheme, the RD-12 loop and test boiler, the boiler steady-state and transient experiments, and the comparison of the model predictions with experimental results. (author)

  16. Transient analysis of a U-tube natural circulation steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Gaikwad, A J; Kumar, Rajesh; Bhadra, Anu; Chakraborty, G; Venkat Raj, V [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai (India)

    1994-06-01

    A computer code has been developed, for transient thermal-hydraulic analysis of proposed 500 MWe PHWR steam generator. The transient behaviour of a nuclear power plant is very much dependent on the steam generator performance, as it provides a thermal linkage between the primary and secondary systems. Study of dynamics of steam generator is essential for over all power plant dynamics as well as design of control systems for steam generator. A mathematical model has been developed for the simulation of thermal-hydraulic phenomena in a U tube natural circulation steam generator. Fluid model is based on one dimensional, nonlinear, single fluid conservation equations of mass, momentum, energy and equation of state. This model includes coupled two phase flow heat transfer and natural circulation. The model accounts for both compressibility and thermal expansion effects. The process simulation and results obtained for transients such as step change in load and total loss of feed water are presented. (author). 5 refs., 7 figs.

  17. Natural circulation cooling in US pressurized water reactors

    International Nuclear Information System (INIS)

    Berta, V.T.; Wilson, G.E.; Boyack, B.E.

    1989-01-01

    The research into the modes of, and heat removed by, natural circulation in PWR systems is reviewed for the purpose of determining the status of this method for off-nominal recovery procedures. The referenced information comes from all facets of the nuclear industry, both domestic and international. The information focuses on recent research (1986--1988); however, pre-1986 research is summarized and referenced. Particular attention is paid to the role of scaling in the experimental facilities and analytical tools. Three modes of natural-circulation cooling are covered: condensation. The conclusion of the review is that the new research reconfirms the pre-1986 conclusion that natural circulation is a viable means of decay heat removal. In addition, the new research sufficiently completes the acquisition of an appropriate experimental data base and the development of system codes to permit the design of valid plant recovery procedures incorporating all three modes of natural circulation. 48 refs., 1 fig., 3 tabs

  18. performance simulation of a natural circulation solar air

    African Journals Online (AJOL)

    User

    in a single glazed flat plate natural circulation solar a prepared in modules .... Nigerian Journal of Technology, used instead of ... boundary associated with the melting the phase ...... Mathematical Modeling of the Thin Layer Drying of Sweet ...

  19. A RELAP5 study to identify flow regime in natural circulation phenomenon

    Energy Technology Data Exchange (ETDEWEB)

    Sabundjian, Gaiane; Torres, Walmir M.; Macedo, Luiz A.; Mesquita, Roberto N.; Andrade, Delvonei A.; Umbehaun, Pedro E.; Conti, Thadeu N.; Masotti, Paulo H.F.; Belchior Junior, Antonio; Angelo, Gabriel, E-mail: gdjian@ipen.b, E-mail: umbehaun@ipen.b, E-mail: wmtorres@ipen.b, E-mail: tnconti@ipen.b, E-mail: rnavarro@ipen.b, E-mail: lamacedo@ipen.b, E-mail: pmasotti@ipen.b, E-mail: abelchior@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    There has been a crescent interest in the scientific community in the study of natural circulation phenomenon. New generation of compact nuclear reactors uses the natural circulation of the fluid as a system of cooling and of residual heat removal in case of accident or shutdown. The objective of this paper is to compare the flow patterns of experimental data and numerical simulation for the natural circulation phenomenon in two-phase flow regime. An experimental circuit built with glass tubes is used for the experiments. Thus, it allows the thermal hydraulic phenomena visualization. There is an electric heater as the heat source, a heat exchanger as the heat sink and an expansion tank to accommodate fluid density excursions. The circuit instrumentation consists of thermocouples and pressure meters to better keep track of the flow and heat transfer phenomena. Data acquisition is performed through a computer interface developed with LABVIEW. The characteristic of the regime is identified using photography techniques. Numerical modeling and simulation is done with the thermal hydraulic code RELAP5, which is widely used for this purpose. This numerical simulation is capable to reproduce some of the flow regimes which are present in the circuit for the natural circulation phenomenon. Comparison between experimental and numerical simulation is presented in this work. (author)

  20. Natural Circulation Characteristics of a Symmetric Loop under Inclined Conditions

    Directory of Open Access Journals (Sweden)

    Xingtuan Yang

    2014-01-01

    Full Text Available Natural circulation is an important process for primary loops of some marine integrated reactors. The reactor works under inclined conditions when severe accidents happen to the ship. In this paper, to investigate the characteristics of natural circulation, experiments were conducted in a symmetric loop under the inclined angle of 0~45°. A CFD model was also set up to predict the behaviors of the loop beyond the experimental scope. Total circulation flow rate decreases with the increase of inclined angle. Meanwhile one circulation is depressed while the other is enhanced, and accordingly the disparity between the branch circulations arises and increases with the increase of inclined angle. Circulation only takes place in one branch circuit at large inclined angle. Also based on the CFD model, the influences of flow resistance distribution and loop configuration on natural circulation are predicted. The numerical results show that to design the loop with the configuration of big altitude difference and small width, it is favorable to reduce the influence of inclination; however too small loop width will cause severe reduction of circulation ability at large angle inclination.

  1. Experimental Study on the Natural Circulation Characteristics in the Primary Loop of the SMART Reactor by using the VISTA Facility

    International Nuclear Information System (INIS)

    Park, Hyun-Sik; Choi, Ki-Yong; Cho, Seok; Yi, Sung-Jae; Park, Choon-Kyung; Chung, Moon-Ki

    2007-01-01

    The SMART uses a two-phase natural circulation in the PRHRS loop to remove the heat from the steam generators to the PRHRS heat exchangers, while a single phase natural circulation occurs in the primary loop to transfer the decay heat from the core to the steam generator. Natural circulation operation with a power range of 20 ∼ 25% was considered for SMART and nowadays the possibility of increasing the power level during the natural circulation operation is being investigated. Previously Park et al. performed several experiments by using the VISTA facility on the thermal-hydraulic characteristics of the PRHRS for the SMART-P, which includes a single-phase natural circulation in the primary loop. From the analysis with the TASS-SMR code it was shown that the reference temperature for the primary steam generator inlet temperature should be increased in order to compensate for the decreased core flow. To investigate the possibility of an increase of the power and reference temperature, it is necessary to get experimental data to characterize the natural circulation phenomena in the primary loop of the SMART. In this paper, the characteristics of natural circulation in the primary loop are experimentally investigated during various operational conditions by using the VISTA facility

  2. Calculation analysis on steady state natural circulation characteristics

    International Nuclear Information System (INIS)

    Wang Fei; Nie Changhua; Huang Yanping

    2005-01-01

    The calculation results of single-phase steady state natural circulation characteristics by using Retran02 code have been presented, good agreement is achieved between the verified calculation result and the experimental data which were conducted at a test facility. Based on the calculation model, some sensibility analyses were made and much deeper understanding for single-phase steady state natural circulation characteristics was obtained. (author)

  3. Cooling Performance of Natural Circulation for a Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Suki; Chun, J. H.; Yum, S. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    This paper deals with the core cooling performance by natural circulation during normal operation and a flow channel blockage event in an open tank-in-pool type research reactor. The cooling performance is predicted by using the RELAP5/ MOD3.3 code. The core decay heat is usually removed by natural circulation to the reactor pool water in open tank-in-pool type research reactors with the thermal power less than several megawatts. Therefore, these reactors have generally no active core cooling system against a loss of normal forced flow. In reactors with the thermal power less than around one megawatt, the reactor core can be cooled down by natural circulation even during normal full power operation. The cooling performance of natural circulation in an open tank-in-pool type research reactor has been investigated during the normal natural circulation and a flow channel blockage event. It is found that the maximum powers without void generation at the hot channel are around 1.16 MW and 820 kW, respectively, for the normal natural circulation and the flow channel blockage event.

  4. Research on enhancement of natural circulation capability in lead–bismuth alloy cooled reactor by using gas-lift pump

    Energy Technology Data Exchange (ETDEWEB)

    Zuo, Juanli, E-mail: Jenyzuo@163.com; Tian, Wenxi, E-mail: wxtian@mail.xjtu.edu.cn; Chen, Ronghua, E-mail: ronghua.chen@stu.xjtu.edu.cn; Qiu, Suizheng; Su, Guanghui, E-mail: ghsu@mail.xjtu.edu.cn

    2013-10-15

    Highlights: • The gas-lift pump has been adopted to enhance the natural circulation capability. • LENAC code is developed in my study. • The calculation results by LENAC code show good agreement with experiment results. • Gas mass flow rate, bubble diameter, rising pipe length are important parameters. -- Abstract: The gas-lift pump has been adopted to enhance the natural circulation capability in the type of lead–bismuth alloy cooled reactors such as Accelerator Driven System (ADS) and Liquid–metal Fast Reactor (LMFR). The natural circulation ability and the system safety are obviously influenced by the two phase flow characteristics of liquid metal–inert gas. In this study, LENAC (LEad bismuth alloy NAtural Circulation capability) code has been developed to evaluate the natural circulation capability of lead–bismuth cooled ADS with gas-lift pump. The drift flow theory, void fraction prediction model and friction pressure drop prediction model have been incorporated into LENAC code. The calculation results by LENAC code show good agreement with experiment results of CIRCulation Experiment (CIRCE) facility. The effects of the gas mass flow rate, void fraction, gas quality, bubble diameter and the rising pipe height or the potential difference between heat exchanger and reactor core on natural circulation capability of gas-lift pump have been analyzed. The results showed that in bubbly flow pattern, for a fixed value of gas mass flow rate, the natural circulation capability increased with the decrease of the bubble diameter. In the bubbly flow, slug flow, churn flow and annular flow pattern, with the gas mass flow rate increasing, the natural circulation capability initially increased and then declined. And the flow parameters influenced the thermal hydraulic characteristics of the reactor core significantly. The present work is helpful for revealing the law of enhancing the natural circulation capability by gas-lift pump, and providing theoretical

  5. Experimental study on thermo-hydraulic instability on reduced-moderation natural circulation BWR concept

    International Nuclear Information System (INIS)

    Watanabe, Noriyuki; Subki, M.H.; Kikura, Hiroshige; Aritomi, Masanori

    2003-01-01

    Reduced-moderation natural circulation BWR has been promoted to solve the recent challenges in BWR nuclear power technology problems as one of advanced small and medium-sized reactors equipped with the passive safety features in conformity with the natural law. However, the elimination of recirculation pumps and a high-density core due to the increase of conversion ratio could cause various thermo-hydraulic instabilities especially during the start-up stage. The occurrences of the thermo-hydraulic instabilities are not desirable and it is one of the main challenges in establishing reduced-moderation natural circulation BWR as a commercial reactor. The purpose of this present study is to experimentally investigate the driving mechanism of the thermo-hydraulic instabilities and the effect of system pressure on the unstable flow patterns. Hence, as the fundamental research for this study, a natural circulation loop that carries boiling fluid with parallel boiling channel has been constructed. Channel gap that has been set at 2 mm in order to simulate reduced-moderation reactor core. Pressure ranges of 0.1 up to 0.7 MPa, input heat flux range of 0 ou to 577 kW/m 2 , and inlet subcooling temperatures of 5, 10, and 15 K respectively, are imposed in the experiments. This experiment clarifies that changes in unstable flow patterns with increase in heat flux can be classified into two in response to system pressure range. In case of atmospheric pressure, unstable flow patters has been classified in beyond order, (1) in-phase geysering, (2) transition oscillation combined with both features of in-phase geysering and natural circulation oscillation, (3) natural circulation oscillation induced by hydrostatic head fluctuation, (4) density wave oscillation, and finally (5) stable boiling two-phase flow. On the other hand, in the system pressure range from 0.2 to 0.7 MPa, unstable patters have been dramatically changed in the following order (1) out-of-phase geysering, (2

  6. Experimental investigation on natural circulation and air-injection enhanced circulation in a simple loop

    International Nuclear Information System (INIS)

    Walter Ambrosini; Nicola Forgione; Francesco Oriolo; Filippo Pellacani; Mariano Tarantino; Claudio Struckmann

    2005-01-01

    Full text of publication follows: Natural circulation represents an interesting phenomenon because of both the complex aspects characterising it and for the widespread application in industry. On the other hand, injection of a gas into a rising branch of a loop represents a means to establish or to enhance a circulation flow, as it occurs in the so-called 'air-lift' loops. Both natural circulation and gas-injection enhanced circulation are presently considered for cooling Accelerator Driven System (ADS) reactors. These are subcritical reactors in which the fission reaction chain is maintained by the injection of neutrons obtained by spallation reactions in a target through a high energy proton beam generated in an external accelerator. The capability of such reactors to be used as incinerators of long lived fission products makes them particularly interesting in the light of the closure of the nuclear fuel cycle. Some of the fluids proposed as coolants for these reactors are liquid metals, with main interest for lead and lead-bismuth eutectic (LBE). Experimental activities are being performed in support to the design of the reactor prototype by different organisations. The university of Pisa, in addition to provide cooperation in these large scale activities performed with LBE has set up a specific experimental program aimed at studying the fundamental mechanisms involved in natural circulation and gas-injection enhanced circulation. The adopted experimental facility consists in a simple loop, having a rectangular lay-out (roughly, 4 m tall and 1 m wide), equipped with a 5 kW, 1 m tall heater, a 2 m long pipe-in-pipe heat exchanger, an air injection device and a separator. The fluid adopted in the tests performed up to now is water, though studies for evaluating the feasibility of the adoption of different fluids have been undertaken. Experimental data reported in previous publications concerning this research were related to a relatively high range of gas

  7. Experimental study of critical heat flux enhancement with hypervapotron structure under natural circulation conditions

    Energy Technology Data Exchange (ETDEWEB)

    Hou, Fangxin [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing (China); Chang, Huajian [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing (China); State Nuclear Power Technology R& D Center (Beijing), Beijing (China); Zhao, Yufeng, E-mail: zhaoyufeng@snptc.com.cn [State Nuclear Power Technology R& D Center (Beijing), Beijing (China); Zhang, Ming; Gao, Tianfang [State Nuclear Power Technology R& D Center (Beijing), Beijing (China); Chen, Peipei [State Power Investment Corporation, Beijing (China)

    2017-05-15

    Highlights: • Natural circulation tests are performed to study the effect of hypervapotron on CHF. • Hypervapotron structure improves CHF under natural circulation conditions. • Visualization data illustrate vapor blanket behavior under subcooled flow conditions. - Abstract: The enhancement of critical heat flux with a hypervapotron structure under natural circulation conditions is investigated in this study. Subcooled flow boiling CHF experiments are performed using smooth and hypervapotron surfaces at different inclination angles under natural circulation conditions. The experimental facility, TESEC (Test of External Vessel Surface with Enhanced Cooling), is designed to conduct CHF experiments in a 30 mm by 61 mm rectangular flow channel with a 200 mm long heated surface along the flow direction. The two-phase flow of subcooled flow boiling on both smooth and hypervapotron heating plates is observed and analyzed by the high-speed visualization technology. The results show that both smooth surface and hypervapotron surface CHF data exhibit a similar trend against inclination angles compared with the CHF results under forced flow condition on the same facility in earlier studies. However, the CHF enhancement of the hypervapotron structure is evidently more significant than the one under forced flow conditions. The experiments also indicate that the natural flow rates are higher with hypervapotron structure. The initiation of CHF is analyzed under transient subcooling and flow rate conditions for both smooth and hypervapotron heating surfaces. An explanation is given for the significant enhancement effect caused by the hypervapotron surface under natural circulation conditions. The visualization data are exhibited to demonstrate the behavior of the vapor blanket at various inclination angles and on different surfaces. The geometric data of the vapor blanket are quantified by an image post-processing method. It is found that the thickness of the vapor blanket

  8. An efficiency booster for energy conversion in natural circulation loops

    International Nuclear Information System (INIS)

    Wang, Dongqing; Jiang, Jin

    2016-01-01

    Highlights: • Low driving power conversion efficiency of natural circulation loops is proved. • The low conversion efficiency leads to low heat transfer capacity of such loops. • An efficiency booster is designed with turbine to increase the efficiency. • Performance of the proposed booster has been numerically simulated. • The booster drastically enhances heat transfer capacity of such loops. - Abstract: In this paper, the capacity of a natural circulation loop for transferring heat from a heat source to a heat sink has been analyzed. It is concluded that the capacity of the natural circulation loop depends on the conversion efficiency of the thermal energy from the heat source to the driving force for the circulation of the flow. The low conversion efficiency leading to weak driving force in such loops has been demonstrated analytically and validated through simulation results. This issue has resulted in a low heat transfer capacity in the circulation loop. To increase the heat transfer capacity, one has to improve this efficiency. To meet such a need, a novel efficiency booster has been developed in this paper. The booster essentially increases the flow driving force and hence significantly improves the overall heat transfer capacity. Design and analysis of this booster have been performed in detail. The performance has been examined through extensive computer simulations. It is concluded that the booster can indeed drastically improve the heat transfer capacity of the natural circulation loop.

  9. An efficiency booster for energy conversion in natural circulation loops

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Dongqing, E-mail: wangdongqing@stu.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shaanxi 710049 (China); Beijing Computational Science Research Center, Beijing 100084 (China); Jiang, Jin, E-mail: jjiang@eng.uwo.ca [Department of Electrical and Computer Engineering, The University of Western Ontario, London, Ontario N6A 5B9 (Canada); Beijing Computational Science Research Center, Beijing 100084 (China)

    2016-08-01

    Highlights: • Low driving power conversion efficiency of natural circulation loops is proved. • The low conversion efficiency leads to low heat transfer capacity of such loops. • An efficiency booster is designed with turbine to increase the efficiency. • Performance of the proposed booster has been numerically simulated. • The booster drastically enhances heat transfer capacity of such loops. - Abstract: In this paper, the capacity of a natural circulation loop for transferring heat from a heat source to a heat sink has been analyzed. It is concluded that the capacity of the natural circulation loop depends on the conversion efficiency of the thermal energy from the heat source to the driving force for the circulation of the flow. The low conversion efficiency leading to weak driving force in such loops has been demonstrated analytically and validated through simulation results. This issue has resulted in a low heat transfer capacity in the circulation loop. To increase the heat transfer capacity, one has to improve this efficiency. To meet such a need, a novel efficiency booster has been developed in this paper. The booster essentially increases the flow driving force and hence significantly improves the overall heat transfer capacity. Design and analysis of this booster have been performed in detail. The performance has been examined through extensive computer simulations. It is concluded that the booster can indeed drastically improve the heat transfer capacity of the natural circulation loop.

  10. Reevaluation of Kori Unit 4 Natural Circulation Test

    Energy Technology Data Exchange (ETDEWEB)

    Yassin, Nassir [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Woo, Sweng Woong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-10-15

    The simulation results showed that the natural circulation flow developed by density difference was capable of removing decay heat from the fuel rod. The maximum pellet centerline temperature of the hot channel showed large margin to the pellet melting temperature. The maximum coolant temperature in the hot channel was well below the saturation temperature. If steam generators provide heat sink to the primary coolant system and thus natural circulation is maintained, the integrity of the fuel in the core can be sustained with large margin. Passive cooling of reactor is inevitable in case of failures in forced cooling system such as loss of electric power for cooling pumps. Fukushima accident showed the importance of the passive core cooling. During the commissioning test of PWRs, natural circulation test is performed to demonstrate the passive core cooling by natural convection. The driving force for coolant flow is developed by the density deference along the loop multiplied by the gravitation. Using the data from 'natural circulation test' and 'RCS flow coast down test' of Kori Unit 4, fuel behavior was reevaluated by FRAPTRAN code. RCS natural circulation test of Kori Unit 4 was reevaluated by FRAPTYRAN simulation to study the fuel behavior during the flow coast down transient and at the equilibrium condition in which decay heat transport and RCS flow were stabilized.

  11. A study of natural circulation cooling using a flow visualization rig

    International Nuclear Information System (INIS)

    Bowman, W.C.; Ferch, R.L.; Omar, A.M.

    1985-01-01

    A flow visualization rig has been built at Monserco Limited to provide visual insight into the thermalhydraulic phenomena which occur during single phase and two phase thermosyphoning in a figure-of-eight heat transport loop. Tests performed with the rig have provided design information for the scaling and instrumentation of a high pressure rig being investigated for simulating CANDU reactor conditions during natural circulation cooling. A videotape was produced, for viewing at this presentation, to show important thermalhydraulic features of the thermosyphoning process. The rig is a standard figure-of-eight loop with two steam generators and three heated channels per pass. An elevated surge tank open to atmosphere was used for pressure control. Two variable speed pumps provided forced circulation for warming up the rig, and for establishing the desired initial conditions for testing. Test rig power could be varied between 0 and 15 kW

  12. Experimental analysis of the natural circulation phenomenon at the monophase system

    International Nuclear Information System (INIS)

    Santos, Thiago A. dos; Stefanni, Giovanni Laranjo de; Conti, Thadeu das Neves

    2011-01-01

    The present work study the phenomenology of the natural circulation, which is the flow circulation without help of any mechanical device. One of the possible application of this study would be a new way of nuclear reactor cooling, and this practice is fundamental for the maintenance and safe of the reactor. For this study, a therma balance in the circuit was performed, which consist of evaluate the behavior of the circuit, observing if not exist excessive energy loss. The balance was made only for power values considered small (up to 4000 W), were the fluid is at the monophasic state. This methodology is extremely important for the evaluation of the equipment and determining therefore if the energy is conserved in order to work with a more complex system such as the two-phase one

  13. Supercritical water natural circulation flow stability experiment research

    Energy Technology Data Exchange (ETDEWEB)

    Ma, Dongliang; Zhou, Tao; Li, Bing [North China Electric Power Univ., Beijing (China). School of Nuclear Science and Engineering; North China Electric Power Univ., Beijing (China). Inst. of Nuclear Thermalhydraulic Safety and Standardization; North China Electric Power Univ., Beijing (China). Beijing Key Lab. of Passive Safety Technology for Nuclear Energy; Huang, Yanping [Nuclear Power Institute of China, Chengdu (China). Science and Technology on Reactor System Design Technology Lab.

    2017-12-15

    The Thermal hydraulic characteristics of supercritical water natural circulation plays an important role in the safety of the Generation-IV supercritical water-cooled reactors. Hence it is crucial to conduct the natural circulation heat transfer experiment of supercritical water. The heat transfer characteristics have been studied under different system pressures in the natural circulation systems. Results show that the fluctuations in the subcritical flow rate (for natural circulation) is relatively small, as compared to the supercritical flow rate. By increasing the heating power, it is observed that the amplitude (and time period) of the fluctuation tends to become larger for the natural circulation of supercritical water. This tends to show the presence of flow instability in the supercritical water. It is possible to observe the flow instability phenomenon when the system pressure is suddenly reduced from the supercritical pressure state to the subcritical state. At the test outlet section, the temperature is prone to increase suddenly, whereas the blocking effect may be observed in the inlet section of the experiment.

  14. PWR hot leg natural circulation modeling with MELCOR code

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jae Hong; Lee, Jong In [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)

    1997-12-31

    Previous MELCOR and SCDAP/RELAP5 nodalizations for simulating the counter-current, natural circulation behavior of vapor flow within the RCS hot legs and SG U-tubes when core damage progress can not be applied to the steady state and water-filled conditions during the initial period of accident progression because of the artificially high loss coefficients in the hot legs and SG U-tubes which were chosen from results of COMMIX calculation and the Westinghouse natural circulation experiments in a 1/7-scale facility for simulating steam natural circulation behavior in the vessel and circulation modeling which can be used both for the liquid flow condition at steady state and for the vapor flow condition at the later period of in-vessel core damage. For this, the drag forces resulting from the momentum exchange effects between the two vapor streams in the hot leg was modeled as a pressure drop by pump model. This hot leg natural circulation modeling of MELCOR was able to reproduce similar mass flow rates with those predicted by previous models. 6 refs., 2 figs. (Author)

  15. PWR hot leg natural circulation modeling with MELCOR code

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jae Hong; Lee, Jong In [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)

    1998-12-31

    Previous MELCOR and SCDAP/RELAP5 nodalizations for simulating the counter-current, natural circulation behavior of vapor flow within the RCS hot legs and SG U-tubes when core damage progress can not be applied to the steady state and water-filled conditions during the initial period of accident progression because of the artificially high loss coefficients in the hot legs and SG U-tubes which were chosen from results of COMMIX calculation and the Westinghouse natural circulation experiments in a 1/7-scale facility for simulating steam natural circulation behavior in the vessel and circulation modeling which can be used both for the liquid flow condition at steady state and for the vapor flow condition at the later period of in-vessel core damage. For this, the drag forces resulting from the momentum exchange effects between the two vapor streams in the hot leg was modeled as a pressure drop by pump model. This hot leg natural circulation modeling of MELCOR was able to reproduce similar mass flow rates with those predicted by previous models. 6 refs., 2 figs. (Author)

  16. A study of the natural biodegradation of two-phase olive mill solid waste during its storage in an evaporation pond.

    Science.gov (United States)

    Borja, R; Sánchez, E; Raposo, F; Rincón, B; Jiménez, A M; Martín, A

    2006-01-01

    A study of the natural biodegradation of two-phase olive mill solid waste (OMSW) during its storage in an evaporation pond was carried out. This system is traditionally used as the final disposal of this waste or constitutes a previous stage before being processed by cogeneration or anaerobic digestion processes. A laboratory-scale pond with a total volume of 520 l was used. The experiment was carried out between December 2002 and August 2003 covering a total period of 269 days. During the experiment, the variations of temperature (outside and inside the pond), the effective volume of the OMSW in the pond, moisture content, total solids (TS), volatile solids (VS), the total chemical oxygen demand (TCOD), total volatile acids (TVA), pH and cumulative methane production as a function of the operation time were evaluated. The experimental results obtained showed that the characteristics of the two-phase OMSW in the pond were dependent on the time of operation and temperature. During the experiment, three periods were clearly observed and differentiated: (a) An initial period with a rapid reduction of moisture content and effective volume, with an increase of TS, VS and TCOD removals, a decrease of pH and an increase of TVA and methane production until day 21. (b) A second period of slow decrease of moisture content and effective volume, due to the decrease of the ambient temperature. This period took place between day 21 and days 120-150 of assay, and it was characterised by a slight change in the OMSW properties. (c) A third period where the characteristics of the waste in the pond changed considerably due to the increase of the temperature outside the pond. The methane gas production also showed an increase due to the increase of methanogenic activity with the increase of temperature. An empirical model to describe the changes of two-phase OMSW characteristics and methane production with the operation time was developed. The proposed model, based on the use of

  17. Study on the phenomena of natural circulation in LMFBR

    International Nuclear Information System (INIS)

    Takeda, Hirofumi; Koga, Tomonari

    1993-01-01

    Decay heat removal with natural circulation is to be introduced to the LMFBR operation under loss of the electric power supply. The natural circulation is highly reliable, but the phenomenon is essentially unstable and subtle, which makes fine prediction difficult. The difficulties of experimental prediction are explained by facts that the phenomena are ruled by the delicate balance between the buoyancy force and the low pressure loss and are influenced by the various parameters such as local geometry, heat capacity and so on. Therefore the similarity rule for the natural circulation has not been fully understood. This study has been conducted to establish the simulation method for the natural circulation phenomena and the detailed phenomena have been reviewed. For the natural circulation in an LMFBR plant, there are no readily available reference velocity and temperature. These values are related only with the heating and cooling rate, the characteristic length and physical properties of the testing fluid. Basic equations were transformed by these values, and dimensionless equations were derived and then two dimensionless numbers, the Gr' number and the Bo' number, were identified. In order to examine the similarity rule for natural circulation we performed experiments using the different scale water models, a 1/20th and a 1/6th model. The temperatures and velocities at typical points were measured in the transient condition with various heating rate as a parameter. Measured temperatures and velocities were transformed to dimensionless forms for comparison and the effects of the Bo' number and the Gr' number were examined. As a result, it was clarified that the effect of the Gr' number is negligibly small but the effect of Bo' number still remained in our experimental range. The Bo' number of an actual plant is within the range of this experiment. Accordingly similitude of the Bo' number becomes important in an experiment to simulate an actual plant. (author)

  18. Theoretical and experimental investigations into natural circulation behaviour in a simulated facility of the Indian PHWR under reduced inventory conditions

    International Nuclear Information System (INIS)

    Satish Kumar, N.V.; Nayak, A.K.; Vijayan, P.K.; Pal, A.K.; Saha, D.; Sinha, R.K.

    2004-01-01

    A theoretical and experimental investigation has been carried out to study natural circulation characteristics of an Indian PHWR under reduced inventory conditions. The theoretical model incorporates a quasi-steady state analysis of natural circulation at different system inventories. It predicts the system flow rate under single-phase and two-phase conditions and the inventory at which reflux condensation occurs. The model predictions were compared with test data obtained from FISBE (facility for integral system behaviour experiments), which simulates the thermal hydraulic behaviour of the Indian 220 MWe PHWR. The experimental results were found to be in close agreement with the predictions. It was also found that the natural circulation could be oscillatory under reduced inventory conditions. (orig.)

  19. Southern Hemisphere extratropical circulation: Recent trends and natural variability

    Science.gov (United States)

    Thomas, Jordan L.; Waugh, Darryn W.; Gnanadesikan, Anand

    2015-07-01

    Changes in the Southern Annular Mode (SAM), Southern Hemisphere (SH) westerly jet location, and magnitude are linked with changes in ocean circulation along with ocean heat and carbon uptake. Recent trends have been observed in these fields but not much is known about the natural variability. Here we aim to quantify the natural variability of the SH extratropical circulation by using Coupled Model Intercomparison Project Phase 5 (CMIP5) preindustrial control model runs and compare with the observed trends in SAM, jet magnitude, and jet location. We show that trends in SAM are due partly to external forcing but are not outside the natural variability as described by these models. Trends in jet location and magnitude, however, lie outside the unforced natural variability but can be explained by a combination of natural variability and the ensemble mean forced trend. These results indicate that trends in these three diagnostics cannot be used interchangeably.

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

    International Nuclear Information System (INIS)

    Blagovechtchenski, A.; Leontieva, V.; Mitrioukhin, A.

    1997-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  2. Validation of SSC using the FFTF natural-circulation tests

    International Nuclear Information System (INIS)

    Horak, W.C.; Guppy, J.G.; Kennett, R.J.

    1982-01-01

    As part of the Super System Code (SSC) validation program, the 100% power FFTF natural circulation test has been simulated using SSC. A detailed 19 channel, 2 loop model was used in SSC. Comparisons showed SSC calculations to be in good agreement with the Fast Flux Test Facility (FFTF), test data. Simulation of the test was obtained in real time

  3. Experimental and analytical study of natural circulation in square loop

    International Nuclear Information System (INIS)

    Moorthi, A.; Prem Sai, K.; Ravi, K.V.

    2015-01-01

    The nuclear safety under station blackout conditions is a major concern in the design of the nuclear reactors. In the case of existing reactors, the heat removal capability of cooling systems under natural circulation conditions is to be ascertained by experiments/analysis. This will ensure the long term core cooling and thereby, the safety of reactor core. Natural circulation occurs when the heat sink is at a higher elevation compared with the heat source. In case, the heat source and the sink are nearly at the same elevation, the difference in the elevations of their thermal centres can provide the elevation head required for natural circulation. An experimental study of natural circulation in the above geometry was carried out. The effect of flow resistance, Heat Source strength (heater power) and elevation difference between the source and the sink on the heat transfer were studied. The results of the experiments were analysed using RELAP5/MOD 3.2 and a good match between the experimental data and RELAP5 predictions is observed. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-31

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

  5. Nuclear reactor lid cooling which can work by natural circulation

    International Nuclear Information System (INIS)

    Wagner, J.

    1985-01-01

    The well-known air cooling of the lid of liquid metal cooled nuclear reactors is improved by the start of natural convection flow ensuring removal of heat in a sufficiently short time, if the blower fails. Go and return branches of the individual cooling circuits are arranged at different heights for this purpose. The circulation is supported by opening valves, which provide a direct path into the reactor building for the cooling air. The draught can be increased by setting up special chimneys. The start of circulation is aided by the temporary opening of another valve. (orig.) [de

  6. Development and experimental evaluation of models for low capillary number two-phase flows in rough walled fractures relevant to natural gradient conditions

    International Nuclear Information System (INIS)

    Glass, R.J.; Yarrington, L.; Nicholl, M.J.

    1997-09-01

    The major results from SNL's Conceptual Model Development and Validation Task (WBS 1.2.5.4.6) as developed through exploration of small scale processes were synthesized in Glass et al. to give guidance to Performance Assessment on improving conceptual models for isothermal flow in unsaturated, fractured rock. There, pressure saturation and relative permeability curves for single fractures were proposed to be a function of both fracture orientation within the gravity field and initial conditions. We refer the reader to Glass et al. for a discussion of the implications of this behavior for Performance Assessment. The scientific research we report here substantiates this proposed behavior. We address the modeling of phase structure within fractures under natural gradient conditions relevant to unsaturated flow through fractures. This phase structure underlies the calculation of effective properties for individual fractures and hence fracture networks as required for Performance Assessment. Standard Percolation (SP) and Invasion Percolation (IP) approaches have been recently proposed to model the underlying phase saturation structures within the individual fractures during conditions of two-phase flow. Subsequent analysis of these structures yields effective two-phase pressure-saturation and relative permeability relations for the fracture. However, both of these approaches yield structures that are at odds with physical reality as we see in experiments and thus effective properties calculated from these structures are in error. Here we develop and evaluate a Modified Invasion Percolation (MIP) approach to better model quasi-static immiscible displacement in fractures. The effects of gravity, contact angle, local aperature field geometry, and local in-plane interfacial curvature between phases are included in the calculation of invasion pressure for individual sites in a discretized aperture field

  7. Detailed evaluation of the natural circulation mass flow rate of water propelled by using an air injection

    International Nuclear Information System (INIS)

    Park, Rae-Joon; Ha, Kwang-Soon; Kim, Jae-Cheol; Hong, Seong-Wan; Kim, Sang-Baik

    2008-01-01

    One-dimensional (1D) air-water two-phase natural circulation flow in the thermohydraulic evaluation of reactor cooling mechanism by external self-induced flow - one-dimensional' (THERMES-1D) experiment has been verified and evaluated by using the RELAP5/MOD3 computer code. Experimental results on the 1D natural circulation mass flow rate of water propelled by using an air injection have been evaluated in detail. The RELAP5 results have shown that an increase in the air injection rate to 50% of the total heat flux leads to an increase in the water circulation mass flow rate. However, an increase in the air injection rate from 50 to 100% does not affect the water circulation mass flow rate, because of the inlet area condition. As the height increases in the air injection part, the void fraction increases. However, the void fraction in the upper part of the air injector maintains a constant value. An increase in the air injection mass flow rate leads to an increase in the local void fraction, but it has no influence on the local pressure. An increase in the coolant inlet area leads to an increase in the water circulation mass flow rate. However, the water outlet area does not have an influence on the water circulation mass flow rate. As the coolant outlet moves to a lower position, the water circulation mass flow rate decreases. (author)

  8. Experimental study of natural circulation flow instability in rectangular channels

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, Tao; Qi, Shi; Song, Mingqiang [North China Electric Power Univ., Beijing (China). School of Nuclear Science and Engineering; Passive Nuclear Safety Technology, Beijing (China). Beijing Key Lab.; Xiao, Zejun [Nuclear, Reactor Thermal Hydraulics Technology, Chengdu (China). CNNC Key Lab.

    2017-05-15

    Experiments of natural circulation flow instability were conducted in rectangular channels with 5 mm and 10 mm wide gaps. Results for different heating powers were obtained. The results showed that the flow will tend to be instable with the growing of heating power. The oscillation period of pressure D-value and volume flow are the same, but their phase positions are opposite. They both can be described by trigonometric functions. The existence of edge position and secondary flow will strengthen the disturbance of fluid flow in rectangle channels, which contributes to heat transfer. The disturbance of bubble and fluid will be strengthened, especially in the saturated boiling section, which make it possible for the mixing flow. The results also showed that the resistance in 5 mm channel is bigger than that in 10 mm channel, it is less likely to form stable natural circulation in the subcooled region.

  9. Liquid Fluoride Salt Experimentation Using a Small Natural Circulation Cell

    Energy Technology Data Exchange (ETDEWEB)

    Yoder Jr, Graydon L [ORNL; Heatherly, Dennis Wayne [ORNL; Williams, David F [ORNL; Elkassabgi, Yousri M. [Texas A& M University, Kingsville; Caja, Joseph [Electrochemical Systems, Inc.; Caja, Mario [ORNL; Jordan, John [Texas A& M University, Kingsville; Salinas, Roberto [Texas A& M University, Kingsville

    2014-04-01

    A small molten fluoride salt experiment has been constructed and tested to develop experimental techniques for application in liquid fluoride salt systems. There were five major objectives in developing this test apparatus: Allow visual observation of the salt during testing (how can lighting be introduced, how can pictures be taken, what can be seen) Determine if IR photography can be used to examine components submerged in the salt Determine if the experimental configuration provides salt velocity sufficient for collection of corrosion data for future experimentation Determine if a laser Doppler velocimeter can be used to quantify salt velocities. Acquire natural circulation heat transfer data in fluoride salt at temperatures up to 700oC All of these objectives were successfully achieved during testing with the exception of the fourth: acquiring velocity data using the laser Doppler velocimeter. This paper describes the experiment and experimental techniques used, and presents data taken during natural circulation testing.

  10. FFTF operating experience with sodium natural circulation: slides included

    Energy Technology Data Exchange (ETDEWEB)

    Burke, T.M.; Additon, S.L.; Beaver, T.R.; Midgett, J.C.

    1981-01-01

    The Fast Flux Test Facility (FFTF) has been designed for passive, back-up, safety grade decay heat removal utilizing natural circulation of the sodium coolant. This paper discusses the process by which operator preparation for this emergency operating mode has been assured, in paralled with the design verification during the FFTF startup and acceptance testing program. Over the course of the test program, additional insights were gained through the testing program, through on-going plant analyses and through general safety evaluations performed throughout the nuclear industry. These insights led to development of improved operator training material for control of decay heat removal during both forced and natural circulation as well as improvements in the related plant operating procedures.

  11. FFTF operating experience with sodium natural circulation: slides included

    International Nuclear Information System (INIS)

    Burke, T.M.; Additon, S.L.; Beaver, T.R.; Midgett, J.C.

    1981-01-01

    The Fast Flux Test Facility (FFTF) has been designed for passive, back-up, safety grade decay heat removal utilizing natural circulation of the sodium coolant. This paper discusses the process by which operator preparation for this emergency operating mode has been assured, in paralled with the design verification during the FFTF startup and acceptance testing program. Over the course of the test program, additional insights were gained through the testing program, through on-going plant analyses and through general safety evaluations performed throughout the nuclear industry. These insights led to development of improved operator training material for control of decay heat removal during both forced and natural circulation as well as improvements in the related plant operating procedures

  12. Natural circulation under variable primary mass inventories at BETHSY facility

    International Nuclear Information System (INIS)

    Bazin, P.; Clement, P.; Deruaz, R.

    1989-01-01

    BETHSY is a high pressure integral test facility which models a 3 loop Framatome PWR with the intent of studying PWR accidents. The BETHSY programme includes both accident transients and tests under successive steady state conditions. So far, tests of the latter type have been especially devoted to situations where natural circulation takes place in the primary coolant system (PCS). Tests 4.1a and 4.1a TC, the results of which are introduced, deal with PCS natural circulation patterns and related heat transport mechanisms under two different core power levels (2 and 5% of nominal power), variable primary mass inventory (100% to 30-40% according to core power) and at two different steam generator liquid levels (standard value and 1 meter). (orig.)

  13. Experimental studies on natural circulation in molten salt loops

    International Nuclear Information System (INIS)

    Srivastava, A.K.; Borgohain, A.; Maheshwari, N.K.; Vijayan, P.K.

    2015-01-01

    Molten salts are increasingly getting attention as a coolant and storage medium in solar thermal power plants and as a liquid fuel, blanket and coolant in Molten Salt Reactors (MSR’s). Two different test facilities named Molten Salt Natural Circulation Loop (MSNCL) and Molten Active Fluoride salt Loop (MAFL) have been setup for thermal hydraulics, instrument development and material related studies relevant to MSR and solar power plants. The working medium for MSNCL is a molten nitrate salt which is a mixture of NaNO 3 and KNO 3 in 60:40 ratio and proposed as one of the coolant option for molten salt based reactor and coolant as well as storage medium for solar thermal power application. On the other hand, the working medium for MAFL is a eutectic mixture of LiF and ThF 4 and proposed as a blanket salt for Indian Molten Salt Breeder Reactor (MSBR). Steady state natural circulation experiments at different power level have been performed in the MSNCL. Transient studies for startup of natural circulation, loss of heat sink, heater trip and step change in heater power have also been carried out in the same. A 1D code LeBENC, developed in-house to simulate the natural circulation characteristics in closed loops, has been validated with the experimental data obtained from MSNCL. Further, LeBENC has been used for Pretest analysis of MAFL. This paper deals with the description of both the loops and experimental studies carried out in MSNCL. Validation of LeBENC along with the pretest analysis of MAFL using the same are also reported in this paper. (author)

  14. Evaluation of system codes for analyzing naturally circulating gas loop

    International Nuclear Information System (INIS)

    Lee, Jeong Ik; No, Hee Cheon; Hejzlar, Pavel

    2009-01-01

    Steady-state natural circulation data obtained in a 7 m-tall experimental loop with carbon dioxide and nitrogen are presented in this paper. The loop was originally designed to encompass operating range of a prototype gas-cooled fast reactor passive decay heat removal system, but the results and conclusions are applicable to any natural circulation loop operating in regimes having buoyancy and acceleration parameters within the ranges validated in this loop. Natural circulation steady-state data are compared to numerical predictions by two system analysis codes: GAMMA and RELAP5-3D. GAMMA is a computational tool for predicting various transients which can potentially occur in a gas-cooled reactor. The code has a capability of analyzing multi-dimensional multi-component mixtures and includes models for friction, heat transfer, chemical reaction, and multi-component molecular diffusion. Natural circulation data with two gases show that the loop operates in the deteriorated turbulent heat transfer (DTHT) regime which exhibits substantially reduced heat transfer coefficients compared to the forced turbulent flow. The GAMMA code with an original heat transfer package predicted conservative results in terms of peak wall temperature. However, the estimated peak location did not successfully match the data. Even though GAMMA's original heat transfer package included mixed-convection regime, which is a part of the DTHT regime, the results showed that the original heat transfer package could not reproduce the data with sufficient accuracy. After implementing a recently developed correlation and corresponding heat transfer regime map into GAMMA to cover the whole range of the DTHT regime, we obtained better agreement with the data. RELAP5-3D results are discussed in parallel.

  15. Development of thermohydraulic software for PWR reactors with natural circulation

    International Nuclear Information System (INIS)

    Chasseur, Alfredo F.; Rauschert, A.; Delmastro, Dario F.

    2009-01-01

    The basics concepts about the development of software for steady state analysis of a reactor with natural circulations, in the primary circuit, are exposed. The reactor type is pressurized light water. The equations, correlations and flux diagrams of the source code of the software developed are shown. The source code of the software was written in FORTRAN 77 making use of modular technique, this save development effort and release of news versions is simplified. (author)

  16. Study on operational aspect of natural circulation HLMC reactor (1)

    International Nuclear Information System (INIS)

    Sienicki, J.J.; Cahalan, J.E.; Spencer, B.W.

    2000-08-01

    The concept of a heavy liquid metal cooled fast reactor that achieves 100% natural circulation heat removal from the core has the potential to attain improved cost competitiveness through extreme simplification, proliferation resistance, and heightened passive safety. The concept offers the potential for simplifications in plant control strategies wherein inherent reactor feedbacks may restore balance between energy release and heat removal from the reactor during operation as well as providing passive reactivity shutdown in the event of transients involving failure to scram. This study was initiated to evaluate the operational characteristics of the 100% natural circulation reactor under normal and transient states using a plant dynamics analysis computer code and to seek design and operational optimization of the concept. In the current Phase I of the project, the stage for the overall study has been prepared. A coupled thermal hydraulics-kinetics plant dynamics analysis code has been developed/modified that has the capabilities to calculate operational and accident transients. Code input has been prepared for the heavy liquid metal cooled natural circulation reactor concept. A preliminary analysis using the plant dynamics code and its input to calculate three illustrative cases relevant to initial startup, shutdown following long-term operation, and change in turbine load demonstrates the capability to analyze typical transient cases. (author)

  17. OPG's approach of crediting natural circulation in outage heat sinks

    International Nuclear Information System (INIS)

    Fung, K.K.; Mackinnon, J.C.

    2001-01-01

    A review of crediting natural circulation as a backup means of removing the reactor core decay heat during an outage in Ontario Power Generation's nuclear stations was completed in 2000. The objective was to define the configurations and conditions under which natural circulation can be confidently credited as an effective heat transport mechanism for use in shutdown heat sink management. The project was an interdisciplinary program, and involved analyses in the areas of heat transport system thermalhydaulics, fuel and fuel channel thermal and mechanical behaviour, radiation physics, and probabilistic risks. The assessment shows that it is economically acceptable to credit natural circulation as a backup means of removing the core decay heat whenever the no fuel failure criteria are met. The economic risks associated with such a potential use decrease with time after shutdown. The waiting times after shutdown when there would be various levels of risks of damaging the pressure tubes and fuel bundles were derived for use in planning maintenance activities so as to minimize the economic risks. (author)

  18. The two-phase flow IPTT method for measurement of nonwetting-wetting liquid interfacial areas at higher nonwetting saturations in natural porous media.

