Pressurized Water Reactor Nuclear Power Plant Research Articles

Active disturbance rejection control of pressurizer pressure during startup process in a pressurized water reactor nuclear power plant

To cope with the participation of power plant peak shaving and realize the diversification of nuclear energy applications including cogeneration, desalination, etc. The application of automatic procedure startup and shutdown system (APS) is an important research direction to improve the safety, economy and flexibility of pressurized water reactor (PWR) nuclear power plant (NPP). But it is difficult to realize automatic control of pressurizer pressure in the PWR NPP startup condition with drastic changes and disturbances. To solve this problem, an active disturbance rejection control (ADRC) algorithm was proposed for the pressure control of the pressurizer. The controller parameters were tuned based on a two-step tuning scheme. Its control performance and disturbance rejection ability were studied and compared with proportional-integral-derivative (PID) controller under different operation conditions. The simulation results showed that the designed controller with fixed parameters can outperform the PID controller with specific optimized parameters on setpoint tracking and disturbance rejection. This study is of great significance for realizing the automatic control of pressurizer pressure in the startup process, and can provide reference for the design of the automatic control system during PWR startup. The controller was applied in typical startup process and achieved satisfactory control performance.

The Effects of Cl− and Selected Deoxidizers on the High-Temperature Corrosion Electrochemistry of Alloy 690 in Nuclear Steam Generator Water

The heat transfer tube in a steam generator serves as a critical heat exchange component in the primary and secondary loops of pressurized water reactor (PWR) nuclear power plants. The corrosion resistance of the heat transfer tube material directly influences the longevity of PWR nuclear power plants. This study investigated the electrochemical corrosion properties of 690 alloy (UNS N06690) in a simulated secondary water environment of PWR, focusing on different chloride ion concentrations and combinations of deoxidizers. The findings reveal a gradual decrease in the corrosion potential of 690 alloy, accompanied by an increase in self-corrosion current and a progressive reduction in the passivation range, ultimately leading to its disappearance as chloride ion concentration rises from 0 µg/L to 500 µg/L. Moreover, the impedance value of the inner film exhibits a declining trend with an increase in chloride ion concentration. Conversely, the resistance value of the outer film remains relatively stable while the size and spacing of oxide particles on the surface of the 690 alloy continuously increase. This observation suggests that chloride ions primarily influence the formation of the inner passivation film, which in turn determines the corrosion resistance of the 690 alloy. Notably, the performance of the 690 alloy is similar when the deoxidizer combination is ammonia (NH3) + erythorbic acid (ERA) or NH3 + hydrazine (N2H4), demonstrating the ability to form a relatively complete passivation film and exhibit improved corrosion resistance compared to NH3+N-isopropyl hydroxylamine, additionally, when the deoxidizer combination is NH3+N2H4, the 690 alloy exhibits lower self-corrosion current density across different chloride ion concentrations, indicating enhanced corrosion resistance.

Discovering the causes for the change of the vibration characteristics of the core support barrel in PWR nuclear power plants: A combined investigation based on ex-core neutron noise analysis and numerical modal analysis

As a key component of pressurized water reactor (PWR) nuclear power plants, the motion of the core support barrel (CSB) has a great impact on the safety of the nuclear reactor. However, the vibration characteristics of the CSB have not yet been fully investigated and may change due to the high doses of radiation and high temperature in the reactor. In this paper, the ex-core neutron noise signals of two PWR units throughout the multiple fuel cycles were analyzed and the vibration frequency of the CSB beam mode was found, for the first time, to drop with time during the operation of PWRs. Then, the statics analysis of the CSB was conducted to preliminarily identify the potential causes. For the further clarification of such vibration frequency drop phenomenon, a numerical assembly model of PWR reactor internals, including the CSB, hold-down ring (HDR), fuel assembly hold-down spring (FAHDS), etc., was proposed with consideration of the frictional contact and fluid-structure interaction for the modal analysis of the CSB. The results showed that the decrease of the FAHDS stiffness is the main cause for the drop of the vibration frequency of the CSB beam mode in real conditions.

