A flow boiling critical heat flux correlation for water and Freon-12 at low mass fluxes

An empirical correlation has been developed for calculating critical heat flux (CHF) at low mass fluxes for vertical upflow in uniformly heated tubes. The correlation is based upon dimensionless groups. It compares favourably with experimental CHF data for both Freon-12 and pressurized water. When solved iteratively in conjunction with the heat balance equation, an overall mean ratio of predicted to experimental CHF of 0.986 was obtained with a root means square (r.m.s.) error of 7.0%, for the 233 low flow rate data sets examined.The boundary between the high flow rate correlation developed in earlier work and the proposed low flow rate correlation can be specified by a dimensionless factor δ₁. For values of δ₁ greater than 0.07, the low flow correlation is valid whereas for values less than 0.07 the high flow correlation applies.Development of this correlation and a means of defining its range of validity enables the prediction of CHF levels to be made over an increased range of coolant flow conditions. This is important in the analysis of postulated loss-of-coolant accidents in water-cooled nuclear reactors.     

Experimental thermohydraulic stability map of a Freon-12 boiling water reactor facility with high exit friction

The natural-circulation characteristics and the density-wave stability characteristics of the natural-circulation Freon-12 facility DESIRE for a specific configuration have been determined by systematically performing experiments in the whole operating range. A large amount of data has been gathered to be used for future benchmarking of computer codes for the calculation of boiling water reactor (BWR) stability. Contrary to expectations, it was found that, for low subcooling values, the stability of the facility improves as the power is increased when keeping the subcooling number constant. This result can serve as a challenging benchmark for models and codes, since it is not in line with the experience from the majority of analytical and numerical models.     

On the occurrence of burnout downstream of a flow obstacle in boiling two-phase upward flow within a vertical annular channel

If a flow obstacle, such as a spacer is placed in a boiling two-phase flow within a channel, the temperature on the surface of the heating tube is severely affected by the existence of the spacer. Under certain conditions, a spacer has a cooling effect, and under other conditions, the spacer causes dryout of the cooling water film on the heating surface. The burnout mechanism, which always occurs upstream of a spacer, however, remains unclear.In a previous paper [Fukano, T., Mori, S., Akamatsu, S., Baba, A., 2002. Relation between temperature fluctuation of a heating surface and generation of drypatch caused by a cylindrical spacer in a vertical boiling two-phase upward flow in a narrow annular channel. Nucl. Eng. Des. 217, 81–90], we reported that the disturbance wave has a significant effect on dryout and burnout occurrence and that a spacer greatly affects the behavior of the liquid film downstream of the spacer.In the present study, we examined in detail the influences of a spacer on the heat transfer and film thickness characteristics downstream of the spacer by considering the result in steam–water and air–water systems. The main results are summarized as follows:

(1) The spacer averages the liquid film in the disturbance wave flow. As a result, dryout tends not to occur downstream of the spacer. This means that large temperature increases do not occur there. However, traces of disturbance waves remain, even if the disturbance waves are averaged by the spacer.

(2) There is a high probability that the location at which burnout occurs is upstream of the downstream spacer, irrespective of the spacer spacing.

(3) The newly proposed burnout occurrence model can explain the phenomena that burnout does occur upstream of the downstream spacer, even if the liquid film thickness t_F m is approximately the same before and behind the spacer.

A contribution to the hydrodynamics of boiling crisis in a forced flow of water

A physical approach is presented for defining the mechanisms responsible for the ‘departure from nucleate boiling’, DNB, in cooling water under forced convection.Based on experimental observations, a hydrodynamic model is proposed. It considers the flow of vapour bubbles away from the heated surface and the counter-current of liquid coolant streaming towards it. The general DNB correlation derived from this model is compared with numerous experimental results reported in literature and with well-known empirical DNB equations. The application to critical heat flux in uniformly heated round tubes and annuli; as well as to non-uniform heat flux profiles, is examined.Due to the scatter inherent at the onset of the boiling crisis, the proposed correlation has been set to fit the lower conservative limit of heat flux density, where DNB is observed rather than to a most probable mean value.     

Influence of a flow obstacle on the occurrence of burnout in boiling two-phase upward flow within a vertical annular channel

When a flow obstruction such as a cylindrical spacer is set in a boiling two-phase flow within an annular channel, the inner tube of which is used as a heater, the temperature on the surface of the heating tube is severely affected by its existence. In some cases, the cylindrical spacer has a cooling effect, and in the other cases it causes the dryout of the cooling water film on the heating surface resulting in the burnout of the heating tube.In the present paper, we have focused our attention on the influence of a flow obstacle on the occurrence of burnout of the heating tube in boiling two-phase flow.The results are summarized as follows:

(1)When the heat flux approaches the burnout condition, the wall temperature on the heating tube fluctuates with a large amplitude. And once the wall temperature exceeds the Leidenfrost temperature, the burnout occurs without exception.

