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

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

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)

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

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%

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

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

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

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)

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.)

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

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

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)

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)

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)

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)

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.

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.

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)

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.

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

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)

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

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)

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.

Fuzzy logic controllers and chaotic natural convection loops

International Nuclear Information System (INIS)

Theler, German

2007-01-01

The study of natural circulation loops is a subject of special concern for the engineering design of advanced nuclear reactors, as natural convection provides an efficient and completely passive heat removal system. However, under certain circumstances thermal-fluid-dynamical instabilities may appear, threatening the reactor safety as a whole.On the other hand, fuzzy logic controllers provide an ideal framework to approach highly non-linear control problems. In the present work, we develop a software-based fuzzy logic controller and study its application to chaotic natural convection loops.We numerically analyse the linguistic control of the loop known as the Welander problem in such conditions that, if the controller were not present, the circulation flow would be non-periodic unstable.We also design a Taka gi-Sugeno fuzzy controller based on a fuzzy model of a natural convection loop with a toroidal geometry, in order to stabilize a Lorenz-chaotic behaviour.Finally, we show experimental results obtained in a rectangular natural circulation loop [es

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.)

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.

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

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.

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)

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)

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)

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.

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

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)

Working regime identification for natural circulation loops by comparative thermalhydraulic analyses with three fluids under identical operating conditions

International Nuclear Information System (INIS)

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

2015-01-01

Highlights: • Thermalhydraulic analyses of NCL to justify the use of supercritical condition. • Mass flow rate of supercritical loop increases with heater power till a maxima. • Supercritical loop suffer from HTD beyond the maxima with jump in fluid temperature. • HTD is pronounced at higher sink temperatures and pressures just above critical. • Supercritical CO_2 is preferred fluid till the HTD and single-phase water afterwards. - Abstract: Computational investigation for comparative thermalhydraulic analyses of rectangular natural circulation loops is performed to propose a guideline for selecting the working fluid and nature of the loop, subcritical or supercritical, under identical levels of operating parameters like pressure, heating power and coolant temperature. A 3-d uniform-diameter loop geometry is developed with horizontal heating and cooling. Heating is provided in constant heat flux mode, whereas cooling is through a constant temperature sink. Due to favourable thermophysical properties and environmental conformity, water, CO_2 and R134a are selected as possible working fluids. Operational parameters are set so as to have sub- to supercritical condition for CO_2, supercritical for R134a and single-phase liquid for water. Mass flow rate for supercritical fluid rapidly increases with heater power, when the fluid is allowed to cross the pseudocritical point during its passage through the heater, and exhibits a maxima. Drastic fall in mass flow rate can be observed beyond the maxima, accompanied by a jump in maximum fluid temperature and a rapid decline in sink-side heat transfer coefficient. That can be identified as heat transfer deterioration in supercritical natural circulation loops, a highly undesirable situation from loop safety point of view. Allowable working range of heater power can be enhanced by increasing system pressure and decreasing sink temperature. For any specified set of operating conditions, CO_2-based supercritical loops

Working regime identification for natural circulation loops by comparative thermalhydraulic analyses with three fluids under identical operating conditions

Energy Technology Data Exchange (ETDEWEB)

Sarkar, Milan K.S.; Basu, Dipankar N., E-mail: dipankar.n.basu@gmail.com

2015-11-15

Highlights: • Thermalhydraulic analyses of NCL to justify the use of supercritical condition. • Mass flow rate of supercritical loop increases with heater power till a maxima. • Supercritical loop suffer from HTD beyond the maxima with jump in fluid temperature. • HTD is pronounced at higher sink temperatures and pressures just above critical. • Supercritical CO{sub 2} is preferred fluid till the HTD and single-phase water afterwards. - Abstract: Computational investigation for comparative thermalhydraulic analyses of rectangular natural circulation loops is performed to propose a guideline for selecting the working fluid and nature of the loop, subcritical or supercritical, under identical levels of operating parameters like pressure, heating power and coolant temperature. A 3-d uniform-diameter loop geometry is developed with horizontal heating and cooling. Heating is provided in constant heat flux mode, whereas cooling is through a constant temperature sink. Due to favourable thermophysical properties and environmental conformity, water, CO{sub 2} and R134a are selected as possible working fluids. Operational parameters are set so as to have sub- to supercritical condition for CO{sub 2}, supercritical for R134a and single-phase liquid for water. Mass flow rate for supercritical fluid rapidly increases with heater power, when the fluid is allowed to cross the pseudocritical point during its passage through the heater, and exhibits a maxima. Drastic fall in mass flow rate can be observed beyond the maxima, accompanied by a jump in maximum fluid temperature and a rapid decline in sink-side heat transfer coefficient. That can be identified as heat transfer deterioration in supercritical natural circulation loops, a highly undesirable situation from loop safety point of view. Allowable working range of heater power can be enhanced by increasing system pressure and decreasing sink temperature. For any specified set of operating conditions, CO{sub 2}-based

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

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.

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)

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.))

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.)

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)

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)

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

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

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

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)

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)

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

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

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

Experimental studies and computational benchmark on heavy liquid metal natural circulation in a full height-scale test loop for small modular reactors

Energy Technology Data Exchange (ETDEWEB)

Shin, Yong-Hoon, E-mail: chaotics@snu.ac.kr [Department of Energy Systems Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 08826 (Korea, Republic of); Cho, Jaehyun [Korea Atomic Energy Research Institute, 111 Daedeok-daero, 989 Beon-gil, Yuseong-gu, Daejeon 34057 (Korea, Republic of); Lee, Jueun; Ju, Heejae; Sohn, Sungjune; Kim, Yeji; Noh, Hyunyub; Hwang, Il Soon [Department of Energy Systems Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 08826 (Korea, Republic of)

2017-05-15

Highlights: • Experimental studies on natural circulation for lead-bismuth eutectic were conducted. • Adiabatic wall boundaries conditions were established by compensating heat loss. • Computational benchmark with a system thermal-hydraulics code was performed. • Numerical simulation and experiment showed good agreement in mass flow rate. • An empirical relation was formulated for mass flow rate with experimental data. - Abstract: In order to test the enhanced safety of small lead-cooled fast reactors, lead-bismuth eutectic (LBE) natural circulation characteristics have been studied. We present results of experiments with LBE non-isothermal natural circulation in a full-height scale test loop, HELIOS (heavy eutectic liquid metal loop for integral test of operability and safety of PEACER), and the validation of a system thermal-hydraulics code. The experimental studies on LBE were conducted under steady state as a function of core power conditions from 9.8 kW to 33.6 kW. Local surface heaters on the main loop were activated and finely tuned by trial-and-error approach to make adiabatic wall boundary conditions. A thermal-hydraulic system code MARS-LBE was validated by using the well-defined benchmark data. It was found that the predictions were mostly in good agreement with the experimental data in terms of mass flow rate and temperature difference that were both within 7%, respectively. With experiment results, an empirical relation predicting mass flow rate at a non-isothermal, adiabatic condition in HELIOS was derived.

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)

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

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

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

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.

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)

Numerical comparison of thermal hydraulic aspects of supercritical carbon dioxide and subcritical water-based natural circulation loop

Energy Technology Data Exchange (ETDEWEB)

Sarkar, Milan Krishna Singhar; Basu, Dipankar Narayan [Dept. of Mechanical Engineering, Indian Institute of Technology Guwahati, Guwahati (India)

2017-02-15

Application of the supercritical condition in reactor core cooling needs to be properly justified based on the extreme level of parameters involved. Therefore, a numerical study is presented to compare the thermalhydraulic performance of supercritical and single-phase natural circulation loops under low-to-intermediate power levels. Carbon dioxide and water are selected as respective working fluids, operating under an identical set of conditions. Accordingly, a three-dimensional computational model was developed, and solved with an appropriate turbulence model and equations of state. Large asymmetry in velocity and temperature profiles was observed in a single cross section due to local buoyancy effect, which is more prominent for supercritical fluids. Mass flow rate in a supercritical loop increases with power until a maximum is reached, which subsequently corresponds to a rapid deterioration in heat transfer coefficient. That can be identified as the limit of operation for such loops to avoid a high temperature, and therefore, the use of a supercritical loop is suggested only until the appearance of such maxima. Flow-induced heat transfer deterioration can be delayed by increasing system pressure or lowering sink temperature. Bulk temperature level throughout the loop with water as working fluid is higher than supercritical carbon dioxide. This is until the heat transfer deterioration, and hence the use of a single-phase loop is prescribed beyond that limit.

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.

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)

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

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)

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.

Study on the stability of a single-phase natural circulation flow in a closed loop. Demonstrative experiments on the higher-mode density wave oscillation

International Nuclear Information System (INIS)

Nishihara, Takashi

1997-01-01

Single-phase natural circulation loops are very important systems driven by the density variation generated thermally and have various applications in energy systems. Many theoretical and experimental works have been carried out on them and it has been known that the oscillatory instability can occur under some conditions. Most of the works on the oscillatory instability have been limited to specific geometry of the loops and they have paid attention only to the instability of fundamental mode, which has the period approximately equal to the item that the fluid goes round the loop, hereinafter referred to as the typical period. The author had applied the linear stability analysis to the simplified rectangular loop to investigate the basic stability characteristics of a natural circulation flow in a closed loop. The results indicate that various higher-mode oscillatory instabilities can be caused with a period approximately equal to one nth of the typical period according to parameters such as the pressure loss coefficient, the locations of a heat source and a heat sink, and so on. In this report, experimental tests were carried out and it was demonstrated that the higher-mode oscillatory instability can be caused with features as predicted in the analysis. The stability analysis was applied to the geometry of the experimental apparatus. The analytical results and those of experiments were compared with regard to the mode and the region of the parameters to be unstable and they have a good agreement qualitatively. (author)

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.

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

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)

Operating experience with gas-bearing circulators in a high-pressure helium loop

International Nuclear Information System (INIS)

Sanders, J.P.; Gat, U.; Young, H.C.

1988-01-01

A high-pressure engineering test loop has been designed and constructed at the Oak Ridge National Laboratory for circulating helium through a test chamber at temperatures to 1,000 deg. C. The purpose of this loop is to determine the thermal and structural performance of proposed components for the primary loops of gas-cooled nuclear reactors. Three gas-bearing circulators, mounted in series, provide a maximum volumetric flow of 0.47 m 3 /s and a maximum head of 78 kJ/kg at operating pressures from 0.1 to 10.7 MPa. Control of gaseous impurities in the circulating gas was the significant operating requirement that dictated the choice of a circulator that is lubricated by the circulating gas. The motor for each circulator is contained within the pressure boundary, and it is cooled by circulating the gas in the motor cavity over water-cooled coils. Each motor is rated at 200 kW at a speed of 23,500 rpm. The circulators have been operated in the loop for more than 5,000 h. The flow of the gas in the loop is controlled by varying the speed of the circulators through the use of individual 250-kVA, solid state power supplies that can be continuously varied in frequency from 50 to 400 Hz. To prevent excessive wear on the gas bearings during startup, the circulator motor accelerates the rotor to 3,000 rpm in less than one second. During operation, no problems associated with the gas bearings, per se, were encountered; however, related problems pointed to design considerations that should be included in future applications of circulators of this type. The primary test that has been conducted in this loop required sustained operation for several weeks without interruption. After a number of unscheduled interruptions, the operating goals were attained. During part of this period, the loop was operated with only two circulators installed in the pressure vessels with a guard installed in the third vessel to protect the closure flange from the gas temperatures. Unattended

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.

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)

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

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

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.

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

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

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)

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

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.)

Experiments on the Heat Transfer and Natural Circulation Characteristics of the Passive Residual Heat Removal System for the Advanced Integral Type Reactor

International Nuclear Information System (INIS)

Park, Hyun-Sik; Choi, Ki-Yong; Cho, Seok; Park, Choon-Kyung; Lee, Sung-Jae; Song, Chul-Hwa; Chung, Moon-Ki; Lee, Un-Chul

2004-01-01

Experiments on the heat transfer characteristics and natural circulation performance of the passive residual heat removal system (PRHRS) for the SMART-P have been performed using the high temperature/high pressure thermal-hydraulic test facility (VISTA). The VISTA facility consists of the primary loop, the secondary loop, the PRHRS loop, and auxiliary systems to simulate the SMART-P, a pilot plant of the SMART. The primary loop is composed of the steam generator (SG) primary side, a simulated core, a main coolant pump, and loop piping, and the PRHRS loop consists of the SG secondary side, a PRHRS heat exchanger, and loop piping. The natural circulation performance of the PRHRS, the heat transfer characteristics of the PRHRS heat exchangers and the emergency cooldown tank (ECT), and the thermal-hydraulic behavior of the primary loop are intensively investigated. The experimental results show that the coolant flows steadily in the PRHRS loop and the heat transfers through the PRHRS heat exchanger and the emergency cooldown tank are sufficient enough to enable the natural circulation of the coolant. The results also show that the core decay heat can be sufficiently removed from the primary loop with the operation of the PRHRS. (authors)

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

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.

Modeling a forced to natural convection boiling test with the program LOOP-W

International Nuclear Information System (INIS)

Carbajo, J.J.

1984-01-01

Extensive testing has been conducted in the Simulant Boiling Flow Visualization (SBFV) loop in which water is boiled in a vertical transparent tube by circulating hot glycerine in an annulus surrounding the tube. Tests ranged from nonboiling forced convection to oscillatory boiling natural convection. The program LOOP-W has been developed to analyze these tests. This program is a multi-leg, one-dimensional, two-phase equilibrium model with slip between the phases. In this study, a specific test, performed at low power where non-boiling forced convection was changed to boiling natural convection and then to non-boiling again, has been modeled with the program LOOP-W

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

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.

