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This study presents the estimation of the electric and magnetic fields level inside a 400 MVA electrical substation. Located in the northeast region of Brazil, the power plant operates with 230 and 69 kV at 60 Hz frequency. The on‐site measurements were performed at previously selected locations covering the whole courtyard and followed the international and national regulatory standards. The electrical and magnetic fields 2‐D contour maps over the entire courtyard are presented, which allow better visualisation and analysis of the exposure levels around all equipment. By using the GoldenSurfer V.13 tool, eight interpolation methods were applied to predict unsampled locations. The plots were made using the method that best fit the sampled dataset after the cross‐validation technique is used. The measurements were taken during six days during a period of 16 h on 741 locations. For this electrical substation, the maps validate the high electric field level close to breakers and switches at 230 kV and high magnetic field close to equipment operating at 69 kV. The comparison between the exposure levels and those established in current guidelines indicate that the electromagnetic environment on this substation is safe for the occupational population as well as for the general public. [ABSTRACT FROM AUTHOR]

The aim of the paper is to perform a comparative analysis of the two power transformation stations (a non-technologized power station and respectively, a re-technologized one), focusing on their description, the main electrical risk factors and the activities of the operational staff. The main risk factors were identified, as well as the circumstances favoring the risk occurrence. The final section includes a summary of the main activities carried out by the operational staff within the two typical power stations. The results will serve developing an innovative management system for occupational safety and health with national applicability. [ABSTRACT FROM AUTHOR]

Utilities' investment in nuclear-generating technology and prior earnings growth were used to predict variations in three measures of plant- operating safety. The pattern of results is consistent with three interpretations: (1) some historically successful utilities that are heavily committed to nuclear power provide leadership; (2) such utilities are gambling with public safety; and (3) a general pattern of purposeful unintended consequences arising from ignorance and self-serving myths combined with the risk propensities of executives and leading to hitherto unidentified threats to public safety. [ABSTRACT FROM AUTHOR]

This paper investigates a virtual power plant (VPP) using the OpenADR 2.0b communication protocol to securely and reliably operate distributed energy resources (DERs) over public Internet infrastructure, providing ancillary services to the transmission system operator (TSO). VPP performance analysis was based on an experimental setup as well as a dynamic behavioural model and numerical simulations. The experimental setup consisted of battery stacks from automated guided vehicles at a harbour container terminal. Within this experimental study, we analysed data exchange utilising the OpenADR protocol over a four-week period between the VPP in Slovenia and the terminal management system in Germany by assessing the selected quality of service parameters – latency (round-trip time), packet loss, retransmissions, bandwidth, amount of traffic and message patterns. The impact of latency on performance of the VPP aggregating battery stacks with other DER types was investigated using numerical simulations. Load profiles for the selected DERs and communication channel models were derived from field measurements. [ABSTRACT FROM AUTHOR]

Highlights • A robust STRUCTure-based 2G-URANS turbulence model is demonstrated. • A rigorous grid convergence study is performed. • The STRUCT turbulence model can adopt URANS like meshes with improved accuracy. • Results from STRUCT model are in good agreement with the DNS data and LES solutions. Abstract Turbulent mixing of hot and cold streams in T-junction geometries is a critical safety issue for power plants, as temperature fluctuations have the potential to lead to high cycle thermal fatigue failures. Due to strong flow deformation in the mixing region, the time-scale separation assumption is not locally respected, and unsteady Reynolds-averaged Navier-Stokes (URANS) models fail to provide sufficient accuracy in predicting the temperature variations. Turbulence resolving methods such as large eddy simulation (LES) can in contrast provide reliable results, while being computationally overly expensive for industrial application. A STRUCTure-based second-generation URANS (2G-URANS) model was recently proposed at MIT, which aims at locally resolving unsteady flow structures; its applicability was demonstrated on a collection of classic flow tests. In the present work, the performance of the STRUCT model is assessed against low Reynolds number DNS data for a squared section T-junction, produced at the THTLAB (Fukushima et al., 2003). The mean and root mean square of the temperature and velocity predictions from the STRUCT solution are evaluated, confirming consistent predictions to the reference DNS results as well as the high-resolution LES simulations. The STRUCT turbulence model demonstrates LES-like accuracy on a URANS-like computational mesh. Spectral analysis of the velocity and temperature variation is also presented and provides consistent prediction of the frequencies of the large turbulent eddies. [ABSTRACT FROM AUTHOR]

Highlights • Based on the distance reject option of the EKNN rule, only normal operating data is needed to construct the CMEW-EKNN model. • An adaptive discounting factor is suggested to tune the EKNN rule, so as to improve the performance of the CMEW-EKNN method. • Based on the framework of evidence theory, the uncertainty of the equipment's operating condition can be described well. Abstract It is essential and challenging to monitor complex industrial processes and thus make an early warning for abnormal conditions, in particular when no fault samples can be observed under unknown uncertainties. To solve this problem, this paper proposes a so-called CMEW-EKNN method, i.e., condition monitoring and early warning method based on the evidential k -nearest neighbor (EKNN) rule in the framework of Evidence Theory. By employing the distance reject option in the EKNN rule, only normal operating data is needed to construct the early warning model. An adaptive discounting factor is adopted to make the early warning boundary adaptive to local distribution characteristics of the training samples, so as to improve both effectiveness and robustness of CMEW-EKNN. Comparisons on two practical applications in power plant demonstrate that the proposed CMEW-EKNN, which adopts the adaptive discounting factor, yields superior fault early warning performance than the PCA-based and FD-kNN fault detection approaches. [ABSTRACT FROM AUTHOR]

