Safety Of Nuclear Power Plants Research Articles

Effects of temperature on the local fracture toughness behavior of Chinese SA508-III welded joint

The structural integrity of welded joints in the reactor pressure vessel (RPV) is directly related to the safety of nuclear power plants. The RPV is made from SA508-III steel in a pressurized water reactor. In this study, we investigated the effects of temperature on the tensile and fracture toughness properties of Chinese SA508-III welded joint in different sampling areas in order to provide reference data for structural integrity assessments of RPVs. The specimens used in tensile and fracture toughness tests were fabricated from the base metal (BM), weld metal (WM), and the heat-affected zone (HAZ) in the welded joint. The representative testing temperatures included the ambient temperature (20 °C), upper shelf temperature (100 °C), and service temperature (320 °C). The results showed that temperature greatly affected the fracture toughness (JIC) values for the SA508-III welded joint. The JIC values for BM and HAZ both decreased remarkably from 20 °C to 320 °C. The fracture morphologies showed that the BM and HAZ in the welded joint exhibited fully ductile fracture at 20 °C, whereas partial cleavage fracture was mixed in ductile fracture mode at 100 °C and 320 °C. The WM exhibited the ductile and cleavage fracture mixed mode at various temperatures, and the JIC values showed slight changes.

КОНЦЕПЦІЯ ДОВІДНИКІВ ДЕФЕКТІВ В ОБЛАДНАННІ ТА ТРУБОПРОВОДАХ ЕНЕРГЕТИЧНИХ ТА ТРАНСПОРТНИХ СИСТЕМ

The paper focuses on the one of the persperctive approaches to the increasing of thje safety of Nuclear Power Plants - Flaw Handbook Concept. Object of study - equipment and piping of Nuclear Power Plants. Purpose of study - the description of the Flaw Handbook Concept and the application of the concept for the specific example. Method of the study - numerical procedures of the finite-element method and fracture mechanics. In the modern economics the optimization of the performance and operation of industry and power systems is of the main importance. The Flaw Handbook Concept is considered in the paper. This concept is applied on the nuclear power plants in the leading states with the aim of the optimization of the procedures of in-service inspection and repair. The main steps of the concept are considered and applied for the specific example. An example of Flaw Handbook using is analysed. The results of the paper can be incorporated into the procedures of in-service inspection for the safety-significant equipment and piping. KEYWORDS: FLAW HANDBOOK, BRITTLE FRACTURE, FATIGUE, FINITE-ELEMENT METHOD, SURGE PIPE.

Application of the dynamic FMEA in the reliability analysis of DCS

Digital distributed instrumentation and control system (DCS) is critical to the safety of nuclear power plants (NPPs). Static analysis methods developed from analog control system are not applicable to DCS due to its enhanced dynamic interactions and complex structure of hardware/software/firmware. The enhanced dynamic interactions of DCS include both sequence and timing factors, which are hardly modelled in the traditional Failure Mode and Effect Analysis (FMEA). In this study, dynamic FMEA (DFMEA) method based on simulation technology is put forward for the design and review of DCS in NPP. DFMEA based on real DCS hardware and software is developed to reveal the dynamic failure paths and failure modes. The results of DFMEA can well support the establishment of the dynamic fault tree/event tree in the review of NPP DCS, which reduces the dependency on the analyst’s experience significantly.

CPR1000 核电厂液态流出物排放系统排放能力分析

CPR1000 nuclear power plant is a nuclear power plant supported by China’s advanced million-kilowatt (1000 mw) pressurized water reactor (PWR) nuclear power technology introduced by China general nuclear power corporation (CGN). The liquid effluent discharge system is an important module of CPR1000 nuclear power plant. The discharge capacity of the liquid effluent discharge system directly affects the operation safety and stability of CPR1000 nuclear power plant. Therefore, this paper takes the liquid effluent discharge system of CPR1000 nuclear power plant as the research object, on the basis of a brief overview of the structure of the liquid effluent discharge system, combined with the technical characteristics of the liquid effluent discharge system, the discharge capacity of the liquid effluent discharge system is further analyzed.