    Science.gov (United States)

    Zhong, Hua; Ouni, Asma El; Lin, Dan; Wang, Bingguo; Brusseau, Mark L

    2016-07-01

    Interfacial areas between nonwetting-wetting (NW-W) liquids in natural porous media were measured using a modified version of the interfacial partitioning tracer test (IPTT) method that employed simultaneous two-phase flow conditions, which allowed measurement at NW saturations higher than trapped residual saturation. Measurements were conducted over a range of saturations for a well-sorted quartz sand under three wetting scenarios of primary drainage (PD), secondary imbibition (SI), and secondary drainage (SD). Limited sets of experiments were also conducted for a model glass-bead medium and for a soil. The measured interfacial areas were compared to interfacial areas measured using the standard IPTT method for liquid-liquid systems, which employs residual NW saturations. In addition, the theoretical maximum interfacial areas estimated from the measured data are compared to specific solid surface areas measured with the N 2 /BET method and estimated based on geometrical calculations for smooth spheres. Interfacial areas increase linearly with decreasing water saturation over the range of saturations employed. The maximum interfacial areas determined for the glass beads, which have no surface roughness, are 32±4 and 36±5 cm -1 for PD and SI cycles, respectively. The values are similar to the geometric specific solid surface area (31±2 cm -1 ) and the N 2 /BET solid surface area (28±2 cm -1 ). The maximum interfacial areas are 274±38, 235±27, and 581±160 cm -1 for the sand for PD, SI, and SD cycles, respectively, and ~7625 cm -1 for the soil for PD and SI. The maximum interfacial areas for the sand and soil are significantly larger than the estimated smooth-sphere specific solid surface areas (107±8 cm -1 and 152±8 cm -1 , respectively), but much smaller than the N 2 /BET solid surface area (1387±92 cm -1 and 55224 cm -1 , respectively). The NW-W interfacial areas measured with the two-phase flow method compare well to values measured using the standard

  19. Verification of RELAP5-3D code in natural circulation loop as function of the initial water inventory

    Science.gov (United States)

    Bertani, C.; Falcone, N.; Bersano, A.; Caramello, M.; Matsushita, T.; De Salve, M.; Panella, B.

    2017-11-01

    High safety and reliability of advanced nuclear reactors, Generation IV and Small Modular Reactors (SMR), have a crucial role in the acceptance of these new plants design. Among all the possible safety systems, particular efforts are dedicated to the study of passive systems because they rely on simple physical principles like natural circulation, without the need of external energy source to operate. Taking inspiration from the second Decay Heat Removal system (DHR2) of ALFRED, the European Generation IV demonstrator of the fast lead cooled reactor, an experimental facility has been built at the Energy Department of Politecnico di Torino (PROPHET facility) to study single and two-phase flow natural circulation. The facility behavior is simulated using the thermal-hydraulic system code RELAP5-3D, which is widely used in nuclear applications. In this paper, the effect of the initial water inventory on natural circulation is analyzed. The experimental time behaviors of temperatures and pressures are analyzed. The experimental matrix ranges between 69 % and 93%; the influence of the opposite effects related to the increase of the volume available for the expansion and the pressure raise due to phase change is discussed. Simulations of the experimental tests are carried out by using a 1D model at constant heat power and fixed liquid and air mass; the code predictions are compared with experimental results. Two typical responses are observed: subcooled or two phase saturated circulation. The steady state pressure is a strong function of liquid and air mass inventory. The numerical results show that, at low initial liquid mass inventory, the natural circulation is not stable but pulsated.

  20. Computer simulation of natural circulation in FFTF secondary loops

    International Nuclear Information System (INIS)

    Beaver, T.R.; Turner, D.M.; Additon, S.L.

    1979-07-01

    A thermal/hydraulic model of the FFTF secondary heat transport loop has been calibrated against transient natural circulation test data collected March to May 1979. The tests verified that the transition to natural convective flow could be effected from near isothermal conditions without excessive cooling at the air dump heat exchangers. Key empirical parameters of pressure drop and heat loss were found to be at 88% and 81% of pretest estimates, respectively. Pretest piping thermal transport and flow calculational models required no further revision to produce good agreement with test data

  1. Analysis of a Natural Circulation in the Reactor Coolant System Following a High Pressure Severe Accident at APR1400

    International Nuclear Information System (INIS)

    Kim, Han Chul; Cho, Yong Jin; Park, Jae Hong; Cho, Song Won

    2011-01-01

    Under a high temperature and pressure condition during a severe accident, hot leg pipes or steam generator tubes could fail due to creep rupture following natural circulation in the Reactor Coolant System (RCS) unless depressurization of the system is performed at a proper time. Natural circulation in the RCS can be a multi-dimensional circulation in the reactor vessel, a partial loop circulation of two-phase flow from the core up to steam generators (SGs), or circulation in the total loop. It can delay the reactor vessel failure time by removing heat from the reactor core. This natural phenomenon can be hardly simulated with a single flow path model for the hot spots of the RCS, since it cannot deal with the counter-current flow. Thus it may estimate accident progression faster than reality, which may cause troubles for optimized implementation of severe accident management strategies. An earlier damage in the RCS other than the reactor pressure vessel may make subsequent behaviors of hydrogen or fission products in the containment quite different from the single reactor vessel failure. Therefore, a RCS model which treats natural circulation is needed to evaluate the RCS response and the safety depressurization strategy in a best-estimate way. The aim of this study is to develop a detailed model which allows natural circulation between the reactor vessel and steam generators through hot legs, based on the existing APR1400 RCS model. The station blackout sequence was selected to be the representative high-pressure scenario. Sensitivity study on the effect of node configuration of the upper plenum and addition of cross flow paths from the upper plenum to the hot legs were carried out. This model is described herein and representative calculation results are presented

  2. Stability analysis on natural circulation boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Metz, Peter

    1999-05-01

    The purpose of the study is a stability analysis of the simplified boiling water reactor concept. A fluid dynamics code, DYNOS, was developed and successfully validated against FRIGG and DESIRE data and a stability benchmark on the Ringhals 1 forced circulation BWR. Three simplified desings were considered in the analysis: The SWRIOOO by Siemens and the SBWR and ESBWR from the General Electric Co. For all three design operational characteristics, i.e. power versus flow rate maps, were calculated. The effects which different geometric and operational parameters, such as the riser height, inlet subcooling etc., have on the characteristics have been investigated. Dynamic simulations on the three simplified design revealed the geysering and the natural circulation oscillations modes only. They were, however, only encountered at pressure below 0.6 MPa. Stability maps for all tree simplified BWRs were calculated and plotted. The study concluded that a fast pressurisation of the reactor vessel is necessary to eliminate the possibility of geysering or natural circulation oscillations mode instability. (au) 26 tabs., 88 ills.

  3. Stability analysis on natural circulation boiling water reactors

    International Nuclear Information System (INIS)

    Metz, Peter

    1999-05-01

    The purpose of the study is a stability analysis of the simplified boiling water reactor concept. A fluid dynamics code, DYNOS, was developed and successfully validated against FRIGG and DESIRE data and a stability benchmark on the Ringhals 1 forced circulation BWR. Three simplified desings were considered in the analysis: The SWRIOOO by Siemens and the SBWR and ESBWR from the General Electric Co. For all three design operational characteristics, i.e. power versus flow rate maps, were calculated. The effects which different geometric and operational parameters, such as the riser height, inlet subcooling etc., have on the characteristics have been investigated. Dynamic simulations on the three simplified design revealed the geysering and the natural circulation oscillations modes only. They were, however, only encountered at pressure below 0.6 MPa. Stability maps for all tree simplified BWRs were calculated and plotted. The study concluded that a fast pressurisation of the reactor vessel is necessary to eliminate the possibility of geysering or natural circulation oscillations mode instability. (au)

  4. An Experimental Study of Natural Circulation in a Loop with Parallel Flow Test Sections

    International Nuclear Information System (INIS)

    Mathisen, R.P.; Eklind, O.

    1965-10-01

    The dynamic behaviour of a natural circulation loop parallel round duct channels has been studied. The test sections were both electrically heated and the power distribution was uniform along the 4300 mm heated length of the 20 mm dia. channels. The inter channel interference and the threshold of flow instability were obtained by using a dynamically calibrated flowmeter in each channel. The pressure was 50 bars and the sub-cooling 6 deg C. The main parameters varied, were the flow restrictions in the one-phase and two-phase sections. The instability data were correlated to the resistance coefficients due to these restrictions. Theoretical calculations for parallel channels in natural circulation have been compared with the experimental results. For the conditions determined by the above mentioned magnitudes, the steady-state computations are in excellent agreement with experiment. The transients are also nearly similar, except for the resonance frequency which for the theoretical case is higher by an amount between 0.3 and 0.5 c.p.s

  5. An Experimental Study of Natural Circulation in a Loop with Parallel Flow Test Sections

    Energy Technology Data Exchange (ETDEWEB)

    Mathisen, R P; Eklind, O

    1965-10-15

    The dynamic behaviour of a natural circulation loop parallel round duct channels has been studied. The test sections were both electrically heated and the power distribution was uniform along the 4300 mm heated length of the 20 mm dia. channels. The inter channel interference and the threshold of flow instability were obtained by using a dynamically calibrated flowmeter in each channel. The pressure was 50 bars and the sub-cooling 6 deg C. The main parameters varied, were the flow restrictions in the one-phase and two-phase sections. The instability data were correlated to the resistance coefficients due to these restrictions. Theoretical calculations for parallel channels in natural circulation have been compared with the experimental results. For the conditions determined by the above mentioned magnitudes, the steady-state computations are in excellent agreement with experiment. The transients are also nearly similar, except for the resonance frequency which for the theoretical case is higher by an amount between 0.3 and 0.5 c.p.s.

  6. Modeling of Natural Self-Pressurized Circulation Circuits

    International Nuclear Information System (INIS)

    Zanocco, Pablo; Gimenez, Marcelo; Delmastro, Dario

    2003-01-01

    In this work, HUARPE code models for self-pressurized two-phase natural convection systems are improved.A drift-flux model is included, allowing the modeling of the relative velocity between phases.The model of steam dome structures is improved, with a thermal resistance scheme, in order to model the material thermal conductivity.This allows improving thermal losses modeling and structures dynamic.'Shape functions' are implemented based on analytic solutions for pressure derivative and density in each node, which allows less diffusive solving schemes, more appropriate for analyzing cases involving density waves phenomena.Finally, pressure evolutions during a pressurization transient are analyzed, comparing the new models and the previous version results.These results are also checked against RELAP code ones, obtained with different dome nodalizations. Moreover, modeling problems are analyzed for each case

  7. In-vessel natural circulation during a hypothetical loss-of-heat-sink accident in the Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Perkins, K.R.; Bari, R.A.; Pratt, W.T.

    1979-05-01

    The capability to remove decay heat from the FFTF core via in-vessel natural circulation has been analyzed for the preboiling phase using a lumped parameter model. The results indicate that boiling will occur in the average fuel assembly for a wide spectrum of initial conditions which appear to be representative of the hypothetical loss-of-heat-sink accident. Two-phase pressure drop calculations indicate that, once the saturation temperature is reached, coolability can only be assured for decay heat levels which are less than 0.5% of the operating power. A review of the limited sodium boiling data indicates that boiling-induced natural circulation may support up to 4% of the operating power, but geometric atypicalities and a large degree of inlet subcooling for the existing data limit the applicability to the loss-of-heat-sink accident in FFTF

  8. Automated scoping methodology for liquid metal natural circulation small reactor

    International Nuclear Information System (INIS)

    Son, Hyung M.; Suh, Kune Y.

    2014-01-01

    Highlights: • Automated scoping methodology for natural circulation small modular reactor is developed. • In-house code is developed to carry out system analysis and core geometry generation during scoping. • Adjustment relations are obtained to correct the critical core geometry out of diffusion theory. • Optimized design specification is found using objective function value. • Convex hull volume is utilized to quantify the impact of different constraints on the scope range. - Abstract: A novel scoping method is proposed that can automatically generate design variable range of the natural circulation driven liquid metal cooled small reactor. From performance requirements based upon Generation IV system roadmap, appropriate structure materials are selected and engineering constraints are compiled based upon literature. Utilizing ASME codes and standards, appropriate geometric sizing criteria on constituting components are developed to ensure integrity of the system during its lifetime. In-house one dimensional thermo-hydraulic system analysis code is developed based upon momentum integral model and finite element methods to deal with non-uniform descritization of temperature nodes for convection and thermal diffusion equation of liquid metal coolant. In order to quickly generate critical core dimensions out of given unit cell information, an adjustment relation that relates the critical geometry estimated from one-group diffusion and that from MCNP code is constructed and utilized throughout the process. For the selected unit cell dimension ranges, burnup calculations are carried out to check the cores can generate energy over the reactor lifetime. Utilizing random method, sizing criteria, and in-house analysis codes, an automated scoping methodology is developed. The methodology is applied to nitride fueled integral type lead cooled natural circulation reactor concept to generate design scopes which satisfies given constraints. Three dimensional convex

  9. Experimental observations of natural circulation flow in the NSTF

    Energy Technology Data Exchange (ETDEWEB)

    Lisowski, Darius D., E-mail: dlisowski@anl.gov; Kraus, Adam R.; Bucknor, Matthew D.; Hu, Rui; Farmer, Mitch T.

    2016-09-15

    A 1/2 scale test facility has been constructed at Argonne National Laboratory to study the heat removal performance and natural circulation flow patterns in a Reactor Cavity Cooling System (RCCS). Our test facility, the Natural convection Shutdown heat removal Test Facility (NSTF), supports the broader goal of developing an inherently safe and fully passive ex-vessel decay heat removal for advanced reactor designs. The project, initiated in 2010 to support the Advanced Reactor Technologies (ART), Small Modular Reactor (SMR), and Next Generation Nuclear Plant (NGNP) programs, has been conducting experimental operations since early 2014. The following paper provides a summary of some primary design features of the 26-m tall test facility along with a description of the data acquisition suite that guides our experimental practices. Specifics of the distributed fiber optic temperature measurements will be discussed, which introduces an unparalleled level of data density that has never before been implemented in a large scale natural circulation test facility. Results from our test series will then be presented, which provide insight into the thermal hydraulic behavior at steady-state and transient conditions for varying heat flux levels and exhaust chimney configuration states.

  10. Two-phase flow

    International Nuclear Information System (INIS)

    Olive, J.

    1990-01-01

    The design, operation and safety of nuclear components requires increasingly accurate knowledge of two-phase flows. This knowledge is also necessary for some studies related to electricity applications. The author presents some concrete examples showing the range of problems and the complexity of the phenomena involved in these types of flows. Then, the basic principles of their numerical modelling are explained, as well as the new tendency to use increasingly local and refined models. The newest computer codes developed at EDF are briefly presented. Experimental studies dealing with twophase flow are also referred to, and their connections to numerical modelling are explained. Emphasis is placed on the major efforts devoted to the development of new test rigs and instrumentation [fr

  11. Methodology for studies of natural circulation in closed circuits

    International Nuclear Information System (INIS)

    Araujo, Rafael de Oliveira Pessoa de

    2009-01-01

    This work presents the results obtained from the analysis of stability of the phenomenon of the natural circulation for one-dimension single-phase flow in a closed loop, by a computer program with the method of finite element. The Navier-Stokes equations in cartesian geometry were used for the balance of mass, momentum and one equation for energy. The formulation has been implemented in a computer code developed at the Nuclear Engineering Institute(IEN-CNEN-RJ) and is now available either for futures analysis or design of nuclear systems. (author)

  12. Natural Circulation Capability Assessments for a Small-medium Reactor

    International Nuclear Information System (INIS)

    Choi, Sun Do

    2010-02-01

    Small-medium reactors have been highly evaluated to have more safe characteristics than those of large reactors. In addition, it could be used for a variety of purposes, such as small-scale power production in mountainous of island area, seawater desalination, regional heating system. For a higher safety, studies about a way of using natural circulation have being conducted around world. CAREM(Argentina), AST- 500(Russia), and NHR-200(china) etc. According to this tendency, REX- 10(Regional Energy rX-10) is designed in Korea for regional heating and small-scale power production. To investigate the thermal-hydraulic behavior of REX-10, we designed Rex-10 Test Facility (RTF), simulating REX-10, by using the scaling law. The scaling ratios of length, volume and power were set with 1/1, 1/50 and 1/50, respectively. The diameter and total length of RTF are 40 cm and approximately 6 m, respectively. The facility is composed of various components, which are a core in the bottom part, a heat exchanger in the middle part, a pressurizer and hot legs in the upper part, and chillers outside the facility. The test instrumentation is also designed to measure temperatures, flow rates, pressures, and pressure drop. The experiment parameters were adopted based on the 1-dimensional approach. There are a variety of parameters which influence natural circulation behavior such as heater power, overall flow resistance parameter, the distance between the center of the heat exchanger and the core. As the experimental geometries are fixed, it is found that the most important parameter is the heater power under the experimental conditions. In addition, to evaluate the effect of heater power, some experiments were conducted at varying heater power condition (from 70 kW to 170 kW) under constant primary pressure (2.0 MPa) and secondary flow rate (4.5 liter per minute). As the results of the experiments, the temperature and flow rate increase with increasing heater power. The flow rate is

  13. 1-D Two-phase Flow Investigation for External Reactor Vessel Cooling

    International Nuclear Information System (INIS)

    Kim, Jae Cheol

    2007-02-01

    During a severe accident, when a molten corium is relocated in a reactor vessel lower head, the RCF(Reactor Cavity Flooding) system for ERVC (External Reactor Vessel Cooling) is actuated and coolants are supplied into a reactor cavity to remove a decay heat from the molten corium. This severe accident mitigation strategy for maintaining a integrity of reactor vessel was adopted in the nuclear power plants of APR1400, AP600, and AP1000. Under the ERVC condition, the upward two-phase flow is driven by the amount of the decay heat from the molten corium. To achieve the ERVC strategy, the two-phase natural circulation in the annular gap between the external reactor vessel and the insulation should be formed sufficiently by designing the coolant inlet/outlet area and gap size adequately on the insulation device. Also the natural circulation flow restriction has to be minimized. In this reason, it is needed to review the fundamental structure of insulation. In the existing power plants, the insulation design is aimed at minimizing heat losses under a normal operation. Under the ERVC condition, however, the ability to form the two-phase natural circulation is uncertain. Namely, some important factors, such as the coolant inlet/outlet areas, flow restriction, and steam vent etc. in the flow channel, should be considered for ERVC design. T-HEMES 1D study is launched to estimate the natural circulation flow under the ERVC condition of APR1400. The experimental facility is one-dimensional and scaled down as the half height and 1/238 channel area of the APR1400 reactor vessel. The air injection method was used to simulate the boiling at the external reactor vessel and generate the natural circulation two-phase flow. From the experimental results, the natural circulation flow rate highly depended on inlet/outlet areas and the circulation flow rate increased as the outlet height as well as the supplied water head increased. On the other hand, the simple analysis using the drift

  14. Transient two-phase flow

    International Nuclear Information System (INIS)

    Hsu, Y.Y.

    1974-01-01

    The following papers related to two-phase flow are summarized: current assumptions made in two-phase flow modeling; two-phase unsteady blowdown from pipes, flow pattern in Laval nozzle and two-phase flow dynamics; dependence of radial heat and momentum diffusion; transient behavior of the liquid film around the expanding gas slug in a vertical tube; flooding phenomena in BWR fuel bundles; and transient effects in bubble two-phase flow. (U.S.)

  15. Numerical simulation of losses along a natural circulation helium loop

    Energy Technology Data Exchange (ETDEWEB)

    Knížat, Branislav, E-mail: branislav.knizat@stuba.sk; Urban, František, E-mail: frantisek.urban@stuba.sk; Mlkvik, Marek, E-mail: marek.mlkvik@stuba.sk; Ridzoň, František, E-mail: frantisek.ridzon@stuba.sk; Olšiak, Róbert, E-mail: robert.olsiak@stuba.sk [Slovak University of Technology in Bratislava, Faculty of Mechanical Engineering, Nám. slobody 17, 812 31 Bratislava, Slovak Republik (Slovakia)

    2016-06-30

    A natural circulation helium loop appears to be a perspective passive method of a nuclear reactor cooling. When designing this device, it is important to analyze the mechanism of an internal flow. The flow of helium in the loop is set in motion due to a difference of hydrostatic pressures between cold and hot branch. Steady flow at a requested flow rate occurs when the buoyancy force is adjusted to resistances against the flow. Considering the fact that the buoyancy force is proportional to a difference of temperatures in both branches, it is important to estimate the losses correctly in the process of design. The paper deals with the calculation of losses in branches of the natural circulation helium loop by methods of CFD. The results of calculations are an important basis for the hydraulic design of both exchangers (heater and cooler). The analysis was carried out for the existing model of a helium loop of the height 10 m and nominal heat power 250 kW.

  16. Effect of ship motions and flow stability in a small marine reactor driven by natural circulation

    International Nuclear Information System (INIS)

    Yoritsune, Tsutomu; Ishida, Toshihisa

    2001-12-01

    By using a small reactor as a power source for investigations and developments under sea, widely expanded activity is expectable. In this case, as for a nuclear reactor, small-size and lightweightness, and simplification of a system are needed with the safety. In JAERI, very small reactors for submersible research vessel (Deep-sea Reactor DRX and submersible Compact Reactor SCR) have been designed on the basis of needs investigation of sea research. Although the reactor is a PWR type, self-pressurization and natural circulation system are adopted in a primary system for small size and lightweightness. The fluid flow condition of the reactor core outlet is designed to be the two-phase with a low quality. Although the flow of a primary system is the two-phase flow with a low quality, the density wave oscillation may occur according to operating conditions. Moreover, since there are ship motions of heaving (the vertical direction acceleration) etc., when a submersible research vessel navigates on the sea surface, the circulation flow of the primary system is directly influenced by this external force. In order to maintain stable operations of the reactor, it is necessary to clarify effects of the flow stability characteristic of the primary coolant system and the external force. Until now, as for the flow stability of a nuclear reactor itself, many research reports have been published including the nuclear-coupled thermal oscillation of BWRs such as LaSalle-2, WNP-2 etc. As for the effect of external force, it is reported that the acceleration change based on a seismic wave affects the reactor core flow and the reactor power in a BWR. On the other hand, also in a PWR, since adoption of natural circulation cooling is considered for a generation 4 reactor, it is thought that the margin of the reactor core flow stability becomes an important parameter in the design. The reactor coolant flow mentioned in this report is the two-phase natural circulation flow coupled with

  17. Effect of natural circulation on RCS depressurization strategy in PWR NPP

    International Nuclear Information System (INIS)

    Zhang Kun; Tong Lili; Cao Xuewu

    2009-01-01

    The natural circulation model of Chinese Qinshan Nuclear Power Plant (NPP) Unit 2 is built using SCDAP/RELAP5 code. Selecting TMLB' accident as the base sequence, this paper analyzes the natural circulation phenomena in high-pressure core melt severe accident. In order to study the effect of natural circulation on RCS depressurization strategy, the accident progressions of RCS depressurization with and without natural circulation are simulated, respectively. According to the results, the natural circulation can delay the initiation of RCS depressurization and the whole accident progression, but it does not evidently influence the results of RCS depressurization. (authors)

  18. Experimental and theoretical study on natural circulation capacity under rolling motion condition

    International Nuclear Information System (INIS)

    Tan Sichao; Gao Puzhen

    2007-01-01

    Effect of rolling motion on natural circulation capacity was studied experimentally and theoretically. Experiments were conducted under the conditions of rolling and unrolling motions. The experimental results show that natural circulation capacity decreases under rolling motion condition. A mathematic model was developed to calculate the natural circulation capacity under rolling motion condition, considering the characteristics of natural circulation, the model was modified. The calculated results agree with experimental data well. Effect of rolling motion on natural circulation was analyzed through calculation and the following conclusions were obtained: (1) The increase of flow resistance coefficient is the main reason that the natural circulation capacity decreases under rolling motion condition; (2) Non-uniform distribution of fluid mass in the pipe has also influence on natural circulation capacity. (author)

  19. Investigation of Natural Circulation Instability and Transients in Passively Safe Small Modular Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ishii, Mamoru [Purdue Univ., West Lafayette, IN (United State

    2016-11-30

    The NEUP funded project, NEUP-3496, aims to experimentally investigate two-phase natural circulation flow instability that could occur in Small Modular Reactors (SMRs), especially for natural circulation SMRs. The objective has been achieved by systematically performing tests to study the general natural circulation instability characteristics and the natural circulation behavior under start-up or design basis accident conditions. Experimental data sets highlighting the effect of void reactivity feedback as well as the effect of power ramp-up rate and system pressure have been used to develop a comprehensive stability map. The safety analysis code, RELAP5, has been used to evaluate experimental results and models. Improvements to the constitutive relations for flashing have been made in order to develop a reliable analysis tool. This research has been focusing on two generic SMR designs, i.e. a small modular Simplified Boiling Water Reactor (SBWR) like design and a small integral Pressurized Water Reactor (PWR) like design. A BWR-type natural circulation test facility was firstly built based on the three-level scaling analysis of the Purdue Novel Modular Reactor (NMR) with an electric output of 50 MWe, namely NMR-50, which represents a BWR-type SMR with a significantly reduced reactor pressure vessel (RPV) height. The experimental facility was installed with various equipment to measure thermalhydraulic parameters such as pressure, temperature, mass flow rate and void fraction. Characterization tests were performed before the startup transient tests and quasi-steady tests to determine the loop flow resistance. The control system and data acquisition system were programmed with LabVIEW to realize the realtime control and data storage. The thermal-hydraulic and nuclear coupled startup transients were performed to investigate the flow instabilities at low pressure and low power conditions for NMR-50. Two different power ramps were chosen to study the effect of startup

  20. Study on reverse flow characteristics under natural circulation in inverted U-tube steam generator

    International Nuclear Information System (INIS)

    Duan Jun; Zhou Tao; Zhang Lei; Hong Dexun; Liu Ping

    2013-01-01

    Natural circulation is important for application in the nuclear power industry. Aiming at the steam generator of AP1000 pressurized water reactor loop, the mathematical model was established to analysis the reverse flow of single-phase water in the inverted U-tubes of a steam generator in a natural circulation system. The length distribution and the mass flow rates in both tubes with normal and reverse flow were determined respectively. The research results show that the reverse flow may result in sharp decrease of gravity pressure head, circulation mass flow rate and heat release rate of natural circulation. It has adverse influence on natural circulation. (authors)

  1. Analysis on natural circulation capacity of the CARR

    Institute of Scientific and Technical Information of China (English)

    TIAN Wenxi; QIU Suizheng; WANG Jiaqiang; SU Guanghui; JIA Dounan; ZHANG Jianwei

    2007-01-01

    The investigation on natural circulation (NC) characteristics of the China Advanced Research Reactor(CARR) is very valuable for practical engineering application and also a key subject for the CARR. In this study, a computer code was developed to calculate the NC capacity of the CARR under different pool water temperatures. Effects of the pool water temperature on NC characteristics were analyzed. The results show that with increasing pool water temperature, the NC flow rate increases while the NC capacity decreases. Based on the computation results and theoretical deduction, a correlation was proposed on predicting the relationship between the NC mass flow and the core power under different conditions. The correlation prediction agrees well with the computational result within ±10% for the maximal deviation. This work is instructive for the actual operation of the CARR.

  2. ESBWR - Robust design for natural circulation and stability performance effectiveness

    Energy Technology Data Exchange (ETDEWEB)

    Alamgir, M. D.; Marquino, W.; Yang, J.; Saha, P.; Fennern, L.; Colby, M. [GE-Hitachi Nuclear Energy, M/C A65, 3901Castle Hayne Road, Wilmington, NC 28401 (United States)

    2012-07-01

    ESBWR is a 4500 MWt Generation III+ natural circulation reactor with an array of robust design features and passive safety systems to deliver highly effective plant performance during normal operation and to keep the reactor safe during postulated transients and accidents. With the submittal of the latest revision of the Design Control Document (DCD) to US Nuclear Regulatory Commission, ESBWR is nearing the completion of the US design certification process. This paper focuses on the natural circulation-driven plant performance aspects during normal operation, and stability evaluation of the robust ESBWR design. The TRACG computer code is used for the analysis of ESBWR plant performance, safety analysis, and stability margins. The paper describes the evaluation of ESBWR stability performance during normal power operation including operation in the Core Power-Feed Water Temperature Operating Domain. For ESBWR the normal power operation condition has the highest power/flow ratio and is limiting from the perspective of stability. The paper includes results from detailed evaluation of the most limiting decay ratio for out-of-phase regional oscillations calculated by perturbing the core inlet flow rate in this out-of-phase mode about the line of symmetry for the azimuthal harmonic mode. The paper also summarizes the ESBWR regional mode stability evaluations during a limiting transient (Loss of Feedwater Heating), and during ATWS (Anticipated Transient without Scram). Nominal decay ratios of limiting Channel oscillation, Core wide oscillation and Regional oscillation are within the maximum acceptance criterion of 0.8, at 95% content and 95% confidence. These stability evaluation results indicate decay ratio is within design limits. The paper also describes the evaluation of ESBWR stability performance during plant startup, and summarizes the defense-in-depth stability solution for ESBWR. (authors)

  3. ESBWR - Robust design for natural circulation and stability performance effectiveness

    International Nuclear Information System (INIS)

    Alamgir, M. D.; Marquino, W.; Yang, J.; Saha, P.; Fennern, L.; Colby, M.

    2012-01-01

    ESBWR is a 4500 MWt Generation III+ natural circulation reactor with an array of robust design features and passive safety systems to deliver highly effective plant performance during normal operation and to keep the reactor safe during postulated transients and accidents. With the submittal of the latest revision of the Design Control Document (DCD) to US Nuclear Regulatory Commission, ESBWR is nearing the completion of the US design certification process. This paper focuses on the natural circulation-driven plant performance aspects during normal operation, and stability evaluation of the robust ESBWR design. The TRACG computer code is used for the analysis of ESBWR plant performance, safety analysis, and stability margins. The paper describes the evaluation of ESBWR stability performance during normal power operation including operation in the Core Power-Feed Water Temperature Operating Domain. For ESBWR the normal power operation condition has the highest power/flow ratio and is limiting from the perspective of stability. The paper includes results from detailed evaluation of the most limiting decay ratio for out-of-phase regional oscillations calculated by perturbing the core inlet flow rate in this out-of-phase mode about the line of symmetry for the azimuthal harmonic mode. The paper also summarizes the ESBWR regional mode stability evaluations during a limiting transient (Loss of Feedwater Heating), and during ATWS (Anticipated Transient without Scram). Nominal decay ratios of limiting Channel oscillation, Core wide oscillation and Regional oscillation are within the maximum acceptance criterion of 0.8, at 95% content and 95% confidence. These stability evaluation results indicate decay ratio is within design limits. The paper also describes the evaluation of ESBWR stability performance during plant startup, and summarizes the defense-in-depth stability solution for ESBWR. (authors)

  4. Gas-liquid two-phase flow behavior in terrain-inclined pipelines for gathering transport system of wet natural gas

    DEFF Research Database (Denmark)

    Yang, Yan; Li, Jingbo; Wang, Shuli

    2018-01-01

    The Volume of Fluid method and Re-Normalisation Group (RNG) k-ε turbulence model were employed to predict the gas-liquid two-phase flow in a terrain-inclined pipeline with deposited liquids. The simulation was carried out in a 22.5 m terrain-inclined pipeline with a 150 mm internal diameter...... on the liquid level under the suction force which caused by the negative pressure around the elbow, and then it touched to the top of the pipe. When the liquid blocked the pipe, the pressure drop between the upstream and downstream of the elbow increased with the increase of the gas velocity. At larger gas...

  5. 3. Workshop for IAEA ICSP on Integral PWR Design Natural Circulation Flow Stability and Thermo-hydraulic Coupling of Containment and Primary System during Accidents. Presentations

    International Nuclear Information System (INIS)

    2012-04-01

    tests were distributed to all ICSP participants for their blind calculations. Objective of ICSP: - Improve understanding of natural circulation phenomena expected to occur in integral type PWRs - Evaluate system code capabilities to predict natural circulation phenomena for integral type PWR, their practicality and efficiency, by simulating an integrated experiment - Supply experimental data for single/two-phase flow natural circulation instability and long-term cooling by natural circulation through coupled primary system and containment - Suggest necessary code improvements or new experiments to reduce uncertainties

  6. Oscillatory two-phase flows

    International Nuclear Information System (INIS)

    Boure, J.A.

    1974-12-01

    Two-phase flow instabilities are classified according to three criteria: the static or dynamic nature of the phenomenon, the necessity or not of a triggering phenomenon, and the pure or compound character of the phenomenon. Tables give the elementary instability phenomena, and the practical types of instability. Flow oscillations (or dynamic instabilities) share a number of characteristics which are dealt with, they are caused by the dynamic interactions between the flow parameters (flow rate, density, pressure, enthalpy and their distributions). Oscillation types are discussed: pure oscillations are density wave oscillations, acoustic oscillations may also occur, various compound oscillations involve either the density wave or the acoustic wave mechanism, interacting with some of the boundary conditions in the device. The analysis of slow oscillations has been made either by means of a simplified model (prediction of the thresholds) or of computer codes. Numerous computer codes are available [fr

  7. Experimental studies in a single-phase parallel channel natural circulation system. Preliminary results

    International Nuclear Information System (INIS)

    Bodkha, Kapil; Pilkhwal, D.S.; Jana, S.S.; Vijayan, P.K.

    2016-01-01

    Natural circulation systems find extensive applications in industrial engineering systems. One of the applications is in nuclear reactor where the decay heat is removed by natural circulation of the fluid under off-normal conditions. The upcoming reactor designs make use of natural circulation in order to remove the heat from core under normal operating conditions also. These reactors employ multiple vertical fuel channels with provision of on-power refueling/defueling. Natural circulation systems are relatively simple, safe and reliable when compared to forced circulation systems. However, natural circulation systems are prone to encounter flow instabilities which are highly undesirable for various reasons. Presence of parallel channels under natural circulation makes the system more complicated. To examine the behavior of parallel channel system, studies were carried out for single-phase natural circulation flow in a multiple vertical channel system. The objective of the present work is to study the flow behavior of the parallel heated channel system under natural circulation for different operating conditions. Steady state and transient studies have been carried out in a parallel channel natural circulation system with three heated channels. The paper brings out the details of the system considered, different cases analyzed and preliminary results of studies carried out on a single-phase parallel channel system.

  8. The Condensation effect on the two-phase flow stability

    International Nuclear Information System (INIS)

    Abdou Mohamed, Hesham Nagah

    2005-01-01

    A one-dimensional analytical model has been developed to be used for the linear analysis of density-wave oscillations in a parallel heated channel and a natural circulation loop.The heater and the riser sections are divided into a single-phase and a two-phase region.The two-phase region is represented by the drift-flux model. The model accounts for aphasic slip and subcooled boiling.The localized friction at the heater and the riser exit is treated considering the two-phase mixture.Also the effects of the condensation in the riser and the change in the system pressure have been studied.The exact equation for the heated channel and the total loop pressure drop is perturbed around the steady state.he stability characteristics of the heated channel and the loop are investigated using the Root finding method criterion.The results are summarized on instability maps in the plane of subcooled boiling number vs. phase change number (i.e., inlet subcooling vs. heater heat flux).The predictions of the model are compared with experimental results published in open literature. The results show that, the treatment effect of localized friction in two-phase mixtures stabilizes the system and improves the agreement of the calculations with the experimental results.For a parallel heated channel, the results indicate a more stable system with high inlet restriction, low outlet restriction, and high inlet velocity. And for a natural circulation loop, an increase in the inlet restriction broadened the range of the continuous circulation mode and stabilized the system, a decrease in the exit restriction or the liquid charging level shifted to the right the range of the continuous circulation mode and stabilized the system and an increase in the riser condensation shifted to the right the range of the continuous circulation mode and stabilized the system.The results show that the model agrees well with the available experimental data. In particular, the results show the significance of

  9. Pumped two-phase heat transfer loop

    Science.gov (United States)

    Edelstein, Fred

    1988-01-01

    A pumped loop two-phase heat transfer system, operating at a nearly constant temperature throughout, includes several independently operating grooved capillary heat exchanger plates supplied with working fluid through independent flow modulation valves connected to a liquid supply line, a vapor line for collecting vapor from the heat exchangers, a condenser between the vapor and the liquid lines, and a fluid circulating pump between the condenser and the heat exchangers.

  10. Natural circulation cooling in a PWR geometry under accident-induced conditions

    International Nuclear Information System (INIS)

    Shimeck, D.J.; Johnsen, G.W.

    1983-01-01

    The characteristics and limits of natural circulation heat rejection over a wide range of conditions were experimentally investigated in a small-scale model of a pressurized water reactor. Conditions that were varied included primary and secondary coolant inventory, decay heat power, and primary noncondensable gas content. The results have defined three distinct modes of natural circulation, their limits and transition points, and the characteristic signatures accompanying natural circulation behavior. Particular emphasis is focused on the limits of natural circulation under severely degraded primary and secondary conditions

  11. Two-phase flow models

    International Nuclear Information System (INIS)

    Delaje, Dzh.

    1984-01-01

    General hypothesis used to simplify the equations, describing two-phase flows, are considered. Two-component and one-component models of two-phase flow, as well as Zuber and Findlay model for actual volumetric steam content, and Wallis model, describing the given phase rates, are presented. The conclusion is made, that the two-component model, in which values averaged in time are included, is applicable for the solving of three-dimensional tasks for unsteady two-phase flow. At the same time, using the two-component model, including values, averaged in space only one-dimensional tasks for unsteady two-phase flow can be solved

  12. Novel modular natural circulation BWR design and safety evaluation

    International Nuclear Information System (INIS)

    Ishii, Mamoru; Shi, Shanbin; Yang, Won Sik; Wu, Zeyun; Rassame, Somboon; Liu, Yang

    2015-01-01

    Highlights: • Introduction of BWR-type natural circulation small modular reactor preliminary design (NMR-50). • Design of long fuel cycle length for the NMR-50. • Design of double passive safety systems for the NMR-50. • RELAP5 analyses of design basis accidents for the NMR-50. - Abstract: The Purdue NMR (Novel Modular Reactor) represents a BWR-type small modular reactor with a significantly reduced reactor pressure vessel (RPV) height. Specifically, it has one third the height of a conventional BWR RPV with an electrical output of 50 MWe. The preliminary design of the NMR-50 including reactor, fuel cycle, and safety systems is described and discussed. The improved neutronics design of the NMR-50 extends the fuel cycle length up to 10 years. The NMR-50 is designed with double passive engineering safety system, which is intended to withstand a prolonged station black out with loss of ultimate heat sink accident such as experienced at Fukushima. In order to evaluate the safety features of the NMR-50, two representative design basis accidents, i.e. main steam line break (MSLB) and bottom drain line break (BDLB), are simulated by using the best-estimate thermal–hydraulic code RELAP5. The RPV water inventory, containment pressure, and the performance of engineering safety systems are investigated for about 33 h after the initiation of the accidents

  13. Analytical and experimental study of liquid metal natural circulation

    Energy Technology Data Exchange (ETDEWEB)

    Chu, H.S.; Kim, C.S.; Suh, K.Y. [Seoul National Univ., Dept. of Nuclear Engineering (Korea, Republic of)

    2001-07-01

    A one-dimensional flow loop model is used to analyze the state of the test loop in the natural circulation condition. Along the low-temperature melting eutectic metal alloy path, the steady-state momentum equation and the energy equation are solved at the one-dimensional lumped nodes. A three-dimensional computational fluid dynamics code, CFX 4.2, is applied to check on the result of first-principle calculation. The major pressure drop results from the four elbows in the test loop. Experiments are planned to cover a wide spectrum of the turbulent regime in the large-diameter piping. The smaller-diameter pipe produces larger temperature difference given the flow velocity. Because the low-temperature melting eutectic metal alloy used in this experiment gets burnable above 150 C, the diameter of the pipe must be large enough to cover the wide velocity range. Given the same velocity, the heater power in the smaller diameter pipe is less than in the larger diameter pipe. This shows that the heat removal is more effective in the large pipe given the flow velocity. (author)

  14. Numerical Simulations and Design Optimization of the PHT Loop of Natural Circulation BWR

    Directory of Open Access Journals (Sweden)

    G. V. Durga Prasad

    2008-01-01

    Full Text Available Mathematical modeling and numerical simulation of natural circulation boiling water reactor (NCBWR are very important in order to study its performance for different designs and various off-design conditions and for design optimization. In the present work, parametric studies of the primary heat transport loop of NCBWR have been performed using lumped parameter models and RELAP5/MOD3.4 code. The lumped parameter models are based on the drift flux model and homogeneous equilibrium mixture (HEM model of two-phase flow. Numerical simulations are performed with both models. Compared to the results obtained from the HEM model, those obtained from the drift flux model are closer to RELAP5. The variations of critical heat flux with various geometric parameters and operating conditions are thoroughly investigated. The material required to construct the primary heat transport (PHT loop of NCBWR has been minimized using sequential quadratic programming. The stability of NCBWR has also been verified at the optimum point.

  15. Applicability of best-estimate analysis TRACE in terms of natural circulation BWR stability

    International Nuclear Information System (INIS)

    Furuya, Masahiro; Ueda, Nobuyuki; Nishi, Yoshihisa

    2011-01-01

    As a part of the international CAMP-Program of the US Nuclear Regulatory Commission (USNRC), the best-estimate code TRACE is validated with the stability database of SIRIUS-N Facility at high pressure. The TRACE code analyzed is version 5 patch level 2. The SIRIUS-N facility simulates thermal-hydraulics of the economic simplified BWR (ESBWR). The oscillation period correlates well with bubble transit time through the chimney region regardless of the system pressure, inlet subcooling and heat flux. Numerical results exhibits type-I density wave oscillation characteristics, since core inlet restriction shifts stability boundary toward the higher inlet subcooling, and chimney exit restriction enlarges instability region and oscillation amplitude. Stability maps in reference to the subcooling and heat flux obtained from the TRACE code agrees with those of the experimental data at 1 MPa. As the pressure increases from 2 MPa to 7.2 MPa, numerical results become much stable than the experimental results. This is because that two-phase frictional loss is underestimate, since the natural circulation flow rate of numerical results is higher by approximately 20% than that of experimental results. (author)

  16. DBSSP - A computer program for simulation of controlled circulation boiler and natural circulation boiler start up behavior

    International Nuclear Information System (INIS)

    Li Bin; Chen Tingkuan; Yang Dong

    2005-01-01

    In this paper, a computer program, Drum Boiler Start-up Simulation Program (DBSSP), is developed for simulating the start up behavior of controlled circulation and natural circulation boilers. The mathematical model developed here is based on the first principles of mass, energy and momentum conservations. In the boiler model, heat transfer in the waterwall, the superheater, the reheater and the economizer is simulated by the distributing parameter method, while heat transfer in the drum and the downcomer is simulated by lumped parameter analysis. The program can provide detailed flow and thermodynamic characteristics of the boiler components. The development of this program is based only on design data, so it can be used for any subcritical, controlled or natural circulation boiler. The simulation results were compared with experimental measurements, and good agreements between them were found. This program is expected to be useful for predicting the characteristics and the performance of controlled circulation and natural circulation boilers during the start up process. It also can be used to optimize a start up system for minimum start up time

  17. Experimental study on the safety of Kyoto University Research Reactor at natural circulation cooling mode

    International Nuclear Information System (INIS)

    Zhang, Jian; Shen, Xiuzhong; Fujihara, Yasuyuki; Sano, Tadafumi; Yamamoto, Toshihiro; Nakajima, Ken

    2015-01-01

    Highlights: • The natural circulation cooling capacity of Kyoto University Research Reactor (KUR) was experimentally investigated. • The distributions of the outlet temperature of the fuel elements under natural circulation operations were measured. • The average temperature rise and the average natural circulation flow velocity in core were calculated. • The safety of KUR under all of the normal operations with natural circulation cooling mode has been analyzed. • The natural circulation flow after the reactor shutdown was confirmed. - Abstract: In this study, the natural circulation cooling capacity of Kyoto University Research Reactor (KUR) is experimentally investigated by measuring the inlet and outlet temperatures of the core under natural circulation operation at various thermal powers ranging from 10 kW to 100 kW and the shutdown state. In view of the uneven power distribution and the resultant inconsistent coolant outlet temperature in the core, eight measuring points located separately in the outlet of the fuel elements were chosen to investigate the distribution of the outlet temperature of the core. The natural circulation cooling capacity represented by the average natural circulation flow velocity in the core is calculated from the temperature difference between the outlet and inlet temperature of the core. The measured outlet temperature of the fuel elements shows a cross-sectional distribution agreeing with the distribution of the thermal output of the fuel elements in the core. Since the measured outlet temperatures decrease quickly in the flow direction in a small local region above the outlet of the core, the mixing of the hot water out of the core with the cold water around the core outlet is found to happen in the small region not more than 5 cm far from the core outlet. The natural circulation flow velocity in the core increases non-linearly with the thermal power. The safety of KUR has been analysed by conservatively estimating the

  18. Flow characteristics of natural circulation in a lead-bismuth eutectic loop

    Institute of Scientific and Technical Information of China (English)

    Chen-Chong Yue; Liu-Li Chen; Ke-Feng Lyu; Yang Li; Sheng Gao; Yue-Jing Liu; Qun-Ying Huang

    2017-01-01

    Lead and lead-alloys are proposed in future advanced nuclear system as coolant and spallation target.To test the natural circulation and gas-lift and obtain thermal-hydraulics data for computational fluid dynamics (CFD) and system code validation,a lead-bismuth eutectic rectangular loop,the KYLIN-Ⅱ Thermal Hydraulic natural circulation test loop,has been designed and constructed by the FDS team.In this paper,theoretical analysis on natural circulation thermal-hydraulic performance is described and the steady-state natural circulation experiment is performed.The results indicated that the natural circulation capability depends on the loop resistance and the temperature and center height differences between the hot and cold legs.The theoretical analysis results agree well with,while the CFD deviate from,the experimental results.

  19. The role of natural circulation in the FFTF [Fast Flux Test Facility] passive safety tests

    International Nuclear Information System (INIS)

    Stover, R.L.; Padilla, A.; Burke, T.M.; Knecht, W.L.

    1987-03-01

    A series of tests were completed at the Fast Flux Test Facility to demonstrate the passive safety characteristics of liquid metal reactors with natural circulation flow. The first test consisted of transition from forced to natural circulation flow at an initial decay power of 0.3%. The second test represented an unprotected loss-of-flow transient to natural circulation from 50% power with the control rods prevented from scramming into the core. The third test was a steady-state, natural circulation condition with core fission powers up ato about 2.3%. Core sodium data and results of single and multi-channel computer models confirmed the reliability and effectiveness of natural circulation flow for liquid metal reactor safety

  20. Mixed convection in a two-phase flow cooling loop

    International Nuclear Information System (INIS)

    Janssens-Maenhout, G.; Daubner, M.; Knebel, J.U.

