Pressurized Water Reactor System Research Articles

Concurrent fault diagnosis of small pressurized water reactors based on long-short term memory networks

The control systems of small pressurized water reactors (SPWR) with complex structures, compact layouts, and variable operating environment may be involved in various types of signal and concurrent faults. Concurrent faults can be taken as two or more single faults occurring simultaneously, which usually cause much larger damage to the system and are more difficult to be diagnosed than single faults. However, their fault diagnosis methods are rarely studied because of the numerous fault types and tremendous diagnostic difficulty. This paper explores the concurrent fault diagnosis method for sensors and actuators in SPWR control systems. An intelligent current fault diagnosis model is developed using long short-term memory network with the training and test datasets generated based on a fault simulation platform of the target SPWR. The test results show that both single and concurrent faults of the SPWR can be diagnosed rapidly in an average of 1.06 s after their occurrence with the classification and diagnosis accuracies reaching up to 96.61% and 97.27%, respectively. Moreover, by injecting different noise signals on the faulty dataset for training and validation, it is shown that the proposed LSTM network has strong noise immunity. This demonstrate the excellent diagnosis accuracy and efficiency of the model under both single and concurrent fault conditions. This study provides valuable guidance for the accuracy diagnosis of complex concurrent faults of nuclear power plants.

A multi-objective co-optimization method of controller parameters for the overall system of small pressurized water reactor

Small pressurized water reactors (SPWRs) normally have complicated and varying working conditions, and become an important direction for nuclear energy development. The overall system of SPWR, including the reactor coolant average temperature control system, the feedwater control system of the once-through steam generator (OTSG), the steam turbine speed control system and the steam dump control system, is used to guarantee the safe and stable operation. The overall system of SPWR is designed with multiple PI controllers. To accomplish intelligent self-tuning of PI controller parameters, reduce the dependence on human engineering experience, and improve the control performance of the overall system, a multi-objective co-optimization method(MMOCO) with PI controller parameters for overall system of SPWR is proposed. Firstly, based on the coordinated control theory and PI control algorithm, the overall system of SPWR is established. Then, a MMOCO for the overall system of SPWR is designed with the Non-Dominated Sorting Genetic Algorithm-II (NSGA-II). Finally, to compare the performance of the controller parameters derived from the MMOCO and the controller parameters obtained via the engineering tuning method, step load decrease transients are selected. The results show that the ability to control the overall system of SPWR is effectively improved after applying the MMOCO.

Implementing Component Degradations into a Modelica Model of an iPWR System to Develop Health Monitoring Techniques

The economic operation of small modular reactors will partly rely on managing and reducing inspection and maintenance activities while supporting new operational paradigms like load-following. Turbine control valves throttle the steam from the steam generator into the steam turbine while maintaining the pressure within the steam generator at a constant set point. Degradation of these components could impact the ability to manage electrical power production. Utilizing the Idaho National Laboratory Hybrid repository and the Oak Ridge National Laboratory TRANSFORM library developed for multiphysics simulations in Dymola/Modelica, an integral pressurized water reactor system was modeled based on the available specifications of the NuScale power module. The effects of various component degradation modes have been implemented into the model in order to simulate faulted plant data during both steady-state and load-following operations. The fault modes resemble different physical fault modes that may occur at an operating nuclear power plant; a leaking turbine control valve and a valve actuator failure due to loss of hydraulic pressure have been implemented. A neural network autoencoder is employed in conjunction with statistical analysis, namely, simple signal thresholding (SST) or sequential probability ratio testing (SPRT), to identify the presence of a fault. Fuzzy logic is additionally employed in a novel and promising manner to classify the state of the system based on the cumulative sum of the neural network residuals. SST and SPRT are both successfully validated using healthy data and proved capable of identifying both fault types; fuzzy logic identified the false positives and classified the faulted data correctly.

