Safety-related Components Research Articles

Effect of test-caused degradation on the unavailability of standby safety components

This paper proposes a safety-critical standby component unavailability model that contains aging effects caused by the elapsed time from installation, component degradation due to surveillance tests, and imperfect maintenance actions. An application of the model to a Motor-Operated Valve and a Motor-Driven Pump involved in the HPIS of a VVER/1000-V446 nuclear power plant is demonstrated and compared with other existing models at component and system levels. In addition, the effects of different unavailability models are reflected in the NPP's risk criterion, i.e., core damage frequency, over five maintenance periods. The results show that, compared with other models that do not simultaneously consider the full effects of degradation and maintenance impacts, the proposed model realistically evaluates the unavailabilities of the safety-related components and the involved systems as a plant age function. Therefore, it can effectively reflect the age-dependent CDF impact of a given testing and maintenance policy in a specified time horizon.

A sequential decision problem formulation and deep reinforcement learning solution of the optimization of O&M of cyber-physical energy systems (CPESs) for reliable and safe power production and supply

The Operation & Maintenance (O&M) of Cyber-Physical Energy Systems (CPESs) is driven by reliable and safe production and supply, that need to account for flexibility to respond to the uncertainty in energy demand and also supply due to the stochasticity of Renewable Energy Sources (RESs); at the same time, accidents of severe consequences must be avoided for safety reasons. In this paper, we consider O&M strategies for CPES reliable and safe production and supply, and develop a Deep Reinforcement Learning (DRL) approach to search for the best strategy, considering the system components health conditions, their Remaining Useful Life (RUL), and possible accident scenarios. The approach integrates Proximal Policy Optimization (PPO) and Imitation Learning (IL) for training RL agent, with a CPES model that embeds the components RUL estimator and their failure process model. The novelty of the work lies in i) taking production plan into O&M decisions to implement maintenance and operate flexibly; ii) embedding the reliability model into CPES model to recognize safety related components and set proper maintenance RUL thresholds. An application, the Advanced Lead-cooled Fast Reactor European Demonstrator (ALFRED), is provided. The optimal solution found by DRL is shown to outperform those provided by state-of-the-art O&M policies.

Rapid Analysis of Aristolochic Acids and Aristolactams in Houttuyniae Herba by LC-MS/MS.

Houttuyniae herba, the Saururaceae plant Houttuynia cordata Thunb., has multiple therapeutic effects, including clearing heat, eliminating toxins, reducing swelling, discharging pus, and relieving stagnation. It has a long history as an edible and medicinal plant in China. Phytochemical studies show that the main constituents include volatile oil, flavonoids, and alkaloids. Aristolactam is a major alkaloid with a structure similar to toxic aristolochic acids. However, there has been no systematic study on aristolochic acids and alkaloids in Houttuyniae herba. Therefore, in this study, an LC-MS/MS method was developed to simultaneously detect seven alkaloids and five aristolochic acids in Houttuyniae herba from different origins. Six alkaloids (O-demethyl nornucifrine, N-nornucifrine, aristololactam AII, aristololactam FI, aristololactam BII, cepharadione B) were found and quantitatively determined in 75 batches of samples. Meanwhile, no aristolochic acids or aristololactams were found in Houttuyniae herba at a limit of detection (LOD) of ≤4 ng/mL. The method developed was fully validated in terms of LOD, limit of quantification (LOQ), linearity, precision, accuracy, and stability. These data clarify the content of the above safety-related components in Houttuyniae herba and provide a reference for further research into its safety.