Fractional order modelling and robust multi-model intelligent controllers’ synthesis for ACP1000 nuclear power plant

In this research work, a third generation Pressurized Water Reactor (PWR) type Nuclear Power Plant (NPP) of 1100 MWe rating called Advanced Chinese PWR (ACP1000) NPP is adopted. The research and development work is mainly encompassed of addition of third steam generator loop, hybrid programming platforms, fractional order modular modelling and intelligent controllers. The ACP1000 NPP is comprised of primary loop, secondary loop and balance of plant (BOP). The ACP1000 nonlinear dynamics is precisely modelled with greater efficiency using fractional order differential equations. The entire plant is modelled into eighteen modules in a fractional order framework in Visual Basic (VB) environment. The eighteen modules are governed by nine controllers. Controllers are configured in a fashion of an adaptive neuro-fuzzy inference system (ANFIS). An innovative multi input single output (MISO) ANFIS virtual instrument (VI) is designed for training and configuring controllers in LabVIEW. The developed MISO ANFIS VI in combination is implemented for multi-input multi output (MIMO) ANFIS controllers in parallel computing framework. The variable transfer between fractional order multi- modular framework in VB (FO- MMF-VB) and robust intelligent ANFIS controllers in LabVIEW (RI-ANFIS-LV) is established in client server configuration and system exec VI for calling external code in VB. This hybrid programming framework of Visual Basic and LabVIEW is actually a development of LabVIEW Visual Basic Integration (LabVb) Toolkit used to evaluate the closed loop transient performance of ACP1000 nuclear power plant. Various parameters are simulated for dynamic transient simulations under load manoeuvring from 100% power to 50% power. The results are analysed and validated against benchmark design data and Preliminary Safety Analysis Report (PSAR). All the results are within the protection systems design limits under normal operating conditions.

Experimental study on the influence of drainage tank structure on separation performance of vertical corrugated plate dryer in PWR

Corrugated plate dryer (CPD) is an important steam-water separation (SWS) devices in the steam generator of Pressurized Water Reactor (PWR) nuclear power plants (NPPs), and the drainage tank (DT) in CPD determines the separation efficiency (SE). This paper designs an experimental bench to study SWS performance of a full-scale CPD commonly used in PWR NPPs. The influence of structure of the DT (including the changes of the height of DT and the number of baffles in the DT) at the lower part of CPD affects the SE, critical airflow velocity (CAV), and pressure drop is studied from the perspective of mechanism. It can be concluded that the increase in the number of baffles inside the DT enhances the CAV and SE of CPD. Meanwhile, increasing the number of baffles has a more pronounced effect on boosting the CAV than that of the height of DT within a specific range, which provides an important reference for the miniaturization design of NPPs.

Optimal Tracking Control for a Discrete Time Nonlinear Nuclear Power System

Recently, increasing attention has been paid to nuclear power control with the appeals of clean energy and demands of power regulation to integrate into the power grid. However, a nuclear power system is a discrete-time (DT) nonlinear and complicated system, where the parameters entangle with intrinsic states. Furthermore, the need for huge computational ability due to the high-level order property in the nuclear reactor model causes many difficulties in the power control of nuclear industries. In this study, a new scheme of optimal tracking control for DT nonlinear nuclear power systems is provided to accomplish the power control of a 2500-MW pressurized water reactor (PWR) nuclear power plant. The proposed approach based on the value iteration method is a novel algorithm in the human intelligence community, which has a basic actor-critic structure with neural networks (NNs). The new approach has some modifications, where the cost function is redefined by leveraging the higher-order polynomial to substitute neural networks in the entire actor critic architecture. Simulation results of the 2500-MW PWR nuclear power plant are given to demonstrate the effectiveness of the developed method.

Advanced Multi-Modeling of PWR Dynamics and Deep Learning based Computational Tool in SIMULINK and LabVIEW

