The Development and Application of the Engineering and Design Analyzer of CAP1400 Nuclear Power Plant

2017 ◽  
Author(s):  
He Yuanlei ◽  
Zhang Qijiang ◽  
Li Xiaoyan
Keyword(s):  
Control System ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Research And Development ◽  
Nuclear Power ◽  
Human Factor ◽  
Typical Application ◽  
Human Factor Engineering ◽  
Engineering Research ◽  
Acceptance Condition

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.

Methodology and Application of Human Reliability Analysis in CAP1400 Nuclear Power Plant

2017 ◽  
Author(s):  
Yongping Qiu ◽  
Jiandong He ◽  
Juntao Hu ◽  
Yucheng Zhuo ◽  
Jie He
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Reliability Analysis ◽  
Nuclear Power ◽  
Power Plants ◽  
Human Factor ◽  
Human Reliability ◽  
Human Reliability Analysis ◽  
Human Factor Engineering ◽  
Human Failure

It is well recognized that humans play an important role in the safety operation of nuclear power plants (NPPs). Usually three types of human interactions (HIs) are defined in the human reliability analysis (HRA) of probabilistic safety assessment (PSA) for NPPs, i.e., pre-initiating event HIs, initiating event-related HIs, and post-initiating event HIs. In this paper, a brief introduction of the HRA methodology for CAP1400 nuclear power plant is first presented, including internal events and external events (mainly internal fire and flooding) HRA. Next, the CAP1400 human failure event quantification content is given with a typical example, and some insights and proposals based on CAP1400 PSA/HRA results are discussed. Finally, the application of HRA in human factor engineering design of CAP1400 is described. The human actions (HAs) most important to safety are identified via a combination of probabilistic and deterministic analyses, and then addressed when conducting the human factor engineering program. The CAP1400 HRA is one of the most important PSA elements and provides fundamental support for CAP1400 PSA and the relevant applications.

Recirculation Pump Speed Control System Upgrades in Nuclear Boiling Water Reactors

2008 ◽  
Author(s):  
James W. Morgan
Keyword(s):  
Control System ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Flow Control ◽  
Nuclear Power ◽  
Boiling Water ◽  
Boiling Water Reactors ◽  
Pump Speed ◽  
Water Reactors ◽  
Generator Sets

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.

Development of a digital turbine control system in a nuclear power plant

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