Ageing management of research reactors instrumentation using neutron radiography applications

This paper introduces the applications of neutron radiography and tomography to improve maintenance and aging management programs in research reactors. Egypt second research reactor (ETRR-2) has an advanced neutron radiography/tomography facility and it is used to shoot and test different objects such as concerning antiquities, industry and applied research in various fields. Neutron radiography/tomography (NR/T) techniques offer ideal opportunities for nondestructive testing (NDT) evaluations. It has been possible to benefit from the applications and use of radiography to help improve maintenance of components and systems as well as aging management programs for the ETRR-2. This work proposes successful applications of NR/T to inspect and test different safety and safety related components such solenoid valves and the shielding tube of the defective fuel instrument to predict both damage and edging of these systems. Also, the NR technique monitored successfully the behavior efficiency of an air filtering unit (AFU) for dust removal from ventilation system in the research reactors. Additionally, we measured the capability of protective maintenance procedures to remove impurities or blockage from a stainless steel filter (SSF) used in water treatment system. NR/T techniques are successfully applied and enhanced both maintenance and ageing management programs of ETRR-2 instrumentation, equipment and systems.

Feasibility Study on the In-Service Testing Program Plan for Safety-Related Dampers of Advanced Power Reactor-1400

Abstract In-service testing (IST) program including test and examination has been emphasized to assess the operability of the safety-related components in nuclear power plants. Since the issuance of ASME N511 in 2007 concerning IST for air treatment, heating, ventilating, and air conditioning (HVAC) systems, the IST program plan for those systems should be needed to establish prior to setting up the detailed program. The proper plan can contribute to enhance the effectiveness of allotment of the resources inevitable for implementation of the program and ultimately save operation and maintenance costs. In order to achieve the purposes, this paper presents the methodologies to establish IST program plan for safety-related dampers including their selection process of the dampers to be included in the program and representative template made by applying the methodologies. In addition, the method to identify testing types per damper for part of selected dampers is also provided in this paper.

Simulation and Mathematical Description of Mechanical Shocks Caused by Brake Actuation on Electric Motors

When a brake engages or releases on an electric motor, a mechanical shock is generated. These so-called brake shocks propagate across the motor housing and the motor shaft, affecting safety relevant mechanical and electronic components. The nature of the interference may be irreversible, i.e. mechanical damage, or reversible, e.g. interference of signal measurement or data transmission. Especially component failures or faulty signal values on rotary encoders are undesirable from a safety point of view. Current shock testing procedures are insufficient to simulate real brake shock characteristics and to identify valid shock limits regarding these shocks. In the first part of this paper, the characteristics of brake shocks are presented and compared to pyroshocks with similar characteristics. Furthermore, it shows that the Pseudo-Velocity Shock Response Spectrum (PVSRS) appears to be the best mathematical method to describe the severity of brake shocks with respect to their potential of damaging encoder components or influencing electrical signals. In the second part a testing machine will be introduced, which is able to generate mechanical shocks with comparable characteristics of real mechanical brake shocks for up to several million cycles. During further research, endurance tests shall be performed with the machine to determine the resilience of safety-related components against mechanical brake shocks. The long-term goal is to define scientifically confirmed test criteria for a standardized shock testing procedure to be applied on safety-related components on electric motors. It is intended to include this testing procedure in an international safety-related standard, like IEC 61800-5-3.

Prediction of J-R curves and thermoelectric power evolution of cast austenitic stainless steels after very long-term aging (200,000 h) at temperatures below 350 °C

Cast austenitic stainless steels (CASS) are materials used to fabricate many important safety-related components in the primary circuits of light water reactors since the early 1970’s. The primary circuit, which transports heated water by nuclear reaction to steam generators, is subjected to in-service temperatures between 285 °C and 325 °C. Under these conditions, CASS undergo thermal aging which may significantly affect their mechanical properties and more especially their fracture toughness. From a metallurgical point of view, the changes in mechanical properties are attributable to several solid-state phase transformation processes including the spinodal decomposition of the ferritic phase and the precipitation of G phase. The kinetics of these phase transformations depends primarily on the time and temperature of aging, secondly on the chemical composition of the heat (chromium, molybdenum, silicon and nickel contents) and thirdly on the heat treatments performed during component manufacturing. The prediction of the long-term behavior of CASS is an important industrial issue for nuclear power plant operators. In the early 1980’s, Electricité De France (EDF) engaged an unparalleled laboratory aging program that is still in progress. Presently, some materials have been aged for more than 200,000 h at low temperatures. The results of this program allow the development of prediction models for Charpy-impact energies, J−R curves and thermoelectric power values much more precise than those proposed so far, based on aging at 400 °C and rarely exceeding 30,000 h. These prediction models are described in this paper.