The reactivity monitoring, prediction, and investigation is the most important parameter to ensure the safety and reliable operation of a nuclear power plant. This parameter is gained further importance in Pressurized Water Reactor (PWR) due to more sophisticated reactivity insertion mechanisms and innovative reactor core fuel loading scheme. Based on deterministic internal and external dynamics and neutronics analysis of Advanced PWR, all the reactivity feedback effects such as Doppler effect, moderator effect, control rod effect, liquid boron effect and reactor poisons effect are investigated, modeled and stochastically optimized using deep artificial intelligence. Advance Pressurized Water Reactor (APWR) of 600 MWe rating (AP-600) is used as a reference reactor model and based on the dynamics of AP-600, an Advanced Pressurized Water Reactor Dynamics and Intelligent Stochastic Optimization based Deterministic Neutronics Simulation (APD-ISO-DNS) Code is developed in the hybrid SIMULINK andLabVIEW environments. AP-600 reactor model is fine-tuned and adjusted for 300 MWe PWR (P-300) and 1070 MWe Advanced Chinese PWR (ACP-1000) using neutronics parameters and operational dynamic data of operating PWR nuclear power plants in Pakistan. Various load reduction transient experiments are conducted and dynamic transient simulations of P-300, AP-600 and ACP-1000 are evaluated in SIMULINK and in LabVIEW environments and found as per design basis.

Some fundamental understandings of Zn-injection water chemistry on material corrosion in pressurized water reactor primary circuit

The optimization of water chemistry in pressurized water reactor (PWR) primary circuit is one of the most effective ways to achieve both safety and economy for operating PWR nuclear power plants (NPPs). Special attention has been paid to fundamental research and engineering application of Zn-injection water chemistry (ZWC) in PWR primary circuit in recent years in China due to the rapid development of nuclear power. The present paper mainly reports the status of PWR NPPs in China and some fundamental understandings of ZWC on material corrosion. Effects of temperature (T), pH at T value, Zn concentration, Zn-injection sequence and lasting time on electrochemical corrosion behavior, oxide film characteristic and stress corrosion cracking (SCC) susceptibility of nuclear-grade austenitic alloys were investigated. A modified point defect model was used to discuss the effects of ZWC on the oxide films in high-temperature pressurized water. Some water chemistry parameters in PWR primary coolant system are proposed for mitigating corrosion and SCC of nuclear-grade austenitic alloys.

Simulation on influences of Al2O3 nanofluids as coolant for nuclear power plant based on modified empirical formula of nanofluids thermal properties at high temperature and high pressure

ABSTRACT The thermal properties of nanofluids are variable with working temperature and pressure and ignored by the existing empirical formula. In the present work, based on the application of nanofluids in pressurized water reactor (PWR) nuclear power plants, different concentrations of Al2O3/H2O nanofluids are prepared for measuring its thermal properties and then modifying empirical formulas at high temperature and high pressure (HTHP). In addition, the primary heat transfer process of PWR nuclear power plants is simulated by numerical method. The results show that, when volume fraction φ within 2.0%, the heat transfer performance of Al2O3 nanofluid increases with the increase of concentrations; however, it is on the contrary when the volume fraction of nanoparticles is larger than 2%. Comparing with water as the coolant, Al2O3 nanofluids with 2% volume fraction give the largest heat transfer enhancement of 103%. Therefore, Al2O3 nanofluids as the coolant can achieve the goals of heat transfer enhancement and safety limit improvement.

Study on the relationship between Fe3O4 fouling and NiFe2O4 oxide layer in the secondary circuit of nuclear steam generator

Fouling is one of the most significant problem in the steam generator (SG) of pressurized water reactor (PWR) nuclear power plant (NPP). To figure out the deposition process, through changing the Fe2+concentration and temperature in the simulated configurations (523 K–603 K), adsorption mechanism of fouling Fe3O4 on steam generator (SG) tube in secondary circuit of pressurized water reactor (PWR) nuclear power plant was investigated by molecular dynamic simulation. The ion concentration distribution, the mean square displacement (MSD), the radial distribution function (RDF), and the energy change before and after adsorption were compared. The influence of the change of temperature and Fe2+ concentration on the corrosion of the SG tube was discussed. Due to the negative surface charge of NiFe2O4, the Fe2+ in the solution did not bond with OH−, but agglomerated with high-pressure high-temperature H2O in the form of triangular bipyramid and octahedron. The diffusion coefficient of OH− increases linearly with the increase of temperature (from 15.28 × 10−5 cm2/s to 18.59 × 10−5 cm2/s) and concentration (from 14.40 × 10−5 cm2/s to 16.06 × 10−5 cm2/s). While the diffusion coefficient of Fe2+reduces, the diffusion coefficient of H2O changes. Simultaneously, Fe3O4 fouling formed on the surface of NiFe2O4 dense oxide layer due to the ion adsorption and interface charges. This work can provide an atomic-scale basis for fouling inhibition in the SG of PWR nuclear power plant.