(2)The trigger of dryout of the water film which causes the burnout is not the nucleate boiling but the evaporation of the base film between disturbance waves.

(3)The burnout never occurs at the downstream side of the spacer. This is because the dryout area downstream of the spacer is rewetted easily by the disturbance waves.

     

Subcooled and low quality flow film boiling of water at atmospheric pressure

This paper describes an experimental study of subcooled and low quality film boiling for water in a vertical tube covering a mass flux range of 50–500 kg m⁻² s⁻¹ and an inlet subcooling range of 5–70°C. Discussion of various observed parametric trends on the film boiling section of the boiling curve is presented. The data are compared with the correlations of Ellion and Hsu.     

The correlation of the experimental data on critical heat loads in forced flow of water heated below the boiling temperature

At low pressures, theory predicts a simplification of the similarity-parameter equation proposed by the writer [2] on the basis of a correlation of the experimental data on critical thermal loads in forced flow of water not heated to boiling, in the pressure range 100–210 atmos.It is shown that it is possible to apply the previously proposed similarity-parameter equation over a wider range of pressures, namely 35–210 atmos.Furthermore an analysis of the published experimental data onq _cr at low pressures confirms the theoretical conclusion about a degeneration of the functional connections between the parameter to be determined and the two determining parameters at water pressures close to atmospheric pressure, owing to which the similarity-parameter equation for this case takes a considerably simpler form. A computational formula obtained on the basis of this equation is recommended for the pressure range 1–15 atmos.O. L. Peskov, N. D. Sergeev, Z. F. Deryugin, and N. A. Gushchina took part in the taking of the measurements and the analysis of the data.     

Transition flow boiling heat transfer to water in a vertical annulus

Transition boiling heat transfer coefficients for water at 25–30 psia flowing upward at low velocities have been obtained Hot mercury, flowing on the inside of a tube with a 0.54 in. o.d. served as the heat source. Water flowed in the annular space between the heat source and an outer glass tube having a 1 in. dia. Thermocouples placed at several elevations within the mercury stream allowed the rate of heat transfer to be determined. The heat transfer coefficients appear consistent with other transition boiling data providing an appropriate allowance is made for the reduction in critical heat flux at high void fractions.     

Predicted impact of core flow rate on the corrosion mitigation effectiveness of hydrogen water chemistry for Kuosheng boiling water reactor

It is currently a common practice that a boiling water reactor (BWR) adopts hydrogen water chemistry (HWC) for mitigating corrosion in structural components in its primary coolant circuit. When the core flow rate (CFR) in a BWR is changed, the coolant residence time in the primary coolant circuit would be different. The concentrations of major redox species (i.e. hydrogen, oxygen, and hydrogen peroxide) in the coolant may accordingly vary due to different durations of radiolysis in the core and other near-core regions. A theoretical model by the name of DEMACE was used in the current study to investigate the impact of various CFRs (from 100% to 80.0%) on the effectiveness of HWC in a domestic BWR. Our analyses indicated that the HWC effectiveness at some locations could be downgraded due to a decrease in CFR. However, a lower CFR was instead beneficial to the corrosion mitigation efficiency of HWC at other locations. The impact of CFR on the HWC effectiveness could vary from location to location in a BWR and eventually from plant to plant.     

Investigation of the two-phase flow in a boiling water reactor using neutron-noise technique

In this paper, the fluctuations of the neutron flux (“neutron noise”) of the Mühleberg BWR are investigated. Above 2 Hz, the noise measured by the in-core neutron detectors is driven exclusively by local fluctuations of the void fraction. Characteristic changes of the neutron-noise signature along the axis can be attributed to changes of flow pattern. By measuring the phase lag between pairs of axially placed neutron detectors, the transit time of the steam between the detectors can be evaluated. The measured transit times are applied to the study of two-phase flow in the core. The neutron-noise method has the advantage of providing in-core information under operational conditions.     