Wide-range and frequent operation changes have become a common phenomenon with thermal power plants in the context of fast penetration of intermittent renewable power. Performance monitoring at transient states is more important to the safety and high efficiency of power plants. Data quality is essential for conducting dynamic performance monitoring, and dynamic data reconciliation (DDR) provides great potential to enhance quality of measured data at transient states. In this paper, a moving window based approach to dynamic data reconciliation is proposed for a real coal-fired power plant using high sampling operational data. Firstly, dynamic characteristics of the system are discussed, taking account of the equipment accumulation in DDR problems. Results of case studies indicate that the data accuracy of measured mass flow parameters are enhanced effectively after DDR, and better results are obtained with the increasing time window size. Comparison with steady state data reconciliation approach is also carried out to indicate the enhanced effect of the DDR approach for the performance monitoring of a real power plant. [ABSTRACT FROM AUTHOR]

Nowadays plant monitoring and control of renewable energy sources can take advantage of the use of unmanned aerial technologies. This manuscript aims to develop an image post-processing tool for remote aerial images able to help operation and maintenance operators in photovoltaic (PV) defects detection using light unmanned aerial vehicles (UAVs). In particular, PV systems deployed in the field in the last ten years show often critical behaviour with a range of failures able to compromise the performance and energy yield of the power plants, thus they can greatly benefit of such novel technologies and tools. The described procedures are here tested on real plant data, with different kind of sensors, in order to find out potential advantages in fast fault detection tasks, using an automatic system for PV plant's mapping. The results of this research will be reported in order to provide an understanding of potential impact of image processing techniques based on UAV in the renewable energy sector. [ABSTRACT FROM AUTHOR]

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Safety behavior and human errors are major concerns for nuclear power plant operators. The present study investigated how nuclear power plant operators’ perceived risk influences the quality of their own work performance in terms of safety behavior and errors. In total, 349 operators from two nuclear power plants in China participated in the present study. We found that perceived risk had a negative linear relationship with safety behavior and a quadratic relationship with errors. Leader support played a moderating role in the relationships between perceived risk, safety behavior, and errors. These results supported the job demands–resources model and provided further evidence for the relationship between perceived risk and outcomes related to safety behavior and errors. Our findings suggest that an effective way to address the issue of high perceived risk is to provide a supportive environment. [ABSTRACT FROM PUBLISHER]

The report presents the results of the development and field testing of a mobile test facility based on a helical magnetic cumulative generator (MCGTF). The system is designed for full-scale modeling of lightning currents to study the safety of power plants of any type, including nuclear power plants. Advanced technologies of high-energy physics for solving both engineering and applied problems underlie this pilot project. The energy from the magnetic cumulative generator (MCG) is transferred to a high-impedance load with high efficiency of more than 50% using pulse transformer coupling. Modeling of the dynamics of the MEG that operates in a circuit with lumped parameters allows one to apply the law of inductance output during operation of the MCG, thus providing the required front of the current pulse in the load without using any switches. The results of field testing of the MCGTF are presented for both the ground loop and the model load. The ground loop generates a load resistance of 2-4 Ω. In the tests, the ohmic resistance of the model load is 10 Ω. It is shown that the current pulse parameters recorded in the resistive-inductive load are close to the calculated values. [ABSTRACT FROM AUTHOR]

Managing the aging of digital control systems ensures that nuclear power plant systems are in adequate safety margins during their life cycles. Software is a core component in the execution of control logic and differs between digital and analog control systems. The hardware aging management for the digital control system is similar to that for the analog system, which has matured over decades of study. However, software aging management is still in the exploratory stage. Software aging evaluation is critical given the higher reliability and safety requirements of nuclear power plants. To ensure effective inputs for reliability assessment, this paper provides the required software aging information during the life cycle. Moreover, the software aging management scheme for safety digital control system is proposed on the basis of collected aging information. [ABSTRACT FROM AUTHOR]

