Natural Circulation Flow Rate Research Articles

Steady characteristic investigation on passive residual heat removal system of Chinese advanced PWR

Thermal-hydraulic characteristic investigation on passive residual heat removal system (PRHRS) of Chinese advanced PWR was conducted to provide input data for PRHRS design and to demonstrate the feasibility of unique design features. A total of 237 sets of test data at steady state have been obtained and the main influence factors on the two-phase natural circulation flow rate and residual heat removal capability were identified. On the basis of theory analysis, a correlation of two-phase natural circulation was obtained, and relative errors of 95% test data were less than ±16%. There is a considerable effect of the system status parameters on the threshold of height between heat source and heat sink, and its correlation of two-phase natural circulation system has been obtained. The steady characteristic research shows that PRHRS has the capability of removing the core decay power through natural circulation.

Analysis on natural circulation capacity of the CARR

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.

Natural circulation flow through water-in-glass evacuated tube solar collectors

Experimental and numerical investigations were undertaken to develop a correlation for natural circulation flow rate through single-ended water-in-glass evacuated tubes mounted over a diffuse reflector. The circulation flow rate was correlated in terms of solar input, tank temperature, collector inclination and tube aspect ratio. The sensitivity of the flow rate correlation to the variation in circumferential heat flux distribution was also investigated.

ICONE15-10439 NATURAL CIRCULATION IN ESBWR

The ESBWR is a natural circulation Boiling Water Reactor (BWR). Natural circulation provides major simplification by removal of the recirculation pumps and associated equipment. It is also synergistic with the passive safety systems. ESBWR is designed for natural circulation with enhanced natural circulation flow rate and greatly improved stability performance relative to operating BWRs at natural circulation conditions. The performance of the ESBWR has been established using the state-of-the-art TRACG computer code, which has been validated against a wide range of test data. Conservative design criteria for stability are easily satisfied. Natural circulation in the ESBWR provides numerous benefits with large margins to thermal design limits.

06/02337 Modeling of solar passive techniques for roof cooling in arid regions: Jain, D. et al. Building and Environment, 2006, 41, (3), 277–287.

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.

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.

Thermal–hydraulic characteristics of a next-generation reactor relying on steam generator secondary side cooling for primary depressurization and long-term passive core cooling

System experiments were conducted at the ROSA-V Large Scale Test Facility (LSTF) for investigation of new safety systems to mitigate consequences of postulated accidents in pressurized water rectors (PWRs). Tested systems included a steam generator (SG) secondary-side automatic depressurization system (SADS) and gravity-driven injection system (GDIS), which are candidates of safety systems for some next-generation PWR designs. The experimental results showed several thermal–hydraulic behaviors typical of these safety systems, including the primary depressurization due to natural circulation cooling, a nonuniform flow behavior among SG U-tubes, an accumulation of the non-condensable gas originally contained in the injected water, liquid holdup in U-tubes due to the countercurrent flow limiting, and long-term passive core cooling with the GDIS injection. From the assessment of the RELAP5/MOD3 code using the present data, it was found that the inability of the code to predict the U-tube nonuniform flow behavior resulted in overprediction of the primary depressurization rate at a pressure less than 1 MPa, and exaggerated oscillation of the natural circulation flow rate in the primary loop.

Experimental Analysis on Mitigation of Thermal Stratification in the Suppression Pool of a Water Wall Type Passive Containment Cooling System

A water wall type passive containment cooling system with an outer pool surrounding the suppression pool is one passive containment cooling system. In the system, a baffle plate in the suppression pool mitigates thermal stratification formed at the vent tube outlet level and enlarges the heat transfer area. The effectiveness of the baffle plate in mitigating thermal stratification was experimentally confirmed; the heat transferred to the outer pool increased about 50% due to a larger high temperature region and a longer effective heat transfer length. The experimental analysis was performed using a three-dimensional thermal-hydraulic analysis program. In the analysis, a laminar flow model and slip conditions on structural walls were used, and the calculated temperature profiles and natural circulation flow rates along the baffle plate agreed with measurements. The model was then judged as a valid and practical tool to evaluate global natural circulation and temperature distributions in a large pool. And i...

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.

RETRAN-02 Comparison of Natural Circulation Flow Rates at Babcock & Wilcox 177-FA Plants

A very important aspect of nuclear steam supply system (NSSS) model development is the process of comparing the computer model results against actual plant responses. Good comparisons will qualify the computer model for specific engineering analyses. Flow rates and decay heat power levels were obtained from planned and unplanned natural circulation events that occurred at Arkansas Nuclear One, Crystal River, Davis-Besse, and Oconee nuclear power plants. A oneloop RETRAN model of the Oconee NSSS is used to attain a spectrum of steady-state equilibrium conditions at different power levels of 25, 50, 75, and 100 MW. The benchmark comparisons are respectable. The comparisons also illustrate the ability of the Babcock and Wilcox raised-loop plant to induce a greater natural circulation flow rate.