Nuclear Power Plant Systems Research Articles

Evaluation of LBB characteristics of candidate materials for main steam line piping in Korea nuclear power plants

Recently, efforts have been made to apply the leak-before-break (LBB) design concept to the main steam line (MSL) piping in secondary systems of Korea standard nuclear power plants (KSNPs). In this study, the microstructures, mechanical properties, and consequential LBB margins of five candidate materials (SA335 Gr.P11/P22, SA508 Gr.3 Cl.2, SA508 Gr.3 Cl.1, and SA508 Gr.1A steels) were evaluated, and strong/weak points of candidate materials for applying MSL piping were analyzed through the comprehensive analysis including microstructure, mechanical properties, LBB margin, usage history, economics, and so on. SA508 Gr.1A steels, which has been used extensively as a piping material in NPPs and has a low fabrication cost, had a good combination of yield strength and fracture toughness. They had LBB margin of 1.11–1.21. SA508 Gr.3 Cl.1 steel had an excellent combination of yield strength and fracture toughness with an LBB margin of 1.64. SA508 Gr.3 Cl.2 steel had the highest LBB margin (1.84) because it had the highest strength; however, it had low fracture toughness and low upper shelf energy (USE). The effect of yield strength on the LBB margin was considerably higher than the effect of fracture toughness. SA508 Gr.3 Steels have high LBB margin owing to the high yield strength, but they have never been used for piping materials in NPPs before. SA335 Gr.P11 and P22 had the LBB margin of 1.1 and 1.23, respectively. However, the elongation and J-R fracture toughness of SA335 Gr.P11 and P22 decreased significantly at 286 °C owing to dynamic strain aging (DSA).

Determination of the main parameters of the process of boric acid droplet entrainment from the WWER reactor in case of an accident

The article presents the results of a calculation study of the process of mass transfer of boric acid from a WWER reactor in the event of an accident. An analysis of the literature data in this field of study was carried out. As a result, it was found that studies of the mass transfer of boric acid were performed with parameters not typical for the emergency operation of passive safety systems of nuclear power plants with WWER. The results of the calculation of the main parameters in the WWER reactor in the event of an accident with a rupture of the main circulation circuit are shown. The calculation analysis allowed us to determine that the maximum diameter of the droplets entrained by the steam is in the range 0.11-0.23 mm.

Dynamic Modeling of Reactor Protection System in Nuclear Power Plant for Reliability Evaluation Based on State Transition Diagram

Reliability assessment of a digital dynamic system using traditional Fault Tree Analysis (FTA) is difficult. This paper addresses the dynamic modeling of safety-critical complex systems such as the digital Reactor Protection System (RPS) in Nuclear Power Plants (NPPs). The digital RPS is a safety system utilized in the NPPs for safe operation and shut-down of the reactor in emergency events. A quantitative evaluation reliability analysis for the digital RPS with 2-out-of-4 architecture using the state transition diagram is presented in this paper. The study assesses the effects of independent hardware failures, Common Cause Failures (CCFs), and software failures on the failure of the RPS through calculating Probability of Failure on Demand (PFD). The results prove the validity of the proposed method in analyzing and evaluating reliability of the digital RPS and also show that the CCFs and longer detection time are the main contributions to the PFD of digital RPS.

Loose parts detection method combining blind deconvolution with support vector machine

The loose parts detection in low signal–noise-ratio (SNR) environments is investigated, which is frequently encountered in safety monitoring of the primary coolant system in nuclear power plant. The objective function method (OFM) based on high-order statistics is firstly applied to the impact response extraction under strong noise background. The parameter optimization of OFM is carried out to improve the solution speed and signal extraction ability. The approach is verified through the steel ball drop experiment. The results show that, compared with the wavelet algorithm, the OFM can suppress the noise more effectively and thus extract the impact response signal submerged in noise better. Secondly, a loose parts detection approach is developed by combining the OFM with the support vector machine. The detection results based on the steel ball drop experiments show that the present method can effectively identify the weak impact response and has good anti-false positive and anti-missing ability.

