Pressurized Water Reactor Power Plant Research Articles

A Multisource Uncertainty Fusion Reliability Evaluation Method for the Control Rod Drive Mechanism of Nuclear Power Plants

The reliability of a pressurized water reactor power plant’s control rod drive mechanism (CRDM) is affected by many factors, such as operation states, unit performance, and dynamic environments. Multiple sources of uncertainties, including random, interval, and fuzzy, exist when analyzing the reliability of CRDMs. Modeling reliability without considering the fusion of multisource uncertainties may result in model distortion and may not provide realistic assessment results. This paper proposes a multisource uncertainty fusion method powered by the margin-based variable transformation and the Bayesian network propagation. The interval and fuzzy variables are transformed into random variables to obtain the corresponding generalized probability density function of the CRDM’s feature parameters. After that, the CRDM’s reliability can be evaluated via probability quantization of the Bayesian network inference result through sampling algorithms. A magnetic-jack type CRDM case is presented to verify the proposed method, and the results show that this method can fuse three types of uncertainty variable in a unified way and effectively obtain reliability evaluation results.

Online intelligent fault diagnosis of redundant sensors in PWR based on artificial neural network

Sensors in the primary circuit of a pressurized water reactor (PWR) are normally designed with redundant structures to improve system safety and reliability. However, reliability of the actual system is often lower than that obtained by theoretical calculation due to the inevitable occurrence of common mode fault (CMF), which is a dependent failure event that can cause multiple failures in redundant channels. CMF may increase the reliability deviation of the system by orders of magnitude and, hence, seriously affects the reliability of the system. To mitigate the CMF of redundant sensors in nuclear power plants, an artificial neural network (ANN) can serve as a data-driven analytic model to monitor sensor parameters, to identify any possible abnormal status of the sensors, and provide an early warning. In this study, by using the high-fidelity dataset obtained in a full-scope PWR simulator as training, validation, and test data, a relevant parameter-based ANN black-box model (RPANN) was established by employing the back-propagation (BP) learning algorithm, which was then defined as an analytic redundancy. Time series-based ANN checking models (TSANNs) were also established for each of the input and output parameters of the RPANN in order to identify its abnormal state based on historical data in the past. When combined with the existing hardware redundancy, the ANN-based analytic redundancy can serve as an online monitoring tool of the hardware status and an online diagnosis strategy for sensor faults. Furthermore, ANN-based analytic redundancy can replace faulty hardware sensors to analytically reconstruct the reading of the monitored sensor parameter without having to reduce the reactor output power or even shut down the reactor for emergency maintenance so that the on-site calibration frequency of hardware sensors in redundant channels can be effectively reduced. This is not only of vital importance in reducing operation and maintenance costs of existing PWR power plants but also plays an important role in building reactor operation schemes with rapid and frequent changes in power output in the future. Simultaneously, the diverse redundancy combining analytic software redundancy and physical hardware redundancy can effectively reduce the threat of CMF of hardware sensors on the operation safety of reactor systems.

Seismic dynamic analysis and structural integrity assessment of the PWR spent fuel pool under beyond design basis earthquake

A series of numerical analyses were carried out to assess the safety of spent fuel pool (SFP) of a pressurized water reactor (PWR) power plant in Taiwan. These analyses performed seismic force evaluation and 3D nonlinear finite element analysis (FEA) for the SFP under beyond design basis earthquake (BDBE). In addition, this study considered various in-structural response spectra (ISRS) with various damping ratios and the reduced factors to account for the partial cracking effects from BDBE. In the seismic force evaluation, the results show that the pool base takes the greatest uniform equivalent loading resulted from the cooling water, racks, and spent fuel components. For the walls, the south wall suffers the greatest equivalent force. The results of 3D FEA indicated that the largest values of stress and strain appear at the bottom corner near south wall of the channel between the main pool and cask loading pit for the overall SFP structure. It is also found that the influences of cracking effects on the seismic force calculation and structural analysis are significant. But comparing the results of numerical analysis by 5% damping ratio with 0.7 reduced factor and 10% damping ratio with 0.5 reduced factor, the impacts on the liner of the SFP are insignificant. Because the cooling water leakage depends on the liner failure, the influence on the leakage is small.

