Pressurized Water Reactor Nuclear Power Research Articles

Impact and tangential composite fretting wear of Zr-4 alloy tubes under random loading conditions

The fuel rod casing tube is a crucial component in pressurized water reactor (PWR) nuclear power plants. Flow-induced vibrations in the circulating water can result in complex alternating fretting wear between the casing tube and the positioning lattice frame. Existing research on fretting wear has primarily focused on unidirectional wear, with limited investigation into composite fretting wear under complex working conditions. This study conducted impact and tangential composite fretting wear tests on Zr-4 alloy tubes under varying constant and random impact loads to examine fretting wear behavior. The findings show that the peak friction coefficients in the impact and tangential composite fretting wear tests were consistent across the three constant load conditions, with higher peak friction coefficients observed under random load conditions and a longer time required to reach these peaks during micromotion tests. Comparative analysis revealed that random impact and tangential composite fretting wear caused the most severe damage. Under constant load conditions, wear damage became more severe with increasing load, transitioning from oxidized and adhesive wear to a peeling layer as the load intensified in the impact and tangential composite fretting wear tests.

Adsorption mechanism of octadecylamine on inner surface of steam generator feed water pipe of pressurized water reactor nuclear power plant

Adsorption mechanism of octadecylamine on inner surface of steam generator feed water pipe of pressurized water reactor nuclear power plant

Coupling of the best-estimate system code and containment analysis code and its application to TMLB’ accident

With the development of advanced pressurized water reactor technology, the thermal-hydraulic coupling effect between the containment and the primary system becomes increasingly tight. In order to meet the demand for integrated safety analysis between the containment and the primary system, this paper investigates a direct coupling method between the best-estimate system code Advanced Reactor Safety Analysis Code and the containment analysis program ATHROC (Analysis of Thermal Hydraulic Response Of Containment). The feasibility of this direct coupling method and the applicability of the coupled program for overall safety analysis are demonstrated using Marviken two-phase flow release experiments. The ATHROC/ARSAC coupled program is employed to analyze the impact of the pressure relief function of the CPR1000 nuclear power plant pressurizer on the behavior of the primary system and containment during the TMLB’ accident. The calculation results indicate that these measures can reduce the pressure of the primary system to the level acceptable by the low-pressure injection system, but at the same time, they cause the pressure in the containment to rise to nearly 0.4 MPa. Therefore, to ensure the structural integrity of the containment, it is necessary for the non-passive hydrogen recombiner to effectively reduce the hydrogen concentration, thereby avoiding additional pressure increase in the containment due to hydrogen deflagration, which could lead to overpressure failure. The findings of this study are of significant reference value for improving the safety performance of thermal-hydraulic systems in operational Gen-II and advanced Gen-III pressurized water reactor nuclear power plants.

Optimal design of circulating water system in pressurized water reactor nuclear power plant

The optimization of the circulating water system of pressurized water reactor nuclear power plants can effectively improve the profit without affecting the primary circuit operation, which is of great significance in enhancing the market competitiveness of nuclear power plants. The circulating water system of the nuclear power plant is modeled, and the circulating water system optimization scheme is formed after the nuclear power unit is reformed from three aspects: condenser heat transfer area, condenser connection mode, and circulating water pump operation mode. The circulating water system optimization calculation is carried out for each scheme by using the minimum annual cost method. Based on the sensitivity analysis of the factors that affect the profits of nuclear power plants, the best circulating water system optimization scheme is finally given for new nuclear power plants and nuclear power plants in operation, which provides a reference for the construction of related nuclear power projects.

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.

