WorldWideScience

Sample records for plant design equipment

  1. Liquid Metal Fast Breeder Reactor plant maintenance and equipment design

    International Nuclear Information System (INIS)

    Swannack, D.L.

    1982-01-01

    This paper provides a summary of maintenance equipment considerations and actual plant handling experiences from operation of a sodium-cooled reactor, the Fast Flux Test Facility (FFTF). Equipment areas relating to design, repair techniques, in-cell handling, logistics and facility services are discussed. Plant design must make provisions for handling and replacement of components within containment or allow for transport to an ex-containment area for repair. The modular cask assemblies and transporter systems developed for FFTF can service major plant components as well as smaller units. The plant and equipment designs for the Clinch River Breeder Reactor (CRBR) plant have been patterned after successful FFTF equipment

  2. Nuclear power plant equipment design and construction rules

    International Nuclear Information System (INIS)

    Boiron, P.

    1983-03-01

    Presentation of the AFCEN (French association for nuclear power plant equipment design and construction rules) working, of its edition activity and of somes of its edited documents such as RCC-C (design and construction rules for PWR power plant fuel assemblies) and RCC-E (design and construction rules for nuclear facility electrical equipments) [fr

  3. Design of equipment management information system for nuclear power plant

    International Nuclear Information System (INIS)

    Wang Chengyuan

    1996-01-01

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

  4. Guideline for design and construction radiation monitoring equipments for Tokai reprocessing plant

    International Nuclear Information System (INIS)

    Miyabe, Kenjiro; Ninomiya, Kazushige; Jin, Kazumi; Morifuji, Masayuki; Nemoto, Kazuhiko; Sato, Akira; Kawai, Keiichi

    1999-12-01

    Various kind of radiation monitoring equipment are used in radiation controlled area at each facility of Tokai reprocessing plant. These equipments have been designed and constructed based on the users requirements, and permitted by governmental regulation office. And, design has been carried out in consideration of the adoption of the new technology and our operational experience. Then, it has been used effectively for the radiation control of the facilities. This report summarizes the technical requirements that should be taken into consideration in the design and installation of radiation monitoring equipments. These requirements are fundamentally applicable when the equipments of the new facilities will be designed or the present instruments will be replaced. (author)

  5. Seismic design of equipment and piping systems for nuclear power plants in Japan

    International Nuclear Information System (INIS)

    Minematsu, Akiyoshi

    1997-01-01

    The philosophy of seismic design for nuclear power plant facilities in Japan is based on 'Examination Guide for Seismic Design of Nuclear Power Reactor Facilities: Nuclear Power Safety Committee, July 20, 1981' (referred to as 'Examination Guide' hereinafter) and the present design criteria have been established based on the survey of governmental improvement and standardization program. The detailed design implementation procedure is further described in 'Technical Guidelines for Aseismic Design of Nuclear Power Plants, JEAG4601-1987: Japan Electric Association'. This report describes the principles and design procedure of the seismic design of equipment/piping systems for nuclear power plant in Japan. (J.P.N.)

  6. Seismic design of equipment and piping systems for nuclear power plants in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Minematsu, Akiyoshi [Tokyo Electric Power Co., Inc. (Japan)

    1997-03-01

    The philosophy of seismic design for nuclear power plant facilities in Japan is based on `Examination Guide for Seismic Design of Nuclear Power Reactor Facilities: Nuclear Power Safety Committee, July 20, 1981` (referred to as `Examination Guide` hereinafter) and the present design criteria have been established based on the survey of governmental improvement and standardization program. The detailed design implementation procedure is further described in `Technical Guidelines for Aseismic Design of Nuclear Power Plants, JEAG4601-1987: Japan Electric Association`. This report describes the principles and design procedure of the seismic design of equipment/piping systems for nuclear power plant in Japan. (J.P.N.)

  7. Research program for seismic qualification of nuclear plant electrical and mechanical equipment. Task 4. Use of fragility in seismic design of nuclear plant equipment. Volume 4

    International Nuclear Information System (INIS)

    Kana, D.D.; Pomerening, D.J.

