Structure Design of Pressure Equipment for Experimental Test on Steam Condensation on the Cold Surface

2012 ◽  
Author(s):  
Lin Yang ◽  
Lingyun Li ◽  
Liyong Han
Keyword(s):  
Cooling System ◽  
Heat Removal ◽  
Structure Design ◽  
Pressurized Water Reactor ◽  
Steam Condensation ◽  
Thermal Measurement ◽  
Cold Surface ◽  
Pressurized Water ◽  
University Of Wisconsin ◽  
Passive Safety System

The advanced pressurized water reactor (APWR) designed by Westinghouse uses a passive safety system which relies on heat removal by condensation to maintain the containment within the design limits of pressure and temperature. Steam condensation inside surface of the containment is one of the most important phenomena during heat removing process in the passive containment cooling system (PCCS). It is very significant for engineering design and code development to study the mechanism of steam condensation on cold surface. There was an experiment done by University of Wisconsin on this subject. However, the pressure equipment cannot support high pressure. In this paper, new pressure equipment was designed. It can support higher pressure and also meet other thermal measurement requirements.

Structure Design of the Pressured Vessel in Experimental Test on Steam Condensation on the Cold Surface for CAP1400

2014 ◽  
Author(s):  
Lin Yang ◽  
Liyong Han
Keyword(s):  
Experimental Test ◽  
Cooling System ◽  
Heat Removal ◽  
Structure Design ◽  
Natural Phenomena ◽  
Pressurized Water Reactor ◽  
Steam Condensation ◽  
Cold Surface ◽  
Pressurized Water ◽  
Passive Safety System

The advanced pressurized water reactor (APWR) uses a passive safety system relying on heat removal by condensation to maintain the containment within the design limits of pressure and temperature. The passive containment cooling system (PCCS) includes many natural phenomena mechanisms. Steam condensation is one of the most important phenomena. It is very significant for engineering designing and code developing to study the mechanism of steam condensation on cold surface. In this paper, the test pressurized vessel in the experimental test on steam condensation on the cold surface for CAP1400 is designed, and the structure pressure is calculated.

Tightness Analysis of Pressured Equipment in Experimental Test on Steam Condensation on the Cold Surface

2013 ◽  
Author(s):  
Lin Yang ◽  
Liyong Han
Keyword(s):  
Experimental Test ◽  
Cooling System ◽  
Passive Safety ◽  
Natural Phenomena ◽  
Pressurized Water Reactor ◽  
Steam Condensation ◽  
Cold Surface ◽  
Pressurized Water ◽  
University Of Wisconsin ◽  
Passive Safety System

To maintain the containment within the design limits of pressure and temperature, the advanced pressurized water reactor (APWR) designed by Westinghouse uses a passive safety system to transfer the heat from inner containment to outside. The passive containment cooling system (PCCS) includes many natural phenomena mechanisms. Steam condensation is one of the most important phenomena. Most heat is removed by steam condensation on inside surface of the containment during the postulated design basic accidents (DBA). It is very significant for engineering designing and code developing to study the mechanism of steam condensation on cold surface. There was an experiment made by University of Wisconsin on it. In this paper, the structure pressure of the pressured equipment is calculated and the tightness is also analyzed.

Analysis of Effects of Buoyancy Force on Steam Condensation With Non-Condensable Gases in a Rectangular Channel

2014 ◽  
Author(s):  
Xian-Mao Wang ◽  
Hua-Jian Chang ◽  
Li-Yong Han
Keyword(s):  
Mixed Convection ◽  
Buoyancy Force ◽  
Nuclear Power ◽  
Cooling System ◽  
Science And Technology ◽  
Rectangular Channel ◽  
Pressurized Water Reactor ◽  
Steam Condensation ◽  
Pressurized Water ◽  
Set Up

With the development of science and technology, some important passive features have been used in nuclear reactors, one of which is passive containment cooling system (PCCS). In the system, steam condensation plays an important role in removing heat from the containment atmosphere during a postulated accident. It has been found that during most time of an accident, the gas regime in the containment will be under natural and mixed convection. Advanced pressurized water reactor (CAP1400), designed by State Nuclear Power Technology Corporation (SNPTC) in China, is one of Chinese national science and technology projects. Since the PCCS has been applied in CAP1400, the study of condensation with non-condensable gases under natural and mixed convection becomes necessary. To have a deeper understanding on the phenomenon of condensation with non-condensable gases under natural and mixed convection, an experiment facility was set up by State Nuclear Power Technology Research & Development Centre (SNPTRD). The test section of the facility is a rectangular channel with one of the walls acting as a condensing plate. The effects of buoyancy force on steam condensation with non-condensable gases are investigated. Also, a CFD model is set up to simulate the process.