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

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.

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

Heat transfer modelling in the vertical tubes of a natural circulation passive containment loop with noncondensable gas

International Nuclear Information System (INIS)

Herranz, L.E.; Munoz-Cobo, J.L.; Tachenko, I.; Sancho, J.; Escriva, A.; Verdu, G.

1994-01-01

One of the key safety systems of the Simplified Boiling Water Reactor (SBWR) of General Electric is the Passive Containment Cooling System (PCCS). This system is designed to behave as a heat sink without need of operator actions in case of a reactor accident. Such a function relies on setting up a natural circulation loop between drywell and wetwell. Along this loop heat is removed by condensing the steam coming from the drywell onto the inner surface of externally cooled vertical tubes. Therefore, a successful design of the condenser requires a good knowledge of the local heat transmission coefficients. In this paper a model of steam condensation into vertical tubes is presented. Based on a modified diffusion boundary layer approach for noncondensables, this model accounts for the effect of shear stress caused by the cocurrent steam-gas mixture on the liquid film thickness. An approximate method to calculate film thickness, avoiding iterative algorithms, has been proposed. At present, this model has been implemented in HTCPIPE code and its results are being checked in terms of local heat transfer coefficients against the experimental data available. A good agreement between measurements and predictions is being observed for tests at atmospheric pressure. Further development and validation of the model is needed to consider aspects such as mist formation, wavy flow and high pressure. (author)

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.)

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

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)

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.

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

Chemical looping reactor system design double loop circulating fluidized bed (DLCFB)

Energy Technology Data Exchange (ETDEWEB)

Bischi, Aldo

2012-05-15

Chemical looping combustion (CLC) is continuously gaining more importance among the carbon capture and storage (CCS) technologies. It is an unmixed combustion process which takes place in two steps. An effective way to realize CLC is to use two interconnected fluidized beds and a metallic powder circulating among them, acting as oxygen carrier. The metallic powder oxidizes at high temperature in one of the two reactors, the air reactor (AR). It reacts in a highly exothermic reaction with the oxygen of the injected fluidising air. Afterwards the particles are sent to the other reactor where the fuel is injected, the fuel reactor (FR). There, they transport heat and oxygen necessary for the reaction with the injected fuel to take place. At high temperatures, the particle's oxygen reacts with the fuel producing Co2 and steam, and the particles are ready to start the loop again. The overall reaction, the sum of the enthalpy changes of the oxygen carrier oxidation and reduction reactions, is the same as for the conventional combustion. Two are the key features, which make CLC promising both for costs and capture efficiency. First, the high inherent irreversibility of the conventional combustion is avoided because the energy is utilized stepwise. Second, the Co2 is intrinsically separated within the process; so there is in principle no need either of extra carbon capture devices or of expensive air separation units to produce oxygen for oxy-combustion. A lot of effort is taking place worldwide on the development of new chemical looping oxygen carrier particles, reactor systems and processes. The current work is focused on the reactor system: a new design is presented, for the construction of an atmospheric 150kWth prototype working with gaseous fuel and possibly with inexpensive oxygen carriers derived from industrial by-products or natural minerals. It consists of two circulating fluidized beds capable to operate in fast fluidization regime; this will increase the

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.)

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

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

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.

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

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

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

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)

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

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)

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

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.

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

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)

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)

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

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

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)

Numerical simulation of a full-loop circulating fluidized bed under different operating conditions

Energy Technology Data Exchange (ETDEWEB)

Xu, Yupeng [National Energy Technology Lab. (NETL), Morgantown, WV (United States); Musser, Jordan M. [National Energy Technology Lab. (NETL), Morgantown, WV (United States); Li, Tingwen [National Energy Technology Lab. (NETL), Morgantown, WV (United States); AECOM, Morgantown, WV (United States); Rogers, William A. [National Energy Technology Lab. (NETL), Morgantown, WV (United States)

2017-10-17

Both experimental and computational studies of the fluidization of high-density polyethylene (HDPE) particles in a small-scale full-loop circulating fluidized bed are conducted. Experimental measurements of pressure drop are taken at different locations along the bed. The solids circulation rate is measured with an advanced Particle Image Velocimetry (PIV) technique. The bed height of the quasi-static region in the standpipe is also measured. Comparative numerical simulations are performed with a Computational Fluid Dynamics solver utilizing a Discrete Element Method (CFD-DEM). This paper reports a detailed and direct comparison between CFD-DEM results and experimental data for realistic gas-solid fluidization in a full-loop circulating fluidized bed system. The comparison reveals good agreement with respect to system component pressure drop and inventory height in the standpipe. In addition, the effect of different drag laws applied within the CFD simulation is examined and compared with experimental results.

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.

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

UPTF loop seal tests and their RELAP simulation

International Nuclear Information System (INIS)

Tuomainen, M.; Tuunanen, J.

1997-01-01

In a pressurized water reactor the loop seals have an effect on the natural circulation. If a loop seal is filled with water it can cause a flow stagnation in the loop during two-phase natural circulation. Also the pressure loss over a filled loop seal is high, which lowers the water level in the core. Tests to investigate the loop seal behaviour were performed on a German Upper Plenum Test Facility (UPTF). The purpose of the tests was to study the amount of water in the loop seal under different steam flow rates. The tests were simulated with RELAP5/MOD3.2. With high steam flow rates the code had problems in simulating the amount of the water remaining in the pump elbow, but in general the agreement between the calculated results and the experimental data was good. (orig.)

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

International Nuclear Information System (INIS)

Yu, Xin-Guo; Choi, Ki-Yong

2015-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-05-15

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

Integrated design of cryogenic refrigerator and liquid-nitrogen circulation loop for HTS cable

Science.gov (United States)

Chang, Ho-Myung; Ryu, Ki Nam; Yang, Hyung Suk

2016-12-01

A new concept of cryogenic cooling system is proposed and investigated for application to long-length HTS cables. One of major obstacles to the cable length of 1 km or longer is the difficulty in circulating liquid nitrogen (LN) along the cables, since the temperature rise and pressure drop of LN flow could be excessively large. This study attempts a breakthrough by integrating the refrigerator with the LN circulation loop in order to eliminate the cryogenic LN pumps, and generate a large LN flow with the power of compressors at ambient temperature. A variety of thermodynamic structures are investigated on standard and modified Claude cycles, where nitrogen is used as refrigerant and the LN circulation loop is included as part of the closed cycle. Four proposed cycles are fully analyzed and optimized with a process simulator (Aspen HYSYS) to evaluate the FOM (figure of merit) and examine the feasibility. The modified dual-pressure cycle cooled with expander stream is recommended for long HTS cables.

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.

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

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

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)

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)

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)

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)

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)

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

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

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)

Investigations on the effect of heater and cooler orientation on the steady state, transient and stability behaviour of single-phase natural circulation in a rectangular loop

International Nuclear Information System (INIS)

Vijayan, P.K.; Bhojwani, V.K.; Bade, M.H.; Sharma, M.; Nayak, A.K.; Saha, D.; Sinha, R.K.

2002-01-01

An instability demonstration facility has been in operation in the heat transfer laboratory of the Reactor Engineering Div. for the past few years. This report deals with the investigations carried out in this facility so far. The facility is essentially a rectangular loop designed to generate single-phase natural circulation data on the steady state and stability behaviour for different orientations of the heat source and the heat sink. Effect of different heat addition paths (i.e. start-up from rest, power raising from initial stable steady and decay of instability due to power step back) and flow direction on the stability behaviour was also studied. The stability map of the system was generated both by the linear and the nonlinear methods

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

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)

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)

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

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)

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

Production circulator fabrication and testing for core flow test loop. Final report, Phase III

Energy Technology Data Exchange (ETDEWEB)

1981-05-01

The performance testing of two production helium circulators utilizing gas film lubrication is described. These two centrifugal-type circulators plus an identical circulator prototype will be arranged in series to provide the helium flow requirements for the Core Flow Test Loop which is part of the Gas-Cooled Fast Breeder Reactor Program (GCFR) at the Oak Ridge National Laboratory. This report presents the results of the Phase III performance and supplemental tests, which were carried out by MTI during the period of December 18, 1980 through March 19, 1981. Specific test procedures are outlined and described, as are individual tests for measuring the performance of the circulators. Test data and run descriptions are presented.

Production circulator fabrication and testing for core flow test loop. Final report, Phase III

International Nuclear Information System (INIS)

1981-05-01

The performance testing of two production helium circulators utilizing gas film lubrication is described. These two centrifugal-type circulators plus an identical circulator prototype will be arranged in series to provide the helium flow requirements for the Core Flow Test Loop which is part of the Gas-Cooled Fast Breeder Reactor Program (GCFR) at the Oak Ridge National Laboratory. This report presents the results of the Phase III performance and supplemental tests, which were carried out by MTI during the period of December 18, 1980 through March 19, 1981. Specific test procedures are outlined and described, as are individual tests for measuring the performance of the circulators. Test data and run descriptions are presented

Thermal hydraulic aspects of steam drum level control philosophy for the natural circulation based heavy water reactor

International Nuclear Information System (INIS)

Gupta, S.K.; Gaikwad, A.J.; Kumar, Rajesh

2004-01-01

From safety considerations advanced nuclear reactors rely more and more on passive systems such as natural circulation for primary heat removal. A natural circulation based water reactor is relatively larger in size so as to reduce flow losses and channel type for proper flow distribution. From the size of steam drum considerations it has to be multi loop but has a common inlet header. Normally the turbine follows the reactor. This paper addresses the thermal hydraulic aspects of the steam drum pressure and level control philosophy for a four drum, natural circulation based, channel type boiling water advanced reactor. Three philosophies may be followed for drum control viz. individual drum control, one control drum approach and an average of all the four drums. For drum pressure control, the steam flow to the turbine is be regulated. A single point pressure control is better than individual drum pressure control. This is discussed in the paper. But the control point has to be at a place down steam the point where all steam line from individual drum meet. This may lead to different pressure in all the four drums depending on the power produced in the respective loops. The difference in pressure cannot be removed even if the four drums are directly connected through pipes. Also the pressure control scheme with/without interconnection is discussed. For level, the control of individual drum may not be normally possible because of common inlet header. As the frictional pressure drops in the large diameter downcomers are small as compared to elevation pressure drops, the level in all the steam drum tend to equalize. Consequently a single representative drum level may be chosen as a control variable for controlling level in all the four drums. But in case, where all the four loops are producing different powers and single point pressure control is effective, the scheme may not work satisfactorily. the level in a drum may depend on the power produced in the loop

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)

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

MES lead bismuth forced circulation loop and test results

International Nuclear Information System (INIS)

Ono, Mikinori; Mine, Tatsuya; Kitano, Teruaki; Kamata, Kin-ya

2003-01-01

Liquid lead-bismuth is a promising material as future reactor coolant or intensive neutron source material for accelerator driven system (ADS). Mitsui Engineering and Shipbuilding Co., Ltd. (MES) completed lead-bismuth coolant (LBC) forced circulation loop in May 2001 and acquired engineering data on economizer, electro magnetic pump, electro magnetic flow meter and so on. For quality control of LBC, oxygen sensor and filtering element are developing using some hydrogen and moisture mixed gases. Structural materials corrosion test for accelerator driver system (ADS) will start soon. And thermal hydraulic test for ADS will start in tree years. (author)

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.)

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.

Influence of cross-sectional ratio of down comer to riser on the efficiency of liquid circulation in loop air lift bubble column

Science.gov (United States)

Yamamoto, Tatsumi; Kawasaki, Hiroyuki; Mori, Hidetoshi

2017-11-01

Loop type bubble columns have good performance of liquid circulation and mass transfer by airlift effect, where the liquid circulation time is an important measurable characteristic parameter. This parameter is affected by the column construction, the aspect ratio of the column, the cross-sectional area ratio of down comer to riser (R), and the superficial gas velocity in the riser (UGR). In this work, the mean gas holdup and the liquid circulation time (TC) have been measured in four types of loop airlift type bubble column: concentric tube internal loop airlift type, rectangular internal loop airlift type, external loop airlift type, external loop airlift with separator. Air and tap water were used as gas and liquid phase, respectively. The results have demonstrated that the mean gas holdup in riser increases in proportion to UGR, and that it in downcomer changes according to the geometric parameters of each bubble column. TC has been found to conform to an empirical equation which depends on UGR and the length of draft tube or division plate in the region of 0.33 < R < 1.

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.

Closed-loop helium circulation system for actuation of a continuously operating heart catheter pump.

Science.gov (United States)

Karabegovic, Alen; Hinteregger, Markus; Janeczek, Christoph; Mohl, Werner; Gföhler, Margit

2017-06-09

Currently available, pneumatic-based medical devices are operated using closed-loop pulsatile or open continuous systems. Medical devices utilizing gases with a low atomic number in a continuous closed loop stream have not been documented to date. This work presents the construction of a portable helium circulation addressing the need for actuating a novel, pneumatically operated catheter pump. The design of its control system puts emphasis on the performance, safety and low running cost of the catheter pump. Static and dynamic characteristics of individual elements in the circulation are analyzed to ensure a proper operation of the system. The pneumatic circulation maximizes the working range of the drive unit inside the catheter pump while reducing the total size and noise production.Separate flow and pressure controllers position the turbine's working point into the stable region of the pressure creation element. A subsystem for rapid gas evacuation significantly decreases the duration of helium removal after a leak, reaching subatmospheric pressure in the intracorporeal catheter within several milliseconds. The system presented in the study offers an easy control of helium mass flow while ensuring stable behavior of its internal components.

