Prediction of Subcooled Water Flow Boiling Critical Heat Flux (CHF) at Low Pressure

2017 ◽  
Author(s):  
Fangxin Hou ◽  
Huajian Chang ◽  
Yufeng Zhao ◽  
Ming Zhang ◽  
Peipei Chen ◽  
...  
Keyword(s):  
Heat Flux ◽  
Critical Heat Flux ◽  
Water Flow ◽  
Pressure Vessel ◽  
High Speed ◽  
Flow Boiling ◽  
Reactor Vessel ◽  
Lower Head ◽  
Inclination Angles ◽  
Reactor Pressure

In vessel retention (IVR) is one of the key severe accident mitigation strategies to maintain reactor pressure vessel (RPV) integrity. IVR designs utilize the reactor pressure vessel lower head to contain molten fuel and rely on external reactor vessel cooling (ERVC) to remove decay heat. The capacity of ERVC is limited by the critical heat flux (CHF) of flow boiling on the outside of the reactor vessel surface. Therefore, the determination of critical heat flux (CHF) is crucial to predict whether the adoption of IVR would be successful in mitigating severe accidents. In 1999, Celeta et.al proposed a superheated layer vapor replenishment model. In this model they postulated that CHF would occur when the superheated layer was occupied by the vapor blanket coming into contact with the heated wall and they successfully predicted the critical heat flux in subcooled water flow boiling under high mass flux, high liquid subcooling and low/medium pressure conditions. To evaluate the practicability of this model in predicting CHF under IVR conditions, CHF experiments were performed under natural circulation conditions on the experiment facility “Test of External Vessel Surface with Enhanced Cooling” (TESEC). Experiments are conducted in a 30 mm wide, 61mm high rectangular flow channel with a 200 mm long heated surface along the flow direction. Two quartz windows are installed at the sidewalls of the flow channel for visualization. In order to simulate various positions of the reactor lower head, experiments at different inclination angles of the test section were conducted. The high speed visualization data at CHF point at various inclination angles were processed and analyzed by a MATLAB code developed by the author. The vapor blanket thickness at various inclination angles was measured from the visualization data and was also predicted by the Celeta model. By using geometry data from high speed images, CHF values were calculated by Celeta model and compared with the experimental results at various inclination angles. Limitations of the Celeta model in adaptation of predicting CHF under IVR conditions were further discussed.

Investigation on Structural Behaviors of Reactor Pressure Vessel With the Effects of Critical Heat Flux and Internal Pressure

2016 ◽  
Vol 139 (2) ◽  
Author(s):  
Jianfeng Mao ◽  
Jianwei Zhu ◽  
Shiyi Bao ◽  
Lijia Luo ◽  
Zengliang Gao
Keyword(s):  
Heat Flux ◽  
Internal Pressure ◽  
Critical Heat Flux ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Wide Range ◽  
Depth Analysis ◽  
Structural Behaviors ◽  
Reactor Pressure

The so-called “in-vessel retention (IVR)” is a severe accident management strategy, which is widely adopted in most advanced nuclear power plants. The IVR mitigation is assumed to be able to arrest the degraded melting core and maintain the structural integrity of reactor pressure vessel (RPV) within a prescribed hour. Essentially, the most dangerous thermal–mechanical loads can be specified as the combination of critical heat flux (CHF) and internal pressure. The CHF is the coolability limits of RPV submerged in water (∼150 °C) and heated internally (∼1327 °C), it results in a sudden transition of boiling crisis from nucleate to film boiling. Accordingly, from a structural integrity perspective, the RPV failure mechanisms span a wide range of structural behaviors, such as melt-through, creep damage, plastic deformation as well as thermal expansion. Furthermore, the geometric discontinuity of RPV created by the local material melting on the inside aggravates the stress concentration. In addition, the internal pressure effect that usually neglected in the traditional concept of IVR is found to be having a significant impact on the total damage evolution, as indicated in the Fukushima accident that a certain pressure (up to 8.0 MPa) still existed inside the RPV. This paper investigates structural behaviors of RPV with the effects of CHF and internal pressure. In achieving this goal, a continuum damage mechanics (CDM) based on the “ductility exhaustion” is adopted for the in-depth analysis.

Experimental Investigation on Critical Heat Flux From Downward-Facing Flat Plate for Different Orientation Angles

2018 ◽  
Author(s):  
K. H. Deng ◽  
Y. Zhang ◽  
C. L. Wang ◽  
Y. P. Zhang ◽  
W. X. Tian ◽  
...  
Keyword(s):  
Heat Flux ◽  
Critical Heat Flux ◽  
Pressure Vessel ◽  
Nuclear Reactor ◽  
Heating Surface ◽  
Severe Accident ◽  
Width Ratio ◽  
Lower Head ◽  
Length Width ◽  
Core Materials

After the severe accident inside a nuclear reactor, the IVR (In-vessel retention) management strategy is an effective way to keep the integrity of pressure vessel and reduce risk of radioactive leakage by holding the damaged core materials through External Reactor Vessel Cooling (ERVS). The damaged core materials aggregate in the lower head of pressure vessel and releasing heat to the lower head. Therefore, it is very important to remove heat timely to keep the integrity of pressure vessel by ERVS. The shape of lower head is hemispherical and the local Critical Heat Flux (CHF) of different parts changed with latitude. In this paper, influence of orientation angles, area and length-width ratio on CHF of plate heating surface for saturated pool boiling is investigate experimentally. The results show that CHF increases with increasing orientation angles and decreasing area, meanwhile, length-width ratio has a significantly effect on CHF.