Study on properties of 18MND5 steel forgings for PWR steam generator

In the pressurized water reactor (PWR) nuclear power plant, the shell material of steam generator is required to have good strength-toughness matching, anti-fatigue performance, and neutron radiation resistance to ensure the long-term safe and reliable service. The manufacturing requirements of Manganese-Nickel-Molybdenum alloy steel forgings of steam generator channel head for Hua-long Pressurized Reactor (HPR1000) and Advanced Passive PWR(AP1000) nuclear power plant were compared. The heat treatment process, chemical composition, mechanical properties, metallographic structure of 18MND5 steel forging and SA 508-III steel forging were analysed and studied. The results show that the performance heat treatment temperature of HPR1000 forging has more stringent controls and with lower element content of C, Mo, P, Si, V, Co than the AP1000 forging. By reducing the C content, HPR1000 forging got better toughness while the strength was ensured.

Unmitigated severe accident analysis for a PWR using MELCOR

After the two major nuclear accidents in history considered to be beyond-design-basis accidents (BDBA) – the Chernobyl Reactor Four explosion in Ukraine (1986) and the Fukushima Daiichi accident in Japan (2011), the United States Nuclear Regulatory Commission and other countries have included in the Final Safety Analysis Report (FSAR), a new chapter (19) dedicated to the Probabilistic Safety Assessment (PSA) and Severe Accident Analysis (SAA), with significant damage to the reactor core. The Safety Analysis Report (SAR) is the most important document used by a regulatory body to assess the adequacy of plant safety at all stages of the life time of a Nuclear Power Plant (NPP), and presents the licensing basis of a plant.This study aims at comparing the progression of severe accidents during a Total Loss of Feed Water Accident (TLOFW) and a TLOFW combined with An Anticipated Transient Without SCRAM (ATWS) reactivity accident for a Pressurized Water Reactor (PWR) type NPP. This paper takes into account the results of the PSA Level 1 for the reference NPP. The calculations were performed with the deterministic computer code MELCOR version 2.2 used to analyze severe accident progression in NPP with a Light Water Reactor (LWR). The paper presents a hydrogen risk assessment – the assessment is deterministic, s a prerequisite before performing PSA Level 2. The hydrogen source is determined, too. Saphiro's diagram is applied to investigate the containment atmosphere condition (inert or combustible) for two scenarios. The main results of the study reveal that the containment integrity of the reference PWR NPP when both accidents occurred is maintained. However, the core is severely damaged. A future work will investigate mitigation strategies to limit the hydrogen risk during accident.

A Global Review of PWR Nuclear Power Plants

Nuclear energy is presented as a real option in the face of the current problem of climate change and the need to reduce CO2 emissions. The nuclear reactor design with the greatest global impact throughout history and which has the most ambitious development plans is the Pressurized Water Reactor (PWR). Thus, a global review of such a reactor design is presented in this paper, utilizing the analysis of (i) technical aspects of the different variants of the PWR design implemented over the past eight years, (ii) the level of implementation of PWR nuclear power plants in the world, and (iii) a life extension scenario and future trends in PWR design based on current research and development (R&D) activity. To develop the second analysis, a statistical study of the implementation of the different PWR variants has been carried out. Such a statistical analysis is based on the operating factor, which represents the relative frequency of reactors operating around the world. The results reflect the hegemony of the western variants in the 300 reactors currently operating, highlighting the North American and French versions. Furthermore, a simulation of a possible scenario of increasing the useful life of operational PWRs up to 60 years has been proposed, seeing that in 2050 the generation capacity of nuclear PWRs power plants will decrease by 50%, and the number of operating reactors by 70%.