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)

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.

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.

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.

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)

Natural Circulation Capability Assessments for a Small-medium Reactor

International Nuclear Information System (INIS)

Choi, Sun Do

2010-02-01

particularly proportional to heater power to the one third. These results show good agreement in terms of the governing equation of steady state natural circulation. In conclusion, it was evaluated that RTF is satisfied with the safety of a natural circulation system. The outlet temperature of the core at 200 kW heater power did not exceed the boiling temperature at 2.0 MPa, and the heat generated in the reactor core was removed by the coolant in the secondary loop

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.)

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

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.

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

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

Experimental study of core thermohydraulics in fast reactors during transition from forced to natural circulation. Influence of inter-wrapper flow

International Nuclear Information System (INIS)

Kamide, H.; Hayashi, K.; Momoi, K.

1997-01-01

The evaluation of core thermohydraulics under natural circulation conditions is important to utilize inherent safety features of Fast Reactors. When heat exchangers of a decay heat removal system are operated in an upper plenum of reactor vessel, cold sodium is provided by the heat exchangers. Core-plenum interactions will occur during a natural circulation condition due to this cold sodium in the upper plenum, e.g., it can penetrate into gap regions between fuel subassemblies (inter-wrapper flow, IWF) and the flow may reverse in low power core channels. These interactions will significantly modify the flow and temperature distributions in the core. Sodium experiments were carried out to study these phenomena. In a test section, seven subassemblies are housed and connected to an upper plenum. The influences of core-plenum interactions on the core thermohydraulics were investigated under steady state conditions and also in transitions from forced to natural circulation. Cooling effects of IWF on the fuel subassemblies were found in spite of natural circulation flow reduction in the primary loop due to temperature decreases in the upper non-heated section in the core. The inter-wrapper flow can effectively cool the core under extreme conditions of low flow rates through the core. (author)

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.

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.

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)

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.

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.

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

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

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)

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

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

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.

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)

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)

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.

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.

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)

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)

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

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)

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

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.)

Verification of RELAP5/MOD3 with theoretical and numerical stability results on single-phase, natural circulation in a simple loop

International Nuclear Information System (INIS)

Ferreri, Juan C.; Ambrosini, Walter

1998-01-01

The theoretical results given by Pierre Welander are used to test the capability of the RELAP5 series of codes to predict instabilities in single-phase flow. These results are related to the natural circulation in a loop formed by two parallel adiabatic tubes with a point heat sink at the top and a point heat source at the bottom. A stability curve may be defined for laminar flow and was extended to consider turbulent flow. By a suitable selection of the ratio of the total buoyancy force in the loop to the friction resistance, the flow may show instabilities. The solution was useful to test two basic numerical properties of the RELAP5 code, namely: a) convergence to steady state flow-rate using a 'lumped parameter' approximation to both the heat source and sink and; b) the effect of nodalization to numerically damp the instabilities. It was shown that, using a single volume to lump the heat source and sink, it was not possible to reach convergence to steady state flow rate when the heated (cooled) length was diminished and the heat transfer coefficient increased to keep constant the total heat transferred to (and removed from) the fluid. An algebraic justification of these results is presented, showing that it is a limitation inherent to the numerical scheme adopted. Concerning the effect of nodalization on the damping of instabilities, it was shown that a 'reasonably fine' discretization led, as expected, to the damping of the solution. However, the search for convergence of numerical and theoretical results was successful, showing the expected nearly chaotic behavior. This search lead to very refined nodalization. The results obtained have also been verified by the use of simple, ad hoc codes. A procedure to assess the effects of nodalization on the prediction of instabilities threshold is outlined in this report. It is based on the experience gained with aforementioned simpler codes. (author)

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

Analysis of a convection loop for GFR post-LOCA decay heat removal

International Nuclear Information System (INIS)

Williams, W.C.; Hejzlar, P.; Saha, P.

2004-01-01

A computer code (LOCA-COLA) has been developed at MIT for steady state analysis of convective heat transfer loops. In this work, it is used to investigate an external convection loop for decay heat removal of a post-LOCA gas-cooled fast reactor (GFR). The major finding is that natural circulation cooling of the GFR is feasible under certain circumstances. Both helium and CO 2 cooled system components are found to operate in the mixed convection regime, the effects of which are noticeable as heat transfer enhancement or degradation. It is found that CO 2 outdoes helium under identical natural circulation conditions. Decay heat removal is found to have a quadratic dependence on pressure in the laminar flow regime and linear dependence in the turbulent flow regime. Other parametric studies have been performed as well. In conclusion, convection cooling loops are a credible means for GFR decay heat removal and LOCA-COLA is an effective tool for steady state analysis of cooling loops. (authors)

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.

On the determination of loop nature in the TEM

International Nuclear Information System (INIS)

Jenkins, M.L.

1997-01-01

The two direct methods available for determining the nature of small clusters, were studied. Experiments in heavy-ion irradiated Ag and Cu, including comparisons with Black-White contrast analysis (B-W), forced us to conclude that use of the 2-1/2D technique for analysis of faulted loops is impossible because of reciprocal-lattice spike and other effects. We used B-W analysis to study the clusters produced by RT heavy-ion irradiation of Cu. In-situ irradiations were performed in Argonne HVEM-Tandem Facility using 40 and 80 keV Kr + ions. Nearly all the analysable first-layer loops were vacancy in nature. Some deeper cluster were also certainly vacancy. There is no unequivocal evidence for interstitial loops under these conditions. The near impossibility of establishing that interstitial loops are not present, is shown by the fact that a large fraction of loops (over 50%) could not be analyzed either because they did not display clear B-W contrast under any condition or sufficiently consistent B-W contrast under several different operating conditions

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

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

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)

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.

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)

Dynamical behaviour of natural convection in closed loops

International Nuclear Information System (INIS)

Ehrhard, P.

1988-04-01

A one dimensional model is presented together with experiments, which describe the natural convective flow in closed loops heated at the bottom and cooled in the upper semicircle. Starting from a single loop, mechanical and thermal coupling with a second loop is discussed. The experiments and the theoretical model both concurrently demonstrate that the investigated natural convection is clearly influenced by non-linear effects. Beside the variety of stable steady flows there are extensive subcritical ranges of convective flow. In these parameter ranges subcritical instabilities of the steady state flow could occur in the presence of finite amplitude disturbances. However, the supercritical, global unstable range is characterized by chaotic histories of the variables of state. Non-symmetric heating generates an imperfect bifurcation out of the steady solution with zero velocity in the loop. This effect stabilizes the flow in the preferred direction. The flow in the opposite direction only remains stable in a small isolated interval of the heating parameter. Furthermore the calculations with the model equations demonstrate that a stable periodic behaviour of the flow is possible in a small parameter window. However, it has not been possible to verify this particular effect in the experiments conducted to date. (orig./GL) [de

Design Construction and Operation of a Supercritical Carbon Dioxide (sCO₂) Loop for Investigation of Dry Cooling and Natural Circulation Potential for Use in Advanced Small Modular Reactors Utilizing sCO₂ Power Conversion Cycles.

Energy Technology Data Exchange (ETDEWEB)

Middleton, Bobby D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Rodriguez, Salvador B. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Carlson, Matthew David [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2015-11-01

This report outlines the work completed for a Laboratory Directed Research and Development project at Sandia National Laboratories from October 2012 through September 2015. An experimental supercritical carbon dioxide (sCO ₂ ) loop was designed, built, and o perated. The experimental work demonstrated that sCO ₂ can be uti lized as the working fluid in an air - cooled, natural circulation configuration to transfer heat from a source to the ultimate heat sink, which is the surrounding ambient environment in most ca ses. The loop was also operated in an induction - heated, water - cooled configuration that allows for measurements of physical parameters that are difficult to isolate in the air - cooled configuration. Analysis included the development of two computational flu id dynamics models. Future work is anticipated to answer questions that were not covered in this project.

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

Experimental and theoretical study on density wave instability in low quality natural circulation loop

Energy Technology Data Exchange (ETDEWEB)

Jia, Hai Jun [Tsinghua Univ., Beijing, BJ (China); Song, Jin Ho [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1997-07-01

A series of thermal-hydraulic experimental research has been performed at INET of Beijing University in two full scale test loops, HRTL-5 and HRTL-200 simulating the HR-5 and HR-200 district heating reactors. The homogeneous equilibrium flow model and the drift flux model are employed to analyze the flow in the HRTL-200 test loop. The frequency domain linear stability analysis program LINTAB has been developed based on the homogeneous equilibrium results from HRTL-200 test facility have been used to validate the LINTAB code. Analysis results using LINTAB showed a good agreement with the test results. (author). 20 refs., 24 figs.

Experimental and theoretical study on density wave instability in low quality natural circulation loop

International Nuclear Information System (INIS)

Jia, Hai Jun; Song, Jin Ho

1997-07-01

A series of thermal-hydraulic experimental research has been performed at INET of Beijing University in two full scale test loops, HRTL-5 and HRTL-200 simulating the HR-5 and HR-200 district heating reactors. The homogeneous equilibrium flow model and the drift flux model are employed to analyze the flow in the HRTL-200 test loop. The frequency domain linear stability analysis program LINTAB has been developed based on the homogeneous equilibrium results from HRTL-200 test facility have been used to validate the LINTAB code. Analysis results using LINTAB showed a good agreement with the test results. (author). 20 refs., 24 figs

Validation of the RELAP5 code for the modeling of flashing-induced instabilities under natural-circulation conditions using experimental data from the CIRCUS test facility

Energy Technology Data Exchange (ETDEWEB)

Kozmenkov, Y. [Helmholtz-Zentrum Dresden-Rossendorf e.V. (FZD), Institute of Safety Research, P.O.B. 510119, D-01324 Dresden (Germany); Institute of Physics and Power Engineering, Obninsk (Russian Federation); Rohde, U., E-mail: U.Rohde@hzdr.de [Helmholtz-Zentrum Dresden-Rossendorf e.V. (FZD), Institute of Safety Research, P.O.B. 510119, D-01324 Dresden (Germany); Manera, A. [Paul Scherrer Institute (Switzerland)

2012-02-15

Highlights: Black-Right-Pointing-Pointer We report about the simulation of flashing-induced instabilities in natural circulation systems. Black-Right-Pointing-Pointer Flashing-induced instabilities are of relevance for operation of pool-type reactors of small power at low pressure. Black-Right-Pointing-Pointer The RELAP5 code is validated against measurement data from natural circulation experiments. Black-Right-Pointing-Pointer The magnitude and frequency of the oscillations were reproduced in good agreement with the measurement data. - Abstract: This paper reports on the use of the RELAP5 code for the simulation of flashing-induced instabilities in natural circulation systems. The RELAP 5 code is intended to be used for the simulation of transient processes in the Russian RUTA reactor concept operating at atmospheric pressure with forced convection of coolant. However, during transient processes, natural circulation with flashing-induced instabilities might occur. The RELAP5 code is validated against measurement data from natural circulation experiments performed 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. It was shown that the RELAP5 code is able to model all relevant phenomena related to flashing induced instabilities. The magnitude and frequency of the oscillations were reproduced in a good agreement with the measurement data. The close correspondence to the experiments was reached by detailed modeling of all components of the CIRCUS facility including the heat exchanger, the buffer vessel and the steam dome at the top of the facility.

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

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)

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

Hydrodynamics of a hybrid circulating fluidized bed reactor with a partitioned loop seal system

Energy Technology Data Exchange (ETDEWEB)

Bae, Dal-Hee; Moon, Jong-Ho; Jin, Gyoung Tae; Shun, Dowon [Korea Institute of Energy Research, Daejeon (Korea, Republic of); Yun, Minyoung; Park, Chan Seung; Norbeck, Joseph M. [University of California, Riverside (United States)

2015-07-15

A circulating fluidized bed (CFB) with a hybrid design has been developed and optimized for steam hydrogasification. The hybrid CFB is composed of a bubbling fluidized bed (BFB) type combustor and a fast fluidized bed (FB) type gasifier. Char is burnt in the combustor and the generated heat is supplied to the gasifier along with the bed materials. Two different types of fluidized beds are connected to each other with a newly developed partitioned loop seal to avoid direct contact between two separate gas streams flowing in each fluidized bed. Gas mixing tests were carried out with Air and Argon in a cold model hybrid CFB to test the loop seal efficiency. Increase in solid inventory in the loop seal can improve the gas separation efficiency. It can be realized at higher gas velocity in fast bed and with higher solid inventory in the loop seal system. In addition, bed hydrodynamics was investigated with varying gas flow conditions and particle sizes in order to obtain a full understanding of changes of solid holdup in the FB. The solid holdup in the FB increased with increasing gas velocity in the BFB. Conversely, increase in gas velocity in the FB contributed to reducing the solid holdup in the FB. It was observed that changing the particle size of bed material does not have a big impact on hydrodynamic parameters.

Hydrodynamics of a hybrid circulating fluidized bed reactor with a partitioned loop seal system

International Nuclear Information System (INIS)

Bae, Dal-Hee; Moon, Jong-Ho; Jin, Gyoung Tae; Shun, Dowon; Yun, Minyoung; Park, Chan Seung; Norbeck, Joseph M.