Critical Heat Flux Measurement and Model for Refrigerant-123 Under Stabilized Flow Conditions in Microchannels

2006 ◽  
Author(s):  
Wai Keat Kuan ◽  
Satish G. Kandlikar
Keyword(s):  
Experimental Data ◽  
Heat Flux ◽  
Critical Heat Flux ◽  
High Speed ◽  
Flow Boiling ◽  
Flux Measurement ◽  
Absolute Error ◽  
Cross Sectional ◽  
Boiling Process ◽  
Parallel Microchannels

The present work is aimed toward understanding the effect of flow boiling stability on critical heat flux (CHF) with Refrigerant-123 (R-123) in microchannel passages. Experimental data and theoretical model to predict the CHF are the focus of this work. The experimental test section has six parallel microchannels with each having a cross sectional area of 1054 × 157 μm2. The effect of flow instabilities in microchannels is investigated using flow restrictors at the inlet of each microchannel to stabilize the flow boiling process and avoid the backflow phenomena. This technique resulted in successfully stabilizing the flow boiling process as seen through a high-speed camera. The present CHF result is found to correlate to mean absolute error (MAE) of 24.1% with a macroscale empirical equation by Katto [13]. A theoretical analysis of flow boiling phenomena revealed that the ratio of evaporation momentum to surface tension forces is an important parameter. For the first time, a theoretical CHF model is proposed using these underlying forces to represent CHF mechanism in microchannels, and its correlation agrees with the experimental data with MAE of 2.5%.

RELAP Simulation on Boiling and Flow Phenomena Under External Reactor Vessel Cooling Condition

2010 ◽  
Author(s):  
Yongchun Li ◽  
Weihua Zhou ◽  
Yanhua Yang ◽  
Bo Kuang ◽  
Xu Cheng
Keyword(s):  
Pressure Vessel ◽  
Flow Boiling ◽  
Reactor Vessel ◽  
Reactor Pressure Vessel ◽  
Severe Accident ◽  
Study Program ◽  
External Cooling ◽  
Flow Phenomena ◽  
Key Issues ◽  
Reactor Pressure

External reactor vessel cooling (ERVC) of the In-vessel retention (IVR) system is widely accepted as a feasible way to remove decay heat from the lower head of the reactor pressure vessel (RPV) under severe accident (SA) conditions. However, some issues relating to ERVC still need to be evaluated before its application, such as boiling and flow phenomena and CHF prediction, etc. To study these key issues, an experimental study program named REPEC (Reactor Pressure Vessel External Cooling) is performed at Shanghai Jiao Tong University. Steady state experiments focusing on flow boiling phenomena investigation are carried out with comprehensive measurements, including temperature distribution, pressure drop and mass flow rate. As a part of studies on boiling mechanism and flow phenomena between RPV and the insulation, the experiment is analyzed and simulated with RELAP code. The code simulation covers most of the experimental cases, and a comparison between simulation results and experimental data are presented and discussed.

Enhancement of RPV Critical Heat Flux by Injection Flow

2017 ◽  
Author(s):  
Wei Chen ◽  
Canhui Sun ◽  
Jun Geng
Keyword(s):  
Heat Flux ◽  
Critical Heat Flux ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Metallic Layer ◽  
Fluid Temperature ◽  
Injection Flow ◽  
Severe Accidents ◽  
Reactor Pressure ◽  
Residual Heat

Under severe accidents, the reactor pressure vessel is flooded with water and the residual heat is removed by two-phase natural circulation through the flow channel between the reactor vessel and thermal insulation. If the heat flux of the outer wall is lower than local critical heat flux, the residual heat can be removed, and if the heat flux of the outer wall is higher than local critical heat flux, the reactor pressure vessel should be molten. For AP-type reactors, like AP1000 and CAP1400, critical heat flux of the reactor pressure vessel is the heat transfer limits of residual heat under severe accidents. Previous studies indicate that after severe accident a two-layer molten pool can be formed, namely metallic layer and oxide layer. Compared with oxide layer, in metallic layer, the heat flux more easily exceeds the heat transfer limits due to its low thermal resistance. In this study, an approach was proposed to enhance local critical heat flux. This approach is expected to be used in local area around reactor pressure wall, like metallic layer, to increase the reactor pressure vessel intact probability under severe accidents. In this new approach, injection flow channels are added to the present flow channel by adding simple flow pipes from insulation near 60 to 80 degree where exceeding critical heat flux is most likely to happen. The fluid flow under external reactor vessel cooling (ERVC) condition is divided into two parts: one part is from downward (0 degree) to upward (90 degree) along the curved reactor pressure vessel and the other part is from injection pipe (about 70 degree) to upward. The fluid temperature from injection pipe is lower than that from downward due to residual heat from the reactor pressure wall. And hence, the local critical heat flux is likely to increase because of inject turbulence and low fluid temperature. An experimental facility is conducted to study the mechanism of injection influence on critical heat flux under normal pressure condition. There are two main loops in this facility: one is main loop while the other is injection loop. The test section is an inclined downward heated rectangular channel with its inclined angle varied from 0 degree to 90 degree. Flow and thermal conditions are listed: in main loop, mass flow velocity ranging from 100kg/m2s to 600kg/m2s with fluid temperature from 90 °C to 105 °C; In injection loop, mass flow velocity ranging from 0 to 600 kg/m2s with fluid temperature from 85 °C to 105 °C. Under the above condition, with and without injection flow, critical heat flux experiments were conducted. It indicates that injection velocity has great effect on critical heat flux, while injection subcooled has little effect. The critical heat flux can be increased by 0.07MW/m2 to 0.33MW/m2 depending on various injection velocities and main loop conditions.

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