Discrete-time sliding mode prescribed performance controller via Kalman filter and disturbance observer for load following of a pressurized water reactor

In practical implementations, most nuclear reactors employ digital controllers to accomplish the load following operation. However, most of the pre-existing works only focus on continuous time controller design, which may yield degraded control performance when the control strategy is employed by a digital controller owing to the finite sampling frequency and the inevitable discrete-time approximations. To this end, this paper proposes a discrete-time sliding mode prescribed performance controller (DTSMPPC) via Kalman filter (KF) and disturbance observer (DO) for a pressurized water reactor (PWR), aiming to achieve an easier implementation on the practical nuclear reactor systems. Specifically, the continuous mathematical model of the PWR system with consideration of model errors and disturbances is first transformed to a discrete-time form by virtue of Euler’s discretization technique. Then, based on this transformed discrete-time model, the KF and DO, which provide the estimates of unmeasured system states (relative density of delayed neutron precursor, average fuel temperature, xenon concentration, and iodine concentration) and uncertainties, respectively, are constructed using the input/output measurement information acquired from the nuclear reactor system merely. By incorporating the observed information provided by the KF and DO into the control framework, the proposed DTSMPPC enables the load following error to fulfill the accurate and robust performance requirement even in the presence of unmeasured system states and uncertainties, while at the same time ensuring the transient and steady-state load following error within a prescribed zone described by performance functions. The system stability and the aforementioned theoretical findings are proved through rigorous analysis and simulations. Simulation results also reveal that the proposed DTSMPPC via the KF and DO yields a superior load following performance compared with some previous control strategies.

CMT and PBL study in the high-pressure phase on the advance core-cooling mechanism experiment with RELAP 5

The passive core cooling system is an important improvement for the emergency core cooling system of pressurized water reactor. However, although there’re studies focus on each component, the study on the integral effect still not sufficient. Especially the study on the core makeup tanks (CMTs). Our simulation study is based on the SBLOCA experiment on Advanced Plant Experiment (APEX) and Advance Core-cooling Mechanism Experiment (ACME) facilities by RELAP 5 MOD 3.2. This study contributes to identify the simulation error and tries to provide some effective way to establish the better describing model on RELAP 5.

Application of new axial power distribution synthesis method using in-core detector signal in digital core protection system

ABSTRACT A new method for the generation of axial power distributions using artificial neural network (ANN) technique and real-time measured planar radial peaking factor, Fxy (hereinafter ‘Live Fxy’) method for digital core protection system of pressurized water reactors is presented. ANN with Simulated Annealing (SA) technique to find global optimum solution was used to calculate core average axial power distribution. In Live Fxy method, axial node-wise Fxys were calculated by multiplying pin/box factors to the measured assembly powers inferred from in-core detector signals without calculating detailed 3D power distributions. The hot pin power distribution for DNBR and LPD was then obtained by multiplying the core average axial power distribution and Fxys at axial nodes. The validation of the method was performed for various core conditions (e.g. core power levels, control rod positions, etc.) of Korean OPR1000 power plant. The result showed a decrease in RMS errors by 2.2% ~ 7.0% (3.6% on average), and the minimum thermal margins for DNBR and LPD were increased by 6.4% and 15.6%, respectively. The application of the method would improve the accuracy of axial power distribution and thermal margin, contributing to the operability and safety of digital core protection system.

Modeling particle deposition in the primary circuit of pressurized water reactors for the OSCAR code

Particulate corrosion products are generated in the primary system of pressurized water reactors (PWRs) by volume precipitation and by erosion of oxides formed on metal surfaces through their uniform corrosion. The activation of corrosion products, mainly 58Co and 60Co (respectively coming from the activation of 58Ni and 59Co) leads to radiation field growth around the primary system, directly impacting system integrity and the radioprotection of nuclear workers. In order to understand and mitigate contamination by activated corrosion products, contamination predictions can be performed using the OSCAR code, which relies on the development of models to describe the numerous and complex interactions at stake. Particulate corrosion products account for a significant portion of corrosion products, as such the deposition/erosion mechanisms have their importance for the overall computation of surface or volume contamination. The aim of this article is to present an updated and inclusive deposition model for particulate corrosion products by taking into account surface interactions. The impact of the new deposition model on contamination predictions is then evaluated and has enabled to reproduce, for the first time using the OSCAR code, the preferential contamination in 58Co in the cold side of the circuit, measured by gamma spectrometry with the EMECC device on commercially PWRs.