2015-01-01

A circulating fluidized bed (CFB) with a hybrid design has been developed and optimized for steam hydrogasification. The hybrid CFB is composed of a bubbling fluidized bed (BFB) type combustor and a fast fluidized bed (FB) type gasifier. Char is burnt in the combustor and the generated heat is supplied to the gasifier along with the bed materials. Two different types of fluidized beds are connected to each other with a newly developed partitioned loop seal to avoid direct contact between two separate gas streams flowing in each fluidized bed. Gas mixing tests were carried out with Air and Argon in a cold model hybrid CFB to test the loop seal efficiency. Increase in solid inventory in the loop seal can improve the gas separation efficiency. It can be realized at higher gas velocity in fast bed and with higher solid inventory in the loop seal system. In addition, bed hydrodynamics was investigated with varying gas flow conditions and particle sizes in order to obtain a full understanding of changes of solid holdup in the FB. The solid holdup in the FB increased with increasing gas velocity in the BFB. Conversely, increase in gas velocity in the FB contributed to reducing the solid holdup in the FB. It was observed that changing the particle size of bed material does not have a big impact on hydrodynamic parameters

Simulations of a Circulating Fluidized Bed Chemical Looping Combustion System Utilizing Gaseous Fuel; Simulation de la combustion en boucle chimique d'une charge gazeuse dans un lit fluidise circulant

Energy Technology Data Exchange (ETDEWEB)

Mahalatkar, K.; Kuhlman, J. [West Virginia University, Dept. of Mechanical and Aerospace Engineering, Morgantown, WV, 26506 (United States); Mahalatkar, K. [ANSYS Inc., 3647 Collins Ferry Road Suite A, Morgantown, WV, 26505 (United States); Kuhlman, J.; Huckaby, E.D.; O' Brien, T. [National Energy Technology Laboratory, 3610 Collins Ferry Road, Morgantown, WV, 26507 (United States)

2011-03-15

Numerical studies using Computational Fluid Dynamics (CFD) have been carried out for a complete circulating fluidized bed chemical looping combustor described in the literature (Abad et al., 2006 Fuel 85, 1174-1185). There have been extensive experimental studies in Chemical Looping Combustion (CLC), however CFD simulations of this concept are quite limited. The CLC experiments that were simulated used methane as fuel. A 2-D continuum model was used to describe both the gas and solid phases. Detailed sub-models to account for fluid-particle and particle-particle interaction forces were included. Global models of fuel and carrier chemistry were utilized. The results obtained from CFD were compared with experimental outlet species concentrations, solid circulation rates, solid mass distribution in the reactors, and leakage and dilution rates. The transient CFD simulations provided a reasonable match with the reported experimental data. (authors)

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)

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)

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

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.

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.

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

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

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)

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.

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.

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

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

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)

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)

Experimental and numerical study on single-phase flow characteristics of natural circulation system with heated narrow rectangular channel under rolling motion condition

International Nuclear Information System (INIS)

Yu, Shengzhi; Wang, Jianjun; Yan, Ming; Yan, Changqi; Cao, Xiaxin

2017-01-01

Highlights: • The phasic difference between flow rate and frictional pressure drop is negligible. • Effect mechanism of rolling motion on flow behaviors of NC is interpreted. • The startup model is proposed and verified. • Steady-state correlations are feasible to predict transient resistance. • The in-house code can simulate instantaneous flow behaviors of NC correctly. - Abstract: Effects of rolling motion on flow characteristics in a natural circulation system were investigated experimentally and numerically. The numerical results from validated code were mainly used to provide detailed information for the discussion and analysis of experimental results. The results indicate that under rolling motion condition, the phasic difference between flow rate and frictional pressure drop of narrow rectangular channel is negligible. Angular acceleration is the eigenvalue for the effects of rolling motion on flow rate under single-phase natural circulation condition. When angular acceleration is approximately equal, even though either the angle or the period of rolling motion is different, peak, trough and time-averaged values of flow rate are approximately equal. Under rolling motion and single-phase natural circulation conditions, the phenomenon that dimensionless time-averaged mass flow rate is smaller than that under steady state condition is controlled by the nonlinear relationship between mass flow rate and the resistance of loop. The factor also causes the result that the absolute difference of dimensionless flow rate between peak and steady state is smaller than that between trough and steady state. The startup model which is proposed in present paper can be used to predict the flow characteristics of single-phase natural circulation system at startup stage of rolling motion favorably. The self-developed code can simulate instantaneous flow characteristics of single-phase natural circulation system under rolling motion and steady state conditions

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 ...

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.

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

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)

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.

Naturalness made easy: two-loop naturalness bounds on minimal SM extensions

Energy Technology Data Exchange (ETDEWEB)

Clarke, Jackson D.; Cox, Peter [ARC Centre of Excellence for Particle Physics at the Terascale,School of Physics, University of Melbourne,Melbourne, 3010 (Australia)

2017-02-24

The main result of this paper is a collection of conservative naturalness bounds on minimal extensions of the Standard Model by (vector-like) fermionic or scalar gauge multiplets. Within, we advocate for an intuitive and physical concept of naturalness built upon the renormalisation group equations. In the effective field theory of the Standard Model plus a gauge multiplet with mass M, the low scale Higgs mass parameter is a calculable function of (MS)-bar input parameters defined at some high scale Λ{sub h}>M. If the Higgs mass is very sensitive to these input parameters, then this signifies a naturalness problem. To sensibly capture the sensitivity, it is shown how a sensitivity measure can be rigorously derived as a Bayesian model comparison, which reduces in a relevant limit to a Barbieri-Giudice-like fine-tuning measure. This measure is fully generalisable to any perturbative EFT. The interesting results of our two-loop renormalisation group study are as follows: for Λ{sub h}=Λ{sub Pl} we find “10% fine-tuning” bounds on the masses of various gauge multiplets of Mloop.

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...

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

Flow rate and temperature characteristics in steady state condition on FASSIP-01 loop during commissioning

Science.gov (United States)

Juarsa, M.; Giarno; Rohman, A. N.; Heru K., G. B.; Witoko, J. P.; Sony Tjahyani, D. T.

2018-02-01

The need for large-scale experimental facilities to investigate the phenomenon of natural circulation flow rate becomes a necessity in the development of nuclear reactor safety management. The FASSIP-01 loop has been built to determine the natural circulation flow rate performance in the large-scale media and aimed to reduce errors in the results for its application in the design of new generation reactors. The commissioning needs to be done to define the capability of the FASSIP-01 loop and to prescribe the experiment limitations. On this commissioning, two scenarios experimental method has been used. The first scenario is a static condition test which was conducted to verify measurement system response during 24 hours without electrical load in heater and cooler, there is water and no water inside the rectangular loop. Second scenario is a dynamics condition that aims to understand the flow rate, a dynamic test was conducted using heater power of 5627 watts and coolant flow rate in the HSS loop of 9.35 LPM. The result of this test shows that the temperature characterization on static test provide a recommendation, that the experiments should be done at night because has a better environmental temperature stability compared to afternoon, with stable temperature around 1°C - 3°C. While on the dynamic test, the water temperature difference between the inlet-outlets in the heater area is quite large, about 7 times the temperature difference in the cooler area. The magnitude of the natural circulation flow rate calculated is much larger at about 300 times compared to the measured flow rate with different flow rate profiles.

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)

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.

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

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)

CFD investigations of natural circulation between the RPV and the cooling pond of VVER-440 type reactors in incidental conditions during maintenance performed with the code CFX-4.3

International Nuclear Information System (INIS)

Legradi, G.; Aszodi, A.

2002-01-01

During the annual maintenance of the VVER-440 type reactors, the RPV, the cooling pond and the transfer pond form a connected flow domain. The reactor is cooled by the natural circulation, which develops in one or two main loops. The cooling pond has its own cooling loops. CFD calculations have been performed with the CFX-4.3 code to investigate whether it is possible to cool the reactor core in case the main loops are lost and other emergency systems are not available. The results point out that the cooling system of the cooling pond is not capable of cooling the reactor core with the present connection. Therefore, modifications of the cooling system are investigated which would make it suitable for removing the remanent heat from the core.(author)

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

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)

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...

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.

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

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)

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)

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)

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)

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.

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)

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

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)

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)

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

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

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.

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

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.

Topologically protected loop flows in high voltage AC power grids

International Nuclear Information System (INIS)

Coletta, T; Delabays, R; Jacquod, Ph; Adagideli, I

2016-01-01

Geographical features such as mountain ranges or big lakes and inland seas often result in large closed loops in high voltage AC power grids. Sizable circulating power flows have been recorded around such loops, which take up transmission line capacity and dissipate but do not deliver electric power. Power flows in high voltage AC transmission grids are dominantly governed by voltage angle differences between connected buses, much in the same way as Josephson currents depend on phase differences between tunnel-coupled superconductors. From this previously overlooked similarity we argue here that circulating power flows in AC power grids are analogous to supercurrents flowing in superconducting rings and in rings of Josephson junctions. We investigate how circulating power flows can be created and how they behave in the presence of ohmic dissipation. We show how changing operating conditions may generate them, how significantly more power is ohmically dissipated in their presence and how they are topologically protected, even in the presence of dissipation, so that they persist when operating conditions are returned to their original values. We identify three mechanisms for creating circulating power flows, (i) by loss of stability of the equilibrium state carrying no circulating loop flow, (ii) by tripping of a line traversing a large loop in the network and (iii) by reclosing a loop that tripped or was open earlier. Because voltages are uniquely defined, circulating power flows can take on only discrete values, much in the same way as circulation around vortices is quantized in superfluids. (paper)

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

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.

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.

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.

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

International Nuclear Information System (INIS)

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

2009-01-01

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

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)

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

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

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)

CaMn0.875Ti0.125O3 as oxygen carrier for chemical-looping combustion with oxygen uncoupling (CLOU)—Experiments in a continuously operating fluidized-bed reactor system

KAUST Repository

Rydé n, Magnus; Lyngfelt, Anders; Mattisson, Tobias

2011-01-01

Particles of the perovskite material CaMn0.875Ti0.125O3 has been examined as oxygen carrier for chemical-looping with oxygen uncoupling, and for chemical-looping combustion of natural gas, by 70h of experiments in a circulating fluidized-bed reactor

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

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.

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)

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

Reactor recirculation pump test loop

International Nuclear Information System (INIS)

Taka, Shusei; Kato, Hiroyuki

1979-01-01

A test loop for a reactor primary loop recirculation pumps (PLR pumps) has been constructed at Ebara's Haneda Plant in preparation for production of PLR pumps under license from Byron Jackson Pump Division of Borg-Warner Corporation. This loop can simulate operating conditions for test PLR pumps with 130 per cent of the capacity of pumps for a 1100 MWe BWR plant. A main loop, primary cooling system, water demineralizer, secondary cooling system, instrumentation and control equipment and an electric power supply system make up the test loop. This article describes the test loop itself and test results of two PLR pumps for Fukushima No. 2 N.P.S. Unit 1 and one main circulation pump for HAZ Demonstration Test Facility. (author)

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.

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.

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

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)

Concepts of self-acting circulation loops for downward heat transfer (reverse thermosiphons)

International Nuclear Information System (INIS)

Dobriansky, Y.

2011-01-01

This paper reviews the scientific and technical knowledge related to general self-acting flow loops (thermosiphons and heat pipes) that transmit heat upwards and self-acting reverse flow loops that transmit heat downwards. This paper classifies the heat and mass transfer processes that take place in general flow loops and analyses the nomenclature applied in the literature. It also presents the principles of operation of sixteen reverse flow loops; four of the loops are powered by an external source of energy, while the remaining loops are self-acting. Of the self-acting loops, vapor was used for heat transfer in seven of them and liquid was used in the remaining ones. Based on the available research results, a list of the advantages and disadvantages of both types of loops is presented.

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.

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.

Capillary-Condenser-Pumped Heat-Transfer Loop

Science.gov (United States)

Silverstein, Calvin C.

1989-01-01

Heat being transferred supplies operating power. Capillary-condenser-pumped heat-transfer loop similar to heat pipe and to capillary-evaporator-pumped heat-transfer loop in that heat-transfer fluid pumped by evaporation and condensation of fluid at heat source and sink, respectively. Capillary condenser pump combined with capillary evaporator pump to form heat exchanger circulating heat-transfer fluids in both loops. Transport of heat more nearly isothermal. Thermal stress in loop reduced, and less external surface area needed in condenser section for rejection of heat to heat sink.

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)

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)

LMFBR with booster pump in pumping loop

International Nuclear Information System (INIS)

Rubinstein, H.J.

1975-01-01

A loop coolant circulation system is described for a liquid metal fast breeder reactor (LMFBR) utilizing a low head, high specific speed booster pump in the hot leg of the coolant loop with the main pump located in the cold leg of the loop, thereby providing the advantages of operating the main pump in the hot leg with the reliability of cold leg pump operation

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

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

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)

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

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)

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

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)

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)

Predictive models of circulating fluidized bed combustors: SO[sub 2] sorption in the CFB loop

Energy Technology Data Exchange (ETDEWEB)

Gidaspow, D.; Therdthianwong, A. (Illinois Inst. of Tech., Chicago, IL (United States). Dept. of Chemical Engineering)

1993-02-01

The overall objective of this investigation is to develop experimentally verified models for circulating fluidized bed (CFB) combustors. Sorption of S0[sub 2] with calcined limestone was studied in a PYROFLOW type CFB loop at conditions approximating those found in a CFB combustor. Initially the CFB loop contained 150 micron CaO particles of a density of 3.3 g/cm[sup 3] and air at 1143[degrees]K and 3.25 atm. Atzero time, air containing 600 ppm SO[sub 2], was introduced into the riser bottom at 1143[degrees]K. The effect of gas velocity, sorbent inventory and inlet pressure on the sorption of SO[sub 2], were studied isothermally by running our hydrodynamic code with the S0[sub 2] sorption conservation of species equation. At a velocity of 5m/sec., reported to be a typical velocity by PYROPOWER, there is reasonably good S0[sub 2] removal. At 10 m/sec the S0[sub 2] removal is poor. The best SO[sub 2], removal is for a velocity of 5 m/s and a high bed inventory, initial bed height, H = 9m. Most of the S0[sub 2] is removed in the first two meters of the reactor. However, the S0[sub 2] removal is not complete at the bed outlet. This is due to mixing. At the left wall of the reactor (wall opposite the solids inlet) the S0[sub 2] removal was poor due to gas bypassing caused by the asymmetrical solids inlet. Simulation of the PYROPOWER loop with a symmetrical inlet gave us an order of magnitude improvement over the conventional PYROPOWER system. These results demonstrate the practical utility of the predictive model that we have developed over the last three years.