    1984-08-01

    The Research Program for Seismic Qualification of Nuclear Plant Electrical and Mechanical Equipment has spanned a period of three years and resulted in seven technical summary reports, each of which have covered in detail the findings of different tasks and subtasks, and have been combined into five NUREG/CR volumes. Volume 4 presents study of the use of fragility concepts in the design of nuclear plant equipment and compares the results of state-of-the-art proof testing with fragility testing

  8. Improvements of seismic design of nuclear power plant equipment

    International Nuclear Information System (INIS)

    Suzuki, Kohei; Takayama, Yoshihiro.

    1997-01-01

    A brief survey and overview of the current research and development in Japan was presented. Particularly, several kinds of new dampers and isolators were developed and those effectiveness were examined by caring out the large-scale vibration test and so on. The evaluation of the energy absorption of these damping devices at the earthquake appeared to be significant. In addition, it must be necessary to investigate the design margin and the failure mode and limit problem to these devices and the nuclear structures and piping supported by those. Mutual exchange of the information related to these technology and research has to be put forward and cooperative works including the international conference on those issues should be promoted. (J.P.N.)

  9. Improvements of seismic design of nuclear power plant equipment

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Kohei [Tokyo Metropolitan Univ., Hachioji (Japan). Faculty of Technology; Takayama, Yoshihiro

    1997-03-01

    A brief survey and overview of the current research and development in Japan was presented. Particularly, several kinds of new dampers and isolators were developed and those effectiveness were examined by caring out the large-scale vibration test and so on. The evaluation of the energy absorption of these damping devices at the earthquake appeared to be significant. In addition, it must be necessary to investigate the design margin and the failure mode and limit problem to these devices and the nuclear structures and piping supported by those. Mutual exchange of the information related to these technology and research has to be put forward and cooperative works including the international conference on those issues should be promoted. (J.P.N.)

  10. Design and Implementation of Equipment for Enhanced Safeguards of a Plutonium Storage in a Reprocessing Plant

    International Nuclear Information System (INIS)

    Richir, P.; Dechamp, L.; Buchet, P.; Dransart, P.; Dzbikowicz, Z.; Peerani, P.; ); Pierssens, L.; Persson, L.; Ancius, D.; Synetos, S.; ); Edmonds, N.; Homer, A.; Benn, K.-A.; Polkey, A.

    2015-01-01

    The Nuclear Security unit (NUSEC) of the Institute for Transuranium Elements (ITU, JRC) was entrusted by DG ENER to design and implement equipment in order to achieve enhanced safeguards of a plutonium dioxide storage located on the MAGNOX reprocessing plant in Sellafield (UK). Enhanced safeguards must lead to a win-win situation for all parties involved. In this case the DG ENER inspectorate will save inspection time, manpower and future financial resources and the operator will have the right to access its storage without the need for inspector presence. To reach this goal, while at the same time taking into account current budget constraints, NUSEC developed applications that use equipment commonly used in the safety and security fields but so far have not been used in safeguards. For instance, two laser scanners are used to detect entry/exit events into and out of the store and to provide the necessary information to an algorithm in order to categorize objects/people passing the scanners, e.g., a Fork Lift Truck, a trolley used to bring in PuO 2 containers, a system used for the dispatch of cans, people, etc. An RFID reader is used to identify equipment duly authorized to access the store. All PuO 2 containers arriving from the production line must be weighed, identified and measured using gamma and neutron detectors before they can be transferred to the store. For this purpose an Unattended Combined Measurement System (UCMS) was designed and manufactured by the JRC in order to do all verification activities using a single instrument. This paper describes the design features of the equipment and its implementation with the support of the Sellafield Ltd. in the framework of the MAGNOX store project. (author)

  11. Improvement of Productivity in TIG Welding Plant by Equipment Design in Orbit

    Science.gov (United States)

    Gnanavel, C.; Saravanan, R.; Chandrasekaran, M.; Jayakanth, J. J.

    2017-03-01

    Measurements and improvements are very indispensable task at all levels of management. Here some samples are, at operator level: Measuring operating parameters to ensure OEE (Overall Equipment Effectiveness) and measuring Q components performance to ensure quality, at supervisory level: measuring operator’s performance to ensure labour utility at managerial level: production and productivity measurements and at top level capital and capacity utilization. An often accepted statement is “Improvement is impossible without measurement”. Measurements often referred as observation. The case study was conducted at Government Boiler factory in India. The scientific approach followed for indentifying non value added activities. Personalised new equipment designed and installed to achieve productivity improvement of 85% for a day. The new equipment can serve 360o around its axis hence it simplified loading and unloading procedures as well as reduce their times and ensured effective space and time.