PWR natural circulation at a small break LOCA

In this paper, a one-dimensional quasi-steady model has been developed to analyse the potential impacts of reduced primary coolant inventory on natural circulation. This model focuses on the development of expressions for the coolant inventory, temperature distributions, and mass flow rate. The model encompasses a range of fluid flow scenarios, including single-phase, combined single-and-two-phase, and two-phase natural circulation, as well as the phenomenon of reflux condensation. The model is also utilized to examine the impact of reactor core power on natural circulation parameters. Comparisons with experimental findings and earlier research are made, highlighting a satisfactory level of agreement with the analytical findings. Based on the findings of this study, it has been observed that within the range of primary coolant inventory of 100–60% and core decay power of 1.5–5% of full power, natural circulation in a PWR type system has the potential to serve as an effective mechanism for transferring core decay heat to the secondary loop. The flow rates of natural circulation in pressurized water reactors (PWRs) are observed to have a significant dependence onthe amount ofliquid inventory, whereas their relationship with power level is comparatively less pronounced. The flow rate of a pressurized water reactor (PWR) system tends to exhibit its highest value within a specific range of inventory, typically falling between 60% and 80%. The transition between the different modes of natural circulation is observed to exhibit a smooth transition, with varying system inventories.

Influence of chromium of low content on the corrosion behavior of SA106B carbon steels in high-temperature pressurized LiOH medium

The corrosion of carbon steel (CS) pipelines widely troubles the primary circuit system of pressurized heavy water reactors. The element chromium (Cr) plays a crucial role in resisting corrosion. However, the mechanism of oxide film formation and dissolution on low Cr-content CS remains unclear. A comparison study in corrosion behavior between SA106B with 0.02 wt% and 0.36 wt% Cr was carried out with weight loss tests, electrochemical measurements, and morphology observations. Immersion tests were performed in deaerated lithium hydroxide solution at 553 K and 6.3 MPa for 30 days. The results showed that although the content of Cr was as low as 0.36 wt%, it exhibited relatively high corrosion resistance to the SA106B with 0.02 wt% Cr. Two layers of oxide film were formed on both the two CSs and chromium oxide was present in the inner oxide layer of 0.36Cr CS. The possible formation mechanism of oxidation film with low content of Cr in high-temperature water was discussed.

Continuous tempering effect induced PWHT alternative technology using wire arc additive manufacturing for application in replacing nuclear pressurized water reactor system repairing: CALPHAD, FEM simulation, and EBSD investigation

This study proposes an effective repair technology using arc additive manufacturing for pressurized water reactors (PWRs) in nuclear power plants (NPP) aimed at avoiding complete replacements and post-weld-heat treatments (PWHTs) of component parts while ensuring safety and reliability. Effective repair technology is defined as economic and process efficiency, because of maintenance costs and radiation exposure, and it is critical in related industries. The technology is designed to relieve the hardness and martensite fraction of the welding heat affected zone (HAZ) of low alloy steels (SA508) in PWRs penetration/nozzles by heat source generated in the WAAM process, thus ensuring structural integrity. In the first layer of wire arc additive manufacturing processes, 89.6% of the martensite phase was formed in the HAZ of SA508, which was significantly reduced to 45.7% due to repetitive thermal behaviors at the third layer. The resulting process dramatically reduced hardness from 450Hv in the initial layer additive manufacturing to 320Hv in 3 layers without additional heat treatments. Moreover, the study quantitatively investigated the martensite starting temperature (Ms) and bainite starting temperature (Bs) and analyzed the microstructure and mechanical behavior of the developed process using thermodynamic calculation (CALPHAD), finite element method (FEM) simulation, and microstructure quantitative analysis by electron backs-catter diffraction (EBSD). The proposed technologies and their quantitative analysis results can be a substantial alternative to the repair technology for penetration/nozzles in nuclear primary water cooling reactor applications, complying with ASME Sec.IX Qw-462.12 and ISO 15614-1.