    2002-03-01

    This report summarizes the numerical simulations using the CFD code CFX4.1 which has additional models for subcooled flow boiling phenomena and the interfacial forces. The improved CFX4.1 code can be applied to the design of boiling induced mixed convection cooling loops in a defined parameter range. The experimental part describes the geysering experiments and the instability effects on the two-phase natural circulation flow. An experimentally validated flow pattern map in the Phase Change Number - Subcooling Number (N PCh - N Sub ) diagram defines the operational range in which flow instabilities such as geysering can be expected. One important perspective of this combined experimental/numerical work, which is in the field of two-phase flow, is its application to the development of accelerator driven systems (ADS). The main objective on an ADS is its potential to transmute minor actinides and long-lived fission products, thus participating in closing the fuel cycle. The development of an ADS is an important issue within the Euratom Fifth FP on Partitioning and Transmutation. One concept of an ADS, which is investigated in more detail within the ''preliminary design study of an experimental ADS'' Project (PDS-XADS) of the Euratom Fifth FP, is the XADS lead-bismuth cooled Experimental ADS of ANSALDO. An essential feature of this concept is the natural circulation of the primary coolant within the reactor pool. The natural circulation, which is driven by the density differences between the blanket and the heat exchanger, is enhanced by the injection of the nitrogen cover gas through spargers located in a riser part just above the blanket. This so-called gas-lift pump system has not been investigated in more detail nor has this gas-lift pump system been numerically/experimentally confirmed. The knowledge gained within the SUCO Programe, i.e. the modelling of the interfacial forces, the experimental work on flow instabilities and the modelling of the interfacial area

  1. Mixed convection in a two-phase flow cooling loop

    Energy Technology Data Exchange (ETDEWEB)

    Janssens-Maenhout, G.; Daubner, M.; Knebel, J.U.

    2002-03-01

    This report summarizes the numerical simulations using the CFD code CFX4.1 which has additional models for subcooled flow boiling phenomena and the interfacial forces. The improved CFX4.1 code can be applied to the design of boiling induced mixed convection cooling loops in a defined parameter range. The experimental part describes the geysering experiments and the instability effects on the two-phase natural circulation flow. An experimentally validated flow pattern map in the Phase Change Number - Subcooling Number (N{sub PCh} - N{sub Sub}) diagram defines the operational range in which flow instabilities such as geysering can be expected. One important perspective of this combined experimental/numerical work, which is in the field of two-phase flow, is its application to the development of accelerator driven systems (ADS). The main objective on an ADS is its potential to transmute minor actinides and long-lived fission products, thus participating in closing the fuel cycle. The development of an ADS is an important issue within the Euratom Fifth FP on Partitioning and Transmutation. One concept of an ADS, which is investigated in more detail within the ''preliminary design study of an experimental ADS'' Project (PDS-XADS) of the Euratom Fifth FP, is the XADS lead-bismuth cooled Experimental ADS of ANSALDO. An essential feature of this concept is the natural circulation of the primary coolant within the reactor pool. The natural circulation, which is driven by the density differences between the blanket and the heat exchanger, is enhanced by the injection of the nitrogen cover gas through spargers located in a riser part just above the blanket. This so-called gas-lift pump system has not been investigated in more detail nor has this gas-lift pump system been numerically/experimentally confirmed. The knowledge gained within the SUCO Programe, i.e. the modelling of the interfacial forces, the experimental work on flow instabilities and the

  2. Natural Circulation Characteristics at Low-Pressure Conditions through PANDA Experiments and ATHLET Simulations

    OpenAIRE

    Paladino, Domenico; Huggenberger, Max; Schäfer, Frank

    2008-01-01

    Natural circulation characteristics at low pressure/low power have been studied by performing experimental investigations and numerical simulations. The PANDA large-scale facility was used to provide valuable, high quality data on natural circulation characteristics as a function of several parameters and for a wide range of operating conditions. The new experimental data allow for testing and improving the capabilities of the thermal-hydraulic computer codes to be used for treating natural c...

  3. An experimental approach to improve the basin type solar still using an integrated natural circulation loop

    International Nuclear Information System (INIS)

    Rahmani, Ahmed; Boutriaa, Abdelouahab; Hadef, Amar

    2015-01-01

    Highlights: • A new experimental approach to improve the conventional solar still performances is proposed. • A passive natural circulation loop is integrated to the conventional solar still. • Natural circulation of humid-air in a closed loop is studied by the present study. • Natural circulation capability in driving air convection in the still was demonstrated. • Air convection created inside the still increase the evaporation heat and mass transfer. - Abstract: In this paper, a new experimental approach is proposed to enhance the performances of the conventional solar still using the natural circulation effect inside the still. The idea consists in generating air flow by a rectangular natural circulation loop appended to the rear side of the still. The proposed still was tested during summer period and the experimental data presented in this paper concerns four typical days. The convective heat transfer coefficient is evaluated and compared with Dunkle’s model. The comparison shows that convective heat transfer is considerably improved by the air convection created inside the still. The natural circulation phenomenon in the still is studied and a good agreement between the experimental data and Vijajan’s laminar correlation is found. Therefore, natural circulation phenomenon is found to have a good effect on the still performances where the still daily productivity is of 3.72 kg/m 2 and the maximum efficiency is of 45.15%

  4. Natural circulation in water cooled nuclear power plants: Phenomena, models, and methodology for system reliability assessments

    International Nuclear Information System (INIS)

    2005-11-01

    In recent years it has been recognized that the application of passive safety systems (i.e. those whose operation takes advantage of natural forces such as convection and gravity), can contribute to simplification and potentially to improved economics of new nuclear power plant designs. Further, the IAEA Conference on The Safety of Nuclear Power: Strategy for the Future which was convened in 1991 noted that for new plants 'the use of passive safety features is a desirable method of achieving simplification and increasing the reliability of the performance of essential safety functions, and should be used wherever appropriate'. Considering the weak driving forces of passive systems based on natural circulation, careful design and analysis methods must be employed to assure that the systems perform their intended functions. To support the development of advanced water cooled reactor designs with passive systems, investigations of natural circulation are an ongoing activity in several IAEA Member States. Some new designs also utilize natural circulation as a means to remove core power during normal operation. In response to the motivating factors discussed above, and to foster international collaboration on the enabling technology of passive systems that utilize natural circulation, an IAEA Coordinated Research Project (CRP) on Natural Circulation Phenomena, Modelling and Reliability of Passive Systems that Utilize Natural Circulation was started in early 2004. Building on the shared expertise within the CRP, this publication presents extensive information on natural circulation phenomena, models, predictive tools and experiments that currently support design and analyses of natural circulation systems and highlights areas where additional research is needed. Therefore, this publication serves both to provide a description of the present state of knowledge on natural circulation in water cooled nuclear power plants and to guide the planning and conduct of the CRP in

  5. Improvement in understanding of natural circulation phenomena in water cooled nuclear power plants

    International Nuclear Information System (INIS)

    Choi, Jong-Ho; Cleveland, John; Aksan, Nusret

    2011-01-01

    Highlights: ► Phenomena influencing natural circulation in passive systems. ► Behaviour in large pools of liquid. ► Effect of non-condensable gas on condensation heat transfer. ► Behaviour of containment emergency systems. ► Natural circulation flow and pressure drop in various geometries. - Abstract: The IAEA has organized a coordinated research project (CRP) on “Natural Circulation Phenomena, Modelling, and Reliability of Passive Systems That Utilize Natural Circulation.” Specific objectives of CRP were to (i) establish the status of knowledge: reactor start-up and operation, passive system initiation and operation, flow stability, 3-D effects, and scaling laws, (ii) investigate phenomena influencing reliability of passive natural circulation systems, (iii) review experimental databases for the phenomena, (iv) examine the ability of computer codes to predict natural circulation and related phenomena, and (v) apply methodologies for examining the reliability of passive systems. Sixteen institutes from 13 IAEA Member States have participated in this CRP. Twenty reference advanced water cooled reactor designs including evolutionary and innovative designs were selected to examine the use of natural circulation and passive systems in their designs. Twelve phenomena influencing natural circulation were identified and characterized: (1) behaviour in large pools of liquid, (2) effect of non-condensable gases on condensation heat transfer, (3) condensation on the containment structures, (4) behaviour of containment emergency systems, (5) thermo-fluid dynamics and pressure drops in various geometrical configurations, (6) natural circulation in closed loop, (7) steam liquid interaction, (8) gravity driven cooling and accumulator behaviour, (9) liquid temperature stratification, (10) behaviour of emergency heat exchangers and isolation condensers, (11) stratification and mixing of boron, and (12) core make-up tank behaviour. This paper summarizes the

  6. Investigations on the thermal-hydraulics of a natural circulation cooled BWR fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Kok, H.V.; Hagen, T.H.J.J. van der; Mudde, R.F. [Delft Univ. of Technology (Netherlands)

    1995-09-01

    A scaled natural circulation loop facility has been built after the Dodewaard Boiling Water Reactor, which is the only operating natural circulation cooled BWR in the world. The loop comprises one fuel assembly, a riser with a downcomer and a condenser with a cooling system. Freon-12 is used as a scaling liquid. This paper reports on the first measurements done with this facility. Quantities like the circulation flow, carry-under and the void-fraction have been measured as a function of power, pressure, liquid level, riser length, condensate temperature and friction factors. The behavior of the circulation flow can be understood by considering the driving force. Special attention has been paid to the carry-under, which has been shown to have a very important impact on the dynamics of a natural circulation cooled BWR.

  7. Analysis of hot leg natural circulation under station blackout severe accident

    International Nuclear Information System (INIS)

    Deng Jian; Cao Xuewu

    2007-01-01

    Under severe accidents, natural circulation flows are important to influence the accident progression and result in a pressurized water reactor (PWR). In a station blackout accident with no recovery of steam generator (SG) auxiliary feedwater (TMLB' severe accident scenario), the hot leg countercurrent natural circulation flow is analyzed by using a severe-accident code, to better understand its potential impacts on the creep-rupture timing among the surge line, the hot leg; and SG tubes. The results show that the natural circulation may delay the failure time of the hot leg. The recirculation ratio and the hot mixing factor are also calculated and discussed. (authors)

  8. Study on scaling law of PWR natural circulation with motion condition

    International Nuclear Information System (INIS)

    Lu Donghua; Xiao Zejun; Chen Bingde

    2009-01-01

    For some nuclear reactors installed on automobiles, boats or deep sea vehicles, it is an important way to investigate their system safety by performing natural circulation experiments under motion condition. This paper studied the natural circulation on moving plants based on work of static natural circulation scaling method. With rigid motion theory, acceleration at each point was obtained on primary system and introduced to momentum equation. Thus a set of motion similar criteria were obtained. Furthermore, equal and unequal height simulation were analyzed. As to the unequal one, non isochronous simulation was needed for displacement and angular acceleration. (authors)

  9. Influence of reactor design on the establishment of natural circulation in pool-type LMFBR

    International Nuclear Information System (INIS)

    Durham, M.E.

    1976-01-01

    The general principles involved in establishing natural circulation in a pool-type liquid metal cooled fast breeder reactor following loss of a.c. supplies are elucidated and the effects of design features by use of the computer code MELANI are quantified. It is shown that natural circulation can provide a feasible means of emergency core cooling in addition to that provided by pony motors. The choice of primary pump rundown time has a significant effect in controlling peak core outlet temperatures in the hypothetical case of natural circulation alone being the core heat removal process. (author)

  10. Study on model of onset of nucleate boiling in natural circulation with subcooled boiling using unascertained mathematics

    Energy Technology Data Exchange (ETDEWEB)

    Zhou Tao [Department of Thermal Engineering, Tsinghua University, Beijing 100084 (China)]. E-mail: zhoutao@mail.tsinghua.edu.cn; Wang Zenghui [Department of Engineering Mechanics, Tsinghua University, Beijing 100084 (China); Yang Ruichang [Department of Thermal Engineering, Tsinghua University, Beijing 100084 (China)

    2005-10-01

    Experiment data got from onset of nucleate boiling (ONB) in natural circulation is analyzed using unascertained mathematics. Unitary mathematics model of the relation between the temperature and onset of nucleate boiling is built up to analysis ONB. Multiple unascertained mathematics models are also built up with the onset of natural circulation boiling equation based on the experiment. Unascertained mathematics makes that affirmative results are a range of numbers that reflect the fluctuation of experiment data more truly. The fluctuating value with the distribution function F(x) is the feature of unascertained mathematics model and can express fluctuating experimental data. Real status can be actually described through using unascertained mathematics. Thus, for calculation of ONB point, the description of unascertained mathematics model is more precise than common mathematics model. Based on the unascertained mathematics, a new ONB model is developed, which is important for advanced reactor safety analysis. It is conceivable that the unascertained mathematics could be applied to many other two-phase measurements as well.

  11. Study on model of onset of nucleate boiling in natural circulation with subcooled boiling using unascertained mathematics

    International Nuclear Information System (INIS)

    Zhou Tao; Wang Zenghui; Yang Ruichang

    2005-01-01

    Experiment data got from onset of nucleate boiling (ONB) in natural circulation is analyzed using unascertained mathematics. Unitary mathematics model of the relation between the temperature and onset of nucleate boiling is built up to analysis ONB. Multiple unascertained mathematics models are also built up with the onset of natural circulation boiling equation based on the experiment. Unascertained mathematics makes that affirmative results are a range of numbers that reflect the fluctuation of experiment data more truly. The fluctuating value with the distribution function F(x) is the feature of unascertained mathematics model and can express fluctuating experimental data. Real status can be actually described through using unascertained mathematics. Thus, for calculation of ONB point, the description of unascertained mathematics model is more precise than common mathematics model. Based on the unascertained mathematics, a new ONB model is developed, which is important for advanced reactor safety analysis. It is conceivable that the unascertained mathematics could be applied to many other two-phase measurements as well

  12. Measurement of liquid level in a natural circulation circuit using an ultrasonic technique

    International Nuclear Information System (INIS)

    Barbosa, Amanda Cardozo; Su, Jian

    2017-01-01

    The measurement by an ultrasonic technique of the water level in the expansion tank of the Natural Circulation Circuit (NCC) of the Experimental Thermo-Hydraulic Laboratory of the Institute of Nuclear Engineering is presented. In the single-phase NCC operation the water level in the expansion tank is stable. However, during the two-phase operation, oscillations occur in the water level due to temperature and vacuum fraction variations. Thus, the development of a technique that allows the measurement of these oscillations, will allow an estimation of the variation of the vacuum fraction of the circuit over time. The experimental set - up was performed on a test bench, using an ultrasonic transducer. The ultrasonic technique used is pulse-echo, in which the same transducer is the transmitter and receiver of the signal. The transducer-shoe assembly is part of an ultrasonic system consisting of an ultrasonic signal generating plate, transducers and a computer (PC) with a program in LabView to control the system. The program is able to calculate the transit time that the ultrasonic signals take to cross the tank base wall, the layer (level) of liquid and return to the transducer. Knowing the speed of the ultrasound in the wall and in the liquid it is possible to calculate the thickness of the wall and the height of the liquid. Measurements were made by filling the tank with a known volume of water and under varying temperature conditions, from room temperature to 90 deg C. The liquid heights are determined and the volume of water calculated by measuring the temperature with a digital thermometer. The volumes measured were highly accurate when compared to the known volumes

  13. Investigation of natural circulation instability and transients in passively safe novel modular reactor

    Science.gov (United States)

    Shi, Shanbin

    The Purdue Novel Modular Reactor (NMR) is a new type small modular reactor (SMR) that belongs to the design of boiling water reactor (BWR). Specifically, the NMR is one third the height and area of a conventional BWR reactor pressure vessel (RPV) with an electric output of 50 MWe. The fuel cycle length of the NMR-50 is extended up to 10 years due to optimized neutronics design. The NMR-50 is designed with double passive engineering safety system. However, natural circulation BWRs (NCBWR) could experience certain operational difficulties due to flow instabilities that occur at low pressure and low power conditions. Static instabilities (i.e. flow excursion (Ledinegg) instability and flow pattern transition instability) and dynamic instabilities (i.e. density wave instability and flashing/condensation instability) pose a significant challenge in two-phase natural circulation systems. In order to experimentally study the natural circulation flow instability, a proper scaling methodology is needed to build a reduced-size test facility. The scaling analysis of the NMR uses a three-level scaling method, which was developed and applied for the design of the Purdue Multi-dimensional Integral Test Assembly (PUMA). Scaling criteria is derived from dimensionless field equations and constitutive equations. The scaling process is validated by the RELAP5 analysis for both steady state and startup transients. A new well-scaled natural circulation test facility is designed and constructed based on the scaling analysis of the NMR-50. The experimental facility is installed with different equipment to measure various thermal-hydraulic parameters such as pressure, temperature, mass flow rate and void fraction. Characterization tests are performed before the startup transient tests and quasi-steady tests to determine the loop flow resistance. The controlling system and data acquisition system are programmed with LabVIEW to realize the real-time control and data storage. The thermal

  14. Measurement of Liquid-Metal Two-Phase Flow with a Dynamic Neutron Radiography

    International Nuclear Information System (INIS)

    Cha, J. E.; Lim, I. C.; Kim, H. R.; Kim, C. M.; Nam, H. Y.; Saito, Y.

    2005-01-01

    The dynamic neutron radiography(DNR) has complementary characteristics to X-ray radiography and is suitable to visualization and measurement of a multi-phase flow research in a metallic duct and liquid metal flow. The flow-field information of liquid metal system is very important for the safety analysis of fast breeder reactor and the design of the spallation target of accelerator driven system. A DNR technique was applied to visualize the flow field in the gas-liquid metal two-phase flow with the HANARO-beam facility. The lead bismuth eutectic and the nitrogen gas were used to construct the two-phase flow field in the natural circulation U-channel. The two-phase flow images in the riser were taken at various combinations of the liquid flow and gas flow with high frame-rate neutron radiography at 1000 fps

  15. Calculation of reverse flow in inverted U-Tubes of steam generator during natural circulation

    International Nuclear Information System (INIS)

    Yang Ruichang; Liu Jinggong; Liu Ruolei; Qin Shiwei; Huang Yanping

    2010-01-01

    The mechanism of reverse flow in inverted U-tubes of steam generators of pressurized water reactors during natural circulation is analyzed by using the full range characteristic curve of parallel U-tubes. A lumped-distributed model to calculate the reverse flow occurred in inverted U-tubes of real steam generators with a large number of U-tubes during natural circulation is developed. The model has the advantages of quick calculation and high accuracy for the analysis of reverse flow in inverted U-tubes of real steam generators with natural circulation. This model has been used to calculate the normal and reverse flows occurred in inverted U-tubes of a steam generator with natural circulation. The comparison of calculated results indicates a well agreement with that predicted by the model in which normal or reverse flow in each individual U-tube is analyzed, which verifies the reliability of the model developed in this paper. (authors)

  16. Natural Circulation Characteristics at Low-Pressure Conditions through PANDA Experiments and ATHLET Simulations

    Directory of Open Access Journals (Sweden)

    Domenico Paladino

    2008-01-01

    Full Text Available Natural circulation characteristics at low pressure/low power have been studied by performing experimental investigations and numerical simulations. The PANDA large-scale facility was used to provide valuable, high quality data on natural circulation characteristics as a function of several parameters and for a wide range of operating conditions. The new experimental data allow for testing and improving the capabilities of the thermal-hydraulic computer codes to be used for treating natural circulation loops in a range with increased attention. This paper presents a synthesis of a part of the results obtained within the EU-Project NACUSP “natural circulation and stability performance of boiling water reactors.” It does so by using the experimental results produced in PANDA and by showing some examples of numerical simulations performed with the thermal-hydraulic code ATHLET.

  17. A generalised correlation for the steady state flow in single-phase natural circulation loops

    International Nuclear Information System (INIS)

    Vijayan, P.K.; Bade, M.H.; Saha, D.; Sinha, R.K.; Venkat Raj, V.

    2000-08-01

    To establish the heat transport capability of natural circulation loops, it is essential to know the flow rate. A generalized correlation for steady state flow valid for uniform and non-uniform diameter loops has been theoretically derived

  18. Status of the IAEA coordinated research project on natural circulation phenomena, modelling, and reliability of passive systems that utilize natural circulation

    International Nuclear Information System (INIS)

    Reyes, J.N. Jr.; Cleveland, J.; Aksan, N.

    2004-01-01

    The International Atomic Energy Agency (IAEA) has established a Coordinated Research Project (CRP) titled ''Natural Circulation Phenomena, Modelling and Reliability of Passive Safety Systems that Utilize Natural Circulation. '' This work has been organized within the framework of the IAEA Department of Nuclear Energy's Technical Working Groups for Advanced Technologies for Light Water Reactors and Heavy Water Reactors (the TWG-LWR and the TWG-HWR). This CRP is part of IAEA's effort to foster international collaborations that strive to improve the economic performance of future water-cooled nuclear power plants while meeting stringent safety requirements. Thus far, IAEA has established 12 research agreements with organizations from industrialized Member States and 3 research contracts with organizations from developing Member States. The objective of the CRP is to enhance our understanding of natural circulation phenomena in water-cooled reactors and passive safety systems. The CRP participants are particularly interested in establishing a natural circulation and passive safety system thermal hydraulic database that can be used to benchmark computer codes for advanced reactor systems design and safety analysis. An important aspect of this CRP relates to developing methodologies to assess the reliability of passive safety systems in advanced reactor designs. This paper describes the motivation and objectives of the CRP, the research plan, and the role of each of the participating organizations. (author)

  19. FFTF primary system transition to natural circulation from low reactor power

    International Nuclear Information System (INIS)

    Bouchey, G.D.; Additon, S.L.; Nutt, W.T.

    1980-01-01

    Plans for reactor and primary loop natural circulation testing in the Fast Flux Test Facility (FFTF) are summarized. Detailed pretest planning with an emphasis on understanding the implications of process noise and model uncertainties for model verification and test acceptance are discussed for a transition to natural circulation in the reactor core and primary heat transport loops from initial conditions of 5% of rated reactor power and 75% of full flow

  20. Experiences with loss of natural circulation events, performed experiments, analysis, computations and development of operational documents

    International Nuclear Information System (INIS)

    Nagy, L.; Varju, A.; Nagy, S.

    1996-01-01

    The refuelling of the unit 4 was started on 18 June, 1988. At the time of the event the reactor was in cold shutdown state, with atmospheric pressure, the reactor head was removed. On June 30 the operational personnel performed a planned switch over of natural circulation from loops 4, 6 to loops 1, 3. In the meantime the effectiveness of the core cooling by natural circulation decreased sharply for about 3 hour-period. After switching over the natural circulation among the loops the operating personnel isolated the loops 4., 6. and started to drain them. Nitrogen used to drain the loops was unintentionally injected into the loops in operation and large amount of primary coolant was pushed out from the SG primary side to the reactor vessel. The operators tried to stop the disturbance of natural circulation by starting the booster pump of make-up system periodically to the working loops. During this injection the personnel performed venting few times to take away the gas-air mixture from the top of the SG primary headers. After all the restoration of the natural circulation was achieved by continuous venting the SG headers. During 1993 annual refuelling outage of Unit 2 at Paks NPP a deterioration of natural circulation in reactor coolant system occurred. A special maintenance task was being performed to repair the cladding of the sealing bellows between the reactor vessel and reactor cavity

  1. Analysis of reverse flow in inverted U-tubes of steam generator under natural circulation condition

    International Nuclear Information System (INIS)

    Yang Ruichang; Liu Ruolei; Liu Jinggong; Qin Shiwei

    2008-01-01

    In this paper, we report on the analysis of reverse flow in inverted U-tubes of a steam generator under natural circulation condition. The mechanism of reverse flow in inverted U-tubes of the steam generator with natural circulation is graphically analyzed by using the full-range characteristic curve of parallel U-tubes. The mathematical model and numerical calculation method for analyzing the reverse flow in inverted U-tubes of the steam generator with natural circulation have been developed. The reverse flow in an inverted U-tube steam generator of a simulated pressurized water reactor with natural circulation in analyzed. Through the calculation, the mass flow rates of normal and reverse flows in individual U-tubes are obtained. The predicted sharp drop of the fluid temperature in the inlet plenum of the steam generator due to reverse flow agrees very well with the experimental data. This indicates that the developed mathematical model and solution method can be used to correctly predict the reverse flow in the inverted U-tubes of the steam generator with natural circulation. The obtained results also show that in the analysis of natural circulation flow in the primary circuit, the reverse flow in the inverted U-tubes of the steam generator must be taken into account. (author)

  2. NATBWR: a steady-state model for natural circulation in boiling-water reactors

    International Nuclear Information System (INIS)

    Healzer, J.M.; Abdollahian, D.

    1983-02-01

    This report documents the NATBWR steady-state BWR natural-circulation model and activities under EPRI Project RP1561-1 to gather data and predict the natural-circulation operation of the BWR. The report is organized into two parts, with the first part describing the NATBWR model and applications of the model to available BWR natural-circulation data and the second part providing user and programming information about the model. Five different operating BWR's were selected to demonstrate the application of the NATBWR model, one of each type from BWR/1 through BWR/4. For each operating plant, the available natural circulation data has been compared to model predictions. In addition to the data predictions, the behavior of the BWR system at reduced inventory, where the system is isolated and scrammed, and cooling provided by natural circulation has been analyzed. Finally, included as an appendix to Part 1 of this report is a discussion of the stability of the BWR system at natural-circulation conditions

  3. Preliminary model validation for integral stability behavior in molten salt natural circulation

    International Nuclear Information System (INIS)

    Cai Chuangxiong; He Zhaozhong; Chen Kun

    2017-01-01

    Passive safety system is an important characteristic of Fluoride-Salt-Cooled High-Temperature Reactor (FHR). In order to remove the decay heat, a direct reactor auxiliary cooling system (DRACS) which uses the passive safety technology is proposed to the FHR as the ultimate heat sink. The DRACS is relying on the natural circulation, so the study of molten salt natural circulation plays an important role at TMSR. A high-temperature molten salt natural circulation test loop has been designed and constructed at the TMSR center of the Chinese Academy of Sciences (CAS) to understand the characteristics of the natural circulation and verify the design model. It adopts nitrate salt as the working fluid to simulate fluoride salts, and uses air as the ultimate heat sink. The test shows the operation very well and has a very nice performance, the Heat transfer coefficients (salt-salt or salt-air), power of the loop, heat loss of molten salt pool (or molten salt pipe or air cooling tower), starting time of the loop, flow rate that can be verified in this loop. A series of experiments have been done and the results show that the experimental data are well matched with the design data. This paper aims at analyzing the molten salt circulation model, studying the characteristics of the natural circulation, and verifying the Integral stability behavior by three different natural circulation experiments. Also, the experiment is going on, and more experiments will been carry out to study the molten salt natural circulation for optimizing the design. (author)

  4. Models development for natural circulation and its transition process in nuclear power plant

    International Nuclear Information System (INIS)

    Yu Lei; Cai Qi; Cai Zhangsheng; Xie Haiyan

    2008-01-01

    On the basis of nuclear power plant (NPP) best-estimate transient analysis code RELAP5/MOD3, the point reactor kinetics model in RELAP5/MOD3 was replaced by the two-group, 3-D space and time dependent neutron kinetic model, in order to exactly analyze the responses of key parameters in natural circulation and its transition process considering the reactivity feedback. The coupled model for three-dimensional physics and thermohydraulics was established and corresponding computing code was developed. Using developed code, natural circulation of NPP and its transiton process were calculated and analyzed. Compared with the experiment data, the calculated results show that its high precise avoids the shortage that the point reactor equation can not reflect the reactivity exactly. This code can be a computing and analysis tool for forced circulation and natural circulation and their transitions. (authors)

  5. Study on natural circulation flow under reactor cavity flooding condition in advanced PWRs

    International Nuclear Information System (INIS)

    Tao Jun; Yang Jiang; Cao Jianhua; Lu Xianghui; Guo Dingqing

    2015-01-01

    Cavity flooding is an important severe accident management measure for the in-vessel retention of a degraded core by external reactor vessel cooling in advanced PWRs. A code simulation study on the natural circulation flow in the gap between the reactor vessel wall and insulation material under cavity flooding condition is performed by using a detailed mechanistic thermal-hydraulic code package RELAP 5. By simulating of an experiment carried out for studying the natural circulation flow for APR1400 shows that the code is applicable for analyzing the circulation flow under this condition. The analysis results show that heat removal capacity of the natural circulation flow in AP1000 is sufficient to prevent thermal failure of the reactor vessel under bounding heat load. Several conclusions can be drawn from the sensitivity analysis. Larger coolant inlet area induced larger natural circulation flow rate. The outlet should be large enough and should not be submerged by the cavity water to vent the steam-water mixture. In the implementation of cavity flooding, the flooding water level should be high enough to provide sufficient natural circulation driven force. (authors)

  6. Hydrodynamics of a natural circulation loop in a scaled-down steam drum-riser-downcomer assembly

    Energy Technology Data Exchange (ETDEWEB)

    Basu, Dipankar N., E-mail: dnbasu@iitg.ernet.in; Patil, N.D.; Bhattacharyya, Souvik; Das, P.K.

    2013-12-15

    Highlights: • Experimental investigation of loop hydrodynamics in a scaled-down simulated AHWR. • Identification of flow regimes and transition analyzing conductance probe signal. • Downcomer flow maximizes with fully developed churn flow and lowest for bubbly flow. • Highest downcomer flow rate is achieved with identical air supply to both risers. • Interaction of varying flow patterns reduces downcomer flow for unequal operation. - Abstract: Complex interactions of different phases, widely varying frictional characteristics of different flow regimes and the involvement of multiple scales of transport make the modelling of a two-phase natural circulation loop (NCL) exceedingly difficult. The knowledge base about the dependency of downcomer flow rate on riser-side flow patterns, particularly for systems with multiple parallel channels is barely developed, necessitating the need for detailed experimentation. The present study focuses on developing a scaled-down test facility relevant to the Advanced Heavy Water Reactor conceived in the atomic energy programme of India to study the hydrodynamics of the NCL using air and water as test fluids. An experimental facility with two risers, one downcomer and a phase-separating drum was fabricated. Conductivity probes and photographic techniques are used to characterize the two phase flow. Normalized voltage signals obtained from the amplified output of conductivity probes and their subsequent analysis through probability distribution function reveal the presence of different two-phase flow patterns in the riser tubes. With the increase in air supply per riser void fraction in the two-phase mixture increases and gradually flow patterns transform from bubbly to fully developed annular through slug, churn and dispersed annular flow regimes. Downcomer flow rate increases rapidly with air supply till a maximum and then starts decreasing due to enhanced frictional forces. However, the maximum value of downcomer water

  7. A state-of-the-art review of recent advances in supercritical natural circulation loops for nuclear applications

    International Nuclear Information System (INIS)

    Sarkar, Milan K.S.; Tilak, Abhilash K.; Basu, Dipankar N.

    2014-01-01

    Highlights: • Review of state-of-the-art on supercritical natural circulation loops. • Summarization of steady-state thermalhydraulic behavior for nuclear applications. • Survey of non-linear and linear stability analyses and parametric effects. • Details of experimental investigations and critical findings form each. • Emphasis on need of further experimentation in future. - Abstract: The concept of supercritical natural circulation loop (SCNCL) is an important inclusion in Generation-IV nuclear reactors. Use of supercritical fluids promises a simplified design, along with higher thermal efficiency for heat transport systems. Characteristics of such loops are markedly different from its single-phase and two-phase counterparts, while carrying quite a few similarities with both as well. Therefore significant number of research studies is carried out on SCNCL in the present millennium and current work presents a state-of-the-art summary of all associated observations. Most of the reported studies are theoretical in nature, with only a limited number of experimental works being reported. A number of indigenous computation codes were developed, while use of commercial software can also be found. Thermal–hydraulic and heat transfer aspects are discussed in details, showing the gradual growth of knowledge and comprehending the influence of various geometric and operating variables on steady-state profile. Water and carbon dioxide are identified as the only fluids considered for analysis both numerically and experimentally. Both time-domain and frequency-domain approach of stability analysis are discussed meticulously. Available experimental works are described, with exhaustive discussion on the novelty of the concerned facility and major observations. Finally a few topics are ear-marked as the possible guidelines for future research

  8. Natural circulating passive cooling system for nuclear reactor containment structure

    Science.gov (United States)

    Gou, Perng-Fei; Wade, Gentry E.

    1990-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  9. Characterization of natural circulation looping of emergency cooling systems in naval and advanced reactors

    International Nuclear Information System (INIS)

    Macedo, Luiz Alberto; Baptista Filho, Benedito Dias

    2000-01-01

    This paper describes the natural circuit looping, resumes the main project characteristics, presents results of the hydraulic characterization, consisting of pressure loss measurements, and presents results from calibration tests of the power and flow measurements and the first experiments in natural circulation. Those experiments comprised transients in natural circulation with application of application of power steps. The results shown a non linear behaviour of the magnetic flow meter and a dependence on the fluid temperature as well. The assembly circuit/instrumentation/data acquisition system is suitable for the research on emergency cooling passive systems

  10. An experimental study on the flow instabilities and critical heat flux under natural circulation

    International Nuclear Information System (INIS)

    Kim, Yun II; Chang, Soon Heung

    2004-01-01

    This study has been carried out to investigate the hydrodynamic stabilities and Critical Heat Flux (CHF) characteristics for the natural and forced circulation. A low pressure experimental loop was constructed, and experiments under various conditions have been performed. In the experiments of the natural circulation, flow oscillations has been observed and the average mass flux under flow oscillation have been measured. Several parameters such as heat flux, the inlet temperature of test section, friction valve opening and riser length have been varied in order to investigate their effects on the flow stability of the natural circulation system. And the CHF data from low flow experiments, namely the natural and forced circulation, have been compared with each other to identify the effects of the flow instabilities on the CHF for the natural circulation mode. The test conditions for the CHF experiments were a low flow of less than 70 kg/m 2 s of water in a vertical round tube with diameter of 0.008 m at near atmospheric pressure. (author)

  11. The development of natural circulation operation support program for ship nuclear power machinery

    International Nuclear Information System (INIS)

    Hao, Jianli; Chen, Wenzhen; Chen, Zhiyun

    2012-01-01

    Highlights: ► The natural circulation under various ocean and ship motion conditions is studied. ► A natural circulation operation support computer program (NCOSP) is developed with Simulink. ► The NCOSP program has the merit of easy input preparation, fast and accurate simulation. ► The NCOSP is suitable for the fast parameter simulation of ship nuclear power machinery. -- Abstract: The existing simulation program of ship nuclear power machinery (SNPM) cannot adequately deal with the natural circulation (NC) operation and the effects of various ocean conditions and ship motion. Aiming at the problem, the natural circulation operation support computer program for SNPM is developed, in which the momentum conservation equation of the primary loop, some heat transfer and flow resistance models and equations are modified for the various ocean conditions and ship motion. The additional pressure loss model and effective height model for the control volume in the gyration movement, simple harmonic rolling and pitching movements are also discussed in the paper. Furthermore, the transient response to load change under NC conditions is analyzed by the developed program. The results are compared with those predicted by the modified RELAP5/mod3.2 code. It is shown that the natural circulation operation support program (NCOSP) is simple in the input preparation, runs fast and has a satisfactory precision, and is therefore very suitable for the operating field support of SNPM under the conditions of NC.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  13. Two-phase flows during a discharge of liquefied gases, initially at saturation. Effect of the nature of the fluid; Ecoulements diphasiques lors de la vidange de gaz liquefies initialement a saturation. Influence de la nature du fluide

    Energy Technology Data Exchange (ETDEWEB)

    Alix, P.

    1997-10-03

    In the case of a confinement loss (breakage of a connection piece) on a pressurized liquefied gas tank, a critical two-phase (liquid-vapour) flow is generated. This thesis is aimed at the validation of models describing these flows with various fluids (water, R 11, methanol, ethyl acetate, pure butane, commercial butane), using a pilot experimental plant. Results show that reduced upstream pressure is the main parameter, thus indicating that a model can be validated using minimal fluids. The homogenous models DEM and HRM appear to be more precise

  14. Theoretical analysis of effect of ocean condition on natural circulation flow

    International Nuclear Information System (INIS)

    Gong Houjun; Yang Xingtuan; Jiang Shengyao; Liu Zhiyong

    2010-01-01

    According to the simulation loop of Integrated natural circulation reactor,the mathematical model of natural circulation in non-inertial reference system is established, and the influence mechanism of ocean condition upon natural circulation is analyzed. Software is programmed to investigate the behaviors in the cases of rolling without heating power, static state with different power and rolling with heating power, and calculation results show that: the inertia force added by rolling causes the periodical fluctuating of the flow rate of channels, but it is not the direct reason of core flow fluctuation. The heave changes the driving head, and causes the same flow rate fluctuation of all channels. Inclining makes the core flow rate decrease, but the change of flow rate of different channels is different.(authors)

  15. Simulation of natural circulation on an integral type experimental facility, MASLWR

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Youngjong; Lim, Sungwon; Ha, Jaejoo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    The OSU MASLWR test facility was reconfigured to eliminate a recurring grounding problem and improve facility reliability in anticipation of conducting an IAEA International Collaborative Standard Problem (ICSP). The purpose of ICSP is to provide experimental data on flow instability phenomena under natural circulation conditions and coupled containment/reactor vessel behavior in integral-type reactors, and to evaluate system code capabilities to predict natural circulation phenomena for integral type PWR, by simulating an integrated experiment. A natural circulation in the primary side during various core powers is analyzed using TASS/SMR code for the integral type experimental facility. The calculation results show higher steady state primary flow than experiment. If it matches the initial flow with experiment, it shows lower primary flow than experiment according to the increase of power. The code predictions may be improved by applying a Reynolds number dependent form loss coefficient to accurately account for unrecoverable pressure losses.

  16. Study of turbulent natural-circulation flow and low-Prandtl-number forced-convection flow

    International Nuclear Information System (INIS)

    Chung, K.S.; Thompson, D.H.

    1980-01-01

    Calculational methods and results are discussed for the coupled energy and momentum equations of turbulent natural circulation flow and low Prandtl number forced convection flow. The objective of this paper is to develop a calculational method for the study of the thermal-hydraulic behavior of coolant flowing in a liquid metal fast breeder reactor channel under natural circulation conditions. The two-equation turbulence model is used to evaluate the turbulent momentum transport property. Because the analogy between momentum transfer and heat transfer does not generally hold for low Prandtl number fluid and natural circulation flow conditions, the turbulent thermal conductivity is calculated independently using equations similar to the two-equation turbulence model. The numerical technique used in the calculation is the finite element method

  17. CIRCUS and DESIRE: Experimental facilities for research on natural-circulation-cooled boiling water reactors

    International Nuclear Information System (INIS)

    Kruijf, W.J.M. de; Haden, T.H.J.J. van der; Zboray, R.; Manera, A.; Mudde, R.F.

    2002-01-01

    At the Delft University of Technology two thermohydraulic test facilities are being used to study the characteristics of Boiling Water Reactors (BWRs) with natural circulation core cooling. The focus of the research is on the stability characteristics of the system. DESIRE is a test facility with freon-12 as scaling fluid in which one fuel bundle of a natural-circulation BWR is simulated. The neutronic feedback can be simulated artificially. DESIRE is used to study the stability of the system at nominal and beyond nominal conditions. CIRCUS is a full-height facility with water, consisting of four parallel fuel channels and four parallel bypass channels with a common riser or with parallel riser sections. It is used to study the start-up characteristics of a natural-circulation BWR at low pressures and low power. In this paper a description of both facilities is given and the research items are presented. (author)

  18. Natural circulation and stratification in the various passive safety systems of the SWR 1000

    International Nuclear Information System (INIS)

    Meseth, J.

    2002-01-01

    In some of the passive safety systems of Siemens' SWR 1000 boiling water reactor (i.e. the emergency condensers and containment cooling condensers), natural circulation is the main effect on both the primary and secondary sides by which optimum system efficiency is achieved. Other passive safety systems of the SWR 1000 require natural circulation on the secondary side only (condensation of steam discharged by the safety and relief valves; cooling of the Reactor Pressure Vessel (RPV) by flooding from the outside in case of core melt), while still other systems require stratification to be effective (i.e. the passive pressure pulse transmitters and steam-driven scram tanks). Complex natural circulation and stratification can take place simultaneously if fluids with different densities are enclosed in a single volume (in a core melt accident, for example, the nitrogen, steam and hydrogen in the containment). Related problems and the solutions thereto planned for the SWR 1000 are reported from the designer's viewpoint. (author)

  19. Simulation experiments for hot-leg U-bend two-phase flow phenomena

    International Nuclear Information System (INIS)

    Ishii, M.; Hsu, J.T.; Tucholke, D.; Lambert, G.; Kataoka, I.

    1986-01-01

    In order to study the two-phase natural circulation and flow termination during a small break loss of coolant accident in LWR, simulation experiments have been performed. Based on the two-phase flow scaling criteria developed under this program, an adiabatic hot leg U-bend simulation loop using nitrogen gas and water and a Freon 113 boiling and condensation loop were built. The nitrogen-water system has been used to isolate key hydrodynamic phenomena from heat transfer problems, whereas the Freon loop has been used to study the effect of phase changes and fluid properties. Various tests were carried out to establish the basic mechanism of the flow termination and reestablishment as well as to obtain essential information on scale effects of parameters such as the loop frictional resistance, thermal center, U-bend curvature and inlet geometry. In addition to the above experimental study, a preliminary modeling study has been carried out for two-phase flow in a large vertical pipe at relatively low gas fluxes typical of natural circulation conditions

  20. Two phase titanium aluminide alloy

    Science.gov (United States)

    Deevi, Seetharama C.; Liu, C. T.

    2001-01-01

    A two-phase titanic aluminide alloy having a lamellar microstructure with little intercolony structures. The alloy can include fine particles such as boride particles at colony boundaries and/or grain boundary equiaxed structures. The alloy can include alloying additions such as .ltoreq.10 at % W, Nb and/or Mo. The alloy can be free of Cr, V, Mn, Cu and/or Ni and can include, in atomic %, 45 to 55% Ti, 40 to 50% Al, 1 to 5% Nb, 0.3 to 2% W, up to 1% Mo and 0.1 to 0.3% B. In weight %, the alloy can include 57 to 60% Ti, 30 to 32% Al, 4 to 9% Nb, up to 2% Mo, 2 to 8% W and 0.02 to 0.08% B.

  1. Analysis of the natural circulation by the computer code RELAP-5

    International Nuclear Information System (INIS)

    Kordis, I.; Mavko, B.; Zeljko, M.

    1984-01-01

    The analysis of the natural circulation is one of the first analysis that was done at IJS with the computer code RELAP 5/MOS 1/CY 018. Specific model of the system was made for the natural circulation. The first 400 s of the transient were analyzed. At that time pumps are rotating only by coolant flow. First results show quite realistic picture of the transient although some changes should be made, especially on the steam generator model due to the unrealistic oscillations of the coolant flow on the secondary side. (author)

  2. Application of noise analysis to stability determination of a natural circulation cooled BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hagen, T.H.J.J. van der; Dam, H. van; Hoogenboom, J.E.; Nissen, W.H.M.; Oosterkamp, W.J.

    1988-01-01

    Experiments were performed on the Dodewaard natural circulation cooled BWR at different conditions. The absolute stability was determined by measuring system responses to control rod and steam flow valve steps. Changes in core stability were studied using the signal of an average power range monitor (APRM) in time domain (auto-correlation function and impulse response) and in frequency domain (power spectral density and peaking factor), the outlet void fraction and variations of the incore coolant velocity. It is shown that the reactor is very stable and that cooling by natural circulation improves load following. Stability monitoring can be performed by all mentioned methods but using APRM signals in frequency domain is preferred.

  3. On Stability of Natural-circulation-cooled Boiling Water Reactors during Start-up (Experimental Results)

    International Nuclear Information System (INIS)

    Manera, A.; Van der Hagen, T.H.J.J.

    2002-01-01

    The characteristics of flashing-induced instabilities, which are of importance during the start-up phase of natural-circulation Boiling Water Reactors (BWRs), are studied. Experiments at typical start-up conditions (low power and low pressure) are carried out on a steam/water natural circulation loop. The mechanism of flashing-induced instability is analyzed in detail and it is found that non-equilibrium between phases and enthalpy transport plays an important role in the instability process. Pressure and steam volume in the steam dome are found to have a stabilizing effect. The main characteristics of the instabilities have been analyzed. (authors)

  4. Stability of single-phase natural circulation with inverted U-tube steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Sanders, J.

    1988-08-01

    For natural circulation it is shown that parallel flow in the tubes of an inverted U-tube stream generator can be, at certain power levels, unstable. A mathematical model, based on one-dimensional Oberbeck-Boussinesq equations, shows that stability can be attained if in some tubes the water flows backward, opposite to the normal flow direction. The results are compared to measurements obtained from the natural circulation test A2-77A in the LOBI-MOD2 integral system test facility.

  5. Assessment of RELAP5/MOD2 against a natural circulation experiment in Nuclear Power Plant Borssele

    International Nuclear Information System (INIS)

    Winters, L.

    1993-07-01

    As part of the ICAP (International Code Assessment and Applications Program) agreement between ECN (Netherlands Energy Research Foundation) and USNRC, ECN has performed a number of assessment calculations for the thermohydraulic system analysis code RELAP5/MOD2/36.05. This document describes the assessment of this computer program versus a natural circulation experiment as conducted at the Borssele Nuclear Power Plant. The results of this comparison show that the code RELAP5/MOD2 predicts well the natural circulation behaviour of Nuclear Power Plant Borssele

  6. Experiments on natural circulation during PWR severe accidents and their analysis

    International Nuclear Information System (INIS)

    Sehgal, B.R.; Stewart, W.A.; Sha, W.T.

    1988-01-01

    Buoyancy-induced natural circulation flows will occur during the early-part of PWR high pressure accident scenarios. These flows affect several key parameters; in particular, the course of such accidents will most probably change due to local failures occurring in the primary coolant system (CS) before substantial core degradation. Natural circulation flow patterns were measured in a one-seventh scale PWR PCS facility at Westinghouse RandD laboratories. The measured flow and temperature distributions are report in this paper. The experiments were analyzed with the COMMIX code and good agreement was obtained between data and calculations. 10 refs., 8 figs., 2 tabs

  7. Analysis of SBO accident and natural circulation of 49-2 swimming pool reactor

    International Nuclear Information System (INIS)

    Wu Yuanyuan; Liu Tiancai; Sun Wei

    2012-01-01

    The transient thermal hydraulic characteristics of 49-2 Swimming Pool Reactor (SPR) were analyzed by RELAP5/MOD3.3 code to verify the capability of natural circulation and minus reactivity feedback for accident mitigation under the condition of station blackout (SBO). Then, the effects on accident consequence and sequence for core channels and primary pumps were briefly discussed. The calculation results show that the reactor can be shutdown by the effect of minus reactivity feedback, and the residual heat can be removed through the stable natural circulation. Therefore, it demonstrates that the 49-2 SPR is safe during the accident of SBO. (authors)

  8. A potential of boiling water power reactors with a natural circulation of a coolant

    International Nuclear Information System (INIS)

    Osmachkin, V.S.; Sokolov, I.N.