Modernization of Power-Generating Units with RBMK-1000 Reactors

A nuclear power plant is a complex technological facility. The introduction of new standards and experience gained in the operation of domestic and foreign nuclear power plants requires the introduction of changes in the operating organization of existing nuclear power plants and improvements in their equipment. A general survey of the modernization of existing nuclear power plants with RBMK-1000 type reactors is presented. Measures for modernization of the power-generating units of nuclear power plants carried out following the accident at the Chernobyl nuclear power plant and at the present time are discussed. Because of modernization, the safety, reliability and economic efficiency of nuclear power plants have improved. The principal focus is on modernization of the reactor plant.

Performance analysis of the SPRHRS for the 1000 MWe class PWR in coping with station black out

In order to improve the safety and reliability of the Chinese nuclear power plants (NPP), an innovative passive nuclear safety design, the secondary passive residual heat removal system (SPRHRS) is proposed for the light water reactors in China. Previous research work has shown that under design-basis accidents, SPRHRS can mitigate the results by partially or completely replacing the traditional emergency water cooling system. However, the Fukushima Daiichi nuclear power plant accident implies necessity of further assessments of NPPs capability to withstand threats resulting from accidents beyond design-basis accidents. In this study, the nodalization of a 1000 MW class PWR with the SPRHRS design was conducted using RELAP5/MOD3.4. Responses of the PWR under the loss of heat sink accident caused by the station blackout (SBO) was analyzed. The results showed that the residual heat can be effectively removed by the SPRHRS under SBO. In addition, sensitivity studies about the effects of the coolant tank volume and the isolation valve opening condition of the SPRHRS on its residual heat removal capability were also performed. The results indicate that these two main design elements have significant influence on the operation characteristics of SPRHRS. For the SPRHRS design in this study, the appropriate value of the coolant tank volume is 400 m3 and the appropriate valve opening signal is set to be triggered when the SG shell side collapsed liquid level drops lower than 20% of its full value. The present work would be instructive for the design of passive safety systems of Chinese NPPs.

МАРКОВСЬКІ МОДЕЛІ ОЦІНЮВАННЯ ФУНКЦІОНАЛЬНОЇ БЕЗПЕКИ ПРОГРАМНО-ТЕХНІЧНИХ КОМПЛЕКСІВ НА САМОДІАГНОСТОВНИХ ПРОГРАМОВНИХ ПЛАТФОРМАХ З УРАХУВАННЯМ ПОМИЛОК ЗАСОБІВ КОНТРОЛЮ

Ensuring the safe operation of nuclear power plants remains one of the most important tasks. An important role in ensuring the safety of nuclear power plants is played by instrumentation and control systems (ICS). Extremely high demands are made on such systems, first of all, on their reliability and functional safety. The object of research and analysis in this work is the Nuclear Island I&C Instrumentation System and Reactor Protection System, which are developed based on programmable RadICS Platform with self-diagnostic. The failure trees of such platforms and ICS based on them were developed. In the next stage, several Markov models of redundant ICS are developed. The article aims to develop and analyze the results of research on Markov models for reliability and safety assessment of ICS based on self-checking programmable platforms. The models take into account errors of checking and diagnostic tools, namely errors associated with identifying failures and failures of the corresponding components of the hardware and software/FPGA and channels of redundant structures. Models have been developed for various options of redundant structures and taking into account the diversity principle for ISC structure and failure rate caused by design defects. Multiple-fragment Markov Models of ICS were also investigated. The scientific novelty lies in the fact that the proposed models take into account an expanded set of parameters of self-diagnosing programmable platforms, ICS, the processes of their use and maintenance. Recommendations on the selection of parameters and structural ICS are formulated.

Effect of two-step tempering treatment on microstructure and impact toughness of bainitic steel for heavy wall thickness reactor pressure vessels

SA508 Gr.3 steel is widely applied for the fabrication of reactor pressure vessels, which require high impact toughness to ensure the safety of nuclear power plants. In this work, a two-step tempering treatment is proposed to manufacture the SA508 Gr.3 steel with high toughness. A pre-tempering at 400 °C was added before the ultimate tempering of intercritical heat treatment process. The ductile-brittle transition temperature of the steel processed with the new heat treatment is significantly decreased by 26 °C. The scanning electron microscopy, electron backscatter diffraction, and transmission electron microscopy characterizations show that the decomposition of M-A constituents into dispersed fine carbide particles and the increase of high-angle grain boundaries contribute to the decrease in ductile-brittle transition temperature. The results suggest that the modified two-step tempering treatment is effective to enhance the impact toughness of SA508 Gr.3 steel.