ESO-based adaptive event-triggered load following control design for a pressurized water reactor with samarium–promethium dynamics

Conventional control systems used in nuclear reactors rely on real-time information transmission between the controllers and the actuators, which is unnecessary and imposes a heavy communication burden. To this end, this paper proposes an extended state observer-based adaptive event-triggered control (ESO-AETC) scheme for the load following problem of a pressurized water reactor (PWR) in the presence of uncertainties and limited communication resources. Different from most of the previous controllers presented in the literature, the samarium–promethium dynamics are considered in the constructed PWR model in this paper. Based on the PWR model with the load-dependent parameters and the samarium–promethium dynamics, an ESO, which can estimate lumped uncertainties and unmeasured system states including the relative density of the delayed neutron precursor, the average fuel temperature, the xenon concentration, the iodine concentration, the samarium concentration, the promethium concentration, and the total reactivity, is designed. With the help of the ESO and the event-triggered techniques, an AETC scheme is developed to avoid the real-time communication in the controller-to-actuator channel of the PWR system. It is rigorously proved that the control error can converge to any prescribed residual set in a finite time according to Lyapunov stability theory, excluding the Zeno behavior. Finally, simulation studies involving comparisons with existing works, estimation performance verification, and sensitivity analysis are conducted to assess the proposed ESO-AETC scheme. Simulation results reveal that (1) relative to the pre-existing model-free controller (MFC) and the disturbance observer-based sliding mode controller (DO-SMC), the proposed ESO-AETC scheme is found to provide better load following performance with at least 56.12% lower maximum absolute error (MAE), at least 15.05% lower integral absolute error (IAE), and 87.58% reduced communication resources in the controller-to-actuator channel, (2) the ESO possesses strong estimation capability for system states and lumped uncertainties, and (3) the proposed ESO-AETC scheme yields a low sensitivity with respect to the change of the controller parameters.

Two-group bubble interfacial area concentration correlation for upward bubbly to churn flows in vertical rod bundle flow channels

The detailed understanding and accurate modeling of the upward two-phase flows in vertical rod bundle flow channels are significant for various industrial applications, such as performing the optimal design and evaluating the transient and accident scenarios for BWR (Boiling Water Reactor) and PWR (Pressurized Water Reactor) systems. The interfacial area concentration (IAC) describing the available interfacial area for the transfer of mass, momentum, and energy between two phases plays an important role in the two-phase flow simulation in the frequently-used thermal-hydraulics system codes. Although various IAC correlations utilizing the drag-based two-group bubble categorization, namely two-group bubble IAC correlations, have been proposed to predict the IAC changes in the two-phase flow ranging from bubbly to churn flows in simple-geometry flow channels, the two-group bubble IAC correlation for rod bundle flow channels is still underdeveloped. This study performed a survey on the existing representative two-group bubble IAC correlations for simple-geometry flow channels and collected 425 experimental IAC data taken in various vertical rod bundle flow channels. The predictive abilities of these two-group bubble IAC correlations were evaluated with the collected experimental IAC data. The evaluations showed that no existing correlations can consistently maintain satisfied predictive ability throughout the upward bubbly to churn flows in vertical rod bundle flow channels. In view of this fact, this study proposed a new two-group IAC correlation for vertical rod bundle flow channels. Specifically, the group-2 bubble void fraction model was proposed by physically analyzing the changing characteristics of the group-2 bubbles in different flow regimes. The Sauter mean diameter model of group-1 bubbles was derived by (a) selecting the distorted small bubble as the representative group-1 bubble for the bubbly, cap-bubbly and churn flows and (b) selecting the spherical small bubble as the representative group-1 bubble for the finely dispersed bubbly flow. The Sauter mean diameter model of group-2 bubbles was proposed by treating the cap-bubble as the representative group-2 bubble and taking into account the bubble deformation caused by the surrounding rods. The proposed two-group bubble IAC correlation has been checked against the collected existing experimental IAC data and the proposed correlation exhibited reliable predictive abilities in the bubbly to churn flows in various rod bundle flow channels with the mean relative deviation of 0.233, 0.212, 0.247, 0.222 and 0.234 for bubbly, finely dispersed bubbly, cap-bubbly, churn and all data, respectively.