    1998-01-01

    The use of the natural circulation of coolant in the boiling water reactors simplifies a reactor control and facilities the service of the equipment components. The moderated core power loads allows the long fuel burnup, good control ability and large water stock set up the enhancement of safety level. That is considered to be very important for isolated regions or small countries. In the paper a high safety level and effectiveness of BWRs with natural circulation are reviewed. The limitations of flow stability and protection measures are being discussed. Some recent efforts in designing of such reactors are described.(author)

  9. Analytical study on thermal-hydraulic behavior of transient from forced circulation to natural circulation in JRR-3

    International Nuclear Information System (INIS)

    Hirano, Masashi; Sudo, Yukio

    1986-01-01

    Transient thermal-hydraulic behaviors of the JRR-3 which is an open-pool type research reactor has been analyzed with the THYDE-P1 code. The focal point is the thermal-hydraulic behaviors related to the core flow reversal during the transition from forced circulation downflow to natural circulation upflow. In the case of a loss-of-coolant accident (LOCA), for example, the core flow reversal is expected to occur just after the water pool isolation from the primary cooling loop with a leak. The core flow reversal should cause a sudden increase in fuel temperature and a steep decrease in the departure-from-nucleate-boiling ratio (DNBR) and the phenomenon is, therefore, very important especially for safety design and evaluation of research reactors. Major purposes of the present work are to clarify physical phenomena during the transient and to identify important parameters affecting the peak fuel temperature and the minimum DNBR. The results calculated with THYDE-P1 assuming the sequences of events of the loss-of-offsite power and LOCA help us to understand the phenomena both qualitatively and quantitatively, with respect to the safety design and evaluation. (author)

  10. Natural circulation studies in a LBE loop for a wide range of temperature

    International Nuclear Information System (INIS)

    Borgohain, A.; Srivastava, A.K.; Jana, S.S.; Maheshwari, N.K.; Kulkarni, R.D.; Vijayan, P.K.; Tewari, R.; Ram, A. Maruthi; Jha, S.K.

    2016-01-01

    Highlights: • A high temperature Lead Bismuth Eutectic loop named as Kilo Temperature Loop (KTL) has been made. • Natural circulation experimental studies were carried out and reported in the range of 200–780 °C. • The experiments at high temperature were carried in inert atmosphere to avoid oxidation of the loop material. • Theoretical studies are carried out to simulate the loop with natural circulation in primary as well as in the secondary side. • The predictions of the code LeBENC used to simulate the natural circulation in the loop are compared with the experimental results. - Abstract: Lead–Bismuth Eutectic (LBE) is increasingly getting more attention as a coolant for advanced reactor systems. It is also the primary coolant of the Compact High Temperature Reactor (CHTR) being designed at Bhabha Atomic Research Centre (BARC). A high temperature liquid metal loop named as Kilo Temperature Loop (KTL) has been installed at BARC for thermal hydraulics, instrument development and material related studies. Natural circulation experimental studies were carried out for the power range of 200–1200 W in the loop. The corresponding LBE flow rate is calculated to be in the range of 0.075–0.12 kg/s. Transient studies for start-up of natural circulation in the loop, loss of heat sink and step power change have also been carried out. The maximum temperature of the loop operated so far is 1100 °C. A computer code named LeBENC has been developed at BARC to simulate the natural circulation characteristics in closed loops. The salient features of the code include ability to handle non-uniform diameter components, axial thermal conduction in fluid and heat losses from the piping to the environment. The code has been modified to take into account of two natural circulation loops in series so that the natural cooling by argon gas in the secondary side of the loop can be simulated. This paper deals with the description of the loop and its operation. The various

  11. An experimental study of two-phase flow instability on two parallel channel with low steam quality

    International Nuclear Information System (INIS)

    Jiang Shengyao; Wu shaorong; Bo Jinhai; Yao Meisheng; Han Bing; Zhang Youjie

    1988-01-01

    An experimental result of two-phase flow instability on two parallel channel natural circulation with low steam quality is presented. The comparison of instability in the single channel and that in parallel channel is given. The effect of unequal inlet resistance coefficient and unequal power on the parallel channel instability is described and the behaviour of instability with equal exit steam quality in the two channel is investigated

  12. Reduced scaling of thermal-hydraulic circuits for studies of PWR reactors natural circulation

    International Nuclear Information System (INIS)

    Botelho, D.A.

    1993-01-01

    The Ishii et al. hydrodynamic similarity criteria for natural circulation were used for scaling reduced models of prototype passive residual heat removal system of a 600 M We PWR. The physical scales of the thermohydraulic parameters obtained presented a reasonable agreement when compared with simplified analytic models of the systems. (author)

  13. Evaluation of High-Pressure RCS Natural Circulations Under Severe Accident Conditions

    International Nuclear Information System (INIS)

    Lee, Byung Chul; Bang, Young Suk; Suh, Nam Duk

    2006-01-01

    Since TMI-2 accident, the occurrence of severe accident natural circulations inside RCS during entire in-vessel core melt progressions before the reactor vessel breach had been emphasized and tried to clarify its thermal-hydraulic characteristics. As one of consolidated outcomes of these efforts, sophisticated models have been presented to explain the effects of a variety of engineering and phenomenological factors involved during severe accident mitigation on the integrity of RCS pressure boundaries, i.e. reactor pressure vessel(RPV), RCS coolant pipe and steam generator tubes. In general, natural circulation occurs due to density differences, which for single phase flow, is typically generated by temperature differences. Three natural circulation flows can be formed during severe accidents: in-vessel, hot leg countercurrent flow and flow through the coolant loops. Each of these flows may be present during high-pressure transients such as station blackout (SBO) and total loss of feedwater (TLOFW). As a part of research works in order to contribute on the completeness of severe accident management guidance (SAMG) in domestic plants by quantitatively assessing the RCS natural circulations on its integrity, this study presents basic approach for this work and some preliminary results of these efforts with development of appropriately detailed RCS model using MELCOR computer code

  14. Integral test facilities for validation of the performance of passive safety systems and natural circulation

    International Nuclear Information System (INIS)

    Choi, J. H.

    2010-10-01

    Passive safety systems are becoming an important component in advanced reactor designs. This has led to an international interest in examining natural circulation phenomena as this may play an important role in the operation of these passive safety systems. Understanding reactor system behaviour is a challenging process due to the complex interactions between components and associated phenomena. Properly scaled integral test facilities can be used to explore these complex interactions. In addition, system analysis computer codes can be used as predictive tools in understanding the complex reactor system behaviour. However, before the application of system analysis computer codes for reactor design, it is capability in making predictions needs to be validated against the experimental data from a properly scaled integral test facility. The IAEA has organized a coordinated research project (CRP) on natural circulation phenomena, modeling and reliability of passive systems that utilize natural circulation. This paper is a part of research results from this CRP and describes representative international integral test facilities that can be used for data collection for reactor types in which natural circulation may play an important role. Example experiments were described along with the analyses of these example cases in order to examine the ability of system codes to model the phenomena that are occurring in the test facilities. (Author)

  15. Adequacy of power-to-volume scaling philosophy to simulate natural circulation in Integral Test Facilities

    International Nuclear Information System (INIS)

    Nayak, A.K.; Vijayan, P.K.; Saha, D.; Venkat Raj, V.; Aritomi, Masanori

    1998-01-01

    Theoretical and experimental investigations were carried out to study the adequacy of power-to-volume scaling philosophy for the simulation of natural circulation and to establish the scaling philosophy applicable for the design of the Integral Test Facility (ITF-AHWR) for the Indian Advanced Heavy Water Reactor (AHWR). The results indicate that a reduction in the flow channel diameter of the scaled facility as required by the power-to-volume scaling philosophy may affect the simulation of natural circulation behaviour of the prototype plants. This is caused by the distortions due to the inability to simulate the frictional resistance of the scaled facility. Hence, it is recommended that the flow channel diameter of the scaled facility should be as close as possible to the prototype. This was verified by comparing the natural circulation behaviour of a prototype 220 MWe Indian PHWR and its scaled facility (FISBE-1) designed based on power-to-volume scaling philosophy. It is suggested from examinations using a mathematical model and a computer code that the FISBE-1 simulates the steady state and the general trend of transient natural circulation behaviour of the prototype reactor adequately. Finally the proposed scaling method was applied for the design of the ITF-AHWR. (author)

  16. Natural circulation analysis for the advanced neutron source reactor refueling process 11

    Energy Technology Data Exchange (ETDEWEB)

    Tucker, R.F.; Dasardhi, S.; Elkassabgi, Y. [Texas A& M Univ., Kingsville, TX (United States); Yoder, G.L. [Oak Ridge National Lab., TN (United States)

    1995-09-01

    During the refueling process of the Advanced Neutron Source Reactor (ANSR), the spent fuel elements must be moved from the primary coolant loop (containing D{sub 2}O), through a heavy water pool, and finally into a light water spent fuel storage area. The present refueling scheme utilizes remote refueling equipment to move the spent fuel elements through a D{sub 2}O filled stack and tunnel into a temporary storage canal. A transfer lock is used to move the spent fuel elements from the D{sub 2}O-filled interim storage canal to a light water pool. Each spent fuel element must be cooled during this process, using either natural circulation or forced convection. This paper presents a summary of the numerical techniques used to analyze natural circulation cooling of the ANSR fuel elements as well as selected results of the calculations. Details of the analysis indicate that coolant velocities below 10 cm/s exist in the coolant channels under single phase natural circulation conditions. Also, boiling does not occur within the channels if power levels are below a few hundred kW when the core transitions to natural circulation conditions.

  17. A comparison of the RELAP5/MOD3 code with the IIST natural circulation experiments

    International Nuclear Information System (INIS)

    Ferng, Y.M.; Lee, C.H.

    1995-01-01

    A series of experiments dealing with variable secondary-side cooling conditions have been conducted at the IIST facility, including the natural circulation experiments under the secondary-side conditions of normal feedwater, loss of feedwater, and full of air. Different cooling conditions at the secondary side directly affect the primary-to-secondary heat transfer and then may influence the heat removal capability of natural circulation in the primary system. The corresponding analytical work is performed using the RELAP5/MOD3 code. Good agreement is reached both qualitatively and quantitatively between the experimental data and calculated results, demonstrating the satisfactory assessment of RELAP5/MOD3 code compared with the IIST natural circulation experiments. The cooling conditions at the secondary side have no significant effect on the heat removal capability of natural circulation as long as sufficient coolant exists on the steam generator secondary side, based on current IIST data and analytical results. Continuous increase of the core temperature and system pressure is also demonstrated experimentally and analytically in the test with the secondary side dry for the sake of deficient heat transfer capability at the steam generator secondary system

  18. Experimental study on convective heat transfer of water flow in a heated tube under natural circulation

    International Nuclear Information System (INIS)

    Yang Ruichang; Liu Ruolei; Zhong Yong; Liu Tao

    2006-01-01

    This paper reports on an experimental study on transitional heat transfer of water flow in a heated vertical tube under natural circulation conditions. In the experiments the local and average heat transfer coefficients were obtained. The experimental data were compared with the predictions by a forced flow correlation available in the literature. The comparisons show that the Nusselt number value in the fully developed region is about 30% lower than the predictions by the forced flow correlation due to flow laminarization in the layer induced by co-current bulk natural circulation and free convection. By using the Rayleigh number Ra to represent the influence of free convection on heat transfer, the empirical correlations for the calculation of local and average heat transfer behavior in the tube at natural circulation have been developed. The empirical correlations are in good agreement with the experimental data. Based on the experimental results, the effect of the thermal entry-length behavior on heat transfer design in the tube under natural circulation was evaluated

  19. Non-linear time series analysis on flow instability of natural circulation under rolling motion condition

    International Nuclear Information System (INIS)

    Zhang, Wenchao; Tan, Sichao; Gao, Puzhen; Wang, Zhanwei; Zhang, Liansheng; Zhang, Hong

    2014-01-01

    Highlights: • Natural circulation flow instabilities in rolling motion are studied. • The method of non-linear time series analysis is used. • Non-linear evolution characteristic of flow instability is analyzed. • Irregular complex flow oscillations are chaotic oscillations. • The effect of rolling parameter on the threshold of chaotic oscillation is studied. - Abstract: Non-linear characteristics of natural circulation flow instabilities under rolling motion conditions were studied by the method of non-linear time series analysis. Experimental flow time series of different dimensionless power and rolling parameters were analyzed based on phase space reconstruction theory. Attractors which were reconstructed in phase space and the geometric invariants, including correlation dimension, Kolmogorov entropy and largest Lyapunov exponent, were determined. Non-linear characteristics of natural circulation flow instabilities under rolling motion conditions was studied based on the results of the geometric invariant analysis. The results indicated that the values of the geometric invariants first increase and then decrease as dimensionless power increases which indicated the non-linear characteristics of the system first enhance and then weaken. The irregular complex flow oscillation is typical chaotic oscillation because the value of geometric invariants is at maximum. The threshold of chaotic oscillation becomes larger as the rolling frequency or rolling amplitude becomes big. The main influencing factors that influence the non-linear characteristics of the natural circulation system under rolling motion are thermal driving force, flow resistance and the additional forces caused by rolling motion. The non-linear characteristics of the natural circulation system under rolling motion changes caused by the change of the feedback and coupling degree among these influencing factors when the dimensionless power or rolling parameters changes

  20. Parametric studies to establish natural circulation in advanced heavy water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bhatia, S K; Dhawan, M L [Reactor Engineering Division, Bhabha Atomic Research Centre, Mumbai (India)

    1994-06-01

    Design of Advanced Heavy Water Reactor (AHWR) is in progress. It consists of vertical pressure tubes with boiling light water coolant flowing through the tubes and heavy water moderator in the calandria. In PHWRs, core heat removal is through forced circulation of the coolant by PHT pumps. In AHWR, no PHT pumps are used and core heat is carried away by natural circulation of the coolant due to density difference between steam/water mixture inside the core and the water region outside the core. This passive means of core heat removal results in a number of benefits viz. (a) extra length of piping, valves, instruments, power supply and control systems for functioning of instruments are eliminated, (b) plant layout is simplified, (c) maintenance of valves and instruments is reduced. Natural circulation in AHWR is achieved by keeping the steam drum at a sufficient height above the core to get the required driving force. The loop height depends on many factors i.e. channel power, V{sub c}/V{sub f} ratio (ratio of coolant volume to fuel volume) and core height. The effect of these parameters on the loop height to establish natural circulation have been studied and presented. (author). 1 ref., 1 fig., 1 tab.

  1. Effect of the inlet throttling on the thermal-hydraulic instability of the natural circulation BWR

    International Nuclear Information System (INIS)

    Furuya, Masahiro; Inada, Fumio; Yoneda, Kimitoshi

    1997-01-01

    Although it is well-established that inlet restriction has a stabilizing for forced circulation BWR, the effect of inlet on the thermal-hydraulic stability of natural circulation BWR remains unknown since increasing inlet restriction affect thermal-hydraulic stability due to reduction of the recirculation flow rate. Therefore experiments have been conducted to investigate the effect of inlet restriction on the thermal-hydraulic stability. A test facility used in this experiments was designed and constructed to have non-dimensional values which are nearly equal to those of natural circulation BWR. Experimental results showed that driving force of the natural circulation was described as a function of heat flux and inlet subcooling independent of inlet restriction. Stability maps in reference to the channel inlet subcooling, heat flux were presented for various inlet restriction which were carried out by an analysis based on the homogeneous flow various using this function. Instability region during the inlet subcooling shifted to the higher inlet subcooling with increasing inlet restriction and became larger with increasing heat flux. (author)

  2. Scaling of two-phase flow transients using reduced pressure system and simulant fluid

    International Nuclear Information System (INIS)

    Kocamustafaogullari, G.; Ishii, M.

    1987-01-01

    Scaling criteria for a natural circulation loop under single-phase flow conditions are derived. Based on these criteria, practical applications for designing a scaled-down model are considered. Particular emphasis is placed on scaling a test model at reduced pressure levels compared to a prototype and on fluid-to-fluid scaling. The large number of similarty groups which are to be matched between modell and prototype makes the design of a scale model a challenging tasks. The present study demonstrates a new approach to this clasical problen using two-phase flow scaling parameters. It indicates that a real time scaling is not a practical solution and a scaled-down model should have an accelerated (shortened) time scale. An important result is the proposed new scaling methodology for simulating pressure transients. It is obtained by considerung the changes of the fluid property groups which appear within the two-phase similarity parameters and the single-phase to two-phase flow transition prameters. Sample calculations are performed for modeling two-phase flow transients of a high pressure water system by a low-pressure water system or a Freon system. It is shown that modeling is possible for both cases for simulation pressure transients. However, simulation of phase change transitions is not possible by a reduced pressure water system without distortion in either power or time. (orig.)

  3. Study of core flow distribution for small modular natural circulation lead or lead-alloy cooled fast reactors

    International Nuclear Information System (INIS)

    Chen, Zhao; Zhao, Pengcheng; Zhou, Guangming; Chen, Hongli

    2014-01-01

    Highlights: • A core flow distribution calculation code for natural circulation LFRs was developed. • The comparison study between the channel method and the CFD method was conducted. • The core flow distribution analysis and optimization design for a 10MW natural circulation LFR was conducted. - Abstract: Small modular natural circulation lead or lead-alloy cooled fast reactor (LFR) is a potential candidate for LFR development. It has many attractive advantages such as reduced capital costs and inherent safety. The core flow distribution calculation is an important issue for nuclear reactor design, which will provide important input parameters to thermal-hydraulic analysis and safety analysis. The core flow distribution calculation of a natural circulation LFR is different from that of a forced circulation reactor. In a forced circulation reactor, the core flow distribution can be controlled and adjusted by the pump power and the flow distributor, while in a natural circulation reactor, the core flow distribution is automatically adjusted according to the relationship between the local power and the local resistance feature. In this paper, a non-uniform heated parallel channel flow distribution calculation code was developed and the comparison study between the channel method and the CFD method was carried out to assess the exactness of the developed code. The core flow distribution analysis and optimization design for a 10MW natural circulation LFR was conducted using the developed code. A core flow distribution optimization design scheme for a 10MW natural circulation LFR was proposed according to the optimization analysis results

  4. Passive safety systems and natural circulation in water cooled nuclear power plants

    International Nuclear Information System (INIS)

    2009-11-01

    Nuclear power produces 15% of the world's electricity. Many countries are planning to either introduce nuclear energy or expand their nuclear generating capacity. Design organizations are incorporating both proven means and new approaches for reducing the capital costs of their advanced designs. In the future most new nuclear plants will be of evolutionary design, often pursuing economies of scale. In the longer term, innovative designs could help to promote a new era of nuclear power. Since the mid-1980s it has been recognized that the application of passive safety systems (i.e. those whose operation takes advantage of natural forces such as convection and gravity), can contribute to simplification and potentially improve economics of new nuclear power plant designs. The IAEA Conference on The Safety of Nuclear Power: Strategy for the Future, which was convened in 1991, noted that for new plants 'the use of passive safety features is a desirable method of achieving simplification and increasing the reliability of the performance of essential safety functions, and should be used wherever appropriate'. Some new designs also utilize natural circulation as a means to remove core power during normal operation. The use of passive systems can eliminate the costs associated with the installation, maintenance, and operation of active systems that require multiple pumps with independent and redundant electric power supplies. However, considering the weak driving forces of passive systems based on natural circulation, careful design and analysis methods must be employed to ensure that the systems perform their intended functions. To support the development of advanced water cooled reactor designs with passive systems, investigations of natural circulation are conducted in several IAEA Member States with advanced reactor development programmes. To foster international collaboration on the enabling technology of passive systems that utilize natural circulation, the IAEA

  5. Experiments on natural circulation of lead-bismuth in the TALL test facility

    International Nuclear Information System (INIS)

    Ma, W.M.; Karbojian, A.; Sehgal, B.R.

    2005-01-01

    Full text of publication follows: Lead-bismuth eutectic (LBE) is a potential candidate coolant for next generation liquid metal reactors due to its favorable properties such as being chemical inert and low melting point, in comparison with sodium and lead considered as coolants in FBRs. Having a high atomic number of LBE allows it be well suited as a spallation target for accelerator-driven systems (ADS) which have been proposed for the transmutation of nuclear waste. Due to its strong buoyancy, the LBE-cooled system should also have significant natural circulation, which is desirable for so-called Generation IV nuclear reactors, which like to employ passive safety and reliability. But so far, very little experimental data have been published on the natural circulation thermal-hydraulics of LBE-cooled systems. Motivated by the increasing interest in LBE-cooled fast reactors and ADS, a test facility called Thermal-hydraulic ADS Lead-bismuth Loop (TALL) was designed and constructed at KTH to investigate the thermalhydraulic characteristics of liquid LBE. The facility consists of a primary loop (LBE loop) and a secondary loop (oil loop). The LBE loop consists of sump tank, core tank, expansion tank, heat exchanger, EM pump, EM flowmeter, electric heaters and instrumentation. The heating of LBE in the core tank and its cooling in the heat exchanger allows natural convection flows as should occur in the prototypic vessel. Recently, our experimental study on natural circulation was performed on the TALL test facility. This paper will present the experimental results and analysis. The facility is of 6.8 m height which is comparable to the full height of the LBE heat exchange circuit in the ANSALDO ADS reactor vessel design, and has been scaled for prototypic (power/volume) ratio to represent the main components. Their LBE volume, flow velocity and heating rates correspond to one tube of the heat exchanger design chosen. During the experiments, the main adjustable

  6. Measurement of two-phase flow variables in a BWR by analysis of in-core neutron detector noise signals

    International Nuclear Information System (INIS)

    Stekelenburg, A.J.C.; Hagen, T.H.J.J. van der

    1996-01-01

    In this paper, the state of the art of the measurement of two-phase flow variables in a boiling water reactor (BWR) by analysis of in-core neutron detector noise signals is given. It is concluded that the neutronic processes involved in neutron noise are quite well understood, but that little is known about the density fluctuations in two-phase flow which are the main cause of the neutron noise. For this reason, the neutron noise measurements, like the well known two-detector velocity measurements, are still difficult to interpret. By analyzing neutron noise measurements in a natural circulation cooled BWR, it is illustrated that, once a theory on the density fluctuations is developed, two-phase flow can be monitored with a single in-core detector. (author). 70 refs, 4 figs

  7. Thermal hydraulic phenomenology for the heating process in a natural circulation facility

    International Nuclear Information System (INIS)

    Torres, Walmir M.; Macedo, Luiz A.; Mesquita, Roberto N.; Masotti, Paulo Henrique F.; Libardi, Rosani Maria P.; Sabundjian, Gaiane; Andrade, Delvonei A.; Umbehaun, Pedro Ernesto; Conti, Thadeu N.; Silva Filho, Mauro F.S.; Melo, Gabriel R.

    2009-01-01

    This work describes thermal hydraulic phenomenology observed for the heating process in a natural circulation facility. Glass made circuit allows observations of the thermal hydraulic processes over several regions. Natural convection, natural circulation, nucleated sub-cooled, saturated boiling and some flow patterns such as, bubbly, slug and churn flow are observed and described. Facility heated and cooled parts are responsible for the natural circulation when in operation. An expansion tank accommodates the fluid density variations due to the temperature changes and void fraction. Instrumentation consists of thermocouples distributed along the circuit. Two differential pressure transducers are used for pressure and level measurements. Instrumentation signals and images are simultaneously acquired to help with phenomenon description. A CCD digital camera at a 250μs shutter speed is used for the images acquisition. Phenomenology described is based on a test under 1.1 x 10 5 W/m 2 of heat flux which corresponds to an electrical heater power of 7000 W and 0.0236 kg/s (85 l/h) of cooling flow rate. (author)

  8. Numerical analysis of the fluid dynamics in a natural circulation loop

    International Nuclear Information System (INIS)

    Angelo, Gabriel

    2013-01-01

    Natural circulation loops apply to many engineering applications such as: water heating solar energy system (thermo-siphons), thermal management of electrical components (voltage converter), geothermal energy, nuclear reactors, etc. In pressurized water nuclear reactors, known as PWR's, the natural circulation loops are employed to ensure passive safety. In critical situations, the heat transfer will occur only by natural convection, without any external control or mechanical devices. This feature is desired and has been considered in modern nuclear reactor projects. This work consists of a numerical study of the natural circulation loop, located at the Instituto de Pesquisas Energeticas e Nucleares / Comissao Nacional de Energia Nuclear in Sao Paulo, Brazil, in order to establish the flow pattern in single phase conditions. The comparison of numerical results to experiments in transient condition revealed significant deviations for the Zero Equation turbulence model. Intermediate deviations for the Eddy Viscosity Turbulence Equation (EVTE), k - ω, SST e SSG models. And the best results are obtained by the k - ε e DES models (with better results for the k - ε model). (author)

  9. Evaluation of transient natural circulation behavior during accident in low power/shutdown condition of YGN units 3/4

    International Nuclear Information System (INIS)

    Bang, Young Seok; Kim, Kap; Seul, Kwang Won; Kim, Hho Jung

    1997-01-01

    A transient natural circulation behavior during a LOCA at hot-standby operation is evaluated for YGN Units 3/4. The plant initial condition is determined within the EOP limitation as suitable to hot-standby mode and the transient scenario is prepared as relevant to evaluation of transient natural circulation. A 0.4% cold leg break with loss of off-site power is calculated with RELAP5/MOD3.2, whose predictability has been verified for SBLOCA natural circulation test, S-NC-8B. Through one hour transient analysis, it is found that the plant has its own decay heat removal capability by natural circulation following a LOCA at hot-standby mode. Additional calculation is performed to investigate an effect of HPSI flow on natural circulation

  10. Development and validation of natural circulation based systems for new WWER designs

    International Nuclear Information System (INIS)

    Kurakov, Y.A.; Dragunov, Y.G.; Podshibiakin, A.K.; Fil, N.S.; Logvinov, S.A.; Sitnik, Y.K.; Berkovich, V.M.; Taranov, G.S.

    2002-01-01

    Elaboration and introduction of NPP designs with improved technical and economic parameters are defined as an important element of the National Program of nuclear power development approved by the Russian Federation Government in 1998. This Program considers the designs of WWER-1000/V-392 and WWER-640/ V-407 power units as the priority projects of the new generation NPPs with increased safety. A number of passive systems based on natural circulation phenomena are used in V-392 and V-407 designs to prevent or mitigate severe accidents. Design basis, configuration and effect of some naturally driven systems of V-392 design sited at Novovoronezh are mainly reflected in the present paper. One of the most important mean for severe accident prevention in V-392 design is so called SPOT - passive heat removal system designed to remove core decay heat in case of station blackout (including failure of all diesel generators). This system extracts the steam from the steam generator, condenses it and returns water to steam generator by natural circulation. The SPOT heat exchangers are cooled by atmospheric air coming by natural circulation through a special direct action control gates which operate passively as well. Extensive experimental investigation of the different aspects of this system operation has been carried out to validate its functioning under real plant conditions. In particular, full-scale section of air heat exchanger-condenser has been tested with natural circulation steam, condensate and air paths modeled. The environment air temperature and steam pressure condensing were varied in the wide range, and the relevant experimental results are being discussed in this paper. The effect of wind velocity and direction to the containment is also checked by the experiments. (author)

  11. Computational simulation of flow and heat transfer in single-phase natural circulation loops

    International Nuclear Information System (INIS)

    Pinheiro, Larissa Cunha

    2017-01-01

    Passive decay heat removal systems based on natural circulation are essential assets for the new Gen III+ nuclear power reactors and nuclear spent fuel pools. The aim of the present work is to study both laminar and turbulent flow and heat transfer in single-phase natural circulation systems through computational fluid dynamics simulations. The working fluid is considered to be incompressible with constant properties. In the way, the Boussinesq Natural Convection Hypothesis was applied. The model chosen for the turbulence closure problem was the k -- εThe commercial computational fluid dynamics code ANSYS CFX 15.0 was used to obtain the numerical solution of the governing equations. Two single-phase natural circulation circuits were studied, a 2D toroidal loop and a 3D rectangular loop, both with the same boundary conditions of: prescribed heat flux at the heater and fixed wall temperature at the cooler. The validation and verification was performed with the numerical data provided by DESRAYAUD et al. [1] and the experimental data provided by MISALE et al. [2] and KUMAR et al. [3]. An excellent agreement between the Reynolds number (Re) and the modified Grashof number (Gr_m), independently of Prandtl Pr number was observed. However, the convergence interval was observed to be variable with Pr, thus indicating that Pr is a stability governing parameter for natural circulation. Multiple steady states was obtained for Pr = 0,7. Finally, the effect of inclination was studied for the 3D circuit, both in-plane and out-of-plane inclinations were verified for the steady state laminar regime. As a conclusion, the Re for the out-of-plane inclination was in perfect agreement with the correlation found for the zero inclination system, while for the in-plane inclined system the results differ from that of the corresponding vertical loop. (author)

  12. Velocity Fields Measurement of Natural Circulation Flow inside a Pool Using PIV Technique

    International Nuclear Information System (INIS)

    Kim, Seok; Kim, Dong Eok; Youn, Young Jung; Euh, Dong Jin; Song, Chul Hwa

    2012-01-01

    Thermal stratification is encountered in large pool of water increasingly being used as heat sink in new generation of advanced reactors. These large pools at near atmospheric pressure provide a heat sink for heat removal from the reactor or steam generator, and the containment by natural circulation as well as a source of water for core cooling. For examples, the PAFS (passive auxiliary feedwater system) is one of the advanced safety features adopted in the APR+ (Advanced Power Reactor Plus), which is intended to completely replace the conventional active auxiliary feedwater system. The PAFS cools down the steam generator secondary side and eventually removes the decay heat from the reactor core by adopting a natural convection mechanism. In a pool, the heat transfer from the PCHX (passive condensation heat exchanger) contributed to increase the pool temperature up to the saturation condition and induce the natural circulation flow of the PCCT (passive condensate cooling tank) pool water. When a heat rod is placed horizontally in a pool of water, the fluid adjacent to the heat rod gets heated up. In the process, its density reduces and by virtue of the buoyancy force, the fluid in this region moves up. After reaching the top free surface, the heated water moves towards the other side wall of the pool along the free surface. Since this heated water is cooling, it goes downward along the wall at the other side wall. Above heater rod, a natural circulation flow is formed. However, there is no flow below heater rod until pool water temperature increases to saturation temperature. In this study, velocity measurement was conducted to reveal a natural circulation flow structure in a small pool using PIV (particle image velocimetry) measurement technique

  13. Experimental facility with two-phase flow and with high concentration of non-condensable gases for research and development of emergency cooling system of advanced nuclear reactors

    International Nuclear Information System (INIS)

    Macedo, Luiz Alberto; Baptista Filho, Benedito Dias

    2006-01-01

    The development of emergency cooling passive systems of advanced nuclear reactors requires the research of some relative processes to natural circulation, in two-phase flow conditions involving condensation processes in the presence of non-condensable gases. This work describes the main characteristics of the experimental facility called Bancada de Circulacao Natural (BCN), designed for natural circulation experiments in a system with a hot source, electric heater, a cold source, heat exchanger, operating with two-phase flow and with high concentration of noncondensable gas, air. The operational tests, the data acquisition system and the first experimental results in natural circulation are presented. The experiments are transitory in natural circulation considering power steps. The distribution of temperatures and the behavior of the flow and of the pressure are analyzed. The experimental facility, the instrumentation and the data acquisition system demonstrated to be adapted for the purposes of research of emergency cooling passive systems, operating with two-phase flow and with high concentration of noncondensable gases. (author)

  14. Self-pressurization analysis of the natural circulation integral nuclear reactor using a new dynamic model

    Directory of Open Access Journals (Sweden)

    Ali Farsoon Pilehvar

    2018-06-01

    Full Text Available Self-pressurization analysis of the natural circulation integral nuclear reactor through a new dynamic model is studied. Unlike conventional pressurized water reactors, this reactor type controls the system pressure using saturated coolant water in the steam dome at the top of the pressure vessel. Self-pressurization model is developed based on conservation of mass, volume, and energy by predicting the condensation that occurs in the steam dome and the flashing inside the chimney using the partial differential equation. A simple but functional model is adopted for the steam generator. The obtained results indicate that the variable measurement is consistent with design data and that this new model is able to predict the dynamics of the reactor in different situations. It is revealed that flashing and condensation power are in direct relation with the stability of the system pressure, without which pressure convergence cannot be established. Keywords: Condensation Power, Flashing Phenomenon, Natural Circulation, Self-Pressurization, Small Modular Reactor

  15. BWR Passive Containment Cooling System by condensation-driven natural circulation

    International Nuclear Information System (INIS)

    Vierow, K.M.; Townsend, H.E.; Fitch, J.R.; Andersen, J.G.M.; Alamgir, M.; Schrock, V.E.

    1991-01-01

    A method of long-term decay heat removal which is safe, reliable, and passive has been incorporated into the design of the Simplified Boiling Water Reactor (SBWR). The primary functions of the Passive Containment Cooling System (PCCS) are to remove heat and maintain the containment pressure below allowable levels following a LOCA. A key component of the PCCS is the PCC condenser unit (PCC). By natural circulation, a steam-nitrogen mixture flows into the PCC heat exchanger, condensate drains to the reactor pressure vessel (RPV), and noncondensables are vented to the suppression chamber (S/C). This analysis focuses on three significant thermal-hydraulic phenomena which occur in the system. Specifically, steam condensation in the presence of a noncondensable, the PCC noncondensable venting and the natural circulation are discussed. Results of TRACG simulations are presented which show that the PCCS performs its intended functions. (author)

  16. Comparison of auxiliary feedwater and EDRS operation during natural circulation of MRX

    International Nuclear Information System (INIS)

    Kim, Jae Hak; Park, Goon Cherl

    1997-01-01

    The MRX is an integral type ship reactor with 100 MWt power, which is designed by Japan Atomic Energy Research Institute. It is characterized by integral type PWR, in-vessel type control rod drive mechanism, water-filled containment vessel and passive decay heat removal system. Marine reactor should have high passive safety. Therefore, in this study, we simulated the loss of flow accident to verify the passive decay heat removal by natural circulation using RETRAN-03 code. auxiliary feed water systems are used for decay heat removal mechanism and results are compared with the loss of flow accident analysis using emergency decay heat removal system by JAERI. Results are very similar to case of EDRS 1 loop operation in JAERI analysis and decay heat is successfully removed by natural circulation

  17. Natural circulation of integrated-type marine reactor at inclined attitude

    International Nuclear Information System (INIS)

    Iyori, Isao; Aya, Izuo; Murata, Hiroyuki; Kobayashi, Michiyuki; Nariai, Hideki

    1987-01-01

    A steady-state single-phase natural circulation test was performed to clarify the effect of inclination by using a model of an integrated-type marine reactor. It was found that several types of flow pattern occur in the natural circulation loop corresponding to the range of inclination angle. Stable flow rates are sustained up to near 90 0 because of the occurrence of a driving force arising from those sections of the facility which were horizontal before the inclination. It was found that the temperature distribution in the steam generator at inclined attitude depends essentially only on the elevation z. The applicability of a one-dimensional analytical model was examined. It was clarified that employment of detailed U-turn flow paths, their correlation, and temperature-distribution function of core is essential for improvement. (orig.)

  18. Frequency analysis for the thermal hydraulic characterization of a natural circulation circuit

    International Nuclear Information System (INIS)

    Torres, Walmir M.; Macedo, Luiz A.; Sabundjian, Gaiane; Andrade, Delvonei A.; Umbehaun, Pedro E.; Conti, Thadeu N.; Mesquita, Roberto N.; Masotti, Paulo H.; Angelo, Gabriel

    2011-01-01

    This paper presents the frequency analysis studies of the pressure signals from an experimental natural circulation circuit during a heating process. The main objective is to identify the characteristic frequencies of this process using fast Fourier transform. Video images are used to associate these frequencies to the observed phenomenology in the circuit during the process. Sub-cooled and saturated flow boiling, heaters vibrations, overall circuit vibrations, chugging and geysering were observed. Each phenomenon has its specific frequency associated. Some phenomena and their frequencies must be avoided or attenuated since they can cause damages to the natural circulation circuit and its components. Special operation procedures and devices can be developed to avoid these undesirable frequencies. (author)

  19. Frequency analysis for the thermal hydraulic characterization of a natural circulation circuit

    Energy Technology Data Exchange (ETDEWEB)

    Torres, Walmir M.; Macedo, Luiz A.; Sabundjian, Gaiane; Andrade, Delvonei A.; Umbehaun, Pedro E.; Conti, Thadeu N.; Mesquita, Roberto N.; Masotti, Paulo H.; Angelo, Gabriel, E-mail: wmtorres@ipen.b, E-mail: lamacedo@ipen.b, E-mail: gdjian@ipen.b, E-mail: delvonei@ipen.b, E-mail: umbehaun@ipen.b, E-mail: tnconti@ipen.b, E-mail: , E-mail: rnavarro@ipen.b, E-mail: pmasotti@ipen.b, E-mail: gabriel.angelo@usp.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    This paper presents the frequency analysis studies of the pressure signals from an experimental natural circulation circuit during a heating process. The main objective is to identify the characteristic frequencies of this process using fast Fourier transform. Video images are used to associate these frequencies to the observed phenomenology in the circuit during the process. Sub-cooled and saturated flow boiling, heaters vibrations, overall circuit vibrations, chugging and geysering were observed. Each phenomenon has its specific frequency associated. Some phenomena and their frequencies must be avoided or attenuated since they can cause damages to the natural circulation circuit and its components. Special operation procedures and devices can be developed to avoid these undesirable frequencies. (author)

  20. Developmental assessment of RELAP5/MOD3 using the semiscale natural circulation tests

    International Nuclear Information System (INIS)

    Carlson, K.E.

    1990-01-01

    A code development effort creating RELAP5/MOD3 from RELAP5/MOD2 has been completed. Upon completion, a developmental assessment task was performed. One of the problems used for the developmental assessment was the Semiscale Natural Circulation Test. Calculated results from RELAP5/MOD3 are compared to measured data and previously calculated results from RELAP5/MOD2. 10 refs., 6 figs., 1 tab

  1. Analyses of natural circulation during a Surry station blackout using SCDAP/RELAP5

    International Nuclear Information System (INIS)

    Bayless, P.D.

    1988-10-01

    The effects of reactor coolant system natural circulation on the response of the Surry nuclear power plant during a station blackout transient were investigated. A TMLB' sequence (loss of all ac power, immediate loss of auxillary feedwater) was simulated from transient initiation until after fuel rod relocation had begun. Integral analyses of the system thermal-hydraulics and the core damage behavior were performed using the SCDAP/RELAP5 computer code and several different models of the plant. Three scoping calculations were performed in which the complexity of the plant model was progressively increased to determine the overall effects of in-vessel and hot leg natural circulation flows on the plant response. The natural circulation flows extended the transient, slowing the core heatup and delaying core damage by transferring energy from the core to structures in the upper plenum and coolant loops. Increased temperatures in the ex-core structures indicated that they may fail, however. Nine sensitivity calculations were then performed to investigate the effects of modeling uncertainties on the multidimensional natural circulation flows and the system response. Creep rupture failure of the pressurizer surge line was predicted to occur in eight of the calculations, with the hot leg failing in the ninth. The failure time was fairly insensitive to the parameters varied. The failures occurred near the time that fuel rod relocation began, well before failure of the reactor vessel would be expected. A calculation was also performed in which creep rupture failure of the surge line was modeled. The subsequent blowdown led to rapid accumulator injection and quenching of the entire core. 18 refs., 105 figs., 17 tabs

  2. Experimental and theoretical studies in Molten Salt Natural Circulation Loop (MSNCL)

    International Nuclear Information System (INIS)

    Srivastava, A.K.; Borgohain, A.; Jana, S.S.; Bagul, R.K.; Singh, R.R.; Maheshwari, N.K.; Belokar, D.G.; Vijayan, P.K.

    2014-12-01

    High Temperature Reactors (HTR) and solar thermal power plants use molten salt as a coolant, as it has low melting point and high boiling point, enabling us to operate the system at low pressure. Molten fluoride salt and molten nitrate salt are proposed as a candidate coolant for High Temperature Reactors (HTR) and solar power plant respectively. BARC is developing a 600 MWth pebble bed high temperature reactor, cooled by natural circulation of fluoride salt and capable of supplying process heat at 1000°C to facilitate hydrogen production by splitting water. Beside this, BARC is also developing a 2MWe solar power tower system using molten nitrate salt. With these requirements, a Molten Salt Natural Circulation Loop (MSNCL) has been designed, fabricated, installed and commissioned in Hall-7, BARC for thermal hydraulic, instrumentation development and material compatibility related studies. Steady state natural circulation experiments with molten nitrate salt (mixture of NaNO 3 and KNO 3 in 60:40 ratio) have been carried out in the loop at different power level. Various transients viz. startup of natural circulation, step power change, loss of heat sink and heater trip has also been studied in the loop. A well known steady state correlation given by Vijayan et. al. has been compared with experimental data. In-house developed code LeBENC has also been validated against all steady state and transient experimental results. The detailed description of MSNCL, steady state and transient experimental results and validation of in-house developed code LeBENC have been described in this report. (author)

  3. Laser anemometry measurements of natural circulation flow in a scale model PWR system

    International Nuclear Information System (INIS)

    Kadambi, J.R.; Schneider, S.J.

    1990-01-01

    This paper reports on experimental studies conducted to investigate the natural circulation of a single-phase fluid in a scale model pressurized water reactor system during a postulated degraded core accident. A half-section of a 1/7 scale model with a plexiglass adiabatic window was used. Water and sulfurhexafluoride (SF 6 ) were used as the fluid. Laser-Doppler anemometry (LDA) was used in marking the velocity measurements along the center plane of the model at five elevations

  4. An experimental investigation of natural circulated air flow in the passive containment cooling system

    International Nuclear Information System (INIS)

    Ryu, S.H.; Oh, S.M.; Park, G.C.

    2004-01-01

    The objective of this study is to investigate the effects of air inlet position and external conditions on the natural circulated air flow rate in a passive containment cooling system of the advanced passive reactor. Experiments have been performed with 1/36 scaled segment type passive containment test facility. The air velocities and temperatures are measured through the air flow path. Also, the experimental results are compared with numerical calculations and show good agreement. (author)

  5. Stability analysis for single-phase liquid metal rectangular natural circulation loops

    International Nuclear Information System (INIS)

    Lu, Daogang; Zhang, Xun; Guo, Chao

    2014-01-01

    Highlights: • The stability for asymmetric liquid metal natural circulation loops is analyzed. • The Na and NaK loops have higher critical Reynolds number than Pb and LBE loops. • Decreasing the ratio of height to width of loop can increase loop stability. • The length of heater would not affect the loop stability obviously. • Adding the length or heat transfer coefficient of cooler can increase loop stability. - Abstract: Natural circulation systems are preferred in some advanced nuclear power plants as they can simplify the designs and improve the inherent safety. The stability and steady-state characteristics of natural circulation are important for the applications of natural circulation loops (NCLs). A linear stability analysis method was used to study the stability behavior of liquid metal NCLs. The influences of the types of working fluids and loop geometry parameters on the stability of NCLs were evaluated. The liquid sodium (Na) loop and sodium–potassium alloy (NaK) loop would be more stable than lead bismuth eutectics (LBE) loop. The pressure drop could stabilize the loop behavior and also lead an increase of operating temperature for the loop. The NCL with a lower aspect ratio (ratio of vertical center distance between the heating and cooling section to the horizontal length of loop) is supposed to be more stable. It was found that the length of heating section would not have an obvious effect on the stability of NCL. However, the loop behavior could be stabilized by adding the length or heat transfer coefficient of the cooling section

  6. Preliminary Study of Single-Phase Natural Circulation for Lab-scaled Molten Salt Application

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Yukyung; Kang, Sarah; Kim, In Guk; Seo, Seok Bin; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of); Park, Seong Dae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Advanced reactors such as MSR (FHR), VHTR and AHTR utilized molten salt as a coolant for efficiency and safety which has advantages in higher heat capacity, lower pumping power and scale compared to liquid metal. It becomes more necessary to study on the characteristics of molten salt. However, due to several characteristics such as high operating temperature, large-scale facility and preventing solidification, satisfying that condition for study has difficulties. Thus simulant fluid was used with scaling method for lab-scale experiment. Scaled experiment enables simulant fluid to simulate fluid mechanics and heat transfer behavior of molten salt on lower operating temperature and reduced scale. In this paper, as a proof test of the scaled experiment, simplified single-phase natural circulation loop was designed in a lab-scale and applied to the passive safety system in advanced reactor in which molten salt is considered as a major coolant of the system. For the application of the improved safety system, prototype was based on the primary loop of the test-scale DRACS, the main passive safety system in FHR, developed at the OSU. For preliminary experiment, single-phase natural circulation under low power was performed. DOWTHERM A and DOWTHERM RP were selected as simulant candidates. Then, study of feasibility with simulant was conducted based on the scaling law for heat transfer characteristics and geometric parameters. Additionally, simulation with MARS code and ANSYS-CFX with the same condition of natural circulation was carried out as verification. For the accurate code simulation, thermo-physical properties of DOWTHERM A and RP were developed and implemented into MARS code. In this study, single-phase natural circulation experiment was performed with simulant oil, DOWTHERM RP, based on the passive safety system of FHR. Feasibility of similarity experiment for molten salt with oil simulant was confirmed by scaling method. In addition, simulation with two

  7. Development of the APR1400 model for countercurrent natural circulation in hot leg and steam generator under station blackout

    International Nuclear Information System (INIS)

    Park, Sang Gil; Kim, Han Chul

    2012-01-01

    In order to analyze severe accident phenomena, Korea Institute of Nuclear Safety (KINS) made a MELCOR model for APR1400 to examine natural circulation and creep rupture failure in the Reactor Coolant System (RCS) under station blackout (SBO). In this study, we are trying to advance the former model to describe natural circulation more accurately. After Fukushima accident, the concerns of severe accident management, assuring the heat removal capability, has risen for the case when the SBO is happened and there are no more electric powers to cool down decay heat. Under SBO there are three kinds of natural circulation which can delay the core heatup. One is in vessel natural circulation in the upper plenum of reactor vessel and the second is countercurrent natural circulation in hot leg through steam generator tubes and the last is full loop natural circulation when the reactor coolant pump loop seal is cleared and reactor coolant pump sealing is damaged by high temperature and high pressure. Among them this study focuses on the countercurrent natural circulation model using MELCOR1.8.6

  8. Development of essential system technologies for advanced reactor - Development of natural circulation analysis code for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Goon Cherl; Park, Ik Gyu; Kim, Jae Hak; Lee, Sang Min; Kim, Tae Wan [Seoul National University, Seoul (Korea)

    1999-04-01

    The objective of this study is to understand the natural circulation characteristics of integral type reactors and to develope the natural circulation analysis code for integral type reactors. This study is focused on the asymmetric 3-dimensional flow during natural circulation such as 1/4 steam generator section isolation and the inclination of the reactor systems. Natural circulation experiments were done using small-scale facilities of integral reactor SMART (System-Integrated Modular Advanced ReacTor). CFX4 code was used to investigate the flow patterns and thermal mixing phenomena in upper pressure header and downcomer. Differences between normal operation of all steam generators and the 1/4 section isolation conditions were observed and the results were used as the data 1/4 section isolation conditions were observed and the results were used as the data for RETRAN-03/INT code validation. RETRAN-03 code was modified for the development of natural circulation analysis code for integral type reactors, which was development of natural circulation analysis code for integral type reactors, which was named as RETRAN-03/INT. 3-dimensional analysis models for asymmetric flow in integral type reactors were developed using vector momentum equations in RETRAN-03. Analysis results using RETRAN-03/INT were compared with experimental and CFX4 analysis results and showed good agreements. The natural circulation characteristics obtained in this study will provide the important and fundamental design features for the future small and medium integral reactors. (author). 29 refs., 75 figs., 18 tabs.