  12. WIPP conceptual design report. Addendum J. Support equipment in the high level waste facility of the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    Rieb, M.J.; Foley, R.S.

    1977-04-01

    The Aerojet Manufacturing Company (AMCO) received a contract in November 1976 to provide consulting services in assisting Holmes and Narver, Incorporated with the conceptual designs, cost estimates, and schedules of equipment used to handle waste casks, to decontaminate waste canisters and to overpack damaged or highly contaminated waste canisters for the Waste Isolation Pilot Plant (WIPP). Also, the layout of the hot cell in which canister handling, overpack and decontamination takes place was to be reviewed along with the time and motion study of the cell operations. This report has been prepared to present the results of the efforts and contains all technical and planning data developed during the program. The contents of this report are presented in three sections: (1) comments on the existing design criteria, equipment conceptual designs, hot cell design and time and motion studies of projected hot cell activities; (2) design descriptions of the equipment concepts and justification for varying from the existing concept (if a variation occurred). Drawings of each concept are provided in Appendix A. These design descriptions and drawings were used as the basis for the cost estimates; and (3) schedule projections and cost estimates for the equipment described in Section 2. Detail cost estimate backup data is provided in Appendix B

  13. Electrical equipment design library

    International Nuclear Information System (INIS)

    1994-01-01

    This book guides the design supervision, construction order for electrical equipment. The contents of this library are let's use electricity like this, leading-in-pole and casual power, electric pole install below 300KVA, electric pole install below 301∼1000KVA, electric pole install exceed 1000KVA, rooftop install exceed 1000KVA, CUBICLE type, 154KV services. It adds an appendix.

  14. Operation monitor for plant equipment

    International Nuclear Information System (INIS)

    Kondo, Tetsufumi; Kanemoto, Shigeru.

    1991-01-01

    In a nuclear power plant, states of each of equipment in the plant are monitored accurately even under such a operation condition that the power is changed. That is, the fundamental idea is based on a model comparison method. A deviation between an output signal upon normal plant state obtained in a forecasting model device and that of the objective equipment in the plant are compared with a predetermined value. The result of the comparison is inputted to an alarm device to alarm the abnormality of the states of the equipment to an operator. The device of the present invention thus constituted can monitor the abnormality of the operation of equipment accurately even under such a condition that a power level fluctuates. As a result, it can remarkably contribute to mitigate operator's monitoring operation under the condition such as during load following operation. (I.S.)

  15. MOPABA-H2 - Computer code for calculation of hydrogen generation and distribution in the equipment of power plants with WWER type reactors in design modes of operation

    International Nuclear Information System (INIS)

    Arkhipov, O.P.; Kharitonov, Yu.V.; Shumskiy, A.M.; Kabakchi, S.A.

    2002-01-01

    With the aim of ensuring the hydrogen explosive-proof situation in the reactor plant, a complex of scientific-and-research work was carried out including the following: revealing the mechanisms of generation and release of hydrogen in the primary equipment components under design operation modes of the reactor plant with WWER; development of calculation procedure and computer code MOPABA-H2 enabling to determine the hydrogen content in RP equipment components under design operation modes. In the process of procedure development it was found out that the calculation of hydrogen content in the plant equipment requires development of the following main mathematical models: radiochemical processes in the primary coolant which has impurities and added special reagents; absorption of the core ionizing radiation by the coolant; steam-zirconium reaction (during design-basis accident of LOCA type); coolant mass transfer over the reactor plant equipment including transition of the phase boundary by the components of the coolant. (author)

  16. The design of in-cell equipment for nuclear fuel reprocessing plant with special reference to the decanners for Pond 5, Sellafield

    International Nuclear Information System (INIS)

    Evans, D.A.

    1987-01-01

    The in-cave equipment has to meet many demanding criteria, which require not only sound and creative design thinking, but a range of design management techniques to ensure the success of the plant. This paper discusses these in some detail in relation to the decanners for Pond 5. (author)

  17. Design of experimental equipment at CRNL

    International Nuclear Information System (INIS)

    Godden, B.

    1976-01-01

    The Plant Design Division provides a design service to the research and development effort at CRNL. Severe constraints, both environmentally and spatially, are placed on the design of special equipment. Several examples are given. Finally, the use of automated drafting systems is described. (author)

  18. PROMSYS, Plant Equipment Maintenance and Inspection Scheduling

    International Nuclear Information System (INIS)

    Morgan, D.L.; Srite, B.E.

