Coordinated Control of a Small Pressurized Water Reactor

2018 ◽  
Author(s):  
Peiwei Sun ◽  
Chong Wang
Keyword(s):  
Control System ◽  
Reactor Core ◽  
Reactor Power ◽  
Pressurized Water Reactors ◽  
Primary System ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Control Rod ◽  
Average Temperature ◽  
Water Reactors

Small Pressurized Water Reactors (SPWR) are different from those of the commercial large Pressurized Water Reactors (PWRs). There are no hot legs and cold legs between the reactor core and the steam generators like in the PWR. The coolant inventory is in a large amount. The inertia of the coolant is large and it takes a long time for the primary system to respond to disturbances. Once-through steam generator is adopted and its water inventory is small. It is very sensitive to disturbances. These unique characteristics challenge the control system design of an SPWR. Relap5 is used to model an SPWR. In the reactor power control system, both the reactor power and the coolant average temperature are regulated by the control rod reactivity. In the feedwater flow control system, the coordination between the reactor and the turbine is considered and coolant average temperature is adopted as one measurable disturbance to balance them. The coolant pressure is adjusted based on the heaters and spray in the pressurizer. The water level in the pressurizer is controlled by the charging flow. Transient simulations are carried out to evaluate the control system performance. When the reactor is perturbed, the reactor can be stabilized under the control system.

Design of Decoupled Mechanical Shim Control System for a Generation III+ Pressurized Water Reactor Based on Feedforward Compensation and Multimodel Approach

2018 ◽  
Author(s):  
Pengfei Wang ◽  
Xinyu Wei ◽  
Fuyu Zhao
Keyword(s):  
Control System ◽  
Power Distribution ◽  
Coupling Effect ◽  
Control Performance ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Decoupling Method ◽  
Control Rod ◽  
Wide Range ◽  
Decoupled Control

The advanced Mechanical Shim (MSHIM) core control strategy employs two separate and independent control rod banks, namely the MSHIM control banks (M-banks) and axial offset (AO) control bank (AO-bank), for automatic reactivity/temperature and axial power distribution control respectively. The M-banks and AO-bank are independently controlled by two closed-loop controllers called the coolant average temperature (Tavg) controller and AO controller. Since the movement of M-banks and AO-bank can both affect the Tavg and AO, the Tavg controller is coupled with the AO controller. In order to avoid the interference between the two controllers, the MSHIM control system adopts an interlock design between them to avoid the simultaneous movement of the M-banks and AO-bank and make sure the priority of the M-bank movement. This design can enhance the stability of the MSHIM control system. However, the control performance is degraded at the same time. In the present study, the feedforward compensation decoupling method and multimodel approach are used to eliminate the coupling effect between the two controllers in the MSHIM control system during a wide range of power maneuvers. A multiple feedforward compensation system is designed with integration of feedforward compensators for the Tavg and AO controllers at five power levels using the multimodel approach. By implementing it in the MSHIM control system, the interlock between the M-banks and AO-bank can be released to realize the independent and decoupled control between Tavg and AO. The effectiveness of the decoupled MSHIM control system is verified by comparing its control performance with that of the original MSHIM control system during typical load change transients of the AP1000 reactor. The obtained results show that superior and decoupled control of Tavg and AO can be achieved with the proposed decoupled MSHIM control system.

Development of Simulation Codes for Some Accidents in Pressurized Water Reactors

2008 ◽  
Author(s):  
X. B. Yang ◽  
G. H. Su ◽  
S. Z. Qiu
Keyword(s):  
Heat Transfer ◽  
Physical Property ◽  
Friction Model ◽  
Physical Models ◽  
Pressurized Water Reactors ◽  
Transfer Model ◽  
Feed Water ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactors

An analysis code has been developed for evaluating the transient thermo-hydraulic behaviors of the pressurized water reactor system. A series of mathematical and physical models is considered in this code, such as the point reactor neutron kinetics model, the heat transfer model, the friction model, the thermo-physical property model and so on. All possible flow and heat transfer conditions in some accidents have been considered and their corresponding models are supplied. Gear’s method is adopted for a better solution to the stiff equations. In this paper, some general accidents in the pressurized water reactors have been investigated, including the station blackout accident (SBO), the loss of flow accident (LOFA), the loss of feed water accident (LOFWA) and the reactivity insertion accident (RIA). The calculated results have been verified by the RELAP5/Mod3 and the results are satisfactory.

Silver—Indium—Cadmium, a Control Rod Material for Pressurized Water Reactors

1962 ◽  
Vol 13 (3) ◽  
pp. 299-300 ◽  
Author(s):  
H. W. Graves ◽  
J. J. Lombardo ◽  
J. S. Theilacker
Keyword(s):  
Pressurized Water Reactors ◽  
Pressurized Water ◽  
Control Rod ◽  
Water Reactors

scholarly journals Investigation Of Rod Control System Reliability Of Pwr Reactors

KnE Energy ◽  
2016 ◽  
Vol 1 (1) ◽  
Author(s):  
Syaiful Bakhri
Keyword(s):  
Control System ◽  
System Reliability ◽  
Nuclear Power ◽  
Power Plants ◽  
Thermal Power ◽  
Fault Tree ◽  
Reactor Core ◽  
Pressurized Water ◽  
Control Rod ◽  
Control Rods

<p class="NoSpacing1"><span lang="IN">The Rod Control System is </span>employed<span lang="IN"> to adjust the position of the control rods in the reactor core </span>which corresponds with <span lang="IN">the thermal power generated in the core </span>as well as <span lang="IN">the electric power generated in the turbine. In a Pressurized Water Reactor (PWR) type nuclear power plants, the control-rod drive </span>employs <span lang="IN">magnetic stepping-type mechanism. This </span>type of <span lang="IN">mechanism consists of a pair of circular coils and latch-style jack with the armature. When the </span>electric <span lang="IN">current </span>is <span lang="IN">supplied to the coils sequentially, the control-rods</span>, which <span lang="IN">are held on the drive shaft</span>, can be driven<span lang="IN"> up</span>ward<span lang="IN"> or down</span>ward<span lang="IN"> in increments. </span>This <span lang="IN">sequential current </span>c<span lang="IN">ontrol</span> drive<span lang="IN"> system is called the Control-Rod Drive Mechanism Control System (CRDMCS) or </span>known also as <span lang="IN">the Rod Control System (RCS). The p</span>urpose of this paper is to investigate the RCS reliability <span lang="IN">of APWR </span>using <span lang="IN">the Fault Tree Analysis (FTA)</span> method<span lang="IN"> since </span>the analysis of reliability which considers<span lang="IN"> the FTA</span> for common CRDM <span lang="IN">can </span>not <span lang="IN">be found</span> in <span lang="IN">any </span>public references. <span lang="IN">The FTA method is used to model the system reliability by developing the fault tree diagram of the system. </span>The<span lang="IN"> results show that the failure of the system is very dependent on the failure of most of the individual systems. However, the failure of the system does not affect the safety of the reactor, since the reactor trips immediately if the system fails. The evaluation results also indicate that the Distribution Panel is the most critical component in the system.</span></p>

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