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)

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

Design of the Natural Circulation Loop and Implementation of DOWTHERM A Properties into MARS-LMR Code

Energy Technology Data Exchange (ETDEWEB)

Shin, Yukyung; Park, Seong Dae; Kang, Sarah; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

2015-05-15

Molten Salt Reactor (MSR), which is one of the generation IV reactors, has an advantage in these requirements. MSR uses a molten salt mixture as the primary coolant, or the fuel itself and it operates on high temperature, so it doesn't need pressurizing. Also, liquid state fuel has an advantage for pyro-processing with easy separation of fission products. These fission products also have relatively short half-lives compared to those of the existing reactors. With these characteristics, MSR can have inherent safety in both direct and indirect sides. Also, MSR can operate at high temperature range, so that it can have the high efficiency to produce electricity. Therefore, research of MSR is meaningful for developing advanced nuclear reactors. FLiBe which is a mixture of lithium fluoride (LiF) and beryllium fluoride (BeF{sub 2}) is used as a primary coolant in MSR and LMR (Liquid Metal cooled Reactor). It has superiority over conventional liquid metal coolant like sodium, because it doesn't react with air or water. thermos-physical properties of DOWTHERM A for MARS-LMR code were made by modifying stg file of existing one. It was based on the process of Moore using 6 output parameters such as specific volume, internal energy, thermal expansion coefficient, isothermal compressibility, specific heat and entropy. With generated stg file (stgdowa.f90) and input file, tpf file (tpfdowa) which includes fluid property tables for MARS-LMR simulation was obtained. For the verification, this tpf file with execution file will be applied to the input deck of our natural circulation design. This work will contribute to researching and developing of MSR and LMR.

RAPK-7. code for calculating mass transfer and corrosion products activation in the circulation loops of water-cooled reactors

International Nuclear Information System (INIS)

Mikhaylov, A.V.; Moryakov, A.V.; Nikitin, A.V.

2012-09-01

The RAPK-7 code was developed to simulate formation of non-irradiated and activated corrosion products, their transport and deposition on inner surfaces of primary components and in primary coolant of water-cooled reactors during their operation on power and after shutdown. The key feature of this code is its particular emphasis on the contamination of circulation loops by radioactive corrosion products of reactor which operates on variable modes. Such reactors typically are: research reactors and their experimental loops, naval nuclear power systems, etc. It's typical for such reactors to have repeated (over the campaign) and frequent variations in power (activating neutron fluxes), thermal-physical, hydrodynamic and other parameters of coolant, intensive water mass exchange between the circulation loop and the pressuriser, etc. The processes of mass-transfer are described by the RAPK-7 code with the use of models similar to those employed by the COTRAN and PACTOLE codes. The circulation circuit is broken down into computation areas. The user will then set the concentrations of water chemistry adjusting additives (alkali, boric acid, ammonia, hydrogen), as well as parameters in each area, such as wall temperature, coolant flow core temperature, pressure, flow rate, velocity, the radial component of coolant flowrate and activating neutron flux density. All the above parameters can be set as time-dependent step functions (bar charts), with independent time steps for each of them. The number of computation areas, the number of time dependencies and the level of detail in their description are limited by computer capabilities only. A 'brake' mode with a single-step change of the required set of parameters is provided to allow for jump-type events, such as replacement of contaminated components with clean ones during core refueling or repairs, emergency injection of boric acid, water mass exchange between the circulation circuit and the pressuriser, etc

High pressure experimental water loop

International Nuclear Information System (INIS)

Grenon, M.

1958-01-01

A high pressure experimental water loop has been made for studying the detection and evolution of cladding failure in a pressurized reactor. The loop has been designed for a maximum temperature of 360 deg. C, a maximum of 160 kg/cm 2 and flow rates up to 5 m 3 /h. The entire loop consists of several parts: a main circuit with a canned rotor circulation pump, steam pressurizer, heating tubes, two hydro-cyclones (one de-gasser and one decanter) and one tubular heat exchanger; a continuous purification loop, connected in parallel, comprising pressure reducing valves and resin pots which also allow studies of the stability of resins under pressure, temperature and radiation; following the gas separator is a gas loop for studying the recombination of the radiolytic gases in the steam phase. The preceding circuits, as well as others, return to a low pressure storage circuit. The cold water of the low pressure storage flask is continuously reintroduced into the high pressure main circuit by means of a return pump at a maximum head of 160 kg /cm 2 , and adjusted to the pressurizer level. This loop is also a testing bench for the tight high pressure apparatus. The circulating pump and the connecting flanges (Oak Ridge type) are water-tight. The feed pump and the pressure reducing valves are not; the un-tight ones have a system of leak recovery. To permanently check the tightness the circuit has been fitted with a leak detection system (similar to the HRT one). (author) [fr

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.

On the Loop Current Penetration into the Gulf of Mexico

Science.gov (United States)

Weisberg, Robert H.; Liu, Yonggang

2017-12-01

The Gulf of Mexico Loop Current generally intrudes some distance into the Gulf of Mexico before shedding an anticyclonic eddy and retreating back to its more direct entry to exit pathway. The control of this aperiodic process remains only partially known. Here we describe the evolution of the Loop Current throughout the era of satellite altimetry, and offer a mechanistic hypothesis on Loop Current intrusion. As a complement to the known effects of Loop Current forcing on the west Florida shelf circulation, we argue that the west Florida shelf, in turn, impacts the Loop Current evolution. A Self-Organizing Map analysis shows that anomalous northward penetrations of the Loop Current into the Gulf of Mexico occur when the eastern side of Loop Current is positioned west from the southwest corner of the west Florida shelf, whereas the more direct inflow to outflow route occurs when the eastern side of the Loop Current comes in contact with the southwest corner of the west Florida shelf. In essence, we argue that the west Florida shelf anchors the Loop Current in its direct path configuration and that farther northward penetration into the Gulf of Mexico occurs when such anchoring is released. To test of this hypothesis heuristically, we estimate that the dissipation and buoyancy work due to known Loop Current forcing of the west Florida shelf circulation (when in contact with the southwest corner) may exceed the pressure work required for the Loop Current to advance against the ambient Gulf of Mexico fluid.Plain Language SummaryThe Gulf of Mexico Loop Current may intrude far into the Gulf of Mexico or take a more direct entry to exit pathway. Such Loop Current behaviors are described using remote observations by satellites, and a heuristic hypothesis on the control of Loop Current intrusion is presented. We argue that energy dissipation and buoyancy work by the west Florida shelf circulation, when the Loop Current contacts the southwest corner of the west Florida shelf

Benchmark analysis of three main circulation pump sequential trip event at Ignalina NPP

International Nuclear Information System (INIS)

Uspuras, E.; Kaliatka, A.; Urbonas, R.

2001-01-01

The Ignalina Nuclear Power Plant is a twin-unit with two RBMK-1500 reactors. The primary circuit consists of two symmetrical loops. Eight Main Circulation Pumps (MCPs) at the Ignalina NPP are employed for the coolant water forced circulation through the reactor core. The MCPs are joined in groups of four pumps for each loop (three for normal operation and one on standby). This paper presents the benchmark analysis of three main circulation pump sequential trip event at RBMK-1500 using RELAP5 code. During this event all three MCPs in one circulation loop at Unit 2 Ignalina NPP were tripped one after another, because of inadvertent activation of the fire protection system. The comparison of calculated and measured parameters led us to establish realistic thermal hydraulic characteristics of different main circulation circuit components and to verify the model of drum separators pressure and water level controllers.(author)

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.

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)

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)

Computational simulation of natural circulation and rewetting experiments using the TRAC/PF1 code; Simulacao computacional dos experimentos: circulacao natural no CTE-150 e remolhamento na ITR utilizando o TRAC-PF1

Energy Technology Data Exchange (ETDEWEB)

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.

On natural circulation in High Temperature Gas-Cooled Reactors and pebble bed reactors for different flow regimes and various coolant gases

International Nuclear Information System (INIS)

Melesed'Hospital, G.

1983-01-01

The use of CO 2 or N 2 (heavy gas) instead of helium during natural circulation leads to improved performance in both High Temperature Gas-Cooled Reactors (HTGR) and in Pebble Bed Reactors (PBR). For instance, the coolant temperature rise corresponding to a coolant pressure level and a rate of afterheat removal could be only 18% with CO 2 as compared to He, for laminar flow in HTGR; this value would be 40% in PBR. There is less difference between HTGR and PBR for turbulent flows; CO 2 is found to be always better than N 2 . These types of results derived from relationships between coolant properties, coolant flow, temperature rise, pressure, afterheat levels and core geometry, are obtained for HTGR and PBR for various flow regimes, both within the core and in the primary loop

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.

Operational characteristics of miniature loop heat pipe with flat evaporator

Energy Technology Data Exchange (ETDEWEB)

Gai, Dongxing; Liu, Zhichun; Liu, Wei; Yang, Jinguo [Huazhong University of Science and Technology, School of Energy and Power Engineering, Wuhan, Hubei (China)

2009-12-15

Loop heat pipes are heat transfer devices whose operating principle is based on the evaporation and condensation of a working fluid, and which use the capillary pumping forces to ensure the fluid circulation. A series of tests have been carried out with a miniature loop heat pipe (mLHP) with flat evaporator and fin-and-tube type condenser. The loop is made of pure copper with stainless mesh wick and methanol as the working fluid. Detailed study is conducted on the start-up reliability of the mLHP at high as well as low heat loads. During the testing of mLHP under step power cycles, the thermal response presented by the loop to achieve steady state is very short. At low heat loads, temperature oscillations are observed throughout the loop. The amplitudes and frequencies of these fluctuations are large at evaporator wall and evaporator inlet. It is expected that the extent and nature of the oscillations occurrence is dependent on the thermal and hydrodynamic conditions inside the compensation chamber. The thermal resistance of the mLHP lies between 0.29 and 3.2 C/W. The effects of different liquid charging ratios and the tilt angles to the start-up and the temperature oscillation are studied in detail. (orig.)

A recent perspective of the Circulation in the Gulf of Mexico

Science.gov (United States)

Candela, J.; Ochoa-de-La-Torre, J. L.; Sheinbaum, J.; Perez-Brunius, P.; Pallas-Sanz, E.; Kolodziejczyk, N.

2013-05-01

The flow through the Gulf of Mexico is an integral part of the North Atlantic Ocean Subtropical Gyre Circulation, known to be forced by the wind over the North Atlantic and by an equivalent contribution from the inter-hemispheric Meridional Overturning Cell. To the North Atlantic Circulation, the Gulf represents an important energy and vorticity sink through the particular behavior of the Loop Current within. Comprehending the structure and dynamics of the Loop Current System (which includes the Yucatan Current (YC), Loop Current (LC), the shedding of anticyclonic Loop Current Eddies (LCE) and peripheral cyclonic gyres) is fundamental for understanding the circulation in the entire Gulf. Within the Gulf, the eastern Loop Current and the western Campeche Bay (CB) regions are characterized by persistent eddy structures, with less structured eddy fields in between. Of these later ones, the northwestern Gulf is a geostrophic turbulence area, constantly perturbed by LCE, which represents, at the same time, an important dissipation and circulation forcing region for the Western Gulf. Important processes recently investigated that will be discussed: 1) The generation and maintenance of the Geostrophic Turbulence field in the north western Gulf. 2) The circulation in deep water induced by the surface geostrophic turbulence field. 3) The generation of intensive jets at depth by interaction of +/- gyres. 4) The generation of coastal trap waves by the interaction of LCEs with the western shelf. 5) The generation of deep topographic Roosby waves by topographic interactions of the LCEs with topography. 5) The characteristics of the Bay of Campeche Circulation, The Campeche Gyre and its interaction with LCEs. 6) The Gulf's response to the passage of hurricanes. 7) The trapping of inertial waves by the LCEs and the related enhanced mixing.

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

Simulation of natural convection in a rectangular loop using finite elements

International Nuclear Information System (INIS)

Pepper, D.W.; Hamm, L.L.; Kehoe, A.B.

1984-01-01

A two-dimensional finite-element analysis of natural convection in a rectangular loop is presented. A psi-omega formulation of the Boussinesque approximation to the Navier-Stokes equation is solved by the false transient technique. Streamlines and isotherms at Ra = 10 4 are shown for three different modes of heating. The results indicate that corner effects should be considered when modeling flow patterns in thermosyphons

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)

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)

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

Main circulator design features for HTR 100, HTR 500 and space heating plants

International Nuclear Information System (INIS)

Engel, J.; Glass, D.