    1986-01-01

    1 - Description of problem or function: PROMSYS is a computer system designed to automate the scheduling of routine maintenance and inspection of plant equipment. This 'programmed maintenance' provides the detailed planning and accomplishment of lubrication, inspection, and similar repetitive maintenance activities which can be scheduled at specified predetermined intervals throughout the year. The equipment items included are the typical pumps, blowers, motors, compressors, automotive equipment, refrigeration units, filtering systems, machine shop equipment, cranes, elevators, motor-generator sets, and electrical switchgear found throughout industry, as well as cell ventilation, shielding, containment, and material handling equipment unique to nuclear research and development facilities. Four related programs are used to produce sorted schedule lists, delinquent work lists, and optional master lists. Five additional programs are used to create and maintain records of all scheduled and unscheduled maintenance history. 2 - Method of solution: Service specifications and frequency are established and stored. The computer program reviews schedules weekly and prints, on schedule cards, instructions for service that is due the following week. The basic output from the computer program comes in two forms: programmed-maintenance schedule cards and programmed-maintenance data sheets. The data sheets can be issued in numerical building, route, and location number sequence as equipment lists, grouped for work assigned to a particular foreman as the foreman's equipment list, or grouped by work charged to a particular work order as the work-order list. Data sheets grouped by equipment classification are called the equipment classification list

  19. Plant-wide integrated equipment monitoring and analysis system

    International Nuclear Information System (INIS)

    Morimoto, C.N.; Hunter, T.A.; Chiang, S.C.

    2004-01-01

    A nuclear power plant equipment monitoring system monitors plant equipment and reports deteriorating equipment conditions. The more advanced equipment monitoring systems can also provide information for understanding the symptoms and diagnosing the root cause of a problem. Maximizing the equipment availability and minimizing or eliminating consequential damages are the ultimate goals of equipment monitoring systems. GE Integrated Equipment Monitoring System (GEIEMS) is designed as an integrated intelligent monitoring and analysis system for plant-wide application for BWR plants. This approach reduces system maintenance efforts and equipment monitoring costs and provides information for integrated planning. This paper describes GEIEMS and how the current system is being upgraded to meet General Electric's vision for plant-wide decision support. (author)

  20. Margins related to equipment design

    International Nuclear Information System (INIS)

    Devos, J.

    1994-01-01

    Safety margins related to design of reactor equipment are defined according to safety regulations. Advanced best estimate methods are proposed including some examples which were computed and compared to experimental results. Best estimate methods require greater computation effort and more material data but give better variable accuracy and need careful experimental validation. Simplified methods compared to the previous are less sensitive to material data, sometimes are more accurate but very long to elaborate

  1. 40 CFR 792.61 - Equipment design.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 31 2010-07-01 2010-07-01 true Equipment design. 792.61 Section 792.61...) GOOD LABORATORY PRACTICE STANDARDS Equipment § 792.61 Equipment design. Equipment used in the... of appropriate design and adequate capacity to function according to the protocol and shall be...

  2. 40 CFR 160.61 - Equipment design.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 23 2010-07-01 2010-07-01 false Equipment design. 160.61 Section 160... LABORATORY PRACTICE STANDARDS Equipment § 160.61 Equipment design. Equipment used in the generation... appropriate design and adequate capacity to function according to the protocol and shall be suitably located...

  3. 21 CFR 58.61 - Equipment design.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 1 2010-04-01 2010-04-01 false Equipment design. 58.61 Section 58.61 Food and... PRACTICE FOR NONCLINICAL LABORATORY STUDIES Equipment § 58.61 Equipment design. Equipment used in the... of appropriate design and adequate capacity to function according to the protocol and shall be...

  4. Improving the hygienic design of closed equipment

    DEFF Research Database (Denmark)

    Friis, Alan; Jensen, Bo Boye Busk

    2005-01-01

    Maintenance of proper hygiene in closed process equipment is in many ways a complex task. The interaction between the physical design and the nature of fluid flow is of main concern. During cleaning the main performance of the flow is to bring cleaning agents in the right doses to all parts of th...... computational fluid dynamics models to be able to predict the cleaning efficiency in especially complex parts of process plants has excellent potentials for desktop improvements and computer pre-validation of the hygienic performance of process plants....