  9. Two-phase ozonation of chlorinated organics

    International Nuclear Information System (INIS)

    Bhattacharyya, D.; Freshour, A.; West, D.

    1995-01-01

    In the last few years the amount of research being conducted in the field of single-phase ozonation has grown extensively. However, traditional aqueous-phase ozonation systems are limited by a lack of selective oxidation potential, low ozone solubility in water, and slow intermediate decomposition rates. Furthermore, ozone may decompose before it can be utilized for pollutant destruction since ozone can be highly unstable in aqueous solutions. Naturally occurring compounds such as NaHCO 3 also affect ozone reactions by inhibiting the formation of OH-free radicals. To compensate for these factors, excess ozone is typically supplied to a reactor. Since ozone generation requires considerable electric power consumption (16 - 24 kWh/kg of O 3 ), attempts to enhance the ozone utilization rate and stability should lead to more efficient application of this process to hazardous waste treatment. To improve the process, ozonation may be more efficiently carried out in a two-phase system consisting of an inert solvent (saturated with O 3 ) contacted with an aqueous phase containing pollutants. The non-aqueous phase must meet the following criteria: (1) non-toxic, (2) very low vapor pressure, (3) high density (for ease of separation), (4) complete insolubility in water, (5) reusability, (6) selective pollutant extractability, (7) high oxidant solubility, and (8) extended O 3 stability. Previously published studies (1) have indicated that a number of fluorinated hydrocarbon compounds fit these criteria. For this project, FC40 (a product of 3M Co.) was chosen due to its low vapor pressure (3 mm Hg) and high specific gravity (1.9). The primary advantages of the FC40 solvent are that it is non-toxic, reusable, has an ozone solubility 10 times that of water, and that 85 % of the ozone remains in the solvent even after 2 hours. This novel two-phase process has been utilized to study the rapid destruction of organic chlorine compounds and organic mixtures

  10. Parametric study of natural circulation flow in molten salt fuel in molten salt reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pauzi, Anas Muhamad, E-mail: Anas@uniten.edu.my [Centre of Nuclear Energy, Universiti Tenaga Nasional (UNITEN), Jalan IKRAM-UNITEN, 43000 Kajang, Selangor (Malaysia); Cioncolini, Andrea; Iacovides, Hector [School of Mechanical, Aerospace, and Civil Engineering (MACE), University of Manchester, Oxford Road, M13 9PL Manchester (United Kingdom)

    2015-04-29

    The Molten Salt Reactor (MSR) is one of the most promising system proposed by Generation IV Forum (GIF) for future nuclear reactor systems. Advantages of the MSR are significantly larger compared to other reactor system, and is mainly achieved from its liquid nature of fuel and coolant. Further improvement to this system, which is a natural circulating molten fuel salt inside its tube in the reactor core is proposed, to achieve advantages of reducing and simplifying the MSR design proposed by GIF. Thermal hydraulic analysis on the proposed system was completed using a commercial computation fluid dynamics (CFD) software called FLUENT by ANSYS Inc. An understanding on theory behind this unique natural circulation flow inside the tube caused by fission heat generated in molten fuel salt and tube cooling was briefly introduced. Currently, no commercial CFD software could perfectly simulate natural circulation flow, hence, modeling this flow problem in FLUENT is introduced and analyzed to obtain best simulation results. Results obtained demonstrate the existence of periodical transient nature of flow problem, hence improvements in tube design is proposed based on the analysis on temperature and velocity profile. Results show that the proposed system could operate at up to 750MW core power, given that turbulence are enhanced throughout flow region, and precise molten fuel salt physical properties could be defined. At the request of the authors and the Proceedings Editor the name of the co-author Andrea Cioncolini was corrected from Andrea Coincolini. The same name correction was made in the Acknowledgement section on page 030004-10 and in reference number 4. The updated article was published on 11 May 2015.

  11. Reliability analysis of 2400 MWth gas-cooled fast reactor natural circulation decay heat removal system

    International Nuclear Information System (INIS)

    Marques, M.; Bassi, C.; Bentivoglio, F.

    2012-01-01

    In support to a PSA (Probability Safety Assessment) performed at the design level on the 2400 MWth Gas-cooled Fast Reactor, the functional reliability of the decay heat removal system (DHR) working in natural circulation has been estimated in two transient situations corresponding to an 'aggravated' Loss of Flow Accident (LOFA) and a Loss of Coolant Accident (LOCA). The reliability analysis was based on the RMPS methodology. Reliability and global sensitivity analyses use uncertainty propagation by Monte Carlo techniques. The DHR system consists of 1) 3 dedicated DHR loops: the choice of 3 loops (3*100% redundancy) is made in assuming that one could be lost due to the accident initiating event (break for example) and that another one must be supposed unavailable (single failure criterion); 2) a metallic guard containment enclosing the primary system (referred as close containment), not pressurized in normal operation, having a free volume such as the fast primary helium expansion gives an equilibrium pressure of 1.0 MPa, in the first part of the transient (few hours). Each dedicated DHR loop designed to work in forced circulation with blowers or in natural circulation, is composed of 1) a primary loop (cross-duct connected to the core vessel), with a driving height of 10 meters between core and DHX mid-plan; 2) a secondary circuit filled with pressurized water at 1.0 MPa (driving height of 5 meters for natural circulation DHR); 3) a ternary pool, initially at 50 C. degrees, whose volume is determined to handle one day heat extraction (after this time delay, additional measures are foreseen to fill up the pool). The results obtained on the reliability of the DHR system and on the most important input parameters are very different from one scenario to the other showing the necessity for the PSA to perform specific reliability analysis of the passive system for each considered scenario. The analysis shows that the DHR system working in natural circulation is

  12. Two-phase flow characteristics in BWRs

    International Nuclear Information System (INIS)

    Katono, Kenichi; Aoyama, Goro; Nagayoshi, Takuji; Yasuda, Kenichi; Nishida, Koji

    2014-01-01

    Reliable prediction of two-phase flow characteristics is important for safety and economy improvements of BWR plants. We have been developing two-phase flow measurement tools and techniques for BWR thermal hydraulic conditions, such as a 3D time-averaged X-ray CT system, an ultrasonic liquid film sensor and a wire-mesh sensor. We applied the developed items in experiments using the multi-purpose steam-water test facility known as HUSTLE, which can simulate two-phase thermal-hydraulic conditions in a BWR reactor pressure vessel, and we constructed a detailed instrumentation database. We validated a 3D two-phase flow simulator using the database and developed the reactor internal two-phase flow analysis system. (author)

  13. Thermal and hydrodynamic characteristics of supercritical CO2 natural circulation in closed loops

    International Nuclear Information System (INIS)

    Chen, Lin; Deng, Bi-Li; Jiang, Bin; Zhang, Xin-Rong

    2013-01-01

    Highlights: ► We model thermosyphon heat transfer and stability with super-/trans-critical turbulence model incorporated. ► Potentials of super-/trans-critical CO 2 thermosyphon are confirmed. ► Three characteristics found: flow instability; high flow rate with density wave; heat transfer discrepancies. ► Major laws of system stability factors are different compared with traditional fluids. ► Traditional thermosyphon flow correlation has its limitations and deserves further development. -- Abstract: Natural convective flow of supercritical fluids has become a hot topic in engineering applications. Natural circulation thermosyphon using supercritical/trans-critical CO 2 can be a potential choice for effectively transportation of heat and mass without pumping devices. This paper presents a series of numerical investigations into the fundamental features in a supercritical/trans-critical CO 2 based natural circulation loop. New heat transport model aiming at trans-critical thermosyphon heat transfer and stability is proposed with supercritical/trans-critical turbulence model incorporated. In this study, the fundamentals include the basic flow and heat transfer behavior of the above loop, the effect of heat source temperature on system stability, the effect of loop diameter on natural convection supercritical CO 2 loop and its coupling effect with heat source temperature and the effect of constant changing heat input condition and system behavior evolution during unsteady input or failure conditions. The fundamental potentials of supercritical/trans-critical CO 2 based natural convection system are confirmed. Basic supercritical CO 2 closed loop flow and heat transfer behaviors are clarified. During this study, the CO 2 loop stability map are also put forward and introduced as an important feature of supercritical CO 2 system. Stability factors of natural convective trans-critical CO 2 flow and its implications on real system control are also discussed in

  14. Oxygen concentration diffusion analysis of lead-bismuth-cooled, natural-circulation reactor

    International Nuclear Information System (INIS)

    Ito, Kei; Sakai, Takaaki

    2001-11-01

    The feasibility study on fast breeder reactors in Japan has been conducted at JNC and related organizations. The Phase-I study has finished in March, 2001. During the Phase-I activity, lead-bismuth eutectic coolant has been selected as one of the possible coolant options and a medium-scale plant, cooled by a lead-bismuth natural circulation flow was studied. On the other side, it is known that lead-bismuth eutectic has a problem of structural material corrosiveness. It was found that oxygen concentration control in the eutectic plays an important role on the corrosion protection. In this report, we have developed a concentration diffusion analysis code (COCOA: COncentration COntrol Analysis code) in order to carry out the oxygen concentration control analysis. This code solves a two-dimensional concentration diffusion equation by the finite differential method. It is possible to simulate reaction of oxygen and hydrogen by the code. We verified the basic performance of the code and carried out oxygen concentration diffusion analysis for the case of an oxygen increase by a refueling process in the natural circulation reactor. In addition, characteristics of the oxygen control system was discussed for a different type of the control system as well. It is concluded that the COCOA code can simulate diffusion of oxygen concentration in the reactor. By the analysis of a natural circulation medium-scale reactor, we make clear that the ON-OFF control and PID control can well control oxygen concentration by choosing an appropriate concentration measurement point. In addition, even when a trouble occurs in the oxygen emission or hydrogen emission system, it observes that control characteristic drops away. It is still possible, however, to control oxygen concentration in such case. (author)

  15. Studies on natural circulation cooling enhancement in a spent fuel in fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Maekawa, Isamu; Akamatsu, Mikio; Toda, Shinichi; Sato, Manabu [Kawasaki Heavy Industries Ltd., Kobe (Japan); Mayumi, Masami

    2001-01-01

    Fast breeder reactor (FBR) has some advantages such as effective application of plutonium, excellent capacity to fire minor-actinides (longer half-life nuclides such as Np, Am, Cm, and so on) contained in radioactive wastes in the reactor to convert their shorter half-life nuclides. However, fuels containing the minor-actinides have a characteristic with higher exotherm and radioactive intensity than those of conventional ones, it is essential at their actual stages to prepare some rational fuel handling systems on their transportation, storage and so forth. In addition, there are few examples on natural circulation heat transfer test of a liquid metal using long sized container. Then, in order to establish an evaluating method on decay-heat removing property of a spent fuel assembly in sodium canister and pot, some natural circulation tests on a long sized container including a quasi pin-bundle structure for a working fluid of lead-bismuth (Pb-Bi) mixture with easier handling than that of sodium was carried out. A specimen could be mounted at optional angles from horizontal to vertical positions so as to evaluate effects of inclined angles. In addition, in order to estimate temperature and flow rate distribution in a long sized container and understand thermal flowing phenomenon in specimen system, numerical analysis using multi-dimensional analysis code was carried out. As a result, it was found that in vertical arrangement system, natural circulation phenomenon is limited at upper portion of the exothermal portion, and its maximum temperature was tested at central portion of top pin-bundle of the exothermal portion. And, it was also found that at horizontal arrangement maximum temperature was 40 centigrade less than that of vertical arrangement, and so forth. (G.K.)

  16. Experimental Observations of Natural Circulation Flow in the NSTF at Steady-State Conditions

    International Nuclear Information System (INIS)

    Lisowski, Darius D.; Farmer, Mitch T.

    2014-01-01

    A ½ scale test facility has been constructed at Argonne National Laboratory (ANL) to study the heat removal performance and natural circulation flow patterns in a Reactor Cavity Cooling System (RCCS). Our test facility, the Natural convection Shutdown heat removal Test Facility (NSTF), supports the broader goal of developing an inherently safe and fully passive ex-vessel decay heat removal for advanced reactor designs. The project, initiated in 2010 to support the Advanced Reactor Concepts (ARC), Small Modular Reactor (SMR), and Next Generation Nuclear Plant (NGNP) programs, has been conducting experimental operations since early 2014. The following paper provides a summary of some primary design features of the 26-m tall test facility along with a description of the data acquisition suite that guides our experimental practices. Specifics of the distributed fiber optic temperature measurements will be discussed, which introduces an unparalleled level of data density that has never before been implemented in a large scale natural circulation test facility. Results from our first test series will then be presented, which provide insight into the thermal hydraulic behavior at steady-state conditions for varying heat flux levels and exhaust chimney configuration states. (author)

  17. Changes in equatorial zonal circulations and precipitation in the context of the global warming and natural modes

    Science.gov (United States)

    Kim, B. H.; Ha, K. J.

    2017-12-01

    The strengthening and westward shift of Pacific Walker Circulation (PWC) is observed during the recent decades. However, the relative roles of global warming and natural variability on the change in PWC unclearly remain. By conducting numerical atmospheric general circulation model (AGCM) experiments using the spatial SST patterns in the global warming and natural modes which are obtained by the multi-variate EOF analysis from three variables including precipitation, sea surface temperature (SST), and divergent zonal wind, we indicated that the westward shift and strengthening of PWC are caused by the global warming SST pattern in the global warming mode and the negative Interdecadal Pacific Oscillation-like SST pattern in the natural mode. The SST distribution of the Pacific Ocean (PO) has more influence on the changes in equatorial zonal circulations and tropical precipitation than that of the Indian Ocean (IO) and Atlantic Ocean (AO). The change in precipitation is also related to the equatorial zonal circulations variation through the upward and downward motions of the circulations. The IO and AO SST anomalies in the global warming mode can affect on the changes in equatorial zonal circulations, but the influence of PO SST disturbs the Indian Walker circulation and Atlantic Walker circulation changes by the IO and AO. The zonal shift of PWC is found to be highly associated with a zonal gradient of SST over the PO through the idealized numerical AGCM experiments and predictions of CMIP5 models.

  18. 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

  19. Prediction about chaotic times series of natural circulation flow under rolling motion

    International Nuclear Information System (INIS)

    Yuan Can; Cai Qi; Guo Li; Yan Feng

    2014-01-01

    The paper have proposed a chaotic time series prediction model, which combined phase space reconstruction with support vector machines. The model has been used to predict the coolant volume flow, in which a synchronous parameter optimization method was brought up based on particle swarm optimization algorithm, since the numerical value selection of related parameter was a key factor for the prediction precision. The average relative error of prediction values and actual observation values was l,5% and relative precision was 0.9879. The result indicated that the model could apply for the natural circulation coolant volume flow prediction under rolling motion condition with high accuracy and robustness. (authors)

  20. Fluid-to-fluid scaling for a gravity- and flashing-driven natural circulation loop

    International Nuclear Information System (INIS)

    Yadigaroglu, G.; Zeller, M.

    1994-01-01

    In certain natural-circulation reactor systems proposed recently, vapor generation takes place by flashing in an adiabatic riser above the core. A step-by-step facility design procedure was used to define suitable scaling criteria for a refrigerant-113 (R-113) experiment simulating the dynamics and stability of such a loop. The fact that vapor generation does not normally take place in the core allows additional flexibility in designing the model; almost perfect simulation can be achieved, mainly by reducing the height of the facility according to the liquid density ratio and scaling for similar void fraction distributions in the prototype and the model. ((orig.))

  1. Effect of marine condition on feature of natural circulation after accident in floating nuclear power plant

    International Nuclear Information System (INIS)

    Yang Fan; Zhang Dan; Tan Changlu; Ran Xu; Yu Hongxing

    2015-01-01

    The incline and swing effect on natural circulation of floating nuclear power plant under site black out (SBO) accident is studied using self-developing marine condition system code RELAP5/MC. It shows that, for floating nuclear power plant under marine condition, the pressurizer fluctuating flow rate, the parallel heat sink (steam generator) have significant influences on the direct passive reactor heat removal (PRHR) system, which is different from other secondary PRHR under marine condition. The flow exchange between the loop and the pressurizer have major effect on cooling capacity for the left side loop. (authors)

  2. 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

  3. Design and measured performance of a solar chimney for natural-circulation solar-energy dryers

    International Nuclear Information System (INIS)

    Ekechukwu, O.V.; Norton, B.

    1995-10-01

    The design and construction of a solar chimney which was undertaken as part of a study on natural-circulation solar-energy dryers is reported. The experimental solar chimney consists of a 5.3m high and 1.64m diameter cylindrical polyethylene-clad vertical chamber, supported structurally by steel framework and draped internally with a selectively-absorbing surface. The performance of the chimney which was monitored extensively with and without the selective surface in place (to study the effectiveness of this design option) is also reported. (author). 14 refs, 7 figs

  4. Fluid-elastic vibration in two-phase cross flow

    International Nuclear Information System (INIS)

    Sasakawa, T.; Serizawa, A.; Kawara, Z.

    2003-01-01

    The present work aims at clarifying the mechanisms of fluid elastic vibration of tube bundles in two-phase cross flow. The experiment is conducted using air-water two-phase flow under atmospheric pressure. The test section is a 1.03m long transparent acrylic square duct with 128 x 128 mm 2 cross section, which consists of 3 rod-rows with 5 rods in each row. The rods are 125mm long aluminum rods with 22 mm in diameter (p/D=1.45). The natural frequency of rod vibration is about 30Hz. The result indicated a diversion of observed trend in vibration behavior depending on two-phase flow patterns either bubbly flow or churn flow. Specifically, in churn flow, the fluid elastic vibration has been observed to occur when the frequency in void fraction fluctuation approached to the natural frequency of the rods, but this was not the case in fluid elastic vibration in bubbly flow. This fact suggests the existence of mechanisms closely coupled with two-phase flow structures depending on the flow patterns, that is, static two-phase character-controlled mechanism in bubbly flow and dynamic character- controlled in churn flow

  5. Two-phase flow in refrigeration systems

    CERN Document Server

    Gu, Junjie; Gan, Zhongxue

    2013-01-01

    Two-Phase Flow in Refrigeration Systems presents recent developments from the authors' extensive research programs on two-phase flow in refrigeration systems. This book covers advanced mass and heat transfer and vapor compression refrigeration systems and shows how the performance of an automotive air-conditioning system is affected through results obtained experimentally and theoretically, specifically with consideration of two-phase flow and oil concentration. The book is ideal for university postgraduate students as a textbook, researchers and professors as an academic reference book, and b

  6. Characteristics of thermal hydraulic stability in a HYPER system with enhanced natural circulation potential

    International Nuclear Information System (INIS)

    Tak, Nam Il; Park, Won S.; Han, Seok Jung

    1999-06-01

    Pb-Bi eutectic chosen as a coolant of HYPER is an excellent heat transfer medium but requires relatively large pumping power. Thus the mixed cooling concept to increase economy and safety is being considered for HYPER. In this cooling concept, a large fraction of total thermal power is carried by natural circulation. However, the mixed cooling concept has been considered for conceptual designs only an it has never been applied to real reactors. The purpose of the present study is to provide simple tools to analyze mixed flow and to examine fundamental stability characteristics of mixed flow. Conventional one-dimensional approaches using mass, momentum, and energy conservation are used to describe a forced circulating flow affected by a large buoyancy force. The results of simple analysis using preliminary design parameters of HYPER show that cooling by mixed flow is possible only when the total pressure loss of system is sufficiently low. The stability behavior of mixed flow in a simple rectangular loop has been studied using numerical solutions of the governing equations. As in the case of natural circulation, three types of flow regions, such as stable, neutrally stable, and unstable regions, were found. The stability map of mixed flow has been obtained using the results of calculations. Forced flow due to the pump is found to increase the stability of the loop, since the stable portion of the stability map is increased. However, the unstable region of the mixed flow does not completely disappear, even though the pump exists. (author). 37 refs., 4 tabs., 23 figs

  7. Unsteady single-phase natural circulation flow mixing prediction using CATHARE three-dimensional capabilities

    Energy Technology Data Exchange (ETDEWEB)

    Salah, Anis Bousbia; Vlassenbroeck, Jacques [Bel V - Subsidiary of the Belgian Federal Agency for Nuclear Contro, Brussels (Belize)

    2017-04-15

    Coolant mixing under natural circulation flow regime constitutes a key parameter that may play a role in the course of an accidental transient in a nuclear pressurized water reactor. This issue has motivated some experimental investigations carried out within the OECD/NEA PKL projects. The aim was to assess the coolant mixing phenomenon in the reactor pressure vessel downcomer and the core lower plenum under several asymmetric steady and unsteady flow conditions, and to provide experimental data for code validations. Former studies addressed the mixing phenomenon using, on the one hand, one-dimensional computational approaches with cross flows that are not fully validated under transient conditions and, on the other hand, expensive computational fluid dynamic tools that are not always justified for large-scale macroscopic phenomena. In the current framework, an unsteady coolant mixing experiment carried out in the Rossendorf coolant mixing test facility is simulated using the three-dimensional porous media capabilities of the thermal–hydraulic system CATHARE code. The current study allows highlighting the current capabilities of these codes and their suitability for reproducing the main phenomena occurring during asymmetric transient natural circulation mixing conditions.

  8. Impact of rapid condensations of large vapor spaces on natural circulation in integral systems

    International Nuclear Information System (INIS)

    Wang, Z.; Almenas, K.; DiMarzo, M.; Hsu, Y.Y.; Unal, C.

    1992-01-01

    In this study we demonstrated that the Interruption-Resumption flow mode (IRM) observed in the University of Maryland 2x4 loop is a unique and effective natural circulation cooling mode. The IRM flow mode consists of a series of large flow cycles which are initiated from a quiescent steady-state flow condition by periodic rapid condensation of large vapor spaces. The significance of this mass/energy transport mechanism is that it cannot be evaluated using the techniques developed for the commonly known density-driven natural circulation cooling mode. We also demonstrated that the rapid condensation mechanism essentially acts as a strong amplifier which will augment small perturbations and will activate several flow phenomena. The interplay of the phenomena involves a degree of randomness. This poses two important implications. First, the study of an isolated flow phenomenon is not sufficient for the understanding of the system-wide IRM fluid movement. Second, the duplication of reactor transients which involves randomness can be achieved only within certain bounds. The modeling of such transients by deterministic computer codes requires recognition of this physical reality. (orig.)

  9. Steady state and linear stability analysis of a supercritical water natural circulation loop

    International Nuclear Information System (INIS)

    Sharma, Manish; Pilkhwal, D.S.; Vijayan, P.K.; Saha, D.; Sinha, R.K.

    2010-01-01

    Supercritical water (SCW) has excellent heat transfer characteristics as a coolant for nuclear reactors. Besides it results in high thermal efficiency of the plant. However, the flow can experience instabilities in supercritical water reactors, as the density change is very large for the supercritical fluids. A computer code SUCLIN using supercritical water properties has been developed to carry out the steady state and linear stability analysis of a SCW natural circulation loop. The conservation equations of mass, momentum and energy have been linearized by imposing small perturbation in flow rate, enthalpy, pressure and specific volume. The equations have been solved analytically to generate the characteristic equation. The roots of the equation determine the stability of the system. The code has been qualitatively assessed with published results and has been extensively used for studying the effect of diameter, height, heater inlet temperature, pressure and local loss coefficients on steady state and stability behavior of a Supercritical Water Natural Circulation Loop (SCWNCL). The present paper describes the linear stability analysis model and the results obtained in detail.

  10. Dynamic model of YGN 3 and 4 steam generators for natural circulation mode

    International Nuclear Information System (INIS)

    Sohn, Jong Joo

    1995-02-01

    A dynamic model for the secondary side of Yonggwang nuclear power plant units 3 and 4 (YGN 3 and 4) steam generator model is developed to improve the accuracy of the present performance analysis code. The new model is based on the one-dimensional three region model to predict the local quality and void fraction distribution along the U-tube length. The local quality concept is used instead of the Wilson bubble rise correlation to simulate the steam generators in the natural circulation mode. The new model can be applicable to the plants in the power operation modes such as load maneuvering transients in which the steam generator internal flow is maintained in the natural circulation mode. To validate the new model, the code predictions are compared with the actual plant data measured for the selected load maneuvering tests performed during the YGN units 3 power ascension test period. The results from the improved model show better agreement with the plant data than those from the present code. Especially, the new model's capability of accurately simulating the steam generator water level behavior during the fast transients such as a large load rejection event is demonstrated

  11. Operating experience of natural circulation core cooling in boiling water reactors

    International Nuclear Information System (INIS)

    Kullberg, C.; Jones, K.; Heath, C.

    1993-01-01

    General Electric (GE) has proposed an advanced boiling water reactor, the Simplified Boiling Water Reactor (SBWR), which will utilize passive, gravity-driven safety systems for emergency core coolant injection. The SBWR design includes no recirculation loops or recirculation pumps. Therefore the SBWR will operate in a natural circulation (NC) mode at full power conditions. This design poses some concerns relative to stability during startup, shutdown, and at power conditions. As a consequence, the NRC has directed personnel at several national labs to help investigate SBWR stability issues. This paper will focus on some of the preliminary findings made at the INEL. Because of the broad range of stability issues this paper will mainly focus on potential geysering instabilities during startup. The two NC designs examined in detail are the US Humboldt Bay Unit 3 BWR-1 plant and Dodewaard plant in the Netherlands. The objective of this paper will be to review operating experience of these two plants and evaluate their relevance to planned SBWR operational procedures. For completeness, experimental work with early natural circulation GE test facilities will also be briefly discussed

  12. Temperature control characteristics analysis of lead-cooled fast reactor with natural circulation

    International Nuclear Information System (INIS)

    Yang, Minghan; Song, Yong; Wang, Jianye; Xu, Peng; Zhang, Guangyu

    2016-01-01

    Highlights: • The LFR temperature control system are analyzed with frequency domain method. • The temperature control compensator is designed according to the frequency analysis. • Dynamic simulation is performed by SIMULINK and RELAP5-HD. - Abstract: Lead-cooled Fast Reactor (LFR) with natural circulation in primary system is among the highlights in advance nuclear reactor research, due to its great superiority in reactor safety and reliability. In this work, a transfer function matrix describing coolant temperature dynamic process, obtained by Laplace transform of the one-dimensional system dynamic model is developed in order to investigate the temperature control characteristics of LFR. Based on the transfer function matrix, a close-loop coolant temperature control system without compensator is built. The frequency domain analysis indicates that the stability and steady-state of the temperature control system needs to be improved. Accordingly, a temperature compensator based on Proportion–Integration and feed-forward is designed. The dynamic simulation of the whole system with the temperature compensator for core power step change is performed with SIMULINK and RELAP5-HD. The result shows that the temperature compensator can provide superior coolant temperature control capabilities in LFR with natural circulation due to the efficiency of the frequency domain analysis method.

  13. Design of an additional heat sink based on natural circulation in pressurized water reactors

    International Nuclear Information System (INIS)

    Frischengruber, Kurt; Solanilla, Roberto; Fernandez, Ricardo; Blumenkrantz, Arnaldo; Castano, Jorge

    1989-01-01

    Residual heat removal through the steam generators in Nuclear Power Plant with pressurized water reactors (PWR) or pressurized heavy water reactors (PHWR in pressured vessel or pressured tube types) requires the maintenance of the steam generator inventory and the availability of and appropriate heat sink, which are based on the operability of the steam generators feedwater system. This paper describes the conceptual design of an assured heat removal system which includes only passive elements and is based on natural circulation. The system can supplement the original systems of the plant. The new system includes a condenser/boiler heat exchanger to condense the steam produced in the steam generator, transferring the heat to the water of an open pool at atmospheric pressure. The condensed steam flows back to the steam generators by natural circulation effects. The performance of an Atucha type PHWR nuclear power station with and without the proposed system is calculated in an emergency power case for the first 5000 seconds after the incident. The analysis shows that the proposed system offers the possibility to cool-down the plant to a low energy state during several hours and avoids the repeated actuation of the primary and secondary system safety valves. (Author) [es

  14. Study of flow instabilities in double-channel natural circulation boiling systems

    International Nuclear Information System (INIS)

    Durga Prasad, Gonella V.; Pandey, Manmohan; Pradhan, Santosh K.; Gupta, Satish K.

    2008-01-01

    Natural circulation boiling systems consisting of parallel channels can undergo different types of oscillations (in-phase or out-of-phase) depending on the geometric parameters and operating conditions. Disturbances in one channel affect the flow in other channels, which triggers thermal-hydraulic oscillations. In the present work, the modes of oscillation under different operating conditions and channel-to-channel interaction during power fluctuations and on-power refueling in a double-channel natural circulation boiling system are investigated. The system is modeled using a lumped parameter mathematical model and RELAP5/MOD3.4. Parametric studies are carried out for an equal-power double-channel system, at different operating conditions, with both the models, and the results are compared. Instabilities, non-linear oscillations, and effects of recirculation loop dynamics and geometric parameters on the mode of oscillations, are studied using the lumped model. The two channels oscillate out-of-phase in Type-I region, but in Type-II region, both the modes of oscillation are observed under different conditions. Channel-to-channel interaction and on-power refueling studies are carried out using the RELAP model. At high powers, disturbances in one channel significantly affect the stability of the other channel. During on-power refueling, a near-stagnation condition or low-velocity reverse flow can occur, the possibility of reverse flow being higher at lower pressures

  15. Thermal-hydraulics stability of natural circulation BWR under startup. Flashing effects

    International Nuclear Information System (INIS)

    Hu, Rui; Kazimi, Mujid S.

    2009-01-01

    To help achieve the necessary natural circulation flow, a fairly long chimney is installed in a boiling natural circulation reactor like the ESBWR. In such systems, thermal-hydraulic stability during low pressure start-up should be examined while considering the flashing induced by the pressure drop in the channel and the chimney due to gravity head. In this work, a BWR stability analysis code in the frequency domain, named FISTAB (Flashing-Induced STability Analysis for BWR), was developed to address the issue of flashing-induced instability. A thermal-hydraulics non-homogeneous equilibrium model (NHEM) based on a drift flux formulation along with a lumped fuel dynamics model is incorporated in the work. The vapor generation rate is derived from the mixture energy conservation equation while considering the effect of flashing. The functionality of the FISTAB code was confirmed by comparison to experimental results from SIRIUS-N facility at CRIEPI, Japan. Both stationary and perturbation results agree well with the experimental results. (author)

  16. Development of a butterfly check valve model under natural circulation conditions

    International Nuclear Information System (INIS)

    Rao, Yuxian; Yu, Lei; Fu, Shengwei; Zhang, Fan

    2015-01-01

    Highlights: • Bases on Lim’s swing check valve model, a butterfly check valve model was developed. • The method to quantify the friction torque T F in Li’s model was corrected. • The developed model was implemented into the RELAP5 code and verified. - Abstract: A butterfly check valve is widely used to prevent a reverse flow in the pipe lines of a marine nuclear power plant. Under some conditions, the natural circulation conditions in particular, the fluid velocity through the butterfly check valve might become too low to hold the valve disk fully open, thereby the flow resistance of the butterfly check valve varies with the location of the valve disk and as a result the fluid flow is significantly affected by the dynamic motion of the valve disk. Simulation of a pipe line that includes some butterfly check valves, especially under natural circulation conditions, is thus complicated. This paper focuses on the development of a butterfly check valve model to enhance the capability of the thermal–hydraulic system code and the developed model is implemented into the RELAP5 code. Both steady-state calculations and transient calculations were carried out for the primary loop system of a marine nuclear power plant and the calculation results are compared with the experimental data for verification purpose. The simulation results show an agreement with the experimental data

  17. Development of a transient calculation model for a closed sodium natural circulation loop

    International Nuclear Information System (INIS)

    Chang, Won Pyo; Ha, Kwi Seok; Jeong, Hae Yong; Heo, Sun; Lee, Yong Bum

    2003-09-01

    A natural circulation loop has usually adopted for a Liquid Metal Reactor (LMR) because of its high reliability. Up-rating of the current KALIMER capacity requires an additional PDRC to the existing PVCS to remove its decay heat under an accident. As the system analysis code currently used for LMR in Korea does not feature a stand alone capability to simulate a closed natural circulation loop, it is not eligible to be applied to PDRC. To supplement its limitation, a steady state calculation model had been developed during the first phase, and development of the transient model has successively carried out to close the present study. The developed model will then be coupled with the system analysis code, SSC-K to assess a long term cooling for the new conceptual design. The incompressibility assumption of sodium which allows the circuit to be modeled with a single loop flow, makes the model greatly simplified comparing with LWR. Some thermal-hydraulic models developed during this study can be effectively applied to other system analysis codes which require such component models, and the present development will also contribute to establishment of a code system for the LMR analysis

  18. Comparison of GAMMA results with experimental data in the naturally circulating gas loop

    International Nuclear Information System (INIS)

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

    2009-01-01

    Natural circulation steady-state data with carbon dioxide and nitrogen are compared to numerical predictions by GAMMA code, which is being developed by KAIST and KAERI. The GAMMA code is a computational tool for predicting various transients those can potentially occur in a high temperature gas cooled reactor. The code has a capability of analyzing multi-dimensional multi-component mixture and includes models for friction, heat transfer, chemical reaction and multi-component molecular diffusion. Natural circulation data with nitrogen and carbon dioxide gases show that the loop operates in a Deteriorated Turbulent Heat Transfer (DTHT) regime which exhibits substantially reduced heat transfer coefficients. The GAMMA code with original heat transfer package predicted conservative results in terms of peak wall temperature. Also, the estimated peak location did not successfully match with the data. Even though GAMMA's original heat transfer package included mixed-convection regime, which is part of DTHT regime, still the results showed that the original heat transfer package performs with insufficient accuracy. After implementing a recently developed correlation and corresponding heat transfer regime map into GAMMA to cover the whole range of deteriorated heat transfer, a better agreement with data was obtained. In addition, RELAP5-MOD3 results are discussed in parallel. (author)

  19. Experiments in a natural circulation loop with supercritical water at low powers

    International Nuclear Information System (INIS)

    Pilkhwal, D.S.; Sharma, Manish; Jana, S.S.; Vijayan, P.K.

    2013-05-01

    Earlier, 1/2 ″ uniform diameter Supercritical Pressure Natural Circulation Loop (SPNL) was set-up in hall-7, BARC for carrying out experiments related to supercritical fluids. The loop is a rectangular loop having two heaters and two coolers. Experiments were carried out with CO 2 under supercritical conditions for various pressures and different combinations of heater and cooler orientations. Since, the design conditions are more severe for supercritical water (SCW) experiments, the loop was modified for SCW by installing new test sections, pressurizer and power supply for operation with supercritical water. Experimental data were generated on steady state, heat transfer and stability under natural circulation conditions for the horizontal heater and horizontal cooler (HHHC) orientation with SCW up to a heater power of 8.5 kW. The flow rate data and instability data were compared with the predictions of in-house developed 1-D code NOLSTA, which showed reasonable agreement. The heat transfer coefficient data were also compared with the predictions of various correlations exhibit peak at bulk temperature lower than that obtained in the experiments. Most of these correlations predicted experimental data well in the pseudo-critical region. However, all correlations are matching well with experimental data beyond the pseudo-critical region. The details of the experimental facility, Experiments carried out and the results presented in this report. (author)

  20. Two-phase flow patterns and their relationship to two-phase heat transfer

    International Nuclear Information System (INIS)

    Hewitt, G.F.

    1977-01-01

    The objective of this lecture was to discuss the general nature of two phase flows, to define the various regimes of flow and to discuss the influence of these regimes on the heat transfer processes taking place. The methods of regime delineation are briefly described and regime descriptions introduced for both vertical and horizontal flows in tubes. ''Flow regime maps'' have been widely used as an aid to determination of the regime which occurs in a given situation. Some of the more widely used maps are described and the limitations of this approach discussed. There have been many attempts to obtain a better phenomenological description of two phase flow patterns. In this lecture, these attempts will be reviewed in the context of the bubble/plug, plug/churn and churn/annular flow transitions in vertical flow. The latter two transitions are related to the flooding/flow reversal phenomena. For horizontal flows, recent work on the onset of slugging will be reviewed. In flows with evaporation or condensation, the situation is influenced by departures from thermodynamic equilibrium and the types of departure observed are discuss briefly. Flow patterns and their relationships with heat transfer regimes are then reviewed for the case of condensation in horizontal tubes and evaporation in vertical tubes

  1. Impact of design options on natural circulation performance of the AFR-300 advanced fast reactor

    International Nuclear Information System (INIS)

    Dunn, F. D.

    2002-01-01

    The AFR-300, Advanced Fast Reactor (300 Mwe), has been proposed as a Generation IV concept. It could also be used to dispose of surplus weapons grade plutonium or as an actinide burner for transmutation of high level radioactive waste. AFR-300 uses metallic fuel and sodium coolant. The design of AFR-300 takes account of the successful design and operation of EBR-II, but the AFR-300 design includes a number of advances such as an advanced fuel cycle, inspectability and improved economics. One significant difference between AFR-300 and EBR-II is that AFR-300 is considerably larger. Another significant difference is that AFR-300 has no auxiliary EM pump in the primary loop to guarantee positive core flow when the main primary pumps are shut down. Thus, one question that has come up in connection with the AFR-300 design is whether natural circulation flow is sufficient to prevent damage to the core if the primary pumps fail. Insufficient natural circulation flow through the core could result in high cladding temperatures and cladding failure due to eutectic penetration of the cladding by the metal fuel. The rate of eutectic penetration of the cladding is strongly temperature dependent, so cladding failure depends on how hot the cladding gets and how long it is at elevated temperatures. To investigate the adequacy of natural circulation flow, a number of pump failure transients and a number of design options have been analyzed with the SASSYS-1 systems analysis code. This code has been validated for natural circulation behavior by analysis of Shutdown Heat Removal Tests performed in EBR-II. The AFR-300 design includes flywheels on the primary pumps to extend the pump coastdown times, and the size of the flywheels can be picked to give optimum coastdown times. One series of transients that has been run consists of protected loss-of-flow transients with various values for the combined moment of inertia of the pump, the motor and the flywheel giving coastdown times from 70

  2. Numerical modeling of supercritical CO{sub 2} natural circulation loop

    Energy Technology Data Exchange (ETDEWEB)

    Archana, V., E-mail: archanav@barc.gov.in [Homi Bhabha National Institute, Mumbai, Maharashtra 400 094 (India); Vaidya, A.M., E-mail: avaidya@barc.gov.in [Bhabha Atomic Research Centre, Mumbai, Maharashtra 400 085 (India); Vijayan, P.K., E-mail: vijayanp@barc.gov.in [Bhabha Atomic Research Centre, Mumbai, Maharashtra 400 085 (India)

    2015-11-15

    Highlights: • Supercritical CO{sub 2} natural circulation loop is modeled by in-house developed 1D and 2D axi-symmetric CFD codes. • Steady state characteristics of VHVC configuration of supercritical CO{sub 2} natural circulation loop are studied over a range of power. • Improved accuracy of predictions by 2D axi-symmetric formulation over 1D formulation is demonstrated. • The validity of correlations used in 1D model such as friction factor and heat transfer correlations is analyzed. • Simulation results shows normal, enhanced and deteriorated heat transfer regimes in supercritical CO{sub 2} natural circulation loop. - Abstract: The objective of this research project is to estimate steady state characteristics of supercritical natural circulation loop (SCNCL) using computational methodology and to compliment on-going experimental investigation of the same at the authors’ organization. For computational investigation, a couple of in-house codes are developed. At first, formulation and a corresponding computer program for the SCNCL based on conservation equations written in 1D framework is developed. Comparison of 1D code results with experimental data showed that, under some operating conditions, there is deviation between computed results and experimental data. To improve predictive capability, it was thought to model the SCNCL using conservation equations in 2D axi-symmetric framework. An existing 2D axi-symmetric code (named NAFA), which was developed and validated for supercritical fluid flow in pipes, is modified for natural circulation loop (NCL) geometry. The modified code, named NAFA-Loop, is subsequently used to compute the steady state characteristics of the SCNCL. These results are compared with experimental data. The steady state characteristics such as the variation of mass flow rate with power, velocity and temperature profiles in heater and cooler are studied using NAFA-Loop. The computed velocity and temperature fields show that the

  3. Two-phased flow component loss data

    International Nuclear Information System (INIS)

    Fairhurst, C.P.

    1983-01-01

    Pressure loss measurements were made for valves and orifice plates under horizontal and vertical two-phase, air/water flow. The results displayed similar trends and were successfully correlated using a semi-empirical approach. (author)

  4. Two-phase flow structure in large diameter pipes

    International Nuclear Information System (INIS)

    Smith, T.R.; Schlegel, J.P.; Hibiki, T.; Ishii, M.

    2012-01-01

    Highlights: ► Local profiles of various quantities measured in large diameter pipe. ► Database for interfacial area in large pipes extended to churn-turbulent flow. ► Flow regime map confirms previous models for flow regime transitions. ► Data will be useful in developing interfacial area transport models for large pipes. - Abstract: Flow in large pipes is important in a wide variety of applications. In the nuclear industry in particular, understanding of flow in large diameter pipes is essential in predicting the behavior of reactor systems. This is especially true of natural circulation Boiling Water Reactor (BWR) designs, where a large-diameter chimney above the core provides the gravity head to drive circulation of the coolant through the reactor. The behavior of such reactors during transients and during normal operation will be predicted using advanced thermal–hydraulics analysis codes utilizing the two-fluid model. Essential to accurate two-fluid model calculations is reliable and accurate computation of the interfacial transfer terms. These interfacial transfer terms can be expressed as the product of one term describing the potential driving the transfer and a second term describing the available surface area for transfer, or interfacial area concentration. Currently, the interfacial area is predicted using flow regime dependent empirical correlations; however the interfacial area concentration is best computed through the use of the one-dimensional interfacial area transport equation (IATE). To facilitate the development of IATE source and sink term models in large-diameter pipes a fundamental understanding of the structure of the two-phase flow is essential. This understanding is improved through measurement of the local void fraction, interfacial area concentration and gas velocity profiles in pipes with diameters of 0.102 m and 0.152 m under a wide variety of flow conditions. Additionally, flow regime identification has been performed to

  5. Two-phase flow induced parametric vibrations in structural systems

    International Nuclear Information System (INIS)

    Hara, Fumio

    1980-01-01

    This paper is divided into two parts concerning piping systems and a nuclear fuel pin system. The significant experimental results concerning the random vibration induced in an L-shaped pipe by air-water two-phase flow and the theoretical analysis of the vibration are described in the first part. It was clarified for the first time that the parametric excitation due to the periodic changes of system mass, centrifugal force and Coriolis force was the mechanism of exciting the vibration. Moreover, the experimental and theoretical analyses of the mechanism of exciting vibration by air-water two-phase flow in a straight, horizontal pipe were carried out, and the first natural frequency of the piping system was strongly related to the dominant frequency of void signals. The experimental results on the vibration of a nuclear fuel pin model in parallel air-water two-phase flow are reported in the latter part. The relations between vibrational strain variance and two-phase flow velocity or pressure fluctuation, and the frequency characteristics of vibrational strain variance were obtained. The theoretical analysis of the dynamic interaction between air-water two-phase flow and a fuel pin structure, and the vibrational instability of fuel pins in alternate air and water slugs or in large bubble flow are also reported. (Kako, I.)

  6. Operation of a cascade air conditioning system with two-phase loop

    Science.gov (United States)

    Feng, Yinshan; Wang, Jinliang; Zhao, Futao; Verma, Parmesh; Radcliff, Thomas D.

    2018-05-29

    A method of operating a heat transfer system includes starting operation of a first heat transfer fluid vapor/compression circulation loop including a fluid pumping mechanism, a heat exchanger for rejecting thermal energy from a first heat transfer fluid, and a heat absorption side of an internal heat exchanger. A first conduit in a closed fluid circulation loop circulates the first heat transfer fluid therethrough. Operation of a second two-phase heat transfer fluid circulation loop is started after starting operation of the first heat transfer fluid circulation loop. The second heat transfer fluid circulation loop transfers heat to the first heat transfer fluid circulation loop through the internal heat exchanger and includes a heat rejection side of the internal heat exchanger, a liquid pump, and a heat exchanger evaporator. A second conduit in a closed fluid circulation loop circulates a second heat transfer fluid therethrough.