Experimental study on the retention of aerosol particles through concrete cracks under high Reynolds number flow

In the event of severe accidents in pressurized water reactor (PWR) nuclear power plants, the potential leakage of radioactive aerosols through containment cracks poses a considerable radioactive hazard to the public. Understanding aerosol transport and retention in cracks helps reduce the conservatism and uncertainty of radioactive hazard assessment. Concrete cracks are recognized as a pivotal pathway for the leakage of radioactive aerosols, and several researchers have undertaken experimental investigations concerning the aerosol transport and retention in concrete cracks. However, the majority of these studies have rather low gas flow Reynolds numbers. In this work, an experimental setup is built to study aerosol transport and retention in concrete cracks under high Reynolds number flow. The TiO2 aerosol with a mass median diameter of 1 μm and two concrete crack specimens are used in experiments. The results of gas flow experiments indicate that the Reynolds number is capable of reaching 10547. Combining the flow experimental data and Suzuki's formula, the equivalent heights of these two crack specimens are approximated as 303.67 μm and 231.48 μm. The experimental results indicate a notably high retention rate of aerosols, exceeding 0.8. Furthermore, under high Reynolds number flow, the retention rate varies over a relatively narrow range, with the larger the equivalent height of the crack resulting in a lower retention rate. The experimental results match well with the mechanistic analysis based on inertial deposition theory, demonstrating the rationality of the inertial deposition theory.

Study of Load Adjustment Strategy for Nuclear Power Units Focusing on Rankine Cycle: Flexibility–Environment–Economy

The demand for the power grid system’s capacity to undergo peak-shaving is increasing as the proportion of renewable energy rises. In China, nuclear power units usually only provide a base load operation in the view of safety and economic considerations, but they do not provide load adjustment services, which undoubtedly increases the pressure of grid load adjustment. In this paper, a novel flexibility load adjustment strategy of the CHP nuclear unit is studied, which is achieved by introducing the thermal storage tank (TST) into the Rankine cycle without changing the output of the nuclear reactor. The AP1000 pressurized water reactor nuclear power unit for combined heat and power is taken as an example, and the thermodynamic model is established through the water vapor equation. Furthermore, the reference system is simulated for the goal of minimizing the imbalance between power supply and demand, and the flexibility–environment–economy benefits are evaluated. The results show that the heat storage/release of the TST may achieve power output flexible adjustment of the nuclear unit, and the power imbalance of the reference energy system is reduced from 1107.99 MWh to 457.24 MWh, a reduction of 58.73%. The introduction of a 600 MWh TST can enable the reference unit to contribute 335 MWh of peak electricity within the reference day. From the perspective of replacing the power generation output increment of coal-fired power units with equal amounts, it can achieve a reduction of 106.09 tons of coal consumption in the case day, which means that 277.73 tons of CO2 emissions can be reduced. The profit of the reference unit can be improved by CHY 70,125 via participating in load adjustment in the case day if following the time-of-use electricity price.

Proposal and Validation of a Diagnosis Method of Fuel Failures in Pressurized Water Reactor Nuclear Power Plants

Based on research into the diagnosis methods for fuel failures in pressurized water reactor nuclear power plants and an analysis of existing problems, this paper proposes an enhanced fuel failure diagnosis method, addressing three key improvements: identification of the occurrence of fuel failure, diagnosis of the number of failed fuel rods, and the size of fuel rod failure. Other factors that may influence the diagnosis results are also identified and discussed. The proposed method is then verified using operational data from in‐service nuclear power plants in China. The verification results demonstrate that the improved diagnosis method accurately identifies fuel failures and exhibits broader applicability.

Research on hydrogen risk prediction in probability safety analysis for severe accidents of nuclear power plants