  5. Perry Nuclear Power Plant Area/Equipment Temperature Monitoring Program

    International Nuclear Information System (INIS)

    McGuire, L.L.

    1991-01-01

    The Perry Nuclear Power Plant Area/Equipment Temperature Monitoring Program serves two purposes. The first is to track temperature trends during normal plant operation in areas where suspected deviations from established environmental profiles exist. This includes the use of Resistance Temperature Detectors, Recorders, and Temperature Dots for evaluation of equipment qualified life for comparison with tested parameters and the established Environmental Design Profile. It also may be used to determine the location and duration of steam leaks for effect on equipment qualified life. The second purpose of this program is to aid HVAC design engineers in determining the source of heat outside anticipated design parameters. Resistance Temperature Detectors, Recorders, and Temperature Dots are also used for this application but the results may include design changes to eliminate the excess heat or provide qualified equipment (cable) to withstand the elevated temperature, splitting of environmental zones to capture accurate temperature parameters, or continued environmental monitoring for evaluation of equipment located in hot spots

  6. Qualification of electric equipments for nuclear power plants

    International Nuclear Information System (INIS)

    Chauvin, G.; Raimondo, E.

    1983-03-01

    Description of the testing equipment, testing methods and standards of the resistance to seisms of electrical equipments (switches, pump motors, electrovalves, ...) for electronuclear power plants in France. Presentation of the French design and construction rules for electrical devices in the domestic and export nuclear market (resistance to thermodynamical and chemical stresses, to seisms, etc...) [fr

  7. Seismic qualification method of equipment for nuclear power plant

    International Nuclear Information System (INIS)

    Kim, J.S.; Choi, T.H.; Sulaimana, R.A.

    1995-01-01

    Safety related equipment installed in Korean Nuclear Power Plants are required to perform a safety function during and after a seismic event. To accomplish this safety function, they must be seismically qualified in accordance with the intent and requirements of the USNRC Reg. Guide 1.100 Rev. 02 and IEEE Std. 344-1987. This paper defines and summarizes acceptable criteria and procedures, based on the Korean experience, for seismic qualification of purchased equipment to be installed in a nuclear power plant. As such the paper is intended to be a concise reference by equipment designers, architectural engineering company and plant owners in uniform implementation of commitments to nuclear regulatory agencies such as the USNRC or Korea Institute of Nuclear Safety (KINS) relating to adequacy of seismic Category 1 equipment. Thus, the paper provides the methodologies which can be used for qualifying equipment for safely related service in Nuclear Power Plants in a cost effective manner

  8. Vitrification process equipment design for the West Valley Demonstration Project

    International Nuclear Information System (INIS)

    Chapman, C.C.; Drosjack, W.P.

    1988-10-01

    The vitrification process and equipment design is nearing completion for the West Valley Project. This report provides the basis and current status for the design of the major vessels and equipment within the West Valley Vitrification Plant. A review of the function and key design features of the equipment is also provided. The major subsystems described include the feed preparation and delivery systems, the melter, the canister handling systems, and the process off-gas system. 11 refs., 33 figs., 4 tabs

  9. Process plant equipment operation, control, and reliability

    CERN Document Server

    Holloway, Michael D; Onyewuenyi, Oliver A

    2012-01-01

    "Process Plant Equipment Book is another great publication from Wiley as a reference book for final year students as well as those who will work or are working in chemical production plants and refinery…" -Associate Prof. Dr. Ramli Mat, Deputy Dean (Academic), Faculty of Chemical Engineering, Universiti Teknologi Malaysia "…give[s] readers access to both fundamental information on process plant equipment and to practical ideas, best practices and experiences of highly successful engineers from around the world… The book is illustrated throughout with numerous black & white p

  10. Optimization on replacement period of plant equipment

    International Nuclear Information System (INIS)

    Kasai, Masao; Asano, Hiromi

    2002-01-01

    Optimization of the replacement period of plant equipment is one of the main items to rationalize the activities on plant maintenance. There are several models to replace the equipment and the formulations for optimizing the replacement period are different among these models. In this study, we calculated the optimum replacement periods for some equipment parts based on the replacement models and found that the optimum solutions are not so largely differ from the replacement models as far as the replacement period is not so large. So we will be able to use the most usable model especially in the early phase of rationalization on plant maintenance, since there are large uncertainties in data for optimization. (author)

  11. Seismic verification of nuclear plant equipment anchorage

    International Nuclear Information System (INIS)

    Lepiece, M.; Van Vyve, J.

