Mass velocity and cold-wall effects on critical heat flux in an advanced light water reactor

The characteristics of Critical Heat Flux (CHF) were investigated for a square array of rod bundles which could possibly be loaded into an integral-type advanced light water reactor. The parametric effects of the mass velocity and the unheated rod were examined by conducting CHF experiments with 5 × 5 test bundles in a Freon-loop. The influence of a cold wall on the CHF was interpreted by introducing a simple phenomenological model which accounts for the influence of a thermal mixing inside the boiling channel. A local parameter CHF correlation applicable to an integral-type reactor was developed from the CHF data base for square-arrayed rod bundles. The local thermal–hydraulic conditions calculated by the subchannel analysis code MATRA were used for the optimization of the correlation coefficients. Correction factors for the low mass velocity, spacer grids, and the non-uniform axial power shapes have been devised which reflected the results of the data assessment and the experimental observations. As a result of the thermal margin evaluation at steady state conditions, it was revealed that the integral-type reactor core has a greater DNBR margin than a typical 1000 MW_e PWR core.     

Critical heat flux prediction in rod bundles under upward low mass flux densities

The analysis of experimental data and results of calculations for heat transfer crisis in heated channels under low upward coolant mass flux densities is presented. This analysis allows the determination of the basic features of the boiling crisis phenomenon. It is shown that the methods currently used for critical heat flux (CHF) prediction have insufficient accuracy in the given range of parameters. A new relationship for the CHF calculation is presented. It should be used for the water–water energy reactor (WWER) and uran–graphite channel reactor—Chernobyl-type (RBMK) rod bundles, and is verified by the test data. The comparison of results obtained by a new CHF correlation and the relationship used in RELAP5/MOD3.1 Code is presented. It is shown that the latter overpredicts the CHF values at atmospheric pressure and for x_cr>0.4 and does not provide conservative estimations for the RBMK fuel bundles.     

A flow boiling burnout correlation for water and Freon-12

An empirical correlation has been developed for calculating critical heat flux (CHF) for vertical upflow in uniformly heated tubes. The correlation is based on parameter groups derived from a dimensional analysis and has been compared with experimental CHF data for Freon-12 and for water. Except for coolant conditions in which (i) mass fluxes are less than 300 kg s⁻¹ m⁻², (ii) dryout qualities are below 10%, or (iii) water pressures are outside the range 3.5 to 12 MPa, the correlation agrees very favourably with the experimental data. The overall mean ratio of calculated to experimental CHF values for 1760 sets of Freon-12 data is 0.992 and the r.m.s. error 3.3%; the corresponding values for 2063 sets of water data are 0.982 and 5.8%. This provides a basis for predicting CHF levels over a wide range of coolant conditions, as required in the analysis of hypothetical loss-of-coolant accidents in water-cooled nuclear reactors.     

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.     

Development of a bundle correction method and its application to predicting CHF in rod bundles

A bundle correction method, based on the conservation laws of mass, energy, and momentum in an open subchannel, is proposed for the prediction of the critical heat flux (CHF) in rod bundles from round tube CHF correlations without detailed subchannel analysis. It takes into account the effects of the enthalpy and mass velocity distributions at subchannel level using the first derivatives of CHF with respect to the independent parameters. Three different CHF correlations for tubes (Groeneveld's CHF table, Katto correlation, and Biasi correlation) have been examined with uniformly heated bundle CHF data collected from various sources. A limited number of CHF data from a non-uniformly heated rod bundle are also evaluated with the aid of Tong's F-factor. The proposed method shows satisfactory CHF predictions for rod bundles both uniform and non-uniform power distributions.     

