Research and development of virtual software for safety control system TXS in nuclear power plant

Author(s):  
Leng Shan ◽  
Cheng Junjie ◽  
Ruan Huifeng ◽  
Liu Yun
Author(s):  
Shihong Ma

Sanmen Nuclear Power Plant is constructing the first AP1000 unit in the world, and DCS system is adopted as the non-safety control system. This paper reviews the experience of Sanmen project, and analyzes the necessity of nuclear power plant Owner’s management for developing the DCS system, as well as how to manage the development effectively.


Author(s):  
He Yuanlei ◽  
Zhang Qijiang ◽  
Li Xiaoyan

In the process of research and development for a new nuclear power plant, it is very necessary to develop a dynamic platform and tools to analyze and verify the plant control & protect system and human factor engineering. Therefor, Shanghai Nuclear Engineering Research and Development Institute (SNERDI) developed the Engineering & Design Analyzer of CAP1400 Nuclear Power Plant (CAP1400 EDA) which provides a dynamic platform environment for analyzing and verifying the control system and human factor engineering of the CAP1400 nuclear power plant, a new Gen III passive nuclear power plant. In this paper, the mechanism and implement approach of the CAP1400 EDA will be mainly introduced, for example, the platform architecture of the EDA, analysis tools integrated in the EDA and CAP1400 nuclear power plant modes based on the EDA. In the meantime, a typical application case based on the CAP1400 EDA will be demonstrated in this paper, for example the capability of the NSSS control system will be verified in a ramp load down & raise operate transient. In this transient process, the NSSS control system of the plant is assessed whether it has the capability to keep the key parameter and state of the plant in an acceptance condition or range. And also other transients such as step load transient, large load transient can be simulated on CAP1400 EDA to verify whether or not the NSSS control systems are properly designed.


Author(s):  
James W. Morgan

The nuclear power industry is faced with determining what to do with equipment and instrumentation reaching obsolescence and selecting the appropriate approach for upgrading the affected equipment. One of the systems in a nuclear power plant that has been a source of poor reliability in terms of replacement parts and control performance is the reactor recirculation pump speed/ flow control system for boiling water reactors (BWR). All of the operating BWR-3 and BWR-4’s use motor-generator sets, with a fluid coupled speed changer, to control the speed of the recirculation water pumps over the entire speed range of the pumps. These systems historically have had high maintenance costs, relative low efficiency, and relatively inaccurate speed control creating unwanted unit de-rates. BWR-5 and BWR-6 recirculation flow control schemes, which use flow control valves in conjunction with two-speed pumps, are also subject to upgrades for improved performance and reliability. These systems can be improved by installing solid-state adjustable speed drives (ASD), also known as variable frequency drives (VFD), in place of the motor-generator sets and the flow control valves. Several system configurations and ASD designs have been considered for optimal reliability and return on investment. This paper will discuss a highly reliable system and ASD design that is being developed for nuclear power plant reactor recirculation water pump controls. Design considerations discussed include ASD topology, controls architecture, accident, transient and hydraulic analyses, potential reactor internals modifications, installation, demolition and economic benefits.


2009 ◽  
Vol 7 (1) ◽  
pp. 67-73 ◽  
Author(s):  
In-Kyu Choi ◽  
Jong-An Kim ◽  
Chang-Ki Jeong ◽  
Joo-Hee Woo ◽  
Ji-Young Choi ◽  
...  

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