    1991-01-01

    More than 60% of the electrical power of Belgium is generated by seven PWR nuclear power plants. For three of them, the electro-mechanical equipment had to be reassessed after ten years of operation, because the seismic requirements were upgraded from 0.1 g to 0.17 g free field ground acceleration. The seismic requalification of the active equipment was a critical problem as the classical methods were too conservative. The approach based on the use of the past experience on the seismic behavior of nonnuclear equipment, chosen and developed by the SQUG, had to be transposed to the Belgian N.P.P. The decision of the accept-ability of equipment relies heavily on the aseismatic capacity of anchorage. The Electrical Power Research Institute (EPRI) developed the procedure and guideline for the demonstration of the aseismatic adequacy of equipment anchorage in a cost-effective and consistent manner, to support the decision by Seismic Review Team. The field inspection procedure to identify the type of fasteners and detect their possible defects and the verification procedure developed to calculate the aseismatic capacity of equipment anchorage on the strength of fasteners, the aseismatic capacity of anchorage and the comparison of the capacity with the demand are reported. (K.I.)

  12. Plant equipment services with laser metrology

    International Nuclear Information System (INIS)

    Hayes, J.H.; Kreitman, P.J.

    1995-01-01

    A new industrial metrology process is now being applied to support PWR Nuclear Plant Steam Generator Replacement Projects. The method uses laser tracking interferometry to perform as built surveys of existing and replacement plant equipment. This method provides precision data with a minimum of setup when compared to alternative methods available. In addition there is no post processing required to ascertain validity. The data is obtained quickly, processed in real time and displayed during the survey in the desired coordinate system. These capabilities make this method of industrial measure ideal for various data acquisition needs throughout the power industry, from internal/external equipment templating to area mapping. Laser tracking interferometry is an improvement on the present use of optical instruments and surveying technique. In order to describe the laser tracking interferometry measurement process, previous methods of templating and surveying are first reviewed

  13. Failure diagnosis aiding device for plant equipment

    International Nuclear Information System (INIS)

    Uhara, Yoshihiko.

    1990-01-01

    The present invention intends to improve the efficiency of trouble shooting for equipments of industrial plants such as nuclear power plants. The device of the present invention comprises an intelligence base and an inference mechanism base. The intelligence base comprises a rule base, an information storing section having a part frame and a working frame and a user's frame. The parts frame contains the failure rate on every parts and data on related operations. The working frame contains the importance and frequency of working. The user's frame contains parameters showing the extent of user's skills. The rule base, the parts frame and the working frame can be selected in accordance with the extent of the user's skill in the inference mechanism. With such a constitution, failures can be checked with the intelligence base in accordance with the knowledges for the failures of the equipments and the extent of user's skill by way of the inference mechanism. (I.S.)

  14. Reduction of Equipment Access Time through Cyber Plant Navigation

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Jang Soo; Goo, Ja Sung; Kim, Yong Yi [Korea Hydro and Nuclera Power Co., Daejeon (Korea, Republic of)

    2012-05-15

    Safe and effective on-the-job training at a nuclear power plant has been gaining its importance in South Korea and in the UAE. As a solution to this, a cyber plant has been developed based on 3D model design data. It allows its users to access equipment and components in a virtual reality without risks or danger of potential radiation exposure and also increases their familiarity with NPP structures. Equipped with navigation functions similar to those of the applications installed in automobiles and smart phones, this application displays the shortest route to reach the target equipment and predicts estimated access time and radiation exposure dose. This application has contributed to the reduction of equipment access time, and therefore has facilitated early response to abnormal conditions, reduced radiation exposure dose, and maximized the effects of OJT at nuclear power plants. This paper will look at the realization of the cyber plant, the operations of the cyber plant, and how cyber plant applications can be applied further

  15. Reduction of Equipment Access Time through Cyber Plant Navigation

    International Nuclear Information System (INIS)

