An experimental investigation of a passive cooling unit for nuclear plant containment

A set of condensation experiments in the presence of noncondensables (e.g. air, helium) was conducted to evaluate the heat removal capacity of a passive cooling unit in a post-accident containment. Condensation heat transfer coefficients on a vertically mounted smooth tube have been obtained for total pressure ranging from 2.48×10⁵ Pa(abs) to 4.55×10⁵ Pa(abs) and air mass fraction ranging from 0.30 to 0.65. An empirical correlation for heat transfer coefficient (h), has been developed in terms of a parameter group made up of steam mole fraction (Xs), total pressure (P_t), temperature difference between bulk gas and wall surface (dT). This correlation covers all data points within 20%. All data points are also in good agreement with the prediction of the diffusion layer model (DLM) with suction and are approximately 2.2 times the Uchida heat transfer correlation. Experiments with an axial shroud around the test tube to model the restriction on radial flow experienced within a tube bundle demonstrated a reduction of the heat transfer coefficient by a factor of about 0.6. The effect of helium (simulating hydrogen) on the heat transfer coefficient was investigated for helium mole fraction in noncondensable gases (X_He/X_nc) at 15, 30 and 60%. It was found that the condensation heat transfer coefficients are generally lower when introducing helium into noncondensable gas. The difference is within 20% of air-only cases when X_He/X_nc is less than 30% and total pressure is less than 4.55×10⁵ Pa(abs). A gas stratification phenomenon was clearly observed for helium mole fraction in excess of 60%.     

Vapour void fraction in subcooled flow boiling

Experimental data on steam void fraction and axial temperature distribution in an annular boiling channel for low mass-flux forced and natural circulation flow of water with inlet subcooling have been obtained. The ranges of variables covered are: mass flux = 1.4 × 10⁴−1.0 × 10⁵ kg/hr m²; heat flux = 4.5 × 10³−7.5 × 10⁴ kcal/hr m²; and inlet subcooling = 10–70°C. The present and literature data match well with the theoretical void predictions using a four-step method similar to that suggested by Zuber and co-workers.     

Condensation in the presence of noncondensable gases

An experimental investigation to examine the effects of surface orientation on the condensation of steam in the presence of noncondensable gas is reported. An air-steam mixture was directed into a rectangular flow-channel over a condensing aluminum surface that has a painted surface finish. The mixture flow was concurrent in all the tests with condensate flow. In this series of experiments, the orientation of the condensing surface was varied from 0–90° (plate surface was facing downwards at 0°), with a variable air-steam mass fraction of 0–0.87, and a mixture velocity of 1–3 m/s. The heat transfer coefficient was measured in addition to making visual observations of the condensation process. It was found that the heat transfer coefficient varied from 100 to 600 W/m² K with the mass fraction of 0.87-0.24 and the maximum heat transfer coefficient of 6200 W/m² K was measured with mass fraction of 0. By tilting the condensing surface from the horizontal to vertical position, the heat transfer coefficient decreased 15 to 25% depending on the mass fraction. With a higher vapor content the effect of the orientation was smaller. This dependence was attributed to the existence of interfacial structure (droplets and ridges) that promoted heat transfer at small inclination angles, when the angle was increased the interface became smoother and heat transfer rates decreased. Heat transfer rates were also observed to increase with flow velocity, vapor content and pressure. The results are compared with some previously published data and a proposed condensation model that showed reasonable agreement with the data trends.     

Wall temperature measurement and heat transfer correlation of turbulent supercritical carbon dioxide flow in vertical circular/non-circular tubes

Experimental data are presented which describe heat transfer characteristics of turbulent supercritical carbon dioxide flow in vertical tubes with circular, triangular, and square cross-sections. Experiments are conducted at a constant pressure of 8 MPa under various conditions such as inlet bulk temperatures ranging from 15 to 32 °C, imposed heat fluxes from 3 to 180 kW/m², and mass velocities from 209 to 1230 kg/m² s. The corresponding Reynolds and Grashof numbers are in the range of 3 × 10⁴ to 1.4 × 10⁵, and 5 × 10⁹ to 4 × 10¹¹, respectively. The test section is composed of an entrance region of 0.6 m long and a heating region of 1.2 m long. Wall temperatures are measured by thermocouples installed at the outer surface of the heating region. In order to identify the effect of the cross-sectional shape on the supercritical heat transfer, wall temperature distributions in the streamwise direction are compared at the same heat flux and mass velocity conditions. Based on the wall temperature data, an improved heat transfer correlation, which can be applicable to both forced convection and mixed convection regimes, is proposed, and compared with previous ones.     