System codes and their necessary power plant nodalizations are an essential step in thermal hydraulic safety analysis. In order to assess the safety of a particular power plant, in addition to the validation and verification of the code, the nodalization of the system needs to be qualified. Since most existing experimental data come from scaled-down facilities, any qualification process must therefore address scale considerations. The Group of Thermal Hydraulic Studies at Technical University of Catalonia has developed a scaling-up methodology (SCUP) for the qualification of full-scale nodalizations through a systematic procedure based on the extrapolation of post-test simulations of Integral Test Facility experiments. In the present work, the SCUP methodology will be employed to qualify the nodalization of the AscóNPP, a Pressurized Water Reactor (PWR), for the reproduction of an important safety phenomenon which is the effectiveness of the Core Exit Temperature (CET) as an Accident Management (AM) indicator. Given the difficulties in placing measurements in the core region, CET measurements are used as a criterion for the initiation of safety operational procedures during accidental conditions in PWR. However, the CET response has some limitation in detecting inadequate core cooling simply because the measurement is not taken in the position where the cladding exposure occurs. In order to apply the SCUP methodology, the OECD/NEA ROSA-2 Test 3, an SBLOCA in the hot leg, has been selected as a starting point. This experiment was conducted at the Large Scale Test Facility (LSTF), a facility operated by the Japanese Atomic Energy Agency (JAEA) and was focused on the assessment of the effectiveness of AM actions triggered by CET measurements. The steps of the SCUP methodology are presented: post-test calculation, scaling effect analysis and design effect analysis. The AscóNPP nodalization has shown to be qualified for the simulation of the involved phenomenology. The final step of the work presented here was to adapt the boundary conditions to a more realistic situation taking place in the AscóNPP. CET and Peak Cladding Temperature (PCT) readings were seen to present large differences similarly as it occurred in the ROSA-2 Test 3. The PCT when all the CET readings were credited was 872 K, however, when only the minimum CET reading was credited the maximum temperature in the core rose to 1053 K. [ABSTRACT FROM AUTHOR]

The safety of Nuclear Power Plants (NPPs) is an important issue that is of concern to all including; regulators, operators, and the general public. Assuring the safety of a NPP is a primary objective by all stakeholders. In the wake of accidents in NPPs recorded in the past such as Chernobyl, TMI, and recently in Fukushima, the need to review existing safety system design and operation as well as performing safety verification of these systems as a means of preventing such accidents in the future is necessary. In this study, we present a framework for achieving safety verification of a NPP with emphasis on using cosimulation with reduced error for real time fault propagation analysis based on Fault Semantic Network in which a multiphysics model is mapped unto fault/risk models, and safety/protection systems of NPPs in order to achieve safety verification based on highest risks and previous accidents. A statistical method is used to reduce errors between simulation results and real time data which is illustrated with a case study from literature. © 2015 American Institute of Chemical Engineers Process Saf Prog 35: 53-60, 2016 [ABSTRACT FROM AUTHOR]

This paper proposes a new hybrid data-driven soft measurement modeling method for power plant sensor condition monitoring and fault diagnosis. The method integrates Generalized Regression Neural Network (GRNN), Mean Impact Value (MIV), Partial-Least Squares Regression (PLSR) and B-Spline transformation techniques. First, the relevant parameters are obtained from mechanism analysis and a GRNN model is built to assess the average contribution rate of each independent variable and filter out the main modeling parameters by method of MIV. Then, the main modeling parameters are modeled with a PLSR method based on the cubic B-Spline transformation, which is an effective approach to the nonlinear modeling and multicollinearity problems. The final reliable model is completed to monitor and diagnose the sensors. Taking the active power sensor of a combined cycle generator unit of Siemens V94.3A as an example, the computational result shows that this modeling approach to sensor measurement data fits well in both accuracy and generalization ability under different conditions. Through fault signs and fault diagnosis methods analysis, this model could accurately identify sensor fault types. Most importantly, only a few model parameters need to be saved, and the model has low computation cost and strong robustness. Therefore the model is more suitable in solving the online real-time monitoring and diagnosis problems. [ABSTRACT FROM AUTHOR]

Safety issues in power plants and their direct proportional impact on the safety regulations are increasing significantly nowadays. As a result, fire cases on the way from delivery on the conveyor belt to the final usage in the plant, education of the human resources and improvement of safety regulations are basic key issues. Therefore, in this research I examined the distribution of fire cases and the role of human responsibility through the example of a Hungarian lignite firing power plant by using foreign publications and research. [ABSTRACT FROM AUTHOR]

The article presents a methodology for decision problems involving equipment maintenance policy selection or design changes (including overall and retrofitting) that can be applied in large power plants. The methodology is based on risk and decision-making concepts aiming at balancing the probability of occurrence of a given failure scenario and the consequences of that failure, expressed in monetary values. Using risk-based policy in maintenance organization, planning, and control is very important in order to reduce costs. [ABSTRACT FROM PUBLISHER]

The technical requirements of a system that continually monitors the condition of the safety barriers of nuclear power plants are substantiated. It is shown for the example of the Rostov NPP that power units with VVER which are operating, under construction, or in the design stage can be equipped with a measuring system for monitoring radiation leaks from the steam generators. This improves safety (by not allowing emergency situations) and increases the installed capacity utilization factor (by eliminating unjustified shutdowns of power units) of operating power plants. [ABSTRACT FROM AUTHOR]