A probabilistic model-based diagnostic framework for nuclear engineering systems

A fault diagnostic framework was investigated in this study for applications in thermal–hydraulic systems of nuclear power plants. The proposed framework consists of quantitative model-based diagnosis, statistical change detection and probabilistic reasoning. The use of physics-based diagnostic models provides high detection sensitivity and allows noise and measurement uncertainty to be incorporated robustly. Performance-related parametric models for each component are constructed based on first principles. Numerical model residuals are generated using the concept of analytical redundancy. Statistical change detection methods are employed to detect non-zero residuals in the presence of uncertainty. The diagnosis task is performed using Bayesian inference to detect and localize possible faults. Application to a single-phase heat exchanger for demonstration showed that the proposed probabilistic framework can provide improved results in comparison with traditional approaches while remaining less sensitive to false alarms in the presence of measurement and modeling uncertainty.

A data-driven framework for identifying important components in complex systems

Complex technical infrastructures are systems of systems characterized by hierarchical structures, made by thousands of mutually interconnected components performing different functions. Given their complexity, it is difficult to derive their functional logic using traditional risk and reliability analysis methods based on engineering knowledge. In this work, we propose to address the problem in an innovative way that makes use of the large amount of data available from monitoring those systems. Specifically, we develop a data-driven framework to identify the critical components of a complex technical infrastructure. The criticality of a component with respect to the safe/failed state of the infrastructure is assessed considering a feature selection technique which employs Random Forest (RF) classification and a feature importance score. The proposed data-driven framework is applied to a nuclear power plant system and a synthetic case study, which mimics the complexity of a technical infrastructure.

An experimental observation of the effects of submicron- and micron-sized mesoporous silica particles on the critical heat flux

Mesoporous particles have attracted attention for nuclear power plant systems, especially as part of a severe accident mitigation strategy to prevent the leakage of radioactive materials. For practical applications to nuclear systems, evaluations of the critical heat flux (CHF) is necessary to investigate the effects of mesoporous silica particles on the thermal safety margin. In this study, thus, pool boiling experiments were conducted using micron-sized porous silica particles to evaluate the effects of porous silica particle deposition on the CHF. Porous particles (0.5, 1, 2 μm) with pore sizes of 2 and 4 nm were prepared. Experiments were conducted under atmospheric pressure and low particle concentrations, ranging from 0.1 to 10 ppmv. Static contact angle measurements and Scanning Electron Microscopy (SEM) images were used for post-experimental surface analyses. As a result, unique CHF trends were observed as a function of the particle size, particle shape, pore size, and concentration. At relatively high concentration of 10 ppmv, CHF enhancement (up to 200 % of pure water) was observed for all particles, except for the 2 μm-sized particles with 4 nm of pores where CHF deterioration by 78 % of pure water was observed. In addition, at low concentration of 0.1 ppmv, CHF values appeared to be lower than those of pure water only for largest porous particles of 2 μm in diameter, not for amorphous ones. A new hypothesis was proposed to interpret the unique CHF trends considering microlayer evaporation theory, porous particle characteristics, and the effects on the particle deposition and the CHF.

Effectiveness analysis of hydrogen control system in AP1000 nuclear power plant under LB-LOCA

The operational data and experience of AP1000 under accidents is insufficiency, it is necessary to conduct an effective analysis of the hydrogen control system to improve the safety of nuclear power plants. Three-dimensional computational fluid dynamics code Gasflow is used to study the severe accident caused by the cold leg LB-LOCA in the 1# steam generation compartment, and gravity injection failed. The hydrogen diffusion process, hydrogen distribution, and hydrogen risk in containment of the AP1000 nuclear power plant is analyzed. According to analysis results, hydrogen diffuse from source term compartment to containment dome space, and then to the bottom, obvious layered distribution is formed in containment. AP1000 hydrogen control system can effectively reduce the hydrogen concentration, eliminate the risk of flame acceleration in containment and deflagration to detonation transition in most compartments, but deflagration to detonation transition may occur in the source term compartment.