Critical heat flux and heat transfer above mixture level under high-pressure boil-off conditions in PWR type and tight-lattice type fuel bundles

Heat transfer tests were conducted in PWR 17 × 17 type and tight-lattice type fuel bundles under high-pressure boil-off (very-low flow, mass fluxes lower than 100 kg/m²s) conditions. There is almost no significant difference in both critical heat flux (CHF) (or dryout point) data and convective heat transfer data above the mixture level between the PWR type and tight-lattice type bundles. The “complete vaporization equation” predicts well the CHF data, i.e. the dryout occurs nearly at the elevation where the thermal-equilibrium quality reaches 1.0. The Groeneveld CHF table used in the RELAP5/MOD3 code should be improved in the region of mass flux between 10 and 100 kg/m²s. The radiative heat transfer has an important contribution to total heat transfer above the mixture level. The Dittus-Boelter correlation, with use of the film temperature in evaluating steam properties, predicts well the convective heat transfer above the mixture level.     

Experimental analysis of heat transfer within the AP600 containment under postulated accident conditions

The new AP600 reactor designed by Westinghouse uses a passive safety system relying on heat removal by condensation to keep the containment within the design limits of pressure and temperature. Even though some research has been done so far in this regard, there are some uncertainties concerning the behavior of the system under postulated accident conditions. In this paper, steam condensation onto the internal surfaces of the AP600 containment walls has been investigated in two scaled vessels with similar aspect ratios to the actual AP600. The heat transfer degradation in the presence of noncondensable gas has been analyzed for different noncondensable mixtures of air and helium (hydrogen simulant). Molar fractions of noncondensables/steam ranged from (0.4–4.0) and helium concentrations in the noncondensable mixture were 0–50% by volume. In addition, the effects of the bulk temperatures, the mass fraction of noncondensable/steam, the cold wall surface temperature, the pressure, noncondensable composition, and the inclination of the condensing surface were studied. It was found that the heat transfer coefficients ranged from 50 to 800 J s⁻¹ K⁻¹ m⁻² with the highest for high wall temperatures at high pressure and low noncondensable molar fractions. The effect of a light gas (helium) in the noncondensable mixture were found to be negligible for concentrations less than approximately 35 molar percent but could result in stratification at higher concentrations. The complete study gives a large and relatively complete data base on condensation within a scaled AP600 containment structure, providing an invaluable set of data against which to validate models. In addition, specific areas requiring further investigation are summarized.     

An investigation on heat transfer characteristics of different pressure steam-water in vertical upward tube

Within the range of pressure from 9 to 30 MPa, mass velocity from 600 to 1200 kg/(m² s), and heat flux at inner wall from 200 to 600 kW/m², experiments have been performed to investigate the heat transfer characteristics of steam-water two-phase flow in vertical upward tube. The outer diameter of the tube is 32 mm, and the wall thickness is 3 mm. Based on results, it was found that Dryout is the main mechanism of the heat transfer deterioration in the sub-critical pressure region. Near the critical pressure, when the heat transfer deterioration occurs, the steam quality of water is lower than that in the sub-critical pressure region, so that DNB is the main mechanism in this pressure region. At supercritical pressure, the heat transfer performance in circular channel is improved and enhanced. Heat transfer deterioration phenomenon is observed when the fluid bulk temperature approaches to the pseudo-critical value. Nusselt correlation of the forced-convection heat transfer in supercritical pressure region has been provided, which can be used to predict heat transfer coefficient of the vertical upward flow in tube.     

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).     

Experimental and analytical study of direct contact condensation of steam in water

The object of this work is to improve our understanding and analysis capability for direct contact condensation of steam in water. Transition criteria between regimes of direct contact condensation have been proposed. A transition criterion for the onset of chugging has been developed from the transient conduction model as well as a criterion for the existence of a condensing jet from the two-layer turbulent eddy transfer model. In order to analyze the effect of non-condensable gas on the chugging boundary, a transient conduction-diffusion model for steam and steam-gas mixtures can be calculated. In particular, the amount of non-condensable gas required to suppress chugging can be quantified by use of this model. The critical gas content is found to be of the order of a few percent. In addition, a methodology is suggested for calculating the product of the interfacial area and the heat transfer coefficient for oscillatory jets. All the models and transition criteria developed are applicable for upward steam injection only with the exception of the methodology used for calculating the heat transfer of high velocity steam jets.     

