Application of NASA Kennedy Space Center system assurance analysis methodology to nuclear power plant systems designs

International Nuclear Information System (INIS)

Page, D.W.

1985-01-01

The Kennedy Space Center (KSC) entered into an agreement with the Nuclear Regulatory Commission (NRC) to conduct a study to demonstrate the feasibility and practicality of applying the KSC System Assurance Analysis (SAA) methodology to nuclear power plant systems designs. In joint meetings of KSC and Duke Power personnel, an agreement was made to select to CATAWBA systems, the Containment Spray System and the Residual Heat Removal System, for the analyses. Duke Power provided KSC with a full set a Final Safety Analysis Reports as well as schematics for the two systems. During Phase I of the study the reliability analyses of the SAA were performed. During Phase II the hazard analyses were performed. The final product of Phase II is a handbook for implementing the SAA methodology into nuclear power plant systems designs. The purpose of this paper is to describe the SAA methodology as it applies to nuclear power plant systems designs and to discuss the feasibility of its application. The conclusion is drawn that nuclear power plant systems and aerospace ground support systems are similar in complexity and design and share common safety and reliability goals. The SAA methodology is readily adaptable to nuclear power plant designs because of it's practical application of existing and well known safety and reliability analytical techniques tied to an effective management information system

Safety analysis of an expert reactor protection system in nuclear power plants

International Nuclear Information System (INIS)

EL-Kafas, A.E.A.E.

1996-01-01

the purpose of the dissertation is to develop a real time expert reactor protection system (ERPS) for operational safety of pressurized water reactor nuclear power plant. The system is developed to diagnose plant failures and for identification of plant transients (with and without scram). for this ERPS. probabilistic safety analysis techniques are used to check the availability and priority of the recommended safety system in case of plant accidents . the real- time information during transients and accidents can be obtained to asses the operator in his decision - making . Also, the ERPS is able to give advice for the reactor operator to take the appropriate corrective action during abnormal situations. The system model consists of the dynamic differential equations for reactor core, pressurizer, steam generator, turbine and generator, piping and plenums. The system of equations can be solved by appropriate codes also displayed directly from sensors of the plant. All scenarios of transients, accidents and fault tress for plant systems are learned to ERPS

A framework for the system-of-systems analysis of the risk for a safety-critical plant exposed to external events

International Nuclear Information System (INIS)

Zio, E.; Ferrario, E.

2013-01-01

We consider a critical plant exposed to risk from external events. We propose an original framework of analysis, which extends the boundaries of the study to the interdependent infrastructures which support the plant. For the purpose of clearly illustrating the conceptual framework of system-of-systems analysis, we work out a case study of seismic risk for a nuclear power plant embedded in the connected power and water distribution, and transportation networks which support its operation. The technical details of the systems considered (including the nuclear power plant) are highly simplified, in order to preserve the purpose of illustrating the conceptual, methodological framework of analysis. Yet, as an example of the approaches that can be used to perform the analysis within the proposed framework, we consider the Muir Web as system analysis tool to build the system-of-systems model and Monte Carlo simulation for the quantitative evaluation of the model. The numerical exercise, albeit performed on a simplified case study, serves the purpose of showing the opportunity of accounting for the contribution of the interdependent infrastructure systems to the safety of a critical plant. This is relevant as it can lead to considerations with respect to the decision making related to safety critical-issues. -- Highlights: ► We consider a critical plant exposed to risk from external events. ► We consider also the interdependent infrastructures that support the plant. ► We use Muir Web as system analysis tool to build the system-of-systems model. ► We use Monte Carlo simulation for the quantitative evaluation of the model. ► We find that the interdependent infrastructures should be considered as they can be a support for the critical plant safety

B plant mission analysis report

International Nuclear Information System (INIS)

Lund, D.P.

1995-01-01

This report further develops the mission for B Plant originally defined in WHC-EP-0722, ''System Engineering Functions and Requirements for the Hanford Cleanup Mission: First Issue.'' The B Plant mission analysis will be the basis for a functional analysis that breaks down the B Plant mission statement into the necessary activities to accomplish the mission. These activities are the product of the functional analysis and will then be used in subsequent steps of the systems engineering process, such as identifying requirements and allocating those requirements to B Plant functions. The information in this mission analysis and the functional and requirements analysis are a part of the B Plant technical baseline

A Production-Rule Analysis System for Nuclear Plant monitoring and emergency response applications

International Nuclear Information System (INIS)

Ragheb, M.; Tsoukalas, L.; McDonough, T.; Parker, M.

1987-01-01

A Production-Rule Analysis System for Nuclear Power Plant Monitoring is presented. The signals generated by the Zion-1 Plant are considered for emergency Response applications. The integrity of the Plant Radiation, the Reactor Coolant, the Fuel Clad, and the Containment Systems, is monitored. Representation of the system is in the form of a goal-tree generating a Knowledge-Base searched by an Inference Engine functioning in the forward-chaining mode. The Gaol-tree is built from Fault-Trees based on plant operational information. The system is implemented on a VAX-8500 and is programmed in OPS-5

Analysis of reactor trips originating in balance of plant systems

International Nuclear Information System (INIS)

Stetson, F.T.; Gallagher, D.W.; Le, P.T.; Ebert, M.W.

1990-09-01

This report documents the results of an analysis of balance-of-plant (BOP) related reactor trips at commercial US nuclear power plants of a 5-year period, from January 1, 1984, through December 31, 1988. The study was performed for the Plant Systems Branch, Office of Nuclear Reactor Regulation, US Nuclear Regulatory Commission. The objectives of the study were: to improve the level of understanding of BOP-related challenges to safety systems by identifying and categorizing such events; to prepare a computerized data base of BOP-related reactor trip events and use the data base to identify trends and patterns in the population of these events; to investigate the risk implications of BOP events that challenge safety systems; and to provide recommendations on how to address BOP-related concerns in regulatory context. 18 refs., 2 figs., 27 tabs

Safety analysis of an expert reactor protection system in nuclear power plants

International Nuclear Information System (INIS)

El-Kafas, A.A.

1997-01-01

The purpose of the dissertation is to develop real time expert reactor protection system (ERPS) for operational safety of pressurized water reactor nuclear power plant. The system is developed to diagnose plant failures and for identification plant transients (with and without scram). For this erps, probabilistic safety analysis techniques are used to check the availability and priority of the recommended safety system in case of plant accidents. The real - time information during transients and accidents can be obtained to assess the operator in his decision - making. Also, the ERPS is able to give advice for the reactor operator to take the appropriate corrective action during abnormal situations. 5-15 figs., 42 refs

Plant air systems safety study: Portsmouth Gaseous Diffusion Plant

International Nuclear Information System (INIS)

1982-05-01

The Portsmouth Gaseous Diffusion Plant Air System facilities and operations are reviewed for potential safety problems not covered by standard industrial safety procedures. Information is presented under the following section headings: facility and process description (general); air plant equipment; air distribution system; safety systems; accident analysis; plant air system safety overview; and conclusion

Thermal Power Plant Performance Analysis

CERN Document Server

2012-01-01

The analysis of the reliability and availability of power plants is frequently based on simple indexes that do not take into account the criticality of some failures used for availability analysis. This criticality should be evaluated based on concepts of reliability which consider the effect of a component failure on the performance of the entire plant. System reliability analysis tools provide a root-cause analysis leading to the improvement of the plant maintenance plan.   Taking in view that the power plant performance can be evaluated not only based on  thermodynamic related indexes, such as heat-rate, Thermal Power Plant Performance Analysis focuses on the presentation of reliability-based tools used to define performance of complex systems and introduces the basic concepts of reliability, maintainability and risk analysis aiming at their application as tools for power plant performance improvement, including: ·         selection of critical equipment and components, ·         defini...

Analysis of high-throughput plant image data with the information system IAP

Directory of Open Access Journals (Sweden)

Klukas Christian

2012-06-01

Full Text Available This work presents a sophisticated information system, the Integrated Analysis Platform (IAP, an approach supporting large-scale image analysis for different species and imaging systems. In its current form, IAP supports the investigation of Maize, Barley and Arabidopsis plants based on images obtained in different spectra.

Control system for technological processes in tritium processing plants with process analysis

International Nuclear Information System (INIS)

Retevoi, Carmen Maria; Stefan, Iuliana; Balteanu, Ovidiu; Stefan, Liviu; Bucur, Ciprian

2005-01-01

Integration of a large variety of installations and equipment into a unitary system for controlling the technological process in tritium processing nuclear facilities appears to be a rather complex approach particularly when experimental or new technologies are developed. Ensuring a high degree of versatility allowing easy modifications in configurations and process parameters is a major requirement imposed on experimental installations. The large amount of data which must be processed, stored and easily accessed for subsequent analyses imposes development of a large information network based on a highly integrated system containing the acquisition, control and technological process analysis data as well as data base system. On such a basis integrated systems of computation and control able to conduct the technological process could be developed as well protection systems for cases of failures or break down. The integrated system responds to the control and security requirements in case of emergency and of the technological processes specific to the industry that processes radioactive or toxic substances with severe consequences in case of technological failure as in the case of tritium processing nuclear plant. In order to lower the risk technological failure of these processes an integrated software, data base and process analysis system are developed, which, based on identification algorithm of the important parameters for protection and security systems, will display the process evolution trend. The system was checked on a existing plant that includes a removal tritium unit, finally used in a nuclear power plant, by simulating the failure events as well as the process. The system will also include a complete data base monitoring all the parameters and a process analysis software for the main modules of the tritium processing plant, namely, isotope separation, catalytic purification and cryogenic distillation

Application of NASA Kennedy Space Center System Assurance Analysis methodology to nuclear power plant systems designs

International Nuclear Information System (INIS)

Page, D.W.

1985-01-01

In May of 1982, the Kennedy Space Center (KSC) entered into an agreement with the NRC to conduct a study to demonstrate the feasibility and practicality of applying the KSC System Assurance Analysis (SAA) methodology to nuclear power plant systems designs. North Carolina's Duke Power Company expressed an interest in the study and proposed the nuclear power facility at CATAWBA for the basis of the study. In joint meetings of KSC and Duke Power personnel, an agreement was made to select two CATAWBA systems, the Containment Spray System and the Residual Heat Removal System, for the analyses. Duke Power provided KSC with a full set of Final Safety Analysis Reports (FSAR) as well as schematics for the two systems. During Phase I of the study the reliability analyses of the SAA were performed. During Phase II the hazard analyses were performed. The final product of Phase II is a handbook for implementing the SAA methodology into nuclear power plant systems designs. The purpose of this paper is to describe the SAA methodology as it applies to nuclear power plant systems designs and to discuss the feasibility of its application. (orig./HP)

Reliability Analysis Study of Digital Reactor Protection System in Nuclear Power Plant

International Nuclear Information System (INIS)

Guo, Xiao Ming; Liu, Tao; Tong, Jie Juan; Zhao, Jun

2011-01-01

The Digital I and C systems are believed to improve a plants safety and reliability generally. The reliability analysis of digital I and C system has become one research hotspot. Traditional fault tree method is one of means to quantify the digital I and C system reliability. Review of advanced nuclear power plant AP1000 digital protection system evaluation makes clear both the fault tree application and analysis process to the digital system reliability. One typical digital protection system special for advanced reactor has been developed, which reliability evaluation is necessary for design demonstration. The typical digital protection system construction is introduced in the paper, and the process of FMEA and fault tree application to the digital protection system reliability evaluation are described. Reliability data and bypass logic modeling are two points giving special attention in the paper. Because the factors about time sequence and feedback not exist in reactor protection system obviously, the dynamic feature of digital system is not discussed

Dynamic analysis of the condensate feedwater system in boiling water reactor plants

International Nuclear Information System (INIS)

Tanji, J.; Omori, T.

1982-01-01

The computer code, CONFAC, has been developed for dynamic analysis of the condensate feedwater system in boiling water reactor plants. This code simulates the hydrodynamics in the piping system, the pump dynamics, and the feedwater controller in order to clarify the system transient characteristics in such cases as pump trip incidents. Code verification was performed by comparison between analytical results and actual plant operational data. Satisfactory agreement was obtained. With the code, appropriate pump start/stop interlocks were estimated for preventing pump cavitation in pump trip incidents

Analysis of a Plant Transcriptional Regulatory Network Using Transient Expression Systems.

Science.gov (United States)

Díaz-Triviño, Sara; Long, Yuchen; Scheres, Ben; Blilou, Ikram

2017-01-01

In plant biology, transient expression systems have become valuable approaches used routinely to rapidly study protein expression, subcellular localization, protein-protein interactions, and transcriptional activity prior to in vivo studies. When studying transcriptional regulation, luciferase reporter assays offer a sensitive readout for assaying promoter behavior in response to different regulators or environmental contexts and to confirm and assess the functional relevance of predicted binding sites in target promoters. This chapter aims to provide detailed methods for using luciferase reporter system as a rapid, efficient, and versatile assay to analyze transcriptional regulation of target genes by transcriptional regulators. We describe a series of optimized transient expression systems consisting of Arabidopsis thaliana protoplasts, infiltrated Nicotiana benthamiana leaves, and human HeLa cells to study the transcriptional regulations of two well-characterized transcriptional regulators SCARECROW (SCR) and SHORT-ROOT (SHR) on one of their targets, CYCLIN D6 (CYCD6).Here, we illustrate similarities and differences in outcomes when using different systems. The plant-based systems revealed that the SCR-SHR complex enhances CYCD6 transcription, while analysis in HeLa cells showed that the complex is not sufficient to strongly induce CYCD6 transcription, suggesting that additional, plant-specific regulators are required for full activation. These results highlight the importance of the system and suggest that including heterologous systems, such as HeLa cells, can provide a more comprehensive analysis of a complex gene regulatory network.

Technical analysis of advanced wastewater-treatment systems for coal-gasification plants

Energy Technology Data Exchange (ETDEWEB)

1981-03-31

This analysis of advanced wastewater treatment systems for coal gasification plants highlights the three coal gasification demonstration plants proposed by the US Department of Energy: The Memphis Light, Gas and Water Division Industrial Fuel Gas Demonstration Plant, the Illinois Coal Gasification Group Pipeline Gas Demonstration Plant, and the CONOCO Pipeline Gas Demonstration Plant. Technical risks exist for coal gasification wastewater treatment systems, in general, and for the three DOE demonstration plants (as designed), in particular, because of key data gaps. The quantities and compositions of coal gasification wastewaters are not well known; the treatability of coal gasification wastewaters by various technologies has not been adequately studied; the dynamic interactions of sequential wastewater treatment processes and upstream wastewater sources has not been tested at demonstration scale. This report identifies key data gaps and recommends that demonstration-size and commercial-size plants be used for coal gasification wastewater treatment data base development. While certain advanced treatment technologies can benefit from additional bench-scale studies, bench-scale and pilot plant scale operations are not representative of commercial-size facility operation. It is recommended that coal gasification demonstration plants, and other commercial-size facilities that generate similar wastewaters, be used to test advanced wastewater treatment technologies during operation by using sidestreams or collected wastewater samples in addition to the plant's own primary treatment system. Advanced wastewater treatment processes are needed to degrade refractory organics and to concentrate and remove dissolved solids to allow for wastewater reuse. Further study of reverse osmosis, evaporation, electrodialysis, ozonation, activated carbon, and ultrafiltration should take place at bench-scale.

Physics, systems analysis and economics of fusion power plants

International Nuclear Information System (INIS)

Ward, D.J.

2006-01-01

Fusion power is being developed because of its large resource base, low environmental impact and high levels of intrinsic safety. It is important, however, to investigate the economics of a future fusion power plant to check that the electricity produced can, in fact, have a market. Using systems code analysis, including costing algorithms, this paper gives the cost of electricity expected from a range of fusion power plants, assuming that they are brought into successful operation. Although this paper does not purport to show that a first generation of fusion plants is likely to be the cheapest option for a future energy source, such plants look likely to have a market in some countries even without taking account of fusion's environmental advantages. With improved technological maturity fusion looks likely to have a widespread potential market particularly if the value of its environmental advantages are captured, for instance through avoiding a carbon tax. (author)

Plant operator performance evaluation system

International Nuclear Information System (INIS)

Ujita, Hiroshi; Fukuda, Mitsuko; Kubota, Ryuji.

1989-01-01

A plant operator performance evaluation system to analyze plant operation records during accident training and to identify and classify operator errors has been developed for the purpose of supporting realization of a training and education system for plant operators. A knowledge engineering technique was applied to evaluation of operator behavior by both even-based and symptom-based procedures, in various situations including event transition due to multiple failures or operational errors. The system classifies the identified errors as to their single and double types based on Swain's error classification and the error levels reflecting Rasmussen's cognitive level, and it also evaluates the effect of errors on plant state and then classifies error influence, using 'knowledge for phenomena and operations', as represented by frames. It has additional functions for analysis of error statistics and knowledge acquisition support of 'knowledge for operations'. The system was applied to a training analysis for a scram event in a BWR plant, and its error analysis function was confirmed to be effective by operational experts. (author)

Situation-assessment and decision-aid production-rule analysis system for nuclear plant monitoring and emergency preparedness

International Nuclear Information System (INIS)

Gvillo, D.; Ragheb, M.; Parker, M.; Swartz, S.

1987-01-01

A Production-Rule Analysis System is developed for Nuclear Plant Monitoring. The signals generated by the Zion-1 Plant are considered. A Situation-Assessment and Decision-Aid capability is provided for monitoring the integrity of the Plant Radiation, the Reactor Coolant, the Fuel Clad, and the Containment Systems. A total of 41 signals are currently fed as facts to an Inference Engine functioning in the backward-chaining mode and built along the same structure as the E-Mycin system. The Goal-Tree constituting the Knowledge Base was generated using a representation in the form of Fault Trees deduced from plant procedures information. The system is constructed in support of the Data Analysis and Emergency Preparedness tasks at the Illinois Radiological Emergency Assessment Center (REAC)

Situation-Assessment And Decision-Aid Production-Rule Analysis System For Nuclear Plant Monitoring And Emergency Preparedness

Science.gov (United States)

Gvillo, D.; Ragheb, M.; Parker, M.; Swartz, S.

1987-05-01

A Production-Rule Analysis System is developed for Nuclear Plant Monitoring. The signals generated by the Zion-1 Plant are considered. A Situation-Assessment and Decision-Aid capability is provided for monitoring the integrity of the Plant Radiation, the Reactor Coolant, the Fuel Clad, and the Containment Systems. A total of 41 signals are currently fed as facts to an Inference Engine functioning in the backward-chaining mode and built along the same structure as the E-Mycin system. The Goal-Tree constituting the Knowledge Base was generated using a representation in the form of Fault Trees deduced from plant procedures information. The system is constructed in support of the Data Analysis and Emergency Preparedness tasks at the Illinois Radiological Emergency Assessment Center (REAC).

Multi-variable systems in nuclear power plant

International Nuclear Information System (INIS)

Collins, G.B.; Howell, J.

1982-01-01

Nuclear power plant are complex multi-variable dynamically interactive systems which employ many facets of systems and control theory in their analysis and design. Whole plant mathematical models must be developed and validated and in addition to their obvious role in control system synthesis and design, they are also widely used for operational constraint and plant malfunction analysis. The need for and scope of an integrated power plant control system is discussed and, as a specific example, the design of an integrated feedwater regulator is reviewed. The multi-variable frequency response analysis employed in the design is described in detail. (author)

Concepts in production ecology for analysis and design of animal and plant-animal production systems

NARCIS (Netherlands)

Ven, van de G.W.J.; Ridder, de N.; Keulen, van H.; Ittersum, van M.K.

2003-01-01

The use of a hierarchy in growth factors (defining, limiting and reducing growth factors), as developed for plant production has shown its usefulness in the analysis and design of plant production systems. This hierarchy presents a theoretical framework for the analysis of biophysical conditions in

Markov modeling and reliability analysis of urea synthesis system of a fertilizer plant

Science.gov (United States)

Aggarwal, Anil Kr.; Kumar, Sanjeev; Singh, Vikram; Garg, Tarun Kr.

2015-12-01

This paper deals with the Markov modeling and reliability analysis of urea synthesis system of a fertilizer plant. This system was modeled using Markov birth-death process with the assumption that the failure and repair rates of each subsystem follow exponential distribution. The first-order Chapman-Kolmogorov differential equations are developed with the use of mnemonic rule and these equations are solved with Runga-Kutta fourth-order method. The long-run availability, reliability and mean time between failures are computed for various choices of failure and repair rates of subsystems of the system. The findings of the paper are discussed with the plant personnel to adopt and practice suitable maintenance policies/strategies to enhance the performance of the urea synthesis system of the fertilizer plant.

archiDART v3.0: A new data analysis pipeline allowing the topological analysis of plant root systems.

Science.gov (United States)

Delory, Benjamin M; Li, Mao; Topp, Christopher N; Lobet, Guillaume

2018-01-01

Quantifying plant morphology is a very challenging task that requires methods able to capture the geometry and topology of plant organs at various spatial scales. Recently, the use of persistent homology as a mathematical framework to quantify plant morphology has been successfully demonstrated for leaves, shoots, and root systems. In this paper, we present a new data analysis pipeline implemented in the R package archiDART to analyse root system architectures using persistent homology. In addition, we also show that both geometric and topological descriptors are necessary to accurately compare root systems and assess their natural complexity.

Comparative Analysis of Power Plant Options for Enhanced Geothermal Systems (EGS

Directory of Open Access Journals (Sweden)

Mengying Li

2014-12-01

Full Text Available Enhanced geothermal systems (EGS extract heat from underground hot dry rock (HDR by first fracturing the HDR and then circulating a geofluid (typically water into it and bringing the heated geofluid to a power plant to generate electricity. This study focuses on analysis, examination, and comparison of leading geothermal power plant configurations with a geofluid temperature from 200 to 800 °C, and also analyzes the embodied energy of EGS surface power plants. The power generation analysis is focused on flash type cycles for using subcritical geofluid (<374 °C and expansion type cycles for using supercritical geofluid (>374 °C. Key findings of this study include: (i double-flash plants have 24.3%–29.0% higher geofluid effectiveness than single-flash ones, and 3%–10% lower specific embodied energy; (ii the expansion type plants have geofluid effectiveness > 750 kJ/kg, significantly higher than flash type plants (geofluid effectiveness < 300 kJ/kg and the specific embodied energy is lower; (iii to increase the turbine outlet vapor fraction from 0.75 to 0.90, we include superheating by geofluid but that reduces the geofluid effectiveness by 28.3%; (iv for geofluid temperatures above 650 °C, double-expansion plants have a 2% higher geofluid effectiveness and 5%–8% lower specific embodied energy than single-expansion ones.

System Definition and Analysis: Power Plant Design and Layout

International Nuclear Information System (INIS)

1996-01-01

This is the Topical report for Task 6.0, Phase 2 of the Advanced Turbine Systems (ATS) Program. The report describes work by Westinghouse and the subcontractor, Gilbert/Commonwealth, in the fulfillment of completing Task 6.0. A conceptual design for critical and noncritical components of the gas fired combustion turbine system was completed. The conceptual design included specifications for the flange to flange gas turbine, power plant components, and balance of plant equipment. The ATS engine used in the conceptual design is an advanced 300 MW class combustion turbine incorporating many design features and technologies required to achieve ATS Program goals. Design features of power plant equipment and balance of plant equipment are described. Performance parameters for these components are explained. A site arrangement and electrical single line diagrams were drafted for the conceptual plant. ATS advanced features include design refinements in the compressor, inlet casing and scroll, combustion system, airfoil cooling, secondary flow systems, rotor and exhaust diffuser. These improved features, integrated with prudent selection of power plant and balance of plant equipment, have provided the conceptual design of a system that meets or exceeds ATS program emissions, performance, reliability-availability-maintainability, and cost goals

Cost-benefit analysis for environmental impacts and radwaste system for nuclear power plant

International Nuclear Information System (INIS)

Mun, K.N.; Yook, C.C.

1982-01-01

During operation of nuclear power plant, radioactive material is inevitably formed. This radioactive material must be safely processed by radwaste system so that essentially zero activity is released to the environment. However zero released activity is not really practicable and population doses resulted from released activity are proportional to total annual cost for the radwaste system. In this study, cost-benefit analysis for the radwaste system of the Korean Nuclear Units 5 and 6 is performed to evaluate the optimization between the total annual cost for the radwaste system and population doses within 80 km from the plants. From the analysis, the following results are obtained; 1. the total population dose is estimated 4.04 x 10 3 man-rem/year, 2. total annual cost for the radwaste system is required $ 1.74 x 10 6 , 3. cost-benefit ratio is estimated $ 429/man-rem. (Author)

The application of plant information system on third Qinshan nuclear power plant

International Nuclear Information System (INIS)

Liu Wangtian

2005-01-01

Plant overall control has been applied in Qinshan Nuclear Power Plant, which enhances the security of plant operation, but it is not enough to improve the technical administration level. In order to integrate the overall information and to improve the technical administration level more. Third Qinshan Nuclear Power Plant applies the plant information system. This thesis introduces the application of plant information system in Third Qinshan Nuclear Power Plant and the effect to the plant after the system is carried into execution, in addition, it does more analysis and exceptions for application of plant information system in the future. (authors)

Module type plant system dynamics analysis code (MSG-COPD). Code manual

International Nuclear Information System (INIS)

Sakai, Takaaki

2002-11-01

MSG-COPD is a module type plant system dynamics analysis code which involves a multi-dimensional thermal-hydraulics calculation module to analyze pool type of fast breeder reactors. Explanations of each module and the methods for the input data are described in this code manual. (author)

Solar Pilot Plant, Phase I. Preliminary design report. Volume II. System description and system analysis. CDRL item 2

Energy Technology Data Exchange (ETDEWEB)

None

1977-05-01

Honeywell conducted a parametric analysis of the 10-MW(e) solar pilot plant requirements and expected performance and established an optimum system design. The main analytical simulation tools were the optical (ray trace) and the dynamic simulation models. These are described in detail in Books 2 and 3 of this volume under separate cover. In making design decisions, available performance and cost data were used to provide a design reflecting the overall requirements and economics of a commercial-scale plant. This volume contains a description of this analysis/design process and resultant system/subsystem design and performance.

Reliability analysis of safety systems of nuclear power plant and utility experience with reliability safeguarding of systems during specified normal operation

International Nuclear Information System (INIS)

Balfanz, H.P.

1989-01-01

The paper gives an outline of the methods applied for reliability analysis of safety systems in nuclear power plant. The main tasks are to check the system design for detection of weak points, and to find possibilities of optimizing the strategies for inspection, inspection intervals, maintenance periods. Reliability safeguarding measures include the determination and verification of the broundary conditions of the analysis with regard to the reliability parameters and maintenance parameters used in the analysis, and the analysis of data feedback reflecting the plant response during operation. (orig.) [de

Methodological aspects of fuel performance system analysis at raw hydrocarbon processing plants

Science.gov (United States)

Kulbjakina, A. V.; Dolotovskij, I. V.

2018-01-01

The article discusses the methodological aspects of fuel performance system analysis at raw hydrocarbon (RH) processing plants. Modern RH processing facilities are the major consumers of energy resources (ER) for their own needs. To reduce ER, including fuel consumption, and to develop rational fuel system structure are complex and relevant scientific tasks that can only be done using system analysis and complex system synthesis. In accordance with the principles of system analysis, the hierarchical structure of the fuel system, the block scheme for the synthesis of the most efficient alternative of the fuel system using mathematical models and the set of performance criteria have been developed on the main stages of the study. The results from the introduction of specific engineering solutions to develop their own energy supply sources for RH processing facilities have been provided.

Analysis of Piping Systems for Life Extension of Heavy Water Plants in India

International Nuclear Information System (INIS)

Mishra, Rajesh K.; Soni, R.S.; Kushwaha, H.S.; Raj, V. Venkat

2002-01-01

Heavy water production in India has achieved many milestones in the past. Two of the successfully running heavy water plants are on the verge of completion of their design life in the near future. One of these two plants, situated at Kota, is a hydrogen sulfide based plant and the other one at Tuticorin is an ammonia-based plant. Various exercises have been planned with an aim to assess the fatigue usage for the various components of these plants in order to extend their life. Considering the process parameters and the past history of the plant performance, critical piping systems and equipment are identified. Analyses have been carried out for these critical piping systems for mainly two kinds of loading, viz. sustained loads and the expansion loads. Static analysis has been carried out to find the induced stress levels due to sustained as well as thermal expansion loading as per the design code ANSI B31.3. Due consideration has been given to the design corrosion allowance while evaluating the stresses due to sustained loads. At the locations where the induced stresses (S L ) due to the sustained loads are exceeding the allowable limits (S h ), exercises have been carried out considering the reduced corrosion allowance value. This strategy is adopted in view of the fact that the thickness measurements carried out at site at various critical locations show a very low rate of corrosion. It has been possible to qualify the system with reduced corrosion allowance values however, it is recommended to keep that location under periodic monitoring. The strategy adopted for carrying out analysis for thermal expansion loading is to qualify the system as per the code allowable value (S a ). If the stresses are more than the allowable value, credit of liberal allowable value as suggested in the code i.e., with the addition of the term (S h -S L ) to the term 0.25 S h , has been taken. However, if at any location, it is found that thermal stress is high, fatigue analysis has

System Evaluation and Economic Analysis of a HTGR Powered High-Temperature Electrolysis Hydrogen Production Plant

International Nuclear Information System (INIS)

McKellar, Michael G.; Harvego, Edwin A.; Gandrik, Anastasia A.

2010-01-01

A design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production has been developed. The HTE plant is powered by a high-temperature gas-cooled reactor (HTGR) whose configuration and operating conditions are based on the latest design parameters planned for the Next Generation Nuclear Plant (NGNP). The current HTGR reference design specifies a reactor power of 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 322 C and 750 C, respectively. The power conversion unit will be a Rankine steam cycle with a power conversion efficiency of 40%. The reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes a steam-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The overall system thermal-to-hydrogen production efficiency (based on the higher heating value of the produced hydrogen) is 40.4% at a hydrogen production rate of 1.75 kg/s and an oxygen production rate of 13.8 kg/s. An economic analysis of this plant was performed with realistic financial and cost estimating assumptions. The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a cost of $3.67/kg of hydrogen assuming an internal rate of return, IRR, of 12% and a debt to equity ratio of 80%/20%. A second analysis shows that if the power cycle efficiency increases to 44.4%, the hydrogen production efficiency increases to 42.8% and the hydrogen and oxygen production rates are 1.85 kg/s and 14.6 kg/s respectively. At the higher power cycle efficiency and an IRR of 12% the cost of hydrogen production is $3.50/kg.

Cost benefit analysis of instrumentation, supervision and control systems for nuclear power plants

International Nuclear Information System (INIS)

Hagen, P.

1973-08-01

A cost benefit analysis is carried out on a BWR type reactor power plant in which an on-line computer performs plant supervision, reporting, logging, calibration and control functions, using display devices and plotters, while an off-line computer is available for bigger jobs such as fuel management calculations. All on-line functions are briefly described and specified. Three types of computer system are considered, a simplex system, a dual computer system and a multi-processor system. These systems are analysed with respect to reliability, back-up instrumentation requirements and costs. While the multiprocessor system gave in all cases the lowest annual failure costs, the margin to the duplex system was so small that hardware, maintenance and software costs would play an important role in making a decision. (JIW)

Application of Functional Analysis on a SCADA System of a Thermal Power Plant

Directory of Open Access Journals (Sweden)

LAKHOUA, M. N.

2009-06-01

Full Text Available The aim of this paper is to present firstly the functionality of a supervisory system for complex processes, and secondly to present the concepts of SCADA (supervisory control and data acquisition systems. A functional analysis technique SADT (Structured Analysis and Design Technique has been applied on an example of a SCADA system of a thermal power plant. This technique allows a functional description of a SCADA system. The paper briefly discusses the functions of a SCADA system and some advantages of the application of functional analysis for the design of a human centered supervisory system. Then the basic principles of the SADT technique applied on the SCADA system are presented. Finally, the different results obtained from the SADT technique are discussed.

Preliminary systems-interaction results from the Digraph Matrix Analysis of the Watts Bar Nuclear Power Plant safety-injection systems

International Nuclear Information System (INIS)

Sacks, I.J.; Ashmore, B.C.; Champney, J.M.; Alesso, H.P.

1983-06-01

This report provides preliminary results generated by a Digraph Matrix Analysis (DMA) for a Systems Interaction analysis performed on the Safety Injection System of the Tennessee Valley Authority Watts Bar Nuclear Power Plant. An overview of DMA is provided along with a brief description of the computer codes used in DMA

Common Cause Failure Analysis for the Digital Plant Protection System

International Nuclear Information System (INIS)

Kagn, Hyun Gook; Jang, Seung Cheol

2005-01-01

Safety-critical systems such as nuclear power plants adopt the multiple-redundancy design in order to reduce the risk from the single component failure. The digitalized safety-signal generation system is also designed based on the multiple-redundancy strategy which consists of more redundant components. The level of the redundant design of digital systems is usually higher than those of conventional mechanical systems. This higher redundancy would clearly reduce the risk from the single failure of components, but raise the importance of the common cause failure (CCF) analysis. This research aims to develop the practical and realistic method for modeling the CCF in digital safety-critical systems. We propose a simple and practical framework for assessing the CCF probability of digital equipment. Higher level of redundancy causes the difficulty of CCF analysis because it results in impractically large number of CCF events in the fault tree model when we use conventional CCF modeling methods. We apply the simplified alpha-factor (SAF) method to the digital system CCF analysis. The precedent study has shown that SAF method is quite realistic but simple when we consider carefully system success criteria. The first step for using the SAF method is the analysis of target system for determining the function failure cases. That is, the success criteria of the system could be derived from the target system's function and configuration. Based on this analysis, we can calculate the probability of single CCF event which represents the CCF events resulting in the system failure. In addition to the application of SAF method, in order to accommodate the other characteristics of digital technology, we develop a simple concept and several equations for practical use

Reliability analysis of repairable safety systems of a reprocessing plant allowing for tolerable system downtimes

International Nuclear Information System (INIS)

Schaefer, H.

1987-01-01

GRS has been engaged in safety analysises of the German Reprocessing Plant for several years. The development and verification of appropriate reliability analysis methods, the generation of data as well as the search for an adequate structural presentation of the results to form a basis of recommendations for technical or administrative measures or contributions to risk oriented evaluations have been or are in the process of being established. In contrast to NPP-studies, the reliability assessment of safety systems of a reprocessing plant is applied to repairable and often relatively small systems allowing for tolerable system downtimes. A sketch of the diverse cooling systems of a vessel containing a selfheating solution is given. The interruption of the cooling function for about one day might be tolerable before boiling will be reached. This interval is suitable for transfer of the solution to a spare vessel or for repairing the failed components, thus restoring the cooling function

Case studies on the feasibility of the transient analysis system STAR in German nuclear power plants

International Nuclear Information System (INIS)

Buettner, W.E.; Felkel, L.; Zapp, A.

1984-01-01

On the basis of distubances which actually have occurred in German nuclear power plants a case-study has been performed to evaluate the feasibility of the computer-based disturbance analysis system STAR. By means of a compact plant simulator the disturbances collected have been remodelled and anlysed, on-line, with the disturbance analysis system STAR. In the last phase of the project experiments have been performed with reactor operators to get their reaction to, and opinion on, computerbased-operator aids. (orig.) [de

System Evaluation and Economic Analysis of a Nuclear Reactor Powered High-Temperature Electrolysis Hydrogen-Production Plant

International Nuclear Information System (INIS)

Harvego, E.A.; McKellar, M.G.; Sohal, M.S.; O'Brien, J.E.; Herring, J.S.

2010-01-01

A reference design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production was developed to provide a basis for comparing the HTE concept with other hydrogen production concepts. The reference plant design is driven by a high-temperature helium-cooled nuclear reactor coupled to a direct Brayton power cycle. The reference design reactor power is 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 540 C and 900 C, respectively. The electrolysis unit used to produce hydrogen includes 4,009,177 cells with a per-cell active area of 225 cm2. The optimized design for the reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes an air-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The inlet air for the air-sweep system is compressed to the system operating pressure of 5.0 MPa in a four-stage compressor with intercooling. The alternating current (AC) to direct current (DC) conversion efficiency is 96%. The overall system thermal-to-hydrogen production efficiency (based on the lower heating value of the produced hydrogen) is 47.1% at a hydrogen production rate of 2.356 kg/s. An economic analysis of this plant was performed using the standardized H2A Analysis Methodology developed by the Department of Energy (DOE) Hydrogen Program, and using realistic financial and cost estimating assumptions. The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a competitive cost. A cost of $3.23/kg of hydrogen was calculated assuming an internal rate of return of 10%.

Development of a nuclear power plant system analysis code

International Nuclear Information System (INIS)

Sim, Suk K.; Jeong, J. J.; Ha, K. S.; Moon, S. K.; Park, J. W.; Yang, S. K.; Song, C. H.; Chun, S. Y.; Kim, H. C.; Chung, B. D.; Lee, W. J.; Kwon, T. S.

1997-07-01

During the period of this study, TASS 1.0 code has been prepared for the non-LOCA licensing and reload safety analyses of the Westinghouse and the Korean Standard Nuclear Power Plants (KSNPP) type reactors operating in Korea. TASS-NPA also has been developed for a real time simulation of the Kori-3/4 transients using on-line graphical interactions. TASS 2.0 code has been further developed to timely apply the TASS 2.0 code for the design certification of the KNGR. The COBRA/RELAP5 code, a multi-dimensional best estimate system code, has been developed by integrating the realistic three-dimensional reactor vessel model with the RELAP5 /MOD3.2 code, a one-dimensional system code. Also, a 3D turbulent two-phase flow analysis code, FEMOTH-TF, has been developed using finite element technique to analyze local thermal hydraulic phenomena in support of the detailed design analysis for the development of the advanced reactors. (author). 84 refs., 27 tabs., 83 figs

Sensitivity Analysis of Wind Plant Performance to Key Turbine Design Parameters: A Systems Engineering Approach; Preprint

Energy Technology Data Exchange (ETDEWEB)

Dykes, K.; Ning, A.; King, R.; Graf, P.; Scott, G.; Veers, P.