Pressurizer system dynamic model for transient control in PWR

Abstract The pressurizer system of pressurized water reactor (PWR) maintains the reactor coolant system pressure during steady-state operation and limits pressure changes during transients. The in/out surge transients will cause pressure variations and they are controlled by either the spray system or the heater system. The spray system is actuated when the pressure exceeds a preset value. The heater system is initiated when the pressure falls below a preset value. The fundamental understanding and a reliable modeling of the pressurizer system behavior under steady state and transient conditions are needed to simulate overall nuclear power plant behavior. In the present study, an algorithm using Python 3.7 is developed to represent the dynamic behavior of the pressurizer system under steady-state and during in/out surge transients. Moreover, RELAP5 code is used to simulate the pressurizer system during the prescribed transients. The analysis and assessment results demonstrate satisfactory control performance during the in/out surge transients that guarantee the safety of PWR during operation. Also, the comparison between Python algorithm and RELAP5 model illustrates the capability and effectiveness of the Python algorithm for dynamic simulation and control.

Effect of organic film on magnetite deposition behavior of alloy 690 steam generator tubes in simulated PWR secondary system

The purpose of this study is to elucidate the effect of a film forming amine (FFA) on the magnetite deposition behavior of Alloy 690 tube in the secondary coolant system of pressurized water reactors. By the addition of octadecylamine (ODA), which is a FFA, an organic film with a petal-like form was evenly formed on the surface of the steam generator (SG) tube. Regardless of the addition of ODA, all simulated deposits were verified as magnetite and had a polyhedral or spherical shape. The amount of deposits on the film-formed SG tube was reduced by approximately 58% compared to the deposits on the as-received SG tube. The influence of the film-formed surface of the SG tube is discussed from the point of view of wettability, zeta potentials, and resistance of the electric double layer.

High-gain disturbance observer-based double closed-loop sliding mode controller for the load following of a pressurized-water reactor

This paper proposes a high-gain disturbance observer-based double closed-loop sliding mode control (HGDO-DCLSMC) strategy for the load following of a pressurized-water reactor (PWR) in the presence of disturbances. The proposed strategy consists of an HGDO and a DCLSMC. The HGDO is developed to estimate lumped disturbances caused by model uncertainties, output disturbances, and unmeasured states. By utilizing the estimation information from the HGDO, the DCLSMC strategy, which includes an inner reactivity loop and an outer power loop, is designed for the PWR system to ensure the load following performance while attenuating the adverse effects of lumped disturbances with feedforward compensation. Even with lumped disturbances and estimation errors of the HGDO, the global asymptotic stability of the overall controlled PWR system is demonstrated through Lyapunov stability theory in a rigorous manner. At last, comparative simulation studies reveal that the proposed HGDO-DCLSMC strategy outperforms a proportional–derivative (PD) controller and a conventional sliding mode controller (SMC).

What Would Happen to a Fusion Power Plant in the Case of Plasma Disruption? A Dynamic Model Simulation

It is a widespread view in the fusion community that steady-state, water-cooled fusion power plants can utilize the power generation systems of conventional pressurized water reactor (PWR) fission plants as is. However, what would happen to a fusion power plant in the case of plasma disruption? The authors constructed a dynamic simulation model of a water-cooled ceramic breeder blanket fusion power plant model on Modelica language [300.0-MW(electric) electrical output/1138-MW(thermal) fusion output] and evaluated the applicability of a PWR power generation system. Simulation results suggest that while the PWR system would function as intended during steady-state operation, the conventional system may not be able to cope with a sudden loss of energy influx in the event of plasma disruption without modification: The PWR system’s steam generator experienced a water overflow in less than 150 s from the plasma disruption.

A Comparative Study on Reliability Analysis Methods for Safety Critical Systems Using Petri-Nets and Dynamic Flowgraph Methodology: A Case Study of Nuclear Power Plant

Safety-critical systems (SCSs) of nuclear power plants (NPPs) are being designed and developed to meet high dependability requirements. Fault tree analysis (FTA) is widely used for risk and reliability analysis of NPPs. However, fault trees (FTs) are static and have only limited capability to represent dynamic systems. FTA is also not capable of modeling non-binary logic and or modelling the system's evolution in time. Dynamic reliability methods are being developed to deal with such limitations. Time series Markov chains and dynamic flowgraph methodology are the dynamic reliability methods alternate to traditional FTA, which can be used for the system performance analysis. In this article, Time series Markov chains and DFM methods, for the system reliability predictions for the SCS of NPP are compared. The benefits of the proposed method are brought out to the traditional methods. The approach is applied on passive residual heat removal system of pressurized heavy water reactor under the station blackout scenario.