1988-01-01

All design alternatives for modern high-temperature reactors have a common circulator concept: It is based on a vertical shaft design with a flying impeller. The circulators are equipped with active magnetic bearings and are driven by induction motors connected to variable-speed static converters. Due to their multiple functions during normal reactor operation and under accident conditions, extremely high requirements are made to safety-relevant circulators, since with the reactor pressurized as well as under depressurized conditions specified delivery heads and flow rates have to be ensured. The use of active magnetic bearings permits to obtain maintenance-free operation and functional safety to an extent which had not been achieved before. Magnetic bearings are therefore provided for the total range including primary gas circulators of a drive power of several MW as well as circulators for helium loops of reactor auxiliary systems. The essential feature for using active magnetic bearings is the retainer bearing technology, preventing contact between rotor and static circulator parts upon unintended deenergisation of the magnets. Results of current experiments are reported. Another aspect to be considered for reliable long-term operation for several decades is the effect of rotor dynamics. The various natural frequencies resulting from torsion and bending modes in view of a drive by a frequency-controlled induction motor have to be considered as well as the specific characteristics of the active magnetic bearings. Special attention has to be directed to the internal cooling loop so as to ensure that reactor temperature excursions in the event of deviation from normal operation can be overcome without damage. For circulator components exposed to temperature fields the design characteristics are determined by combining experimental and analytical methods. The coordination of all component parts is currently being optimized on a prototype circulator whose detailed

Effect of Mixing on Microorganism Growth in Loop Bioreactors

Directory of Open Access Journals (Sweden)

A. M. Al Taweel

2012-01-01

Full Text Available The impact of mixing on the promotion of microorganism growth rate has been analyzed using a multiphase forced-circulation pipe-loop reactor model capable of identifying conditions under which it is possible to convert natural gas into Single-Cell Protein. The impact of mixing in the interphase mass transfer was found to exert a critical role in determining the overall productivity of the bioreactor, particularly at the high cell loadings needed to reduce the capital costs associated with the large-scale production needed for the production of relatively low-value SCP in a sustainable manner.

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)

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)

Cryogenic Loop Heat Pipes for the Cooling of Small Particle Detectors at CERN

OpenAIRE

Pereira, H; Haug, F; Silva, P; Wu, J; Koettig, T

2010-01-01

The loop heat pipe (LHP) is among the most effective heat transfer elements. Its principle is based on a continuous evaporation/condensation process and its passive nature does not require any mechanical devices such as pumps to circulate the cooling agent. Instead a porous wick structure in the evaporator provides the capillary pumping forces to drive the fluid [1]. Cryogenic LHP are investigated as potential candidates for the cooling of future small-scale particle detectors and upgrades of...

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)

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)

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

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)

Effect of steam quality on two—phase flow in a netural circulation loop

Institute of Scientific and Technical Information of China (English)

贾海军; 吴少融; 等

1996-01-01

Test pressures are 1.0-4.0MPa,heating powers 27-190kW,inlet subcoolings 5-80℃,water used as coolant,and steam quality at the outlet of test section is less than 0.05,These test conditions cover the parameters for a typical 200MW heating reactor.The experimental results show that the stema quality is the dominant factor in a natural circulation system with low pressure and low steam quality about the effect of system pressure,heating power and inlet subcooling on the flow rate,relative oscilatroy amplitude and oscilatory region of flow rate.

Effect of non-condensable gas on steady-state operation of a loop thermosyphon

International Nuclear Information System (INIS)

He, Jiang; Lin, Guiping; Bai, Lizhan; Miao, Jianyin; Zhang, Hongxing; Wang, Lu

2014-01-01

Non-condensable gas (NCG) generated inside two-phase heat transfer devices can adversely affect the thermal performance and limit the lifetime of such devices. In this work, extensive experimental investigation of the effect of NCG on the steady-state operation of an ammonia-stainless steel loop thermosyphon was conducted. In the experiments, nitrogen was injected into the loop thermosyphon as NCG, and the thermal performance of the loop thermosyphon was tested at different NCG inventories, heat loads applied to the evaporator and condenser cooling conditions, i.e. natural air cooling or circulating ethanol cooling. Experimental results reveal that NCG elevates the steady-state operating temperature of the evaporator, especially when the loop thermosyphon is operating in the low temperature range; meanwhile, the more NCG exists in the loop thermosyphon, the higher the operating temperature of the evaporator, and the lower the reservoir temperature. In addition, the existence of NCG results in the decrease of the overall thermal conductance of the loop thermosyphon, and the overall thermal conductance under the ethanol cooling condition may be even lower than that under the air cooling condition when the heat load is smaller than a certain value. Finally, the experimental results are theoretically analysed and explained. (authors)

MEASUREMENTS OF HYDRQDYNAMIC INSTABILITIES, FLOWOSCILLATIONS AND BURNOUT IN A NATURAL CIRCULATIONLOOP

International Nuclear Information System (INIS)

Becker, Kurt M.; Mathisen, R.P.; Eklind, O.; Norman, B.

1964-01-01

The hydrodynamic stability and the burnout conditions for flow of boiling water have been studied in a natural circulation loop in the pressure range from 10 to 70 atg. The test section was a round, duct of 20 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, increasing throttling before the test section, but decreases with increasing inlet sub-cooling and increasing throttling after the test section. The measured thresholds of instability compared well with the analytical results by Jahnberg. For an inlet sub-cooling temperature of about 2 deg C the measured burnout steam qualities were low by a factor of about 1.3 compared to forced circulation data obtained with the same test section. At higher sub-cooling temperatures the discrepancy between forced and natural circulation data increased, so that at Δt sub = 16 deg C, the natural circulation data were low by a factor of about 2.5. However, by applying inlet throttling of the flow the burnout values approached and finally coincided with the forced circulation data

Simulations of a Circulating Fluidized Bed Chemical Looping Combustion System Utilizing Gaseous Fuel Simulation de la combustion en boucle chimique d’une charge gazeuse dans un lit fluidisé circulant

Directory of Open Access Journals (Sweden)

Mahalatkar K.

2011-05-01

Full Text Available Numerical studies using Computational Fluid Dynamics (CFD have been carried out for a complete circulating fluidized bed chemical looping combustor described in the literature (Abad et al., 2006 Fuel 85, 1174-1185. There have been extensive experimental studies in Chemical Looping Combustion (CLC, however CFD simulations of this concept are quite limited. The CLC experiments that were simulated used methane as fuel. A 2-D continuum model was used to describe both the gas and solid phases. Detailed sub-models to account for fluid-particle and particleparticle interaction forces were included. Global models of fuel and carrier chemistry were utilized. The results obtained from CFD were compared with experimental outlet species concentrations, solid circulation rates, solid mass distribution in the reactors, and leakage and dilution rates. The transient CFD simulations provided a reasonable match with the reported experimental data. Des études numériques de simulation des écoulements (CFD ont été réalisées sur un lit fluidisé circulant opérant en combustion par boucle chimique (CLC décrit dans la littérature (Abad et al., 2006 Fuel 85, 1174-1185. Si de nombreuses études expérimentales ont été conduites pour étudier le procédé CLC, les études concernant la simulation des écoulements par CFD de ce concept sont très limitées. Le système de combustion en boucle chimique simulé dans cette étude concerne la combustion d’une charge gazeuse (méthane. Un modèle 2-D à deux phases continues a été utilisé pour décrire les phases gaz et solide avec des sous-modèles détaillés pour décrire les forces d’interactions entre fluideparticule et particule-particule. Des modèles cinétiques globaux ont été intégrés pour décrire les réactions de combustion et de transformation du matériau transporteur d’oxygène. Les résultats obtenus par CFD ont été comparés aux concentrations expérimentales mesurées des diff

Hydrogen production from natural gas using an iron-based chemical looping technology: Thermodynamic simulations and process system analysis

International Nuclear Information System (INIS)

Kathe, Mandar V.; Empfield, Abbey; Na, Jing; Blair, Elena; Fan, Liang-Shih

2016-01-01

Highlights: • Design of iron-based chemical looping process using moving bed for H_2 from CH_4. • Auto-thermal operation design using thermodynamic rationale for 90% carbon capture. • Cold gas efficiency: 5% points higher than Steam Methane Reforming baseline case. • Net thermal efficiency: 6% points higher than Steam Methane Reforming baseline case. • Sensitivity analysis: Energy recovery scheme, operating pressure, no carbon capture. - Abstract: Hydrogen (H_2) is a secondary fuel derived from natural gas. Currently, H_2 serves as an important component in refining operations, fertilizer production, and is experiencing increased utilization in the transportation industry as a clean combustion fuel. In recent years, industry and academia have focused on developing technology that reduces carbon emissions. As a result, there has been an increase in the technological developments for producing H_2 from natural gas. These technologies aim to minimize the cost increment associated with clean energy production. The natural gas processing chemical looping technology, developed at The Ohio State University (OSU), employs an iron-based oxygen carrier and a novel gas–solid counter-current moving bed reactor for H_2 production. Specifically, this study examines the theoretical thermodynamic limits for full conversion of natural gas through iron-based oxygen carrier reactions with methane (CH_4), by utilizing simulations generated with ASPEN modeling software. This study initially investigates the reducer and the oxidizer thermodynamic phase diagrams then derives an optimal auto-thermal operating condition for the complete loop simulation. This complete loop simulation is initially normalized for analysis on the basis of one mole of carbon input from natural gas. The H_2 production rate is then scaled to match that of the baseline study, using a full-scale ASPEN simulation for computing cooling loads, water requirements and net parasitic energy consumption. The

Dynamic simulation of a circulating fluidized bed boiler system part I: Description of the dynamic system and transient behavior of sub-models

International Nuclear Information System (INIS)

Kim, Seong Il; Choi, Sang Min; Yang, Jong In

2016-01-01

Dynamic performance simulation of a CFB boiler in a commercial-scale power plant is reported. The boiler system was modeled by a finite number of heat exchanger units, which are sub-grouped into the gas-solid circulation loop, the water-steam circulation loop, and the inter-connected heat exchangers blocks of the boiler. This dynamic model is an extension from the previously reported performance simulation model, which was designed to simulate static performance of the same power plant, where heat and mass for each of the heat exchanger units were balanced for the inter-connected heat exchanger network among the fuel combustion system and the water-steam system. Dynamic performance simulation was achieved by calculating the incremental difference from the previous time step, and progressing for the next time step. Additional discretization of the heat exchanger blocks was necessary to accommodate the dynamic response of the water evaporation and natural circulation as well as the transient response of the metal temperature of the heat exchanger elements. Presentation of the simulation modeling is organized into two parts; system configuration of the model plant and the general approach of the simulation are presented along with the transient behavior of the sub-models in Part I. Dynamic sub-models were integrated in terms of the mass flow and the heat transfer for simulating the CFB boiler system. Dynamic simulation for the open loop response was performed to check the integrated system of the water-steam loop and the solid-gas loop of the total boiler system. Simulation of the total boiler system which includes the closed-loop control system blocks is presented in the following Part II

Dynamic simulation of a circulating fluidized bed boiler system part I: Description of the dynamic system and transient behavior of sub-models

Energy Technology Data Exchange (ETDEWEB)

Kim, Seong Il; Choi, Sang Min; Yang, Jong In [Dept. of Mechanical Engineering, Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2016-12-15

Dynamic performance simulation of a CFB boiler in a commercial-scale power plant is reported. The boiler system was modeled by a finite number of heat exchanger units, which are sub-grouped into the gas-solid circulation loop, the water-steam circulation loop, and the inter-connected heat exchangers blocks of the boiler. This dynamic model is an extension from the previously reported performance simulation model, which was designed to simulate static performance of the same power plant, where heat and mass for each of the heat exchanger units were balanced for the inter-connected heat exchanger network among the fuel combustion system and the water-steam system. Dynamic performance simulation was achieved by calculating the incremental difference from the previous time step, and progressing for the next time step. Additional discretization of the heat exchanger blocks was necessary to accommodate the dynamic response of the water evaporation and natural circulation as well as the transient response of the metal temperature of the heat exchanger elements. Presentation of the simulation modeling is organized into two parts; system configuration of the model plant and the general approach of the simulation are presented along with the transient behavior of the sub-models in Part I. Dynamic sub-models were integrated in terms of the mass flow and the heat transfer for simulating the CFB boiler system. Dynamic simulation for the open loop response was performed to check the integrated system of the water-steam loop and the solid-gas loop of the total boiler system. Simulation of the total boiler system which includes the closed-loop control system blocks is presented in the following Part II.

Vibration and acoustic signatures of the water circulation pump in the pressurised LMR fuel element test loop at IPEN

International Nuclear Information System (INIS)

Holland, L.

1985-01-01

This report presents results of vibration and acoustic field measurements made on the water circulating pump in the IPEN - CNEN/Sao Paulo pressurised water loop. The use of such measurements to monitor the vibration of coolant circulating pumps of light water reactors is indicated. Measurements were made for defined water flows and pressures varying between 5 bar/5.22 ls sup(-1) and 40 bar/17,42ls sup(-1). Analyses of various recordings of two accelerometer signals and 1 microphone signal were made principally in the frequency range 0-5 KHz using a Nicolet 660 A Fourier analyser. Results of these analyses indicate that CPSD distributions might be more sensitive indicators of changes in pump operating conditions than the more frequently used PSD distributions. In addition, as an indicador of changing pump conditions the acoustic-vibration signal pair is perhaps a more sensitive indicator than the vibration-vibration signal pair. While coherence distributions are elearly sensitive to changing pump conditions, trends in the change of these distributions were not readily identified. It is recommended that more detailed analyses be made using pattern recognition techniques in conjunction with frequency zooming. (Author) [pt

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.

