Development of a PID-Fuzzy controller in the water level control of a pressurizer of a nuclear reactor

It is well known that safety in the operation of nuclear power plants is a primary requirement because a failure of this system can result in serious problems to the environment. A nuclear reactor has several systems that help keep it in normal operation, within safety margins. Many of these systems operate in the control of variable quantities in the primary circuit of a reactor. However, nuclear reactors are nonlinear physical systems, and this introduces a complexity in the control strategies. Among several mechanisms in the thermal-hydraulic system of a reactor that actuate as a controller, the pressurizer is the component responsible for absorbing pressure variations that occur in the primary circuit. This work aims at the development of a PID controller (Proportional Integral Derivative) based on fuzzy logic to operate in a pressurizer of a nuclear Pressurized Water Reactor. A Fuzzy Controller was developed using the process of fuzzification, inference, and defuzzification of the variables of interest to a pressurizer, then this controller was coupled to a PID Controller building a PID Controller, but oriented by Fuzzy logic. Subsequently, the PID-Fuzzy Controller was experimentally validated in a Simulation Plant in which transients like those in a PWR were conducted. The PID parameters were analyzed and adjusted for better responses and results. The results of the validation were also compared to simple controllers (on / off).

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Component and Integration Test of an FPGA-Based PWR Protection Sub-System Using UVM

Volume 1: Operations and Maintenance, Engineering, Modifications, Life Extension, Life Cycle and Balance of Plant; I&C, Digital Controls, and Influence of Human Factors ◽

10.1115/icone25-66526 ◽

2017 ◽

Author(s):

Xuanxuan Shui ◽

Yichun Wu ◽

Junyi Zhou ◽

Yuanfeng Cai

Keyword(s):

Nuclear Power ◽

Power Industry ◽

Nuclear Power Industry ◽

Normal Operation ◽

Static Timing Analysis ◽

Specific Test ◽

Pressurized Water ◽

Safety Margins ◽

Integration Test ◽

Test Process

Field programmable gate arrays (FPGAs) have drawn wide attention from nuclear power industry for digital instrument and control applications (DI&C), because it’s much easier and simpler than microprocessor-based applications, which makes it more reliable. FPGAs can also enhance safety margins of the plant with potential possibility for power upgrading at normal operation. For these reasons, more and more nuclear power corporations and research institutes are treating FPGA-based protection system as a technical alternative. As nuclear power industry requires high reliability and safety for DI&C Systems, the development method and process should be fully verified and validated. For this reason, to improve the application of FPGA in NPP I&C system, the specific test methods are critical for the developers and regulators. However, current international standards and research reports, like IEC 62566 and NUREG/CR-7006, which have demonstrated the life circle of the development of FPGA-based safety critical DI&C in NPPs, but the specific test requirements and methods which are significant to the developers are not provided. In this paper, the whole test process of a pressurized water reactor (PWR) protection sub-system (Primary Coolant Flow Low Protection, Over Temperature Delta T Protection, Over Power Delta T Protection) is described, including detail component and integration tests. The Universal Verification Methodology (UVM) based on System Verilog class libraries is applied to establish the verification test platform. All these tests are conducted in a simulation environment. The test process is driven by the test coverage which includes code coverages (i.e., Statement, Branch, Condition and Expression, Toggle, Finite State Machine) and function coverage. Specifically, Register Transaction Level (RTL) simulation is conducted for Component tests, while RTL simulation, Gate Level simulation, Timing simulation and Static timing analysis are conducted for the integration test. The issues (e.g., the floating point calculation, FPGA resource allocation and optimization) arose in the test process are also analyzed and discussed, which can be references for the developers in this area. The component and integration tests are part of the Verification and Validation (V&V) work, which should be done by the V&V team separated from the development team. The testing method could assure the test results reliable and authentic. It is practical and useful for the development and V&V of FPGA-based safety DI&C systems.

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A hybrid PSO-PID approach for trajectory tracking application of a liquid level control process

An International Journal of Optimization and Control Theories & Applications (IJOCTA) ◽

10.11121/ijocta.01.2015.00244 ◽

2015 ◽

Vol 5 (2) ◽

pp. 63-73 ◽

Cited By ~ 1

Author(s):

Turker Tekin Erguzel

Keyword(s):

Water Level ◽

Pid Controller ◽

Nuclear Power ◽

Power Plants ◽

Control Process ◽

Nonlinear Process ◽

Reference Trajectory ◽

Proportional Integral Derivative ◽

Level Control ◽

Steam Generators

Water level control is a crucial step for steam generators (SG) which are widely used to control the temperature of nuclear power plants. The control process is therefore a challenging task to improve the performance of water level control system. The performance assessment is another consideration to underline. In this paper, in order to get better control of water level, the nonlinear process was first expressed in terms of a transfer function (TF), a proportional-integral-derivative (PID) controller was then attached to the model. The parameters of the PID controller was finally optimized using particle swarm optimization (PSO). Simulation results indicate that the proposed approach can make an effective tracking of a given level set or reference trajectory.