Modeling of low-pressure subcooled boiling flow of water via the homogeneous MUSIG approach

Applying a three-dimensional two-fluid model coupled with homogeneous multiple size group (MUSIG) approach, numerical simulations of upward subcooled boiling flow of water at low pressure were performed on the computational fluid dynamics (CFD) code CFX-10 with user defined FORTRAN program. A modified bubble departure diameter correlation based on the Unal's semi-mechanistic model and the empirical correlation of Tolubinski and Kostanchuk was developed. The water boiling flow experiments at low pressure in a vertical concentric annulus from reference were used to validate the models. Moreover, the influences of the non-drag force on the radial void fraction distribution were investigated, including lift force, turbulent dispersion force and wall lubrication force. Good quantitative agreement with the experimental data is obtained, including the local distribution of bubble diameter, void fraction, and axial liquid velocity. The results indicate that the local bubble diameter first increases and then decreases due to the effect of bubble breakup and coalescence, and has the maximum bubble diameter along the radial direction. Especially, the peak void fraction phenomenon in the vicinity of the heated wall is predicted at low pressure, which is developed from the wall repulsive force between vapor bubbles and heated wall. Nevertheless, there is a high discrepancy for the prediction of the local axial vapor velocity.     

Spectral shift rod for the boiling water reactor

A Boiling Water Reactor core concept has been proposed using a new fuel component called spectral shift rod (SSR). The SSR is a new type of water rod in which a water level is formed during core operation. The water level can be controlled by the core recirculation flow rate. By using SSRs, the reactor can be operated with all control rods withdrawn through the operation cycle as well as that a much larger natural uranium saving is possible due to spectral shift operation than in current BWRs. The steady state and transient characteristics of the SSRs have been examined by experiments and analyses to certify the feasibility. In a reference design, a four times larger spectral shift width as for the current BWR has been obtained.     

Methods for leak detection for KWU pressurized and boiling water reactors

Leakage monitoring is an essential criterion to rule out the possibility of double ended pipe rupture in the primary coolant system. Subcritical cracks can be detecred with a considerable margin before they extend to critical crack lengths resulting in spontaneous failure. In those KWU PWRs which went into operation recently, a Leakage Monitoring System was installed that is based on thermodynamic analysis. It utilizes the following measured parameters: dew point temperature, accumulated condensate inside aircoolers, air temperature, sump water level, gully monitoring. In KWU's BWRs, although the measurement concept has to be slightly changed because of a different design of buildings and components, the same instrumentation will be used. Besides this installed monitoring system, different approaches like acoustic leak detection systems or the application of moisture sensitive instrumentation have been considered. Both systems have been successfully tested.     

A generalized flow correlation for two-phase natural circulation loops

A generalized correlation has been proposed to estimate the steady-state flow in two-phase natural circulation loops. The steady-state governing equations for homogeneous equilibrium model, viz. continuity, momentum and energy equations have been solved to obtain the dimensionless flow rate as a function of a modified Grashof number and a geometric number. To establish the validity of this correlation, two-phase natural circulation flow rate data from five different loops have been tested with the proposed correlation and found to be in good agreement.     

The critical heat flux for boiling water in tubes

An analysis of the experimental results obtained by various authors on critical heat flux is carried out by using nondimensional criteria. Recommendations are given for the numerical methods of determining values of the critical heat flux in the case of a steam-water mixture, underheated to saturation in tubes and in ring-shaped and plane slotted channels.     

A mechanistic critical heat flux model for wide range of subcooled and low quality flow boiling

A mechanistic model to predict a critical heat flux (CHF) over a wide operating range in the subcooled and low quality flow boiling has been proposed based on a concept of the bubble coalescence in the wall bubbly layer. The conservation equations of mass, energy and momentum, together with appropriate constitutive relations, are solved analytically to derive the CHF formula. The model is characterized by an introduction of the drag force due to wall-attached bubbles roughness in the momentum balance, which determines the limiting transverse interchange of mass flux crossing the interface of the wall bubbly layer and core. Comparison between the predictions by the proposed model and the experimental CHF data shows good agreement over a wide range of parameters for both light water and fusion reactors operating conditions. The model correctly accounts for the effects of flow variables such as pressure, mass flux and inlet subcooling as well as geometry parameters.     

A semiempirical prediction of the decay ratio for the boiling water reactors start-up process

During the start-up of a commercial boiling water reactor (BWR), the power and the coolant flow are continuously monitored. In order to prevent power instability events, the decay ratio (DR) could also be monitored. The process can be made safer if the operator could anticipate the DR too. DR depends on the power, the flow and many other quantities such as axial and radial neutron flux distribution, feed water temperature, void fraction, etc. A simple relationship for DR is derived. Three independent variables seem to be enough: the power, the flow and a single parameter standing for all other quantities which affect the DR. The relationship is validated with data from commercial BWR start-ups. A practical procedure for the start-up of a BWR is designed; it could help preventing instability events.