Research on the inlet preswirl effect of clearance flow in canned motor reactor coolant pump

For a pressurized water reactor power plant, the reactor coolant pump (RCP) is a kernel component. And for a canned motor RCP, the rotor system's properties determines its safety. The liquid coolant inside the canned motor RCP fills clearance between the metal shields of rotor and stator, forming a lengthy clearance flow. The influence of inlet preswirl on rotordynamic coefficients of clearance flow in canned motor RCP and their effects on the rotordynamic characteristics of the pump are numerically and experimentally investigated in this work. A quasi-steady state computational fluid dynamics (CFD) method has been used to investigate the influence of inlet preswirl. A vertical experiment rig has also been established for this purpose. Rotordynamic coefficients on different inlet preswirl ratios (IR) are obtained through CFD and experiment. Results show that the cross-coupled stiffness of the clearance flow would change significantly with inlet preswirl, but other rotordynamic coefficients would not change significantly with inlet preswirl. For the case of clearance flow between the stator and rotor cans, influence of inlet preswirl is not so significant as the IR is not large enough.

Pressurizer water level control with estimation of primary circuit coolant mass in nuclear power plants via robust observer based dynamic sliding mode control

Pressurizer is one of the most important components in a pressurized water nuclear power plant. Since it is responsible for coolant mass balance and non-boiling heat transfer in the primary circuit. The pressurizer water level control system in a nuclear power plant with the Pressurized Water Reactor (PWR) is responsible for the 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 such as the coolant leakage in the primary circuit pipeline system. In the process of PWR power plant operation, the incorrect pressurizer water level may disturb pressure control or may cause damage to electric heaters which could threaten plant security and stability. In modern reactors, standard Proportional–Integral–Derivative (PID) controllers are used to control the water level in a pressurizer. In this paper, for the first time, in order to improve the performance and transient response of the pressurizer control system, an observer-based dynamic sliding mode control using Lyapunov-approach is designed and applied for the level control problem of the pressurizer in a pressurized water nuclear power plant based on the validated nonlinear model. The proposed method exhibits the desired dynamic properties during the entire output tracking process independent of perturbations. The comparison between the Dynamic Sliding Mode Control (DSMC) and the PID controller shows a significant improvement in water set-point tracking and increased ability in disturbance rejection for the proposed observer-based DSMC. Simulation results confirm the improvement in the transient response obtained by using the proposed controller.

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%.

EcSOD in the Health Benefits of Exercise

The coupling of the Qeshm island desalination plant with different energy sources was economically evaluated by using a Desalination Economic Evaluation Program (DEEP) package. The capacity of this plant was 18,000 m3/day and its energy source was 250 MW (th) gas power plant. Six energy source plants, namely, coal, oil, gas, combined cycle (CC), pressurized water reactor (PWR), and pebble-bed modular reactor (PBMR), were coupled with this desalination plant. The estimated annual electrical costs indicated that the optimum fossil power plant was CC with 97.72 $/kWh, which was 30% higher than 67.42 and 64.63 $/kWh in PWR and PBMR power plants, respectively. The total annual water costs and the total annual costs of coupling with the desalination plant of CC, PWR, and PMMR plants were approximately 1.78, 1.49, and 0.77 $/m3, and 94.74, 49.23, and 49.1 m$, respectively. The construction costs of these power plants are approximately 131, 510 and 553 M$, respectively, On the basis of the contraction times, costs, and lifetimes of these power plants 25, 40 and 60 years for CC, PBMR and PWR, we concluded that CC, PBMR and PWR are short-term, medium-term and long-term strategy to generate electricity and couple with desalination plants, respectively.

Coefficient of friction evolution with temperature under fretting wear for FeCrAl fuel cladding candidate

A leading cause of fuel failures in U.S. pressurized water reactor power plants is grid-to-rod fretting. One of the major variables affecting fretting wear is the coefficient of friction (COF). Reliability of the fuel cladding is influenced by the COF and thus it is important to know how the material's COF will be affected by temperature transients. Kanthal advanced powder metallurgy technology or APMT, a FeCrAl steel alloy, is being investigated as a commercial alternative to conventional fuel cladding in a nuclear reactor due to its potentially favorable performance under accident conditions. This leaves APMT steel fretting performance to be examined. Tests were performed to examine the evolution of the COF with temperature under fretting. The contact was simulated with both a cylindrical and a rectangular specimen over a line contact area. Confocal scanning laser microscopy was used to obtain a 3D map of the surface, which was used to calculate the wear rate coefficient of the samples. The wear rate coefficient was compared to wear volume to show the effects of the COF and temperature on wear. Energy Dispersive Spectroscopy was performed to qualitatively describe the microchemical changes the material underwent in the fretted region. Nanoindentation was used to compare microstructural properties inside and outside the fretting scar after fretting at different temperatures. COF measurements showed that APMT steel's COF and wear decreases with temperature up to 350 °C.