2014-02-01

This paper introduces the development of a new software framework for research, design, and development of wind energy systems which is meant to 1) represent a full wind plant including all physical and nonphysical assets and associated costs up to the point of grid interconnection, 2) allow use of interchangeable models of varying fidelity for different aspects of the system, and 3) support system level multidisciplinary analyses and optimizations. This paper describes the design of the overall software capability and applies it to a global sensitivity analysis of wind turbine and plant performance and cost. The analysis was performed using three different model configurations involving different levels of fidelity, which illustrate how increasing fidelity can preserve important system interactions that build up to overall system performance and cost. Analyses were performed for a reference wind plant based on the National Renewable Energy Laboratory's 5-MW reference turbine at a mid-Atlantic offshore location within the United States.

Design and simulation of a plant control system for a GCFR demonstration plant

International Nuclear Information System (INIS)

Estrine, E.A.; Greiner, H.G.

1980-02-01

A plant control system is being designed for a 300 MW(e) Gas Cooled Fast Breeder Reactor (GCFR) demonstration plant. Control analysis is being performed as an integral part of the plant design process to ensure that control requirements are satisfied as the plant design evolves. Plant models and simulations are being developed to generate information necessary to further define control system requirements for subsequent plant design iterations

Disturbance analysis in nuclear power plants

International Nuclear Information System (INIS)

Sillamaa, M.A.

Disturbance analysis is any systematic procedure that helps an operator determine what has failed. This paper describes the typical information currently provided in CANDU power plants to help the operator respond to a disturbance. It presents a simplified model of how an operator could get into trouble, and briefly reviews development work on computerized disturbance analysis systems for nuclear power plants being done in various countries including Canada. Disturbance analysis systems promise to be useful tools in helping operators improve their response to complex situations. However, the originality and complexity of the work for a disturbance analysis system and the need to develop operator confidence and management support require a 'walk before you run' approach

Proceedings: Power Plant Electric Auxiliary Systems Workshop

International Nuclear Information System (INIS)

1992-06-01

The EPRI Power Plant Electric Auxiliary Systems Workshop, held April 24--25, 1991, in Princeton, New Jersey, brought together utilities, architect/engineers, and equipment suppliers to discuss common problems with power plant auxiliary systems. Workshop participants presented papers on monitoring, identifying, and solving problems with auxiliary systems. Panel discussions focused on improving systems and existing and future plants. The solutions presented to common auxiliary system problems focused on practical ideas that can enhance plant availability, reduce maintenance costs, and simplify the engineering process. The 13 papers in these proceedings include: Tutorials on auxiliary electrical systems and motors; descriptions of evaluations, software development, and new technologies used recently by electric utilities; an analysis of historical performance losses caused by power plant auxiliary systems; innovative design concepts for improving auxiliary system performance in future power plants

System Evaluation and Life-Cycle Cost Analysis of a Commercial-Scale High-Temperature Electrolysis Hydrogen Production Plant

Energy Technology Data Exchange (ETDEWEB)

Edwin A. Harvego; James E. O' Brien; Michael G. McKellar

2012-11-01

Results of a system evaluation and lifecycle cost analysis are presented for a commercial-scale high-temperature electrolysis (HTE) central hydrogen production plant. The plant design relies on grid electricity to power the electrolysis process and system components, and industrial natural gas to provide process heat. The HYSYS process analysis software was used to evaluate the reference central plant design capable of producing 50,000 kg/day of hydrogen. The HYSYS software performs mass and energy balances across all components to allow optimization of the design using a detailed process flow sheet and realistic operating conditions specified by the analyst. The lifecycle cost analysis was performed using the H2A analysis methodology developed by the Department of Energy (DOE) Hydrogen Program. This methodology utilizes Microsoft Excel spreadsheet analysis tools that require detailed plant performance information (obtained from HYSYS), along with financial and cost information to calculate lifecycle costs. The results of the lifecycle analyses indicate that for a 10% internal rate of return, a large central commercial-scale hydrogen production plant can produce 50,000 kg/day of hydrogen at an average cost of $2.68/kg. When the cost of carbon sequestration is taken into account, the average cost of hydrogen production increases by $0.40/kg to $3.08/kg.

Improvement on reliability of control system in power plant

International Nuclear Information System (INIS)

Taguchi, S.; Mizumoto, T.; Hirose, Y.; Kashiwai, J.; Takami, I.; Shono, M.; Roji, Y.; Kizaki, S.

1985-01-01

Studies made of Japanese PWR operating experiences have revealed that failures in the control system are the primary causes of unscheduled shutdowns. An attempt has, therefore, been made to improve the reliability of the control system in order to raise the plant reliability. The following are the procedures applied to solve the issue; study of operating experiences, fault tree analysis and failure mode and effects analysis. Improvement measures are developed for the control system whose failure threatens to cause the plant trip during the plant life. These systems are the main feedwater control system, rod control system, pressurizer control system and main steam control system in the primary control system. As a result, the plant unavailability is expected to be reduced significantly by applying the improvements. The improvements are applied to the plants under construction and the operating plants in co-operation with utilities and vendors. (author)

Seismic analysis of the reactor coolant system of PWR nuclear power plants

International Nuclear Information System (INIS)

Borsoi, L.; Sollogoub, P.

1986-01-01

For safety considerations, seismic analyses are performed of the Reactor Coolant System (R.C.S.) of PWR Plants. After a brief description of the R.C.S. and R.C.S. operation, the paper presents the two types of analysis used to determine the effect of earthquake on the R.C.S.: modal spectral analysis and nonlinear time history analysis. The paper finally shows how seismic loadings are combined with other types of loadings and illustrates how the consideration of seismic loads affects R.C.S. design [fr

Proposal of a concept and reliability analysis for a fusion plant magnet protection system

International Nuclear Information System (INIS)

Schnauder, H.; Pamfilie, E.

1993-05-01

The unavailability for the current switch down in case of a demand in the magnet coils of a fusion demonstration plant must be decreased by a few orders of magnitude as compared to the one of experimental facilities. The safety requirements to prevent initiation of event sequences which might lead to the release of radioactivity and energy by the plant must be fulfilled with the same standards as applied in a normally applicable plant. On the basis of this proven technology a general usable magnet protection system will be proposed, which achieves some considerable improvements in the failure detectability as compared to the conventional protection systems. It will be demonstrated by fault tree analysis that the principal demands on safety can be satisfied by that approach. The improvements are achieved by the use of an additional microprocessor supported system for failure detection without being used for initiation of any safety related actions. An influence on a safety action by the additional system therefore is excluded. (orig.) [de

Seismic analysis with FEM for fuel transfer system of PWR nuclear power plant

International Nuclear Information System (INIS)

Jia Xiaofeng; Liu Pengliang; Bi Xiangjun; Ji Shunying

2012-01-01

In the PWR nuclear power plant, the function of the fuel transfer system (FTS) is to transfer the fuel assembly between the reactor building and the fuel building. The seismic analysis of the transfer system structure should be carried out to ensure the safety under OBE and SSE. Therefore, the ANASYS 12.0 software is adopted to construct the finite element analysis model for the fuel transfer system in a million kilowatt nuclear power plant. For the various configurations of FTS in the operating process, the stresses of the main structures, such as the transfer tube, fuel assembly container, fuel conveyor car, lifting frame in the reactor building, lifting frame in the fuel building, support and guide structure of conveyor car and the lifting frame in both buildings, are computed. The stresses are combined with the method of square root of square sum (SRSS) and assessed under various seismic conditions based on RCCM code, the results of the assessment satisfy the code. The results show that the stresses of the fuel transfer system structure meet the strength requirement, meanwhile, it can withstand the earthquake well. (authors)

Modular Trough Power Plant Cycle and Systems Analysis

Energy Technology Data Exchange (ETDEWEB)

Price, H.; Hassani, V.

2002-01-01

This report summarizes an analysis to reduce the cost of power production from modular concentrating solar power plants through a relatively new and exciting concept that merges two mature technologies to produce distributed modular electric power in the range of 500 to 1,500 kWe. These are the organic Rankine cycle (ORC) power plant and the concentrating solar parabolic (CSP) trough technologies that have been developed independent of each other over many years.

LISSAT Analysis of a Generic Centrifuge Enrichment Plant

International Nuclear Information System (INIS)

Lambert, H; Elayat, H A; O'Connell, W J; Szytel, L; Dreicer, M

2007-01-01

The U.S. Department of Energy (DOE) is interested in developing tools and methods for use in designing and evaluating safeguards systems for current and future plants in the nuclear power fuel cycle. The DOE is engaging several DOE National Laboratories in efforts applied to safeguards for chemical conversion plants and gaseous centrifuge enrichment plants. As part of the development, Lawrence Livermore National Laboratory has developed an integrated safeguards system analysis tool (LISSAT). This tool provides modeling and analysis of facility and safeguards operations, generation of diversion paths, and evaluation of safeguards system effectiveness. The constituent elements of diversion scenarios, including material extraction and concealment measures, are structured using directed graphs (digraphs) and fault trees. Statistical analysis evaluates the effectiveness of measurement verification plans and randomly timed inspections. Time domain simulations analyze significant scenarios, especially those involving alternate time ordering of events or issues of timeliness. Such simulations can provide additional information to the fault tree analysis and can help identify the range of normal operations and, by extension, identify additional plant operational signatures of diversions. LISSAT analyses can be used to compare the diversion-detection probabilities for individual safeguards technologies and to inform overall strategy implementations for present and future plants. Additionally, LISSAT can be the basis for a rigorous cost-effectiveness analysis of safeguards and design options. This paper will describe the results of a LISSAT analysis of a generic centrifuge enrichment plant. The paper will describe the diversion scenarios analyzed and the effectiveness of various safeguards systems alternatives

An advanced computational algorithm for systems analysis of tokamak power plants

International Nuclear Information System (INIS)

Dragojlovic, Zoran; Rene Raffray, A.; Najmabadi, Farrokh; Kessel, Charles; Waganer, Lester; El-Guebaly, Laila; Bromberg, Leslie

2010-01-01

A new computational algorithm for tokamak power plant system analysis is being developed for the ARIES project. The objective of this algorithm is to explore the most influential parameters in the physical, technological and economic trade space related to the developmental transition from experimental facilities to viable commercial power plants. This endeavor is being pursued as a new approach to tokamak systems studies, which examines an expansive, multi-dimensional trade space as opposed to traditional sensitivity analyses about a baseline design point. The new ARIES systems code consists of adaptable modules which are built from a custom-made software toolbox using object-oriented programming. The physics module captures the current tokamak physics knowledge database including modeling of the most-current proposed burning plasma experiment design (FIRE). The engineering model accurately reflects the intent and design detail of the power core elements including accurate and adjustable 3D tokamak geometry and complete modeling of all the power core and ancillary systems. Existing physics and engineering models reflect both near-term as well as advanced technology solutions that have higher performance potential. To fully assess the impact of the range of physics and engineering implementations, the plant cost accounts have been revised to reflect a more functional cost structure, supported by an updated set of costing algorithms for the direct, indirect, and financial cost accounts. All of these features have been validated against the existing ARIES-AT baseline case. The present results demonstrate visualization techniques that provide an insight into trade space assessment of attractive steady-state tokamaks for commercial use.

Promises in intelligent plant control systems

International Nuclear Information System (INIS)

Otaduy, P.J.

1987-01-01

The control system is the brain of a power plant. The traditional goal of control systems has been productivity. However, in nuclear power plants the potential for disaster requires safety to be the dominant concern, and the worldwide political climate demands trustworthiness for nuclear power plants. To keep nuclear generation as a viable option for power in the future, trust is the essential critical goal which encompasses all others. In most of today's nuclear plants the control system is a hybrid of analog, digital, and human components that focuses on productivity and operates under the protective umbrella of an independent engineered safety system. Operation of the plant is complex, and frequent challenges to the safety system occur which impact on their trustworthiness. Advances in nuclear reactor design, computer sciences, and control theory, and in related technological areas such as electronics and communications as well as in data storage, retrieval, display, and analysis have opened a promise for control systems with more acceptable human brain-like capabilities to pursue the required goals. This paper elaborates on the promise of futuristic nuclear power plants with intelligent control systems and addresses design requirements and implementation approaches

Food-Energy Interactive Tradeoff Analysis of Sustainable Urban Plant Factory Production Systems

OpenAIRE

Li-Chun Huang; Yu-Hui Chen; Ya-Hui Chen; Chi-Fang Wang; Ming-Che Hu

2018-01-01

This research aims to analyze the food–energy interactive nexus of sustainable urban plant factory systems. Plant factory systems grow agricultural products within artificially controlled growing environment and multi-layer vertical growing systems. The system controls the supply of light, temperature, humidity, nutrition, water, and carbon dioxide for growing plants. Plant factories are able to produce consistent and high-quality agricultural products within less production space for urban a...

Suitability review of FMEA and reliability analysis for digital plant protection system and digital engineered safety features actuation system

Energy Technology Data Exchange (ETDEWEB)

Kim, I. S.; Kim, T. K.; Kim, M. C.; Kim, B. S.; Hwang, S. W.; Ryu, K. C. [Hanyang Univ., Seoul (Korea, Republic of)

2000-11-15

Of the many items that should be checked out during a review stage of the licensing application for the I and C system of Ulchin 5 and 6 units, this report relates to a suitability review of the reliability analysis of Digital Plant Protection System (DPPS) and Digital Engineered Safety Features Actuation System (DESFAS). In the reliability analysis performed by the system designer, ABB-CE, fault tree analysis was used as the main methods along with Failure Modes and Effect Analysis (FMEA). However, the present regulatory technique dose not allow the system reliability analysis and its results to be appropriately evaluated. Hence, this study was carried out focusing on the following four items ; development of general review items by which to check the validity of a reliability analysis, and the subsequent review of suitability of the reliability analysis for Ulchin 5 and 6 DPPS and DESFAS L development of detailed review items by which to check the validity of an FMEA, and the subsequent review of suitability of the FMEA for Ulchin 5 and 6 DPPS and DESFAS ; development of detailed review items by which to check the validity of a fault tree analysis, and the subsequent review of suitability of the fault tree for Ulchin 5 and 6 DPPS and DESFAS ; an integrated review of the safety and reliability of the Ulchin 5 and 6 DPPS and DESFAS based on the results of the various reviews above and also of a reliability comparison between the digital systems and the comparable analog systems, i.e., and analog Plant Protection System (PPS) and and analog Engineered Safety Features Actuation System (ESFAS). According to the review mentioned above, the reliability analysis of Ulchin 5 and 6 DPPS and DESFAS generally satisfies the review requirements. However, some shortcomings of the analysis were identified in our review such that the assumed test periods for several equipment were not properly incorporated in the analysis, and failures of some equipment were not included in the

Effect Analysis of Digital I and C Systems on Plant Safety based on Fault-Tree Analysis

International Nuclear Information System (INIS)

Lee, Seung Jun; Jung, Wondea

2014-01-01

Deterioration and an inadequate supply of components of analog I and C systems have led to inefficient and costly maintenance. Moreover, since the fast evolution of digital technology has enabled more reliable functions to be designed for NPP safety, the transition from analog to digital has been accelerated. Owing to the distinguishable characteristics of digital I and C systems, a reliability analysis of digital systems has become an important element of a probabilistic safety assessment (PSA). Digital I and C systems have unique characteristics such as fault-tolerant techniques and software. However, these features have not been properly considered yet in most NPP PSA models. The effect of digital I and C systems should be evaluated by comparing them to that of analog I and C systems. Before installing a digital I and C system, even though it is expected that the plant safety can be improved through the advantageous features of digital I and C systems, it should be validated whether the total NPP safety is better than analog systems or is the same at least. In this work, the fault-tree (FT) technique, which is most widely used in a PSA, was used to compare the effects of analog and digital I and C systems. From a case study, the results of plant safety were compared. In this work, the effect of a digital RPS was evaluated by comparing it to that of an analog RPS based on the FT models. In the evaluation results, it was observed that digital RPS has a positive effect on reducing the system unavailability. The analysis results can be used for the development of a guide for evaluating digital I and C systems and reliability requirements

Modular Modeling System (MMS) code: a versatile power plant analysis package

International Nuclear Information System (INIS)

Divakaruni, S.M.; Wong, F.K.L.

1987-01-01

The basic version of the Modular Modeling System (MMS-01), a power plant systems analysis computer code jointly developed by the Nuclear Power and the Coal Combustion Systems Divisions of the Electric Power Research Institute (EPRI), has been released to the utility power industry in April 1983 at a code release workshop held in Charlotte, North Carolina. Since then, additional modules have been developed to analyze the Pressurized Water Reactors (PWRs) and the Boiling Water Reactors (BWRs) when the safety systems are activated. Also, a selected number of modules in the MMS-01 library have been modified to allow the code users more flexibility in constructing plant specific systems for analysis. These new PWR and BWR modules constitute the new MMS library, and it includes the modifications to the MMS-01 library. A year and half long extensive code qualification program of this new version of the MMS code at EPRI and the contractor sites, back by further code testing in an user group environment is culminating in the MMS-02 code release announcement seminar. At this seminar, the results of user group efforts and the code qualification program will be presented in a series of technical sessions. A total of forty-nine papers will be presented to describe the new code features and the code qualification efforts. For the sake of completion, an overview of the code is presented to include the history of the code development, description of the MMS code and its structure, utility engineers involvement in MMS-01 and MMS-02 validations, the enhancements made in the last 18 months to the code, and finally the perspective on the code future in the fossil and nuclear industry

PWR systems transient analysis

International Nuclear Information System (INIS)

Kennedy, M.F.; Peeler, G.B.; Abramson, P.B.

1985-01-01

Analysis of transients in pressurized water reactor (PWR) systems involves the assessment of the response of the total plant, including primary and secondary coolant systems, steam piping and turbine (possibly including the complete feedwater train), and various control and safety systems. Transient analysis is performed as part of the plant safety analysis to insure the adequacy of the reactor design and operating procedures and to verify the applicable plant emergency guidelines. Event sequences which must be examined are developed by considering possible failures or maloperations of plant components. These vary in severity (and calculational difficulty) from a series of normal operational transients, such as minor load changes, reactor trips, valve and pump malfunctions, up to the double-ended guillotine rupture of a primary reactor coolant system pipe known as a Large Break Loss of Coolant Accident (LBLOCA). The focus of this paper is the analysis of all those transients and accidents except loss of coolant accidents

Fault tree analysis for reactor systems

International Nuclear Information System (INIS)

Crosetti, P.A.

1971-01-01

Reliability analysis is playing an increasingly important role in quantitative assessment of system performance for assuring nuclear safety, improving plant performance and plant life, and reducing plant operating costs. The complexity of today's nuclear plants warrant the use of techniques which will provide a comprehensive evaluation of systems in their total context. In particular, fault tree analysis with probability evaluation can play a key role in assuring nuclear safety, in improving plant performance and plant life, and in reducing plant operating costs. The technique provides an all inclusive, versatile mathematical tool for analyzing complex systems. Its application can include a complete plant as well as any of the systems and subsystems. Fault tree analysis provides an objective basis for analyzing system design, performing trade-off studies, analyzing common mode failures, demonstrating compliance with AEC requirements, and justifying system changes or additions. The logic of the approach makes it readily understandable and, therefore, it serves as an effective visibility tool for both engineering and management. (U.S.)

RETRAN code analysis of Tsuruga-2 plant chemical volume control system (CVCS) reactor coolant leakage incident

International Nuclear Information System (INIS)

Kawai, H.

2001-01-01

JAPC purchased RETRAN, a program for transient thermal hydraulic analysis of complex fluid flow system, from the U.S. Electric Power Research Institute in 1992. Since then, JAPC has been utilizing RETRAN to evaluate safety margins of actual plant operation, in coping with troubles (investigating trouble causes and establishing countermeasures), and supporting reactor operation (reviewing operational procedures etc.). In this paper, a result of plant analysis performed on a CVCS reactor primary coolant leakage incident which occurred at JAPC's Tsuruga-2 plant (4-loop PWR, 3423 MWt, 1160 MW) on July 12 of 1999 and, based on the result, we made a plan to modify our operational procedure for reactor primary coolant leakage events in order to make earlier plant shutdown and this reduced primary coolant leakage. (author)

Function analysis of nuclear power plants for developing of man-machine interface system for Korean next generation reactor

Energy Technology Data Exchange (ETDEWEB)

Goo, In Soo; Kim, Jang Yyul; Kim, Jung Soo; Kim, Chang Hoi; Na, Nan Joo; Park, Keun Ok; Park, Won Man; Park, Jae Chang; Suh, Sang Moon; Oh, In Suk; Lee, Dong Young; Lee, Yong Hee; Cha, Kyung Ho; Chun, Se Woo; Hur, Sup [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Jung, Kyung Hoon [Korea Electric Power Co., Seoul (Korea, Republic of); Park, Yeon Sik; Lee, Bum Joo [Korea Power Engineering Company, Inc., Seoul (Korea, Republic of)

1995-02-01

In this study, we developed the methodology and implementation plant of function analysis and performed function analysis, which is one of the major activities for the development of Man-Machine Interface System of the KNGR. Identifying the functions of existing plants followed by structuring the functions, we established functions structured at the conceptual and middle levels. This structure was further checked if it would satisfy regulatory requirements and developed to include the aspects of plant performance and other plant features to emphasize its practicality for the application to the design. (Author) 13 refs., 8 figs., 4 tabs.

Function analysis of nuclear power plants for developing of man-machine interface system for Korean next generation reactor

International Nuclear Information System (INIS)

Goo, In Soo; Kim, Jang Yyul; Kim, Jung Soo; Kim, Chang Hoi; Na, Nan Joo; Park, Keun Ok; Park, Won Man; Park, Jae Chang; Suh, Sang Moon; Oh, In Suk; Lee, Dong Young; Lee, Yong Hee; Cha, Kyung Ho; Chun, Se Woo; Hur, Sup; Jung, Kyung Hoon; Park, Yeon Sik; Lee, Bum Joo

1995-02-01

In this study, we developed the methodology and implementation plant of function analysis and performed function analysis, which is one of the major activities for the development of Man-Machine Interface System of the KNGR. Identifying the functions of existing plants followed by structuring the functions, we established functions structured at the conceptual and middle levels. This structure was further checked if it would satisfy regulatory requirements and developed to include the aspects of plant performance and other plant features to emphasize its practicality for the application to the design. (Author) 13 refs., 8 figs., 4 tabs

PlantCV v2: Image analysis software for high-throughput plant phenotyping

Directory of Open Access Journals (Sweden)

Malia A. Gehan

2017-12-01

Full Text Available Systems for collecting image data in conjunction with computer vision techniques are a powerful tool for increasing the temporal resolution at which plant phenotypes can be measured non-destructively. Computational tools that are flexible and extendable are needed to address the diversity of plant phenotyping problems. We previously described the Plant Computer Vision (PlantCV software package, which is an image processing toolkit for plant phenotyping analysis. The goal of the PlantCV project is to develop a set of modular, reusable, and repurposable tools for plant image analysis that are open-source and community-developed. Here we present the details and rationale for major developments in the second major release of PlantCV. In addition to overall improvements in the organization of the PlantCV project, new functionality includes a set of new image processing and normalization tools, support for analyzing images that include multiple plants, leaf segmentation, landmark identification tools for morphometrics, and modules for machine learning.

The plant-window system

International Nuclear Information System (INIS)

Wood, R.T.; Mullens, J.A.; Naser, J.A.

1995-01-01

Power plant data, and the information that can be derived from it, provide the link to the plant through which the operations, maintenance and engineering staff understand and manage plant performance. The increasing use of computer technology in the U.S. nuclear power industry has greatly expanded the capability to obtain, analyze, and present data about the plant to station personnel. However, it is necessary to transform the vast quantity of available data into clear, concise, and coherent information that can be readily accessed and used throughout the plant. This need can be met by an integrated computer workstation environment that provides the necessary information and software applications, in a manner that can be easily understood and used, to the proper users throughout the plant. As part of a Cooperative Research and Development Agreement with the Electric Power Research Institute, the Oak Ridge National Laboratory has developed functional requirements for a Plant-Wide Integrated Environment Distributed On Workstations (Plant-Window) System. The Plant-Window System (PWS) can serve the needs of operations, engineering, and maintenance personnel at nuclear power stations by providing integrated data and software applications (e.g., monitoring, analysis, diagnosis, and control applications) within a common environment. The PWS requirements identify functional capabilities and provide guidelines for standardized hardware, software, and display interfaces to define a flexible computer environment that permits a tailored implementation of workstation capabilities and facilitates future upgrades

Nitrogen-system safety study: Portsmouth Gaseous Diffusion Plant

International Nuclear Information System (INIS)

1982-07-01

The Department of Energy has primary responsibility for the safety of operations at DOE-owned nuclear facilities. The guidelines for the analysis of credible accidents are outlined in DOE Order 5481.1. DOE has requested that existing plant facilities and operations be reviewed for potential safety problems not covered by standard industrial safety procedures. This review is being conducted by investigating individual facilities and documenting the results in Safety Study Reports which will be compiled to form the Existing Plant Final Safety Analysis Report which is scheduled for completion in September, 1984. This Safety Study documents the review of the Plant Nitrogen System facilities and operations and consists of Section 4.0, Facility and Process Description, and Section 5.0, Accident Analysis, of the Final Safety Analysis Report format. The existing nitrogen system consists of a Superior Air Products Company Type D Nitrogen Plant, nitrogen storage facilities, vaporization facilities and a distribution system. The system is designed to generate and distribute nitrogen gas used in the cascade for seal feed, buffer systems, and for servicing equipment when exceptionally low dew points are required. Gaseous nitrogen is also distributed to various process auxiliary buildings. The average usage is approximately 130,000 standard cubic feet per day

Nuclear power plant control room crew task analysis database: SEEK system. Users manual

International Nuclear Information System (INIS)

Burgy, D.; Schroeder, L.

1984-05-01

The Crew Task Analysis SEEK Users Manual was prepared for the Office of Nuclear Regulatory Research of the US Nuclear Regulatory Commission. It is designed for use with the existing computerized Control Room Crew Task Analysis Database. The SEEK system consists of a PR1ME computer with its associated peripherals and software augmented by General Physics Corporation SEEK database management software. The SEEK software programs provide the Crew Task Database user with rapid access to any number of records desired. The software uses English-like sentences to allow the user to construct logical sorts and outputs of the task data. Given the multiple-associative nature of the database, users can directly access the data at the plant, operating sequence, task or element level - or any combination of these levels. A complete description of the crew task data contained in the database is presented in NUREG/CR-3371, Task Analysis of Nuclear Power Plant Control Room Crews (Volumes 1 and 2)

Reliability analysis of self-supply system of V-1 nuclear power plant

International Nuclear Information System (INIS)

Kuklik, B.

The results are summarized of the fault tree analysis of the V-1 power plant self-consumption system. The 6 kV busbars providing power for the main circulating pumps, the steam generator feed pumps and other important components including the 0.4 kV busbars are of the highest importance for nuclear safety. A fault tree analysis was also made of the emergency core cooling system of the reactor. Dangerous faults are defined and fault trees are developed. A brief description is given of the calculation algorithm for a digital computer. Some results are discussed. The calculated reliability of the emergency core cooling system is 10 5 years, of the 6 kV busbars it is 6.6x10 4 years. In case of a permanent or a long-term outage of the 220 kV stand-bye power supply, the system reliability is reduced to 7x10 2 years. (Z.M.)

System seismic analysis of an innovative primary system for a large pool type LMFBR plant

International Nuclear Information System (INIS)

Pan, Y.C.; Wu, T.S.; Cha, B.K.; Burelbach, J.; Seidensticker, R.

1984-01-01

The system seismic analysis of an innovative primary system for a large pool type liquid metal fast breeder reactor (LMFBR) plant is presented. In this primary system, the reactor core is supported in a way which differs significantly from that used in previous designs. The analytical model developed for this study is a three-dimensional finite element model including one-half of the primary system cut along the plane of symmetry. The model includes the deck and deck mounted components,the reactor vessel, the core support structure, the core barrel, the radial neutron shield, the redan, and the conical support skirt. The sodium contained in the primary system is treated as a lumped mass appropriately distributed among various components. The significant seismic behavior as well as the advantages of this primary system design are discussed in detail

Reliability analysis of microcomputer boards and computer based systems important to safety of nuclear plants

International Nuclear Information System (INIS)

Shrikhande, S.V.; Patil, V.K.; Ganesh, G.; Biswas, B.; Patil, R.K.

2010-01-01

Computer Based Systems (CBS) are employed in Indian nuclear plants for protection, control and monitoring purpose. For forthcoming CBS, Reactor Control Division has designed and developed a new standardized family of microcomputer boards qualified to stringent requirements of nuclear industry. These boards form the basic building blocks of CBS. Reliability analysis of these boards is being carried out using analysis package based on MIL-STD-217Plus methodology. The estimated failure rate values of these standardized microcomputer boards will be useful for reliability assessment of these systems. The paper presents reliability analysis of microcomputer boards and case study of a CBS system built using these boards. (author)

A new balance-of-plant model for the SASSYS-1 LMR systems analysis code

International Nuclear Information System (INIS)

Briggs, L.L.

1989-01-01

A balance-of-plant model has been added to the SASSYS-1 liquid metal reactor systems analysis code. Until this addition, the only waterside component which SASSYS-1 could explicitly model was the water side of a steam generator, with the remainder of the water side represented by boundary conditions on the steam generator. The balance-of-plant model is based on the model used for the sodium side of the plant. It will handle subcooled liquid water, superheated steam, and saturated two-phase fluid. With the exception of heated flow paths in heaters, the model assumes adiabatic conditions along flow paths; this assumption simplifies the solution procedure while introducing very little error for a wide range of reactor plant problems. Only adiabatic flow is discussed in this report. 3 refs., 4 figs

Analysis of reactor trips involving balance-of-plant failures

International Nuclear Information System (INIS)

Seth, S.; Skinner, L.; Ettlinger, L.; Lay, R.

1986-01-01

The relatively high frequency of plant transients leading to reactor trips at nuclear power plants in the US is of economic and safety concern to the industry. A majority of such transients is due to failures in the balance-of-plant (BOP) systems. As a part of a study conducted for the US Nuclear Regulatory Commission, Mitre has carried out a further analysis of the BOP failures associated with reactor trips. The major objectives of the analysis were to examine plant-to-plant variations in BOP-related trips, to understand the causes of failures, and to determine the extent of any associated safety system challenges. The analysis was based on the Licensee Event Reports submitted on all commercial light water reactors during the 2-yr period, 1984-1985

Chemistry management system for nuclear power plants

International Nuclear Information System (INIS)

Nagasawa, Katsumi; Maeda, Katsuji

1998-01-01

Recently, the chemistry management in the nuclear power plants has been changing from the problem solution to the predictive diagnosis and maintenance. It is important to maintain the integrity of plant operation by an adequate chemistry control. For these reasons, many plant operation data and chemistry analysis data should be collected and treated effectively to evaluate chemistry condition of the nuclear power plants. When some indications of chemistry anomalies occur, quick and effective root cause evaluation and countermeasures should be required. The chemistry management system has been developed as to provide sophisticate chemistry management in the nuclear power plants. This paper introduces the concept and functions of the chemistry management system for the nuclear power plants. (author)

Coal handling system structural analysis for modifications or plant life extension

International Nuclear Information System (INIS)

Dufault, A.; Weider, F.; Doyle, P.

1989-01-01

One neglected aspect of plant modification or life extension is the extent to which previous projects may have affected the integrity of existing structures. During the course of a project to backfit fire protection facilities to existing coal handling systems, it was found that past modifications had added loads to existing coal handling structures which exceeded the available design margin. This paper describes the studies that discovered the original problem areas, as well as the detailed analysis and design considerations used to repair these structures

Development of the simulation system IMPACT for analysis of nuclear power plant severe accidents

International Nuclear Information System (INIS)

Naitoh, Masanori; Ujita, Hiroshi; Nagumo, Hiroichi

1997-01-01

The Nuclear Power Engineering Corporation (NUPEC) has initiated a long-term program to develop the simulation system IMPACT for analysis of hypothetical severe accidents in nuclear power plants. IMPACT employs advanced methods of physical modeling and numerical computation, and can simulate a wide spectrum of senarios ranging from normal operation to hypothetical, beyond-design-basis-accident events. Designed as a large-scale system of interconnected, hierarchical modules, IMPACT's distinguishing features include mechanistic models based on first principles and high speed simulation on parallel processing computers. The present plan is a ten-year program starting from 1993, consisting of the initial one-year of preparatory work followed by three technical phases: Phase-1 for development of a prototype system; Phase-2 for completion of the simulation system, incorporating new achievements from basic studies; and Phase-3 for refinement through extensive verification and validation against test results and available real plant data

System identification of the Arabidopsis plant circadian system

Science.gov (United States)

Foo, Mathias; Somers, David E.; Kim, Pan-Jun

2015-02-01

The circadian system generates an endogenous oscillatory rhythm that governs the daily activities of organisms in nature. It offers adaptive advantages to organisms through a coordination of their biological functions with the optimal time of day. In this paper, a model of the circadian system in the plant Arabidopsis (species thaliana) is built by using system identification techniques. Prior knowledge about the physical interactions of the genes and the proteins in the plant circadian system is incorporated in the model building exercise. The model is built by using primarily experimentally-verified direct interactions between the genes and the proteins with the available data on mRNA and protein abundances from the circadian system. Our analysis reveals a great performance of the model in predicting the dynamics of the plant circadian system through the effect of diverse internal and external perturbations (gene knockouts and day-length changes). Furthermore, we found that the circadian oscillatory rhythm is robust and does not vary much with the biochemical parameters except those of a light-sensitive protein P and a transcription factor TOC1. In other words, the circadian rhythmic profile is largely a consequence of the network's architecture rather than its particular parameters. Our work suggests that the current experimental knowledge of the gene-to-protein interactions in the plant Arabidopsis, without considering any additional hypothetical interactions, seems to suffice for system-level modeling of the circadian system of this plant and to present an exemplary platform for the control of network dynamics in complex living organisms.

Seismic analysis of nuclear power plants

International Nuclear Information System (INIS)

Halbritter, A.L.

1984-01-01

Nuclear Power Plants require exceptional safety guarantees which are reflected in a rigorous control of the employed materials, advanced construction technology, sophisticated methods of analysis and consideration of non conventional load cases such as the earthquake loading. In this paper, the current procedures used in the seismic analysis of Nuclear Power Plants are presented. The seismic analysis of the structures has two objectives: the determination of forces in the structure in order to design it against earthquakes and the generation of floor response spectra to be used in the design of mechanical and electrical components and piping systems. (Author) [pt

The assisting system for uranium enrichment plant operation

International Nuclear Information System (INIS)

Nakazawa, Hiroaki; Yamamoto, Fumio

1990-01-01

We have been developing an operation assisting system, partially supported by AI system, for uranium enrichment plant. The AI system is a proto-type system aiming a final one which can be applied to any future large uranium enrichment plant and also not only to specific operational area but also to complex and multi-phenomenon operational area. An existing AI system, for example facility diagnostic system that utilizes the result of CCT analysis as knowledge base, has weakness in flexibility and potentiality. To build AI system, we have developed the most suitable knowledge representations using deep knowledge for each facility or operation of uranium enrichment plant. This paper describes our AI proto-type system adopting several knowledge representations that can represent an uranium enrichment plant's operation with deep knowledge. (author)

The System 80+ Standard Plant design control document. Volume 21

International Nuclear Information System (INIS)

1997-01-01

This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume contains parts 1--10 of section 19 (Probabilistic Risk Assessment) of the ADM Design and Analysis. Topics covered are: methodology; initiating event evaluation; accident sequence determination; data analysis; systems analysis; external events analysis; shutdown risk assessment; accident sequence quantification; and sensitivity analysis. Also included in this volume are Appendix 19.8A Shutdown Risk Assessment and Appendix A to Appendix 19.8A Request for Information

The software safety analysis based on SFTA for reactor power regulating system in nuclear power plant

International Nuclear Information System (INIS)

Liu Zhaohui; Yang Xiaohua; Liao Longtao; Wu Zhiqiang

2015-01-01

The digitalized Instrumentation and Control (I and C) system of Nuclear power plants can provide many advantages. However, digital control systems induce new failure modes that differ from those of analog control systems. While the cost effectiveness and flexibility of software is widely recognized, it is very difficult to achieve and prove high levels of dependability and safety assurance for the functions performed by process control software, due to the very flexibility and potential complexity of the software itself. Software safety analysis (SSA) was one way to improve the software safety by identify the system hazards caused by software failure. This paper describes the application of a software fault tree analysis (SFTA) at the software design phase. At first, we evaluate all the software modules of the reactor power regulating system in nuclear power plant and identify various hazards. The SFTA was applied to some critical modules selected from the previous step. At last, we get some new hazards that had not been identified in the prior processes of the document evaluation which were helpful for our design. (author)

Nuclear plants gain integrated information systems

International Nuclear Information System (INIS)

Villavicencio-Ramirez, A.; Rodriquez-Alvarez, J.M.

1994-01-01

With the objective of simplifying the complex mesh of computing devices employed within nuclear power plants, modern technology and integration techniques are being used to form centralized (but backed up) databases and distributed processing and display networks. Benefits are immediate as a result of the integration and the use of standards. The use of a unique data acquisition and database subsystem optimizes the high costs of engineering, as this task is done only once for the life span of the system. This also contributes towards a uniform user interface and allows for graceful expansion and maintenance. This article features an integrated information system, Sistema Integral de Informacion de Proceso (SIIP). The development of this system enabled the Laguna Verde Nuclear Power plant to fully use the already existing universe of signals and its related engineering during all plant conditions, namely, start up, normal operation, transient analysis, and emergency operation. Integrated systems offer many advantages over segregated systems, and this experience should benefit similar development efforts in other electric power utilities, not only for nuclear but also for other types of generating plants

Plant Phenotype Characterization System

Energy Technology Data Exchange (ETDEWEB)

Daniel W McDonald; Ronald B Michaels

2005-09-09

This report is the final scientific report for the DOE Inventions and Innovations Project: Plant Phenotype Characterization System, DE-FG36-04GO14334. The period of performance was September 30, 2004 through July 15, 2005. The project objective is to demonstrate the viability of a new scientific instrument concept for the study of plant root systems. The root systems of plants are thought to be important in plant yield and thus important to DOE goals in renewable energy sources. The scientific study and understanding of plant root systems is hampered by the difficulty in observing root activity and the inadequacy of existing root study instrumentation options. We have demonstrated a high throughput, non-invasive, high resolution technique for visualizing plant root systems in-situ. Our approach is based upon low-energy x-ray radiography and the use of containers and substrates (artificial soil) which are virtually transparent to x-rays. The system allows us to germinate and grow plant specimens in our containers and substrates and to generate x-ray images of the developing root system over time. The same plant can be imaged at different times in its development. The system can be used for root studies in plant physiology, plant morphology, plant breeding, plant functional genomics and plant genotype screening.