The risk of hydrogen combustion after the Fukushima nuclear accident has always been a topic of concern for the nuclear power industry. In the hydrogen risk assessment of probability safety analysis (PSA) in nuclear power plants, the traditional methods by using lumped parameter program method is fast but may have large uncertainties, while the newly developed CFD analysis method is more accurate but has not yet been widely applied to the hydrogen risk analysis of PSA due to a variety of factors. The lumped parameter analysis program MAAP is used firstly to obtain the hydrogen parameters under the severe accident of small LOCA in this paper, based on China's third-generation large-scale pressurized water reactor (PWR) nuclear power plant HPR1000, and then the probability of deflagaration to detonation (DDT) for hydrogen risk is analyzed with uncertainty. Secondly, the CFD software GASFLOW program is used to analyze the same accident sequence to obtain more accurate hydrogen distribution and other parameters, and at the same time to obtain the DDT probability value of hydrogen risk. The hydrogen distribution obtained by CFD calculation can be used to guide the uncertainty value of the lumped parameter program to produce more accurate DDT value of hydrogen risk. The analysis results show that there is a certain uncertainty in hydrogen risk assessment when using lumped parameter program method, which can be corrected by combining the gas distribution analysis results given by CFD software, to obtain a more accurate and reliable probability value, so as to provide a more accurate reference for PSA analysis and to improve the overall safety of the nuclear power plant.

The Study of Multiaxial Loading and Damage to the Structure and Materials in the PWR Steam Generator of Nuclear Reactor

In Pressurized Water Reactor (PWR) Nuclear Power Plants (NPPs), the steam generator is crucial for transferring heat from the primary to secondary cooling systems, vital for steam production to drive turbines, and central to nuclear power safety. This study explores recent research on multi-axial loading, structural integrity, and material durability in PWR steam generators, shedding light on key factors affecting these systems. Common corrosion-related degradation in steam generators often arises from design, material, and water chemistry factors. However, the shift to All Volatile Treatment (AVT), the development of advanced material alloys, and enhanced water quality control in primary and secondary systems have significantly reduced instances of steam generator degradation. These findings promise to enhance the reliability and safety of steam generators in future nuclear applications.

Application of the method of vacuum withdrawal inert radioactive gases to control of fuel rods leaking of nuclear reactors WWER types while operation

Currently, a comprehensive approach is being developed for the integrity control of fuel rod claddings in pressurized water reactor (PWR) nuclear power plants. One of the key stages is the reliable detection of leaking fuel rods within the reactor's active zone during power operation. This study examines the general trends in the activity of fission product radionuclides in the primary coolant of various generations and designs of PWRs. The experience of Novovoronezh Nuclear Power Plant is considered regarding the method of sampling and separation of inert radioactive gases from the primary coolant. A comparison is made between different methods of measuring inert radioactive gases in the primary coolant: direct spectrometry, cryogenic method, and vacuum extraction method. The issue of reliable detection of fuel rod leakage during reactor operation is discussed, using the example of Unit 5 at Novovoronezh Nuclear Power Plant. Comparative results are presented for the analysis of the integrity of fuel rod claddings based on key reference radionuclides. The most reliable results indicating fuel rod leakage were obtained through the control of specific activities and analysis of changes in the ratio of inert radioactive gases using the vacuum extraction method. Control of the specific activity of inert radioactive gases in the primary coolant enables a highly reliable assessment of the condition of physical barriers. The proposed method of sampling and separation of inert radioactive gases demonstrates high reliability and convergence of results, serving as an additional tool for integrity control of fuel rod claddings during power operation. Having representative data on the specific activities of inert radioactive gases allows for a reliable assessment of the condition of fuel rod claddings within the reactor core.

System-level modeling, analysis and coordinated control design for the pressurized water reactor nuclear power system

A flexible operation of the nuclear power system (NPS) over a large operating range is emphasized when it is used as marine power. The current study exposes two bottlenecks in this field: (1) lack of the system-level high-quality dynamic model whose structure is simple enough for control design limits the implementation of the model-based controller; (2) nonlinearity problem due to the large load variation poses challenges for NPS to maintain its performance over a wide operating range. For these reasons, a first-hand simplified nonlinear model is developed and validated for the pressurized water reactor nuclear power system (PWRNPS) through mixed modeling methods of mechanism analysis and data identification. An optimal 50% full-power compromise steady-state operation scheme is first proposed for PWRNPS to balance its primary coolant temperature and secondary steam pressure. Using the simplified nonlinear model as the predictive model, a novel nonlinear model predictive controller (NMPC) is designed to realize the wide-range flexible operation of PWRNPS. Especially, an integrating tracking error is innovatively added to the cost function of NMPC to achieve offset-free control. Two groups of simulations on a complex PWRNPS process developed on the MATLAB/SIMULINK platform demonstrate the effectiveness of the proposed methods.