Assessment of fuel-rod meltdown in a severe accident at Bushehr nuclear power plant (BNPP)

After the Fukushima disaster, interest in the evaluation of severe accidents in nuclear power plants and off-site consequences has significantly increased. Because experimental studies are difficult to conduct, computational methods play a substantial role in accident analysis. In this study, a severe accident in the Bushehr pressurized water reactor power plant caused by a station blackout with a total loss of alternating current power supply has been evaluated. This analysis presents the in-core damage of fuel rods and the release of fission products as well as the thermal hydraulic response of the station components during the loss of active emergency cooling systems. In this manner, a perfect model of the Bushehr nuclear power plant using the MELCOR code is prepared. The accident progression is simulated, and the thermal responses of the fuels and hydraulic components are presented. It is shown that, without operator intervention, steam generators will become dry in approximately 3000 s, and the heat sink of the reactor will be lost. The simulation results show that at approximately 8600 s, the upper parts of the core start melting. This model calculates the shortest available time for accident prevention and proves that the time available is sufficient for operator manual action to prevent a nuclear disaster.

Risk Assessment of Mediterranean Sea Waters on Cooling System of Nuclear Power Plant

The study evaluated the impacts such as corrosion and scale formation which may result from, using the once-through pass cooling system of a suggested nuclear power at different selected coastal regions of Alexandria coast. (Abu-Qir, Sidi Kerrir, Estern Harbour, El Max and Sidi Abd Elrhaman). The study had used The Langelier Saturation Index (LSI), which is a measure of a solution’s ability to dissolve or deposit calcium carbonate, which is often used as an indicator of the corrosivity of water. Like the Langelier saturation index, the Ryznar stability index (RSI) is also used to predicate the formation of scale and also the tendency for corrosion due to the concentration of calcium carbonate in the selected water system. Both of the two indexes ((LSI) and (RSI)) had been used to evaluate the internal environmental impact which may be occurred when used once-through pass system is as a cooling method of a coastal pressurized water reactor power plant. The study results showed that the calculated LSI and RSI values for selected regions indicated that, there was a low tendency for scale formation and possibility of corrosion at these selected study areas, so using nickel copper alloys in condenser is recommended to prevent any possibility of attack of corrosion besides the using of chemical treatment method of inlet cooling water system.

Fretting wear comparison of cladding materials for reactor fuel cladding application

Relative motion between the fuel rods and fuel assembly spacer grids can lead to excessive fuel rod wear and, in some cases, to fuel rod failure. Based on industry data, such grid-to-rod-fretting is a significant cause of fuel failures in U.S. pressurized water reactor power plants. Kanthal advanced powder metallurgy technology or APMT, an FeCrAl steel alloy, and a braided SiC fiber, Chemical Vapor Infiltration SiC matrix (SiC/SiC) cladding by General Atomics are possible alternatives to conventional fuel cladding in a nuclear reactor due to their favorable performance under accident conditions. Tests were performed to examine the reliability of the cladding candidates and a conventional cladding, Zircaloy-4, under dry fretting conditions at elevated temperature. The contact was simulated with a rectangular and a cylindrical specimen over a line contact area. Confocal scanning laser microscopy was used to obtain a 3D map of the surface, which was in turn used for wear and work rate calculations on the samples. The wear rate coefficient was used as a measure of the performance and wear under fretting. Additionally, Energy Dispersive Spectroscopy was performed to qualitatively describe the microchemical changes the material undergoes during fretting. While APMT steel and SiC/SiC can perform favorably in loss of coolant accident scenarios, they also need to perform well when compared to Zircaloy-4 with respect to fretting wear. Wear coefficient measurements showed that APMT steel performs favorably in comparison to Zircaloy-4 with respect to fretting wear.