Magnetite (Fe3O4) and nickel ferrite (NiFe2O4) zeta potential measurements at high temperature: Part II – Results, study of the influence of temperature, boron concentration and lithium concentration on the zeta potential

Predicting the deposition of typical corrosion products (CPs) particles formed in the primary system of pressurized water reactors (PWRs) is important for system integrity and the radioprotection of nuclear workers. Such corrosion products foul heat transfer surfaces, promote localized corrosion and, when made radioactive in the reactor core, transport throughout the coolant system and give rise to radiation fields around components. Accurately predicting CPs’ propensity to transport and deposit around the system entails knowing their zeta potentials, quantities that until now have been unavailable. The zeta potentials of magnetite and nickel ferrite particles between 20 °C and 240 °C have been measured in the chemical conditions representative of an operating cycle of the primary system of PWRs. The measurements were performed via the streaming potential method as described in a previous paper – Part I. The measured values increased with temperature but decreased with increasing concentrations of boron and lithium. They are suitable for predicting radiation field growth around components of a typical PWR system.

Magnetite (Fe3O4) and nickel ferrite (NiFe2O4) zeta potential measurements at high temperature: Part I—Design, materials and preliminary characterization of an apparatus implementing the streaming potential method

An apparatus for measuring the streaming potential of particles at high temperature has been designed and tested in a recirculating water loop operating at temperatures up to 300 °C. The ultimate objective is to derive values of the zeta potential of corrosion-product particles under the conditions of the primary coolant of pressurized water reactors (PWRs), quantities that until now have been unavailable. They are required for modeling the transport and radioactivation of particulate oxides around the coolant system and the consequential growth of radiation fields around components. The experimental technique employed, measuring the potential difference across a packed bed of particles through which the coolant flows, involved fabricating a tube of temperature-resistant polymer that insulated the particles from the containing stainless steel pressure boundary of the loop. The particle bed was contained between porous ceramic membranes and the whole sealed with shrunk-on PTFE to avoid by-passing flows. Platinum wires, inserted into the loop through Conax fittings and insulated from contact with it by coils of oxidized zirconium wire, served as electrodes upstream and downstream of the bed to measure the potential difference The detailed design and testing of the apparatus containing magnetite powder to verify its correct functioning under operating conditions are described. Since measuring streaming potential involved obtaining the voltage difference across the particle bed alone at the same time as the pressure drop through it from the flowing coolant was measured, the preparatory tests first established baseline values, from measurements on an empty test section, which were subtracted from the measurements on the full test section. Also, immediately before each streaming potential determination, a rapid measurement provided the overall electrical resistance of the test section. The initial measurements of the streaming potential of magnetite, between 20 and 240 °C in a solution containing 1800 ppm of boron and 3.5 ppm of lithium, led to corresponding values of zeta potential. The values show that the apparatus provided reproducible results. The results of the complete series of tests under conditions of the primary system of PWRs, involving nickel ferrite as well as magnetite particles, are the subject of a separate paper.

Numerical study of blowdown period of double-end break loss of coolant accident in PWR based on RELAP5

Loss of coolant accident (LOCA) is one of the most severe accidents of nuclear power plants (NPP). During the early blowdown period, the system is in a high-pressure gradient and high-velocity gradient state. Therefore, the hydrodynamic forces and safety analysis are highly related to calculating thermal-hydraulic parameters in the flow field. This paper studied the factors and modeling methods that affect the calculation of thermal-hydraulic characteristics under large break LOCA conditions. Firstly, the reactor core and primary system of a typical two-loop pressurized water reactor (PWR) NPP is modeled using the system analysis code RELAP5 Mod4.0. The normal operation state is obtained after the steady calculation. Then, transient calculation of double-end break LOCA is implemented considering the effects of break opening time and the pressurizer. The influence of the results on the calculation of hydrodynamic force has also been analyzed. Finally, the modeling method of reactor pressurized vessel (RPV) downcomer is presented. It is found that the flow field can be better simulated when the downcomer is modeled with parallel channels and several transverse junctions. Thus, this paper provides meaningful suggestions for modeling and simulation in the early blowdown period under LOCA conditions and contributes to further hydrodynamic load calculation and reactor safety analysis.