A numerical investigation of natural convection heat transfer in horizontal spent-fuel storage cask

Natural convection heat transfer in a horizontally placed dry spent-fuel storage cask is numerical investigated. The commercial computational fluid dynamics (CFD) code, -3.2 is used and the laminar and turbulent model are employed. The numerical predictions obtained are compared with the experimental data reported by Nishimura et al. [J. Nucl. Sci. Technol. 33 (1996) 821]. The computational results corresponding to laminar model agree well with the experimental data, but the calculated results of turbulent model are higher. The velocity pattern and the isotherms are drawn. With the increasing of Rayleigh number, the heat transfer in the cask changes from conduction dominant mode to convection dominant mode. In the condition of Ra_m=1.3×10⁹, turbulent model prevails. The convective heat transfer is so strong that almost all temperature changes take place in the region near the wall of the cask. The Rayleigh number Ra_m and the Nusselt number Nu_m characterized by maximum temperature difference are defined to depict the heat transfer characteristics. It is found laminar and turbulent models predict the same trend but different value. The flow patterns in the cask can be divided to three regimes. In these three regimes, modified Nusselt numbers are proportional to the 0.7, 0.25 and 0 power of the modified Rayleigh number, respectively.     

Investigation of pre- and post-dryout heat transfer of steam-water two-phase flow in a rod bundle

Pre- and post-dryout heat transfer experiments were performed for steam-water two-phase flow in a 5 × 5 rod bundle under conditions of total mass fluxes from 80 to 320 kg/m²s, inlet qualities from 0.1 to 0.8, heat fluxes from 3 to 26 W/cm² and a pressure of 3 MPa. Heater rod surface temperatures or heat transfer coefficients predicted by several correlations were compared with experimental data with emphasis on the applicability of the correlations to the present experimental conditions which were pertinent to thermal-hydraulic conditions during a LOCA in a nuclear reactor. The Chen and Biorge et al. correlations underestimated heat transfer coefficients in the pre-dryout region. The Varone-Rohsenow prediction which accounted for the thermal nonequilibrium effect, calculated heater rod surface temperatures relatively well in the post-dryout region over the whole region of the present experimental conditions. The Dittus-Boelter and Groeneveld correlations predicted heater rod surface temperatures relatively well in the post-dryout region under high total mass flux conditions, but underestimated considerably under low total mass flux conditions.     

Power up-grading study for the first Egyptian research reactor

In the present work, power up-grading study is performed, for the first Egyptian Research Reactor (ET-RR-1), using the present fuel basket with 4×4 fuel rods, (17.5 mm pitch), and a proposed fuel basket with 5×5 fuel rods, (14.0 mm pitch), without violating the thermal hydraulic safety criteria. These safety criteria are; fuel centerline temperature (fuel melting), clad surface temperature (surface boiling), outlet coolant temperature, and maximum heat flux (critical heat flux ratio). Different thermal reactor powers (2–10 MW) and different core coolant flow rates (450, 900, 1350 m³ h⁻¹) are considered. The thermal hydraulic analysis was performed using the subchannel code COBRA-IIIC for the estimation of temperatures, coolant velocities and critical heat flux. The neutronic calculations were performed using WIMS-D4 code with 5 — group neutron cross section library. These cross sections were adapted to use in the two-dimensional (2-D) diffusion code DIXY for core calculations. The study concluded that ET-RR-1 power can be upgraded safely up to 4 MW with the present 4×4-fuel basket and with the proposed 5×5-fuel basket up to 5 MW with the present coolant flow rate (900 m³ h⁻¹). With the two fuel arrays, the reactor power can be upgraded to 6 MW with coolant flow rate of 1350 m³ h⁻¹ without violating the safety criterion. It is also concluded that, loading the ET-RR-1 core with the proposed fuel basket (5×5) increases the excess reactivity of the reactor core than the present 4×4 fuel matrix with equal U-235 mass load and gave better fuel economy of fuel utilization.     