Exergy and exergoeconomic analyses of a novel integration of a 1000 MW pressurized water reactor power plant and a gas turbine cycle through a superheater

Combined cycles are used for various reasons, including to increase the efficiency of power generation system. In this study, a gas turbine cycle is combined with a pressurized water reactor (PWR) power plant to increase the total plant efficiency. In this novel cycle, saturated steam produced in the steam generators of the nuclear power plant is superheated by the hot combustion gases exiting the gas turbine. An exergoeconomic analysis is carried out and the effects of compressor pressure ratio and gas turbine inlet temperature are investigated on the net power output, the first and second law efficiencies, the total cost rate and the specific cost of the produced work. The results show that there is an optimum pressure ratio for each gas turbine inlet temperature. The combined cycle total cost rate and the specific cost of the produced work for a gas turbine inlet temperature of 1500 K and a compressor pressure ratio of 13 are determined to be 41,882 $/h and 31.63 $/MWh, respectively.

The FAC Experimental Research on Vapor-water Two-phase Flow Piping System of CAP1400 Secondary Loop

CAP1400 power plant is one of the advanced Third-generation Pressurized Water Reactor power plant. To ensure the safety and stability of the power plant secondary loop during operation, the corrosion characteristic of CAP1400 secondary loop should been investigated. This paper set up a test facility to study the vapor-water two-phase flow corrosion status, which could be seen in the piping system after the control valve of heater drain line. The key parameters, such as pressure drop, temperature, and surface topography have been observed and measured, and the different corrosion characteristic in straight pipe, elbow, tee, and reducer also have been acquired in the tests. The test results and analysis could provide the basis for the corrosion resistance design of the drain pipelines in the CAP1400 nuclear power plant.

Economic evaluation of Qeshm island MED-desalination plant coupling with different energy sources including fossils and nuclear power plants

The coupling of the Qeshm island desalination plant with different energy sources was economically evaluated by using a Desalination Economic Evaluation Program (DEEP) package. The capacity of this plant was 18,000m3/day and its energy source was 250MW (th) gas power plant. Six energy source plants, namely, coal, oil, gas, combined cycle (CC), pressurized water reactor (PWR), and pebble-bed modular reactor (PBMR), were coupled with this desalination plant. The estimated annual electrical costs indicated that the optimum fossil power plant was CC with 97.72$/kWh, which was 30% higher than 67.42 and 64.63$/kWh in PWR and PBMR power plants, respectively. The total annual water costs and the total annual costs of coupling with the desalination plant of CC, PWR, and PMMR plants were approximately 1.78, 1.49, and 0.77$/m3, and 94.74, 49.23, and 49.1m$, respectively. The construction costs of these power plants are approximately 131, 510 and 553M$, respectively, On the basis of the contraction times, costs, and lifetimes of these power plants 25, 40 and 60years for CC, PBMR and PWR, we concluded that CC, PBMR and PWR are short-term, medium-term and long-term strategy to generate electricity and couple with desalination plants, respectively.

Control rod drop hydrodynamic analysis for a pressurized water reactor

During operation of nuclear power reactors, reactivity initiated accidents can take place such as a control rod drop. If this occurs, the reactivity increases significantly and leads to an enhancement in power, fuel temperature and damage of reactor eventually. Exact assessment of these accidents depends on the hydrodynamic information. In this research, it is tried to simulate the unsteady flow field around the control rod for a pressurized water reactor power plant. In order to simulate the flow field around the control rod inside the guide tube, averaged Navier–Stokes equations accompanied by the layering dynamic mesh strategy have been used. The information exchange between the two computational stationary and moving grids, the computational grid around the control rod and the grid next to the guide tube, has been taken place through the interface. It was concluded that the time duration of control rod to reach the bottom of the core depends on the leakage. It was also observed that the velocity and acceleration of the control rod would be reduced by decreasing leakage flow rate and in certain leakages, the acceleration of the control rod approaches zero due to equilibrium conditions. During this research, a correlation based on the achieved data was proposed which would provide useful information on the relation between the leakage and the time for control rod to reach the bottom of the core.