Design and comparative strength analysis of wheel rims of a lightweight electric vehicle using Al6063 T6 and Al5083 aluminium alloys

Purpose: Use of aluminium alloys in critical parts of a vehicle is common since they can combine the two important properties of a material those are being strength and lightweight. The aim in this research is to guide to design process of a wheel with taking example of an electric race vehicle implementation. Design/methodology/approach: In this study, the fatigue strengths of wheels produced for a two-person racing electric vehicle (Demobil09) are evaluated by calculating maximum distortion energy criterion (Von Mises) with Finite Element Analysis. Findings: Aluminium alloy wheels are crucial safety related components and are subjected to static and dynamic loads directly. Using FEA results, the weight and equivalent stress of the wheel are both reduced. So, the energy consumption is also decreased. Modal frequencies of the wheel models are determined. Research limitations/implications: In this paper, the materials analysed are AL6063 T6 and Al5083 aluminium alloys. Different materials can be analysed in future works. Practical implications: This paper is focusing on how to reduce the energy consumption of a two-person electric vehicle concentrating on reducing the weight of vehicle wheels. The vehicle is more technological than mass production cars since it is an electric race car which uses a hub motor, the body and chassis are produced using carbon polymer composites and all electronic units are designed and produced. Although its specialities it has homologated safety equipment like seats and safety belts. Originality/value: All boundary conditions must be analysed in details and a strength analysis must be conducted during design of the wheels for different load cases to ensure the strength of a wheel while keeping the weight as low as possible. In this complex process, this paper can give some clues to designers for strengths and weights of the designs since three different wheel forms are evaluated for reducing energy consumption of the vehicle.

Seismic analysis based structure integrity assessment of steam generator in fast breeder reactor

Abstract Earthquakes, low-probability high-risk events which brings out the deficiency in the design and construction of components and buildings. Many seismic design codes given by various standards aims to reduce the damage during earthquake or to avoid serious accidents. In case of a Nuclear Power Plant, all safety related structures, systems and components are designed to with stand a very low probability of earthquake. This work gives the details of seismic analysis by Time history method for a typical Steam Generator (SG) used in sodium cooled fast breeder reactor. Steam generator of a sodium cooled reactor have sodium in shell side and water in tube side unlike water cooled nuclear reactors, where tube side and shell side is water. The failure of tubes in sodium cooled fast reactor steam generator result in leaking of sodium to water side causing sodium water reaction which is explosive in nature. The large energy release in sodium water reaction will damage the complete structure of SG and connected equipments causing shutdown of the plant. Hence it is very important to ensure the safety of the SG at all operating conditions even during event of severe earthquake. In this work the structural integrity of SG is analysed for events of earthquakes viz. Operating Basis Earthquake (OBE) and Safe Shutdown Earthquake (SSE) by performing seismic analysis by Time history method. The stresses at various critical locations are estimated and checked against limits given by ASME Section –III NC rules

Preliminary Reliability Allocation Method for Fusion Facility Based on Probabilistic Safety Assessment