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.

Energy and Exergy Analyses of a PWR-Type Nuclear Power Plant Coupled with an ME-TVC-MED Desalination System

Electricity–water cogeneration power plants are an important tool for advancing sustainable water treatment technologies because they provide a cost-effective and environmentally friendly solution for meeting the energy and water needs of communities. By integrating power and water production, these technologies can reduce carbon emissions and help mitigate the impact of climate change. This work deals with the energy and exergy analysis of a cogeneration plant for electrical power generation and water desalination using real operational data. The power side is a pressurized water reactor (PWR) nuclear power plant (NPP), while the desalination side is a multi-effect distillation (MED) system with a thermo-vapor compressor (TVC) plant coupled with a conventional multi-effect plant (ME-TVC-MED). A mathematical model was implemented in MATLAB software and validated through a comparison with previously published research. The exergy analysis was carried out based on the second law of thermodynamics to evaluate the irreversibility of the plant and the subsystems. In this study, the components of the sub-systems were analyzed separately to identify and quantify the component that has a high loss of energy and exergy. According to the energy and exergy analyses, the highest source of irreversibility occurs in the reactor core with 50% of the total exergy destruction. However, turbines, steam generators, and condensers also contribute to energy loss. Further, the thermodynamic efficiency of the cogeneration plant was obtained as 35.38%, which is more effective than other systems. In the ME-TVC-MED desalination unit, the main sources of energy losses are located in the evaporators and the thermo-compressor (about 50% and 36%, respectively). Moreover, the exergetic efficiency of the ME-TVC-MED unit was found to be low at 6.43%, indicating a high degree of technical inefficiency in the desalination process. Therefore, many opportunities exist to improve the performance of the cogeneration system.

Flexible operation of the pressurized water reactor nuclear power system using multi-model predictive control over a wide nonlinear operating range

The nonlinearity problem caused by the large load variation is a significant challenge for the pressurized water reactor nuclear power system (PWRNPS) to realize its flexible operation over a wide operating range when it participates in peak regulation in the power grid. Thus, this work proposes a novel multi-model predictive control (MMPC) strategy for PWRNPS to address its nonlinearity problem in multivariable control within a wide range of conditions, in which the coupling effect between the average coolant temperature and the steam turbine power is considered. Given the insufficient understanding of the nonlinearity of PWRNPS in previous work, an in-depth investigation in this regard is performed to support the control design, in which the nonlinearity of PWRNPS at the system level is analyzed both on quantitative and qualitative aspects for the first time. Considering the captured uneven nonlinearity distribution of PWRNPS, a new approach integrating hierarchical clustering with gap metric is proposed to guide the decomposition of the operating space of PWRNPS systematically, based on which MMPC is designed to realize the nonlinearity control of PWRNPS under the large load variation. Three types of simulations demonstrate the effectiveness of the proposed methods.

System design and operation optimization on the hybrid system with nuclear power, concentrated solar, and thermal storage

A hybrid system with nuclear power, concentrated solar and thermal storage is proposed to enhance the peak shaving ability of pressurized water reactor nuclear power plant, and operation strategies are provided aiming at operation flexibility and energy efficiency. The hybrid system is thermodynamically analyzed under the design condition. Results show that the energy efficiency of the hybrid system is 35.74%, and the equivalent solar-to-power efficiency is 24.05% under the design condition. Then, the hybrid system’s performances under off-design conditions are studied with changing direct normal irradiance (DNI) and output power. Results show that when DNI increases from 500 W/m2 to 900 W/m2, the equivalent solar-to-power efficiency of the system increases by 6.25%. When the core power keeps constant, the peak shaving depth of the hybrid system can be 25%. This study may provide guidance for improving the flexibility and energy efficiency of pressurized water reactor nuclear power plant.