Experimental Research on Steam Condensation on Cold Surface: Facility Design

2014 ◽  
Author(s):  
Li-Yong Han ◽  
Lin Yang ◽  
Shan Zhou ◽  
Shen Wang ◽  
Chun-Lai Tian ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Test Section ◽  
Cooling System ◽  
Theoretical Research ◽  
Test Facility ◽  
Measurement Technology ◽  
Ambient Conditions ◽  
Steam Condensation ◽  
Cold Surface ◽  
Process Gas

The passive containment cooling system (PCCS) of the 3rd generation APWR utilizes natural phenomena to transfer the heat released from the reactor to the environment during postulated designed basic accidents. Steam condensation on the inner surface of the containment shell is one of the most dominate mechanism to keep the ambient conditions within the design limits. Extensive experiment and theoretical research shows condensation is a complex process, gas pressure, film temperature and velocity of the gas have impact on the heat transfer coefficient. To span the expected range of conditions and provide proper model for evaluating the condensation heat transfer process, SCOPE test facility was designed by State Nuclear Power Technology Research & Development Centre (SNPTRD) in various conditions anticipated the operating range of CAP1400 in accident conditions. Pressurized test section with a rectangular flowing channel was used, with one of the walls cooled to maintain low temperature for condensing, supplying systems was designed for different pressures, gas temperatures, velocities and coolant water temperatures. Facility components, test section structure, supplying systems and measurement technology were described in this paper, also results of some pre-tests was introduce to show property of the facility.

An Experimental Investigation of High Level Vibration on the Residual Heat Removal Line of a Pressurized Water Reactor

2009 ◽  
Author(s):  
Jeffrey A. Brown ◽  
Robert D. Blevins ◽  
H. Joseph Fernando
Keyword(s):  
Heat Removal ◽  
Coupled System ◽  
Acoustic Resonance ◽  
Acoustic Mode ◽  
Side Branch ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
High Level ◽  
Residual Heat

This paper presents the results of a scaled aero acoustic test that modeled a side branch resonance observed in the residual heat removal suction line of a large pressurized water reactor. Resolution of the acoustic resonance was sought by detuning the eddy shedding frequency from the fundamental side branch acoustic mode. The specific physical modifications and their ability to detune the coupled system are presented.

Experimental Research on Non-Condensable Gases Effects in Passive Decay Heat Removal System

2016 ◽  
Author(s):  
Yang Liu ◽  
Haijun Jia ◽  
Li Weihua
Keyword(s):  
High Pressure ◽  
Heat Removal ◽  
Volume Ratio ◽  
Test Facility ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Decay Heat ◽  
Optimizing Design ◽  
Heat Removal System ◽  
Removal System

Passive decay heat removal (PDHR) system is important to the safety of integral pressurized water reactor (IPWR). In small break LOCA sequence, the depressurization of the reactor pressure vessel (RPV) is achieved by the PDHR that remove the decay heat by condensing steam directly through the SGs inside the RPV at high pressure. The non-condensable gases in the RPV significantly weaken the heat transfer capability of PDHR. This paper focus on the non-condensable gas effects in passive decay heat removal system at high pressure. A series of experiments are conducted in the Institute of Nuclear and New Energy Technology test facility with various heating power and non-condensable gas volume ratio. The results are significant to the optimizing design of the PDHR and the safety operation of the IPWR.

Sensitivity Study of LOCA-Generated Debris on the Pressure Drop of Fuel Assembly

2017 ◽  
Author(s):  
Hammad Aslam Bhatti ◽  
Zhangpeng Guo ◽  
Weiqian Zhuo ◽  
Shahroze Ahmed ◽  
Da Wang ◽  
...  
Keyword(s):  
Pressure Drop ◽  
Fuel Assembly ◽  
Cooling System ◽  
Safety Issue ◽  
Cooling Water ◽  
Sensitivity Study ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Core Cooling