Analysis of design of auxiliary system of Booshehr Nuclear Power Plant

International Nuclear Information System (INIS)

Naseh Hasanzadeh, M.

1999-01-01

Power plant's internal auxiliary system has an important role in its safety operation. Because of the decay heat and safety aspects in the nuclear power plants, this role is more important. In this thesis, operation of the nuclear power plant with PWR reactor is studied and deferent nuclear systems described. In the next section all electrical loads in the Booshehr Nuclear Power Plant identified and feeding methods of each load is determined. by use of the single line diagram of the internal auxiliary system, the nominal rating of all electrical devices as transformers, inverters, Ups, diesel generators and etc. is determined. In the following, short circuit calculations performed and by above conclusion, rating values of circuit breakers is determined. At last the starting problems of electrical motors is studied and the results of motor's behavior at starting moment is discussed

Introducing WISDEM:An Integrated System Modeling for Wind Turbines and Plant (Presentation)

Energy Technology Data Exchange (ETDEWEB)

Dykes, K.; Graf, P.; Scott, G.; Ning, A.; King, R.; Guo, Y.; Parsons, T.; Damiani, R.; Felker, F.; Veers, P.

2015-01-01

The National Wind Technology Center wind energy systems engineering initiative has developed an analysis platform to leverage its research capabilities toward integrating wind energy engineering and cost models across wind plants. This Wind-Plant Integrated System Design & Engineering Model (WISDEM) platform captures the important interactions between various subsystems to achieve a better National Wind Technology Center wind energy systems engineering initiative has developed an analysis platform to leverage its research capabilities toward integrating wind energy engineering and cost models across wind plants. This Wind-Plant Integrated System Design & Engineering Model (WISDEM) platform captures the important interactions between various subsystems to achieve a better understanding of how to improve system-level performance and achieve system-level cost reductions. This work illustrates a few case studies with WISDEM that focus on the design and analysis of wind turbines and plants at different system levels.

The plant expert system (PLEXSYS) development environment

International Nuclear Information System (INIS)

Hashemi, S.; Patterson, L.; Jeffery, M.; Delashmutt, L.

1989-06-01

The PLEXSYS software engineering tool provides an environment with which utility engineers can build and use expert systems for power plant applications. PLEXSYS provides the engineer with access to many powerful Artificial Intelligence methodologies, while retaining an engineering frame of reference and minimizing the need for a formal background in computer science. The principle concept is that the description and understanding of power plant systems centers on graphical forms such as piping and instrumentation diagrams and electrical line diagrams, which define a graphics-based model of plant knowledge that is common to many applications. PLEXSYS provides a model editor that allows the user to construct and modify models of hydraulic, electrical, and information systems in terms of elementary components and their interconnections. Analysis of the resulting schematic models is provided by several functions that perform network analysis, schematic browsing, mathematical modeling and customization of the user interface. 41 figs., 1 tab

RETRAN code analysis of Tsuruga-2 plant chemical volume control system (CVCS) reactor coolant leakage incident

International Nuclear Information System (INIS)

Kawai, Hiroshi

2002-01-01

In the Chemical Volume Control System (CVCS) reactor primary coolant leakage incident, which occurred in Tsuruga-2 (4-loop PWR, 3,423 MWt, 1,160 MWe) on July 12, 1999, it took about 14 hours before the leakage isolation. The delayed leakage isolation and a large amount of leakage have become a social concern. Effective procedure modification was studied. Three betterments were proposed based on a qualitative analysis to reduce the pressure and temperature of the primary loop as fast as possible by the current plant facilities while maintaining enough subcooling of the primary loop. I analyzed the incident with RETRAN code in order to quantitatively evaluate the leakage reduction when these betterments are adopted. This paper is very new because it created a typical analysis method for PWR plant behavior during plant shutdown procedure which conventional RETRAN transient analyses rarely dealt with. Also the event time is very long. To carry out this analysis successfully, I devised new models such as an Residual Heat Removal System (RHR) model etc. and simplified parts of the conventional model. Based on the analysis results, I confirmed that leakage can be reduced by about 30% by adopting these betterments. Then the Japan Atomic Power Company (JAPC) modified the operational procedure for reactor primary coolant leakage events adopting these betterments. (author)

Energy analysis and projecting of power plants

International Nuclear Information System (INIS)

Jirlow, K.

1975-01-01

Energy analysis aims at a better explanation of energy flow and energy exchange at different production processes. In this report the energy budget is analysed for separate nuclear power plants and for expanding systems of power plants. A mathematical model is developed for linear and exponential expanding of nuclear power. The profitableness for nuclear power plants in Sweden is considered to be good. (K.K.)

Methodology and application of surrogate plant PRA analysis to the Rancho Seco Power Plant: Final report

International Nuclear Information System (INIS)

Gore, B.F.; Huenefeld, J.C.

1987-07-01

This report presents the development and the first application of generic probabilistic risk assessment (PRA) information for identifying systems and components important to public risk at nuclear power plants lacking plant-specific PRAs. A methodology is presented for using the results of PRAs for similar (surrogate) plants, along with plant-specific information about the plant of interest and the surrogate plants, to infer important failure modes for systems of the plant of interest. This methodology, and the rationale on which it is based, is presented in the context of its application to the Rancho Seco plant. The Rancho Seco plant has been analyzed using PRA information from two surrogate plants. This analysis has been used to guide development of considerable plant-specific information about Rancho Seco systems and components important to minimizing public risk, which is also presented herein

Dynamic analysis of Boushehr Nuclear Power Plant in connected to grid system

International Nuclear Information System (INIS)

Karimi Fard, A.

1999-01-01

Models of generating with the pressurized water reactor (PWR) have been developed for simulating. the plant dynamics under system disturbances. These models include power plant, energy sources, turbine, transmission system and control system such as Avr and govern and other local control devices. Simulink toolbox of Matlab software is used for simulations. The study is mainly based on the Bushehr Nuclear Power Plants (BNPP) parameters. Assuming that BNPP is connected to infinite bus with double tie line. Four cases are studied to examine the internal dynamic behavior of BNPP. First and second cases are used to load following studies in nuclear power plant. Another cases are used to study the dynamic behavior after short circuit fault and line outages in transmission systems. The results discussed in the thesis

Thermoeconomic analysis of power plants. Final report

Energy Technology Data Exchange (ETDEWEB)

Tsatsaronis, G.; Winhold, M.

1984-08-01

In this report, the concept of exergy and the general methodology of the exergetic analysis and the thermoeconomic (combined exergetic and economic) analysis of energy conversion systems are presented. The THESIS (THermodynamic and Economc SImulation System) computer program used for these analyses is briefly described. Detailed mass, energy, exergy and money balances for a reference steam power plant (Harry Allen Station) are shown. The effect of the most important process parameters on the overall efficiency is investigated. A year-by-year and a levelized revenue requirement analysis are presented. The costs of exergy losses are compared with the capital costs and other expenses due to owning and operating each particular plant component. The question whether it is profitable to reduce the exergy losses by increasing these costs and vice versa is investigated. A cost sensitivity analysis including the effect of coal price and average annual capacity factor is performed. The methodology applied in this report appears to be useful in analyzing and evaluating energy conversion systems. The analyses presented here allow identification and evaluation of the inefficiencies and the opportunities for improvement of an energy conversion process. Results indicate that modifications in certain process parameters can lead to a decrease in the cost of electricity produced by the reference plant.

Interim reliability-evaluation program: analysis of the Browns Ferry, Unit 1, nuclear plant. Appendix B - system descriptions and fault trees

International Nuclear Information System (INIS)

Mays, S.E.; Poloski, J.P.; Sullivan, W.H.; Trainer, J.E.; Bertucio, R.C.; Leahy, T.J.

1982-07-01

This report describes a risk study of the Browns Ferry, Unit 1, nuclear plant. The study is one of four such studies sponsored by the NRC Office of Research, Division of Risk Assessment, as part of its Interim Reliability Evaluation Program (IREP), Phase II. This report is contained in four volumes: a main report and three appendixes. Appendix B provides a description of Browns Ferry, Unit 1, plant systems and the failure evaluation of those systems as they apply to accidents at Browns Ferry. Information is presented concerning front-line system fault analysis; support system fault analysis; human error models and probabilities; and generic control circuit analyses

PSA effect analysis of a design modification of the auxiliary feedwater system for a Westinghouse type plant

International Nuclear Information System (INIS)

Bae, Yeon Kyoung; Lee, Eun Chan

2012-01-01

The auxiliary feedwater system is an important system used to mitigate most accidents considered in probabilistic safety assessment (PSA). The reference plant has produced electric power for about thirty years. Due to age related deterioration and lack of parts, a turbine driven auxiliary feedwater pump (TD AFWP), some valves, and piping of the auxiliary feedwater system should be replaced. This change includes relocation of some valves, installation of valves for maintenance of the steam generator, and a new cross tie line. According to the design change, the Final Safety Analysis Report (FSAR) has been revised. Therefore, this design modification affects the PSA. It is thus necessary to assess the improvement of plant safety. In this paper, the impact of the design change of the auxiliary feedwater system on the PSA is assessed. The results demonstrate that this modification considering the plant safety decreased the total CDF

Risk analysis of highly combustible gas storage, supply, and distribution systems in PWR plants

International Nuclear Information System (INIS)

Simion, G.P.; VanHorn, R.L.; Smith, C.L.; Bickel, J.H.; Sattison, M.B.; Bulmahn, K.D.

1993-06-01

This report presents the evaluation of the potential safety concerns for pressurized water reactors (PWRs) identified in Generic Safety Issue 106, Piping and the Use of Highly Combustible Gases in Vital Areas. A Westinghouse four-loop PWR plant was analyzed for the risk due to the use of combustible gases (predominantly hydrogen) within the plant. The analysis evaluated an actual hydrogen distribution configuration and conducted several sensitivity studies to determine the potential variability among PWRs. The sensitivity studies were based on hydrogen and safety-related equipment configurations observed at other PWRs within the United States. Several options for improving the hydrogen distribution system design were identified and evaluated for their effect on risk and core damage frequency. A cost/benefit analysis was performed to determine whether alternatives considered were justifiable based on the safety improvement and economics of each possible improvement

PGSB/MIPS PlantsDB Database Framework for the Integration and Analysis of Plant Genome Data.

Science.gov (United States)

Spannagl, Manuel; Nussbaumer, Thomas; Bader, Kai; Gundlach, Heidrun; Mayer, Klaus F X

2017-01-01

Plant Genome and Systems Biology (PGSB), formerly Munich Institute for Protein Sequences (MIPS) PlantsDB, is a database framework for the integration and analysis of plant genome data, developed and maintained for more than a decade now. Major components of that framework are genome databases and analysis resources focusing on individual (reference) genomes providing flexible and intuitive access to data. Another main focus is the integration of genomes from both model and crop plants to form a scaffold for comparative genomics, assisted by specialized tools such as the CrowsNest viewer to explore conserved gene order (synteny). Data exchange and integrated search functionality with/over many plant genome databases is provided within the transPLANT project.

Reliability of emergency ac power systems at nuclear power plants

International Nuclear Information System (INIS)

Battle, R.E.; Campbell, D.J.

1983-07-01

Reliability of emergency onsite ac power systems at nuclear power plants has been questioned within the Nuclear Regulatory Commission (NRC) because of the number of diesel generator failures reported by nuclear plant licensees and the reactor core damage that could result from diesel failure during an emergency. This report contains the results of a reliability analysis of the onsite ac power system, and it uses the results of a separate analysis of offsite power systems to calculate the expected frequency of station blackout. Included is a design and operating experience review. Eighteen plants representative of typical onsite ac power systems and ten generic designs were selected to be modeled by fault trees. Operating experience data were collected from the NRC files and from nuclear plant licensee responses to a questionnaire sent out for this project

System analysis and mass balance of a caloric power plant

International Nuclear Information System (INIS)

Schachermayer, E.; Rechberger, H.; Brunner, P.H.; Maderner, W.

1995-05-01

The present feasibility study analyses the possibilities of examining and presenting material and energy aspects of the interaction between the power plant and the region. A method has been worked out to establish an overall system called power plant and its periphery which combines the thermal power plant Voitsberg 3, the supplying coal mine Oberdorf, and the ash dumping site Karlschacht 2. One goal of this study was therefore, to find a way of combining the different steps 'mining' , 'power plant' , and 'ash dumping' to an overall system considering both qualitative and quantitative aspects. Another goal of the present feasibility study was to develop methods to consider the impacts of the power plant and its periphery to the environment and the use of resources. For this purpose the region is analysed from two different points of view: - the region defined by the 'product'; its size is defined by the number of inhabitants the power plant supplies with the product 'electricity'. The region defined by the area, in which the specific impacts of the power plant on air, water, soil etc. have to be examined; its size will therefore vary considerably according to the subject of investigation. Based on data provided by the power plant operators a rough estimate could be made to what extent the power plant is responsible for the air pollutant emissions in the region defined by the 'product'. According to these data the part of pollution caused by the power plant amounts to 10% for NOx and particulate matter, to 40% for SO 2 , and to 70% for CO 2 . Furthermore, it can be seen, that the material fluxes of selenium and mercury have to be surveyed in more detail. The ashes dumped in Karlschacht 2 contain a large amount of heavy metals, mainly arsenic, selenium, and mercury. These elements occur in a much higher concentration than in the earth crust. (author)

Introduction of Electrical System Simulation and Analysis Used in Korean Nuclear Power Plants

International Nuclear Information System (INIS)

Kim, Sang Hak; Jeong, Woo Sung

2015-01-01

The purpose of this paper is to introduce the simulation methods and tools to analyse and predict the performance of the electric power distribution system for nuclear power plants (NPPs) in Korea. Electrical System design engineers are to evaluate the load flow, bus voltage profiles, short circuit levels, motor starting, and fast bus transfer under various plant operating conditions and to verify the adequacy of power distribution System for a reliable power supply to plant loads under various disturbances which could jeopardize a safe and reliable operation of nuclear power plants. (authors)

Safety analysis code input automation using the Nuclear Plant Data Bank

International Nuclear Information System (INIS)

Kopp, H.; Leung, J.; Tajbakhsh, A.; Viles, F.

1985-01-01

The Nuclear Plant Data Bank (NPDB) is a computer-based system that organizes a nuclear power plant's technical data, providing mechanisms for data storage, retrieval, and computer-aided engineering analysis. It has the specific objective to describe thermohydraulic systems in order to support: rapid information retrieval and display, and thermohydraulic analysis modeling. The Nuclear Plant Data Bank (NPBD) system fully automates the storage and analysis based on this data. The system combines the benefits of a structured data base system and computer-aided modeling with links to large scale codes for engineering analysis. Emphasis on a friendly and very graphically oriented user interface facilitates both initial use and longer term efficiency. Specific features are: organization and storage of thermohydraulic data items, ease in locating specific data items, graphical and tabular display capabilities, interactive model construction, organization and display of model input parameters, input deck construction for TRAC and RELAP analysis programs, and traceability of plant data, user model assumptions, and codes used in the input deck construction process. The major accomplishments of this past year were the development of a RELAP model generation capability and the development of a CRAY version of the code

Plant dynamics studies towards design of plant protection system for PFBR

Energy Technology Data Exchange (ETDEWEB)

Natesan, K., E-mail: natesan@igcar.gov.in [Nuclear and Safety Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Kasinathan, N.; Velusamy, K.; Selvaraj, P.; Chellapandi, P. [Nuclear and Safety Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India)

2012-09-15

Highlights: Black-Right-Pointing-Pointer Analysis of various design basis events in a fast breeder reactor towards design of plant protection system. Black-Right-Pointing-Pointer Plant dynamic modeling of a sodium cooled fast breeder reactor. Black-Right-Pointing-Pointer Selection of optimum set of plant parameters for considering best plant availability. - Abstract: Prototype fast breeder reactor (PFBR) is a 500 MWe (1250 MWt) liquid sodium cooled pool type reactor currently under construction in India. For a safe and efficient operation of the plant, it is necessary that the reactor is protected from all the transients that may occur in the plant. In order to accomplish this, adequate number of SCRAM parameters is required in the plant protection system with reliable instrumentation. For identifying the SCRAM parameters, the neutronic and thermal hydraulic responses of the plant for various possible events need to be established. Towards this, a one dimensional plant dynamics code DYANA-P has been developed with thermal hydraulic models for reactor core, hot and cold pools, intermediate heat exchangers, pipelines, steam generator, primary sodium circuits and secondary sodium circuits. The code also incorporates neutron kinetics and reactivity feedback models. By a comprehensive plant dynamics study an optimum list of SCRAM parameters and the maximum permissible response time for various instruments used for deriving them have been arrived at.

A nuclear power plant system engineering workstation

International Nuclear Information System (INIS)

Mason, J.H.; Crosby, J.W.

1989-01-01

System engineers offer an approach for effective technical support for operation and maintenance of nuclear power plants. System engineer groups are being set up by most utilities in the United States. Institute of Nuclear Power operations (INPO) and U.S. Nuclear Regulatory Commission (NRC) have endorsed the concept. The INPO Good Practice and a survey of system engineer programs in the southeastern United States provide descriptions of system engineering programs. The purpose of this paper is to describe a process for developing a design for a department-level information network of workstations for system engineering groups. The process includes the following: (1) application of a formal information engineering methodology, (2) analysis of system engineer functions and activities; (3) use of Electric Power Research Institute (EPRI) Plant Information Network (PIN) data; (4) application of the Information Engineering Workbench. The resulting design for this system engineer workstation can provide a reference for design of plant-specific systems

Elevated service water temperature systems analysis for a nuclear power plant

International Nuclear Information System (INIS)

Lewis, T.; Hurt, W.

1992-01-01

This paper describes analyses performed to support the evaluation of the effects of elevated Service Water (SW) temperatures on the operation of a Pressurized Water Reactor. The purpose of the analyses is to provide justification of continued plant operation with SW temperatures up to 5 degrees F (3 degrees C) above the original temperature design limit. The study involved evaluation of the following major components or plant transients: Containment Design Basis Accident (DBA), Emergency Diesel Generator (EDG), Plant Cooldown, Engineered Safety Feature (ESF) Room Coolers, Engineered Safety Feature Pumps, and Assessment for Impact on Normal Operation. The principal objective was related to raising the design maximum temperature of the SW system from 95 degrees F (35 degrees C) to 100 degrees F (38 degrees C). since the Service Water system is safety related, an serves a plant during both normal and design basis conditions, a wide variety of components must be analyzed under various operating modes. The evaluation of systems and components affected by elevated SW temperature is presented, along with conclusions

Optimal control systems in hydro power plants

International Nuclear Information System (INIS)

Babunski, Darko L.

2012-01-01

The aim of the research done in this work is focused on obtaining the optimal models of hydro turbine including auxiliary equipment, analysis of governors for hydro power plants and analysis and design of optimal control laws that can be easily applicable in real hydro power plants. The methodology of the research and realization of the set goals consist of the following steps: scope of the models of hydro turbine, and their modification using experimental data; verification of analyzed models and comparison of advantages and disadvantages of analyzed models, with proposal of turbine model for design of control low; analysis of proportional-integral-derivative control with fixed parameters and gain scheduling and nonlinear control; analysis of dynamic characteristics of turbine model including control and comparison of parameters of simulated system with experimental data; design of optimal control of hydro power plant considering proposed cost function and verification of optimal control law with load rejection measured data. The hydro power plant models, including model of power grid are simulated in case of island ing and restoration after breakup and load rejection with consideration of real loading and unloading of hydro power plant. Finally, simulations provide optimal values of control parameters, stability boundaries and results easily applicable to real hydro power plants. (author)

Numerical analysis on the performance of solar chimney power plant system

International Nuclear Information System (INIS)

Xu Guoliang; Ming Tingzhen; Pan Yuan; Meng Fanlong; Zhou Cheng

2011-01-01

Power generating technology based on renewable energy resources will definitely become a new trend of future energy utilization. Numerical simulations on air flow, heat transfer and power output characteristics of a solar chimney power plant model with energy storage layer and turbine similar to the Spanish prototype were carried out in this paper, and mathematical model of flow and heat transfer for the solar chimney power plant system was established. The influences of solar radiation and pressure drop across the turbine on the flow and heat transfer, output power and energy loss of the solar chimney power plant system were analyzed. The numerical simulation results reveal that: when the solar radiation and the turbine efficiency are 600 W/m 2 and 80%, respectively, the output power of the system can reach 120 kW. In addition, large mass flow rate of air flowing through the chimney outlet become the main cause of energy loss in the system, and the collector canopy also results in large energy loss.

Operator psychological selection system for nuclear power plant

International Nuclear Information System (INIS)

He Xuhong; Huang Xiangrui

2004-01-01

Based on a detailed job analysis of nuclear power plant operator including operation procedures analysis, interview with personnel familiar with operator job, and 9 events happened in the past in the plant involved operator error analysis, several operator work characteristics and performance influence factors are obtained. According to these specific characteristics and factors, referring to the psychological selection research results in the other related critical occupational fields, a full psychological selection system of nuclear power plant operator is forwarded in this paper, including 21 dimensions in 3 facets as general psychological ability, personality and psychological healthy. Practical measurement methods for the proposed selection dimensions are discussed in the end

Expert systems for assisting the analysis of hazards

International Nuclear Information System (INIS)

Evrard, J.M.; Martinez, J.M.; Souchet, Y.

1990-01-01

The advantage of applying expert systems in the analysis of safety in the operation of nuclear power plants is discussed. Expert systems apply a method based on a common representation of nuclear power plants. The main steps of the method are summarized. The applications given concern in the following fields: the analysis of hazards in the electric power supplies of a gas-graphite power plant; the evaluation of the availability of safety procedures in a PWR power plant; the search for the sources of leakage in a PWR power plant. The analysis shows that expert systems are a powerful tool in the study of safety of nuclear power plants [fr

Plant specific PTS analysis of Kori Unit 1

Energy Technology Data Exchange (ETDEWEB)

Sung-Yull, Hong; Changheui, Jang; Ill-Seok, Jeong [Korea Eletric Power Research Inst., Daejon (Korea, Republic of); Tae-Eun, Jin [Korea Power Engineering Company, Yonging (Korea, Republic of)

1997-09-01

Currently, a nuclear PLIM (Plant Lifetime Management) program is underway in Korea to extend the operation life of Kori-1 which was originally licensed for 30 years. For the life extension of nuclear power plants, the residual lives of major components should be evaluated for the extended operation period. According to the residual life evaluation of reactor pressure vessel, which was classified as one of the major components crucial to life extension, it was found by screening analysis that reference PTS temperature would exceed screening criteria before the target extended operation years. In order to deal with this problem, a plant-specific PTS analysis for Kori-1 RPV has been initiated. In this paper, the relationship between PTS analysis and Kori-1 PLIM program is briefly described. The plant-specific PTS analysis covers system transient analysis, downcomer mixing analysis, and probabilistic fracture mechanics analysis to check the integrity or RPV during various PTS transients. The step-by-step procedure of the analysis will be described in detail. Finally, various issues regarding RPV materials and its integrity will be briefly mentioned, and their implications on Kori-1 PTS analysis will be discussed. Despite of the screening analysis result concern, it is now expected that Kori-1 PTS issues can be handled through the plant-specific PTS analysis. (author). 14 refs, 4 figs, 2 tabs.

KKS:The new power plant characterisation system in practice

International Nuclear Information System (INIS)

Becker, G.; Bohr, E.; Wildberg, D.

1982-01-01

These use of the new power plant characterisation system (KKS) in power plant operation is discussed. KKS intends to ensure reliable characterisation of plant components. It must be read, interpreted, communicated directly or by telephone and written down. An analysis under the aspects of information psychology and ergonomy has shown that the findings of ergonomy and information psychology have hardly been considerd in the design of KKS. The conclusions of this analysis are investigated with regard to their consequences for KKS application in power plant operation. Possibilities of improvement are discussed. (orig./HP) [de

Optimized production planning model for a multi-plant cultivation system under uncertainty

Science.gov (United States)

Ke, Shunkui; Guo, Doudou; Niu, Qingliang; Huang, Danfeng

2015-02-01

An inexact multi-constraint programming model under uncertainty was developed by incorporating a production plan algorithm into the crop production optimization framework under the multi-plant collaborative cultivation system. In the production plan, orders from the customers are assigned to a suitable plant under the constraints of plant capabilities and uncertainty parameters to maximize profit and achieve customer satisfaction. The developed model and solution method were applied to a case study of a multi-plant collaborative cultivation system to verify its applicability. As determined in the case analysis involving different orders from customers, the period of plant production planning and the interval between orders can significantly affect system benefits. Through the analysis of uncertain parameters, reliable and practical decisions can be generated using the suggested model of a multi-plant collaborative cultivation system.

A study on expert system applications for nuclear power plant

International Nuclear Information System (INIS)

Huh, Young Hwan; Kim, Yeong Jin; Park, Nam Seog; Dong, In Sook; Choi, In Seon

1987-12-01

The application of artificial intelligence techniques to nuclear power plants such as expert systems is rapidly emerging. expert systems can contribute significantly to the availability and the improved operation and safety of nuclear power plants. The objective of the project is to develop an expert system in a selected application area in the nuclear power plants. This project will last for 3 years. The first year's tasks are: - Information collection and literature survey on expert systems. - Analysis of several applicable areas for applying AI technologies to the nuclear power plants. - Conceptual design of a few selected domains. - Selection of hardware and software tools for the development of the expert system

Mathematical modeling and fuzzy availability analysis for serial processes in the crystallization system of a sugar plant

Science.gov (United States)

Aggarwal, Anil Kr.; Kumar, Sanjeev; Singh, Vikram

2017-03-01

The binary states, i.e., success or failed state assumptions used in conventional reliability are inappropriate for reliability analysis of complex industrial systems due to lack of sufficient probabilistic information. For large complex systems, the uncertainty of each individual parameter enhances the uncertainty of the system reliability. In this paper, the concept of fuzzy reliability has been used for reliability analysis of the system, and the effect of coverage factor, failure and repair rates of subsystems on fuzzy availability for fault-tolerant crystallization system of sugar plant is analyzed. Mathematical modeling of the system is carried out using the mnemonic rule to derive Chapman-Kolmogorov differential equations. These governing differential equations are solved with Runge-Kutta fourth-order method.

The System 80+ Standard Plant design control document. Volume 17

International Nuclear Information System (INIS)

1997-01-01

This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume contains parts 2-7 and appendix 15A for section 15 (Accident Analysis) of the ADM Design and Analysis. Topics covered in these parts are: decrease in heat removal; decrease in RCS flow rate; power distribution anomalies; increase in RCS inventory; decrease in RCS inventory; release of radioactive materials. The appendix covers radiological release models. Also contained here are five technical specifications for section 16 (Technical Specifications) of the ADM Design and Analysis. They are: TS 1.0 Use and Applications; TS 2.0 Safety Limits; TS 3.0 LCO Availability; TS 3.1 Reactivity Control; and TS 3.2 Power Distribution

The System 80+ Standard Plant design control document. Volume 10

International Nuclear Information System (INIS)

1997-01-01

This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume contains Appendices 6A, 6B, and 6C for section 6 (Engineered Safety Features) of the ADM Design and Analysis. Also, parts 1--5 of section 7 (Instrumentation and Control) of the ADM Design and Analysis are covered. The following information is covered in these parts: introduction; reactor protection system; ESF actuation system; system required for safe shutdown; and safety-related display instrumentation

Computerized ECT data analysis system

International Nuclear Information System (INIS)

Miyake, Y.; Fukui, S.; Iwahashi, Y.; Matsumoto, M.; Koyama, K.

1988-01-01

For the analytical method of the eddy current testing (ECT) of steam generator tubes in nuclear power plants, the authors have developed the computerized ECT data analysis system using a large-scale computer with a high-resolution color graphic display. This system can store acquired ECT data up to 15 steam generators, and ECT data can be analyzed immediately on the monitor in dialogue communication with a computer. Analyzed results of ECT data are stored and registered in the data base. This system enables an analyst to perform sorting and collecting of data under various conditions and obtain the results automatically, and also to make a plan of tube repair works. This system has completed the test run, and has been used for data analysis at the annual inspection of domestic plants. This paper describes an outline, features and examples of the computerized eddy current data analysis system for steam generator tubes in PWR nuclear power plants

Exergetic Analysis of a Novel Solar Cooling System for Combined Cycle Power Plants

Directory of Open Access Journals (Sweden)

Francesco Calise

2016-09-01

Full Text Available This paper presents a detailed exergetic analysis of a novel high-temperature Solar Assisted Combined Cycle (SACC power plant. The system includes a solar field consisting of innovative high-temperature flat plate evacuated solar thermal collectors, a double stage LiBr-H2O absorption chiller, pumps, heat exchangers, storage tanks, mixers, diverters, controllers and a simple single-pressure Combined Cycle (CC power plant. Here, a high temperature solar cooling system is coupled with a conventional combined cycle, in order to pre-cool gas turbine inlet air in order to enhance system efficiency and electrical capacity. In this paper, the system is analyzed from an exergetic point of view, on the basis of an energy-economic model presented in a recent work, where the obtained main results show that SACC exhibits a higher electrical production and efficiency with respect to the conventional CC. The system performance is evaluated by a dynamic simulation, where detailed simulation models are implemented for all the components included in the system. In addition, for all the components and for the system as whole, energy and exergy balances are implemented in order to calculate the magnitude of the irreversibilities within the system. In fact, exergy analysis is used in order to assess: exergy destructions and exergetic efficiencies. Such parameters are used in order to evaluate the magnitude of the irreversibilities in the system and to identify the sources of such irreversibilities. Exergetic efficiencies and exergy destructions are dynamically calculated for the 1-year operation of the system. Similarly, exergetic results are also integrated on weekly and yearly bases in order to evaluate the corresponding irreversibilities. The results showed that the components of the Joule cycle (combustor, turbine and compressor are the major sources of irreversibilities. System overall exergetic efficiency was around 48%. Average weekly solar collector

Probability analysis of nuclear power plant hazards

International Nuclear Information System (INIS)

Kovacs, Z.

1985-01-01

The probability analysis of risk is described used for quantifying the risk of complex technological systems, especially of nuclear power plants. Risk is defined as the product of the probability of the occurrence of a dangerous event and the significance of its consequences. The process of the analysis may be divided into the stage of power plant analysis to the point of release of harmful material into the environment (reliability analysis) and the stage of the analysis of the consequences of this release and the assessment of the risk. The sequence of operations is characterized in the individual stages. The tasks are listed which Czechoslovakia faces in the development of the probability analysis of risk, and the composition is recommended of the work team for coping with the task. (J.C.)

The System 80+ Standard Plant design control document. Volume 15

International Nuclear Information System (INIS)

1997-01-01

This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume contains all five parts of section 12 (Radiation Protection) of the ADM Design and Analysis. Topics covered are: ALARA exposures; radiation sources; radiation protection; dose assessment; and health physics program. All six parts and appendices A and B for section 13 (Conduct of Operations) of the ADM Design and Analysis are also contained in this volume. Topics covered are: organizational structure; training program; emergency planning; review and audit; plant procedures; industrial security; sabotage protection (App 13A); and vital equipment list (App 13B)

BWR plant dynamic analysis code BWRDYN user's manual

International Nuclear Information System (INIS)

Yokobayashi, Masao; Yoshida, Kazuo; Fujiki, Kazuo

1989-06-01

Computer code BWRDYN has been developed for thermal-hydraulic analysis of a BWR plant. It can analyze the various types of transient caused by not only small but also large disturbances such as operating mode changes and/or system malfunctions. The verification of main analytical models of the BWRDYN code has been performed with measured data of actual BWR plant. Furthermore, the installation of BOP (Balance of Plant) model has made it possible to analyze the effect of BOP on reactor system. This report describes on analytical models and instructions for user of the BWRDYN code. (author)

FPGA-Based Plant Protection System

Energy Technology Data Exchange (ETDEWEB)

Lee, Yoon Hee; Ha, Jae Hong; Kim, Hang Bae [KEPCO E and C, Daejeon (Korea, Republic of)

2011-08-15

This paper relates to a plant protection system which detects non-permissible conditions and determines initiation of protective actions for nuclear power plants (NPPs). Conventional plant protection systems were designed based on analog technologies. It is well known that existing protection systems for NPPs contain many components which are becoming obsolete at an increasing rate. Nowadays maintenance and repair for analog-based plant protection systems may be difficult as analog parts become obsolete or difficult to obtain. Accordingly, as an alternative to the analog technology, the digitalisation of the plant protection system was required. Recently digital plant protection systems which include programmable logic controllers (PLCs) and/or computers have been introduced. However PLC or computer-based plant protection systems use an operating system and application software, and so they may result in a common mode failure when a problem occurs in the operating system or application software. Field Programmable Gate Arrays (FPGAs) are highlighted as an alternative to conventional protection or control systems. The paper presents the design of a four-channel plant protection system whose protection functions are implemented in FPGAs without any central processing unit or operating system.

FPGA-Based Plant Protection System

International Nuclear Information System (INIS)

Lee, Yoon Hee; Ha, Jae Hong; Kim, Hang Bae

2011-01-01

This paper relates to a plant protection system which detects non-permissible conditions and determines initiation of protective actions for nuclear power plants (NPPs). Conventional plant protection systems were designed based on analog technologies. It is well known that existing protection systems for NPPs contain many components which are becoming obsolete at an increasing rate. Nowadays maintenance and repair for analog-based plant protection systems may be difficult as analog parts become obsolete or difficult to obtain. Accordingly, as an alternative to the analog technology, the digitalisation of the plant protection system was required. Recently digital plant protection systems which include programmable logic controllers (PLCs) and/or computers have been introduced. However PLC or computer-based plant protection systems use an operating system and application software, and so they may result in a common mode failure when a problem occurs in the operating system or application software. Field Programmable Gate Arrays (FPGAs) are highlighted as an alternative to conventional protection or control systems. The paper presents the design of a four-channel plant protection system whose protection functions are implemented in FPGAs without any central processing unit or operating system

Interaction of electromagnetic pulse with commercial nuclear-power-plant systems

Energy Technology Data Exchange (ETDEWEB)

Ericson, D.M. Jr.; Strawe, D.F.; Sandberg, S.J.; Jones, V.K.; Rensner, G.D.; Shoup, R.W.; Hanson, R.J.; Williams, C.B.

1983-02-01

This study examines the interaction of the electromagnetic pulse from a high altitude nuclear burst with commercial nuclear power plant systems. The potential vulnerability of systems required for safe shutdown of a specific nuclear power plant are explored. EMP signal coupling, induced plant response and component damage thresholds are established using techniques developed over several decades under Defense Nuclear Agency sponsorship. A limited test program was conducted to verify the coupling analysis technique as applied to a nuclear power plant. The results are extended, insofar as possible, to other nuclear plants.

Interaction of electromagnetic pulse with commercial nuclear-power-plant systems

International Nuclear Information System (INIS)

Ericson, D.M. Jr.; Strawe, D.F.; Sandberg, S.J.; Jones, V.K.; Rensner, G.D.; Shoup, R.W.; Hanson, R.J.; Williams, C.B.

1983-02-01

This study examines the interaction of the electromagnetic pulse from a high altitude nuclear burst with commercial nuclear power plant systems. The potential vulnerability of systems required for safe shutdown of a specific nuclear power plant are explored. EMP signal coupling, induced plant response and component damage thresholds are established using techniques developed over several decades under Defense Nuclear Agency sponsorship. A limited test program was conducted to verify the coupling analysis technique as applied to a nuclear power plant. The results are extended, insofar as possible, to other nuclear plants

The System 80+ Standard Plant design control document. Volume 19

International Nuclear Information System (INIS)

1997-01-01

This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume contains five technical specification bases that are part of Appendix 16 A of the ADM Design and Analysis. They are: TS B3.3 Instrumentation Bases; TS B3.4 RCS Bases; TS B3.5 ECCS Bases; TS B3.6 Containment Systems Bases; and TS B3.7 Plant Systems Bases

Control system upgrades support better plant economics

International Nuclear Information System (INIS)

De Grosbois, J.; Hepburn, A.; Storey, H.; Basso, R.; Kumar, V.

2002-01-01

This paper (second in the series, see [1]) provides insight on how nuclear plants can achieve better efficiencies and reduced operations and maintenance (O and M) costs through focused control system upgrades. An understanding of this relationship is necessary to properly assess the economics of plant refurbishment decisions. Traditional economic feasibility assessment methods such as benefit cost analysis (BCA), internal rate of return (IRR), benefit cost ratios (B/C), or payback analysis are often performed without full consideration of project alternatives, quantified benefits, and life cycle costing. Consideration must be given to not only capital cost and project risk, but also to the potential economic benefits of new technology and added functionality offered by plant upgrades. Recent experience shows that if upgrades are focused on priority objectives, and are effectively implemented, they can deliver significant payback over the life of the plant, sometimes orders of magnitude higher than their initial capital cost. The following discussion explores some of the key issues and rationale behind upgrade decisions and their impact on improved plant efficiency and reduced O and M costs. A subsequent paper will explain how the justification for these improvements can be captured in an economic analysis and feasibility study to support strategic decision-making in a plant refurbishment context. (author)

Assessing nuclear power plant safety and recovery from earthquakes using a system-of-systems approach

International Nuclear Information System (INIS)

Ferrario, E.; Zio, E.