  7. Correlation of Steam Generator Mixing Parameters for Severe Accident Hot-Leg Natural Circulation

    Energy Technology Data Exchange (ETDEWEB)

    Liao, Yehong; Guentay, Salih [Paul Scherrer Institut, Villigen PSI, CH-5232 (Switzerland)

    2008-07-01

    Steam generator inlet plenum mixing phenomenon with hot-leg counter-current natural circulation during a PWR station blackout severe accident is one of the important processes governing which component will fail first as a result of thermal challenge from the circulating gas with high temperature and pressure. Since steam generator tube failure represents bypass release of fission product from the reactor to environment, study of inlet plenum mixing parameters is important to risk analysis. Probability distribution functions of individual mixing parameter should be obtained from experiments or calculated by analysis. In order to perform sensitivity studies of the synergetic effects of all mixing parameters on the severe accident-induced steam generator tube failure, the distribution and correlation of these mixing parameters must be known to remove undue conservatism in thermal-hydraulic calculations. This paper discusses physical laws governing three mixing parameters in a steady state and setups the correlation among these mixing parameters. The correlation is then applied to obtain the distribution of one of the mixing parameters that has not been given in the previous CFD analysis. Using the distributions and considering the inter-dependence of the three mixing parameters, three sensitivity cases enveloping the mixing parameter uncertainties are recommended for the plant analysis. (authors)

  8. Circulation of Coxiella burnetii in a Naturally Infected Flock of Dairy Sheep: Shedding Dynamics, Environmental Contamination, and Genotype Diversity

    OpenAIRE

    Joulié, A.; Laroucau, K.; Bailly, X.; Prigent, M.; Gasqui, P.; Lepetitcolin, E.; Blanchard, B.; Rousset, E.; Sidi-Boumedine, K.; Jourdain, E.

    2015-01-01

    Q fever is a worldwide zoonosis caused by Coxiella burnetii. Domestic ruminants are considered to be the main reservoir. Sheep, in particular, may frequently cause outbreaks in humans. Because within-flock circulation data are essential to implementing optimal management strategies, we performed a follow-up study of a naturally infected flock of dairy sheep. We aimed to (i) describe C. burnetii shedding dynamics by sampling vaginal mucus, feces, and milk, (ii) assess circulating strain divers...

  9. Fundamental study on thermo-hydraulics during start-up in natural circulation boiling water reactors, (1)

    International Nuclear Information System (INIS)

    Aritomi, Masanori; Chiang Jing-Hsien; Takahashi, Tohru; Wataru, Masumi; Mori, Michitsugu.

    1992-01-01

    Recently, many concepts, in which passive and simplified functions are actively adapted, have been proposed for the next generation LWRs. The natural circulation BWR is one such considered from the requirements for next generation LWRs as compared with current BWRs. It is pointed out from this consideration that a thermo-hydraulic instability, which may appear during start-up, greatly influences concept feasibility because its occurence makes operation for raising power output difficult. Thermo-hydraulic instabilities are investigated experimentally under conditions simulating normal and abnormal start-up processes. It is clarified that three kinds of thermo-hydraulic instabilities may occur during start-up in the natural circulation BWR according to its procedure and reactor configuration, which are (1) geysering induced by condensation, (2) natural circulation instability induced by hydrostatic head fluctuation in steam separators and (3) density wave instability. Driving mechanisms of the geysering and the natural circulation instability, which have never understood enough, are inferred from the results. Finally, the difference of thermo-hydraulic behavior during start-up processes between thermal natural circulation boilers and the Dodewaard reactor is discussed. (author)

  10. Feasibility analysis of the Primary Loop of Pool-Type Natural Circulating Nuclear Reactor Dedicated to Seawater Desalination

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Woonho; Jeong, Yong Hoon [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    In this study, the feasibility of natural circulation was evaluated for the reference plant AHR400 (Advanced Heating Reactor 400MWth). AHR400 is a pool-type desalination-dedicated nuclear reactor. As a consequence, AHR400 has low operating pressure and temperature which provides large safety margin. Removal of the reactor coolant pump from the AHR400 will enforce integrity of the reactor vessel and passive safety feature. Therefore, the study also tried to find out optimized primary loop design to achieve total natural circulation of the coolant. Natural circulation capacity of the primary loop of the desalination dedicated nuclear reactor AHR400 was evaluated. It was concluded that to remove RCP from the AHR400 and operates the reactor only by natural circulation of the coolant is impossible. Decreased core power as half make removal of RCP possible with 15m central height difference between the core and IHXs. Furthermore, validation and modification of pressure loss coefficients by small-scaled natural circulation experiment at a pool-type reactor would provide more accurate results.

  11. Steam drum level control studies of a natural circulation multi loop reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Rajesh; Contractor, A.D.; Srivastava, Abhishek; Lele, H.G. [Bhabha Atomic Research Centre, Trombay, Mumbai (India). Reactor Safety Div.; Vaze, K.K. [Bhabha Atomic Research Centre, Trombay, Mumbai (India). Reactor Design and Development Group

    2013-12-15

    The proposed heavy water moderated and light water cooled pressure tube type boiling water reactor works on natural circulation at all power levels. It has parallel inter-connected loops with 452 boiling channels in the main heat transport system configuration. These multiple (four) interconnected loops influence the steam drum level control adversely through the common reactor inlet header. Alternate design studies made earlier for efficient control of SD levels have shown favorable results. This has lead to explore further the present scheme with the compartmentalization of CRIH into four compartments catering to four loops separately. The conventional 3-element level control has been found to be working satisfactorily. The interconnections between ECCS header and inlet header compartments have also increased the safety margin for various LOCA and design basis events. The paper deals with the SD level control aspects for this novel MHT configuration which has been analyzed for various PIEs (Postulated Initiating Events) and found to be satisfactory. (orig.)

  12. Analysis of natural circulation stability in a low pressure thermohydraulic test loop

    International Nuclear Information System (INIS)

    Jafari, J.; D'Auria, F.; Kazeminejad, H.; Davilu, H.

    2002-01-01

    This paper discusses an instability study of a natural circulation (NC) loop performed with the aid of Relap5 thermal-hydraulic system code. This loop has been designed and constructed for the analysis of relevant thermohydraulic parameters of a nuclear reactor. In this study, the main parameters for the stability of NC are identified and characterized through the execution of proper code runs. The obtained stability boundary (SB) in the dimensionless Zuber- Sub-cooling plane is compared with the SB reported in referenced literature. The agreement of predicted NC stability boundaries with the results of independent studies demonstrates both the capability of the mentioned code in assessing NC loop stability and the quality of the performed calculations.(author)

  13. Steady state and transient heat transfer on molten salt natural circulation loop

    International Nuclear Information System (INIS)

    Kudariyawar, Jayaraj Y.; Vaidya, A.M.; Maheshwari, N.K.; Satyamurthy, P.

    2016-01-01

    In this work, heat transfer characteristics of Molten Salt Natural Circulation Loop (MSNCL) are studied using 3D CFD simulations. Molten Nitrate salt, NaNO_3+KNO_3 (60:40 ratio by weight), is used as a fluid in MSNCL. In the MSNCL, in heater section, flow is developing and also mixed convection flow regime exists. The local Nusselt number variation in heater is calculated from computed data and is compared with that from Boelter correlation. Steady state heat transfer characteristics are obtained using CFD simulations. Transient heat transfer characteristics in the oscillatory flow formed in MSNCL with horizontal heater configuration are also studied and are found to be different as compared to vertical heater configuration. (author)

  14. Modeling of natural circulation and assessment of passive safety system performance (activities current status)

    International Nuclear Information System (INIS)

    Bykov, M.A.; Kryuchkov, M.E.

    2011-01-01

    Conclusions under the analysis of natural circulation: • Verifying calculations by means of CORSAR/GP code for values of power of a heater W = 1.5 and 2.5 kw have been done. Comparison with experimental data has been spent. • The received results do not coincide from a physical part of sight - in experiment change of a sign on the flow rate is observed during a problem, in calculations the flow rate is unequivocal. • However numerical values for the first variant (W = 1.5 kw) are close enough to the experimental. • For the second variant (W = 2.5 kw) are observed appreciable divergences. Researches in this direction will proceed. As is planned to continue researches in the field of uncertainty of parameters for this facilities

  15. Experimental Investigation of a Natural Circulation Solar Domestic Water Heater Performance under Standard Consumption Rate

    DEFF Research Database (Denmark)

    Kolaei, Alireza Rezania; Taherian, H.; Ganji, D. D.

    2012-01-01

    This paper reports experimental studies on the performance of a natural circulation solar water heater considering the weather condition of a city in north of Iran. The tests are done on clear and partly cloudy days. The variations of storage tank temperature due to consumption from the tank, daily...... consumption influence on the solar water heater efficiency, and on the input temperature of the collector are studied and the delivered daily useful energy has been obtained. The results show that by withdrawing from storage tank, the system as well as its collector efficiency will increase. Considering...... the value of the coefficient FRUL and τα, which are obtained experimentally as 6.03 and 0.83 respectively, average. monthly total load that is covered by this solar water heating system is estimated....

  16. Natural circulation in a VVER reactor geometry: Experiments with the PACTEL facility and Cathare simulations

    Energy Technology Data Exchange (ETDEWEB)

    Raussi, P.; Kainulainen, S. [Lappeenranta Univ. of Technology, Lappeenranta (Finland); Kouhia, J. [VTT Energy, Lappeenranta (Finland)

    1995-09-01

    There are some 40 reactors based on the VVER design in use. Database available for computer code assessment for VVER reactors is rather limited. Experiments were conducted to study natural circulation behaviour in the PACTEL facility, a medium-scale integral test loop patterned after VVER pressurized water reactors. Flow behaviour over a range of coolant inventories was studied with a small-break experiment. In the small-break experiments, flow stagnation and system repressurization were observed when the water level in the upper plenum fell below the entrances to the hot legs. The cause was attributed to the hot leg loop seals, which are a unique feature of the VVER geometry. At low primary inventories, core cooling was achieved through the boiler-condenser mode. The experiment was simulated using French thermalhydraulic system code CATHARE.

  17. Studies of deep water formation and circulation in the Weddell Sea using natural and anthropogenic tracers

    International Nuclear Information System (INIS)

    Schlosser, Peter; Bayer, Reinhold

    1991-01-01

    The application of natural and anthropogenic trace substances in oceanographic studies of the Weddell Sea is reviewed. The potential of some steady-state and transient tracers (tritium, CFC-11 and CFC-12, 18 O, and helium isotopes) for studies of deep water formation and circulation is discussed on the basis of data sets collected mainly on cruises of R/V 'Polastern' to the Weddell Sea during the 1980s. CFC/ tritium ratio dating of young water masses is applied to estimate mean age and transit times of water involved in Weddell Sea Bottom Water formation. The history of the CFC-11/tritium ratio through time is derived for Weddell Sea shelf waters. (author). 36 refs.; 18 figs

  18. Computational simulation of natural circulation and rewetting experiments using the TRAC/PF1 code

    International Nuclear Information System (INIS)

    Silva, J.D. da.

    1994-05-01

    In this work the TRAC code was used to simulate experiments of natural circulation performed in the first Brazilian integral test facility at (COPESP), Sao Paulo and a rewetting experiment in a single tube test section carried out at CDTN, Belo Horizonte, Brazil. In the first simulation the loop behavior in two transient conditions with different thermal power, namely 20 k W and 120 k W, was verified in the second one the quench front propagation, the liquid mass collected in the carry over measuring tube and the wall temperature at different elevations during the flooding experiment was measured. A comparative analysis, for code consistency, shows a good agreement between the code results and experimental data, except for the quench from velocity. (author). 15 refs, 19 figs, 12 tabs

  19. Numerical Investigation of Startup Instabilities in Parallel-Channel Natural Circulation Boiling Systems

    Directory of Open Access Journals (Sweden)

    S. P. Lakshmanan

    2010-01-01

    Full Text Available The behaviour of a parallel-channel natural circulation boiling water reactor under a low-pressure low-power startup condition has been studied numerically (using RELAP5 and compared with its scaled model. The parallel-channel RELAP5 model is an extension of a single-channel model developed and validated with experimental results. Existence of in-phase and out-of-phase flashing instabilities in the parallel-channel systems is investigated through simulations under equal and unequal power boundary conditions in the channels. The effect of flow resistance on Type-I oscillations is explored. For nonidentical condition in the channels, the flow fluctuations in the parallel-channel systems are found to be out-of-phase.

  20. Natural Circulation Level Optimization and the Effect during ULOF Accident in the SPINNOR Reactors

    International Nuclear Information System (INIS)

    Abdullah, Ade Gafar; Su'ud, Zaki; Kurniadi, Rizal; Kurniasih, Neny; Yulianti, Yanti

    2010-01-01

    Natural circulation level optimization and the effect during loss of flow accident in the 250 MWt MOX fuelled small Pb-Bi Cooled non-refueling nuclear reactors (SPINNOR) have been performed. The simulation was performed using FI-ITB safety code which has been developed in ITB. The simulation begins with steady state calculation of neutron flux, power distribution and temperature distribution across the core, hot pool and cool pool, and also steam generator. When the accident is started due to the loss of pumping power the power distribution and the temperature distribution of core, hot pool and cool pool, and steam generator change. Then the feedback reactivity calculation is conducted, followed by kinetic calculation. The process is repeated until the optimum power distribution is achieved. The results show that the SPINNOR reactor has inherent safety capability against this accident.

  1. Heat transfer characteristics of horizontal steam generators under natural circulation conditions

    International Nuclear Information System (INIS)

    Hyvaerinen, J.

    1996-01-01

    This paper deals with the heat transfer characteristics of horizontal steam generators, particularly under natural circulation (decay heat removal) conditions on the primary side. Special emphasis is on the inherent features of horizontal steam generator behaviour. A mathematical model of the horizontal steam generator primary side is developed and qualitative results are obtained analytically. A computer code, called HSG, is developed to solve the model numerically, and its predictions are compared with experimental data. The code is employed to obtain for VVER 440 steam generators quantitative results concerning the dependence of primary-to-secondary heat transfer efficiency on the primary side flow rate, temperature and secondary level. It turns out that the depletion of the secondary inventory leads to an inherent limitation of the decay energy removal in VVER steam generators. The limitation arises as a consequence of the steam generator tube bundle geometry. As an example, it is shown that the grace period associated with pressurizer safety valve opening during a station black-out is 2 1/2-3 hours instead of the 5-6 hours reported in several earlier studies. (However, the change in core heat-up timing is much less-about 1 h at most.) The heat transfer limitation explains the fact that, in the Greifswald VVER 440 station black-out accident in 1975, the steam generators never boiled dry. In addition, the stability of single-phase natural circulation is discussed and insights on the modelling of horizontal steam generators with general-purpose thermal-hydraulic system codes are also presented. (orig.)

  2. Transient computational fluid dynamics analysis of emergency core cooling injection at natural circulation conditions

    Energy Technology Data Exchange (ETDEWEB)

    Scheuerer, Martina, E-mail: Martina.Scheuerer@grs.de [Gesellschaft fuer Anlagen- und Reaktorsicherheit, Forschungsinstitute, 85748 Garching (Germany); Weis, Johannes, E-mail: Johannes.Weis@grs.de [Gesellschaft fuer Anlagen- und Reaktorsicherheit, Forschungsinstitute, 85748 Garching (Germany)

    2012-12-15

    Highlights: Black-Right-Pointing-Pointer Pressurized thermal shocks are important phenomena for plant life extension and aging. Black-Right-Pointing-Pointer The thermal-hydraulics of PTS have been studied experimentally and numerically. Black-Right-Pointing-Pointer In the Large Scale Test Facility a loss of coolant accident was investigated. Black-Right-Pointing-Pointer CFD software is validated to simulate the buoyancy driven flow after ECC injection. - Abstract: Within the framework of the European Nuclear Reactor Integrated Simulation Project (NURISP), computational fluid dynamics (CFD) software is validated for the simulation of the thermo-hydraulics of pressurized thermal shocks. A proposed validation experiment is the test series performed within the OECD ROSA V project in the Large Scale Test Facility (LSTF). The LSTF is a 1:48 volume-scaled model of a four-loop Westinghouse pressurized water reactor (PWR). ROSA V Test 1-1 investigates temperature stratification under natural circulation conditions. This paper describes calculations which were performed with the ANSYS CFD software for emergency core cooling injection into one loop at single-phase flow conditions. Following the OECD/NEA CFD Best Practice Guidelines (Mahaffy, 2007) the influence of grid resolution, discretisation schemes, and turbulence models (shear stress transport and Reynolds stress model) on the mixing in the cold leg were investigated. A half-model was used for these simulations. The transient calculations were started from a steady-state solution at natural circulation conditions. The final calculations were obtained in a complete model of the downcomer. The results are in good agreement with data.

  3. Computational stability appraisal of rectangular natural circulation loop: Effect of loop inclination

    International Nuclear Information System (INIS)

    Krishnani, Mayur; Basu, Dipankar N.

    2017-01-01

    Highlights: • Computational model developed for single-phase rectangular natural circulation loop. • Role of loop inclination to vertical on thermalhydraulic stability is explored. • Inclination has strong stabilizing effect due to lower effective gravitation force. • Increase in tilt angle reduces settling time and highest amplitude of oscillation. • An angle of 15° is suggested for the selected loop geometry. - Abstract: Controlling stability behavior of single-phase natural circulation loops, without significantly affecting its steady-state characteristics, is a topic of wide research interest. Present study explores the role of loop inclination on a particular loop geometry. Accordingly a 3D computational model of a rectangular loop is developed and transient conservation equations are solved to obtain the temporal variation in flow parameters. Starting from the quiescent state, simulations are performed for selected sets of operating conditions and also with a few selected inclination angles. System experiences instability at higher heater powers and also with higher sink temperatures. Inclination is found to have a strong stabilizing influence owing to the reduction in the effective gravitational acceleration and subsequent decline in local buoyancy effects. The settling time and highest amplitude of oscillations substantially reduces for a stable system with a small inclination. Typically-unstable systems can also suppress the oscillations, when subjected to tilting, within a reasonable period of time. It is possible to stabilize the loop within shorter time span by increasing the tilt angle, but at the expense of reduction in steady-state flow rate. Overall a tilt angle of 15° is suggested for the selected geometry. Results from the 3D model is compared with the predictions from an indigenous 1D code. While similar qualitative influence of inclination is observed, the 1D model predicts early appearance of the stability threshold and hence hints

  4. Transient computational fluid dynamics analysis of emergency core cooling injection at natural circulation conditions

    International Nuclear Information System (INIS)

    Scheuerer, Martina; Weis, Johannes

    2012-01-01

    Highlights: ► Pressurized thermal shocks are important phenomena for plant life extension and aging. ► The thermal-hydraulics of PTS have been studied experimentally and numerically. ► In the Large Scale Test Facility a loss of coolant accident was investigated. ► CFD software is validated to simulate the buoyancy driven flow after ECC injection. - Abstract: Within the framework of the European Nuclear Reactor Integrated Simulation Project (NURISP), computational fluid dynamics (CFD) software is validated for the simulation of the thermo-hydraulics of pressurized thermal shocks. A proposed validation experiment is the test series performed within the OECD ROSA V project in the Large Scale Test Facility (LSTF). The LSTF is a 1:48 volume-scaled model of a four-loop Westinghouse pressurized water reactor (PWR). ROSA V Test 1-1 investigates temperature stratification under natural circulation conditions. This paper describes calculations which were performed with the ANSYS CFD software for emergency core cooling injection into one loop at single-phase flow conditions. Following the OECD/NEA CFD Best Practice Guidelines (Mahaffy, 2007) the influence of grid resolution, discretisation schemes, and turbulence models (shear stress transport and Reynolds stress model) on the mixing in the cold leg were investigated. A half-model was used for these simulations. The transient calculations were started from a steady-state solution at natural circulation conditions. The final calculations were obtained in a complete model of the downcomer. The results are in good agreement with data.

  5. Numerical simulation of shell-side heat transfer and flow of natural circulation heat exchanger

    International Nuclear Information System (INIS)

    Xue Ruojun; Deng Chengcheng; Li Chaojun; Wang Mingyuan

    2012-01-01

    In order to analyze the influence on the heat transfer and flow characteristics of the heat exchanger model of different solving models and structures, a variety of transformation to the model equivalent for the heat exchanger was studied. In this paper, Fluent software was used to simulate the temperature-field and flow-field of the equivalent model, and investigate its heat-transferring and flow characteristics. Through comparative analysis of the distribution of temperature-field and flow-field for different models, the heat-transferring process and natural convection situation of heat exchanger were deeply understood. The results show that the temperature difference between the inside and outside of the natural circulation heat exchanger tubes is larger and the flow is more complex, so the turbulence model is the more reasonable choice. Asymmetry of tubes position makes the flow and heat transfer of the fluid on both sides to be dissymmetrical and makes the fluid interaction, and increases the role of natural convection. The complex structure of heat exchanger makes the flow and heat transfer of the fluid on both sides to be irregular to some extent when straight tubes into C-bent are transformed, and all these make the turbulence intensity increase and improve the effect of heat transfer. (authors)

  6. Two phase cooling for superconducting magnets

    International Nuclear Information System (INIS)

    Eberhard, P.H.; Gibson, G.A.; Green, M.A.; Ross, R.R.; Smits, R.G.

    1986-01-01

    Comments on the use of two phase helium in a closed circuit tubular cooling system and some results obtained with the TPC superconducting magnet are given. Theoretical arguments and experimental evidence are given against a previously suggested method to determine helium two phase flow regimes. Two methods to reduce pressure in the magnet cooling tubes during quenches are discussed; 1) lowering the density of helium in the magnet cooling tubes and 2) proper location of pressure relief valves. Some techniques used to protect the refrigerator from too much cold return gas are also mentioned

  7. Two phase cooling for superconducting magnets

    International Nuclear Information System (INIS)

    Eberhard, P.H.; Gibson, G.A.; Green, M.A.; Ross, R.R.; Smits, R.G.; Taylor, J.D.; Watt, R.D.

    1986-01-01

    Comments on the use of two phase helium in a closed circuit tubular cooling system and some results obtained with the TPC superconducting magnet are given. Theoretical arguments and experimental evidence are given against a previously suggested method to determine helium two phase flow regimes. Two methods to reduce pressure in the magnet cooling tubes during quenches are discussed; (1) lowering the density of helium in the magnet cooling tubes and (2) proper location of pressure relief valves. Some techniques used to protect the refrigerator from too much cold return gas are also mentioned. 10 refs., 1 fig., 5 tabs

  8. The role of heater thermal response in reactor thermal limits during oscillartory two-phase flows

    Energy Technology Data Exchange (ETDEWEB)

    Ruggles, A.E.; Brown, N.W. [Univ. of Tennessee, Knoxville, TN (United States); Vasil`ev, A.D. [Nuclear Safety Institute, Moscow, (Russian Federation); Wendel, M.W. [Oak Ridge National Lab., TN (United States)

    1995-09-01

    Analytical and numerical investigations of critical heat flux (CHF) and reactor thermal limits are conducted for oscillatory two-phase flows often associated with natural circulation conditions. It is shown that the CHF and associated thermal limits depend on the amplitude of the flow oscillations, the period of the flow oscillations, and the thermal properties and dimensions of the heater. The value of the thermal limit can be much lower in unsteady flow situations than would be expected using time average flow conditions. It is also shown that the properties of the heater strongly influence the thermal limit value in unsteady flow situations, which is very important to the design of experiments to evaluate thermal limits for reactor fuel systems.

  9. RETRAN-02 analysis of upper head cooling during controlled natural circulation cooldown of Yankee Nuclear Power Station

    International Nuclear Information System (INIS)

    Fujita, N.; Helrich, R.E.; Bergeron, P.A.

    1982-01-01

    RETRAN-02 is particularly well-suited for investigating the fluid conditions in the upper head during a natural circulation cooldown. The RETRAN input model was developed with four basic objectives: (1) accurate description of the upper head cooling mechanisms; (2) proper simulation of natural circulation; (3) respresentations of operator actions required to proceed from full-power to shutdown-cooling-system conditions using both automatic and manual controls; and (4) reduction of the computer cost of simulating this evolution of approximately 10-hour duration. The response of the upper head fluid temperature calculated by RETRAN was in close agreement with measured data obtained from a natural circulation cooldown experiment performed for the Connecticut Yankee Plant, whose design is very similar to the Yankee Nuclear Power Station

  10. Analysis of a natural draught tower in the circulation seawater system of nuclear power plant of Laguna Verde

    International Nuclear Information System (INIS)

    Tijerina S, F.; Vargas A, A.

    2009-10-01

    The analysis of a natural draught tower in open circuit for the cooling system of seawater circulation on the nuclear power plant of Laguna Verde, it is based on conditions of 2027 MWt and 2317 MWt, where the flows of circulation water system hardly vary and whose purpose will be, to cool the seawater circulation. The circulation water system is used as heat drain in main condenser of turbo generator to condense the nuclear vapor. The annual average temperature in the seawater at present is of 26 C to the entrance to circulation water system and it is vary in accordance with the time of year. The mean temperature of leaving of circulation water system to the sea is of 41 C. Having a cooling tower to reduce the entrance temperature to the circulation water system, it improves the efficiency of thermal transfer in condenser, it improves the vacuum in condenser giving more operative margin to avoid condenser losses by air entrances and nuclear power plant shutdowns, as well as for to improve the efficiency of operative balance of nuclear power plant, also it prevents the impact in thermal transfer efficiency in condenser by the climatic change. (Author)

  11. Nonlinear dynamics of two-phase flow

    International Nuclear Information System (INIS)

    Rizwan-uddin

    1986-01-01

    Unstable flow conditions can occur in a wide variety of laboratory and industry equipment that involve two-phase flow. Instabilities in industrial equipment, which include boiling water reactor (BWR) cores, steam generators, heated channels, cryogenic fluid heaters, heat exchangers, etc., are related to their nonlinear dynamics. These instabilities can be of static (Ledinegg instability) or dynamic (density wave oscillations) type. Determination of regions in parameters space where these instabilities can occur and knowledge of system dynamics in or near these regions is essential for the safe operation of such equipment. Many two-phase flow engineering components can be modeled as heated channels. The set of partial differential equations that describes the dynamics of single- and two-phase flow, for the special case of uniform heat flux along the length of the channel, can be reduced to a set of two coupled ordinary differential equations [in inlet velocity v/sub i/(t) and two-phase residence time tau(t)] involving history integrals: a nonlinear ordinary functional differential equation and an integral equation. Hence, to solve these equations, the dependent variables must be specified for -(nu + tau) ≤ t ≤ 0, where nu is the single-phase residence time. This system of nonlinear equations has been solved analytically using asymptotic expansion series for finite but small perturbations and numerically using finite difference techniques

  12. Two phase transitions in Nuclear Physics

    International Nuclear Information System (INIS)

    Bes, D.R.

    1985-01-01

    The status of the art of the problem associated with two phase transitions in the nuclear matter, viz.: the disappearance of the nuclear superfluiditiy with the raising of the rotation velocity and the appearance of an octupolar deformation in the actinide zone, is presented. (L.C.) [pt

  13. Two-phase flow in fractured rock

    International Nuclear Information System (INIS)

    Davies, P.; Long, J.; Zuidema, P.

    1993-11-01

    This report gives the results of a three-day workshop on two-phase flow in fractured rock. The workshop focused on two-phase flow processes that are important in geologic disposal of nuclear waste as experienced in a variety of repository settings. The goals and objectives of the workshop were threefold: exchange information; describe the current state of understanding; and identify research needs. The participants were divided into four subgroups. Each group was asked to address a series of two-phase flow processes. The following groups were defined to address these processes: basic flow processes; fracture/matrix interactions; complex flow processes; and coupled processes. For each process, the groups were asked to address these four issues: (1) describe the two-phase flow processes that are important with respect to repository performance; (2) describe how this process relates to the specific driving programmatic issues given above for nuclear waste storage; (3) evaluate the state of understanding for these processes; and (4) suggest additional research to address poorly understood processes relevant to repository performance. The reports from each of the four working groups are given here

  14. Experimental study of the transition from forced to natural circulation in EBR-II at low power and flow

    International Nuclear Information System (INIS)

    Gillette, J.L.; Singer, R.M.; Tokar, J.V.; Sullivan, J.E.

    1979-01-01

    A series of tests have been conducted in EBR-II which studied the dynamics of the transition from forced to natural circulation flow in a liquid-metal-cooled fast breeder reactor (LMFBR). Each test was initiated by abruptly tripping an electromagnetic pump which supplies 5 to 6% of the normal full operational primary flow rate. The ensuing flow coast-down reached a minimum value after which the flow increased as natural circulation was established. The effects of secondary system flow through the intermediate heat exchanger and reactor decay power level on the minimum in-core flow rates and maximum in-core temperatures were examined

  15. The low-power low-pressure flow resonance in a natural circulation cooled boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hagen, T.H.J.J. van der; Stekelenburg, A.J.C. [Delft Univ. of Technology (Netherlands)

    1995-09-01

    The last few years the possibility of flow resonances during the start-up phase of natural circulation cooled BWRs has been put forward by several authors. The present paper reports on actual oscillations observed at the Dodewaard reactor, the world`s only operating BWR cooled by natural circulation. In addition, results of a parameter study performed by means of a simple theoretical model are presented. The influence of relevant parameters on the resonance characteristics, being the decay ratio and the resonance frequency, is investigated and explained.

  16. Benchmark Simulation of Natural Circulation Cooling System with Salt Working Fluid Using SAM

    Energy Technology Data Exchange (ETDEWEB)

    Ahmed, K. K.; Scarlat, R. O.; Hu, R.

    2017-09-03

    Liquid salt-cooled reactors, such as the Fluoride Salt-Cooled High-Temperature Reactor (FHR), offer passive decay heat removal through natural circulation using Direct Reactor Auxiliary Cooling System (DRACS) loops. The behavior of such systems should be well-understood through performance analysis. The advanced system thermal-hydraulics tool System Analysis Module (SAM) from Argonne National Laboratory has been selected for this purpose. The work presented here is part of a larger study in which SAM modeling capabilities are being enhanced for the system analyses of FHR or Molten Salt Reactors (MSR). Liquid salt thermophysical properties have been implemented in SAM, as well as properties of Dowtherm A, which is used as a simulant fluid for scaled experiments, for future code validation studies. Additional physics modules to represent phenomena specific to salt-cooled reactors, such as freezing of coolant, are being implemented in SAM. This study presents a useful first benchmark for the applicability of SAM to liquid salt-cooled reactors: it provides steady-state and transient comparisons for a salt reactor system. A RELAP5-3D model of the Mark-1 Pebble-Bed FHR (Mk1 PB-FHR), and in particular its DRACS loop for emergency heat removal, provides steady state and transient results for flow rates and temperatures in the system that are used here for code-to-code comparison with SAM. The transient studied is a loss of forced circulation with SCRAM event. To the knowledge of the authors, this is the first application of SAM to FHR or any other molten salt reactors. While building these models in SAM, any gaps in the code’s capability to simulate such systems are identified and addressed immediately, or listed as future improvements to the code.

  17. Operation of a forced two phase cooling system on a large superconducting magnet

    International Nuclear Information System (INIS)

    Green, M.A.; Burns, W.A.; Eberhard, P.H.; Gibson, G.H.; Pripstein, M.; Ross, R.R.; Smits, R.G.; Taylor, J.D.; Van Slyke, H.

    1980-05-01

    This paper describes the operation of a forced two phase cooling system on a two meter diameter superconducting solenoid. The magnet is a thin high current density superconducting solenoid which is cooled by forced two phase helium in tubes around the coil. The magnet, which is 2.18 meters in diameter and 3.4 meters long, has a cold mass of 1700 kg. The two phase cooling system contains less than 300 liters of liquid helium, most of which is contained in a control dewar. This paper describes the operating characteristics of the LBL two phase forced cooling system during cooldown and warm up. The paper presents experimental data on operations of the magnet using either a helium pump or the refrigerator compressor to circulate two phase helium through the superconducting coil cooling tubes

  18. Thermal characteristics and performance of Ag-water nanofluid: Application to natural circulation loops

    International Nuclear Information System (INIS)

    Koca, Halil Dogacan; Doganay, Serkan; Turgut, Alpaslan

    2017-01-01

    Highlights: • Thermal conductivity and viscosity of Ag-water nanofluid were measured. • Thermal performance of Ag-water nanofluid was compared with water. • Effectiveness enhanced up to 11% with 1 wt% Ag-water nanofluid. • Effectiveness of Ag-water nanofluid samples increased with inclination angle. • Ag-water nanofluid has potential to be used in flat-plate solar collectors. - Abstract: The goal of this study is to investigate the thermal conductivity, viscosity and thermal performance in a single-phase natural circulation mini loop of Ag-water nanofluid which can be a potential working fluid for natural convective flat-plate solar collectors. The silver-water nanofluid with 5 wt% concentration, which contains also polyvinylpyrrolidone (PVP) with 1.25 wt%, was purchased. Then, the sample was diluted with de-ionized water to four different concentrations of 0.25, 0.5, 0.75 and 1 wt%. Thermal conductivity and viscosity were measured by 3ω method and Brookfield rheometer, respectively. An effectiveness factor was used to define the thermal performance of Ag-water nanofluids for different inclination angles and heating powers. The results showed that nanofluid samples are thermally less conductive than the literature, at ambient temperature (23 °C). The viscosity of nanofluid decreases significantly with increasing temperature and increases with increasing concentration. Our measurements appear to be more compatible with PVP solution results available in the literature. Effectiveness is enhanced up to 11% with 1 wt% concentrated nanofluid compared to de-ionized water and the effectiveness of the mini loop indicates an enhancement with increase in inclination angle and particle concentration at whole applied power. According to obtained results, it is concluded that Ag-water nanofluid has a promising potential to be used in natural convective flat-plate solar collector.

  19. An introduction to two-phase flows

    International Nuclear Information System (INIS)

    Lemonnier, Herve

    2006-01-01

    This course aims at proposing the necessary background for a rational approach to two-phase flows which are notably present in numerous industrial devices and equipment designed to perform energy transfer or mass transfer. The first part proposes a phenomenological approach to main two-phase flow structures and presents their governing variables. The second part presents some proven measurement techniques. The third part focuses on modelling. It recalls the equation elaboration techniques which are based on basic principles of mechanics and thermodynamics and on the application of different averaging operators to these principles. Some useful models are then presented such as models of pressure loss in a duct. The last chapter addresses some fundamental elements of heat transfers in ebullition and condensation

  20. Apparatus for monitoring two-phase flow

    Science.gov (United States)

    Sheppard, John D.; Tong, Long S.

    1977-03-01

    A method and apparatus for monitoring two-phase flow is provided that is particularly related to the monitoring of transient two-phase (liquid-vapor) flow rates such as may occur during a pressurized water reactor core blow-down. The present invention essentially comprises the use of flanged wire screens or similar devices, such as perforated plates, to produce certain desirable effects in the flow regime for monitoring purposes. One desirable effect is a measurable and reproducible pressure drop across the screen. The pressure drop can be characterized for various known flow rates and then used to monitor nonhomogeneous flow regimes. Another useful effect of the use of screens or plates in nonhomogeneous flow is that such apparatus tends to create a uniformly dispersed flow regime in the immediate downstream vicinity. This is a desirable effect because it usually increases the accuracy of flow rate measurements determined by conventional methods.

  1. Geometrical automata for two phase flow simulation

    International Nuclear Information System (INIS)

    Herrero, V.; Guido-Lavalle, G.; Clausse, A.

    1996-01-01

    An automaton is an entity defined by a mathematical state which changes following iterative rules representing the interaction with the neighborhood. A model of automata for two-phase flow simulation consisting in a field of disks which are allowed to change their radii and move in a plane is presented. The model is more general than the classical cellular automata in two respects: (1) the grid of cellular automata is dismissed in favor of a trajectory generator; and (2) the rules of interaction involve parameters intended to represent some of the most relevant variables governing the actual physical interactions between phases. Computational experiments show that the algorithm captures the essential physics underlying two-phase flow problems such as bubbly-slug pattern transition and void fraction development along tubes. A comparison with experimental data of void fraction profiles is presented, showing excellent agreement. (orig.)

  2. Review of two-phase water hammer

    International Nuclear Information System (INIS)

    Beuthe, T.G.

    1997-01-01

    In a thermalhydraulic system like a nuclear power plant, where steam and water mix and are used to transport large amounts of energy, there is a potential to create two-phase water hammer. Large water hammer pressure transients are a threat to piping integrity and represent an important safety concern. Such events may cause unscheduled plant down time. The objective of this review is to provide a summary of the information on two-phase water hammer available in the open literature with particular emphasis on water hammer occurrences in nuclear power plants. Past reviews concentrated on studies concerned with preventing water hammer. The present review focuses on the fundamental experimental, analytical, and modelling studies. The papers discussed here were chosen from searches covering up to July 1993. (author)

  3. Modelling aspects of two phase flow

    International Nuclear Information System (INIS)

    Mayinger, F.

    1977-01-01

    In two phase flow scaling is much more limited to very narrowly defined physical phenomena than in single phase fluids. For complex and combined phenomena it can be achieved not by using dimensionless numbers alone but in addition a detailed mathematical description of the physical problem - usually in the form of a computer program - must be available. An important role plays the scaling of the thermodynamic data of the modelling fluid. From a literature survey and from own scaling experiments the conclusion can be drawn that Freon is a quite suitable modelling fluid for scaling steam-water mixtures. However, whithout a theoretical description of the phenomena nondimensional numbers for scaling two phase flow must be handled very carefully. (orig.) [de

  4. Apparatus for monitoring two-phase flow

    International Nuclear Information System (INIS)

    Sheppard, J.D.; Tong, L.S.

    1977-01-01

    A method and apparatus for monitoring two-phase flow is provided that is particularly related to the monitoring of transient two-phase (liquid-vapor) flow rates such as may occur during a pressurized water reactor core blow-down. The present invention essentially comprises the use of flanged wire screens or similar devices, such as perforated plates, to produce certain desirable effects in the flow regime for monitoring purposes. One desirable effect is a measurable and reproducible pressure drop across the screen. The pressure drop can be characterized for various known flow rates and then used to monitor nonhomogeneous flow regimes. Another useful effect of the use of screens or plates in nonhomogeneous flow is that such apparatus tends to create a uniformly dispersed flow regime in the immediate downstream vicinity. This is a desirable effect because it usually increases the accuracy of flow rate measurements determined by conventional methods. 3 claims, 9 figures

  5. MODELING THE AMBIENT CONDITION EFFECTS OF AN AIR-COOLED NATURAL CIRCULATION SYSTEM

    Energy Technology Data Exchange (ETDEWEB)

    Hu, Rui; Lisowski, Darius D.; Bucknor, Matthew; Kraus, Adam R.; Lv, Qiuping

    2017-07-02

    empirical model was also implemented in the computational models of the NSTF using both RELAP5-3D and STARCCM+ codes. Accounting for the effects of ambient conditions, simulations from both codes predicted the natural circulation flow rates very well.

  6. Study of nonequilibrium dispersed two phase flow

    International Nuclear Information System (INIS)

    Reyes, J.N. Jr.

    1986-01-01

    Understanding the behavior of liquid droplets in a superheated steam environment is essential to the accurate prediction of nuclear fuel rod surface temperatures during the blowdown and reflood phase of a loss-of-coolant-accident (LOCA). In response to this need, this treatise presents several original and significant contributions to the field of thermofluid physics. The research contained herein presents a statistical derivation of the two-phase mass, momentum, and energy-conservation equations using a droplet continuity equation analogous to that used in the Kinetic Theory of Gases. Unlike the Eulerian volume and time-averaged conservation equations generally used to describe dispersed two-phase flow behavior, this statistical averaging approach results in an additional mass momentum or energy term in each of the respective conservation equations. Further, this study demonstrates that current definitions of the volumetric vapor generation rate used in the mass conservation equation are inappropriate results under certain circumstances. The mass conservation equation derived herein is used to obtain a new definition for the volumetric vapor-generation rate. Last, a simple two phase phenomenological model, based on the statistically averaged conservation equations, is presented and solved analytically. It is shown that the actual quality and vapor temperature, under these circumstances, depend on a single dimensionless group

  7. Review of two-phase instabilities

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Han Ok; Seo, Han Ok; Kang, Hyung Suk; Cho, Bong Hyun; Lee, Doo Jeong

    1997-06-01

    KAERI is carrying out a development of the design for a new type of integral reactors. The once-through helical steam generator is important design features. The study on designs and operating conditions which prevent flow instability should precede the introduction of one-through steam generator. Experiments are currently scheduled to understand two-phase instability, evaluate the effect of each design parameter on the critical point, and determine proper inlet throttling for the prevention of instability. This report covers general two-phase instability with review of existing studies on this topics. The general classification of two phase flow instability and the characteristics of each type of instability are first described. Special attention is paid to BWR core flow instability and once-through steam generator instability. The reactivity feedback and the effect of system parameters are treated mainly for BWR. With relation to once-through steam generators, the characteristics of convective heating and dryout point oscillation are first investigated and then the existing experimental studies are summarized. Finally chapter summarized the proposed correlations for instability boundary conditions. (author). 231 refs., 5 tabs., 47 figs

  8. Application of Melcor code for the calculo of TMLB sequence in PWR with natural circulating into the vessel

    International Nuclear Information System (INIS)

    Marten-Fuertes, F.

    1995-01-01

    The use of computer codes to analyze the phenomena of severe accidents is very important to take decisions in Nuclear Safety. This paper presents the MELCOR code used to calculate the TMLB sequence of PWR with natural circulation into the vessels. The main goal of this code is its application for the PSA (probabilistic safety analysis)

  9. Evaluation method for core thermohydraulics during natural circulation in fast reactors numerical predictions of inter-wrapper flow

    International Nuclear Information System (INIS)

    Kamide, H.; Kimura, N.; Miyakoshi, H.; Nagasawa, K.

    2001-01-01

    Decay heat removal using natural circulation is one of the important functions for the safety of fast reactors. As a decay heat removal system, direct reactor auxiliary cooling system has been selected in current designs of fast reactors. In this design, dumped heat exchanger provides cold sodium and it covers the reactor core outlet. The cold sodium can penetrate into the gap region between the subassemblies. This gap flow is referred as inter-wrapper flow (IWF). A numerical estimation method for such natural circulation phenomena in a reactor core has been developed, which models each subassembly as a rectangular duct with gap region between the subassemblies and also the upper plenum in a reactor vessel. This numerical simulation method was verified based on experimental data of a sodium test using 7- subassembly core model and also a water test which simulates IWF using the 1/12 sector model of a reactor core. We applied the estimation method to the natural circulation in a 600 MW class fast reactor. The temperature in the core strongly depended on IWF, flow redistribution in the core, and inter-subassembly heat transfer. It is desired for prediction methods on the natural circulation to simulate these phenomena. (author)

  10. Thermal analysis of lithium cooled natural circulation loop module for fuel rod testing in the Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Eyler, L.L.; Kim, D.; Stover, R.L.; Beaver, T.R.

    1987-01-01

    Maximum heat removal capability of a lithium cooled natural circulation fuel rod test module design is determined. Loop geometry is optimized within limitations of design specifications for nominal operation temperatures, materials, and test module environment. Results provide test module operation limits and range of potential uncertainties. 3 refs., 12 figs

  11. Computational simulation of the natural circulation occurring in an experimental test section of a pool type research reactor

    International Nuclear Information System (INIS)

    Nascimento, Francisco R.T. do; Lima Junior, Carlos A.S.; Oliveira, Andre F.S. de; Affonso, Renato R.W.; Faccini, Jose L.H.; Moreira, Maria L.

    2015-01-01

    The present work presents a computational simulation of the natural circulation phenomenon developing in an experimental test section of a pool type research reactor. The test section has been designed using a reduced scale in height 1:4.7 in relation to a pool type 30 MW research reactor prototype. It comprises a cylindrical vessel, which is opened to atmosphere, and representing the reactor pool; a natural circulation pipe, a lower plenum, and a heater containing electrical resistors in rectangular plate format, which represents the fuel elements, with a chimney positioned on the top of the resistor assembly. In the computational simulation, it was used a commercial CFD software, without any turbulence model. Besides, in the presence of the natural circulation, a laminar flow has been assumed and the equations of the mass conservation, momentum and energy were solved by the finite element method. In addition, the results of the simulation are presented in terms of velocities and temperatures differences, respectively: at inlet and outlet of the heater and of the natural circulation pipe. (author)

  12. Computational simulation of the natural circulation occurring in an experimental test section of a pool type research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nascimento, Francisco R.T. do; Lima Junior, Carlos A.S.; Oliveira, Andre F.S. de; Affonso, Renato R.W.; Faccini, Jose L.H.; Moreira, Maria L., E-mail: rogerio.tdn@gmail.com, E-mail: souzalima_ca@ien.gov.br, E-mail: oliveira.afelipe@gmail.com, E-mail: raoniwa@yahoo.com.br, E-mail: faccini@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    The present work presents a computational simulation of the natural circulation phenomenon developing in an experimental test section of a pool type research reactor. The test section has been designed using a reduced scale in height 1:4.7 in relation to a pool type 30 MW research reactor prototype. It comprises a cylindrical vessel, which is opened to atmosphere, and representing the reactor pool; a natural circulation pipe, a lower plenum, and a heater containing electrical resistors in rectangular plate format, which represents the fuel elements, with a chimney positioned on the top of the resistor assembly. In the computational simulation, it was used a commercial CFD software, without any turbulence model. Besides, in the presence of the natural circulation, a laminar flow has been assumed and the equations of the mass conservation, momentum and energy were solved by the finite element method. In addition, the results of the simulation are presented in terms of velocities and temperatures differences, respectively: at inlet and outlet of the heater and of the natural circulation pipe. (author)

  13. Natural Circulation High Pressure Loop Dynamics Around Operating Point, Tests and Modelling With Retran 02

    International Nuclear Information System (INIS)

    Masriera, N.A; Doval, A.S; Mazufri, C.M

    2000-01-01

    The Natural Circulation High Pressure Loop (CAPCN) reproduces in scale all the one-dimensional thermal-hydraulic phenomena occurring in the primary loop of CAREM-25 reactor.It plays an important role in the qualification process of calculating computer codes.This facility demanded to develop several technological solutions in order to achieve the measuring and control quality required by that process.This engineering and experimental development allowed completing the first stage of dynamic tests during 1998.The trends of recorded data were systematically evaluated in terms of the deviations of main variables in response to different perturbations.By this analysis a group of eight transients was selected, providing a Minimum Representative Set (MRS) of dynamic tests, allowing the evaluation of all dynamic phenomena.Each of these transients was simulated with RETRAN-02, using a spreadsheet to facilitate the consistent elaboration and modification of input files.Comparing measured data and computer simulations, it may be concluded that it is possible to reproduce the dynamic response of all the transients with a level of approximation quite homogeneous and generally acceptable.It is possible to identify the detailed physical models that fit better the dynamic phenomena, and which of the limitations of RETRAN code are more relevant

  14. Cold-season atmospheric response to the natural variability of the Atlantic meridional overturning circulation

    Energy Technology Data Exchange (ETDEWEB)

    Gastineau, Guillaume; Frankignoul, Claude [LOCEAN/IPSL, Universite Pierre et Marie Curie, 4 place Jussieu, BP100, Paris Cedex 05 (France)

    2012-07-15

    The influence of the natural variability of the Atlantic meridional overturning circulation (AMOC) on the atmosphere is studied in multi-centennial simulations of six global climate models, using Maximum Covariance Analysis (MCA). In all models, a significant but weak influence of the AMOC changes is found during the Northern Hemisphere cold-season, when the ocean leads the atmosphere by a few years. Although the oceanic pattern slightly varies, an intensification of the AMOC is followed in all models by a weak sea level pressure response that resembles a negative phase of the North Atlantic Oscillation (NAO). The signal amplitude is typically 0.5 hPa and explains about 10% of the yearly variability of the NAO in all models. The atmospheric response seems to be due primarily due to an increase of the heat loss along the North Atlantic Current and the subpolar gyre, associated with an AMOC-driven warming. Sea-ice changes appear to be less important. The stronger heating is associated to a southward shift of the lower-tropospheric baroclinicity and a decrease of the eddy activity in the North Atlantic storm track, which is consistent with the equivalent barotropic perturbation resembling the negative phase of the NAO. This study thus provides some evidence of an atmospheric signature of the AMOC in the cold-season, which may have some implications for the decadal predictability of climate in the North Atlantic region. (orig.)

