Natural Circulation Experiments Research Articles

Pre- and post-test calculations to natural circulation experiments at the integral test facility ISB-VVER using the thermalhydraulic code ATHLET

In 1995 at the integral test facility ISB-VVER in Elektrogorsk near Moscow natural circulation experiments were performed, which were scientifically accompanied by the Forschungszentrum Rossendorf. These experiments were the first of this kind at a test facility, which models VVER-1000 thermalhydraulics. Using the code ATHLET which is being developed by ‘Gesellschaft für Anlagen und Reaktorsicherheit’, pre- and post-test calculations were done to determine the thermalhydraulic events to be expected and to define and tune the boundary conditions of the test. The conditions found for natural circulation instabilities and cold leg loop seal clearing could be confirmed by the tests. Besides the thermalhydraulic standard measuring system, the facility was equipped with needle shaped conductivity probes for measuring the local void fractions.

Evaluation of Inherent Distortions in the IIST Facility Using the RELAP5/MOD3 Code

The Institute of Nuclear Energy Research (INER) integral system test (IIST) facility is a reduced-height, reduced-pressure test facility constructed at INER that is used to simulate the thermal hydraulics of the Maanshan nuclear power plant (NPP). A small-scaled facility is not capable of simulating all the physical phenomena of an NPP because the behavior of an NPP during accidents is very complicated. Proper scaling then plays an important role in the design of a test facility to ensure the usefulness and applicability of experimental data obtained from a small-scaled facility. However, distortions caused by necessary compromises in the design and construction of a small-scaled test facility exist. The analysis here evaluates whether the inherent distortions in the IIST facility will distort the thermal-hydraulic behaviors of a natural-circulation experiment and influence the usefulness and applicability of the experimental data. Based on the current calculations, the IIST experimental results are found to be partially distorted. Appropriate consideration of and correction for these distortion effects are needed before the results of the IIST natural-circulation experiments can be used to reliably investigate the Maanshan NPP behavior expected by way of an appropriate scale-up procedure.

A Comparison of the RELAP5/MOD3 Code with the IIST Natural Circulation Experiments

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.

Investigation of Scale-Up Capability of RELAP5/MOD3 and Scaling Distortion in the IIST Facility Via Natural Circulation Experiments

During the past years, a number of reduced-scale test facilities have been constructed to investigate the physical phenomena of transients or accidents occurring in nuclear power plants. Since the behavior of a nuclear power plant is complicated, it is quite impossible for a small-scaled facility to simulate all the physical phenomena during the transient process. But, by way of proper scaling, most of the important aspects of transient behavior can be simulated. Calculations using RELAP5/MOD3 investigate whether most of the key thermal-hydraulic phenomena observed in the Institute of Nuclear Energy Research Integral System Test (IIST) facility can be expected in a prototype plant. When compared with experimental data, the calculated results of two different scale models show reasonable agreement with the natural circulation transients. The scale-up capability of RELAP5/MOD3 is demonstrated by simulating the single-phase and two-phase natural circulation transients. Also, the scaling distortions in the heat transfer areas of the IIST facility do not strongly distort the thermal-hydraulic behavior of experimental data.

Influence of the position and number of decay heat exchangers on the thermal hydraulics of a slab test facility: A comparison of analytical and experimental data

To assess the capability of passive decay heat removal systems of an advanced pool-type liquid-metal cooled reactor, natural circulation experiments have been performed to investigate the in-vessel cooling modes caused by the position and number of decay heat exchangers in operation. The rather simple slab test facility AQUARIUS is equipped with an electrically heated core and decay heat exchangers. Four different arrangements of heat exchangers are under consideration to study the temperature distribution and the flow behavior in the apparatus during both symmetrical and asymmetrical heat removal. The experiments have been carried out in water under laminar flow conditions. Temperatures have been measured under quasi-steady-state conditions. The observed flow paths have been documented photographically. In case of asymmetrical heat removal, especially when a single heat exchanger is operated in one of both upper plena, the temperature distribution and the flow behavior are different from symmetrical cooling modes. A comparison of analytically predicted temperatures using the COMMIX-2(V) computer program with experimental data shows reasonably matches the findings. The results of the numerically determined velocity fields are in good agreement with the visual observations.

Numerical simulation of natural circulation experiments conducted at the IIST facility

Abstract Natural circulation plays an important role in long-term cooling of pressurized water reactors (PWRs) under small break loss-of-coolant accidents. Recently, natural circulation experiments have been conducted at the Institute of Nuclear Energy Research integral system test (IIST) facility, which is used to simulate the Westinghouse three-loop Maanshan PWR. A numerical simulation is presented to investigate the natural circulation phenomena of the IIST facility with the RELAP5/MOD3 code. The calculated results are in good agreement with the experimental data of the single-phase natural circulation both quantitatively and qualitatively. The influences of power level and system pressure on natural circulation can also be predicted by the current model. Based on the two-phase natural circulation data, the calculated flow rate history is similar to that obtained from the experiment.

