Operating experience with automatic reactor-power control system at Obninsk Atomic Power Plant employing signals from in-core self-powered detectors

1979 ◽  
Vol 47 (4) ◽  
pp. 820-823
Author(s):  
M. G. Mitel'man ◽  
N. D. Rozenblyum ◽  
V. B. Tregubov ◽  
Yu. M. Shpiposkikh ◽  
A. I. Shtyfurko
Keyword(s):  
Control System ◽  
Power Plant ◽  
Power Control ◽  
Operating Experience ◽  
Reactor Power ◽  
Atomic Power Plant ◽  
Power Control System ◽  
Automatic Reactor ◽  
Self Powered ◽  
Core Self

Performance of automatic reactor-power control system of the first atomic power plant with correction signals derived from in-reactor transducers

1980 ◽  
Vol 49 (1) ◽  
pp. 486-488
Author(s):  
P. T. Potapenko ◽  
V. N. Sarylov ◽  
F. F. Voskresenskii ◽  
V. G. Dunaev ◽  
A. P. Shulekin
Keyword(s):  
Control System ◽  
Power Plant ◽  
Power Control ◽  
Reactor Power ◽  
Atomic Power Plant ◽  
Power Control System ◽  
Automatic Reactor

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 .

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.

scholarly journals Design of power control system using SMES and SVC for fusion power plant

2008 ◽  
Vol 97 ◽  
pp. 012271
Author(s):  
K Niiyama ◽  
T Yagai ◽  
M Tsuda ◽  
T Hamajima
Keyword(s):  
Control System ◽  
Power Plant ◽  
Power Control ◽  
Fusion Power ◽  
Fusion Power Plant ◽  
Power Control System

Power Control System Design for a Heat Pipe Cooled Reactor

2021 ◽  
Author(s):  
Haowei Sun ◽  
Peiwei Sun
Keyword(s):  
Control System ◽  
Power Control ◽  
Heat Pipe ◽  
Transfer Functions ◽  
Reactor Power ◽  
Simulation Platform ◽  
Power Control System ◽  
Dynamic Linking ◽  
Task Program ◽  
System Program

Abstract Due to the small size, the high-power density and the strong environmental adaptability, the heat pipe cooled reactor has a broad application prospect in the fields of space exploration, underwater vehicle power system and other fields[1]. It will be one of the subversive technologies which will change the future nuclear power development. The safe operation of the heat pipe cooled reactor requires an effective power control system. 3KeyMaster is a comprehensive real-time simulation analysis platform and the control system can be built by using its logic module. The interface program can be developed in its task program, and other program can be integrated in this platform[2]. The system program of the heat pipe cooled reactor has compiled as a dynamic linking library. Through the task program of the 3KeyMaster simulation platform to call the dynamic linking library by using C++ program, the system program is coupled with the 3KeyMaster simulation platform. The modules of 3KeyMaster platform exchange data with the system program by shared memory. Through the responses of the temperature due to step change in the reactor power obtained by 3KeyMaster, the transfer functions are obtained by using the system identification module in MATLAB & Simulink. The power disturbance is introduced, and simulation is carried out on the 3KeyMaster. The changes of the important parameters are analyzed to obtain its dynamic characteristics. Based on the dynamic analysis on the system responses, the control strategy for the power control system is proposed. The power control system is implemented in MATLAB & Simulink. The control system parameters are tuned to determine a group of suitable control parameters[3]. Finally, the control performance of power control system is evaluated by simulations.

scholarly journals Design and Verification of Reactor Power Control Based on Stepped Dynamic Matrix Controller

2019 ◽  
Vol 2019 ◽  
pp. 1-11
Author(s):  
Xu Zhang ◽  
Zhiguang Deng ◽  
Jun Li ◽  
Youwei Yang ◽  
Quan Ma ◽  
...  
Keyword(s):  
Control System ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Power Control ◽  
Nuclear Power ◽  
Reactor Power ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Dynamic Matrix ◽  
Adjustment Time

As key equipment in nuclear power plant, the reactor power control system is adopted to strictly control and regulate the reactor power of a PWR (pressurized water reactor) in a nuclear power plant. A well-optimized predictive control algorithm based on SDMC (stepped dynamic matrix controller) is developed and introduced in this paper and applied to the power regulation of a reactor power model. In addition, the test and verification of this application is conducted by two different methods and devices: the virtual verification platform and the physical DCS (digital control system). The result of the verification suggests that the application of SDMC gains a better performance in the maximum dynamic deviation, adjustment time, overshoot, and so on.

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