管径与倾角对管束外含空气蒸汽冷凝传热影响研究

通过对不同管径和倾角的3×3管束开展管外含空气蒸汽冷凝试验,研究了传热管管径和倾角影响管束外含空气蒸汽冷凝传热的基本规律。结果表明:管径和倾角的影响在不同压力范围内具有明显差异。在压力0.8 MPa以下,冷凝传热系数总体随管径和倾角的减小而增大,管径12 mm、0°倾角传热管的冷凝传热系数较管径19 mm、90°倾角的冷凝传热系数最大可增加29%。在压力0.8 MPa以上,冷凝传热系数随管径的减小而减小,最大可降低18%;随倾角的减小先减小后增大,在约60°倾角时,冷凝传热系数最小。      

JR curves of the low alloy steel 20 MnMoNi 5 5 with two different sulphur contents in oxygen-containing high temperature water at 240°C

J_R curves of the low alloy steel 20 MnMoNi 5 5 with two different sulphur contents (0.003 and 0.011 wt.%) were determined at 240°C in oxygen-containing high temperature water as well as in air. The tests were performed by the single-specimen unloading compliance technique at load line displacement rates from 1 × 10⁻⁴ down to 1 × 10⁻⁶ mm s⁻¹ on 20% side-grooved 2T CT specimens in an autoclave testing facility at an oxygen content of 8 ppm and a pressure of 7 MPa under quasi-stagnant flow conditions.In the case of testing in high temperature water, remarkably lower J_R curves than in air at the same load line displacement rate (1 × 10⁻⁴ mm s⁻¹) were obtained. A decrease in the load line displacement rate as well as an increase in the sulphur content of the steel caused a reduction of the J_R curves. At the fastest load line displacement rate a stretch zone could be detected fractographically on the specimens tested in air and in high temperature water and consequently J_i could be determined. When testing in high temperature water, the J_i value of the higher sulphur material type decreases from 45 N mm⁻¹ in air to 3 N mm⁻¹, much more than that of the optimized material type from 51 N mm⁻¹ in air to 20 N mm⁻¹ at 1 × 10⁻⁴ mm s⁻¹.     

Experimental Study on Heat Removal Characteristics for Water Wall Type Passive Containment Cooling System

To evaluate the heat removal capability of a water wall type cooling system, which is one passive containment cooling system (PCCS), the thermal hydraulic behavior in the suppression pool (S/P) and the outer pool (O/P, flat plate water wall) have been investigated experimentally. The following results were obtained. (1) A thermal stratification boundary, which separates the pools into the upper high temperature and lower low temperature regions, was formed just below the vent tube outlet. (2) Convection heat transfer characteristics in the S/P and O/P along the primary containment vessel (PCV) wall had no significant differences and were those of natural convection. Correlation of the natural convection heat transfer up to the Ra number of 2×10¹⁴ was obtained. (3) Vertical variations of local condensation heat transfer coefficients under a noncondensable gas presence were within ±10% of the average value for the 4.7 m heat transfer length. An experimental correlation for the average condensation heat transfer coefficients was obtained as a function of steam and noncondensable gas mass ratio. (4) An analytical model to evaluate the system performance of the water wall type PCCS was verified. (5) A baffle plate concept to mitigate thermal stratification at the vent outlet and to enlarge the high temperature region in the S/P was considered as a means to improve heat release capability. Thermal hydraulics with a baffle plate were examined, and effectiveness of the baffle plate to improve the heat release capability was confirmed.     

A power perturbation technique for the measurement of fuel-to-coolant heat transfer coefficients in fast reactors

This technique provides a method of obtaining average fuel to coolant heat transfer coefficients for individual fuel subassemblies in fast reactors. A series of experiments on the UK prototype fast reactor (PFR) over the period 1977–1979 have demonstrated that the technique is simple, requires no special instrumentation other than thermocouples to monitor coolant outlet temperatures, and the measurement can be made during normal reactor operation. Thus it is possible to determine how heat transfer coefficients change with operating conditions and with the degree of burn-up in the fuel.The analysis of a single experiment is presented to illustrate the technique. This was conducted at a single reduced power level of 200 thermal megawatts for two different primary coolant flow rates, both steady fractions of the maximum (0.88 and 0.47). Cyclic and single-step perturbations of about 10% amplitude were impressed on the steady power and the delayed coolant temperature response at subassembly outlets was monitored. Burn-ups in the subassemblies ranged between 1.0% and 4.7%. From the measured delays at the two flows it was possible to determine the fuel time-constant and hence the fuel-to-coolant heat transfer coefficient. It was also shown that a simple, lumped-element, heat transfer model can be used to obtain sufficiently accurate estimates from measurements at just one coolant flow.Fuel surface-to-coolant thermal conductances (i.e. gap conductances) were subsequently derived from the heat transfer coefficients. These ranged between 2.4 kW m⁻² K⁻¹ and 3.3 kW m⁻² K⁻¹ with the smaller conductances being obtained for those fuel elements with the larger degree of burn-up. These values are lower than expected but consistent with a higher than expected value for the negative power coefficient of reactivity feedback which has been observed at reduced power.     

