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.

Influence of Sump on Containment Thermal Hydraulics: Synthesis of the TOSQAN Tests

2015 ◽  
Vol 1 (4) ◽  
Author(s):  
Emmanuel Porcheron ◽  
Pascal Lemaitre ◽  
Amandine Nuboer
Keyword(s):  
Nuclear Power ◽  
Severe Accident ◽  
Water Evaporation ◽  
Plant Water ◽  
Pressurized Water Reactor ◽  
Steam Condensation ◽  
Pressurized Water ◽  
Noncondensable Gases ◽  
Nuclear Core ◽  
Hydraulic Conditions

During the course of a severe accident in a nuclear power plant, water can be collected in the sump containment through steam condensation on walls, cooling circuit leak, and by spray systems activation. Therefore, the sump can become a place of heat and mass exchanges through water evaporation and steam condensation, which influences the distribution of hydrogen released in containment during nuclear core degradation. The objective of this paper is to present the analysis of semi-analytical experiments on sump interaction between containment atmosphere for typical accidental thermal hydraulic conditions in a pressurized water reactor (PWR). Tests are conducted in the TOSQAN facility developed by the Institut de Radioprotection et de Sûreté Nucléaire in Saclay. The TOSQAN facility is particularly well adapted to characterize the distribution of gases in a containment vessel. A tests’ grid was defined to investigate the coupled effect of the sump evaporation with wall condensation, for air steam conditions, with noncondensable gases (He, SF6), and for steady and transient states (two depressurization tests).

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.

Technology Lecture

1983 ◽  
Vol 385 (1789) ◽  
pp. 241-251 ◽  
Keyword(s):  
Nuclear Power ◽  
Task Force ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Nuclear Accidents ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Turbine Generators ◽  
Set Up ◽  
Regulatory Commission

The Central Electricity Generating Board propose to build a pressurized water reactor at Sizewell in Suffolk. The PWR Task Force was set up in June 1981 to provide a communications centre for developing firm design proposals for this reactor. These were to follow the Standardized Nuclear Unit Power Plant System designed by Bechtel for the Westinghouse nuclear steam supply system for reactors built in the United States. Changes were required to the design to accommodate, for example, the use of two turbine generators and to satisfy British safety requirements. Differences exist between the British and American licensing procedures. In the U.K. the statutory responsibility for the safety of a nuclear power station rests unambiguously with the Generating Boards. In the U.S.A. the Nuclear Regulatory Commission issues detailed written instructions, which must be followed precisely. Much of the debate on the safety of nuclear powrer focuses on the risks of big nuclear accidents. It is necessary to explain to the public what, in a balanced perspective, the risks of accidents actually are. The vocabulary used in the nuclear power industry contributes to the misunderstanding and fear felt by the general public. The long-term consequences of big nuclear accidents can be presented in terms of reduction in life expectancy, increased chance of cancer or the equivalent pattern of compulsory cigarette smoking.

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.

Post Small-Break LOCA Long Term Core Cooling Performance in ACME Test Facility

2016 ◽  
Author(s):  
Zhanfei Qi ◽  
Sheng Zhu
Keyword(s):  
Nuclear Power ◽  
Cooling System ◽  
Safety Margin ◽  
Test Facility ◽  
Passive Safety ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Safety Features ◽  
Core Cooling

CAP1400 Pressurized Water Reactor is developed by China’s State Nuclear Power Technology Corporation (SNPTC) based on the passive safety concept and advanced system design. The Advanced Core-cooling Mechanism Experiment (ACME) integral effect test facility, which was constructed at Tsinghua University, represents a 1/3-scale height of CAP1400 RCS and passive safety features. It is designed to simulate the performance of CAP1400 passive core cooling system in the small break loss of coolant accidents (SBLOCA) for design certification, safety review and safety analysis code development. The Long Term Core Cooling (LTCC) post-LOCA could be simulated by ACME as well. A series of test cases with various break sizes and locations with post-LOCA LTCC period were conducted in ACME facility. This paper describes the post-LOCA LTCC test conducted in ACME test facility. The LTCC phenomena in different cases are very similar. It’s found that the interval that switching from IRWST injection to sump recirculation has the least safety margin. However, it’s shown that the post-LOCA LTCC in ACME could be well maintained by passive core cooling system according to the test results even though the recirculation water level in ACME IRWST-2 is lower than the containment recircualtion level in CAP1400 conservatively.

The Failure Analysis and Processing of Digital Reactor Protection System

2013 ◽  
Author(s):  
Jun Zhao ◽  
Xing Zhou ◽  
Jin Hu ◽  
Yanling Yu
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Phase 1 ◽  
Protection System ◽  
Pressurized Water Reactor ◽  
Communication Failure ◽  
Pressurized Water ◽  
Power Failure ◽  
Self Test ◽  
Reactor Protection System

The Qinshan Nuclear Power Plant phase 1 unit (QNPP-1) has a power rating of 320 MWe generated by a pressurized water reactor that was designed and constructed by China National Nuclear Corporation (CNNC). The TELEPERM XS I&C system (TXS) is to be implemented to transform analog reactor protection system (RPS) in QNPP-1. The paper mainly describes the function, structure and characteristic of RPS in QNPP-1. It focuses on the outstanding features of digital I&C, such as strong online self-test capability, the degradation of the voting logic processing, interface improvements and CPU security. There are some typical failures during the operation of reactor protection system in QNPP-1. The way to analyze and process the failures is different from analog I&C. The paper summarizes typical failures of the digital RPS in the following types: CPU failure, communication failure, power failure, Input and output (IO) failure. It discusses the cause, risk and mainly processing points of typical failure, especially CPU and communication failures of the digital RPS. It is helpful for the maintenance of the system. The paper covers measures to improve the reliability of related components which has been put forward effective in Digital reactor protection system in QNPP-1. It will be valuable in nuclear community to improve the reliability of important components of nuclear power plants.

Thermal-Hydraulic Studies of the Steam Separation in Horizontal Steam Generator at PGV Test Facility

2009 ◽  
Author(s):  
Yuriy V. Parfenov ◽  
Oleg I. Melikhov ◽  
Vladimir I. Melikhov ◽  
Ilya V. Elkin
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Test Facility ◽  
Evolutionary Development ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Tube Bundles ◽  
Horizontal Steam Generator ◽  
Calculation Results ◽  
Steam Separation

A new design of nuclear power plant (NPP) with pressurized water reactor “NPP-2006” was developed in Russia. It represents the evolutionary development of the designs of NPPs with VVER-1000 reactors. Horizontal steam generator PGV-1000 MKP with in-line arrangement of the tube bundles will be used in “NPP-2006”. PGV test facility was constructed at the Electrogorsk Research and Engineering Center on NPP Safety (EREC) to investigate the process of the steam separation in steam generator. The description of the PGV test facility and tests, which will be carried out at the facility in 2009, are presented in this paper. The experimental results will be used for verification of the 3D thermal-hydraulic code STEG, which is developed in EREC. STEG pretest calculation results are presented in the paper.

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