Online estimation approach of the steam specific enthalpy for wet steam turbines in nuclear power plants

Wet steam turbines are widely used in pressurized water reactor nuclear power plants. The specific enthalpy of the wet saturated steam is an indispensable variable for the thermodynamic balance and performance monitoring of the wet steam turbines, but it is rarely online measured in practice. In this study, a model-based online estimation approach of the steam specific enthalpy for the wet steam turbine in nuclear power plants is proposed. The wet steam turbine is divided into several stage groups. In each stage group, the steam expansion process, the moisture separation process, and the steam extraction process are captured in mechanism modeling at first. Then, by solving the model, the specific enthalpy and the dryness fraction of the wet saturated steam are obtained. Using the historical operation data of a 1000 MW pressurized water reactor unit, the model-based pseudo-online simulations and the measured data are compared which suggests a high accuracy of the proposed approach.

Experimental research and theoretical analysis of flow instability in advanced secondary passive residual heat removal system

To cope with station blackout (SBO) accident in the pressurized water reactor (PWR) nuclear power plant (NPP), the advanced secondary passive residual heat removal system (ASP) was developed. A newly advanced secondary passive heat removal system test facility (ASPTF) for generation III NPP was designed and constructed to obtain the thermal hydraulic characteristics of ASP. The flow instability in ASP was observed in ASPTF, which was more likely occurred under low heating power and system pressure conditions. The effects of different influencing factors on flow instability in ASP were obtained. Then, the theoretical analysis of flow instability in ASP was conducted. Based on the experimental and theoretical analysis results, the optimization and improvement measures of ASP were summarized.

Design and implementation of cloud platform for nuclear accident simulation

Introduction: To meet the multi-user, cross-time-and-space, cross-platform online demand of work, and professional training teaching in nuclear reactor safety analysis under the normalization of Coronavirus Disease 2019.Method: Taking the nuclear accident simulation software PCTRAN as an example, this study adopts cloud computing technology to build the NasCloud, a nuclear accident simulation cloud platform based on Browser/Server architecture, and successfully realizes multi-user, cross-time-and-space, cross-platform applications. Targeting the AP1000, a pressurized water reactor nuclear power plant, the simulation of cold-leg Small Break Loss of Coolant Accident and cold-leg Large Break Loss of Coolant Accident were carried out to verify the correctness of the NasCloud’s accident simulation function.Results: The result shows that the simulation functions and results of the NasCloud in multi-terminal are consistent with the single version of PCTRAN. At the same time, the platform has high scalability, concurrency and security characteristics.Discussion: Therefore, the nuclear accident simulation cloud platform built in this study can provide solutions for the work and training of nuclear reactor safety analysis, and provide reference for other engineering design and simulation software cloud to computing transformation.

Experimental and numerical studies on shield building on non-rock site

Fluid-structure interaction (FSI) and soil-structure interaction (SSI) affect the seismic response of third-generation pressurized water reactor (PWR) nuclear power plants on non-rock site. In this paper, two 1/50 scaled simplified models were designed and a series of shaking table tests were carried to study the seismic responses of models considering FSI and SSI effect. Dynamic characteristics, acceleration, strain, hydrodynamic pressure and experimental phenomenon of the two models were analyzed. Moreover, the numerical analysis was performed and numerical results were compared with test data. The results show that the presence of model soil reduces the system’s frequency, and the presence and sloshing of water reduce the seismic response of model structure. The reasonableness and feasibility of the method used for the numerical simulation is verified. This study could provide a reference for analyzing the seismic response of prototype structure of nuclear power plant.