The construction of ITER indicates that the development of fusion energy has entered the engineering stage. Reliability assurance, which is essential in the design and safety supervision of a new reactor, is an urgent problem to be solved in the process of fusion reactor engineering. But, up to now, the reliability work of a fusion device has not considered the security of the fusion reactor. This paper proposes a method to establish a reliability index for fusion safety–related components. The reliability index is a useful indicator to evaluate the reliability of the system, and it is also the regulatory basis for regulatory authorities. First, this paper gives the recommended values of probability safety goals for a fusion reactor. Second, the reliability requirements for the fusion safety system that meet the safety goals are calculated. In this part, a Probability Safety Assessment (PSA) is adopted to establish the risk models and calculate the undesired consequences of a fusion reactor. Based on the PSA analysis results, a risk-informed approach is used to categorize the plant structures, systems, and components of the fusion reactor as four categories according to their safety significance. Last, the reliability index of the safety-related components is given based on the results of the risk-informed safety categorization and PSA analysis results. The validation of the reliability index system is still being studied, and this work is expected to support the reliability evaluation and safety supervision of a fusion reactor in the future.

Human reliability in non-destructive inspections of nuclear power plant components: modeling and analysis

Non-destructive inspection (NDI) is one of the key elements in ensuring quality of engineering systems and their safe use. This inspection is a very complex task, during which the inspectors have to rely on their sensory, perceptual, cognitive, and motor skills. It requires high vigilance once it is often carried out on large components, over a long period of time, and in hostile environments and restriction of workplace. A successful NDI requires careful planning, choice of appropriate NDI methods and inspection procedures, as well as qualified and trained inspection personnel. A failure of NDI to detect critical defects in safety-related components of nuclear power plants, for instance, may lead to catastrophic consequences for workers, public and environment. Therefore, ensuring that NDI is reliable and capable of detecting all critical defects is of utmost importance. Despite increased use of automation in NDI, human inspectors, and thus human factors, still play an important role in NDI reliability. Human reliability is the probability of humans conducting specific tasks with satisfactory performance. Many techniques are suitable for modeling and analyzing human reliability in NDI of nuclear power plant components, such as FMEA (Failure Modes and Effects Analysis) and THERP (Technique for Human Error Rate Prediction). An example by using qualitative and quantitative assessesments with these two techniques to improve typical NDI of pipe segments of a core cooling system of a nuclear power plant, through acting on human factors issues, is presented.

An investigation on effects of spot face geometry on fatigue life of drum brake’s lever

Heavy commercial vehicles, which used for construction works, mostly have drum brakes. There are three types of drum brakes named S cam brakes, Z cam brakes and Wedge brakes, which are safety related components for operation of the vehicle. Fatigue life of brake components is critical. One of the most important component is lever which transfers actuation force to camshaft. In this study, spot face geometry of the lever was examined. FEA model multilinear material model was prepared using ANSYS. Two types of specimens were analyzed by static structural FEM under load of 15 kN. Effects of this three-dimensional stress analysis on fatigue life compared experimentally. All the experimental results were plotted and explained by using Weibull Analysis which is one of the widely used lifetime distributions in reliability engineering.

Fluid-elastic instability evaluation for reactor vessel internals with structural interaction

Provided fluid velocity reaches a specific threshold value, the vibrational amplitude of the corresponding structure increases rapidly and becomes unstable. Even though nuclear industrial concerns have been focused primarily on heat exchangers, steam generators and pipes, evaluation of flow-induced vibration is also required for safety-related components such as reactor vessel internals (RVIs). For this matter, the fluid-elastic instability of RVIs in a typical 1000 MWe nuclear power plant (NPP) was examined by enhanced methods taking into account numerical capability and uncertainties. First, computational fluid dynamics (CFD) analyses of the whole geometry were conducted to assess the flow characteristics with and without consideration of fuel assembly under various NPP conditions. Based on the resulting data, upper guide structure (UGS) was selected as the most critical among five RVI subcomponents. Second, detailed modal analyses of the UGS and attached multiple tubes were performed using both added mass and fluid-structure interaction models to find out natural frequencies and corresponding mode shapes, and to investigate effect of a flooded structure. Third, fluid-elastic instability was evaluated based upon the results from the CFD and modal analyses, and compared with a stability diagram. Thereby, it was confirmed that the RVIs considered in this study had sufficient safety margin against the fluid-elastic instability under representative conditions. Moreover, the calculated stability ratios increased when either the simple model was employed without consideration of the fuel assembly as the current common practice or fluid-structure interaction effect was incorporated.

