WorldWideScience

Sample records for nuclear system components

  1. Ventilation systems and components of nuclear power plants

    International Nuclear Information System (INIS)

    1997-01-01

    The most important radiation and nuclear safety requirements for the design and manufacture of nuclear power plant ventilation systems and components are presented in the guide. Also the regulatory activities of the Finnish Centre for Radiation and Nuclear Safety (STUK) as regards the ventilation systems and components are explained. Documents and data which shall be submitted to STUK during the various phases of the regulatory procedure relating to the design, construction, commissioning and operation of the nuclear power plants are presented. (13 refs.)

  2. Safety classification of nuclear power plant systems, structures and components

    International Nuclear Information System (INIS)

    1992-01-01

    The Safety Classification principles used for the systems, structures and components of a nuclear power plant are detailed in the guide. For classification, the nuclear power plant is divided into structural and operational units called systems. Every structure and component under control is included into some system. The Safety Classes are 1, 2 and 3 and the Class EYT (non-nuclear). Instructions how to assign each system, structure and component to an appropriate safety class are given in the guide. The guide applies to new nuclear power plants and to the safety classification of systems, structures and components designed for the refitting of old nuclear power plants. The classification principles and procedures applying to the classification document are also given

  3. Requirements for containment system components in CANDU nuclear power plants

    International Nuclear Information System (INIS)

    1988-02-01

    This Standard specifies the requirements and establishes the rules for design, fabrication, and installation of pressure-retaining containment system components. In this Standard the term 'components' includes non registered items

  4. Requirements for containment system components in CANDU nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1988-02-01

    This Standard specifies the requirements and establishes the rules for design, fabrication, and installation of pressure-retaining containment system components. In this Standard the term `components` includes non registered items.

  5. Basic components of a national control system for nuclear materials

    International Nuclear Information System (INIS)

    Rabot, G.

    1986-01-01

    The paper presents the different aspects related to the organization and the functioning of a national control and accounting system for nuclear materials. The legal aspects and the relations with the IAEA are included

  6. Human as a component of a nuclear material safeguard system

    International Nuclear Information System (INIS)

    Morgan, D.E.; Schechter, R.S.

    1978-01-01

    Many human vigilance experiments are summarized and principles are extracted which should be useful in designing and evaluating a nuclear material safeguard system. A human is a poor observer and is not a dependable part of any man-machine system when required to function as an observer. There are a few techniques which improve his performance by providing feedback. A conceptual model is presented which is helpful in design and evaluation of systems. There is some slight experimental support for the model. Finally, some techniques of time study and statistical control charting will be useful as a means of detecting nuclear diversion attempts

  7. The condition monitoring system of turbine system components for nuclear power plants

    International Nuclear Information System (INIS)

    Ono, Shigetoshi

    2013-01-01

    The thermal and nuclear power plants have been imposed a stable supply of electricity. To certainly achieve this, we built the plant condition monitoring system based on the heat and mass balance calculation. If there are some performance changes on the turbine system components of their power plants, the heat and mass balance of the turbine system will change. This system has ability to detect the abnormal signs of their components by finding the changes of the heat and mass balance. Moreover we note that this system is built for steam turbine cycle operating with saturated steam conditions. (author)

  8. Supervision of electrical and instrumentation systems and components at nuclear facilities

    International Nuclear Information System (INIS)

    1986-01-01

    The general guidelines for the supervision of nuclear facilities carried out by the Finnish Centre for Radiation and Nuclear Safety (STUK) are set forth in the guide YVL 1.1. This guide shows in more detail how STUK supervises the electrical and instrumentation systems and components of nuclear facilities

  9. Lithuanian requirements for ageing management of systems and components important to safety of nuclear power plant

    International Nuclear Information System (INIS)

    Ramanauskiene, A.

    2000-01-01

    In this paper the Lithuanian requirements for ageing management of systems and components important to safety of Ignalina nuclear power plant (two RBMK-1500 water-cooled graphite moderated channel-type power reactors) are presented

  10. General requirements for pressure-retaining systems and components in CANDU nuclear power plants

    International Nuclear Information System (INIS)

    1991-11-01

    This standard specifies the general requirements for the design, fabrication and installation of pressure-retaining systems, components, and their supports in CANDU nuclear power plants. (16 figs., 2 tabs., 25 refs.)

  11. Development of the software for the component reliability database system of Korean nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Han, Sang Hoon; Kim, Seung Hwan; Choi, Sun Young [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-03-01

    A study was performed to develop the system for the component reliability database which consists of database system to store the reliability data and softwares to analyze the reliability data.This system is a part of KIND (Korea Information System for Nuclear Reliability Database).The MS-SQL database is used to stores the component population data, component maintenance history, and the results of reliability analysis. Two softwares were developed for the component reliability system. One is the KIND-InfoView for the data storing, retrieving and searching. The other is the KIND-CompRel for the statistical analysis of component reliability. 4 refs., 13 figs., 7 tabs. (Author)

  12. A computer-controlled electronic system for the ultrasonic NDT of components for nuclear power stations

    International Nuclear Information System (INIS)

    Rehrmann, M.; Harbecke, D.

    1987-01-01

    The paper describes an automatic ultrasonic testing system combined with a computer-controlled electronics system, called IMPULS I, for the non-destructive testing of components of nuclear reactors. The system can be used for both in-service inspection and for inspection during the manufacturing process. IMPUL I has more functions and less components than conventional ultrasonic systems, and the system gives good reproducible test results and is easy to operate. (U.K.)

  13. Steam generators: critical components in nuclear steam supply systems

    Energy Technology Data Exchange (ETDEWEB)

    Stevens-Guille, P D

    1974-02-28

    Steam generators are critical components in power reactors. Even small internal leaks result in costly shutdowns for repair. Surveys show that leaks have affected one half of all water-cooled reactors in the world with steam generators. CANDU reactors have demonstrated the highest reliability. However, AECL is actively evolving new technology in design, manufacture, inspection and operation to maintain reliability. (auth)

  14. Spain's nuclear components industry

    International Nuclear Information System (INIS)

    Kaibel, E.

    1985-01-01

    Spanish industrial participation in supply of components for nuclear power plants has grown steadily over the last fifteen years. The share of Spanish companies in work for the five second generation nuclear power plants increased to 50% of total capital investments. The necessity to maintain Spanish technology and production in the nuclear field is emphasized

  15. Application of the Safety Classification of Structures, Systems and Components in Nuclear Power Plants

    International Nuclear Information System (INIS)

    2016-04-01

    This publication describes how to complete tasks associated with every step of the classification methodology set out in IAEA Safety Standards Series No. SSG-30, Safety Classification of Structures, Systems and Components in Nuclear Power Plants. In particular, how to capture all the structures, systems and components (SSCs) of a nuclear power plant to be safety classified. Emphasis is placed on the SSCs that are necessary to limit radiological releases to the public and occupational doses to workers in operational conditions This publication provides information for organizations establishing a comprehensive safety classification of SSCs compliant with IAEA recommendations, and to support regulators in reviewing safety classification submitted by licensees

  16. On the classification of structures, systems and components of nuclear research and test reactors

    International Nuclear Information System (INIS)

    Mattar Neto, Miguel

    2009-01-01

    The classification of structures, systems and components of nuclear reactors is a relevant issue related to their design because it is directly associated with their safety functions. There is an important statement regarding quality standards and records that says Structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed. The definition of the codes, standards and technical requirements applied to the nuclear reactor design, fabrication, inspection and tests may be seen as the main result from this statement. There are well established guides to classify structures, systems and components for nuclear power reactors such as the Pressurized Water Reactors but one can not say the same for nuclear research and test reactors. The nuclear reactors safety functions are those required to the safe reactor operation, the safe reactor shutdown and continued safe conditions, the response to anticipated transients, the response to potential accidents and the control of radioactive material. So, it is proposed in this paper an approach to develop the classification of structures, systems and components of these reactors based on their intended safety functions in order to define the applicable set of codes, standards and technical requirements. (author)

  17. Safety philosophy and design principles for systems and components of nuclear power plant: external event

    International Nuclear Information System (INIS)

    Lopes, J.P.G.

    1986-01-01

    In nuclear power plants, some systems and components are designed to withstand external impacts. Such systems and components are those which have to perform their functions even during and after the occurrences of an earthquake, for example, fulfilling the safety objectives and avoiding the release of radioactive material to the environment. The aim of this report is to introduce the safety philosophy and design principles for systems/components to perform their functions during and after the occurrence of an earthquake, as applied by NUCLEN for Angra 2 and 3. (Author) [pt

  18. Experimental Study of Nuclear Security System Components for Achieving the Intrusion Process via Sensor's Network System

    International Nuclear Information System (INIS)

    EL-Kafas, A.A.

    2011-01-01

    Cluster sensors are one of nuclear security system components which are used to detect any intrusion process of the nuclear sites. In this work, an experimental measuring test for sensor performance and procedures are presented. Sensor performance testing performed to determine whether a particular sensor will be acceptable in a proposed design. We have access to a sensors test field in which the sensor of interest is already properly installed and the parameters have been set to optimal levels by preliminary testing. The glass-breakage (G.B) and open door (O.D) sensors construction, operation and design for the investigated nuclear site are explained. Intrusion tests were carried out inside the field areas of the sensors to evaluate the sensor performance during the intrusion process. Experimental trials were performed for achieving the intrusion process via sensor network system. The performance and intrusion senses of cluster sensors inside the internal zones was recorded and evaluated. The obtained results explained that the tested and experimented G.B sensors have a probability of detection P (D) value 65% founded, and 80% P (D) of Open-door sensor

  19. Reliability analysis of nuclear component cooling water system using semi-Markov process model

    International Nuclear Information System (INIS)

    Veeramany, Arun; Pandey, Mahesh D.

    2011-01-01

    Research highlights: → Semi-Markov process (SMP) model is used to evaluate system failure probability of the nuclear component cooling water (NCCW) system. → SMP is used because it can solve reliability block diagram with a mixture of redundant repairable and non-repairable components. → The primary objective is to demonstrate that SMP can consider Weibull failure time distribution for components while a Markov model cannot → Result: the variability in component failure time is directly proportional to the NCCW system failure probability. → The result can be utilized as an initiating event probability in probabilistic safety assessment projects. - Abstract: A reliability analysis of nuclear component cooling water (NCCW) system is carried out. Semi-Markov process model is used in the analysis because it has potential to solve a reliability block diagram with a mixture of repairable and non-repairable components. With Markov models it is only possible to assume an exponential profile for component failure times. An advantage of the proposed model is the ability to assume Weibull distribution for the failure time of components. In an attempt to reduce the number of states in the model, it is shown that usage of poly-Weibull distribution arises. The objective of the paper is to determine system failure probability under these assumptions. Monte Carlo simulation is used to validate the model result. This result can be utilized as an initiating event probability in probabilistic safety assessment projects.

  20. Development of web-based integrity evaluation system for primary components in a nuclear power plant

    International Nuclear Information System (INIS)

    Lee, S.M.; Kim, J.C.; Choi, J.B.; Kim, Y.J.; Choi, S.N.; Jang, K.S.; Hong, S.Y.

    2004-01-01

    A nuclear power plant is composed of a number of primary components. Maintaining the integrity of these components is one of the most critical issues in nuclear industry. In order to maintain the integrity of these primary components, a complicated procedure is required including periodical in-service inspection, failure assessment, fracture mechanics analysis, etc. Also, experts in different fields have to co-operate to resolve the integrity issues on the basis of inspection results. This integrity evaluation process usually takes long, and thus, is detrimental for the plant productivity. Therefore, an effective safety evaluation system is essential to manage integrity issues on a nuclear power plant. In this paper, a web-based integrity evaluation system for primary components in a nuclear power plant is proposed. The proposed system, which is named as WEBIES (web-based integrity evaluation system), has been developed in the form of 3-tier system architecture. The system consists of three servers; application program server, user interface program server and data warehouse server. The application program server includes the defect acceptance analysis module and the fracture mechanics analysis module which are programmed on the basis of ASME sec. XI, appendix A. The data warehouse server provides data for the integrity evaluation including material properties, geometry information, inspection data and stress data. The user interface program server provides information to all co- workers in the field of integrity evaluation. The developed system provides engineering knowledge-based information and concurrent and collaborative working environment through internet, and thus, is expected to raise the efficiency of integrity evaluation procedures on primary components of a nuclear power plant. (orig.)

  1. Development of web-based integrity evaluation system for primary components in a nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Lee, S.M.; Kim, J.C.; Choi, J.B.; Kim, Y.J. [SAFE Research Center, Sungkyunkwan Univ., Suwon (Korea); Choi, S.N.; Jang, K.S.; Hong, S.Y. [Korea Electronic Power Research Inst., Daejeon (Korea)

    2004-07-01

    A nuclear power plant is composed of a number of primary components. Maintaining the integrity of these components is one of the most critical issues in nuclear industry. In order to maintain the integrity of these primary components, a complicated procedure is required including periodical in-service inspection, failure assessment, fracture mechanics analysis, etc. Also, experts in different fields have to co-operate to resolve the integrity issues on the basis of inspection results. This integrity evaluation process usually takes long, and thus, is detrimental for the plant productivity. Therefore, an effective safety evaluation system is essential to manage integrity issues on a nuclear power plant. In this paper, a web-based integrity evaluation system for primary components in a nuclear power plant is proposed. The proposed system, which is named as WEBIES (web-based integrity evaluation system), has been developed in the form of 3-tier system architecture. The system consists of three servers; application program server, user interface program server and data warehouse server. The application program server includes the defect acceptance analysis module and the fracture mechanics analysis module which are programmed on the basis of ASME sec. XI, appendix A. The data warehouse server provides data for the integrity evaluation including material properties, geometry information, inspection data and stress data. The user interface program server provides information to all co- workers in the field of integrity evaluation. The developed system provides engineering knowledge-based information and concurrent and collaborative working environment through internet, and thus, is expected to raise the efficiency of integrity evaluation procedures on primary components of a nuclear power plant. (orig.)

  2. Air-conditioning and ventilation systems and components of nuclear facilities

    International Nuclear Information System (INIS)

    2006-01-01

    The Guide defines the requirements for the design, implementation and operation of the air-conditioning and ventilation systems of nuclear facilities belonging to safety classes 3 and 4, and for the related documents to be submitted to STUK (Radiation and Nuclear Safety Authority, Finland). Furthermore, the Guide describes the inspections of air-conditioning and ventilation systems to be conducted by STUK during construction and operation of the facilities. As far as systems and components belonging to safety class 2 are concerned, STUK sets additional requirements case by case. In general, air-conditioning systems refer to systems designed to manage the indoor air cleanness, temperature, humidity and movement. In some rooms of a nuclear power plant, ventilation systems are also used to prevent radioactive materials from spreading outside the rooms. Guide YVL1.0 defines the safety principles concerning the air-conditioning and ventilation of nuclear power plants. Guide YVL2.0 gives the requirements for the design of nuclear power plant systems. In addition, YVLGuide groups 3, 4, 5 and 7 deal with the requirements for air-conditioning and ventilation systems with regard to the mechanical equipment, fire prevention, electrical systems, instrumentation and control technology, and the restriction of releases. The rules and regulations issued by the Ministry of the Environment and the Ministry of the Interior (RakMK, the Finnish building code) concerning the design and operation of air-conditioning and ventilation systems and the related fire protection design bases also apply to nuclear facilities. Exhaust gas treatment systems, condenser vacuum systems of boiling water reactor plants and leak collection systems are excluded from the scope of this Guide

  3. Advancements in the design of safety-related systems and components of the MARS nuclear plant

    International Nuclear Information System (INIS)

    Caira, M.; Caruso, G.; Naviglio, A.; Sorabella, L.; Farello, C.E.

    1992-01-01

    In the paper, the advancements in the design of safety-related systems and components of the MARS nuclear plant, equipped with a 600 MW th PWR, are described. These advancements are due to the special safety features of this plant, which relies completely on inherent and passive safety. In particular, the new steps of the design of the innovative, completely passive, and with an unlimited autonomy Emergency core Cooling System are described, together with the characteristics of the last version of the steam generator, developed in a new design involving disconnecting components, for a fast erection and an easy maintenance. (author)

  4. Nuclear power plant systems, structures and components and their safety classification

    International Nuclear Information System (INIS)

    2000-01-01

    The assurance of a nuclear power plant's safety is based on the reliable functioning of the plant as well as on its appropriate maintenance and operation. To ensure the reliability of operation, special attention shall be paid to the design, manufacturing, commissioning and operation of the plant and its components. To control these functions the nuclear power plant is divided into structural and functional entities, i.e. systems. A systems safety class is determined by its safety significance. Safety class specifies the procedures to be employed in plant design, construction, monitoring and operation. The classification document contains all documentation related to the classification of the nuclear power plant. The principles of safety classification and the procedures pertaining to the classification document are presented in this guide. In the Appendix of the guide, examples of systems most typical of each safety class are given to clarify the safety classification principles

  5. Maintenance Management Support Systems for component aging estimation at nuclear power plants

    International Nuclear Information System (INIS)

    Shimizu, Shunichi; Ando, Yasumasa; Morioka, Toshihiko; Okuzumi, Naoaki

    1991-01-01

    Maintenance Management Support Systems (MMSSs) for nuclear power plants have been developed using component aging estimation methods and decision tree analysis for maintenance planning. The former evaluates actual component reliability through statistical analysis on field maintenance data. The latter provides preventive maintenance (PM) planning guidance using heuristic expert knowledge and estimated reliability parameters. The following aspects have been investigated: (1) A systematic and effective method of managing components/parts design information and field maintenance data (2) A method for estimating component aging based on a statistical analysis of field maintenance data (3) A method for providing PM planning guidance using estimated component reliability/performance parameters and decision tree analysis. Based on these investigations, two MMSSs were developed. One deals with 'general maintenance data', which are common to all component types and are amenable to common data handling. The other system deals with 'specific maintenance data', which are specific to an individual component type. Both systems provide PM planning guidance for PM cycles propriety and the PM work priority. The function of these systems were verified using simulated maintenance data. (author)

  6. Assessment and management of ageing of major nuclear power plant components important to safety: Metal components of BWR containment systems

    International Nuclear Information System (INIS)

    2000-10-01

    At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of the major NPP components (e.g. caused by unanticipated phenomena and by operating, maintenance or manufacturing errors) can jeopardize plant safety and also plant life. Ageing in these NPPs must therefore be effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling within acceptable limits the ageing degradation and wear-out of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This TECDOC is one in a series of reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers, and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety, which was issued by the IAEA in 1992. The current practices for the assessment of safety margins (fitness for service) and the inspection, monitoring and mitigation of ageing degradation of selected components of Canada deuterium-uranium (CANDU) reactors, boiling water reactors (BWRs), pressurized water reactors (PWRs), and water moderated, water cooled energy reactors (WWERs) are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs, and to provide a common technical basis for dialogue between plant operators and regulators when dealing with age related licensing issues. The guidance reports are directed toward technical experts from NPPs and from regulatory, plant design, manufacturing and technical support organizations dealing with specific

  7. Fabricating nuclear components

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    Activities of the Nuclear Engineering Division of Vickers Ltd., particularly fabrication of long slim tubular components for power reactors and the construction of irradiation loops and rigs, are outlined. The processes include hydraulic forming for fabrication of various types of tubes and outer cases of fuel transfer buckets, various specialised welding operations including some applications of the TIG process, and induction brazing of specialised assemblies. (U.K.)

  8. Views on seismic design standardization of structures, systems and components of nuclear facilities

    International Nuclear Information System (INIS)

    Reddy, G.R.

    2011-01-01

    Structures, Systems and Components (SSCs) of nuclear facilities have to be designed for normal operating loads such as dead weight, pressure, temperature etc., and accidental loads such as earthquakes, floods, extreme, wind air craft impact, explosions etc. Manmade accidents such as aircraft impact, explosions etc., sometimes may be considered as design basis event and sometimes taken care by providing administrative controls. This will not be possible in the case of natural events such as earthquakes, flooding, extreme winds etc. Among natural events earthquakes are considered as most devastating and need to be considered as design basis event which has certain annual frequency specified in design codes. For example nuclear power plants are designed for a seismic event has 10000 year return period. It is generally felt that design of SSCs for earthquake loads is very time consuming and expensive. Conventional seismic design approaches demands for large number of supports for systems and components. This results in large space occupation and in turn creates difficulties for maintenance and in service inspection of systems and components. In addition, complete exercise of design need to be repeated for plants being located at different sites due to different seismic demands. However, advanced seismic response control methods will help to standardize the seismic design meeting the safety and economy. These methods adopt passive, semi active and active devices, and base isolators to control the seismic response. In nuclear industry, it is advisable to go for passive devices to control the seismic responses. Ideally speaking, these methods will make the designs made for normal loads can also satisfy the seismic demand without calling for change in material, geometry, layout etc. in the SSCs. This paper explain the basic ideas of seismic response control methods, demonstrate the effectiveness of control methods through case studies and eventually give the procedure to

  9. Component nuclear containment structure

    International Nuclear Information System (INIS)

    Harstead, G.A.

    1979-01-01

    The invention described is intended for use primarily as a nuclear containment structure. Such structures are required to surround the nuclear steam supply system and to contain the effects of breaks in the nuclear steam supply system, or i.e. loss of coolant accidents. Nuclear containment structures are required to withstand internal pressure and temperatures which result from loss of coolant accidents, and to provide for radiation shielding during operation and during the loss of coolant accident, as well as to resist all other applied loads, such as earthquakes. The nuclear containment structure described herein is a composite nuclear containment structure, and is one which structurally combines two previous systems; namely, a steel vessel, and a lined concrete structure. The steel vessel provides strength to resist internal pressure and accommodate temperature increases, the lined concrete structure provides resistance to internal pressure by having a liner which will prevent leakage, and which is in contact with the concrete structure which provides the strength to resist the pressure

  10. Lifetime management for mechanical systems, structures and components in nuclear power plants

    International Nuclear Information System (INIS)

    Roos, E.; Herter, K.-H.; Schuler, X.

    2006-01-01

    Guidelines, codes and standards contain regulations and requirements with respect to the quality of mechanical systems, structures and components (SSC) of nuclear power plants. These concern safe operation during the total lifetime (lifetime management), safety against ageing phenomena (ageing management) as well as proof of integrity (e.g. break exclusion or avoidance of fracture). Within this field the ageing management is a key element. Depending on the safety-relevance of the SSC under observation including preventive maintenance various tasks are required in particular to clarify the mechanisms which contribute system-specifically to the damage of the components and systems and to define their controlling parameters which have to be monitored and checked. Appropriate continuous or discontinuous measures are to be considered in this connection. The approach to ensure a high standard of quality in operation and the management of the technical and organisational aspects are demonstrated and explained

  11. Use of expert systems in the structural safety assessment of of pressurized nuclear components

    International Nuclear Information System (INIS)

    Jovanovic, A.; Sturm, D.

    1990-01-01

    The paper describes research currently performed at MPA Stuttgart on development of expert systems and application of artificial intelligence methods and techniques, for structural safety assessment of power plant pressurized components. The research is done as an extension of preceding and existing large research programs of MPA, in the domain of structural safety of components. In this preceding research a waste amount of practical engineering knowledge and experience has been accumulated: development in the direction of AI-based systems is a way to use this knowledge more efficiently in future research and in the nuclear power plant practice. Applications on which the current research is focussed are expert systems applied for the leak-before-break analysis for the structural safety evaluation in high temperature regimes

  12. Multi-component Self-Consistent Nuclear Energy System: On proliferation resistance aspect

    International Nuclear Information System (INIS)

    Shmelev, A.; Saito, M; Artisyuk, V.

    2000-01-01

    Self-Consistent Nuclear Energy System (SCNES) that simultaneously meets four requirements: energy production, fuel production, burning of radionuclides and safety is targeted at harmonization of nuclear energy technology with human environment. The main bulk of SCNES studies focus on a potential of fast reactor (FR) in generating neutron excess to keep suitable neutron balance. Proliferation resistance was implicitly anticipated in a fuel cycle with co-processing of Pu, minor actinides (MA) and some relatively short-lived fission products (FP). In a contrast to such a mono-component system, the present paper advertises advantage of incorporating accelerator and fusion driven neutron sources which could drastically improve characteristics of nuclear waste incineration. What important is that they could help in creating advanced Np and Pa containing fuels with double protection against uncontrolled proliferation. The first level of protection deals with possibility to approach long life core (LLC) in fission reactors. Extending the core life-time to reactor-time is beneficial from the proliferation resistance viewpoint since LLC would not necessarily require fuel management at energy producing site, with potential advantage of being moved to vendor site for spent fuel refabrication. Second level is provided by the presence of substantial amounts of 238 Pu and 232 U in these fuels that makes fissile nuclides in them isotopically protected. All this reveals an important advantage of a multi-component SCNES that could draw in developing countries without elaborated technological infrastructure. (author)

  13. Nuclear Plant Aging Research (NPAR) program plan: Components, systems, and structures

    International Nuclear Information System (INIS)

    1987-09-01

    The nuclear plant aging research described in this plan is intended to resolve issues related to the aging and service wear of equipment and systems and major components at commercial reactor facilities and their possible impact on plant safety. Emphasis has been placed on identification and characterization of the mechanisms of material and component degradation during service and evaluation of methods of inspection, surveillance, condition monitoring, and maintenance as means of mitigating such effects. Specifically, the goals of the program are as follows: (1) to identify and characterize aging and service wear effects which, if unchecked, could cause degradation of equipment, a systems, and major components and thereby impair plant safety; (2) to identify methods of inspection, surveillance, and monitoring, or of evaluating residual life of equipment, systems, and major components, which will ensure timely detection of significant aging effects prior to loss of safety function; and (3) to evaluate the effectiveness of storage, maintenance, repair, and replacement practices in mitigating the rate and extent of degradation caused by aging and service wear

  14. Computer-aided stress analysis system for nuclear plant primary components

    International Nuclear Information System (INIS)

    Murai, Tsutomu; Tokumaru, Yoshio; Yamazaki, Junko.

    1980-06-01

    Generally it needs a vast quantity of calculation to make the stress analysis reports of nuclear plant primary components. In Japan, especially, stress analysis reports are under obligation to make for each plant. In Mitsubishi Heavy Industries, Ltd., We have been making great efforts to rationalize the process of analysis for about these ten years. As the result of rationalization up to now, a computer-aided stress analysis system using graphic display, graphic tablet, data file, etc. was accomplished and it needs us only the least hand work. In addition we developed a fracture safety analysis system. And we are going to develop the input generator system for 3-dimensional FEM analysis by graphics terminals in the near future. We expect that when the above-mentioned input generator system is accomplished, it will be possible for us to solve instantly any case of problem. (author)

  15. Advanced Surveillance, Diagnostic and Prognostic Techniques in Monitoring Structures, Systems and Components in Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-09-15

    direction of research, development and demonstration in this area. The technologies discussed in this project are intended to establish the state of the art in surveillance, diagnostics and prognostics (SDP) technologies for equipment and process health monitoring in nuclear facilities. It is also intended to identify technology gaps and research needs of the nuclear industry in the area of SDP. The report draws on the conventional SDP technologies, as well as the latest tools, algorithms and techniques that have emerged over the last few years, especially in enabling technologies including fast data acquisition, data storage, data qualification and data analysis algorithms, such as empirical and physical modelling techniques. These new tools have made it possible to identify problems earlier and with better resolution. The significance of the material presented in this report is that it contributes not only to the current needs of the nuclear industry but also to the design improvements of the next generation of reactors. For example, the nuclear industry is currently striving to operate the plants for up to 80 years or more, as the value of nuclear assets has risen in recent years, resulting partly from environmental concerns with fossil energy production, as well as increased future demand for base load electricity. This long term operation (LTO) or life extension goal of the nuclear industry has stimulated renewed interest in more frequent monitoring of equipment to guard against ageing effects, not to mention the economic benefits that SDP implementation can produce, and contributions to radiation exposure that is as low as reasonably achievable, reduction of human errors, and optimized maintenance. Together with capabilities that enhance situational awareness, the technologies described in this report will enable more holistic management of plant structures, systems and components (SSCs), maintain high capacity factor in LTO and enable higher levels of safe operation

  16. Advanced Surveillance, Diagnostic and Prognostic Techniques in Monitoring Structures, Systems and Components in Nuclear Power Plants

    International Nuclear Information System (INIS)

    2013-01-01

    direction of research, development and demonstration in this area. The technologies discussed in this project are intended to establish the state of the art in surveillance, diagnostics and prognostics (SDP) technologies for equipment and process health monitoring in nuclear facilities. It is also intended to identify technology gaps and research needs of the nuclear industry in the area of SDP. The report draws on the conventional SDP technologies, as well as the latest tools, algorithms and techniques that have emerged over the last few years, especially in enabling technologies including fast data acquisition, data storage, data qualification and data analysis algorithms, such as empirical and physical modelling techniques. These new tools have made it possible to identify problems earlier and with better resolution. The significance of the material presented in this report is that it contributes not only to the current needs of the nuclear industry but also to the design improvements of the next generation of reactors. For example, the nuclear industry is currently striving to operate the plants for up to 80 years or more, as the value of nuclear assets has risen in recent years, resulting partly from environmental concerns with fossil energy production, as well as increased future demand for base load electricity. This long term operation (LTO) or life extension goal of the nuclear industry has stimulated renewed interest in more frequent monitoring of equipment to guard against ageing effects, not to mention the economic benefits that SDP implementation can produce, and contributions to radiation exposure that is as low as reasonably achievable, reduction of human errors, and optimized maintenance. Together with capabilities that enhance situational awareness, the technologies described in this report will enable more holistic management of plant structures, systems and components (SSCs), maintain high capacity factor in LTO and enable higher levels of safe operation

  17. Time dependent unavailability analysis of nuclear safety systems considering periodically tested components

    International Nuclear Information System (INIS)

    Goes, Alexandre Gromann de Araujo

    1988-01-01

    It is of utmost importance to have a computer code in order to analyze how different parameters (like test duration time) affect the unavailability of safety systems of nuclear. In this context, a study was performed in order to evaluate the model employed by the FRANTIC computer code, which performs detailed calculations on the contribution to the system unavailability originated by hardware failures, component tests and repairs, aiming at considering the influence of different test schemes on the system unavailability. It was shown, by means of the results attained that the numerical model used by the FRANTIC code and the analytical model proposed by APOSTOLAKIS and CHU (4) give unavailability values much similar when the component tests are supposed to be perfect. When a test is supposed to be imperfect (that is, when it may induce a test is supposed to be imperfect (that is, when it may induce a failure on the component being tested), the analytical model presents more conservative results. (author)

  18. Nuclear plant reliability data system. 1979 annual reports of cumulative system and component reliability

    International Nuclear Information System (INIS)

    1979-01-01

    The primary purposes of the information in these reports are the following: to provide operating statistics of safety-related systems within a unit which may be used to compare and evaluate reliability performance and to provide failure mode and failure rate statistics on components which may be used in failure mode effects analysis, fault hazard analysis, probabilistic reliability analysis, and so forth

  19. 10 CFR 50.69 - Risk-informed categorization and treatment of structures, systems and components for nuclear...

    Science.gov (United States)

    2010-01-01

    ..., systems and components for nuclear power reactors. (a) Definitions. Risk-Informed Safety Class (RISC)-1... functions. Risk-Informed Safety Class (RISC)-2 structures, systems and components (SSCs) means nonsafety-related SSCs that perform safety significant functions. Risk-Informed Safety Class (RISC)-3 structures...

  20. Fusion technology development: first wall/blanket system and component testing in existing nuclear facilities

    International Nuclear Information System (INIS)

    Hsu, P.Y.S.; Bohn, T.S.; Deis, G.A.; Judd, J.L.; Longhurst, G.R.; Miller, L.G.; Millsap, D.A.; Scott, A.J.; Wessol, D.E.

    1980-12-01

    A novel concept to produce a reasonable simulation of a fusion first wall/blanket test environment employing an existing nuclear facility, the Engineering Test Reactor at the Idaho National Engineering Laboratory, is presented. Preliminary results show that an asymmetric, nuclear test environment with surface and volumetric heating rates similar to those expected in a fusion first wall/blanket or divertor chamber surface appears feasible. The proposed concept takes advantage of nuclear reactions within the annulus of an existing test space (15 cm in diameter and approximately 100 cm high) to provide an energy flux to the surface of a test module. The principal reaction considered involves 3 He in the annulus as follows: n + 3 He → p + t + 0.75 MeV. Bulk heating in the test module is accomplished by neutron thermalization, gamma heating, and absorption reactions involving 6 Li in the blanket breeding region. The concept can be extended to modified core configurations that will accommodate test modules of different sizes and types. It makes possible development testing of first wall/blanket systems and other fusion components on a scale and in ways not otherwise available until actual high-power fusion reactors are built

  1. Nuclear fuel cycle system simulation tool based on high-fidelity component modeling

    Energy Technology Data Exchange (ETDEWEB)

    Ames, David E.,

    2014-02-01

    The DOE is currently directing extensive research into developing fuel cycle technologies that will enable the safe, secure, economic, and sustainable expansion of nuclear energy. The task is formidable considering the numerous fuel cycle options, the large dynamic systems that each represent, and the necessity to accurately predict their behavior. The path to successfully develop and implement an advanced fuel cycle is highly dependent on the modeling capabilities and simulation tools available for performing useful relevant analysis to assist stakeholders in decision making. Therefore a high-fidelity fuel cycle simulation tool that performs system analysis, including uncertainty quantification and optimization was developed. The resulting simulator also includes the capability to calculate environmental impact measures for individual components and the system. An integrated system method and analysis approach that provides consistent and comprehensive evaluations of advanced fuel cycles was developed. A general approach was utilized allowing for the system to be modified in order to provide analysis for other systems with similar attributes. By utilizing this approach, the framework for simulating many different fuel cycle options is provided. Two example fuel cycle configurations were developed to take advantage of used fuel recycling and transmutation capabilities in waste management scenarios leading to minimized waste inventories.

  2. Multi-component nuclear energy system to meet requirement of self-consistency

    International Nuclear Information System (INIS)

    Saito, Masaki; Artisyuk, Vladimir; Shmelev, Anotolii; Korovin, Yorii

    2000-01-01

    Environmental harmonization of nuclear energy technology is considered as an absolutely necessary condition in its future successful development for peaceful use. Establishment of Self-Consistent Nuclear Energy System, that simultaneously meets four requirements - energy production, fuel production, burning of radionuclides and safety, strongly relies on the neutron excess generation. Implementation of external non-fission based neutron sources into fission energy system would open the possibility of approaching Multicomponent Self-Consistent Nuclear Energy System with unlimited fuel resources, zero radioactivity release and high protection against uncontrolled proliferation of nuclear materials. (author)

  3. Case study on the use of PSA methods: Determining safety importance of systems and components at nuclear power plants

    International Nuclear Information System (INIS)

    1991-04-01

    This case study emphasizes the step of probabilistic safety assessment (PSA) regarding identification of systems and components important to nuclear plant safety. An importance analysis involves combining information that is both qualitative and probabilistic in nature to generate a numerical ranking to determine the system and/or component failures that dominate the risk. Such a ranking can suggest where hardware, software, human factors and component design changes can be implemented to improve plant safety. Examples of using ranking methodology are described. A qualitative ranking criteria is discussed for components and systems that are not included in a PSA. 18 refs, 7 figs, 18 tabs

  4. Development of an integrated database management system to evaluate integrity of flawed components of nuclear power plant

    International Nuclear Information System (INIS)

    Mun, H. L.; Choi, S. N.; Jang, K. S.; Hong, S. Y.; Choi, J. B.; Kim, Y. J.

    2001-01-01

    The object of this paper is to develop an NPP-IDBMS(Integrated DataBase Management System for Nuclear Power Plants) for evaluating the integrity of components of nuclear power plant using relational data model. This paper describes the relational data model, structure and development strategy for the proposed NPP-IDBMS. The NPP-IDBMS consists of database, database management system and interface part. The database part consists of plant, shape, operating condition, material properties and stress database, which are required for the integrity evaluation of each component in nuclear power plants. For the development of stress database, an extensive finite element analysis was performed for various components considering operational transients. The developed NPP-IDBMS will provide efficient and accurate way to evaluate the integrity of flawed components

  5. Effects of composition on properties in an 11-component nuclear waste glass system

    International Nuclear Information System (INIS)

    Chick, L.A.; Piepel, G.F.; Mellinger, G.B.; May, R.P.; Gray, W.J.; Buckwalter, C.Q.

    1981-09-01

    Ninety simplified nuclear waste glass compositions within an 11-component oxide composition matrix were tested for crystallinity, viscosity, volatility, and chemical durability. Empirical models of property response as a function of glass composition were developed using statistical experimental design and modeling techniques. A new statistical technique was developed to calculate the effects of oxide components on each property. Independent melts were used to check the prediction accuracy of the models

  6. Requirements for class 1, 2, and 3 pressure-retaining systems and components in CANDU nuclear power plants

    International Nuclear Information System (INIS)

    1991-09-01

    This third edition of CAN/CSA-N285.1 supersedes the 1981 and 1975 editions. It provides the specific requirements for design, fabrication, and installation of Class 1, 2 and 3 pressure-retaining systems and components in CANDU nuclear power plants, and over pressure protection of the heat transport system. The general requirements for pressure-retaining systems and components are given in CSA Standard CAN/CSA-N285.0, with which Class 1, 2 and 3 systems and components must also comply

  7. Development of a web-based fatigue life evaluation system for primary components in a nuclear power plant

    International Nuclear Information System (INIS)

    Seo, Hyong Won; Lee, Sang Min; Choi, Jae Boong; Kim, Young Jin; Choi, Sung Nam; Jang, Ki Sang; Hong, Sung Yull

    2004-01-01

    A nuclear power plant is composed of a number of primary components. Maintaining the integrity of these components is one of the most critical issues in nuclear industry. In order to maintain the integrity of these primary components, a complicated procedure is required including regular in-service inspection, failure assessment, fracture mechanics analysis, etc. Also, experts in different fields have to co-operate to resolve the integrity issues on the basis of inspection results. This integrity evaluation process usually takes long, and thus, is detrimental for the plant productivity. Therefore, an effective safety evaluation system is essential to manage the integrity issues on a nuclear power plant. In this paper, a web-based fatigue life evaluation system for primary components in nuclear power plant is proposed. This system provides engineering knowledge-based information and concurrent and collaborative working environment through internet, and thus, is expected to raise the efficiency of integrity evaluation procedures on primary components of a nuclear power plant

  8. Developments of integrity evaluation technology for pressurized components in nuclear power plant and IT based integrity evaluation system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Jin; Choi, Jae Boong; Shim, Do Jun [Sungkyunkwan Univ., Seoul (Korea, Republic of)] (and others)

    2003-03-15

    The objective of this research is to develop an efficient evaluation technology and to investigate applicability of newly-developed technology, such as internet-based cyber platform, to operating power plants. Development of efficient evaluation systems for Nuclear Power Plant components, based on structural integrity assessment techniques, are increasingly demanded for safe operation with the increasing operating period of Nuclear Power Plants. The following five topics are covered in this project: development of assessment method for wall-thinned nuclear piping based on pipe test; development of structural integrity program for steam generator tubes with cracks of various shape; development of fatigue life evaluation system for mam components of NPP; development of internet-based cyber platform and integrity program for primary components of NPP; effect of aging on strength of dissimilar welds.

  9. Automated ultrasonic inspection of nuclear plant components

    International Nuclear Information System (INIS)

    Baron, J.A.; Dolbey, M.P.

    1982-01-01

    For reasons of safety and efficiency, automated systems are used in performing ultrasonic inspection of nuclear components. An automated system designed specifically for the inspection of headers in a nuclear plant is described. In-service inspection results obtained with this system are shown to correlate with pre-service inspection results obtained by manual methods

  10. Development of integrity evaluation technology for pressurized components in nuclear power plant and IT based integrity evaluation system

    International Nuclear Information System (INIS)

    Kim, Young Jin; Choi, Jae Boong; Shim, Do Jun

    2004-02-01

    The objective of this research is to develop on efficient integrity evaluation technology and to investigate the applicability of the newly-developed technology such as internet-based cyber platform etc. to Nuclear Power Plant(NPP) components. The development of an efficient structural integrity evaluation system is necessary for safe operation of NPP as the increase of operating periods. Moreover, material test data as well as emerging structural integrity assessment technology are also needed for the evaluation of aged components. The following five topics are covered in this project: development of the wall-thinning evaluation program for nuclear piping; development of structural integrity evaluation criteria for steam generator tubes with cracks of various shape; development of fatigue life evaluation system for major components of NPP; ingegration of internet-based cyber platform and integrity evaluation program for primary components of NPP; effects of aging on strength of dissimilar welds

  11. Development of integrity evaluation technology for pressurized components in nuclear power plant and IT based integrity evaluation system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Jin; Choi, Jae Boong; Shim, Do Jun [Sungkyunkwan Univ., Seoul (Korea, Republic of)] (and others)

    2004-02-15

    The objective of this research is to develop on efficient integrity evaluation technology and to investigate the applicability of the newly-developed technology such as internet-based cyber platform etc. to Nuclear Power Plant(NPP) components. The development of an efficient structural integrity evaluation system is necessary for safe operation of NPP as the increase of operating periods. Moreover, material test data as well as emerging structural integrity assessment technology are also needed for the evaluation of aged components. The following five topics are covered in this project: development of the wall-thinning evaluation program for nuclear piping; development of structural integrity evaluation criteria for steam generator tubes with cracks of various shape; development of fatigue life evaluation system for major components of NPP; ingegration of internet-based cyber platform and integrity evaluation program for primary components of NPP; effects of aging on strength of dissimilar welds.

  12. EBaLM-THP - A neural network thermohydraulic prediction model of advanced nuclear system components

    International Nuclear Information System (INIS)

    Ridluan, Artit; Manic, Milos; Tokuhiro, Akira

    2009-01-01

    In lieu of the worldwide energy demand, economics and consensus concern regarding climate change, nuclear power - specifically near-term nuclear power plant designs are receiving increased engineering attention. However, as the nuclear industry is emerging from a lull in component modeling and analyses, optimization for example using ANN has received little research attention. This paper presents a neural network approach, EBaLM, based on a specific combination of two training algorithms, error-back propagation (EBP), and Levenberg-Marquardt (LM), applied to a problem of thermohydraulics predictions (THPs) of advanced nuclear heat exchangers (HXs). The suitability of the EBaLM-THP algorithm was tested on two different reference problems in thermohydraulic design analysis; that is, convective heat transfer of supercritical CO 2 through a single tube, and convective heat transfer through a printed circuit heat exchanger (PCHE) using CO 2 . Further, comparison of EBaLM-THP and a polynomial fitting approach was considered. Within the defined reference problems, the neural network approach generated good results in both cases, in spite of highly fluctuating trends in the dataset used. In fact, the neural network approach demonstrated cumulative measure of the error one to three orders of magnitude smaller than that produce via polynomial fitting of 10th order

  13. Identification of seismically risk-sensitive systems and components in nuclear power plants: feasibility study

    International Nuclear Information System (INIS)

    Azarm, M.; Boccio, J.; Farahzad, P.

    1983-06-01

    An approach for the identification of risk-sensitive components in a nuclear power plant during and after a seismic event is described. Application of the methodology to two hypothetical power plants - a Boiling Water Reactor and a Pressurized Water Reactor - are presented and the results are given in tabular and graphical form. Conclusions drawn and lessons learned through the course of this study, based on the relative importance of various accident scenarios and sensitivity analyses, are discussed. In addition, the areas that may need further investigation are identified

  14. High temperature brazing of primary-system components in the nuclear field

    International Nuclear Information System (INIS)

    Belicic, M.; Fricker, H.W.; Iversen, K.; Leukert, W.

    1981-01-01

    Apart from the well-known welding procedures, high-temperature brazing is successfully applied in the manufacture of primary components in the field of nuclear reactor construction. This technique is applied in all cases where apart from sufficient resistance and high production safety importance is laid on dimensional stability without subsequent mechanical processing of the components. High-temperature brazing is therefore very important in the manufacture of fuel rod spacers or control rod guide tubes. In this context, during one brazing process many brazing seams have to be produced in extremely narrow areas and within small tolerances. As basic materials precipitation hardening alloys with a high nickel percentage, austenitic Cr-Ni-steels or the zirconium alloy Zry 4 are used. Generally applied are: boron free nickel or zirconium brazing filler metals. (orig.)

  15. Modeling Chilled-Water Storage System Components for Coupling to a Small Modular Reactor in a Nuclear Hybrid Energy System

    Science.gov (United States)

    Misenheimer, Corey Thomas

    The intermittency of wind and solar power puts strain on electric grids, often forcing carbonbased and nuclear sources of energy to operate in a load-follow mode. Operating nuclear reactors in a load-follow fashion is undesirable due to the associated thermal and mechanical stresses placed on the fuel and other reactor components. Various Thermal Energy Storage (TES) elements and ancillary energy applications can be coupled to nuclear (or renewable) power sources to help absorb grid instabilities caused by daily electric demand changes and renewable intermittency, thereby forming the basis of a candidate Nuclear Hybrid Energy System (NHES). During the warmer months of the year in many parts of the country, facility air-conditioning loads are significant contributors to the increase in the daily peak electric demand. Previous research demonstrated that a stratified chilled-water storage tank can displace peak cooling loads to off-peak hours. Based on these findings, the objective of this work is to evaluate the prospect of using a stratified chilled-water storage tank as a potential TES reservoir for a nuclear reactor in a NHES. This is accomplished by developing time-dependent models of chilled-water system components, including absorption chillers, cooling towers, a storage tank, and facility cooling loads appropriate for a large office space or college campus, as a callable FORTRAN subroutine. The resulting TES model is coupled to a high-fidelity mPower-sized Small Modular Reactor (SMR) Simulator, with the goal of utilizing excess reactor capacity to operate several sizable chillers in order to keep reactor power constant. Chilled-water production via single effect, lithium bromide (LiBr) absorption chillers is primarily examined in this study, although the use of electric chillers is briefly explored. Absorption chillers use hot water or low-pressure steam to drive an absorption-refrigeration cycle. The mathematical framework for a high-fidelity dynamic

  16. Operating experiences with passive systems and components in German nuclear power plants

    International Nuclear Information System (INIS)

    Maqua, M.

    1996-01-01

    Operating experience with passive systems and components is limited to the equipment installed in existing NPPs. In German power plants, this experience is available for equipment of the IAEA categories A, C and D. The presentation is focused on typical examples out of these three categories. An overview is given on the number of reported events and typical failure modes. Selected failures are discussed in detail. 1 ref., 6 figs, 7 tabs

  17. Operating experiences with passive systems and components in German nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Maqua, M [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Koeln (Germany)

    1996-12-01

    Operating experience with passive systems and components is limited to the equipment installed in existing NPPs. In German power plants, this experience is available for equipment of the IAEA categories A, C and D. The presentation is focused on typical examples out of these three categories. An overview is given on the number of reported events and typical failure modes. Selected failures are discussed in detail. 1 ref., 6 figs, 7 tabs.

  18. Structural analysis of nuclear components

    International Nuclear Information System (INIS)

    Ikonen, K.; Hyppoenen, P.; Mikkola, T.; Noro, H.; Raiko, H.; Salminen, P.; Talja, H.

    1983-05-01

    THe report describes the activities accomplished in the project 'Structural Analysis Project of Nuclear Power Plant Components' during the years 1974-1982 in the Nuclear Engineering Laboratory at the Technical Research Centre of Finland. The objective of the project has been to develop Finnish expertise in structural mechanics related to nuclear engineering. The report describes the starting point of the research work, the organization of the project and the research activities on various subareas. Further the work done with computer codes is described and also the problems which the developed expertise has been applied to. Finally, the diploma works, publications and work reports, which are mainly in Finnish, are listed to give a view of the content of the project. (author)

  19. Structural Component Fabrication and Characterization of Advanced Radiation Resistant ODS Steel for Next Generation Nuclear Systems

    International Nuclear Information System (INIS)

    Noh, Sang Hoon; Kim, Young Chun; Jin, Hyun Ju; Choi, Byoung Kwon; Kang, Suk Hoon; Kim, Tae Kyu

    2016-01-01

    In a sodium-cooled fast reactor (SFR), the coolant outlet temperature and peak temperature of the fuel cladding tube will be about 545 .deg. C and 700 .deg. C with 250 dpa of a very high neutron dose rate. To realize this system, it is necessary to develop an advanced structural material having high creep and irradiation resistance at high temperatures. Austenitic stainless steel may be one of the candidates because of good strength and corrosion resistance at the high temperatures, however irradiation swelling severely occurred to 120dpa at high temperatures and this eventually leads to a decrease of the mechanical properties and dimensional stability. Advanced radiation resistant ODS steel (ARROS) has been newly developed for the in-core structural components in SFR, which has very attractive microstructures to achieve both superior creep and radiation resistances at high temperatures [4]. Nevertheless, the use of ARROS as a structural material essentially requires the fabrication technology development for component parts such as sheet, plate and tube. In this study, plates and tubes were tentatively fabricated with a newly developed alloy, ARROS. Microstructures as well as mechanical properties were also investigated to determine the optimized condition of the fabrication processes.

  20. A support vector machine integrated system for the classification of operation anomalies in nuclear components and systems

    International Nuclear Information System (INIS)

    Rocco S, Claudio M.; Zio, Enrico

    2007-01-01

    A support vector machine (SVM) approach to the classification of transients in nuclear power plants is presented. SVM is a machine-learning algorithm that has been successfully used in pattern recognition for cluster analysis. In the present work, single- and multiclass SVM are combined into a hierarchical structure for distinguishing among transients in nuclear systems on the basis of measured data. An example of application of the approach is presented with respect to the classification of anomalies and malfunctions occurring in the feedwater system of a boiling water reactor. The data used in the example are provided by the HAMBO simulator of the Halden Reactor Project

  1. Incipient Fault Detection and Isolation of Field Devices in Nuclear Power Systems Using Principal Component Analysis

    International Nuclear Information System (INIS)

    Kaistha, Nitin; Upadhyaya, Belle R.

    2001-01-01

    An integrated method for the detection and isolation of incipient faults in common field devices, such as sensors and actuators, using plant operational data is presented. The approach is based on the premise that data for normal operation lie on a surface and abnormal situations lead to deviations from the surface in a particular way. Statistically significant deviations from the surface result in the detection of faults, and the characteristic directions of deviations are used for isolation of one or more faults from the set of typical faults. Principal component analysis (PCA), a multivariate data-driven technique, is used to capture the relationships in the data and fit a hyperplane to the data. The fault direction for each of the scenarios is obtained using the singular value decomposition on the state and control function prediction errors, and fault isolation is then accomplished from projections on the fault directions. This approach is demonstrated for a simulated pressurized water reactor steam generator system and for a laboratory process control system under single device fault conditions. Enhanced fault isolation capability is also illustrated by incorporating realistic nonlinear terms in the PCA data matrix

  2. Intelligent Component Monitoring for Nuclear Power Plants

    International Nuclear Information System (INIS)

    Tsoukalas, Lefteri

    2010-01-01

    Reliability and economy are two major concerns for a nuclear power generation system. Next generation nuclear power reactors are being developed to be more reliable and economic. An effective and efficient surveillance system can generously contribute toward this goal. Recent progress in computer systems and computational tools has made it necessary and possible to upgrade current surveillance/monitoring strategy for better performance. For example, intelligent computing techniques can be applied to develop algorithm that help people better understand the information collected from sensors and thus reduce human error to a new low level. Incidents incurred from human error in nuclear industry are not rare and have been proven costly. The goal of this project is to develop and test an intelligent prognostics methodology for predicting aging effects impacting long-term performance of nuclear components and systems. The approach is particularly suitable for predicting the performance of nuclear reactor systems which have low failure probabilities (e.g., less than 10 -6 year -1 ). Such components and systems are often perceived as peripheral to the reactor and are left somewhat unattended. That is, even when inspected, if they are not perceived to be causing some immediate problem, they may not be paid due attention. Attention to such systems normally involves long term monitoring and possibly reasoning with multiple features and evidence, requirements that are not best suited for humans.

  3. Experimental qualification of nuclear components

    Energy Technology Data Exchange (ETDEWEB)

    Alliot, P; Fronte, T; Genty, F [FRAMATOME - Cedex 16, Paris la Defense (France)

    1988-07-01

    In the process of showing the adequacy of the seismic design of French PWR reactor, Fermat has repeatedly used dynamic testing on actual nuclear reactor components both on site and in manufacturing shops. The objective and results of a few representative examples of this on-site experimental verification are presented in this paper: the experimental dynamic analysis of a manipulator crane; the investigation of the seismic behaviour of fuel storage racks equipped with aseismic bearing devices. Difficulties to select satisfactory testing methods are also discussed for the particular case of the electrical cabinets. (author)

  4. Experimental qualification of nuclear components

    International Nuclear Information System (INIS)

    Alliot, P.; Fronte, T.; Genty, F.

    1988-01-01

    In the process of showing the adequacy of the seismic design of French PWR reactor, Fermat has repeatedly used dynamic testing on actual nuclear reactor components both on site and in manufacturing shops. The objective and results of a few representative examples of this on-site experimental verification are presented in this paper: the experimental dynamic analysis of a manipulator crane; the investigation of the seismic behaviour of fuel storage racks equipped with aseismic bearing devices. Difficulties to select satisfactory testing methods are also discussed for the particular case of the electrical cabinets. (author)

  5. Nuclear power plant component protection

    International Nuclear Information System (INIS)

    Michel, E.; Ruf, R.; Dorner, H.

    1976-01-01

    Described is a nuclear power plant installation which includes a concrete biological shield forming a pit in which a reactor pressure vessel is positioned. A steam generator on the outside of the shield is connected with the pressure vessel via coolant pipe lines which extend through the shield, the coolant circulation being provided by a coolant pump which is also on the outside of the shield. To protect these components on the outside of the shield and which are of mainly or substantially cylindrical shape, semicylindrical concrete segments are interfitted around them to form complete outer cylinders which are retained against outward separation radially from the components, by rings of high tensile steel which may be interspaced so closely that they provide, in effect, an outer steel cylinder. The invention is particularly applicable to pressurized-water coolant reactor installations

  6. IPRDS: component histories and nuclear plant aging

    International Nuclear Information System (INIS)

    Borkowski, R.J.; Kahl, W.K.

    1984-01-01

    A comprehensive assessment of nuclear power plant component operating histories, maintenance histories, and design and fabrication details is essential to understanding aging phenomena. As part of the In-Plant Reliability Data System (IPRDS), an attempt is being made to collect and analyze such information from a sampling of US nuclear power plants. Utilizing the IPRDS, one can reconstruct the failure history of the components and gain new insight into the causes and modes of failures resulting from normal or premature aging. This information assembled from the IPRDS can be combined with operating histories and postservice component inspection results for cradle-to-grave assessments of component aging under operating conditions. A comprehensive aging assessment can then be used to provide guidelines for improving the detection, monitoring, and mitigation of aging-related failures

  7. Behavior of spent nuclear fuel and storage system components in dry interim storage.

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, A.B. Jr.; Gilbert, E.R.; Guenther, R.J.

    1982-08-01

    Irradiated nuclear fuel has been handled under dry conditions since the early days of nuclear reactor operation, and use of dry storage facilities for extended management of irradiated fuel began in 1964. Irradiated fuel is currently being stored dry in four types of facilities: dry wells, vaults, silos, and metal casks. Essentially all types of irradiated nuclear fuel are currently stored under dry conditions. Gas-cooled reactor (GCR) and liquid metal fast breeder reactor (LMFBR) fuels are stored in vaults and dry wells. Certain types of fuel are being stored in licensed dry storage facilities: Magnox fuel in vaults in the United Kingdom and organic-cooled reactor (OCR) fuel in silos in Canada. Dry storage demonstrations are under way for Zircaloy-clad fuel from boiling water reactors BWR's, pressurized heavy-water reactors (PHWRs), and pressurized water reactors (PWRs) in all four types of dry storage facilities. The demonstrations and related hot cell and laboratory tests are directed toward expanding the data base and establishing a licensing basis for dry storage of water reactor fuel. This report reviews the scope of dry interim storage technology, the performance of fuel and facility materials, the status of programs in several countries to license dry storage of water reactor fuel, and the characteristics of water reactor fuel that relate to dry storage conditions.

  8. Behavior of spent nuclear fuel and storage-system components in dry interim storage

    International Nuclear Information System (INIS)

    Johnson, A.B. Jr.; Gilbert, E.R.; Guenther, R.J.