A study into the modes of the VVER-1000 RCP starting in an earlier inoperative loop

To simulate the mode of the RCP starting in an earlier inoperative loop, KORSAR/GP, a code supporting coupled numerical modeling of neutronic and thermal-hydraulic transients in a VVER reactor plant in operating and emergency conditions, was chosen as the computational tool. Studying these modes using thermal-hydraulic codes makes it possible to analyze the course of transients and certain emergency processes without using commercial test procedures, which contributes to laying the groundwork for addressing the issues involved in ensuring the reliability, operating safety and efficiency of nuclear power plants. Increased requirements to the safety of NPPs identify the need for avoiding excessive conservatism in the analysis based on which requirements to safety systems are formulated, as well as for enhancing the knowledge of the regularities of thermal-hydraulic transients based on advanced computer programs (or codes) designed for improved computational analysis of non-stationary thermal hydraulics in the water-cooled reactor circulation circuits in emergency and transient modes relying on inhomogeneous non-equilibrium mathematical models of two-phase flows and on a detailed description of the physical transient regularities. The purpose of the study is to analyze computationally the starting of a VVER-1000 RCP in an earlier inoperative loop at different reactor plant power values. To do this, one requires to develop the VVER-1000 reactor primary circuit computational pattern to model the transient taking place as one RCP is started, to conduct a further analysis, and to compare the key monitored reactor coolant and core parameters (power, temperature, flow rate, etc.).

Cause and effect analysis of failures at Russian nuclear power plants between years 1992 and 2018

Multiple systems determine safe operation of every nuclear power plant. Elements of these system may periodically fail due to some reasons. This article contains annual failure statistics analysis of Russian nuclear power plants between years 1992 and 2018. It was also assessed the effectiveness of nuclear power plant safety improvement measures, aimed on failure elimination.

Обоснование безопасности и эффективности эксплуатации АЭС при выборе параметров технического обслуживания и ремонта

Обоснование безопасности и эффективности эксплуатации АЭС при выборе параметров технического обслуживания и ремонта

Best estimate plus uncertainty analysis of small break loss of coolant accidents with different break sizes for CPR1000

The best estimate plus uncertainty (BEPU) is an advanced method which has been widely adopted in nuclear power plant safety analysis. In order to investigate the transient behaviors of the pressurized water reactor CPR1000 during the cold leg small break loss of coolant accident (SBLOCA) at different break sizes, BEPU analysis of the SBLOCA was carried out in this study, and the sizes of the break were 0.2%, 0.4%, 0.6%, 0.8% and 1.0% of the cross-sectional area of the cold leg, respectively. The minimum collapsed level (MCL) in the active core section was determined to be the figure of merit (FOM). The base case calculations, uncertainty propagation calculations and sensitivity analysis calculations were performed with the NUSOL-SYS code. Based on the results, the transient behaviors were analyzed, the lower 95/95 tolerance limits of the MCL at different break sizes were quantified, and the important input parameters were determined.

Study on multi-unit level 3 PSA to understand a characteristics of risk in a multi-unit context

Since the Fukushima Daiichi accident in 2011, concerns for the safety of multi-unit Nuclear Power Plant (NPP) sites have risen. This is because more than 70% of NPP sites are multi-unit sites that have two or more NPP units and a multi-unit accident occurred for the first time. After this accident, Probability Safety Assessment (PSA) has been considered in many countries as one of the tools to quantitatively assess the safety for multi-unit NPP sites. One of the biggest concerns for a multi-unit accident such as Fukushima is that the consequences (health and economic) will be significantly higher than in the case of a single-unit accident. However, many studies on multi-unit PSA have focused on Level 1 & 2 PSA, and there are many challenges in terms of public acceptance due to various speculations without an engineering background. In this study, two kinds of multi-unit Level 3 PSA for multi-unit site have been carried out. The first case was the estimation of multi-unit risk with conservative assumptions to investigate the margin between multi-unit risk and QHO, and the other was to identify the effect of time delays in releases between NPP units on the same site. Through these two kinds of assessments, we aimed at investigating the level of multi-unit risk and understanding the characteristics of risk in a multi-unit context.