2014-01-01

We adopt a ‘system-of-systems’ framework of analysis, previously presented by the authors, to include the interdependent infrastructures which support a critical plant in the study of its safety with respect to the occurrence of an earthquake. We extend the framework to consider the recovery of the system of systems in which the plant is embedded. As a test system, we consider the impacts produced on a nuclear power plant (the critical plant) embedded in the connected power and water distribution, and transportation networks which support its operation. The Seismic Probabilistic Risk Assessment of such system of systems is carried out by Hierarchical modeling and Monte Carlo simulation. First, we perform a top-down analysis through a hierarchical model to identify the elements that at each level have most influence in restoring safety, adopting the criticality importance measure as a quantitative indicator. Then, we evaluate by Monte Carlo simulation the probability that the nuclear power plant enters in an unsafe state and the time needed to recover its safety. The results obtained allow the identification of those elements most critical for the safety and recovery of the nuclear power plant; this is relevant for determining improvements of their structural/functional responses and supporting the decision-making process on safety critical-issues. On the test system considered, under the given assumptions, the components of the external and internal water systems (i.e., pumps and pool) turn out to be the most critical for the safety and recovery of the plant. - Highlights: • We adopt a system-of-system framework to analyze the safety of a critical plant exposed to risk from external events, considering also the interdependent infrastructures that support the plant. • We develop a hierarchical modeling framework to represent the system of systems, accounting also for its recovery. • Monte Carlo simulation is used for the quantitative evaluation of the

Cooperation of nuclear, thermal and hydroelectric power plants in the power system

International Nuclear Information System (INIS)

1984-01-01

The conference heard 36 papers of which 23 were incorporated in INIS. The subjects discussed were: the development of power industry in Czechoslovakia, methods of statistical analysis of data regarding nuclear power plant operation, the incorporation of WWER nuclear power plants in the power supply system, the standardization of nuclear power plants, the service life of components, use of nuclear energy sources, performance of the reactor accident protection system, the use of nuclear power and heating plants in Hungary, risk analysis, optimization of nuclear power plants, accidents caused by leakage of the primary and secondary circuit. (J.P.)

Hybrid intelligent monironing systems for thermal power plant trips

Science.gov (United States)

Barsoum, Nader; Ismail, Firas Basim

2012-11-01

Steam boiler is one of the main equipment in thermal power plants. If the steam boiler trips it may lead to entire shutdown of the plant, which is economically burdensome. Early boiler trips monitoring is crucial to maintain normal and safe operational conditions. In the present work two artificial intelligent monitoring systems specialized in boiler trips have been proposed and coded within the MATLAB environment. The training and validation of the two systems has been performed using real operational data captured from the plant control system of selected power plant. An integrated plant data preparation framework for seven boiler trips with related operational variables has been proposed for IMSs data analysis. The first IMS represents the use of pure Artificial Neural Network system for boiler trip detection. All seven boiler trips under consideration have been detected by IMSs before or at the same time of the plant control system. The second IMS represents the use of Genetic Algorithms and Artificial Neural Networks as a hybrid intelligent system. A slightly lower root mean square error was observed in the second system which reveals that the hybrid intelligent system performed better than the pure neural network system. Also, the optimal selection of the most influencing variables performed successfully by the hybrid intelligent system.

Life-cycle analysis of renewable energy systems

DEFF Research Database (Denmark)

Sørensen, Bent

1994-01-01

An imlementation of life-cycle analysis (LCA) for energy systems is presented and applied to two renewable energy systems (wind turbines and building-integrated photovoltaic modules) and compared with coal plants......An imlementation of life-cycle analysis (LCA) for energy systems is presented and applied to two renewable energy systems (wind turbines and building-integrated photovoltaic modules) and compared with coal plants...

The System 80+ Standard Plant design control document. Volume 23

International Nuclear Information System (INIS)

1997-01-01

This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume contains part 16 References and Appendix 19 A Design Alternatives for section 19 (Probabilistic Risk Assessment) of the ADM Design and Analysis. Also covered is section 20 Unresolved Safety Issues of the ADM Design and Analysis. Finally sections 1--6 of the ADM Emergency Operations Guidelines are contained in this volume. Information covered in these sections include: standard post-trip actions; diagnostic actions; reactor trip recovery guideline; LOCA recovery; SG tube rupture recovery

An artificial intelligence approach towards disturbance analysis in nuclear power plants

International Nuclear Information System (INIS)

Lindner, A.; Klebau, J.; Fielder, U.; Baldeweg, F.

1987-01-01

The scale and degree of sophistication of technological plants, e.g. nuclear power plants, have been essentially increased during the last decades. Conventional disturbance analysis systems have proved to work successfully in wellknown situations. But in cases of emergencies, the operator staff needs a more advanced assistance in realizing diagnosis and therapy control. The significance of introducing artificial intelligence methods in nuclear power technology is emphasized. Main features of the on-line disturbance analysis system SAAP-2 are reported about. It is being developed for application in nuclear power plants. 9 refs. (author)

Estimation of reliability on digital plant protection system in nuclear power plants using fault simulation with self-checking

International Nuclear Information System (INIS)

Lee, Jun Seok; Kim, Suk Joon; Seong, Poong Hyun

2004-01-01

Safety-critical digital systems in nuclear power plants require high design reliability. Reliable software design and accurate prediction methods for the system reliability are important problems. In the reliability analysis, the error detection coverage of the system is one of the crucial factors, however, it is difficult to evaluate the error detection coverage of digital instrumentation and control system in nuclear power plants due to complexity of the system. To evaluate the error detection coverage for high efficiency and low cost, the simulation based fault injections with self checking are needed for digital instrumentation and control system in nuclear power plants. The target system is local coincidence logic in digital plant protection system and a simplified software modeling for this target system is used in this work. C++ based hardware description of micro computer simulator system is used to evaluate the error detection coverage of the system. From the simulation result, it is possible to estimate the error detection coverage of digital plant protection system in nuclear power plants using simulation based fault injection method with self checking. (author)

Plutonium finishing plant safety systems and equipment list

International Nuclear Information System (INIS)

Bergquist, G.G.

1995-01-01

The Safety Equipment List (SEL) supports Analysis Report (FSAR), WHC-SD-CP-SAR-021 and the Plutonium Finishing Plant Operational Safety Requirements (OSRs), WHC-SD-CP-OSR-010. The SEL is a breakdown and classification of all Safety Class 1, 2, and 3 equipment, components, or system at the Plutonium Finishing Plant complex

Computer aided plant engineering: An analysis and suggestions for computer use

International Nuclear Information System (INIS)

Leinemann, K.

1979-09-01

To get indications to and boundary conditions for computer use in plant engineering, an analysis of the engineering process was done. The structure of plant engineering is represented by a network of substaks and subsets of data which are to be manipulated. Main tool for integration of CAD-subsystems in plant engineering should be a central database which is described by characteristical requirements and a possible simple conceptual schema. The main features of an interactive system for computer aided plant engineering are shortly illustrated by two examples. The analysis leads to the conclusion, that an interactive graphic system for manipulation of net-like structured data, usable for various subtasks, should be the base for computer aided plant engineering. (orig.) [de

Safety and security analysis for distributed control system in nuclear power plants

International Nuclear Information System (INIS)

Lu Zhigang; Liu Baoxu

2011-01-01

The Digital Distributed Control System (DCS) is the core that manages all monitoring and operation tasks in a Nuclear Power Plant (NPP). So, Digital Distributed Control System in Nuclear Power Plant has strict requirements for control and automation device safety and security due to many factors. In this article, factors of safety are analyzed firstly, while placing top priority on reliability, quality of supply and stability have also been carefully considered. In particular, advanced digital and electronic technologies are adopted to maintain sufficient reliability and supervisory capabilities in nuclear power plants. Then, security of networking and information technology have been remarked, several design methodologies considering the security characteristics are suggested. Methods and technologies of this article are being used in testing and evaluation for a real implement of a nuclear power plant in China. (author)

Nuclear power plant personnel qualifications and training: TAPS: the task analysis profiling system. Volume 2

International Nuclear Information System (INIS)

Jorgensen, C.C.

1985-06-01

This report discusses an automated task analysis profiling system (TAPS) designed to provide a linking tool between the behaviors of nuclear power plant operators in performing their tasks and the measurement tools necessary to evaluate their in-plant performance. TAPS assists in the identification of the entry-level skill, knowledge, ability and attitude (SKAA) requirements for the various tasks and rapidly associates them with measurement tests and human factors principles. This report describes the development of TAPS and presents its first demonstration. It begins with characteristics of skilled human performance and proceeds to postulate a cognitive model to formally describe these characteristics. A method is derived for linking SKAA characteristics to measurement tests. The entire process is then automated in the form of a task analysis computer program. The development of the program is detailed and a user guide with annotated code listings and supporting test information is provided

PlantPAN: Plant promoter analysis navigator, for identifying combinatorial cis-regulatory elements with distance constraint in plant gene groups

Directory of Open Access Journals (Sweden)

Huang Hsien-Da

2008-11-01

Full Text Available Abstract Background The elucidation of transcriptional regulation in plant genes is important area of research for plant scientists, following the mapping of various plant genomes, such as A. thaliana, O. sativa and Z. mays. A variety of bioinformatic servers or databases of plant promoters have been established, although most have been focused only on annotating transcription factor binding sites in a single gene and have neglected some important regulatory elements (tandem repeats and CpG/CpNpG islands in promoter regions. Additionally, the combinatorial interaction of transcription factors (TFs is important in regulating the gene group that is associated with the same expression pattern. Therefore, a tool for detecting the co-regulation of transcription factors in a group of gene promoters is required. Results This study develops a database-assisted system, PlantPAN (Plant Promoter Analysis Navigator, for recognizing combinatorial cis-regulatory elements with a distance constraint in sets of plant genes. The system collects the plant transcription factor binding profiles from PLACE, TRANSFAC (public release 7.0, AGRIS, and JASPER databases and allows users to input a group of gene IDs or promoter sequences, enabling the co-occurrence of combinatorial transcription factor binding sites (TFBSs within a defined distance (20 bp to 200 bp to be identified. Furthermore, the new resource enables other regulatory features in a plant promoter, such as CpG/CpNpG islands and tandem repeats, to be displayed. The regulatory elements in the conserved regions of the promoters across homologous genes are detected and presented. Conclusion In addition to providing a user-friendly input/output interface, PlantPAN has numerous advantages in the analysis of a plant promoter. Several case studies have established the effectiveness of PlantPAN. This novel analytical resource is now freely available at http://PlantPAN.mbc.nctu.edu.tw.

A new tool for risk analysis and assessment in petrochemical plants

Directory of Open Access Journals (Sweden)

El-Arkam Mechhoud

2016-09-01

Full Text Available The aim of our work was the implementation of a new automated tool dedicated to risk analysis and assessment in petrochemical plants, based on a combination of two analysis methods: HAZOP (HAZard and OPerability and FMEA (Failure Mode and Effect Analysis. Assessment of accident scenarios is also considered. The principal advantage of the two analysis methods is to speed-up hazard identification and risk assessment and forecast the nature and impact of such accidents. Plant parameters are analyzed under a graphical interface to facilitate the exploitation of our developed approach. This automated analysis brings out the different deviations of the operating parameters of any system in the plant. Possible causes of these deviations, their consequences and preventive actions are identified. The result is risk minimization and dependability enhancement of the considered system.

Cause analysis and suggestion of urea consumption in denitrification system of coal-fired power plant

Science.gov (United States)

Zhang, Xueying; Dong, Ruifeng; Guo, Yang; Wang, Fangfang; Yang, Shuo

2018-02-01

In the daily operation of many power plants, the urea consumption of denitration system is much more than normal. Therefore, the process of site testing and laboratory analysis are carried out. Several suggestions are given out. (1) The position of sampling hole on the exit flue of denitrification system should be redesigned. (2) The denitrification optimization and adjustment should be carried out based on the technical specifications for the operation system. (3) The flue gas CEMS system for single point sampling should be transformed into two or three point sampling mode. (4) When the coal - fired unit is shutting down, examine the ammonia injection and nozzle branch, in order to improve the operation reliability of denitration system.

The System 80+ Standard Plant design control document. Volume 2

International Nuclear Information System (INIS)

1997-01-01

This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume covers the following information of the CDM: (2.8) Steam and power conversion; (2.9) Radioactive waste management; (2.10) Tech Support Center; (2.11) Initial test program; (2.12) Human factors; and sections 3, 4, and 5. Also covered in this volume are parts 1--6 of section 1 (General Plant Description) of the ADM Design and Analysis

Integrated piping structural analysis system

International Nuclear Information System (INIS)

Motoi, Toshio; Yamadera, Masao; Horino, Satoshi; Idehata, Takamasa

1979-01-01

Structural analysis of the piping system for nuclear power plants has become larger in scale and in quantity. In addition, higher quality analysis is regarded as of major importance nowadays from the point of view of nuclear plant safety. In order to fulfill to the above requirements, an integrated piping structural analysis system (ISAP-II) has been developed. Basic philosophy of this system is as follows: 1. To apply the date base system. All information is concentrated. 2. To minimize the manual process in analysis, evaluation and documentation. Especially to apply the graphic system as much as possible. On the basis of the above philosophy four subsystems were made. 1. Data control subsystem. 2. Analysis subsystem. 3. Plotting subsystem. 4. Report subsystem. Function of the data control subsystem is to control all information of the data base. Piping structural analysis can be performed by using the analysis subsystem. Isometric piping drawing and mode shape, etc. can be plotted by using the plotting subsystem. Total analysis report can be made without the manual process through the reporting subsystem. (author)

Development of a dynamical systems model of plant programmatic performance on nuclear power plant safety risk

International Nuclear Information System (INIS)

Hess, Stephen M.; Albano, Alfonso M.; Gaertner, John P.

2005-01-01

Application of probabilistic risk assessment (PRA) techniques to model nuclear power plant accident sequences has provided a significant contribution to understanding the potential initiating events, equipment failures and operator errors that can lead to core damage accidents. Application of the lessons learned from these analyses has resulted in significant improvements in plant operation and safety. However, this approach has not been nearly as successful in addressing the impact of plant processes and management effectiveness on the risks of plant operation. The research described in this paper presents an alternative approach to addressing this issue. In this paper we propose a dynamical systems model that describes the interaction of important plant processes on nuclear safety risk. We discuss development of the mathematical model including the identification and interpretation of significant inter-process interactions. Next, we review the techniques applicable to analysis of nonlinear dynamical systems that are utilized in the characterization of the model. This is followed by a preliminary analysis of the model that demonstrates that its dynamical evolution displays features that have been observed at commercially operating plants. From this analysis, several significant insights are presented with respect to the effective control of nuclear safety risk. As an important example, analysis of the model dynamics indicates that significant benefits in effectively managing risk are obtained by integrating the plant operation and work management processes such that decisions are made utilizing a multidisciplinary and collaborative approach. We note that although the model was developed specifically to be applicable to nuclear power plants, many of the insights and conclusions obtained are likely applicable to other process industries

Availability Performance Analysis of Thermal Power Plants

Science.gov (United States)

Bhangu, Navneet Singh; Singh, Rupinder; Pahuja, G. L.

2018-03-01

This case study presents the availability evaluation method of thermal power plants for conducting performance analysis in Indian environment. A generic availability model has been proposed for a maintained system (thermal plants) using reliability block diagrams and fault tree analysis. The availability indices have been evaluated under realistic working environment using inclusion exclusion principle. Four year failure database has been used to compute availability for different combinatory of plant capacity, that is, full working state, reduced capacity or failure state. Availability is found to be very less even at full rated capacity (440 MW) which is not acceptable especially in prevailing energy scenario. One of the probable reason for this may be the difference in the age/health of existing thermal power plants which requires special attention of each unit from case to case basis. The maintenance techniques being used are conventional (50 years old) and improper in context of the modern equipment, which further aggravate the problem of low availability. This study highlights procedure for finding critical plants/units/subsystems and helps in deciding preventive maintenance program.

The System 80+ Standard Plant design control document. Volume 20

International Nuclear Information System (INIS)

1997-01-01

This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume contains 2 technical specifications bases as part of Appendix 16 A Tech Spec Bases. They are TS B3.8 Electrical Power Technical Systems Bases and TS B3.9 Refueling Operations Bases. All 3 parts of section 17 (QA) and all 10 parts of section 18 (Human Factors) of the ADM Design and Analysis are contained in this volume. Topics covered in section 17 are: design phase QA; operations phase QA; and design phase reliability assurance. Topics covered by section 18 are: design team organization; design goals; design process; functional task analysis; control room configuration; information presentation; control and monitoring; verification and validation; and review documents

Plant experience with an expert system for alarm diagnosis

International Nuclear Information System (INIS)

Gimmy, K.L.

1986-01-01

An expert system called Diagnosis of Multiple Alarms (DMA) is in routine use at four nuclear reactors operated by the DuPont Company. The system is wired to plant alarm annunciators and does event-tree analysis to see if a pattern exists. Any diagnosis is displayed to the plant operator and the corrective procedure to be followed is also identified. The display is automatically superseded if a higher priority diagnosis is made. The system is integrated with operator training and procedures. Operating results have been positive. DMA has diagnosed several hard-to-locate small leaks. There have been some false diagnosis, and realistic plant environments must be considered in such expert systems. 2 refs., 5 figs

Development of user interface to support automatic program generation of nuclear power plant analysis by module-based simulation system

International Nuclear Information System (INIS)

Yoshikawa, Hidekazu; Mizutani, Naoki; Nakaya, Ken-ichiro; Wakabayashi, Jiro

1988-01-01

Module-based Simulation System (MSS) has been developed to realize a new software work environment enabling versatile dynamic simulation of a complex nuclear power system flexibly. The MSS makes full use of modern software technology to replace a large fraction of human software works in complex, large-scale program development by computer automation. Fundamental methods utilized in MSS and developmental study on human interface system SESS-1 to help users in generating integrated simulation programs automatically are summarized as follows: (1) To enhance usability and 'communality' of program resources, the basic mathematical models of common usage in nuclear power plant analysis are programed as 'modules' and stored in a module library. The information on usage of individual modules are stored in module database with easy registration, update and retrieval by the interactive management system. (2) Target simulation programs and the input/output files are automatically generated with simple block-wise languages by a precompiler system for module integration purpose. (3) Working time for program development and analysis in an example study of an LMFBR plant thermal-hydraulic transient analysis was demonstrated to be remarkably shortened, with the introduction of an interface system SESS-1 developed as an automatic program generation environment. (author)

Analysis of a natural draught tower in the circulation seawater system of nuclear power plant of Laguna Verde

International Nuclear Information System (INIS)

Tijerina S, F.; Vargas A, A.

2009-10-01

The analysis of a natural draught tower in open circuit for the cooling system of seawater circulation on the nuclear power plant of Laguna Verde, it is based on conditions of 2027 MWt and 2317 MWt, where the flows of circulation water system hardly vary and whose purpose will be, to cool the seawater circulation. The circulation water system is used as heat drain in main condenser of turbo generator to condense the nuclear vapor. The annual average temperature in the seawater at present is of 26 C to the entrance to circulation water system and it is vary in accordance with the time of year. The mean temperature of leaving of circulation water system to the sea is of 41 C. Having a cooling tower to reduce the entrance temperature to the circulation water system, it improves the efficiency of thermal transfer in condenser, it improves the vacuum in condenser giving more operative margin to avoid condenser losses by air entrances and nuclear power plant shutdowns, as well as for to improve the efficiency of operative balance of nuclear power plant, also it prevents the impact in thermal transfer efficiency in condenser by the climatic change. (Author)

The System 80+ Standard Plant design control document. Volume 18

International Nuclear Information System (INIS)

1997-01-01

This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume contains the following technical specifications of section 16 (Technical Specifications) of the ADM Design and Analysis: TS 3.3 Instrumentation; TS 3.4 Reactor Coolant System; TS 3.5 Emergency Core Cooling System; TS 3.6 Containment Systems; TS 3.7 Plant Systems; TS 3.8 Electrical Power Systems; TS 3.9 Refueling Operations; TS 4.0 Design Features; TS 5.0 Administrative Controls. Appendix 16 A Tech Spec Bases is also included. It contains the following: TS B2.0 Safety Limits Bases; TS B3.0 LCO Applicability Bases; TS B3.1 Reactivity Control Bases; TS B3.2 Power Distribution Bases

Risk-oriented management of system configurations during power plant operation

International Nuclear Information System (INIS)

Kafka, P.

1991-01-01

The paper explains the purpose, design, and functions of an on-line database called risk monitor, which is developed on the basis of a risk model mapping an entire plant and the functions of the system configurations by a probabilistic and risk-oriented approach in a computerized simulation model. The frequency of occurrence of the various failure modes is computed and specified by fault tree analysis and event tree analysis. The monitoring system is a diagnostic tool for power plant operators. (DG) [de

Adapting a reactor safety assessment system for specific plants

International Nuclear Information System (INIS)

Ballard, T.L.; Cordes, G.A.

1991-01-01

The Reactor Safety Assessment System (RSAS) is an expert system being developed by the Idaho National Engineering Laboratory, the University of Maryland (UofM) and US Nuclear Regulatory Commission (NRC) for use in the NRC Operations center. RSAS is designed to help the Reactor Safety Team monitor and project core status during an emergency at a licensed nuclear power plant. Analysis uses a hierarchical plant model based on equipment availability and automatically input parametric plant information. There are 3 families of designs of pressurized water reactors and 75 plants using modified versions of the basic design. In order to make an RSAS model for each power plant, a generic model for a given plant type is used with differences being specified by plant specific files. Graphical displays of this knowledge are flexible enough to handle any plant configuration. A variety of tools have been implemented to make it easy to modify a design to fit a given plant while minimizing chance for error. 3 refs., 4 figs

A meta-analysis of plant facilitation in coastal dune systems: responses, regions, and research gaps.

Science.gov (United States)

Castanho, Camila de Toledo; Lortie, Christopher J; Zaitchik, Benjamin; Prado, Paulo Inácio

2015-01-01

Empirical studies in salt marshes, arid, and alpine systems support the hypothesis that facilitation between plants is an important ecological process in severe or 'stressful' environments. Coastal dunes are both abiotically stressful and frequently disturbed systems. Facilitation has been documented, but the evidence to date has not been synthesized. We did a systematic review with meta-analysis to highlight general research gaps in the study of plant interactions in coastal dunes and examine if regional and local factors influence the magnitude of facilitation in these systems. The 32 studies included in the systematic review were done in coastal dunes located in 13 countries around the world but the majority was in the temperate zone (63%). Most of the studies adopt only an observational approach to make inferences about facilitative interactions, whereas only 28% of the studies used both observational and experimental approaches. Among the factors we tested, only geographic region mediates the occurrence of facilitation more broadly in coastal dune systems. The presence of a neighbor positively influenced growth and survival in the tropics, whereas in temperate and subartic regions the effect was neutral for both response variables. We found no evidence that climatic and local factors, such as life-form and life stage of interacting plants, affect the magnitude of facilitation in coastal dunes. Overall, conclusions about plant facilitation in coastal dunes depend on the response variable measured and, more broadly, on the geographic region examined. However, the high variability and the limited number of studies, especially in tropical region, indicate we need to be cautious in the generalization of the conclusions. Anyway, coastal dunes provide an important means to explore topical issues in facilitation research including context dependency, local versus regional drivers of community structure, and the importance of gradients in shaping the outcome of net

A meta-analysis of plant facilitation in coastal dune systems: responses, regions, and research gaps

Directory of Open Access Journals (Sweden)

Camila de Toledo Castanho

2015-02-01

Full Text Available Empirical studies in salt marshes, arid, and alpine systems support the hypothesis that facilitation between plants is an important ecological process in severe or ‘stressful’ environments. Coastal dunes are both abiotically stressful and frequently disturbed systems. Facilitation has been documented, but the evidence to date has not been synthesized. We did a systematic review with meta-analysis to highlight general research gaps in the study of plant interactions in coastal dunes and examine if regional and local factors influence the magnitude of facilitation in these systems. The 32 studies included in the systematic review were done in coastal dunes located in 13 countries around the world but the majority was in the temperate zone (63%. Most of the studies adopt only an observational approach to make inferences about facilitative interactions, whereas only 28% of the studies used both observational and experimental approaches. Among the factors we tested, only geographic region mediates the occurrence of facilitation more broadly in coastal dune systems. The presence of a neighbor positively influenced growth and survival in the tropics, whereas in temperate and subartic regions the effect was neutral for both response variables. We found no evidence that climatic and local factors, such as life-form and life stage of interacting plants, affect the magnitude of facilitation in coastal dunes. Overall, conclusions about plant facilitation in coastal dunes depend on the response variable measured and, more broadly, on the geographic region examined. However, the high variability and the limited number of studies, especially in tropical region, indicate we need to be cautious in the generalization of the conclusions. Anyway, coastal dunes provide an important means to explore topical issues in facilitation research including context dependency, local versus regional drivers of community structure, and the importance of gradients in shaping

A Design of Ginseng Planting Environment Monitoring System Based on WSN

Directory of Open Access Journals (Sweden)

Xin Ding

2014-03-01

Full Text Available Through the analysis of ginseng products industry chain, this paper designs and implements ginseng planting environment monitoring system. The system realized data collection and detection of ginseng planting environment in real time by using wireless sensor, transmission of environmental parameters in real time by using GPRS wireless transmission module, and video monitor and alarm of ginseng land by using unattended machine. It is the foundation of information transformation of ginseng products industry chain based on the Internet of Things. The experiment of ginseng planting base in Fusong indicates the system can offer support of original data for scientific cultivation of ginseng, comprehensive analysis of ginseng products and propaganda of ginseng brand.

An Automated and Continuous Plant Weight Measurement System for Plant Factory.

Science.gov (United States)

Chen, Wei-Tai; Yeh, Yu-Hui F; Liu, Ting-Yu; Lin, Ta-Te

2016-01-01

In plant factories, plants are usually cultivated in nutrient solution under a controllable environment. Plant quality and growth are closely monitored and precisely controlled. For plant growth evaluation, plant weight is an important and commonly used indicator. Traditional plant weight measurements are destructive and laborious. In order to measure and record the plant weight during plant growth, an automated measurement system was designed and developed herein. The weight measurement system comprises a weight measurement device and an imaging system. The weight measurement device consists of a top disk, a bottom disk, a plant holder and a load cell. The load cell with a resolution of 0.1 g converts the plant weight on the plant holder disk to an analog electrical signal for a precise measurement. The top disk and bottom disk are designed to be durable for different plant sizes, so plant weight can be measured continuously throughout the whole growth period, without hindering plant growth. The results show that plant weights measured by the weight measurement device are highly correlated with the weights estimated by the stereo-vision imaging system; hence, plant weight can be measured by either method. The weight growth of selected vegetables growing in the National Taiwan University plant factory were monitored and measured using our automated plant growth weight measurement system. The experimental results demonstrate the functionality, stability and durability of this system. The information gathered by this weight system can be valuable and beneficial for hydroponic plants monitoring research and agricultural research applications.

An Automated and Continuous Plant Weight Measurement System for Plant Factory

Directory of Open Access Journals (Sweden)

Wei-Tai eChen

2016-03-01

Full Text Available In plant factories, plants are usually cultivated in nutrient solution under a controllable environment. Plant quality and growth are closely monitored and precisely controlled. For plant growth evaluation, plant weight is an important and commonly used indicator. Traditional plant weight measurements are destructive and laborious. In order to measure and record the plant weight during plant growth, an automated measurement system was designed and developed herein. The weight measurement system comprises a weight measurement device and an imaging system. The weight measurement device consists of a top disk, a bottom disk, a plant holder and a load cell. The load cell with a resolution of 0.1 g converts the plant weight on the plant holder disk to an analogue electrical signal for a precise measurement. The top disk and bottom disk are designed to be durable for different plant sizes, so plant weight can be measured continuously throughout the whole growth period, without hindering plant growth. The results show that plant weights measured by the weight measurement device are highly correlated with the weights estimated by the stereo-vision imaging system; hence, plant weight can be measured by either method. The weight growth of selected vegetables growing in the National Taiwan University plant factory were monitored and measured using our automated plant growth weight measurement system. The experimental results demonstrate the functionality, stability and durability of this system. The information gathered by this weight system can be valuable and beneficial for hydroponic plants monitoring research and agricultural research applications.

Noise and vibration analysis system

International Nuclear Information System (INIS)

Johnsen, J.R.; Williams, R.L.

1985-01-01

The analysis of noise and vibration data from an operating nuclear plant can provide valuable information that can identify and characterize abnormal conditions. Existing plant monitoring equipment, such as loose parts monitoring systems (LPMS) and neutron flux detectors, may be capable of gathering noise data, but may lack the analytical capability to extract useful meanings hidden in the noise. By analyzing neutron noise signals, the structural motion and integrity of core components can be assessed. Computer analysis makes trending of frequency spectra within a fuel cycle and from one cycle to another a practical means of core internals monitoring. The Babcock and Wilcox Noise and Vibration Analysis System (NVAS) is a powerful, compact system that can automatically perform complex data analysis. The system can acquire, process, and store data, then produce report-quality plots of the important parameter. Software to perform neutron noise analysis and loose parts analysis operates on the same hardware package. Since the system is compact, inexpensive, and easy to operate, it allows utilities to perform more frequency analyses without incurring high costs and provides immediate results

Development of web based performance analysis program for nuclear power plant turbine cycle

International Nuclear Information System (INIS)

Park, Hoon; Yu, Seung Kyu; Kim, Seong Kun; Ji, Moon Hak; Choi, Kwang Hee; Hong, Seong Ryeol

2002-01-01

Performance improvement of turbine cycle affects economic operation of nuclear power plant. We developed performance analysis system for nuclear power plant turbine cycle. The system is based on PTC (Performance Test Code), that is estimation standard of nuclear power plant performance. The system is developed using Java Web-Start and JSP(Java Server Page)

Nuclear power plant diagnostic system

International Nuclear Information System (INIS)

Prokop, K.; Volavy, J.

1982-01-01

Basic information is presented on diagnostic systems used at nuclear power plants with PWR reactors. They include systems used at the Novovoronezh nuclear power plant in the USSR, at the Nord power plant in the GDR, the system developed at the Hungarian VEIKI institute, the system used at the V-1 nuclear power plant at Jaslovske Bohunice in Czechoslovakia and systems of the Rockwell International company used in US nuclear power plants. These diagnostic systems are basically founded on monitoring vibrations and noise, loose parts, pressure pulsations, neutron noise, coolant leaks and acoustic emissions. The Rockwell International system represents a complex unit whose advantage is the on-line evaluation of signals which gives certain instructions for the given situation directly to the operator. The other described systems process signals using similar methods. Digitized signals only serve off-line computer analyses. (Z.M.)

A modular approach to modeling power plant systems

International Nuclear Information System (INIS)

Yee, N.S.

1990-01-01

This paper reports on power plants which are large, non-linear systems with numerous interactions between its component parts. In the analysis of such complex systems, dynamic simulation is recognized as a powerful method of keeping track of the myriad of interactions. A simulation can be used to answer the what if questions that are asked when replacing components, changing operational procedures, or training operators. While there are many applications for the simulation of power plant components and systems, its use is often discouraged because it can be difficult and expensive. Power plant engineering is itself a multi-disciplinary field involving fluid mechanics, heat transfer, thermodynamics, chemical engineering, nuclear engineering, and electrical engineering. Simulation requires, in addition, knowledge in model formulation, computer programming and numerical solution of differential equations

Nuclear power plant fire protection: philosophy and analysis

International Nuclear Information System (INIS)

Berry, D.L.

1980-05-01

This report combines a fire severity analysis technique with a fault tree methodology for assessing the importance to nuclear power plant safety of certain combinations of components and systems. Characteristics unique to fire, such as propagation induced by the failure of barriers, have been incorporated into the methodology. By applying the resulting fire analysis technique to actual conditions found in a representative nuclear power plant, it is found that some safety and nonsafety areas are both highly vulnerable to fire spread and impotant to overall safety, while other areas prove to be of marginal importance. Suggestions are made for further experimental and analytical work to supplement the fire analysis method

The System 80+ Standard Plant design control document. Volume 11

International Nuclear Information System (INIS)

1997-01-01

This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume covers parts 6 and 7 and appendix 7A for section 7 (Instrumentation and Control) of the ADM Design and Analysis. The topics covered by these are: other systems required for safety; control systems not required by safety; and CMF evaluation of limiting faults. Parts 1--3 of section 8 (Electric Power) of the ADM are also included in this volume. Topics covered by these parts are: introduction; offsite power system; and onsite power system

Internal event analysis of Laguna Verde Unit 1 Nuclear Power Plant. System Analysis; Analisis de Eventos Internos para la Unidad 1 de la Central Nucleoelectrica de Laguna Verde. Analisis de sistemas

Energy Technology Data Exchange (ETDEWEB)

Huerta B, A; Aguilar T, O; Nunez C, A; Lopez M, R [Comision Nacional de Seguridad Nuclear y Salvaguardias, 03000 Mexico D.F. (Mexico)

1993-07-01

The Level 1 results of Laguna Verde Nuclear Power Plant PRA are presented in the {sup I}nternal Event Analysis of Laguna Verde Unit 1 Nuclear Power Plant{sup ,} CNSNS-TR-004, in five volumes. The reports are organized as follows: CNSNS-TR-004 Volume 1: Introduction and Methodology. CNSNS-TR-004 Volume 2: Initiating Event and Accident Sequences. CNSNS-TR-004 Volume 3: System Analysis. CNSNS-TR-004 Volume 4: Accident Sequence Quantification and Results. CNSNS-TR-004 Volume 5: Appendices A, B and C. This volume presents the results of the system analysis for the Laguna Verde Unit 1 Nuclear Power Plant. The system analysis involved the development of logical models for all the systems included in the accident sequence event tree headings, and for all the support systems required to operate the front line systems. For the Internal Event analysis for Laguna Verde, 16 front line systems and 5 support systems were included. Detailed fault trees were developed for most of the important systems. Simplified fault trees focusing on major faults were constructed for those systems that can be adequately represent,ed using this kind of modeling. For those systems where fault tree models were not constructed, actual data were used to represent the dominant failures of the systems. The main failures included in the fault trees are hardware failures, test and maintenance unavailabilities, common cause failures, and human errors. The SETS and TEMAC codes were used to perform the qualitative and quantitative fault tree analyses. (Author)

A Hydroponic Co-cultivation System for Simultaneous and Systematic Analysis of Plant/Microbe Molecular Interactions and Signaling.

Science.gov (United States)

Nathoo, Naeem; Bernards, Mark A; MacDonald, Jacqueline; Yuan, Ze-Chun

2017-07-22

An experimental design mimicking natural plant-microbe interactions is very important to delineate the complex plant-microbe signaling processes. Arabidopsis thaliana-Agrobacterium tumefaciens provides an excellent model system to study bacterial pathogenesis and plant interactions. Previous studies of plant-Agrobacterium interactions have largely relied on plant cell suspension cultures, the artificial wounding of plants, or the artificial induction of microbial virulence factors or plant defenses by synthetic chemicals. However, these methods are distinct from the natural signaling in planta, where plants and microbes recognize and respond in spatial and temporal manners. This work presents a hydroponic cocultivation system where intact plants are supported by metal mesh screens and cocultivated with Agrobacterium. In this cocultivation system, no synthetic phytohormone or chemical that induces microbial virulence or plant defense is supplemented. The hydroponic cocultivation system closely resembles natural plant-microbe interactions and signaling homeostasis in planta. Plant roots can be separated from the medium containing Agrobacterium, and the signaling and responses of both the plant hosts and the interacting microbes can be investigated simultaneously and systematically. At any given timepoint/interval, plant tissues or bacteria can be harvested separately for various "omics" analyses, demonstrating the power and efficacy of this system. The hydroponic cocultivation system can be easily adapted to study: 1) the reciprocal signaling of diverse plant-microbe systems, 2) signaling between a plant host and multiple microbial species (i.e. microbial consortia or microbiomes), 3) how nutrients and chemicals are implicated in plant-microbe signaling, and 4) how microbes interact with plant hosts and contribute to plant tolerance to biotic or abiotic stresses.

ANALYSIS OF THE ENERGY SYSTEM BALANCE EFFICIENCY PROVIDED WITH THE DIFFERENT GROUPS OF GENERATING PLANTS

Directory of Open Access Journals (Sweden)

O. Maksymovа

2017-12-01

Full Text Available Currently methods of efficiency analysis are being developed and applied, based on optimization tasks for various types and modes. Usually, the optimization criterion for these objectives is efficiency that can be calculated in various ways, for which there is no concurrent views. The target function based on minimization of given cost that allows comparing options with the same useful effect is used to search for the best indicators of power plants operated within the system. Marginal costs on the amount of difference in the useful effect are introduced to the target function in case of various useful effects. The criterion of selecting the best power plant from an economic point of view is the difference between the reduced costs of the considered and the basic options, but this approach does not allow using the results for long-term projections. Such approach depends on the situation and does not reflect the real costs. The value of the target function to optimize the effectiveness of the technical-economic method is not "marginal" and does not allow assessing the impact of various processes on the overall option efficiency. Therefore, the development of the efficiency criterion that considers the changing needs of the energy system is relevant for analyzing the power plant.

Systems analysis of the CANDU 3 Reactor

Energy Technology Data Exchange (ETDEWEB)

Wolfgong, J.R.; Linn, M.A.; Wright, A.L.; Olszewski, M.; Fontana, M.H. [Oak Ridge National Lab., TN (United States)

1993-07-01

This report presents the results of a systems failure analysis study of the CANDU 3 reactor design; the study was performed for the US Nuclear Regulatory Commission. As part of the study a review of the CANDU 3 design documentation was performed, a plant assessment methodology was developed, representative plant initiating events were identified for detailed analysis, and a plant assessment was performed. The results of the plant assessment included classification of the CANDU 3 event sequences that were analyzed, determination of CANDU 3 systems that are ``significant to safety,`` and identification of key operator actions for the analyzed events.