    1982-08-01

    Irradiated nuclear fuel has been handled under dry conditions since the early days of nuclear reactor operation, and use of dry storage facilities for extended management of irradiated fuel began in 1964. Irradiated fuel is currently being stored dry in four types of facilities: dry wells, vaults, silos, and metal casks. Essentially all types of irradiated nuclear fuel are currently stored under dry conditions. Gas-cooled reactor (GCR) and liquid metal fast breeder reactor (LMFBR) fuels are stored in vaults and dry wells. Certain types of fuel are being stored in licensed dry storage facilities: Magnox fuel in vaults in the United Kingdom and organic-cooled reactor (OCR) fuel in silos in Canada. Dry storage demonstrations are under way for Zircaloy-clad fuel from boiling water reactors BWR's, pressurized heavy-water reactors (PHWRs), and pressurized water reactors (PWRs) in all four types of dry storage facilities. The demonstrations and related hot cell and laboratory tests are directed toward expanding the data base and establishing a licensing basis for dry storage of water reactor fuel. This report reviews the scope of dry interim storage technology, the performance of fuel and facility materials, the status of programs in several countries to license dry storage of water reactor fuel, and the characteristics of water reactor fuel that relate to dry storage conditions

  9. Lubrication of nuclear reactor components

    International Nuclear Information System (INIS)

    Wild, E.; Mack, K.J.

    1978-01-01

    Safe operation of liquid metal cooled nuclear reactors requires a knowledge of the tribological behaviour of contacting components at high temperatures with slow relative movement at high frictional loads in a chemically aggressive environment. Experiments have been performed on various material combinations in liquid sodium and argon. Because of the small sliding movements, hydrodynamic lubrication is not expected and thus surface finish is an important factor. Tests have been performed on brushed, ground and lapped surfaces. Among the material combinations tested a CrC-coating on a 1.4961 stainless steel substrate performed well. Friction coefficients of 0.35-0.5 in argon and 0.1-1.2 in liquid sodium were recorded. (author)

  10. Experiences from Loviisa Nuclear Power Station concerning the decontamination of steam generators and primary system components

    International Nuclear Information System (INIS)

    Jaernstroem, R.

    1989-01-01

    Loviisa 1 and 2 are 465 MWe PWR units of the Soviet type VVER-440. Loviisa 1 has been in commercial operation since spring 1977 and Loviisa 2 from the beginning of 1980. Decontamination of primary circuit components - even big ones as steam generators - can be performed in an efficient and quick way with good results and resonable expences. Total costs for decontamination of the two steam generators including planning, construction, documentation, operation, chemicals etc. did not rise above 100,000.00 dollars. (author) 6 figs., 2 tabs

  11. Quality assurance grading criteria for plant systems and components: Results from a pilot plant project at Grand Gulf Nuclear Station. Final report

    International Nuclear Information System (INIS)

    Parkinson, W.J.

    1995-12-01

    As part of the original design of a nuclear power plant, the NSSS vendor, architect/engineer and utility identified structures, systems and components (SSCs) as safety related and assigned them to a Q-list. A Q-list is usually very large, e.g. 75,000 components, which creates large ongoing annual operating costs for the utility. Operating experience and the greater knowledge of plant systems safety accumulated during the past 20 years have suggested that many components are not truly important to safety and do not warrant the Q-classification and the associated costs. The completion of Probabilistic Safety Analyses (PSAs) for many nuclear power plants has contributed to this greater knowledge. This report describes a practical application of PSA technology to modify the existing QA program at the Grand Gulf Nuclear Station. Section 1 introduces the term, QA Safety Significant (QASS), and relates it to the existing term, ''safety related''. Section 2 describes six deterministic criteria as a basis for classifying systems as QASS or non-QASS. An expert panel reviewed 421 systems at Grand Gulf Nuclear Station and identified 42 of them as QASS. All components in non-QASS systems are classified as non-QASS. For QASS systems, Section 3 describes five deterministic criteria for classifying components as QASS or non-QASS. By using these two sets of criteria, the expert panel found that the number of components requiring full QA compliance could be reduced by 24%. These results are summarized in Section 4

  12. 4. Nuclear power plant component failures

    International Nuclear Information System (INIS)

    1990-01-01

    Nuclear power plant component failures are dealt with in relation to reliability in nuclear power engineering. The topics treated include classification of failures, analysis of their causes and impacts, nuclear power plant failure data acquisition and processing, interdependent failures, and human factor reliability in nuclear power engineering. (P.A.). 8 figs., 7 tabs., 23 refs

  13. Study of a simplified method of evaluating the economic maintenance importance of components in nuclear power plant system

    International Nuclear Information System (INIS)

    Aoki, Takayuki; Takagi, Toshiyuki; Kodama, Noriko

    2014-01-01

    Safety risk importance of components in nuclear power plants has been evaluated based on the probabilistic risk assessment and used for the decisions in various plant managements. But economic risk importance of the components has not been discussed very much. Therefore, this paper discusses risk importance of the components from the viewpoint of plant economic efficiency and proposes a simplified evaluation method of the economic risk importance (or economic maintenance importance). As a result of consideration, the followings were obtained. (1) A unit cost of power generation is selected as a performance indicator and can be related to a failure rate of components in nuclear power plant which is a result of maintenance. (2) The economic maintenance importance has to major factors, i.e. repair cost at component failure and production loss associated with plant outage due to component failure. (3) The developed method enables easy understanding of economic impacts of plant shutdown or power reduction due to component failures on the plane which adopts the repair cost in vertical axis and the production loss in horizontal axis. (author)

  14. 2-component heating systems

    Energy Technology Data Exchange (ETDEWEB)

    Radtke, W

    1987-03-01

    The knowledge accumulated only recently of the damage to buildings and the hazards of formaldehyde, radon and hydrocarbons has been inducing louder calls for ventilation, which, on their part, account for the fact that increasing importance is being attached to the controlled ventilation of buildings. Two-component heating systems provide for fresh air and thermal comfort in one. While the first component uses fresh air blown directly and controllably into the rooms, the second component is similar to the Roman hypocaustic heating systems, meaning that heated outer air is circulating under the floor, thus providing for hot surfaces and thermal comfort. Details concerning the two-component heating system are presented along with systems diagrams, diagrams of the heating system and tables identifying the respective costs. Descriptions are given of the two systems components, the fast heat-up, the two-component made, the change of air, heat recovery and control systems. Comparative evaluations determine the differences between two-component heating systems and other heating systems. Conclusive remarks are dedicated to energy conservation and comparative evaluations of costs. (HWJ).

  15. Development of a web-based aging monitoring system for an integrity evaluation of the major components in a nuclear power plant

    International Nuclear Information System (INIS)

    Choi, Jae-Boong; Yeum, Seung-Won; Ko, Han-Ok; Kim, Young-Jin; Kim, Hong-Key; Choi, Young-Hwan; Park, Youn-Won

    2010-01-01

    Structural and mechanical components in a nuclear power plant are designed to operate for its entire service life. Recently, a number of nuclear power plants are being operated beyond their design life to produce more electricity without shutting down. The critical issue in extending a lifetime is to maintain the level of safety during the extended operation period while satisfying the international regulatory standards. However, only a small portion of these components are of great importance for a significant aging degradation which would deeply affect the long-term safety and reliability of the related facilities. Therefore, it is beneficial to build a monitoring system to measure an aging status. While a number of integrity evaluation systems have been developed for NPPs, a real-time aging monitoring system has not been proposed yet . This paper proposes an expert system for the integrity evaluation of nuclear power plants based on a Web-based Reality Environment (WRE). The proposed system provides the integrity assessment for the major mechanical components of a nuclear power plant under concurrent working environments. In the WRE, it is possible for users to understand a mechanical system such as its size, geometry, coupling condition etc. In conclusion, it is anticipated that the proposed system can be used for a more efficient integrity evaluation of the major components subjected to an aging degradation.

  16. Development of a web-based aging monitoring system for an integrity evaluation of the major components in a nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jae-Boong, E-mail: boong33@skku.ed [SAFE Research Centre, School of Mechanical Engineering, Sungkyunkwan University, 300 Chunchun-dong, Jangan-gu, Suwon, Kyonggi-do 440-746 (Korea, Republic of); Yeum, Seung-Won; Ko, Han-Ok; Kim, Young-Jin [SAFE Research Centre, School of Mechanical Engineering, Sungkyunkwan University, 300 Chunchun-dong, Jangan-gu, Suwon, Kyonggi-do 440-746 (Korea, Republic of); Kim, Hong-Key; Choi, Young-Hwan; Park, Youn-Won [Korea Institute of Nuclear Safety, 34 Gwahak-ro, Yusong-ku, Teajon 305-338 (Korea, Republic of)

    2010-01-15

    Structural and mechanical components in a nuclear power plant are designed to operate for its entire service life. Recently, a number of nuclear power plants are being operated beyond their design life to produce more electricity without shutting down. The critical issue in extending a lifetime is to maintain the level of safety during the extended operation period while satisfying the international regulatory standards. However, only a small portion of these components are of great importance for a significant aging degradation which would deeply affect the long-term safety and reliability of the related facilities. Therefore, it is beneficial to build a monitoring system to measure an aging status. While a number of integrity evaluation systems have been developed for NPPs, a real-time aging monitoring system has not been proposed yet . This paper proposes an expert system for the integrity evaluation of nuclear power plants based on a Web-based Reality Environment (WRE). The proposed system provides the integrity assessment for the major mechanical components of a nuclear power plant under concurrent working environments. In the WRE, it is possible for users to understand a mechanical system such as its size, geometry, coupling condition etc. In conclusion, it is anticipated that the proposed system can be used for a more efficient integrity evaluation of the major components subjected to an aging degradation.

  17. Development of a three dimensional elastic plastic analysis system for the integrity evaluation of nuclear power plant components

    International Nuclear Information System (INIS)

    Huh, Nam Su; Im, Chang Ju; Kim, Young Jin; Pyo, Chang Ryul; Park, Chi Yong

    2000-01-01

    In order to evaluate the integrity of nuclear power plant components, the analysis based on fracture mechanics is crucial. For this purpose, finite element method is popularly used to obtain J-integral. However, it is time consuming to design the finite element model of a cracked structure. Also, the J-integral should by verified by alternative methods since it may differ depending on the calculation method. The objective of this paper is to develop a three-dimensional elastic-plastic J-integral analysis system which is named as EPAS program. The EPAS program consists of an automatic mesh generator for a through-wall crack and a surface crack, a solver based on ABAQUS program, and a J-integral calculation program which provides DI (Domain Integral) and EDI (Equivalent Domain Integral) based J-integral calculation. Using the EPAS program, an optimized finite element model for a cracked structure can be generated and corresponding J-integral can be obtained subsequently

  18. IPRDS - Component histories and nuclear plant aging

    International Nuclear Information System (INIS)

    Borkowski, R.J.; Kahl, W.K.

    1984-01-01

    A comprehensive assessment of nuclear power plant component operating histories, maintenance histories, and design and fabrication details is essential to understanding aging phenomena. As part of the In-Plant Reliability Data System (IPRDS), an attempt is being made to collect and analyze such information from a sampling of U.S. nuclear power plants. Utilizing the IPRDS, one can reconstruct the failure history of the components and gain new insight into the causes and modes of failures resulting from normal or premature aging. This information assembled from the IPRDS can be combined with operating histories and postservice component inspection results for ''cradle-to-grave'' assessments of component aging under operating conditions. A comprehensive aging assessment can then be used to provide guidelines for improving the detection, monitoring, and mitigation of aging-related failures. The examples chosen for this paper illustrate two aging-related areas: the effects of an improved preventive maintenance policy in mitigating aging of a feedwater pump and the identification of reoccuring failures in parts of diesel generators

  19. Test to prove the resistance to incidents of components of electric and control systems in the safety containment of nuclear power plants

    International Nuclear Information System (INIS)

    1982-01-01

    The marginal program for proving the suitability of safety-relevant components of electric and control systems in the safety containment during a loss-of-coolant incident is described. Variant test conditions are established in the component-specific test program. Special attention has been paid to the representation of the course of pressure and temperature for the performance test of the valve room of the Nuclear Power Plant Philippsburg 2. (DG) [de

  20. Methodology of aging management in structures, systems and components of a nuclear power plant and its application to a pilot program in Laguna Verde

    International Nuclear Information System (INIS)

    Jarvio C, G.; Fernandez S, G.

    2009-10-01

    From its origin the nuclear power plants confront the effects of time and of environment, giving as result the aging of its structures, systems and components. In this document the general process is described for the establishment of Aging Management Program developed by IAEA. Following the program methodology is guaranteed that a nuclear power plant manages the aging effects appropriately and to make decisions for its solution, assuring the characteristic functions of structures, systems and components of same nuclear power plant. On the other hand, the implantation of an aging management program constitutes the base for development of a licence renovation program, like it can be the specific case of the Central Laguna Verde Units 1 and 2. (Author)

  1. Development of a web-based monitoring system using operation parameters for the main component in nuclear power plants

    International Nuclear Information System (INIS)

    Son, Dong Chan; An, Kung Il; Hong, Suk Young; Lee, Jeong Soo; Jung, Duk Jin; Shin, Sun Hee; Son, So Hee

    2004-02-01

    The frequency of the damage is increasing, which is caused by the fatigue, according to the increase of running of nuclear power plants. So we need to acquire the reliance of design data to estimate the fatigue and damage of major machinery that might happen as time-dependent crack growth characterization. The research is focused on keeping operating record of nuclear power plants about major machinery which consists of a nuclear reactor pressure boarder on each excessive operating condition including normal operating and extraordinary operating by estimating fracture mechanical movements on real time and fatigue about major nuclear power plants machinery, which are acquired the pressure and temperature data. For further details about the scope and contents of R and D are following. Development of H/W that is necessary to acquire operating real time data of heating and hydraulic power. Selection of a safety variable about major system by each type (the four NPP, all unit). Communication protocol development for connecting between CARE system data base server and fatigue monitoring system data base server. Development of connecting database for controlling and storing of heating and hydraulic power operating data. Real time monitoring system development based on Web using JAVA

  2. Development of a web-based monitoring system using operation parameters for the main component in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Son, Dong Chan; An Kang Il; Hong, Suk Young; Lee, Jeong Soo; Lee, Kwang Yeol; Shin, Sun Hee; Lee, Chun Wha; Son, So Hee [Daesang Information Technology Co., Ltd., Seoul (Korea, Republic of)

    2003-03-15

    The frequency of the damage is increasing, which is caused by the fatigue, according to the increase of running of nuclear power plants. So we need to acquire the reliance of design data to estimate the fatigue and damage of major machinery that might happen as time-dependent crack growth characterization. The research is focused on keeping operating record of nuclear power plants about major machinery which consists of a nuclear reactor pressure boarder on each excessive operating condition including normal operating and extraordinary operating by estimating fracture mechanical movements on real time and fatigue about major nuclear power plants machinery, which are acquired the pressure and temperature data. For further details about the scope and contents of R and D are following: development of H/W that is necessary to acquire operating real time data of heating and hydraulic power, selection of a safety variable about major system by each type (the fourth unit), communication protocol development for connecting between CARE system data base server and fatigue monitoring system data base server, development of connecting database for controlling and storing of heating and hydraulic power operating data, real time monitoring system development based on web using JAVA.

  3. Development of a web-based monitoring system using operation parameters for the main component in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Son, Dong Chan; An, Kung Il; Hong, Suk Young; Lee, Jeong Soo; Jung, Duk Jin; Shin, Sun Hee; Son, So Hee [Daesang Information Technology Co., Ltd., Seoul (Korea, Republic of)

    2004-02-15

    The frequency of the damage is increasing, which is caused by the fatigue, according to the increase of running of nuclear power plants. So we need to acquire the reliance of design data to estimate the fatigue and damage of major machinery that might happen as time-dependent crack growth characterization. The research is focused on keeping operating record of nuclear power plants about major machinery which consists of a nuclear reactor pressure boarder on each excessive operating condition including normal operating and extraordinary operating by estimating fracture mechanical movements on real time and fatigue about major nuclear power plants machinery, which are acquired the pressure and temperature data. For further details about the scope and contents of R and D are following. Development of H/W that is necessary to acquire operating real time data of heating and hydraulic power. Selection of a safety variable about major system by each type (the four NPP, all unit). Communication protocol development for connecting between CARE system data base server and fatigue monitoring system data base server. Development of connecting database for controlling and storing of heating and hydraulic power operating data. Real time monitoring system development based on Web using JAVA.

  4. Periodic inspection of CANDU nuclear power plant containment components

    International Nuclear Information System (INIS)

    1989-09-01

    This Standard is one in a series intended to provide uniform requirements for CANDU nuclear power plants. It provides requirements for the periodic inspection of containment components including the containment pressure suppression systems

  5. Nuclear component horizontal seismic restraint

    International Nuclear Information System (INIS)

    Snyder, G.J.

    1988-01-01

    In a nuclear reactor having a reactor vessel, a reactor guard vessel, a thermal insulation shell and a horizontal seismic restraint, a restraint is described comprising: a. a first ring on the wall of the reactor vessel; b. a second ring on the wall of the reactor guard vessel in alignment with the first ring; c. a first block attached to the second ring proximate the first ring so as to provide a predetermined clearance between the first block and the first ring which is reduced to zero during thermal expansion; d. motion limit means extending through an aperture in the thermal insulation shell in alignment with the second ring and the first block; the e. a second block attached to the motion limit means proximate the second ring and in alignment the first block so as to provide a predetermined clearance between the second block and the second ring which is reduced to zero during thermal expansion

  6. Electronic components and systems

    CERN Document Server

    Dennis, W H

    2013-01-01

    Electronic Components and Systems focuses on the principles and processes in the field of electronics and the integrated circuit. Covered in the book are basic aspects and physical fundamentals; different types of materials involved in the field; and passive and active electronic components such as capacitors, inductors, diodes, and transistors. Also covered in the book are topics such as the fabrication of semiconductors and integrated circuits; analog circuitry; digital logic technology; and microprocessors. The monograph is recommended for beginning electrical engineers who would like to kn

  7. Trend and pattern analysis of failures of main feedwater system components in United States commercial nuclear power plants

    International Nuclear Information System (INIS)

    Gentillon, C.D.; Meachum, T.R.; Brady, B.M.

    1987-01-01

    The goal of the trend and pattern analysis of MFW (main feedwater) component failure data is to identify component attributes that are associated with relatively high incidences of failure. Manufacturer, valve type, and pump rotational speed are examples of component attributes under study; in addition, the pattern of failures among NPP units is studied. A series of statistical methods is applied to identify trends and patterns in failures and trends in occurrences in time with regard to these component attributes or variables. This process is followed by an engineering evaluation of the statistical results. In the remainder of this paper, the characteristics of the NPRDS that facilitate its use in reliability and risk studies are highlighted, the analysis methods are briefly described, and the lessons learned thus far for improving MFW system availability and reliability are summarized (orig./GL)

  8. LWR nuclear power plant component failures

    International Nuclear Information System (INIS)

    Schmidt, W.H.

    1980-10-01

    An analysis of the most significant light water reactor (LWR) nuclear power plant component failures, from information in the computerized Nuclear Safety Information Center (NSIC) data bank, shows that for both pressurized water reactor (PWR) and boiling water reactor (BWR) plants the component category most responsible for reactor shutdowns is valves. Next in importance for PWR shutdowns is steam generators followed by seals of all kinds. For BWR plants, seals, and pipes and pipe fittings are the second and third most important component failure categories which lead to reactor shutdown. The data are for records extending from early 1972 through September 1978. A list of the most significant component categories and a breakdown of the number of component citations for both PWR and BWR reactor types are presented

  9. Commissioning of qualification of structures, systems and components for seismic and environmental loads of CIRENE nuclear power plant

    International Nuclear Information System (INIS)

    Bianchi, A.; Gatti, F.; Muzzi, F.; Zola, M.; De Pasquali, F.

    1993-01-01

    On behalf of the Italian National Electricity Board (ENEL) concerning the commissioning of qualification of structures, systems and components of CIRENE NPP, ISMES performed a technical surveillance on the documentation concerning the environmental and seismic qualification of the safety related systems and experimental activities (dynamic and static tests) on plant buildings. The aims of the work were: - the evaluation of the qualification carried out (by test, by analysis, by combination of analysis and test) on the equipment and system, compared with the requirements of the ENEL technical specifications and the most recent international regulations; - the experimental determination of modal quantities (frequencies, damping, mode shapes) of the structures and, in the case of reactor building, the complex impedance of the soil for supporting the analytical work. The present paper deals with the criteria, the system and the results concerning the technical surveillance and with the characteristics and the results of the experimental tests

  10. Analysis of failed nuclear plant components

    Science.gov (United States)

    Diercks, D. R.

    1993-12-01

    Argonne National Laboratory has conducted analyses of failed components from nuclear power- gener-ating stations since 1974. The considerations involved in working with and analyzing radioactive compo-nents are reviewed here, and the decontamination of these components is discussed. Analyses of four failed components from nuclear plants are then described to illustrate the kinds of failures seen in serv-ice. The failures discussed are (1) intergranular stress- corrosion cracking of core spray injection piping in a boiling water reactor, (2) failure of canopy seal welds in adapter tube assemblies in the control rod drive head of a pressurized water reactor, (3) thermal fatigue of a recirculation pump shaft in a boiling water reactor, and (4) failure of pump seal wear rings by nickel leaching in a boiling water reactor.

  11. Analysis of failed nuclear plant components

    International Nuclear Information System (INIS)

    Diercks, D.R.

    1993-01-01

    Argonne National Laboratory has conducted analyses of failed components from nuclear power-generating stations since 1974. The considerations involved in working with an analyzing radioactive components are reviewed here, and the decontamination of these components is discussed. Analyses of four failed components from nuclear plants are then described to illustrate the kinds of failures seen in service. The failures discussed are (1) intergranular stress-corrosion cracking of core spray injection piping in a boiling water reactor, (2) failure of canopy seal welds in adapter tube assemblies in the control rod drive head of a pressurized water reactor, (3) thermal fatigue of a recirculation pump shaft in a boiling water reactor, and (4) failure of pump seal wear rings by nickel leaching in a boiling water reactor

  12. Analysis of failed nuclear plant components

    International Nuclear Information System (INIS)

    Diercks, D.R.

    1992-07-01

    Argonne National Laboratory has conducted analyses of failed components from nuclear power generating stations since 1974. The considerations involved in working with and analyzing radioactive components are reviewed here, and the decontamination of these components is discussed. Analyses of four failed components from nuclear plants are then described to illustrate the kinds of failures seen in service. The failures discussed are (a) intergranular stress corrosion cracking of core spray injection piping in a boiling water reactor, (b) failure of canopy seal welds in adapter tube assemblies in the control rod drive head of a pressure water reactor, (c) thermal fatigue of a recirculation pump shaft in a boiling water reactor, and (d) failure of pump seal wear rings by nickel leaching in a boiling water reactor

  13. In-service diagnostic systems of steam generators, pressurizers and other components of WWER type nuclear power plants

    International Nuclear Information System (INIS)

    Matal, O.

    1988-01-01

    A detailed description is presented of the systems of vibration inspections and systems of determining residual service life, implemented as in-service diagnostic systems for steam generators and pressurizers at the Dukovany nuclear power plant. Low temperature accelerometers of the KD or KS type and high temperature accelerometers CA 91 are used as vibration sensors. In the system of vibration inspection a total of 64 vibration measuring chains of Czechoslovak make and design are installed in the power plant. Systems are being built for determining residual service life which consist of 75 special chains for heat monitoring with thermocouples installed on selected assemblies of the steam generators and the pressurizers serving to monitor and evaluate heat stress. Also included in the system for determining residual service life are 16 routes for water withdrawal from steam generators. Their purpose is to make in-service determinations of places of biggest concentrations of impurities in secondary water, to determine the biggest local chemical exposure of primary collector and heat exchange tube materials and to optimize the size and place of leachate withdrawal. (Z.M.). 2 figs., 2 tabs., 15 refs

  14. Method to classify the safety class of Structure, System and Components in a Defueled Condition of Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong-Hak; Jeon, Dang-Hee [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    During pre-decommissioning phase, licensing and engineering work need to change the design basis of the plant such as safety analysis report, downgrade of systems, technical specifications and program and procedures to change of NPP condition from in an operation condition to in a defueled condition. The many systems to need to operate in an operational condition will not be operated during in a defueled condition and the function of systems will be changed from in an operation condition to in a defueled condition. So a downgrade of systems may be needed and reclassifying the safety class of structure, system and component (SSC) may be conducted. By the reclassification of SSC, activity related with quality assurance and maintenance of SSC is affected. In this paper, the method to reclassify SSC in a defueled condition is studied. The many systems to need to operate in an operational condition will not be operated during in a defueled condition and the function of systems will be changed from in an operation condition to in a defueled condition. The operation of NPP during a defueled condition need to conduct licensing and engineering work need to change the design basis of the plant optimize by downgrading systems and reclassifying the safety class of SSC. In this paper, the method to reclassify safety class for a defueled condition is studied.

  15. Structural mechanics of nuclear plant components

    International Nuclear Information System (INIS)

    Roche, R.

    1986-10-01

    Sound structural analysis are needed for designing safe and reliable components, hence his play is very important in nuclear industry. This report is a provisional writing on the good practice in structural mechanics. Emphasis is put on non elastic analysis, damage appraisal, fatigue, fracture mechanics and also on elevated temperature behaviour [fr

  16. Modular core component support for nuclear reactor

    International Nuclear Information System (INIS)

    Finch, L.M.; Anthony, A.J.

    1975-01-01

    The core of a nuclear reactor is made up of a plurality of support modules for containing components such as fuel elements, reflectors and control rods. Each module includes a component support portion located above a grid plate in a low-pressure coolant zone and a coolant inlet portion disposed within a module receptacle which depends from the grid plate into a zone of high-pressure coolant. Coolant enters the module through aligned openings within the receptacle and module inlet portion and flows upward into contact with the core components. The modules are hydraulically balanced within the receptacles to prevent expulsion by the upward coolant forces. (U.S.)

  17. Evaluation and mitigation of the degradation by corrosion in the components of the service water system of a nuclear power plant

    International Nuclear Information System (INIS)

    Salaices A, E.; Salaices, M.; Ovando, R.

    2005-01-01

    One of the main problems that face the nuclear power stations is the degradation by corrosion in the service water systems. The corrosion causes lost substantial in energy generation and a high cost in maintenance and repairs. In this work, the results of a study of the degradation by the MIC mechanisms (microorganisms influenced corrosion), incrustations in heat exchangers and erosion for solid particles in the components of a typical service water system of a nuclear plant are presented. Diverse mitigation options are analyzed for these mechanisms. In the analysis, it was used the CHECWORKS-CWA code to carry out the evaluation of the degradation so much as well as the mitigation of the caused damage. The results are presented in susceptibility indexes and degradation rates component-by-component. A significant decrement could be observed in the susceptibility to MIC when changing the operation conditions of stagnated flow to continuous flow. With respect to the erosion by solid particles, it was found a significant reduction of the damage it when adding filters to the system. Finally, in the case of the heat exchangers, it is shown that one of the more viable options to diminish incrustations and existent calcium deposits it is the reduction of the pH of the service water. (Author)

  18. Polyphophoinositides components of plant nuclear membranes

    International Nuclear Information System (INIS)

    Hendrix, K.W.; Boss, W.F.

    1987-01-01

    The polyphosphoinositides, phosphatidylinositol monophosphate (PIP) and phosphatidylinositol bisphosphate (PIP 2 ), have been shown to be important components in signal transduction in many animal cells. Recently, these lipids have been found to be associated with plasma membrane but not microsomal membrane isolated from fusogenic wild carrot cells; however, in that study the lipids of the nuclear membrane were not analyzed. Since polyphosphoinositides had been shown to be associated with the nuclear membranes as well as the plasma membrane in some animal cells, it was important to determine whether they were associated with plant nuclear membranes as well. Cells were labeled for 18h with [ 3 H] inositol and the nuclei were isolated by a modification of the procedure of Saxena et al. Preliminary lipid analyses indicate lower amount of PIP and PIP 2 in nuclear membranes compared to whole protoplasts. This suggests that the nuclear membranes of carrot cells are not enriched in PIP and PIP 2 ; however, the Triton X-100 used during the nuclear isolation procedure may have affected the recovery of the lipids. Experiments are in progress to determine the effects of Triton X-100 on lipid extraction

  19. Technology development for special nuclear components

    International Nuclear Information System (INIS)

    Sanatkumar, A.

    1994-01-01

    One of the attractive features of Candu Pressurised Heavy Water Reactor design which influenced the decision to make it the foundation of our nuclear power programme, is that its main components (calandria, end shields, coolant channel components) are relatively simple - in comparison with reactor pressure vessel and associated components of Boiling Water Reactors or Pressurised Water Reactors - and considered to be within the scope of manufacture of developing countries. Over the last two decades, India has been very successful in technology development in many important and critical areas. We are now about to launch the construction of the first 500 MWe PHWR project at Tarapur. In this context, this paper focuses attention on some of the aspects relating to self-reliance in design, engineering and manufacture of these special components as currently perceived. (author). 3 refs

  20. Analysis of changes in the fuel component of the cost of electricity in the transition to a closed fuel cycle in nuclear power system

    Energy Technology Data Exchange (ETDEWEB)

    Gurin, Andrey V. [National Research Centre ' ' Kurchatov Institute' ' , Moscow (Russian Federation); Alekseev, P.N.

    2017-09-15

    This paper presents a study of scenarios of transition to a closed fuel cycle in the system of nuclear power, built basing on resource availability requirements at the stage of full life-cycle reactors. Conventionally, there are three main scenarios for the development of nuclear energy: with VVER reactors operating in an open fuel cycle; with VVER reactors operating in a closed fuel cycle; and co-operating VVER and BN, operating in a closed fuel cycle. For the considered scenarios, a quantitative estimation of change in time of material balances were performed, including spent fuel balance, balance of plutonium, reprocessed and depleted uranium, radioactive waste, and the analysis of the fuel component of the cost of electricity.

  1. Analysis of changes in the fuel component of the cost of electricity in the transition to a closed fuel cycle in nuclear power system

    International Nuclear Information System (INIS)

    Gurin, Andrey V.; Alekseev, P.N.

    2017-01-01

    This paper presents a study of scenarios of transition to a closed fuel cycle in the system of nuclear power, built basing on resource availability requirements at the stage of full life-cycle reactors. Conventionally, there are three main scenarios for the development of nuclear energy: with VVER reactors operating in an open fuel cycle; with VVER reactors operating in a closed fuel cycle; and co-operating VVER and BN, operating in a closed fuel cycle. For the considered scenarios, a quantitative estimation of change in time of material balances were performed, including spent fuel balance, balance of plutonium, reprocessed and depleted uranium, radioactive waste, and the analysis of the fuel component of the cost of electricity.

  2. Design, maintenance and lifetime of nuclear components

    International Nuclear Information System (INIS)

    Noel, R.L.; Eisenhut, D.G.; Carey, J.J.; Reynes, L.J.

    1989-01-01

    Division D of SMiRT deals with experience feedback relating to the in-service behavior of nuclear components, the design and construction of this equipment, its maintenance and the evaluation and management of its lifetime. The nuclear industry now having reached maturity, with more than 300 units in service worldwide, these problems are now of predominant importance to the activity of the industry and in its development programs. This applies particularly to the problems relating to the lifetime of nuclear plants, problems which are rightly of such concern both to the utilities, in view of the enormous investments involved, and also to the safety authorities. These contributions have been reviewed for the purpose of analyzing the essential points. This analysis highlights the considerable advances achieved during the recent decades in design and maintenance methods and practices. It also identifies the areas in which progress still remains to be made

  3. Nondestructive testing of nuclear reactor components integrity

    International Nuclear Information System (INIS)

    Mala, M.; Miklos, M.

    2011-01-01

    Nuclear energy must respond to current challenges in the energy market. The significant parameters are increase of the nuclear fuel price, closed fuel cycle, reduction and safe and the final disposal of high level radioactive waste. Nowadays, the discussions on suitable energy mix are taking place not only here in Czech Republic, but also in many other European countries. It is necessary to establish an appropriate ratio among the production of electricity from conventional, nuclear and renewable energy sources. Also, it is necessary to find ways how to streamline the economy, central part of the nuclear fuel cycle and thereby to increase the competitiveness of nuclear energy. This streamlining can be carried out by improving utilization of existing nuclear fuel with maintaining a high degree of nuclear facilities safety. Increasing operational reliability and safety together with increasing utilization of nuclear fuel place increasing demands on monitoring of changes during fuel burnup. The potential fuel assembly damages in light water reactors are prevented by the introduction of new procedures and programs of the fuel assembly monitoring. One of them is the Post Irradiation Inspection Program (PIIP) which is a good tool for monitoring of chemical regime impact on the fuel assembly cladding behavior. Main nondestructive techniques that are used at nuclear power plants for the fuel assembly integrity evaluation are ultrasonic measurements, eddy current measurements, radiographic testing, acoustic techniques and others. Ultrasonic system is usual tool for leak fuel rod evaluation and it is also used at Temelin NPP. Since 2009, Temelin NPP has cooperated with Research Center Rez Ltd in frame of PIIP program at both units WWER 1000. This program was established for US VVantage6 fuel assemblies and also it continues for Russian TVSA-T fuel assemblies. (author)

  4. IAEA guidance on ageing management for nuclear power plants. Guidance on effective management of the physical ageing of systems, structures and components important to safety for nuclear power plants. Overview. Programmatic guidelines. Component specific guidelines. Review guidelines. Version 1, 2002

    International Nuclear Information System (INIS)

    2002-01-01

    Operational experience shows that excellent plant safety and excellent performance go hand in hand, and that they are achieved by effective leadership and management that includes a unified approach to safety and production. This is also applicable to ageing management. Effective ageing management leads to both enhanced plant safety and enhanced performance and is a prerequisite for long service life. The IAEA project on Safety Aspects of NPP Ageing has produced since 1990 a comprehensive set of programmatic and component specific guidelines on managing ageing, while providing an interactive environment for information exchange and co-operation among practitioners, and has assisted Member States in the application of the guidelines through the provision of training and advice. The objective of the CD-ROM is to preserve the IAEA's guidance on ageing management and to facilitate its retrieval, updating, extension and dissemination in order to help increase the effectiveness of ageing management at nuclear power plants

  5. Effect of high frequency content of uniform hazard response spectra on nuclear power plant structures, systems and components

    Energy Technology Data Exchange (ETDEWEB)

    Usmani, A. [Amec Foster Wheeler, Toronto, ON (Canada); Baughman, P.D. [Paul D. Baughman Consulting, Exeter, NH (United States)

    2015-07-01

    The Uniform Hazard Spectrum (UHS) is developed from a probabilistic seismic hazard assessment and represents a response spectrum for which the amplitude at each frequency has a specified and uniform (equal) probability of exceedance. The high spectral acceleration at high frequencies in the UHS can result mainly from small non-damaging low energy earthquakes. Historically Canadian and U.S. nuclear power plants have been designed using the standard shape spectrum given in CSA N289.3 or USNRC Regulatory Guide 1.60, which have maximum spectral accelerations in the lower (2-10 Hz.) frequency range. The impact of the high frequency content of UHS on the nuclear power plant SSCs is required to be assessed. This paper briefly describes the methodologies used for screening and evaluation of the effects of UHS high frequency content on the nuclear power SSCs that have been designed using the CSA N289.3 standard shape spectrum. (author)

  6. HTGR nuclear heat source component design and experience

    International Nuclear Information System (INIS)

    Peinado, C.O.; Wunderlich, R.G.; Simon, W.A.

    1982-05-01

    The high-temperature gas-cooled reactor (HTGR) nuclear heat source components have been under design and development since the mid-1950's. Two power plants have been designed, constructed, and operated: the Peach Bottom Atomic Power Station and the Fort St. Vrain Nuclear Generating Station. Recently, development has focused on the primary system components for a 2240-MW(t) steam cycle HTGR capable of generating about 900 MW(e) electric power or alternately producing high-grade steam and cogenerating electric power. These components include the steam generators, core auxiliary heat exchangers, primary and auxiliary circulators, reactor internals, and thermal barrier system. A discussion of the design and operating experience of these components is included

  7. Modelization of cooling system components

    Energy Technology Data Exchange (ETDEWEB)

    Copete, Monica; Ortega, Silvia; Vaquero, Jose Carlos; Cervantes, Eva [Westinghouse Electric (Spain)

    2010-07-01

    In the site evaluation study for licensing a new nuclear power facility, the criteria involved could be grouped in health and safety, environment, socio-economics, engineering and cost-related. These encompass different aspects such as geology, seismology, cooling system requirements, weather conditions, flooding, population, and so on. The selection of the cooling system is function of different parameters as the gross electrical output, energy consumption, available area for cooling system components, environmental conditions, water consumption, and others. Moreover, in recent years, extreme environmental conditions have been experienced and stringent water availability limits have affected water use permits. Therefore, modifications or alternatives of current cooling system designs and operation are required as well as analyses of the different possibilities of cooling systems to optimize energy production taking into account water consumption among other important variables. There are two basic cooling system configurations: - Once-through or Open-cycle; - Recirculating or Closed-cycle. In a once-through cooling system (or open-cycle), water from an external water sources passes through the steam cycle condenser and is then returned to the source at a higher temperature with some level of contaminants. To minimize the thermal impact to the water source, a cooling tower may be added in a once-through system to allow air cooling of the water (with associated losses on site due to evaporation) prior to returning the water to its source. This system has a high thermal efficiency, and its operating and capital costs are very low. So, from an economical point of view, the open-cycle is preferred to closed-cycle system, especially if there are no water limitations or environmental restrictions. In a recirculating system (or closed-cycle), cooling water exits the condenser, goes through a fixed heat sink, and is then returned to the condenser. This configuration

  8. Research on development model of nuclear component based on life cycle management

    International Nuclear Information System (INIS)

    Bao Shiyi; Zhou Yu; He Shuyan

    2005-01-01

    At present the development process of nuclear component, even nuclear component itself, is more and more supported by computer technology. This increasing utilization of the computer and software has led to the faster development of nuclear technology on one hand and also brought new problems on the other hand. Especially, the combination of hardware, software and humans has increased nuclear component system complexities to an unprecedented level. To solve this problem, Life Cycle Management technology is adopted in nuclear component system. Hence, an intensive discussion on the development process of a nuclear component is proposed. According to the characteristics of the nuclear component development, such as the complexities and strict safety requirements of the nuclear components, long-term design period, changeable design specifications and requirements, high capital investment, and satisfaction for engineering codes/standards, the development life-cycle model of nuclear component is presented. The development life-cycle model is classified at three levels, namely, component level development life-cycle, sub-component development life-cycle and component level verification/certification life-cycle. The purposes and outcomes of development processes are stated in detailed. A process framework for nuclear component based on system engineering and development environment of nuclear component is discussed for future research work. (authors)

  9. Indigenous procurement of nuclear components at Tarapur (Paper No. 013)

    International Nuclear Information System (INIS)

    Verma, D.K.; Moss, V.J.

    1987-02-01

    The Tarapur Atomic Power Station (TAPS) was the first nuclear power station in developing countries and the first twin BWR units in the world. The Station has two units of boiling water reactor of very early design; along with its turbo-generator and supporting systems; constructed by M/s. I.G.E. on turnkey basis. Based on vendor recommendations initial operating spares for 5 years of operation were purchased from original equipment manufacturers. This does not call for the participation of the ultimate user; in the design, development, manufacture and quality control and user's participation remained confined to assemble the acceptable component(s) procured from original source in the assembly. As early as 1972, Plant initiated indigenising the nuclear components by gradually increasing the contribution of indigenous industry with due participation of the departmental agencies. Procurement of nuclear components requires development of engineering to an extent; where interphase communication between TAPS and counterpart indigenous industry is practicable to motivate them. Feedback from operation and maintenance practices is also utilised effectively. For some of the components initial sample were developed at TAPS and subsequently bulk fabrication was taken by industry. This paper describes manufacture, quality control during the process of manufacture and procurement of indigenous nuclear components relevant to Tarapur Atomic Power Station. (author)

  10. Monitoring ageing of components in nuclear plants

    International Nuclear Information System (INIS)

    Fritz, M.R.

    1992-01-01

    There are several mechanisms of ageing or damage in nuclear components, of which the best known can be classified into three categories: generalized damage mechanisms (wear, corrosion, erosion,...), local damage phenomena (fatigue, corrosion,...) and material property degradations. For ageing evaluation, the first requirement is a good understanding of the damage mechanisms and the determination of the kinetic laws and major influencing factors. When these factors are measurable physical parameters, ageing monitoring and periodic evaluation of damage level become possible. From the set of tools available for ageing evaluation, four are presented here in more detail: the transient logging procedure, the defect injuriousness analysis, the fatigue meter, the probabilistic approach to structural integrity. (author)

  11. Monitoring ageing of components in nuclear plants

    Energy Technology Data Exchange (ETDEWEB)

    Fritz, M R [FRAMATOME, Paris (France)

    1992-07-01

    There are several mechanisms of ageing or damage in nuclear components, of which the best known can be classified into three categories: generalized damage mechanisms (wear, corrosion, erosion,...), local damage phenomena (fatigue, corrosion,...) and material property degradations. For ageing evaluation, the first requirement is a good understanding of the damage mechanisms and the determination of the kinetic laws and major influencing factors. When these factors are measurable physical parameters, ageing monitoring and periodic evaluation of damage level become possible. From the set of tools available for ageing evaluation, four are presented here in more detail: the transient logging procedure, the defect injuriousness analysis, the fatigue meter, the probabilistic approach to structural integrity. (author)

  12. Nuclear reactor sealing system

    International Nuclear Information System (INIS)

    McEdwards, J.A.

    1983-01-01

    A liquid metal-cooled nuclear reactor sealing system is disclosed. The nuclear reactor includes a vessel sealed at its upper end by a closure head. The closure head comprises at least two components, one of which is rotatable; and the two components define an annulus therebetween. The sealing system includes at least a first and second inflatable seal disposed in series in an upper portion of the annulus. The system further includes a dip seal extending into a body of insulation located adjacent a bottom portion of the closure head. The dip seal comprises a trough formed by a lower portion of one of the components, and a seal blade pendently supported from the other component and extending downwardly into the trough. A body of liquid metal is contained in the trough which submerges a portion of the seal blade. The seal blade is provided with at least one aperture located above the body of liquid metal for providing fluid communication between the annulus intermediate the dip seal and the inflatable seals, and a body of cover gas located inside the vessel. There also is provided means for introducing a purge gas into the annulus intermediate the inflatable seals and the seal blade. The purge gas is introduced in an amount sufficient to substantially reduce diffusion of radioactive cover gas or sodium vapor up to the inflatable seals. The purge gas mixes with the cover gas in the reactor vessel where it can be withdrawn from the vessel for treatment and recycle to the vessel

  13. Analysis of availability, functional integration and remote maintenance for the design of critical components and systems in nuclear fusion technology

    OpenAIRE

    Fernández Berceruelo, Iván

    2016-01-01

    Junto a las conocidas ventajas de la Fusión Nuclear frente a otras tecnologías energéticas: seguridad inherente por principio físico, gestión del impacto ambiental y disponibilidad a largo plazo de combustibles primarios (deuterio, litio), la competitividad de la energía generada por fusión determinará su capacidad de penetración en el mercado. Para lograrla será necesario revisar rigurosamente las opciones de diseño existentes, de modo que se justifique sólidamente la elección de una solució...

  14. Interactive visualization system to analyze corrugated millimeter-waveguide component of ECH in nuclear fusion with FDTD simulation

    International Nuclear Information System (INIS)

    Kashima, N; Nakamura, H; Kubo, S; Tamura, Y; Ito, A M

    2014-01-01

    We have simulated distribution of electromagnetic waves through the system composed of miter bends by Finite-Difference Time-Domain (FDTD) simulation. We develop the interactive visualization system using a new interactive GUI system which is composed of the virtual reality system and android tablet to analyze the FDTD simulation. The effect of the waveguide system with grooves have been investigated to quantitatively by visualization system. Comparing waveguide system with grooves and without grooves, grooves have been confirmed to suppress the surface current at the metal surface. The surface current at complex shape such as the miter bend have been investigated

  15. Manufacturing of nuclear power components in CDM

    International Nuclear Information System (INIS)

    Krishnan, J.; Jawale, S.B.

    2002-01-01

    Full text: In the nuclear research programme in India, Dr. H.J. Bhabha, the architecture of the Indian Nuclear programme felt a need for proto-type development and precision manufacturing facility to fulfill the requirements of mechanical components in establishing the manufacturing capability for the successful and self sustained nuclear programme. Centre for Design and Manufacture (CDM) hitherto known as CWS was established in 1964 to cater to the specific requirements of DAE and other associated units like ISRO, DRDO. Since then CDM has made multiple technological achievements and changes towards high quality products. The acquisition of up-to-date machines during High-Tech facility under VIII Plan project and Advance Precision Fabrication facility under IX Plan project has changed the capability of CDM towards CAD, CAM, CAE and CNC machining centres. Considering the rapid growth in the design and manufacturing, it was renamed as Centre for Design and Manufacture in March 2002, with the mission of quality output through group effort and team work

  16. NHI Component Technical Readiness Evaluation System

    International Nuclear Information System (INIS)

    Sherman, S.; Wilson, Dane F.; Pawel, Steven J.

    2007-01-01

    A decision process for evaluating the technical readiness or maturity of components (i.e., heat exchangers, chemical reactors, valves, etc.) for use by the U.S. DOE Nuclear Hydrogen Initiative is described. This system is used by the DOE NHI to assess individual components in relation to their readiness for pilot-scale and larger-scale deployment and to drive the research and development work needed to attain technical maturity. A description of the evaluation system is provided, and examples are given to illustrate how it is used to assist in component R and D decisions.

  17. Development of life evaluation technology for nuclear power plant components

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Jin [Sungkyunkwan Univ., Seoul (Korea, Republic of); Kwon, J. D. [Yeungnam Univ., Gyeongsan (Korea, Republic of); Kang, K. J. [Chonnam National Univ., Gwangju (Korea, Republic of)] (and others)

    2001-03-15

    This research focuses on development of reliable life evaluation technology for nuclear power plant (NPP) components, and is divided into two parts, development of life evaluation systems for pressurized components and evaluation of applicability of emerging technology to operating plants. For the development of life evaluation system for nuclear pressure vessels, the following seven topics are covered: development of expert systems for integrity assessment of pressurized components, development of integrity evaluation systems of steam generator tubes, prediction of failure probability for NPP components based on probabilistic fracture mechanics, development of fatigue damage evaluation technique for plant life extension, domestic round robin analysis for pressurized thermal shock of reactor vessels, domestic round robin analysis of constructing P--T limit curves for reactor vessels, and development of data base for integrity assessment. For evaluation of applicability of emerging technology to operating plants, on the other hand, the following eight topics are covered: applicability of the Leak-Before-Break analysis to Cast S/S piping, collection of aged material tensile and toughness data for aged Cast S/S piping, finite element analyses for load carrying capacity of corroded pipes, development of Risk-based ISI methodology for nuclear piping, collection of toughness data for integrity assessment of bi-metallic joints, applicability of the Master curve concept to reactor vessel integrity assessment, measurement of dynamic fracture toughness, and provision of information related to regulation and plant life extension issues.

  18. Nuclear systems

    CERN Document Server

    Todreas, Neil E

    2011-01-01

    Principal Characteristics of Power ReactorsIntroductionPower CyclesPrimary Coolant SystemsReactor CoresFuel AssembliesAdvanced Water- and Gas-Cooled Reactors (Generation III And III+)Advanced Thermal and Fast Neutron Spectrum Reactors (Generation IV)ReferencesProblemsThermal Design Principles and ApplicationIntroductionOverall Plant Characteristics Influenced by Thermal Hydraulic ConsiderationsEnergy Production and Transfer ParametersThermal Design LimitsThermal Design MarginFigures of Merit for Core Thermal PerformanceThe Inverted Fuel ArrayThe Equivalent Annulus ApproximationReferencesProble

  19. SNR coolant system components

    International Nuclear Information System (INIS)

    De Haas Van Dorsser, A.H.; Mausbeck, H.

    1976-01-01

    The DEBENELUX prototype fast reactor power plant SNR 300 at Kalkar has a loop-type heat transfer system similar to that of the prototype LMFBR plants in the USA and Japan. There exist three 257 MW/sub th/ primary sodium loops, each with a hot leg centrifugal pump and three 85.6 MW/sub th/ intermediate heat exchangers in parallel. From there the heat is transferred to the steam generators via three secondary sodium loops with one cold leg sodium circulating pump in each. At a nominal reactor outlet temperature of 819 0 K and a turbine inlet power of 771 MW/sub th/ super heated steam of 166 bar and 733 0 K is produced, giving rise to a plant rating of 327 MW/sub e/ gross. The primary and secondary loops are described in detail

  20. Generic nuclear power plant component failure data bank

    International Nuclear Information System (INIS)

    Araujo Goes, A.G. de; Gibelli, S.M.O.

    1988-11-01

    This report consist in the development of a generic nuclear power plant component failure data bank. This data bank was implemented in a PC-XT microcomputer, IBM compatible, using the Open Access II program. Generic failure data tables for Westinghouse nuclear power plants and for general PWR power plants are presented. They are the final product of a research which included a preselection and a selection of data collected from the available sources in the library of CNEN (National Nuclear Energy Commission) and from the CIN/CNEN (Neclear Information Center). Futhermore, a proposal of evaluating models of average failure rates of pumps and valves are also presented. Through the electronic data bank one can easily have a generic view of failure rate ranges as well as failure models foe a certain component. It is very importante to develop procedures to collect and store generic failure data that can be quickly accessed, in order to update the Probabilistic Safety Study of Angra-1 and to used in studies which may have component failures of nuclear power plant safety systems. In the future, data specialization can be achieved by means of statistical calculations involving specific data collected from the operational experience of Angra-1 nuclear power plant and the generic data bank. (author) [pt

  1. Development of the magnescope as an instrument for in situ evaluation of steel components of nuclear systems

    International Nuclear Information System (INIS)

    Jiles, D.C.; Bi, Y.; Biner, S.B.

    1997-08-01

    Fatigue damage causes continuous, cumulative microstructural changes in materials and the magnetic properties of steels are sensitive to these microstructural changes. The work therefore focused on the relationship between fatigue damage and the measured magnetic properties of different steels under a variety of fatigue conditions. The project also investigated the feasibility and applicability of magnetic inspection techniques for non-destructive evaluation of fatigue damage. From the results of a series of fatigue tests, conducted on different steels under both low-cycle and high-cycle fatigue conditions, the magnetic properties, such as coercivity, remanence and Barkhausen effect, were found to change systematically with fatigue damage. The magnetic properties showed significant changes, especially during early stage of the fatigue and also at the end of the fatigue lifetime. An approximately linear relationship between the mechanical modulus and magnetic remanence was observed and was explained by a model developed in this study to describe the dynamic changes in magnetic and mechanical properties. The results of this research demonstrated that magnetic measurements are suitable for non-destructive evaluation of fatigue damage in steels such as A533B steel and Cr-Mo steels. The magnetic measurement techniques have been incorporated into instrumentation for in-situ evaluation of steel structures and components

  2. Development of life evaluation technology for nuclear power plant components

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Jin; Kim, Yun Jae; Choi, Jae Boong [Sungkyunkwan Univ., Seoul (Korea, Republic of)] (and others)

    2002-03-15

    This project focuses on developing reliable life evaluation technology for nuclear power plant components, and is divided into two parts, development of a life evaluation system for nuclear pressure vessels and evaluation of applicability of emerging technology to operating plants. For the development of life evaluation system for nuclear pressure vessels, the following seven topics are covered in this project: defect assessment method for steam generator tubes, development of fatigue monitoring system, assessment of corroded pipes, domestic round robin analysis for constructing P-T limit curve for RPV, development of probabilistic integrity assessment technique, effect of aging on strength of dissimilar welds, applicability of LBB to cast stainless steel, and development of probabilistic piping fracture mechanics.

  3. Component reliability for electronic systems

    CERN Document Server

    Bajenescu, Titu-Marius I

    2010-01-01

    The main reason for the premature breakdown of today's electronic products (computers, cars, tools, appliances, etc.) is the failure of the components used to build these products. Today professionals are looking for effective ways to minimize the degradation of electronic components to help ensure longer-lasting, more technically sound products and systems. This practical book offers engineers specific guidance on how to design more reliable components and build more reliable electronic systems. Professionals learn how to optimize a virtual component prototype, accurately monitor product reliability during the entire production process, and add the burn-in and selection procedures that are the most appropriate for the intended applications. Moreover, the book helps system designers ensure that all components are correctly applied, margins are adequate, wear-out failure modes are prevented during the expected duration of life, and system interfaces cannot lead to failure.

  4. Nuclear plant aging research - an overview (electrical and mechanical components)

    International Nuclear Information System (INIS)

    Vora, J.P.

    1985-01-01

    As the operating nuclear power plants advance in age there must be a conscious national and international effort to understand the influence and safety implications of aging and service wear of components and structures in nuclear power plants and develop measures which are practical and cost effective for timely mitigation of aging degradation that could significantly affect plant safety. The Office of Nuclear Regulatory Research has, therefore, initiated a multi-year, multi-disciplinary program on Nuclear Plant Aging Research (NPAR). The overall goals identified for the program are as follows: 1) to identify and characterize aging and service wear effects associated with electrical and mechanical components, interfaces, and systems whose failure could impair plant safety; 2) to identify and recommend methods of inspection, surveillance and condition monitoring of electrical and mechanical components and systems which will be effective in detecting significant aging effects prior to loss of safety function so that timely maintenance and repair or replacement can be implemented; and, 3) to identify and recommend acceptable maintenance practices which can be undertaken to mitigate the effects of aging and to diminish the rate and extent of degradation caused by aging and service wear. The specific research activities to be implemented to achieve these goals are described

  5. Component Reification in Systems Modelling

    DEFF Research Database (Denmark)

    Bendisposto, Jens; Hallerstede, Stefan

    When modelling concurrent or distributed systems in Event-B, we often obtain models where the structure of the connected components is specified by constants. Their behaviour is specified by the non-deterministic choice of event parameters for events that operate on shared variables. From a certain......? These components may still refer to shared variables. Events of these components should not refer to the constants specifying the structure. The non-deterministic choice between these components should not be via parameters. We say the components are reified. We need to address how the reified components get...... reflected into the original model. This reflection should indicate the constraints on how to connect the components....

  6. Implementation of microelectronic components in nuclear application

    International Nuclear Information System (INIS)

    Ashour, M.A

    1997-01-01

    As the next logical step in the evolution of programmable devices, Field programmable interconnect components (FPIC) bring the benefits of programmability to the system-level by enabling totally p rogrammable hardware . Continuing what was started by programmable memories twenty years ago and then enhanced by programmable logic ten years later, programmable interconnect holds the key to complete system programmability. History has shown that flexibility is the key benefit realized by programmable technologies (see figure 1). Initially used in a lab environment for design verification purposes, programmable technologies enhance development and ease of experimentation. As experience by more users is accumulated, performances improves and component prices are reduced, applications rapidly expand to address highly flexible and quickly implemented final manufactured products. With similar attributes of it's programmable predecessors, FPIC technology provides an attractive solution to the design verification problems of today and the manufacturing challenges of tomorrow

  7. Digital Components in Swedish NPP Power Systems

    International Nuclear Information System (INIS)

    Karlsson, Mattias; Eriksson, Tage

    2015-01-01

    Swedish nuclear power plants have over the last 20 years of operation modernised or exchanged several systems and components of the electrical power system. Within these works, new components based on digital technology have been employed in order to realize functionality that was previously achieved by using electro-mechanical or analogue technology. Components and systems such as relay protection, rectifiers, inverters, variable speed drives and diesel-generator sets are today equipped with digital components. Several of the systems and components fulfil functions with a safety-role in the NPP. Recently, however, a number of incidents have occurred which highlight deficiencies in the design or HMI of the equipment, which warrants questions whether there are generic problems with some applications of digital components that needs to be addressed. The use of digital components has presented cost effective solutions, or even the only available solution on the market enabling a modernisation. The vast majority of systems using digital components have been operating without problems and often contribute to improved safety but the challenge of non-detectable, or non-identifiable, failure modes remain. In this paper, the extent to which digital components are used in Swedish NPP power systems will be presented including a description of typical applications. Based on data from maintenance records and fault reports, as well as interviews with designers and maintenance personnel, the main areas where problems have been encountered and where possible risks have been identified will be described. The paper intends to investigate any 'tell-tales' that could give signals of unwanted behaviour. Furthermore, particular benefits experienced by using digital components will be highlighted. The paper will also discuss the safety relevance of these findings and suggest measures to improve safety in the application of digital components in power systems. (authors)

  8. IR-360 nuclear power plant safety functions and component classification

    Energy Technology Data Exchange (ETDEWEB)

    Yousefpour, F., E-mail: fyousefpour@snira.co [Management of Nuclear Power Plant Construction Company (MASNA) (Iran, Islamic Republic of); Shokri, F.; Soltani, H. [Management of Nuclear Power Plant Construction Company (MASNA) (Iran, Islamic Republic of)

    2010-10-15

    The IR-360 nuclear power plant as a 2-loop PWR of 360 MWe power generation capacity is under design in MASNA Company. For design of the IR-360 structures, systems and components (SSCs), the codes and standards and their design requirements must be determined. It is a prerequisite to classify the IR-360 safety functions and safety grade of structures, systems and components correctly for selecting and adopting the suitable design codes and standards. This paper refers to the IAEA nuclear safety codes and standards as well as USNRC standard system to determine the IR-360 safety functions and to formulate the principles of the IR-360 component classification in accordance with the safety philosophy and feature of the IR-360. By implementation of defined classification procedures for the IR-360 SSCs, the appropriate design codes and standards are specified. The requirements of specific codes and standards are used in design process of IR-360 SSCs by design engineers of MASNA Company. In this paper, individual determination of the IR-360 safety functions and definition of the classification procedures and roles are presented. Implementation of this work which is described with example ensures the safety and reliability of the IR-360 nuclear power plant.