Natural circulation experiments with a VVER reactor geometry

Abstract Experiments were conducted to study natural circulation behaviour in the PACTEL facility, a medium-scale integral test loop patterned after Soviet-designed VVER pressurized water reactors. Using stepwise inventory reduction and small-break experiments, primary loop flow behaviour was studied over a range of coolant inventories. The tests revealed a trend toward decreasing primary side mass flow rate with inventory. 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 and coolant flow into the hot legs changed from single to two-phase flow. The cause of this flow interruption was attributed to the hot leg loop seals, which are a unique feature of the VVER geometry. Finally, an experiment was conducted to demonstrate how loop seal refilling behaviour at low coolant inventories depends upon the steam flow rate through an individual hot leg. It was shown that loop seal refilling results when low steam velocities permit countercurrent flow in the upflow side of the loop seal.

An application of a higher order finite difference method to a natural convection experiment in the hot plenum of an LMFBR

A two dimensional thermal-hydraulic analysis of a natural circulation experiment has been performed to evaluate the effectiveness of a higher order finite difference method for solving the Navier-Stokes and the energy equations. In the method, the convection terms appearing in each equation are solved by the Method of Characteristics using the third order Lagrange type polynomial as the interpolation function, and an iterative procedure is applied to solve the time derivative terms of each equation stably with second order accuracy. The analytical results have been compared with an experiment in which the temperature and the velocity distributions in the plenum region were measured with their fluctuations, and it was shown that the higher order finite difference method could simulate natural convection phenomena involving fluctuations well.

STEAM CONDENSATION AND LIQUID HOLDUP IN STEAM GENERATOR U-TUBES DURING OSCILLATORY NATURAL CIRCULATION

Flow oscillations during two-phase natural-circulation experiments for pressurized water reactors (PWRs) with inverted U-tube steam generators occur at high pressure and at a primary inventory range between two-phase circulation and reflux heat removal. These oscillations, originating in the steam generator U-tubes, consist of alternating phases of (1) interrupted (or stalled) flow and (2) rapid two-phase circulation. Using integral system data from LOBI-MOD2 experiments, it is demonstrated that the predominant mechanism that governs the duration of the interrupted flow phase is steam condensation at the top of the U-tubes. The frequency is not governed by mechanisms related to limitation of countercur-rent flow (CCFL). The article also discusses the possible mechanisms of liquid holdup in the vertical U-tube region, necessary to have a closed condensation space. This may be CCFL but could also be simply related to the water inventory in the circuit. Furthermore, the influence of pressure is discussed, sh...

Investigation into the occurrence of backflow effects in a heated rectangular channel during natural circulation

To address the capability of passive decay heat removal systems of an advanced liquid-metal-cooled nuclear reactor, natural circulation experiments were performed to investigate the bouyancy-driven backflow phenomenon in the mixed convection flow of a cooling channel. In the AQUARIUS half-slab test facility, the core is represented by a vertical channel formed by two parallel electrically heated walls. At the top end, the channel is connected to an upper plenum where the fluid is cooled by an immersion cooler. Fluid velocities and temperatures were measured under quasi-steady-state laminar conditions. The experiments indicate that for low heat flux densities and large channel widths, cold water from the upper plenum flows down and penetrates the heated channel where the flow direction is upward. A comparison of analytical predictions using the COMMIX-2(V) three-dimensional computer code against experimental AQUARIUS data shows reasonable overall agreement. The numerical results reflect that backflow effects were not predicted for the investigated problem.

Scaling of the accuracy of the Relap5/mod2 code

This paper presents an attempt to derive uncertainty values in the prediction of BWR and PWR transient scenarios. The small break LOCA counterpart tests performed in the BWR simulators Piper-one, Fist and Rosa-III, and natural circulation experiments performed in the PWR simulators Lobi, Spes and Lstf, constitute the basis of the activity. The application of Relap5/mod2 to the analyses of the above experiments, the evaluation of the comparison between predicted results and measured data, and the calculation of the BWR and PWR plants scenarios, were fundamental in achieving the proposed goal. The main result of the activity is constituted by the development of a methodology suitable for deriving uncertainty values of code calculations. The values reported for the uncertainty should be considered as the result of a demonstrative pilot application of the methodology.

Characterization of instabilities during two-phase natural circulation in typical PWR conditions

Strong oscillations in fluid velocities and densities were measured in the LOBI test facility during natural circulation experiments. A sort of “siphon condensation” occurs in the U-tubes of steam generators when primary-side mass inventory reaches roughly 64% of the initial value. The present paper deals with the characterization of the phenomenon considering flooding and condensation dynamics of U-tubes; RELAP5/MOD2 calculations made it possible to select system parameters affecting the oscillation characteristics. An attempt was made to evaluate the possibility of instability occurring in real plant situations.