A circulating loop tester for liquid alloyed metal of lead-bismuth

International Nuclear Information System (INIS)

Kitano, Teruaki; Ono, Mikinori; Kamata, Kinya

2002-01-01

Mitsui Engineering and Shipbuilding Co., Ltd. (MES) had focused to merits of this lead-bismuth alloy, to actively carry out many works on this field such as an experience of development of heat exchanger at industrial level of intercourse with IPPE (Institute of Physics and Power Engineering) in Russia with an experience of using results for 80 years on coolant for nuclear reactor. Before about 20 years, MES developed a heat exchanger for installation at a lead-zinc separation process in a refinery in Japan under cooperation of the Mitsui Metal and Mine Co., Ltd., to deliver it for a power generation system at the Hachinohe refinery. As the heat exchanger aims at control of cooling in the separation process, it also contributes to power generation of about 4,300 kW, and now it continues to separate and contribute to self-power generation in the refinery. The heat exchanger is filled with the liquid alloyed metal of lead-bismuth for an intermediate thermal medium in its casing. The metal has some merits such as inactivity to air and water, high boiling point (1,700 centigrade), almost no volume change at its coagulation, and its minus reactivity coefficient. However, the metal has some problems to be solved, such as its steel corrosion, its purification, and control technology. To grow up lead-bismuth technology to a nuclear energy technology in Japan, the lead-bismuth circulating loop tester was produced on May, 2001, to establish application technology on this system to nuclear energy technology in Japan. (G.K.)

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

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

Most advanced nuclear power plant designs adopted several kinds of passive systems. Natural circulation is used as a key driving force for many passive systems and even for core heat removal during normal operation such as NuScale, CAREM, ESBWR and Indian AHWR designs. Simulation of natural circulation phenomena is very challenging since the driving force of it is weak compared to forced circulation and involves a coupling between primary system and containment for integral type reactor. The IAEA ICSP (International Collaborative Standard Problem) on 'Integral PWR Design Natural Circulation Flow Stability and Thermo-hydraulic Coupling of Containment and Primary System during Accidents' was proposed within the CRP on 'Natural Circulation Phenomena, Modelling, and Reliability of Passive Systems that utilize Natural Circulation'. Oregon State University (OSU) of USA offered to host this ICSP. This ICSP plans to conduct the following experiments and blind/open simulations with system codes: 1. Quasi-steady state operation with different core power levels: Conduct quasi-steady state operation with step-wise increase of core power level in order to observe single phase natural circulation flow according to power level. The experimental facility and operating conditions for an integral PWR will be used. 2. Thermo-hydraulic Coupling between Primary system and Containment: Conduct a loss of feedwater transient with subsequent ADS blowdown and long term cooling to determine the progression of a loss of feedwater transient by natural circulation through primary and containment systems. These tests would examine the blowdown phase as well as the long term cooling using sump natural circulation by coupling the primary to containment systems. This data could be used for the evaluation of system codes to determine if they model specific phenomena in an accurate manner. OSU completed planned two ICSP tests in July 2011 and real initial and boundary conditions measured from the

Comparison of numerical results with experimental data for single-phase natural convection in an experimental sodium loop

International Nuclear Information System (INIS)

Ribando, R.J.

1979-01-01

A comparison is made between computed results and experimental data for a single-phase natural convection test in an experimental sodium loop. The test was conducted in the Thermal-Hydraulic Out-of-Reactor Safety (THORS) facility, an engineering-scale high temperature sodium loop at the Oak Ridge National Laboratory (ORNL) used for thermal-hydraulic testing of simulated Liquid Metal Fast Breeder Reactor (LMFBR) subassemblies at normal and off-normal operating conditions. Heat generation in the 19 pin assembly during the test was typical of decay heat levels. The test chosen for analysis in this paper was one of seven natural convection runs conducted in the facility using a variety of initial conditions and testing parameters. Specifically, in this test the bypass line was open to simulate a parallel heated assembly and the test was begun with a pump coastdown from a small initial forced flow. The computer program used to analyze the test, LONAC (LOw flow and NAtural Convection) is an ORNL-developed, fast-running, one-dimensional, single-phase, finite-difference model used for simulating forced and free convection transients in the THORS loop

Natural Ores as Oxygen Carriers in Chemical Looping Combustion

Energy Technology Data Exchange (ETDEWEB)

Tian, Hanjing; Siriwardane, Ranjani; Simonyi, Thomas; Poston, James

2013-08-01

Chemical looping combustion (CLC) is a combustion technology that utilizes oxygen from oxygen carriers (OC), such as metal oxides, instead of air to combust fuels. The use of natural minerals as oxygen carriers has advantages, such as lower cost and availability. Eight materials, based on copper or iron oxides, were selected for screening tests of CLC processes using coal and methane as fuels. Thermogravimetric experiments and bench-scale fixed-bed reactor tests were conducted to investigate the oxygen transfer capacity, reaction kinetics, and stability during cyclic reduction/oxidation reaction. Most natural minerals showed lower combustion capacity than pure CuO/Fe{sub 2}O{sub 3} due to low-concentrations of active oxide species in minerals. In coal CLC, chryscolla (Cu-based), magnetite, and limonite (Fe-based) demonstrated better reaction performances than other materials. The addition of steam improved the coal CLC performance when using natural ores because of the steam gasification of coal and the subsequent reaction of gaseous fuels with active oxide species in the natural ores. In methane CLC, chryscolla, hematite, and limonite demonstrated excellent reactivity and stability in 50-cycle thermogravimetric analysis tests. Fe{sub 2}O{sub 3}-based ores possess greater oxygen utilization but require an activation period before achieving full performance in methane CLC. Particle agglomeration issues associated with the application of natural ores in CLC processes were also studied by scanning electron microscopy (SEM).

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)

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)

Design and study of Engineering Test Facility - Helium Circulator

International Nuclear Information System (INIS)

Jiang Huijing; Ye Ping; Zhao Gang; Geng Yinan; Wang Jie

2015-01-01

Helium circulator is one of the key equipment of High-temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM). In order to simulate most normal and accident operating conditions of helium circulator in HTR-PM, a full scale, rated flow rate and power, engineering test loop, which was called Engineering Test Facility - Helium Circulator (ETF-HC), was designed and established. Two prototypes of helium circulator, which was supported by Active Magnetic Bearing (AMB) or sealed by dry gas seals, would be tested on ETF-HC. Therefore, special interchangeable design was under consideration. ETF-HC was constructed compactly, which consisted of eleven sub-systems. In order to reduce the flow resistance of the circuit, special ducts, elbows, valves and flowmeters were selected. Two stages of heat exchange loops were designed and a helium - high pressure pure water heat exchanger was applied to ensure water wouldn't be vaporized while simulating accident conditions. Commissioning tests were carried out and operation results showed that ETF-HC meets the requirement of helium circulator operation. On this test facility, different kinds of experiments were supposed to be held, including mechanical and aerodynamic performance tests, durability tests and so on. These tests would provide the features and performance of helium circulator and verify its feasibility, availability and reliability. (author)

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)

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)

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.)

Computational fluid dynamics simulation for chemical looping combustion of coal in a dual circulation fluidized bed

International Nuclear Information System (INIS)

Su, Mingze; Zhao, Haibo; Ma, Jinchen

2015-01-01

Highlights: • CFD simulation of a 5 kW_t_h CLC reactor of coal was conducted. • Gas leakage, flow pattern and combustion efficiency of the reactor was analyzed. • Optimal condition was achieved based on operation characteristics understanding. - Abstract: A dual circulation fluidized bed system is widely accepted for chemical looping combustion (CLC) for enriching CO_2 from the utilization of fossil fuels. Due to the limitations of the measurement, the details of multiphase reactive flows in the interconnected fluidized bed reactors are difficult to obtain. Computational Fluid Dynamics (CFD) simulation provides a promising method to understand the hydrodynamics, chemical reaction, and heat and mass transfers in CLC reactors, which are very important for the rational design, optimal operation, and scaling-up of the CLC system. In this work, a 5 kW_t_h coal-fired CLC dual circulation fluidized bed system, which was developed by our research group, was first simulated for understanding gas leakage, flow pattern and combustion efficiency. The simulation results achieved good agreement with the experimental measurements, which validates the simulation model. Subsequently, to improve the combustion efficiency, a new operation condition was simulated by increasing the reactor temperature and decreasing the coal feeding. An improvement in the combustion efficiency was attained, and the simulation results for the new operation condition were also validated by the experimental measurements in the same CLC combustor. All of the above processes demonstrated the validity and usefulness of the simulation results to improve the CLC reactor operation.

Preliminary test results and CFD analysis for moderator circulation test at Korea

Energy Technology Data Exchange (ETDEWEB)

Kim, H.T. [Korea Atomic Energy Research Inst., Daejeon (Korea, Republic of); Im, S.H.; Sung, H.J. [Korea Advanced Inst. of Science and Tech., Daejeon (Korea, Republic of); Seo, H.; Bang, I.C. [Ulsan National Inst. of Science and Tech., Ulsan (Korea, Republic of)

2014-07-01

Korea Atomic Energy Research Institute (KAERI) is carrying out a scaled-down moderator test program to simulate the CANDU6 moderator circulation phenomena during steady state operation and accident conditions. This research program includes the construction of the Moderator Circulation Test (MCT) facility, production of the validation data for self-reliant CFD tools, and development of optical measurement system using the Particle Image Velocimetry (PIV). The MCT facility includes a primary circulation loop (pipe lines, a primary side pump, a heat exchanger, valves, flow meters) and a secondary side loop (pipe lines, a secondary side pump, and an external cooling tower). The loop leakage test and non-heating test are performed in the present work. In the present work the PIV technique is used to measure the velocity distributions in the scaled moderator tank of MCT under iso-thermal test conditions. The preliminary PIV measurement data are obtained and compared with CFX code predictions. (author)

Preliminary test results and CFD analysis for moderator circulation test at Korea

International Nuclear Information System (INIS)

Kim, H.T.; Im, S.H.; Sung, H.J.; Seo, H.; Bang, I.C.

2014-01-01

Korea Atomic Energy Research Institute (KAERI) is carrying out a scaled-down moderator test program to simulate the CANDU6 moderator circulation phenomena during steady state operation and accident conditions. This research program includes the construction of the Moderator Circulation Test (MCT) facility, production of the validation data for self-reliant CFD tools, and development of optical measurement system using the Particle Image Velocimetry (PIV). The MCT facility includes a primary circulation loop (pipe lines, a primary side pump, a heat exchanger, valves, flow meters) and a secondary side loop (pipe lines, a secondary side pump, and an external cooling tower). The loop leakage test and non-heating test are performed in the present work. In the present work the PIV technique is used to measure the velocity distributions in the scaled moderator tank of MCT under iso-thermal test conditions. The preliminary PIV measurement data are obtained and compared with CFX code predictions. (author)

Bed system performance in helium circulation mode

Energy Technology Data Exchange (ETDEWEB)

Kim, Yean Jin; Jung, Kwang Jin; Ahn, Do Hee; Chung, Hong Suk [UST, Daejeon (Korea, Republic of); Kang, Hee Suk [KAERI, Daejeon (Korea, Republic of); Yun, Sei Hun [NFRI, Deajeon (Korea, Republic of)

2016-05-15

As a part of the International Thermonuclear Experimental Reactor (ITER) Project, We have conducted an experiment for storing hydrogen to depleted uranium and zirconium cobalt. The helium blanket effect has been observed in experiments using metal hydrides. The collapse of the hydrogen isotopes are accompanied by the decay heat and helium-3. Helium-3 dramatically reduces the hydrogen isotope storage capacity by surrounding the metal. This phenomenon is called a helium blanket effect. In addition the authors are working on the recovery and removal techniques of helium-3. In this paper, we discuss the equipment used to test the helium blanket effect and the results of a helium circulation experiment. The helium-3 produced surrounds the storage material surface and thus disturbs the reaction of the storage material and the hydrogen isotope. Even if the amount of helium-3 is small, the storage capacity of the SDS bed significantly drops. This phenomenon is the helium blanket effect. To resolve this phenomenon, a circulating loop was introduced. Using a circulating system, helium can be separated from the storage material. We made a helium loop that includes a ZrCo bed. Then using a metal bellows pump, we tested the helium circulation.