  9. IR-360 nuclear power plant safety functions and component classification

    International Nuclear Information System (INIS)

    Yousefpour, F.; Shokri, F.; Soltani, H.

    2010-01-01

    The IR-360 nuclear power plant as a 2-loop PWR of 360 MWe power generation capacity is under design in MASNA Company. For design of the IR-360 structures, systems and components (SSCs), the codes and standards and their design requirements must be determined. It is a prerequisite to classify the IR-360 safety functions and safety grade of structures, systems and components correctly for selecting and adopting the suitable design codes and standards. This paper refers to the IAEA nuclear safety codes and standards as well as USNRC standard system to determine the IR-360 safety functions and to formulate the principles of the IR-360 component classification in accordance with the safety philosophy and feature of the IR-360. By implementation of defined classification procedures for the IR-360 SSCs, the appropriate design codes and standards are specified. The requirements of specific codes and standards are used in design process of IR-360 SSCs by design engineers of MASNA Company. In this paper, individual determination of the IR-360 safety functions and definition of the classification procedures and roles are presented. Implementation of this work which is described with example ensures the safety and reliability of the IR-360 nuclear power plant.

  10. Chemical immobilization of fission products reactive with nuclear reactor components

    International Nuclear Information System (INIS)

    Grossman, L.N.; Kaznoff, A.I.; Clukey, H.V.

    1975-01-01

    This invention teaches a method of immobilizing deleterious fission products produced in nuclear fuel materials during nuclear fission chain reactions through the use of additives. The additives are disposed with the nuclear fuel materials in controlled quantities to form new compositions preventing attack of reactor components, especially nuclear fuel cld, by the deleterious fission products. (Patent Office Record)

  11. Safety surveillance of activities on nuclear pressure components in China

    International Nuclear Information System (INIS)

    Li Ganjie; Li Tianshu; Yan Tianwen

    2005-01-01

    The nuclear pressure components, which perform the nuclear safety functions, are one of the key physical barriers for nuclear safety. For the national strategy on further development of nuclear power and localization of nuclear pressure components, there still exist some problems in preparedness on the localization. As for the technical basis, what can not be overlooked is the management. Aiming at the current problems, National Nuclear Safety Administration (NNSA) has taken measures to strengthen the propagation and popularization of nuclear safety culture, adjust the review and approval policies for nuclear pressure components qualification license, establish more stringent management requirements, and enhance the surveillance of activities on nuclear pressure equipment. Meanwhile, NNSA has improved the internal management and the regulation efficiency on nuclear pressure components. At the same time, with the development and implementation of 'Rules on the Safety Regulation for Nuclear Safety Important Components' to be promulgated by the State Council of China, NNSA will complete and improve the regulation on nuclear pressure components and other nuclear equipment. (authors)

  12. Probabilistic methods in nuclear power plant component ageing analysis

    International Nuclear Information System (INIS)

    Simola, K.

    1992-03-01

    The nuclear power plant ageing research is aimed to ensure that the plant safety and reliability are maintained at a desired level through the designed, and possibly extended lifetime. In ageing studies, the reliability of components, systems and structures is evaluated taking into account the possible time- dependent decrease in reliability. The results of analyses can be used in the evaluation of the remaining lifetime of components and in the development of preventive maintenance, testing and replacement programmes. The report discusses the use of probabilistic models in the evaluations of the ageing of nuclear power plant components. The principles of nuclear power plant ageing studies are described and examples of ageing management programmes in foreign countries are given. The use of time-dependent probabilistic models to evaluate the ageing of various components and structures is described and the application of models is demonstrated with two case studies. In the case study of motor- operated closing valves the analysis are based on failure data obtained from a power plant. In the second example, the environmentally assisted crack growth is modelled with a computer code developed in United States, and the applicability of the model is evaluated on the basis of operating experience

  13. Quality assurance during the manufacture of nuclear power plant components

    International Nuclear Information System (INIS)

    Mueller, J.

    1976-01-01

    Apart from the special requirements of quality assurance in the production of components for the nuclear industry, in particular nuclear power stations, the author discusses special methods of quality control in the testing of welded joints. (TK) [de

  14. Remote nuclear green pellet processing system

    International Nuclear Information System (INIS)

    Cellier, Francis.

    1980-01-01

    An automated system for manufacturing nuclear fuel pellets for use in nuclear fuel elements of nuclear power reactors is described. The system comprises process components arranged vertically but not directly under each other within a single enclosure. The vertical-lateral arrangement provides for gravity flow of the product from one component to the next and for removal of each component without interference with the other components. The single enclosure eliminates time consuming transfer between separate enclosures of each component while providing three-sided access to the component through glove ports. (auth)

  15. Definition of criteria and characteristics for the deterministic evaluation of re-design component suitability for the use in reactor instrumentation and control systems of nuclear power plants

    International Nuclear Information System (INIS)

    Arians, Robert; Arnold, Simone; Lindner, Falk; Mbonjo, Herve; Quester, Claudia; Sommer, Dagmar

    2015-03-01

    Diversity is one of the key concepts in the challenge to improve the robustness of digi-tal instrumentation and control (I and C) systems important to safety against common cause failures. In the cause of this project, a diversity matrix was established that can be used as a basis in the assessment of the diversity of digital I and C systems or their components. The matrix comprises diversity criteria which are structured according to the life cycle of I and C systems and their components, and shows their applicability to the technical components and additional items of a generic digital I and C system.

  16. Component aging and reliability trends in Loviisa Nuclear Power Plant

    International Nuclear Information System (INIS)

    Jankala, K.E.; Vaurio, J.K.

    1989-01-01

    A plant-specific reliability data collection and analysis system has been developed at the Loviisa Nuclear Power Plant to perform tests for component aging and analysis of reliability trends. The system yields both mean values an uncertainty distribution information for reliability parameters to be used in the PSA project underway and in living-PSA applications. Several different trend models are included in the reliability analysis system. Simple analytical expressions have been derived from the parameters of these models, and their variances have been obtained using the information matrix. This paper is focused on the details of the learning/aging models and the estimation of their parameters and statistical accuracies. Applications to the historical data of the Loviisa plant are presented. The results indicate both up- and down-trends in failure rates as well as individuality between nominally identical components

  17. Nuclear information access system

    International Nuclear Information System (INIS)

    Ham, C. H.; Yang, M. H.; Yoon, S. W.

    1998-01-01

    The energy supply in the countries, which have abundant energy resources, may not be affected by accepting the assertion of anti-nuclear and environment groups. Anti-nuclear movements in the countries which have little energy resources may cause serious problem in securing energy supply. Especially, it is distinct in Korea because she heavily depends on nuclear energy in electricity supply(nuclear share in total electricity supply is about 40%).The cause of social trouble surrounding nuclear energy is being involved with various circumstances. However, it is very important that we are not aware of the importance of information access and prepared for such a situation from the early stage of nuclear energy's development. In those matter, this paper analyzes the contents of nuclear information access system in France and Japan which have dynamic nuclear development program and presents the direction of the nuclear access regime through comparing Korean status and referring to progresses of the regime

  18. KHIC's experience in the design and fabrication of nuclear components

    International Nuclear Information System (INIS)

    Suh, S.-C.

    1992-01-01

    Since 1980, Korea Heavy Industries ampersand Construction Company, Ltd. (KHIC) has specialized in the design and equipment supply for nuclear power facilities in Korea. In April 1987, KHIC became the prime contractor for the construction of Yonggwang 3 ampersand 4 (YGN 3 ampersand 4) nuclear power project. Accordingly, KHIC's technological self-reliance capability for the manufacturing processes of the primary system equipment and components has increased from 18% during the initial stage of Yonggwang 1 ampersand 2 (YGN 1 ampersand 2) project to 63% for YGN 3 ampersand 4 project. Self-reliance capability for the secondary system equipment and components has increased from 28% to 84% during the same period of time as well. The ultimate goal is to achieve complete and total assurance that our products are of the finest quality in the nuclear industry in the world market. Henceforth, we will be able to guarantee complete customer satisfaction and reliability of our products with safety assurance and leading edge technology

  19. Overview moderator material for nuclear reactor components

    International Nuclear Information System (INIS)

    Mairing Manutu Pongtuluran; Hendra Prihatnadi

    2009-01-01

    In order for a reactor design is considered acceptable absolute technical requirement is fulfilled because the most important part of a reactor design. Safety considerations emphasis on the handling of radioactive substances emitted during the operation of a reactor and radioactive waste handling. Moderator material is a layer that interacts directly with neutrons split the nuclear fuel that will lead to changes in physical properties, nuclear properties, mechanical properties and chemical properties. Reviews moderator of this time is of the types of moderator is often used to meet the requirements as nuclear material. (author)

  20. Development of stretcher component robots for rescue against nuclear disaster

    International Nuclear Information System (INIS)

    Iwano, Yuki; Osuka, Koichi; Amano, Hisanori

    2006-01-01

    This paper studies the rescue robots to rescue people in an area polluted with radioactive leakage in nuclear power institutions. In particular, we propose the rescue system which consists of a group of small mobile robots. First, small traction robots set the posture of the fainted victims to carry easily, and carry them to the safety space with the mobile robots for the stretcher composition. In this paper, we confirm that the stretcher component robots could transport and convey a 40 [kg] dummy doll. And, we also show an application usage of stretcher robot. (author)

  1. Off-line programming and simulation in handling nuclear components

    International Nuclear Information System (INIS)

    Baker, C.P.

    1993-10-01

    IGRIP was used to create a simulation of the robotic workcell design for handling components at the PANTEX nuclear arms facility. This initial simulation identified problems with the customer's proposed worker layout, and allowed a correction to be proposed. Refinement of the IGRIP simulation allowed the design and construction of a workcell mock-up and accurate off-line programming of the system. IGRIP's off-line programming capabilities are being used to develop the motion control code for the workcell. PNLs success in this area suggests that simulation and off-line programming may be valuable tools for developing robotics in some automation resistant industries

  2. Development of in-service inspection plans for nuclear components at the Surry 1 nuclear power station

    International Nuclear Information System (INIS)

    Vo, T.V.; Simonen, F.A.; Doctor, S.R.; Smith, B.W.; Gore, B.F.

    1993-01-01

    As part of the nondestructive evaluation reliability program sponsored by the US Nuclear Regulatory Commission at Pacific Northwest Laboratory, a methodology has been developed for establishing in-service inspection priorities of nuclear power plant components. The method uses results of probabilistic risk assessment in conjunction with the techniques of failure modes and effects analysis to identify and prioritize the most risk-important systems and components for inspection at nuclear power plants. Surry nuclear power station unit 1 was selected for demonstrating the methodology. The specific systems selected for analysis were the reactor pressure vessel, the reactor coolant, the low pressure injection including the accumulators, and the auxiliary feedwater. The results provide a risk-based ranking of components that can be used to establish a prioritization of the components and a basis for developing improved in-service inspection plans at nuclear power plants

  3. Future nuclear systems technology

    International Nuclear Information System (INIS)

    Brooks, H.

    1979-01-01

    Five directions can be identified for evolution of nuclear systems, possibly a sixth. These are, first, and perhaps most important, toward a means of extending fissile resources through improvement of the efficiency of their use; second, improvements in nuclear safety; third, reduction in the environmental impacts of nuclear electric power generation, particularly water requirements; fourth, improvements in proliferation resistance of the nuclear fuel cycle; and fifth, improvements in economics. And added in a sixth, and somewhat more speculative direction, the use of nuclear power for purposes other than the direct generation of electricity

  4. Power components behavior under nuclear radiations

    International Nuclear Information System (INIS)

    Jaureguy, J.C.; Azais, B.

    1989-01-01

    Many apparatus, either fixed or on-board of vehicles, use power converters. The most common scheme includes chopper with bipolar transistors. In case of nuclear radiations, these equipments may be severely damaged. Depending on the disturbance level, the need for changes in power transistor technology has to be considered or not [fr

  5. Projection and analysis of nuclear components

    International Nuclear Information System (INIS)

    Heeschen, U.

    1980-01-01

    The classification and the types of analysis carried out in pipings for quality control and safety of nuclear power plants, are presented. The operation and emergency conditions with emphasis of possible simplifications of calculations are described. (author/M.C.K.) [pt

  6. Five-Axis Ultrasonic Additive Manufacturing for Nuclear Component Manufacture

    Science.gov (United States)

    Hehr, Adam; Wenning, Justin; Terrani, Kurt; Babu, Sudarsanam Suresh; Norfolk, Mark

    2017-03-01

    Ultrasonic additive manufacturing (UAM) is a three-dimensional metal printing technology which uses high-frequency vibrations to scrub and weld together both similar and dissimilar metal foils. There is no melting in the process and no special atmosphere requirements are needed. Consequently, dissimilar metals can be joined with little to no intermetallic compound formation, and large components can be manufactured. These attributes have the potential to transform manufacturing of nuclear reactor core components such as control elements for the High Flux Isotope Reactor at Oak Ridge National Laboratory. These components are hybrid structures consisting of an outer cladding layer in contact with the coolant with neutron-absorbing materials inside, such as neutron poisons for reactor control purposes. UAM systems are built into a computer numerical control (CNC) framework to utilize intermittent subtractive processes. These subtractive processes are used to introduce internal features as the component is being built and for net shaping. The CNC framework is also used for controlling the motion of the welding operation. It is demonstrated here that curved components with embedded features can be produced using a five-axis code for the welder for the first time.

  7. Modeling the degradation of nuclear components

    International Nuclear Information System (INIS)

    Stock, D.; Samanta, P.; Vesely, W.

    1993-01-01

    This paper describes component level reliability models that use information on degradation to predict component reliability, and which have been used to evaluate different maintenance and testing policies. The models are based on continuous time Markov processes, and are a generalization of reliability models currently used in Probabilistic Risk Assessment. An explanation of the models, the model parameters, and an example of how these models can be used to evaluate maintenance policies are discussed

  8. Inspection and repair of nuclear components

    International Nuclear Information System (INIS)

    Lahner, K.; Poetz, F.

    1993-01-01

    Despite careful design, manufacturing and operation, some of the important safety-relevant components show deterioration with time. Because of activation and contamination of these components, their inspection and repair has to be performed with manipulators. Some sophisticated manipulators are described, built by ABB Reaktor and used for inspection, maintenance and repair of PWR steam generators, fuel alignment pins, core baffle former bolts and reactor pressure vessel head penetrations. (Z.S.) 7 figs

  9. Statistics of Shared Components in Complex Component Systems

    Science.gov (United States)

    Mazzolini, Andrea; Gherardi, Marco; Caselle, Michele; Cosentino Lagomarsino, Marco; Osella, Matteo

    2018-04-01

    Many complex systems are modular. Such systems can be represented as "component systems," i.e., sets of elementary components, such as LEGO bricks in LEGO sets. The bricks found in a LEGO set reflect a target architecture, which can be built following a set-specific list of instructions. In other component systems, instead, the underlying functional design and constraints are not obvious a priori, and their detection is often a challenge of both scientific and practical importance, requiring a clear understanding of component statistics. Importantly, some quantitative invariants appear to be common to many component systems, most notably a common broad distribution of component abundances, which often resembles the well-known Zipf's law. Such "laws" affect in a general and nontrivial way the component statistics, potentially hindering the identification of system-specific functional constraints or generative processes. Here, we specifically focus on the statistics of shared components, i.e., the distribution of the number of components shared by different system realizations, such as the common bricks found in different LEGO sets. To account for the effects of component heterogeneity, we consider a simple null model, which builds system realizations by random draws from a universe of possible components. Under general assumptions on abundance heterogeneity, we provide analytical estimates of component occurrence, which quantify exhaustively the statistics of shared components. Surprisingly, this simple null model can positively explain important features of empirical component-occurrence distributions obtained from large-scale data on bacterial genomes, LEGO sets, and book chapters. Specific architectural features and functional constraints can be detected from occurrence patterns as deviations from these null predictions, as we show for the illustrative case of the "core" genome in bacteria.

  10. Statistics of Shared Components in Complex Component Systems

    Directory of Open Access Journals (Sweden)

    Andrea Mazzolini

    2018-04-01

    Full Text Available Many complex systems are modular. Such systems can be represented as “component systems,” i.e., sets of elementary components, such as LEGO bricks in LEGO sets. The bricks found in a LEGO set reflect a target architecture, which can be built following a set-specific list of instructions. In other component systems, instead, the underlying functional design and constraints are not obvious a priori, and their detection is often a challenge of both scientific and practical importance, requiring a clear understanding of component statistics. Importantly, some quantitative invariants appear to be common to many component systems, most notably a common broad distribution of component abundances, which often resembles the well-known Zipf’s law. Such “laws” affect in a general and nontrivial way the component statistics, potentially hindering the identification of system-specific functional constraints or generative processes. Here, we specifically focus on the statistics of shared components, i.e., the distribution of the number of components shared by different system realizations, such as the common bricks found in different LEGO sets. To account for the effects of component heterogeneity, we consider a simple null model, which builds system realizations by random draws from a universe of possible components. Under general assumptions on abundance heterogeneity, we provide analytical estimates of component occurrence, which quantify exhaustively the statistics of shared components. Surprisingly, this simple null model can positively explain important features of empirical component-occurrence distributions obtained from large-scale data on bacterial genomes, LEGO sets, and book chapters. Specific architectural features and functional constraints can be detected from occurrence patterns as deviations from these null predictions, as we show for the illustrative case of the “core” genome in bacteria.

  11. Coherent systems with multistate components

    International Nuclear Information System (INIS)

    Caldarola, L.

    1980-01-01

    The basic rules of the Boolean algebra with restrictions on variables are briefly recalled. This special type of Boolean algebra allows one to handle fault trees of systems made of multistate (two or more than two states) components. Coherent systems are defined in the case of multistate components. This definition is consistent with that originally suggested by Barlow in the case of binary (two states) components. The basic properties of coherence are described and discussed. Coherent Boolean functions are also defined. It is shown that these functions are irredundant, that is they have only one base which is at the same time complete and irredundant. However, irredundant functions are not necessarily coherent. Finally a simplified algorithm for the calculation of the base of a coherent function is described. In the case that the function is not coherent, the algorithm can be used to reduce the size of the normal disjunctive form of the function. This in turn eases the application of the Nelson algorithm to calculate the complete base of the function. The simplified algorithm has been built in the computer program MUSTAFA-1. In a sample case the use of this algorithm caused a reduction of the CPU time by a factor of about 20. (orig.)

  12. Nuclear reactor trip system

    International Nuclear Information System (INIS)

    Cook, B.M.

    1982-01-01

    Each parameter of the processes of a nuclear reactor and components operatively associated with it is monitored by a set of four like sensors. A trip system normally operates on a ''two out four'' configuration; i.e., to trip the reactor it is necessary that at least two sensors of a set sense an off-normal parameter. This assumes that all sensors are in normal operating condition. However, when a sensor is in test or is subject to maintenance or is defective or disabled, the ''two out of four''configuration would be reduced to a ''one out of three'' configuration because the affected sensor is taken out of service. This would expose the system to the possibility that a single sensor failure, which may be spurious, will cause a trip of the reactor. To prevent this, it is necessary that the affected sensor be bypassed. If only one sensor is bypassed, the system operates on a ''two out of three'' configuration. With two sensors bypassed, the sensing of an off-normal parameter by a third sensor trips the reactor. The by-pass circuit also disables the circuit coupling the by-passed sensor to the trip circuit. (author)

  13. Component Control System for a Vehicle

    Science.gov (United States)

    Fraser-Chanpong, Nathan (Inventor); Spain, Ivan (Inventor); Dawson, Andrew D. (Inventor); Bluethmann, William J. (Inventor); Lee, Chunhao J. (Inventor); Vitale, Robert L. (Inventor); Guo, Raymond (Inventor); Waligora, Thomas M. (Inventor); Akinyode, Akinjide Akinniyi (Inventor); Reed, Ryan M. (Inventor)

    2016-01-01

    A vehicle includes a chassis, a modular component, and a central operating system. The modular component is supported by the chassis. The central operating system includes a component control system, a primary master controller, and a secondary master controller. The component control system is configured for controlling the modular component. The primary and secondary master controllers are in operative communication with the component control system. The primary and secondary master controllers are configured to simultaneously transmit commands to the component control system. The component control system is configured to accept commands from the secondary master controller only when a fault occurs in the primary master controller.

  14. Preinspection of nuclear power plant systems

    International Nuclear Information System (INIS)

    1975-01-01

    The general plans of the systems affecting the safety of the nuclear power plants are accepted by the Institute of Radiation Protection (IRP) on the basis of the preinspection of the systems. This is the prerequisite of the preinspection of the structures and components belonging to these systems. Exceptionally, when separately agreed, the IRP may perform the preinspection of a separate structure or component, although the preinspection documentation of the whole system, e.g. the nuclear heat generating system, has not been accepted. This guide applies to the nuclear power plant systems that have been defined to be preinspected in the classification document accepted by the IRP

  15. In-plant reliability data base for nuclear power plant components: data collection and methodology report

    International Nuclear Information System (INIS)

    Drago, J.P.; Borkowski, R.J.; Pike, D.H.; Goldberg, F.F.

    1982-07-01

    The development of a component reliability data for use in nuclear power plant probabilistic risk assessments and reliabiilty studies is presented in this report. The sources of the data are the in-plant maintenance work request records from a sample of nuclear power plants. This data base is called the In-Plant Reliability Data (IPRD) system. Features of the IPRD system are compared with other data sources such as the Licensee Event Report system, the Nuclear Plant Reliability Data system, and IEEE Standard 500. Generic descriptions of nuclear power plant systems formulated for IPRD are given

  16. Method to chemically decontaminate nuclear reactor components

    International Nuclear Information System (INIS)

    Bertholdt, H.O.

    1984-01-01

    The large decontamination of components of the primary circuit of activated corrosion products in the oxide layer of the structure materials firstly involves an approx. 1 hour oxidation treatment with alkali permanganate solution. Following intermediate rinsing with deionate, they are etched with an inhibited citrate-oxalate solution for 5-20 hours. This is followed by post-treatment with a citric acid/H2O2 solution containing suspended fiber particles. (orig./PW)

  17. Investigation on legislation necessity of qualification of quality assurance auditors for civilian nuclear components

    International Nuclear Information System (INIS)

    Zhu Hong

    2004-01-01

    The paper discusses the actual state and legislation necessity of administration for qualification of quality assurance auditors engaging in nuclear component activities in our country, and presents the tentative idea for establishing qualification system of quality assurance auditors. (author)

  18. Component codification and identification systems

    International Nuclear Information System (INIS)

    Pannenbaecker, K.

    1977-01-01

    The lecture covers the codification in power stations during the erection phase and commercial operation phase. A diagram gives a survey. There are three basic-codifications for application; 1) Kraftwerk-Kennzeichen-System (KKS) for marking each component in orientated systems, for marking electrical orientated positions in cubicals, switch gears etc. and for marking rooms in buildings; 2) Ordnungssystem (OS) for cost calculation and ordering; 3) Unterlagenarten-Schluessel (UAS) for letters, reports etc. and for documentation. The OS is developed on the principle of cost account number and is therefore close to the organization of each supplier and his special form of design and constrution. KKS has only to mark hardware. Therefore all German owners, consultants, authorities and suppliers develop KKS together and conform to it in DIN 407119. (ORU) [de

  19. Safety aspects of nuclear power plant component aging

    International Nuclear Information System (INIS)

    Conte, M.; Deletre, G.; Henry, J.Y.

    1988-01-01

    The safety of nuclear plants depends on the capacity of the systems they are composed to perform the functions they were designed for. The identification and understanding of phenomena liable to degrade this operational capacity thus constitute one of the safety problems for which allowance must be made at the earliest stage of a project. Aging, a natural and hence unavoidable process affecting all the components of an installation, was identified at a very early stage as being one of these phenomena. The investigation and implementation of solutions to the safety problems associated to aging make it necessary to: defining the domain in which the consequences of aging are to be evaluated, identifying the parameters involved, identifying the components sensitive to these parameters, understanding the mechanisms which govern its evolution. The results of qualification tests, and of tests and checks carried out at different stages of construction and operation, as well as allowance for operating experience, constitute the necessary basis for establishing or improving the regulatory requirements. The procedures for validating components and systems of the installation are also drawn up on the basis of these tests. Finally, the actions initiated within the scope of research and development programmes supply the additional data necessary for such validation, and provide the indispensable support for knowledge improvement

  20. Mobile nuclear power systems

    International Nuclear Information System (INIS)

    Andersson, B.

    1988-11-01

    This report is meant to present a general survey of the mobile nuclear power systems and not a detailed review of their technical accomplishments. It is based in published material mainly up to 1987. Mobile nuclear power systems are of two fundamentally different kinds: nuclear reactors and isotopic generators. In the reactors the energy comes from nuclear fission and in the isotopic generators from the radioactive decay of suitable isotopes. The reactors are primarily used as power sourves on board nuclear submarines and other warships but have also been used in the space and in remote places. Their thermal power has ranged from 30 kWth (in a satellite) to 175 MWth (on board an aircraft carrier). Isotopic generators are suitable only for small power demands and have been used on board satellites and spaceprobes, automatic weatherstations, lighthouses and marine installations for navigation and observation. (author)

  1. Cooling system for auxiliary reactor component

    International Nuclear Information System (INIS)

    Fujihira, Tomoko.

    1991-01-01

    A cooling system for auxiliary reactor components comprises three systems, that is, two systems of reactor component cooling water systems (RCCW systems) and a high pressure component cooling water system (HPCCW system). Connecting pipelines having partition valves are intervened each in a cooling water supply pipeline to an emmergency component of each of the RCCW systems, a cooling water return pipeline from the emmergency component of each of the RCCW systems, a cooling water supply pipeline to each of the emmergency components of one of the RCCW system and the HPCCW system and a cooling water return pipeline from each of the emmergency components of one of the RCCW system and the HPCCW system. With such constitution, cooling water can be supplied also to the emmergency components in the stand-by system upon periodical inspection or ISI, thereby enabling to improve the backup performance of the emmergency cooling system. (I.N.)

  2. Nuclear propulsion systems engineering

    International Nuclear Information System (INIS)

    Madsen, W.W.; Neuman, J.E.: Van Haaften, D.H.

    1992-01-01

    The Nuclear Energy for Rocket Vehicle Application (NERVA) program of the 1960's and early 1970's was dramatically successful, with no major failures during the entire testing program. This success was due in large part to the successful development of a systems engineering process. Systems engineering, properly implemented, involves all aspects of the system design and operation, and leads to optimization of theentire system: cost, schedule, performance, safety, reliability, function, requirements, etc. The process must be incorporated from the very first and continued to project completion. This paper will discuss major aspects of the NERVA systems engineering effort, and consider the implications for current nuclear propulsion efforts

  3. Modeling fabrication of nuclear components: An integrative approach

    Energy Technology Data Exchange (ETDEWEB)

    Hench, K.W.

    1996-08-01

    Reduction of the nuclear weapons stockpile and the general downsizing of the nuclear weapons complex has presented challenges for Los Alamos. One is to design an optimized fabrication facility to manufacture nuclear weapon primary components in an environment of intense regulation and shrinking budgets. This dissertation presents an integrative two-stage approach to modeling the casting operation for fabrication of nuclear weapon primary components. The first stage optimizes personnel radiation exposure for the casting operation layout by modeling the operation as a facility layout problem formulated as a quadratic assignment problem. The solution procedure uses an evolutionary heuristic technique. The best solutions to the layout problem are used as input to the second stage - a simulation model that assesses the impact of competing layouts on operational performance. The focus of the simulation model is to determine the layout that minimizes personnel radiation exposures and nuclear material movement, and maximizes the utilization of capacity for finished units.

  4. The magnet components database system

    International Nuclear Information System (INIS)

    Baggett, M.J.; Leedy, R.; Saltmarsh, C.; Tompkins, J.C.

    1990-01-01

    The philosophy, structure, and usage MagCom, the SSC magnet components database, are described. The database has been implemented in Sybase (a powerful relational database management system) on a UNIX-based workstation at the Superconducting Super Collider Laboratory (SSCL); magnet project collaborators can access the database via network connections. The database was designed to contain the specifications and measured values of important properties for major materials, plus configuration information (specifying which individual items were used in each cable, coil, and magnet) and the test results on completed magnets. These data will facilitate the tracking and control of the production process as well as the correlation of magnet performance with the properties of its constituents. 3 refs., 10 figs

  5. The magnet components database system

    International Nuclear Information System (INIS)

    Baggett, M.J.; Leedy, R.; Saltmarsh, C.; Tompkins, J.C.

    1990-01-01

    The philosophy, structure, and usage of MagCom, the SSC magnet components database, are described. The database has been implemented in Sybase (a powerful relational database management system) on a UNIX-based workstation at the Superconducting Super Collider Laboratory (SSCL); magnet project collaborators can access the database via network connections. The database was designed to contain the specifications and measured values of important properties for major materials, plus configuration information (specifying which individual items were used in each cable, coil, and magnet) and the test results on completed magnets. The data will facilitate the tracking and control of the production process as well as the correlation of magnet performance with the properties of its constituents. 3 refs., 9 figs

  6. Two component memory of Rotstein effect in nuclear emulsions

    International Nuclear Information System (INIS)

    Gushchin, E.M.; Lebedev, A.N.; Somov, S.V.; Timofeev, M.K.; Tipografshchik, G.I.

    1991-01-01

    Two sharply differing memory components - fast and slow -are simultaneously detected during investigation into the controlled mode of fast charged particle detection in simple nuclear emulsions, with the emulsion trace sensitivity, corresponding to these components, being about 5 time different. The value of memory time is T m ≅40 μs for fast memory and T m ≅3.5 ms for the slow one. The detection of two Rotstein effect memory components confirms the correctness of the trap model

  7. Nuclear imaging system

    International Nuclear Information System (INIS)

    Barrett, H.H.; Horrigan, F.A.

    1975-01-01

    This invention relates to a nuclear imaging system for mapping the source of high energy nuclear particles from a living organ which has selectively absorbed a radioactive compound by spatially coding the energy from the source in a Fresnel pattern on a detector and decoding the detector output to prouce an image of the source. The coding is produced by a Fresnel zone plate interposed between the nuclear energy source and the detector whose position is adjustable with respect to the detector to focus the slices of the nuclear source on the detector. By adjusting the zone plate to a plurality of positions, data from a plurality of cross-sectional slices are produced from which a three-dimensional image of the nuclear source may be obtained. (Patent Office Record)

  8. Proof of fatigue strength of ferritic and austenitic nuclear components

    Energy Technology Data Exchange (ETDEWEB)

    Roos, E.; Herter, K.H.; Schuler, X.; Weissenberg, T. [Materialpruefungsanstalt, Univ. Stuttgart (Germany)

    2009-07-01

    For the construction, design and operation of nuclear components and systems the appropriate technical codes and standards provide material data, detailed stress analysis procedures and a design philosophy which guarantees a reliable behaviour of the structural components throughout the specified lifetime. Especially for cyclic stress evaluation the different codes and standards provide different fatigue analyses procedures to be performed considering the various mechanical and thermal loading histories and geometric complexities of the components. For the fatigue design curves used as limiting criteria the influence of different factors like e.g., environment, surface finish and temperature must be taken into consideration in an appropriate way. Fatigue tests were performed with low alloy steels as well as with Nb- and Ti-stabilized German austenitic stainless steels in air and simulated high temperature boiling water reactor environment. The experimental results are compared and valuated with the mean data curves in air as well as with mean data curves under high temperature water environment published in the international literature. (orig.)

  9. Coupling component systems towards systems of systems

    OpenAIRE

    Autran , Frédéric; Auzelle , Jean-Philippe; Cattan , Denise; Garnier , Jean-Luc; Luzeaux , Dominique; Mayer , Frédérique; Peyrichon , Marc; Ruault , Jean-René

    2008-01-01

    International audience; Systems of systems (SoS) are a hot topic in our "fully connected global world". Our aim is not to provide another definition of what SoS are, but rather to focus on the adequacy of reusing standard system architecting techniques within this approach in order to improve performance, fault detection and safety issues in large-scale coupled systems that definitely qualify as SoS, whatever the definition is. A key issue will be to secure the availability of the services pr...

  10. Nuclear power systems: Their safety

    International Nuclear Information System (INIS)

    Myers, L.C.

    1993-01-01

    Mankind utilizes energy in many forms and from a variety of sources. Canada is one of a growing number of countries which have chosen to embrace nuclear-electric generation as a component of their energy systems. As of August 1992 there were 433 power reactors operating in 35 countries and accounting for more than 15% of the world's production of electricity. In 1992, thirteen countries derived at least 25% of their electricity from nuclear units, with France leading at nearly 70%. In the same year, Canada produced about 16% of its electricity from nuclear units. Some 68 power reactors are under construction in 16 countries, enough to expand present generating capacity by close to 20%. No human endeavour carries the guarantee of perfect safety and the question of whether or not nuclear-electric generation represents an 'acceptable' risk to society has long been vigorously debated. Until the events of late April 1986, nuclear safety had indeed been an issue for discussion, for some concern, but not for alarm. The accident at the Chernobyl reactor in the USSR has irrevocably changed all that. This disaster brought the matter of nuclear safety back into the public mind in a dramatic fashion. This paper discusses the issue of safety in complex energy systems and provides brief accounts of some of the most serious reactor accidents which have occurred to date. (author). 7 refs

  11. The maintenance optimization of structural components in nuclear power plants

    International Nuclear Information System (INIS)

    Bryla, P.; Ardorino, F.; Aufort, P.; Jacquot, J.P.; Magne, L.; Pitner, P.; Verite, B.; Villain, B.; Monnier, B.

    1997-10-01

    An optimization process, called 'OMF-Structures', is developed by Electricite de France (EDF) in order to extend the current 'OMF' Reliability Centered Maintenance to piping structural components. The Auxiliary Feedwater System of a 900 MW French nuclear plant has been studied in order to lay the foundations of the method. This paper presents the currently proposed principles of the process. The principles of the OMF-Structures process include 'Risk-Based Inspection' concepts within an RCM process. Two main phases are identified: The purpose of the first phase is to select the risk-significant failure modes and associated elements. This phase consists of two major steps: potential consequences evaluation and reliability performance evaluation. The second phase consists of the definition of preventive maintenance programs for piping elements that are associated with risk-significant failure modes. (author)

  12. Romanian network for structural integrity assessment of nuclear components

    International Nuclear Information System (INIS)

    Roth, Maria; Constantinescu, Dan Mihai; Brad, Sebastian; Ducu, Catalin

    2008-01-01

    Full text: Based of the Romanian option to develop and operate nuclear facilities, using as model the networks created at European level and taking into account the international importance of the structural integrity assessments for lifetime extension of the nuclear components, a national Project started since 2005 in the framework of the National Program 'Research of Excellence', Modulus I 2006-2008, managed by the Ministry of Education and Research. Entitled 'Integrated Network for Structural Integrity Monitoring of Critical Components in Nuclear Facilities', with the acronym RIMIS, the Project had two main objectives: - to elaborate a procedure applicable to the structural integrity assessment of the critical components used in Romanian nuclear facilities; - to integrate the national networking in a similar one, at European level, to enhance the scientific significance of Romanian R and D organizations as well as to increase the contribution to solving one of the major issue of the nuclear field. The paper aimed to present the activities performed in the Romanian institutes, involved in the Project, the final results obtained as part of the R and D activities, including experimental, theoretical and modeling ones regarding structural integrity assessment of nuclear components employed in CANDU type reactors. Also the activity carried out in the framework of the NULIFE network, created at European level of the FP6 Program and sustained by the RIMIS network will be described. (authors)

  13. IAEA Mission to Onagawa Nuclear Power Station to Examine the Performance of Systems, Structures and Components Following the Great East Japanese Earthquake and Tsunami, Onagawa and Tokyo, Japan, 30 July - 11 August 2012. IAEA Mission Report

    International Nuclear Information System (INIS)

    2012-01-01

    To strengthen global nuclear safety, the IAEA Action Plan on Nuclear Safety (1) recommends the use of IAEA technical peer review services for plant safety, in the light of the accident at TEPCO's Fukushima Dai-ichi Nuclear Power Plant, and (2) encourages that Member States promptly use IAEA review services to gather and disseminate information on the performance of their nuclear power plants (NPPs) and the performance of the designed protective measures against site specific extreme natural hazards and to utilize the lessons learned in the enhancement of NPP safety worldwide. The Government of Japan and the IAEA have concurred to deploy a mission to Onagawa Nuclear Power Station (NPS), owned and operated by Tohoku Electric Power Co., Inc. (Tohoku EPCo), with the objective of gathering information, during the course of a two-week period on site. This included collecting data on the performance of the structures, systems and components of the Onagawa NPS, in the 11 March 2011 Great East Japan Earthquake (GEJE) and its major aftershocks, as well as compiling the information gathered in a seismic experience database for future use by the Member States to gauge the performance of their facilities against external hazards. The Onagawa NPS has three boiling water reactors (units); with the first unit operating for the last twenty-eight years. Unit 1 began commercial operation in June 1984. Unit 2 began commercial operation in July 1995 and Unit 3 began commercial operation in January 2002. The three units have a combined electric generation capacity of 2,174 Megawatts. Situated on the eastern coast of Japan facing the Pacific Ocean, the Onagawa NPS was the closest nuclear power station to the epicentre of the enormous M9.0 GEJE. Due to its proximity to the earthquake source, the plant experienced very high levels of ground motion -the strongest shaking that any nuclear power plant has ever experienced from an earthquake. The plant shut down safely. The mission objective

  14. An approach to safety problems relating to ageing of nuclear power plant components

    International Nuclear Information System (INIS)

    Conte, M.; Deletre, G.; Henry, J.Y.; Le Meur, M.

    1989-10-01

    The safety of nuclear power plants, in France, is discussed. The attention is focused on the ageing phenomena, as a potential cause of the degradation of the systems functional capabilities. The allowance for ageing in design and its importance on safety, are analyzed. The understanding of phenomena relating to ageing and the components surveillance, are considered. As the effective ageing on the components of nuclear power plants is not fully understood, technical improvements and more accurate analysis are required

  15. Nuclear component design ontology building based on ASME codes

    International Nuclear Information System (INIS)

    Bao Shiyi; Zhou Yu; He Shuyan

    2005-01-01

    The adoption of ontology analysis in the study of concept knowledge acquisition and representation for the nuclear component design process based on computer-supported cooperative work (CSCW) makes it possible to share and reuse numerous concept knowledge of multi-disciplinary domains. A practical ontology building method is accordingly proposed based on Protege knowledge model in combination with both top-down and bottom-up approaches together with Formal Concept Analysis (FCA). FCA exhibits its advantages in the way it helps establish and improve taxonomic hierarchy of concepts and resolve concept conflict occurred in modeling multi-disciplinary domains. With Protege-3.0 as the ontology building tool, a nuclear component design ontology based ASME codes is developed by utilizing the ontology building method. The ontology serves as the basis to realize concept knowledge sharing and reusing of nuclear component design. (authors)

  16. Quality assurance in nuclear fuel element component supply

    International Nuclear Information System (INIS)

    Jenkins, B.P.

    1987-01-01

    The paper describes the application of Quality Assurance to nuclear fuel element component supply. The Quality Assurance programme includes integrated procurement, purchasing, surveillance and receipt inspection functions. Purchasing policy is based on a consistent preference for competitive tendering. Multiple sourcing is used to encourage competitive pricing and increase security of supply. A receipt inspection facility is maintained to ensure the high product quality levels demanded by the nuclear industry. (U.K.)

  17. Seismic fragility of nuclear power plant components (Phase II)

    International Nuclear Information System (INIS)

    Bandyopadhyay, K.K.; Hofmayer, C.H.; Kassir, M.K.; Pepper, S.E.

    1990-02-01

    As part of the Component Fragility Program which was initiated in FY 1985, three additional equipment classes have been evaluated. This report contains the fragility results and discussions on these equipment classes which are switchgear, I and C panels and relays. Both low and medium voltage switchgear assemblies have been considered and a separate fragility estimate for each type is provided. Test data on cabinets from the nuclear instrumentation/neutron monitoring system, plant/process protection system, solid state protective system and engineered safeguards test system comprise the BNL data base for I and C panels (NSSS). Fragility levels have been determined for various failure modes of switchgear and I ampersand C panels, and the deterministic results are presented in terms of test response spectra. In addition, the test data have been evaluated for estimating the respective probabilistic fragility levels which are expressed in terms of a median value, an uncertainty coefficient, a randomness coefficient and an HCLPF value. Due to a wide variation of relay design and the fragility level, a generic fragility level cannot be established for relays. 7 refs., 13 figs., 12 tabs

  18. Nuclear analysis techniques as a component of thermoluminescence dating

    Energy Technology Data Exchange (ETDEWEB)

    Prescott, J.R.; Hutton, J.T.; Habermehl, M.A. [Adelaide Univ., SA (Australia); Van Moort, J. [Tasmania Univ., Sandy Bay, TAS (Australia)

    1996-12-31

    In luminescence dating, an age is found by first measuring dose accumulated since the event being dated, then dividing by the annual dose rate. Analyses of minor and trace elements performed by nuclear techniques have long formed an essential component of dating. Results from some Australian sites are reported to illustrate the application of nuclear techniques of analysis in this context. In particular, a variety of methods for finding dose rates are compared, an example of a site where radioactive disequilibrium is significant and a brief summary is given of a problem which was not resolved by nuclear techniques. 5 refs., 2 tabs.

  19. Nuclear analysis techniques as a component of thermoluminescence dating

    Energy Technology Data Exchange (ETDEWEB)

    Prescott, J R; Hutton, J T; Habermehl, M A [Adelaide Univ., SA (Australia); Van Moort, J [Tasmania Univ., Sandy Bay, TAS (Australia)

    1997-12-31

    In luminescence dating, an age is found by first measuring dose accumulated since the event being dated, then dividing by the annual dose rate. Analyses of minor and trace elements performed by nuclear techniques have long formed an essential component of dating. Results from some Australian sites are reported to illustrate the application of nuclear techniques of analysis in this context. In particular, a variety of methods for finding dose rates are compared, an example of a site where radioactive disequilibrium is significant and a brief summary is given of a problem which was not resolved by nuclear techniques. 5 refs., 2 tabs.

  20. Nuclear Systems Kilopower Overview

    Science.gov (United States)

    Palac, Don; Gibson, Marc; Mason, Lee; Houts, Michael; McClure, Patrick; Robinson, Ross

    2016-01-01

    The Nuclear Systems Kilopower Project was initiated by NASAs Space Technology Mission Directorate Game Changing Development Program in fiscal year 2015 to demonstrate subsystem-level technology readiness of small space fission power in a relevant environment (Technology Readiness Level 5) for space science and human exploration power needs. The Nuclear Systems Kilopower Project consists of two elements. The primary element is the Kilopower Prototype Test, also called the Kilopower Reactor Using Stirling Technology(KRUSTY) Test. This element consists of the development and testing of a fission ground technology demonstrator of a 1 kWe fission power system. A 1 kWe system matches requirements for some robotic precursor exploration systems and future potential deep space science missions, and also allows a nuclear ground technology demonstration in existing nuclear test facilities at low cost. The second element, the Mars Kilopower Scalability Study, consists of the analysis and design of a scaled-up version of the 1 kWe reference concept to 10 kWe for Mars surface power projected requirements, and validation of the applicability of the KRUSTY experiment to key technology challenges for a 10 kWe system. If successful, these two elements will lead to initiation of planning for a technology demonstration of a 10 kWe fission power capability for Mars surface outpost power.

  1. Canisters and nonfuel components at commercial nuclear reactors

    International Nuclear Information System (INIS)

    Gibbard, K.; Thorpe, J.; Moore, R.S.

    1995-01-01

    The Energy Information Administration of the U.S. Department of Energy (DOE) collects data annually from commercial nuclear power reactors via the Nuclear Fuel Data survey, Form RW-859. Over the past three years, the survey has collected data on the quantities and types of nonfuel components and on the quantities and contents of canisters in storage at reactor sites. This paper focuses on the annual changes in the data, specific implications of these changes, and lessons that should be applied to future revisions of the study. The total number of canisters reported by utilities for each year from 1986 to 1993 is listed. Changes in the quantities of nonfuel components report by General Reactors from 1992 to 1993 are also provided. Comparisons of canister and nonfuel components components data from year to year and from reactor to reactor point out that survey questions on these topics have been interpreted differently by reactor personnel

  2. NUCLEBRAS' installations for tests of nuclear power plants components

    International Nuclear Information System (INIS)

    Vasconcelos Paiva, I.P. de; Horta, J.A.L.; Avelar Esteves, F. de; Pinheiro, R.B.

    1983-05-01

    The reasons for NUCLEBRAS' Nuclear Technology Development Center to implement a laboratory for supporting Brazilian manufacturers, giving to them the means for performing functional tests of industrial products, are presented. A brief description of the facilities under construction: the Components Test Loop and the Facility for Testing N.P.P. Components under Accident Conditions, and of other already in operation, is given, as well as its objectives and main technical characteristics. Some test results already obtained are also presented. (Author) [pt

  3. Nuclebras' installations for performance tests of nuclear power plants components

    International Nuclear Information System (INIS)

    Vasconcelos Paiva, I.P. de; Avelar Esteves, F. de; Horta, J.A.L.; Resende, M.F.R.; Pinheiro, R.B.

    1984-01-01

    The reasons for Nuclebras' Nuclear Technology Development Center to implement a laboratory for supporting Brazilian manufactures, giving to them the means for performing functional tests of industrial products, are presented. A brief description of facilities under construction: the components Test Loop and Facility for Testing N.P.P. components under Accident conditions, and other already in operation, as well as its objectives and main technical characteristics. Some test results had already obtained are also presented. (Author) [pt

  4. Improving nuclear envelope dynamics by EBV BFRF1 facilitates intranuclear component clearance through autophagy.

    Science.gov (United States)

    Liu, Guan-Ting; Kung, Hsiu-Ni; Chen, Chung-Kuan; Huang, Cheng; Wang, Yung-Li; Yu, Cheng-Pu; Lee, Chung-Pei

    2018-02-26

    Although a vesicular nucleocytoplasmic transport system is believed to exist in eukaryotic cells, the features of this pathway are mostly unknown. Here, we report that the BFRF1 protein of the Epstein-Barr virus improves vesicular transport of nuclear envelope (NE) to facilitate the translocation and clearance of nuclear components. BFRF1 expression induces vesicles that selectively transport nuclear components to the cytoplasm. With the use of aggregation-prone proteins as tools, we found that aggregated nuclear proteins are dispersed when these BFRF1-induced vesicles are formed. BFRF1-containing vesicles engulf the NE-associated aggregates, exit through from the NE, and putatively fuse with autophagic vacuoles. Chemical treatment and genetic ablation of autophagy-related factors indicate that autophagosome formation and autophagy-linked FYVE protein-mediated autophagic proteolysis are involved in this selective clearance of nuclear proteins. Remarkably, vesicular transport, elicited by BFRF1, also attenuated nuclear aggregates accumulated in neuroblastoma cells. Accordingly, induction of NE-derived vesicles by BFRF1 facilitates nuclear protein translocation and clearance, suggesting that autophagy-coupled transport of nucleus-derived vesicles can be elicited for nuclear component catabolism in mammalian cells.-Liu, G.-T., Kung, H.-N., Chen, C.-K., Huang, C., Wang, Y.-L., Yu, C.-P., Lee, C.-P. Improving nuclear envelope dynamics by EBV BFRF1 facilitates intranuclear component clearance through autophagy.

  5. Study of wet blasting of components in nuclear power stations

    International Nuclear Information System (INIS)

    Hall, J.

    1999-12-01

    This report looks at the method of wet blasting radioactive components in nuclear power stations. The wet blaster uses pearl shaped glass beads with the dimensions of 150-250 μm mixed with water as blasting media. The improved design, providing outer operator's positions with proper radiation protection and more efficient blasting equipment has resulted in a lesser dose taken by the operators. The main reason to decontaminate components in nuclear power plants is to enable service on these components. On components like valves, pump shafts, pipes etc. oxides form and bind radiation. These components are normally situated at some distance from the reactor core and will mainly suffer from radiation from so called activation products. When a component is to be decontaminated it can be decontaminated to a radioactive level where it will be declassified. This report has found levels ranging from 150-1000 Bq/kg allowing declassification of radioactive materials. This difference is found between different countries and different organisations. The report also looks at the levels of waste generated using wet blasting. This is done by tracking the contamination to determine where it collects. It is either collected in the water treatment plant or collected in the blasting media. At Barsebaeck the waste levels, from de-contaminating nearly 800 components in one year, results in a waste volume of about 0,250 m 3 . This waste consists of low and medium level waste and will cost about 3 600 EURO to store. The conclusions of the report are that wet blasting is an indispensable way to treat contaminated components in modern nuclear power plants. The wet blasting equipment can be improved by using a robot enabling the operators to remotely treat components from the outer operator's positions. There they will benefit from better radiation protection thus further reduce their taken dose. The wet blasting equipment could also be used to better control the levels of radioactivity on

  6. Metal plutonium conversion to components of nuclear reactor fuel

    International Nuclear Information System (INIS)

    Subbotin, V.G.; Panov, A.V.; Mashirev, V.P.

    2000-01-01

    Capabilities of different technologies for plutonium conversion to the fuel components of nuclear reactors are studied. Advantages and shortcomings of aqueous and nonaqueous methods of plutonium treatment are shown. Proposals to combine and coordinate efforts of world scientific and technological community in solving problems concerning plutonium of energetic and weapon origin treatment were put forward. (authors)

  7. Fracture mechanics and fatigue evaluation of nuclear reactor components

    International Nuclear Information System (INIS)

    Mattar Neto, Miguel; Andrade, Arnaldo H.P. de; Maneschy, Eduardo

    1995-01-01

    This paper presents a theoretical study available in the available literature for evaluation the environmental effects on the lifetime of nuclear power plant components. The author's motivation is to provide some technical tools to identify what research development could be done in this area

  8. Metal plutonium conversion to components of nuclear reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Subbotin, V.G.; Panov, A.V. [Russian Federal Nuclear Center, ALL-Russian Science and Research, Institute of Technical Physics, Snezhinsk (Russian Federation); Mashirev, V.P. [ALL-Russian Science and Research Institute of Chemical Technology, Moscow (Russian Federation)

    2000-07-01

    Capabilities of different technologies for plutonium conversion to the fuel components of nuclear reactors are studied. Advantages and shortcomings of aqueous and nonaqueous methods of plutonium treatment are shown. Proposals to combine and coordinate efforts of world scientific and technological community in solving problems concerning plutonium of energetic and weapon origin treatment were put forward. (authors)

  9. Structural integrity monitoring of critical components in nuclear facilities

    International Nuclear Information System (INIS)

    Roth, Maria; Constantinescu, Dan Mihai; Brad, Sebastian; Ducu, Catalin; Malinovschi, Viorel

    2007-01-01

    Full text: The paper presents the results obtained as part of the Project 'Integrated Network for Structural Integrity Monitoring of Critical Components in Nuclear Facilities', RIMIS, a research work underway within the framework of the Ministry of Education and Research Programme 'Research of Excellence'. The main objective of the Project is to constitute a network integrating the national R and D institutes with preoccupations in the structural integrity assessment of critical components in the nuclear facilities operating in Romania, in order to elaborate a specific procedure for this field. The degradation mechanisms of the structural materials used in the CANDU type reactors, operated by Unit 1 and Unit 2 at Cernavoda (pressure tubes, fuel elements sheaths, steam generator tubing) and in the nuclear facilities relating to reactors of this type as, for instance, the Hydrogen Isotopes Separation facility, will be investigated. The development of a flexible procedure will offer the opportunity to extend the applications to other structural materials used in the nuclear field and in the non-nuclear fields as well, in cooperation with other institutes involved in the developed network. The expected results of the project will allow the integration of the network developed at national level in the structures of similar networks operating within the EU, the enhancement of the scientific importance of Romanian R and D organizations as well as the increase of our country's contribution in solving the major issues of the nuclear field. (authors)

  10. 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

  11. Nuclear fuel preheating system

    International Nuclear Information System (INIS)

    Andrea, C.

    1975-01-01

    A nuclear reactor new fuel handling system which conveys new fuel from a fuel preparation room into the reactor containment boundary is described. The handling system is provided with a fuel preheating station which is adaptd to heat the new fuel to reactor refueling temperatures in such a way that the fuel is heated from the top down so that fuel element cladding failure due to thermal expansions is avoided. (U.S.)

  12. Detection of instrument or component failures in a nuclear plant by Luenberger observers

    International Nuclear Information System (INIS)

    Wilburn, N.P.; Colley, R.W.; Alexandro, F.J.; Clark, R.N.

    1985-01-01

    A diagnostic system, which will distinguish between instrument failures (flowmeters, etc.) and component failures (valves, filters, etc.) that show the same symptoms, has been developed for nuclear Plants using Luenberger observers. Luenberger observers are online computer based modules constructed following the technology of Clark [3]. A seventh order model of an FFTF subsystem was constructed using the Advanced Continuous Simulation Language (ACSL) and was used to show through simulation that Luenberger observers can be applied to nuclear systems

  13. Component configuration control system development at EBR-II

    International Nuclear Information System (INIS)

    Monson, L.R.; Stratton, R.C.

    1984-01-01

    One ofthe major programs being pursued by the EBR-II Division of Argonne National Laboratory is to improve the reliability of plant control and protection systems. This effort involves looking closely at the present state of the art and needs associated with plant diagnostic, control and protection systems. One of the areas of development at EBR-II involves a component configuration control system (CCCS). This system is a computerized control and planning aid for the nuclear power operator

  14. Nuclear reactor monitoring system

    International Nuclear Information System (INIS)

    Drummond, C.N.; Bybee, R.T.; Mason, F.L.; Worsham, H.J.

    1976-01-01

    The invention pertains to an improved monitoring system for the neutron flux in a nuclear reactor. It is proposed to combine neutron flux detectors, a thermoelement, and a background radiation detector in one measuring unit. The spatial arrangement of these elements is fixed with great exactness; they are enclosed by an elastic cover and are brought into position in the reactor with the aid of a bent tube. The arrangement has a low failure rate and is easy to maintain. (HP) [de

  15. Nuclear reactor shutdown system

    International Nuclear Information System (INIS)

    Mangus, J.D.; Cooper, M.H.

    1982-01-01

    An improved nuclear reactor shutdown system is described comprising a temperature sensitive device connected to control the electric power supply to a magnetic latch holding a body of a neutron absorbing material. The temperature sensitive device is exposed to the reactor coolant so that when the reactor coolant temperature rises above a specific level, the temperature sensitive device will cause deenergization of the magnetic latch to allow the body of neutron absorbing material to enter the reactor core. (author)

  16. Gyrotron: an ECH system component

    International Nuclear Information System (INIS)

    Loring, C.M.; Eason, H.O.; Kimrey, H.D.; White, T.L.; Jory, H.R.; Evans, S.J.