Angra 2, the second Brazilian nuclear power plant, began the commercial operation in 2001. The plant is a pressurized water reactor (PWR) type with electrical output of about 1350 MW. In the present work, the thermal hydraulic RELAP5-3D code was used to develop a model of this reactor. The model was performed using geometrical and material data from the Angra 2 final safety analysis report (FSAR). Simulations of the reactor behavior during steady state and loss of coolant accident were performed. Results of temperature distribution within the core, inlet and outlet coolant temperatures, coolant mass flow, and other parameters have been compared with the reference data and demonstrated to be in good agreement with each other. This study demonstrates that the developed RELAP5-3D model is capable of reproducing the thermal hydraulic behavior of the Angra 2 PWR and it can contribute to the process of the plant safety analysis. [ABSTRACT FROM AUTHOR]

The paper aims to conduct a case study on the materials used in electrical contacts associated with the switchgear used in installations falling within the low-voltage electrical station configuration.The landmark of the paper was the major influence that has the contact resistance of the switchgear contacts on their reliability and lifetime. [ABSTRACT FROM AUTHOR]

Control valves are one of the key steam turbine components that guarantee operational safety in a power plant. There are two aerodynamic aspects, which are the current focus for the development of Alstom's valves. One is the reduction of the aerodynamic loss to increase the efficiency of the power plant. The other is operational flexibility, which is increasingly required to react faster to load requirements from the electric grid. This is becoming more important as power generation becomes increasingly decentralized, with a growing contribution from renewable energy sources. For this reason, a fast control loop is required for valve operation, which depends on an accurate linearization of the valve characteristic. In this paper the flow fields in an existing steam control valve have been analyzed and subsequently optimized using CFD techniques. The approach specifically designed for drilled strainers is further illustrated. Following the validation of the baseline design with model testing, an improved diffuser has been designed using CFD analysis and the resulting performance benefit has been confirmed with further testing. The grid frequency support requires control valve throttling. For this reason, an accurate prediction of the linearization table is extremely important to support the required response time limits. Further numerical work has been carried out with various opening positions of the valve, leading to an improved valve linearization characteristic. It is demonstrated that the numerical prediction of the linearization curve agrees very well with data obtained from operating power plants. [ABSTRACT FROM AUTHOR]

The operator of a power block must prove its safety to the state authority. This article deals with the modeling of safety systems using the fault tree analysis method, powered by RiskSpectrum software. Based on real data, the importance analysis of several safety functions is shown. As a result of a quantitative analysis, the influence of several safety system components on the total availability of the protection system is described. The statistics of a number of components that influence 90% of the unavailability of the safety system and the component with the greatest importance measure from each safety function is also shown. [ABSTRACT FROM PUBLISHER]

Highlights: [•] A protection system for vertical probes in closed loop geothermal systems is proposed. [•] In the annulus between probes and protection system there is a liquid. [•] This system leads to an increase in the heat transfer rate. [•] TRCM scheme, with addiction of protection system, is proposed for single U tube borehole. [ABSTRACT FROM AUTHOR]

Twenty soil profile samples and fourteen surface soil samples collected from the vicinity of the lignite-fired power plants in the Megalopolis basin (Greece) were analysed for their natural radionuclide concentration and 137Cs, since fossil fuels are associated with naturally occurring radioactive materials and hence with radiological impact. No significant enhancement of surface soil radioactivity levels in the vicinity of lignite-fired plants was observed. A downcore decreasing trend of 137Cs was observed in a number of cores reflecting its atmospheric origin, whereas the uniform distribution observed in a number of other cores gave information on the mechanical alteration of the soil. The average dose rate value was found to be 63 ± 22 nGy h−1, while the annual average effective dose from the terrestrial gamma radiation was found to be 0.08 ± 0.03 mSv. [ABSTRACT FROM AUTHOR]

Abstract: Analysis of the influence of thermal transients during steam turbine start-up and shut-down, with superimposed creep loading, not only affects the safe operation of power plant but also economic and environmental concerns. As a powerful tool, constitutive material models allow a direct coupling to the finite element method to simulate components with complex structures combined with a parallel evolution of deformation and damage under multiaxial loadings. Constitutive material models were developed in the past two decades up to the threshold of their application. Their preparation and application is limited to a few particularly important cases because of the high computational complexity involved. Their applicability in industrial practice for lifetime calculations depends on the performance of calculation of large finite element models, particularly under cyclic loadings. This paper describes a proposal to solve this problem by developing a sequential local extrapolation based calculation reduction method of internal variables to accelerate the calculation of deformation and lifetime of large components. The effectiveness of this method is shown by recalculation of validation experiments, which were performed with cruciform specimens of a modern 10%Cr forged steel under complex thermo-mechanical service-type loadings. [Copyright &y& Elsevier]

There are four operating nuclear power plant (NPP) units in Finland. The Teollisuuden Voima (TVO) power company has two 840 MWe BWR units supplied by Asea-Atom at the Olkiluoto site. The Fortum corporation (formerly IVO) has two 500 MWe VVER 440/213 units at the Loviisa site. In addition, a 1600 MWe European Pressurized Water Reactor supplied by AREVA NP (formerly the Framatome ANP-Siemens AG Consortium) is under construction at the Olkiluoto site. Recently, the Finnish Parliament ratified the government Decision in Principle that the utilities' applications to build two new NPP units are in line with the total good of the society. The Finnish utilities, Fenno power company, and TVO company are in progress of qualifying the type of the new nuclear builds. In Finland, risk-informed applications are formally integrated in the regulatory process of NPPs that are already in the early design phase and these are to run through the construction and operation phases all through the entire plant service time. A plant-specific full-scope probabilistic risk assessment (PRA) is required for each NPP. PRAs shall cover internal events, area events (fires, floods), and external events such as harsh weather conditions and seismic events in all operating modes. Special attention is devoted to the use of various risk-informed PRA applications in the licensing of Olkiluoto 3 NPP. [ABSTRACT FROM AUTHOR]