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Thermal Hydraulic Analysis of a Nuclear Reactor due to Loss of Coolant Accident with and without Emergency Core Cooling System

10.38032/jea.2020.02.004 ◽

2020 ◽

Vol 01 (02) ◽

pp. 53-60

Author(s):

Pronob Deb Nath ◽

Kazi Mostafijur Rahman ◽

Md. Abdullah Al Bari

Keyword(s):

Nuclear Reactor ◽

Cooling System ◽

Computational Study ◽

Reactor Core ◽

Primary Circuit ◽

Feed Water ◽

Pressurized Water ◽

Loss Of Coolant Accident ◽

Loss Of Coolant ◽

Core Cooling

This paper evaluates the thermal hydraulic behavior of a pressurized water reactor (PWR) when subjected to the event of Loss of Coolant Accident (LOCA) in any channel surrounding the core. The accidental break in a nuclear reactor may occur to circulation pipe in the main coolant system in a form of small fracture or equivalent double-ended rupture of largest pipe connected to primary circuit line resulting potential threat to other systems, causing pressure difference between internal parts, unwanted core shut down, explosion and radioactivity release into environment. In this computational study, LOCA for generation III+ VVER-1200 reactor has been carried out for arbitrary break at cold leg section with and without Emergency Core Cooling System (ECCS). PCTRAN, a thermal hydraulic model-based software developed using real data and computational approach incorporating reactor physics and control system was employed in this study. The software enables to test the consequences related to reactor core operations by monitoring different operating variables in the system control bar. Two types of analysis were performed -500% area break at cold leg pipe due to small break LOCA caused by malfunction of the system with and without availability of ECCS. Thermal hydraulic parameters like, coolant dynamics, heat transfer, reactor pressure, critical heat flux, temperature distribution in different sections of reactor core have also been investigated in the simulation. The flow in the reactor cooling system, steam generators steam with feed-water flow, coolant steam flow through leak level of water in different section, power distribution in core and turbine were plotted to analyze their behavior during the operations. The simulation showed that, LOCA with unavailability of Emergency Core Cooling System (ECCS) resulted in core meltdown and release of radioactivity after a specific time.

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The application of Deep Reinforcement Learning in Coordinated Control of Nuclear Reactors

Journal of Physics Conference Series ◽

10.1088/1742-6596/2113/1/012030 ◽

2021 ◽

Vol 2113 (1) ◽

pp. 012030

Author(s):

Jing Li ◽

Yanyang Liu ◽

Xianguo Qing ◽

Kai Xiao ◽

Ying Zhang ◽

...

Keyword(s):

Control System ◽

Reinforcement Learning ◽

Nuclear Power ◽

Power Plants ◽

Nuclear Reactor ◽

Coordinated Control ◽

Level Control ◽

Simulation Experiments ◽

Reinforcement Learning Model ◽

Reactor Control System

Abstract The nuclear reactor control system plays a crucial role in the operation of nuclear power plants. The coordinated control of power control and steam generator level control has become one of the most important control problems in these systems. In this paper, we propose a mathematical model of the coordinated control system, and then transform it into a reinforcement learning model and develop a deep reinforcement learning control algorithm so-called DDPG algorithm to solve the problem. Through simulation experiments, our proposed algorithm has shown an extremely remarkable control performance.

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Performances evaluation and comparison of PID controller and fuzzy logic controller for process liquid level control

2015 15th International Conference on Control, Automation and Systems (ICCAS) ◽

10.1109/iccas.2015.7364848 ◽

2015 ◽

Cited By ~ 6

Author(s):

Deepa Shivshant Bhandare ◽

N. R. Kulkarni

Keyword(s):

Fuzzy Logic ◽

Pid Controller ◽

Fuzzy Logic Controller ◽

Liquid Level ◽

Level Control ◽

Performances Evaluation ◽

Liquid Level Control

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Radionuclide Composition of Airborne Discharges of European NPPS With WWER, Pressurized Water Reactor, and Boiling Water Reactor Facilities

Journal of Nuclear Engineering and Radiation Science ◽

10.1115/1.4047044 ◽

2020 ◽

Vol 7 (1) ◽

Author(s):

Angelina-Nataliya V. Vukolova ◽

Andrei A. Rusinkevich

Keyword(s):