The coolant of emergency core cooling system (ECCS), for long-term core cooling (LTCC), comes from the containment sump under the loss-of-coolant accident (LOCA). In the event of LOCA, within the containment of the pressurized water reactor (PWR), thermal insulation of piping and other materials in the vicinity of the break could be dislodged. A fraction of these dislodged insulation and other materials would be transported to the floor of the containment by coolant. Some of these debris might get through strainer and eventually accumulate over the suction sump screens of the emergency core cooling systems (ECCS). So, these debris like fibrous glass, fibrous wool, chemical precipitates and other particles cause pressure drop across the sump screen to increase, affecting the cooling water recirculation. As to address this safety issue, the downstream effect tests were performed over full-scale mock up fuel assembly. Sensitivity studies on pressure drop through LOCA-generated debris, deposited on fuel assembly, were performed to evaluate the effects of debris type and flowrate. Fibrous debris is the most crucial material in terms of causing pressure drop, with fibrous wool (FW) debris being more efficacious than fibrous glass (FG) debris.

PKL Tests on Heterogeneous Inherent Boron Dilution Following SB-LOCA: Applicability to Reactor Scale

2010 ◽  
Author(s):  
Klaus Umminger ◽  
Simon Philipp Schollenberger ◽  
Se´bastien Cornille ◽  
Claire Agnoux ◽  
Delphine Quintin ◽  
...  
Keyword(s):  
Cooling System ◽  
Natural Circulation ◽  
Limited Range ◽  
Test Facility ◽  
Test Results ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
The Core ◽  
Physical Phenomena ◽  
Boron Dilution

In the course of a small break LOCA in a Pressurized Water Reactor (PWR) the flow regime in the Reactor Cooling System (RCS) passes through a number of different phases and the filling level may decrease down to the point where the decay heat is transferred to the secondary side under Reflux-Condenser (RC) conditions. During RC, the steam formed in the core condensates in the Steam Generator (SG) U-tubes. For a limited range of break size and configuration, a continuous accumulation of condensate may cause the formation of boron-depleted slugs. If natural circulation reestablishes, as the RCS is refilled, boron-depleted slugs might be transported to the Reactor Pressure Vessel (RPV) and to the core. To draw conclusions on the risk of boron dilution processes in SB-LOCA transients, two important issues, the limitation of slug size and the onset of Natural Circulation (NC) have to be assessed on the basis of experimental data, as system Thermal-Hydraulic codes are limited in their capability to replicate the complex physical phenomena involved. The OECD PKL III tests were performed at AREVA’s PKL test facility in Erlangen, Germany, to evaluate important phases of the boron dilution transient in PWRs. Several integral and separate effect tests were conducted, addressing the inherent boron dilution issue. The PKL III integral transient test runs provide sufficient data to state major conclusions on the formation and maximum possible size of the boron-depleted slugs, their boron concentration and their transport into the RPV with the restart of NC. Some of these conclusions can be applied to reactor scale. It has to be mentioned, that even though this paper is based on PKL test results obtained within the OECD PKL project, the conclusions of this paper reflect the views of the authors and not necessarily of all the members of the OECD PKL project.

Experimental Comparison on the Multi-Hole Steam Spraying Condensation Heat Transfer Characteristics

2018 ◽  
Author(s):  
Yuhao Zhang ◽  
Daogang Lu ◽  
Bin Ouyang ◽  
Yonglong Yuan
Keyword(s):  
Heat Transfer ◽  
High Speed ◽  
Experimental Results ◽  
Experimental Comparison ◽  
Pressurized Water Reactor ◽  
Heat Transfer Characteristics ◽  
Pressurized Water ◽  
Transfer Characteristics ◽  
Passive Safety System ◽  
Contact Condensation

In the third generation pressurized water reactor AP1000 plant, the Automatic Depressurization System (ADS) is one of the most important passive safety system. As the steam continues to discharge into the IRWST, the high-temperature and high-pressure steam condenses in the IRWST intensely and rapidly. In the present work, multi-hole spargers DCC experimental bench has been built to study the condensation and mixing phenomena in the accidental depressurization events in AP1000. The key parameters including the temperature, velocity, and flow patterns are monitored by the thermocouples, PIV technique, and high speed camera, etc. Based on the experimental results, the heat transfer characteristics of the steam-jet Direct Contact Condensation (DCC) in the ADS depressurization process were analyzed. The experimental comparison shows that the spraying steam behaviors are closely related to the steam mass flux, which determined the ejection patterns. The spraying steam is condensed rapidly in the core region with the flow regime locating in chugging, transitional chugging, and stable ejection stages. The 88-hole experimental results fit well with Fukuda’s condensation oscillation correlations, whereas the 2-hole experimental results are slightly lower than the values predicted by traditional single hole correlation.

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