Detailed and verified NPP model for safe interconnection with electrical grid

A precise understanding for nuclear power plant (NPP) response to grid disturbances is required to ensure safe and reliable operation for both NPP and power grid. NPP performance is studied by dynamic modeling and simulation, since it is not practical to study it by experiments. In this work, a mathematical model for complete pressurized water reactor NPP is represented based on MATLAB/Simulink. First, the main parts of the model (reactor, steam generator, pressurizer, and steam turbine) are studied and verified individually. Then, those parts are connected together in a highly controlled system to test the overall plant behavior under load changes. The results of simulation indicate that the plant is capable to regulate its output according to electrical grid demand and reach the steady-state conditions. Thus, a high degree of safe interface between the electric grid and NPP is achieved.

TRACE simulations of IIST cooldown experiments

Abstract IIST (Institute of Nuclear Energy Research Integral System Test) facility is a RHRP (Reduced-Height and Reduced-Pressure) facility for the three-loop PWR (Pressurized Water Reactor) NPP (Nuclear Power Plant) Maanshan. This test facility was established in 1992 for safety studies of the Maanshan NPP. TRACE is an advanced thermal hydraulic code which is still developed and updated by U.S. NRC. So, validation of specific models against experiments is still necessary. In this paper, cooldown experiments in IIST facility are used to validate the TRACE models. Additionally, the results of TRACE calculations are compared to results of RELAP5 calculations. This comparison shows that the calculations of TRACE are in good agreement to the data of IIST experiments and to the calculational results of RELAP5. So, TRACE can be used to simulate cooldown transients.

Deep rectifier neural network applied to the accident identification problem in a PWR nuclear power plant

A Nuclear Power Plant (NPP) is an advanced and complex power generation system that requires a high level of safety to operate. So, when confronted with an anomalous event, such as an accident or a transient scenario, if the event is not correctly identified quickly enough, it can spread rapidly in the system leading the plant to a risky situation. Because of this, various statistical methods and machine learning techniques have been employed aiming at developing a support system able to identify the occurring event in a short amount of time. This article shows the performance of the Deep Rectifier Neural Network (DRNN) in the nuclear accident identification problem. For this, a 5-layer DRNN with rectifier linear unit (ReLu) activation function in neurons of the hidden layers was developed, and 12 postulated accidents plus the normal operation condition were used for a Brazilian pressurized water reactor (PWR). The 5-Layer DRNN demonstrated good generalization capacity compared with other techniques at the literature, achieving a precision of 99.03% for the validation dataset.

Simulation study of frequency control characteristics of a generation III + nuclear power plant

The AP1000, as an advanced Generation III + pressurized water reactor (PWR) nuclear power plant (NPP), is designed to have better transient response capability, higher security and economy compared with the traditional Generation II PWR NPPs. This is partly due to the implementation of the advanced Mechanical Shim (MSHIM) core control strategy that eliminates the need for boron concentration adjustments during load follow operations. Since the world’s first AP1000 NPP is still under construction, there is no actual operating experience and data. This paper studies the dynamic responses and control characteristics of the AP1000 NPP during frequency control transients by modeling and simulation. First, a complete set of nonlinear mathematical models for the AP1000 NPP, including those of the nuclear steam supply system (NSSS) and balance of plant (BOP) system, was established. The S-function modules for these mathematical models and the modules of corresponding control systems were developed in MATLAB/Simulink. Then, by coupling all the modules according to the “reactor-following-turbine” operation mode, a full-scope, dynamic simulation platform of the AP1000 NPP was developed. After validation of the platform against the best-estimate analysis code CENTS developed by the Westinghouse under two typical 10% of full power (FP) step load change transients, it was used for the dynamic simulation and study of frequency control characteristics of the AP1000 NPP. Typical frequency control transients with ± 3%FP step load changes in power grid were simulated. Dynamic responses of key parameters of the NSSS and BOP system were presented and studied to assess and evaluate the load follow capability, economy and safety of the AP1000 NPP during these transients. Moreover, the impact of key primary frequency control parameters on the dynamic response characteristics of the AP1000 NPP and the control effect of system frequency was also studied, and the recommended values of the frequency deadband and bias coefficient have been given for the world’s first AP1000 NPP under construction in Sanmen, Zhejiang province of China. The simulation and analysis results in the present study contribute to the digestion and absorption of the frequency control characteristics of the AP1000 and can also offer certain guidance for the engineering practice.