Corrosion Behavior on Ni-base Alloys Applied with Nano-TiO2 in High Temperature Caustic Water

In Pressurized Water Reactor (PWR) power plants, Alloy 600, a Ni-Cr-Fe alloy, is used for steam generator tube materials. However, these tubes have experienced a lot of corrosion problems during their operation times. Several chemicals have been investigated for inhibitors of stress corrosion cracking (SCC) in the tube materials. Titanium dioxide (TiO2), which is sonochemically treated in water, was tested to evaluate the inhibition effect of SCC using a reverse U-bend specimen of alloy 600 under the condition of a 10% NaOH solution and a temperature of 315oC. The TiO2 particle size was reduced by ultrasonic power, and the morphology of the TiO2 powder was observed by a transmission electron microscope (TEM) based on the ultrasonic processing times. The particle size of TiO2 affects the SCC rate of the tube materials, and shows an improvement in resistance on SCC when decreasing the particle size of TiO2. When the nano-sized inhibitor is applied, the property of the oxide layer is changed to a more dense composition. The chemical composition in the oxide layer was analyzed using an X-ray photoelectron spectroscope (XPS) with a variation in the ultrasonic process times, and the difference in the Ti-compound structure between the oxide surface and inner layer of oxide was compared using XPS data. Keywords: Corrosion inhibitor, nanoparticle, sonochemistry, steam generator tube, stress corrosion cracking (SCC), titanium dioxide (TiO2).

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

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.

Economic Evaluation on the MOX Fuel in the Closed Fuel Cycle

The mixed oxide (MOX) fuel is one of the most important fuels for the advanced reactors in the future. It is flexible to be applied either in the thermal reactor like pressurized water reactor (PWR) or in the fast reactor (FR). This paper compares the two approaches from the view of fuel cost. Two features are involved. (1) The cost of electricity (COE) is investigated based on the simulation of realistic operation of a practical PWR power plant and a typical fast breeder reactor design. (2) A new economic analysis model is established, considering the discount rate and the revenue of the reprocessed plutonium besides the traditional costs in the processes of fuel cycle. The sensitivity of COE to the changing parameters is also analyzed. The results show that, in the closed fuel cycle, the fuel cost of applying MOX fuels in the FBR is about 25% lower than that in the PWR at the current operating and fuel cycle level.

Optimization design for mechanical seals in reactor coolant pumps based on a fluid–solid strong-interaction model

The study on the mechanism and performance of the mechanical seals in reactor coolant pumps (RCPs) is very important for the safe operations of pressurized water reactor power plants. By exploring the operating mechanism of the first seal of the hydrostatic mechanical seal in RCPs, an analytical fluid–solid strong-interaction model of the seal is proposed in this article. The model holds that the mechanical deformations of the seal assembly are dominated by the deflections of the seal rings, and this idea is demonstrated by the numerical simulation result of a fluid–solid interaction (FSI) model. Using the analytical FSI model, the regularity that the leakage rate of the first seal varies with the differential pressure in a RCP is obtained, and compared with the operational data, which is used to verify the model. Based on the understanding of the reliability of the seal, a dimensionless parameter Λ that acts as an attribute to the reliability is proposed in this article. Using the analytical FSI model and Λ as the optimization algorithm and optimization object, respectively, the optimum designs about the seal faceplateconfigurations are performed. Also, the specific optimization conclusions are given simultaneously.

Fluid-solid strong-interaction model of mechanical seals in reactor coolant pumps

The studies on the mechanisms and performances of the mechanical seals in reactor coolant pumps are very important for the safe operations of the pressurized water reactor power plants. Based on the hydrostatic mechanical seal in reactor coolant pumps, an analytical fluid-solid strong-interaction model is proposed in this paper. According to the design features and operational principles of the seal, an analytical method to calculate the mechanical deformation of the seal assembly is developed based on the ring deformation theory. A strong-interaction algorithm combining the analysis of the mechanical deformation of the seal assembly and flow field between the seal faceplates is utilized, in which the three kinds of equations including the fluid domain, solid domain and coupling action are constituted in the same equations set and all the variables are solved simultaneously. So the analytical fluid-solid strong-interaction model used for the seal is built. Moreover, the model is verified by the experimental results. Based on the model, the design parameters of the seal are studied. Two different conditions of the general case and fixed seal leakage rate are discussed respectively, and the regularities that the seal behaviors are affected by the parameters of the holding screws on the clamp rings and seal faceplates are obtained. The research results can provide a theoretical basis for performance analysis, design and assemblage of the seal. Compared to the numerical methods, the proposed model has the unique advantages of high efficiency, convenience and easy application of constraints.