Preliminary Hazard Analysis applied to Uranium Hexafluoride - UF6 production plant

International Nuclear Information System (INIS)

Tomzhinsky, David; Bichmacher, Ricardo; Braganca Junior, Alvaro; Peixoto, Orpet Jose

1996-01-01

The purpose of this paper is to present the results of the Preliminary hazard Analysis applied to the UF 6 Production Process, which is part of the UF 6 Conversion Plant. The Conversion Plant has designed to produce a high purified UF 6 in accordance with the nuclear grade standards. This Preliminary Hazard Analysis is the first step in the Risk Management Studies, which are under current development. The analysis evaluated the impact originated from the production process in the plant operators, members of public, equipment, systems and installations as well as the environment. (author)

Fire hazard analysis at the first unit of the Ignalina nuclear power plant: 1. Analysis of fire prevention and ventilation systems and secondary effects

International Nuclear Information System (INIS)

Poskas, P.; Simonis, V.; Zujus, R. and others

2004-01-01

Evaluation of the fire prevention and ventilation systems and the secondary effects on safety at the Ignalina NPP from the point of view of fire hazard using computerized system is presented. Simplified screening algorithms for fire prevention, ventilation and the evaluation of secondary effects are developed, which allow accelerating fire hazard analysis at the Ignalina NPP. The analysis indicated that the fire prevention systems practically meet the national requirements and international recommendations for fire prevention. But it is necessary to introduce in separate rooms the measures improving fire prevention to guarantee the effective functioning of the ventilation systems and the reduction of the influence of secondary effects on safety. Computerized system of fire prevention and ventilation systems and evaluation of secondary effects on safety can be easily applied for fire hazard analysis at different big plants. (author)

Application of 3-dimensional CAD modeling system in nuclear plants

International Nuclear Information System (INIS)

Suwa, Minoru; Saito, Shunji; Nobuhiro, Minoru

1990-01-01

Until now, the preliminary work for mutual components in nuclear plant were readied by using plastic models. Recently with the development of computer graphic techniques, we can display the components on the graphics terminal, better than with use of plastic model and actual plants. The computer model can be handled, both telescopically and microscopically. A computer technique called 3-dimensional CAD modeling system was used as the preliminary work and design system. Through application of this system, database for nuclear plants was completed in arrangement step. The data can be used for piping design, stress analysis, shop production, testing and site construction, in all steps. In addition, the data can be used for various planning works, even after starting operation of plant. This paper describes the outline of the 3-dimensional CAD modeling system. (author)

Tele-maintenance 'intelligent' system for technical plants result management

International Nuclear Information System (INIS)

Concetti, Massimo; Cuccioletta, Roberto; Fedele, Lorenzo; Mercuri, Giampiero

2009-01-01

The management of technical plant for productivity and safety is generally a complex activity, particularly when many plants distributed in the territory are considered (i.e. the more and more frequent case of outsourced plants maintenance by specialized companies), granted quality and cost results are required (i.e. the case of some rather innovative contract solutions) and the technology involved is heterogeneous and innovative (i.e. electro-mechanical plants). In order to efficiently achieve the above aims an 'intelligent' maintenance-management system for the distant monitoring and controlling by a remote control center has been developed. The so-called GrAMS (granted availability management system) system is conceived to give to organizations involved in technical-industrial plants management the possibility to tend to a 'well-known availability' and 'zero-failures' management. In particular, this study deals with the diagnostic aspects and safety level of technical plants (such as elevators, thermo-technical plants, etc.), and with the involvement of ad hoc designed software analysis tools based on neural networks and reliability indicators. Part of the research dealing with the tele-maintenance intelligent system has been financed by the Italian High Institute for Safety (ISPESL) and led to the development of a pre-industrial prototype whose realization and testing is here described

PUREX Plant deactivation mission analysis report

International Nuclear Information System (INIS)

Lund, D.P.

1995-01-01

The purpose of the PUREX Deactivation Project mission analysis is to define the problem to be addressed by the PUREX mission, and to lay the ground work for further system definition. The mission analysis is an important first step in the System Engineering (SE) process. This report presents the results of the PUREX Deactivation Project mission analysis. The purpose of the PUREX Deactivation Project is to prepare PUREX for Decontamination and Decommissioning within a five year time frame. This will be accomplished by establishing a passively safe and environmentally secure configuration of the PUREX Plant, that can be preserved for a 10-year horizon. During deactivation, appropriate portions of the safety envelop will be maintained to ensure deactivation takes place in a safe and regulatory compliant manner

Techno-Economic Assessment of Redundancy Systems for a Cogeneration Plant

Directory of Open Access Journals (Sweden)

Majid Mohd Amin Abd

2014-07-01

Full Text Available The use of distributed power generation has advantage as well as disadvantage. One of the disadvantages is that the plant requires a dependable redundancy system to provide back up of power during failure of its power generation equipment. This paper presents a study on techno-economic assessment of redundancy systems for a cogeneration plant. Three redundancy systems were investigated; using public utility, generator set and gas turbine as back up during failures. Results from the analysis indicate that using public utility provides technical as well as economic advantages in comparison to using generator set or turbine as back up. However, the economic advantage of the public utility depends on the frequency of failures the plant will experience as well on the maximum demand charge. From the break even analysis of the understudied plant, if the number of failures exceeds 3 failures per year for the case of maximum demand charge of RM56.80, it is more economical to install a generator set as redundancy. The study will be useful for the co-generator operators to evaluate the feasibility of redundancy systems.

Development of Nuclear Plant Specific Analysis Simulators with ATLAS

International Nuclear Information System (INIS)

Jakubowski, Z.; Draeger, P.; Horche, W.; Pointner, W.

2006-01-01

The simulation software ATLAS, based on the best-estimate code ATHLET, has been developed by the GRS for a range of applications in the field of nuclear plant safety analysis. Through application of versatile simulation tools and graphical interfaces the user should be able to analyse with ATLAS all essential accident scenarios. Detailed analysis simulators for several German and Russian NPPs are being constructed on the basis of ATLAS. An overview of the ATLAS is presented in the paper, describing its configuration, functions performed by main components and relationships among them. A significant part of any power plant simulator are the balance-of-plant (BOP) models, not only because all the plant transients and non-LOCA accidents can be initiated by operation of BOP systems, but also because the response of the plant to transients or accidents is strongly influenced by the automatic operation of BOP systems. Modelling aspects of BOP systems are shown in detail, also the interface between the process model and BOP systems. Special emphasis has been put on the BOP model builder based on the methodology developed in the GRS. The BOP modeler called GCSM-Generator is an object oriented tool which runs on the online expert system G2. It is equipped with utilities to edit the BOP models, to verification them and to generate a GCSM code, specific for the ATLAS. The communication system of ATLAS presents graphically the results of the simulation and allows interactively influencing the execution of the simulation process (malfunctions, manual control). Displays for communications with simulated processes and presentation of calculations results are also presented. In the framework of the verification of simulation models different tools are used e.g. the PC-codes MATHCAD for the calculation and documentation, ATLET-Input-Graphic for control of geometry data and the expert system G2 for development of BOP-Models. The validation procedure and selected analyses results

Flood risk analysis procedure for nuclear power plants

International Nuclear Information System (INIS)

Wagner, D.P.

1982-01-01

This paper describes a methodology and procedure for determining the impact of floods on nuclear power plant risk. The procedures are based on techniques of fault tree and event tree analysis and use the logic of these techniques to determine the effects of a flood on system failure probability and accident sequence occurrence frequency. The methodology can be applied independently or as an add-on analysis for an existing risk assessment. Each stage of the analysis yields useful results such as the critical flood level, failure flood level, and the flood's contribution to accident sequence occurrence frequency. The results of applications show the effects of floods on the risk from nuclear power plants analyzed in the Reactor Safety Study

Plant monitor system

International Nuclear Information System (INIS)

Scarola, K.; Jamison, D.; Manazir, R.; Rescori, R.; Harmon, D.

1991-01-01

An advanced control room complex for a nuclear power plant, including a discrete indicator and alarm system which is nuclear qualified for rapid response to changes in plant parameters and a component control system which together provide a discrete monitoring and control capability at a panel in the control room. A separate data processing system, which need not be nuclear qualified, provides integrated and overview information to the control room and to each panel, through CRTs and a large, overhead integrated process status overview board. The discrete indicator and alarm system and the data processing system receive inputs from common plant sensors and validate the sensor outputs to arrive at a representative value of the parameter for use by the operator during both normal and accident conditions, thereby avoiding the need for him to assimilate data from each sensor individually. The integrated process status board is at the apex of an information hierarchy that extends through four levels and provides access at each panel to the full display hierarchy. The control room panels are preferably of a modular construction, permitting the definition of inputs and outputs, the main machine interface, and the plant specific algorithms, to proceed in parallel with the fabrication of the panels, the installation of the equipment and the generic testing thereof. (author)

Operational margin monitoring system for boiling water reactor power plants

International Nuclear Information System (INIS)

Fukutomi, S.; Takigawa, Y.

1992-01-01

This paper reports on an on-line operational margin monitoring system which has been developed for boiling water reactor power plants to improve safety, reliability, and quality of reactor operation. The system consists of a steady-state core status prediction module, a transient analysis module, a stability analysis module, and an evaluation and guidance module. This system quantitatively evaluates the thermal margin during abnormal transients as well as the stability margin, which cannot be evaluated by direct monitoring of the plant parameters, either for the current operational state or for a predicted operating state that may be brought about by the intended operation. This system also gives operator guidance as to appropriate or alternate operations when the operating state has or will become marginless

Abnormal transient analysis by using PWR plant simulator, (2)

International Nuclear Information System (INIS)

Naitoh, Akira; Murakami, Yoshimitsu; Yokobayashi, Masao.

1983-06-01

This report describes results of abnormal transient analysis by using a PWR plant simulator. The simulator is based on an existing 822MWe power plant with 3 loops, and designed to cover wide range of plant operation from cold shutdown to full power at EOL. In the simulator, malfunctions are provided for abnormal conditions of equipment failures, and in this report, 17 malfunctions for secondary system and 4 malfunctions for nuclear instrumentation systems were simulated. The abnormal conditions are turbine and generator trip, failure of condenser, feedwater system and valve and detector failures of pressure and water level. Fathermore, failure of nuclear instrumentations are involved such as source range channel, intermediate range channel and audio counter. Transient behaviors caused by added malfunctions were reasonable and detail information of dynamic characteristics for turbine-condenser system were obtained. (author)

Accident analysis for nuclear power plants

International Nuclear Information System (INIS)

2002-01-01

calculation results. This safety report also discusses various factors that need to be considered to ensure that the accident analysis is of an acceptable quality. The report is intended for use primarily by analyses coordinating, performing or reviewing accident analyses for NPPs, on both the utility and regulatory sides. The report will also be of use as a background document for relevant IAEA activities, such as training courses and workshops. While the main body of the report does not focus exclusively on a single reactor type, the examples provided in the annexes are related mostly to the accident analysis of NPPs with pressurized water reactors. The report: Applies to both NPPs being built and operating plants; deals with internal events in reactors or in their associated process systems; thus the emphasis is on the physical transient behaviour of reactors and their systems, including reactor containment; discusses both best estimate and conservative accident analyses; covers design basis accidents as well as beyond design basis accidents, although the design basis accidents are covered in greater detail; focuses on thermohydraulic aspects of safety analysis; neutronic, structural and radiological aspects are also covered to some extent; covers the course of an accident from the initiating event up to source term estimation. The main body of the report is intended to be as generally applicable as possible to all reactor types

BEAP profiles as rapid test system for status analysis and early detection of process incidents in biogas plants.

Science.gov (United States)

Refai, Sarah; Berger, Stefanie; Wassmann, Kati; Hecht, Melanie; Dickhaus, Thomas; Deppenmeier, Uwe

2017-03-01

A method was developed to quantify the performance of microorganisms involved in different digestion levels in biogas plants. The test system was based on the addition of butyrate (BCON), ethanol (ECON), acetate (ACON) or propionate (PCON) to biogas sludge samples and the subsequent analysis of CH 4 formation in comparison to control samples. The combination of the four values was referred to as BEAP profile. Determination of BEAP profiles enabled rapid testing of a biogas plant's metabolic state within 24 h and an accurate mapping of all degradation levels in a lab-scale experimental setup. Furthermore, it was possible to distinguish between specific BEAP profiles for standard biogas plants and for biogas reactors with process incidents (beginning of NH 4 + -N inhibition, start of acidification, insufficient hydrolysis and potential mycotoxin effects). Finally, BEAP profiles also functioned as a warning system for the early prediction of critical NH 4 + -N concentrations leading to a drop of CH 4 formation.

ITER plant systems

International Nuclear Information System (INIS)

Kolbasov, B.; Barnes, C.; Blevins, J.

1991-01-01

As part of a series of documents published by the IAEA that summarize the results of the Conceptual Design Activities for the ITER project, this publication describes the conceptual design of the ITER plant systems, in particular (i) the heat transport system, (ii) the electrical distribution system, (iii) the requirements for radioactive equipment handling, the hot cell, and waste management, (iv) the supply system for fluids and operational chemicals, (v) the qualitative analyses of failure scenarios and methods of burn stability control and emergency shutdown control, (vi) analyses of tokamak building functions and design requirements, (vii) a plant layout, and (viii) site requirements. Refs, figs and tabs

Thermodynamic and exergoeconomic analysis of a cement plant: Part I – Methodology

International Nuclear Information System (INIS)

Atmaca, Adem; Yumrutaş, Recep

2014-01-01

Highlights: • Energy, exergy and exergoeconomic analysis of a complete cement plant have been investigated. • The first and second law efficiencies based on the energy and exergy analysis are defined for the entire cement plant. • The specific energy consumption of the whole sections of the cement plant have been analyzed. • The specific manufacturing costs of farine, clinker and cement have been determined by the cost analysis. - Abstract: The energy, exergy and exergoeconomic analysis of a cement factory has been studied within two parts. This paper is the first part of the study which includes the thermodynamic and exergoeconomic methodology and formulations developed for such a comprehensive and detailed analysis. The second part of this study is about the application of the developed formulation which considers an actual cement plant located in Gaziantep, Turkey. The energy consumption by the cement industry is about 5% of the total global industrial energy consumption. It is also one of the world’s largest industrial sources of CO 2 emissions. In this paper, a cement plant is considered with all main manufacturing units. Mass, energy, and exergy balances are applied to each system. The first and second law efficiencies based on the energy and exergy analysis and performance assessment parameters are defined for the entire cement plant. The formulations for the cost of products, and cost formation and allocation within the system are developed based on exergoeconomic analysis. In order to obtain the optimal marketing price of cement and to decrease specific energy consumption of the whole plant, the cost analysis formulated here have substantial importance

Development of plant maintenance systems

International Nuclear Information System (INIS)

Tomita, Jinji; Ike, Masae; Nakayama, Kenji; Kato, Hisatomo

1989-01-01

Toshiba is making active efforts for the continuing improvement of reliability and maintainability of operating nuclear power plants. As a part of these efforts, the company has developed new maintenance administration systems, diagnostic monitoring facilities for plant equipments, computer-aided expert systems, and remote-controlled machines for maintenance work. The maintenance administration systems provide efficient work plans and data acquisition capabilities for the management of personnel and equipments involved in nuclear power plant maintenance. The plant diagnostic facilities monitor and diagnose plant conditions for preventive maintenance, as well as enabling rapid countermeasures to be carried out should they be required. Expert systems utilizing artificial intelligence (AI) technology are also employed. The newly developed remote-controlled machines are useful tools for the maintenance inspection of equipment which can not be easily accessed. (author)

Comparative exergetic performance analysis for certain thermal power plants in Serbia

Directory of Open Access Journals (Sweden)

Mitrović Dejan M.

2016-01-01

Full Text Available Traditional methods of analysis and calculation of complex thermal systems are based on the first law of thermodynamics. These methods use energy balance for a system. In general, energy balances do not provide any information about internal losses. In contrast, the second law of thermodynamics introduces the concept of exergy, which is useful in the analysis of thermal systems. Exergy is a measure for assessing the quality of energy, and allows one to determine the location, cause, and real size of losses incurred as well as residues in a thermal process. The purpose of this study is to comparatively analyze the performance of four thermal power plants from the energetic and exergetic viewpoint. Thermodynamic models of the plants are developed based on the first and second law of thermodynamics. The primary objectives of this paper are to analyze the system components separately and to identify and quantify the sites having largest energy and exergy losses. Finally, by means of these analyses, the main sources of thermodynamic inefficiencies as well as a reasonable comparison of each plant to others are identified and discussed. As a result, the outcomes of this study can provide a basis for the improvement of plant performance for the considered thermal power plants.

Determination of microelement distribution in different components of soil-plant system

International Nuclear Information System (INIS)

Luu Viet Hung; Maslov, O.D.; Gustova, M.V.; Trinh Thi Thu My; Phung Khac Nam Ho

2011-01-01

Leaves, stem, and roots of two shrub types: tea (Camellia sinensis), sweet leaf (Sauropus androgynus) and two herb types: vetiver grass (Vetiveria zizanioides L.Nash), maize (Zea mays L) and Thucuc soil where the plants grow, were collected to be studied. Contents of 22 elements in the samples were determined by three methods: XRFA (X-Ray Fluorescence Analysis), GAA (Gamma Activation Analysis), and tracking method to study distribution of these elements in plants and soil-plant relationship. The study was carried out at the Flerov Laboratory of Nuclear Reactions, JINR, Dubna. Distribution of the elements in the soil-plant system was studied

RAMI analysis for DEMO HCPB blanket concept cooling system

Energy Technology Data Exchange (ETDEWEB)

Dongiovanni, Danilo N., E-mail: danilo.dongiovanni@enea.it [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati (Italy); Pinna, Tonio [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati (Italy); Carloni, Dario [KIT, Institute of Neutron Physics and Reactor Technology (INR) – KIT (Germany)

2015-10-15

Highlights: • RAMI (reliability, availability, maintainability and inspectability) preliminary assessment for HCPB blanket concept cooling system. • Reliability block diagram (RBD) modeling and analysis for HCPB primary heat transfer system (PHTS), coolant purification system (CPS), pressure control system (PCS), and secondary cooling system. • Sensitivity analysis on system availability performance. • Failure models and repair models estimated on the base of data from the ENEA fusion component failure rate database (FCFRDB). - Abstract: A preliminary RAMI (reliability, availability, maintainability and inspectability) assessment for the HCPB (helium cooled pebble bed) blanket cooling system based on currently available design for DEMO fusion power plant is presented. The following sub-systems were considered in the analysis: blanket modules, primary cooling loop including pipework and steam generators lines, pressure control system (PCS), coolant purification system (CPS) and secondary cooling system. For PCS and CPS systems an extrapolation from ITER Test Blanket Module corresponding systems was used as reference design in the analysis. Helium cooled pebble bed (HCPB) system reliability block diagrams (RBD) models were implemented taking into account: system reliability-wise configuration, operating schedule currently foreseen for DEMO, maintenance schedule and plant evolution schedule as well as failure and corrective maintenance models. A simulation of plant activity was then performed on implemented RBDs to estimate plant availability performance on a mission time of 30 calendar years. The resulting availability performance was finally compared to availability goals previously proposed for DEMO plant by a panel of experts. The study suggests that inherent availability goals proposed for DEMO PHTS system and Tokamak auxiliaries are potentially achievable for the primary loop of the HCPB concept cooling system, but not for the secondary loop. A

Development of the simulation system {open_quotes}IMPACT{close_quotes} for analysis of nuclear power plant severe accidents

Energy Technology Data Exchange (ETDEWEB)

Naitoh, Masanori; Ujita, Hiroshi; Nagumo, Hiroichi [Nuclear Power Corp. (Japan)] [and others

1997-07-01

The Nuclear Power Engineering Corporation (NUPEC) has initiated a long-term program to develop the simulation system {open_quotes}IMPACT{close_quotes} for analysis of hypothetical severe accidents in nuclear power plants. IMPACT employs advanced methods of physical modeling and numerical computation, and can simulate a wide spectrum of senarios ranging from normal operation to hypothetical, beyond-design-basis-accident events. Designed as a large-scale system of interconnected, hierarchical modules, IMPACT`s distinguishing features include mechanistic models based on first principles and high speed simulation on parallel processing computers. The present plan is a ten-year program starting from 1993, consisting of the initial one-year of preparatory work followed by three technical phases: Phase-1 for development of a prototype system; Phase-2 for completion of the simulation system, incorporating new achievements from basic studies; and Phase-3 for refinement through extensive verification and validation against test results and available real plant data.

Fire risk analysis for nuclear power plants: Methodological developments and applications

International Nuclear Information System (INIS)

Kazarians, M.; Apostolakis, G.; Siv, N.O.

1985-01-01

A methodology to quantify the risk from fires in nuclear power plants is described. This methodology combines engineering judgment, statistical evidence, fire phenomenology, and plant system analysis. It can be divided into two major parts: (1) fire scenario identification and quantification, and (2) analysis of the impact on plant safety. This article primarily concentrates on the first part. Statistical analysis of fire occurrence data is used to establish the likelihood of ignition. The temporal behaviors of the two competing phenomena, fire propagation and fire detection and suppression, are studied and their characteristic times are compared. Severity measures are used to further specialize the frequency of the fire scenario. The methodology is applied to a switchgear room of a nuclear power plant

Pressure fluctuation analysis for charging pump of chemical and volume control system of nuclear power plant

Directory of Open Access Journals (Sweden)

Chen Qiang

2016-01-01

Full Text Available Equipment Failure Root Cause Analysis (ERCA methodology is employed in this paper to investigate the root cause for charging pump’s pressure fluctuation of chemical and volume control system (RCV in pressurized water reactor (PWR nuclear power plant. RCA project task group has been set up at the beginning of the analysis process. The possible failure modes are listed according to the characteristics of charging pump’s actual pressure fluctuation and maintenance experience during the analysis process. And the failure modes are analysed in proper sequence by the evidence-collecting. It suggests that the gradually untightened and loosed shaft nut in service should be the root cause. And corresponding corrective actions are put forward in details.

Graphical analysis of French nuclear power plant production date

Energy Technology Data Exchange (ETDEWEB)

Jourdan, J.P. [Electricite de France (EDF), Projet Production EPR 1, 93 - Saint-Denis (France)

2001-07-01

The analysis of values of plant production uses here an original method of graphical analysis. This method clarifies various difficulties of analysing big experience feedback databases among which the language interpretation and distinctions between scarce events and multi-annual events. In general, the method shows the logical processes that production values obey (pure chance logic, administrative logic, and willpower) This method of graphical analysis provides a tool to observe and question in a concrete way so that each person involved can put the events in which he played a role into the general context of other plants. It is a deductible method to improve this big and complex system. (author)

Graphical analysis of French nuclear power plant production date

International Nuclear Information System (INIS)

Jourdan, J.P.

2001-01-01

The analysis of values of plant production uses here an original method of graphical analysis. This method clarifies various difficulties of analysing big experience feedback databases among which the language interpretation and distinctions between scarce events and multi-annual events. In general, the method shows the logical processes that production values obey (pure chance logic, administrative logic, and willpower) This method of graphical analysis provides a tool to observe and question in a concrete way so that each person involved can put the events in which he played a role into the general context of other plants. It is a deductible method to improve this big and complex system. (author)

Global biogeography of mating system variation in seed plants.

Science.gov (United States)

Moeller, David A; Briscoe Runquist, Ryan D; Moe, Annika M; Geber, Monica A; Goodwillie, Carol; Cheptou, Pierre-Olivier; Eckert, Christopher G; Elle, Elizabeth; Johnston, Mark O; Kalisz, Susan; Ree, Richard H; Sargent, Risa D; Vallejo-Marin, Mario; Winn, Alice A

2017-03-01

Latitudinal gradients in biotic interactions have been suggested as causes of global patterns of biodiversity and phenotypic variation. Plant biologists have long speculated that outcrossing mating systems are more common at low than high latitudes owing to a greater predictability of plant-pollinator interactions in the tropics; however, these ideas have not previously been tested. Here, we present the first global biogeographic analysis of plant mating systems based on 624 published studies from 492 taxa. We found a weak decline in outcrossing rate towards higher latitudes and among some biomes, but no biogeographic patterns in the frequency of self-incompatibility. Incorporating life history and growth form into biogeographic analyses reduced or eliminated the importance of latitude and biome in predicting outcrossing or self-incompatibility. Our results suggest that biogeographic patterns in mating system are more likely a reflection of the frequency of life forms across latitudes rather than the strength of plant-pollinator interactions. © 2017 John Wiley & Sons Ltd/CNRS.

The Functional Resonance Analysis Method for a systemic risk based environmental auditing in a sinter plant: A semi-quantitative approach

International Nuclear Information System (INIS)

Patriarca, Riccardo; Di Gravio, Giulio; Costantino, Francesco; Tronci, Massimo

2017-01-01

Environmental auditing is a main issue for any production plant and assessing environmental performance is crucial to identify risks factors. The complexity of current plants arises from interactions among technological, human and organizational system components, which are often transient and not easily detectable. The auditing thus requires a systemic perspective, rather than focusing on individual behaviors, as emerged in recent research in the safety domain for socio-technical systems. We explore the significance of modeling the interactions of system components in everyday work, by the application of a recent systemic method, i.e. the Functional Resonance Analysis Method (FRAM), in order to define dynamically the system structure. We present also an innovative evolution of traditional FRAM following a semi-quantitative approach based on Monte Carlo simulation. This paper represents the first contribution related to the application of FRAM in the environmental context, moreover considering a consistent evolution based on Monte Carlo simulation. The case study of an environmental risk auditing in a sinter plant validates the research, showing the benefits in terms of identifying potential critical activities, related mitigating actions and comprehensive environmental monitoring indicators. - Highlights: • We discuss the relevance of a systemic risk based environmental audit. • We present FRAM to represent functional interactions of the system. • We develop a semi-quantitative FRAM framework to assess environmental risks. • We apply the semi-quantitative FRAM framework to build a model for a sinter plant.

The Functional Resonance Analysis Method for a systemic risk based environmental auditing in a sinter plant: A semi-quantitative approach

Energy Technology Data Exchange (ETDEWEB)

Patriarca, Riccardo, E-mail: riccardo.patriarca@uniroma1.it; Di Gravio, Giulio; Costantino, Francesco; Tronci, Massimo

2017-03-15

Environmental auditing is a main issue for any production plant and assessing environmental performance is crucial to identify risks factors. The complexity of current plants arises from interactions among technological, human and organizational system components, which are often transient and not easily detectable. The auditing thus requires a systemic perspective, rather than focusing on individual behaviors, as emerged in recent research in the safety domain for socio-technical systems. We explore the significance of modeling the interactions of system components in everyday work, by the application of a recent systemic method, i.e. the Functional Resonance Analysis Method (FRAM), in order to define dynamically the system structure. We present also an innovative evolution of traditional FRAM following a semi-quantitative approach based on Monte Carlo simulation. This paper represents the first contribution related to the application of FRAM in the environmental context, moreover considering a consistent evolution based on Monte Carlo simulation. The case study of an environmental risk auditing in a sinter plant validates the research, showing the benefits in terms of identifying potential critical activities, related mitigating actions and comprehensive environmental monitoring indicators. - Highlights: • We discuss the relevance of a systemic risk based environmental audit. • We present FRAM to represent functional interactions of the system. • We develop a semi-quantitative FRAM framework to assess environmental risks. • We apply the semi-quantitative FRAM framework to build a model for a sinter plant.

Wind power plant system services

DEFF Research Database (Denmark)

Basit, Abdul; Altin, Müfit

Traditionally, conventional power plants have the task to support the power system, by supplying power balancing services. These services are required by the power system operators in order to secure a safe and reliable operation of the power system. However, as in the future the wind power...... is going more and more to replace conventional power plants, the sources of conventional reserve available to the system will be reduced and fewer conventional plants will be available on-line to share the regulation burden. The reliable operation of highly wind power integrated power system might...... then beat risk unless the wind power plants (WPPs) are able to support and participate in power balancing services. The objective of this PhD project is to develop and analyse control strategies which can increase the WPPs capability to provide system services, such as active power balancing control...

Recommendations on the choice of gas analysis equipment for systems of continuous monitoring and accounting of emissions from thermal power plants

Science.gov (United States)

Kondrat'eva, O. E.; Roslyakov, P. V.; Burdyukov, D. A.; Khudolei, O. D.; Loktionov, O. A.

2017-10-01

According to Federal Law no. 219-FZ, dated July 21, 2014, all enterprises that have a significant negative impact on the environment shall continuously monitor and account emissions of harmful substances into the atmospheric air. The choice of measuring equipment that is included in continuous emission monitoring and accounting systems (CEM&ASs) is a complex technical problem; in particular, its solution requires a comparative analysis of gas analysis systems; each of these systems has its advantages and disadvantages. In addition, the choice of gas analysis systems for CEM&ASs should be maximally objective and not depend on preferences of separate experts and specialists. The technique of choosing gas analysis equipment that was developed in previous years at Moscow Power Engineering Institute (MPEI) has been analyzed and the applicability of the mathematical tool of a multiple criteria analysis to choose measuring equipment for the continuous emission monitoring and accounting system have been estimated. New approaches to the optimal choice of gas analysis equipment for systems of the continuous monitoring and accounting of harmful emissions from thermal power plants have been proposed, new criteria of evaluation of gas analysis systems have been introduced, and weight coefficients have been determined for these criteria. The results of this study served as a basis for the Preliminary National Standard of the Russian Federation "Best Available Technologies. Automated Systems of Continuous Monitoring and Accounting of Emissions of Harmful (Polluting) Substances from Thermal Power Plants into the Atmospheric Air. Basic Requirements," which was developed by the Moscow Power Engineering Institute, National Research University, in cooperation with the Council of Power Producers and Strategic Electric Power Investors Association and the All-Russia Research Institute for Materials and Technology Standardization.

The System 80+ Standard Plant Information Management System

Energy Technology Data Exchange (ETDEWEB)

Turk, R.S.; Bryan, R.E. [ABB Combuions Engineering Nuclear Systems (United States)

1998-07-01

Historically, electric nuclear power plant owners, following the completion of construction and startup, have been left with a mountain of hard-copy documents and drawings. Hundreds of thousands of hours are spent searching for relevant documents and, in most cases, the documents found require many other documents and drawings to fully understand the design basis. All too often the information is incomplete, and eventually becomes obsolete. In the U.S., utilities spend millions of dollars to discover design basis information and update as-built data for each plant. This information must then be stored in an easily accessed usable form to assist satisfy regulatory requirements and to improve plant operating efficiency. ABB Combustion Engineering Nuclear Systems (ABB-CE) has an active program to develop a state-of-the-art Plant Information Management System (IMS) for its advanced light water reactor, the System 80+TM Standard Plant Design. This program is supported by ABB's Product Data Management (PDM) and Computer Aided Engineering (CAE) efforts world wide. This paper describes the System 80+ plant IMS and how it will be used during the entire life cycle of the plant. (author)

The System 80+ Standard Plant Information Management System

International Nuclear Information System (INIS)

Turk, R.S.; Bryan, R.E.

1998-01-01

Historically, electric nuclear power plant owners, following the completion of construction and startup, have been left with a mountain of hard-copy documents and drawings. Hundreds of thousands of hours are spent searching for relevant documents and, in most cases, the documents found require many other documents and drawings to fully understand the design basis. All too often the information is incomplete, and eventually becomes obsolete. In the U.S., utilities spend millions of dollars to discover design basis information and update as-built data for each plant. This information must then be stored in an easily accessed usable form to assist satisfy regulatory requirements and to improve plant operating efficiency. ABB Combustion Engineering Nuclear Systems (ABB-CE) has an active program to develop a state-of-the-art Plant Information Management System (IMS) for its advanced light water reactor, the System 80+TM Standard Plant Design. This program is supported by ABB's Product Data Management (PDM) and Computer Aided Engineering (CAE) efforts world wide. This paper describes the System 80+ plant IMS and how it will be used during the entire life cycle of the plant. (author)

Operating plant safety analysis needs

International Nuclear Information System (INIS)

Young, M.Y.; Love, D.S.

1992-01-01

The primary objective for nuclear power station owners is to operate and manage their plants safely. However, there is also a need to provide economical electric power, which requires that the unit be operated as efficiently as possible, consistent with the safety requirements. The objectives cited above can be achieved through the identification and use of available margins inherent in the plant design. As a result of conservative licensing and analytical approaches taken in the past, many of these margins may be found in the safety analysis limits within which plants currently operate. Improvements in the accuracy of the safety analysis, and a more realistic treatment of plant initial and boundary conditions, can make this margin available for a variety of uses which enhance plant performance, help to reduce O and M costs, and may help to extend licensed operation. Opportunities for improvement exist in several areas in the accident analysis normally performed for Chapter 15 of the FSAR. For example, recent modifications to the ECCS rule, 10CFR50.46 and Appendix K, allow use of margins previously unavailable in the analysis of the Loss of Coolant Accident (LOCA). To take advantage of this regulatory change, new methods are being developed to analyze both the large and small break loss of coolant accident (LOCA). As this margin is used, enhancements in the analysis of other transients will become necessary. The paper discusses accident analysis methods, future development needs, and analysis margin utilization in specific accident scenarios

Performance Analysis of an Updraft Tower System for Dry Cooling in Large-Scale Power Plants

Directory of Open Access Journals (Sweden)

Haotian Liu

2017-11-01

Full Text Available An updraft tower cooling system is assessed for elimination of water use associated with power plant heat rejection. Heat rejected from the power plant condenser is used to warm the air at the base of an updraft tower; buoyancy-driven air flows through a recuperative turbine inside the tower. The secondary loop, which couples the power plant condenser to a heat exchanger at the tower base, can be configured either as a constant-pressure pump cycle or a vapor compression cycle. The novel use of a compressor can elevate the air temperature in the tower base to increases the turbine power recovery and decrease the power plant condensing temperature. The system feasibility is evaluated by comparing the net power needed to operate the system versus alternative dry cooling schemes. A thermodynamic model coupling all system components is developed for parametric studies and system performance evaluation. The model predicts that constant-pressure pump cycle consumes less power than using a compressor; the extra compression power required for temperature lift is much larger than the gain in turbine power output. The updraft tower system with a pumped secondary loop can allow dry cooling with less power plant efficiency penalty compared to air-cooled condensers.

Localization of equipment for digital plant protection system

Energy Technology Data Exchange (ETDEWEB)

Koo, I. S.; Park, H. Y.; Lee, C. K. and others

2000-10-01

The objective of this project lies on the development of design requirements, establishment of structure and manufacture procedures, development of the software verification and validation(V and V) techniques of the digital plant protection system. The functional requirements based on the analog protection system and digital design requirements are introduced, the processor and system bus for safety grade equipment are selected and the interface requirements and the design specification have been developed in order to manufacture the quick prototype of the digital plant protection system. The selection guidelines of parts, software development and coding and testing for digital plant protection system have been performed through manufacturing the quick prototype based on the developed design specification. For the software verification and validation, the software review plan and techniques of verification and validation have been researched. The digital validation system is developed in order to verify the quick prototype. The digital design requirements are reviewed by the software safety plan and V and V plans. The formal methods for verifying the safety-grade software are researched, then the methodology of formal analysis and testing have been developed.

Localization of equipment for digital plant protection system

International Nuclear Information System (INIS)

Koo, I. S.; Park, H. Y.; Lee, C. K. and others

2000-10-01

The objective of this project lies on the development of design requirements, establishment of structure and manufacture procedures, development of the software verification and validation(V and V) techniques of the digital plant protection system. The functional requirements based on the analog protection system and digital design requirements are introduced, the processor and system bus for safety grade equipment are selected and the interface requirements and the design specification have been developed in order to manufacture the quick prototype of the digital plant protection system. The selection guidelines of parts, software development and coding and testing for digital plant protection system have been performed through manufacturing the quick prototype based on the developed design specification. For the software verification and validation, the software review plan and techniques of verification and validation have been researched. The digital validation system is developed in order to verify the quick prototype. The digital design requirements are reviewed by the software safety plan and V and V plans. The formal methods for verifying the safety-grade software are researched, then the methodology of formal analysis and testing have been developed

Application of the failure modes and effects analysis technique to theemergency cooling system of an experimental nuclear power plant

International Nuclear Information System (INIS)

Conceicao Junior, Osmar

2009-01-01

This study consists on the application of the Failure Modes and EffectsAnalysis (FMEA), a hazard identification and a risk assessment technique, tothe Emergency Cooling System (ECS) of an experimental nuclear power plant,which is responsible for mitigating the consequences of an eventual loss ofcoolant accident on the Pressurized Water Reactor (PWR). Such analysisintends to identify possible weaknesses on the design of the system andpropose some improvements in order to maximize its reliability. To achievethis goal a detailed study of the system was carried on (through itstechnical documentation), the correspondent reliability block diagram wasobtained, the FMEA analysis was executed and, finally, some suggestions werepresented. (author)

Design of Nuclear Power Plant Online Monitoring System

International Nuclear Information System (INIS)

An, Sang-ha; Jeong, Yong-hoon; Chang, Soon-heung; Lee, Song-kyu

2007-01-01

Statistical Quality Control techniques have been applied to many aspects of industrial engineering. An application to nuclear power plant maintenance and control is also presented that can greatly improve plant safety. As a demonstration of such an approach, a specific system is analyzed: the reactor coolant pumps (RCPs) and the fouling resistance of heat exchanger. This research uses Shewart X-bar, R charts, Cumulative Sum charts (CUSUM), and Sequential Probability Ratio Test (SPRT) to analyze the process for the state of statistical control. And the Control Chart Analyzer (CCA) has been made to support these analyses that can make a decision of error in process. The analysis shows that statistical process control methods can be applied as an early warning system capable of identifying significant equipment problems well in advance of traditional control room alarm indicators. Such a system would provide operators with enough time to respond to possible emergency situations and thus improve plant safety and reliability

ChemAND - a system health monitor for plant chemistry

Energy Technology Data Exchange (ETDEWEB)

Turner, C.W.; Mitchel, G.R.; Tosello, G.; Balakrishnan, P.V.; McKay, G.; Thompson, M. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Dundar, Y.; Bergeron, M.; Laporte, R. [Hydro-Quebec, Groupe Chimie, Centrale Nucleaire Gentilly-2, Gentilly, Quebec (Canada)

2001-03-01

Effective management of plant systems throughout their lifetime requires much more than data acquisition and display - it requires that the plant's system health be continually monitored and managed. AECL has developed a System Health Monitor called ChemAND for CANDU plant chemistry. ChemAND, a Chemistry ANalysis and Diagnostic system, monitors key chemistry parameters in the heat transport system, moderator-cover gas, annulus gas, and the steam cycle during full-power operation. These parameters can be used as inputs to models that calculate the effect of current plant operating conditions on the present and future health of the system. Chemistry data from each of the systems are extracted on a regular basis from the plant's Historical Data Server and are sorted according to function, e.g., indicators for condenser in-leakage, air in-leakage, heavy water leakage into the annulus gas, fuel failure, etc. Each parameter is conveniently displayed and is trended along with its alarm limits. ChemAND currently includes two analytical models developed for the balance-of-plant. The first model, ChemSolv, calculates crevice chemistry conditions in the steam generator (SG) from either the SG blowdown chemistry conditions or from a simulated condenser leak. This information can be used by plant staff to evaluate the susceptibility of the SG tubes to crevice corrosion. ChemSolv also calculates chemistry conditions throughout the steam-cycle system as determined by the transport of volatile species such as ammonia, hydrazine, morpholine, and oxygen. The second model, SLUDGE, calculates the deposit loading and distribution in the SG as a function of time, based on concentrations of corrosion product in the final feedwater for both normal and start-up conditions. Operations personnel can use this information to predict where to inspect and when to clean. (author)

ChemAND - a system health monitor for plant chemistry

International Nuclear Information System (INIS)

Turner, C.W.; Mitchell, G.R.; Tosello, G.; Balakrishnan, P.V.; McKay, G.; Thompson, M.; Dundar, Y.; Bergeron, M.; Laporte, R.