Experimental study on the effect of angles and positions of mixing vanes on CHF in a 2 × 2 rod bundle with working fluid R-134a

In this paper, the CHF experiment on the effect of angles and position of mixing vanes was performed in a 2 × 2 rod bundle. The test section had rectangular geometry in which four rod, each with a diameter of 9.5 mm, were inserted. The rod-to-rod gap was 3.15 mm, and the rod-to-wall gap was 1.575 mm. It was vertically installed in the test loop and was uniformly heated by electricity. The heating length was 1.125 m. The working fluid was R-134a. The mass flux ranged from 1000 to 1800 kg/m². The test pressure ranged from 14.67 to 25.67 bar. CHF data in the 2 × 2 rod bundle without a mixing vane were compared to the Bowring correlation and a CHF look-up table at equivalent hydraulic diameter. For this comparison, Katto's fluid-to-fluid model is applied. The results had a good agreement with error rates of 16 and 20%. In the CHF experiment with the mixing vanes with various angles, the angles of the mixing vanes were 20–40°. The CHF enhancement ratio (CER) was largest at 30°. CHF was enhanced up to 19%. A CHF experiment on the position of the mixing vane was also performed. In the experiment on the position of mixing vane, CER was reduced with increasing distance between grid and CHF location because swirl flow decayed. We also performed the CHF experiment on mixing vane developed by KAIST.     

Critical heat flux at high velocity channel flow with high subcooling

A quantitative analysis of critical heat flux (CHF) in heated channels under high mass flux with high subcooling was successfully carried out by applying a new flow model to the existing CHF model of a macro-water-sublayer on the heated wall and steam blankets over it. The CHF correlation proposed could correctly predict the existing experimental data for circular tubes of 0.33–4 mm in diameter with mass flux of 124–90 000 kg (m² s)⁻¹ and inlet water subcooling of 35–210 K at 0.1–7.1 MPa, resulting in CHF of 4.2–224 MW m⁻², and for rectangular channels of 3–20 mm gap with a mass flux of 940–27 000 kg (m² s)⁻¹ and inlet water subcooling of 13–166 K at 0.1–3.0 MPa, resulting in CHF of 2.0–62 MW m⁻². An error of the CHF correlation has also been estimated.     

Flooding limited CHF in a vertical 3 × 3 rod bundle with non-uniform axial heat flux

KAERI has performed an experimental study on the critical heat flux (CHF) under zero flow conditions with a non-uniformly heated 3 × 3 rod bundle. Experimental conditions are in the range of a system pressure from 0.50 to 14.96 MPa and inlet water subcooling enthalpies from 68 to 352 kJ/kg. The test section used in the present experiments consisted of a vertical flow channel, upper and lower plenums, and a non-uniformly heated 3 × 3 rod bundle. The experimental results show that the CHFs in low-pressure conditions are somewhat scattered within a narrow range. As the system pressure increases, however, the CHFs show a good parametric trend. The CHFs occur in the upper region of the heated section, but the locations of the detected CHFs move gradually in a downward direction with the increase of the system pressure. Even though the effects of the inlet water subcooling enthalpies and system pressure of the flooding CHF are relatively smaller than those of the flow boiling CHF, the CHF increases by increasing the inlet water subcooling enthalpies. Several existing correlations for the countercurrent flooding CHF based on Wallis's flooding correlation and Kutateladze's criterion for the onset of flooding are compared with the CHF data obtained in the present experiments to examine the applicability of the correlations.     

An utilization of liquid sublayer dryout mechanism in predicting critical heat flux under low pressure and low velocity conditions in round tubes