    1981-01-01

    The gyrotron, or electron-cyclotron maser, in the form of a gyromonotron, is being developed as a source of millimeter wave energy for fusion plasma heating. The characteristics of this high power, high efficiency electron tube are described in terms of the requirements for the beam power supply system, the mechanical support system, the cooling system, the focusing and tuning magnets, and the waveguide system. Requirements of power level and transmission efficiency dictate the use of oversize waveguide. The implications, both to the user and to the interaction mechanisms in the gyrotron, of the use of oversize waveguide are treated. The effects of variations of various operating parameters upon the gyrotron's power output and stability are also discussed. Data from gyrotron development and system operation are used where appropriate

  17. Two component systems: physiological effect of a third component.

    Directory of Open Access Journals (Sweden)

    Baldiri Salvado

    Full Text Available Signal transduction systems mediate the response and adaptation of organisms to environmental changes. In prokaryotes, this signal transduction is often done through Two Component Systems (TCS. These TCS are phosphotransfer protein cascades, and in their prototypical form they are composed by a kinase that senses the environmental signals (SK and by a response regulator (RR that regulates the cellular response. This basic motif can be modified by the addition of a third protein that interacts either with the SK or the RR in a way that could change the dynamic response of the TCS module. In this work we aim at understanding the effect of such an additional protein (which we call "third component" on the functional properties of a prototypical TCS. To do so we build mathematical models of TCS with alternative designs for their interaction with that third component. These mathematical models are analyzed in order to identify the differences in dynamic behavior inherent to each design, with respect to functionally relevant properties such as sensitivity to changes in either the parameter values or the molecular concentrations, temporal responsiveness, possibility of multiple steady states, or stochastic fluctuations in the system. The differences are then correlated to the physiological requirements that impinge on the functioning of the TCS. This analysis sheds light on both, the dynamic behavior of synthetically designed TCS, and the conditions under which natural selection might favor each of the designs. We find that a third component that modulates SK activity increases the parameter space where a bistable response of the TCS module to signals is possible, if SK is monofunctional, but decreases it when the SK is bifunctional. The presence of a third component that modulates RR activity decreases the parameter space where a bistable response of the TCS module to signals is possible.

  18. CANDU nuclear power system

    International Nuclear Information System (INIS)

    1981-01-01

    This report provides a summary of the components that make up a CANDU reactor. Major emphasis is placed on the CANDU 600 MW(e) design. The reasons for CANDU's performance and the inherent safety of the system are also discussed

  19. Study of wet blasting of components in nuclear power stations

    Energy Technology Data Exchange (ETDEWEB)

    Hall, J

    1999-12-01

    This report looks at the method of wet blasting radioactive components in nuclear power stations. The wet blaster uses pearl shaped glass beads with the dimensions of 150-250 {mu}m mixed with water as blasting media. The improved design, providing outer operator's positions with proper radiation protection and more efficient blasting equipment has resulted in a lesser dose taken by the operators. The main reason to decontaminate components in nuclear power plants is to enable service on these components. On components like valves, pump shafts, pipes etc. oxides form and bind radiation. These components are normally situated at some distance from the reactor core and will mainly suffer from radiation from so called activation products. When a component is to be decontaminated it can be decontaminated to a radioactive level where it will be declassified. This report has found levels ranging from 150-1000 Bq/kg allowing declassification of radioactive materials.This difference is found between different countries and different organisations. The report also looks at the levels of waste generated using wet blasting. This is done by tracking the contamination to determine where it collects. It is either collected in the water treatment plant or collected in the blasting media. At Barsebaeck the waste levels, from de-contaminating nearly 800 components in one year, results in a waste volume of about 0,250 m{sup 3}. This waste consists of low and medium level waste and will cost about 3 600 EURO to store. The conclusions of the report are that wet blasting is an indispensable way to treat contaminated components in modern nuclear power plants. The wet blasting equipment can be improved by using a robot enabling the operators to remotely treat components from the outer operator's positions. There they will benefit from better radiation protection thus further reduce their taken dose. The wet blasting equipment could also be used to better control the levels of

  20. On estimating the fracture probability of nuclear graphite components

    International Nuclear Information System (INIS)

    Srinivasan, Makuteswara

    2008-01-01

    The properties of nuclear grade graphites exhibit anisotropy and could vary considerably within a manufactured block. Graphite strength is affected by the direction of alignment of the constituent coke particles, which is dictated by the forming method, coke particle size, and the size, shape, and orientation distribution of pores in the structure. In this paper, a Weibull failure probability analysis for components is presented using the American Society of Testing Materials strength specification for nuclear grade graphites for core components in advanced high-temperature gas-cooled reactors. The risk of rupture (probability of fracture) and survival probability (reliability) of large graphite blocks are calculated for varying and discrete values of service tensile stresses. The limitations in these calculations are discussed from considerations of actual reactor environmental conditions that could potentially degrade the specification properties because of damage due to complex interactions between irradiation, temperature, stress, and variability in reactor operation

  1. Nuclear reactor safety system

    International Nuclear Information System (INIS)

    Ball, R.M.; Roberts, R.C.

    1983-01-01

    The invention provides a safety system for a nuclear reactor which uses a parallel combination of computer type look-up tables each of which receives data on a particular parameter (from transducers located in the reactor system) and each of which produces the functional counterpart of that particular parameter. The various functional counterparts are then added together to form a control signal for shutting down the reactor. The functional counterparts are developed by analysis of experimental thermal and hydraulic data, which are used to form expressions that define safe conditions

  2. Nuclear reactor safety systems

    International Nuclear Information System (INIS)

    Ball, R.M.; Roberts, R.C.

    1980-01-01

    A safety system for shutting down a nuclear reactor under overload conditions is described. The system includes a series of parallel-connected computer memory type look-up tables each of which receives data on a particular reactor parameter and in each of which a precalculated functional value for that parameter is stored indicative of the percentage of maximum reactor load that the parameter contributes. The various functional values corresponding to the actual measured parameters are added together to provide a control signal used to shut down the reactor under overload conditions. (U.K.)

  3. Short-range components of nuclear forces: Experiment versus mythology

    International Nuclear Information System (INIS)

    Kukulin, V. I.; Platonova, M. N.

    2013-01-01

    The present-day situation around the description of various (central, spin-orbit, and tensor) components of short-range nuclear forces is discussed. A traditional picture of these interactions based on the idea of one-meson exchange is contrasted against numerous results of recent experiments. As is shown in the present study, these results often deviate strongly from the predictions of traditional models. One can therefore state that such models are inapplicable to describing short-range nuclear forces and that it is necessary to go over from a traditional description to some alternative QCD-based (or QCD-motivated) picture. This means that, despite the widespread popularity of traditional concepts of short-range nuclear forces and their applicability in many particular cases, these concepts are not more than scientific myths that show their inconsistency when analyzed from the viewpoint of the modern experiment

  4. Some current engineering topics in nuclear power plant components

    International Nuclear Information System (INIS)

    Amana, M.

    1977-01-01

    An analysis based on the principle of fracture mechanics, is presented for several engineering problems occuring in nuclear power plant components. The specific problems covered are: underclad cracking; stress corrosion cracking; cracks in HAZ of nozzle weld; feedwater nozzle corner crack; shift of transition temperature due to neutron irradiation; LWR local-ECC thermal shock experiment; and design and material selection of RPV in terms of fracture mechanics. (B.R.H.)

  5. Nuclear Power Plant Mechanical Component Flooding Fragility Experiments Status

    Energy Technology Data Exchange (ETDEWEB)

    Pope, C. L. [Idaho State Univ., Pocatello, ID (United States); Savage, B. [Idaho State Univ., Pocatello, ID (United States); Johnson, B. [Idaho State Univ., Pocatello, ID (United States); Muchmore, C. [Idaho State Univ., Pocatello, ID (United States); Nichols, L. [Idaho State Univ., Pocatello, ID (United States); Roberts, G. [Idaho State Univ., Pocatello, ID (United States); Ryan, E. [Idaho State Univ., Pocatello, ID (United States); Suresh, S. [Idaho State Univ., Pocatello, ID (United States); Tahhan, A. [Idaho State Univ., Pocatello, ID (United States); Tuladhar, R. [Idaho State Univ., Pocatello, ID (United States); Wells, A. [Idaho State Univ., Pocatello, ID (United States); Smith, C. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-07-24

    This report describes progress on Nuclear Power Plant mechanical component flooding fragility experiments and supporting research. The progress includes execution of full scale fragility experiments using hollow-core doors, design of improvements to the Portal Evaluation Tank, equipment procurement and initial installation of PET improvements, designation of experiments exploiting the improved PET capabilities, fragility mathematical model development, Smoothed Particle Hydrodynamic simulations, wave impact simulation device research, and pipe rupture mechanics research.

  6. Experience on environmental qualification of safety-related components for Darlington Nuclear Generating Station

    International Nuclear Information System (INIS)

    Yu, A.S.; Kukreti, B.M.

    1987-01-01

    The proliferation of Nuclear Power Plant safety concerns has lead to increasing attention over the Environmental Qualification (EQ) of Nuclear Power Plant Safety-Related Components to provide the assurance that the safety related equipment will meet their intended functions during normal operation and postulated accident conditions. The environmental qualification of these components is also a Licensing requirement for Darlington Nuclear Generating Station. This paper provides an overview of EQ and the experience of a pilot project, in the qualification of the Main Moderator System safety-related functions for the Darlington Nuclear Generating Station currently under construction. It addresses the various phases of qualification from the identification of the EQ Safety-Related Components List, definition of location specific service conditions (normal, adbnormal and accident), safety-related functions, Environmental Qualification Assessments and finally, an EQ system summary report for the Main Moderator System. The results of the pilot project are discussed and the methodology reviewed. The paper concludes that the EQ Program developed for Darlington Nuclear Generating Station, as applied to the qualification of the Main Moderator System, contained all the elements necessary in the qualification of safety-related equipment. The approach taken in the qualification of the Moderator safety-related equipment proves to provide a sound framework for the qualification of other safety-related components in the station

  7. In-service inspection of electronics components, circuits and nuclear radiation detectors

    International Nuclear Information System (INIS)

    Darbhe, M.D.

    2002-01-01

    A nuclear reactor is a complex process plant. Like a nuclear power plant, the research reactors also employ various nuclear and process systems, the scope and number of such systems being plant-specific. In-service inspection of these systems is an important requirement and is applied at various levels of their constituent units such as detectors, electronics components, circuits and integrated systems. The sensors used cover a wide range such as neutronic, radiation, process (pressure, temperature, flow, level) and many others. The present discussion is limited to neutronic and radiation detectors. The electronic components used normally consist of passive components like resistors, capacitors, semiconductor components like diodes, transistors, analog integrated circuits and digital integrated circuits and electromagnetic relays, to name a few. In order to have a comprehensive surveillance and ISI plan, over the entire plant life, it is necessary to understand various mechanisms, which degrade the performance of these systems. These are discussed initially and later various ISI methods that are used on component-circuit or system level, to ensure optimum system performance, are discussed. The computerised systems, because of hardware and software considerations, have to be given special attention, and the same are discussed briefly

  8. Aging management and PLEX in Swiss nuclear power plants and prioritization of safety class 2 and 3 components

    International Nuclear Information System (INIS)

    Fuchs, R.; Stejskal, J.

    2000-01-01

    In this presentation ageing management of systems and components important to safety of the Swiss nuclear power plants are presented. Status of electrical components, status of mechanical components as well as status of civil structures are reviewed. The scheme of the high pressure core spray system is included

  9. System diagnostics using qualitative analysis and component functional classification

    International Nuclear Information System (INIS)

    Reifman, J.; Wei, T.Y.C.

    1993-01-01

    A method for detecting and identifying faulty component candidates during off-normal operations of nuclear power plants involves the qualitative analysis of macroscopic imbalances in the conservation equations of mass, energy and momentum in thermal-hydraulic control volumes associated with one or more plant components and the functional classification of components. The qualitative analysis of mass and energy is performed through the associated equations of state, while imbalances in momentum are obtained by tracking mass flow rates which are incorporated into a first knowledge base. The plant components are functionally classified, according to their type, as sources or sinks of mass, energy and momentum, depending upon which of the three balance equations is most strongly affected by a faulty component which is incorporated into a second knowledge base. Information describing the connections among the components of the system forms a third knowledge base. The method is particularly adapted for use in a diagnostic expert system to detect and identify faulty component candidates in the presence of component failures and is not limited to use in a nuclear power plant, but may be used with virtually any type of thermal-hydraulic operating system. 5 figures

  10. Nuclear power system

    International Nuclear Information System (INIS)

    Yampolsky, J.S.; Cavallaro, L.; Paulovich, K.F.; Schleicher, R.W.

    1989-01-01

    This patent describes an inherently safe modular nuclear power system for producing electrical power at acceptable efficiency levels using working fluids at relatively low temperatures and pressures. The system comprising: a reactor module for heating a first fluid; a heat exchanger module for transferring heat from the first fluid to a second fluid; a first piping system effecting flow of the first fluid in a first fluid circuit successively through the reactor module and the heat exchanger module; a power conversion module comprising a turbogenerator driven by the second fluid, and means for cooling the second fluid upon emergence thereof from the turbogenerator; a second piping system comprising means for effecting flow of the second fluid in a second fluid circuit successively through the heat exchanger module and the power conversion module; and a plurality of pits for receiving the modules

  11. Nuclear criticality information system

    International Nuclear Information System (INIS)

    Koponen, B.L.; Hampel, V.E.

    1981-01-01

    The nuclear criticality safety program at LLNL began in the 1950's with a critical measurements program which produced benchmark data until the late 1960's. This same time period saw the rapid development of computer technology useful for both computer modeling of fissile systems and for computer-aided management and display of the computational benchmark data. Database management grew in importance as the amount of information increased and as experimental programs were terminated. Within the criticality safety program at LLNL we began at that time to develop a computer library of benchmark data for validation of computer codes and cross sections. As part of this effort, we prepared a computer-based bibliography of criticality measurements on relatively simple systems. However, it is only now that some of these computer-based resources can be made available to the nuclear criticality safety community at large. This technology transfer is being accomplished by the DOE Technology Information System (TIS), a dedicated, advanced information system. The NCIS database is described

  12. Hybrid solar lighting distribution systems and components

    Science.gov (United States)

    Muhs, Jeffrey D [Lenoir City, TN; Earl, Dennis D [Knoxville, TN; Beshears, David L [Knoxville, TN; Maxey, Lonnie C [Powell, TN; Jordan, John K [Oak Ridge, TN; Lind, Randall F [Lenoir City, TN

    2011-07-05

    A hybrid solar lighting distribution system and components having at least one hybrid solar concentrator, at least one fiber receiver, at least one hybrid luminaire, and a light distribution system operably connected to each hybrid solar concentrator and each hybrid luminaire. A controller operates all components.

  13. Tests of qualification of national components of nuclear power plants under design basis accident

    International Nuclear Information System (INIS)

    Mesquita, A.Z.

    1990-01-01

    With the purpose of qualifying national components of nuclear power plants, whose working must be maintained during and after an accident, the Thermohydraulic Division of CDTN have done tests to check the equipment stability, under Design Basis Accident conditions. Until this moment, the following components were tested: electrical junction boxes (connectors); coating systems for wall, inside cover and steel containment; hydraulics components of personnel and equipment airlock. This work describes the test instalation, the tests performed and its results. The components tested, in a general way, fulfil the specified requirements. (author) [pt

  14. The nuclear reactor systems

    International Nuclear Information System (INIS)

    Bacher, P.

    2008-01-01

    This paper describes the various nuclear reactor systems, starting with the Generation II, then the present development of the Generation III and the stakes and challenges of the future Generation IV. Some have found appropriate to oppose reactor systems or generations one to another, especially by minimizing the enhancements of generation III compared to generation II or by expecting the earth from generation IV (meaning that generation III is already obsolete). In the first part of the document (chapter 2), some keys are given to the reader to develop its proper opinion. Chapter 3 describes more precisely the various reactor systems and generations. Chapter 4 discusses the large industrial manoeuvres around the generation III, and the last chapter gives some economical references, taking into account, for the various means of power generation, the impediments linked to climate protection

  15. Nuclear reactor power supply system

    International Nuclear Information System (INIS)

    Cook, B.M.

    1982-01-01

    The redundant signals from the sensor assemblies measuring the process parameters of a nuclear reactor power supply are transmitted each in its turn to a protection system which operates to actuate the protection apparatus for signals indicating off-process conditions. Each sensor assembly includes a number of like sensors measuring the same parameters. The sets of process signals derived from the sensor assemblies are each in its turn transmitted from the protection system to the control system which impresses control signals on the reactor or its components to counteract the tendency for conditions to drift off-normal status requiring operation of the protection system. A parameter signal selector prevents a parameter signal which differs from the other parameter signals of the set by more than twice the allowable variation from passing to the control system. Test signals are periodically impressed by a test unit on a selected pair of a selection unit and control channels. This arrangement eliminates the possibility that a single component failure which may be spurious will cause an inadvertent trip of the reactor during test. (author)

  16. OCSEGen: Open Components and Systems Environment Generator

    Science.gov (United States)

    Tkachuk, Oksana

    2014-01-01

    To analyze a large system, one often needs to break it into smaller components.To analyze a component or unit under analysis, one needs to model its context of execution, called environment, which represents the components with which the unit interacts. Environment generation is a challenging problem, because the environment needs to be general enough to uncover unit errors, yet precise enough to make the analysis tractable. In this paper, we present a tool for automated environment generation for open components and systems. The tool, called OCSEGen, is implemented on top of the Soot framework. We present the tool's current support and discuss its possible future extensions.

  17. Process information systems in nuclear reprocessing

    International Nuclear Information System (INIS)

    Jaeschke, A.; Keller, H.; Orth, H.

    1987-01-01

    On a production management level, a process information system in a nuclear reprocessing plant (NRP) has to fulfill conventional operating functions and functions for nuclear material surveillance (safeguards). Based on today's state of the art of on-line process control technology, the progress in hardware and software technology allows to introduce more process-specific intelligence into process information systems. Exemplified by an expert-system-aided laboratory management system as component of a NRP process information system, the paper demonstrates that these technologies can be applied already. (DG) [de

  18. Intelligent robot / manipulator systems for NDT of primary components in nuclear power plants; Intelligente Roboter / Handhabungssysteme fuer die Pruefung von Primaerkreiskomponenten in Kernkraftwerken

    Energy Technology Data Exchange (ETDEWEB)

    Dirauf, F.; Gottfried, R.; Bauer, R. [Siemens AG, KWU, Erlangen (Germany)

    1999-08-01

    The inspection robot developed by Siemens KWU for BWR reactor pressure vessel inspection has a mounting plate with exchangeable parts so as to fit into the various profiles of the vertical guiding tracks at the different pressure vessels. The robot is a versatile device also due to its variable cinematic equipment and thus can be used for any task hitherto performed by the available manipulators. For BWR pressure vessel testing, a novel, compact probe system equipped with five radiation assemblies has been combined with the ultrasonic SAPHIR probe. Until now, NDE of the RPV nozzles in BWRs has been carried out from the outside of the component. The newly developed manipulator of Siemens for inspection of the RPV nozzles can be moved to the nozzles either by manipulating arms or by a floating device, and is fixed to the nozzles by means of pneumatic suckers. Due to the modular design, probe arrays can be exchanged according to nozzle size or structural profiles to be tested. The mobile testing robot SISTAR for PWR pressure vessels consists of a floating cylinder platform that is moved under water to the target position by popellers or by ropes. It is self-adjusting for taking horizontal position and is held in position in the center of the RPV by means of a radial arrangement of legs automatically and synchronously extending from the robot. The platform can be equipped with one or two manipulator arms, depending on the testing task. (orig./CB) [Deutsch] Der von Siemens KWU neuentwickelte Pruefroboter fuer SWR-Reaktordruckbehaelter besitzt einen auswechselbaren Grundwagen, so dass er an die unterschiedlichen Profile der vertikalen Fuehrungsschienen in den einzelnen Kraftwerken angepasst werden kann. Er ist damit auch aufgrund seiner variablen Kinematik universell einsetzbar und kann saemtliche herkoemmliche Manipulatoren ersetzen. In Verbindung mit dem US-Pruefgeraet SAPHIR wurde fuer die SWR-Reaktordruckbehaelter ein neuartiges kompaktes Pruefkopfsystem mit fuenf

  19. Canisters and nonfuel components at commercial nuclear reactors

    International Nuclear Information System (INIS)

    Gibbard, K.; Disbrow, J.

    1994-01-01

    This paper discusses detailed data on canisters and nonfuel components (NFC) at US commercial nuclear power reactors. A wide variety of NFC have been reported on the Form RW-859, open-quotes Nuclear Fuel Dataclose quotes survey. They may have been integral with an assembly, noncanistered in baskets, destined for disposal as low-level radioactive waste, or stored in canisters. Similarly, data on the family of canistered spent nuclear fuel (SNF) in storage pools was compiled. Approximately 85 percent of the 40,194 pieces of nonfuel assembly (NFA) hardware reported were integral with an assembly. This represents data submitted by 95 of the 107 reactors in 10 generic assembly classes. In addition, a total of 286 canisters have been reported as being in storage pools as of December 31, 1992. However, an additional 264 open baskets were also reported to contain miscellaneous SNF and nonfuel materials, garbage and debris. All of these 286 canisters meet the dimensional envelope requirements specified for disposal for open-quotes standard fuelclose quotes under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (10 CFR 961); most of the baskets do not

  20. Nuclear reactor refueling system

    International Nuclear Information System (INIS)

    Wade, E.E.

    1978-01-01

    A system for transferring fuel assemblies between a nuclear reactor core and a fuel storage area while the fuel assembies remain completely submerged in a continuous body of coolant is described. The system comprises an in-vessel fuel transfer machine located inside the reactor vessel and an ex-vessel fuel transfer machine located in a fuel storage tank. The in-vessel fuel transfer machine comprises two independently rotatable frames with a pivotable fuel transfer apparatus disposed on the lower rotatable frame. The ex-vessel fuel transfer machine comprises one frame with a pivotable fuel transfer apparatus disposed thereon. The pivotable apparatuses are capable of being aligned with each other to transfer a fuel assembly between the reactor vessel and fuel storage tank while the fuel assembly remains completely submerged in a continuous body of coolant. 9 claims, 7 figures

  1. CORROSION ISSUES ASSOCIATED WITH AUSTENITIC STAINLESS STEEL COMPONENTS USED IN NUCLEAR MATERIALS EXTRACTION AND SEPARATION PROCESSES

    Energy Technology Data Exchange (ETDEWEB)

    Mickalonis, J.; Louthan, M.; Sindelar, R.

    2012-12-17

    This paper illustrated the magnitude of the systems, structures and components used at the Savannah River Site for nuclear materials extraction and separation processes. Corrosion issues, including stress corrosion cracking, pitting, crevice corrosion and other corrosion induced degradation processes are discussed and corrosion mitigation strategies such as a chloride exclusion program and corrosion release testing are also discussed.

  2. Structured Performance Analysis for Component Based Systems

    OpenAIRE

    Salmi , N.; Moreaux , Patrice; Ioualalen , M.

    2012-01-01

    International audience; The Component Based System (CBS) paradigm is now largely used to design software systems. In addition, performance and behavioural analysis remains a required step for the design and the construction of efficient systems. This is especially the case of CBS, which involve interconnected components running concurrent processes. % This paper proposes a compositional method for modeling and structured performance analysis of CBS. Modeling is based on Stochastic Well-formed...

  3. Comparison between Japan and the United States in the frequency of events in equipment and components at nuclear power plants

    International Nuclear Information System (INIS)

    Shimada, Yoshio

    2007-01-01

    The Institute of Nuclear Safety System, Incorporated (INSS) conducted trend analyses until 2005 to compare the frequency of events in certain electrical components and instrumentation components at nuclear power plants between Japan and the United States. The results revealed that events have occurred approximately an order of magnitude less often in Japan than in the United States. This paper compared Japan and the United States in more detail in terms of how often events - events reported under the reporting standards of the Nuclear Information Archive (NUCIA) or the Institute of Nuclear Power Operations (INPO) - occurred in electrical components, instrumentation components and mechanical components at nuclear power plants. The results were as follows: (1) In regard to electrical components and instrumentation components, events have occurred one-eighth less frequently in Japan than in the United States, suggesting that the previous results were correct. (2) Events have occurred more often in mechanical components than electrical components and instrumentation components in both Japan and the United States, and there was a smaller difference in the frequency of events in mechanical components between the two countries. (3) Regarding mechanical components, it was found that events in the pipes for critical systems and equipment, such as reactor coolant systems, emergency core cooling systems, instrument and control systems, ventilating and air-conditioning systems, and turbine equipment, have occurred more often in Japan than in the United States. (4) The above observations suggest that there is little scope for reducing the frequency of events in electrical components and instrumentation components, but that mechanical components such as pipes for main systems like emergency core cooling systems and turbine equipment in the case of PWRs, could be improved by re-examining inspection methods and intervals. (author)

  4. EPRI research on component aging and nuclear plant life extension

    International Nuclear Information System (INIS)

    Sliter, G.E.; Carey, J.J.

    1985-01-01

    This paper first describes several research efforts sponsored by the Electric Power Research Institute (EPRI) that examine the aging degradation of organic materials and the nuclear plant equipment in which they appear. This research includes a compendium of material properties characterizing the effects of thermal and radiation aging, shake table testing to evaluate the effects of aging on the seismic performance of electrical components, and a review of condition monitoring techniques applicable to electrical equipment. Also described is a long-term investigation of natural versus artificial aging using reactor buildings as test beds. The paper then describes how the equipment aging research fits into a broad-scoped EPRI program on nuclear plant life extension. The objective of this program is to provide required information, technology, and guidelines to enable utilities to significantly extend operating life beyond the current 40-year licensed term

  5. Nuclear fuel cycle system analysis

    International Nuclear Information System (INIS)

    Ko, W. I.; Kwon, E. H.; Kim, S. G.; Park, B. H.; Song, K. C.; Song, D. Y.; Lee, H. H.; Chang, H. L.; Jeong, C. J.

    2012-04-01

    The nuclear fuel cycle system analysis method has been designed and established for an integrated nuclear fuel cycle system assessment by analyzing various methodologies. The economics, PR(Proliferation Resistance) and environmental impact evaluation of the fuel cycle system were performed using improved DB, and finally the best fuel cycle option which is applicable in Korea was derived. In addition, this research is helped to increase the national credibility and transparency for PR with developing and fulfilling PR enhancement program. The detailed contents of the work are as follows: 1)Establish and improve the DB for nuclear fuel cycle system analysis 2)Development of the analysis model for nuclear fuel cycle 3)Preliminary study for nuclear fuel cycle analysis 4)Development of overall evaluation model of nuclear fuel cycle system 5)Overall evaluation of nuclear fuel cycle system 6)Evaluate the PR for nuclear fuel cycle system and derive the enhancement method 7)Derive and fulfill of nuclear transparency enhancement method The optimum fuel cycle option which is economical and applicable to domestic situation was derived in this research. It would be a basis for establishment of the long-term strategy for nuclear fuel cycle. This work contributes for guaranteeing the technical, economical validity of the optimal fuel cycle option. Deriving and fulfillment of the method for enhancing nuclear transparency will also contribute to renewing the ROK-U.S Atomic Energy Agreement in 2014

  6. Systems integration processes for space nuclear electric propulsion systems

    International Nuclear Information System (INIS)

    Olsen, C.S.; Rice, J.W.; Stanley, M.L.

    1991-01-01

    The various components and subsystems that comprise a nuclear electric propulsion system should be developed and integrated so that each functions ideally and so that each is properly integrated with the other components and subsystems in the optimum way. This paper discusses how processes similar to those used in the development and intergration of the subsystems that comprise the Multimegawatt Space Nuclear Power System concepts can be and are being efficiently and effectively utilized for these purposes. The processes discussed include the development of functional and operational requirements at the system and subsystem level; the assessment of individual nuclear power supply and thruster concepts and their associated technologies; the conduct of systems integration efforts including the evaluation of the mission benefits for each system; the identification and resolution of concepts development, technology development, and systems integration feasibility issues; subsystem, system, and technology development and integration; and ground and flight subsystem and integrated system testing

  7. Software Quality Assurance for Nuclear Safety Systems

    International Nuclear Information System (INIS)

    Sparkman, D R; Lagdon, R

    2004-01-01

    The US Department of Energy has undertaken an initiative to improve the quality of software used to design and operate their nuclear facilities across the United States. One aspect of this initiative is to revise or create new directives and guides associated with quality practices for the safety software in its nuclear facilities. Safety software includes the safety structures, systems, and components software and firmware, support software and design and analysis software used to ensure the safety of the facility. DOE nuclear facilities are unique when compared to commercial nuclear or other industrial activities in terms of the types and quantities of hazards that must be controlled to protect workers, public and the environment. Because of these differences, DOE must develop an approach to software quality assurance that ensures appropriate risk mitigation by developing a framework of requirements that accomplishes the following goals: (sm b ullet) Ensures the software processes developed to address nuclear safety in design, operation, construction and maintenance of its facilities are safe (sm b ullet) Considers the larger system that uses the software and its impacts (sm b ullet) Ensures that the software failures do not create unsafe conditions Software designers for nuclear systems and processes must reduce risks in software applications by incorporating processes that recognize, detect, and mitigate software failure in safety related systems. It must also ensure that fail safe modes and component testing are incorporated into software design. For nuclear facilities, the consideration of risk is not necessarily sufficient to ensure safety. Systematic evaluation, independent verification and system safety analysis must be considered for software design, implementation, and operation. The software industry primarily uses risk analysis to determine the appropriate level of rigor applied to software practices. This risk-based approach distinguishes safety

  8. Ageing study of protection automation components of Olkiluoto nuclear power plant

    International Nuclear Information System (INIS)

    Simola, K.; Haenninen, S.

    1993-07-01

    A study on ageing of reactor protection system of the Olkiluoto nuclear power plant is described. The objective of the study was to present an ageing analysis approach and apply in to the automation chains of reactor protection system of the Olkiluoto nuclear power plant. The study includes the measuring instrumentation, the protection logics, and the control electronics of some pumps and valves. The analysis is based on the information collected on the structure of the system, environmental conditions and maintenance practices of components, and operating experience. Based on this information, the possible ageing effects of equipment and their safety significance are evaluated. (orig.). (15 refs., 16 figs., 12 tabs.)

  9. Project of mechanical components for nuclear power plants

    International Nuclear Information System (INIS)

    Amaral, J.A.R. do; Farias Brito David, D. de

    1984-01-01

    The equipment foreseen to be part of a nuclear power plant must show high quality and safety due to the presence of radioactivity. Besides the perfect functioning during the rigid operating conditions, some postulated loadings are foreseen, like earthquake and loss of coolant accidents, which must be also considered in the design of the components. The design and calculation's concept and development, the interactions with the piping and civil designs, as well as their influences in the licensing process with the authorities are described. (Author) [pt

  10. Design and structural calculation of nuclear power plant mechanical components

    International Nuclear Information System (INIS)

    Amaral, J.A.R. do

    1986-01-01

    The mechanical components of a nuclear power plant must show high quality and safety due to the presence of radioactivity. Besides the perfect functioning during the rigid operating conditions, some postulated loadings are foreseen, like earthquake and loss of coolant accidents, which must be also considered in the design. In this paper, it is intended to describe the design and structural calculations concept and development, the interactions with the piping and civil designs, as well as their influences in the licensing process with the authorities. (Author) [pt

  11. The fabrication of nozzles for nuclear components by welding

    International Nuclear Information System (INIS)

    Moraes, M.M.; Krausser, P.; Echeverria, J.A.V.

    1986-01-01

    A nozzle with medium outside diameter of 1000 mm and medium thickness of 150 mm composed integrally by deposited metal by submerged-arc (wire S3NiMo1, 0.5mm) was fabricated in NUCLEP. The nondestructive, mechanical, metallographic and chemical testing carried out in a test sample made by the same procedure and welding parameters, showed results according to specifications established for primary components for nuclear power plants, and the tests presented mechanical properties and tenacity better than similar nozzle samples. This nozzle is cheapest concerning to importations, in respecting to its forged similar. (M.C.K.) [pt

  12. Systems with randomly failing repairable components

    DEFF Research Database (Denmark)

    Der Kiureghian, Armen; Ditlevsen, Ove Dalager; Song, Junho

    2005-01-01

    Closed-form expressions are derived for the steady-state availability, mean rate of failure, mean duration of downtime and reliability of a general system with randomly and independently failing repairable components. Component failures are assumed to be homogeneous Poisson events in time and rep...

  13. The 'Pole Nucleaire Bourgogne' for developing the nuclear components industry

    International Nuclear Information System (INIS)

    Kottmann, G.

    2012-01-01

    The 'Pole Nucleaire Bourgogne' (PNB) is a high-technology and heavy industries cluster in Burgundy with an international calling. It aims at innovating, educating and federating in order to place the French nuclear industry in a leading position. PNB gathers 76 small-, and medium-sized enterprises, most of them operating in the metal sector, in design and in the control/measuring sector. The aim of PNB is to make enterprises work and cooperate on specific topics according to their sectors of activities and their skills. PNB has identified 3 domains of strategical innovations: -) ecological manufacturing and durability of heavy components, -) controls for high performance components, and -) maintenance and dismantling techniques in hostile environments. The various industry sectors represented in PNB allows a cross-fertilization between high-tech industries (aeronautics, energy, transportation)

  14. Prevent recurrence of nuclear disaster (4). Future tasks in the field of structure and components

    International Nuclear Information System (INIS)

    Okamoto, Koji; Takagi, Toshiyuki; Ueda, Susumu

    2012-01-01

    Structure and components subcommittee under the special committee of seismic safety of nuclear power stations of the Atomic Energy Society of Japan discussed future activities related with technical problems of seismic design of structures, components and piping system and evaluation of seismic effects in collaboration with the Japan Society of Mechanical Engineers. These problems were arranged by 'logic of seismic safety' and tabulated just enough, and then their roadmap was prepared. This article described selected relevant problems and discussed safety margins of seismic design and their related problems, referring to state of countermeasures and evaluated results of nuclear power stations after Great East Japan Earthquake occurred in March 11, 2011. Main problems were related with seismic safety margins of structure and components, consideration of ground motion index, rationalization and upgrade of seismic design, application of new technology, integrity evaluation of structure and components after or at earthquake, and upgrade of seismic probabilistic risk assessment methodology. (T. Tanaka)

  15. Ranking of risk significant components for the Davis-Besse Component Cooling Water System

    International Nuclear Information System (INIS)

    Seniuk, P.J.

    1994-01-01

    Utilities that run nuclear power plants are responsible for testing pumps and valves, as specified by the American Society of Mechanical Engineers (ASME) that are required for safe shutdown, mitigating the consequences of an accident, and maintaining the plant in a safe condition. These inservice components are tested according to ASME Codes, either the earlier requirements of the ASME Boiler and Pressure Vessel Code, Section XI, or the more recent requirements of the ASME Operation and Maintenance Code, Section IST. These codes dictate test techniques and frequencies regardless of the component failure rate or significance of failure consequences. A probabilistic risk assessment or probabilistic safety assessment may be used to evaluate the component importance for inservice test (IST) risk ranking, which is a combination of failure rate and failure consequences. Resources for component testing during the normal quarterly verification test or postmaintenance test are expensive. Normal quarterly testing may cause component unavailability. Outage testing may increase outage cost with no real benefit. This paper identifies the importance ranking of risk significant components in the Davis-Besse component cooling water system. Identifying the ranking of these risk significant IST components adds technical insight for developing the appropriate test technique and test frequency

  16. Solid State Lighting Reliability Components to Systems

    CERN Document Server

    Fan, XJ

    2013-01-01

    Solid State Lighting Reliability: Components to Systems begins with an explanation of the major benefits of solid state lighting (SSL) when compared to conventional lighting systems including but not limited to long useful lifetimes of 50,000 (or more) hours and high efficacy. When designing effective devices that take advantage of SSL capabilities the reliability of internal components (optics, drive electronics, controls, thermal design) take on critical importance. As such a detailed discussion of reliability from performance at the device level to sub components is included as well as the integrated systems of SSL modules, lamps and luminaires including various failure modes, reliability testing and reliability performance. This book also: Covers the essential reliability theories and practices for current and future development of Solid State Lighting components and systems Provides a systematic overview for not only the state-of-the-art, but also future roadmap and perspectives of Solid State Lighting r...

  17. Fabrication of high performance components for Indian nuclear reactors

    International Nuclear Information System (INIS)

    Jayaraj, R.N.

    2011-01-01

    Nuclear Fuel Complex (NFC), a Unit of the Department of Atomic Energy (DAE) has been engaged for well over three-and-half decades in the manufacture of fuels for Pressurized Heavy Water Reactors (PHWRs) and Boiling Water Reactors (BWRs). All the fuel assembly components, like, fuel clad tubes, end plugs, spacers, spacer grids etc. are also being manufactured at NFC in Zirconium alloy material. Apart from the regular production of these components and finished fuel assemblies, NFC has also been engaged in the production of Zirconium alloy reactor core structurals, like, pressure tubes, calandria tubes, garter springs and reactivity control mechanisms for PHWRs and square channels for BWRs. While all these structural components are produced through standardized flow sheets, there have been continuous innovations carried out in the processes to meet the ever increasing end-use characteristics laid down by the utilities. The paper enumerates various aspects of different technologies developed at NFC for the manufacture of high performance components for reactor applications

  18. Application of NUREG/CR-5999 interim fatigue curves to selected nuclear power plant components

    International Nuclear Information System (INIS)

    Ware, A.G.; Morton, D.K.; Nitzel, M.E.

    1995-03-01

    Recent test data indicate that the effects of the light water reactor (LWR) environment could significantly reduce the fatigue resistance of materials used in the reactor coolant pressure boundary components of operating nuclear power plants. Argonne National Laboratory has developed interim fatigue curves based on test data simulating LWR conditions, and published them in NUREG/CR-5999. In order to assess the significance of these interim fatigue curves, fatigue evaluations of a sample of the components in the reactor coolant pressure boundary of LWRs were performed. The sample consists of components from facilities designed by each of the four U.S. nuclear steam supply system vendors. For each facility, six locations were studied, including two locations on the reactor pressure vessel. In addition, there are older vintage plants where components of the reactor coolant pressure boundary were designed to codes that did not require an explicit fatigue analysis of the components. In order to assess the fatigue resistance of the older vintage plants, an evaluation was also conducted on selected components of three of these plants. This report discusses the insights gained from the application of the interim fatigue curves to components of seven operating nuclear power plants

  19. Application of environmentally-corrected fatigue curves to nuclear power plant components

    International Nuclear Information System (INIS)

    Ware, A.G.; Morton, D.K.; Nitzel, M.E.

    1996-01-01

    Recent test data indicate that the effects of the light water reactor (LWR) environment could significantly reduce the fatigue resistance of materials used in the reactor coolant pressure boundary components of operating nuclear power plants. Argonne National Laboratory has developed interim fatigue curves based on test data simulating LWR conditions, and published them in NUREG/CR-5999. In order to assess the significance of these interim fatigue curves, fatigue evaluations of a sample of the components in the reactor coolant pressure boundary of LWRs were performed. The sample consists of components from facilities designed by each of the four US nuclear steam supply system vendors. For each facility, six locations were studied including two locations on the reactor pressure vessel. In addition, there are older vintage plants where components of the reactor coolant pressure boundary were designed to codes that did not require an explicit fatigue analysis of the components. In order to assess the fatigue resistance of the older vintage plants, an evaluation was also conducted on selected components of three of these plants. This paper discusses the insights gained from the application of the interim fatigue curves to components of seven operating nuclear power plants

  20. Hot gas path component cooling system

    Science.gov (United States)

    Lacy, Benjamin Paul; Bunker, Ronald Scott; Itzel, Gary Michael

    2014-02-18

    A cooling system for a hot gas path component is disclosed. The cooling system may include a component layer and a cover layer. The component layer may include a first inner surface and a second outer surface. The second outer surface may define a plurality of channels. The component layer may further define a plurality of passages extending generally between the first inner surface and the second outer surface. Each of the plurality of channels may be fluidly connected to at least one of the plurality of passages. The cover layer may be situated adjacent the second outer surface of the component layer. The plurality of passages may be configured to flow a cooling medium to the plurality of channels and provide impingement cooling to the cover layer. The plurality of channels may be configured to flow cooling medium therethrough, cooling the cover layer.

  1. System, structure, and component evaluation for life-cycle management

    International Nuclear Information System (INIS)

    Hanley, N.E.; Banerjee, A.K.; Woods, P.B.; Perrin, J.S.; Marian, F.A.

    1992-01-01

    In recent years, many nuclear organizations and utilities have studied the possibility of extending the service life of nuclear power plants beyond the original license period. From these studies, recommendations have resulted for maintaining the option of future decisions concerning operating license renewal. Several of the recommendations are considered beneficial to the management and operation of nuclear plants in meeting many of their near-term goals. In 1986, Public Service Electric and Gas (PSE and G) concluded that a full-scale nuclear plant license renewal program for their Salem 1 and 2 and Hope Creek nuclear stations was not cost-effective at that time. Rather, it would be better served if the nuclear plant life extension (PLEX) option were maintained for future consideration. To help plan for the life extension option, a strategic 5-yr life cycle management (LCM) program was begun. In support of the LCM program, evaluations for the following Salem structures and components were performed: (1) intake structures, (2) reactor vessel support, (3) containment liner, and (4) containment structure (below grade). This paper discusses the systems, structures, and components (SSC) evaluation methodology and, as an example, discusses the evaluation performed for reactor vessel support

  2. Nuclear system vaporization

    International Nuclear Information System (INIS)

    Bougault, R.; Brou, R.; Colin, J.; Cussol, D.; Durand, D.; Le Brun, C.; Lecolley, J.F.; Lopez, O.; Louvel, M.; Nakagawa, T.; Peter, J.; Regimbart, R.; Steckmeyer, J.C.; Tamain, B.; Vient, E.; Yuasa-Nakagawa, K.; Wieloch, A.

    1998-01-01

    A particular case of the hot nuclei de-excitation is the total nuclear dislocation into light particles (n, p, d, t, 3 He and α). Such events were first observed at bombarding energies lower than 100 MeV/nucleon due to high detection performances of the INDRA multidetector. The light system Ar + Ni was studied at several bombarding energies ranging from 32 to 95 MeV/nucleon. The events associated to a total vaporization of the system occur above the energy threshold of ∼ 50 MeV/nucleon. A study of the form of these events shows that we have essentially two sources. The excitation energy of these sources may be determined by means of the kinematic properties of their de-excitation products. A preliminary study results in excitation energy values of the order 10 - 14 MeV/nucleon. The theoretical calculation based on a statistical model modified to take into account high excitation energies and excited levels in the lightest nuclei predicts that the vaporization of the two partner nuclei in the Ar + Ni system takes place when the excitation energy exceeds 12 MeV/nucleon what is qualitatively in agreement with the values deduced from calorimetric analysis

  3. Nuclear medicine imaging system

    Science.gov (United States)

    Bennett, Gerald W.; Brill, A. Bertrand; Bizais, Yves J. C.; Rowe, R. Wanda; Zubal, I. George

    1986-01-01

    A nuclear medicine imaging system having two large field of view scintillation cameras mounted on a rotatable gantry and being movable diametrically toward or away from each other is disclosed. In addition, each camera may be rotated about an axis perpendicular to the diameter of the gantry. The movement of the cameras allows the system to be used for a variety of studies, including positron annihilation, and conventional single photon emission, as well as static orthogonal dual multi-pinhole tomography. In orthogonal dual multi-pinhole tomography, each camera is fitted with a seven pinhole collimator to provide seven views from slightly different perspectives. By using two cameras at an angle to each other, improved sensitivity and depth resolution is achieved. The computer system and interface acquires and stores a broad range of information in list mode, including patient physiological data, energy data over the full range detected by the cameras, and the camera position. The list mode acquisition permits the study of attenuation as a result of Compton scatter, as well as studies involving the isolation and correlation of energy with a range of physiological conditions.

  4. Canadian programs on understanding and managing aging degradation of nuclear power plant components

    International Nuclear Information System (INIS)

    Chadha, J.A.; Pachner, J.

    1989-06-01

    Maintaining adequate safety and reliability of nuclear power plants and nuclear power plant life assurance and life extension are growing in importance as nuclear plants get older. Age-related degradation of plant components is complex and not fully understood. This paper provides an overview of the Canadian approach and the main activities and their results towards understanding and managing age-related degradation of nuclear power plant components, structures and systems. A number of pro-active programs have been initiated to anticipate, detect and mitigate potential aging degradation at an early stage before any serious impact on plant safety and reliability. These programs include Operational Safety Management Program, Nuclear Plant Life Assurance Program, systematic plant condition assessment, refurbishment and upgrading, post-service examination and testing, equipment qualification, research and development, and participation in the IAEA programs on safety aspects of nuclear power plant aging and life extension. A regulatory policy on nuclear power plants is under development and will be based on the domestic as well as foreign and international studies and experience

  5. Seismic fragility of nuclear power plant components. Phase I

    International Nuclear Information System (INIS)

    Bandyopadhyay, K.K.; Hofmayer, C.H.

    1986-06-01

    As part of the Component Fragility Research Program, sponsored by the US Nuclear Regulatory Commission, BNL is involved in establishing seismic fragility levels for various nuclear power plant equipment by identifying, collecting and analyzing existing test data from various sources. In Phase I of this program, BNL has reviewed approximately seventy test reports to collect fragility or high level test data for switchgears, motor control centers and similar electrical cabinets, valve actuators and numerous electrical devices of various manufacturers and models. This report provides an assessment and evaluation of the data collected in Phase I. The fragility data for medium voltage and low voltage switchgears and motor control centers are analyzed using the test response spectra (TRS) as a measure of the fragility level. The analysis reveals that fragility levels can best be described by a group of TRS curves corresponding to various failure modes. The lower-bound curve indicates the initiation of malfunctioning or structural damage; whereas, the upper-bound curve corresponds to overall failure of the equipment based on known failure modes. High level test data for some components are included in the report. These data indicate that some components are inherently strong and do not exhibit any failure mode even when tested at the vibration limit of a shake table. The common failure modes are identified in the report. The fragility levels determined in this report have been compared with those used in the PRA and Seismic Margin Studies. It appears that the BNL data better correlate with the HCLPF (High Confidence of a Low Probability of Failure) level used in Seismic Margin Studies and can improve this level as high as 60% for certain applications. Specific recommendations are provided for proper application of BNL fragility data to other studies

  6. Nuclear data information system for nuclear materials

    International Nuclear Information System (INIS)

    Fujita, Mitsutane; Noda, Tetsuji; Utsumi, Misako

    1996-01-01

    The conceptual system for nuclear material design is considered and some trials on WWW server with functions of the easily accessible simulation of nuclear reactions are introduced. Moreover, as an example of the simulation on the system using nuclear data, transmutation calculation was made for candidate first wall materials such as 9Cr-2W steel, V-5Cr-5Ti and SiC in SUS316/Li 2 O/H 2 O(SUS), 9Cr-2W/Li 2 O/H 2 O(RAF), V alloy/Li/Be(V), and SiC/Li 2 ZrO 3 /He(SiC) blanket/shield systems based on ITER design model. Neutron spectrum varies with different blanket/shield compositions. The flux of low energy neutrons decreases in order of V< SiC< RAF< SUS blanket/shield systems. Fair amounts of W depletion in 9Cr-2W steel and the increase of Cr content in V-5Cr-5Ti were predicted in SUS or RAF systems. Concentration change in W and Cr is estimated to be suppressed if Li coolant is used in place of water. Helium and hydrogen production are not strongly affected by the different blanket/shield compositions. (author)

  7. Thermionic nuclear reactor systems

    International Nuclear Information System (INIS)

    Kennel, E.B.

    1986-01-01

    Thermionic nuclear reactors can be expected to be candidate space power supplies for power demands ranging from about ten kilowatts to several megawatts. The conventional ''ignited mode'' thermionic fuel element (TFE) is the basis for most reactor designs to date. Laboratory converters have been built and tested with efficiencies in the range of 7-12% for over 10,000 hours. Even longer lifetimes are projected. More advanced capabilities are potentially achievable in other modes of operation, such as the self-pulsed or unignited diode. Coupled with modest improvements in fuel and emitter material performance, the efficiency of an advanced thermionic conversion system can be extended to the 15-20% range. Advanced thermionic power systems are expected to be compatible with other advanced features such as: (1) Intrinsic subcritically under accident conditions, ensuring 100% safety upon launch abort; (2) Intrinsic low radiation levels during reactor shutdown, allowing manned servicing and/or rendezvous; (3) DC to DC power conditioning using lightweight power MOSFETS; and (4) AC output using pulsed converters

  8. Sensor Failure Detection of FASSIP System using Principal Component Analysis

    Science.gov (United States)

    Sudarno; Juarsa, Mulya; Santosa, Kussigit; Deswandri; Sunaryo, Geni Rina

    2018-02-01

    In the nuclear reactor accident of Fukushima Daiichi in Japan, the damages of core and pressure vessel were caused by the failure of its active cooling system (diesel generator was inundated by tsunami). Thus researches on passive cooling system for Nuclear Power Plant are performed to improve the safety aspects of nuclear reactors. The FASSIP system (Passive System Simulation Facility) is an installation used to study the characteristics of passive cooling systems at nuclear power plants. The accuracy of sensor measurement of FASSIP system is essential, because as the basis for determining the characteristics of a passive cooling system. In this research, a sensor failure detection method for FASSIP system is developed, so the indication of sensor failures can be detected early. The method used is Principal Component Analysis (PCA) to reduce the dimension of the sensor, with the Squarred Prediction Error (SPE) and statistic Hotteling criteria for detecting sensor failure indication. The results shows that PCA method is capable to detect the occurrence of a failure at any sensor.

  9. Countermeasure technologies against materials deterioration of nuclear power plant components

    International Nuclear Information System (INIS)

    2004-09-01

    This report was tentative safety standard on countermeasure technologies against materials deterioration of nuclear power plant components issued in 2004 on the base of the testing data obtained until March 2004, which was to be applied for technical evaluation for lifetime management of aged plants and preventive maintenance or repair of neutron irradiated components such as core shrouds and jet pumps. In order to prevent stress corrosion cracks (SCCs) of austenitic stainless steel welds of reactor components, thermal surface modification using laser beams was used on neutron irradiated materials with laser cladding or surface melting process methods by limiting heat input according to amount of accumulated helium so as to prevent crack initiation caused by helium bubble growth and coalescence. Laser cladding method of laser welding using molten sleeve set inside pipe surface to prevent SCCs of nickel-chromium-iron alloy welds, alloy 690 cladding method using tungsten inert gas (TIG) welding to prevent SCCs of nickel-chromium-iron alloy welds for dissimilar joints of pipes, and laser surface solid solution heat treatment method of laser irradiation on surfaces to prevent SCCs of austenitic stainless steel welds were also included as repair technologies. (T. Tanaka)

  10. Probabilistic approaches to life prediction of nuclear plant structural components

    International Nuclear Information System (INIS)

    Villain, B.; Pitner, P.; Procaccia, H.

    1996-01-01

    In the last decade there has been an increasing interest at EDF in developing and applying probabilistic methods for a variety of purposes. In the field of structural integrity and reliability they are used to evaluate the effect of deterioration due to aging mechanisms, mainly on major passive structural components such as steam generators, pressure vessels and piping in nuclear plants. Because there can be numerous uncertainties involved in a assessment of the performance of these structural components, probabilistic methods. The benefits of a probabilistic approach are the clear treatment of uncertainly and the possibility to perform sensitivity studies from which it is possible to identify and quantify the effect of key factors and mitigative actions. They thus provide information to support effective decisions to optimize In-Service Inspection planning and maintenance strategies and for realistic lifetime prediction or reassessment. The purpose of the paper is to discuss and illustrate the methods available at EDF for probabilistic component life prediction. This includes a presentation of software tools in classical, Bayesian and structural reliability, and an application on two case studies (steam generator tube bundle, reactor pressure vessel). (authors)

  11. Probabilistic approaches to life prediction of nuclear plant structural components

    International Nuclear Information System (INIS)

    Villain, B.; Pitner, P.; Procaccia, H.

    1996-01-01

    In the last decade there has been an increasing interest at EDF in developing and applying probabilistic methods for a variety of purposes. In the field of structural integrity and reliability they are used to evaluate the effect of deterioration due to aging mechanisms, mainly on major passive structural components such as steam generators, pressure vessels and piping in nuclear plants. Because there can be numerous uncertainties involved in an assessment of the performance of these structural components, probabilistic methods provide an attractive alternative or supplement to more conventional deterministic methods. The benefits of a probabilistic approach are the clear treatment of uncertainty and the possibility to perform sensitivity studies from which it is possible to identify and quantify the effect of key factors and mitigative actions. They thus provide information to support effective decisions to optimize In-Service Inspection planning and maintenance strategies and for realistic lifetime prediction or reassessment. The purpose of the paper is to discuss and illustrate the methods available at EDF for probabilistic component life prediction. This includes a presentation of software tools in classical, Bayesian and structural reliability, and an application on two case studies (steam generator tube bundle, reactor pressure vessel)

  12. Thermally activated, single component epoxy systems

    KAUST Repository

    Unruh, David A.

    2011-08-23

    A single component epoxy system in which the resin and hardener components found in many two-component epoxies are combined onto the same molecule is described. The single molecule precursor to the epoxy resin contains both multiple epoxide moieties and a diamine held latent by thermally degradable carbamate linkages. These bis-carbamate "single molecule epoxies" have an essentially infinite shelf life and access a significant range in curing temperatures related to the structure of the carbamate linkages used. © 2011 American Chemical Society.

  13. Thermally activated, single component epoxy systems

    KAUST Repository

    Unruh, David A.; Pastine, Stefan J.; Moreton, Jessica C.; Frechet, Jean

    2011-01-01

    A single component epoxy system in which the resin and hardener components found in many two-component epoxies are combined onto the same molecule is described. The single molecule precursor to the epoxy resin contains both multiple epoxide moieties and a diamine held latent by thermally degradable carbamate linkages. These bis-carbamate "single molecule epoxies" have an essentially infinite shelf life and access a significant range in curing temperatures related to the structure of the carbamate linkages used. © 2011 American Chemical Society.

  14. Evaluation and mitigation of the degradation by corrosion in the components of the service water system of a nuclear power plant; Evaluacion y mitigacion de la degradacion por corrosion en los componentes del sistema de agua de servicio de una planta nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Salaices A, E.; Salaices, M.; Ovando, R. [IIE, Av. Reforma 113 Col. Palmira, 62490 Cuernavaca, Morelos (Mexico)]. e-mail: sal@iie.org.mx

    2005-07-01

    One of the main problems that face the nuclear power stations is the degradation by corrosion in the service water systems. The corrosion causes lost substantial in energy generation and a high cost in maintenance and repairs. In this work, the results of a study of the degradation by the MIC mechanisms (microorganisms influenced corrosion), incrustations in heat exchangers and erosion for solid particles in the components of a typical service water system of a nuclear plant are presented. Diverse mitigation options are analyzed for these mechanisms. In the analysis, it was used the CHECWORKS-CWA code to carry out the evaluation of the degradation so much as well as the mitigation of the caused damage. The results are presented in susceptibility indexes and degradation rates component-by-component. A significant decrement could be observed in the susceptibility to MIC when changing the operation conditions of stagnated flow to continuous flow. With respect to the erosion by solid particles, it was found a significant reduction of the damage it when adding filters to the system. Finally, in the case of the heat exchangers, it is shown that one of the more viable options to diminish incrustations and existent calcium deposits it is the reduction of the pH of the service water. (Author)

  15. 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.)

  16. Assessments of nuclear systems

    International Nuclear Information System (INIS)

    Ekholm, R.