The Kaunertal power station has been in operation since 1965. During one of the regular inspections of the facility, it was noticed that the entire pressure shaft and surge tank must be completely renewed due to the long years of wear and tear. The renewal is necessary as a result of the heavy hydraulic loading from power station peak operation and due to the fact that parts of the pressure shaft were constructed in a then unknown landslip and have since been subjected to additional loading from creep movement of the rock mass. The route of the new pressure shaft will run below the identified landslip with a sufficient safety factor, further south than the existing pressure shaft. The implementation of the project with its heart, the steel lined pressure shaft approx. 1, 430 m long with a diameter of 4.3 m, to be constructed inside a PC segmental lining installed with the progress of the double-shield TBM, will take place in 2012 to 2015. Das Kaunertalkraftwerk ist seit 1965 in Betrieb. Im Zuge regelmäßig durchgeführter Anlagenüberprüfungen wurde festgestellt, dass der gesamte Kraftabstieg (Druckschacht und Wasserschloss) infolge seiner langjährigen Beanspruchung umfassend erneuert werden muss. Der Erneuerungsbedarf ist einerseits auf die starke hydraulische Belastung im Rahmen des Spitzenkraftwerksbetriebs zurückzuführen, andererseits auf den Umstand, dass der Druckschacht seinerzeit auf Teilstrecken in einer lokalen, beim Bau nicht erkannten Massenbewegung errichtet wurde und seitdem einer zusätzlichen Beanspruchung aus Kriechbewegungen des Untergrunds ausgesetzt ist. Die Trasse des neuen Druckschachts wird mit ausreichendem Sicherheitsabstand nun unterhalb der identifizierten Massenbewegung und weiter südlich als der bestehende Druckschacht geführt. Die Realisierung des Projekts mit seinem Herzstück, dem gepanzerten ca. 1.430 m langen Druckschacht mit 4,3 m Durchmesser, der innerhalb einer mit Doppelschild-TBM hergestellten Tübbingvorauskleidung errichtet wird, erfolgt in den Jahren 2012 bis 2015. [ABSTRACT FROM AUTHOR]

Abstract: Safety assessment of thermal power plants (TPP) is an important means to ensure the safety of production in thermal power production enterprises. Modern information technology can play an important role in TPP safety assessment. The evaluation of power plant systems relies, to a large extent, on the knowledge and experience of the experts undertaking the task. Case-based reasoning (CBR) is introduced for the safety assessment of TPP since it models expertise through experience management. This paper provides a case-based approach for the Management System safety assessment decision making of TPP (MSSATPP). We introduce a case matching method named CBR-Grey, which integrates the Delphi approach and grey system theory. Based on this method, we implement a prototype of case-based knowledge system (CBRSYS-TPP) for the evaluation decision making of the panel of experts. Our experimental results based on a real-world TPP safety assessment data set show that CBRSYS-TPP has high accuracy and systematically good performance. [Copyright &y& Elsevier]

The article concerns the construction of the Kopswerk II pressure tunnel and associated works. Important details of the construction and construction sequence are described, particularly the logistics of a large site in the high mountains. In addition to a summary of the various drill and blast tunnels, the article concentrates on the TBM drive and the drilling and grouting works in the pressure tunnel. Der Beitrag befasst sich mit den Bauarbeiten für den Druckstollen des Kopswerkes II und dessen Nebenanlagen. Dabei werden die Eckpfeiler der Bauausführung und des Bauablaufes, insbesondere auch die Logistik einer Baustelle im Hochgebirge beschrieben. Neben einem Abriss über die verschiedenen Sprengvortriebe wird vor allem über den TBM-Vortrieb und die Bohr- und Injektionsarbeiten im Druckstollen berichtet. [ABSTRACT FROM AUTHOR]

The damage of unallowable blast induced vibrations on the newly constructed concrete structures in the excavations of the Karoun III project results in a decrease in the safety of its structures. Therefore prediction and control of the vibrations are of importance in the study area. To predict the vibrations in this area, two main approaches were used and the results were interpreted and compared. The vibrations were first predicted using several widely used empirical models. Then, an artificial neural network (ANN) was used. In the ANN model, maximum charge per delay, total charge per round, distance from blast site, direction of firing, blast hole length, number of blast holes, total delay in milliseconds, number of delay intervals and average specific charge were taken into consideration as input parameters. However, the ANN approach showed very high correlation and very small estimation error, the best empirical model resulted in a reliable correlation and an acceptable error of estimation. By considering all aspects the empirical approach was used in practice. Finally a plot based on standard values of vibrations was proposed for the safety of the structures. [ABSTRACT FROM PUBLISHER]