Nonuniform Steam Generator U-Tube Flow Distribution during Natural Circulation Tests in ROSA-IV Large Scale Test Facility

Natural circulation experiments were conducted in a large-scale ( scale in volume) full-height simulator of a Westinghouse-type pressurized water reactor. This facility has two steam generators eac...

Nonuniform Steam Generator U-Tube Flow Distribution During Natural Circulation Tests in ROSA-IV Large Scale Test Facility

Natural circulation experiments were conducted in a large-scale ( scale in volume) full-height simulator of a Westinghouse-type pressurized water reactor. This facility has two steam generators each containing 141 full-size U-tubes of 9 different heights. Transition of the natural circulation mode was observed in the experiments as the primary side mass inventory was decreased. Three major circulation modes were observed: single-phase liquid natural circulation, two-phase natural circulation, and reflux condensation. For all these circulation modes, and during the transitions between the modes, the mass flow distribution among the steam generator U-tubes was significantly nonuniform. The longer U-tubes indicated reversed flow at higher primary side mass inventories and also tended to empty earlier than the shorter U-tubes when the primary side mass inventory was decreased.

The results of 5% small-break LOCA tests and natural circulation tests at the ROSA-IV LSTF

Five 5% small-break loss-of-coolant accident (SBLOCA) experiments and two natural circulation experiments were conducted at the ROSA-IV Large Scale Test Facility (LSTF). The liquid holdup in the upflow side of steam generator (SG) U-tubes temporarily depressed the core collapsed liquid level below the bottom of core during the loop seal clearing in the cold-leg break SBLOCA tests. This phenomena was affected by the core power and core bypass but was affected little by the actuation of the high pressure injection system. Overall thermal-hydraulic phenomena in a loop seal line break test was similar to that of cold-leg break tests, however, the liquid holdup phenomena played a little role. In a hot-leg break test a temporary but rapid depression of the core liquid level was observed immediately after the initiation of accumulator injection which caused condensation and depressurization in the cold leg. The change of natural circulation flow rate with the decrease of primary system mass inventory was qualitatively the same as observed in Semiscale, LOBI and PKL. The SG effective overall heat transfer coefficient below the secondary-side collapsed liquid level was weakly dependent on the secondary side liquid level and the core power. The measured minimum heat transfer coefficient was 1.7 kW/m 2K for the full secondary side mass inventory.

BWR Natural Circulation Experiments at ROSA-III under Low-Power and Reduced Inventory Conditions

BWR Natural Circulation Experiments at ROSA-III under Low-Power and Reduced Inventory Conditions

BWR natural circulation experiments at ROSA-III under low-power and reduced inventory conditions.

The natural circulation core cooling is important during small break loss-of-coolant accidents and transients involving loss of forced-circulation core cooling. This paper presents results of natural circulation experiments conducted in the ROSA-III (Rig of Safety Assessment No. 3) facility under reduced mass inventory conditions. Three distinct modes of BWR natural circulation were identified in the experiments: the primary circulation mode, the internal circulation mode and the open-loop (core uncovery) mode. One-dimensional quasi-steady analytical model was developed for the internal natural circulation mode to determine the core uncovery criterion. The model predicted well the two-phase mixture level inside the core shroud for quasi-steady state natural circulation at pressures of 7.35 and 2.06 MPa, and scaled core powers below 20%, and was used to predict BWR natural circulation behaviors.

Engineering methods of analyzing natural-circulation conditions in systems of BN type

The authors report calculations on modes of transition to natural circulation (NC) in systems of BN-600 type on the basis of an engineering method applied to the entire plant. A basic purpose of devising the method was to obtain reliable results on the basis of comparison with experiment. The BN-600 was used in a series of NC experiments by special methods. Initially, experiments were made with only slight core heating. The final stage was direct transition to NC after the emergency protection had operated from a power level of half the nominal value. The model describes a plant nominally having two heat-transfer branches; one has a set of loops in which an accident occurs and the other consists of the other loops, where there are known perturbations in the secondary circuit. The perturbations from the third circuit (steam-water and air one) are represented as piece-wise-linear functions.

Survey of Natural-Circulation Cooling in U.S. Pressurized Water Reactors

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.

Verification study of the FORE-2M nuclear/thermal-hydraulic analysis computer code

The verification of the LMFBR core transient performance code, FORE-2M, was performed in two steps. Different components of the computation (individual models) were verified by comparing with analytical solutions and with results obtained from other conventionally accepted computer codes (e.g., TRUMP, LIFE, etc.). For verification of the integral computation method of the code, experimental data in TREAT, SEFOR and natural circulation experiments in EBR-II were compared with the code calculations. Good agreement was obtained for both of these steps. Confirmation of the code verification for undercooling transients is provided by comparisons with the recent FFTF natural circulation experiments.