    1978-01-01

    Assessments of competing energy systems are gaining increased importance as a means for an optimal choice of energy source for each specific major application considering the growing energy needs and the shortage of supply. However it is important to make sure that the assessments reflect scientific facts rather than private interests. If this is not achieved, scientists will lose credibility and one will lose the basis for political decisions. It is concluded that to accomplish the globally justified needs for thousands of nuclear reactors soon after the year 2000 and to save a maximum of lives with a minimum of environmental impact, emphasis must be put on low energy costs and on a good fuel and capital resource utilization. This goal can be best accomplished by expendient introduction of the fast breeders and of promising advanced reactors. The gas cooled breeder and the high temperature reactor have outstanding short and long terms merits on this respect, but are not enjoying the financial support that they deserve. (UK)

  17. Preloading of bolted connections in nuclear reactor component supports

    Energy Technology Data Exchange (ETDEWEB)

    Yahr, G T

    1984-10-01

    A number of failures of threaded fasteners in nuclear reactor component supports have been reported. Many of those failures were attributed to stress corrosion cracking. This report discusses how stress corrosion cracking can be avoided in bolting by controlling the maximum bolt preloads so that the sustained stresses in the bolts are below the level required to cause stress corrosion cracking. This is a basic departure from ordinary bolted joint design where the only limits on preload are on the minimum preload. Emphasis is placed on the importance of detailed analysis to determine the acceptable range of preload and the selection of a method for measuring the preload that is sufficiently accurate to ensure that the preload is actually within the acceptable range. Procedures for determining acceptable preload range are given, and the accuracy of various methods of measuring preload is discussed.

  18. Preloading of bolted connections in nuclear reactor component supports

    International Nuclear Information System (INIS)

    Yahr, G.T.

    1984-10-01

    A number of failures of threaded fasteners in nuclear reactor component supports have been reported. Many of those failures were attributed to stress corrosion cracking. This report discusses how stress corrosion cracking can be avoided in bolting by controlling the maximum bolt preloads so that the sustained stresses in the bolts are below the level required to cause stress corrosion cracking. This is a basic departure from ordinary bolted joint design where the only limits on preload are on the minimum preload. Emphasis is placed on the importance of detailed analysis to determine the acceptable range of preload and the selection of a method for measuring the preload that is sufficiently accurate to ensure that the preload is actually within the acceptable range. Procedures for determining acceptable preload range are given, and the accuracy of various methods of measuring preload is discussed

  19. Dynamic analysis and qualification test of nuclear components

    International Nuclear Information System (INIS)

    Kim, B.K.; Lee, C.H.; Park, S.H.; Kim, Y.M.; Kim, B.S.; Kim, I.G.; Chung, C.W.; Kim, Y.M.

    1981-01-01

    This report contains the study on the dynamic characteristics of Wolsung fuel rod and on the dynamic balancing of rotating machinery to evaluate the performance of nuclear reactor components. The study on the dynamic characteristics of Wolsung fuel rod was carried out by both experimental and theoretical methods. Forced vibration testing of actual Wolsung fuel rod using sine sweep and sine dwell excitation was conducted to find the dynamic and nonlinear characteristics of the fuel rod. The data obtained by the test were used to analyze the nonlinear impact characteristics of the fuel rod which has a motion-constraint stop in the center of the rod. The parameters used in the test were the input force level of the exciter, the clearance gap between the fuel rod and the motion constraints, and the frequencies. Test results were in good agreement with the analytical results

  20. SIMODIS - a software package for simulating nuclear reactor components

    International Nuclear Information System (INIS)

    Guimaraes, Lamartine; Borges, Eduardo M.

    2000-01-01

    In this paper it is presented the initial development effort in building a nuclear reactor component simulation package. This package was developed to be used in the MATLAB simulation environment. It uses the graphical capabilities from MATLAB and the advantages of compiled languages, as for instance FORTRAN and C ++ . From the MATLAB it takes the facilities for better displaying the calculated results. From the compiled languages it takes processing speed. So far models from reactor core, UTSG and OTSG have been developed. Also, a series a user-friendly graphical interfaces have been developed for the above models. As a by product a set of water and sodium thermal and physical properties have been developed and may be used directly as a function from MATLAB, or by being called from a model, as part of its calculation process. The whole set was named SIMODIS, which stands for SIstema MODular Integrado de Simulacao. (author)

  1. Components and renewal parts in the nuclear power industry

    International Nuclear Information System (INIS)

    Clark, T.F. Jr.

    1986-01-01

    This paper indicates that the nuclear parts industry has been forced to make major investments in time, personnel and financial resources in order to solve short term/emergency procurement problems. What is required, as was previously indicated, is a coordinated industry-wide effort toward long range planning and implementation of a program that addresses these issues. The industry is developing programs directed toward inventory optimization and ''innovative-creative'' financing of manufacturing inventory/work-in-process in an effort to significantly reduce delivery lead times. Product transition, utilization of cancelled plant equipment, equipment qualification programs, and dedication of commercially manufactured/procured parts and components for safety related application continue to be major elements of our program to support current utility requirements

  2. Service life monitoring of the main components at the Temelin nuclear power plant

    International Nuclear Information System (INIS)

    Hahn, J.; Vincour, D.

    2007-01-01

    Knowledge and experience gained from the introduction and periodical implementation of life assessment of the major components of the Temelin nuclear power plant is summarized. The initial Soviet technical design of the plant did not incorporate lifetime monitoring and evaluation, therefore it was completed with demonstrative strength and lifetime calculations from Czech companies. Moreover, a Westinghouse primary circuit diagnosis and monitoring system, including the monitoring of temperature and pressure cycles for low-cycle fatigue evaluation, was installed at the plant. The DIALIFE code for the calculation of mainly the low-cycle fatigue of the key pressure components, was developed and installed subsequently as a superstructure to the monitoring system. (author)

  3. Component reliability criticality or importance metrics for systems with degrading components

    NARCIS (Netherlands)

    Peng, H.; Coit, D.W.; Feng, Q.

    2012-01-01

    This paper proposes two new importance measures: one new importance measure for systems with -independent degrading components, and another one for systems with -correlated degrading components. Importance measures in previous research are inadequate for systems with degrading components because

  4. Understanding Nuclear Safety Culture: A Systemic Approach

    International Nuclear Information System (INIS)

    Afghan, A.N.

    2016-01-01

    The Fukushima accident was a systemic failure (Report by Director General IAEA on the Fukushima Daiichi Accident). Systemic failure is a failure at system level unlike the currently understood notion which regards it as the failure of component and equipment. Systemic failures are due to the interdependence, complexity and unpredictability within systems and that is why these systems are called complex adaptive systems (CAS), in which “attractors” play an important role. If we want to understand the systemic failures we need to understand CAS and the role of these attractors. The intent of this paper is to identify some typical attractors (including stakeholders) and their role within complex adaptive system. Attractors can be stakeholders, individuals, processes, rules and regulations, SOPs etc., towards which other agents and individuals are attracted. This paper will try to identify attractors in nuclear safety culture and influence of their assumptions on safety culture behavior by taking examples from nuclear industry in Pakistan. For example, if the nuclear regulator is an attractor within nuclear safety culture CAS then how basic assumptions of nuclear plant operators and shift in-charges about “regulator” affect their own safety behavior?

  5. LIRA - License Renewal Assistant an expert system advisor for system and component screening

    International Nuclear Information System (INIS)

    Wood, R.M.; DeLuke, R.J.; Lu, Yi; Catron, S.R.

    1992-01-01

    In developing a license renewal application for a nuclear power plant, it is necessary to identify those systems and components for which age-related degradation must be evaluated and addressed in detail. One approach, used in the Monticello Lead Plant project, is to screen all plant systems and components, based on criteria developed by the Nuclear Utility Management and Resources Council (NUMARC). This paper describes an expert system developed as an assistant in the application of the screening methodology. 4 refs., 5 figs., 1 tab

  6. Fault diagnosis of main coolant pump in the nuclear power station based on the principal component analysis

    International Nuclear Information System (INIS)

    Feng Junting; Xu Mi; Wang Guizeng

    2003-01-01

    The fault diagnosis method based on principal component analysis is studied. The fault character direction storeroom of fifteen parameters abnormity is built in the simulation for the main coolant pump of nuclear power station. The measuring data are analyzed, and the results show that it is feasible for the fault diagnosis system of main coolant pump in the nuclear power station

  7. Aseismic foundation system for nuclear power stations

    International Nuclear Information System (INIS)

    Jolivet, F.; Richli, M.

    1977-01-01

    The aseismic foundation system, as described in this paper, is a new development, which makes it possible to build standard nuclear power stations in areas exposed to strong earthquakes. By adopting proven engineering concepts in design and construction of components, great advantages are achieved in the following areas: safety and reliability; efficiency; design schedule; cost. The need for an aseismic foundation system will arise more and more, as a large part of nuclear power station sites are located in highly seismic zones or must meet high intensity earthquake criteria due to the lack of historic data. (Auth.)

  8. Component aging evaluation with expert systems

    International Nuclear Information System (INIS)

    Wiesemann, J.S.; Maguire, H.T. Jr.

    1988-01-01

    The age degradation of components involves a complex relationship between a variety of variables. These relationships are typically modeled using probabilistic and deterministic analyses. These methods depend upon a formal understanding of the underlying degradation mechanisms and a database of experience which allows statistical analyses to extract numerical trends. At present, not all age degradation mechanisms are adequately modeled and available data for age degradation is in most cases insufficient. In addition, these methods tend to focus upon answers to isolated questions (e.g., What is the component failure rate?) rather than the more pertinent questions concerning operations and maintenance (e.g., should the component be replaced at the next outage). Fortunately, knowledge in the form of personal experience does exist which allows plant personnel to make decisions concerning operations and maintenance. This knowledge can be modeled using expert systems. This paper discusses CAGES (Component Aging Expert System). It combines expert rules (heuristics), probabilistic models, and deterministic models to make evaluations of component aging; predict the implications for component life extension, operational readiness, maintenance effectiveness, and safety, and make recommendations for maintenance and operation

  9. Diesel engine management systems and components

    CERN Document Server

    2014-01-01

    This reference book provides a comprehensive insight into todays diesel injection systems and electronic control. It focusses on minimizing emissions and exhaust-gas treatment. Innovations by Bosch in the field of diesel-injection technology have made a significant contribution to the diesel boom. Calls for lower fuel consumption, reduced exhaust-gas emissions and quiet engines are making greater demands on the engine and fuel-injection systems. Contents History of the diesel engine.- Areas of use for diesel engines.- Basic principles of the diesel engine.- Fuels: Diesel fuel.- Fuels: Alternative fuels.- Cylinder-charge control systems.- Basic principles of diesel fuel-injection.- Overview of diesel fuel-injection systems.- Fuel supply to the low pressure stage.- Overview of discrete cylinder systems.- Unit injector system.- Unit pump system.- Overview of common-rail systems.- High pressure components of the common-rail system.- Injection nozzles.- Nozzle holders.- High pressure lines.- Start assist systems.-...

  10. Perspective : component tracking on the Nova system

    International Nuclear Information System (INIS)

    MacDonald, S.

    1999-01-01

    The issue of introducing Component Tracking as a service to natural gas producers, shippers and straddle plant operators was discussed. Approximately 39 companies in the industry were contacted by consultants at Nova Gas Transmission in an effort to assess if introducing this service would add value to individual producers. The numerous implications that may have to be dealt with if Component Tracking is introduced were also described. Component Tracking would provide an equitable approach to the allocation of molecules in the gas stream, and could provide producers with the ability to avoid capital outlay in field plants by alternatively contracting for recovery of the liquids at the straddle plants. Component Tracking is to be voluntary and each shipper would be able to decide whether to utilize the service at each of their receipt points onto the Nova system

  11. Development of a Versatile Ultrasonic Internal Pipe/Vessel Component Monitor for In-Service Inspection of Nuclear Reactor Components

    Energy Technology Data Exchange (ETDEWEB)

    Searfass, Clifford T. [Structural Integrity Associates, Inc., State College, PA (United States); Malinowski, Owen M. [Structural Integrity Associates, Inc., State College, PA (United States); Van Velsor, Jason K. [Structural Integrity Associates, Inc., State College, PA (United States)

    2015-03-22

    The stated goal of this work was to develop a versatile system which could accurately measure vessel and valve internal vibrations and cavitation formation under in-service conditions in nuclear power plants, ultrasonically. The developed technology will benefit the nuclear power generation industry by allowing plant operators to monitor valve and vessel internals during operation. This will help reduce planned outages and plant component failures. During the course of this work, Structural Integrity Associates, Inc. gathered information from industry experts that target vibration amplitudes to be detected should be in the range of 0.001-in to 0.005-in (0.025-mm to 0.127-mm) and target vibration frequency ranges which should be detected were found to be between 0-Hz and 300-Hz. During the performed work, an ultrasonic measuring system was developed which utilized ultrasonic pulse-echo time-of-flight measurements to measure vibration frequency and amplitude. The developed system has been shown to be able to measure vibration amplitudes as low as 0.0008-in (0.020-mm) with vibration frequencies in the range of 17-Hz to 1000-Hz. Therefore, the developed system was able to meet the industry needs for vibration measurement. The developed ultrasonic system was also to be able to measure cavitation formation by monitoring the received ultrasonic time- and frequency-domain signals. This work also demonstrated the survivability of commercially available probes at temperatures up to 300-F for several weeks.

  12. Scottish Nuclear's information systems strategy

    International Nuclear Information System (INIS)

    Inglis, P.

    1991-01-01

    Scottish Nuclear, the company which has owned and operated Scotland's nuclear power generating capacity since privatization, inherited a substantial amount of computer hardware and software from its predecessor, the South of Scotland Electricity Board. Each of the two power stations, Torness and Hunterston, were using Digital Vax clusters as the Scottish Nuclear company was formed. This had a major influence on the information systems strategy which has subsequently been adopted. (UK)

  13. 4+ Dimensional nuclear systems engineering

    International Nuclear Information System (INIS)

    Suh, Kune Y.

    2009-01-01

    Nuclear power plants (NPPs) require massive quantity of data during the design, construction, operation, maintenance and decommissioning stages because of their special features like size, cost, radioactivity, and so forth. The system engineering thus calls for a fully integrated way of managing the information flow spanning their life cycle. This paper proposes digital systems engineering anchored in three dimensional (3D) computer aided design (CAD) models. The signature in the proposal lies with the four plus dimensional (4 + D) Technology TM , a critical know how for digital management. ESSE (Engineering Super Simulation Emulation) features a 4 + D Technology TM for nuclear energy systems engineering. The technology proposed in the 3D space and time plus cost coordinates, i.e. 4 + D, is the backbone of digital engineering in the nuclear systems design and management. Dased on an integrated 3D configuration management system, ESSE consists of solutions JANUS (Junctional Analysis Neodynamic Unit SoftPower), EURUS (Engineering Utilities Research Unit SoftPower), NOTUS (Neosystemic Optimization Technical Unit SoftPower), VENUS (Virtual Engineering Neocybernetic Unit SoftPower) and INUUS (Informative Neographic Utilities Unit SoftPower). NOTUS contributes to reducing the construction cost of the NPPs by optimizing the component manufacturing procedure and the plant construction process. Planning and scheduling construction projects can thus benefit greatly by integrating traditional management techniques with digital process simulation visualization. The 3D visualization of construction processes and the resulting products intrinsically afford most of the advantages realized by incorporating a purely schedule level detail based the 4 + D system. Problems with equipment positioning and manpower congestion in certain areas can be visualized prior to the actual operation, thus preventing accidents and safety problems such as collision between two machines and losses in

  14. Integrating the pastoral component in agricultural systems

    Directory of Open Access Journals (Sweden)

    Paulo César de Faccio Carvalho

    2018-03-01

    Full Text Available ABSTRACT This paper aims to discuss the impact of the introduction of pastures and grazing animals in agricultural systems. For the purposes of this manuscript, we focus on within-farm integrated crop-livestock systems (ICLS, typical of Southern Brazil. These ICLS are designed to create and enhance the synergisms and emergent properties have arisen from agricultural areas where livestock activities are integrated with crops. We show that the introduction of the crop component will affect less the preceding condition than the introduction of the livestock component. While the introduction of crops in pastoral systems represents increasing diversity of the plant component, the introduction of animals would represent the entry of new flows and interactions within the system. Thus, given the new complexity levels achieved from the introduction of grazing, the probability of arising emergent properties is theoretically much higher. However, grazing management is vital in determining the success or failure of such initiative. The grazing intensity practiced during the pasture phase would affect the canopy structure and the forage availability to animals. In adequate and moderate grazing intensities, it is possible to affirm that livestock combined with crops (ICLS has a potential positive impact. As important as the improvements that grazing animals can generate to the soil-plant components, the economic resilience remarkably increases when pasture rotations are introduced compared with purely agriculture systems, particularly in climate-risk situations. Thus, the integration of the pastoral component can enhance the sustainable intensification of food production, but it modifies simple, pure agricultural systems into more complex and knowledge-demanding production systems.

  15. Radiation control system of nuclear power plants

    International Nuclear Information System (INIS)

    Kapisovsky, V.; Kosa, M.; Melichar, Z.; Moravek, J.; Jancik, O.

    1977-01-01

    The SYRAK system is being developed for in-service radiation control of the V-1 nuclear power plant. Its basic components are an EC 1010 computer, a CAMAC system and communication means. The in-service release of radionuclides is measured by fuel can failure detection, by monitoring rare gases in the coolant, by gamma spectrometric coolant monitoring and by iodine isotopes monitoring in stack disposal. (O.K.)

  16. Innovative nuclear energy systems roadmap

    International Nuclear Information System (INIS)

    2007-12-01

    Developing nuclear energy that is sustainable, safe, has little waste by-product, and cannot be proliferated is an extremely vital and pressing issue. To resolve the four issues through free thinking and overall vision, research activities of 'innovative nuclear energy systems' and 'innovative separation and transmutation' started as a unique 21st Century COE Program for nuclear energy called the Innovative Nuclear Energy Systems for Sustainable Development of the World, COE-INES. 'Innovative nuclear energy systems' include research on CANDLE burn-up reactors, lead-cooled fast reactors and using nuclear energy in heat energy. 'Innovative separation and transmutation' include research on using chemical microchips to efficiently separate TRU waste to MA, burning or destroying waste products, or transmuting plutonium and other nuclear materials. Research on 'nuclear technology and society' and 'education' was also added in order for nuclear energy to be accepted into society. COE-INES was a five-year program ending in 2007. But some activities should be continued and this roadmap detailed them as a rough guide focusing inventions and discoveries. This technology roadmap was created for social acceptance and should be flexible to respond to changing times and conditions. (T. Tanaka)

  17. Operating Experience Insights into Pipe Failures for Electro-Hydraulic Control and Instrument Air Systems in Nuclear Power Plant. A Topical Report from the Component Operational Experience, Degradation and Ageing Programme

    International Nuclear Information System (INIS)

    2015-01-01

    Structural integrity of piping systems is important for plant safety and operability. In recognition of this, information on degradation and failure of piping components and systems is collected and evaluated by regulatory agencies, international organisations (e.g. OECD/NEA and IAEA) and industry organisations worldwide to provide systematic feedback for example to reactor regulation and research and development programmes associated with non-destructive examination (NDE) technology, in-service inspection (ISI) programmes, leak-before-break evaluations, risk-informed ISI, and probabilistic safety assessment (PSA) applications involving passive component reliability. Several OECD member countries have agreed to establish the OECD/NEA 'Component Operational Experience, Degradation and Ageing Programme' (CODAP) to encourage multilateral co-operation in the collection and analysis of data relating to degradation and failure of metallic piping and non-piping metallic passive components in commercial nuclear power plants. The scope of the data collection includes service-induced wall thinning, part through-wall cracks, through-wall cracks with and without active leakage, and instances of significant degradation of metallic passive components, including piping pressure boundary integrity. The OECD/NEA Committee on the Safety of Nuclear Installations (CSNI) acts as an umbrella committee of the Project. CODAP is the continuation of the 2002-2011 'OECD/NEA Pipe Failure Data Exchange Project' (OPDE) and the Stress Corrosion Cracking Working Group of the 2006-2010 'OECD/NEA Stress Corrosion Cracking and Cable Ageing Project' (SCAP). OPDE was formally launched in May 2002. Upon completion of the third term (May 2011), the OPDE project was officially closed to be succeeded by CODAP. SCAP was enabled by a voluntary contribution from Japan. It was formally launched in June 2006 and officially closed with an international workshop held in Tokyo in May

  18. Dynamic interaction of components, structure, and foundation of nuclear power facilities

    International Nuclear Information System (INIS)

    Pajuhesh, J.; Hadjian, A.H.

    1977-01-01

    A solution is formulated for the dynamic analysis of structures and components with different stiffness and damping characteristics, including the consideration of soil-structure interaction effects. Composite structures are often analysed approximately, in particular with regards to damping. For example, the reactor and other equipment in nuclear power plant structures are often analysed by assuming them uncoupled from the supporting structures. To achieve a better accuracy, the coupled system is hereby analysed as a composite component-structure-soil system. To demonstrate the assembly technique, two examples are considered: (a) a steel structure sitting on a concrete stem and linked by a steel bridge to another concrete structure, and (b) an actual model of a nuclear power plant containment structure. (Auth.)

  19. Survey of artificial intelligence methods for detection and identification of component faults in nuclear power plants

    International Nuclear Information System (INIS)

    Reifman, J.

    1997-01-01

    A comprehensive survey of computer-based systems that apply artificial intelligence methods to detect and identify component faults in nuclear power plants is presented. Classification criteria are established that categorize artificial intelligence diagnostic systems according to the types of computing approaches used (e.g., computing tools, computer languages, and shell and simulation programs), the types of methodologies employed (e.g., types of knowledge, reasoning and inference mechanisms, and diagnostic approach), and the scope of the system. The major issues of process diagnostics and computer-based diagnostic systems are identified and cross-correlated with the various categories used for classification. Ninety-five publications are reviewed

  20. Detecting and mitigating aging in component cooling water systems

    International Nuclear Information System (INIS)

    Lofaro, R.J.

    1991-01-01

    The time-dependent effects of aging on component cooling water (CCW) systems in nuclear power plants has been studied and documented as part of a research program sponsored by the US Nuclear Regulatory Commission. It was found that age related degradation leads to failures in the CCW system which can result in an increase in system unavailability, if not properly detected and mitigated. To identify effective methods of managing this degradation, information on inspection, monitoring, and maintenance practices currently available was obtained from various operating plants and reviewed. The findings were correlated with the most common aging mechanisms and failure modes and a compilation of aging detection and mitigation practices was formulated. This paper discusses the results of this work

  1. Detecting and mitigating aging in component cooling water systems

    International Nuclear Information System (INIS)

    Lofaro, R.J.; Aggarwal, S.

    1992-01-01

    The time-dependent effects of aging on component cooling water (CCW) systems in nuclear power plants has been studied and documented as part of a research program sponsored by the US Nuclear Regulatory Commission. It was found that age related degradation leads to failures in the CCW system which can result in an increase in system unavailability, if not properly detected and mitigated. To identify effective methods of managing this degradation, information on inspection, monitoring, and maintenance practices currently available was obtained from various operating plants and reviewed. The findings were correlated with the most common aging mechanisms and failure modes, and a compilation of aging detection and mitigation practices was formulated. This paper discusses the results of this work

  2. Aspects for selection of materials and fabrication processes for nuclear component manufacturing

    International Nuclear Information System (INIS)

    Pernstich, K.

    1980-01-01

    For components of the Nuclear steam supply System of Light Water Reactors an extremely high safety standard is required. These requirements only can be met by adequate selection of materials and fabrication processes and their proper application in combination with strict quality assurance and control measurements. A general overview of the basic aspects to be considered in this connection is presented together with an indication of the present state of art for the main materials and fabrication processes. (author) [pt

  3. Nuclear Medicine National Headquarter System

    Data.gov (United States)

    Department of Veterans Affairs — The Nuclear Medicine National HQ System database is a series of MS Excel spreadsheets and Access Database Tables by fiscal year. They consist of information from all...

  4. BUSINESS PROCESS MANAGEMENT SYSTEMS TECHNOLOGY COMPONENTS ANALYSIS

    Directory of Open Access Journals (Sweden)

    Andrea Giovanni Spelta

    2007-05-01

    Full Text Available The information technology that supports the implementation of the business process management appproach is called Business Process Management System (BPMS. The main components of the BPMS solution framework are process definition repository, process instances repository, transaction manager, conectors framework, process engine and middleware. In this paper we define and characterize the role and importance of the components of BPMS's framework. The research method adopted was the case study, through the analysis of the implementation of the BPMS solution in an insurance company called Chubb do Brasil. In the case study, the process "Manage Coinsured Events"" is described and characterized, as well as the components of the BPMS solution adopted and implemented by Chubb do Brasil for managing this process.

  5. System for Cooling of Electronic Components

    Science.gov (United States)

    Vasil'ev, L. L.; Grakovich, L. P.; Dragun, L. A.; Zhuravlev, A. S.; Olekhnovich, V. A.; Rabetskii, M. I.

    2017-01-01

    Results of computational and experimental investigations of heat pipes having a predetermined thermal resistance and a system based on these pipes for air cooling of electronic components and diode assemblies of lasers are presented. An efficient compact cooling system comprising heat pipes with an evaporator having a capillary coating of a caked copper powder and a condenser having a developed outer finning, has been deviced. This system makes it possible to remove, to the ambient air, a heat flow of power more than 300 W at a temperature of 40-50°C.

  6. Security Hardened Cyber Components for Nuclear Power Plants: Phase I SBIR Final Technical Report

    Energy Technology Data Exchange (ETDEWEB)

    Franusich, Michael D. [SpiralGen, Inc., Pittsburgh, PA (United States)

    2016-03-18

    SpiralGen, Inc. built a proof-of-concept toolkit for enhancing the cyber security of nuclear power plants and other critical infrastructure with high-assurance instrumentation and control code. The toolkit is based on technology from the DARPA High-Assurance Cyber Military Systems (HACMS) program, which has focused on applying the science of formal methods to the formidable set of problems involved in securing cyber physical systems. The primary challenges beyond HACMS in developing this toolkit were to make the new technology usable by control system engineers and compatible with the regulatory and commercial constraints of the nuclear power industry. The toolkit, packaged as a Simulink add-on, allows a system designer to assemble a high-assurance component from formally specified and proven blocks and generate provably correct control and monitor code for that subsystem.

  7. A survival analysis on critical components of nuclear power plants

    International Nuclear Information System (INIS)

    Durbec, V.; Pitner, P.; Riffard, T.

    1995-06-01

    Some tubes of heat exchangers of nuclear power plants may be affected by Primary Water Stress Corrosion Cracking (PWSCC) in highly stressed areas. These defects can shorten the lifetime of the component and lead to its replacement. In order to reduce the risk of cracking, a preventive remedial operation called shot peening was applied on the French reactors between 1985 and 1988. To assess and investigate the effects of shot peening, a statistical analysis was carried on the tube degradation results obtained from in service inspection that are regularly conducted using non destructive tests. The statistical method used is based on the Cox proportional hazards model, a powerful tool in the analysis of survival data, implemented in PROC PHRED recently available in SAS/STAT. This technique has a number of major advantages including the ability to deal with censored failure times data and with the complication of time-dependant co-variables. The paper focus on the modelling and a presentation of the results given by SAS. They provide estimate of how the relative risk of degradation changes after peening and indicate for which values of the prognostic factors analyzed the treatment is likely to be most beneficial. (authors). 2 refs., 3 figs., 6 tabs

  8. Radio frequency system for nuclear fusion

    International Nuclear Information System (INIS)

    Kozeki, Shoichiro; Sagawa, Norimoto; Takizawa, Teruhiro

    1987-01-01

    The importance of radio frequency waves has been increasing in the area of nuclear fusion since they are indispensable for heating of plasma, etc. This report outlines radio frequency techniques used for nuclear fusion and describes the development of radio frequency systems (radio frequency plasma heating system and current drive system). Presently, in-depth studies are underway at various research institutes to achieve plasma heating by injection of radio frequency electric power. Three ranges of frequencies, ICRF (ion cyclotron range of frequency), LHRF (lower hybrid range of frequency) and ECRF (electron cyclotron range of frequency), are considered promissing for radio frequency heating. Candidate waves for plasma current driving include ECW (electron cyclotron wave), LHW (lower hybrid wave), MSW (magnetic sound wave), ICW (ion cyclotron wave) with minority component, and FW (fast wave). FW is the greatest in terms of current drive efficiency. In general, a radio frequency system for nuclear fusion consists of a radio frequency power source, transmission/matching circuit component and plasma connection component. (Nogami, K.)

  9. Risk-based priorities for inspection of nuclear pressure boundary components at selected LWRs

    International Nuclear Information System (INIS)

    Vo, T.V.; Simonen, F.A.; Gore, B.F.; Doctor, S.R.; Smith, B.W.

    1990-03-01

    Data from existing probabilistic risk assessments for eight representative nuclear power plants were used to identify and prioritize the most relevant systems to plant safety. The objective was to assess current in-service inspection requirements for pressure boundary systems and components, and to develop recommendations for improvements. This study demonstrates the feasibility of using risk-based methods to develop plant-specific inspection plans. Results for the eight representative plants also indicate generic trends that suggest improvements in current inspection plans now based on priorities set in accordance with code definitions of Class 1, 2, and 3 systems. 2 refs., 4 figs., 5 tabs

  10. Risk-based priorities for inspection of nuclear pressure boundary components at selected LWRs

    International Nuclear Information System (INIS)

    Vo, T.V.; Simonen, F.A.; Gore, B.F.; Doctor, S.R.; Smith, B.W.

    1990-01-01

    Data from existing probabilistic risk assessments for eight representative nuclear power plants were used to identify and prioritize the most relevant systems to plant safety. The objective of this paper is to assess current in-service inspection requirements for pressure boundary systems and components, and to develop recommendations for improvements. This study demonstrates the feasibility of using risk-based methods to develop plant-specific inspection plans. Results for the eight representative plants also indicate generic trends that suggest improvements in current inspection plans now based on priorities set in accordance with code definitions of Class 1, 2, and 3 systems

  11. Cladding Effects on Structural Integrity of Nuclear Components

    International Nuclear Information System (INIS)

    Sattari-Far, Iradi; Andersson, Magnus

    2006-06-01

    different clad components. Measurement of cladding residual stresses in a decommissioned reactor pressure vessel head, which was exposed to service conditions (pressure test, temperature, neutron irradiation, etc.), and the results from the cladding in a cut-out-piece, which did not experience any service or test pressure, basically showed similar profiles. Considering the low scatter and the reproducible data, the hole-drilling technique is recommended in measurement of the peak of the cladding residual stresses. The profile and magnitude of the cladding residual stresses depend mainly upon cladding composition, cladding thickness, clad component geometry and clad component temperature. The peak of the cladding residual stresses is actually about 2-3 mm under the surface of the clad layer, and values in the range of 150 and 500 MPa are reported. Fracture assessments on different clad components at different loading conditions reveal that fracture assessments based on LEFM and ASME Kk curve lead to unrealistic conservative results, and the cladding residual stresses are of importance for surface crack behaviour, especially under cold loads. The NESC projects have shown that the Master Curve methodology can give good predictions of the conducted experiments. It is reasonable to assume a peak value of cladding residual stresses in the whole clad layer to be equal to the yield strength of the cladding material (around 300 MPa) at room temperature. Providing that the clad component has received PWHT, it can be assumed no residual stresses in the underlying base material. For the nuclear pressure vessel, it is also reasonable to assume that the cladding stress free temperature is at the operation temperature of the vessel (around 300 deg C). It has been shown that the cladding residual stresses have negligible influence on subclad crack behaviour in clad components (receiving PWHT). It has also been shown that the crack growth for subclad cracks would be towards the base

  12. Cladding Effects on Structural Integrity of Nuclear Components

    Energy Technology Data Exchange (ETDEWEB)

    Sattari-Far, Iradi; Andersson, Magnus [lnspecta Technology AB, Stockholm (Sweden)

    2006-06-15

    measurement of different clad components. Measurement of cladding residual stresses in a decommissioned reactor pressure vessel head, which was exposed to service conditions (pressure test, temperature, neutron irradiation, etc.), and the results from the cladding in a cut-out-piece, which did not experience any service or test pressure, basically showed similar profiles. Considering the low scatter and the reproducible data, the hole-drilling technique is recommended in measurement of the peak of the cladding residual stresses. The profile and magnitude of the cladding residual stresses depend mainly upon cladding composition, cladding thickness, clad component geometry and clad component temperature. The peak of the cladding residual stresses is actually about 2-3 mm under the surface of the clad layer, and values in the range of 150 and 500 MPa are reported. Fracture assessments on different clad components at different loading conditions reveal that fracture assessments based on LEFM and ASME Kk curve lead to unrealistic conservative results, and the cladding residual stresses are of importance for surface crack behaviour, especially under cold loads. The NESC projects have shown that the Master Curve methodology can give good predictions of the conducted experiments. It is reasonable to assume a peak value of cladding residual stresses in the whole clad layer to be equal to the yield strength of the cladding material (around 300 MPa) at room temperature. Providing that the clad component has received PWHT, it can be assumed no residual stresses in the underlying base material. For the nuclear pressure vessel, it is also reasonable to assume that the cladding stress free temperature is at the operation temperature of the vessel (around 300 deg C). It has been shown that the cladding residual stresses have negligible influence on subclad crack behaviour in clad components (receiving PWHT). It has also been shown that the crack growth for subclad cracks would be towards the

  13. FFTF Heat Transport System (HTS) component and system design

    International Nuclear Information System (INIS)

    Young, M.W.; Edwards, P.A.

    1980-01-01

    The FFTF Heat Transport Systems and Components designs have been completed and successfully tested at isothermal conditions up to 427 0 C (800 0 F). General performance has been as predicted in the design analyses. Operational flexibility and reliability have been outstanding throughout the test program. The components and systems have been demonstrated ready to support reactor powered operation testing planned later in 1980

  14. Nuclear fusion system

    International Nuclear Information System (INIS)

    Dow, W.G.

    1981-01-01

    The invention pertains to the method and apparatus for the confining of a stream of fusible positive ions at values of density and high average kinetic energy, primarily of tightly looping motions, to produce nuclear fusion at a useful rate; more or less intimately mixed with the fusible ions will be lowerenergy electrons at about equal density, introduced solely for the purpose of neutralizing the positive space charge of the ions

  15. Nuclear system test simulator

    International Nuclear Information System (INIS)

    Sawyer, S.D.; Hill, W.D.; Wilson, P.A.; Steiner, W.M.

    1987-01-01

    A transportable test simulator is described for a nuclear power plant. The nuclear power plant includes a control panel, a reactor having actuated rods for moving into and out of a reactor for causing the plant to operate, and a control rod network extending between the control panel and the reactor rods. The network serially transmits command words between the panel and rods, and has connecting interfaces at preselected points remote from the control panel between the control panel and rods. The test simulator comprises: a test simulator input for transport to and connection into the network at at least one interface for receiving the serial command words from the network. Each serial command includes an identifier portion and a command portion; means for processing interior of the simulator for the serial command words for identifying that portion of the power plant designated in the identifier portion and processing the word responsive to the command portion of the word after the identification; means for generating a response word responsive to the command portion; and output means for sending and transmitting the response word to the nuclear power plant at the interface whereby the control panel responds to the response word

  16. Recent advances in the TIG welding process and the application of the welding of nuclear components

    International Nuclear Information System (INIS)

    Lucas, W.; Males, B.O.

    1982-01-01

    Recent advances in the field of precision arc welding techniques and infacilities for production of nuclear power plant components arc presented. Of the precision welding techniques, pulsed TIG welding, pulsed plasma arc welding, hot-wire TIG welding, and pulsed inert-gas metal-arc welding. In the field of weld cladding, GMA plasma welding is cited as an alternative to submerged-arc welding with a strip electrode. Transistors and computer-controlled welding systems get a special mention. Applications of TIG welding in the UK are cited, e.g. welding of components for the AGR nuclear power plant and construction of equipment for repair work in feedwater pipes of the MAGNOX reactor. (orig.) [de

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

    International Nuclear Information System (INIS)

    Vasconcelos, Vanderley de; Soares, Wellington Antonio; Marques, Raíssa Oliveira; Silva Júnior, Silvério Ferreira da; Raso, Amanda Laureano

    2017-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Vasconcelos, Vanderley de; Soares, Wellington Antonio; Marques, Raíssa Oliveira; Silva Júnior, Silvério Ferreira da; Raso, Amanda Laureano, E-mail: vasconv@cdtn.br, E-mail: soaresw@cdtn.br, E-mail: raissaomarques@gmail.com, E-mail: silvasf@cdtn.br, E-mail: amandaraso@hotmail.com [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2017-07-01

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

  19. Marketing information system - concept and components

    OpenAIRE

    Domazet Ivana S.

    2003-01-01

    Current conditions for the carrying out of business activities are being characterized by an intensive changing of the business surrounding, as well as by a need for a flexible adapting to the newly-created conditions. Dynamisation of economic ambience, based on the principles of a propulsive marketing economy is affirming a business philosophy which will be based on requests of the strategic marketing. Vital component of every successful company is information system, which helps data acquis...

  20. Narrow gap mechanised arc welding in nuclear components manufactured by AREVA NP

    International Nuclear Information System (INIS)

    Peigney, A.

    2007-01-01

    Nuclear components require welds of irreproachable and reproducible quality. Moreover, for a given welding process, productivity requirements lead to reduce the volume of deposited metal and thus to use narrow gap design. In the shop, narrow gap Submerged Arc Welding process (SAW) is currently used on rotating parts in flat position for thicknesses up to 300 mm. Welding is performed with one or two wires in two passes per layer. In Gas Tungsten Arc Welding process (GTAW), multiple applications can be found because this process presents the advantage of allowing welding in all positions. Welding is performed in one or two passes per layer. The process is used in factory and on the nuclear sites for assembling new components but also for replacing components and for repairs. Presently, an increase of productivity of the process is sought through the use of hot wire and/or two wires. Concerning Gas Metal Arc Welding process (GMAW), its use is growing for nuclear components, including narrow gap applications. This process, limited in its applications in the past on account of the defects it generated, draws benefit from the progress of the welding generators. Then it is possible to use this efficient process for high security components such as those of nuclear systems. It is to be noted that the process is applicable in the various welding positions as it is the case for GTAW, while being more efficient than the latter. This paper presents the state of the art in the use of narrow gap mechanised arc welding processes by AREVA NP units. (author) [fr

  1. Component design challenges for the ground-based SP-100 nuclear assembly test

    International Nuclear Information System (INIS)

    Markley, R.A.; Disney, R.K.; Brown, G.B.

    1989-01-01

    The SP-100 ground engineering system (GES) program involves a ground test of the nuclear subsystems to demonstrate their design. The GES nuclear assembly test (NAT) will be performed in a simulated space environment within a vessel maintained at ultrahigh vacuum. The NAT employs a radiation shielding system that is comprised of both prototypical and nonprototypical shield subsystems to attenuate the reactor radiation leakage and also nonprototypical heat transport subsystems to remove the heat generated by the reactor. The reactor is cooled by liquid lithium, which will operate at temperatures prototypical of the flight system. In designing the components for these systems, a number of design challenges were encountered in meeting the operational requirements of the simulated space environment (and where necessary, prototypical requirements) while also accommodating the restrictions of a ground-based test facility with its limited available space. This paper presents a discussion of the design challenges associated with the radiation shield subsystem components and key components of the heat transport systems

  2. Life Cycle Management Managing the Aging of Critical Nuclear Plant Components

    International Nuclear Information System (INIS)

    Meyer, Theodore A.; Elder, G. Gary; Llovet, Ricardo

    2002-01-01

    Life Cycle Management is a structured process to manage equipment aging and long-term equipment reliability for nuclear plant Systems, Structures and Components (SSCs). The process enables the identification of effective repair, replace, inspect, test and maintenance activities and the optimal timing of the activities to maximize the economic value to the nuclear plant. This paper will provide an overview of the process and some of the tools that can be used to implement the process for the SSCs deemed critical to plant safety and performance objectives. As nuclear plants strive to reduce costs, extend life and maximize revenue, the LCM process and the supporting tools summarized in this paper can enable development of a long term, cost efficient plan to manage the aging of the plant SSCs. (authors)

  3. Influence of nuclear radiation and laser beams on optical fibers and components

    Directory of Open Access Journals (Sweden)

    Pantelić Slađana N.

    2011-01-01

    Full Text Available The influence of nuclear radiation and particles has been the object of investigation for a long time. For new materials and systems the research should be continued. Human activities in various environments, including space, call for more detailed research. The role of fibers in contemporary communications, medicine, and industry increases. Fibers, their connections and fused optics components have one type of tasks - the transmission of information and power. The other type of tasks is reserved for fiber lasers: quantum generators and amplifiers. The third type of tasks is for fiber sensors, including high energy nuclear physics. In this paper we present some chosen topics in the mentioned areas as well as our experiments with nuclear radiation and laser beams to fiber and bulk materials of various nature (glass, polymer, metallic, etc..

  4. Power conditioning for space nuclear reactor systems

    Science.gov (United States)

    Berman, Baruch

    1987-01-01

    This paper addresses the power conditioning subsystem for both Stirling and Brayton conversion of space nuclear reactor systems. Included are the requirements summary, trade results related to subsystem implementation, subsystem description, voltage level versus weight, efficiency and operational integrity, components selection, and shielding considerations. The discussion is supported by pertinent circuit and block diagrams. Summary conclusions and recommendations derived from the above studies are included.

  5. Nuclear power - an inevitable component of a sustainable energy mix

    International Nuclear Information System (INIS)

    Mesarovic, M.

    2000-01-01

    Nuclear power plants already add consequential amounts of energy to the global energy supply and continue to offer advantages for large additions of capacity. If increased, the nuclear share in world's energy mix would reduce the environmental damages as well as the climate change threats caused by the use of fossil fuels, thus providing an essential element of sustainable development. Such a potential contribution of nuclear power on large scale in a sustainable energy mix is considered, with its actual burdens and challenges discussed. Sustainable energy development with or without nuclear power is presented, with public acceptance of nuclear energy and global warming issues discussed in more details. (author)

  6. Nuclear database management systems

    International Nuclear Information System (INIS)

    Stone, C.; Sutton, R.

    1996-01-01

    The authors are developing software tools for accessing and visualizing nuclear data. MacNuclide was the first software application produced by their group. This application incorporates novel database management and visualization tools into an intuitive interface. The nuclide chart is used to access properties and to display results of searches. Selecting a nuclide in the chart displays a level scheme with tables of basic, radioactive decay, and other properties. All level schemes are interactive, allowing the user to modify the display, move between nuclides, and display entire daughter decay chains

  7. Commercial Off-the-Shelf (COTS) Components and Enterprise Component Information System (eCIS)

    Energy Technology Data Exchange (ETDEWEB)

    John Minihan; Ed Schmidt; Greg Enserro; Melissa Thompson

    2008-06-30

    The purpose of the project was to develop the processes for using commercial off-the-shelf (COTS) parts for WR production and to put in place a system for implementing the data management tools required to disseminate, store, track procurement, and qualify vendors. Much of the effort was devoted to determining if the use of COTS parts was possible. A basic question: How does the Nuclear Weapons Complex (NWC) begin to use COTS in the weapon Stockpile Life Extension Programs with high reliability, affordability, while managing risk at acceptable levels? In FY00, it was determined that a certain weapon refurbishment program could not be accomplished without the use of COTS components. The elements driving the use of COTS components included decreased cost, greater availability, and shorter delivery time. Key factors that required implementation included identifying the best suppliers and components, defining life cycles and predictions of obsolescence, testing the feasibility of using COTS components with a test contractor to ensure capability, as well as quality and reliability, and implementing the data management tools required to disseminate, store, track procurement, and qualify vendors. The primary effort of this project then was to concentrate on the risks involved in the use of COTS and address the issues of part and vendor selection, procurement and acceptance processes, and qualification of the parts via part and sample testing. The Enterprise Component Information System (eCIS) was used to manage the information generated by the COTS process. eCIS is a common interface for both the design and production of NWC components and systems integrating information between SNL National Laboratory (SNL) and the Kansas City Plant (KCP). The implementation of COTS components utilizes eCIS from part selection through qualification release. All part related data is linked across an unclassified network for access by both SNL and KCP personnel. The system includes not

  8. Natural versus artificial aging of nuclear power plant components

    International Nuclear Information System (INIS)

    Shaw, M.T.

    1992-01-01

    This program seeks to understand the aging of polymeric materials, in cables and other components in nuclear reactor containment, by comparing aging processes for a variety of materials under natural conditions with those under the accelerated laboratory conditions used in qualification. The first five-year phase has been completed in what is planned as a long-term study of up to 40 years. Data from the program can be used as a basis of forecasting more realistic lifetimes in reactor service. The program is of critical importance for utilities both for the safe operation of plants and for minimizing the cost of periodic replacements upon expiration of originally predicted qualified life. The first five-year period has involved the selection and acquisition of test specimens, their preparation for placement in the containment, the selection of plants and locations for the specimens, the establishment of methods for monitoring radiation and temperature levels at each site, development of plans for scheduled removals, test method development, and testing of the specimens by physical and mechanical methods. Specimens have been subjected to short-term accelerated aging, as well as to reactor containment aging for up to five years. They consist of many types of polymers in products of several different manufacturers. Environmental conditions cover a wide range of temperature and radiation levels at 17 locations in 9 reactors of participating utilities. Initial results, which include tests of special cases subject to 8 years of reactor aging at Northeast Utilities, indicate several instances of changes having statistical significance in density or tensile properties due to containment service, but none of these changes are large enough to be of any concern. 12 refs., 19 figs., 21 tabs

  9. Nuclear reactor insulation and preheat system

    International Nuclear Information System (INIS)

    Wampole, N.C.

    1978-01-01

    An insulation and preheat system is disclosed for preselected components of a fluid cooled nuclear reactor. A gas tight barrier or compartment of thermal insulation surrounds the selected components and includes devices to heat the internal atmosphere of the ocmpartment. An external surface of the compartment of enclosure is cooled, such as by a circulating fluid. The heating devices provide for preheating of the components, as well as maintenance of a temperature sufficient to ensure that the reactor coolant fluid will not solidify during shutdown. The external cooling limits the heat transferred to other plant structures, such as supporting concrete and steel. The barrier is spaced far enough from the surrounded components so as to allow access for remote or manual inspection, maintenance, and repair

  10. Virtual enterprise model for the electronic components business in the Nuclear Weapons Complex

    Energy Technology Data Exchange (ETDEWEB)

    Ferguson, T.J.; Long, K.S.; Sayre, J.A. [Sandia National Labs., Albuquerque, NM (United States); Hull, A.L. [Sandia National Labs., Livermore, CA (United States); Carey, D.A.; Sim, J.R.; Smith, M.G. [Allied-Signal Aerospace Co., Kansas City, MO (United States). Kansas City Div.

    1994-08-01

    The electronic components business within the Nuclear Weapons Complex spans organizational and Department of Energy contractor boundaries. An assessment of the current processes indicates a need for fundamentally changing the way electronic components are developed, procured, and manufactured. A model is provided based on a virtual enterprise that recognizes distinctive competencies within the Nuclear Weapons Complex and at the vendors. The model incorporates changes that reduce component delivery cycle time and improve cost effectiveness while delivering components of the appropriate quality.

  11. A pilot application of risk-based methods to establish in-service inspection priorities for nuclear components at Surry Unit 1 Nuclear Power Station

    International Nuclear Information System (INIS)

    Vo, T.; Gore, B.; Simonen, F.; Doctor, S.

    1994-08-01

    As part of the Nondestructive Evaluation Reliability Program sponsored by the US Nuclear Regulatory Commission, the Pacific Northwest Laboratory is developing a method that uses risk-based approaches to establish in-service inspection plans for nuclear power plant components. This method uses probabilistic risk assessment (PRA) results and Failure Modes and Effects Analysis (FEMA) techniques to identify and prioritize the most risk-important systems and components for inspection. The Surry Nuclear Power Station Unit 1 was selected for pilot applications of this method. The specific systems addressed in this report are the reactor pressure vessel, the reactor coolant, the low-pressure injection, and the auxiliary feedwater. The results provide a risk-based ranking of components within these systems and relate the target risk to target failure probability values for individual components. These results will be used to guide the development of improved inspection plans for nuclear power plants. To develop inspection plans, the acceptable level of risk from structural failure for important systems and components will be apportioned as a small fraction (i.e., 5%) of the total PRA-estimated risk for core damage. This process will determine target (acceptable) risk and target failure probability values for individual components. Inspection requirements will be set at levels to assure that acceptable failure probabilistics are maintained

  12. Determination of ABOS 1-3+ system components belonging to the scope of license application of Paks Nuclear Power Plant Unit 1 for extension of service life, designated for the review, and verification of completeness of the scope

    International Nuclear Information System (INIS)

    Biro, Agnes Janosine; Tanits, Katalin Baumann-ne; Gosi, Peter; Kovacs, Andras; Ratkai, Sandor

    2012-01-01

    It is one major requirement of licensing the extension of design service life to determine the systems, structures and components that belong to the scope of licensing. According to the domestic regulatory requirements the ABOS 1-3 safety class components, the non safety system components of seismic safety class 3 and those non safety class components whose failure would occur due to its unmanaged ageing process and which may jeopardize safety class components with the released medium shall be involved into the scope of licensing of service life extension (SLE). In the task the components for the scope of SLE licensing of Unit 1 was determined using and, if necessary, further developing the tools provided by and exploiting, verifying and, as appropriate, supplementing the data included in the central technical database (IMR/MDM) of the NPP. As basis for determination of the scope the systems, structures and individual components categorized into safety class in the Final Safety Report were taken. Digitalized mechanical technological schemes were also used in determining the components of the systems fulfilling safety functions and in verifying the completeness. In order to assign the components belonging to the fulfillment of the function of the systems and to review the scope, the digitalization of the ABOS 2-3 electric and the ABOS 2 I and C circuit diagrams and distributor single-line diagrams and the processing and analysis of the digitalized data was performed. The ABOS + scope components were verified by walkdown. The completed component lists were compared to the components of the SLE licensing scope of the IMR/MDM database and the necessary supplementation, correction of the IMR/MDM data was also performed. In order to identify the components requiring review during licensing, also the active/passive safety function fulfillment modes were determined for every component of the licensing scope for Unit 1, which is now regarded as complete. As the results of

  13. Risk and safety analysis of nuclear systems

    CERN Document Server

    Lee, John C

    2011-01-01

    The book has been developed in conjunction with NERS 462, a course offered every year to seniors and graduate students in the University of Michigan NERS program. The first half of the book covers the principles of risk analysis, the techniques used to develop and update a reliability data base, the reliability of multi-component systems, Markov methods used to analyze the unavailability of systems with repairs, fault trees and event trees used in probabilistic risk assessments (PRAs), and failure modes of systems. All of this material is general enough that it could be used in non-nuclear a

  14. Predictive based monitoring of nuclear plant component degradation using support vector regression

    International Nuclear Information System (INIS)

    Agarwal, Vivek; Alamaniotis, Miltiadis; Tsoukalas, Lefteri H.

    2015-01-01

    Nuclear power plants (NPPs) are large installations comprised of many active and passive assets. Degradation monitoring of all these assets is expensive (labor cost) and highly demanding task. In this paper a framework based on Support Vector Regression (SVR) for online surveillance of critical parameter degradation of NPP components is proposed. In this case, on time replacement or maintenance of components will prevent potential plant malfunctions, and reduce the overall operational cost. In the current work, we apply SVR equipped with a Gaussian kernel function to monitor components. Monitoring includes the one-step-ahead prediction of the component's respective operational quantity using the SVR model, while the SVR model is trained using a set of previous recorded degradation histories of similar components. Predictive capability of the model is evaluated upon arrival of a sensor measurement, which is compared to the component failure threshold. A maintenance decision is based on a fuzzy inference system that utilizes three parameters: (i) prediction evaluation in the previous steps, (ii) predicted value of the current step, (iii) and difference of current predicted value with components failure thresholds. The proposed framework will be tested on turbine blade degradation data.

  15. Thermochemical modelling of multi-component systems

    International Nuclear Information System (INIS)

    Sundman, B.; Gueneau, C.

    2015-01-01

    Computational thermodynamic, also known as the Calphad method, is a standard tool in industry for the development of materials and improving processes and there is an intense scientific development of new models and databases. The calculations are based on thermodynamic models of the Gibbs energy for each phase as a function of temperature, pressure and constitution. Model parameters are stored in databases that are developed in an international scientific collaboration. In this way, consistent and reliable data for many properties like heat capacity, chemical potentials, solubilities etc. can be obtained for multi-component systems. A brief introduction to this technique is given here and references to more extensive documentation are provided. (authors)

  16. The system of nuclear material control of Kazakhstan

    International Nuclear Information System (INIS)

    Yeligbayeva, G.Zh.

    2001-01-01

    Full text: The State system for nuclear material control consists of three integral components. The efficiency of each is to guarantee the non-proliferation regime in Kazakhstan. The components are the following: accounting, export and import control and physical protection of nuclear materials. First, the implementation of the goals of accounting and control bring into force, by the organization of the system for accounting and measurement of nuclear materials to determine present quantity. Organizing the accounting for nuclear material at facilities will ensure the efficiency of accountancy and reporting information. This defines the effectiveness of the state system for the accounting for the Kazakhstan's nuclear materials. Currently, Kazakhstan's nuclear material is fully safeguarded in designated secure locations. Kazakhstan has a nuclear power plant, 4 research reactors and a fuel fabrication plant. The governmental information system for nuclear materials control consist of two level: Governmental level - KAEA collects reports from facilities and prepares the reports for International Atomic Energy Agency, keeping of supporting documents and other necessary information, a data base of export and import, a data base of nuclear material inventory. Facility level - registration and processing information from key measurement points, formation the facility's nuclear materials accounting database. All facilities have computerized systems. Currently, all facilities are safeguarded under IAEA safeguarding standards, through IAEA inspections. Annually, IAEA verifies all nuclear materials at all Kazakhstan nuclear facilities. The government reporting system discloses the existence of all nuclear material and its transfer intended for interaction through the export control system and the nuclear control accounting system. Nuclear material export is regulated by the regulations of the Nuclear Export Control Law. The standard operating procedure is the primary means for

  17. Seismic proving tests on the reliability for large components and equipment of nuclear power plants

    International Nuclear Information System (INIS)

    Ohno, Tokue; Tanaka, Nagatoshi

    1988-01-01

    Since Japan has destructive earthquakes frequently, the structural reliability for large components and equipment of nuclear power plants are rigorously required. They are designed using sophisticated seismic analyses and have not yet encountered a destructive earthquake. When nuclear power plants are planned, it is very important that the general public understand the structural reliability during and after an earthquake. Seismic Proving Tests have been planned by Ministry of International Trade and Industry (Miti) to comply with public requirement in Japan. A large-scale high-performance vibration table was constructed at Tasted Engineering Laboratory of Nuclear Power Engineering Test Center (NU PEC), in order to prove the structural reliability by vibrating the test model (of full scale or close to the actual size) in the condition of a destructive earthquake. As for the test models, the following four items were selected out of large components and equipment important to the safety: Reactor Containment Vessel; Primary Coolant Loop or Primary Loop Recirculation System; Reactor Pressure Vessel; and Reactor Core Internals. Here is described a brief of the vibration table, the test method and the results of the tests on PWR Reactor Containment Vessel and BWR Primary Loop Recirculation System (author)

  18. Computer systems and nuclear industry

    International Nuclear Information System (INIS)

    Nkaoua, Th.; Poizat, F.; Augueres, M.J.

    1999-01-01

    This article deals with computer systems in nuclear industry. In most nuclear facilities it is necessary to handle a great deal of data and of actions in order to help plant operator to drive, to control physical processes and to assure the safety. The designing of reactors requires reliable computer codes able to simulate neutronic or mechanical or thermo-hydraulic behaviours. Calculations and simulations play an important role in safety analysis. In each of these domains, computer systems have progressively appeared as efficient tools to challenge and master complexity. (A.C.)

  19. Marketing information system - concept and components

    Directory of Open Access Journals (Sweden)

    Domazet Ivana S.

    2003-01-01

    Full Text Available Current conditions for the carrying out of business activities are being characterized by an intensive changing of the business surrounding, as well as by a need for a flexible adapting to the newly-created conditions. Dynamisation of economic ambience, based on the principles of a propulsive marketing economy is affirming a business philosophy which will be based on requests of the strategic marketing. Vital component of every successful company is information system, which helps data acquisition and analysis, and its transfer into information, which is then forwarded to users and management. Bearing that in mind, we have created marketing information system, which collects relevant data, analyses it and then produces information, which can be used in planning, implementation and control, those making your company more efficient.

  20. Some applications of capacitance technology in nuclear reactor components inspections

    International Nuclear Information System (INIS)

    Walton, H.

