scholarly journals DESIGN ANALYSIS OF A PREPACKAGED NUCLEAR POWER PLANT (1000 eKW). VOLUME I. PRIMARY AND SECONDARY SYSTEM DESIGN

1959 ◽  
Author(s):  
Not Given Author
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
System Design ◽  
Nuclear Power ◽  
Design Analysis ◽  
Secondary System

Study for Application of Safety Soft Control and Information System in Highly-Integrated Control Rooms of CPR1000 NPP

2013 ◽  
Author(s):  
Xuegang Zhang ◽  
Wei Liu ◽  
Hai Chang ◽  
Jianbo Wen ◽  
Yiqian Wu ◽  
...  
Keyword(s):  
Information System ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
System Design ◽  
Nuclear Power ◽  
Integrated Control ◽  
Continuous Control ◽  
Design Basis ◽  
Control Rooms ◽  
Soft Control

For most of the newly built nuclear power plants, the computerized main control rooms (MCR) are adopted. The soft control, the typical feature of computerized Human-Interface System (HIS) in the computerized main control room and mediated by software rather than by direct physical connections, is comprised of safety and non-safety control interface which provides the operators with manual control for component-level, and allows both continuous control of plant process and discrete control of components in nuclear power plant. The safety soft control and information system (SSCIS) is used to give the safety commands to and check the immediate response of the safety process. This paper describes the application of the system design basis, functionality, communication, operation faceplate and system modes for SSCIS which is firstly introduced in CPR1000 nuclear power plant. The design criteria and basic design features of SSCIS is developed to be as the design basis of the design implementation. The ISG-04 ‘Highly-Integrated Control Rooms-Communications issues (HICRc)’ provides acceptable methods for addressing SSCIS communications in digital I&C system design. The NUREG0700 ‘Human-System Interface Design Review Guidelines’ is applied as reference for human factor engineering requirement in the SSCIS design. And the SSCIS design has also fully considered the possible customer usual practice.

Mechanical and Hydraulic Loads on the Supports of the Rotary Machine Due to Shaft Seizure in a Nuclear Power Plant

2004 ◽  
Author(s):  
S. K. Gupta ◽  
B. Chatterjee ◽  
Rajesh Kumar
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Heat Transport ◽  
Transport System ◽  
Nuclear Power ◽  
Inertial Load ◽  
Secondary System ◽  
Rotating Shaft ◽  
Rotary Machine ◽  
Major Equipment

In a nuclear power plant there are two major equipment with high mechanical inertia, which have a rotating shaft. These are Pumps in the Primary Heat Transport System and Turbine in the secondary system. In both cases, the shaft seizure leads to transfer of very large loads to the supports. These supports, if not designed for seizure loads may fail. If the supports fail, there is a good possibility of a missile generated and hit the safety equipment. Seizure loads in these machines have three components namely mechanical inertial load, electrical load and hydraulic load. While the electrical and hydraulic loads have a limited peak value, the inertial load depends on the seizure time. For the normal observed seizures the three have a similar order of magnitude during seizure. As the casing is overdesigned the combined load is experienced by the supports. The pump of the Primary Heat Transport System (PHTS) of a nuclear power plant is centrifugal type run by an induction motor. If the pump shaft seizes, the seizure load will be experienced by the support structure. Due to the presence of the flywheel, the total moment of inertia of the pump motor assembly is quite high. Hence the resisting torque may be higher than the support’s design torque. Besides, the electric torque will continue to be applied as the motor trip on the overload current is delayed by several seconds as the corresponding relay is a thermal relay. Seizure torque will depend on pump seizure time. Lesser the seizure time, higher would be the load on the pump supports. The turbine in the secondary system has a large inertia due to blades. In case of a seizure the generator is tripped in hundreds of milliseconds. The load experienced by turbine supports due to seizure is significantly enhanced in the first few seconds due to sustained steam supply before it is cut off. This paper discusses the estimation of the three types of loads during seizure of the shafts in the pumps and turbine. It also discusses the possible safety consequences of these loads.

Coupling of RELAP and LabVIEW to Model a Nuclear Power Plant System Realistically

2014 ◽  
Author(s):  
J. C. Pack ◽  
Z. Fu ◽  
F. Aydogan
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Virtual Instrument ◽  
Nuclear Power ◽  
System Modeling ◽  
Loop Model ◽  
Secondary System ◽  
Loss Of Coolant Accident ◽  
Accident Scenarios ◽  
Computer Simulators

Within the study and design of a nuclear power plant extensive system modeling is necessary to determine how the reactor is going to perform in any given situation, not only in the normal performance of the reactor but also transients including unanticipated transients without scram and hypothetical accidents. One of the difficulties in the performance of this modeling is that there are often separate programs used to model the primary and other loops in multiple loop systems. When the modeling requires no interaction between the loops, this method is adequate but in many of these scenarios an understanding of the interaction loops is crucial especially in the case of transients including accident scenarios. However, each loop is generally modeled individually and there is no feedback effect between loops. The purpose of this article is to demonstrate how this coupling between the primary and secondary system of a typical PWR can be performed. The primary and secondary sides of the PWR are modeled with Reactor Excursion and Leak Analysis Program (RELAP5) and Laboratory Virtual Instrument Engineering Workbench (LabVIEW) computer simulators respectively. Primary loop model includes a four loops PWR. The coupling between RELAP5 and LabVIEW has been executed with steady state and transients, in this case a loss of coolant accident (LOCA). The results of the coupling have been compared with the typical RELAP5 results without coupling.

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