scholarly journals Design of digital nuclear power small reactor once-through steam generator control system

2022 ◽  
Author(s):  
Hong Qian ◽  
Mingyao Zou
Keyword(s):  
Control System ◽  
Steam Generator ◽  
Nuclear Power

Research on Fuzzy Decoupled and Coordinated Control of Nuclear Power Plant

2010 ◽  
Author(s):  
Wei Zhang ◽  
Guoqing Xia ◽  
Ming Ai ◽  
Minyu Fu ◽  
Hegao Cai
Keyword(s):  
Control System ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Generator ◽  
Nuclear Power ◽  
Coordinated Control ◽  
Steam Pressure ◽  
Momentum Conservation ◽  
Dynamic Feature ◽  
Decoupled Control

The mathematic models of reactor, once-through steam generator and turbine are built based on the mass, energy and momentum conservation theorem. Because of serious coupling and different dynamic characteristic, the coordinated control that solves big system problem is presented to apply into the nuclear power plant after researching deeply the variety feature and coupling relation of primary parameters of the nuclear power plant. The coordinated control system is filled with manage control, coordinated control and bottom controller. The simulation is processed by changing turbine load. Compared with non-coordinated control system, the coordinated control system improves briefly the dynamic feature of nuclear power plant. The fuzzy decoupled control strategy between once-through steam generator and turbine is proposed. The fuzzy decoupled frame including a compensator and design method of the decoupled compensator are given, the fuzzy rules are applied in the decoupled compensator. Finally, a fuzzy decoupled control system is designed in detail with a two inputs and two outputs’ system, which is applied in the coordinated control system of the nuclear power plant. The simulation results show that the coordinated control system based on the decoupled strategies is better than the coordinated control system, which weakens the couple connection, reduces the fluctuation of exit steam pressure by adjusting the feedwater flux.

scholarly journals Evolved Fuzzy Control System for a Steam Generator

2010 ◽  
Vol 5 (2) ◽  
pp. 179 ◽  
Author(s):  
Daniela Hossu ◽  
Ioana Făgărășan ◽  
Andrei Hossu ◽  
Sergiu St. Iliescu
Keyword(s):  
Control System ◽  
Water Level ◽  
Steam Generator ◽  
Measurement Errors ◽  
Nuclear Power ◽  
Power Plants ◽  
Fuzzy Model ◽  
Operating Conditions ◽  
Steam Generators ◽  
Control Approach

Poor control of steam generator water level is the main cause of unexpected shutdowns in nuclear power plants. Particularly at low powers, it is a difficult task due to shrink and swell phenomena and flow measurement errors. In addition, the steam generator is a highly complex, nonlinear and time-varying system and its parameters vary with operating conditions. Therefore, there is a need to systematically investigate the problem of controlling the water level in the steam generator in order to prevent such costly reactor shutdowns. The objective of this paper is to design, evaluate and implement a water level controller for steam generators based on a fuzzy model predictive control approach. An original concept of modular evolved control system, seamless and with gradual integration into the existent control system is proposed as base of implementation of the presented system.

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.

scholarly journals A Conceptual Approach to Eliminate Bypass Release of Fission Products by In-Containment Relief Valve under SGTR Accident

2018 ◽  
Vol 2018 ◽  
pp. 1-12
Author(s):  
Taeseok Kim ◽  
Wonjun Choi ◽  
Joongoo Jeon ◽  
Nam Kyung Kim ◽  
Hoichul Jung ◽  
...  
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Fission Products ◽  
Mitigation Strategies ◽  
Severe Accident ◽  
Base Case ◽  
Conceptual Approach ◽  
Relief Valve ◽  
Radioactive Nuclides

During a hypothesized severe accident, a containment building is designed to act as a final barrier to prevent release of fission products to the environment in nuclear power plants. However, in a bypass scenario of steam generator tube rupture (SGTR), radioactive nuclides can be released to environment even if the containment is not ruptured. Thus, thorough mitigation strategies are needed to prevent such unfiltered release of the radioactive nuclides during SGTR accidents. To mitigate the consequence of the SGTR accident, this study was conducted to devise a conceptual approach of installing In-Containment Relief Valve (ICRV) from steam generator (SG) to the free space in the containment building and it was simulated by MELCOR code for numerical analysis. Simulation results show that the radioactive nuclides were not released to the environment in the ICRV case. However, the containment pressure increased more than the base case, which is a disadvantage of the ICRV. To minimize the negative effects of the ICRV, the ICRV linked to Reactor Drain Tank (RDT) and cavity flooding was performed. Because the overpressurization of containment is due to heat of ex-vessel corium, only cavity flooding was effective for depressurization. The conceptual design of the ICRV is effective in mitigating the SGTR accident.

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