    1985-01-01

    The paper considers application of a capacitance measuring system that has overcome many of the original contraints, such as sensitivity to cable length, induced electric field and high acoustic noise, and illustrates the ease of use with examples of proven capability in severe environments of high temperature or high radiation. The Capacitance Displacement Transducer (CDT) measuring principle was originally developed as a working technique during the early years of full-scale, on-load refuelling trials performed in the Windscale Civil Advanced Gas-Cooled Reactor (CAGR) test rig where it was necessary to measure the vibrational behaviour of fuel components in simulated reactor conditions. At that time, 1968-1969, no instrumentation existed that would measure displacement in the range 0 to 100 mms to an accuracy of 25x10 -3 mms, without physical contact, at temperatures of 600 0 C in high velocity gas, in high acoustic noise fields of 150 db's over cable lengths approaching 100 metres. The principles incorporated in the CDT overcome all these problems. The advantages inherent in this system have been extended to metrology applications in more recent years by the further development of the electronics to enable linear displacement measurement to be obtained between two capacitance plates whose separation varies, either by plate movement or by surface irregularity. This principle has been used to good effect in novel applications associated with the inspection of nominally inaccessible internal tube surfaces

  1. Transport and repair of contaminated nuclear components - liabilities and insurance

    International Nuclear Information System (INIS)

    Brunego, C.; Deprimoz, J.; Engelhard, M.

    1983-01-01

    The nuclear park has been constructed fairly recently and has not yet required large-scale maintenance efforts; however account should now be taken of the fact that periodic checks of nuclear power plants will imply systematic transfers of irradiated or contaminated materials outside the plants. In this context, the paper reviews the nuclear third party liability regime under the Paris Convention and the Euratom directives on radiation protection. It then describes the cover offered by insurance pools in several European countries. (NEA) [fr

  2. Thermal Components Boost Performance of HVAC Systems

    Science.gov (United States)

    2012-01-01

    As the International Space Station (ISS) travels 17,500 miles per hour, normal is having a constant sensation of free-falling. Normal is no rain, but an extreme amount of shine.with temperatures reaching 250 F when facing the Sun. Thanks to a number of advanced control systems onboard the ISS, however, the interior of the station remains a cool, comfortable, normal environment where astronauts can live and work for extended periods of time. There are two main control systems on the ISS that make it possible for humans to survive in space: the Thermal Control System (TCS) and the Environmental Control and Life Support system. These intricate assemblies work together to supply water and oxygen, regulate temperature and pressure, maintain air quality, and manage waste. Through artificial means, these systems create a habitable environment for the space station s crew. The TCS constantly works to regulate the temperature not only for astronauts, but for the critical instruments and machines inside the spacecraft as well. To do its job, the TCS encompasses several components and systems both inside and outside of the ISS. Inside the spacecraft, a liquid heat-exchange process mechanically pumps fluids in closed-loop circuits to collect, transport, and reject heat. Outside the ISS, an external system circulates anhydrous ammonia to transport heat and cool equipment, and radiators release the heat into space. Over the years, NASA has worked with a variety of partners.public and private, national and international. to develop and refine the most complex thermal control systems ever built for spacecraft, including the one on the ISS.

  3. Advanced nuclear systems. Review study

    International Nuclear Information System (INIS)

    Liebert, Wolfgang; Glaser, Alexander; Pistner, Christoph; Baehr, Roland; Hahn, Lothar

    1999-04-01

    The task of this review study is to from provide an overview of the developments in the field of the various advanced nuclear systems, and to create the basis for more comprehensive studies of technology assessment. In an overview the concepts for advanced nuclear systems pursued worldwide are subdivided into eight subgroups. A coarse examination raster (set pattern) is developed to enable a detailed examination of the selected systems. In addition to a focus on enhanced safety features, further aspects are also taken into consideration, like the lowering of the proliferation risk, the enhancement of the economic competitiveness of the facilities and new usage possibilities (for instance concerning the relaxation of the waste disposal problem or the usage of alternative fuels to uranium). The question about the expected time span for realization and the discussion about the obstacles on the way to a commercially usable reactor also play a substantial role as well as disposal requirements as far as they can be presently recognized. In the central chapter of this study, the documentation of the representatively selected concepts is evaluated as well as existing technology assessment studies and expert opinions. In a few cases where this appears to be necessary, according technical literature, further policy advisory reports, expert statements as well as other relevant sources are taken into account. Contradictions, different assessments and dissents in the literature as well as a few unsettled questions are thus indicated. The potential of advanced nuclear systems with respect to economical and societal as well as environmental objectives cannot exclusively be measured by the corresponding intrinsic or in comparison remarkable technical improvements. The acceptability of novel or improved systems in nuclear technology will have to be judged by their convincing solutions for the crucial questions of safety, nuclear waste and risk of proliferation of nuclear weapons

  4. Some experience from seismic check-ups of components of Mochovce nuclear power plant

    International Nuclear Information System (INIS)

    Masopust, R.

    1987-01-01

    The first Czechoslovak nuclear power plant with the so-called partial anti-seismic design will be built in Mochovce. The evaluation of seismic resistance is prescribed only for equipment and systems which secure the safe reactor shutdown, the withdrawal of residual heat and prevent uncontrolled release of radioactivity into the environment. The following variants were compared in the calculation analysis of the primary loop of the WWER-440 reactor for the Mochovce nuclear power plant: the seismically unsecured loop of a usual design for WWER-440 nuclear power plants, the loop provided with mechanical or hydraulic dampers and the loop provided with viscose shock absorbers. The tests showed that technically most suitable is the use of viscose shock absorbers which do not completely block the movement of the system during the earthquake but absorb it intensively. The viscose shock absorbers are also much cheaper than the dampers. Briefly described is experience with the experimental evaluation of the seismic resistance of components of the Mochovce nuclear power plant. Great difficulty was encountered by the non-existence in Czechoslovakia of a seismic table allowing simultaneous excitation in the vertical and horizontal directions. (Z.M.). 18 refs

  5. Performance of materials in the component cooling water systems of pressurized water reactors

    International Nuclear Information System (INIS)

    Lee, B.S.

    1993-01-01

    The component cooling water (CCW) system provides cooling water to several important loads throughout the plant under all operating conditions. An aging assessment CCW systems in pressurized water reactors (PWRs) was conducted as part of Nuclear Plant Aging Research Program (NPAR) instituted by the US Nuclear Regulatory Commission. This paper presents some of the results on the performances of materials in respect of their application in CCW Systems. All the CCW system failures reported to the Nuclear Plant Reliability Data System (NPRDS) from January 1988 to June 1990 were reviewed; it is concluded that three of the main contributors to CCW system failures are valves, pumps, and heat exchangers. This study identified the modes and causes of failure for these components; most of the causes for the aging-related failures could be related to the performance of materials. Also, in this paper the materials used for these components are reviewed, and there aging mechanisms under CCW system conditions are discussed

  6. NDE during precision manufacturing of pressure components in nuclear reactors

    International Nuclear Information System (INIS)

    Raj, Baldev; Venkataram, B.; Chellapandi, P.

    2010-01-01

    Energy is the critical enabler for all social and economic developments and growth of civilization. For a nation to be energy secure, it should have a balanced and healthy energy basket with a varied mix of energy sources in right proportions depending on the resources of the country. It is now a well realized fact that nuclear energy is an inevitable option that should be present in energy basket of nuclear mature countries. This is due the fact that nuclear power has proved to be (a) capable of generating electricity safely on a large-scale with price stability over long periods of time satisfying a modern economy's significant demand for electricity that must be available round-the-clock; and (b) it is environmentally benign and provides a clean energy source with minimum of green house gas emissions. Internationally, about I 696 electricity is derived from nuclear power. In the Indian context, the contribution from nuclear power currently is about 3%, which needs to be enhanced by 4 fold by 2030 and 10 fold by 2050 if India is to sustain its current gross domestic product. NDE intertwined with materials, manufacturing technology and total life cycle management are crucial to safe and economic nuclear power.

  7. Micro rapid prototyping system for micro components

    International Nuclear Information System (INIS)

    Li Xiaochun; Choi Hongseok; Yang Yong

    2002-01-01

    Similarities between silicon-based micro-electro-mechanical systems (MEMS) and Shape Deposition Manufacturing (SDM) processes are obvious: both integrate additive and subtractive processes and use part and sacrificial materials to obtain functional structures. These MEMS techniques are two-dimensional (2-D) processes for a limited number of materials while SDM enables the building of parts that have traditionally been impossible to fabricate because of their complex shapes or of their variety in materials. This work presents initial results on the development of a micro rapid prototyping system that adapts SDM methodology to micro-fabrication. This system is designed to incorporate microdeposition and laser micromachining. In the hope of obtaining a precise microdeposition, an ultrasonic-based micro powder-feeding mechanism was developed in order to form thin patterns of dry powders that can be cladded or sintered onto a substrate by a micro-sized laser beam. Furthermore, experimental results on laser micromachining using a laser beam with a wavelength of 355 nm are also presented. After further improvement, the developed micro manufacturing system could take computer-aided design (CAD) output to reproduce 3-D heterogeneous micro-components from a wide selection of materials

  8. Nuclear excited power generation system

    International Nuclear Information System (INIS)

    Parker, R.Z.; Cox, J.D.

    1989-01-01

    A power generation system is described, comprising: a gaseous core nuclear reactor; means for passing helium through the reactor, the helium being excited and forming alpha particles by high frequency radiation from the core of the gaseous core nuclear reactor; a reaction chamber; means for coupling chlorine and hydrogen to the reaction chamber, the helium and alpha particles energizing the chlorine and hydrogen to form a high temperature, high pressure hydrogen chloride plasma; means for converting the plasma to electromechanical energy; means for coupling the helium back to the gaseous core nuclear reactor; and means for disassociating the hydrogen chloride to form molecular hydrogen and chlorine, to be coupled back to the reaction chamber in a closed loop. The patent also describes a power generation system comprising: a gaseous core nuclear reactor; means for passing hydrogen through the reactor, the hydrogen being excited by high frequency radiation from the core; means for coupling chlorine to a reaction chamber, the hydrogen energizing the chlorine in the chamber to form a high temperature, high pressure hydrogen chloride plasma; means for converting the plasma to electromechanical energy; means for disassociating the hydrogen chloride to form molecular hydrogen and chlorine, and means for coupling the hydrogen back to the gaseous core nuclear reactor in a closed loop

  9. Prognostic Health Monitoring System: Component Selection Based on Risk Criteria and Economic Benefit Assessment

    International Nuclear Information System (INIS)

    Pham, Binh T.; Agarwal, Vivek; Lybeck, Nancy J.; Tawfik, Magdy S.

    2012-01-01

    Prognostic health monitoring (PHM) is a proactive approach to monitor the ability of structures, systems, and components (SSCs) to withstand structural, thermal, and chemical loadings over the SSCs planned service lifespan. The current efforts to extend the operational license lifetime of the aging fleet of U.S. nuclear power plants from 40 to 60 years and beyond can benefit from a systematic application of PHM technology. Implementing a PHM system would strengthen the safety of nuclear power plants, reduce plant outage time, and reduce operation and maintenance costs. However, a nuclear power plant has thousands of SSCs, so implementing a PHM system that covers all SSCs requires careful planning and prioritization. This paper therefore focuses on a component selection that is based on the analysis of a component's failure probability, risk, and cost. Ultimately, the decision on component selection depends on the overall economical benefits arising from safety and operational considerations associated with implementing the PHM system. (author)

  10. Condition Based Prognostics of Passive Components - A New Era for Nuclear Power Plant Life Management

    International Nuclear Information System (INIS)

    Bakhtiari, S.; Mohanty, S.; Prokofiev, I.; Tregoning, R.

    2012-01-01

    As part of a research project sponsored by the U.S. NRC, Argonne National Laboratory (ANL) conducted scoping studies to identify viable and promising sensors and techniques for in-situ inspection and real-time monitoring of degradation in nuclear power plant (NPP) systems, structures, and components (SSC). Significant advances have been made over the past two decades toward development of online monitoring (OLM) techniques for detection, diagnostics, and prognostics of degradation in active nuclear power plant (NPP) components (e.g., pumps, valves). However, early detection of damage and degradation in safety-critical passive components, (e.g. piping, tubing pressure vessel), is challenging, and will likely remain so for the foreseeable future. Ensuring the structural integrity of the reactor pressure vessel (RPV) and piping systems in particular is a prerequisite to long term safe operation of NPPs. The current practice is to implement inservice inspection (ISI) and preventive maintenance programs. While these programs have generally been successful, they are limited in that information is only obtained during plant outages. Additionally, these inspections, often the critical path in the outage schedule, are costly, time consuming, and involve potentially high dose to nondestructive examination/evaluation (NDE) personnel. A viable plant-wide on-line structural health monitoring program for continuous and automatic monitoring of critical SSCs could be a more effective approach for guarding against unexpected failures. Specifically, OLM information about the current condition of the SSCs could be input to an online prognostics (OLP) system to forecast their remaining useful life in real time. This paper provides an overview of scoping studies performed at ANL on assessing the viability of OLM and OLP systems for real time and automated monitoring and remaining of condition and the remaining useful life of passive components in NPPs. (author)

  11. Effect of component aging on PWR control rod drive systems

    International Nuclear Information System (INIS)

    Grove, E.; Gunther, W.; Sullivan, K.

    1992-01-01

    An aging assessment of PWR control rod drive (CRD) systems has been completed as part of the US NRC Nuclear Plant Aging Research (NPAR) Program. The design, construction, maintenance, and operation of the Babcock ampersand Wilcox (B ampersand W), Combustion Engineering (CE), and Westinghouse (W) systems were evaluated to determine the potential for degradation as each system ages. Operating experience data were evaluated to identify the predominant failure modes, causes, and effects. This, coupled with an assessment of the materials of construction and operating environment, demonstrate that each design is subject to degradation, which if left unchecked, could affect its safety function as the plant ages. An industry survey, conducted with the assistance of EPRI and NUMARC, identified current CRD system maintenance and inspection practices. The results of this survey indicate that some plants have performed system modifications, replaced components, or augmented existing preventive maintenance practices in response to system aging. The survey results also supported the operating experience data, which concluded that the timely replacement of degraded components, prior to failure, was not always possible using existing condition monitoring techniques. The recommendations presented in this study also include a discussion of more advanced monitoring techniques, which provide trendable results capable of detecting aging

  12. INIS - International Nuclear Information System

    International Nuclear Information System (INIS)

    1995-01-01

    The paper presents International Nuclear Information System (INIS): history of its development; INIS support products (INIS Reference Series, Friendly Inputting of Bibliographic Records software); INIS output products (INIS Atomindex, magnetic tapes, online service, database on CD-ROM, microfiche service); INIS philosophy; input of INIS database by subject areas; and examples of INIS input

  13. Recent space nuclear power systems

    International Nuclear Information System (INIS)

    Takizuka, Takakazu; Yasuda, Hideshi; Hishida, Makoto

    1991-01-01

    For the advance of mankind into the space, the power sources of large output are indispensable, and it has been considered that atomic energy is promising as compared with solar energy and others. Accordingly in USA and USSR, the development of the nuclear power generation systems for space use has been carried out since considerable years ago. In this report, the general features of space nuclear reactors are shown, and by taking the system for the SP-100 project being carried out in USA as the example, the contents of the recent design regarding the safety as an important factor are discussed. Moreover, as the examples of utilizing space nuclear reactors, the concepts of the power source for the base on the moon, the sources of propulsive power for the rockets used for Mars exploration and others, the remote power transmission system by laser in the space and so on are explained. In September, 1988, the launching of a space shuttle of USA was resumed, and the Jupiter explorer 'Galileo' and the space telescope 'Hubble' were successfully launched. The space station 'Mir' of USSR has been used since February, 1986. The history of the development of the nuclear power generation systems for space use is described. (K.I.)

  14. Natural circulating passive cooling system for nuclear reactor containment structure

    Science.gov (United States)

    Gou, Perng-Fei; Wade, Gentry E.

    1990-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  15. Passive cooling system for nuclear reactor containment structure

    Science.gov (United States)

    Gou, Perng-Fei; Wade, Gentry E.

    1989-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  16. Reliability of sprinkler systems. Exploration and analysis of data from nuclear and non-nuclear installations

    International Nuclear Information System (INIS)

    Roenty, V.; Keski-Rahkonen, O.; Hassinen, J.P.

    2004-12-01

    Sprinkler systems are an important part of fire safety of nuclear installations. As a part of effort to make fire-PSA of our utilities more quantitative a literature survey from open sources worldwide of available reliability data on sprinkler systems was carried out. Since the result of the survey was rather poor quantitatively, it was decided to mine available original Finnish nuclear and non-nuclear data, since nuclear power plants present a rather small device population. Sprinklers are becoming a key element for the fire safety in modern, open non-nuclear buildings. Therefore, the study included both nuclear power plants and non-nuclear buildings protected by sprinkler installations. Data needed for estimating of reliability of sprinkler systems were collected from available sources in Finnish nuclear and non-nuclear installations. Population sizes on sprinkler system installations and components therein as well as covered floor areas were counted individually from Finnish nuclear power plants. From non-nuclear installations corresponding data were estimated by counting relevant things from drawings of 102 buildings, and plotting from that sample needed probability distributions. The total populations of sprinkler systems and components were compiled based on available direct data and these distributions. From nuclear power plants electronic maintenance reports were obtained, observed failures and other reliability relevant data were selected, classified according to failure severity, and stored on spreadsheets for further analysis. A short summary of failures was made, which was hampered by a small sample size. From non-nuclear buildings inspection statistics from years 1985.1997 were surveyed, and observed failures were classified and stored on spreadsheets. Finally, a reliability model is proposed based on earlier formal work, and failure frequencies obtained by preliminary data analysis of this work. For a model utilising available information in the non-nuclear

  17. Nuclear chromodynamics: applications of QCD to relativistic multiquark systems

    International Nuclear Information System (INIS)

    Brodsky, S.J.; Ji, C.R.

    1984-07-01

    We review the applications of quantum chromodynamics to nuclear multiquark systems. In particular, predictions are given for the deuteron reduced form factor in the high momentum transfer region, hidden color components in nuclear wavefunctions, and the short distance effective force between nucleons. A new antisymmetrization technique is presented which allows a basis for relativistic multiquark wavefunctions and solutions to their evolution to short distances. Areas in which conventional nuclear theory conflicts with QCD are also briefly reviewed. 48 references

  18. Procurement and quality control of components important to safety in nuclear engineering projects

    International Nuclear Information System (INIS)

    Zhang Zhihua; Zhang Yiyun

    2006-01-01

    The procurement and quality control of components is a very important work in the nuclear engineering. This paper introduces the project management techniques, such as how to make a plan of components purchase in nuclear engineering. This paper discussed the classification of components, evaluation of the potential suppliers, invitation of bids, exchange of design details with the suppliers, quality assurance and quality assurance audit, and the equipment checks before acceptance and some engineering experiences. (authors)

  19. Nuclear reactor safety system

    International Nuclear Information System (INIS)

    Sato, Takashi.

    1979-01-01

    Purpose: To allow sufficient removal of radioactive substance released in the reactor containment shell upon loss of coolants accidents thus to sufficiently decrease the exposure dose to human body. Constitution: A clean-up system is provided downstream of a heat exchanger and it is branched into a pipeway to be connected to a spray nozzle and further connected by way of a valve to a reactor container. After the end of sudden transient changes upon loss of coolants accidents, the pool water stored in the pressure suppression chamber is purified in the clean-up system and then sprayed in the dry-well by way of a spray nozzle. The sprayed water dissolves to remove water soluble radioactive substances floating in the dry-well and then returns to the pressure suppression chamber. Since radioactive substances in the dry-well can thus removed rapidly and effectively and the pool water can be reused, public hazard can also be decreased. (Horiuchi, T.)

  20. Conductivity of two-component systems

    Energy Technology Data Exchange (ETDEWEB)

    Kuijper, A. de; Hofman, J.P.; Waal, J.A. de [Shell Research BV, Rijswijk (Netherlands). Koninklijke/Shell Exploratie en Productie Lab.; Sandor, R.K.J. [Shell International Petroleum Maatschappij, The Hague (Netherlands)

    1996-01-01

    The authors present measurements and computer simulation results on the electrical conductivity of nonconducting grains embedded in a conductive brine host. The shapes of the grains ranged from prolate-ellipsoidal (with an axis ratio of 5:1) through spherical to oblate-ellipsoidal (with an axis ratio of 1:5). The conductivity was studied as a function of porosity and packing, and Archie`s cementation exponent was found to depend on porosity. They used spatially regular and random configurations with aligned and nonaligned packings. The experimental results agree well with the computer simulation data. This data set will enable extensive tests of models for calculating the anisotropic conductivity of two-component systems.

  1. Aging of metal components in US nuclear reactors

    International Nuclear Information System (INIS)

    Mayfield, M.E.; Strosnider, J.R.

    1998-01-01

    This paper presents an overview of the aging of metal components in U.S. Light Water Reactors. The types of degradation being experienced in components such as the pressure vessel, piping, reactor internals, and steam generators, and the programs being implemented to manage the degradation are discussed. (author)

  2. On Nuclear Molecules Built up from sup 1 sup 3 sup 2 Sn Components

    CERN Document Server

    Swiatecki, W J

    2003-01-01

    The possible existence of nuclear quasi-molecules built up from sup 1 sup 3 sup 2 Sn components is investigated. The crucial question is whether the extra stability of the doubly magic sup 1 sup 3 sup 2 Sn nuclei makes them sufficiently rigid to be able to withstand the strains imposed by their mutual interactions. It is argued that if the simplest quasi-molecular dumbbell configuration were found to be (meta-)stable, then triangular and even tetrahedral structures might have comparable barriers against disintegration and comparable spontaneous fission lifetimes. These are estimated using simplifying assumptions. As regards the dumbbell's stability, one may relate this to the existence of a potential energy pocket in the deformation energy landscape of a fissioning sup 2 sup 6 sup 4 Fm nucleus, and to the presence of ''bimodal'' fission in heavy Fm isotopes. Further experimental and theoretical studies of such systems may be relevant for answering the question concerning nuclear quasi-molecules.

  3. Expert systems and nuclear safety

    International Nuclear Information System (INIS)

    Beltracchi, L.

    1990-01-01

    The US Nuclear Regulatory Commission (NRC) and the Electric Power Research Institute have initiated a broad-based exploration of means to evaluate the potential applications of expert systems in the nuclear industry. This exploratory effort will assess the use of expert systems to augment the diagnostic and decision-making capabilities of personnel with the goal of enhancing productivity, reliability, and performance. The initial research effort is the development and documentation of guidelines for verifying and validating (V and V) expert systems. An initial application of expert systems in the nuclear industry is to aid operations and maintenance personnel in decision-making tasks. The scope of the decision aiding covers all types of cognitive behavior consisting of skill, rule, and knowledge-based behavior. For example, procedure trackers were designed and tested to support rule-based behavior. Further, these systems automate many of the tedious, error-prone human monitoring tasks, thereby reducing the potential for human error. The paper version of the procedure contains the knowledge base and the rules and thus serves as the basis of the design verification of the procedure tracker. Person-in-the-loop tests serve as the basis for the validation of a procedure tracker. When conducting validation tests, it is important to ascertain that the human retains the locus of control in the use of the expert system

  4. Concept of a new method for fatigue monitoring of nuclear power plant components

    International Nuclear Information System (INIS)

    Zafosnik, M.; Cizelj, L.

    2007-01-01

    Fatigue is one of the well-understood aging mechanisms affecting mechanical components in many industrial facilities including nuclear power plants. Operational experience of nuclear power plants worldwide to date confirmed adequate design of safety related components against fatigue. In some cases however, for example when the plant life extension is envisioned, it may be very useful to monitor the remaining fatigue life of safety related components. Nuclear power plants components are classified into safety classes regarding their importance in mitigating the consequences of hypothetic accidents. Service life of components subjected to fatigue loading can be estimated with Usage Factor uk. A concept of the new method aiming both at monitoring the current state of the component and predicting its remaining lifetime in the life-extension conditions is presented. The method is based on determination of partial Usage Factor of components in which operating transients will be considered and compared to design transients. (author)

  5. Automatic acoustic and vibration monitoring system for nuclear power plants

    International Nuclear Information System (INIS)

    Tothmatyas, Istvan; Illenyi, Andras; Kiss, Jozsef; Komaromi, Tibor; Nagy, Istvan; Olchvary, Geza

    1990-01-01

    A diagnostic system for nuclear power plant monitoring is described. Acoustic and vibration diagnostics can be applied to monitor various reactor components and auxiliary equipment including primary circuit machinery, leak detection, integrity of reactor vessel, loose parts monitoring. A noise diagnostic system has been developed for the Paks Nuclear Power Plant, to supervise the vibration state of primary circuit machinery. An automatic data acquisition and processing system is described for digitalizing and analysing diagnostic signals. (R.P.) 3 figs

  6. Safety device and machine system of nuclear power plant

    International Nuclear Information System (INIS)

    1978-10-01

    It introduces principle and kinds of heat power including heat balance and nuclear power. It explains a lot of technical terms about the nuclear power system, which are primary loop, reactor, steam generator, primary coolant pump and pressurizer in PWR, chemical and volume control system, component cooling system, safety injection system, and spent fuel cooling and storage system in auxiliary system, liquid solid and gaseous waste disposal system in radwaste disposal, gland sealing system, turbine instrumentation, turning gear, hydrogen cooling system, condenser, feedwater heater, degenerate heater, auxiliary heat exchanger, centrifugal pump, rotary reciprocating and tank and pressure vessel.

  7. Safety assessment of primary system components at the USNRC

    Energy Technology Data Exchange (ETDEWEB)

    Serpan, C Z; Chen, C Y; Taboada, A

    1988-12-31

    This document deals with the safety assessment in nuclear reactor components at the USNRC. The USNRC regulations and requirements concerning nuclear reactor design and operations are presented, together with guides and standards which describe how the actions should be implemented. The safety assessment relies on fracture analysis and Non Destructive Examination (NDE). (TEC).

  8. Nuclear systems of level measurement

    International Nuclear Information System (INIS)

    Lara, A.J.; Cabrera, M.J.

    1992-01-01

    In the industry there are processes in which is necessary to maintain the products level controlled which are handled for their transformation. The majority of such processes and by the operation conditions, they do not admit measure systems of level of invasive type then the application of nuclear techniques for level measurement results a big aid in these cases, since all the system installation is situated beyond frontiers of vessels that contain the product for measuring. In the Department of Nuclear Technology Applications of Mexican Petroleum Institute was developed a level measurement system by gamma rays transmission which operates in the Low Density Polyethylene plant of Petrochemical Complex Escolin at Poza Rica, Veracruz, Mexico. (Author)

  9. Nuclear power plant annunciator systems

    International Nuclear Information System (INIS)

    Rankin, W.L.

    1983-08-01

    Analyses of nuclear power plant annunciator systems have uncovered a variety of problems. Many of these problems stem from the fact that the underlying philosophy of annunciator systems have never been elucidated so as to impact the initial annunciator system design. This research determined that the basic philosophy of an annunciator system should be to minimize the potential for system and process deviations to develop into significant hazards. In order to do this the annunciator system should alert the operators to the fact that a system or process deviation exists, inform the operators as to the priority and nature of the deviation, guide the operators' initial responses to the deviation, and confirm whether operators responses corrected the deviation. Annunciator design features were analyzed to determine to what degree they helped the system meet the functional criteria, the priority for implementing specific design features, and the cost and ease of implementing specific design features

  10. The manufacturing of components for the nuclear industry

    International Nuclear Information System (INIS)

    Fogarty, John

    The experience of one company in the Canadian nuclear industry, a prime supplier of end fittings for CANDU type reactors, is described. Many factors such as work flow and continuity, financing, quality control, and export trade, are dealt with. (E.C.B.)

  11. RSE-M: In-Service Inspection Rules for Mechanical Components of PWR Nuclear Islands

    International Nuclear Information System (INIS)

    2016-01-01

    The RSE-M code defines in-service inspection operations. It applies to pressure equipment used in PWR plants, as well as spare parts for such equipment. The RSE-M code does not apply to equipment made from materials other than metal. It is based on the RCC-M code for requirements relating to the design and fabrication of mechanical components. Use: The inspection rules specified in the RSE-M code describe the standard requirements of best practice within the French nuclear industry, based on its own feedback from operating several nuclear units and partly supplemented with requirements stipulated by French regulations. To date, the 58 units in France's nuclear infrastructure enforce the in-service inspection rules of the RSE-M code. Operation of 30 commissioned units in China's nuclear infrastructure, corresponding to the M310, CPR-1000 and CPR-600 reactors, is based on the RSE-M code (since 2007, use of AFCEN codes has been required by NNSA for Generation II+ reactors). Contents of the 2016 Edition: Volume I - Rules: Section A - General rules, Section B - Specific rules for class 1 components, Section C - Specific rules for class 2 or 3 components, Section D - Specific rules for components not assigned to any particular RSE-M class; Volume II - Appendices 1 to 8: Appendices 1.0 to 1.9: supporting appendices for the general requirements, Appendix 2.1: appendix associated with chap. 2000 Requalifications, Hydraulic Proof Tests and Hydraulic Tests, Appendices 4.1 to 4.4: appendices associated with chap. 4000 Examination techniques, Appendices 5.1 to 5.8 and RPP2: appendices associated with chap. 5000 Mechanical and Materials, Appendices 8.1 to 8.2: appendices associated with chap. 8000 Maintenance Operations; Volume III: Appendix 3.1 - Visit tables: main primary and secondary systems, EPR pre-service inspection program, Class 2 or 3 vessels; Appendix 3.2 - Inspection Plans For Non-Nuclear Pressure Equipment

  12. Evaluation of Nuclear Hydrogen Production System

    International Nuclear Information System (INIS)

    Park, Won Seok; Park, C. K.; Park, J. K. and others

    2006-04-01

    The major objective of this work is tow-fold: one is to develop a methodology to determine the best VHTR types for the nuclear hydrogen demonstration project and the other is to evaluate the various hydrogen production methods in terms of the technical feasibility and the effectiveness for the optimization of the nuclear hydrogen system. Both top-tier requirements and design requirements have been defined for the nuclear hydrogen system. For the determination of the VHTR type, a comparative study on the reference reactors, PBR and PBR, was conducted. Based on the analytic hierarchy process (AHP) method, a systematic methodology has been developed to compare the two VHTR types. Another scheme to determine the minimum reactor power was developed as well. Regarding the hydrogen production methods, comparison indices were defined and they were applied to the IS (Iodine-Sulfur) scheme, Westinghouse process, and the, high-temperature electrolysis method. For the HTE, IS, and MMI cycle, the thermal efficiency of hydrogen production were systematically evaluated. For the IS cycle, an overall process was identified and the functionality of some key components was identified. The economy of the nuclear hydrogen was evaluated, relative to various primary energy including natural gas coal, grid-electricity, and renewable. For the international collaborations, two joint research centers were established: NH-JRC between Korea and China and NH-JDC between Korea and US. Currently, several joint researches are underway through the research centers

  13. The effects of aging on electrical and I ampersand C components: Results of US Nuclear Plant Aging Research

    International Nuclear Information System (INIS)

    Aggarwal, S.K.; Gunther, W.E.

    1993-01-01

    The US NRC's hardware oriented engineering research program for plant aging and degradation monitoring has achieved results in the area of electrical, control, and instrumentation (ECI) components used in nuclear power plants (NPPs). The principal goals of the program, known as the Nuclear Power Plant Aging Research (NPAR) Program, are to understand the effects of age-related degradation in NPPs and how to manage and mitigate them effectively. This paper describes how these goals have been achieved for key ECI components used in the safety systems of NPPs. The status of relevant on-going and planned research projects is also provided

  14. The effects of aging on electrical and I ampersand C components: Results of US nuclear plant aging research

    International Nuclear Information System (INIS)

    Aggarwal, S.K.; Gunther, W.E.

    1991-01-01

    The US NRC's hardware oriented engineering research program for plant aging and degradation monitoring has achieved results in the area of electrical, control, and instrumentation (ECI) components used in nuclear power plants (NPPs). The principal goals of the program, known as the Nuclear Power Plant Aging Research (NPAR) Program, are to understand the effects of age-related degradation in NPPs and how to manage and mitigate them effectively. This paper describes how these goals have been achieved for key ECI components used in the safety systems of NPPs. The status of relevant on-going and planned research projects is also provided

  15. Reliability Data Handbook for Piping Components in Nordic Nuclear Power Plants - R Book, Phase 2

    International Nuclear Information System (INIS)

    Hedtjaern Swaling, Vidar; Olsson, Anders

    2011-02-01

    This report presents results of a long research and development project financed by the regulatory body Straalsaekerhetsmyndigheten (SSM) (former SKI), the Swedish nuclear power plant licensees. The report presents a harmonized method for estimating Reliability Data for Piping Components in ASME code class 1 and 2 piping components (R-Book). Data in the R-Book is measured based on 'data driven' strategy. This first version of the R-Book comprises rupture frequencies and failure rates for all systems where ASME Code Class 1 or 2 events could be found in the OECD OPDE database. Nordic and Non-Nordic data are presented separately. Worldwide experience data is used to set up the relevant calculation cases, i.e. intersections of attributes for which there are at least one event present

  16. ITER nuclear components, preparing for the construction and R and D results

    Energy Technology Data Exchange (ETDEWEB)

    Ioki, K. E-mail: iokik@itereu.de; Akiba, M.; Barabaschi, P.; Barabash, V.; Chiocchio, S.; Daenner, W.; Elio, F.; Enoeda, M.; Ezato, K.; Federici, G.; Gervash, A.; Grebennikov, D.; Jones, L.; Kajiura, S.; Krylov, V.; Kuroda, T.; Lorenzetto, P.; Maruyama, S.; Merola, M.; Miki, N.; Morimoto, M.; Nakahira, M.; Ohmori, J.; Onozuka, M.; Rozov, V.; Sato, K.; Strebkov, Yu.; Suzuki, S.; Tanchuk, V.; Tivey, R.; Utin, Yu

    2004-08-01

    Progress has been made in the preparation of the procurement specifications for key nuclear components of ITER. Detailed design of the vacuum vessel (VV) and in-vessel components is being performed to consider fabrication methods and non-destructive tests (NDT). R and D activities are being carried out on vacuum vessel UT inspection with waves launched at an angle of 20 deg. or 30 deg. , on flow distribution tests of a two-channel model, on fabrication and testing of FW mock-ups and panels, on the blanket flexible support as a complete system including the housing, on the blanket co-axial pipe connection with guard vacuum for leak detection, and on divertor vertical target prototypes. The results give confidence in the validity of the design and identify possibilities of attractive alternate fabrication methods.

  17. ITER nuclear components, preparing for the construction and R&D results

    Science.gov (United States)

    Ioki, K.; Akiba, M.; Barabaschi, P.; Barabash, V.; Chiocchio, S.; Daenner, W.; Elio, F.; Enoeda, M.; Ezato, K.; Federici, G.; Gervash, A.; Grebennikov, D.; Jones, L.; Kajiura, S.; Krylov, V.; Kuroda, T.; Lorenzetto, P.; Maruyama, S.; Merola, M.; Miki, N.; Morimoto, M.; Nakahira, M.; Ohmori, J.; Onozuka, M.; Rozov, V.; Sato, K.; Strebkov, Yu; Suzuki, S.; Tanchuk, V.; Tivey, R.; Utin, Yu

    2004-08-01

    Progress has been made in the preparation of the procurement specifications for key nuclear components of ITER. Detailed design of the vacuum vessel (VV) and in-vessel components is being performed to consider fabrication methods and non-destructive tests (NDT). R&D activities are being carried out on vacuum vessel UT inspection with waves launched at an angle of 20° or 30°, on flow distribution tests of a two-channel model, on fabrication and testing of FW mock-ups and panels, on the blanket flexible support as a complete system including the housing, on the blanket co-axial pipe connection with guard vacuum for leak detection, and on divertor vertical target prototypes. The results give confidence in the validity of the design and identify possibilities of attractive alternate fabrication methods.

  18. Concrete component aging and its significance relative to life extension of nuclear power plants

    International Nuclear Information System (INIS)

    Naus, D.J.

    1986-09-01

    The objectives of this study are to (1) expand upon the work that was initiated in the first two Electric Power Research Institute studies relative to longevity and life extension considerations of safety-related concrete components in light-water reactor (LWR) facilities and (2) provide background that will logically lead to subsequent development of a methodology for assessing and predicting the effects of aging on the performance of concrete-based materials and components. These objectives are consistent with Nuclear Plant Aging Research (NPAR) Program goals: (1) to identify and characterize aging and service wear effects that, if unchecked, could cause degradation of structures, components, and systems and, thereby, impair plant safety; (2) to identify methods of inspection, surveillance, and monitoring or of evaluating residual life of structures, components, and systems that will ensure timely detection of significant aging effects before loss of safety function; and (3) to evaluate the effectiveness of storage, maintenance, repair, and replacement practices in mitigating the rate and extent of degradation caused by aging and service wear

  19. Nuclear technology databases and information network systems

    International Nuclear Information System (INIS)

    Iwata, Shuichi; Kikuchi, Yasuyuki; Minakuchi, Satoshi

    1993-01-01

    This paper describes the databases related to nuclear (science) technology, and information network. Following contents are collected in this paper: the database developed by JAERI, ENERGY NET, ATOM NET, NUCLEN nuclear information database, INIS, NUclear Code Information Service (NUCLIS), Social Application of Nuclear Technology Accumulation project (SANTA), Nuclear Information Database/Communication System (NICS), reactor materials database, radiation effects database, NucNet European nuclear information database, reactor dismantling database. (J.P.N.)

  20. Emerging nuclear energy systems and nuclear weapon proliferation

    International Nuclear Information System (INIS)

    Gsponer, A.; Sahin, S.; Jasani, B.

    1983-01-01

    Generally when considering problems of proliferation of nuclear weapons, discussions are focused on horizontal proliferation. However, the emerging nuclear energy systems currently have an impact mainly on vertical proliferation. The paper indicates that technologies connected with emerging nuclear energy systems, such as fusion reactors and accelerators, enhance the knowledge of thermonuclear weapon physics and will enable production of military useful nuclear materials (including some rare elements). At present such technologies are enhancing the arsenal of the nuclear weapon states. But one should not forget the future implications for horizontal proliferation of nuclear weapons as some of the techniques will in the near future be within the technological and economic capabilities of non-nuclear weapon states. Some of these systems are not under any international control. (orig.) [de

  1. Modern technical diagnostic system for the main components of powerful turbine generator

    International Nuclear Information System (INIS)

    Ezovit, G.P.; Uglyarenko, V.P.; Burlaka, S.I.; Goroz, N.I.; Orinin, S.E.; Komaritsa, V.N.; Zav'yalov, D.N.; Mazurenko, O.A.

    2011-01-01

    The modern diagnostic system to monitor the technical state of a powerful turbine generator is considered. This system permits the detection of defects in its main components and cooling system at the early stage of their development, prevention of damage and, as a consequence, emergency shutdown of nuclear power units

  2. Evaluation methods for corrosion damage of components in cooling systems of nuclear power plants by coupling analysis of corrosion and flow dynamics (1). Major targets and development strategies of the evaluation methods

    International Nuclear Information System (INIS)

    Naitoh, Masanori; Uchida, Shunsuke; Koshizuka, Seiichi; Ninokata, Hisashi; Hiranuma, Naoki; Dosaki, Koji; Nishida, Koji; Akiyama, Minoru; Saitoh, Hiroaki

    2008-01-01

    Problems in major components and structural materials in nuclear power plants have often been caused by flow induced vibration and corrosion and their overlapping effects. In order to establish safe and reliable plant operation, future problems for structural materials should be predicted based on combined analyses of flow dynamics and corrosion and they should be mitigated before becoming serious issues for plant operation. Three approaches have been prepared for predicting future problems in structural materials: 1. Computer program packages for predicting future corrosion fatigue on structural materials, 2. Computer program packages for predicting future corrosion damage on structural materials, and 3. Computer program packages for predicting wall thinning caused by flow accelerated corrosion. General features of evaluation methods and their computer packages, technical innovations required for their development, and application plans for the developed approaches for plant operation are introduced in this paper. (author)

  3. Approach to testing fusion components in existing nuclear facilities

    International Nuclear Information System (INIS)

    Hsu, P.Y.; Miller, L.G.; Longhurst, G.R.; Masson, L.S.; Kulcinski, G.L.

    1980-01-01

    The concept presented makes use of the fast spectrum in the Engineering Test Reactor (ETR) at the Idaho National Engineering Laboratory (INEL). Preliminary results show that an asymmetric, nuclear test environment with particle and radiant energy fluxes impinging on a first wall/blanket or divertor surface appears feasible in a neutron/gamma field not greatly different from that seen by a representative first wall/blanket module

  4. Applications of ultrasonic phased array technique during fabrication of nuclear tubing and other components for the Indian nuclear power program

    International Nuclear Information System (INIS)

    Kapoor, K.

    2015-01-01

    Ultrasonic phased array technique has been applied in fabrication of nuclear fuel and structural at NFC. The integrity of the nuclear fuel and structural components is most crucial as they are exposed to severe environment during operation leading to rapid degradation of its properties during its lifecycle. Nuclear Fuel Complex has mandate for the fabrication of the nuclear fuel and core structurals for Indian PHWRs/BWR, sub-assemblies for the PFBR and steam generator tubing for PFBR and PHWRs which are the most critical materials for the Indian Nuclear Power program. NDE during fabrication of these materials is thus most crucial as it provides the confidence to the designer for safe operation during its lifetime. Many of these techniques have to be developed in-house to meet unique requirements of high sensitivity, resolution and shape of the components. Some of the advancements in the NDE during the fabrication include use of ultrasonic phased array which is detailed in this paper

  5. RCC-M - Design and Conception Rules for Mechanical Components of PWR Nuclear Islands

    International Nuclear Information System (INIS)

    2007-01-01

    The design and construction rules applicable to mechanical components of PWR Nuclear Islands (RCC-M) are a part of the collection of design and construction rules for nuclear power plants. It covers the rules applicable to the design and manufacture of pressure boundaries of mechanical equipment of pressurized water reactors (PWR). The pressure components subject to the RCC-M are specified in A 4000. They include the reactor fluid systems (primary, secondary and auxiliary systems) and other components which are not subject to pressure: vessel internals, supports for pressure components subject to the RCC-M, nuclear island storage tanks. When a pressure equipment is subject to the RCC-M, all its elements subject to pressure are also, in accordance with the provisions of A 4000, and these elements are the same class as the component. In this case all the provisions of the RCC-M are applicable: design, procurement, manufacture, inspection and pressure testing. Elements which are not subject to pressure and which are subject to the RCC-M may be covered within the Code by limited specific provisions (procurement of materials for example). The other rules applicable to this equipment must be in contractual form. The assemblies comprising pressure equipment assembled by a manufacturer to constitute an integrated and functional whole, shall be subject to the rules indicated in this Code. Main objectives of Code Requirements are to ensure the integrity and mechanical stability over the equipment design life. Function ability and operability of equipment are not directly addressed in the Code. The RCC-M contributes to ensuring compliance with regulatory requirements. These requirements depend on the applicable regulatory context. The RCC-M is representative of the state of the art as concerns the design and manufacture of PWR components, ensuring an overall safety level tested through experience. The RCC-M consists of five sections, which provide rules for the design and

  6. Diagnosing component faults in a generic nuclear power plant using counterfactual and temporal reasoning

    International Nuclear Information System (INIS)

    Oehrstroem, P.; Nielsen, F.R.; Pedersen, S.A.

    1992-01-01

    The subject of main interest is the logical and epistemological aspects of diagnostic reasoning. The aim was to understand the role of conditionals and causality in this respect. A model of causal and temporal reasoning was developed and evaluated in a controlled but complex setting. The generic nuclear power plant was used as a test ground. The coherence and scope of a logical theory of diagnostic reasoning was studied in order to discover whether the theory constitutes an adequate tool for making correct diagnoses of component faults in a generic nuclear power plant. A diagnosing system based on the CIMP system was run on a computer model of a nuclear power plant, various errors were then introduced. The aim of the diagnosis is mainly explanation and only partly repair. The causal field defines a conceptual framework within which the diagnostic purpose is given and within which various diagnostic possibilities and causal relationships are given, here with regard to error detection in a control room. The causal field is tacitly given and related to the operator's training and experience. The logical aspects of the problem of the diagnosis is described. The computer model is described and the symptom language is introduced. The process of reasoning about the possible diagnosis is presented. The utilization of ideas similiar to the heuristic classification is discussed. A data base command language for manipulating lists of symptoms is described and the design of a CIMP user interface for symptom language visualization is outlined. (AB)

  7. PERFORMANCE OF ALTERNATIVE COMPONENT PRICING SYSTEMS FOR PORK

    OpenAIRE

    Brorsen, B. Wade; Akridge, Jay T.; Boland, Michael A.; Mauney, Sean; Forrest, John C.

    1998-01-01

    One method of implementing value-based marketing is a component pricing system. This research develops and evaluates alternative component pricing systems for pork. Two electronic technologies for estimating carcass components (optical probe and electromagnetic scanner) were evaluated on two sets of data representing different populations. Model accuracy increased as additional components were added.

  8. Risk and safety analysis of nuclear systems

    National Research Council Canada - National Science Library

    Lee, John C; McCormick, Norman J

    2011-01-01

    ...), and failure modes of systems. All of this material is general enough that it could be used in non-nuclear applications, although there is an emphasis placed on the analysis of nuclear systems...

  9. US Army Nuclear Burst Detection System (NBDS)

    International Nuclear Information System (INIS)

    Glaser, R.F.

    1980-07-01

    The Nuclear Burst Detection System (NBDS) was developed to meet the Army requirements of an unattended, automatic nuclear burst reporting system. It provides pertinent data for battlefield commanders on a timely basis with high reliability

  10. Cable handling system for use in a nuclear reactor

    International Nuclear Information System (INIS)

    Crosgrove, R.O.; Larson, E.M.; Moody, E.

    1982-01-01

    A cable handling system for use in an installation such as a nuclear reactor is disclosed herein along with relevant portions of the reactor which, in a preferred embodiment, is a liquid metal fast breeder reactor. The cable handling system provides a specific way of interconnecting certain internal reactor components with certain external components, through an assembly of rotatable plugs. Moreover, this is done without having to disconnect these components from one another during rotation of the plugs and yet without interfering with other reactor components in the vicinity of the rotating plugs and cable handling system

  11. Nuclear plant-aging research on reactor protection systems

    International Nuclear Information System (INIS)

    Meyer, L.C.

    1988-01-01

    This report presents the rsults of a review of the Reactor Trip System (RTS) and the Engineered Safety Feature Actuating System (ESFAS) operating experiences reported in Licensee Event Reports (LER)s, the Nuclear Power Experience data base, Nuclear Plant Reliability Data System, and plant maintenance records. Our purpose is to evaluate the potential significance of aging, including cycling, trips, and testing as contributors to degradation of the RTS and ESFAS. Tables are presented that show the percentage of events for RTS and ESFAS classified by cause, components, and subcomponents for each of the Nuclear Steam Supply System vendors. A representative Babcock and Wilcox plant was selected for detailed study. The US Nuclear Regulatory Commission's Nuclear Plant Aging Research guidelines were followed in performing the detailed study that identified materials susceptible to aging, stressors, environmental factors, and failure modes for the RTS and ESFAS as generic instrumentation and control systems. Functional indicators of degradation are listed, testing requirements evaluated, and regulatory issues discussed

  12. Study on the establishment of effective nuclear export system

    International Nuclear Information System (INIS)

    Kim, Byung Koo; So, Dong Sup; Baik, Dae Hyun; Kwack, Eun Ho; Shin, Jang Soo; Yoon, Wan Ki; Park, Wan Soo; Kim, Hyun Tae.

    1997-02-01

    To improve Korean nuclear export control system, the modification of the present export license procedure for the nuclear equipment and materials and the classification of control items and their related technologies are required. And it is also necessary to make a database of the original countries who have the right of prior consent. For the efficient export control of LWR items to DPRK, it is desirable to manage the export license scheme of nuclear reactor facility as a total package, and to prepare a control regime for the retransfer of nuclear reactor component such as reactor coolant pump and nuclear fuel whose technologies are not self-reliant. It is especially essential to prepare a systematic procedure for the supply of nuclear equipment and materials to DPRK in order to meet international guidelines of NSG and others through an accord on the nuclear cooperation between Republic of Korea (ROK) and DPRK. The principal elements to be included in the accord are the range of cooperation, the restriction within the peaceful uses, prior consent right in case of retransfer of important nuclear reactor components and of storage, transfer and changes of nuclear fuels, application of safeguards to the supplied Trigger list items, physical protection of nuclear material, requirement of the return of nuclear equipment and materials, and restriction right for the suspension or termination of the agreement. (author). 40 refs., 5 tabs., 8 figs

  13. For establishment on nuclear disaster prevention system

    International Nuclear Information System (INIS)

    Anon.

    2000-01-01

    For increasing requirement of peoples for review of nuclear disaster countermeasure at a chance of the JCO critical accident, the Japanese Government newly established the 'Special Measure Act on Nuclear Disaster Countermeasure', which was enacted on July 16, 2000. The nuclear business relatives such as electric power company and so forth established the Business program on nuclear disaster prevention in nuclear business relatives' after their consultation with local communities at their construction, under their co-operation. Simultaneously, the electric power industry field decided to intend to provide some sufficient countermeasures to incidental formation of nuclear accident such as start of the Co-operative agreement on nuclear disaster prevention among the nuclear business relatives' and so forth. Here were described on nuclear safety and disaster prevention, nuclear disaster prevention systems at the electric power industry field, abstract on 'Business program on nuclear disaster prevention in nuclear business relatives', preparation of technical assistance system for nuclear disaster prevention, executive methods and subjects on nuclear disaster prevention at construction areas, recent business on nuclear disaster prevention at the Nuclear Technical Center, and subjects on establishment of nuclear disaster prevention system. (G.K.)

  14. Analysis of appraisal tool of system security engineering capability maturity based on component

    International Nuclear Information System (INIS)

    Liu Zhenghai; Yang Xiaohua; Zou Shuliang; Liu Yachun; Xiao Jiantian; Liu Zhiming

    2012-01-01

    Spent Fuel Reprocessing is a part of nuclear fuel cycle and is the inevitably choice of nuclear power sustainable development. Reprocessing needs to face with radiological, criticality, chemical hazards. Besides using the tradition appraisal methods based on the security goals, it is a beneficial supplement that using the appraisal method of system security engineering capability maturity model based on the process. Experts should check and approve large numbers of documents during the appraisal based on system security engineering capability maturity model, so it is necessary that developing a tool to assist the expert to complete the appraisal. The method of developing software based on component is highly effective, nimble and reliable. Component technology is analyzed, the methods of extraction model domain components and general components is introduced, and the appraisal system is developed based on component technology. (authors)

  15. Nuclear maintenance and management system

    International Nuclear Information System (INIS)

    Yamaji, Yoshihiro; Abe, Norihiko

    2000-01-01

    The Mitsubishi Electric Co., Ltd. has developed to introduce various computer systems for desk-top business assistance in a power plant such as system isolation assisting system, operation parameter management system, and so on under aiming at business effectiveness since these ten and some years. Recently, by further elapsed years of the plants when required for further cost reduction and together with change of business environment represented by preparation of individual personal computer, further effectiveness, preparation of the business environment, and upgrading of maintenance in power plant business have been required. Among such background, she has carried out various proposals and developments on construction of a maintenance and management system integrated the business assistant know-hows and the plant know-hows both accumulated previously. They are composed of three main points on rationalization of business management and document management in the further effectiveness, preparation of business environment, TBM maintenance, introduction of CBM maintenance and introduction of maintenance assistance in upgrading of maintenance. Here was introduced on system concepts aiming at the further effectiveness of the nuclear power plant business, preparation of business environment, upgrading of maintenance and maintenance, and so on, at a background of environment around maintenance business in the nuclear power plants (cost-down, highly elapsed year of the plant, change of business environment). (G.K)

  16. Nuclear-powered artificial heart system

    International Nuclear Information System (INIS)

    Pouchot, W.D.; Lehrfeld, D.

    1976-01-01

    As reported to the 9th IECEC, a bench model version of a nuclear-powered artificial heart system to be used as a replacement for the natural heart was constructed and tested as part of a broader U.S. ERDA program. A report is given of the system design and integration, bench testing, and field support equipment of an implantable and advanced version of the bench model incorporating some of the component developments reported to the 10th IECEC. The basic elements of the system are a 32-watt Pu-238 heat source, a Stirling engine thermal converter, a coupling mechanism, and a mechanical blood pump drive actuating, alternatively, two artificial ventricles of polymeric material. As tested on the bench using a mock circulation, the system provides approximately 9 liters/minute at 120/80 mm Hg aortic pressure. At 190/145 mm Hg aortic pressure, the maximum flow decreases to about 7 liters/minute

  17. Nuclear electronic components of surface contamination monitor based on multi-electrode proportional counter

    International Nuclear Information System (INIS)

    Du Xiangyang; Zhang Yong; Han Shuping; Rao Xianming; Fang Jintu

    2001-01-01

    The nuclear electronic components applying in Portal Monitor and Hands and Feet Surface Contamination Monitor were based on modern integrated circuit are introduced. The detailed points in circuit design and manufacturing technique are analyzed

  18. Crankshaft and component adequacy: Update of analysis and testing developed for nuclear standby engines

    International Nuclear Information System (INIS)

    Anon.

    1987-01-01

    This book contains eight selections. Some of the topics are: reliability improvement of diesels in nuclear standby applications, diesel engine crankshaft torsional vibrations, pendulum dampers, transportation fatalities,and diesel component life predictions

  19. Nuclear power component in foresight on energy in Poland

    International Nuclear Information System (INIS)

    Szczurek, J.; Chwaszczewski, S.; Czerski, P.; Luszcz, M.

    2007-01-01

    On behalf of Ministry of Science and Higher Education, the first technology foresight study on future developments in the energy sector is being conducted in Poland. The study aimed to identify energy-related technologies, scenarios, and a mix of energy sources and infrastructure developments that will ensure security of energy supply for Poland. This paper provides a short description of the methodology applied as well as preliminary results and findings of all subtasks of the foresight study referring to the perspective of nuclear power option in Poland, embracing a time horizon of 24 years. (author)

  20. Recent progress of welding technology applied for nuclear components

    International Nuclear Information System (INIS)

    Kobayashi, T.; Hoshino, T.; Koide, H.; Yamamoto, T.; Takahashi, T.; Hashimoto, T.

    2005-01-01

    More than 30 years have been passed since the first nuclear power plant was in operation. Various needs for welding technology have been emerged and the technology has been developed. This paper first describes the key technologies in BWR power plants and then introduces ones in PWR power plants. Welding techniques are introduced in detail. Applications of arc welding, gas tungsten arc welding, electroslag welding, electron beam welding are explained. In order to avoid stress corrosion cracking, water jet and laser peening techniques are used. (author)

  1. Application of colony complex algorithm to nuclear component optimization design

    International Nuclear Information System (INIS)

    Yan Changqi; Li Guijing; Wang Jianjun

    2014-01-01

    Complex algorithm (CA) has got popular application to the region of nuclear engineering. In connection with the specific features of the application of traditional complex algorithm (TCA) to the optimization design in engineering structures, an improved method, colony complex algorithm (CCA), was developed based on the optimal combination of many complexes, in which the disadvantages of TCA were overcame. The optimized results of benchmark function show that CCA has better optimizing performance than TCA. CCA was applied to the high-pressure heater optimization design, and the optimization effect is obvious. (authors)

  2. NCIS - a Nuclear Criticality Information System (overview)

    International Nuclear Information System (INIS)

    Koponen, B.L.; Hampel, V.E.

    1983-07-01

    A Nuclear Criticality Information System (NCIS) is being established at the Lawrence Livermore National Laboratory (LLNL) in order to serve personnel responsible for safe storage, transport, and handling of fissile materials and those concerned with the evaluation and analysis of nuclear, critical experiments. Public concern for nuclear safety provides the incentive for improved access to nuclear safety information

  3. Storage, handling and movement of fuel and related components at nuclear power plants

    International Nuclear Information System (INIS)

    1979-01-01

    The report describes in general terms the various operations involved in the handling of fresh fuel, irradiated fuel, and core components such as control rods, neutron sources, burnable poisons and removable instruments. It outlines the principal safety problems in these operations and provides the broad safety criteria which must be observed in the design, operation and maintenance of equipment and facilities for handling, transferring, and storing nuclear fuel and core components at nuclear power reactor sites

  4. Failure modes of safety-related components at fires on nuclear power plants

    International Nuclear Information System (INIS)

    Aaslund, A.