Stability and steady state analysis of control and safety systems of Nuclear Power Plants

Control Systems are the first layer of safety in Nuclear Power Plants, the failure of which invokes Safety Systems. The failure of safety systems can lead to radiation exposure to the public. Therefore, it is essential to ensure the stability of such systems. This paper proposes an effective methodology for stability and steady state analysis of control and safety systems of Nuclear Power Plants. The methodology includes Petri net modeling and analysis of its dynamic behavior. Despite of its simplicity, it is very accurate. The application of the proposed techniques is shown and validated on a reactor protection system, which is under operation in six Nuclear Power Plants.

Availability analysis of safety-critical and control systems of NPP using stochastic modeling

Non-functional requirements are essentially important and play vital role for applications ranging from safety-critical systems (SCS) to simple gaming applications to ensure their quality. SCS demands not only for safe and reliable systems but systems those remain safe and available while under attacks. Availability analysis approaches include, but are not limited to cluster technique, Markov based chain models, Reliability Block Diagrams (RBD), Fault Tree Analysis (FTA) and Flow Network. The classical approaches fail to account for the comprehensive and accurate analysis of the diverse characteristics such as temporal behavior of systems, uncertainty in system behavior and failure data, functional dependencies among components and multiple failure modes for components or systems. This paper presents a novel approach for the availability analysis of a Digital Feed Water Control System (DFWCS) of nuclear power plant, which considers the maintenance and repair of the main-steam safety valves. The approach will be useful when no operational profile data is available for that. The system has been modeled using Stochastic Petri Net capturing all the system requirements along with the partial failures of its subsystems and common-cause failures and analyzed using TimeNet tool. The proposed methodology proves to be efficient and overcomes the limitations of the traditional approaches and the Markov model approach as it computes the state-transition probabilities, rather than assuming or qualitatively assessing it.

Capabilities of State-of-the-Art Computational Tools to Justify the Safety of Nuclear Energy

The main aspects limiting the widespread development of nuclear energy—the consequences of possible severe accidents and the problem of radioactive waste management—are considered. It is shown that modern computational tools and digital technologies can successfully solve the problems of substantiating and ensuring the safety of nuclear facilities, including modeling the states and processes occurring in a reactor installation, and the entire complex of nuclear power plant (NPP) systems, the spread of contaminants in emergency situations, the choice and justification of solutions on decommissioning nuclear and radiation hazardous facilities, and the disposal of radioactive waste (RW).

Calculation and Analysis of Stable Operation of Feed water Pumps for Floating Nuclear Power Stations under Marine Conditions

Feed water pumps play an important role in floating nuclear power plant system equipment, and their safe and reliable operation directly affects the safety and reliability of the entire power plant. The movement status of floating nuclear power plants under sea conditions will change, and additional inertial forces are generated on the pipelines of the water supply system, which may cause cavitation of the feed water pump. This paper deduces the additional force on the fluid at the inlet pipe of the feed pump under the marine environment, establishes a mathematical model of the fluid flow at the inlet pipe of the feed pump under the influence of ocean conditions, and calculates the stable operation of the feed pump under the ocean conditions once in a century according to the layout plan. The requirements of effective cavitation allowance and necessary cavitation allowance provide theoretical calculation basis for design and layout optimization of feed pumps for floating nuclear power plants.