From a theoretical assessment of extensive critical heat flux (CHF) data under low pressure and low velocity (LPLV) conditions, it was found out that lots of CHF data would not be well predicted by a normal annular film dryout (AFD) mechanism, although their flow patterns were identified as annular–mist flow. To predict these CHF data, a liquid sublayer dryout (LSD) mechanism has been newly utilized in developing the mechanistic CHF model based on each identified CHF mechanism. This mechanism postulates that the CHF occurrence is caused by dryout of the thin liquid sublayer resulting from the annular film separation or breaking down due to nucleate boiling in annular film or hydrodynamic fluctuation. In principle, this mechanism well supports the experimental evidence of residual film flow rate at the CHF location, which can not be explained by the AFD mechanism. For a comparative assessment of each mechanism, the CHF model based on the LSD mechanism is developed together with that based on the AFD mechanism. The validation of these models is performed on the 1406 CHF data points ranging over P=0.1–2 MPa, G=4–499 kg m⁻² s⁻¹, L/D=4–402. This model validation shows that 1055 and 231 CHF data are predicted within ±30 error bound by the LSD mechanism and the AFD mechanism, respectively. However, some CHF data whose critical qualities are <0.4 or whose tube length-to-diameter ratios are <70 are considerably overestimated by the CHF model based on the LSD mechanism. These overestimations seem to be caused by an inadequate CHF mechanism classification and an insufficient consideration of the flow instability effect on CHF. Further studies for a new classification criterion screening the CHF data affected by flow instabilities as well as a new bubble detachment model for LPLV conditions, are needed to improve the model accuracy.     

Some problems for bundle CHF prediction based on CHF measurements in simple flow geometries

A comparison of critical heat flux (CHF) fuel bundles data with CHF data obtained in simple flow geometries was made. The base for the comparison was primary experimental data obtained in annular, circular, rectangular, triangular, and dumb-bell shaped channels cooled with water and R-134a. The investigated range of flow parameters (pressure, mass flux, and critical quality) in R-134a was chosen to be equivalent to modern nuclear reactor water flow conditions (p=7 and 10 MPa, G=350–5000 kg (m² s)⁻¹, x_cr=−0.1–1). The proper scaling laws were applied to convert the data from water to R-134a equivalent conditions and vise versa. The effects of flow parameters (p, G, x_cr) and the effects of geometric parameters (D, L) were evaluated during comparison. The comparison showed that no one simple flow geometry can be used for accurate and reliable bundle CHF prediction in wide range of flow parameters based on local (critical) conditions approach. The comparison also showed that the limiting critical quality phenomenon is unique characteristic for each flow geometry which depends on many factors: flow conditions (pressure and mass flux), geometrical parameters (diameter or surface curvature, gap size, etc.), flow obstructions (spacers, appendages, turbulizers, etc.) and others.     

圆管临界热流密度的流体模化

为了评价现有的临界热流密度(CHF)流体模化技术,在氟里昂-12热工实验装置上完成了φ202.5圆管CHF实验。在此基础上,用不同的数据源,比较了Ahmad补偿失真模型、鲁钟琪模型、Groeneveld模型和Stevens-Kirby经验比例因子模型的预测精度和适用范围,为复杂流道中的CHF流体模化研究奠定了基础。     

Experimental research progress on critical heat flux of Chinese PWR

One of the most important requirements in the design of pressurized water reactor (PWR) is to avoid the occurrence of critical heat flux (CHF). The design criteria for PWR specify that they must be operated at a certain percentage below CHF at all times and locations so as to the cladding temperature of fuel element at safe values. So in the process of safety assessment, CHF is one of important thermal-hydraulic parameters limiting the available power, whose size directly affects safety and economy of PWR nuclear power plant. This paper deals with a summary of experimental research progress on CHF of Chinese PWR. It mainly presents CHF experimental researches of Φ10 fuel assembly, CHF experimental researches of standard fuel assembly, and CHF experimental progress of non-uniform heated rod bundles. It should be emphasized that it also presents experimental research programs on CHF of Chinese advanced fuel assembly with self-reliance copyright. All CHF data obtained will be used for design improvement of Chinese PWR and R&D program of New Generation 1000 MWe PWR.     