  15. Estimating steady state and transient characteristics of molten salt natural circulation loop using CFD

    Energy Technology Data Exchange (ETDEWEB)

    Kudariyawar, J.Y. [Homi Bhabha National Institue, Mumbai (India); Vaidya, A.M.; Maheshwari, K.K.; Srivastava, A.K. [Reactor Engineering Division, Bhabha Atomic Research Center, Mumbai (India); Satyamurthy, P. [ATDS, Bhabha Atomic Research Center, Mumbai (India)

    2015-03-15

    The steady state and transient characteristics of a molten salt natural circulation loop (NCL) are obtained by 3D CFD simulations. The working fluid is a mixture of NaNO{sub 3} and KNO{sub 3} in 60:40 ratio. Simulation is performed using PHOENICS CFD software. The computational domain is discretized by a body fitted grid generated using in-built mesh generator. The CFD model includes primary side. Primary side fluid is subjected to heat addition in heater section, heat loss to ambient (in piping connecting heater and cooler) and to secondary side (in cooler section). Reynolds Averaged Navier Stokes equations are solved along with the standard k-ε turbulence model. Validation of the model is done by comparing the computed steady state Reynolds number with that predicted by various correlations proposed previously. Transient simulations were carried out to study the flow initiations transients for different heater powers and different configurations. Similarly the ''power raising'' transient is computed and compared with in-house experimental data. It is found that, using detailed information obtained from 3D transient CFD simulations, it is possible to understand the physics of oscillatory flow patterns obtained in the loop under certain conditions.

  16. Experimental verification of the horizontal steam generator boil-off transfer degradation at natural circulation

    Energy Technology Data Exchange (ETDEWEB)

    Hyvaerinen, J. [Finnish Centre for Radiation and Nuclear Safety, Helsinki (Finland); Kouhia, J. [VTT Energy, Lappeenranta (Finland)

    1997-12-31

    The presentation summarises the highlights of experimental results obtained for VVER type horizontal steam generator heat transfer, primary side flow pattern, and mixing in the hot collector during secondary side boil-off with primary at single-phase natural circulation. The experiments were performed using the PACTEL facility with Large Diameter (LD) steam generator models, with collector instrumentation designed specifically for these tests. The key findings are as follows: (1) the primary to secondary heat transfer degrades as the secondary water inventory is depleted, following closely the wetted tube area; (2) a circulatory flow pattern exists in the tube bundle, resulting in reversed flow (from cold to the hot collector) in the lower part of the tube bundle, and continuous flow through the upper part, including the tubes that have already dried out; and (3) mixing of the hot leg flow entering the hot collector and reversed, cold, tube flow remains confined within the collector itself, extending only a row or two above the elevation at which tube flow reversal has taken place. 6 refs.

  17. Uncertainty correlation in stochastic safety analysis of natural circulation decay heat removal of liquid metal reactor

    International Nuclear Information System (INIS)

    Takata, Takashi; Yamaguchi, Akira

    2009-01-01

    Since various uncertainties of input variables are involved and nonlinearly-correlated in the Best Estimate (BE) plant dynamics code, it is of importance to evaluate the importance of input uncertainty to the computational results and to estimate the accuracy of the confidence level of the results. In order to estimate the importance and the accuracy, the authors have applied the stochastic safety analysis procedure using the Latin Hypercube sampling method to Liquid Metal Reactor (LMR) natural circulation Decay Heat Removal (DHR) phenomenon in the present paper. 17 input variables are chosen for the analyses and 5 influential variables, which affect the maximum coolant temperature at the core in a short period of time (several tens seconds), are selected to investigate the importance by comparing with the full-scope parametric analysis. As a result, it has been demonstrated that a comparative small number of samples is sufficient enough to estimate the dominant input variable and the confidence level. Furthermore, the influence of the sampling method on the accuracy of the upper tolerance limit (confidence level of 95%) has been examined based on the Wilks' formula. (author)

  18. A review of investigations on flow instabilities in natural circulation boiling loops

    International Nuclear Information System (INIS)

    Gonella V Durga Prasad; Manmohan Pandey; Manjeet S Kalra

    2005-01-01

    Full text of publication follows: Steam generation systems are subjected to flow instabilities due to parametric fluctuations, inlet conditions etc., which may result in mechanical vibrations of components (called flow induced vibrations) and system control problems. Analysis of these instabilities in natural circulation boiling loops is very important for the safety of nuclear reactors and other boiling systems. This paper presents the state of the art in this area by reviewing over 100 contributions made in the past 30 years. A large number of experimental and numerical investigations have been conducted to study and understand the conditions for inception of flow instabilities, parametric effects of instabilities, and the system behavior under such conditions. Work done on instabilities due to channel thermal-hydraulics as well as neutronics-thermohydraulics coupling has been reviewed. Different methods of analysis used by researchers and results obtained by them have been discussed. Various numerical techniques adopted and computer codes developed have also been discussed. The knowledge obtained from the investigations made in the past three decades has been summarized to present the state of the art of the understanding of flow instabilities. (authors)

  19. Development of natural circulation small and medium sized boiling water reactor: HSBWR-600

    International Nuclear Information System (INIS)

    Miki, Minoru; Horiuchi, Tetsuo; Yoshimoto, Yuichiro; Sumida, Isao; Murase, Michio; Akita, Minoru; Niino, Tsuyoshi

    1988-01-01

    In nuclear power generation, the development of large reactors has been promoted as the main energy source in Japan. However, world economy entered low growth age, and the growth of electric power demand slowed down. Accordingly, attention has been paid to the medium and small reactors that can cope with whatever needs by serializing their types in addition to the nuclear power plants of medium output matching to electric power demand. In order to cope with these new needs, the economical efficiency of medium and small reactors must be as close as possible to that of large reactors, and as the countermeasures to the demerits due to small size, those must be made into the plants having simplified systems and the safety easily acceptable to public. Hitachi Ltd. plans to develop the natural circulation type medium and small BWRs of 600 NWe output class, HSBWR-600, on the basis of the nuclear power plant technology based on the rich results of design and operation of BWRs obtained so far, and to rank them as one of the BWR series. The target of their development design, the circumstance of their development, the core design and the thermo-hydraulic characteristics, the reactor pressure vessel and in-core structures, the safety design, system design, building layout and the evaluation are reported. (Kako, I.)

  20. Integral nuclear power reactor with natural coolant circulation. Investigation of passive RHR system

    International Nuclear Information System (INIS)

    Samoilov, O.B.; Kuul, V.S.; Malamud, V.A.; Tarasov, G.I.

    1996-01-01

    The development of a small power (up to 240 MWe) integral PWR for nuclear co-generation power plants has been carried out. The distinctive features of this advanced reactor are: primary circuit arrangement in a single pressure vessel; natural coolant circulation; passive safety systems with self-activated control devices; use of a second (guard) vessel housing the reactor; favourable conditions for the most severe accident management. A passive steam condensing channel has been developed which is activated by the direct action of the primary circuit pressure without an automatic controlling action or manual intervention for emergency cooling of an integral reactor with an in-built pressurizer. In an emergency situation as pressure rises in the reactor a self-activated device blows out non-condensable gases from the condenser tube bundle and returns them in the steam-condensing mode of the operation with the returing primary coolant condensate into the reactor. The thermo-physical test facility is constructed and the experimental development of the steam-condensing channels is performed aiming at the verification of mathematical models for these channels operation in integral reactors both at loss-of-heat removal and LOCA accidents. (orig.)

  1. Study on natural circulation characteristics of an IPWR under inclined and rolling condition

    Energy Technology Data Exchange (ETDEWEB)

    He, Lihui [College of Computer Science and Information Technology, Harbin Normal University, Harbin (China); Wang, Bing [Fundamental Science on Nuclear Safety and Simulation Technology Laboratory, Harbin Engineering University, Harbin (China); Xia, Genglei, E-mail: xiagenglei@163.com [Fundamental Science on Nuclear Safety and Simulation Technology Laboratory, Harbin Engineering University, Harbin (China); Peng, Minjun [Fundamental Science on Nuclear Safety and Simulation Technology Laboratory, Harbin Engineering University, Harbin (China)

    2017-06-15

    Highlights: • An ocean-based thermal-hydraulic analysis code was developed based on RELAP5 codes. • The inclination condition can reduce the mass flow rate of reactor core. • The system parameters asymmetry increases with the increasing inclination angle. • Flow oscillation of different loops cancel each other due to the symmetrical arrangement of the reactor. • The off-center roll axis location can break the symmetry and enlarge fluctuation amplitude of the core flow rate. - Abstract: An ocean-based thermal-hydraulic system analysis code was developed based on RELAP5/MOD3 code by adding additional force model of ocean condition and control volume coordinate solver model. The natural circulation operation characteristics of integrated pressurized water reactor (IPWR) under ocean conditions were studied and the effects of inclination and rolling motions were analyzed. The results conclude that, the inclination condition can reduce the mass flow rate of reactor core and lead to inconsistent coolant flow rates of the left and right loops, furthermore, it affects the heat transfer of once-through steam generators (OTSGs). In the case of rolling motion, the additional pressure drop of the loop is dominated by tangential force, and flow oscillation of different loops cancel each other due to the symmetrical arrangement of the reactor. The off-center roll axis location, the combination of the inclination and rolling motion, both can break the thermal-hydraulic symmetry among different loops and enlarge fluctuation amplitude of the core flow rate.

  2. Experimental verification of the horizontal steam generator boil-off transfer degradation at natural circulation

    Energy Technology Data Exchange (ETDEWEB)

    Hyvaerinen, J [Finnish Centre for Radiation and Nuclear Safety, Helsinki (Finland); Kouhia, J [VTT Energy, Lappeenranta (Finland)

    1998-12-31

    The presentation summarises the highlights of experimental results obtained for VVER type horizontal steam generator heat transfer, primary side flow pattern, and mixing in the hot collector during secondary side boil-off with primary at single-phase natural circulation. The experiments were performed using the PACTEL facility with Large Diameter (LD) steam generator models, with collector instrumentation designed specifically for these tests. The key findings are as follows: (1) the primary to secondary heat transfer degrades as the secondary water inventory is depleted, following closely the wetted tube area; (2) a circulatory flow pattern exists in the tube bundle, resulting in reversed flow (from cold to the hot collector) in the lower part of the tube bundle, and continuous flow through the upper part, including the tubes that have already dried out; and (3) mixing of the hot leg flow entering the hot collector and reversed, cold, tube flow remains confined within the collector itself, extending only a row or two above the elevation at which tube flow reversal has taken place. 6 refs.

  3. Experimental verification of the horizontal steam generator boil-off transfer degradation at natural circulation

    International Nuclear Information System (INIS)

    Hyvaerinen, J.; Kouhia, J.

    1997-01-01

    The presentation summarises the highlights of experimental results obtained for VVER type horizontal steam generator heat transfer, primary side flow pattern, and mixing in the hot collector during secondary side boil-off with primary at single-phase natural circulation. The experiments were performed using the PACTEL facility with Large Diameter (LD) steam generator models, with collector instrumentation designed specifically for these tests. The key findings are as follows: (1) the primary to secondary heat transfer degrades as the secondary water inventory is depleted, following closely the wetted tube area; (2) a circulatory flow pattern exists in the tube bundle, resulting in reversed flow (from cold to the hot collector) in the lower part of the tube bundle, and continuous flow through the upper part, including the tubes that have already dried out; and (3) mixing of the hot leg flow entering the hot collector and reversed, cold, tube flow remains confined within the collector itself, extending only a row or two above the elevation at which tube flow reversal has taken place

  4. Core Power Limits For A Lead-Bismuth Natural Circulation Actinide Burner Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Davis, Cliff Bybee; Kim, D.; Todreas, N. E.; Mujid S. Kazimi

    2002-04-01

    The Idaho National Engineering and Environmental Laboratory and Massachusetts Institute of Technology are investigating the suitability of lead-bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The design being considered here is a pool type reactor that burns actinides and utilizes natural circulation of the primary coolant, a conventional steam power conversion cycle, and a passive decay heat removal system. Thermal-hydraulic evaluations of the actinide burner reactor were performed to determine allowable core power ratings that maintain cladding temperatures below corrosion-established temperature limits during normal operation and following a loss-of-feedwater transient. An economic evaluation was performed to optimize various design parameters by minimizing capital cost. The transient power limit was initially much more restrictive than the steady-state limit. However, enhancements to the reactor vessel auxiliary cooling system for transient decay heat removal resulted in an increased power limit of 1040 MWt, which was close to the steady-state limit. An economic evaluation was performed to estimate the capital cost of the reactor and its sensitivity to the transient power limit. For the 1040 MWt power level, the capital cost estimate was 49 mills per kWhe based on 1999 dollars.

  5. Experimental study of flow instability and CHF in a natural circulation system with subcooled boiling

    International Nuclear Information System (INIS)

    Yang, R.C.; Shi, D.Q.; Lu, Z.Q.; Zheng, R.C.; Wang, Y.

    1996-01-01

    Experimental study has been performed to investigate flow instability and critical heat flux (CHF) in a natural circulation system with subcooled boiling. In the experiments three kinds of heated sections were used. Freon-12 was used as the working medium. The experiments show which one of the two phenomena, flow instability and CHF condition, may first occur in the system depends on not only the heat input power to the heated section and the parameters of the working medium, but also the construction of the heated section. The occurrence of the flow instability mainly depends on the total heat input power to the heated section and the CHF condition is mainly caused by the local heat flux of the heated section. In the experiments two kinds of flow instability, flow instability with high frequency and flow instability with low frequency, were found. But they all belong to density wave instability. The influence of the parameters of the working medium on the onset of the flow instability and CHF condition in the system were investigated. The stability boundaries were determined through the experiments. By means of dimensional analysis of integral equations, a common correlation describing the threshold condition of onset of the flow instability was obtained

  6. Thermo-hydraulic instability of natural circulation BWRs at low pressure star-up. Experimental estimation of instability region with test facility considering scaling law

    International Nuclear Information System (INIS)

    Inada, F.; Furuya, M.; Yasuo, A.; Tabata, H.; Yoshioka, Y.; Kim, H.T.

    1995-01-01

    In natural circulation BWRs developed for advanced light water reactors with simplified passive safety systems, thermo-hydraulic stability should be confirmed especially at low pressure start-up. In this paper, nondimensional parameters to estimate the hydrodynamic stability to reactors at low pressure start-up were obtained by transformation of the basic equations of drift-flux model in the two-phase region into nondimensional form. A test facility based on these parameters was then constructed. The height of the test facility is 70% of SBWR and many nondimensional test facility parameters are almost the same as those of the reactor. Reactor stability was estimated experimentally. Stability maps below 0.5MPa were obtained on the heat flux - channel inlet subcooling place. It was found that there were two stability boundaries, between which the flow became unstable. Flow was stable in the high and low channel inlet subcooling regions. Typical conditions of SBWR at low pressure start-up were noted in the high channel inlet subcooling stable region. The heat flux at typical SBWR start-up was about one fifth that of the stability boundary. Though some nondimensional parameters of the test facility did not exactly agree with those of SBWR, it was suggested that the flow in SBWR was stable below 0.5MPa because of the large margin. (author)

  7. Analytical evaluation on dynamical response characteristics of reduced-moderation water reactor with tight-lattice core under natural circulation core cooling

    International Nuclear Information System (INIS)

    Ishikawa, Nobuyuki; Okubo, Tsutomu

    2009-01-01

    The time-domain analyses with TRAC-BF1 code were performed for clarifying the dynamical response characteristics of the reduced-moderation water reactor (RMWR) with tight-lattice core configuration. The response characteristics were evaluated based on the step response basically utilized for dynamical system evaluation. As for the most fundamental dynamical characteristics, the channel flow response characteristics of single fuel assembly were evaluated. In the evaluation, the appropriate single-phase pressure drop setting at the inlet orifice was determined in terms of response stability from the design viewpoint. In addition, from the investigation on the relation of the response and transit time of coolant, it is confirmed that the channel flow response of RMWR is dominated by the transit time of vapor phase resulting from a high void fraction operation condition. As for a natural circulation flow response, it is clarified that the response is strongly influenced by the effect of two-phase pressure loss owing to a high void fraction condition. The reactor power response with reactivity feedback shows quite stable response characteristics on account of the small absolute value of void reactivity coefficient.

  8. Neutron radiography experiments for verification of soluble boron mixing and transport modeling under natural circulation conditions

    International Nuclear Information System (INIS)

    Morlang, M.M.; Feltus, M.A.

    1996-01-01

    The use of neutron radiography for visualization of fluid flow through flow visualization modules has been very successful. Current experiments at the Penn State Breazeale Reactor serve to verify the mixing and transport of soluble boron under natural flow conditions as would be experienced in a pressurized water reactor. Different flow geometries have been modeled including holes, slots, and baffles. Flow modules are constructed of aluminum box material 1 1/2 inches by 4 inches in varying lengths. An experimental flow system was built which pumps fluid to a head tank and natural circulation flow occurs from the head tank through the flow visualization module to be radio-graphed. The entire flow system is mounted on a portable assembly to allow placement of the flow visualization module in front of the neutron beam port. A neutron-transparent fluor-inert fluid is used to simulate water at different densities. Boron is modeled by gadolinium oxide powder as a tracer element, which is placed in a mixing assembly and injected into the system a remotely operated electric valve, once the reactor is at power. The entire sequence is recorded on real-time video. Still photographs are made frame-by-frame from the video tape. Computers are used to digitally enhance the video and still photographs. The data obtained from the enhancement will be used for verification of simple geometry predictions using the TRAC and RELAP thermal-hydraulic codes. A detailed model of a reactor vessel inlet plenum, downcomer region, flow distribution area and core inlet is being constructed to model the APGOO plenum. Successive radiography experiments of each section of the model under identical conditions will provide a complete vessel / core model for comparison with the thermal-hydraulic codes

  9. Neutron radiography experiments for verification of soluble boron mixing and transport modeling under natural circulation conditions

    International Nuclear Information System (INIS)

    Feltus, M.A.; Morlang, G.M.

    1996-01-01

    The use of neutron radiography for visualization of fluid flow through flow visualization modules has been very successful. Current experiments at the Penn State Breazeale Reactor serve to verify the mixing and transport of soluble boron under natural flow conditions as would be experienced in a pressurized water reactor. Different flow geometries have been modeled including holes, slots, and baffles. Flow modules are constructed of aluminum box material 1 1/2 inches by 4 inches in varying lengths. An experimental flow system was built which pumps fluid to a head tank and natural circulation flow occurs from the head tank through the flow visualization module to be radiographed. The entire flow system is mounted on a portable assembly to allow placement of the flow visualization module in front of the neutron beam port. A neutron-transparent fluorinert fluid is used to simulate water at different densities. Boron is modeled by gadolinium oxide powder as a tracer element, which is placed in a mixing assembly and injected into the system by remote operated electric valve, once the reactor is at power. The entire sequence is recorded on real-time video. Still photographs are made frame-by-frame from the video tape. Computers are used to digitally enhance the video and still photographs. The data obtained from the enhancement will be used for verification of simple geometry predictions using the TRAC and RELAP thermal-hydraulic codes. A detailed model of a reactor vessel inlet plenum, downcomer region, flow distribution area and core inlet is being constructed to model the AP600 plenum. Successive radiography experiments of each section of the model under identical conditions will provide a complete vessel/core model for comparison with the thermal-hydraulic codes

  10. Online sequential condition prediction method of natural circulation systems based on EOS-ELM and phase space reconstruction

    International Nuclear Information System (INIS)

    Chen, Hanying; Gao, Puzhen; Tan, Sichao; Tang, Jiguo; Yuan, Hongsheng

    2017-01-01

    Highlights: •An online condition prediction method for natural circulation systems in NPP was proposed based on EOS-ELM. •The proposed online prediction method was validated using experimental data. •The training speed of the proposed method is significantly fast. •The proposed method can achieve good accuracy in wide parameter range. -- Abstract: Natural circulation design is widely used in the passive safety systems of advanced nuclear power reactors. The irregular and chaotic flow oscillations are often observed in boiling natural circulation systems so it is difficult for operators to monitor and predict the condition of these systems. An online condition forecasting method for natural circulation system is proposed in this study as an assisting technique for plant operators. The proposed prediction approach was developed based on Ensemble of Online Sequential Extreme Learning Machine (EOS-ELM) and phase space reconstruction. Online Sequential Extreme Learning Machine (OS-ELM) is an online sequential learning neural network algorithm and EOS-ELM is the ensemble method of it. The proposed condition prediction method can be initiated by a small chunk of monitoring data and it can be updated by newly arrived data at very fast speed during the online prediction. Simulation experiments were conducted on the data of two natural circulation loops to validate the performance of the proposed method. The simulation results show that the proposed predication model can successfully recognize different types of flow oscillations and accurately forecast the trend of monitored plant variables. The influence of the number of hidden nodes and neural network inputs on prediction performance was studied and the proposed model can achieve good accuracy in a wide parameter range. Moreover, the comparison results show that the proposed condition prediction method has much faster online learning speed and better prediction accuracy than conventional neural network model.

  11. The effects of phototherapy on the numbers of circulating natural killer cells and T lymphocytes in psoriasis.

    LENUS (Irish Health Repository)

    Tobin, A M

    2009-04-01

    The innate immune system is believed to be important in the pathogenesis of psoriasis and natural killer (NK) have been found in increased numbers in psoriatic plaques. Alterations in the numbers of NK cells in peripheral blood have been reported. We investigated the effect of phototherapy on levels of peripheral NK cells and lymphocytes in patients with psoriasis. In nine patients whom we followed before, during and after narrowband ultraviolet B (UVB) treatment there were no differences in the numbers of circulating lymphocytes, lymphocyte subsets or cells expressing NK markers and controls. Treatment with narrowband UVB did, however, significantly lower circulating CD4 counts which gradually recovered posttreatment.

  12. Coupling two-phase fluid flow with two-phase darcy flow in anisotropic porous media

    KAUST Repository

    Chen, J.

    2014-06-03

    This paper reports a numerical study of coupling two-phase fluid flow in a free fluid region with two-phase Darcy flow in a homogeneous and anisotropic porous medium region. The model consists of coupled Cahn-Hilliard and Navier-Stokes equations in the free fluid region and the two-phase Darcy law in the anisotropic porous medium region. A Robin-Robin domain decomposition method is used for the coupled Navier-Stokes and Darcy system with the generalized Beavers-Joseph-Saffman condition on the interface between the free flow and the porous media regions. Obtained results have shown the anisotropic properties effect on the velocity and pressure of the two-phase flow. 2014 Jie Chen et al.

  13. Coupling Two-Phase Fluid Flow with Two-Phase Darcy Flow in Anisotropic Porous Media

    Directory of Open Access Journals (Sweden)

    Jie Chen

    2014-06-01

    Full Text Available This paper reports a numerical study of coupling two-phase fluid flow in a free fluid region with two-phase Darcy flow in a homogeneous and anisotropic porous medium region. The model consists of coupled Cahn-Hilliard and Navier-Stokes equations in the free fluid region and the two-phase Darcy law in the anisotropic porous medium region. A Robin-Robin domain decomposition method is used for the coupled Navier-Stokes and Darcy system with the generalized Beavers-Joseph-Saffman condition on the interface between the free flow and the porous media regions. Obtained results have shown the anisotropic properties effect on the velocity and pressure of the two-phase flow.

  14. Two-phase flow and heat transfer under low gravity

    Science.gov (United States)

    Frost, W.

    1981-01-01

    Spacelab experiment to investigate two-phase flow patterns under gravity uses a water-air mixture experiment. Air and water are circulated through the system. The quality or the mixture or air-water is controlled. Photographs of the test section are made and at the same time pressure drop across the test section is measured. The data establishes a flow regime map under reduced gravity conditions with corresponding pressure drop correlations. The test section is also equipped with an electrical resistance heater in order to allow a flow boiling experiment to be carried out using Freon II. High-speed photographs of the test section are used to determine flow patterns. The temperature gradient and pressure drop along the duct can be measured. Thus, quality change can be measured, and heat transfer calculated.

  15. Two-phase flow dynamics in ECC

    International Nuclear Information System (INIS)

    Albraaten, P.J.

    1981-07-01

    The present report summarizes the achievements within the project ''Two-phase Systems and ECC''. The results during 1978 - 1980 are accounted for in brief as they have been documented in earlier reports. The results during the first half of 1981 are accounted for in greater detail. They contain a new model for the Basset force and test runs with this model using the test code RISQUE. Furthermore, test runs have been performed with TRAC-PD2 MOD 1. This code was implemented on Edwards Pipe Blowdown experiment (a standard test case) and UC-Berkeley Reflooding experiment (a non-standard test case.) (Auth.)

  16. Two phase cooling for superconducting magnets

    International Nuclear Information System (INIS)

    Eberhard, P.H.; Gibson, G.A.; Green, M.A.; Ross, R.R.; Smits, R.G.; Taylor, J.D.; Watt, R.D.

    1985-08-01

    A closed circuit tubular cooling system for superconducting magnets offers advantages of limiting boiloff and containing high pressures during quenches. Proper location of automatic valves to lower pressures and protect the refrigerator in the event of quenches is described. Theoretical arguments and exprimental evidence are given against a previously suggested method to determine He two phase flow regimes. If loss of flow occurs due to some types of refrigeration failure and transfer lines have enough heat leak to warm up, quenches are induced when the flow is restored. Examples are taken from experience with the TPC magnet

  17. Stability monitoring of a natural-circulation-cooled boiling water reactor

    International Nuclear Information System (INIS)

    Hagen, T.H.J.J. van der.

    1989-01-01

    Methods for monitoring the stability of a boiling water reactor (BWR) are discussed. Surveillance of BWR stability is of importance as problems were encountered in several large reactors. Moreover, surveying stability allows plant owners to operate at high power with acceptable stability margins. The results of experiments performed on the Dodewaard BWR (the Netherlands) are reported. This type reactor is cooled by natural circulation, a cooling principle that is also being considered for new reactor designs. The stability of this reactor was studied both with deterministic methods and by noise analysis. Three types of stability are distinguished and were investigated separately: reactor-kinetic stability, thermal-hydraulic stability and total-plant stability. It is shown that the Dodewaard reactor has very large stability margins. A simple yet reliable stability criterion is introduced. It can be derived on-line from thhe noise signal of ex-vessel neutron detectors during normal operation. The sensitivity of neutron detectors to in-core flux perturbations - reflected in the field-of-view of the detector - was calculated in order to insure proper stability surveillance. A novel technique is presented which enables the determination of variations of the in-core coolant velocity by noise correlation. The velocity measured was interpreted on the basis of experiments performed on the air/water flow in a model of a BWR coolant channel. It appeared from this analysis that the velocity measured was much higher than the volume-averaged water and air velocities and the volumetric flux. The applicability of the above-mentioned technique to monitoring of local channel-flow stability was tested. It was observed that stability effects on the coolant velocity are masked by other effects originating from the local flow pattern. Experimental and theoretical studies show a shorter effective fuel time constant in a BWR than was assumed. (author). 118 refs.; 73 figs.; 21 tabs

  18. Boron dilution transients during natural circulation flow in PWR-Experiments and CFD simulations

    Energy Technology Data Exchange (ETDEWEB)

    Hoehne, Thomas [Forschungszentrum Dresden-Rossendorf (FZD)-Institute of Safety Research, P.O. Box 510119, D-01314 Dresden (Germany)], E-mail: T.Hoehne@fzd.de; Kliem, Soeren; Rohde, Ulrich; Weiss, Frank-Peter [Forschungszentrum Dresden-Rossendorf (FZD)-Institute of Safety Research, P.O. Box 510119, D-01314 Dresden (Germany)

    2008-08-15

    Coolant mixing in the cold leg, downcomer and the lower plenum of pressurized water reactors is an important phenomenon mitigating the reactivity insertion into the core. Therefore, mixing of the de-borated slugs with the ambient coolant in the reactor pressure vessel was investigated at the four loops 1:5 scaled Rossendorf coolant mixing model (ROCOM) mixing test facility. In particular thermal hydraulics analyses have shown, that weakly borated condensate can accumulate in the pump loop seal of those loops, which do not receive a safety injection. After refilling of the primary circuit, natural circulation in the stagnant loops can re-establish simultaneously and the de-borated slugs are shifted towards the reactor pressure vessel (RPV). In the ROCOM experiments, the length of the flow ramp and the initial density difference between the slugs and the ambient coolant was varied. From the test matrix experiments with 0 resp. 2% density difference between the de-borated slugs and the ambient coolant were used to validate the CFD software ANSYS CFX. To model the effects of turbulence on the mean flow a higher order Reynolds stress turbulence model was employed and a mesh consisting of 6.4 million hybrid elements was utilized. Only the experiments and CFD calculations with modeled density differences show stratification in the downcomer. Depending on the degree of density differences the less dense slugs flow around the core barrel at the top of the downcomer. At the opposite side, the lower borated coolant is entrained by the colder safety injection water and transported to the core. The validation proves that ANSYS CFX is able to simulate appropriately the flow field and mixing effects of coolant with different densities.

  19. Microgravity Two-Phase Flow Transition

    Science.gov (United States)

    Parang, M.; Chao, D.

    1999-01-01

    Two-phase flows under microgravity condition find a large number of important applications in fluid handling and storage, and spacecraft thermal management. Specifically, under microgravity condition heat transfer between heat exchanger surfaces and fluids depend critically on the distribution and interaction between different fluid phases which are often qualitatively different from the gravity-based systems. Heat transfer and flow analysis in two-phase flows under these conditions require a clear understanding of the flow pattern transition and development of appropriate dimensionless scales for its modeling and prediction. The physics of this flow is however very complex and remains poorly understood. This has led to various inadequacies in flow and heat transfer modeling and has made prediction of flow transition difficult in engineering design of efficient thermal and flow systems. In the present study the available published data for flow transition under microgravity condition are considered for mapping. The transition from slug to annular flow and from bubbly to slug flow are mapped using dimensionless variable combination developed in a previous study by the authors. The result indicate that the new maps describe the flow transitions reasonably well over the range of the data available. The transition maps are examined and the results are discussed in relation to the presumed balance of forces and flow dynamics. It is suggested that further evaluation of the proposed flow and transition mapping will require a wider range of microgravity data expected to be made available in future studies.

  20. Turbulence in two-phase flows

    International Nuclear Information System (INIS)

    Sullivan, J.P.; Houze, R.N.; Buenger, D.E.; Theofanous, T.G.

    1981-01-01

    Hot film Anemometry and Laser Doppler Velocimetry have been employed in this work to study the turbulence characteristics of Bubbly and Stratified two-phase flows, respectively. Extensive consistency checks were made to establish the reliability and hence the utility of these experimental techniques for the measurement of turbulence in two-phase flows. Buoyancy-driven turbulence in vertical bubbly flows has been identified experimentally and correlated in terms of a shear velocity superposition approach. This approach provides a criterion for the demarcation of the buoyancy-driven turbulence region from the wall shear-generated turbulence region. Our data confirm the roughly isotropic behavior expected for buoyancy-driven turbulence. Upgrading of our experimental system will permit investigations of the wall-shear dominated regime (i.e., isotropy, superposition approach, etc.). The stratified flow data demonstrate clearly that the maximum in the mean velocity profile does not coincide with the zero shear plane, indicating the existence of a negative eddy viscosity region. Previous studies do not take into account this difference and thus they yield incorrect friction factor data in addition to certain puzzling behavior in the upper wall region. The conditioned turbulence data in the wavy region indicate interesting trends and that an appropriate normalization of intensities must take into account the shear velocity at the interfacial (wavy) region

  1. Two Phase Flow Simulation Using Cellular Automata

    International Nuclear Information System (INIS)

    Marcel, C.P.

    2002-01-01

    The classical mathematical treatment of two-phase flows is based on the average of the conservation equations for each phase.In this work, a complementary approach to the modeling of these systems based on statistical population balances of aut omata sets is presented.Automata are entities defined by mathematical states that change following iterative rules representing interactions with the neighborhood.A model of automata for two-phase flow simulation is presented.This model consists of fie lds of virtual spheres that change their volumes and move around a certain environment.The model is more general than the classical cellular automata in two respects: the grid of cellular automata is dismissed in favor of a trajectory generator, and the rules of interaction involve parameters representing the actual physical interactions between phases.Automata simulation was used to study unsolved two-phase flow problems involving high heat flux rates. One system described in this work consists of a vertical channel with saturated water at normal pressure heated from the lower surface.The heater causes water to boil and starts the bubble production.We used cellular automata to describe two-phase flows and the interaction with the heater.General rule s for such cellular automata representing bubbles moving in stagnant liquid were used, with special attention to correct modeling of different mechanisms of heat transfer.The results of the model were compared to previous experiments and correlations finding good agreement.One of the most important findings is the confirmation of Kutateladze's idea about a close relation between the start of critical heat flux and a change in the flow's topology.This was analyzed using a control volume located in the upper surface of the heater.A strong decrease in the interfacial surface just before the CHF start was encountered.The automata describe quite well some characteristic parameters such as the shape of the local void fraction in the

  2. Experimental investigation of effect of spacer on two phase turbulent mixing rate in subchannels of pressure tube type BWR

    Energy Technology Data Exchange (ETDEWEB)

    Verma, Shashi Kant; Sinha, S.L. [National Institute of Technology, Raipur (India). Mechanical Engineering Dept.; Chandraker, D.K. [Bhabha Atomic Research Centre, Mumbai (India). Reactor Design and Development Group

    2017-11-15

    Turbulent mixing rate between adjacent subchannels in a two-phase flow has been known to be strongly dependent on the flow pattern. The most important aspect of turbulent motion is that the velocity and pressure at a fixed point do not remain constant with time even in steady state but go through very irregular high frequency fluctuations. These fluctuations influence the diffusion of scalar and vector quantities. The Advanced Heavy Water Reactor (AHWR) is a vertical pressure tube type, heavy water moderated and boiling light water cooled natural circulation based reactor. The fuel bundle of AHWR contains 54 fuel rods set in three concentric rings of 12, 18 and 24 fuel rods. This fuel bundle is divided into number of imaginary interacting flow channel called subchannels. Alteration from single phase to two phase flow situation occurs in reactor rod bundle with raise in power. The two phase flow regimes like bubbly, slug-churn, and annular flow are generally encountered in reactor rod bundle. Prediction of thermal margin of the reactor has necessitated the investigation of turbulent mixing rate of coolant between these subchannels under these flow regimes. Thus, it is fundamental to estimate the effect of spacer grids on turbulent mixing between subchannels of AHWR rod bundle.

  3. Numerical simulation of the two-phase flows in a hydraulic coupling by solving VOF model

    International Nuclear Information System (INIS)

    Luo, Y; Zuo, Z G; Liu, S H; Fan, H G; Zhuge, W L

    2013-01-01

    The flow in a partially filled hydraulic coupling is essentially a gas-liquid two-phase flow, in which the distribution of two phases has significant influence on its characteristics. The interfaces between the air and the liquid, and the circulating flows inside the hydraulic coupling can be simulated by solving the VOF two-phase model. In this paper, PISO algorithm and RNG k–ε turbulence model were employed to simulate the phase distribution and the flow field in a hydraulic coupling with 80% liquid fill. The results indicate that the flow forms a circulating movement on the torus section with decreasing speed ratio. In the pump impeller, the air phase mostly accumulates on the suction side of the blades, while liquid on the pressure side; in turbine runner, air locates in the middle of the flow passage. Flow separations appear near the blades and the enclosing boundaries of the hydraulic coupling

  4. Interfacial structures in downward two-phase bubbly flow

    International Nuclear Information System (INIS)

    Paranjape, S.S.; Kim, S.; Ishii, M.; Kelly, J.

    2003-01-01

    Downward two-phase flow was studied considering its significance in view of Light Water Reactor Accidents (LWR) such as Loss of Heat Sink (LOHS) by feed water loss or secondary pipe break. The flow studied, was an adiabatic, air-water, co-current, vertically downward two-phase flow. The experimental test sections had internal hydraulic diameters of 25.4 mm and 50.8 mm. Flow regime map was obtained using the characteristic signals obtained from an impedance void meter, employing neural network based identification methodology to minimize the subjective judgment in determining the flow regimes. A four sensor conductivity probe was used to measure the local two phase flow parameters, which characterize the interfacial structures. The local time averaged two-phase flow parameters measured were: void fraction (α), interfacial area concentration (a i ), bubble velocity (v g ), and Sauter mean diameter (D Sm ). The flow conditions were from the bubbly flow regime. The local profiles of these parameters as well as their axial development revealed the nature of the interfacial structures and the bubble interaction mechanisms occurring in the flow. Furthermore, this study provided a good database for the development of the interfacial area transport equation, which dynamically models the changes in the interfacial area along the flow field. An interfacial area transport equation was developed for downward flow based on that developed for the upward flow, with certain modifications in the bubble interaction terms. The area averaged values of the interfacial area concentration were compared with those predicted by the interfacial area transport model. (author)

  5. Modeling of two-phase slug flow

    International Nuclear Information System (INIS)

    Fabre, J.; Line, A.

    1992-01-01

    When gas and liquid flow in a pipe, over a range of flow rates, a flow pattern results in which sequences of long bubbles, almost filling the pipe cross section, are successively followed by liquid slugs that may contain small bubbles. This flow pattern, usually called slug flow, is encountered in numerous practical situations, such as in the production of hydrocarbons in wells and their transportation in pipelines; the production of steam and water in geothermal power plants; the boiling and condensation in liquid-vapor systems of thermal power plants; emergency core cooling of nuclear reactors; heat and mass transfer between gas and liquid in chemical reactors. This paper provides a review of two phase slug flow modeling

  6. A study of thermal-hydraulic requirements for increasing the power rates for natural-circulation boiling water reactors

    International Nuclear Information System (INIS)

    Yasuo, A.; Inada, F.; Hidaka, M.

    1992-01-01

    In this paper, the feasibility of higher power rates for natural-circulation boiling water reactors (BWRs) is studied with the objective of examining the flexibility of the plant power rate in constructing such plants to cope with the increasing demand for electricity. By applying existing one-dimensional design codes, the riser heights necessary to meet two major thermal-hydraulic requirements, i.e., critical power and core stability, are systematically calculated. Several restrictions on the maximum diameter and height of the pressure vessel are also considered because these restrictions could make construction impossible or drastically increase the construction costs. A very simple map of the dominant parameters for higher power rates is obtained. It is concluded that natural-circulation BWRs of >1000 MW (electric) will be feasible within the restrictions considered here

  7. Neutronic design for a 100MWth Small modular natural circulation lead or lead-alloy cooled fast reactors core

    International Nuclear Information System (INIS)

    Chen, C.; Chen, H.; Zhang, H.; Chen, Z.; Zeng, Q.

    2015-01-01

    Lead or lead-alloy cooled fast reactor with good fuel proliferation and nuclear waste transmutation capability, as well as high security and economy, is a great potential for the development of fourth-generation nuclear energy systems. Small natural circulation reactor is an important technical route lead cooled fast reactors industrial applications, which has been chosen as one of the three reference technical for solution lead or lead-alloy cooled fast reactors by GIF lead-cooled fast reactor steering committee. The School of Nuclear Science and Technology of USTC proposed a small 100MW th natural circulation lead cooled fast reactor concept called SNCLFR-100 based realistic technology. This article describes the SNCLFR-100 reactor of the overall technical program, core physics calculation and analysis. The results show that: SNCLFR-100 with good neutronic and safety performance and relevant design parameters meet the security requirements with feasibility. (author)

  8. Natural circulation cooldown analysis for Yonggwang 3 and 4 per US NRC BTP RSB 5-1 requirements

    International Nuclear Information System (INIS)

    Seo, J.T.; Ko, C.S.; Ro, T.S.; Simoni, L.P.

    2004-01-01

    The Natural Circulation Cooldown (NCC) analysis from normal operations to shutdown cooling entry conditions for Yonggwang units 3 and 4 (YGN 3 and 4) was performed within the requirements of U.S. Nuclear Regulatory Commission (NRC) Branch Technical Position (BTP) RSB 5-1. The results showed that the YGN 3 and 4 can be cooled and depressurized to the shutdown entry conditions (350 deg F, 410 psia) within 16 hours under natural circulation condition requiring only 78% of the minimum condensate water storage capacity in conformance with BTP RSB 5-1 requirements. The results also demonstrated that the safety grade Reactor Coolant Gas Vent System (RCGVS) has sufficient capacity for the RCS depressurization as well as for the steam void control in the reactor vessel upper head region. (author)

  9. Numerical analysis of the fluid dynamics in a natural circulation loop; Analise numerica da dinamica do escoamento em circuitos de circulacao natural

    Energy Technology Data Exchange (ETDEWEB)

    Angelo, Gabriel

    2013-07-01

    Natural circulation loops apply to many engineering applications such as: water heating solar energy system (thermo-siphons), thermal management of electrical components (voltage converter), geothermal energy, nuclear reactors, etc. In pressurized water nuclear reactors, known as PWR's, the natural circulation loops are employed to ensure passive safety. In critical situations, the heat transfer will occur only by natural convection, without any external control or mechanical devices. This feature is desired and has been considered in modern nuclear reactor projects. This work consists of a numerical study of the natural circulation loop, located at the Instituto de Pesquisas Energeticas e Nucleares / Comissao Nacional de Energia Nuclear in Sao Paulo, Brazil, in order to establish the flow pattern in single phase conditions. The comparison of numerical results to experiments in transient condition revealed significant deviations for the Zero Equation turbulence model. Intermediate deviations for the Eddy Viscosity Turbulence Equation (EVTE), k - {omega}, SST e SSG models. And the best results are obtained by the k - {epsilon} e DES models (with better results for the k - {epsilon} model). (author)

  10. Computational simulation of flow and heat transfer in single-phase natural circulation loops; Simulacao computacional de escoamento e transferencia de calor em circuitos de circulacao natural monofasica

    Energy Technology Data Exchange (ETDEWEB)

    Pinheiro, Larissa Cunha

    2017-07-01

    Passive decay heat removal systems based on natural circulation are essential assets for the new Gen III+ nuclear power reactors and nuclear spent fuel pools. The aim of the present work is to study both laminar and turbulent flow and heat transfer in single-phase natural circulation systems through computational fluid dynamics simulations. The working fluid is considered to be incompressible with constant properties. In the way, the Boussinesq Natural Convection Hypothesis was applied. The model chosen for the turbulence closure problem was the k -- εThe commercial computational fluid dynamics code ANSYS CFX 15.0 was used to obtain the numerical solution of the governing equations. Two single-phase natural circulation circuits were studied, a 2D toroidal loop and a 3D rectangular loop, both with the same boundary conditions of: prescribed heat flux at the heater and fixed wall temperature at the cooler. The validation and verification was performed with the numerical data provided by DESRAYAUD et al. [1] and the experimental data provided by MISALE et al. [2] and KUMAR et al. [3]. An excellent agreement between the Reynolds number (Re) and the modified Grashof number (Gr{sub m}), independently of Prandtl Pr number was observed. However, the convergence interval was observed to be variable with Pr, thus indicating that Pr is a stability governing parameter for natural circulation. Multiple steady states was obtained for Pr = 0,7. Finally, the effect of inclination was studied for the 3D circuit, both in-plane and out-of-plane inclinations were verified for the steady state laminar regime. As a conclusion, the Re for the out-of-plane inclination was in perfect agreement with the correlation found for the zero inclination system, while for the in-plane inclined system the results differ from that of the corresponding vertical loop. (author)

  11. Two-phase flow models in unbounded two-phase critical flows

    International Nuclear Information System (INIS)

    Celata, G.P.; Cumo, M.; Farello, G.E.

    1985-01-01

    With reference to a Loss-of-Coolant Accident in Light Water Reactors, an analysis of the unbounded two-phase critical flow (i.e. the issuing two-phase jet) has been accomplished. Considering jets external shape, obtained by means of photographic pictures; pressure profiles inside the jet, obtained by means of a movable ''Pitot;'' and jet phases distribution information, obtained by means of X-rays pictures; a characterization of the flow pattern in the unbounded region of a two-phase critical flow is given. Jets X-ray pictures show the existence of a central high density ''core'' gradually evaporating all around, which gives place to a characteristic ''dartflow'' the length of which depends on stagnation thermodynamic conditions

  12. Heat transfer experiments and correlations for natural and forced circulations of water in rod bundles at low Reynolds numbers

    International Nuclear Information System (INIS)

    Kim, Sung-Ho; El-Genk, Mohamed S.; Rubio, Reuben A.; Bryson, James W.; Foushee, Fabian C.