Numerical investigation of film boiling heat transfer on the horizontal surface in an oscillating system with low frequencies

Abstract Film boiling is of great importance in nuclear safety as it directly influences the integrity of nuclear fuel in case of accidents involving loss of coolant. Recently, nuclear power plant safety under earthquake conditions has received much attention. However, to the best of our knowledge, there are no existing studies reporting film boiling in an oscillating system. Most previous studies for film boiling were performed on stationary systems. In this study, numerical simulations were performed for saturated film boiling of water on a horizontal surface under low frequencies to investigate the effect of system oscillation on film boiling heat transfer. A coupled level-set and volume-of-fluid method was used to track the interface between the vapor and liquid phases. With a fixed oscillation amplitude, overall, heat transfer decreases with oscillation frequency. However, there is a frequency region in which heat transfer remains nearly constant. This lock-on phenomenon occurs when the oscillation frequency is near the natural bubble release frequency. With a fixed oscillation frequency, heat transfer decreases with oscillation amplitude. With a fixed maximum amplitude of the additional gravity, heat transfer is affected little by the combination of oscillation amplitude and frequency.

Study on failure mechanism of prestressed concrete containments following a loss of coolant accident

A loss of coolant accident (LOCA) is considered as a severe accident in the safety of nuclear power plant. LOCA has significant negative influences on the containment structures. Because the characteristics of temperature variation are instantaneous, the thermal field in the inner wall of the containment has a remarkably non-uniform distribution, which leads to complex mechanical behavior. In this study, the distribution of the thermal field of a prestressed concrete containment vessel (PCCV) at different times under a LOCA is analyzed. On this basis, the mechanical behavior of the PCCV under the coupling effect of thermal load and internal pressure are systematically investigated, and the failure mechanism and failure mode of the structure are revealed. The results show that the structure fails with a large deformation near the largest hole and evident expansion at the dome considering the thermal load. However, the deformation develops deficiently under the high temperature of the LOCA. Finally, design suggestions are proposed for a containment under thermal–pressure coupling action.

Formuler l’action publique en termes de tests. Les stress tests européens comme réponse aux crises financières et nucléaires

In Europe, stress tests are used by institutional leaders as a tool for regulating the banking and nuclear sectors. They were first used in 2009 in the banking sector in the wake of the sovereign debt crisis and in 2011 in the nuclear sector following the Fukushima accident. Where in the former case their purpose is to convince investors of bank stability, in the latter they seek to reassure the European public as to the safety of nuclear power plants. Both explicitly aim to offer an “objective and transparent” evaluation of technical objects. This evaluation defines “crisis” and the ways for resolving it in well-delimited terms. Stress tests result in an expansion of the domain of European institutional intervention, which is centralized in the case of banking and distributed in that of nuclear power. While doing so, they also eliminate alternate ways of conceptualizing crisis and possible responses thereto.

A Robust Cybersecurity Solution Platform Architecture for Digital Instrumentation and Control Systems in Nuclear Power Facilities

Most nuclear power plants (NPPs) under construction or under design are expected to deploy largely digital instrumentation and control (I&C) systems. Current fleets are increasingly looking toward converting to digital I&C systems due to the advantages of precise control, economic operation, and ease of procurement over conventional analog I&C systems. With all the benefits digital I&C systems bring, challenging cybersecurity concerns are introduced as well. Cyberattacks targeted at industrial control systems have grown in both frequency and capability in recent years. Despite efforts to air-gap digital I&C systems, NPPs can be vulnerable to these cyberattacks, as evidenced by recent cyber incidents at nuclear facilities. Cybersecurity of NPPs should be addressed in three complementary thrusts: cyberattack prevention, detection, and response. Considering the requirements of digital I&C systems, we propose a novel cybersecurity solution platform that consists of a data collection and extraction system, a multilayer cyberattack detection system, a cause analysis system with dynamic risk assessment, a cyberattack response system, and a main control room display system. This architecture also promotes cooperation between information technology experts and the operation technology team to improve cybersecurity by integrating process data together with traditional host system and network data in a unified platform. This paper presents the proposed cybersecurity architecture and demonstrates its efficacy with a simulated cyberattack on a cyber-physical system testbed. Together with traditional intrusion prevention methods and rule-based intrusion detection systems, this platform provides a solution for prevention, detection, and response to cyberattacks that is congruous with the defense-in-depth strategies of other NPP safety and security systems.