    2000-03-01

    Probabilistic assessment methods can be used to identify specific plant vulnerabilities. Application of such methods can also facilitate selection among system design alternatives available for safety enhancements. The quality of assessment results is however strongly dependent on realistic and accurate input data for modelling of system component behaviour and failure modes during conditions to be assessed. Use of conservative input data may not lead to results providing guidance on safety upgrades. Adequate input data for probabilistic assessments seems to be lacking for at least failure modes of some electrical components when exposed to a fire. This report presents an attempt to improve the situation with respect to such input data. In order to take advantage of information in existing documentation of fire incident occurrences some of the lessons learned from the fire at Browns Ferry Nuclear Power Plant on March 22, 1975 are discussed in this report. Also a summary of results from different fire tests of electrical cables presented in a fire risk analysis report is a part of the references. The failure modes used to describe fire-induced damage are 'open circuit' and 'hot short' which seems to be commonly accepted terms within the branch. Definitions of the terms are included in the report. Effects of the failure modes when occurring in some of the channels of the reactor protection system are discussed with respect to the existing design of the reactor protection system at Ringhals 2 nuclear power unit. Experiences from the Browns Ferry fire and results from fire tests of electrical cables indicate that the dominating failure mode for electrical cables is 'open circuit'. An 'open circuit' failure leads to circuit disjunction and loss of continuity. The circuit can no longer transmit its signal or power. When affecting channels of the reactor protection system an 'open circuit' failure can cause extensive inadvertent actions of safety related equipment

  5. Qualification of electronic components for use in nuclear power plants

    International Nuclear Information System (INIS)

    Zorrilla, J.; Antonaccio, E; Luraschi, C.; Rodriguez, F.; Ranalli, J.; Ponce, M.; Dotro, R.; Guinda, J.

    2013-01-01

    There are a large number of instrument subjected to different service condition in a NPP. For instance different instruments can be found working in environment where the dose rate goes from negligible levels up to very harsh radiation levels. When technical specification and or equipment purchasing should be carried out it is possible to find the total leak of qualified instrument. In this context there is a need of dedicated qualification. In this work two different radiation resistance for two different I&C equipment/component were studies. The first I&C equipment was an LVDT (liner variable differential transformer). This equipment was tested while it was actuated in a strong gamma field in order to evaluate possible electromagnetic interferences a number of cycles equivalent to one year of service. After that the component was subjected to accelerated radiation aging and then actuated test under gamma field were carried out. The second I&C component to be tested was an (author)

  6. 10 CFR 110.26 - General license for the export of nuclear reactor components.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false General license for the export of nuclear reactor components. 110.26 Section 110.26 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) EXPORT AND IMPORT OF... Denmark Finland France Germany Greece Indonesia Ireland Italy Japan Latvia Lithuania Luxembourg...

  7. Aging techniques and qualified life for safety system components

    International Nuclear Information System (INIS)

    Weaver, W.W.

    1980-01-01

    Presently, the qualified life objective for Class IE safety system components in nuclear power plants is somewhat of a subjective engineering judgment. When the desired qualified life is ascertained, there are other choices that must be made (which may be influenced by the desired qualified life) such as selecting the aging procedure to use in the qualification process. Adding complexity to the situation is the fact that there are some limitations in aging techniques at the present time. This article presents (1) a discussion of the limitations in aging procedures, (2) the general philosophy of qualification, and (3) a proposed method for specifying a desired qualified life, which uses a probabilistic approach. The probabilistic approach proposed in item 3 can be applied to natural aging programs and eventually to accelerated aging once the present technical difficulties are overcome

  8. The selection of field component reliability data for use in nuclear safety studies

    International Nuclear Information System (INIS)

    Coxson, B.A.; Tabaie, Mansour

    1990-01-01

    The paper reviews the user requirements for field component failure data in nuclear safety studies, and the capability of various data sources to satisfy these requirements. Aspects such as estimating the population of items exposed to failure, incompleteness, and under-reporting problems are discussed. The paper takes as an example the selection of component reliability data for use in the Pre-Operational Safety Report (POSR) for Sizewell 'B' Power Station, where field data has in many cases been derived from equipment other than that to be procured and operated on site. The paper concludes that the main quality sought in the available data sources for such studies is the ability to examine failure narratives in component reliability data systems for equipment performing comparable duties to the intended plant application. The main benefit brought about in the last decade is the interactive access to data systems which are adequately structured with regard to the equipment covered, and also provide a text-searching capability of quality-controlled event narratives. (author)

  9. Lifetime assessment and lifetime management for key components of nuclear power plants

    International Nuclear Information System (INIS)

    Dou Yikang; Sun Hanhong; Qu Jiadi

    2000-01-01

    On the bases of investigation on recent development of plant lifetime management in the world, the author gives some points of view on how to establish plant lifetime assessment (PLA) and management (PLM) systems for Chinese nuclear power plants. The main points lie in: 1) safety regulatory organizations, utilities and R and D institutes work cooperatively for PLA and PLM; 2) PLA and PLM make a interdependent cycle, which means that a good PLM system ensures authentic input for PLA, while veritable PLA provides valuable feedback for PLM improvement; 3) PLA and PLM should be initiated for some key components. The author also analyzes some important problems to be tackled in PLA and PLM from the view angle of a R and D institute

  10. Manufacture of piping components for nuclear power plants

    International Nuclear Information System (INIS)

    Bartecek, R.

    1983-01-01

    Hammer forging of hollow forging ingots, extrusion and elestroslag remelting may be used for the manufacture of large pipes. Technologies have been developed for the manufacture of elbows based on various types of forming. These procedures mainly include the hydraulic pressing of elbows from tubes and the pressing of symmetrical halves of elbows with subsequent welding. The hammer forging of valves, cross pieces, etc., has been replaced by forging and pressing. In order to prevent failures from occurring in the pipes during operation of nuclear power plants, pipes are being made of larger forgings, which reduces the number of welds. This improves the quality of the pipes, reduces production and assembly costs and is metal-saving. (E.S.)

  11. Electron beam welding: study of process capabilities and limitations towards development of nuclear components

    International Nuclear Information System (INIS)

    Vadolia, Gautam; Singh, Kongkham Premjit

    2015-01-01

    Electron beam (EB) welding technology is an established and widely adopted technique in nuclear research and development area. Electron Beam welding is thought of as a candidate process for ITER Vacuum Vessel Fabrication. Dhruva Reactor @ BARC, Mumbai and Niobium Superconducting accelerator Cavitity @ BARC has adopted the EB welding technique as a fabrication route. The highly concentrated energy input of the electron beam has added the advantages over the conventional welding as being less HAZ and provided smooth and clean surface. EB Welding has also been used for the joining of various reactive and refractory materials. EB system as heat source has also been used for vacuum brazing application. The Welding Institute (TWI) has demonstrated that EBW is potentially suitable to produce high integrity joints in 50 mm pure copper. TWI has also examined 150 kV Reduced Pressure Electron Beam (RPEB) gun in welding 140 mm and 147 mm thickness Nuclear Reactor Pressure Vessel Steel (SA 508 grade). EBW in 10 mm thick SS316 plates were studied at IPR and results were encouraging. In this paper, the pros and cons and role of electron beam process will be studied to analyze the importance of electron beam welding in nuclear components fabrication. Importance of establishing the high precision Wire Electro Discharge Machining (WEDM) facility will also be discussed. (author)

  12. Unattended nuclear systems for local energy supply

    International Nuclear Information System (INIS)

    Lynch, G.F.; Bancroft, A.R.; Hilborn, J.W.; McDougall, D.S.; Ohta, M.M.

    1988-02-01

    This paper describes recent developments in a small nuclear heat and electricity production system - the SLOWPOKE Energy System - that make it possible to locate the system close to the load, and that could have a major impact on local energy supply. The most important unique features arising from these developments are walk-away safety and the ability to operate in an unattended mode. Walk-away safety means that radiological protection is provided by intrinsic characteristics and does not depend on either engineered safety systems or operator intervention. This, in our view, is essential to public acceptance. The capability for unattended operation results from self-regulation; however, the performance can be remotely monitored. The SLOWPOKE Energy System consists of a water-filled pool, operating at atmospheric pressure, which cools and moderates a beryllium-reflected thermal reactor that is fuelled with 100 to 400 kg of low-enriched uranium. The pool water also provides shielding from radioactive materials trapped in the fuel. Heat is drawn from the pool and transferred either to a building hot-water distribution system or to an organic liquid which is converted to vapour to drive a turbine-generator unit. Heating loads between 2 qnd 10 MWt, and electrical loads up to 1 MWe can be satisfied. SLOWPOKE is a dramatic departure from conventional nuclear power reactors. Its nuclear heat source is intrinsically simple, having only one moving part: a solid neutron absorber which is slowly withdrawn from the reactor to balance the fuel burnup. Its power is self-regulated and excessive heat production cannot occur, even for the most severe combinations of system failure. Cooling of the fuel is assured by natural physical processes that do not depend on mechanical components such as pumps. These intrinsic characteristics assure public safety and ultra high reliability

  13. Unattended nuclear systems for local energy supply

    International Nuclear Information System (INIS)

    Lynch, G.F.; Bancroft, A.R.; Hilborn, J.W.; McDougall, D.S.; Ohta, M.M.

    1986-10-01

    This paper describes recent developments in a small nuclear heat and electricity production system - the SLOWPOKE energy system - that make it possible to locate the system close to the load, and that could have a major impact on local energy supply. The most important unique features arising from these developments are walk-away safety and the ability to operate in an unattended mode. Walk-away safety means that radiological protection is provided by intrinsic characteristics and does not depend on either engineered safety systems or operator intervention. This, in our view, is essential to public acceptance. The capability for unattended operation results from self-regulation, however the performance can be remotely monitored. The SLOWPOKE energy system consists of a water-filled pool, operating at atmospheric pressure, which cools and moderates a beryllium-reflected thermal reactor that is fuelled with 100 to 400 kg of low enriched uranium. The pool water also provides shielding from radioactive materials trapped in the fuel. Heat is drawn from the pool and transferred either to a building hot-water distribution system or to an organic liquid which is converted to vapour to drive a turbine-generator unit. Heating loads between 2 and 10 MWt, and electrical loads up to 1 MWe can be satisfied. SLOWPOKE is a dramatic departure from conventional nuclear power reactors. Its nuclear heat source is intrinsically simple, having only one moving part: a solid neutron absorber which is slowly withdrawn from the reactor to balance the fuel burnup. Its power is self-regulated and excessive heat production cannot occur, even for the most severe combinations of system failure. Cooling of the fuel is assured by natural physical processes that do not depend on mechanical components such as pumps. These intrinsic characteristics assure public safety and ultra high reliability. (author)

  14. Development of expert system for structural design of FBR components

    International Nuclear Information System (INIS)

    Ueda, Hiroyoshi; Uno, Masayoshi; Ogawa, Hiroshi; Shimakawa, Takashi; Yoshimura, Shinobu; Yagawa, Genki.

    1995-01-01

    The characteristics of structural design processes for nuclear components can be summarized as follows : (1) Many engineers belonging to different fields are working in parallel, exchanging a huge amount of data and information. (2) A final solution is determined after a number of iterative design processes. (3) Solutions have to be examined many times based on sophisticated design codes. (4) Sophisticated calculation methods such as the finite element method are frequently utilized, and experts' knowledge on such analyses plays important roles in the design process. Taking these issues into consideration, a new expert system for structural design is developed in the present study. Here, the object-oriented data flow mechanism and the blackboard model are utilized to systematize structural design processes in a computer. An automated finite element calculation module is implemented, and experts' knowledge is stored in knowledge base. In addition, a new algorithm is employed to automatically draw the design window, which is defined as an area of permissible solutions in a design parameter space. The developed system is successfully applied to obtain the design windows of four components selected from the demonstration FBR structures. (author)

  15. Reliability of piping system components. Volume 1: Piping reliability - A resource document for PSA applications

    Energy Technology Data Exchange (ETDEWEB)

    Nyman, R; Erixon, S; Tomic, B; Lydell, B

    1995-12-01

    SKI has undertaken a multi-year research project to establish a comprehensive passive component failure database, validate failure rate parameter estimates and establish a model framework for integrating passive component failures in existing PSAs. Phase 1 of the project produced a relational database on worldwide piping system failure events in the nuclear and chemical industries. This phase 2 report gives a graphical presentation of piping system operating experience, and compares key failure mechanisms in commercial nuclear power plants and chemical process industry. Inadequacies of traditional PSA methodology are addressed, with directions for PSA methodology enhancements. A data-driven-and-systems-oriented analysis approach is proposed to enable assignment of unique identities to risk-significant piping system component failures. Sufficient operating experience does exist to generate quality data on piping failures. Passive component failures should be addressed by today`s PSAs to allow for aging analysis and effective, on-line risk management. 111 refs, 36 figs, 20 tabs.

  16. Reliability of piping system components. Volume 1: Piping reliability - A resource document for PSA applications

    International Nuclear Information System (INIS)

    Nyman, R.; Erixon, S.; Tomic, B.; Lydell, B.

    1995-12-01

    SKI has undertaken a multi-year research project to establish a comprehensive passive component failure database, validate failure rate parameter estimates and establish a model framework for integrating passive component failures in existing PSAs. Phase 1 of the project produced a relational database on worldwide piping system failure events in the nuclear and chemical industries. This phase 2 report gives a graphical presentation of piping system operating experience, and compares key failure mechanisms in commercial nuclear power plants and chemical process industry. Inadequacies of traditional PSA methodology are addressed, with directions for PSA methodology enhancements. A data-driven-and-systems-oriented analysis approach is proposed to enable assignment of unique identities to risk-significant piping system component failures. Sufficient operating experience does exist to generate quality data on piping failures. Passive component failures should be addressed by today's PSAs to allow for aging analysis and effective, on-line risk management. 111 refs, 36 figs, 20 tabs

  17. Foam decontamination of large nuclear components before dismantling

    International Nuclear Information System (INIS)

    Costes, J.R.; Sahut, C.

    1998-01-01

    Following some simple theoretical considerations, the authors show that foam compositions can be advantageously circulated them for a few hours in components requiring decontamination before dismantling. The technique is illustrated on six large ferritic steel valves, then on austenitic steel heat exchangers for which the Ce(III)/Ce(IV) redox pair was used to dissolve the chromium; Ce(III) was reoxidized by ozone injection into the foam vector gas. Biodegradable surfactants are sued in the process; tests have shown that the foaming power disappears after a few days, provided the final radioactive liquid waste is adjusted to neutral pH, allowing subsequent coprecipitation of concentration treatment. (author)

  18. Reliability for systems of degrading components with distinct component shock sets

    International Nuclear Information System (INIS)

    Song, Sanling; Coit, David W.; Feng, Qianmei

    2014-01-01

    This paper studies reliability for multi-component systems subject to dependent competing risks of degradation wear and random shocks, with distinct shock sets. In practice, many systems are exposed to distinct and different types of shocks that can be categorized according to their sizes, function, affected components, etc. Previous research primarily focuses on simple systems with independent failure processes, systems with independent component time-to-failure, or components that share the same shock set or type of shocks. In our new model, we classify random shocks into different sets based on their sizes or function. Shocks with specific sizes or function can selectively affect one or more components in the system but not necessarily all components. Additionally the shocks from the different shock sets can arrive at different rates and have different relative magnitudes. Preventive maintenance (PM) optimization is conducted for the system with different component shock sets. Decision variables for two different maintenance scheduling problems, the PM replacement time interval, and the PM inspection time interval, are determined by minimizing a defined system cost rate. Sensitivity analysis is performed to provide insight into the behavior of the proposed maintenance policies. These models can be applied directly or customized for many complex systems that experience dependent competing failure processes with different component shock sets. A MEMS (Micro-electro mechanical systems) oscillator is a typical system subject to dependent and competing failure processes, and it is used as a numerical example to illustrate our new reliability and maintenance models

  19. Dynamic interactions of components, structure, and foundation of nuclear power facilities

    International Nuclear Information System (INIS)

    Pajuhesh, J.; Hadjian, A.H.

    1977-01-01

    A solution is formulated for the dynamic analysis of structures and components with different stiffness and damping characteristics, including the consideration of soil-structure interaction effects. Composite structures are often analysed approximately, in particular with regards to damping. For example, the reactor and other equipment in nuclear power plant structures are often analysed by assuming them uncoupled from the supporting structures. To achieve a better accuracy, the coupled system is hereby analysed as a composite component-structure-soil system. Although derivation of mass and stiffness matrices for the component-structure-soil system is a simple problem, the determination of the damping characteristics of such a system is more complex. This emphasis on the proper evaluation of system damping is warranted on the grounds that, when resonance conditions occur, the response amplitude is governed to a significant degree by the system damping. The damping information is usually available for each sub-structure separately with its based fixed or devoid of rigid-body modes of motion. The rigid-body motions are often free of damping resistance but sometimes, such as in the case of soil-structure interaction, or in the case of aerodynamic resistance, are uniquely defined. The composite damping matrix for the complete structure is hereby derived from the above-mentioned information. Thus, the damping matrix is first obtained for the free-free model of each sub-structure (the model containing the structural degrees of freedom together with rigid-body modes of motion), and then the submatrices for the free-free models are assembled to form the composite damping matrix in acccordance with an assembly technique relating the sub-structure coordinates to the global coordinates of the composite structure

  20. Systems approach to nuclear waste glass development

    International Nuclear Information System (INIS)

    Jantzen, C.M.

    1986-01-01

    Development of a host solid for the immobilization of nuclear waste has focused on various vitreous wasteforms. The systems approach requires that parameters affecting product performance and processing be considered simultaneously. Application of the systems approach indicates that borosilicate glasses are, overall, the most suitable glasses for the immobilization of nuclear waste. Phosphate glasses are highly durable; but the glass melts are highly corrosive and the glasses have poor thermal stability and low solubility for many waste components. High-silica glasses have good chemical durability, thermal stability, and mechanical stability, but the associated high melting temperatures increase volatilization of hazardous species in the waste. Borosilicate glasses are chemically durable and are stable both thermally and mechanically. The borosilicate melts are generally less corrosive than commercial glasses, and the melt temperature miimizes excessive volatility of hazardous species. Optimization of borosilicate waste glass formulations has led to their acceptance as the reference nuclear wasteform in the United States, United Kingdom, Belgium, Germany, France, Sweden, Switzerland, and Japan

  1. Detection and mitigation of aging and service wear effects of nuclear power plant components in Canada

    International Nuclear Information System (INIS)

    Pachner, J.

    1987-07-01

    In Canada, the operational safety management of nuclear power plants employs methods which are intended to prevent, detect, correct and mitigate system and component failures from any cause, including the effects of aging and service wear degradation. The paper gives an overview of the application of these methods in the detection and mitigation of aging effects before they impact on plant safety and production reliability. Regulatory audits of these methods, to ensure that an acceptable level of plant safety is maintained by the nuclear power plant licensees, are also described. The methods are: a preventive maintenance program, Significant Event Reporting system, and a reliability based assessment of performance of safety related systems. The above methods are discussed and illustrated by examples. The soundness of the approach has been proven by the results achieved in 163 reactor-years of operation. Present and future developments include reviews of current monitoring, testing and inspection methods to ensure that appropriate time variant parameters (capable of revealing aging degradation before loss of functional capability) are monitored, and reviews of the effectiveness of existing maintenance programs and methods in mitigating aging and service wear effects

  2. Establishment of nuclear data system

    International Nuclear Information System (INIS)

    Chang, Jong Hwa; Kim, J. D.; Oh, S. Y.; Lee, Y. O.; Gil, C. S.; Cho, Y. S.

    1997-01-01

    Fission fragment data have been collected and added to the existing nuclear database system. A computer program was written for generating on-line graphs of energy-dependent neutron reaction cross section. This program deals with about 300 major nuclides and serves on the internet. As a part of nuclear data evaluation works, the covariance data for neutron cross section of structural nuclides were evaluated. Also the elastic and inelastic cross sections were evaluated by using ABAREX and EGNASH2 code. In the field of nuclear data processing, a cross section library for TWODANT code for liquid metal reactor was generated and validated against Russian and French critical reactors. The resonance data for Pu-242 in CASMO-3 library were updated. In addition, continuous-energy libraries for MCNP were generated from ENDF/B-VI.2, JEF-2.2 and JENDL-3.2. These libraries were validated against the results from a series of critical experiments at HANARO. (author). 87 refs., 29 tabs., 23 figs

  3. Nuclear material control systems for nuclear power plants

    International Nuclear Information System (INIS)

    1975-06-01

    Paragraph 70.51(c) of 10 CFR Part 70 requires each licensee who is authorized to possess at any one time special nuclear material in a quantity exceeding one effective kilogram to establish, maintain, and follow written material control and accounting procedures that are sufficient to enable the licensee to account for the special nuclear material in his possession under license. While other paragraphs and sections of Part 70 provide specific requirements for nuclear material control systems for fuel cycle plants, such detailed requirements are not included for nuclear power reactors. This guide identifies elements acceptable to the NRC staff for a nuclear material control system for nuclear power reactors. (U.S.)

  4. Development of life evaluation technology for nuclear power plant components

    Energy Technology Data Exchange (ETDEWEB)

    Song, Sung Jin; Kim, Young Hwan; Shin, Hyun Jae [Sungkwunkwan Univ., Seoul (Korea, Republic of); Lee, Hyang Beom [Soongsil Univ., Seoul (Korea, Republic of); Shin, Young Kil [Kunsan National Univ., Gunsan (Korea, Republic of); Chung, Hyun Jo [Wonkwang Univ., Iksan (Korea, Republic of); Park, Ik Keun; Park, Eun Soo [Seoul National University of Technology, Seoul (Korea, Republic of)

    2001-03-15

    Retaining reliabilities of nondestructive testing is essential for the life-time maintenance of nuclear power plant. In order to Improve reliabilities of ultrasonic testing and eddy current testing, the following five subjects were carried out in this study: development of BEM analysis technique for ECT of SG tube, development of neural network technique for the intelligent analysis of ECT flaw signals of SG tubes, development of RFECT technology for the inspection of SG tube, FEM analysis of ultrasonic scattering field and evaluation of statistical reliability of PD-RR test of ultrasonic testing. As results, BEM analysis of eddy current signal, intelligent analysis of eddy current signal using neural network, and FEM analysis of remote field eddy current testing have been developed for the inspection of SG tubes. FEM analysis of ultrasonic waves in 2-dimensional media and evaluation of statistical reliability of ultrasonic testing with PD-RR test also have been carried out for the inspection of weldments. Those results can be used to Improve reliability of nondestructive testing.

  5. Development of life evaluation technology for nuclear power plant components

    International Nuclear Information System (INIS)

    Song, Sung Jin; Kim, Young Hwan; Shin, Hyun Jae; Lee, Hyang Beom; Shin, Young Kil; Chung, Hyun Jo; Park, Ik Keun; Park, Eun Soo

    2001-03-01

    Retaining reliabilities of nondestructive testing is essential for the life-time maintenance of nuclear power plant. In order to Improve reliabilities of ultrasonic testing and eddy current testing, the following five subjects were carried out in this study: development of BEM analysis technique for ECT of SG tube, development of neural network technique for the intelligent analysis of ECT flaw signals of SG tubes, development of RFECT technology for the inspection of SG tube, FEM analysis of ultrasonic scattering field and evaluation of statistical reliability of PD-RR test of ultrasonic testing. As results, BEM analysis of eddy current signal, intelligent analysis of eddy current signal using neural network, and FEM analysis of remote field eddy current testing have been developed for the inspection of SG tubes. FEM analysis of ultrasonic waves in 2-dimensional media and evaluation of statistical reliability of ultrasonic testing with PD-RR test also have been carried out for the inspection of weldments. Those results can be used to Improve reliability of nondestructive testing

  6. Nuclear Analyses of Indian LLCB Test Blanket System in ITER

    Science.gov (United States)

    Swami, H. L.; Shaw, A. K.; Danani, C.; Chaudhuri, Paritosh

    2017-04-01

    Heading towards the Nuclear Fusion Reactor Program, India is developing Lead Lithium Ceramic Breeder (LLCB) tritium breeding blanket for its future fusion Reactor. A mock-up of the LLCB blanket is proposed to be tested in ITER equatorial port no.2, to ensure the overall performance of blanket in reactor relevant nuclear fusion environment. Nuclear analyses play an important role in LLCB Test Blanket System design & development. It is required for tritium breeding estimation, thermal-hydraulic design, coolants process design, radioactive waste management, equipment maintenance & replacement strategies and nuclear safety. The nuclear behaviour of LLCB test blanket module in ITER is predicated in terms of nuclear responses such as tritium production, nuclear heating, neutron fluxes and radiation damages. Radiation shielding capability of LLCB TBS inside and outside bio-shield was also assessed to fulfill ITER shielding requirements. In order to supports the rad-waste and safety assessment, nuclear activation analyses were carried out and radioactivity data were generated for LLCB TBS components. Nuclear analyses of LLCB TBS are performed using ITER recommended nuclear analyses codes (i.e. MCNP, EASY), nuclear cross section data libraries (i.e. FENDL 2.1, EAF) and neutronic model (ITER C-lite v.l). The paper describes a comprehensive nuclear performance of LLCB TBS in ITER.

  7. Nuclear material statistical accountancy system

    International Nuclear Information System (INIS)

    Argentest, F.; Casilli, T.; Franklin, M.

    1979-01-01

    The statistical accountancy system developed at JRC Ispra is refered as 'NUMSAS', ie Nuclear Material Statistical Accountancy System. The principal feature of NUMSAS is that in addition to an ordinary material balance calcultation, NUMSAS can calculate an estimate of the standard deviation of the measurement error accumulated in the material balance calculation. The purpose of the report is to describe in detail, the statistical model on wich the standard deviation calculation is based; the computational formula which is used by NUMSAS in calculating the standard deviation and the information about nuclear material measurements and the plant measurement system which are required as data for NUMSAS. The material balance records require processing and interpretation before the material balance calculation is begun. The material balance calculation is the last of four phases of data processing undertaken by NUMSAS. Each of these phases is implemented by a different computer program. The activities which are carried out in each phase can be summarised as follows; the pre-processing phase; the selection and up-date phase; the transformation phase, and the computation phase

  8. INIS - International Nuclear Information System

    International Nuclear Information System (INIS)

    Nevyjel, A.

    1983-10-01

    The International Nuclear Information System is operated by the IAEA in close cooperation with its participating countries. Each country is responsible for the acquisition of the literature published within its boundaries. These data are collected by the INIS secretariat in Vienna and the resulting comprehensive data base is available for all member states. On behalf of Austrian Federal Chancellor's Office the Austrian Research Centre Seibersdorf operates the Austrian INIS-Center, which offers information services in form of retrospective searches and current awareness services. (Author) [de

  9. Supersymmetry for nuclear cluster systems

    International Nuclear Information System (INIS)

    Levai, G.; Cseh, J.; Van Isacker, P.

    2001-01-01

    A supersymmetry scheme is proposed for nuclear cluster systems. The bosonic sector of the superalgebra describes the relative motion of the clusters, while its fermionic sector is associated with their internal structure. An example of core+α configurations is discussed in which the core is a p-shell nucleus and the underlying superalgebra is U(4/12). The α-cluster states of the nuclei 20 Ne and 19 F are analysed and correlations between their spectra, electric quadrupole transitions, and one-nucleon transfer reactions are interpreted in terms of U(4/12) supersymmetry. (author)

  10. Nuclear detection systems in traffic

    International Nuclear Information System (INIS)

    Farkas, T.; Pernicka, L.; Svec, A.

    2005-01-01

    Illicit trafficking in nuclear materials (nuclear criminality) has become a problem, due to the circulation of a high number of radioactive sources caused by the changes of the organisational infrastructures to supervise these material within the successor states of the former Soviet Union. Aim of this paper is to point out the technical requirements and the practicability of an useful monitoring system at preselected traffic check points (railway and highway border crossings, industrial sites entry gates, international airports). The ITRAP lab test was designed to work as strict benchmark to qualify border monitoring systems 67 with very low false alarm rates, in addition the minimum sensitivity to give an alarm has been defined for fix-installed systems, pocket type and hand held instruments. For the neutron tests a special prepared Californium source ( 252 Cf) was used to simulate the weapons plutonium. The source is shielded against gamma radiation, use a moderator and provides the required neutron rate of 20000 n/s at 2 rn distance. To test the false alarm rate (rate of false positive ) the same test facility , under the same background conditions, was used but without a radioactive test source. The ITRAP lab tests for the fix-installed systems started at May 1998 and first results were given in September 1998. Only 2 of 14 fix-installed monitoring systems could fulfil the minimum requirement for neutron detection. 7 of 14 fix-installed monitoring systems (50%) passed the ITRAP lab test. The analytical method developed and used for certification of installed radiation monitors in the Slovak Institute of Metrology consists in measurement of radiation activity of selected radionuclide in defined conditions. (authors)

  11. Seismic fragility of nuclear power plant components (Phase 2): A fragility handbook on eighteen components

    International Nuclear Information System (INIS)

    Bandyopadhyay, K.K.; Hofmayer, C.H.; Kassir, M.K.; Shteyngart, S.

    1991-06-01

    Fragility estimates of seven equipment classes were published in earlier reports. This report presents fragility analysis results from eleven additional equipment categories. The fragility levels are expressed in probabilistic terms. For users' convenience, this concluding report includes a summary of fragility results of all eighteen equipment classes. A set of conversion factors based on judgment is recommended for use of the information for early vintage equipment. The knowledge gained in conducting the Component Fragility Program and similar other programs is expected to provide a new direction for seismic verification and qualification of equipment. 15 refs., 12 tabs

  12. Nuclear Systems (NS): Technology Demonstration Unit (TDU)

    Data.gov (United States)

    National Aeronautics and Space Administration — The Nuclear Systems Project demonstrates nuclear power technology readiness to support the goals of NASA's Space Technology Mission Directorate. To this end, the...

  13. Application of risk-based methods for inspection of nuclear power plant components

    International Nuclear Information System (INIS)

    Balkey, K.R.

    1992-01-01

    In-service inspections (ISIs) can play a significant role in minimizing equipment and structural failures. All aspects of inspections, i.e., objectives, method, timing, and the acceptance criteria for detected flaws can affect the probability of component failure. Where ISI programs exist, they are primarily based on prior experience and engineering judgment. At best, some include an implicit consideration of risk (probability of failure multiplied by consequence). Since late 1988, a multidisciplined American Society of Mechanical Engineers (ASME) Research Task Force on Risk-Based Inspection Guidelines has been addressing the general question of how to formally incorporate risk considerations into plans and requirements for the ISI of components and structural systems. The task force and steering committee that guided the project have concluded that appropriate analytical methods exist for evaluating and quantifying risks associated with pressure boundary and structural failures. With the support of about a dozen industry and government organizations, the research group has recommended a general methodology for establishing a risk-based inspection program that could be applied to any nuclear system or structural system

  14. Chloroplast two-component systems: evolution of the link between photosynthesis and gene expression.

    Science.gov (United States)

    Puthiyaveetil, Sujith; Allen, John F

    2009-06-22

    Two-component signal transduction, consisting of sensor kinases and response regulators, is the predominant signalling mechanism in bacteria. This signalling system originated in prokaryotes and has spread throughout the eukaryotic domain of life through endosymbiotic, lateral gene transfer from the bacterial ancestors and early evolutionary precursors of eukaryotic, cytoplasmic, bioenergetic organelles-chloroplasts and mitochondria. Until recently, it was thought that two-component systems inherited from an ancestral cyanobacterial symbiont are no longer present in chloroplasts. Recent research now shows that two-component systems have survived in chloroplasts as products of both chloroplast and nuclear genes. Comparative genomic analysis of photosynthetic eukaryotes shows a lineage-specific distribution of chloroplast two-component systems. The components and the systems they comprise have homologues in extant cyanobacterial lineages, indicating their ancient cyanobacterial origin. Sequence and functional characteristics of chloroplast two-component systems point to their fundamental role in linking photosynthesis with gene expression. We propose that two-component systems provide a coupling between photosynthesis and gene expression that serves to retain genes in chloroplasts, thus providing the basis of cytoplasmic, non-Mendelian inheritance of plastid-associated characters. We discuss the role of this coupling in the chronobiology of cells and in the dialogue between nuclear and cytoplasmic genetic systems.

  15. Order in large and chaos in small components of nuclear wave functions

    International Nuclear Information System (INIS)

    Soloviev, V.G.

    1992-06-01

    An investigation of the order and chaos of the nuclear excited states has shown that there is order in the large and chaos in the small quasiparticle or phonon components of the nuclear wave functions. The order-to-chaos transition is treated as a transition from the large to the small components of the nuclear wave function. The analysis has shown that relatively large many-quasiparticle components of the wave function at an excitation energy (4-8)MeV may exist. The large many-quasiparticle components of the wave functions of the neutron resonances are responsible for enhanced E1-, M1- and E2-transition probabilities from neutron resonance to levels lying (1-2)MeV below them. (author)

  16. Calculations of atomic magnetic nuclear shielding constants based on the two-component normalized elimination of the small component method

    Science.gov (United States)

    Yoshizawa, Terutaka; Zou, Wenli; Cremer, Dieter

    2017-04-01

    A new method for calculating nuclear magnetic resonance shielding constants of relativistic atoms based on the two-component (2c), spin-orbit coupling including Dirac-exact NESC (Normalized Elimination of the Small Component) approach is developed where each term of the diamagnetic and paramagnetic contribution to the isotropic shielding constant σi s o is expressed in terms of analytical energy derivatives with regard to the magnetic field B and the nuclear magnetic moment 𝝁 . The picture change caused by renormalization of the wave function is correctly described. 2c-NESC/HF (Hartree-Fock) results for the σiso values of 13 atoms with a closed shell ground state reveal a deviation from 4c-DHF (Dirac-HF) values by 0.01%-0.76%. Since the 2-electron part is effectively calculated using a modified screened nuclear shielding approach, the calculation is efficient and based on a series of matrix manipulations scaling with (2M)3 (M: number of basis functions).

  17. Computation of the mechanical behaviour of nuclear reactor components

    International Nuclear Information System (INIS)

    Brosi, S.; Niffenegger, M.; Roesel, R.; Reichlin, K.; Duijvestijn, A.

    1994-01-01

    A possible limiting factor of the service life of a reactor is the mechanical load carrying margin, i.e. the excess of the load carrying capacity over the actual loading, of the central, heavy section components. This margin decreases during service but, for safety reasons, may not fall below a critical value. Therefore, it is essential to check and to control continuously the factors which cause the decrease. The reasons for the decrease are shown at length and in detail in an example relating to the test which almost achieved failure of a pipe emanating from a reactor pressure vessel, weakened by an artificial crack and undergoing a water-hammer loading. The latter was caused by a sudden valve closure supposed to follow upon a break far downstream. The computational and experimental difficulties associated with the simultaneous occurrence of an extreme weakening and an extreme loading in an already rather complicated geometry are explained. It is concluded that available computational tools and present know-how are sufficient to simulate the behaviour under such conditions as would prevail in normal service, and even to analyse departures from them, as long as not all difficulties arise simultaneously. (author) figs., tabs., refs

  18. Applications of Advanced Electromagnetics Components and Systems

    CERN Document Server

    Kouzaev, Guennadi A

    2013-01-01

    This text, directed to the microwave engineers and Master and PhD students, is on the use of electromagnetics to the development and design of advanced integrated components distinguished by their extended field of applications. The results of hundreds of authors scattered in numerous journals and conference proceedings are carefully reviewed and classed.  Several chapters are to refresh the knowledge of readers in advanced electromagnetics. New techniques are represented by compact electromagnetic–quantum equations which can be used in modeling of microwave-quantum integrated circuits of future In addition, a topological method to the boundary value problem analysis is considered with the results and examples.  One extended chapter is for the development and design of integrated components for extended bandwidth applications, and the technology and electromagnetic issues of silicon integrated transmission lines, transitions, filters, power dividers, directional couplers, etc are considered. Novel prospec...

  19. Composite type nuclear power system

    International Nuclear Information System (INIS)

    Nakamoto, Koichiro.

    1993-01-01

    The present invention realizes a high thermal efficiency by heating steams at the exit of a steam generator of a nuclear power plant to high temperature by a thermal super-heating boiler. That is, a thermal superheating boiler is disposed between the steam generator and a turbogenerator to heat steams from the steam generator and supply them to the turbogenerator. In this case, it may be possible that feedwater superheating boiler pipelines to the steam generator are caused to pass through the thermal superheating boiler so that they also have a performance of heating feedwater. If the system of the present invention is used, it is possible to conduct base load operation by nuclear power and a load following operation by controlling the thermal superheating boiler. Further, a hydrogen producing performance is applied to the thermal superheating boiler to produce hydrogen when electric power load is lowered. An internally sustaining type operation method can be conducted of burning hydrogen by the superheating boiler upon increased electric power load. As a result, a power generation system which has an excellent economical property and can easily cope with the load following operation can be attained. (I.S.)

  20. Components of Maternal Healthcare Delivery System Contributing to ...

    African Journals Online (AJOL)

    Components of Maternal Healthcare Delivery System Contributing to Maternal Deaths ... transcripts were analyzed using a directed approach to content analysis. Excerpts were categorized according to three main components of the maternal ...

  1. Nuclear integrated database and design advancement system

    International Nuclear Information System (INIS)

    Ha, Jae Joo; Jeong, Kwang Sub; Kim, Seung Hwan; Choi, Sun Young.

    1997-01-01

    The objective of NuIDEAS is to computerize design processes through an integrated database by eliminating the current work style of delivering hardcopy documents and drawings. The major research contents of NuIDEAS are the advancement of design processes by computerization, the establishment of design database and 3 dimensional visualization of design data. KSNP (Korea Standard Nuclear Power Plant) is the target of legacy database and 3 dimensional model, so that can be utilized in the next plant design. In the first year, the blueprint of NuIDEAS is proposed, and its prototype is developed by applying the rapidly revolutionizing computer technology. The major results of the first year research were to establish the architecture of the integrated database ensuring data consistency, and to build design database of reactor coolant system and heavy components. Also various softwares were developed to search, share and utilize the data through networks, and the detailed 3 dimensional CAD models of nuclear fuel and heavy components were constructed, and walk-through simulation using the models are developed. This report contains the major additions and modifications to the object oriented database and associated program, using methods and Javascript.. (author). 36 refs., 1 tab., 32 figs

  2. Methods and means of the radioisotope flaw detection of the nuclear power reactors components

    International Nuclear Information System (INIS)

    Dekopov, A.S.; Majorov, A.N.; Firsov, V.G.

    1979-01-01

    Methods and means are considered for the radioisotopic flaw detection of the nuclear reactors pressure vessels and structural components of the reactor circuit. Methods of control are described as in the technological process of fabrication of the power reactors assemblies as during the systematic-preventive repair of the nuclear power station equipment during exploitation. Methodological base is given of the technology of radiation control of welded joints of the pressure vessel branch piper of the WWER-440 and WWER-1000 reactors in the process of assembling and exploitation and joining pipes with the pipe-plate of the steamgenerator in the process of fabrication. Methods of the radioisotope flaw detection in the process of exploitation take into consideration the influence of the radioisotope background, and ensure obtaining of the demanded by the rules of control, sensitivity. Methods of control of welded joints of the steamgenerator of nuclear power plants are based on the simultaneous examination of all joints with application of the shaped radiographic plate-holders. Special gamma-flaw-detection equipment is developed for control of the welded joints of the main branch-pipes. Design peculiarities are given of the installation for flaw detection. These installations are equipped with the system for emergency return of the radiation source into the storage position from the position for exposure. They have automatic exposure-meters for determination of the exposure time. Successfull exploitation of such installations in the Finland during assembling equipment for the nuclear reactor of the nuclear power plant ''Loviisa-1'' and in the USSR on the Novovoronezh nuclear power plant has shown possibility for detection of flaws having dimensions about 1% of the equipment used. For control of welded joints of pipes with pipe-plates at the steam generators, portable flaw-detectors are used. Sensitivity of these flaw-detectors towards detection of the wire standards has

  3. Transference of know-how for the fabrication of heavy components for nuclear power reactors

    International Nuclear Information System (INIS)

    Meier, F.

    1977-01-01

    1) Heavy components for nuclear power reactors. Reactor pressure vessels with total weight of 540 tons; steam generators: heat exchangers with U-type tube bundles, total weight 420 tons. 2) Choice of know-how recipient. Technical criteria, i.e. manufacturing facilities, existing quality assurance system, location of the workshops, possibilities for training, infrastructures. 3. Measures for transferring know-how to a newly established company. Planning and erection of the factory: organisational set up of the company; personnel selection and training; transfer of documentation; transfer of know-how that cannot be transferred in a written form. 4) Contracts for assuring the transfer of know-how. Stipulation of mutual rights and obligations of the know-how owner and receiver in individual contracts: engineering services contract, technical information contract, personnel training contract, license contract. (orig.) [de

  4. Fatigue analysis for analytically overloaded piping components and valves in nuclear power plants

    International Nuclear Information System (INIS)

    Charalambus, B.

    1992-01-01

    Lately, in connection with life extension aspects of power plants, an increasingly accurate determination of the lifetime of components in nuclear stations is being required. In order to assess reliably current fatigue levels in piping systems, variables such as pressure, temperature, and resultant force and moment transients as well as analytical methods which take into account the real operational history must be considered. This paper presents a method for analyzing the transient heat transfer between fluid and pipe wall in order to investigate effects which until now have been assumed conservatively to be caused by a sudden jump in temperature. Further, an example is given showing that the K e factor approach in current design codes for performing simplified elastic-plastic fatigue analyses is conservative. (orig.)

  5. Gasoline engine management systems and components

    CERN Document Server

    2015-01-01

    The call for environmentally compatible and economical vehicles necessitates immense efforts to develop innovative engine concepts. Technical concepts such as gasoline direct injection helped to save fuel up to 20 % and reduce CO2-emissions. Descriptions of the cylinder-charge control, fuel injection, ignition and catalytic emission-control systems provides comprehensive overview of today´s gasoline engines. This book also describes emission-control systems and explains the diagnostic systems. The publication provides information on engine-management-systems and emission-control regulations. Contents History of the automobile.- Basics of the gasoline engine.- Fuels.- Cylinder-charge control systems.- Gasoline injection systems over the years.- Fuel supply.- Manifold fuel injection.- Gasoline direct injection.- Operation of gasoline engines on natural gas.- Ignition systems over the years.- Inductive ignition systems.- Ignition coils.- Spark plugs.- Electronic control.- Sensors.- Electronic control unit.- Exh...

  6. Application of Component Technology to E-commerce System

    Institute of Scientific and Technical Information of China (English)

    ZHU Jianfeng

    2004-01-01

    At present E-commerce system tends to become more complex, and traditional system designing methods can not fufil the need of E-commerce system, thus requiring an effective methodas solution. With this concern, this paper introduces some concepts of component technology, then brings forward the new connotation and basic features of component technology through the analysis of its technological character. This paper finally discusses the application of component technology to E-commerce system.

  7. The development of component-based information systems

    CERN Document Server

    Cesare, Sergio de; Macredie, Robert

    2015-01-01

    This work provides a comprehensive overview of research and practical issues relating to component-based development information systems (CBIS). Spanning the organizational, developmental, and technical aspects of the subject, the original research included here provides fresh insights into successful CBIS technology and application. Part I covers component-based development methodologies and system architectures. Part II analyzes different aspects of managing component-based development. Part III investigates component-based development versus commercial off-the-shelf products (COTS), includi

  8. Advanced nuclear systems in comparison

    International Nuclear Information System (INIS)

    Brogli, R.; Foskolos, K.; Goetzmann, C.; Kroeger, W.; Stanculescu, A.; Wydler, P.

    1996-09-01

    This study aims at a comparison of future reactor concepts, paying particular attention to aspects of safety, of the fuel cycle, the economics, the experience-base and the state of development. Representative examples of typical development lines, that could possibly be 'of interest' within a time horizon of 50 years were selected for comparison. This can be divided into three phases: - Phase I includes the next 10 years and will be characterised mainly by evolutionary developments of light water reactors (LWR) of large size; representative: EPR, - Phase II: i.e. the time between 2005 and 2020 approximately, encompasses the forecasted doubling of today's world-wide installed nuclear capacity; along with evolutionary reactors, innovative systems like AP600, PIUS, MHTGR, EFR will emerge, - Phase III covers the time between 2020 and 2050 and is characterised by the issue of sufficient fissile material resources; novel fast reactor systems including hybrid systems can, thus, become available; representatives: IFR, EA, ITER (the latter being). The evaluated concepts foresee partly different fuel cycles. Fission reactors can be operated in principle on the basis of either a Uranium-Plutonium-cycle or a Thorium-Uranium-cycle, while combinations of these cycles among them or with other reactor concepts than proposed are possible. With today's nuclear park (comprising mainly LWRs), the world-wide plutonium excess increases annually by about 100 t. Besides strategies based on reprocessing like: - recycling in thermal and fast reactors with mixed oxide fuels, - plutonium 'burning' in reactors with novel fuels without uranium or in 'hybrid' systems, allowing a reduction of this excess, direct disposal of spent fuel elements including their plutonium content ('one-through') is being considered. (author) figs., tabs., 32 refs

  9. Integrated network for structural integrity monitoring of critical components in nuclear facilities, RIMIS

    International Nuclear Information System (INIS)

    Roth, Maria; Constantinescu, Dan Mihai; Brad, Sebastian; Ducu, Catalin; Malinovschi, Viorel

    2008-01-01

    The round table aims to join specialists working in the research area of the Romanian R and D Institutes and Universities involved in structural integrity assessment of materials, especially those working in the nuclear field, together with the representatives of the end user, the Cernavoda NPP. This scientific event will offer the opportunity to disseminate the theoretical, experimental and modelling activities, carried out to date, in the framework of the National Program 'Research of Excellence', Module I 2006-2008, managed by the National Authority for Scientific Research. Entitled 'Integrated Network for Structural Integrity Monitoring of Critical Components in Nuclear Facilities, RIMIS, the project has two main objectives: 1. - to elaborate a procedure applicable to the structural integrity assessment of critical components used in Romanian nuclear facilities (CANDU type Reactor, Hydrogen Isotopes Separation installations); 2. - to integrate the national networking into a similar one of European level, and to enhance the scientific significance of Romanian R and D organisations as well as to increase the contribution in solving major issues of the nuclear field. The topics of the round table will be focused on: 1. Development of a Structural Integrity Assessment Methodology applicable to the nuclear facilities components; 2. Experimental investigation methods and procedures; 3. Numeric simulation of nuclear components behaviour; 4. Further activities to finalize the assessment procedure. Also participations and contributions to sustain the activity in the European Network NULIFE, FP6 will be discussed. (authors)

  10. Recirculation system for nuclear reactors

    International Nuclear Information System (INIS)

    Braun, H. E.; Dollard, W. J.; Tower, S. N.

    1980-01-01

    A recirculation system for use in pressurized water nuclear reactors to increase the output temperature of the reactor coolant, thereby achieving a significant improvement in plant efficiency without exceeding current core design limits. A portion of the hot outlet coolant is recirculated to the inlets of the peripheral fuel assemblies which operate at relatively low power levels. The outlet temperature from these peripheral fuel assemblies is increased to a temperature above that of the average core outlet. The recirculation system uses external pumps and introduces the hot recirculation coolant to the free space between the core barrel and the core baffle, where it flows downward and inward to the inlets of the peripheral fuel assemblies. In the unlikely event of a loss of coolant accident, the recirculation system flow path through the free space and to the inlets of the fuel assemblies is utilized for the injection of emergency coolant to the lower vessel and core. During emergency coolant injection, the emergency coolant is prevented from bypassing the core through the recirculation system by check valves inserted into the recirculation system piping

  11. Conceptual benefits of passive nuclear power plants and their effect on component design

    International Nuclear Information System (INIS)

    DeVine, J.C. Jr.

    1996-01-01

    Today, nearly ten years after the advanced light water reactor (ALWR) Program was conceived by US utility leaders, and a decade and a half since a new nuclear power plant was ordered in the US, the ALWR passive plant is coming into its own. This design concept, a midsized simplified light water reactor, features extremely reliable passive systems for accident prevention and mitigation and combines proven experience with state-of-the-art engineering and human factors. It is now emerging as the front runner to become the next generation reactor in the US and perhaps around the world. Although simple and straightforward in concept, the passive plant is in many respects a significant departure from previous trends in reactor engineering. Successful implementation of this concept presents numerous challenges to the designers of passive plant systems and components. This paper provides a brief history of the ALWR program, it outlines the ALWR passive plant design objectives and principles, and it summarizes with examples their implications on component design. (orig.)

  12. Power system protection 1 principles and components

    CERN Document Server

    Association, Electricity Training

    1995-01-01

    The worldwide growth in demand for electricity has forced the pace of developments in electrical power system design to meet consumer needs for reliable, secure and cheap supplies. Power system protection, as a technology essential to high quality supply, is widely recognised as a specialism of growing and often critical importance, in which power system needs and technological progress have combined to result in rapid developments in policy and practice in recent years. In the United Kingdom, the need for appropriate training in power system protection was recognised in the early 1960s with t

  13. Nuclear reactor system for ABWR

    International Nuclear Information System (INIS)

    Miyano, Hiroshi; Kitagawa, Koji

    1997-01-01

    Various tests and measurements were performed during the pre-operational test run of Unit No. 6 of The Tokyo Electric Power Co., Inc.'s Kashiwazaki-Kariwa Nuclear Power Station, the first advanced boiling water reactor (ABWR) unit in the world, and the design and performance adequacy of the ABWR were confirmed. The realization of the ABWR in Japan took about 20 years. It was decided that technologies for the reactor internal pump (RIP) and the fine-motion control rod drive (FMCRD), which had been applied in Europe, would be incorporated in the ABWR aiming at simplification of its structure and operation. These main components were evaluated, modified and verified in consideration of the unique Japanese environment, such as seismic conditions, through a joint study program with Japanese utilities as well as an improvement and standardization program in cooperation with the government. In addition to incorporating RIP and FMCRD technologies, the ABWR also has improved features in terms of the design of the reactor pressure vessel and internals, as well as automated servicing equipment for the RIP, FMCRD, and primary containment vessel. (author)

  14. Fieldable Nuclear Material Identification System

    International Nuclear Information System (INIS)

    Radle, James E.; Archer, Daniel E.; Carter, Robert J.; Mullens, James Allen; Mihalczo, John T.; Britton, Charles L. Jr.; Lind, Randall F.; Wright, Michael C.

    2010-01-01

    The Fieldable Nuclear Material Identification System (FNMIS), funded by the NA-241 Office of Dismantlement and Transparency, provides information to determine the material attributes and identity of heavily shielded nuclear objects. This information will provide future treaty participants with verifiable information required by the treaty regime. The neutron interrogation technology uses a combination of information from induced fission neutron radiation and transmitted neutron imaging information to provide high confidence that the shielded item is consistent with the host's declaration. The combination of material identification information and the shape and configuration of the item are very difficult to spoof. When used at various points in the warhead dismantlement sequence, the information complimented by tags and seals can be used to track subassembly and piece part information as the disassembly occurs. The neutron transmission imaging has been developed during the last seven years and the signature analysis over the last several decades. The FNMIS is the culmination of the effort to put the technology in a usable configuration for potential treaty verification purposes.

  15. New generation nuclear microprobe systems

    International Nuclear Information System (INIS)

    Jamieson, David N.

    2001-01-01

    Over the past 20 years, the minimum probe size for nuclear microscopy has stayed around 1 μm with only a few groups reporting a sub-micron probe size around 0.5 μm. No breakthroughs in nuclear microprobe design have been forthcoming to produce dramatic improvements in spatial resolution. The difficulties of breaking the constraints that are preventing reduction of the probe size have been well recognised in the past. Over the past 5 years it has become clear that some of these constraints may not be as limiting as first thought. For example, chromatic aberration clearly is not as significant as implied from first-order ion optics calculations. This paper reviews the constraints in view of the increased understanding of the past 5 years and looks at several new approaches, presently being evaluated in Melbourne and elsewhere, on how to make progress. These approaches include modified RF ion sources for improved beam brightness and exploitation of relaxed constraints on some lens aberrations allowing the use of high demagnification probe forming lens systems

  16. Estimation of component failure rates for PSA on nuclear power plants 1982-1997

    International Nuclear Information System (INIS)

    Kirimoto, Yukihiro; Matsuzaki, Akihiro; Sasaki, Atsushi

    2001-01-01

    Probabilistic safety assessment (PSA) on nuclear power plants has been studied for many years by the Japanese industry. The PSA methodology has been improved so that PSAs for all commercial LWRs were performed and used to examine for accident management.On the other hand, most data of component failure rates in these PSAs were acquired from U.S. databases. Nuclear Information Center (NIC) of Central Research Institute of Electric Power Industry (CRIEPI) serves utilities by providing safety- , and reliability-related information on operation and maintenance of the nuclear power plants, and by evaluating the plant performance and incident trends. So, NIC started a research study on estimating the major component failure rates at the request of the utilities in 1988. As a result, we estimated the hourly-failure rates of 47 component types and the demand-failure rates of 15 component types. The set of domestic component reliability data from 1982 to 1991 for 34 LWRs has been evaluated by a group of PSA experts in Japan at the Nuclear Safety Research Association (NSRA) in 1995 and 1996, and the evaluation report was issued in March 1997. This document describes the revised component failure rate calculated by our re-estimation on 49 Japanese LWRs from 1982 to 1997. (author)

  17. Classification of transportation packaging and dry spent fuel storage system components according to importance to safety

    International Nuclear Information System (INIS)

    Tyacke, M.J.; McConnell, J.W. Jr.; Ayers, A.L. Jr.; O'Connor, S.C.; Jankovich, J.P.

    1996-01-01

    The Idaho National Engineering Laboratory prepared a technical report for the Office of Nuclear Material Safety and Safeguards of the US Nuclear Regulatory Commission, entitled Classification of Transportation Packaging and Dry Spent Fuel Storage System Components According to Importance to Safety, NUREG/CR-6407. This paper provides the results of that report. It also presents the graded approach for classification of components used in transportation packagings and dry spent fuel storage systems. This approach provides a method for identifying the classification of components according to importance to safety within transportation packagings and dry spent fuel storage systems. Record retention requirements are discussed to identify the documentation necessary to validate that the individual components were fabricated in accordance with their assigned classification. A review of the existing regulations pertaining to transportation packagings and dry storage systems was performed to identify current requirements. The general types of transportation packagings and dry storage systems are identified. The methodology used in this paper is based on Regulatory Guide 7.10, Establishing Quality Assurance Programs for Packaging Used in the Transport of Radioactive Material. This paper also includes a list of generic components for each of the general types of transportation packagings and spent fuel storage systems, with a classification category assigned to each component. Several examples concerning the safety importance of components are presented

  18. DIGITAL SUBSTATION COMPONENT SYSTEM "SMART GRID"

    Directory of Open Access Journals (Sweden)

    V.I. Vasilchenko

    2014-12-01

    Full Text Available New production technologies of modern control systems have moved from the stage of research and experimentation into the stage of practical use. Modern communication standards for the exchange of information are developed and introduced. Digital devices, protectors and automation are widely used. There has been substantial development of hardware and software of control systems.

  19. Novel Devices and Components for THz Systems

    Science.gov (United States)

    2014-04-25

    effectively . The cage rod system is again 30 mm...often used to reflect RF radiation, or create Faraday cages , this concept can be applied to the FPI mirrors. Wire meshes will simulate solid conductors...Single peice construction should resist leaning more effectively . The cage rod system is again 30 mm. 175 wire-mesh-mirrors. Fixing the rocking

  20. Reactor component inventory system at FFTF

    International Nuclear Information System (INIS)

    Ordonez, C.R.; Redekopp, R.D.; Reed, E.A.