Preventing cascading failure of electric power protection systems in nuclear power plant

Cascading failure is the main cause of large blackouts in electrical power systems; this paper analyzes a cascading failure in Hanbit nuclear power plant unit two (2) caused by a circuit breaker (CB) operation failure. This malfunction has been expanded to the loss of offsite power (LOOP). In this study, current practices are reviewed and then the methodologies of how to prevent cascading failures in protection power systems are introduced. An overview on the implementation of IEC61850 GOOSE messaging-based zone selective interlocking (ZSI) scheme as key solution is proposed. In consideration of ZSI blocking time, all influencing factors such as circuit breaker opening time, relay I/O response time and messages travelling time in the communication network should be taken into account. The purpose of this paper is to elaborate on the effect of cascading failure in NPP electrical power protection system and propose preventive actions for this failures. Finally, the expected advantages and challenges are elaborated.

CFD Investigation With Variable Orientation of Orifice at Throat in Self-Priming Venturi Scrubber

There is wide application of venturi scrubber in Containment Filtered Venting System (CFVS) of Nuclear Power Plants (NPP) in order to eliminate the gaseous contaminants from polluted gas in Reactor Core of nuclear power plant. The different orientations of orifice at particular angle are used in same and reverse direction of flow respectively. Computational Fluid Dynamics (CFD) results of these arrangements are compared with the orifice pipe which is normal to the direction of gas flow. The pressure drop and velocity at throat affect the performance of venturi. A saleable software “ANSYS WORKBENCH” tool has been used for this investigation effort. Velocity field of gas flow is obtained by using K-Є turbulence model. The momentum conservation principle is used to calculate Gas velocity distribution. Euler-Euler approach is used to recognize performance of fluid dynamics inside the venturi. Pressure drop and liquid to gas ratio for various mass flow rate of gas is determined by CFD analysis of Self priming Venturi Scrubber.

A further study on fracture resistance evaluation of nuclear pipes using CP specimens

The leak-before-break (LBB) concept has been widely used to design the piping system in nuclear power plant, which is based on the fracture resistance (J-R) curves of cracked nuclear pipes. However, accurate fracture resistance evaluation of nuclear pipe is always a tough task. Full-scale nuclear pipe tests are costly and difficulty, and the specimens listed in test standards (e.g. ASTM E1820) are usually too conservative. Thus, compact pipe (CP) specimen has been developed to evaluate the J-R curves of nuclear pipes with high accuracy but much lower cost than full-scale pipe tests. But, the practical application of CP specimen still asks for more theoretical and experimental research. In this study, the assembly of CP specimen was modified to improve the assembly convenience and ensure a more uniform loading condition. Instead of load-line displacement (LLD), crack mouth opening displacement (CMOD) based J-formulation was proposed to determine the J-R curves of modified CP specimens considering more reliable evaluation of CMOD. In addition, the differences of constraint effect among CT specimen, CP specimens and full-scale pipes are distinguished using equivalent plastic strain (εp) isoline regions, which agrees quite well with the results of fracture resistance tests.

Comparison of bi-directional multi-wave alternating magnetic field effect on ferromagnetic nanofluid flow in a circular pipe under laminar flow conditions

Nanofluids have been attracting huge attention because of their heat transfer enhancement capabilities. Furthermore, magnetic field effect has been being researched recently. By reason of further heat transfer enhancement potential, constant and alternating magnetic fields have been utilized in the present work. Forced convection heat transfer of Fe3O4/water nanofluid flow in a straight pipe under constant and alternating magnetic field effect has been investigated experimentally. Experiments were performed under laminar flow regime (1122 < Re < 2124) and constant heat flux was applied externally on the pipe surface. It is aimed to study effect of different parameters such as Reynolds number, volume concentration of nanoparticle (0 ≤ ϕ ≤ 0.05), constant magnetic field (B = 0.3 T), alternating magnetic field with different wave types (sinus, square and triangle) and different frequencies (2, 5 and 15 Hz) of alternating magnetic field on the convective heat transfer. Experimental results showed that the constant magnetic field offers 13% convective heat transfer enhancement compared to the absence of a magnetic field. On the other hand, the alternating magnetic field increases the convective heat transfer in the pipe up to 35%. Furthermore, lower frequencies of the alternating magnetic field have been more effective in convective heat transfer enhancement. Square wave type alternating magnetic field steps forward in the aspect of convective heat transfer enhancement rate among the other wave types. The alternating magnetic field applications look promising in the future for increasing energy efficiency, and it can also be implemented in heat exchangers, solar collectors, emergency heat removal systems in nuclear power plants.