Abstract: Safety in nuclear power plants is one of the main problems in design. Steam generator reliability and performance are serious concerns in the operation of pressurized water reactors. Evidence shows that in the cold collector where tubes are connected to it probability of cracking exists. In this research stress distribution in the collector body at different operation condition of WWER-1000 nuclear power plant was analyzed and the collector stability has been investigated by analytical fracture mechanics and finite element method evaluations. Results show that the stresses in the collector body are in the elastic range and the stress intensity factors are lower than critical values. The analytical calculation shows the stability of semi-elliptical cracks near to expansion region of collector body during operation. [Copyright &y& Elsevier]

Abstract: The regulation of nuclear power plant (NPP) is evolving in a direction to harmonize probabilistic safety criteria in the near future. The utilities will not only have to demonstrate that they are operating below a target risk level but also to demonstrate that the unavailability of some of the critical safety systems are below a specified level. In order to satisfy the Technical Specification and Maintenance (TS&M) requirements in a cost effective manner multi-objective optimization of TS&M requirements is of profound interest. The constrained multi-objective optimization of the TS&M requirements of a nuclear power plant (NPP) based on risk and cost gives the pareto-optimal solutions, from which the utility can pick suitable decision variables. The paper presents a multi objective genetic algorithm (GA) technique to investigate a trade-off between risk and cost both at the system and the plant level for Loss of Coolant Accident (LOCA) and Main Steam Line Break (MSLB) as initiating events in a NPP. [Copyright &y& Elsevier]

The health-related damages associated with emissions from coal-fired power plants can vary greatly across facilities as a function of plant, site, and population characteristics, but the degree of variability and the contributing factors have not been formally evaluated. In this study, we modeled the monetized damages associated with 407 coal-fired power plants in the United States, focusing on premature mortality from fine particulate matter (PM2.5). We applied a reduced-form chemistry-transport model accounting for primary PM2.5 emissions and the influence of sulfur dioxide (SO2) and nitrogen oxide (NO x) emissions on secondary particulate formation. Outputs were linked with a concentration-response function for PM2.5-related mortality that incorporated nonlinearities and model uncertainty. We valued mortality with a value of statistical life approach, characterizing and propagating uncertainties in all model elements. At the median of the plant-specific uncertainty distributions, damages across plants ranged from $30,000 to $500,000 per ton of PM2.5, $6,000 to $50,000 per ton of SO2, $500 to $15,000 per ton of NO x, and $0.02 to $1.57 per kilowatt-hour of electricity generated. Variability in damages per ton of emissions was almost entirely explained by population exposure per unit emissions (intake fraction), which itself was related to atmospheric conditions and the population size at various distances from the power plant. Variability in damages per kilowatt-hour was highly correlated with SO2 emissions, related to fuel and control technology characteristics, but was also correlated with atmospheric conditions and population size at various distances. Our findings emphasize that control strategies that consider variability in damages across facilities would yield more efficient outcomes. [ABSTRACT FROM AUTHOR]

Abstract: Hydrogen safety issues and especially hydrogen hazard''s address are key points to remove any safety-related barrier in the implementation process of hydrogen energy systems. Demonstrative systems based on hydrogen technologies represent a clear contribution to the task of showing the feasibility of the new technologies and their beneficial capabilities among the public. In this paper, the safety features of the first hydrogen energy demonstrative plant conceived in Latin America are analyzed. The facilities, located in the village of Pico Truncado, Patagonia, Argentina, serve to gain day-to-day experience in the production, storage, distribution, conversion and use of hydrogen in several applications. The plant uses electrolysis to generate pure hydrogen from renewable primary sources, taking advantage of the installed wind power capacity that is continually growing in the region. The installations were designed to accomplish with two primary objectives: total safety assurance and minimization of human errors. Some details of the plant, including a general layout, are presented here, in addition with design criteria, hydrogen hazards, structural precautions, gas monitoring system, existing regulations and safety requirements. [Copyright &y& Elsevier]

This study presents a comparative assessment of severe accident risks in the energy sector, based on the historical experience of fossil (coal, oil, natural gas, and LPG [Liquefied Petroleum Gas]) and hydro chains contained in the comprehensive Energy-related Severe Accident Database (ENSAD), as well as Probabilistic Safety Assessment (PSA) for the nuclear chain. Full energy chains were considered because accidents can take place at every stage of the chain. Comparative analyses for the years 1969-2000 included a total of 1870 severe (≥ 5 fatalities) accidents, amounting to 81,258 fatalities. Although 79.1% of all accidents and 88.9% of associated fatalities occurred in less developed, non-OECD countries, industrialized OECD countries dominated insured losses (78.0%), reflecting their substantially higher insurance density and stricter safety regulations. Aggregated indicators and frequency-consequence (F-N) curves showed that energy-related accident risks in non-OECD countries are distinctly higher than in OECD countries. Hydropower in non-OECD countries and upstream stages within fossil energy chains are most accident-prone. Expected fatality rates are lowest for Western hydropower and nuclear power plants; however, the maximum credible consequences can be very large. Total economic damages due to severe accidents are substantial, but small when compared with natural disasters. Similarly, external costs associated with severe accidents are generally much smaller than monetized damages caused by air pollution. [ABSTRACT FROM AUTHOR]