2001-01-01

Effective management of plant systems throughout their lifetime requires much more than data acquisition and display-it requires that the plant's system health be continually monitored and managed. AECL has developed a System Health Monitor called ChemAND for CANDU plant chemistry. ChemAND, a Chemistry ANalysis and Diagnostic system, monitors key chemistry parameters in the heat transport system, moderator-cover gas, annulus gas, and the steam cycle during full-power operation. These parameters can be used as inputs to models that calculate the effect of current plant operating conditions on the present and future health of the system. Chemistry data from each of the systems are extracted on a regular basis from the plant's Historical Data Server and are sorted according to function, e.g., indicators for condenser in-leakage, air in-leakage, heavy water leakage into the annulus gas, fuel failure, etc. Each parameter is conveniently displayed and is trended along with its alarm limits. ChemAND currently includes two analytical models developed for the balance-of-plant. The first model, ChemSolv, calculates crevice chemistry conditions in the steam generator (SG) from either the SG blowdown chemistry conditions or from a simulated condenser leak. This information can be used by plant staff to evaluate the susceptibility of the SG tubes to crevice corrosion. ChemSolv also calculates chemistry conditions throughout the steam cycle system, as determined by the transport of volatile species such as ammonia, hydrazine, morpholine, and oxygen. The second model, SLUDGE, calculates the deposit loading and distribution in the SG as a function of time, based on concentrations of corrosion product in the final feedwater for both normal and start-up conditions. Operations personnel can use this information to predict where to inspect and when to clean. (author)

ChemAND - a system health monitor for plant chemistry

International Nuclear Information System (INIS)

Turner, C.W.; Mitchel, G.R.; Tosello, G.; Balakrishnan, P.V.; McKay, G.; Thompson, M.; Dundar, Y.; Bergeron, M.; Laporte, R.

2001-03-01

Effective management of plant systems throughout their lifetime requires much more than data acquisition and display - it requires that the plant's system health be continually monitored and managed. AECL has developed a System Health Monitor called ChemAND for CANDU plant chemistry. ChemAND, a Chemistry ANalysis and Diagnostic system, monitors key chemistry parameters in the heat transport system, moderator-cover gas, annulus gas, and the steam cycle during full-power operation. These parameters can be used as inputs to models that calculate the effect of current plant operating conditions on the present and future health of the system. Chemistry data from each of the systems are extracted on a regular basis from the plant's Historical Data Server and are sorted according to function, e.g., indicators for condenser in-leakage, air in-leakage, heavy water leakage into the annulus gas, fuel failure, etc. Each parameter is conveniently displayed and is trended along with its alarm limits. ChemAND currently includes two analytical models developed for the balance-of-plant. The first model, ChemSolv, calculates crevice chemistry conditions in the steam generator (SG) from either the SG blowdown chemistry conditions or from a simulated condenser leak. This information can be used by plant staff to evaluate the susceptibility of the SG tubes to crevice corrosion. ChemSolv also calculates chemistry conditions throughout the steam-cycle system as determined by the transport of volatile species such as ammonia, hydrazine, morpholine, and oxygen. The second model, SLUDGE, calculates the deposit loading and distribution in the SG as a function of time, based on concentrations of corrosion product in the final feedwater for both normal and start-up conditions. Operations personnel can use this information to predict where to inspect and when to clean. (author)

Invariant methods for an ensemble-based sensitivity analysis of a passive containment cooling system of an AP1000 nuclear power plant

International Nuclear Information System (INIS)

Di Maio, Francesco; Nicola, Giancarlo; Borgonovo, Emanuele; Zio, Enrico

2016-01-01

Sensitivity Analysis (SA) is performed to gain fundamental insights on a system behavior that is usually reproduced by a model and to identify the most relevant input variables whose variations affect the system model functional response. For the reliability analysis of passive safety systems of Nuclear Power Plants (NPPs), models are Best Estimate (BE) Thermal Hydraulic (TH) codes, that predict the system functional response in normal and accidental conditions and, in this paper, an ensemble of three alternative invariant SA methods is innovatively set up for a SA on the TH code input variables. The ensemble aggregates the input variables raking orders provided by Pearson correlation ratio, Delta method and Beta method. The capability of the ensemble is shown on a BE–TH code of the Passive Containment Cooling System (PCCS) of an Advanced Pressurized water reactor AP1000, during a Loss Of Coolant Accident (LOCA), whose output probability density function (pdf) is approximated by a Finite Mixture Model (FMM), on the basis of a limited number of simulations. - Highlights: • We perform the reliability analysis of a passive safety system of Nuclear Power Plant (NPP). • We use a Thermal Hydraulic (TH) code for predicting the NPP response to accidents. • We propose an ensemble of Invariant Methods for the sensitivity analysis of the TH code • The ensemble aggregates the rankings of Pearson correlation, Delta and Beta methods. • The approach is tested on a Passive Containment Cooling System of an AP1000 NPP.

Semi-Markov reliability analysis of alternating systems in a nuclear power plant

International Nuclear Information System (INIS)

Lee, K.N.; Cho, N.Z.

1992-01-01

Nuclear power plant operations that follow current testing and maintenance requirements sometimes result in inadvertent reactor trips, and operating staffs devote a significant amount of time and effort in complying with these requirements. Significant benefits could result from changes in current technical specifications. In this paper the benefits and impacts of changes in allowed outage times (AOTs) and surveillance test intervals (STIs) are evaluated for an alternative system that consists of multiple trains and whose operation is alternated train by train. because of testing and AOT requirements, the alternating system exhibits semi-Markovian characteristics that change states in accordance with a Markov process but take an arbitrarily distributed amount of time between changes. The state probabilities are quantified by memorizing the necessary number of past state probabilities. Two measures of plant performance, namely, core damage probability and plant unavailability (reactor downtime), were calculated for the evaluation of AOT and STI. Results indicate that there is an optimal point that gives the lowest core damage probability and that the methodology developed in this study can be applied to existing alternating systems to evaluate accurately the various alternatives in the technical specifications

Support analysis for safety analysis development for CANDU nuclear power plant

International Nuclear Information System (INIS)

Bedreaga, L.; Florescu, Gh.; Apostol, M.; Nitoi, M.

2004-01-01

Probabilistic Safety Assessment analysis (PSA) is a technique used to assess the safety of a nuclear power plant. Assessments of the nuclear plant systems/components from safety point of view consist in accomplishment of a lot of support analyses that are the base for the main analysis, in order to evaluate the impact of occurrences of abnormal states for these systems. Evaluation of initiating events frequency and components failure rate is based on underlying probabilistic theory and mathematic statistics. Some of these analyses are detailed analyses and are known very well in PSA. There are also some analyses, named support analyses for PSA, which are very important but less applicable because they involve a huge human effort and hardware facilities to accomplish. The usual methods applicable in PSA such as input data extracted from the specific documentation (operation procedures, testing procedures, maintenance procedures and so on) or conservative evaluation provide a high level of uncertainty for both input and output data. The paper describes support analysis required to improve the certainty level in evaluation of reliability parameters and also in the final results (either risk, reliability or safety assessment). (author)

Remote inspection system for nuclear power plants

International Nuclear Information System (INIS)

Inagaki, K.; Fujii, M.; Doi, A.; Harima, T.

1977-01-01

A remote inspection system for nuclear power plants was constructed based on an analysis of inspections performed by an operator on patrol. This system consists of an operator's console and a remote station. The remote station, equipped with five kinds of sensors, is steered along the inspection route by a photoelectric guiding system or may be manually controlled from an operator's console in a main control room. Signals for control and inspection data are multiplexed and transmitted through a coaxial cable

Exergy analysis of a circulating fluidized bed boiler cogeneration power plant

International Nuclear Information System (INIS)

Gürtürk, Mert; Oztop, Hakan F.

2016-01-01

Highlights: • Analysis of energy and exergy for a cogeneration power plant have been performed. • This plant has circulating fluidized bed boiler. • Energy and exergy efficiencies of the boiler are obtained as 84.65% and 29.43%, respectively. • Exergy efficiency of the plant was calculated as 20%. - Abstract: In this study, energy and exergy analysis of a cogeneration power plant have been performed. The steam which is produced by the cogeneration power plant is used for salt production and most important part of the cogeneration power plant is the circulation fluidized bed boiler. Energy and exergy efficiency of the circulation fluidized bed boiler were found as 84.65% and 29.43%, respectively. Exergy destruction of the circulation fluidized bed boiler was calculated as 21789.39 kW and 85.89% of exergy destruction in the plant. The automation system of the cogeneration power plant is insufficient. Exergy efficiency of the plant was calculated as 20%. Also, some design parameters increasing energy losses were determined.

Application of analysis technology in nuclear plant

International Nuclear Information System (INIS)

Takaoka, Keiko; Miura, Hiromi; Umeda, Kenji

1996-01-01

Recently, thanks to the rapid improvement of EWS performance, the authors have been able to carry out design evaluation comparatively, easily, utilizing computational fluid dynamics (CFD) technology. The Nuclear Plant Engineering Department has carried out some analyses in the past several years with the main purpose of evaluating the design of nuclear reactor internals. These studies included ''Thermal Hydraulic Analysis for Top Plenum'' and ''Flow Analysis for Lower Plenum''. It is considered to be a special matter in thermal hydraulic analysis of the top plenum that temperature distribution has been estimated with a relatively small number of meshes by means of an imaginary spray nozzle, and in the flow analysis for the lower plenum that flow distribution has been found to change largely, depending on the reactor internals. One of the ways to confirm the safety of nuclear plants, detailed structural analysis, is required for all possible combinations of transient and load conditions during operation. In particular, it is very important to clarify the thermal stress behavior under operating conditions and to evaluate fatigue analysis in accordance with the Code Requirements. However, it is very complicated and it takes a lot of time. A new system was developed which can operate continuously all of the definitions of the analytical model, the analyzation of pressurized thermal and external stress, and editing reports. In this paper, the authors introduce this system and apply it to a pressurized water reactor

Analysis of synchronous and induction generators used at hydroelectric power plant

Science.gov (United States)

Diniş, C. M.; Popa, G. N.; lagăr, A.

2017-01-01

In this paper is presented an analysis of the operating electric generators (synchronous and induction) within a small capacity hydroelectric power plant. Such is treated the problem of monitoring and control hydropower plant using SCADA systems. Have been carried an experimental measurements in small hydropower plant for different levels of water in the lake and various settings of the operating parameters.

Plant utilization against digestive system disorder in Southern Assam, India.

Science.gov (United States)

Choudhury, Prakash Roy; Choudhury, Manabendra Dutta; Ningthoujam, Sanjoy Singh; Mitra, Abhijit; Nath, Deepa; Talukdar, Anupam Das

2015-12-04

Being one of the most common types of life threatening diseases in Southern Assam, India, the digestive system disorders (DSD) have gained much attention in recent decades. Traditional beliefs and inadequate income of mass population result in the use of alternative phytotherapies to treat the diseases. The present paper documents the medicinal knowledge and utilization of plants for treatment of digestive system disorders in Southern Assam, India by Disease Consensus Index (DCI). It also determines the most suitable plant species used to treat digestive system disorders in the study area. The study was based on ethnomedicinal field survey covering a period of 1 year from 2014-2015. The ethnomedicinal information was collected by using semi-structured questionnaires from different traditional Bengali people having knowledge on medicinal plants. Collected data were analyzed by calculating DCI. During the survey, 29 informants were interviewed and a total of 49 plants under 46 genera belonging to 33 families were listed. Data analysis revealed that Litsea glutinosa, Momordica charantia, Andrographis paniculata, Lawsonia inermis, Cleome viscosa, Psidium guajava, Ageratum conyzoides, Cuscuta reflexa, Cynodon dactylon and Carica papaya are the most prominent plants among the people of Southern Assam for treating DSD. This explorative survey emphasizes the need to preserve and document the traditional healing practices for managing DSD inviting for more imminent scientific research on the plants to determine their efficacy as well as safety. With the help of statistical analysis (DCI), we propose 10 priority plants for DSD in present work. Systematic pharmacological study with these plants may contribute significant result. Copyright © 2015 Elsevier Ireland Ltd. All rights reserved.

Multielement neutron-activation analysis of plants and fertilizers

International Nuclear Information System (INIS)

Srapenyants, R.A.; Saveliev, I.B.

1977-01-01

The development of an automated technique for simultaneous multielement activation analysis of plants and fertilizers for the macronutrient elements N, P, K, Ca, Mg, Cl, and Si is presented. The developed universal NAA is based on the installation manufactured and supplied by Sames, France. The components of the automatic installation for neutron activation analysis are: neutron generator; a pneumatic transfer system; a scintillation crystal detector; a spectrometer rack including a basic multichannel analyser; a control panel for the neutron generator and pneumatic transfer system; a computer and teletype. On the basis of analytical procedures, algorithms and software, the first automatic (computer based) installation for multielement analyses of plants and fertilizers has been completed and is in routine use in the agrochemical and plant breeding research program in the Soviet Union. The proposed technique together with the full automatic real-time process of measurement and processing of data by computer, provides a throughput of 250-500 samples (1250-2500 elements determinations) per 8-hour shift, with the accuracy of +-3%; for N and +-5%; for P, K, Mg, Cl and +-15% for Ca. (T.G.)

A landscape simulation system for power plants

International Nuclear Information System (INIS)

Yoshinaga, Toshiaki; Yoshida, Miki; Usami, Yoshiaki.

1997-01-01

As scenes of power plants give many influences to environments, the plants that harmonized with the environments are demanded. We developed a landscape simulation system for the plants by using computer graphics technologies. This system has functions to generate realistic images about plant buildings and environments. Since the system contains information of ridge lines in addition to usual terrain data, the terrain shapes are expressed more precisely. Because the system enables users to visualize plant construction plans, the advance evaluations of plant scenes become possible. We regard this system as useful for environmental assessment of power plants. (author)

Radiation risk analysis of tritium in PWR plants

International Nuclear Information System (INIS)

Yang Maochun; Wang Shimin

1999-03-01

Tritium is a common radionuclide in PWR nuclear power plant. In the normal operation conditions, its radiation risk to plant workers is the internal radiation exposure when tritium existing in air as HTO (hydrogen tritium oxide) is breathed in. As the HTO has the same physical and chemical characteristics as water, the main way that HTO entering the air is by evaporation. There are few opening systems in Nuclear Power Plant, the radiation risk of tritium mainly exists near the area of spent fuel pit and reactor pit. The highest possible radiation risk it may cause--the maximum concentration in air is the level when equilibrium is established between water and air phases for tritium. The author analyzed the relationship among the concentration of HTO in water, in air and the water temperature when equilibrium is established, the equilibrated HTO concentration in air increases with HTO concentration in water and water temperature. The analysis revealed that at 30 degree C, the equilibrated HTO concentration in air might reach 1 DAC (derived air concentration) when the HTO concentration in water is 28 GBq/m 3 . Owing to the operation of plant ventilation systems and the existence of moisture in the input air of the ventilation, the practical tritium concentration in air is much lower than its equilibrated levels, the radiation risk of tritium in PWR plant is quite limited. In 1997, Daya Bay Nuclear Power Plant's practical monitoring result of the HTO concentration in the air of the nuclear island and the urine of workers supported this conclusion. Based on this analysis, some suggestions to the reduction of tritium radiation risk were made

Integrated systems analysis of the PIUS reactor

International Nuclear Information System (INIS)

Fullwood, F.; Kroeger, P.; Higgins, J.

1993-11-01

Results are presented of a systems failure analysis of the PIUS plant systems that are used during normal reactor operation and postulated accidents. This study was performed to provide the NRC with an understanding of the behavior of the plant. The study applied two diverse failure identification methods, Failure Modes Effects ampersand Criticality Analysis (FMECA) and Hazards ampersand Operability (HAZOP) to the plant systems, supported by several deterministic analyses. Conventional PRA methods were also used along with a scheme for classifying events by initiator frequency and combinations of failures. Principal results of this study are: (a) an extensive listing of potential event sequences, grouped in categories that can be used by the NRC, (b) identification of support systems that are important to safety, and (c) identification of key operator actions

Integrated systems analysis of the PIUS reactor

Energy Technology Data Exchange (ETDEWEB)

Fullwood, F.; Kroeger, P.; Higgins, J. [Brookhaven National Lab., Upton, NY (United States)] [and others

1993-11-01

Results are presented of a systems failure analysis of the PIUS plant systems that are used during normal reactor operation and postulated accidents. This study was performed to provide the NRC with an understanding of the behavior of the plant. The study applied two diverse failure identification methods, Failure Modes Effects & Criticality Analysis (FMECA) and Hazards & Operability (HAZOP) to the plant systems, supported by several deterministic analyses. Conventional PRA methods were also used along with a scheme for classifying events by initiator frequency and combinations of failures. Principal results of this study are: (a) an extensive listing of potential event sequences, grouped in categories that can be used by the NRC, (b) identification of support systems that are important to safety, and (c) identification of key operator actions.

Application of the failure modes and effects analysis technique to the emergency cooling system of an experimental nuclear power plant

International Nuclear Information System (INIS)

Conceicao Junior, Osmar; Silva, Antonio Teixeira e

2009-01-01

This study consists on the application of the failure modes and effects analysis (FMEA), a hazard identification and a risk assessment technique, to the emergency cooling system (ECS), of an experimental nuclear power plant. The choice of this technique was due to its detailed analysis of each component of the system, enabling the identification of all possible ways of failure and its related consequences (in order of importance), allowing the designer to improve the system, maximizing its security and reliability. Through the application of this methodology, it could be observed that the ECS is an intrinsically safe system, in spite of the modifications proposed. (author)

Dynamic Analysis of a Hybrid Energy Storage System (H-ESS Coupled to a Photovoltaic (PV Plant

Directory of Open Access Journals (Sweden)

Linda Barelli

2018-02-01

Full Text Available Nowadays energy storage is strongly needed to allow grid safety and stability due to the wide penetration of renewable plants. Mainly economic and technological issues impede a relevant integration of conventional storage devices in the energy system. In this scenario, the hybridization of different storage technologies can be a techno-economic solution useful to overcome these issues and promote their diffusion. Hybridization allows multi-operation modes of the Energy Storage System (ESS, merging the positive features of base-technologies and extending their application ranges. This paper provides a dynamic analysis of a hybrid energy storage system (H-ESS consisting of a flywheel and a battery pack coupled to a photovoltaic generation plant and a residential load up to 20 kW. A dynamic model of the overall micro-grid (MG was developed implementing the H-ESS preliminary sizing and a suitable management algorithm. The instantaneous behavior of each component was evaluated. A brief summary of the MG performance at different weather and load conditions was provided together with a characterization of the impact of power fluctuations on the battery current and on the power exchange with the grid.

Information system for IAEA inspectors at a centrifuge enrichment plant

International Nuclear Information System (INIS)

Baker, A.L.; Tape, J.W.; Picard, R.R.; Strittmatter, R.B.

1985-01-01

An information system has been developed to aid International Atomic Energy Agency (IAEA) inspectors at the Portsmouth Gas Centrifuge Plant in the US. This system is designed to provide the inspectors with data storage, data analysis, and data evaluation and decision capabilities with minimal impact on the plant operations. The techniques and methodologies developed for this specific case are described with discussion of their general applicability to IAEA inspections at all types of facilities. 7 refs

Computerized operator support system with new man-machine interface for BWR power plants

International Nuclear Information System (INIS)

Monta, K.; Naito, N.; Sugawara, M.; Sato, N.; Mori, N.; Tai, I.; Fukumoto, A.; Tsuchida, M.

1984-01-01

Improvement of the man-machine interface of nuclear power plants is an important contribution to the further enhancement of operational safety. In addition, recent advances in computer technology seem to offer the greatest opportunity to date for achieving improvement in the man-machine interface. The development of a computerized operator support system for BWRs has been undertaken since 1980 with the support of the Japanese Government. The conceptual design of this system is based on the role of the operators. The main functions are standby system management, disturbance analysis and post-trip operational guidance. The objective of the standby system management is to monitor the standby status of the engineered safety feature during normal operation to assure its proper functioning at the onset of emergency situations. The disturbance analysis system detects disturbances in the plant in their early stages and informs the plant operators about, for example, the cause of the disturbances, the plant status and possible propagations. Consequently, operators can take corrective actions to prevent unnecessary plant shutdown. The objective of the post trip operational guide is to support operators in diagnosis and corrective action after a plant trip. Its functions are to monitor the performance of the engineered safety feature, to identify the plant status and to guide the appropriate corrective action to achieve safe plant shutdown. The information from the computerized operator support system is supplied to operators through a colour CRT operator console. The authors have evaluated the performance of various new man-machine interfacing tools and proposed a new operator console design. A prototype system has been developed and verification/validation is proceeding with a BWR plant simulator. (author)

Development of the safety analysis system - SAIS - application to the Kola nuclear power plant

International Nuclear Information System (INIS)

Balfanz, H.P.; Fuhrmann, C.; Neumann, L.; Rumpf, J.; Kubintsev, B.; Marakulin, I.; Shevelev, V.; Terekhov, I.

1995-01-01

The project was started in July 1991 and finished by the end of 1993. It was aimed at adapting SAIS to WWER-440/W213 plant design and operational procedures, demonstrating the possibilities of SAIS for safety evaluation and examining the system by plant personnel and a PSA team. The project covered -the use of data form as well as fault and event tree methods of SAIS, - a probabilistic assessment of the high pressure injection system in case of a small break loss-of-coolant accident, - a quantification of human error probabilities for operator actions to cool down the primary circuit in case of a small break loss-of-coolant accident when the high pressure injection system has failed and - a comparison of Russian and German operational procedures and safety documents used in the probabilistic analyses of the SAIS-Kola project. As a main result SAIS was found to be an appropriate tool to give assistance to the plant personnel on safety evaluation of the plant within the frame of reconstruction measures and on the establishment of a qualified reliability data collection system at Kola NPP. (orig./HP) [de

Industry Application ECCS / LOCA Integrated Cladding/Emergency Core Cooling System Performance: Demonstration of LOTUS-Baseline Coupled Analysis of the South Texas Plant Model

Energy Technology Data Exchange (ETDEWEB)

Zhang, Hongbin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Szilard, Ronaldo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Epiney, Aaron [Idaho National Lab. (INL), Idaho Falls, ID (United States); Parisi, Carlo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Vaghetto, Rodolfo [Texas A & M Univ., College Station, TX (United States); Vanni, Alessandro [Texas A & M Univ., College Station, TX (United States); Neptune, Kaleb [Texas A & M Univ., College Station, TX (United States)

2017-06-01

Under the auspices of the DOE LWRS Program RISMC Industry Application ECCS/LOCA, INL has engaged staff from both South Texas Project (STP) and the Texas A&M University (TAMU) to produce a generic pressurized water reactor (PWR) model including reactor core, clad/fuel design and systems thermal hydraulics based on the South Texas Project (STP) nuclear power plant, a 4-Loop Westinghouse PWR. A RISMC toolkit, named LOCA Toolkit for the U.S. (LOTUS), has been developed for use in this generic PWR plant model to assess safety margins for the proposed NRC 10 CFR 50.46c rule, Emergency Core Cooling System (ECCS) performance during LOCA. This demonstration includes coupled analysis of core design, fuel design, thermalhydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results. Within this context, a multi-physics best estimate plus uncertainty (MPBEPU) methodology framework is proposed.

Quantitative evaluation of physical protection system in nuclear power plant

International Nuclear Information System (INIS)

Sun Yahua; Li Bin; Li Shiju

2009-01-01

Based on the prompt detection analysis, this paper introduced one analysis model of intrusion path in nuclear power plant by means of morphology analysis and developed the evaluation software for path model analysis of physical protection system. Quantitative analysis on three elements (detection, delay, and response) of physical protection system was presented with an imaginary intrusion event example in Mac Arthur nuclear center. The results indicated that the path prompt detection analysis worked effectively to find the weak point of the physical protection system in NPP, and meantime we can also get the high cost-effectiveness improved measures. It is an effective approach to evaluate the overall performance of the system. (authors)

Application of disturbance analysis methodology to safety related transients in the electrical systems of a nuclear power plant. Report UCLA-ENG-8056

International Nuclear Information System (INIS)

Guarro, S.; Okrent, D.

1981-08-01

The present study tries to address the question of whether or not the computerized on-line procedures known under the name of DAS (Disturbance Analysis System) can be usefully and successfully applied to provide timely diagnostics and operational suggestions during the occurrence of a major electrical transient in the auxiliary systems of a nuclear power plant. The perspective of the study is from the plant-safety point of view. A short definition of DAS methodology features and capabilities is presented. A discussion of some of the problems of a general nature that are encountered in DAS safety-oriented applications are also included. The event insufficient power on both emergency buses, with reference to a particular plant dsign (San Onofre 1), is presented. Some transients that have recently occurred in the power supply systems of operating plants are examined. Whether or not a DAS could have successfully dealt with such occurrences is considered

Application of disturbance analysis methodology to safety related transients in the electrical systems of a nuclear power plant. Report UCLA-ENG-8056

Energy Technology Data Exchange (ETDEWEB)

Guarro, S.; Okrent, D.

1981-08-01

The present study tries to address the question of whether or not the computerized on-line procedures known under the name of DAS (Disturbance Analysis System) can be usefully and successfully applied to provide timely diagnostics and operational suggestions during the occurrence of a major electrical transient in the auxiliary systems of a nuclear power plant. The perspective of the study is from the plant-safety point of view. A short definition of DAS methodology features and capabilities is presented. A discussion of some of the problems of a general nature that are encountered in DAS safety-oriented applications are also included. The event insufficient power on both emergency buses, with reference to a particular plant dsign (San Onofre 1), is presented. Some transients that have recently occurred in the power supply systems of operating plants are examined. Whether or not a DAS could have successfully dealt with such occurrences is considered.

Protocol: optimising hydroponic growth systems for nutritional and physiological analysis of Arabidopsis thaliana and other plants

Science.gov (United States)

2013-01-01

Background Hydroponic growth systems are a convenient platform for studying whole plant physiology. However, we found through trialling systems as they are described in the literature that our experiments were frequently confounded by factors that affected plant growth, including algal contamination and hypoxia. We also found the way in which the plants were grown made them poorly amenable to a number of common physiological assays. Results The drivers for the development of this hydroponic system were: 1) the exclusion of light from the growth solution; 2) to simplify the handling of individual plants, and 3) the growth of the plant to allow easy implementation of multiple assays. These aims were all met by the use of pierced lids of black microcentrifuge tubes. Seed was germinated on a lid filled with an agar-containing germination media immersed in the same solution. Following germination, the liquid growth media was exchanged with the experimental solution, and after 14-21 days seedlings were transferred to larger tanks with aerated solution where they remained until experimentation. We provide details of the protocol including composition of the basal growth solution, and separate solutions with altered calcium, magnesium, potassium or sodium supply whilst maintaining the activity of the majority of other ions. We demonstrate the adaptability of this system for: gas exchange measurement on single leaves and whole plants; qRT-PCR to probe the transcriptional response of roots or shoots to altered nutrient composition in the growth solution (we demonstrate this using high and low calcium supply); producing highly competent mesophyll protoplasts; and, accelerating the screening of Arabidopsis transformants. This system is also ideal for manipulating plants for micropipette techniques such as electrophysiology or SiCSA. Conclusions We present an optimised plant hydroponic culture system that can be quickly and cheaply constructed, and produces plants with similar

Effect Analysis of Faults in Digital I and C Systems of Nuclear Power Plants

Energy Technology Data Exchange (ETDEWEB)

Lee, Seung Jun; Jung, Won Dea [KAERI, Dajeon (Korea, Republic of); Kim, Man Cheol [Chung-Ang University, Seoul (Korea, Republic of)

2014-08-15

A reliability analysis of digital instrumentation and control (I and C) systems in nuclear power plants has been introduced as one of the important elements of a probabilistic safety assessment because of the unique characteristics of digital I and C systems. Digital I and C systems have various features distinguishable from those of analog I and C systems such as software and fault-tolerant techniques. In this work, the faults in a digital I and C system were analyzed and a model for representing the effects of the faults was developed. First, the effects of the faults in a system were analyzed using fault injection experiments. A software-implemented fault injection technique in which faults can be injected into the memory was used based on the assumption that all faults in a system are reflected in the faults in the memory. In the experiments, the effect of a fault on the system output was observed. In addition, the success or failure in detecting the fault by fault-tolerant functions included in the system was identified. Second, a fault tree model for representing that a fault is propagated to the system output was developed. With the model, it can be identified how a fault is propagated to the output or why a fault is not detected by fault-tolerant techniques. Based on the analysis results of the proposed method, it is possible to not only evaluate the system reliability but also identify weak points of fault-tolerant techniques by identifying undetected faults. The results can be reflected in the designs to improve the capability of fault-tolerant techniques.

Effect analysis of faults in digital I and C systems of nuclear power plants

International Nuclear Information System (INIS)

Lee, Seung Jun

2014-01-01

A reliability analysis of digital instrumentation and control (I and C) systems in nuclear power plants has been introduced as one of the important elements of a probabilistic safety assessment because of the unique characteristics of digital I and C systems. Digital I and C systems have various features distinguishable from those of analog I and C systems such as software and fault-tolerant techniques. In this work, the faults in a digital I and C system were analyzed and a model for representing the effects of the faults was developed. First, the effects of the faults in a system were analyzed using fault injection experiments. A software-implemented fault injection technique in which faults can be injected into the memory was used based on the assumption that all faults in a system are reflected in the faults in the memory. In the experiments, the effect of a fault on the system output was observed. In addition, the success or failure in detecting the fault by fault-tolerant functions included in the system was identified. Second, a fault tree model for representing that a fault is propagated to the system output was developed. With the model, it can be identified how a fault is propagated to the output or why a fault is not detected by fault-tolerant techniques. Based on the analysis results of the proposed method, it is possible to not only evaluate the system reliability but also identify weak points of fault-tolerant techniques by identifying undetected faults. The results can be reflected in the designs to improve the capability of fault-tolerant techniques. (author)

Plant growth and gas balance in a plant and mushroom cultivation system

Science.gov (United States)

Kitaya, Y.; Tani, A.; Kiyota, M.; Aiga, I.

1994-11-01

In order to obtain basic data for construction of a plant cultivation system incorporating a mushroom cultivation subsystem in the CELSS, plant growth and atmospheric CO2 balance in the system were investigated. The plant growth was promoted by a high level of CO2 which resulted from the respiration of the mushroom mycelium in the system. The atmospheric CO2 concentration inside the system changed significantly due to the slight change in the net photosynthetic rate of plants and/or the respiration rate of the mushroom when the plant cultivation system combined directly with the mushroom cultivation subsystem.

Plant Systems Biology (editorial)

Science.gov (United States)

In June 2003, Plant Physiology published an Arabidopsis special issue devoted to plant systems biology. The intention of Natasha Raikhel and Gloria Coruzzi, the two editors of this first-of-its-kind issue, was ‘‘to help nucleate this new effort within the plant community’’ as they considered that ‘‘...

Application of the methodology of safety probabilistic analysis to the modelling the emergency feedwater system of Juragua nuclear power plant

International Nuclear Information System (INIS)

Troncoso, M.; Oliva, G.

1993-01-01

The application of the methodology developed in the framework of the national plan of safety probabilistic analysis (APS) to the emergency feed water system for the failures of small LOCAS and external electrical supply loss in the nuclear power plant is illustrated in this work. The facilities created by the ARCON code to model the systems and its documentation are also expounded

A critical review on energy, exergy, exergoeconomic and economic (4-E analysis of thermal power plants

Directory of Open Access Journals (Sweden)

Ravinder Kumar

2017-02-01

Full Text Available The growing energy supply, demand has created an interest towards the plant equipment efficiency and the optimization of existing thermal power plants. Also, a thermal power plant dependency on fossil fuel makes it a little bit difficult, because of environmental impacts has been always taken into consideration. At present, most of the power plants are going to be designed by the energetic performance criterion which is based on the first law of thermodynamics. Sometimes, the system energy balance is not sufficient for the possible finding of the system imperfections. Energy losses taking place in a system can be easily determined by using exergy analysis. Hence, it is a powerful tool for the measurement of energy quality, thereby helps to make complex thermodynamic systems more efficient. Nowadays, economic optimization of plant is also a big problem for researchers because of the complex nature. At a viewpoint of this, a comprehensive literature review over the years of energy, exergy, exergoeconomic and economic (4-E analysis and their applications in thermal power plants stimulated by coal, gas, combined cycle and cogeneration system have been done thoroughly. This paper is addressed to those researchers who are doing their research work on 4-E analysis in various thermal power plants. If anyone extracts an idea for the development of the concept of 4-E analysis using this article, we will achieve our goal. This review also indicates the scope of future research in thermal power plants.

TOSHIBA CAE system for nuclear power plant

International Nuclear Information System (INIS)

Machiba, Hiroshi; Sasaki, Norio

1990-01-01

TOSHIBA aims to secure safety, increase reliability and improve efficiency through the engineering for nuclear power plant using Computer Aided Engineering (CAE). TOSHIBA CAE system for nuclear power plant consists of numbers of sub-systems which had been integrated centering around the Nuclear Power Plant Engineering Data Base (PDBMS) and covers all stage of engineering for nuclear power plant from project management, design, manufacturing, construction to operating plant service and preventive maintenance as it were 'Plant Life-Cycle CAE System'. In recent years, TOSHIBA has been devoting to extend the system for integrated intelligent CAE system with state-of-the-art computer technologies such as computer graphics and artificial intelligence. This paper shows the outline of CAE system for nuclear power plant in TOSHIBA. (author)

Study on aging management of fire protection system in nuclear power plant

International Nuclear Information System (INIS)

Fang Huasong; Du Yu; Li Jianwen; Shi Haining; Tu Fengsheng

2010-01-01

Fire prevention, fire fighting and fire automatic alarms are three aspects which be included in fire protection system in nuclear power plants. The fire protection system can protect personnel, equipment etc in the fire, so their performance will have a direct influence on the safe operation in nuclear power plants. The disabled accidents caused by aging have happened continuously with the extension of time in the fire protection system, which is the major security risk during the running time in nuclear power plants. In view of the importance of fire protection system and the severity of aging problems, the aging are highly valued by the plant operators and related organizations. Though the feedback of operating experience in nuclear power plant, the impact of the fire-fighting equipment aging on system performance and reliability be assessed, the aging sensitive equipment be selected to carry out the aging analysis and to guide the management and maintenance to guarantee the healthy operation in life time of fire protection system in nuclear power plant. (authors)

Development of basic system for sensor calibration support in nuclear power plants

International Nuclear Information System (INIS)

Kusumi, Naohiro; Ohga, Yukiharu; Fukuda, Mitsuko; Ishizaki, Yuuichi; Koyama, Mikio; Maeda, Akihiko

2004-01-01

It is strongly desirable to reduce maintenance costs and shorten the time of periodic inspections in nuclear power plants. Therefore, it is important to reduce the amount of maintenance work during the inspection. In Japan, sensor calibration is usually performed at every periodic inspection, and the sensor calibration requires a large amount of work. A system for sensor calibration support has been developed to reduce sensor calibration work. The system is composed of two subsystems: a statistical analysis subsystem and a drift detection subsystem, as well as a human-machine interface, which offers support information. The statistical analysis subsystem supports the decision of the sensor calibration intervals based on the statistical analysis of sensor calibration data. There is the possibility that sensor drift increases beyond an allowance value before the sensor calibration intervals determined by the statistical analysis subsystem because of malfunctions, etc. To cope with this, the drift detection subsystem detects the sensor drift online during the plant operation. By combining the statistical analysis subsystem and the drift detection subsystem, a reliable sensor calibration support system is realized. The basic system composed of two subsystems was developed and evaluated using real plant data. The results showed that the sensor calibration intervals can be extended beyond current intervals and that the system is capable of detecting the sensor drift online. (author)

The Development of a Fault Tree Model for Balance of Plant System

International Nuclear Information System (INIS)

Hwang, Mee Jeong; Park, Jin Hee; Lim, Ho Gon

2011-01-01

In this paper, we propose a fault tree modeling method for BOP (balance of plant) system to develop a combined risk model and trip model, and the application plans of the developed model. Where, the trip means the reactor trip and turbine and generator trip. We have usually modeled the safety-related systems and their supporting systems to assess the risk analysis of a nuclear power plant. However, the BOP system.s condition change induces the risk change. That is, the BOP system.s condition is relevant to plants. performance and affects to the plant risk. The existing model for BOP systems is a simplified system model or SPV (Single-point vulnerability) evaluation model. However, these models are not effective enough to use for the plant's performance evaluation. Also, lately an integrated decision-making framework is required for risk-informed applications. The methods for monitoring the performance of a nuclear power plant differ from the purpose. For example, MSPI (mitigating system performance index) and MR (maintenance rule) use different methods and indexes to monitor the performance. Therefore, for consistent decision-making, it is necessary to develop a risk assessment model including a systems model inducing reactor trip. The system.s model inducing reactor trip and turbine/generator trip is defined as the 'trip model'

Plant experience with check valves in passive systems

Energy Technology Data Exchange (ETDEWEB)

Pahladsingh, R R [GKN Joint Nuclear Power Plant, Dodewaard (Netherlands)

1996-12-01

In the design of the advanced nuclear reactors there is a tendency to introduce more passive safety systems. The 25 year old design of the GKN nuclear reactor is different from the present BWR reactors because of some special features, such as the Natural Circulation - and the Passive Isolation Condenser system. When reviewing the design, one can conclude that the plant has 25 years of experience with check valves in passive systems and as passive components in systems. The result of this experience has been modeled in a plant-specific ``living PSA`` for the plant. A data-analysis has been performed on components which are related to the safety systems in the plant. As part of this study also the check valves have been taken in consideration. At GKN, the check valves have shown to be reliable components in the systems and no catastrophic failures have been experienced during the 25 years of operation. Especially the Isolation Condenser with its operation experience can contribute substantially to the insight of check valves in stand-by position at reactor pressure and operating by gravity under different pressure conditions. With the introduction of several passive systems in the SBWR-600 design, such as the Isolation Condensers, Gravity Driven Cooling, and Suppression Pool Cooling System, the issue of reliability of check valves in these systems is actual. Some critical aspects for study in connection with check valves are: What is the reliability of a check valve in a system at reactor pressure, to open on demand; what is the reliability of a check valve in a system at low pressure (gravity), to open on demand; what is the reliability of a check valve to open/close when the stand-by check wave is at zero differential pressure. The plant experience with check valves in a few essential safety systems is described and a brief introduction will be made about the application of check valves in the design of the new generation reactors is given. (author). 6 figs, 1 tab.