    1988-01-01

    Experimental heat transfer studies were conducted for fully developed forced and natural flows of water through seven uniformly heated rod bundles, triangularly arrayed with P/D = 1.25, 1.38, and 1.5. In forced circulation experiments, Re ranged from 80 to 50,000 and Pr from 3 to 8.5, while in natural circulation, Re varied from 260 to 2,000, and Ra q from 8 x 10 8 to 2.5 x 10 8 . The forced flow data fell into the two basic flow regimes: turbulent and laminar flow. At the transition between these regimes, Re, which varied from 2,200 for P/D = 1.25 to 5,500 for P/D = 1.5, increased linearly with P/D. The heat transfer data for turbulent flow was within ±15 percent of Weisman's correlation, which was developed for fully developed turbulent flow in rod bundles at Re > 25,000. The laminar flow data showed the dependence of Nu on Re to be weaker than that for turbulent flow, but the exponent of Re increased with P/D: Nu = A Re B Pr 1/3 , where A is equal to 1.061, 0.511, and 0.346 for P/D = 1.25, 1.38 and 1.5, respectively, and B is a linear function of P/D (B = 0.797 P/D - 0.656). Natural circulation data indicated that rod spacing only slightly affected heat transfer, and Nu increased proportionally to Ra 0.25 ; Nu = 0.272 Ra q 0.25 . The application of the results to SNL's ACRR indicated that although the core is cooled by natural convection, either the natural circulation correlation or the forced turbulent flow correlation can be used to accurately predict the single phase heat transfer coefficient in the ACRR. These results were concluded because of the high Rayleigh and Reynolds numbers in the ACRR. The ACRR operates near the boundary between mixed and forced turbulent flow regimes: consequently, achieving the high heat transfer coefficient was possible with natural circulation. (author)

  13. Numerical flow analyses of a two-phase hydraulic coupling

    Energy Technology Data Exchange (ETDEWEB)

    Hur, N.; Kwak, M.; Moshfeghi, M. [Sogang University, Seoul (Korea, Republic of); Chang, C.-S.; Kang, N.-W. [VS Engineering, Seoul (Korea, Republic of)

    2017-05-15

    We investigated flow characteristics in a hydraulic coupling at different charged water conditions and speed ratios. Hence, simulations were performed for three-dimensional two-phase flow by using the VOF method. The realizable k-ε turbulence model was adopted. To resolve the interaction of passing blades of the primary and secondary wheels, simulations were conducted in the unsteady framework using a sliding grid technique. The results show that the water-air distribution inside the wheel is strongly dependent upon both amount of charged water and speed ratio. Generally, air is accumulated in the center of the wheel, forming a toroidal shape wrapped by the circulating water. The results also show that at high speed ratios, the solid-body-like rotation causes dry areas on the periphery of the wheels and, hence, considerably decreases the circulating flow rate and the transmitted torque. Furthermore, the momentum transfer was investigated through the concept of a mass flux triangle based on the local velocity multiplied by the local mixture density instead of the velocity triangle commonly used in a single-phase turbomachine analysis. Also, the mass fluxes along the radius of the coupling in the partially charged and fully charged cases were found to be completely different. It is shown that the flow rate at the interfacial plane and also the transmitted torque are closely related and are strongly dependent upon both the amount of charged water and speed ratio. Finally, a conceptual categorization together with two comprehensive maps was provided for the torque transmission and also circulating flow rates. These two maps in turn exhibit valuable engineering information and can serve as bases for an optimal design of a hydraulic coupling.

  14. Kinetics of two phase fuel reflected reactors

    International Nuclear Information System (INIS)

    Buzano, M.L.; Corno, S.E.; Mattioda, F.

    2000-01-01

    In the present work a self-consistent mathematical model for the local dynamics of a quite particular class of fission reactors has been developed and solved. These devices consist of an innermost multiplying region, in which a significant fraction of the fissile fuel is diluted into a liquid phase, while the complementary fuel fraction operates as a standing solid matrix. This unconventional active region is surrounded by a standard peripheral reflector. For cooling purposes, the fluid fraction of the fuel needs to be circulated through external heat exchangers. The pump driven circulation causes the delayed neutron precursors, dissolved inside the fluid phase, to be spatially homogenized in the core volume well before decaying, while a continuous removal of precursor nuclei from the core takes place as a consequence of the outside circulation. Furthermore, the fraction of the extracted precursors still surviving after the solenoidal trip through the heat exchangers is continuously reinserted into the core. A new type of dynamical model is required to account for these unusual technological features. The mathematical structure of the evolution model presented in this paper consists of a system of integro-differential-difference equations, whose solution is derived in closed-form, by means of fully analytical techniques. Many dynamics and safety features of reactors of this type can be clarified a priori, upon inspection of the mathematical properties of the solution of the model. The rigorous time-eigenvalue generating equation can be explicitly established in the present theoretical context, together with the evaluation of any kind of transients. A short survey on the possible fields of application of these reactors is also presented

  15. Assessment of a general methodology for the analysis of natural circulation stability with water at supercritical pressure

    International Nuclear Information System (INIS)

    Debrah, K. S.

    2014-07-01

    To advance nuclear energy to meet future energy needs, the concept of Super Critical Water-Cooled Reactor (SCWR) as part or Generation IV (Gen IV) reactors was introduced with plans to deploy by 2030. Supercritical water-cooled reactors pose new challenges in stability and natural circulation phenomena at supercritical pressures because of the strong variability of thermodynamic and thermo-physical properties. ln this research, included in the frame work of the International Atomic Energy Agency (lAEA) fellowship and Coordinated Research Project (CRP) on H eat transfer Behavior and Thermo hydraulics Codes Testing for SCWRs , the natural circulation H 2 O experimental data at supercritical pressures of 25 MPa obtained at the China Institute of Atomic Energy (CIAE) of China, was used to evaluate the predictions of different system codes: RELAP5/MOD3.3, STAR-CCM+ as well as three (3) different and independent developed in-house codes (Ishii-sup loop, NCLoop T ran and NCLoop L ine). Stability analyses of an idealized loop (loop equivalent to CIAE natural circulation loop) of uniform diameter equivalent to the CIAE natural circulation loop at 25 MPa was performed using RELAP5 and an in-house code (Ishii-sup Loop). It was found for both RELAP and Ishii-sup Loop that, when heat structures are accounted for in models equipped with heat transfer and friction correlations for 'normal' fluids, the comparison with experimental data is not completely satisfactory because the observed experimental oscillations were delayed in simulation. It has also been found that the stability margin was slightly earlier than the peak of the flow rate-power curve at a given inlet enthalpy. Results from STAR-CCM+ was also compared with results obtained with RELAP5 and the in-house code of NCLoop. Even though STAR-CCM+ predicted a lower flow rate than the in-house codes, all codes exhibited the ability to predict the instability and results from all codes compared favorably. Stability

  16. Coupling two-phase fluid flow with two-phase darcy flow in anisotropic porous media

    KAUST Repository

    Chen, J.; Sun, S.; Chen, Z.

    2014-01-01

    in the free fluid region and the two-phase Darcy law in the anisotropic porous medium region. A Robin-Robin domain decomposition method is used for the coupled Navier-Stokes and Darcy system with the generalized Beavers-Joseph-Saffman condition

  17. Two-Phase Quality/Flow Meter

    Science.gov (United States)

    Moerk, J. Steven (Inventor); Youngquist, Robert C. (Inventor); Werlink, Rudy J. (Inventor)

    1999-01-01

    A quality and/or flow meter employs a capacitance probe assembly for measuring the dielectric constant of flow stream, particularly a two-phase flow stream including liquid and gas components.ne dielectric constant of the flow stream varies depending upon the volume ratios of its liquid and gas components, and capacitance measurements can therefore be employed to calculate the quality of the flow, which is defined as the volume ratio of liquid in the flow to the total volume ratio of gas and liquid in the flow. By using two spaced capacitance sensors, and cross-correlating the time varying capacitance values of each, the velocity of the flow stream can also be determined. A microcontroller-based processing circuit is employed to measure the capacitance of the probe sensors.The circuit employs high speed timer and counter circuits to provide a high resolution measurement of the time interval required to charge each capacitor in the probe assembly. In this manner, a high resolution, noise resistant, digital representation of each of capacitance value is obtained without the need for a high resolution A/D converter, or a high frequency oscillator circuit. One embodiment of the probe assembly employs a capacitor with two ground plates which provide symmetry to insure that accurate measurements are made thereby.

  18. Two-phase flux simulations by robots

    International Nuclear Information System (INIS)

    Barrera, F.D.

    1997-01-01

    Two-Phase flow systems are studied following the statistical formulation, which takes into account the bubble population balances. This is done by means of automata simulation. Geometrical automata are associated to the dispersed phase, and are represented by discs on the plane, resembling bubbles moving in a fluid environment. Following pre-determined rules, the automata evolve, and useful statistical information about their interaction is obtained. This information is applied in the present work to study the mechanisms that induce bubble coalescence. Models for one and two sized automata are presented. It was found that in the case of the model for one size, the probability of interaction among bubbles and the pair correlation function depends not only on the void fraction, but also on the number of elements of the dispersed phase. A correlation for the collision probability between two bubbles is obtained, and this result was extended to the pair correlation function. For the case of systems with two characteristic sizes, a model was formulated for analyzing the interaction among bubbles of the two groups. The interaction of bubbles for one and two sized systems were related by a symmetry factor, which shows the dependence of the interaction among bubbles with the size distribution. By means of the automata simulation, the phenomena of bubble confinement and screening were characterized. It was found that the first phenomenon is stronger in systems with greater distance among bubbles, and that the second effect increases with void fraction and bubble number. (author)

  19. Constitutive equations for two-phase flows

    International Nuclear Information System (INIS)

    Boure, J.A.

    1974-12-01

    The mathematical model of a system of fluids consists of several kinds of equations complemented by boundary and initial conditions. The first kind equations result from the application to the system, of the fundamental conservation laws (mass, momentum, energy). The second kind equations characterize the fluid itself, i.e. its intrinsic properties and in particular its mechanical and thermodynamical behavior. They are the mathematical model of the particular fluid under consideration, the laws they expressed are so called the constitutive equations of the fluid. In practice the constitutive equations cannot be fully stated without reference to the conservation laws. Two classes of model have been distinguished: mixture model and two-fluid models. In mixture models, the mixture is considered as a single fluid. Besides the usual friction factor and heat transfer correlations, a single constitutive law is necessary. In diffusion models, the mixture equation of state is replaced by the phasic equations of state and by three consitutive laws, for phase change mass transfer, drift velocity and thermal non-equilibrium respectively. In the two-fluid models, the two phases are considered separately; two phasic equations of state, two friction factor correlations, two heat transfer correlations and four constitutive laws are included [fr

  20. Numerical calculation of two-phase flows

    International Nuclear Information System (INIS)

    Travis, J.R.; Harlow, F.H.; Amsden, A.A.

    1975-06-01

    The theoretical study of time-varying two-phase flow problems in several space dimensions introduces such a complicated set of coupled nonlinear partial differential equations that numerical solution procedures for high-speed computers are required in almost all but the simplest examples. Efficient attainment of realistic solutions for practical problems requires a finite- difference formulation that is simultaneously implicit in the treatment of mass convection, equations of state, and the momentum coupling between phases. Such a method is described, the equations on which it is based are discussed, and its properties are illustrated by means of examples. In particular, the capability for calculating physical instabilities and other time-varying dynamics, at the same time avoiding numerical instability is emphasized. The computer code is applicable to problems in reactor safety analysis, the dynamics of fluidized dust beds, raindrops or aerosol transport, and a variety of similar circumstances, including the effects of phase transitions and the release of latent heat or chemical energy. (U.S.)

  1. Condensation in a two-phase pool

    International Nuclear Information System (INIS)

    Duffey, R.B.; Hughes, E.D.

    1991-01-01

    We consider the case of vapor condensation in a liquid pool, when the heat transfer is controlled by heat losses through the walls. The analysis is based on drift flux theory for phase separation in the pool, and determines the two-phase mixture height for the pool. To our knowledge this is the first analytical treatment of this classic problem that gives an explicit result, previous work having established the result for the evaporative case. From conservation of mass and energy in a one-dimensional steady flow, together with a void relation between the liquid and vapor fluxes, we determine the increase in the mixture level from the base level of the pool. It can be seen that the thermal and hydrodynamic influences are separable. Thus, the thermal influence of the wall heat transfer appears through its effect on the condensing length L*, so that at high condensation rates the pool is all liquid, and at low rates overflows (the level swell or foaming effect). Similarly, the phase separation effect hydrodynamically determines the height via the relative velocity of the mixture to the entering flux. We examine some practical applications of this result to level swell in condensing flows, and also examine some limits in ideal cases

  2. Study of natural circulation for the design of a research reactor using computational fluid dynamics and evolutionary computation techniques

    International Nuclear Information System (INIS)

    Oliveira, Andre Felipe da Silva de

    2012-01-01

    Safety is one of the most important and desirable characteristics in a nuclear plant Natural circulation cooling systems are noted for providing passive safety. These systems can be used as mechanism for removing the residual heat from the reactor, or even as the main cooling system for heated sections, such as the core. In this work, a computational fluid dynamics (CFD) code called CFX is used to simulate the process of natural circulation in a research reactor pool after its shutdown. The physical model studied is similar to the Open Pool Australian Light water reactor (OPAL), and contains the core, cooling pool, reflecting tank, circulation pipes and chimney. For best computing performance, the core region was modeled as a porous medium, where the parameters were obtained from a separately detailed CFD analysis. This work also aims to study the viability of the implementation of Differential Evolution algorithm for optimization the physical and operational parameters that, obeying the laws of similarity, lead to a test section on a reduced scale of the reactor pool.

  3. Simulation of single phase instability behaviour in a rectangular natural circulation loop using RELAP5/ MOD 3.2 computer code

    International Nuclear Information System (INIS)

    Sharma, Manish; Pilkhwal, D.S.; Vijayan, P.K.; Saha, D.; Sinha, R.K.

    2002-06-01

    Occurrence of instability in natural circulation loops can lead to problems in control and occurrence of critical heat flux (CHF) during low flow periods. Remaining within an identified stable zone operation is therefore desirable. Natural circulation loops can pass through an unstable zone during start-up and power raising. In the present work RELAPS / MOD 3.2 computer code has been used to simulate the unstable oscillatory behavior observed in a rectangular natural circulation loop having horizontal heater and horizontal cooler (HHHC) orientation. The results were compared with the experimental data. This report describes the nodalization scheme adopted tor this work and results of the analysis in detail. (author)

  4. Natural and anthropogenic radionuclide distributions in the Nansen Basin, Artic Ocean: Scavenging rates and circulation timescales

    Science.gov (United States)

    Kirk Cochran, J.; Hirschberg, David J.; Livingston, Hugh D.; Buesseler, Ken O.; Key, Robert M.

    Determination of the naturally occurring radionuclides 232Th, 230Th, 228 Th and 210Pb, and the anthropogenic radionuclides 241Am, 239,240Pu, 134Cs and 137Cs in water samples collected across the Nansen Basin from the Barents Sea slope to the Gakkel Ridge provides tracers with which to characterize both scavenging rates and circulation timescales in this portion of the Arctic Ocean. Large volume water samples (˜ 15001) were filtered in situ to separate particulate (> 0.5 μm) and dissolved Th isotopes and 241Am. Thorium-230 displays increases in both particulate and dissolved activities with depth, with dissolved 230Th greater and particulate 230Th lower in the deep central Nansen Basin than at the Barents Sea slope. Dissolved 228Th activities also are greater relative to 228Ra, in the central basin. Residence times for Th relative to removal from solution onto particles are ˜1 year in surface water, ˜10 years in deep water adjacent to the Barents Sea slope, and ˜20 years in the Eurasian Basin Deep Water. Lead-210 in the central basin deep water also has a residence time of ˜20 years with respect to its removal from the water column. This texture of scavenging is reflected in distributions of the particle-reactive anthropogenic radionuclide 241Am, which shows higher activities relative to Pu in the central Nansen Basin than at the Barents Sea slope. Distributions Of 137Cs show more rapid mixing at the basin margins (Barents Sea slope in the south, Gakkel Ridge in the north) than in the basin interior. Cesium-137 is mixed throughout the water column adjacent to the Barents Sea slope and is present in low but detectable activities in the Eurasian Basin Deep Water in the central basin. At the time of sampling (1987) the surface water at all stations had been labeled with 134Cs released in the 1986 accident at the Chernobyl nuclear power station. In the ˜1 year since the introduction of Chernobyl 134Cs to the Nansen Basin, it had been mixed to depths of ˜800 m at

  5. Natural circulation data and methods for advanced water cooled nuclear power plant designs. Proceedings of a technical committee meeting

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-04-01

    The complex set of physical phenomena that occur in a gravity environment when a geometrically distinct heat sink and heat source are connected by a fluid flow path can be identified as natural circulation (NC). No external sources of mechanical energy for the fluid motion are involved when NC is established. Within the present context, natural convection is used to identify the phenomena that occur when a heat source is put in contact with a fluid. Therefore, natural convection characterizes a heat transfer regime that constitutes a subset of NC phenomena. This report provides the presented papers and summarizes the discussions at an IAEA Technical Committee Meeting (TCM) on Natural Circulation Data and Methods for innovative Nuclear Power Plant Design. While the planned scope of the TCM involved all types of reactor designs (light water reactors, heavy water reactors, gas-cooled reactors and liquid metal-cooled reactors), the meeting participants and papers addressed only light water reactors (LWRs) and heavy water reactors (HWRs). Furthermore, the papers and discussion addressed both evolutionary and innovative water cooled reactors, as defined by the IAEA. The accomplishment of the objectives of achieving a high safety level and reducing the cost through the reliance on NC mechanisms, requires a thorough understanding of those mechanisms. Natural circulation systems are usually characterized by smaller driving forces with respect to the systems that use an external source of energy for the fluid motion. For instance, pressure drops caused by vertical bends and siphons in a given piping system, or heat losses to environment are a secondary design consideration when a pump is installed and drives the flow. On the contrary, a significant influence upon the overall system performance may be expected due to the same pressure drops and thermal power release to the environment when natural circulation produces the coolant flow. Therefore, the level of knowledge for

  6. Natural circulation data and methods for advanced water cooled nuclear power plant designs. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    2002-04-01

    The complex set of physical phenomena that occur in a gravity environment when a geometrically distinct heat sink and heat source are connected by a fluid flow path can be identified as natural circulation (NC). No external sources of mechanical energy for the fluid motion are involved when NC is established. Within the present context, natural convection is used to identify the phenomena that occur when a heat source is put in contact with a fluid. Therefore, natural convection characterizes a heat transfer regime that constitutes a subset of NC phenomena. This report provides the presented papers and summarizes the discussions at an IAEA Technical Committee Meeting (TCM) on Natural Circulation Data and Methods for innovative Nuclear Power Plant Design. While the planned scope of the TCM involved all types of reactor designs (light water reactors, heavy water reactors, gas-cooled reactors and liquid metal-cooled reactors), the meeting participants and papers addressed only light water reactors (LWRs) and heavy water reactors (HWRs). Furthermore, the papers and discussion addressed both evolutionary and innovative water cooled reactors, as defined by the IAEA. The accomplishment of the objectives of achieving a high safety level and reducing the cost through the reliance on NC mechanisms, requires a thorough understanding of those mechanisms. Natural circulation systems are usually characterized by smaller driving forces with respect to the systems that use an external source of energy for the fluid motion. For instance, pressure drops caused by vertical bends and siphons in a given piping system, or heat losses to environment are a secondary design consideration when a pump is installed and drives the flow. On the contrary, a significant influence upon the overall system performance may be expected due to the same pressure drops and thermal power release to the environment when natural circulation produces the coolant flow. Therefore, the level of knowledge for

  7. Circulating concentrations of insulin-like growth factor-1 in dogs with naturally occurring mitral regurgitation

    DEFF Research Database (Denmark)

    Pedersen, Henrik Duelund; Falk, Bo Torkel; Häggström, Jens

    2005-01-01

    Insulin-like growth factor-1 (IGF-1), which mediates most effects of growth hormone, has effects on cardiac mass and function, and plays an important role in the regulation of vascular tone. In humans, an inverse relationship between degree of heart failure (HF) and circulating IGF-1 concentrations...... has been found in several studies. In dogs with HF, few studies have focused on IGF-1. We examined circulating IGF-1 concentrations in dogs with mitral regurgitation (MR) caused by myxomatous mitral valve disease. Study 1 included 88 Cavalier King Charles Spaniels (CKCSs) with a broad range...... of asymptomatic MR (median serum IGF-1: 76.7 µg/L; 25-75 percentile, 59.8-104.9 µg/L). As expected, standard body weight and percentage under- or overweight correlated directly with IGF-1. MR (assessed in 4 different ways) did not correlate with IGF-1. In study 2, 28 dogs with severe MR and stable, treated...

  8. Two-phase Heating in Flaring Loops

    Science.gov (United States)

    Zhu, Chunming; Qiu, Jiong; Longcope, Dana W.

    2018-03-01

    We analyze and model a C5.7 two-ribbon solar flare observed by the Solar Dynamics Observatory, Hinode, and GOES on 2011 December 26. The flare is made of many loops formed and heated successively over one and half hours, and their footpoints are brightened in the UV 1600 Å before enhanced soft X-ray and EUV missions are observed in flare loops. Assuming that anchored at each brightened UV pixel is a half flaring loop, we identify more than 6700 half flaring loops, and infer the heating rate of each loop from the UV light curve at the footpoint. In each half loop, the heating rate consists of two phases: intense impulsive heating followed by a low-rate heating that is persistent for more than 20 minutes. Using these heating rates, we simulate the evolution of their coronal temperatures and densities with the model of the “enthalpy-based thermal evolution of loops.” In the model, suppression of thermal conduction is also considered. This model successfully reproduces total soft X-ray and EUV light curves observed in 15 passbands by four instruments GOES, AIA, XRT, and EVE. In this flare, a total energy of 4.9 × 1030 erg is required to heat the corona, around 40% of this energy is in the slow-heating phase. About two-fifths of the total energy used to heat the corona is radiated by the coronal plasmas, and the other three fifth transported to the lower atmosphere by thermal conduction.

  9. Ostwald ripening in two-phase mixtures

    International Nuclear Information System (INIS)

    Voorhees, P.W.

    1982-01-01

    Experimental measurements of the temperature of a rapidly solidified solid-liquid mixture have been made over a range of volume fractions solid 0.23 to 0.95. These experiments demonstrate the viability of measuring the change in interfacial curvature with time via precision thermometry. The experimental measurements also indicate that there is no radical change in interface morphology over a wide range of volume fractions solid. A solution to the multi-particle diffusion problem (MDP) has been constructed through the use of potential theory. The solution to the MDP was used to describe the diffusion field within a coarsening two-phase mixture consisting of dispersed spherical second-phase particles. Since this theory is based upon the MDP, interparticle diffusional interactions are specifically included in the treatment. As a result, the theory yields, for the first time, insights into the influence of the local distribution of curvature on a particle's coarsening rate. The effect of interparticle interactions on the collective behavior of an ensemble of coarsening particles was also investigated. It was found that any arbitrary distribution of particle radii will tend to a specific time independent distribution when the particle radii are scaled by the average particle radius. Furthermore, it was determined that with increasing volume fraction of coarsening phase, these time independent distributions become broader and more symmetric. It was also found that the ripening kinetics, as measured by the growth rate of the average particle size, increases by a factor of five upon increasing the volume fraction of coarsening phase from zero to 0.5

  10. Phenomenological studies of two-phase flow processes for nuclear waste isolation

    International Nuclear Information System (INIS)

    Pruess, K.; Finsterle, S.; Persoff, P.; Oldenburg, C.

    1994-01-01

    The US civilian radioactive waste management program is unique in its focus on a site in the unsaturated zone, at Yucca Mountain, Nevada. Two-phase flow phenomena can also play an important role in repositories beneath the water table where gas is generated by corrosion, hydrolysis, and biological degradation of the waste packages. An integrated program has been initiated to enhance our understanding of two-phase flow behavior in fractured rock masses. The studies include two-phase (gas-liquid) flow experiments in laboratory specimens of natural rock fractures, analysis and modeling of heterogeneity and instability effects in two-phase flow, and design and interpretation of field experiments by means of numerical simulation. We present results that identify important aspects of two-phase flow behavior on different space and time scales which are relevant to nuclear waste disposal in both unsaturated and saturated formations

  11. Analyses of liquid-gas two-phase flow in fermentation tanks

    International Nuclear Information System (INIS)

    Toi, Takashi; Serizawa, Akimi; Takahashi, Osamu; Kawara, Zensaku; Gofuku, Akio; Kataoka, Isao.

    1993-01-01

    The understanding of two-phase flow is one of the important problems for both design and safety analyses of various engineering systems. For example, the flow conditions in beer fermentation tanks have an influence on the quality of production and productivity of tank. In this study, a two-dimensional numerical calculation code based on the one-pressure two-fluid model is developed to understand the circulation structure of low quality liquid-gas two-phase flows induced by bubble plume in a tank. (author)

  12. Parallel two-phase-flow-induced vibrations in fuel pin model

    International Nuclear Information System (INIS)

    Hara, Fumio; Yamashita, Tadashi

    1978-01-01

    This paper reports the experimental results of vibrations of a fuel pin model -herein meaning the essential form of a fuel pin from the standpoint of vibration- in a parallel air-and-water two-phase flow. The essential part of the experimental apparatus consisted of a flat elastic strip made of stainless steel, both ends of which were firmly supported in a circular channel conveying the two-phase fluid. Vibrational strain of the fuel pin model, pressure fluctuation of the two-phase flow and two-phase-flow void signals were measured. Statistical measures such as power spectral density, variance and correlation function were calculated. The authors obtained (1) the relation between variance of vibrational strain and two-phase-flow velocity, (2) the relation between variance of vibrational strain and two-phase-flow pressure fluctuation, (3) frequency characteristics of variance of vibrational strain against the dominant frequency of the two-phase-flow pressure fluctuation, and (4) frequency characteristics of variance of vibrational strain against the dominant frequency of two-phase-flow void signals. The authors conclude that there exist two kinds of excitation mechanisms in vibrations of a fuel pin model inserted in a parallel air-and-water two-phase flow; namely, (1) parametric excitation, which occurs when the fundamental natural frequency of the fuel pin model is related to the dominant travelling frequency of water slugs in the two-phase flow by the ratio 1/2, 1/1, 3/2 and so on; and (2) vibrational resonance, which occurs when the fundamental frequency coincides with the dominant frequency of the two-phase-flow pressure fluctuation. (auth.)

  13. Developing two-phase flow modelling concepts for rock fractures

    Energy Technology Data Exchange (ETDEWEB)

    Keto, V. (Fortum Nuclear Services Oy, Espoo (Finland))

    2010-01-15

    The Finnish nuclear waste disposal company, Posiva Oy, is planning an underground repository for spent nuclear fuel to be constructed on the island of Olkiluoto on the south-west coast of Finland. One element of the site investigations conducted at Olkiluoto is the excavation of the underground rock characterisation facility (ONKALO) that will be extended to the final disposal depth (approximately -400 m). The bedrock around the excavated tunnel volume is fully saturated with groundwater, which water commonly contains a mixture of dissolved gases. These gases remain dissolved due to the high hydrostatic pressure. During tunnel excavation work the natural hydrostatic pressure field is disturbed and the water pressure will decrease close to the atmospheric pressure in the immediate vicinity of the tunnel. During this pressure drop two-phase flow conditions (combined flow of both water and gas) may develop in the vicinity of the underground opening, as the dissolved gas is exsoluted under the low pressure (the term exsolution refers here to release of the dissolved gas molecules from the water phase into a separate gas phase). This report steers towards concept development for numerical two-phase flow modeling for fractured rock. The focus is on the description of gas phase formation process under disturbed hydraulic conditions by exsolution of dissolved gases from groundwater, and on understanding the effects of a possibly formed gas phase on groundwater flow conditions in rock fractures. A mathematical model of three mutually coupled nonlinear partial differential equations for two-phase flow is presented and corresponding constitutional relationships are introduced and discussed. Illustrative numerical simulations are performed in a simplified setting using COMSOL Multiphysics 3.5a - software package. Shortcomings and conceptual problems are discussed. (orig.)

  14. Developing two-phase flow modelling concepts for rock fractures

    International Nuclear Information System (INIS)

    Keto, V.

    2010-01-01

    The Finnish nuclear waste disposal company, Posiva Oy, is planning an underground repository for spent nuclear fuel to be constructed on the island of Olkiluoto on the south-west coast of Finland. One element of the site investigations conducted at Olkiluoto is the excavation of the underground rock characterisation facility (ONKALO) that will be extended to the final disposal depth (approximately -400 m). The bedrock around the excavated tunnel volume is fully saturated with groundwater, which water commonly contains a mixture of dissolved gases. These gases remain dissolved due to the high hydrostatic pressure. During tunnel excavation work the natural hydrostatic pressure field is disturbed and the water pressure will decrease close to the atmospheric pressure in the immediate vicinity of the tunnel. During this pressure drop two-phase flow conditions (combined flow of both water and gas) may develop in the vicinity of the underground opening, as the dissolved gas is exsoluted under the low pressure (the term exsolution refers here to release of the dissolved gas molecules from the water phase into a separate gas phase). This report steers towards concept development for numerical two-phase flow modeling for fractured rock. The focus is on the description of gas phase formation process under disturbed hydraulic conditions by exsolution of dissolved gases from groundwater, and on understanding the effects of a possibly formed gas phase on groundwater flow conditions in rock fractures. A mathematical model of three mutually coupled nonlinear partial differential equations for two-phase flow is presented and corresponding constitutional relationships are introduced and discussed. Illustrative numerical simulations are performed in a simplified setting using COMSOL Multiphysics 3.5a - software package. Shortcomings and conceptual problems are discussed. (orig.)

  15. Vapor Compressor Driven Hybrid Two-Phase Loop, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — This Small Business Innovation Research Phase I project will demonstrate a vapor compressor driven hybrid two-phase loop technology. The hybrid two-phase loop...

  16. A review of modern advances in analyses and applications of single-phase natural circulation loop in nuclear thermal hydraulics

    International Nuclear Information System (INIS)

    Basu, Dipankar N.; Bhattacharyya, Souvik; Das, P.K.

    2014-01-01

    Highlights: • Comprehensive review of state-of-the-art on single-phase natural circulation loops. • Detailed discussion on growth in solar thermal system and nuclear thermal hydraulics. • Systematic development in scaling methodologies for fabrication of test facilities. • Importance of numerical modeling schemes for stability assessment using 1-D codes. • Appraisal of current trend of research and possible future directions. - Abstract: A comprehensive review of single-phase natural circulation loop (NCL) is presented here. Relevant literature reported since the later part of 1980s has been meticulously surveyed, with occasional obligatory reference to a few pioneering studies originating prior to that period, summarizing the key observations and the present trend of research. Development in the concept of buoyancy-induced flow is discussed, with introduction to flow initiation in an NCL due to instability. Detailed discussion on modern advancement in important application areas like solar thermal systems and nuclear thermal hydraulics are presented, with separate analysis for various reactor designs working on natural circulation. Identification of scaling criteria for designing lab-scale experimental facilities has gone through a series of modification. A systematic analysis of the same is presented, considering the state-of-the-art knowledge base. Different approaches have been followed for modeling single-phase NCLs, including simplified Lorenz system mostly for toroidal loops, 1-D computational modeling for both steady-state and stability characterization and 3-D commercial system codes to have a better flow visualization. Methodical review of the relevant studies is presented following a systematic approach, to assess the gradual progression in understanding of the practical system. Brief appraisal of current research interest is reported, including the use of nanofluids for fluid property augmentation, marine reactors subjected to rolling waves

  17. A review of modern advances in analyses and applications of single-phase natural circulation loop in nuclear thermal hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Basu, Dipankar N., E-mail: dipankar.n.basu@gmail.com [Department of Mechanical Engineering, Indian Institute of Technology Guwahati, Guwahati 781039 (India); Bhattacharyya, Souvik; Das, P.K. [Department of Mechanical Engineering, Indian Institute of Technology Kharagpur, Kharagpur 721302 (India)

    2014-12-15

    Highlights: • Comprehensive review of state-of-the-art on single-phase natural circulation loops. • Detailed discussion on growth in solar thermal system and nuclear thermal hydraulics. • Systematic development in scaling methodologies for fabrication of test facilities. • Importance of numerical modeling schemes for stability assessment using 1-D codes. • Appraisal of current trend of research and possible future directions. - Abstract: A comprehensive review of single-phase natural circulation loop (NCL) is presented here. Relevant literature reported since the later part of 1980s has been meticulously surveyed, with occasional obligatory reference to a few pioneering studies originating prior to that period, summarizing the key observations and the present trend of research. Development in the concept of buoyancy-induced flow is discussed, with introduction to flow initiation in an NCL due to instability. Detailed discussion on modern advancement in important application areas like solar thermal systems and nuclear thermal hydraulics are presented, with separate analysis for various reactor designs working on natural circulation. Identification of scaling criteria for designing lab-scale experimental facilities has gone through a series of modification. A systematic analysis of the same is presented, considering the state-of-the-art knowledge base. Different approaches have been followed for modeling single-phase NCLs, including simplified Lorenz system mostly for toroidal loops, 1-D computational modeling for both steady-state and stability characterization and 3-D commercial system codes to have a better flow visualization. Methodical review of the relevant studies is presented following a systematic approach, to assess the gradual progression in understanding of the practical system. Brief appraisal of current research interest is reported, including the use of nanofluids for fluid property augmentation, marine reactors subjected to rolling waves

  18. Steam drum level dynamics in a multiple loop natural circulation system of a pressure-tube type boiling water reactor

    International Nuclear Information System (INIS)

    Jain, Vikas; Kulkarni, P.P.; Nayak, A.K.; Vijayan, P.K.; Saha, D.; Sinha, R.K.

    2011-01-01

    Highlights: → We have highlighted the problem of drum level dynamics in a multiple loop type NC system using RELAP5 code. → The need of interconnections in steam and liquid spaces close to drum is established. → The steam space interconnections equalize pressure and liquid space interconnections equalize level. → With this scheme, the system can withstand anomalous conditions. → However, the controller is found to be inevitable for inventory balance. - Abstract: Advanced Heavy Water Reactor (AHWR) is a pressure tube type boiling water reactor employing natural circulation as the mode of heat removal under all the operating conditions. Main heat transport system (MHTS) of AHWR is essentially a multi-loop natural circulation system with all the loops connected to each other. Each loop of MHTS has a steam drum that provides for gravity based steam-water separation. Steam drum level is a very critical parameter especially in multi-loop natural circulation systems as large departures from the set point may lead to ineffective separation of steam-water or may affect the driving head. However, such a system is susceptible to steam drum level anomalies under postulated asymmetrical operating conditions among the different quadrants of the core like feedwater flow distribution anomaly among the steam drums or power anomaly among the core quadrants. Analyses were carried out to probe such scenarios and unravel the underlying dynamics of steam drum level using system code RELAP5/Mod3.2. In addition, a scheme to obviate such problem in a passive manner without dependence on level controller was examined. It was concluded that steam drums need to be connected in the liquid as well as steam space to make the system tolerant to asymmetrical operating conditions.

  19. Experimental and numerical investigation on two-phase flow instabilities

    Energy Technology Data Exchange (ETDEWEB)

    Ruspini, Leonardo Carlos

    2013-03-01

    Two-phase flow instabilities are experimentally and numerically studied within this thesis. In particular, the phenomena called Ledinegg instability, density wave oscillations and pressure drop oscillations are investigated. The most important investigations regarding the occurrence of two-phase flow instabilities are reviewed. An extensive description of the main contributions in the experimental and analytical research is presented. In addition, a critical discussion and recommendations for future investigations are presented. A numerical framework using a hp-adaptive method is developed in order to solve the conservation equations modelling general thermo-hydraulic systems. A natural convection problem is analysed numerically in order to test the numerical solver. Moreover, the description of an adaptive strategy to solve thermo-hydraulic problems is presented. In the second part of this dissertation, a homogeneous model is used to study Ledinegg, density wave and pressure drop oscillations phenomena numerically. The dynamic characteristics of the Ledinegg (flow excursion) phenomenon are analysed through the simulation of several transient examples. In addition, density wave instabilities in boiling and condensing systems are investigated. The effects of several parameters, such as the fluid inertia and compressibility volumes, on the stability limits of Ledinegg and density wave instabilities are studied, showing a strong influence of these parameters. Moreover, the phenomenon called pressure drop oscillations is numerically investigated. A discussion of the physical representation of several models is presented with reference to the obtained numerical results. Finally, the influence of different parameters on these phenomena is analysed. In the last part, an experimental investigation of these phenomena is presented. The designing methodology used for the construction of the experimental facility is described. Several simulations and a non

  20. Thermo-fluid dynamics of two-phase flow

    CERN Document Server

    Ishii, Mamoru; Ishii, Mamoru; Ishii, M

    2006-01-01

    Provides a very systematic treatment of two phase flow problems from a theoretical perspectiveProvides an easy to follow treatment of modeling and code devlopemnt of two phase flow related phenomenaCovers new results of two phase flow research such as coverage of fuel cells technology.

  1. Effects of HIV infection and ART on phenotype and function of circulating monocytes, natural killer, and innate lymphoid cells.

    Science.gov (United States)

    Nabatanzi, Rose; Cose, Stephen; Joloba, Moses; Jones, Sarah Rowland; Nakanjako, Damalie

    2018-03-15

    HIV infection causes upregulation of markers of inflammation, immune activation and apoptosis of host adaptive, and innate immune cells particularly monocytes, natural killer (NK) and innate lymphoid cells (ILCs). Although antiretroviral therapy (ART) restores CD4 T-cell counts, the persistent aberrant activation of monocytes, NK and ILCs observed likely contributes to the incomplete recovery of T-cell effector functions. A better understanding of the effects of HIV infection and ART on the phenotype and function of circulating monocytes, NK, and ILCs is required to guide development of novel therapeutic interventions to optimize immune recovery.

  2. The light water integral reactor with natural circulation of the coolant at supercritical pressure B-500 SKDI

    International Nuclear Information System (INIS)

    Silin, V.A.; Voznesensky, V.A.; Afrov, A.M.

    1993-01-01

    Pressure increase in the primary circuit over the critical value gives a possibility to construct the B-500SKDI (500 MWe) lightwater integral reactor with natural circulation of the coolant in the vessel with a diameter less than 5 m. The given reactor has a high safety level, simple operability, its specific capital cost and fuel expenditure being lower as compared to a conventional PWR. The development of the reactor is carried out taking into consideration verified technical decisions of current NPPs on the basis of Russian LWR technology. (orig.)

  3. Two phase flow problems in power station boilers

    International Nuclear Information System (INIS)

    Firman, E.C.

    1974-01-01

    The paper outlines some of the waterside thermal and hydrodynamic phenomena relating to design and operation of large boilers in central power stations. The associated programme of work is described with an outline of some results already obtained. By way of introduction, the principal features of conventional and nuclear drum boilers and once-through nuclear heat exchangers are described in so far as they pertain to this area of work. This is followed by discussion of the relevant physical phenomena and problems which arise. For example, the problem of steam entrainment from the drum into the tubes connecting it to the furnace wall tubes is related to its effects on circulation and possible mechanisms of tube failure. Other problems concern the transient associated with start-up or low load operation of plant. The requirement for improved mathematical representation of steady and dynamic performance is mentioned together with the corresponding need for data on heat transfer, pressure loss, hydrodynamic stability, consequences of deposits, etc. The paper concludes with reference to the work being carried out within the C.E.G.B. in relation to the above problems. The facilities employed and the specific studies being made on them are described: these range from field trials on operational boilers to small scale laboratory investigations of underlying two phase flow mechanisms and include high pressure water rigs and a freon rig for simulation studies

  4. Summary on experimental methods for statistical transient analysis of two-phase gas-liquid flow

    International Nuclear Information System (INIS)

    Delhaye, J.M.; Jones, O.C. Jr.

    1976-06-01

    Much work has been done in the study of two-phase gas-liquid flows. Although it has been recognized superficially that such flows are not homogeneous in general, little attention has been paid to the inherent discreteness of the two-phase systems. Only relatively recently have fluctuating characteristics of two-phase flows been studied in detail. As a result, new experimental devices and techniques have been developed for use in measuring quantities previously ignored. This report reviews and summarizes most of these methods in an effort to emphasize the importance of the fluctuating nature of these flows and as a guide to further research in this field

  5. Closed Loop Two-Phase Thermosyphon of Small Dimensions: a Review of the Experimental Results

    Science.gov (United States)

    Franco, Alessandro; Filippeschi, Sauro

    2012-06-01

    A bibliographical review on the heat and mass transfer in gravity assisted Closed Loop Two Phase Thermosyphons (CLTPT) with channels having a hydraulic diameter of the order of some millimetres and input power below 1 kW is proposed. The available experimental works in the literature are critically analysed in order to highlight the main results and the correlation between mass flow rate and heat input in natural circulation loops. A comparison of different experimental apparatuses and results is made. It is observed that the results are very different among them and in many cases the experimental data disagree with the conventional theory developed for an imposed flow rate. The paper analyses the main differences among the experimental devices and try to understand these disagreements. From the present analysis it is evident that further systematic studies are required to generate a meaningful body of knowledge of the heat and mass transport mechanism in these devices for practical applications in cooling devices or energy systems.

  6. Theoretical study and experimental investigation of mixed and natural circulation in LMFBR core subassemblies

    International Nuclear Information System (INIS)

    Leteinturier, D.; Blanc, D.; Menant, B.; Basque, G.

    1980-02-01

    A presentation is made of theoretical and experimental studies carried out in France on mixed and natural convection in LMFBR wire wrapped bundles. Two codes are described, one for mixed convection THERNAT and the other for natural convection BACCHUS. THe related experimental program FETUNA, with electrically heated bundles in sodium loops, is also presented

  7. Study on thermalhydraulics of natural circulation decay heat removal in FBR. Experiment with water of typical reactor trip in the demonstration FBR

    International Nuclear Information System (INIS)

    Koga, Tomonari; Murakami, Takahiro; Eguchi, Yuzuru

    2010-01-01

    Intending to enhance safety and to reduce costs, an FBR plant is being developed in Japan. In relies solely on natural circulation of the primary cooling loop to remove a decay heat of the core after reactor trips. A water test was carried out to advance the development. The test used a 1/10 reduced scale model simulating the core and cooling systems. The experiments simulated representative accidents from steady state to decay heat removal through reactor trip and clarified thermal-hydraulic issues on the thermal circulation performance. Some modifications of the system design were proposed for solving serious problems of natural circulation. An improved design complying with the suggestions will make it possible for natural circulation of the cooling systems to remove the decay heat of the core without causing and unstable or unpredictable change. (author)

  8. Reliability Assessment of 2400 MWth Gas-Cooled Fast Reactor Natural Circulation Decay Heat Removal in Pressurized Situations

    Directory of Open Access Journals (Sweden)

    C. Bassi

    2008-01-01

    Full Text Available As the 2400 MWth gas-cooled fast reactor concept makes use of passive safety features in combination with active safety systems, the question of natural circulation decay heat removal (NCDHR reliability and performance assessment into the ongoing probabilistic safety assessment in support to the reactor design, named “probabilistic engineering assessment” (PEA, constitutes a challenge. Within the 5th Framework Program for Research and Development (FPRD of the European Community, a methodology has been developed to evaluate the reliability of passive systems characterized by a moving fluid and whose operation is based on physical principles, such as the natural circulation. This reliability method for passive systems (RMPSs is based on uncertainties propagation into thermal-hydraulic (T-H calculations. The aim of this exercise is finally to determine the performance reliability of the DHR system operating in a “passive” mode, taking into account the uncertainties of parameters retained for thermal-hydraulical calculations performed with the CATHARE 2 code. According to the PEA preliminary results, exhibiting the weight of pressurized scenarios (i.e., with intact primary circuit boundary for the core damage frequency (CDF, the RMPS exercise is first focusing on the NCDHR performance at these T-H conditions.

  9. Investigation of some locally water-soluble natural polymers as circulation loss control agents during oil fields drilling

    Directory of Open Access Journals (Sweden)

    A.M. Alsabagh

    2014-03-01

    Full Text Available Eliminating or controlling lost circulation during drilling process is costly and time-consuming. Polymers play an important role in mud loss control for their viscosity due to their high molecular weight. In this paper, three natural cellulosic polymers (carboxymethyl cellulose, guar gum and potato starch were investigated as lost circulation control material by measuring different filtration parameters such as; spurt loss, fluid loss and permeability plugging tester value according to the American Petroleum Institute (API standard. The experiments were conducted in a permeability plugging apparatus (PPA at a differential pressure of 100 and 300 psi, using 10, 60 and 90 ceramic discs. From the obtained data, it was found that the 0.1% from the carboxymethyl cellulose exhibited the best results in the filtration parameters among 0.3% guar gum and 0.6% potato starch. At the same time the carboxymethyl cellulose (CMC enhanced the rheological properties of the drilling mud better than the two other used natural polymers in the term of gel strength, thixotropy, plastic and apparent viscosity. These results were discussed in the light of the adsorption and micellar formation.

  10. CFD Analysis of the Primary Cooling System for the Small Modular Natural Circulation Lead Cooled Fast Reactor SNRLFR-100

    Directory of Open Access Journals (Sweden)

    Pengcheng Zhao

    2016-01-01

    Full Text Available Small modular reactor (SMR has drawn wide attention in the past decades, and Lead cooled fast reactor (LFR is one of the most promising advanced reactors which are able to meet the safety economic goals of Gen-IV nuclear energy systems. A small modular natural circulation lead cooled fast reactor-100 MWth (SNRLFR-100 is being developed by University of Science and Technology of China (USTC. In the present work, a 3D CFD model, primary heat exchanger model, fuel pin model, and point kinetic model were established based on some reasonable simplifications and assumptions, the steady-state natural circulation characteristics of SNCLFR-100 primary cooling system were discussed and illustrated, and some reasonable suggestions were proposed for the reactor’s thermal-hydraulic and structural design. Moreover, in order to have a first evaluation of the system behavior in accident conditions, an unprotected loss of heat sink (ULOHS transient simulation at beginning of the reactor cycle (BOC has been analyzed and discussed based on the steady-state simulation results. The key temperatures of the reactor core are all under the safety limits at transient state; the reactor has excellent thermal-hydraulic performance.

  11. On the pulsatile nature of intracranial and spinal CSF-circulation demonstrated by MR imaging

    International Nuclear Information System (INIS)

    Greitz, D.; Franck, A.; Nordell, B.

    1993-01-01

    Cerebrospinal fluid (CSF) flow was studied in 24 healthy volunteers using gated MR phase imaging. The subarachnoid space (SAS) was divided into 5 compartments depending on the magnitude of the pulsatile CSF flows: a high velocity compartment in the area of the brain stem and spinal cord, 2 slow ones at the upper and lower extremes of the SAS, and finally 2 intermediate velocity compartments in between. The main pulsatile spinal flow channel had a meandering pattern. The extraventricular CSF-circulation can be explained by pulsatile CSF flow without the necessity of assuming existence of a net flow. A successive time offset during the cardiac cycle has been found in the fronto-occipital direction of the interplay between the arterial expansion, brain expansion, volume changes of the CSF spaces and of the veins. It is proposed to name this time offset the intrac