Power Plants ◽

Boiling Water ◽

Normal Operation ◽

Boiling Water Reactor ◽

Pressurized Water Reactor ◽

Pressurized Water ◽

Water Reactor ◽

Operation Conditions ◽

Discharge Activity ◽

Radionuclide Composition

Abstract The article presents the analysis of the data on radionuclide composition of airborne discharges of 52 European nuclear power plants (NPPs) with water–water energetic reactor facilities (WWER), pressurized water reactor facilities (PWR), and boiling water reactor facilities (BWR) under normal operation conditions. It contains lists of radionuclides, registered in discharges of researched NPPs, and gives estimation of contributions of radionuclides, forming the discharge, into total activity of discharge and into total effective dose, created by the discharge activity. It was determined that the maximal contribution into discharge activity of all researched NPPs make noble gases, tritium, and carbon-14, while the latter is the main dose-making radionuclide.

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Comparison of state feedback and PID control of pressurizer water level in nuclear power plant

Archives of Control Sciences ◽

10.2478/acsc-2013-0028 ◽

2013 ◽

Vol 23 (4) ◽

pp. 455-471 ◽

Cited By ~ 1

Author(s):

Mariusz Czapliński ◽

Paweł Sokólski ◽

Kazimierz Duzinkiewicz ◽

Robert Piotrowski ◽

Tomasz Rutkowski

Keyword(s):

Power Plant ◽

Nuclear Power Plant ◽

Water Level ◽

Nuclear Power ◽

State Feedback ◽

Reference Value ◽

Primary Circuit ◽

Pipeline System ◽

Level Control ◽

Pressurized Water

Abstract The pressurizer water level control system in nuclear power plant with pressurized water reactor (PWR) is responsible for coolant mass balance. The main control goal is to stabilize the water level at a reference value and to suppress the effect of time-varying disturbances (e.g. coolant leakage in primary circuit pipeline system). In the process of PWR power plant operation incorrect water level may disturb pressure control or may cause damage to electric heaters which could threaten plant security and stability. In modern reactors standard PID controllers are used to control water level in a pressurizer. This paper describes the performance of state feedback integral controller (SFIC) with reduced-order Luenberger state observer designed for water level control in a pressurizer and compares it to the standard PID controller. All steps from modeling of a pressurizer through control design to implementation and simulation testing in Matlab/Simulink environment are detailed in the paper.

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ANALYTICAL SIMULATION OF A PWR USING MELCOR

Brazilian Journal of Radiation Sciences ◽

10.15392/bjrs.v8i3a.1357 ◽

2021 ◽

Vol 8 (3A) ◽

Author(s):

Maritza Rodríguez Gual ◽

Nathalia N. Araújo ◽

Marcos C. Maturana

Keyword(s):

Nuclear Power ◽

Power Plants ◽

Cooling System ◽

Nuclear Regulatory Commission ◽

Severe Accident ◽

Normal Operation ◽

Detailed Knowledge ◽

Pressurized Water ◽

Severe Accidents ◽

Reference Plant

After the two most significant nuclear accidents in history – the Chernobyl Reactor Four explosion in Ukraine(1986) and the Fukushima Daiichi accident in Japan (2011) –, the Final Safety Analysis Report (FSAR) included a new chapter (19) dedicated to the Probabilistic Safety Assessment (PSA) and Severe Accident Analysis (SAA), covering accidents with core melting. FSAR is the most important document for licensing of siting, construction, commissioning and operation of a nuclear power plant. In the USA, the elaboration of the FSAR chapter 19 is according to the review and acceptance criteria described in the NUREG-0800 and U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.200. The same approach is being adopted in Brazil by National Nuclear Energy Commission (CNEN). Therefore, the FSAR elaboration requires a detailed knowledge of severe accident phenomena and an analysis of the design vulnerabilities to the severe accidents, as provided in a PSA – e.g., the identification of the initiating events involving significant Core Damage Frequency (CDF) are made in the PSA Level 1. As part of the design and certification activities of a plant of reference, the Laboratory of Risk Analysis, Evaluating and Management (LabRisco), located in the University of São Paulo (USP), Brazil, has been preparing a group of specialists to model the progression of severe accidents in Pressurized Water Reactors (PWR), to support the CNEN regulatory expectation – since Brazilian Nuclear Power Plants (NPP), i.e., Angra 1, 2 and 3, have PWR type, the efforts of the CNEN are concentrated on accidents at this type of reactor. The initial investigation objectives were on completing the detailed input data for a PWR cooling system model using the U.S. NRC MELCOR 2.2 code, and on the study of the reference plant equipment behavior – by comparing this model results and the reference plant normal operation main parameters, as modeled with RELAP5/MOD2 code.

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