Design study of plant system for the fusion experimental reactor (FER)

International Nuclear Information System (INIS)

Iida, Hiromasa; Kuroda, Hideo; Yamada, Masao; Suzuki, Tatsushi; Honda, Tsutomu; Ohmura, Hiroshi; Itoh, Shinichi.

1986-11-01

This report describes design study results of the FER plant system. The purpose of this study is to have an image of the FER plant system as a whole by designing major auxiliary systems, reactor building and maintenance and radwaste desposal systems. The major auxiliary systems include tritium, cooling, evacuation and fueling systems. For these each systems, flowdiagrams are studied and designs of devices and pipings are conducted. In the reactor building design, layout of the above auxiliary systems in the building is studied with careful zoning concept by the radiation level. Structural integrity of the reactor building is also studied including seismic analysis. In the design of the maintenance and radwaste system flowdiagram of failed reactor components is developed and transfer vehicles and buildings are designed. Finally assuming JAERI Naka site as the reactor site layout of the whole FER plant system is developed. (author)

Structural and stress analysis of nuclear piping systems

International Nuclear Information System (INIS)

Hata, Hiromichi

1982-01-01

The design of the strength of piping system is important in plant design, and its outline on the example of PWRs is reported. The standards and guides concerning the design of the strength of piping system are shown. The design condition for the strength of piping system is determined by considering the requirements in the normal operation of plants and for the safety design of plants, and the loads in normal operation, testing, credible accident and natural environment are explained. The methods of analysis for piping system are related to the transient phenomena of fluid, piping structure and local heat conduction, and linear static analysis, linear time response analysis, nonlinear time response analysis, thermal stress analysis and fluid transient phenomenon analysis are carried out. In the aseismatic design of piping system, it is desirable to avoid the vibration together with a building supporting it, and as a rule, to make it into rigid structure. The piping system is classified into high temperature and low temperature pipings. The formulas for calculating stress and the allowable condition, the points to which attention must be paid in the design of piping strength and the matters to be investigated hereafter are described. (Kako, I.)

A plant resource and experiment management system based on the Golm Plant Database as a basic tool for omics research

Directory of Open Access Journals (Sweden)

Selbig Joachim

2008-05-01

Full Text Available Abstract Background For omics experiments, detailed characterisation of experimental material with respect to its genetic features, its cultivation history and its treatment history is a requirement for analyses by bioinformatics tools and for publication needs. Furthermore, meta-analysis of several experiments in systems biology based approaches make it necessary to store this information in a standardised manner, preferentially in relational databases. In the Golm Plant Database System, we devised a data management system based on a classical Laboratory Information Management System combined with web-based user interfaces for data entry and retrieval to collect this information in an academic environment. Results The database system contains modules representing the genetic features of the germplasm, the experimental conditions and the sampling details. In the germplasm module, genetically identical lines of biological material are generated by defined workflows, starting with the import workflow, followed by further workflows like genetic modification (transformation, vegetative or sexual reproduction. The latter workflows link lines and thus create pedigrees. For experiments, plant objects are generated from plant lines and united in so-called cultures, to which the cultivation conditions are linked. Materials and methods for each cultivation step are stored in a separate ACCESS database of the plant cultivation unit. For all cultures and thus every plant object, each cultivation site and the culture's arrival time at a site are logged by a barcode-scanner based system. Thus, for each plant object, all site-related parameters, e.g. automatically logged climate data, are available. These life history data and genetic information for the plant objects are linked to analytical results by the sampling module, which links sample components to plant object identifiers. This workflow uses controlled vocabulary for organs and treatments. Unique

System implementation of hazard analysis and critical control points (HACCP) in a nitrogen production plant

International Nuclear Information System (INIS)

Barrantes Salazar, Alexandra

2014-01-01

System of hazard analysis and critical control points are deployed in a production plant of liquid nitrogen. The fact that the nitrogen has become a complement to food packaging to increase shelf life, or provide a surface that protect it from manipulation, has been the main objective. Analysis of critical control points for the nitrogen production plant has been the adapted methodology. The knowledge of both the standard and the production process, as well as the on site verification process, have been necessary. In addition, all materials and/or processing units that are found in contact with the raw material or the product under study were evaluated. Such a way that the intrinsic risks of each were detected, from the physical, chemical and biological points of view according to the origin or pollution source. For each found risk was evaluated the probability of occurrence according to the frequency and gravity of it, with these variables determined was achieved the definition of the type of risk detected. In the cases that was presented a greater risk or critical, these were subjected decision tree; with which is concluded the non determination of critical control points. However, for each one of them were established the maximum permitted limits. To generate each of the results it has literature or scientific reference of reliable provenance, where is indicated properly the support of the evaluated matter. In a general way, the material matrix and the process matrix are found without critical control points; so that the project is concluded in the analysis, and it has to generate without the monitoring system and verification. To increase this project is suggested in order to cover the packaging system of gaseous nitrogen, due to it was delimited to liquid nitrogen. Furthermore, the liquid nitrogen is a 100% automated and closed process so the introduction of contaminants is very reduced, unlike the gaseous nitrogen process. (author) [es

Digital Processor Module Reliability Analysis of Nuclear Power Plant

International Nuclear Information System (INIS)

Lee, Sang Yong; Jung, Jae Hyun; Kim, Jae Ho; Kim, Sung Hun

2005-01-01

The system used in plant, military equipment, satellite, etc. consists of many electronic parts as control module, which requires relatively high reliability than other commercial electronic products. Specially, Nuclear power plant related to the radiation safety requires high safety and reliability, so most parts apply to Military-Standard level. Reliability prediction method provides the rational basis of system designs and also provides the safety significance of system operations. Thus various reliability prediction tools have been developed in recent decades, among of them, the MI-HDBK-217 method has been widely used as a powerful tool for the prediction. In this work, It is explained that reliability analysis work for Digital Processor Module (DPM, control module of SMART) is performed by Parts Stress Method based on MIL-HDBK-217F NOTICE2. We are using the Relex 7.6 of Relex software corporation, because reliability analysis process requires enormous part libraries and data for failure rate calculation

Experience on a BWR plant diagnosis system

International Nuclear Information System (INIS)

Tanabe, A.; Kawai, K.; Hashimoto, Y.

1981-01-01

It is important to watch plant dynamics and equipment condition for avoiding a big transient or avoiding damage to a system by equipment failure. After the TMI accident the necessity of a diagnosis system has been recognized and such development activities have become of primary importance in many organizations. A diagnosis system has two kinds of function. One is the early detection of an anomaly before detection by a conventional instrumentation system. The other is appropriate instruction after alarm or scram has occurred. The authors have been developing the former system by a noise analysis technique and a feasibility study has been undertaken in recent years as a joint research programme of several electric power companies and the Toshiba Corporation. A prototype diagnosis system has been installed on a BWR plant in Japan. This diagnosis system concerns reactor core, jet pumps and three main control systems. Many data from normal operation have been accumulated using this system and a variation pattern of normal noise data is clarified. On this basis, anomally detection criteria have been determined using statistical decision theory. It is confirmed that this system performance is satisfactory, and that the system will be of great use for surveillance of core and control systems without artificial disturbances. (author)

Building of a CAD system for instrumentation and control system of nuclear power plant

International Nuclear Information System (INIS)

Ma Zhicai; Hu Chunping; Zhang Dongsheng

2012-01-01

Base on the analysis of deign documents and process, a database for instrumentation and control system design can be developed with a popular desktop relational database management system (RDBMS). With the RDBMS, an instrumentation and control system CAD system can be built unitizing database link feature of popular CAD software, with the function of management of design data, output of list and forms. and design of drawings. A CAD system of this kind has been used in the design practice of nuclear power plant. With this system, it is shown that, the consistency of information has been controlled and the load on the engineer has been significantly reduced. The methodology used here can also be used in the CAD system for CAP1000 and CAP1400 plant. (authors) series

Design of an operator support system for online maintenance at nuclear power plant

International Nuclear Information System (INIS)

Chu Yongyue; Li Huwei; Gao Qiang; Yi Yan; Yang Ming

2013-01-01

Online maintenance based on reliability centered management is pivotal for the safe and economical operation of Nuclear Power Plant (NPP). This paper presents an operator support system through which the operators can effectively manage plant configuration and identify the weaknesses in plant operation. The proposed operator support system is based on the GO-FLOW, which is a success-oriented availability analysis methodology and can be used for evaluating phased missions. In this paper, the design of the proposed operator support system is introduced through a case study of the Auxiliary Feed Water System (AFWS). (author)

Development of a computerized operator support system for BWR power plant

International Nuclear Information System (INIS)

Monta, K.; Sekimizu, K.; Sato, N.; Araki, T.; Mori, N.

1985-01-01

A computerized operator support system for BWR power plant has been developed since 1980 supported by the Japanese government. The main functions of the systems are post trip operational guidance, disturbance analysis, standby system management, operational margin monitoring and control rod operational guidance. The former two functions aim at protection against incidents during operation of nuclear power plants and the latter three functions aim at their prevention. As the final stage of the development, these functions are combined with the plant supervision function and are organized as an advanced man-machine interface for BWR power plant. During the above process, operator task analyses are performed to enable synthesis of these support functions for right fit to operator tasks and to realize a hierarchical structure for CRT displays for right fit to operators cognitive needs. (author)

Development of support system for nuclear power plant piping

International Nuclear Information System (INIS)

Horino, Satoshi

1987-01-01

Ishikawajima-Harima Heavy Industries Co., Ltd. has advanced the development of Integrated Nuclear Plant Piping System (INUPPS) for nuclear power plants since 1980, and continued its improvement up to now. This time as its component, a piping support system (PISUP) has been developed. The piping support system deals with the structures such as piping supports and the stands for maintenance and inspection, and as for standard supporting structures, it builds up automatically the structures including the selection of optimum members by utilizing the standard patterns in cooperation with the piping design system including piping stress analysis. As for the supporting structures deviating from the standard, by amending a part of the standard patterns in dialogue from, structures can be built up. By using the data produced in this way, this system draws up consistently a design book, production management data and so on. From the viewpoint of safety, particular consideration is given to the aseismatic capability of nuclear power plants, and piping is fundamentally designed regidly to avoid resonance. It is necessary to make piping supports so as to have sufficient strength and rigidity. The features of the design of piping supports for nuclear power plant, the basic concept of piping support system, the constitution of the software and hardware, the standard patterns and the structural patterns of piping support system and so on are described. (Kako, I.)

Analysis of the options - rationale for servomanipulator maintenance in future reprocessing plants

International Nuclear Information System (INIS)

Herndon, J.N.; Hamel, W.R.

1984-04-01

The currently available maintenance systems which can be applied, in various combinations, to large-volume in-cell maintenance operations are: (1) contact, (2) overhead cranes, (3) power manipulators, (4) mechanical master/slave manipulators, and (5) servomanipulators. The requirements for reprocessing plant maintenance are reviewed, and the capabilities of remote maintenance systems are described. A basic trade-off analysis of these remote maintenance systems considering reprocessing plant requirements is given. Justification is given for selecting the overhead crane/servomanipulator-based maintenance concept as the option most desirable for future large reprocessing plant maintenance. 23 references, 6 tables

Energetic-economic dynamic computational analysis of plants with small capacity - Gera model

International Nuclear Information System (INIS)

Storfer, A.F.; Demanboro, A.C. de; Campello, C.A.G.B.

1990-01-01

A methodology and mathematical model for energy and economic analysis of hydroelectric power plants with low and medium capacity are presented. This methodology will be used for isolated or integrated hydroelectric power plants, including plants that part of their produced energy will go to the site market and part will go to regional electric system. (author)

Exergetic and environmental analysis of a pulverized coal power plant

International Nuclear Information System (INIS)

Restrepo, Álvaro; Miyake, Raphael; Kleveston, Fábio; Bazzo, Edson

2012-01-01

This paper presents the results of exergetic and environmental analysis of a typical pulverized coal power plant located in Brazil. The goal was to quantify both the exergy destruction and the environmental impact associated with a thermal power plant. The problem boundary consists of the entire coal delivery route, including mining and beneficiation, transport, pre-burning processes and the power plant. The used data were obtained mainly from field measurements taken in all system processes, from mining to the power plant. The study focused only on the operation period. Previous works have shown that the construction and decommissioning periods contribute less than 1% of the environmental impact. The exergetic analysis was based on the second law of thermodynamics while the environmental analysis was based on life cycle assessment (LCA) using SimaPro 7.2, focussing on the climate change and acidification impact categories. The CO 2 -eq emission was 1300 kg per MWh. The highest degree of environmental impact occurred during the combustion process. The exergetic and environmental analysis provides a tool to evaluate irreversibilities and the environmental impact, identifying the most significant stages and equipment of the entire power generation process. -- Highlights: ► Exergetic and environmental analysis of a typical Brazilian PC power plant. ► Environmental impact associated with the mining, transport and thermal power plant. ► Life cycle assessment used for environmental analysis. ► Acidification impact category evaluated using Eco-indicator 99. ► Climate change impact evaluation using (Global Warming Potential) GWP 100a.

Development of testing system for the thermo-mechanical fatigue crack analysis of nuclear power plant pipes

International Nuclear Information System (INIS)

Lee, Ho Jin; Kim, Maan Won; Lee, Bong Sang

2003-12-01

Fatigue crack growth analysis plays an important role in the structural integrity assessment or the service life calculation of the nuclear power plant pipes. To obtain the material properties as a basic data to achieve an accurate crack growth analysis, a lot of tests and numerical crack growth simulations have been done for decades. The BS 7910 or the ASME Boiler and Pressure Vessel Code Section XI, generally used to evaluate crack growth behavior, were made under the based on simple stress states or at the evaluated isothermal temperature. It is well known that the ASME code could sometimes give so conservative results in some cases of which the cracked components are experiencing with cyclic thermal shock. In this report, we suggested a method for the life assessment of a crack embedded in nuclear power plant pipes under the thermal-mechanical fatigue loads. We here use the numerical method to get the temperature history for thermal- mechanical fatigue crack growth test. And then we can calculate the remaining life time of the pipe by using the fracture mechanics and the test results together. For this purpose, we constructed a thermal-mechanical fatigue crack growth testing system. We also gave a lot of review about recent researches in the experimental field of thermal-mechanical fatigue analysis

EXERGETIC ANALYSIS OF A COGENERATION POWER PLANT

Directory of Open Access Journals (Sweden)

Osvaldo Manuel Nuñez Bosch

2016-07-01

Full Text Available Cogeneration power plants connected to industrial processes have a direct impact on the overall efficiency of the plant and therefore on the economic results. Any modification to the thermal outline of these plants must first include an exergetic analysis to compare the benefits it can bring the new proposal. This research is performed to a cogeneration plant in operation with an installed electrical capacity of 24 MW and process heat demand of 190 MW, it shows a study made from the Second Law of Thermodynamics. Exergetic evaluation of each component of the plant was applied and similarly modified cogeneration scheme was evaluated. The results illustrate that the exergy losses and irreversibilities are completely different from one subsystem to another. In general, the total exergy destruction represented 70,7% from the primary fuel exergy. Steam generator was the subsystem with the highest irreversibility of the plant with 54%. It was demonstrated that the increase of the steam parameters lead to reduce exergy destruction and exergy efficiency elevation. The suppression of the reduction system and the adding of an extraction-condensing steam turbine produce the same effect and contribute to drop off the electrical consumption from the grid.

Implementation of New Reactivity Measurement System and New Reactor Noise Analysis Equipment in a VVER-440 Nuclear Power Plant

Science.gov (United States)

Vegh, János; Kiss, Sándor; Lipcsei, Sándor; Horvath, Csaba; Pos, István; Kiss, Gábor

2010-10-01

The paper deals with two recently developed, high-precision nuclear measurement systems installed at the VVER-440 units of the Hungarian Paks NPP. Both developments were motivated by the reactor power increase to 108%, and by the planned plant service time extension. The first part describes the RMR start-up reactivity measurement system with advanced services. High-precision picoampere meters were installed at each reactor unit and measured ionization chamber current signals are handled by a portable computer providing data acquisition and online reactivity calculation service. Detailed offline evaluation and analysis of reactor start-up measurements can be performed on the portable unit, too. The second part of the paper describes a new reactor noise diagnostics system using state-of-the-art data acquisition hardware and signal processing methods. Details of the new reactor noise measurement evaluation software are also outlined. Noise diagnostics at Paks NPP is a standard tool for core anomaly detection and for long-term noise trend monitoring. Regular application of these systems is illustrated by real plant data, e.g., results of standard reactivity measurements during a reactor startup session are given. Noise applications are also illustrated by real plant measurements; results of core anomaly detection are presented.

A technical analysis for cogeneration systems with potential applications in twelve California industrial plants. [energy saving heat-electricity utility systems

Science.gov (United States)

Moretti, V. C.; Davis, H. S.; Slonski, M. L.

1978-01-01

In a study sponsored by the State of California Energy Resources Conservation and Development Commission, 12 industrial plants in five utility districts were surveyed to assess the potential applications of the cogeneration of heat and electricity in California industry. Thermodynamic calculations were made for each plant in determining the energy required to meet the existing electrical and steam demands. The present systems were then compared to conceptual cogeneration systems specified for each plant. Overall energy savings were determined for the cogeneration applications. Steam and gas turbine topping cycle systems were considered as well as bottoming cycle systems. Types of industries studied were: pulp and paper, timber, cement, petroleum refining, enhanced oil recovery, foods processing, steel and glass

Developments in data acquisition systems with LabView datalogging and supervisory control module for tritium removal plant, with data base and process analysis

International Nuclear Information System (INIS)

Moraru, Carmen Maria; Stefan, Iuliana; Balteanu, Ovidiu; Stefan, Liviu; Bucur, Ciprian; Hartescu, Florin

2006-01-01

Full text: The implementation of the new trends for tritium processing nuclear plants, and especially those with an experimental character or of new technology development, shows a very high complexity due to issues raised by the integration of a high diversity of instrumentation and equipment into a unitary control system of the technological process. Keeping the system's flexibility is a demand of the experimental plants for which the change of configuration, process and parameters is something usual. The big amount of data that needs to be monitored, stored and accessed for ulterior analyses demands the achievement of an information network where the data acquiring, control and analysis systems of the technological process can be integrated with a data base system. Thus, integrated computing and control systems needed for the control of the technological process will be executed, to be continued with the execution of failure protection system, by choosing methods corresponding to the technological processes within the tritium processing nuclear plants. (authors)

Nexant Parabolic Trough Solar Power Plant Systems Analysis; Task 1: Preferred Plant Size, 20 January 2005 - 31 December 2005

Energy Technology Data Exchange (ETDEWEB)

Kelly, B.

2006-07-01

The Rankine cycles for commercial parabolic trough solar projects range in capacity from 13.5 MWe at the Solar Electric Generating Station I (SEGS I) plant, to a maximum of 89 MWe at the SEGS VIII/IX plants. The series of SEGS projects showed a consistent reduction in the levelized energy cost due to a combination of improvements in collector field technology and economies of scale in both the Rankine cycle and the operation and maintenance costs. Nonetheless, the question of the optimum Rankine cycle capacity remains an open issue. The capacities of the SEGS VIII/IX plants were limited by Federal Energy Regulatory Commission and Public Utility Regulatory Policy Act requirements to a maximum net output of 80 MWe. Further improvements in the Rankine cycle efficiency, and economies of scale in both the capital and the operating cost, should be available at larger plant sizes. An analysis was conducted to determine the effect of Rankine cycle capacities greater than 80 MWe on the levelized energy cost. The study was conducted through the following steps: (1) Three gross cycle capacities of 88 MWe, 165 MWe, and 220 MWe were selected. (2) Three Rankine cycle models were developed using the GateCycle program. The models were based on single reheat turbine cycles, with main steam conditions of 1,450 lb{sub f}/in{sup 2} and 703 F, and reheat steam conditions of 239 lb{sub f}/in{sup 2} and 703 F. The feedwater heater system consisted of 5 closed heaters and 1 open deaerating heater. The design condenser pressure was 2.5 in. HgA. (3) The optimization function within Excelergy was used to determine the preferred solar multiple for each plant. Two cases were considered for each plant: (a) a solar-only project without thermal storage, and (b) a solar-fossil hybrid project, with 3 hours of thermal storage and a heat transport fluid heater fired by natural gas. (4) For each of the 6 cases, collector field geometries, heat transport fluid pressure losses, and heat transport pump

Seismic margin reviews of nuclear power plants: Identification of important functions and systems

International Nuclear Information System (INIS)

Prassinos, P.G.; Moore, D.L.; Amico, P.J.

1987-01-01

The results from the review of the seven utility-sponsored seismic PRAs plus the Zion SSMRP have been used to develop some insights regarding the importance of various systems and functions to seismic margins. By taking this information and combining it with the fragility insights we can develop some functional/systemic screening guideline for margin studies. This screening approach will greatly reduce the scope of the analysis. It is possible only to come to conclusions regarding the importance of plant systems and safety functions for PWRs, for which six plants were studied. For PWRs, it is possible to categorize plant safety functions as belonging to one of two groups, one of which is important to the assessment of seismic margins and one of which is not. The important functional group involves only two functions that must be considered for estimating seismic margin. These two functions are shutting down the nuclear reaction and providing cooling to the reactor core in the time period immediately following the seismic event (that is, the injection phase or pre-residual heat removal time period). It is possible to reasonably estimate the seismic margin of the plant by performing a study only involving the analysis of the plant systems and structure which are required in order to perform the two functions. Such analysis must include an assessment of a complete set of seismic initiating events. (orig./HP)

Design of equipment management information system for nuclear power plant

International Nuclear Information System (INIS)

Wang Chengyuan

1996-01-01

The author describes the ideas and practical method for need analysis, system function dividing, code design, program design and network disposition of equipment purchase management system of nuclear power plant during building, from the view of engineering investment control, schedule control and quality control

Insights from a reliability review of digital plant protection system

International Nuclear Information System (INIS)

Kim, I.S.; Hwang, S.W.; Kim, B.S.; Jeong, C.H.; Oh, S.H.

2001-01-01

The full text follows: As part of the design efforts for Ulchin nuclear power plant units 5 and 6 of Korea, a reliability analysis of digital plant protection system (DPPS) was performed by ABB-CE. An independent review of the DPPS reliability analysis was undertaken by Hanyang University to assist Korea Institute of Nuclear Safety (KINS), the nuclear regulatory body of Korea, in evaluating the design acceptability of the digital system. The DPPS is designed to encompass both reactor trip function and ESFAS (engineered safety feature actuation system) initiation function. The major methods used by the ABB-CE to assess the Ulchin 5-6 DPPS reliability are failure mode and effect analysis (FMEA) and fault tree analysis. Hence, our independent review was conducted focusing on: -) the establishment of review criteria based on various sources, such as the standard review plan of KINS, 10CFR50 Appendix A, IEEE standards 279, 577, and 603; -) the suitability of the FMEA and fault tree analysis for the Ulchin 5-6 DPPS, including the specific methods used (e.g., for human reliability analysis and common-cause failure analysis), the assumptions made (e.g., with respect to test frequency and watchdog timer coverage), and the data employed (e.g., CCF parameter, human error probability, and processor failure rate); and -) the design acceptability of the DPPS especially as compared to the analog plant protection system from a reliability and safety perspective. The paper shall also discuss key issues requiring further in-depth investigation, such as reliability of programmable logic controllers (PLCs), coverage factor of watchdog timers, and susceptibility of redundant digital units to common cause failure. Sensitivity analyses were carried out to investigate the impact of several parameters of special interest, like the coverage factor of watchdog timer and human error probability (e.g. an operator error to manually trip the reactor, or to mis-calibrate the trip parameters) on

ERP-IV-A program for transient thermal-hydraulic analysis of PWR plant

International Nuclear Information System (INIS)

Dai Anguo; Tang Jiahuan; Qian Huifu; Gao Zhikang

1987-12-01

The author deal with the descriptions of physical model of transient process in PWR plant and the function of ERP-IV (ERR-IV Transient Thermo-Hydraulic Analysis Code). The code has been developed for safety analysis and design transient. The code is characterized by the multi-loop long-term, short term, wide-range plant simulation with the capability to analyze natural circulation condition. The description of ERP-IV includes following parts: reactor, primary coolant loops, pressurizer, steam generators, main steam system, turbine, feedwater system, steam dump, relive valves, and safety valves in secondary side, etc.. The code can use for accident analysis, such as loss of all A.C. power to power plant auxiliaries (a station blackout), loss of normal feedwater, loss of load, loss of condenser vacuum and other events causing a turbine trip, complete loss of forced reactor coolant flow, uncontrolled rod cluster control assembly bank withdrawal. It can also be used for accident analysis of the emergency and limiting conditions, such as feedwater line break and main steam line rupture. It can also be utilized as a tool for system design studies, component design, setpoint studies and design transition studies, etc

Reliability analysis techniques in power plant design

International Nuclear Information System (INIS)

Chang, N.E.

1981-01-01

An overview of reliability analysis techniques is presented as applied to power plant design. The key terms, power plant performance, reliability, availability and maintainability are defined. Reliability modeling, methods of analysis and component reliability data are briefly reviewed. Application of reliability analysis techniques from a design engineering approach to improving power plant productivity is discussed. (author)

Functional analysis for complex systems of nuclear fusion plant

International Nuclear Information System (INIS)

Pinna, Tonio; Dongiovanni, Danilo Nicola; Iannone, Francesco

2016-01-01

Highlights: • Functional analysis for complex systems. • Functional Flow Block Diagrams (FFBD). • IDEFØ diagrams. • Petri Net algorithm - Abstract: In system engineering context, a functional analysis is the systematic process of identifying, describing and correlating the functions a system must perform in order to be successful at any foreseen life-cycle phase or operational state/mode. By focusing on what the system must do disregarding the implementation, the functional analysis supports an unbiased system requirement allocation analysis. The system function architecture is defined in terms of process, protection (interlock) or nuclear safety functions. Then, the system functions are analyzed from several points of view in order to highlight the various pieces of information defining the way the system is designed to accomplish its mission as defined in the system requirement documents. The process functional flow is identified and represented by Functional Flow Block Diagrams (FFBD) while the system function interfaces are identified and represented by IDEFØ diagrams. Function interfaces are defined as relationships across identified functions in terms of function input (from other functions or requirements), output (added value or outcome of the function), controls (from other functions or systems) and mechanisms necessary to fulfill the function. The function architecture is further detailed by considering for each function: a) the phase of application, b) the actions performed c) the controlled variable and control actions to be foreseen in the implementation of the functions, d) the system involved in the control action, e) the equipment involved in the function, f) the requirements allocated to the function. The methodology here presented are suggested for the designing of fusion facilities and reactors already from the first phases of the pre-conceptual design, as it is now for DEMO.

A New Digital Imaging and Analysis System for Plant and Ecosystem Phenological Studies

Science.gov (United States)

Ramirez, G.; Ramirez, G. A.; Vargas, S. A., Jr.; Luna, N. R.; Tweedie, C. E.

2015-12-01

Over the past decade, environmental scientists have increasingly used low-cost sensors and custom software to gather and analyze environmental data. Included in this trend has been the use of imagery from field-mounted static digital cameras. Published literature has highlighted the challenge scientists have encountered with poor and problematic camera performance and power consumption, limited data download and wireless communication options, general ruggedness of off the shelf camera solutions, and time consuming and hard-to-reproduce digital image analysis options. Data loggers and sensors are typically limited to data storage in situ (requiring manual downloading) and/or expensive data streaming options. Here we highlight the features and functionality of a newly invented camera/data logger system and coupled image analysis software suited to plant and ecosystem phenological studies (patent pending). The camera has resulted from several years of development and prototype testing supported by several grants funded by the US NSF. These inventions have several unique features and functionality and have been field tested in desert, arctic, and tropical rainforest ecosystems. The system can be used to acquire imagery/data from static and mobile platforms. Data is collected, preprocessed, and streamed to the cloud without the need of an external computer and can run for extended time periods. The camera module is capable of acquiring RGB, IR, and thermal (LWIR) data and storing it in a variety of formats including RAW. The system is full customizable with a wide variety of passive and smart sensors. The camera can be triggered by state conditions detected by sensors and/or select time intervals. The device includes USB, Wi-Fi, Bluetooth, serial, GSM, Ethernet, and Iridium connections and can be connected to commercial cloud servers such as Dropbox. The complementary image analysis software is compatible with all popular operating systems. Imagery can be viewed and

Development and demonstration of near-real-time accounting systems for reprocessing plants

International Nuclear Information System (INIS)

Cobb, D.D.; Hakkila, E.A.; Dayem, H.A.; Shipley, J.P.; Baker, A.L.

1981-01-01

A program to develop and demonstrate near-real-time accounting systems for reprocessing plants has been active at Los Alamos since 1976. The technology has been developed through modeling and simulation of process operation and measurement systems and evaluation of these data using decision analysis techniques. Aspects of near-real-time systems have been demonstrated successfully at the AGNS reprocessng plant as part of a joint study of near-real-time accounting

Application of expert system to nuclear power plant operation and guidance system

International Nuclear Information System (INIS)

Goto, M.; Takada, Y.

1990-01-01

For a nuclear power plant, it is important that an expert system supplies useful information to the operator to meet the increasing demand for high-level plant operation. It is difficult to build a user-friendly expert system that supplies useful information in real time using existing general-purpose expert system shells. Therefore a domain-specific expert system shell with a useful knowledge representation for problem-solving in nuclear power plant operation was selected. The Plant Table (P/T) representation format was developed for description of a production system for nuclear power plant operation knowledge. The P/T consists of plant condition representation designed to process multiple inputs and single output. A large number of operation inputs for several plant conditions are divided into 'timing conditions', 'preconditions' and 'completion conditions' to facilitate knowledge-base build-up. An expert system for a Nuclear Power Plant Operation and Guidance System utilizing the P/T was developed to assist automatic plant operation and surveillance test operation. In these systems, automatic plant operation signals to the plant equipment and operation guidance messages to the operators are both output based on the processing and assessment of plant operation conditions by the P/T. A surveillance test procedure guide for major safety-related systems, such as those for emergency core cooling systems, is displayed on a CRT (Cathode Ray Tube) and test results are printed out. The expert system for a Nuclear Power Plant Operation and Guidance System has already been successfully applied to Japanese BWR plants

Nuclear power plant safe operation principles and some topics concerning systems reliability analysis

International Nuclear Information System (INIS)

Borsky, M.; Kreim, R.; Stanicek, J.

1997-01-01

General safety criteria are specified, and nuclear power plant equipment is classified into systems either important or unimportant for nuclear safety. The former class is subdivided into safety systems and safety related systems. The safety requirements concern earthquakes, storms, fires, floods, man-induced events, and equipment failures. The actual state of systems important for safety is described. (M.D.)

Chapter 15. Plant pathology and managing wildland plant disease systems

Science.gov (United States)

David L. Nelson

2004-01-01

Obtaining specific, reliable knowledge on plant diseases is essential in wildland shrub resource management. However, plant disease is one of the most neglected areas of wildland resources experimental research. This section is a discussion of plant pathology and how to use it in managing plant disease systems.

Simulation modeling and analysis of a complex system of a thermal power plant

Directory of Open Access Journals (Sweden)

Sorabh Gupta

2009-09-01

Full Text Available The present paper deals with the opportunities for the modeling of flue gas and air system of a thermal power plant by making the performance evaluation using probabilistic approach. The present system of thermal plant under study consists of four subsystems with three possible states: full working, reduced capacity working and failed. Failure and repair rates for all the subsystems are assumed to be constant. Formulation of the problem is carried out using Markov Birth-Death process using probabilistic approach and a transition diagram represents the operational behavior of the system. Interrelationship among the full working and reduced working has been developed. A probabilistic model has been developed, considering some assumptions. Data in feasible range are selected from a survey of thermal plant and the effect of each subsystem on the system availability is tabulated in the form of availability matrices, which provides various performance/availability levels for different combinations of failure and repair rates of all subsystems. Based upon various availability values obtained in availability matrices and graphs of failure/repair rates of different subsystems, performance and optimum values of failure/repair rates for maximum availability, of each subsystem is analyzed and then maintenance priorities are decided for all subsystems.

The computer program system for structural design of nuclear power plants

International Nuclear Information System (INIS)

Aihara, S.; Atsumi, K.; Sasagawa, K.; Satoh, S.

1979-01-01

In recent days, the design method of the Nuclear Power Plant has become more complex than in the past. The Finite Element Method (FEM) applied for analysis of Nuclear Power Plants, especially requires more computer use. The recent computers have made remarkable progress, so that in design work manpower and time necessary for analysis have been reduced considerably. However, instead the arrangement of outputs have increased tremendously. Therefore, a computer program system was developed for performing all of the processes, from data making to output arrangement, and rebar evaluations. This report introduces the computer program system pertaining to the design flow of the Reactor Building. (orig.)

The System 80+ Standard Plant design control document. Volume 1

International Nuclear Information System (INIS)

1997-01-01

This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume covers the DCD introduction and contains sections 1 and parts 1--7 of section 2 of the CDM. Parts 1--7 included the following: (2.1) Design of SSC; (2.2) Reactor; (2.3) RCS and connected systems; (2.4) Engineered Safety Features; (2.5) Instrumentation and Control; (2.6) Electric Power; and (2.7) Auxiliary Systems

Economic Analysis of a Nuclear Reactor Powered High-Temperature Electrolysis Hydrogen Production Plant

International Nuclear Information System (INIS)

E. A. Harvego; M. G. McKellar; M. S. Sohal; J. E. O'Brien; J. S. Herring

2008-01-01

A reference design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production was developed to provide a basis for comparing the HTE concept with other hydrogen production concepts. The reference plant design is driven by a high-temperature helium-cooled nuclear reactor coupled to a direct Brayton power cycle. The reference design reactor power is 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 540 C and 900 C, respectively. The electrolysis unit used to produce hydrogen includes 4,009,177 cells with a per-cell active area of 225 cm2. The optimized design for the reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes an air-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The inlet air for the air-sweep system is compressed to the system operating pressure of 5.0 MPa in a four-stage compressor with intercooling. The alternating-current, AC, to direct-current, DC, conversion efficiency is 96%. The overall system thermal-to-hydrogen production efficiency (based on the lower heating value of the produced hydrogen) is 47.12% at a hydrogen production rate of 2.356 kg/s. An economic analysis of this plant was performed using the standardized H2A Analysis Methodology developed by the Department of Energy (DOE) Hydrogen Program, and using realistic financial and cost estimating assumptions. The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a competitive cost. A cost of $3.23/kg of hydrogen was calculated assuming an internal rate of return of 10%

GiA Roots: software for the high throughput analysis of plant root system architecture

Science.gov (United States)

2012-01-01

Background Characterizing root system architecture (RSA) is essential to understanding the development and function of vascular plants. Identifying RSA-associated genes also represents an underexplored opportunity for crop improvement. Software tools are needed to accelerate the pace at which quantitative traits of RSA are estimated from images of root networks. Results We have developed GiA Roots (General Image Analysis of Roots), a semi-automated software tool designed specifically for the high-throughput analysis of root system images. GiA Roots includes user-assisted algorithms to distinguish root from background and a fully automated pipeline that extracts dozens of root system phenotypes. Quantitative information on each phenotype, along with intermediate steps for full reproducibility, is returned to the end-user for downstream analysis. GiA Roots has a GUI front end and a command-line interface for interweaving the software into large-scale workflows. GiA Roots can also be extended to estimate novel phenotypes specified by the end-user. Conclusions We demonstrate the use of GiA Roots on a set of 2393 images of rice roots representing 12 genotypes from the species Oryza sativa. We validate trait measurements against prior analyses of this image set that demonstrated that RSA traits are likely heritable and associated with genotypic differences. Moreover, we demonstrate that GiA Roots is extensible and an end-user can add functionality so that GiA Roots can estimate novel RSA traits. In summary, we show that the software can function as an efficient tool as part of a workflow to move from large numbers of root images to downstream analysis. PMID:22834569

Environmental cost/benefit analysis for fusion power plants

International Nuclear Information System (INIS)

Young, J.R.

1976-11-01

This document presents a cost/benefit analysis of use of fusion power plants early in the 21st century. The first section describes the general formulation of the analysis. Included are the selection of the alternatives to the fusion reactor, selection of the power system cases to be compared, and a general comparison of the environmental effects of the selected alternatives. The second section compares the cumulative environmental effects from 2010 to 2040 for the primary cases of the power system with and without fusion reactors. The third section briefly illustrates the potential economic benefits if fusion reactors produce electricity at a lower unit cost than LMFBRs can. The fourth section summarizes the cost/benefit analysis

Developments of HTGR thermofluid dynamic analysis codes and HTGR plant dynamic simulation code

International Nuclear Information System (INIS)

Tanaka, Mitsuhiro; Izaki, Makoto; Koike, Hiroyuki; Tokumitsu, Masashi

1983-01-01

In nuclear power plants as well as high temperature gas-cooled reactor plants, the design is mostly performed on the basis of the results after their characteristics have been grasped by carrying out the numerical simulation using the analysis code. Also in Kawasaki Heavy Industries Ltd., on the basis of the system engineering accumulated with gas-cooled reactors since several years ago, the preparation and systematization of analysis codes have been advanced, aiming at lining up the analysis codes for heat transferring flow and control characteristics, taking up HTGR plants as the main object. In this report, a part of the results is described. The example of the analysis applying the two-dimensional compressible flow analysis codes SOLA-VOF and SALE-2D, which were developed by Los Alamos National Laboratory in USA and modified for use in Kawasaki, to HTGR system is reported. Besides, Kawasaki has developed the control characteristics analyzing code DYSCO by which the change of system composition is easy and high versatility is available. The outline, fundamental equations, fundamental algorithms and examples of application of the SOLA-VOF and SALE-2D, the present status of system characteristic simulation codes and the outline of the DYSCO are described. (Kako, I.)

ChemANDTM - a system health monitor for plant chemistry

International Nuclear Information System (INIS)

Turner, C.W.; Mitchel, G.R.; Balakrishnan, P.V.; Tosello, G.