Condition Monitoring and Diagnostics of Nuclear Mechanical Components

Typically, there exist two pressure boundary systems in a PWR(Pressurized Water Reactor) NPP(Nuclear Power Plant). One is the reactor coolant system(RCS) called as the primary system or Nuclear Steam Supply System(NSSS). The other is balance of power(BOP) called as the secondary system. The key mechanical components of the primary system is mainly comprised of reactor pressure vessel(RPV), pressurizer(PZR), steam generator(SG), reactor coolant pump(RCP) and pipings. The NIMS(NSSS Integrity Monitoring System) has been developed to monitor and diagnose the structural safety of the NSSS components on an on-line basis. In this paper, the Korean NIMS developed by KAERI is introduced. It is comprised of four independent sub-systems such as IVMS(Internal Vibration Monitoring System), LPMS(Loose Part Monitoring System), ALMS(Acoustic Leakage Monitoring System), and RCPVMS(Reactor Coolant Pump Vibration Monitoring System). The IVMS has been developed for early detection of the degradation of the preload condition of the reactor internal components by measuring the change of vibratory modal frequencies of the core barrel assembly. The LPMS is to monitor the presence of a loosened or detached metallic object within the reactor coolant system using the vibration sensors installed on the surface of the system. The primary purpose of the ALMS is to monitor coolant leakage at the potential leak regions such as the reactor vessel, welded region in pipings, and valves, etc. The second purpose is to detect initiation of crack on the surface of the pressure boundary of the reactor coolant system. The RCPVMS is to provide diagnostic information for detecting such symptoms as the shaft crack, the RCP misalignment, rotor unbalance, and the abnormalities of support bearings.

Parametric optimization of steam cycle in PWR nuclear power plant using improved genetic-simplex algorithm

A parametric optimization framework for the steam cycle of a typical pressurized water reactor (PWR) nuclear power plants (NPP) (i.e. Daya Bay nuclear plant) is proposed. The approach determines the optimal operation parameters of the steam cycle to maximize the power output with primary loop remaining unchanged. Thermodynamic model of the steam cycle system has been established to determine the power output under different operating conditions, and the availability and accuracy of the model are investigated. In addition, an improved genetic-simplex algorithm (IGSA) is put forward by using modified search strategies and integrating genetic algorithm (GA) with simplex algorithm (SA). Performance comparison between the improved algorithm and the original ones is performed by solving the optimization test problems. The improved algorithm enables the handling of nonlinear constrained optimization problem because of dramatically enhanced search capability. Based on the model, the parametric optimization for the steam cycle of Daya Bay PWR nuclear plant is carried out using the IGSA, GA and SA, respectively. In the optimized schemes it has been found that the IGSA can yield increase in power output as much as 23.8MW by an average of 236 iterations over current practice, and the corresponding thermal efficiency increases from 33.87% to 34.69%. In comparison, even more iterations have been executed, the maximum power output increments gained by the GA and SA are 21.4MW and 18.7MW, respectively. The results show that the new algorithm IGSA is superior to original ones. It also demonstrates that the proposed approach can effectively increase the efficiency of the PWR nuclear power plant without burning any additional fuel or replacing the devices of the system.

Numerical investigation on cavitation within letdown orifice of PWR nuclear power plant

The present study numerically investigates the influence of three parameters (including outlet pressure, the circular bead of orifice inlet and the shape change of orifice outlet induced by cavitation erosion) on the cavitation characteristics within letdown orifice of pressurized water reactor (PWR) nuclear power plant. In order to understand the determinants of cavitation within letdown orifice, the influences of the mentioned parameters on cavitation characteristics are revealed. Correspondingly, the formation and variation of low pressure zones are analysed. Meanwhile, areas with high risk of cavitation erosion are displayed and confirmed through the examination in nuclear power plant. From the above, it can be concluded: Cavitation would not occur within letdown orifice unless the outlet pressure decreases to a certain value when inlet pressure remains unchanged; once cavitation has been formed within letdown orifice, the degree of cavitation would increase with the decrease of outlet pressure. The variation of circular bead could change both the location and the degree of cavitation. The shape change of orifice outlet induced by cavitation erosion would enhance the degree of subsequent cavitation around the discharge end of letdown orifice; in addition, this effect would be strengthened with the increase of shape change.