Fully developed turbulent convection of Lead Bismuth Eutectic in the elementary cell of the NACIE-UP Fuel Pin Bundle

One of the requirements for achieving high levels of safety in fourth generation nuclear power plants, is that core thermal-hydraulics can be simulated numerically at a good level of accuracy. To this aim, detailed validation of turbulent heat transfer models needs to be carried out and discussed, in relevant flow and heat transfer configurations, and for low Prandtl-number fluids.The focus of this research is the turbulent, fully developed convection in a heated rod bundle with a triangular arrangement, and a pitch-to-diameter ratio P/D=1.4. Statistics derived from a set of Direct Numerical Simulations (DNS) at the moderate Reynolds number of Re=8290 are compared to Reynolds Averaged Navier-Stokes (RANS) solutions where the closure is provided by the two-equation model SST k-ω and by the Reynolds-Stress Model. Comparisons are drawn for forced flow and in mixed convection conditions (Ri≈0.25). Global quantities extracted from experiments performed in very similar conditions are also compared against the numerical results. Profiles of the turbulent Prandtl number about the unit flow cell are also displayed and discussed.This work clarifies through detailed comparison that if, on the one hand, only the Reynolds-Stress model is able to reasonably capture important flow features like secondary flow components, on the other hand also the SST k-ω two-equation model considered is acceptably accurate in predicting the integral quantities of interest.

Temperature field in gas-cooled reactor core in transient conditions under different approaches to mass flow profiling

Positive effect of profiling the gas-cooled reactor core within the framework of the GT-MHR project was investigated in (Podgorny and Kuzevanov 2017, Kuzevanov and Podgorny 2017, 2018). The necessity arises to supplement already implemented analysis of equilibrium conditions of core operation with investigation of effects of profiling on the temperature field in transient modes of reactor core operation. The present paper is dedicated to the investigation of development of transients in gas-cooled nuclear reactor core subject to the implementation of different principles of core profiling. Investigation of transients in reactor core represents complex problem, solution of which by conducting direct measurements is beyond the resources available to the authors. Besides the above, numerical simulation based on advanced CFD software complexes (ANSYS 2016, 2016a, 2016b, Shaw 1992, Anderson et al. 2009, Petrila and Trif 2005, Mohammadi and Pironneau 1994) is also fairly demanding in terms of required computer resources. The algorithm for calculating temperature fields using the model where the reactor core is represented as the solid medium with gas voids was developed by the authors and the assumption was made that heat transfer due to molecular heat conductivity can be described by thermal conductivity equation written for continuous medium with thermal physics parameters equivalent to respective parameters of porous object in order to get the possibility of obtaining prompt solutions of this type of problems. Computer code for calculating temperature field in gas-cooled reactor in transient operation modes was developed based on the suggested algorithm. Proprietary computation code was verified by comparing the results of numerous calculations with results of CFD-modeling of respective transients in the object imitating the core of gas-cooled nuclear reactor. The advantage of the developed computer code is the possibility of real-time calculation of evolution of conditions in complex configurations of gas-cooled reactor cores with different channel diameters. This allows using the computer code in the calculations of transients in loops of reactor facility as a whole, in particular for developing reactor simulators. Results are provided of calculations of transients for reactor core imitating the core of gas-cooled nuclear reactor within the framework of GT-MHR project performed for different approaches to profiling coolant mass flow. Results of calculations unambiguously indicate the significant difference of temperature regimes during transients in the reactor core with and without profiling and undeniable enhancement of reliability of nuclear reactor (Design of the Reactor Core 2005, International Safeguards 2014, Safety of Nuclear Power Plants 2014) with profiling of coolant mass flow in the reactor core as a whole.

Application of the FDTD-based simulation code VSTL REV to the lightning surge analysis of a nuclear power plant

The installation of effective lightning protection is necessary to protect power equipment and sensitive electronic devices and guarantee human safety in nuclear power plants. To evaluate the effectiveness of countermeasures against lightning, it is necessary to predict lightning surge phenomena. Recently, the finite-difference time-domain (FDTD) method, which is one of the full-wave numerical approaches, has been applied for analyzing electromagnetic transient phenomena in three-dimensional structures such as lightning protection systems and buildings or grounding structures such grounding grids. In this study, we model buildings and grounding structures of a nuclear power plant using the FDTD-based surge simulation code VSTL REV, and we study the effect of soil resistivities on the step voltages around the reactor building and the voltages induced on grounding buses drawn into an auxiliary building and the metal sheaths of coaxial cables in the case of a direct lightning strike to the nuclear power plant.