1999-07-01

Effective management of plant systems throughout their lifetime requires much more than data acquisition and display - it requires that the plant's system health be continually monitored and managed. AECL has developed a System Health Monitor called ChemAND for CANDU plant chemistry. ChemAND, a Chemistry ANalysis and Diagnostic system, monitors key chemistry parameters in the heat transport system, moderator-cover gas, annulus gas, and the steam cycle during full-power operation and feeds these parameters to models that calculate the effect of current plant operating conditions on the present and future health of the system. Chemistry data from each of the systems are extracted on a regular basis from the plant's Historical Data Server and are sorted according to function, e.g., indicators for condenser in-leakage, air in-leakage, heavy water leakage into the annulus gas, fuel failure, etc. Each parameter is conveniently displayed and is trended along with its alarm limits. ChemAND currently has two analytical models developed for the balance-of-plant. CHEMSOLV calculates crevice chemistry conditions in the steam generator (SG) from either the SG blowdown chemistry conditions or from a simulated condenser leak. This information will be used by operations personnel to evaluate the potential for SG tube corrosion in the crevice region. CHEMSOLV also calculates chemistry conditions throughout the steam-cycle system, as determined by the transport of volatile species such as ammonia, hydrazine, morpholine, and oxygen. A second model, SLUDGE, calculates the deposit loading in the SG as a function of time, based on concentrations of corrosion product in the final feedwater and plant operating conditions. Operations personnel can use this information to predict where to inspect and when to clean. In a future development, SLUDGE will track deposit loading arising from start-up crud bursts and will be used in conjunction with the thermohydraulics code, THIRST, to predict

Canadian CANDU plant historical data systems. A review and look to the future

International Nuclear Information System (INIS)

Deverno, M.; De Grosbois, J.; Pothier, H.

1996-01-01

As part of several CANDU Owner's Group (COG) projects, AECL has conducted a review of current approaches and investigated solutions to plant process data collection, management, and use. Emphasis was placed on understanding the existing functionality and uses of plant data systems, their future needs and benefits. The result is a vision of a plant-wide Historical Data System (HDS) providing seamless access to both near real-time and historical data, user tool-kits for data visualization and analysis, and data management of the larger volume of data acquired during the life of a plant. HDS technology is critical to the implementation of technical surveillance and analysis, predictive and preventative maintenance programs, and other efforts necessary to enhance plant safety, availability, production and productivity. HDS technology will lead to higher capability and capacity factors while minimizing operations, maintenance, and administration (OM and A) costs. (author)

Canadian CANDU plant historical data systems. A review and look to the future

Energy Technology Data Exchange (ETDEWEB)

Deverno, M; De Grosbois, J [Atomic Energy of Canada Limited, Chalk River, ON (Canada); Pothier, H [Atomic Energy of Canada Limited, Mississauga, ON (Canada); and others

1997-12-31

As part of several CANDU Owner`s Group (COG) projects, AECL has conducted a review of current approaches and investigated solutions to plant process data collection, management, and use. Emphasis was placed on understanding the existing functionality and uses of plant data systems, their future needs and benefits. The result is a vision of a plant-wide Historical Data System (HDS) providing seamless access to both near real-time and historical data, user tool-kits for data visualization and analysis, and data management of the larger volume of data acquired during the life of a plant. HDS technology is critical to the implementation of technical surveillance and analysis, predictive and preventative maintenance programs, and other efforts necessary to enhance plant safety, availability, production and productivity. HDS technology will lead to higher capability and capacity factors while minimizing operations, maintenance, and administration (OM and A) costs. (author) 15 refs., 7 figs.

Risk-informed decision-making analysis for the electrical raceway fire barrier systems on a BWR-4 plant

International Nuclear Information System (INIS)

Wu, Ching-Hui; Lin, Tsu-Jen; Kao, Tsu-Mu; Chen, Chyn-Rong

2003-01-01

This paper describes a risk-informed decision-making approach used to resolve the fire barrier issue in a BWR-4 nuclear plant where Appendix R separation requirements cannot be met without installing additional fire protection features such as electrical raceway fire barrier system. The related risk measures in CDF (core damage frequency) and LERF (large early release frequency) of the fire barrier issue can be determined by calculating the difference in plant risks between various alternative cases and that met the requirement of the Appendix R. In some alternative cases, additional early-detection and fast-response fire suppression systems are suggested. In some other cases, cable re-routing of some improper layout of non-safety related cables are required. Sets of fire scenarios are re-evaluated more detailed by reviewing the cable damage impact for the BWR-4 plant. The fire hazard model, COMPBRM III-e, is used in this study and the dominant results in risk measures are benchmarked with the CFD code, FDS 2.0, to ensure that the risk impact of fire barrier is estimated accurately in the risk-informed decision making. The traditional deterministic qualitative methods, such as defense-in-depth, safety margin and post-fire safety shutdown capability are also proceeded. The value-impact analysis for proposed alternatives of fire wrapping required by Appendix R has been completed for technical basis of the exemption on Appendix R application. The outcome of the above analysis should be in compliance with the regulatory guidelines (RG) 1.174 and 1.189 for the applications in the risk-informed decision-making of the fire wrapping issues. (author)

Integration of a nuclear power plant in electrical systems, alternative programs, optimization

International Nuclear Information System (INIS)

Souza, J.A.M. de.

1991-01-01

The problem of integration of a nuclear power plants in a electrical power system, to support the power demand of the system, and mainly also support the power demand at the critical period, I.E., peak demands, is analysed. The factors considered in this analysis are: the demand structure of the region, the availability of others power plants in the electrical net and the capacity factor. (author)

Station black out analysis for CANDU 6 plant

International Nuclear Information System (INIS)

Baburajan, P.K.; Rao, R.S.; Gupta, S.K.

2011-01-01

As part of International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP), 'Benchmarking severe accident computer codes for pressurised heavy water reactor applications', thermal hydraulic analysis of severe accident station black out (SBO) is carried out for a generic CANDU 6 plant. The CRP is conducted in order to improve severe accident analysis capability for heavy water reactors (HWRs) through the benchmarking exercise. The plant simulation is carried out using RELAP5/Mod3.4 best estimate system thermal hydraulic code. The total thermal power of the plant is 2064 MW. There are 380 fuel channels in the core, 12 fuel bundles per channel and each bundle assembly has 37 fuel elements. The primary heat transport system (HTS) consists of two loops. Each loop consist of inlet and exit headers, feeder lines, fuel channels, hot leg and cold leg of steam generator, pumps, pump suction and discharge lines. Ninety five fuel channels in each pass of the loop are simulated as a single channel. The steam generator as the secondary side heat sink consists of annulus down-comer, riser, steam separator, steam drum, steam header and steam lines. Fuel channels (pressure tube) and calandria tube are simulated using SCDAPSIM to study the severe accident code behaviour. The SBO transient is initiated after obtaining the steady state conditions. Present analysis is carried out till the pressure tube failure. Analysis results show that the secondary inventory is lost in about 6500 seconds of the transient. The primary inventory is lost in 10370 seconds of the transient and subsequently the pressure tube failure is predicted as the tube wall temperature exceeded 900 K. Further analysis is to be carried out by incorporating changes in the calandria model and including the modeling of calandria vault and containment. (author)

Weibull distribution in reliability data analysis in nuclear power plant

International Nuclear Information System (INIS)

Ma Yingfei; Zhang Zhijian; Zhang Min; Zheng Gangyang

2015-01-01

Reliability is an important issue affecting each stage of the life cycle ranging from birth to death of a product or a system. The reliability engineering includes the equipment failure data processing, quantitative assessment of system reliability and maintenance, etc. Reliability data refers to the variety of data that describe the reliability of system or component during its operation. These data may be in the form of numbers, graphics, symbols, texts and curves. Quantitative reliability assessment is the task of the reliability data analysis. It provides the information related to preventing, detect, and correct the defects of the reliability design. Reliability data analysis under proceed with the various stages of product life cycle and reliability activities. Reliability data of Systems Structures and Components (SSCs) in Nuclear Power Plants is the key factor of probabilistic safety assessment (PSA); reliability centered maintenance and life cycle management. The Weibull distribution is widely used in reliability engineering, failure analysis, industrial engineering to represent manufacturing and delivery times. It is commonly used to model time to fail, time to repair and material strength. In this paper, an improved Weibull distribution is introduced to analyze the reliability data of the SSCs in Nuclear Power Plants. An example is given in the paper to present the result of the new method. The Weibull distribution of mechanical equipment for reliability data fitting ability is very strong in nuclear power plant. It's a widely used mathematical model for reliability analysis. The current commonly used methods are two-parameter and three-parameter Weibull distribution. Through comparison and analysis, the three-parameter Weibull distribution fits the data better. It can reflect the reliability characteristics of the equipment and it is more realistic to the actual situation. (author)

Seismic analysis of nuclear piping system

International Nuclear Information System (INIS)

Shrivastava, S.K.; Pillai, K.R.V.; Nandakumar, S.

1975-01-01

To illustrate seismic analysis of nuclear power plant piping, a simple piping system running between two floors of the reactor building is assumed. Reactor building floor response is derived from time-history method. El Centre earthquake (1940) accelerogram is used for time-history analysis. The piping system is analysed as multimass lumped system. Behaviour of the pipe during the said earthquake is discussed. (author)

DNA Damage Repair System in Plants: A Worldwide Research Update.

Science.gov (United States)

Gimenez, Estela; Manzano-Agugliaro, Francisco

2017-10-30

Living organisms are usually exposed to various DNA damaging agents so the mechanisms to detect and repair diverse DNA lesions have developed in all organisms with the result of maintaining genome integrity. Defects in DNA repair machinery contribute to cancer, certain diseases, and aging. Therefore, conserving the genomic sequence in organisms is key for the perpetuation of life. The machinery of DNA damage repair (DDR) in prokaryotes and eukaryotes is similar. Plants also share mechanisms for DNA repair with animals, although they differ in other important details. Plants have, surprisingly, been less investigated than other living organisms in this context, despite the fact that numerous lethal mutations in animals are viable in plants. In this manuscript, a worldwide bibliometric analysis of DDR systems and DDR research in plants was made. A comparison between both subjects was accomplished. The bibliometric analyses prove that the first study about DDR systems in plants (1987) was published thirteen years later than that for other living organisms (1975). Despite the increase in the number of papers about DDR mechanisms in plants in recent decades, nowadays the number of articles published each year about DDR systems in plants only represents 10% of the total number of articles about DDR. The DDR research field was done by 74 countries while the number of countries involved in the DDR & Plant field is 44. This indicates the great influence that DDR research in the plant field currently has, worldwide. As expected, the percentage of studies published about DDR systems in plants has increased in the subject area of agricultural and biological sciences and has diminished in medicine with respect to DDR studies in other living organisms. In short, bibliometric results highlight the current interest in DDR research in plants among DDR studies and can open new perspectives in the research field of DNA damage repair.

Development of advanced secondary chemistry monitoring system for Korea nuclear power plants

International Nuclear Information System (INIS)

Lee, Sang Hak; Kim, Chung Tae

1997-01-01

Water chemistry control is one of the most important tasks in order to maintain the reliability of plant equipments and extend the operating life of the plant. KEPCO and KOPEC developed a computerized tool for this purpose -ASCMS (advanced secondary chemistry monitoring system) which is able to monitor and diagnose the secondary water chemistry. A prototype system had been installed at KORI 3 nuclear power plant since April 1993 in order to evaluate the system performance. After the implementation of enhancements identified during the testing of the prototype, we have developed the advanced secondary monitoring system, ASCMS which is installed at 5 nuclear power plants and has been under operations since April 1997. The ASCMS comprises PC subsystem designed for data acquisition, data analysis, and data diagnosis. The ASCMS will provide overall information related to steam generator secondary side water chemistry problems and improve plant availability, reduce radiation exposure to workers, and reduce operating and maintenance costs. 6 figs

Analysis on nuclear power plant control room system design and improvement based on human factor engineering

International Nuclear Information System (INIS)

Gao Feng; Liu Yanzi; Sun Yongbin

2014-01-01

The design of nuclear power plant control room system is a process of improvement with the implementation of human factor engineering theory and guidance. The method of implementation human factor engineering principles into the nuclear power plant control room system design and improvement was discussed in this paper. It is recommended that comprehensive address should be done from control room system function, human machine interface, digital procedure, control room layout and environment design based on the human factor engineering theory and experience. The main issues which should be paid more attention during the control room system design and improvement also were addressed in this paper, and then advices and notices for the design and improvement of the nuclear power plant control room system were afforded. (authors)

Integration of acoustic emission systems within Integri-TechTM analysis system for structural health monitoring of pressurised engineering plant

International Nuclear Information System (INIS)

Ghouri, A A; Galbraith, Walter; Pierce, S Gareth; Gachagan, Anthony; Rafferty, Steven; Pickwell, Andy

2015-01-01

The aim of this Acoustic Emission (AE) based Structural Health Monitoring project is to enable accurate location of AE sources in pressurised engineering plant and to use AE source location data to establish defect locations for use within Integri-Tech TM ; a finite element based analysis, monitoring and fitness for service assessment system. Integri-Tech TM is a windows based system which carries out combined analysis and assessment providing fatigue life and remnant life calculations and inspection priorities presenting the results in an accessible web portal format. The software uses finite element stress models created in the companion software Model Wizard. The AE monitoring system that has been developed can be used with an array of up to four AE broad band sensor channels with associated signal processing. Using a flexible approach in MATLAB, the authors have developed algorithms which were used for analysing the received AE signals to extract information about the nature and location of the source. The ability to carry out source location and possibly perform real time monitoring (detecting cracking as it occurs) is attractive feature of the AE system developed for this project. The time of arrival (TOA) data was used by Integri-Tech TM software to calculate source location using its own built-in algorithm, and this was verified independently using a MATLAB approach. (paper)

Security challenges in designing I and C systems for nuclear power plant

International Nuclear Information System (INIS)

Behera, Rajendra Prasad; Jayanthi, T.; Madhusoodanan, K.; Satya Murty, S.A.V.

2016-01-01

Geographically distributed instrumentation and control (I and C) systems in any nuclear power plant (NPP) facilitate the operator with remote access to real-time data and issue supervisory command to remote control devices deployed in the field. The increased connectivity to plant communication network has exposed I and C systems to security vulnerabilities both in terms of physical and logical access. For example, denial-of service and fault induction attack can disrupt the operation of I and C systems by delaying or blocking the flow of data through plant communication network. The design process of I and C system is quite challenging since an engineer has to consider both safety and security features implemented in hardware and software components of the system. This paper analyzes attack taxonomy based on available data and presents Security Tree Analysis (STA) technique towards building safe and secures I and C systems for Nuclear Power Plant. (author)

Developments in power plant cooling systems

International Nuclear Information System (INIS)

Agarwal, N.K.

1993-01-01

A number of cooling systems are used in the power plants. The condenser cooling water system is one of the most important cooling systems in the plant. The system comprises a number of equipment. Plants using sea water for cooling are designed for the very high corrosion effects due to sea water. Developments are taking place in the design, materials of construction as well as protection philosophies for the various equipment. Power optimisation of the cycle needs to be done in order to design an economical system. Environmental (Protection) Act places certain limitations on the effluents from the plant. An attempt has been made in this paper to outline the developing trends in the various equipment in the condenser cooling water systems used at the inland as well as coastal locations. (author). 5 refs., 6 refs

Analysis of combustion turbine inlet air cooling systems applied to an operating cogeneration power plant

International Nuclear Information System (INIS)

Chacartegui, R.; Jimenez-Espadafor, F.; Sanchez, D.; Sanchez, T.

2008-01-01

In this work, combustion turbine inlet air cooling (CTIAC) systems are analyzed from an economic outlook, their effects on the global performance parameters and the economic results of the power plant. The study has been carried out on a combined cogeneration system, composed of a General Electric PG 6541 gas turbine and a heat recovery steam generator. The work has been divided into three parts. First, a revision of the present CTIAC technologies is shown, their effects on power plant performance and evaluation of the associated investment and maintenance costs. In a second phase of the work, the cogeneration plant was modelled with the objective of evaluating the power increase and the effects on the generated steam and the thermal oil. The cogeneration power plant model was developed, departing from the recorded operational data of the plant in 2005 and the gas turbine model offered by General Electric, to take into consideration that, in 2000, the gas turbine had been remodelled and the original performance curves should be corrected. The final objective of this model was to express the power plant main variables as a function of the gas turbine intake temperature, pressure and relative humidity. Finally, this model was applied to analyze the economic interest of different intake cooling systems, in different operative ranges and with different cooling capacities

The System 80+ Standard Plant design control document. Volume 24

International Nuclear Information System (INIS)

1997-01-01

This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume contains sections 7--11 of the ADM Emergency Operations Guidelines. Topics covered are: excess steam demand recovery; loss of all feedwater; loss of offsite power; station blackout recovery; and functional recovery guideline. Appendix A Severe Accident Management Guidelines and Appendix B Lower Mode Operational Guidelines are also included

Interface between computational fluid dynamics (CFD) and plant analysis computer codes

International Nuclear Information System (INIS)

Coffield, R.D.; Dunckhorst, F.F.; Tomlinson, E.T.; Welch, J.W.

1993-01-01

Computational fluid dynamics (CFD) can provide valuable input to the development of advanced plant analysis computer codes. The types of interfacing discussed in this paper will directly contribute to modeling and accuracy improvements throughout the plant system and should result in significant reduction of design conservatisms that have been applied to such analyses in the past

Vibration analysis in nuclear power plant using neural networks

International Nuclear Information System (INIS)

Loskiewicz-Buczak, A.; Alguindigue, I.E.

1993-01-01

Vibration monitoring of components in nuclear power plants has been used for a number of years. This technique involves the analysis of vibration data coming from vital components of the plant to detect features which reflect the operational state of machinery. The analysis leads to the identification of potential failures and their causes, and makes it possible to perform efficient preventive maintenance. This paper documents the authors' work on the design of a vibration monitoring methodology enhanced by neural network technology. This technology provides an attractive complement to traditional vibration analysis because of the potential of neural networks to handle data which may be distorted or noisy. This paper describes three neural networks-based methods for the automation of some of the activities related to motion and vibration monitoring in engineering systems

State-of-the-art report on accident analysis and risk analysis of reprocessing plants in European countries

International Nuclear Information System (INIS)

Nomura, Yasushi

1985-12-01

This report summarizes informations obtained from America, England, France and FRG concerning methodology, computer code, fundamental data and calculational model on accident/risk analyses of spent fuel reprocessing plants. As a result, the followings are revealed. (1) The system analysis codes developed for reactor plants can be used for reprocessing plants with some code modification. (2) Calculational models and programs have been developed for accidental phenomenological analyses in FRG, but with insufficient data to prove them. (3) The release tree analysis codes developed in FRG are available to estimate radioactivity release amount/probability via off-gas/exhaustair lines in the case of accidents. (4) The computer codes developed in America for reactor-plant environmental transport/safety analyses of released radioactivity can be applied to reprocessing facilities. (author)

On exergy analysis of industrial plants and significance of ambient temperature

Energy Technology Data Exchange (ETDEWEB)

Rian, Berit

2011-07-01

The exergy analysis has been a relatively mature theory for more than 30 years. However, it is not that developed in terms of procedures for optimizing systems, which partly explains why it is not that common. Misconceptions and prejudices, even among scientists, are also partly to blame.The main objective of this work was to contribute to the development of an understanding and methodology of the exergy analysis. The thesis was mainly based on three papers, two of which provided very different examples from existing industrial systems in Norway, thus showing the societal perspective in terms of resource utilization and thermodynamics. The last paper and the following investigation were limited to certain aspects of ambient conditions. Two Norwegian operational plants have been studied, one operative for close to 30 years (Kaarstoe steam production and distribution system), while the other has just started its expected 30 years of production (Snoehvit LNG plant). In addition to mapping the current operational status of these plants, the study of the Kaarstoe steam production and distribution system concluded that the potential for increasing the thermodynamic performance by rather cautious actions was significant, whereas the study of the Snoehvit LNG plant showed the considerable profit which the Arctic location provided in terms of reduced fuel consumption. The significance of the ambient temperature led to the study of systems with two ambient bodies (i.e. ambient water and ambient air) of different temperatures, here three different systems were investigated: A regenerative steam injection gas turbine (RSTIG), a simple Linde air liquefaction plant (Air Liq) and an air-source heat pump water heater (HPWH). In particular, the effect of the chosen environment on exergy analysis was negligible for RSTIG, modest for Air Liq and critical for HPWH. It was found that the amount of exergy received from the alternative ambient body, compared to the main exergy flow of

Simulation analysis of capacity and performance improvement in wastewater treatment plants: Case study of Alexandria eastern plant

Science.gov (United States)

Moursy, Aly; Sorour, Mohamed T.; Moustafa, Medhat; Elbarqi, Walid; Fayd, Mai; Elreedy, Ahmed

2018-05-01

This study concerns the upgrading of a real domestic wastewater treatment plant (WWTP) supported by simulation. The main aims of this work are to: (1) decide between two technologies to improve WWTP capacity and its nitrogen removal efficiency; membrane bioreactor (MBR) and integrated fixed film activated sludge (IFAS), and (2) perform a cost estimation analysis for the two proposed solutions. The model used was calibrated based on data from the existing WWTP, namely, Eastern plant and located in Alexandria, Egypt. The activated sludge model No. 1 (ASM1) was considered in the model analysis by GPS-X 7 software. Steady-state analysis revealed that high performances corresponded to high compliance with Egyptian standards were achieved by the two techniques; however, MBR was better. Nonetheless, the two systems showed poor nitrogen removal efficiency according to the current situation, which reveals that the plant needs a modification to add an anaerobic treatment unit before the aerobic zone.

Some Sensitivity Studies of Chemical Transport Simulated in Models of the Soil-Plant-Litter System

Energy Technology Data Exchange (ETDEWEB)

Begovich, C.L.

2002-10-28

Fifteen parameters in a set of five coupled models describing carbon, water, and chemical dynamics in the soil-plant-litter system were varied in a sensitivity analysis of model response. Results are presented for chemical distribution in the components of soil, plants, and litter along with selected responses of biomass, internal chemical transport (xylem and phloem pathways), and chemical uptake. Response and sensitivity coefficients are presented for up to 102 model outputs in an appendix. Two soil properties (chemical distribution coefficient and chemical solubility) and three plant properties (leaf chemical permeability, cuticle thickness, and root chemical conductivity) had the greatest influence on chemical transport in the soil-plant-litter system under the conditions examined. Pollutant gas uptake (SO{sub 2}) increased with change in plant properties that increased plant growth. Heavy metal dynamics in litter responded to plant properties (phloem resistance, respiration characteristics) which induced changes in the chemical cycling to the litter system. Some of the SO{sub 2} and heavy metal responses were not expected but became apparent through the modeling analysis.

Conceptual design of an integrated information system for safety related analysis of nuclear power plants (IRIS Phase 1)

International Nuclear Information System (INIS)

Hofer, K.; Zehnder, P.; Galperin, A.

1994-01-01

This report deals with a conceptual design of an integrated information management system, called PSI-IRIS, as needed to assist the analysts for reactor safety related investigations on Swiss nuclear power plants within the project STARS. Performing complicated engineering analyses of an NPP requires storage and manipulation of a large amount of information, both data and knowledge. This information is characterized by its multi-disciplinary nature, complexity, and diversity. The problems caused by inefficient and lengthy manual operations involving the data flow management within the framework of the safety related analysis of an NPP, can be solved by applying computer aided engineering (CAE) principles. These principles are the basis for the design of the integrated information management system PSI-IRIS presented in this report. The basic idea is to create a computerized environment, which includes both database and functional capabilities. The database of the PSI-IRIS consists of two parts, an NPP generic database (GDB) and a collection of analysis results (CASE L IB). The GDB includes all technical plant data and information needed to generate input decks for all computer codes utilized within the STARS project. The CASE L IB storage contains the accumulated knowledge, input decks, and result files of the NPP transient analyses. Considerations and analysis of the data types and the required data manipulation capabilities as well as operational requirements resulted in the choice of an object-oriented database management system (OODBMS) as a development platform for solving the software engineering problems. Several advantages of OODBMS's over conventional relational database management systems were found of crucial importance, especially providing the necessary flexibility for different data types and the potential for extensibility. (author) 15 figs., tabs., 20 refs

Summary of component reliability data for probabilistic safety analysis of Korean standard nuclear power plant

International Nuclear Information System (INIS)

Choi, S. Y.; Han, S. H.

2004-01-01

The reliability data of Korean NPP that reflects the plant specific characteristics is necessary for PSA of Korean nuclear power plants. We have performed a study to develop the component reliability DB and S/W for component reliability analysis. Based on the system, we had have collected the component operation data and failure/repair data during plant operation data to 1998/2000 for YGN 3,4/UCN 3,4 respectively. Recently, we have upgraded the database by collecting additional data by 2002 for Korean standard nuclear power plants and performed component reliability analysis and Bayesian analysis again. In this paper, we supply the summary of component reliability data for probabilistic safety analysis of Korean standard nuclear power plant and describe the plant specific characteristics compared to the generic data

Grey relational analysis for evaluating the effects of different rates of wine lees-derived biochar application on a plant-soil system with multi-metal contamination.

Science.gov (United States)

Xu, Min; Zhu, Qihong; Wu, Jun; He, Yan; Yang, Gang; Zhang, Xiaohong; Li, Li; Yu, Xiaoyu; Peng, Hong; Wang, Lilin

2018-03-01

In this study, grey relational analysis (GRA) was used to investigate the effects of different application rates of wine lees-derived biochar on a plant-soil system with multi-metal contamination. A pot experiment was conducted to determine rice growth in multi-metal-contaminated soil amended with samples of wine lees-derived biochar, and 47 indicators (including soil properties, microbial activity, and plant physiology) were selected as evaluation indexes to assess the plant-soil system. The results indicated that higher wine lees-derived biochar application rates (2% W/W) were favorable for soil fertility, the bioconcentration factor (BF), and the mobility factor (MF, %) (with the exception of Cr, Zn, and Hg), but an application of 1% produced the highest plant growth, enzymatic activities, and bacterial diversity. The richness of the bacterial communities was reduced in the soil amended with the wine lees-derived biochar. According to the GRA assessment, the 1% application rate of wine lees-derived biochar was more suitable for restoring the holistic plant-soil system than were the application rates of 0, 0.5, and 2% (W/W). Furthermore, this study shows that GRA is a useful method for evaluating plant-soil systems.

The Plant-Window system: A flexible, expandable computing environment for the integration of power plant activities

International Nuclear Information System (INIS)

Wood, R.T.; Mullens, J.A.; Naser, J.A.

1994-01-01

Power plant data, and the information that can be derived from it, provide the link to the plant through which the operations, maintenance and engineering staff understand and manage plant performance. The increasing use of computer technology in the US nuclear power industry has greatly expanded the capability to obtain, analyze, and present data about the plant to station personnel. However, it is necessary to transform the vast quantity of available data into clear, concise, and coherent information that can be readily accessed and used throughout the plant. This need can be met by an integrated computer workstation environment that provides the necessary information and software applications, in a manner that can be easily understood and used, to the proper users throughout the plant. As part of a Cooperative Research and Development Agreement with the Electric Power Research Institute, the Oak Ridge National Laboratory has developed functional requirements for a Plant-Wide Integrated Environment Distributed on Workstations (Plant-Window) System. The Plant-Window System (PWS) can serve the needs of operations, engineering, and maintenance personnel at nuclear power stations by providing integrated data and software applications (e.g., monitoring, analysis, diagnosis, and control applications) within a common environment. The PWS requirements identify functional capabilities and provide guidelines for standardized hardware, software, and display interfaces to define a flexible computer environment that permits a tailored implementation of workstation capabilities and facilitates future upgrades

Utilizing primary energy savings and exergy destruction to compare centralized thermal plants and cogeneration/trigeneration systems

International Nuclear Information System (INIS)

Espirito Santo, Denilson Boschiero do; Gallo, Waldyr Luiz Ribeiro

2017-01-01

Rising energy conversion processes efficiencies reduces CO_2 emissions and global warming implications. Decentralized electricity production through cogeneration/trigeneration systems can save primary energy if it operates with high efficiency. High efficiency is obtained when the system produces electricity and a substantial amount of the energy rejected by the prime mover is used to meet site thermal demands. Environmental concerns and international agreements are directing governments of different countries to incentive high efficiency solutions. Centralized thermal plants and cogeneration/trigeneration efficiency are compared through efficiency indicators using the first law of thermodynamics and the second law of thermodynamics. This paper proposes the use of the primary energy savings analysis and the exergy destruction analysis to compare decentralized power production through cogeneration/trigeneration systems and centralized thermal plants. The analysis concluded that both methods achieve the same results if the thermal efficiency indicator is used to compare the methods. The analysis also revealed that trigeneration systems with the same energy input are comparable with quite different thermal efficiency centralized thermal plants. Case 1 is comparable to a 53% thermal efficiency power plant and case 2 is comparable to a 77% thermal efficiency power plant. - Highlights: • Trigeneration and thermal plants are compared using PES and exergy destruction. • The thermal efficiency indicator is used to compare both methods. • The same equivalent thermal efficiency is achieved by both methods. • Same energy input trigeneration is similar to different thermal efficiency plants. • Evaluated trigeneration are comparable to a 53–77% thermal efficiency power plant.

Analysis of Plant Breeding on Hadoop and Spark

Directory of Open Access Journals (Sweden)

Shuangxi Chen

2016-01-01

Full Text Available Analysis of crop breeding technology is one of the important means of computer-assisted breeding techniques which have huge data, high dimensions, and a lot of unstructured data. We propose a crop breeding data analysis platform on Spark. The platform consists of Hadoop distributed file system (HDFS and cluster based on memory iterative components. With this cluster, we achieve crop breeding large data analysis tasks in parallel through API provided by Spark. By experiments and tests of Indica and Japonica rice traits, plant breeding analysis platform can significantly improve the breeding of big data analysis speed, reducing the workload of concurrent programming.

Balance of Plant System Analysis and Component Design of Turbo-Machinery for High Temperature Gas Reactor Systems

Energy Technology Data Exchange (ETDEWEB)

Ballinger, Ronald G. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Wang, Chun Yun [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Kadak, Andrew [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Todreas, Neil [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Mirick, Bradley [Concepts, Northern Engineering and Research, Woburn, MA (United States); Demetri, Eli [Concepts, Northern Engineering and Research, Woburn, MA (United States); Koronowski, Martin [Concepts, Northern Engineering and Research, Woburn, MA (United States)

2004-08-30

power conversion system have been verified with an industry-standard general thermal-fluid code Flownet. With respect to the dynamic model, bypass valve control and inventory control have been used as the primary control methods for the power conversion system. By performing simulation using the dynamic model with the designed control scheme, the combination of bypass and inventory control was optimized to assure system stability within design temperature and pressure limits. Bypass control allows for rapid control system response while inventory control allows for ultimate steady state operation at part power very near the optimum operating point for the system. Load transients simulations show that the indirect, three-shaft arrangement gas turbine power conversion system is stable and controllable. For the indirect cycle the intermediate heat exchanger (IHX) is the interface between the reactor and the turbomachinery systems. As a part of the design effort the IHX was identified as the key component in the system. Two technologies, printed circuit and compact plate-fin, were investigated that have the promise of meeting the design requirements for the system. The reference design incorporates the possibility of using either technology although the compact plate-fin design was chosen for subsequent analysis. The thermal design and parametric analysis with an IHX and recuperator using the plate-fin configuration have been performed. As a three-shaft arrangement, the turbo-shaft sets consist of a pair of turbine/compressor sets (high pressure and low pressure turbines with same-shaft compressor) and a power turbine coupled with a synchronous generator. The turbines and compressors are all axial type and the shaft configuration is horizontal. The core outlet/inlet temperatures are 900/520 C, and the optimum pressure ratio in the power conversion cycle is 2.9. The design achieves a plant net efficiency of approximately 48%.

Balance of Plant System Analysis and Component Design of Turbo-Machinery for High Temperature Gas Reactor Systems

International Nuclear Information System (INIS)

Ballinger, Ronald G.; Chunyun Wang; Kadak, Andrew; Todreas, Neil

2004-01-01

power conversion system have been verified with an industry-standard general thermal-fluid code Flownet. With respect to the dynamic model, bypass valve control and inventory control have been used as the primary control methods for the power conversion system. By performing simulation using the dynamic model with the designed control scheme, the combination of bypass and inventory control was optimized to assure system stability within design temperature and pressure limits. Bypass control allows for rapid control system response while inventory control allows for ultimate steady state operation at part power very near the optimum operating point for the system. Load transients simulations show that the indirect, three-shaft arrangement gas turbine power conversion system is stable and controllable. For the indirect cycle the intermediate heat exchanger (IHX) is the interface between the reactor and the turbomachinery systems. As a part of the design effort the IHX was identified as the key component in the system. Two technologies, printed circuit and compact plate-fin, were investigated that have the promise of meeting the design requirements for the system. The reference design incorporates the possibility of using either technology although the compact plate-fin design was chosen for subsequent analysis. The thermal design and parametric analysis with an IHX and recuperator using the plate-fin configuration have been performed. As a three-shaft arrangement, the turbo-shaft sets consist of a pair of turbine/compressor sets (high pressure and low pressure turbines with same-shaft compressor) and a power turbine coupled with a synchronous generator. The turbines and compressors are all axial type and the shaft configuration is horizontal. The core outlet/inlet temperatures are 900/520 C, and the optimum pressure ratio in the power conversion cycle is 2.9. The design achieves a plant net efficiency of approximately 48%

Towards systems biology of the gravity response of higher plants -multiscale analysis of Arabidopsis thaliana root growth

Science.gov (United States)

Palme, Klaus; Aubry, D.; Bensch, M.; Schmidt, T.; Ronneberger, O.; Neu, C.; Li, X.; Wang, H.; Santos, F.; Wang, B.; Paponov, I.; Ditengou, F. A.; Teale, W. T.; Volkmann, D.; Baluska, F.; Nonis, A.; Trevisan, S.; Ruperti, B.; Dovzhenko, A.

Gravity plays a fundamental role in plant growth and development. Up to now, little is known about the molecular organisation of the signal transduction cascades and networks which co-ordinate gravity perception and response. By using an integrated systems biological approach, a systems analysis of gravity perception and the subsequent tightly-regulated growth response is planned in the model plant Arabidopsis thaliana. This approach will address questions such as: (i) what are the components of gravity signal transduction pathways? (ii) what are the dynamics of these components? (iii) what is their spatio-temporal regulation in different tis-sues? Using Arabidopsis thaliana as a model-we use root growth to obtain insights in the gravity response. New techniques enable identification of the individual genes affected by grav-ity and further integration of transcriptomics and proteomics data into interaction networks and cell communication events that operate during gravitropic curvature. Using systematic multiscale analysis we have identified regulatory networks consisting of transcription factors, the protein degradation machinery, vesicle trafficking and cellular signalling during the gravire-sponse. We developed approach allowing to incorporate key features of the root system across all relevant spatial and temporal scales to describe gene-expression patterns and correlate them with individual gene and protein functions. Combination of high-resolution microscopy and novel computational tools resulted in development of the root 3D model in which quantitative descriptions of cellular network properties and of multicellular interactions important in root growth and gravitropism can be integrated for the first time.

Logic generator and validator for the disturbance analysis and surveillance system

International Nuclear Information System (INIS)

Szabo, J.

1983-01-01

The on-line monitoring of a power plant (or any process plant) consists of three primary functions: data acquisition, data analysis, and data presentation. Data acquisition is the process by which the basic set of physical data from plant components is obtained. In the data analysis, or logic, phase a specific data analysis computer code determines the meaning of the acquired physical data. The results of this logical analysis may either be presented to a plant operator who will then have to activate the controls systems accordingly or they can be used to automatically activate the same control systems. This work relates to second of the above three functions by presenting advanced methods for generating reliable data analysis computer codes

Development of A Plant Navigation System

International Nuclear Information System (INIS)

Furuta, Tomihiko; Nakagawa, Tsuneo; Kubota, Ryuji; Ikeda, Kouji

1998-01-01

A 'Plant Navigation System (PNS)' is under development to assist nuclear power plant (NPP) operators by automatically displaying the plant situation and plant operational procedures on a CRT screen when abnormalities occur. The operation procedures given in a symptom-oriented manual are expressed in a tree - type flowchart (modified PAD). The optimum operation procedure for an NPP is selected automatically using built-in diagnostic logics based on the current status of the NPP. Concerning the plant situation, the PNS displays important information only on the current status of the NPP. A prototype PNS system is being constructed. (authors)

Lithium Ion Batteries Ageing Analysis when used in a PV Power Plants

DEFF Research Database (Denmark)

Beltran, H.; Swierczynski, Maciej Jozef; Aparicio, N.

2012-01-01

This paper analyzes the integration of lithium ion (Li-ion) batteries into large scale grid-connected PV plants. It performs a systematic analysis on both the operation improvement obtained by a PV+ES power plant and the ageing experienced by the Li-ion batteries used as Energy Storage (ES) system...... when operating under different energy management strategies (EMS). In this paper, the PV+ES power plant structure is presented and the selection of Li-on batteries as ES system (ESS) is justified. Moreover, the simulation model used for studying the Li-ion battery ageing is explained and tested...

Monitoring support system for nuclear power plant

International Nuclear Information System (INIS)

Higashikawa, Yuichi; Kubota, Rhuji; Tanaka, Keiji; Takano, Yoshiyuki

1996-01-01

The nuclear power plants in Japan reach to 49 plants and supply 41.19 million kW in their installed capacities, which is equal to about 31% of total electric power generation and has occupied an important situation as a stable energy supplying source. As an aim to keeping safe operation and working rate of the power plants, various monitoring support systems using computer technology, optical information technology and robot technology each advanced rapidly in recent year have been developed to apply to the actual plants for a plant state monitoring system of operators in normal operation. Furthermore, introduction of the emergent support system supposed on accidental formation of abnormal state of the power plants is also investigated. In this paper, as a monitoring system in the recent nuclear power plants, design of control panel of recent central control room, introduction to its actual plant and monitoring support system in development were described in viewpoints of improvement of human interface, upgrade of sensor and signal processing techniques, and promotion of information service technique. And, trend of research and development of portable miniature detector and emergent monitoring support system are also introduced in a viewpoint of labor saving and upgrade of the operating field. (G.K.)