Absorption of a volatile organic compound by a jet loop reactor with circulation of a surfactant solution: Performance evaluation

International Nuclear Information System (INIS)

Park, Byungjoon; Hwang, Geelsu; Haam, Seungjoo; Lee, Changha; Ahn, Ik-Sung; Lee, Kyoungjoo

2008-01-01

Biofiltration shows high efficiency for the removal of industrial waste gases and reliable operational stability at low investment and operating cost, especially when the VOC concentration is low, such as 100 ppmv (μL L -1 ) or less. However, it has been reported that the abrupt change in VOC concentrations leads to the failure of the biofilter. Hence, the pretreatment of waste gases is necessary to ensure the stable operation of the biofilter. The objective of this study is to develop a jet loop reactor (JLR) with circulation of a surfactant solution to lower the concentration of VOCs, especially hydrophobic VOCs. Toluene and Tween 81 were used as a model industrial waste gas and a surfactant, respectively. Among several non-ionic surfactants tested, Tween 81 showed the most rapid dissolution of toluene. When a JLR is replaced with fresh Tween 81 solution (0.3% w/v) every hour, it successfully absorbed for 48 h over 90% of the toluene in an inlet gas containing toluene at 1000 ppmv (μL L -1 ) or less. Therefore, JLR with circulation of a surfactant solution is believed to ensure the stable operation of the biofilter even with the unexpected increase in the VOC concentrations

Experimental study on the effect of fill ratio on an R744 two-phase thermosyphon loop

International Nuclear Information System (INIS)

Tong, Zhen; Liu, Xiao-Hua; Li, Zhen; Jiang, Yi

2016-01-01

Highlights: • Performance of R744 two-phase thermosyphon loop is experimentally analyzed. • There are usually some fluids that circulate in the loop without changing phase. • Maximum heat transfer ability is achieved at the fill ratio around 100%. • Lowest driving temperature difference is achieved at the fill ratio around 62%. • Thermosyphon loop with a lower fill ratio is more likely to fluctuate at small heat loads. - Abstract: As a natural, environmentally friendly fluid with excellent thermodynamic and transport properties, carbon dioxide is an effective alternative refrigerant. This paper describes an experiment conducted on an R744-based two-phase thermosyphon loop (TPTL). With different fill ratios of 45~151%, the effect of fill ratio on the working performance of the R744 TPTL is investigated. To maintain the conservation of momentum, part of the fluid circulates in the loop without changing phase; this part of the fluid may be liquid, vapor, or both liquid and vapor depending on the fill ratio. This is how the R744 TPTL self-adjusts among different heat loads. The experimental results show that the working state of the R744 TPTL has a lot to do with the fill ratio. With a low fill ratio, the TPTL is more likely to fluctuate under small heat loads. When the fill ratio is around 100%, the TPTL reaches its maximum heat transfer ability, and when the fill ratio is around 62%, the lowest driving temperature difference is achieved. Considering that the fill ratio's effect on the driving temperature difference is not very significant and that pursuing maximum heat transfer ability is more meaningful, a fill ratio of around 100% is recommended.

Comparison of numerical results with experimental data for single-phase natural convection in an experimental sodium loop. [LMFBR

Energy Technology Data Exchange (ETDEWEB)

Ribando, R.J.

1979-01-01

A comparison is made between computed results and experimental data for a single-phase natural convection test in an experimental sodium loop. The test was conducted in the Thermal-Hydraulic Out-of-Reactor Safety (THORS) facility, an engineering-scale high temperature sodium loop at the Oak Ridge National Laboratory (ORNL) used for thermal-hydraulic testing of simulated Liquid Metal Fast Breeder Reactor (LMFBR) subassemblies at normal and off-normal operating conditions. Heat generation in the 19 pin assembly during the test was typical of decay heat levels. The test chosen for analysis in this paper was one of seven natural convection runs conducted in the facility using a variety of initial conditions and testing parameters. Specifically, in this test the bypass line was open to simulate a parallel heated assembly and the test was begun with a pump coastdown from a small initial forced flow. The computer program used to analyze the test, LONAC (LOw flow and NAtural Convection) is an ORNL-developed, fast-running, one-dimensional, single-phase, finite-difference model used for simulating forced and free convection transients in the THORS loop.

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)

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

SImulated Dodewaard ASsembly: Developments in loop-design

International Nuclear Information System (INIS)

Graaf, R. van de.

1992-03-01

A computer program was written to calculate void-fraction, flow rate, system circulation time and pressure drops for SIDAS under natural circulation conditions. In this program the thermohydraulic behaviour of the loop is simulated. Taking into account for the large uncertainty in two-phase frictional pressure drops, the chimney length is calculated, together with the length of the tubes which connect the downcomer at assembly height with the assembly inlet in a roundabout way. Tube diameter is chosen such that the frictional pressure losses are negligible. Using the results, it was decided to construct the chimney 'telescopically' (consisting of a fixed part and a movable part) in order to influence the driving force. Calculations of the enthalpy of the condensed vapour flow for various system conditions have shown that it is impractical to use this flow to lower the temperature of the total downcomer flow at the necessary subcooling temperature. It is therefore decided to use the condensor flow only for lowering the total downcomer flow enthalpy at saturation enthalpy and to establish the necessary subcooling separately by cooling of the flow in the connecting tubes. (orig.)

SImulated Dodewaard ASsembly: Developments in loop-design

Energy Technology Data Exchange (ETDEWEB)

Graaf, R van de

1992-03-01

A computer program was written to calculate void-fraction, flow rate, system circulation time and pressure drops for SIDAS under natural circulation conditions. In this program the thermohydraulic behaviour of the loop is simulated. Taking into account for the large uncertainty in two-phase frictional pressure drops, the chimney length is calculated, together with the length of the tubes which connect the downcomer at assembly height with the assembly inlet in a roundabout way. Tube diameter is chosen such that the frictional pressure losses are negligible. Using the results, it was decided to construct the chimney `telescopically` (consisting of a fixed part and a movable part) in order to influence the driving force. Calculations of the enthalpy of the condensed vapour flow for various system conditions have shown that it is impractical to use this flow to lower the temperature of the total downcomer flow at the necessary subcooling temperature. It is therefore decided to use the condensor flow only for lowering the total downcomer flow enthalpy at saturation enthalpy and to establish the necessary subcooling separately by cooling of the flow in the connecting tubes. (orig.).

The Brownian loop soup

OpenAIRE

Lawler, Gregory F.; Werner, Wendelin

2003-01-01

We define a natural conformally invariant measure on unrooted Brownian loops in the plane and study some of its properties. We relate this measure to a measure on loops rooted at a boundary point of a domain and show how this relation gives a way to ``chronologically add Brownian loops'' to simple curves in the plane.

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.

Antimony measurement in high pressure reactor water loop

International Nuclear Information System (INIS)

Svarc, V.; Dudjakova, K.; Martykan, M.; Sus, F.; Kysela, J.

2005-01-01

RVS 3 loop is highly contaminated due to antimony main circulating pump bearings. Antimony is one of the major contaminant in PWR units. Different technologies to remove Sb from the systems have been tried. (N.T.)

Advanced multi-evaporator loop thermosyphon

International Nuclear Information System (INIS)

Mameli, M.; Mangini, D.; Vanoli, G.F.T.; Araneo, L.; Filippeschi, S.; Marengo, M.

2016-01-01

A novel prototype of multi-evaporator closed loop thermosyphon is designed and tested at different heaters position, inclinations and heat input levels, in order to prove that a peculiar arrangement of multiple heaters may be used in order to enhance the flow motion and consequently the thermal performance. The device consists in an aluminum tube (Inner/Outer tube diameter 3.0 mm/5.0 mm), bent into a planar serpentine with five U-turns and partially filled with FC-72, 50% vol. The evaporator zone is equipped with five heated patches (one for each U-turn) in series with respect to the flow path. In the first arrangement, heaters are wrapped on each bend symmetrically, while in the second layout heaters are located on the branch just above the U-turn, non-symmetrical with respect to the gravity direction, in order to promote the fluid circulation in a preferential direction. The condenser zone is cooled by forced air and equipped with a 50 mm transparent section for the flow pattern visualization. The non-symmetrical heater arrangement effectively promotes a stable fluid circulation and a reliable operation for a wider range of heat input levels and orientations with respect to the symmetrical case. In vertical position, the heat flux dissipation exceeds the pool boiling heat transfer limit for FC-72 by 75% and the tube wall temperatures in the evaporator zone are kept lower than 80 °C. Furthermore, the heat flux capability is up to five times larger with respect to the other existing wickless heat pipe technologies demonstrating the attractiveness of the new concept for electronic cooling thermal management. - Highlights: • A novel passive heat transfer device named Multi-Evaporator Loop Thermosyphon is tested. • The loop is investigated at different heating patterns, inclinations and heat power levels. • The non-symmetrical heating configuration promotes the fluid circulation within the loop. • The performance in terms of maximum heat flux exceeds the

Heat transfer in a one-dimensional mixed convection loop

International Nuclear Information System (INIS)

Kim, Min Joon; Lee, Yong Bum; Kim, Yong Kyun; Kim, Jong Man; Nam, Ho Yun

1999-01-01

Effects of non-uniform heating in the core and additional forced circulation during decay heat removal operation are studied with a simplified mixed convection loop. The heat transfer coefficient is calculated analytically and measured experimentally. The analytic solution obtained from a one-dimensional heat equation is found to agree well with the experimental results. The effects of the non-uniform heating and the forced circulation are discussed

Hardware-in-the-loop-simulation using the example of a compressed-natural-gas hybrid; Hardware-in-the-Loop-Simulation am Beispiel eines Erdgas-Hybridfahrzeugs

Energy Technology Data Exchange (ETDEWEB)

Boehm, M.; Mauk, T.; Reuss, H.C. [Univ. Stuttgart (Germany)

2008-07-01

This paper deals with the development of hybrid specific software for a compressed natural gas hybrid. After the presentation of the project the software development process is being illustrated. A two phase hardware-in-the-loop simulation is an important part of it. The vehicle model and the modelling process will be described. (orig.)

WWER type reactor primary loop imitation on large test loop facility in MARIA reactor

International Nuclear Information System (INIS)

Moldysh, A.; Strupchevski, A.; Kmetek, Eh.; Spasskov, V.P.; Shumskij, A.M.

1982-01-01

At present in Poland in cooperation with USSR a nuclear water loop test facility (WL) in 'MARIA' reactor in Sverke is under construction. The program objective is to investigate processes occuring in WWER reactor under emergency conditions, first of all after the break of the mainprimary loop circulation pipe-line. WL with the power of about 600 kW consists of three major parts: 1) an active loop, imitating the undamaged loops of the WWER reactor; 2) a passive loop assignedfor modelling the broken loop of the WWER reactor; 3) the emergency core cooling system imitating the corresponding full-scale system. The fuel rod bundle consists of 18 1 m long rods. They were fabricated according to the standard WWER fuel technology. In the report some general principles of WWERbehaviour imitation under emergency conditions are given. They are based on the operation experience obtained from 'SEMISCALE' and 'LOFT' test facilities in the USA. A description of separate modelling factors and criteria effects on the development of 'LOCA'-type accident is presented (the break cross-section to the primary loop volume ratio, the pressure differential between inlet and outlet reactor chambers, the pressure drop rate in the loop, the coolant flow rate throuh the core etc.). As an example a comparison of calculated flow rate variations for the WWER-1000 reactor and the model during the loss-of-coolant accident with the main pipe-line break at the core inlet is given. Calculations have been carried out with the use of TECH'-M code [ru

Theory of energy level and its application in water-loop heat pump system

International Nuclear Information System (INIS)

Yu, Qi Dong

2017-01-01

Highlights: • Novel theory of saving energy and its application in water loop heat pump. • Reverse energy caused by units to water loop and its solution. • New method for determining the energy-saving range of water loop heat pump. • Capacity model of auxiliary heat source and its size for all building types. • Advice for reducing total energy consumption of water loop heat pump. - Abstract: It is a difficult problem to how to determine the reverse energy caused by units to water loop when a water-loop heat pump (WLHP) is in cooling and heating simultaneous mode, which not only has a great impact on energy-saving rate but also decides the use of auxiliary heat source in winter. This paper presents a theory of energy level to improve the research on WLHP system by using the relationship among building, circulating water and units. In this theory, the circulating water replaces building load as a new method to convert the reverse energy into energy change of circulating water and the equation of energy level also is built to determine the energy-saving range of WLHP system and report the capacity model of auxiliary heat source for all building types. An office building with different auxiliary powers is tested to analyze system operation characteristic and the effect of auxiliary heat source on unit and system and the results validate previous conclusions and suggest that an energy balance should be considered between units and auxiliary power to improve overall operation.

Predictive models of circulating fluidized bed combustors: SO{sub 2} sorption in the CFB loop. Fourteenth technical progress report

Energy Technology Data Exchange (ETDEWEB)

Gidaspow, D.; Therdthianwong, A. [Illinois Inst. of Tech., Chicago, IL (United States). Dept. of Chemical Engineering

1993-02-01

The overall objective of this investigation is to develop experimentally verified models for circulating fluidized bed (CFB) combustors. Sorption of S0{sub 2} with calcined limestone was studied in a PYROFLOW type CFB loop at conditions approximating those found in a CFB combustor. Initially the CFB loop contained 150 micron CaO particles of a density of 3.3 g/cm{sup 3} and air at 1143{degrees}K and 3.25 atm. Atzero time, air containing 600 ppm SO{sub 2}, was introduced into the riser bottom at 1143{degrees}K. The effect of gas velocity, sorbent inventory and inlet pressure on the sorption of SO{sub 2}, were studied isothermally by running our hydrodynamic code with the S0{sub 2} sorption conservation of species equation. At a velocity of 5m/sec., reported to be a typical velocity by PYROPOWER, there is reasonably good S0{sub 2} removal. At 10 m/sec the S0{sub 2} removal is poor. The best SO{sub 2}, removal is for a velocity of 5 m/s and a high bed inventory, initial bed height, H = 9m. Most of the S0{sub 2} is removed in the first two meters of the reactor. However, the S0{sub 2} removal is not complete at the bed outlet. This is due to mixing. At the left wall of the reactor (wall opposite the solids inlet) the S0{sub 2} removal was poor due to gas bypassing caused by the asymmetrical solids inlet. Simulation of the PYROPOWER loop with a symmetrical inlet gave us an order of magnitude improvement over the conventional PYROPOWER system. These results demonstrate the practical utility of the predictive model that we have developed over the last three years.