Reduced optimal controller design for the nuclear reactor power control system

2011 ◽  
pp. 84-86
Author(s):  
H. Zarabadipour ◽  
H. Emadi
Keyword(s):  
Control System ◽  
Power Control ◽  
Nuclear Reactor ◽  
Reactor Power ◽  
Reduced Order Model ◽  
Optimal Controller ◽  
Balance Model ◽  
Order Model ◽  
Reduced Order ◽  
Power Control System

The power control system of a nuclear reactor is one of the key systems that concern the safe operation of the plant. Much attention is paid to the power control systems’ performance of nuclear reactor in engineering. The goal of this paper is apply balance model reductionto derive reduced order model and then design the reduced optimal controller for nuclear reactor power system.The simulation results with reduced-order model and with optimized controller show that the proposed technique is improved .

scholarly journals Reduced Order Model of Position Control System

2011 ◽  
pp. 113-115
Author(s):  
Yogesh V. Hote ◽  
A. N. Jha ◽  
J. R. P. Gupta
Keyword(s):  
Control System ◽  
Position Control ◽  
Open Loop ◽  
Simple Approach ◽  
Reduced Order Model ◽  
Order Model ◽  
Reduced Order ◽  
Reduced Order Modelling ◽  
Position Control System

In this paper, simple approach is proposed to determine reduced order model of a unstable open-loop position control system. This approach is based on Krishnamurthy’s approach on Routh criterion on reduced order modelling. The results are simulated in Matlab environment.

A POD reduced-order model for resolving the neutron transport problems of nuclear reactor

2020 ◽  
Vol 149 ◽  
pp. 107799
Author(s):  
Yue Sun ◽  
Junhe Yang ◽  
Yahui Wang ◽  
Zhuo Li ◽  
Yu Ma
Keyword(s):  
Nuclear Reactor ◽  
Neutron Transport ◽  
Reduced Order Model ◽  
Order Model ◽  
Reduced Order ◽  
Transport Problems

Reactor Kinetics in a Loss-of-Forced-Cooling (LOFC) Test of HTGRs

2012 ◽  
Author(s):  
Kuniyoshi Takamatsu ◽  
Kazuhiro Sawa
Keyword(s):  
Control System ◽  
High Temperature ◽  
Power Control ◽  
Reactor Power ◽  
Reactor Kinetics ◽  
The Core ◽  
Power Control System ◽  
Stop Test ◽  
High Temperature Reactor ◽  
Forced Cooling

The High-Temperature Engineering Test Reactor (HTTR) is the first High-Temperature Gas-cooled Reactor (HTGR) with a thermal power of 30 MW and a maximum reactor outlet coolant temperature of 950 °C; it was built at the Oarai Research and Development Center of JAEA. At present, test studies are being conducted using the HTTR to improve HTGR technologies in collaboration with domestic industries that also contribute to foreign projects for the acceleration of HTGR development worldwide. To improve HTGR technologies, advanced analysis techniques are currently under development using data obtained with the HTTR, which include reactor kinetics, thermal hydraulics, safety evaluation, and fuel performance evaluation data (including the behavior of fission products). In this study, a three gas circulator trip test and a vessel cooling system (VCS) stop test were performed as a loss of forced cooling (LOFC) test to demonstrate the inherent safety features of HTGR. The VCS stop test involved stopping the VCS located outside the reactor pressure vessel to remove the residual heat of the reactor core as soon as the three gas circulators are tripped. All three gas circulators were tripped at 9, 24 and 30 MW. The primary coolant flow rate was reduced from the rated 45 t/h to 0 t/h. Control rods (CRs) were not inserted into the core and the reactor power control system was not operational. In fact, the three gas circulator tripping test at 9 MW has already been performed in a previous study. However, the results cannot be disclosed to the public because of a confidentiality agreement. Therefore, we cannot refer to the difference between the analytical and test results. We determined that the reactor power immediately decreases to the decay heat level owing to the negative reactivity feedback effect of the core, although the reactor shutdown system was not operational. Moreover, the temperature distribution in the core changes slowly because of the high heat capacity due to the large amount of core graphite. Core dynamics analysis of the LOFC test for the HTTR was performed. The relationship among the reactivities (namely, Doppler, moderator temperature, and xenon reactivities) affecting recriticality time and reactor peak power level as well as total reactivity was addressed. Furthermore, the analytical results for a reactor transient of hundred hours are presented. Based on the results, emergency operating procedures can be developed for the case of a loss of coolant accident in HTGR when the CRs are not inserted into the core and the reactor power control system is not operational. The analytical results will be used in the design and construction of the Kazakhstan High-Temperature Reactor and the realization of commercial Very High-Temperature Reactor systems.

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