Abstract: This study couples structural analyses of four typical U.S. nuclear power plant containments with existing probabilistic risk assessment (PRA) models to assess the effects of degradation. This method is used to determine the increase in the early release frequencies (risk) due to postulated cases of corrosion in the steel liners and shells, as well as other forms of degradation. [Copyright &y& Elsevier]

Abstract: This paper is concerned with the mechanistic modeling and theoretical/computational analysis of flow and heat transfer in future Generation-IV Supercritical Water-Cooled Reactors (SCWRs). The issues discussed in the paper include: the development of analytical models of the properties of supercritical water and the application of a full three-dimensional (3-D) computational modeling framework to simulate fluid flow and heat transfer in SCWRs. Several results of calculations are shown including the evaluation of water properties (density, specific heat, thermal conductivity, viscosity, and Prandtl number) near the pseudo-critical temperature for various supercritical pressures and the CFD predictions using the NPHASE computer code. It is demonstrated that the proposed approach is very promising for future mechanistic analyses of SCWR thermal-hydraulics and safety. [Copyright &y& Elsevier]

Abstract: Advanced boiling water reactor (ABWR) plants have achieved an excellent operating performance since the first ABWR plant started its commercial operation in 1996. Based on the ABWR technology, progress has been made towards a next generation ABWR, AB1600 in Toshiba. The AB1600 plant aims at meeting the demand for the replacement of the current BWRs, which is expected to be realized by 2020 and beyond. In the AB1600 design, therefore, further improvements in economic and reliability aspects have been pursued by incorporating several new technologies. The reactor power is uprated to 1600MWe from 1350MWe of the first ABWR plant in order to benefit from economy of scale. A large fuel assembly with high power density is adopted in order to reduce both of the capital cost and maintenance cost associated with refueling by decreasing the number of fuel assemblies and control rod drives. The AB1600 safety system design employs a hybrid safety system, which consists of both active and passive systems for the design basis and the beyond-design basis accidents, to enhance the safety of the plant. As a countermeasure against severe accidents, the passive systems for the functions of decay heat removal, coolant injection and molten core debris cooling are incorporated. [Copyright &y& Elsevier]

Abstract: AREVA NP has developed an innovative boiling water reactor (BWR) SWR-1000 in close cooperation with German nuclear utilities and with support from various European partners. This Generation III+ reactor design marks a new era in the successful tradition of BWR and, with a net electrical output of approximately 1250MWe, is aimed at ensuring competitive power generating costs compared to gas and coal fired stations. It is particularly suitable for countries whose power networks cannot facilitate large power plants. At the same time, the SWR-1000 meets the highest safety standards, including control of core melt accidents. These objectives are met by supplementing active safety systems with passive safety equipment of various designs for accident detection and control and by simplifying systems needed for normal plant operation on the basis of past operating experience. The plant is also protected against airplane crash loads. The functional capabilities and capacities of all new systems and components were successfully tested under realistic and conservative boundary conditions in large-scale test facilities in Finland, Switzerland and Germany. In general, the SWR-1000 design is based on well-proven analytical codes and design tools validated for BWR applications through recalculation of relevant experiments and independent licensing activities performed by authorities or their experts. The overview of used analytical codes and design tools as well as performed experimental validation programs is presented. Effective implementation of passive safety systems is demonstrated through the numerical simulation of transients and loss of coolant accidents (LOCAs) as well as through analytical simulation of a severe accident associated with the core melt. In the LOCA simulation presented the existing active core flooding systems were not used for emergency control: only passive systems were relevant for the analyses. Despite this – no core heat-up occurred. In the case of reactor core melting numerically is demonstrated that the molten core debris would be retained inside the reactor vessel due to the effective passive external water cooling of the vessel, keeping it completely intact. A short construction period of just 48 months from first concrete to provisional take over, flexible fuel cycle lengths of between 12 and 24 months and a high fuel discharge burn-up all contribute towards meeting economic goals. Realistic average availability for a plant lifetime of 60 years and 12 months cycle is 94.5%. Systems and plant design were reviewed by expert groups of European utilities. With the SWR-1000, AREVA NP has developed a design concept for a BWR plant that is now ready for commercial deployment and which fully meets the most stringent international requirements in terms of nuclear safety and nuclear regulatory. [Copyright &y& Elsevier]