Experimental study of the characteristics of the flow in the first rows of tube banks

This paper presents the experimental study of the flow instabilities in the first rows of tube banks. The study is performed using hot wire anemometry technique in an aerodynamic channel as well as flow visualizations in a water channel. In the wind channel three tube banks with square arrangement and pitch to diameter ratios P/D = 1.26, 1.4 and 1.6 were studied. The Reynolds number range for the velocities measurements, computed with the tube diameter and the flow velocity in the narrow gap between tubes was 7 × 10⁴–8 × 10⁴. Continuous and discrete wavelets were applied to decompose the velocity results, thus allowing the analysis of phenomena in time–frequency domain. Visualizations in a water channel complemented the analysis of the hot wire results. For this purpose, dye was injected in the flow in the water channel with a tube bank with P/D = 1.26. The range of the Reynolds number of the experiments was 3 × 10⁴–4 × 10⁴. The main results show the presence of instabilities, generated after the second row of the tube bank, which propagates to the interior of the bank. In the resulting flow, the three orthogonal components are equally significant. The three-dimensional behavior of the flow is responsible for a mass redistribution inside the bank that leads to velocity values not expected for the studied geometry, according to the known literature. The resulting flow process can be interpreted as a secondary flow which is characteristic of tube banks.     

A critical heat flux approach for square rod bundles using the 1995 Groeneveld CHF table and bundle data of heat transfer research facility

The critical heat flux (CHF) approach using CHF look-up tables has become a widely accepted CHF prediction technique. In these approaches, the CHF tables are developed based mostly on the data bank for flow in circular tubes. A set of correction factors was proposed by Groeneveld et al. [Groeneveld, D.C., Cheng, S.C., Doan, T., 1986. 1986 AECL-UO Critical Heat Flux lookup table. Heat Transf. Eng. 7(1–2), 46] to extend the application of the CHF table to other flow situations including flow in rod bundles. The proposed correction factors are based on a limited amount of data not specified in the original paper. The CHF approach of Groeneveld and co-workers is extensively used in the thermal hydraulic analysis of nuclear reactors. In 1996, Groeneveld et al. proposed a new CHF table to predict CHF in circular tubes [Groeneveld, D.C., et al., 1996. The 1995 look-up table for Critical Heat Flux. Nucl. Eng. Des. 163(1), 23]. In the present study, a set of correction factors is developed to extend the applicability of the new CHF table to flow in rod bundles of square array. The correction factors are developed by minimizing the statistical parameters of the ratio of the measured and predicted bundle CHF data from the Heat Transfer Research Facility. The proposed correction factors include: the hydraulic diameter factor (K_hy), the bundle factor (K_bf), the heated length factor (K_hl), the grid spacer factor (K_sp), the axial flux distribution factors (K_nu), the cold wall factor (K_cw) and the radial power distribution factor (K_rp). The value of constants in these correction factors is different when the heat balance method (HBM) and direct substitution method (DSM) are adopted to predict the experimental results of HTRF. With the 1995 Groeneveld CHF Table and the proposed correction factors, the average relative error is 0.1 and 0.0% for HBM and DSM, respectively, and the root mean square (RMS) error is 31.7% in DSM and 17.7% in HBM for 9852 square array data points of HTRF.     

Simulations of turbulence, heat transfer and mixing across narrow gaps between rod-bundle subchannels

Turbulent flow and temperature fields were determined numerically in a rectangular duct containing a heated rod. As the spacing δ between the rod and the duct wall decreased from 0.10D (D is the rod diameter) to 0.03D, coherent turbulent kinetic energy and temperature fluctuations dramatically increased in the gap region, but, for δ = 0.01D, coherent fluctuations essentially disappeared. As δ/D → 0, the frequency of coherent fluctuations decreased and cross-gap mixing weakened, contrary to predictions based on extrapolated available empirical correlations.     

Experimental data on burn-out in a simulated BWR fuel bundle

Critical heat flux (CHF) experiments have been carried out on a 16-rod test section having the typical geometry of boiling water reactor (BWR) fuel elements and in particular a 366 cm length. Heat fluxes were uniform, both axially and radially. The tests were carried out for the CNEN Plutonium Program on CISE's 8 MW IETI-3 facility, at 71 kg/cm² abs, mass velocities of 12–200 g/cm² s and inlet sub-cooling of 15–180°C. Each corner rod was instrumented with four separate thermocouples to detect nnd locate the initiation of CHF, while the other rods were instrumented with four-junction thermopiles.     