    1985-02-01

    A reliable inventory control system was developed at the Fast Flux Test Facility (FFTF) to keep track of the occupancy of 900 refueling facility locations, to compile historical data on the movement of each reactor assembly, and to simulate assembly moves. The simulate capability is valuable because it allows verification of documents before they are issued for use in the plant, and eliminates the possibility of planning illegal or impossible moves. The system is installed on a UNIVAC 1100 computer and is maintained using a data base management system by Sperry Univac called MAPPER

  1. Expert system for accelerator single-freedom nonlinear components

    International Nuclear Information System (INIS)

    Wang Sheng; Xie Xi; Liu Chunliang

    1995-01-01

    An expert system by Arity Prolog is developed for accelerator single-freedom nonlinear components. It automatically yields any order approximate analytical solutions for various accelerator single-freedom nonlinear components. As an example, the eighth order approximate analytical solution is derived by this expert system for a general accelerator single-freedom nonlinear component, showing that the design of the expert system is successful

  2. Remote system for counting of nuclear pulses

    International Nuclear Information System (INIS)

    Nieves V, J.A.; Garcia H, J.M.; Aguilar B, M.A.

    1999-01-01

    In this work, it is describe technically the remote system for counting of nuclear pulses, an integral system of the project radiological monitoring in a petroleum distillation tower. The system acquires the counting of incident nuclear particles in a nuclear detector which process this information and send it in serial form, using the RS-485 toward a remote receiver, which can be a Personal computer or any other device capable to interpret the communication protocol. (Author)

  3. European nuclear data studies for fast systems

    International Nuclear Information System (INIS)

    Rullhusen, P.; Hambsch, F.-J.; Mondelaers, W.; Plompen, A.J.M.; Schillebeeckx, P.

    2010-01-01

    Nuclear data needs for fast systems are highlighted and the following projects are described: Joint European research projects: MUSE Experiments for Sub-critical Neutronics Validation; High- and Intermediate Energy Nuclear Data for ADS (HINDAS); and the Time-Of-Flight facility for Nuclear Data Measurements for ADS (n T OF N D A DS); European Research Programme for the Transmutation of High Level Nuclear Waste in an Accelerator Driven System (EUROTRANS-NUDATRA); and CANDIDE; Programmes for transnational access to experimental facilities in Europe: European Facilities for Nuclear Data Measurements (EFNUDAT); Neutron Data Measurements at IRMM (NUDAME); European facility for innovative reactor and transmutation neutron data (EUFRAT) (P.A.)

  4. Integrated nuclear and radiation protection systems

    International Nuclear Information System (INIS)

    Oprea, I.; Oprea, M.; Stoica, V.; Cerga, V.; Pirvu, V.; Badea, E.

    1993-01-01

    A multifunctional radiation monitoring equipment, flexible and capable to meet virtually environmental radiation monitoring, activity measurement and computational requirements, for nuclear laboratories has been designed. It can be used as a radiation protection system, for radionuclide measurement in isotope laboratories, nuclear technology, health physics and nuclear medicine, nuclear power stations and nuclear industry. The equipment is able to measure, transmit and record gamma dose rate and isotope activities. Other parameters and functions are optionally available, such as: self-contained alarm level, system self-test, dose integrator, syringe volume calculation for a given dose corrected for decay, calibration factor, 99 Mo assays performing and background subtraction

  5. EDF ageing management program of nuclear components: a safety and economical issue

    International Nuclear Information System (INIS)

    Faidy, C.

    2005-01-01

    Ageing management of Nuclear Power Plants is an essential issue for utilities, in term of safety and availability and corresponding economical consequences. Practically all nuclear countries have developed a systematic program to deal with ageing of components on their plants. This paper presents the ageing management program developed by EDF and that are compared with different other approaches in other countries (IAEA guidelines and GALL report). The paper presents a general overview of the programs, the major results, recommendations and conclusions. (author)

  6. Artificial heart system thermal insulation component development

    International Nuclear Information System (INIS)

    Svedberg, R.C.; Buckman, R.W. Jr.

    1975-01-01

    A concentric cup vacuum multifoil insulation system has been selected by virtue of its size, weight, and thermal performance to insulate the hot radioisotope portion of the thermal converter of an artificial implantable heart system. A factor of 2 improvement in thermal performance, based on the heat loss per number of foil layers (minimum system weight and volume) has been realized over conventional spiral wrapped multifoil vacuum insulation. This improvement is the result of the concentric cup construction to maintain a uniform interfoil spacing and the elimination of corner heat losses. Based on external insulation system dimensions (surface area in contact with host body), heat losses of 0.019 W/ cm 2 at 1140 0 K (1600 0 F) and 0.006 W/cm 2 at 920 0 K (1200 0 F) have been achieved. Factors which influence thermal performance of the nickel foil concentric cup insulation system include the number of cups, configuration and method of application of zirconia (ZrO 2 ) spacer material, system pressure, emittance of the cups, and operating temperature

  7. Nuclear power project management information system

    International Nuclear Information System (INIS)

    Zou Lailong; Zhang Peng; Xiao Ziyan; Chun Zengjun; Huang Futong

    2001-01-01

    Project Management Information System is an important infrastructure facility for the construction and operation of Nuclear Power Station. Based on the practice of Lingao nuclear power project management information system (NPMIS), the author describes the NPMIS design goals, system architecture and software functionality, points out the outline issues during the development and deployment of NPMIS

  8. Results of an aging-related failure survey of light water safety systems and components

    International Nuclear Information System (INIS)

    Meale, B.M.; Satterwhite, D.G.; MacDonald, P.E.

    1988-01-01

    The collection and evaluation of operating experience data are necessary in determining the effects of aging on the safety of operating nuclear plants. This paper presents the final results of a two-year research effort evaluating aging impacts on components in light water reactor systems. This research was performed as a part of the Nuclear Plant Aging Research program, sponsored by the US Nuclear Regulatory Commission. Two unique types of data analyses were performed. In the first, an aging-survey study, aging-related failure data for fifteen light water reactor systems were obtained from the Nuclear Plant Reliability Data System (NPRDS). These included safety, support, and power conversion systems. A computerized sort of these records classified each record into one of five generic categories, based on the utility's choice of the failure's NPRDS cause category. Systems and components within the systems that were most affected by aging were identified. In the second analysis, information on aging-related reported causes of failures was evaluated for component failures reported to NPRDS for auxiliary feedwater, high pressure injection, service water, and Class 1E electrical power distribution systems. 3 refs., 13 figs., 4 tabs

  9. Nuclear data evaluation method and evaluation system

    International Nuclear Information System (INIS)

    Liu Tingjin

    1995-01-01

    The evaluation methods and Nuclear Data Evaluation System have been developed in China. A new version of the system has been established on Micro-VAX2 computer, which is supported by IAEA under the technology assistance program. The flow chart of Chinese Nuclear Data Evaluation System is shown out. For last ten years, the main efforts have been put on the double differential cross section, covariance data and evaluated data library validation. The developed evaluation method and Chinese Nuclear Data Evaluation System have been widely used at CNDC and in Chinese Nuclear Data Network for CENDL. (1 tab., 15 figs.)

  10. Automated accounting systems for nuclear materials

    International Nuclear Information System (INIS)

    Erkkila, B.

    1994-01-01

    History of the development of nuclear materials accounting systems in USA and their purposes are considered. Many present accounting systems are based on mainframe computers with multiple terminal access. Problems of future improvement accounting systems are discussed

  11. RCC-M: Design and construction rules for mechanical components of PWR nuclear islands

    International Nuclear Information System (INIS)

    2017-01-01

    AFCEN's RCC-M code concerns the mechanical components designed and manufactured for pressurized water reactors (PWR). It applies to pressure equipment in nuclear islands in safety classes 1, 2 and 3, and certain non-pressure components, such as vessel internals, supporting structures for safety class components, storage tanks and containment penetrations. RCC-M covers the following technical subjects: sizing and design, choice of materials and procurement. Fabrication and control, including: associated qualification requirements (procedures, welders and operators, etc.), control methods to be implemented, acceptance criteria for detected defects, documentation associated with the different activities covered, and quality assurance. The design, manufacture and inspection rules defined in RCC-M leverage the results of the research and development work pioneered in France, Europe and worldwide, and which have been successfully used by industry to design and build PWR nuclear islands. AFCEN's rules incorporate the resulting feedback. Use: France's last 16 nuclear units (P'4 and N4); 4 CP1 reactors in South Africa (2) and Korea (2); 44 M310 (4), CPR-1000 (28), CPR-600 (6), HPR-1000 (4) and EPR (2) reactors in service or undergoing construction in China; 4 EPR reactors in Europe: Finland (1), France (1) and UK (2). Content: Section I - nuclear island components, subsection 'A': general rules, subsection 'B': class 1 components, subsection 'C': class 2 components, subsection 'D': class 3 components, subsection 'E': small components, subsection 'G': core support structures, subsection 'H': supports, subsection 'J': low pressure or atmospheric storage tanks, subsection 'P': containment penetration, subsection 'Q': qualification of active mechanical components, subsection 'Z': technical appendices; section II - materials; section III - examination

  12. Establishment of nuclear data evaluation system (I)

    International Nuclear Information System (INIS)

    Chang, Jong Hwa; Lee, Chang Kun; Kim, Jeong Do; Kim, Young Sik; Kim, Young Jin; Kim, Hyung Guk; Kil, Chung Sup; Kim, Kang Suk

    1994-08-01

    Nuclear data is fundamental data for development of new type of nuclear, upgrade of nuclear fuel, treatment of radwaste, research on fusion reactor, radioisotope usage, and nuclear medical therapy. Nuclear data is produced with experiments. However rack of experimental data for thousands of nuclides and various reaction types makes it essential to do statistical evaluation and theoretical interpolation. This study is intended to join international cooperation after establishing domestic basis for nuclear data evaluation work. This project is the first year of five year plan to do followings: 1) Establishment of database system to collect experimental data, 2) Setup of computer assistance system for evaluation work, 3) Verification of established system by test evaluation of selected nuclide reaction. The system has a collection of mass data of nuclides, computer codes for test evaluation of total cross section of 0-16 and collection of EXFOR format data for 0-16. This system will be improved continuously on next years. (Author)

  13. 11-th International conference Nuclear power safety and nuclear education - 2009. Abstracts. Part 1. Session: Safety of nuclear technology; Innovative nuclear systems and fuel cycle; Nuclear knowledge management

    International Nuclear Information System (INIS)

    2009-01-01

    The book includes abstracts of the 11-th International conference Nuclear power safety and nuclear education - 2009 (29 Sep - 2 Oct, 2009, Obninsk). Problems of safety of nuclear technology are discussed, innovative nuclear systems and fuel cycles are treated. Abstracts on professional education for nuclear power and industry are presented. Nuclear knowledge management are discussed

  14. Planned reliability in the transport and installation of large nuclear components

    International Nuclear Information System (INIS)

    Bieler, L.

    1988-01-01

    The transport and installation of heavy and bulky large components require detailed planning of all jobs and activities, trained and experienced personnel and corresponding technical equipment for reliable and quality-assured implementation. The correct approach to the planning and implementation of such transports and installations has been confirmed by years of successful performance of these jobs e.g. in reactor pressure vessels and steam generators for nuclear power plants. Large components for nuclear power plants are truly extreme examples but will be all the better suited for demonstrating the problems inherent in transport and installation. (orig.) [de

  15. Catalog of components for electric and hybrid vehicle propulsion systems

    Science.gov (United States)

    Eissler, H. C.

    1981-01-01

    This catalog of commercially available electric and hybrid vehicle propulsion system components is intended for designers and builders of these vehicles and contains 50 categories of components. These categories include those components used between the battery terminals and the output axle hub, as well as some auxiliary equipment. An index of the components and a listing of the suppliers and their addresses and phone numbers are included.

  16. Imprecise system reliability and component importance based on survival signature

    International Nuclear Information System (INIS)

    Feng, Geng; Patelli, Edoardo; Beer, Michael; Coolen, Frank P.A.

    2016-01-01

    The concept of the survival signature has recently attracted increasing attention for performing reliability analysis on systems with multiple types of components. It opens a new pathway for a structured approach with high computational efficiency based on a complete probabilistic description of the system. In practical applications, however, some of the parameters of the system might not be defined completely due to limited data, which implies the need to take imprecisions of component specifications into account. This paper presents a methodology to include explicitly the imprecision, which leads to upper and lower bounds of the survival function of the system. In addition, the approach introduces novel and efficient component importance measures. By implementing relative importance index of each component without or with imprecision, the most critical component in the system can be identified depending on the service time of the system. Simulation method based on survival signature is introduced to deal with imprecision within components, which is precise and efficient. Numerical example is presented to show the applicability of the approach for systems. - Highlights: • Survival signature is a novel way for system reliability and component importance • High computational efficiency based on a complete description of system. • Include explicitly the imprecision, which leads to bounds of the survival function. • A novel relative importance index is proposed as importance measure. • Allows to identify critical components depending on the service time of the system.

  17. A methodology for on-line fatigue life monitoring of Indian nuclear power plant components

    International Nuclear Information System (INIS)

    Mukhopadhyay, N.K.; Dutta, B.K.; Kushawaha, H.S.

    1992-01-01

    Fatigue is one of the most important aging effects of nuclear power plant components. Information about accumulation of fatigue helps in assessing structural degradation of the components. This assists in-service inspection and maintenance and may also support future life extension program of a plant. In the present report a methodology is being proposed for monitoring on line fatigue life of nuclear power plant components using available plant instrumentations. Major factors affecting fatigue life of a nuclear power plant components are the fluctuations of temperature, pressure and flow rate. Green's function technique is used in on line fatigue monitoring as computation time is much less than finite element method. A code has been developed which computes temperature and stress Green's functions in 2-D and axisymmetric structure by finite element method due to unit change in various fluid parameters. A post processor has also been developed which computes the temperature and stress responses using corresponding Green's functions and actual fluctuation in fluid parameters. In this post processor, the multiple site problem is solved by superimposing single site Green's function technique. It is also shown that Green's function technique is best suited for on line fatigue life monitoring of nuclear power plant components. (author). 6 refs., 43 figs

  18. Recycle and reuse of materials and components from waste streams of nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    2000-01-01

    All nuclear fuel cycle processes utilize a wide range of equipment and materials to produce the final products they are designed for. However, as at any other industrial facility, during operation of the nuclear fuel cycle facilities, apart from the main products some byproducts, spent materials and waste are generated. A lot of these materials, byproducts or some components of waste have a potential value and may be recycled within the original process or reused outside either directly or after appropriate treatment. The issue of recycle and reuse of valuable material is important for all industries including the nuclear fuel cycle. The level of different materials involvement and opportunities for their recycle and reuse in nuclear industry are different at different stages of nuclear fuel cycle activity, generally increasing from the front end to the back end processes and decommissioning. Minimization of waste arisings and the practice of recycle and reuse can improve process economics and can minimize the potential environmental impact. Recognizing the importance of this subject, the International Atomic Energy Agency initiated the preparation of this report aiming to review and summarize the information on the existing recycling and reuse practice for both radioactive and non-radioactive components of waste streams at nuclear fuel cycle facilities. This report analyses the existing options, approaches and developments in recycle and reuse in nuclear industry

  19. Advanced Reconnaissance System Component Reliability Study

    Science.gov (United States)

    1956-07-31

    dielectrics. Gaseous dielectrics such as sulphur hexafluoride and ’ fluorocarbons at two to three atmospheres. Fluorinated liquid dielectrics. 3) The...limits. (2) determine compatibility with varnish treatments, (3) compatibility in a complete insulation system. Mechanical and thermal limits of...of a varnish to have good • adhersion, provide an element of flexibility and be chemically compatible with’the wire it is impregnating.. Factors of

  20. Algorithmic fault tree construction by component-based system modeling

    International Nuclear Information System (INIS)

    Majdara, Aref; Wakabayashi, Toshio

    2008-01-01

    Computer-aided fault tree generation can be easier, faster and less vulnerable to errors than the conventional manual fault tree construction. In this paper, a new approach for algorithmic fault tree generation is presented. The method mainly consists of a component-based system modeling procedure an a trace-back algorithm for fault tree synthesis. Components, as the building blocks of systems, are modeled using function tables and state transition tables. The proposed method can be used for a wide range of systems with various kinds of components, if an inclusive component database is developed. (author)

  1. The information system of the Spanish nuclear power plants: DACNE

    International Nuclear Information System (INIS)

    Diez M, Jose E.

    1995-01-01

    DACNE information system is a set of two databases aimed at collecting and retrieving information related to operation of the Spanish nuclear power plants. The first one is the Operation Events Database and the second is the Reliability Components Database. The system was designed and developed by UNESA and came into operation early in 1989. A significant amount of data is currently stored in the system available for information exchange and for supporting operational programs. (author). 6 figs., 4 tabs

  2. Nuclear reactor component populations, reliability data bases, and their relationship to failure rate estimation and uncertainty analysis

    International Nuclear Information System (INIS)

    Martz, H.F.; Beckman, R.J.

    1981-12-01

    Probabilistic risk analyses are used to assess the risks inherent in the operation of existing and proposed nuclear power reactors. In performing such risk analyses the failure rates of various components which are used in a variety of reactor systems must be estimated. These failure rate estimates serve as input to fault trees and event trees used in the analyses. Component failure rate estimation is often based on relevant field failure data from different reliability data sources such as LERs, NPRDS, and the In-Plant Data Program. Various statistical data analysis and estimation methods have been proposed over the years to provide the required estimates of the component failure rates. This report discusses the basis and extent to which statistical methods can be used to obtain component failure rate estimates. The report is expository in nature and focuses on the general philosophical basis for such statistical methods. Various terms and concepts are defined and illustrated by means of numerous simple examples

  3. System Risk Balancing Profiles: Software Component

    Science.gov (United States)

    Kelly, John C.; Sigal, Burton C.; Gindorf, Tom

    2000-01-01

    The Software QA / V&V guide will be reviewed and updated based on feedback from NASA organizations and others with a vested interest in this area. Hardware, EEE Parts, Reliability, and Systems Safety are a sample of the future guides that will be developed. Cost Estimates, Lessons Learned, Probability of Failure and PACTS (Prevention, Avoidance, Control or Test) are needed to provide a more complete risk management strategy. This approach to risk management is designed to help balance the resources and program content for risk reduction for NASA's changing environment.

  4. Multilayer electronic component systems and methods of manufacture

    Science.gov (United States)

    Thompson, Dane (Inventor); Wang, Guoan (Inventor); Kingsley, Nickolas D. (Inventor); Papapolymerou, Ioannis (Inventor); Tentzeris, Emmanouil M. (Inventor); Bairavasubramanian, Ramanan (Inventor); DeJean, Gerald (Inventor); Li, RongLin (Inventor)

    2010-01-01

    Multilayer electronic component systems and methods of manufacture are provided. In this regard, an exemplary system comprises a first layer of liquid crystal polymer (LCP), first electronic components supported by the first layer, and a second layer of LCP. The first layer is attached to the second layer by thermal bonds. Additionally, at least a portion of the first electronic components are located between the first layer and the second layer.

  5. Nuclear energy in Canada: the CANDU system

    International Nuclear Information System (INIS)

    Robertson, J.A.L.

    1979-10-01

    Nuclear electricity in Canada is generated by CANDU nuclear power stations. The CANDU reactor - a unique Canadian design - is fuelled by natural uranium and moderated by heavy water. The system has consistently outperformed other comparable nuclear power systems in the western world, and has an outstanding record of reliability, safety and economy. As a source of energy it provides the opportunity for decreasing our dependence on dwindling supplies of conventional fossil fuels. (auth)

  6. 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)

  7. Nuclear Space Power Systems Materials Requirements

    International Nuclear Information System (INIS)

    Buckman, R.W. Jr.

    2004-01-01

    High specific energy is required for space nuclear power systems. This generally means high operating temperatures and the only alloy class of materials available for construction of such systems are the refractory metals niobium, tantalum, molybdenum and tungsten. The refractory metals in the past have been the construction materials selected for nuclear space power systems. The objective of this paper will be to review the past history and requirements for space nuclear power systems from the early 1960's through the SP-100 program. Also presented will be the past and present status of refractory metal alloy technology and what will be needed to support the next advanced nuclear space power system. The next generation of advanced nuclear space power systems can benefit from the review of this past experience. Because of a decline in the refractory metal industry in the United States, ready availability of specific refractory metal alloys is limited

  8. Component-Based Approach in Learning Management System Development

    Science.gov (United States)

    Zaitseva, Larisa; Bule, Jekaterina; Makarov, Sergey

    2013-01-01

    The paper describes component-based approach (CBA) for learning management system development. Learning object as components of e-learning courses and their metadata is considered. The architecture of learning management system based on CBA being developed in Riga Technical University, namely its architecture, elements and possibilities are…

  9. International Nuclear Information System in Malaysia

    International Nuclear Information System (INIS)

    Samsurdin Ahamad

    1984-01-01

    Practice of the International Nuclear Information System (INIS) in Malaysia is reviewed. The Nuclear Energy Unit, a participating representative of Malaysia, holds the responsibilities of disseminating information through this system. Its available services relevant to the aims of INIS are discussed

  10. Containments for consolidated nuclear steam systems

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1978-01-01

    A containment system for a consolidated nuclear steam system incorporating a nuclear core, steam generator and reactor coolant pumps within a single pressure vessel is described which is designed to provide radiation shielding and pressure suppression. Design details, including those for the dry well and wet well of the containment, are given. (UK)

  11. Feature-based component model for design of embedded systems

    Science.gov (United States)

    Zha, Xuan Fang; Sriram, Ram D.

    2004-11-01

    An embedded system is a hybrid of hardware and software, which combines software's flexibility and hardware real-time performance. Embedded systems can be considered as assemblies of hardware and software components. An Open Embedded System Model (OESM) is currently being developed at NIST to provide a standard representation and exchange protocol for embedded systems and system-level design, simulation, and testing information. This paper proposes an approach to representing an embedded system feature-based model in OESM, i.e., Open Embedded System Feature Model (OESFM), addressing models of embedded system artifacts, embedded system components, embedded system features, and embedded system configuration/assembly. The approach provides an object-oriented UML (Unified Modeling Language) representation for the embedded system feature model and defines an extension to the NIST Core Product Model. The model provides a feature-based component framework allowing the designer to develop a virtual embedded system prototype through assembling virtual components. The framework not only provides a formal precise model of the embedded system prototype but also offers the possibility of designing variation of prototypes whose members are derived by changing certain virtual components with different features. A case study example is discussed to illustrate the embedded system model.

  12. Croatian National System of Nuclear Materials Control

    International Nuclear Information System (INIS)

    Biscan, R.

    1998-01-01

    In the process of economic and technological development of Croatia by using or introducing nuclear power or in the case of international co-operation in the field of peaceful nuclear activities, including international exchange of nuclear material, Croatia should establish and implement National System of Nuclear Materials Control. Croatian National System of accounting for and control of all nuclear material will be subjected to safeguards under requirements of Agreement and Additional Protocol between the Republic of Croatia and the International Atomic Energy Agency (IAEA) for the Application of Safeguards in Connection with the Treaty on the Non-Proliferation of Nuclear Weapons (NPT). The decision by NPT parties at the 1995 NPT Review and Extension Conference to endorse the Fullscope IAEA Safeguards Standard (FSS) as a necessary precondition of nuclear supply means that states are obliged to ensure that the recipient country has a FSS agreement in place before any nuclear transfer can take place (Ref. 1). The FSS standard of nuclear supply is a central element of the Nuclear Suppliers Group (NSG) Guidelines which the NSG adopted in 1992 and should be applied to members and non-members of the NSG. The FSS standard of nuclear supply in general allows for NPT parties or countries which have undertaken the same obligations through other treaty arrangements, to receive favourable treatment in nuclear supply arrangements. However, the Iraqi experience demonstrate that trade in nuclear and dual-use items, if not properly monitored, can contribute to a nuclear weapons program in countries acting contrary to their non-proliferation obligation. Multilateral nuclear export control mechanisms, including the FSS supply standard, provide the basis for co-ordination and standardisation of export control measures. (author)

  13. Experience with Nuclear Medicine Information System

    Directory of Open Access Journals (Sweden)

    Bilge Volkan-Salanci

    2012-12-01

    Full Text Available Objective: Radiology information system (RIS is basically evolved for the need of radiologists and ignores the vital steps needed for a proper work flow of Nuclear Medicine Department. Moreover, CT/MRI oriented classical PACS systems are far from satisfying Nuclear Physicians like storing dynamic data for reprocessing and quantitative analysis of colored images. Our purpose was to develop a workflow based Nuclear Medicine Information System (NMIS that fulfills the needs of Nuclear Medicine Department and its integration to hospital PACS system. Material and Methods: Workflow in NMIS uses HL7 (health level seven and steps include, patient scheduling and retrieving information from HIS (hospital information system, radiopharmacy, acquisition, digital reporting and approval of the reports using Nuclear Medicine specific diagnostic codes. Images and dynamic data from cameras of are sent to and retrieved from PACS system (Corttex© for reprocessing and quantitative analysis. Results: NMIS has additional functions to the RIS such as radiopharmaceutical management program which includes stock recording of both radioactive and non-radioactive substances, calculation of the radiopharmaceutical dose for individual patient according to body weight and maximum permissible activity, and calculation of radioactivity left per unit volume for each radionuclide according their half lives. Patient scheduling and gamma camera patient work list settings were arranged according to specific Nuclear Medicine procedures. Nuclear Medicine images and reports can be retrieved and viewed from HIS. Conclusion: NMIS provides functionality to standard RIS and PACS system according to the needs of Nuclear Medicine. (MIRT 2012;21:97-102

  14. Evaluation of the Waste Isolation Pilot Plant classification of systems, structures and components

    International Nuclear Information System (INIS)

    1985-07-01

    A review of the classification system for systems, structures, and components at the Waste Isolation Pilot Plant (WIPP) was performed using the WIPP Safety Analysis Report (SAR) and Bechtel document D-76-D-03 as primary source documents. The regulations of the US Nuclear Regulatory Commission (NRC) covering ''Disposal of High level Radioactive Wastes in Geologic Repositories,'' 10 CFR 60, and the regulations relevant to nuclear power plant siting and construction (10 CFR 50, 51, 100) were used as standards to evaluate the WIPP design classification system, although it is recognized that the US Department of Energy (DOE) is not required to comply with these NRC regulations in the design and construction of WIPP. The DOE General Design Criteria Manual (DOE Order 6430.1) and the Safety Analysis and Review System for AL Operation document (AL 54f81.1A) were reviewed in part. This report includes a discussion of the historical basis for nuclear power plant requirements, a review of WIPP and nuclear power plant classification bases, and a comparison of the codes and standards applicable to each quality level. Observations made during the review of the WIPP SAR are noted in the text of this reoport. The conclusions reached by this review are: WIPP classification methodology is comparable to corresponding nuclear power procedures. The classification levels assigned to WIPP systems are qualitatively the same as those assigned to nuclear power plant systems

  15. The safety related aspects of pressure components in nuclear power plants

    International Nuclear Information System (INIS)

    Lindackers, K.H.

    1979-01-01

    Over the last two years the safety philosophy for nuclear power plants in the Federal Republic of Germany has changed considerably, as everyone working in the field perceives. The original and appropriate philosophy of risk minimalisation through graduated safety barriers has been more and more replaced by the utopian goal of total prevention of any damage. The reasons for this development are discussed briefly especially regarding pressure components. The very numerous pressure components of a nuclear power station are not all of equal importance with respect to safety. Although considerable efforts have been made, it has not been possible, to date, to achieve an agreement between operators, manufacturers, licensing authorities, independent experts, and other specialists about the safety related classification of the manifold pressure bearing parts in nuclear power stations. The background of this extremely regrettable situation is explained. In the last part of the paper the author suggests a simple and clear safety philosophy for pressure components in nuclear power stations. This philosophy is orientated both on Safety Regulations of the Radiation Protection Decree ('Strahlenschutzverordnung') of the 13th October 1976 and on the Safety Criteria for Nuclear Power Stations from 21st October 1977. Only a simple, clear framework can make a contribution to the further improvement of the already exceptional safety of nuclear facilities and to the removal of obstacles in the licensing procedure which, taken as a whole, tie up skilled personnel to a senseless degree, involve considerable financial expenditure, and have no relevance for the safety of nuclear power plants. (orig.) [de

  16. A philosophy for space nuclear systems safety

    International Nuclear Information System (INIS)

    Marshall, A.C.

    1992-01-01

    The unique requirements and contraints of space nuclear systems require careful consideration in the development of a safety policy. The Nuclear Safety Policy Working Group (NSPWG) for the Space Exploration Initiative has proposed a hierarchical approach with safety policy at the top of the hierarchy. This policy allows safety requirements to be tailored to specific applications while still providing reassurance to regulators and the general public that the necessary measures have been taken to assure safe application of space nuclear systems. The safety policy used by the NSPWG is recommended for all space nuclear programs and missions

  17. Nuclear Reactor Component Code CUPID-I: Numerical Scheme and Preliminary Assessment Results

    International Nuclear Information System (INIS)

    Cho, Hyoung Kyu; Jeong, Jae Jun; Park, Ik Kyu; Kim, Jong Tae; Yoon, Han Young

    2007-12-01

    A component scale thermal hydraulic analysis code, CUPID (Component Unstructured Program for Interfacial Dynamics), is being developed for the analysis of components of a nuclear reactor, such as reactor vessel, steam generator, containment, etc. It adopted three-dimensional, transient, two phase and three-field model. In order to develop the numerical schemes for the three-field model, various numerical schemes have been examined including the SMAC, semi-implicit ICE, SIMPLE, Row Scheme and so on. Among them, the ICE scheme for the three-field model was presented in the present report. The CUPID code is utilizing unstructured mesh for the simulation of complicated geometries of the nuclear reactor components. The conventional ICE scheme that was applied to RELAP5 and COBRA-TF, therefore, were modified for the application to the unstructured mesh. Preliminary calculations for the unstructured semi-implicit ICE scheme have been conducted for a verification of the numerical method from a qualitative point of view. The preliminary calculation results showed that the present numerical scheme is robust and efficient for the prediction of phase changes and flow transitions due to a boiling and a flashing. These calculation results also showed the strong coupling between the pressure and void fraction changes. Thus, it is believed that the semi-implicit ICE scheme can be utilized for transient two-phase flows in a component of a nuclear reactor

  18. Nuclear Reactor Component Code CUPID-I: Numerical Scheme and Preliminary Assessment Results

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Hyoung Kyu; Jeong, Jae Jun; Park, Ik Kyu; Kim, Jong Tae; Yoon, Han Young

    2007-12-15

    A component scale thermal hydraulic analysis code, CUPID (Component Unstructured Program for Interfacial Dynamics), is being developed for the analysis of components of a nuclear reactor, such as reactor vessel, steam generator, containment, etc. It adopted three-dimensional, transient, two phase and three-field model. In order to develop the numerical schemes for the three-field model, various numerical schemes have been examined including the SMAC, semi-implicit ICE, SIMPLE, Row Scheme and so on. Among them, the ICE scheme for the three-field model was presented in the present report. The CUPID code is utilizing unstructured mesh for the simulation of complicated geometries of the nuclear reactor components. The conventional ICE scheme that was applied to RELAP5 and COBRA-TF, therefore, were modified for the application to the unstructured mesh. Preliminary calculations for the unstructured semi-implicit ICE scheme have been conducted for a verification of the numerical method from a qualitative point of view. The preliminary calculation results showed that the present numerical scheme is robust and efficient for the prediction of phase changes and flow transitions due to a boiling and a flashing. These calculation results also showed the strong coupling between the pressure and void fraction changes. Thus, it is believed that the semi-implicit ICE scheme can be utilized for transient two-phase flows in a component of a nuclear reactor.

  19. The structural aging assessment program: ranking methodology for CANDU nuclear generating station concrete components

    International Nuclear Information System (INIS)

    Philipose, K.E.; Muhkerjee, P.K.; McColm, E.J.

    1997-01-01

    Most of the major structural components in CANDU nuclear generating stations are constructed of reinforced concrete. Although passive in nature, these structures perform many critical safety functions in the operation of each facility. Aging can affect the structural capacity and integrity of structures. The reduction in capacity due to aging is not addressed in design codes. Thus a program is warranted to monitor the aging of safety-related CANDU plant structures and to prioritize those that require maintenance and repairs. Prioritization of monitoring efforts is best accomplished by focusing on those structures judged to be the most critical to plant performance and safety. The safety significance of each sub-element and its degradation with time can be evaluated using a numerical rating system. This will simplify the utility's efforts, thereby saving maintenance costs while providing a higher degree of assurance that performance is maintained. This paper describes the development of a rating system (ranking procedure) as part of the Plant Life Management of CANDU generating station concrete structures and illustrates its application to an operating plant. (author)

  20. Testing and operation of nuclear air-cleaning systems in Qinshan NPP

    International Nuclear Information System (INIS)

    Yang Lin

    1993-01-01

    The components of nuclear air-cleaning system, system function, operational mode and the performance of cleaning components in Qinshan Nuclear Power Plant are described. The items, purpose, methods and results of in-place testing after the installation are also described in detail. The in-place testing verifies that nuclear air-cleaning systems in Qinshan Nuclear Power Plant are reliable and high effective. It also describes the points of the operational management. It is shown that the good operational management is the key which developed prescription function of nuclear air-cleaning systems. At present, Qinshan Nuclear Power Plant will be in full power, the normal operation of the system is satisfied with the demand of safe operation in Qinshan Nuclear Power Company

  1. Nuclear Material Control and Accountability System Effectiveness Tool (MSET)

    International Nuclear Information System (INIS)

    Powell, Danny H.; Elwood, Robert H. Jr.; Roche, Charles T.; Campbell, Billy J.; Hammond, Glenn A.; Meppen, Bruce W.; Brown, Richard F.

    2011-01-01

    A nuclear material control and accountability (MC and A) system effectiveness tool (MSET) has been developed in the United States for use in evaluating material protection, control, and accountability (MPC and A) systems in nuclear facilities. The project was commissioned by the National Nuclear Security Administration's Office of International Material Protection and Cooperation. MSET was developed by personnel with experience spanning more than six decades in both the U.S. and international nuclear programs and with experience in probabilistic risk assessment (PRA) in the nuclear power industry. MSET offers significant potential benefits for improving nuclear safeguards and security in any nation with a nuclear program. MSET provides a design basis for developing an MC and A system at a nuclear facility that functions to protect against insider theft or diversion of nuclear materials. MSET analyzes the system and identifies several risk importance factors that show where sustainability is essential for optimal performance and where performance degradation has the greatest impact on total system risk. MSET contains five major components: (1) A functional model that shows how to design, build, implement, and operate a robust nuclear MC and A system (2) A fault tree of the operating MC and A system that adapts PRA methodology to analyze system effectiveness and give a relative risk of failure assessment of the system (3) A questionnaire used to document the facility's current MPC and A system (provides data to evaluate the quality of the system and the level of performance of each basic task performed throughout the material balance area (MBA)) (4) A formal process of applying expert judgment to convert the facility questionnaire data into numeric values representing the performance level of each basic event for use in the fault tree risk assessment calculations (5) PRA software that performs the fault tree risk assessment calculations and produces risk importance

  2. T-book. Reliability data of components in Nordic nuclear power plants. 6. ed

    International Nuclear Information System (INIS)

    2005-01-01

    The main objective of the T-Book is to provide reliability data for the unavailability computations that are made for each component that is considered in the compulsory, probabilistic safety assessments (PSA) of nuclear power plants. As the use of PSA is large in the normal safety work at the NPPs, there is a need for easily accessible and reliable failure data. The failure characteristics presented in the T-Book are primarily based on the failure reports stored in the central database TUD and the Licensee Event Reports delivered to the Swedish Nuclear Power Inspectorate (SKI). Fortunately, the TUD database was started already in the middle of the seventies by the Swedish power companies. In 1981, the Finnish power company TVO, operating two reactor units of Swedish design, joined the data collection system. Before the TUD data are statistically treated they are carefully examined with respect to the consistency and correctness. This T-Book comprises only critical failures, i.e. failures that stops the function of components or leads to repair. The first edition of the T-Book was issued in 1982 encompassing operational statistics from 21 reactor years. The second edition was published 1985, based on operating data covering about 40 reactor years. The T-Book 3 was published in 1992 and included data up to the operating year 1987 (108 reactor years). Edition 4 was published 1994 containing information up to and including 1992 (178 reactor years). Edition 5 was published year 2000 containing information up to and including 1996 (234 reactor years). This edition 6 contains information including year 2002 (315 reactor years). At the same time as the amount of data has increased with the successive editions of the T-Book there has been a continuous work to improve the methods for the statistical inference and related program tools, required to derive the reliability parameters from the operational data in the database. Already in the initial edition there was a Bayesian

  3. GPU Nuclear Corporation's radiation exposure management system

    International Nuclear Information System (INIS)

    Slobodien, M.J.; Bovino, A.A.; Perry, O.R.; Hildebrand, J.E.

    1984-01-01

    GPU Nuclear Corporation has developed a central main frame (IBM 3081) based radiation exposure management system which provides real time and batch transactions for three separate reactor facilities. The structure and function of the data base are discussed. The system's main features include real time on-line radiation work permit generation and personnel exposure tracking; dose accountability as a function of system and component, job type, worker classification, and work location; and personnel dosemeter (TLD and self-reading pocket dosemeters) data processing. The system also carries the qualifications of all radiation workers including RWP training, respiratory protection training, results of respirator fit tests and medical exams. A warning system is used to prevent non-qualified persons from entering controlled areas. The main frame system is interfaced with a variety of mini and micro computer systems for dosemetry, statistical and graphics applications. These are discussed. Some unique dosemetry features which are discussed include assessment of dose for up to 140 parts of the body with dose evaluations at 7,300 and 1000 mg/cm 2 for each part, tracking of MPC hours on a 7 day rolling schedule; automatic pairing of TLD and self-reading pocket dosemeter values, creation and updating of NRC Forms 4 and 5, generation of NRC required 20.407 and Reg Guide 1.16 reports. As of July 1983, over 20 remote on-line stations were in use with plans to add 20-30 more by May 1984. The system provides response times for on-line activities of 2-7 seconds and 23 1/2 hours per day ''up time''. Examples of the various on-line and batch transactions are described

  4. A pilot application of risk-informed methods to establish inservice inspection priorities for nuclear components at Surry Unit 1 Nuclear Power Station. Revision 1

    International Nuclear Information System (INIS)

    Vo, T.V.; Phan, H.K.; Gore, B.F.; Simonen, F.A.; Doctor, S.R.

    1997-02-01

    As part of the Nondestructive Evaluation Reliability Program sponsored by the US Nuclear Regulatory Commission, the Pacific Northwest National Laboratory has developed risk-informed approaches for inservice inspection plans of nuclear power plants. This method uses probabilistic risk assessment (PRA) results to identify and prioritize the most risk-important components for inspection. The Surry Nuclear Power Station Unit 1 was selected for pilot application of this methodology. This report, which incorporates more recent plant-specific information and improved risk-informed methodology and tools, is Revision 1 of the earlier report (NUREG/CR-6181). The methodology discussed in the original report is no longer current and a preferred methodology is presented in this Revision. This report, NUREG/CR-6181, Rev. 1, therefore supersedes the earlier NUREG/CR-6181 published in August 1994. The specific systems addressed in this report are the auxiliary feedwater, the low-pressure injection, and the reactor coolant systems. The results provide a risk-informed ranking of components within these systems

  5. Treatment of core components from nuclear power plants with PWR and BWR reactors - 16043

    International Nuclear Information System (INIS)

    Viermann, Joerg; Friske, Andreas; Radzuweit, Joerg

    2009-01-01

    During operation of a Nuclear Power Plant components inside the RPV get irradiated. Irradiation has an effect on physical properties of these components. Some components have to be replaced after certain neutron doses or respectively after a certain operating time of the plant. Such components are for instance water channels and control rods from Boiling Water Reactors (BWR) or control elements, poisoning elements and flow restrictors from Pressurized Water Reactors (PWR). Most of these components are stored in the fuel pool for a certain time after replacement. Then they have to be packaged for further treatment or for disposal. More than 25 years ago GNS developed a system for disposal of irradiated core components which was based on a waste container suitable for transport, storage and disposal of Intermediate Level Waste (ILW), the so-called MOSAIK R cask. The MOSAIK R family of casks is subject of a separate presentation at the ICEM 09 conference. Besides the MOSAIK R cask the treatment system developed by GNS comprised underwater shears to cut the components to size as well as different types of equipment to handle the components, the shears and the MOSAIK R casks in the fuel pool. Over a decade of experience it showed that this system although effective needed improvement for BWR plants where many water channels and control rods had to be replaced after a certain operating time. Because of the large numbers of components the time period needed to cut the components in the pool had a too big influence on other operational work like rearranging of fuel assemblies in the pool. The system was therefore further developed and again a suitable cask was the heart of the solution. GNS developed the type MOSAIK R 80 T, a cask that is capable to ship the unsegmented components with a length of approx. 4.5 m from the Power plants to an external treatment centre. This treatment centre consisting of a hot cell installation with a scrap shear, super-compactor and a heavy

  6. Consideration of a design optimization method for advanced nuclear power plant thermal-hydraulic components

    International Nuclear Information System (INIS)

    Ridluan, Artit; Tokuhiro, Akira; Manic, Milos; Patterson, Michael; Danchus, William

    2009-01-01

    In order to meet the global energy demand and also mitigate climate change, we anticipate a significant resurgence of nuclear power in the next 50 years. Globally, Generation III plants (ABWR) have been built; Gen' III+ plants (EPR, AP1000 others) are anticipated in the near term. The U.S. DOE and Japan are respectively pursuing the NGNP and MSFR. There is renewed interest in closing the fuel cycle and gradually introducing the fast reactor into the LWR-dominated global fleet. In order to meet Generation IV criteria, i.e. thermal efficiency, inherent safety, proliferation resistance and economic competitiveness, plant and energy conversion system engineering design have to increasingly meet strict design criteria with reduced margin for reliable safety and uncertainties. Here, we considered a design optimization approach using an anticipated NGNP thermal system component as a Case Study. A systematic, efficient methodology is needed to reduce time consuming trial-and-error and computationally-intensive analyses. We thus developed a design optimization method linking three elements; that is, benchmarked CFD used as a 'design tool', artificial neural networks (ANN) to accommodate non-linear system behavior and enhancement of the 'design space', and finally, response surface methodology (RSM) to optimize the design solution with targeted constraints. The paper presents the methodology including guiding principles, an integration of CFD into design theory and practice, consideration of system non-linearities (such as fluctuating operating conditions) and systematic enhancement of the design space via application of ANN, and a stochastic optimization approach (RSM) with targeted constraints. Results from a Case Study optimizing the printed circuit heat exchanger for the NGNP energy conversion system will be presented. (author)

  7. Development of nuclear material accountancy control system

    International Nuclear Information System (INIS)

    Hirosawa, Naonori; Kashima, Sadamitsu; Akiba, Mitsunori

    1992-01-01

    PNC is developing a wide area of nuclear fuel cycle. Therefore, much nuclear material with a various form exists at each facility in the Works, and the controls of the inventory changes and the physical inventories of nuclear material are important. Nuclear material accountancy is a basic measure in safeguards system based on Non-Proliferation Treaty (NPT). In the light of such importance of material accountancy, the data base of nuclear material control and the material accountancy report system for all facilities has been developed by using the computer. By this system, accountancy report to STA is being presented certainly and timely. Property management and rapid corresponding to various inquiries can be carried out by the data base system which has free item searching procedure. (author)

  8. Future development of nuclear energy systems

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-03-01

    Nuclear energy development in Japan has passed about 30 years, and reaches to a step to supply about 35 % of total electric power demand. However, together with globalization of economic and technical development, its future progressing method is required for its new efforts. Among such conditions, when considering a state of future type nuclear energy application, its contribution to further environmental conservation and international cooperation is essential, and it is required for adoption to such requirement how it is made an energy source with excellent economics.The Research Committee on 'Engineering Design on Nuclear Energy Systems' established under recognition in 1998 has been carried out some discussions on present and future status of nuclear energy development. And so forth under participation of outer specialists. Here were summarized on two year's committee actions containing them and viewpoints of nuclear industries, popularization of nuclear system technology, and so forth. (G.K.)

  9. Nuclear fuel element leak detection system

    International Nuclear Information System (INIS)

    John, C.D. Jr.

    1978-01-01

    Disclosed is a leak detection system integral with a wall of a building used to fabricate nuclear fuel elements for detecting radiation leakage from the nuclear fuel elements as the fuel elements exit the building. The leak detecting system comprises a shielded compartment constructed to withstand environmental hazards extending into a similarly constructed building and having sealed doors on both ends along with leak detecting apparatus connected to the compartment. The leak detecting system provides a system for removing a nuclear fuel element from its fabrication building while testing for radiation leaks in the fuel element

  10. 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)

  11. Pulse shaping amplifier (PSA) for nuclear spectroscopy system

    International Nuclear Information System (INIS)

    Lombigit, L.; Maslina Mohd Ibrahim; Nolida Yusup; Nur Aira Abdul Rahman; Yong, C.F.

    2014-01-01

    Pulse Shaping Amplifier (PSA) is an essential components in nuclear spectroscopy system. This networks have two functions; to shape the output pulse and performs noise filtering. In this paper, we describes procedure for design and development of a pulse shaping amplifier which can be used for nuclear spectroscopy system. This prototype was developed using high performance electronics devices and assembled on a FR4 type printed circuit board. Performance of this prototype was tested by comparing it with an equivalent commercial spectroscopy amplifier (Model SILENA 7611). The test results show that the performance of this prototype is comparable to the commercial spectroscopic amplifier. (author)

  12. Report on the Regulators Experience of NDT Qualification for In-service Inspection of Nuclear Components

    International Nuclear Information System (INIS)

    2003-08-01

    In November 1992, the Nuclear Regulators Working Group (NRWG) decided to set up a task force on qualification of non-destructive testing (NDT) systems for pre and in-service inspection of light water reactors. The first task was to agree on the philosophy and principles governing the qualification of techniques, equipment, software, procedures, and personnel for NDT to be used for the inspection of structural components that are important to safety in nuclear power plants; and to establish a common view on essential aspects of NDT qualifications. The first task, which also included a comparison of the common views of the European regulators with the qualification approach outlined in Appendix VIII to Section XI of the ASME Code, was completed in 1996. The result was published in the report ''Common position of European regulators on qualification of NDT systems for pre- and in-service inspection of light water reactor components''2. In parallel, the European nuclear power industries had set up a working group, the European Network for Inspection Qualification (ENIQ), to discuss and agree on how to perform inspection qualifications. In 1995, ENIQ finalized its first version of ''European methodology for qualification of non-destructive tests''3. A second version 4 was then published in 1997. This second version is in relatively close agreement with the principles given in the regulators common position document. With these two basic documents, a platform was established for the further development of qualification strategies in the European countries. The second task of the NRWG Task Force was to follow and evaluate the first ENIQ pilot study from a regulatory point of view. The objective of this pilot study was to explore ways of how to apply the European qualification methodology and to test its feasibility. The pilot study commenced late 1996 and was planned to be finalized a year later. Depending on unforeseen difficulties, the pilot study has been delayed

  13. Multi-level predictive maintenance for multi-component systems

    International Nuclear Information System (INIS)

    Nguyen, Kim-Anh; Do, Phuc; Grall, Antoine

    2015-01-01

    In this paper, a novel predictive maintenance policy with multi-level decision-making is proposed for multi-component system with complex structure. The main idea is to propose a decision-making process considered on two levels: system level and component one. The goal of the decision rules at the system level is to address if preventive maintenance actions are needed regarding the predictive reliability of the system. At component level the decision rules aim at identifying optimally a group of several components to be preventively maintained when preventive maintenance is trigged due to the system level decision. Selecting optimal components is based on a cost-based group improvement factor taking into account the predictive reliability of the components, the economic dependencies as well as the location of the components in the system. Moreover, a cost model is developed to find the optimal maintenance decision variables. A 14-component system is finally introduced to illustrate the use and the performance of the proposed predictive maintenance policy. Different sensitivity analysis are also investigated and discussed. Indeed, the proposed policy provides more flexibility in maintenance decision-making for complex structure systems, hence leading to significant profits in terms of maintenance cost when compared with existing policies. - Highlights: • A predictive maintenance policy for complex structure systems is proposed. • Multi-level decision process based on prognostic results is proposed. • A cost-based group importance measure is introduced for decision-making. • Both positive and negative dependencies between components are investigated. • A cost model and Monte Carlo simulation are developed for optimization process.

  14. Component Configuration Control System: an application of logic programming

    International Nuclear Information System (INIS)

    Stratton, R.C.; Town, G.G.

    1985-01-01

    A computer application system is described which provides nuclear reactor power plant operators with an improved decision support system. This system combines traditional computer applications such as graphics display with artificial intelligence methodologies such as reasoning and diagnosis so as to improve plant operability. This paper discusses the issues, and a solution, involved with the system integration of applications developed using traditional and artificial intelligence languages

  15. Progress on Plant-Level Components for Nuclear Fuel Recycling: Commonality

    International Nuclear Information System (INIS)

    De Almeida, Valmor F.

    2011-01-01

    Progress made in developing a common mathematical modeling framework for plant-level components of a simulation toolkit for nuclear fuel recycling is summarized. This ongoing work is performed under the DOE Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, which has an element focusing on safeguards and separations (SafeSeps). One goal of this element is to develop a modeling and simulation toolkit for used nuclear fuel recycling. The primary function of the SafeSeps simulation toolkit is to enable the time-dependent coupling of separation modules and safeguards tools (either native or third-party supplied) that simulate and/or monitor the individual separation processes in a separations plant. The toolkit integration environment will offer an interface for the modules to register in the toolkit domain based on the commonality of diverse unit operations. This report discusses the source of this commonality from a combined mathematical modeling and software design perspectives, and it defines the initial basic concepts needed for development of application modules and their integrated form, that is, an application software. A unifying mathematical theory of chemical thermomechanical network transport for physicochemical systems is proposed and outlined as the basis for developing advanced modules. A program for developing this theory from the underlying first-principles continuum thermomechanics will be needed in future developments; accomplishment of this task will enable the development of a modern modeling approach for plant-level models. Rigorous, advanced modeling approaches at the plant-level can only proceed from the development of reduced (or low-order) models based on a solid continuum field theory foundation. Such development will pave the way for future programmatic activities on software verification, simulation validation, and model uncertainty quantification on a scientific basis; currently, no satisfactory foundation exists for

  16. Progress on Plant-Level Components for Nuclear Fuel Recycling: Commonality

    Energy Technology Data Exchange (ETDEWEB)

    de Almeida, Valmor F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2011-08-15

    Progress made in developing a common mathematical modeling framework for plant-level components of a simulation toolkit for nuclear fuel recycling is summarized. This ongoing work is performed under the DOE Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, which has an element focusing on safeguards and separations (SafeSeps). One goal of this element is to develop a modeling and simulation toolkit for used nuclear fuel recycling. The primary function of the SafeSeps simulation toolkit is to enable the time-dependent coupling of separation modules and safeguards tools (either native or third-party supplied) that simulate and/or monitor the individual separation processes in a separations plant. The toolkit integration environment will offer an interface for the modules to register in the toolkit domain based on the commonality of diverse unit operations. This report discusses the source of this commonality from a combined mathematical modeling and software design perspectives, and it defines the initial basic concepts needed for development of application modules and their integrated form, that is, an application software. A unifying mathematical theory of chemical thermomechanical network transport for physicochemical systems is proposed and outlined as the basis for developing advanced modules. A program for developing this theory from the underlying first-principles continuum thermomechanics will be needed in future developments; accomplishment of this task will enable the development of a modern modeling approach for plant-level models. Rigorous, advanced modeling approaches at the plant-level can only proceed from the development of reduced (or low-order) models based on a solid continuum field theory foundation. Such development will pave the way for future programmatic activities on software verification, simulation validation, and model uncertainty quantification on a scientific basis; currently, no satisfactory foundation exists for

  17. The Location GNSS Modules for the Components of Proteus System

    Science.gov (United States)

    Brzostowski, K.; Darakchiev, R.; Foks-Ryznar, A.; Sitek, P.

    2012-01-01

    The Proteus system - the Integrated Mobile System for Counterterrorism and Rescue Operations is a complex innovative project. To assure the best possible localization of mobile components of the system, many different Global Navigation Satellite System (GNSS) modules were taken into account. In order to chose the best solution many types of tests were done. Full results and conclusions are presented in this paper. The idea of measurements was to test modules in GPS Standard Positioning Service (SPS) with EGNOS system specification according to certain algorithms. The tests had to answer the question: what type of GNSS modules should be used on different components with respect to specific usage of Proteus system. The second goal of tests was to check the solution quality of integrated GNSS/INS (Inertial Navigation System) and its possible usage in some Proteus system components.

  18. Understanding and Managing Aging of Spent Nuclear Fuel and Facility Components in Wet Storage

    International Nuclear Information System (INIS)

    Johnson, A. B.

    2007-01-01

    Storage of nuclear fuel after it has been discharged from reactors has become the leading spent fuel management option. Many storage facilities are being required to operate longer than originally anticipated. Aging is a term that has emerged to focus attention on the consequences of extended operation on systems, structures, and components that comprise the storage facilities. The key to mitigation of age-related degradation in storage facilities is to implement effective strategies to understand and manage aging of the facility materials. A systematic approach to preclude serious effects of age-related degradation is addressed in this paper, directed principally to smaller facilities (test and research reactors). The first need is to assess the materials that comprise the facility and the environments that they are subject to. Access to historical data on facility design, fabrication, and operation can facilitate assessment of expected materials performance. Methods to assess the current condition of facility materials are summarized in the paper. Each facility needs an aging management plan to define the scope of the management program, involving identification of the materials that need specific actions to manage age-related degradation. For each material identified, one or more aging management programs are developed and become part of the plan Several national and international organizations have invested in development of comprehensive and systematic approaches to aging management. A method developed by the US Nuclear Regulatory Commission is recommended as a concise template to organize measures to effectively manage age-related degradation of storage facility materials, including the scope of inspection, surveillance, and maintenance that is needed to assure successful operation of the facility over its required life. Important to effective aging management is a staff that is alert for evidence of materials degradation and committed to carry out the aging

  19. PML nuclear body component Sp140 is a novel autoantigen in primary biliary cirrhosis.

    Science.gov (United States)

    Granito, Alessandro; Yang, Wei-Hong; Muratori, Luigi; Lim, Mark J; Nakajima, Ayako; Ferri, Silvia; Pappas, Georgios; Quarneti, Chiara; Bianchi, Francesco B; Bloch, Donald B; Muratori, Paolo

    2010-01-01

    Some patients with primary biliary cirrhosis (PBC) have antinuclear antibodies (ANAs). These ANAs include the "multiple nuclear dots" (MND) staining pattern, targeting promyelocytic leukemia protein (PML) nuclear body (NB) components, such as "speckled 100-kD" protein (Sp100) and PML. A new PML NB protein, designated as Sp140, was identified using serum from a PBC patient. The aim of this study was to analyze the immune response against Sp140 protein in PBC patients. We studied 135 PBC patients and 157 pathological controls with type 1 autoimmune hepatitis, primary sclerosing cholangitis, and systemic lupus erythematosus. We used indirect immunofluorescence and a neuroblastoma cell line expressing Sp140 for detecting anti-Sp140 antibodies, and a commercially available immunoblot for detecting anti-Sp100 and anti-PML antibodies. Anti-Sp140 antibodies were present in 20 (15%) PBC patients but not in control samples, with a higher frequency in antimitochondrial antibody (AMA)-negative cases (53 vs. 9%, P<0.0001). Anti-Sp140 antibodies were found together with anti-Sp100 antibodies in all but one case (19 of 20, 90%) and with anti-PML antibodies in 12 (60%) cases. Anti-Sp140 positivity was not associated with a specific clinical feature of PBC. Our study identifies Sp140 as a new, highly specific autoantigen in PBC for the first time. The very frequent coexistence of anti-Sp140, anti-Sp100 and anti-PML antibodies suggests that the NB is a multiantigenic complex in PBC and enhances the diagnostic significance of these reactivities, which are particularly useful in AMA-negative cases.

  20. Robotic control architecture development for automated nuclear material handling systems

    International Nuclear Information System (INIS)

    Merrill, R.D.; Hurd, R.; Couture, S.; Wilhelmsen, K.

    1995-02-01

    Lawrence Livermore National Laboratory (LLNL) is engaged in developing automated systems for handling materials for mixed waste treatment, nuclear pyrochemical processing, and weapon components disassembly. In support of these application areas there is an extensive robotic development program. This paper will describe the portion of this effort at LLNL devoted to control system architecture development, and review two applications currently being implemented which incorporate these technologies