Abstract: The digitalized Instrumentation and Control (I&C) system of Nuclear power plants can provide more powerful overall operation capability, and user friendly man-machine interface. The operator can obtain more information through digital I&C system. However, while I&C system being digitalized, three issues are encountered: (1) software common-cause failure, (2) the interaction failure between operator and digital instrumentation and control system interface, and (3) the non-detectability of software failure. These failures might defeat defense echelons, and make the Diversity and Defense-in-Depth (D3) analysis be more difficult. This work developed an integrated methodology to evaluate nuclear power plant safety effect by interactions between operator and digital I&C system, and then propose improvement recommendations. This integrated methodology includes component-level software fault tree, system-level sequence-tree method and nuclear power plant computer simulation analysis. Software fault tree can clarify the software failure structure in digital I&C systems. Sequence-tree method can identify the interaction process and relationship among operator and I&C systems in each D3 echelon in a design basis event. Nuclear power plant computer simulation analysis method can further analyze the available backup facilities and allowable manual action duration for the operator when the digital I&C fail to function. Applying this methodology to evaluate the performance of digital nuclear power plant D3 design, could promote the nuclear power plant operation safety. The operator can then trust the nuclear power plant than before, when operating the highly automatic digital I&C facilities. [Copyright &y& Elsevier]

It is very difficult for operators to predict the progression of a loss of coolant accident (LOCA) because nuclear plant operators are provided with only partial information during the accident or they may have insufficient time to analyze the data despite being provided with considerable information. Therefore, its break location should be identified and the break size should be predicted accurately in order to provide the operators and technical support personnel with important and valuable information needed to successfully manage the accident. In this paper, support vector machines (SVMs) are used to identify the break location of a LOCA and predict the break size using the support vector classification (SVC) and support vector regression (SVR), which are well-known application areas of SVMs. The SVR models to predict the break size were optimized using a genetic algorithm. The inputs to the SVMs are the time-integrated values obtained by integrating the measurement signals in a short time interval after a reactor scram. The results showed that the proposed algorithm identified the break locations of LOCAs without fault and predicted the break size accurately. [ABSTRACT FROM AUTHOR]

The article presents a review of an investigation study on accident management in VVER-1000 reactor type which was conducted in the framework of a European Commission funded project. It looks into wide-ranged analysis of accident management in VVER-1000, with main regard to the qualification of the computational tool utilized in the study. The proposal for an optimal accident management strategy for this kind of nuclear power plant which involved the depressurization of steam generators and primary system depressurization is also discussed.

Abstract: In-core temperature measurements are pivotal in maintaining nuclear reactors in a safe state of operation. Thermocouples serve as the liaison in ensuring this safe state. The realization of the thermocouple''s full potential is hindered by the fact that thermocouples cannot be situated in areas with high radiation fields. Radiation has the potential of generating voltages in the thermocouple wires, hence producing an error in the temperature transmitter output. In this paper, a mathematical model is developed to quantify the effect that radiation from the Canada Deuterium Uranium (CANDU) Nuclear Power Plants (NPPs) has on the thermocouple temperature reading. Subsequently, a method to offset the effect of radiation on the thermocouple is proposed. Simulation is performed to verify the effectiveness of the proposed system. [Copyright &y& Elsevier]

Risk criterion is a term that may distinguish between what is considered as an acceptable level of safety and what is not. One of the ways of determining quantitative risk criteria for temporary changes in a nuclear power plant considering probabilistic safety assessment is presented. Risk criteria are based on timing and on time duration of the change. Several examples of temporary changes in a nuclear power plant were examined to evaluate the criteria. Results show that it is possible to determine a set of risk criteria for temporary changes. Risk criteria can represent a standpoint for risk-informed decision making. [ABSTRACT FROM AUTHOR]

Renewed interest in precursor analysis has shown that the evaluation of near misses is an interdisciplinary effort, fundamental within the life of an organization for reducing operational risks and enabling accident prevention. The practice of precursor analysis has been a part of nuclear power plant regulation in the United States for over 25 years. During this time, the models used in the analysis have evolved from simple risk equations to quite complex probabilistic risk assessments. But, one item that has remained constant over this time is that the focus of the analysis has been on modeling the scenario using the risk model (regardless of the model sophistication) and then using the results of the model to determine the severity of the precursor incident. We believe that evaluating precursors in this fashion could be a shortcoming since decision making during the incident is not formally investigated. Consequently, we present the idea for an evaluation procedure that enables one to integrate current practice with the evaluation of decisions made during the precursor event. The methodology borrows from technologies both in the risk analysis and the decision analysis realms. We demonstrate this new methodology via an evaluation of a U.S. precursor incident. Specifically, the course of the incident is represented by the integration of a probabilistic risk assessment model (i.e., the risk analysis tool) with an influence diagram and the corresponding decision tree (i.e., the decision analysis tools). The results and insights from the application of this new methodology are discussed. [ABSTRACT FROM AUTHOR]

The construction of the Philippine's first nuclear power plant by Westinghouse has come to symbolize the corruption and cronyism of the Marcos' years. The plant has created so much controversy that it has yet to operate, in a country that desperately needs electrical power. The entire affair suggests that American multinationals cannot simply adopt the business practices of the host country, but instead must carefully analyze the political and business environments of their foreign based operations, and then take into account the long-term ethical implications of their decisions. The case also suggests that government needs to play a more active role in regulating multinationals, especially when sophisticated and potentially dangerous technologies are involved. [ABSTRACT FROM AUTHOR]