Operator Support System Design forthe Operation of RSG-GAS Research Reactor

The components of RSG-GAS main control room are facing the problem of material ageing and technology obsolescence as well, and therefore the need for modernization and refurbishment are essential. The modernization in control room can be applied on the operator support system which bears the function in providing information for assisting the operator in conducting diagnosis and actions. The research purpose is to design an operator support system for RSG-GAS control room. The design was developed based on the operator requirement in conducting task operation scenarios and the reactor operation characteristics. These scenarios include power operation, low power operation and shutdown/scram reactor. The operator support system design is presented in a single computer display which contains structure and support system elements e.g. operation procedure, status of safety related components and operational requirements, operation limit condition of parameters, alarm information, and prognosis function. The prototype was developed using LabView software and consisted of components structure and features of the operator support system. Information of each component in the operator support system need to be completed before it can be applied and integrated in the RSG-GAS main control room.

Verification and validation of explicit moving particle simulation method for application to internal flooding analysis in nuclear reactor building

ABSTRACT The Fukushima nuclear accident raised the importance of flooding study in nuclear reactor buildings. It is known that the external flooding can be attributed to natural causes, while the internal flooding is caused by the piping ruptures, tank failure or the actuation of fire suppression systems. The building flooding process can damage the safety-related components and systems, which needs to study carefully. In order to simulate this phenomenon which is accompanied by complex flow with free surface, a particle method based on Lagrangian approach named explicit moving particle simulation (EMPS) method, is employed in this analysis. Verification and validation analyses are carried out. The verification problems are a hydrostatic analysis and a water spreading to investigate the differences of the particle wall and polygon wall boundary models, while the validation studies of two experiments of dam-break induced flooding show good agreements. Afterwards, the internal flooding process in AP1000 is simulated by assuming a break of the in-containment refueling water storage tank as an example. The results achieved so far indicate that the EMPS method is capable to simulate the internal flooding process in the nuclear reactor buildings.

Research on the Cross Section Precision of High-strength Steel Tube with Rectangular Section in Rotary Draw Bending

For the demand of energy conservation and security improvement, high-strength steel (HSS) is increasingly being used to produce safety related automotive components. However, cross-section distortion occurs easily in bending of HSS tube with rectangular section (RS), affecting the forming precision. HSS BR1500HS tube by rotary draw bending is taken as the study object and a description method of cross-section distortion is proposed in this paper. The influence on cross-section precision of geometric parameters including cross-section position, thickness of tube, bend radius etc. are studied by experiment. Besides, simulation of the rotary draw bending of HSS tube with rectangular section by ABAQUS are carried out and compared to the experiment. The results by simulation agree well with the experiment and show that the cross-section is approximately trapezoidal after distortion; the maximum of distortion exists at 45 ∼ 60° of the bending direction; and the absolute and relative distortion values increase with the decreasing of tube thickness or bending radius. Therefore, the results can provide a reference for the design of geometric parameters of HSS tube with rectangular section in rotary draw bending.

Sensor Integration in Castings Made of Aluminum - New Approaches for Direct Sensor Integration in Aluminum High Pressure Die Casting