CHF prediction in nuclear fuel elements by using round tube data

The 1995 CHF table for uniformly heated round tubes, developed jointly by Canadian and Russian researchers, has been used for the prediction of critical heat flux (CHF) in 5×5 test sections simulating fuel elements of pressurized water reactors. Comparisons between measured and calculated CHF indicates that the table with an appropriate diameter correction can be applied to rod bundles of the type considered in this study. The relation for the diameter correction factor was derived from the CHF data. The tolerance limits associated with the departure from nucleate boiling ratio (DNBR) are evaluated by using statistical analysis.     

An experimental study on the critical heat flux for low flow of water in a non-uniformly heated vertical rod bundle over a wide range of pressure conditions

An experimental study of the critical heat flux (CHF) has been performed for a water flow in a non-uniformly heated vertical 3 × 3 rod bundle under low flow and a wide range of pressure conditions. The experiment was especially focused on the parametric trends of the CHF and the applicability of the conventional CHF correlations to a return-to-power conditions of a main steam line break accident whose conditions might be a low mass flux, intermediate pressure, and a high inlet subcooling. The effects of the mass flux and pressure on the CHF are relatively large and complicated in the low pressure conditions. At a high mass flux or a low critical quality, the local heat flux at the CHF location sharply decreases with an increasing local critical quality. However, at a low mass flux or a high critical quality, the local heat flux at the CHF location shows a nearly constant value regardless of the increase of the critical quality. The CHF data at the very low mass flux conditions are correlated well by the churn-to-annular flow transition criterion or the flow reversal phenomena. Several conventional CHF correlations predict the present return-to-power CHF data with reasonable accuracies. However, the prediction capabilities become worse in a very low mass flux of below about 100 kg/(m² s).     

CFD investigation of vertical rod bundles of supercritical water-cooled nuclear reactor

The commercial CFD code STAR-CD v4.02 is used as the numerical simulation tool for the supercritical water-cooled nuclear reactor (SCWR). The numerical simulation is based on the real full 3D rod bundles’ geometry of the nuclear reactors. For satisfying the near-wall resolution of y⁺ ≤ 1, the structure mesh with the stretched fine mesh near wall is employed. The validation of the numerical simulation for mesh generation strategy and the turbulence model for the heat transfer of supercritical water is carried out to compare with 3D tube experiments. After the validation, the same mesh generation strategy and the turbulence model are employed to study three types of the geometry frame of the real rod bundles. Through the numerical investigations, it is found that the different arrangement of the rod bundles will induce the different temperature distribution at the rods’ walls. The wall temperature distributions are non-uniform along the wall and the values depend on the geometry frame. At the same flow conditions, downward flow gets higher wall temperature than upward flow. The hexagon geometry frame has the smallest wall temperature difference comparing with the others. The heat transfer is controlled by P/D ratio of the bundles.     

Critical heat flux of forced convection boiling in an oscillating acceleration field — III. Reduction mechanism of CHF in subcooled flow boiling

To investigate the effect of variation in acceleration on the critical heat flux (CHF) in subcooled flow boiling, a photographic study was made. The test section was an internally heated vertical annulus with a glass shroud, in which Freon-113 flowed upwardly. The observation was made at a pressure of 3 bar, a mass flux of 920 kg/m²s, an inlet subcooling 45 K and a slightly lower heat flux level than steady CHF. The vertical acceleration was oscillated with amplitude of 0.3g_e and a period of 6 s.At low apparent gravitational acceleration, bubbles generated on the heated surface moved longer along the surface without detachment and coalesced with other bubbles to form large vapor slugs. This causes early CHF, the mechanism of which is dry-out of the liquid film existing between the heated surface and vapor slugs.