Fluence dependence of the surface roughness of InP after bombardment

InP(1 0 0) surfaces were sputtered under ultrahigh vacuum conditions by 5 keV ions at an angle of incidence of 41° to the sample normal. The fluence, , used in this study, varied from 1 × 10¹⁴ to 5 × 10¹⁸ cm⁻². The surface topography was investigated using field emission scanning electron microscopy (FE-SEM) and atomic force microscopy (AFM). At the lower fluences ( 5 × 10¹⁶ cm⁻²) only conelike features appeared, similar in shape as was found for noble gas ion bombardment of InP. At the higher fluences, ripples also appeared on the surface. The bombardment-induced topography was quantified using the rms roughness. This parameter showed a linear relationship with the logarithm of the fluence. A model is presented to explain this relationship. The ripple wavelength was also determined using a Fourier transform method. These measurements as a function of fluence do not agree with the predictions of the Bradley–Harper theory.     

Analysis of flow distribution a PWR fuel rod bundle model containng A 90% blockage

An experimental investigation, covering a Reynolds number range from 2 × 10³ to 3.5 × 10⁴, was conducted to study the velocity and turbulence intensity distributions due to the presence of a blockage in an unheated 7 × 7 rod bundle. The blockage configuration, consisting of a 4 × 4 rod array, created a maximum flow area reduction of 90% in the central nine subchannels. The blockage sleeve length was 38.3 × rod diameter and the 90% blockage zone length extended for 16.4 × rod diameter. The results showed that upstream of the blockage, the flow was not influenced by the blockage until it reached the location where the inlet taper section of the swelling started. At the downstream end, the flow disturbance was extensive and persisted over a distance of about 83 rod diameters. Compared to the downstream velocity profiles, the turbulence intensity measurements however showed a faster recovery from the blockage influence. At the higher Reynolds number, velocity profiles calculated using the COBRA subchannel computer code compared consistently with the experimental data. The general flow behaviour of the various subchannels was reasonably well predicted. However, at low Reynolds number, due mainly to the frictional form loss calculation scheme in COBRA and uncertainty in the flow transition, the flow diversion due to the blockage to the surrounding unblocked subchannels was overestimated. The influence of the degree of recovery from the rod swelling on the flow was also studied using COBRA.     

Fluid-to-fluid modeling of critical heat flux in 4 × 4 rod bundles

Fluid-to-fluid modeling of critical heat flux (CHF) is to simulate the CHF behaviors for water by employing low cost modeling fluid, and the flow scaling factor is the key to apply the technique to fuel bundles. The CHF experiments in 4×4 rod bundles have been carried out in Freon-12 loop in equivalent nuclear reactor water conditions (P=10.0–16.0 MPa, G=488.0–2100.0 kg/m² s, X_cr=−0.20–0.30). The models in fluid-to-fluid modeling of CHF is verified by the CHF data for Freon-12 obtained in the experiment and the CHF correlation for water obtained by Nuclear Power Institute of China (NPIC) in the same 4×4 rod bundles. It has been found that the S.Y. Ahmad Compensation Distortion model, the Lu Zhongqi model, the Groeneveld model and Stevens–Kirby model overpredict the bundles CHF values for water. Then an empirical correlation of flow scaling factor is proposed. Comparison of the CHF data in two kinds of test sections for Freon-12, in which the distance of the last grid away the end of heated length is different, shows that the spacer grid, which is located at 20 mm away from the end of the heated length, has evidently influenced on the CHF value in the 4×4 rod bundles for Freon-12. This is different from that for water, and the need for further work is required.     

Simulating test for thermal mixing in the hot gas chamber of the HTR-10

The helium coolant at the outlet of the pebble bed core of the 10 MW High Temperature Gas-cooled Reactor-Test Module exhibits a severe radial temperature deviation. In order to avoid damages at the downstream components due to alternating thermal loads such as the steam generator, a hot gas chamber is especially designed to solve the problem. Thermal mixing performance of the coolant in the hot gas chamber is experimentally investigated on a 1:1.5 scale model by air. The experimental result shows that within the Reynolds number range of 1.4×10⁵–5.8×10⁵, the hot gas chamber with a radial mixer reaches excellent thermal mixing of the coolant of about 94%. The flow resistance coefficient for the hot gas chamber is also presented.