The use of sensors for detection, measurement and evaluation of mechanical and thermal loads is well known and essential for the implementation of »Structural Health Monitoring« (SHM). For this, sensors are mainly used for condition monitoring of mechanical loads and their impact to the castings state, which is a decisive advantage for safety-related components. The use of sensors on the surface of metallic components, in particular of cast metal components made of aluminum, is still limited to the use of strain gauges. They are usually applied on the surface of the cast metal components and get fixed by adhesives. The idea is, to integrate the sensors directly during the aluminum casting process. Since integrated sensors are naturally protected against chemical and mechanical influences, furthermore the load can be measured directly at the point of interest inside the component. Measurement data can be recorded and provide a good data basis for future calculations and dimensioning of components, which is known as »Data Mining« or »Industrial Data Space«. New technology and material combinations, which allow the fabrication of sensors capable of withstanding force and temperature during the integration process in aluminum casting, are investigated.In this paper, the design and fabrication of a strain gauge printed on an aluminum sheet is shown. These sensor sheets get integrated in aluminum during high pressure die casting (HPDC) in a way that a specimen is build up. The specimen is characterized in a fatigue bending test and the sensor data was read permanently during this test. It is shown that the new approach with printed thick film sensors on aluminum substrate sheets works properly to withstand the heavy thermal conditions during high pressure die casting. The fabricated sensor is able to sense the mechanical tiredness and detects the fatigue of the metal matrix. This a first step to use such material integrated sensors in structural health monitoring applications.

Ultrasonic Inspection and Data Analysis of Glass- and Carbon-Fibre-Reinforced Plastics

Non-destructive testing (NDT) helps to find material defects without having an influence on the material itself. It is applied as a method of quality control, for online structural health monitoring, and for inspection of safety related components. Due to the ability of automation and a simple test setup ultrasonic testing is one major NDT technique next to several existing options. Whereas contact technique allows the use of higher frequencies of some MHz and phased array focusing, air-coupled ultrasonic testing (ACUT) shows different advantages. Most significant for ACUT is the absence of any coupling fluid and an economical test procedure respective time and costs. Both contact technique and ACUT have been improved and enhanced during the past years. One important enhancement is the development of airborne transducers based on ferroelectrets, like charged cellular polypropylene (cpp), which makes the application of any matching layers being mandatory in conventional piezoelectric transducers unnecessary. In this contribution we show ultrasonic inspection results of specimens made of carbon- and glass-fibre-reinforced plastic. These specimens include defects represented by drill holes and artificial delaminations of various size and depth. We compare inspection results achieved by using contact technique to those achieved by ACUT. For ACUT, conventional piezoelectric transducers and transducers based on cpp were used, both focused as well as non-focused types. Contact inspections were performed with a multi-channel matrix array probe. Once the inspection data is recorded it can be analysed in order to detect and evaluate defects in the specimen. We present different analysing strategies and compare these regarding detection rate and sizing of defects.

Process Safety Architecture System Neutral Solution Comparison

There are many different solutions of safety systems on the marked. We will see the difference and advantage of the architecture of safety related components also the corresponding definition in the international safety standards. The aim is to give an overview of the safety life cycle from the analysis, realization to the operation phase and the basic knowledge for future decisions of safety related solutions.

Seismic Isolation of Reactor Assembly for a Fixed Base Accelerator Driven System Reactor Building

In the SILER (Seismic-Initiated events risk mitigation in LEad-cooled Reactors) Project, it is interesting to apply seismic isolation technology for the reactor assembly of the fixed base reactor building for ADS (Acceleration Driven System) heavy liquid reactor MYRRHA (Multipurpose Hybrid Research Reactor for High-Tech Application) which contains the most critical safety related components, such as reactor vessel, safe shutdown and control rod mechanisms, primary heat exchangers, primary pumps, spallation target assembly and fuel assemblies, etc. The purpose of this paper is to investigate the possibility of an application of a partial seismic isolation to the safety critical components only, here, the reactor assembly. This paper presents the preliminary analysis results of the isolated reactor assembly and compares these with those of seismic isolated ADS reactor building. The analysis results show the reduction of the seismic acceleration response but the increase of the relative displacement for the reactor assembly. Some safety issues, especially, coolant's incapable covering the reactor core make difficult to apply for the partial seismic isolation of the ADS reactor assembly due to large relative displacement occurring the partial isolation system. Further study on the partial seismic isolation application of the critical safety components are also discussed.