Quantitative measures of thoroughness of FBD simulations for PLC-based digital I&C system

Simulation is a widely used functional verification method for FBD programs of PLC-based digital I&C system in nuclear power plants. It is difficult, however, to estimate the thoroughness (i.e., effectiveness or quality) of a simulation in the absence of any clear measure for the estimation. This paper proposes two sets of structural coverage adequacy criteria for the FBD simulation, toggle coverage and modified condition/decision coverage, which can estimate the thoroughness of simulation scenarios for FBD programs, as recommended by international standards for functional safety. We developed two supporting tools to generate numerous simulation scenarios and to measure automatically the coverages of the scenarios. The results of our experiment on five FBD programs demonstrated that the measures and tools can help software engineers estimate the thoroughness and improve the simulation scenarios quantitatively.

Safety and Reliability Assessment of NPP Instrumentation and Control Systems Considering Different Communication Architectures

Safety and reliability assessment of instrumentation and control (I&amp;C) systems used in different safety-critical industries is a responsible and challenging task. Different assessment models recommended by international and national regulatory documents and used by experts worldwide still have disadvantages and limitations. Therefore, studies of assessment model improvements and refinements are essential. This paper proposes that the assessment models be improved by taking into account different architectures of communications both between different systems and within one particular system. In most models, communication lines are considered absolutely reliable, but the analysis performed shows that the communications should be necessarily addressed. Several analytical models are described to assess the reliability of safety-critical systems for nuclear power plants with different communication options.

A Software Fault Tree Analysis Technique for Formal Requirement Specifications of Nuclear Reactor Protection Systems

Rigorous safety demonstration of software, as well as systems, is required when developing digital reactor protection systems in nuclear power plants. Various safety analysis techniques try to identify, analyze, and find remedies for potential hazards at each stage of software development life-cycle. This paper proposes a software fault tree analysis technique for software requirements written in the NuSCR formal specification language. The proposed method can mechanically construct software fault trees and calculate minimal cut-sets, encompassing timing constraints of multi-cycles, from NuSCR formal specifications. We also improved the fault tree construction and analysis tool into “NuFTA 2.0” to cope with multi-cycled executions. The case study with a preliminary version of requirements specification for a Korean nuclear power plant in operation shows the proposed technique’s effectiveness and applicability to other V&Vs such as simulation.

The Effect of Different Branch Angles and Different Branch Pipe Sizes on the Onset Law of Liquid Entrainment

The T-junction is widely used in nuclear power plant systems. In order to ensure the normal operation of nuclear power plant, it is very important to study the onset point of T-junction liquid entrainment. Previous researches on the onset point of liquid entrainment have been lucubrated, but most of them are for small branch pipes. This paper is about the onset law of large branch pipe. In addition, the effect of branch angle and branch size on the onset point of liquid entrainment was analyzed by comparing with other branch pipe sizes in this paper. The onset phenomenon of liquid entrainment is described and the effect of angle on onset is studied in this research. It is found that the effect of branch angle on the onset point of entrainment is consistent with that of other sizes: as the angle between the branch pipe and the horizontal direction decreases, the onset needs greater Frg on condition of the same hb. The reason is that the decrease of the inertial force in the vertical direction acting on the liquid phase by the gas phase as the angle decreases. In addition, by comparing with the onset point law of liquid entrainment with different branch pipe sizes, we found the effect of branch pipe size on the onset is that the smaller branch pipe needs the greater gas velocity when onset occurs. The reason is that there is different flow resistance of different sizes of branch pipe.