    Suh, Jang Soo; Goo, Ja Sung; Kim, Yong Yi

    2012-01-01

    Safe and effective on-the-job training at a nuclear power plant has been gaining its importance in South Korea and in the UAE. As a solution to this, a cyber plant has been developed based on 3D model design data. It allows its users to access equipment and components in a virtual reality without risks or danger of potential radiation exposure and also increases their familiarity with NPP structures. Equipped with navigation functions similar to those of the applications installed in automobiles and smart phones, this application displays the shortest route to reach the target equipment and predicts estimated access time and radiation exposure dose. This application has contributed to the reduction of equipment access time, and therefore has facilitated early response to abnormal conditions, reduced radiation exposure dose, and maximized the effects of OJT at nuclear power plants. This paper will look at the realization of the cyber plant, the operations of the cyber plant, and how cyber plant applications can be applied further

  16. A guide to qualification of electrical equipment for nuclear power plants. Final report, November 1983

    International Nuclear Information System (INIS)

    Marion, A.; Lamken, D.; Harrall, T.; Kasturi, S.; Holzman, P.; Carfagno, S.; Thompson, D.; Boyer, B.; Hanneman, H.; Rule, W.

    1983-09-01

    Equipment qualification demonstrates that nuclear power plant equipment can perform its safety function - that despite age or the adverse conditions of a design basis accident the equipment can work as needed. This report is a guide to the overall process of electrical equipment qualification. It should interest those who design such equipment, those who buy it, or test it, and even those who install and maintain it. (author)

  17. A proposal of nuclear fusion power plant equipped with SMES

    International Nuclear Information System (INIS)

    Natsukawa, Tatsuya; Makamura, Hirokazu; Molinas, Marta; Nomura, Shinichi; Tsuji-Iio, Shunji; Shimada, Ryuichi

    2000-01-01

    When we intend to operate the nuclear fusion power plant (NFPP) under the economically efficient conditions as an independent power plant, it is desirable that the generated electric power should be sent to network according to the power demand. With such strategy being expanded, some energy storage system is available. In this paper, NFPP equipped with the superconducting magnetic energy storage system (SMES) as electric power storage device is proposed. The advantages of NFPP equipped with SMES are discussed and a case study of 500 MW NFPP equipped with 6 GWh SMES is done with estimating its operational value. For SMES coil, the concept of Force Balanced Coil (FBC) is applied and 6 GWh class FBC is briefly designed

  18. Hot conditioning equipment conceptual design report

    Energy Technology Data Exchange (ETDEWEB)

    Bradshaw, F.W., Westinghouse Hanford

    1996-08-06

    This report documents the conceptual design of the Hot Conditioning System Equipment. The Hot conditioning System will consist of two separate designs: the Hot Conditioning System Equipment; and the Hot Conditioning System Annex. The Hot Conditioning System Equipment Design includes the equipment such as ovens, vacuum pumps, inert gas delivery systems, etc.necessary to condition spent nuclear fuel currently in storage in the K Basins of the Hanford Site. The Hot Conditioning System Annex consists of the facility of house the Hot Conditioning System. The Hot Conditioning System will be housed in an annex to the Canister Storage Building. The Hot Conditioning System will consist of pits in the floor which contain ovens in which the spent nuclear will be conditioned prior to interim storage.

  19. Hot conditioning equipment conceptual design report

    International Nuclear Information System (INIS)

    Bradshaw, F.W.

    1996-01-01

    This report documents the conceptual design of the Hot Conditioning System Equipment. The Hot conditioning System will consist of two separate designs: the Hot Conditioning System Equipment; and the Hot Conditioning System Annex. The Hot Conditioning System Equipment Design includes the equipment such as ovens, vacuum pumps, inert gas delivery systems, etc.necessary to condition spent nuclear fuel currently in storage in the K Basins of the Hanford Site. The Hot Conditioning System Annex consists of the facility of house the Hot Conditioning System. The Hot Conditioning System will be housed in an annex to the Canister Storage Building. The Hot Conditioning System will consist of pits in the floor which contain ovens in which the spent nuclear will be conditioned prior to interim storage

  20. Design by analysis of composite pressure equipment

    International Nuclear Information System (INIS)

    Durand, S.; Mallard, H.

    2004-01-01

    Design by analysis has been particularly pointed out by the european pressure equipment directive. Advanced mechanical analysis like finite element method are used instead of classical design by formulas or graphs. Structural behaviour can be understood by the designer. Design by analysis of metallic pressure equipments is widely used. Material behaviour or limits analysis is based on sophisticated approach (elasto-plastic analysis,..). Design by analysis of composite pressure equipments is not systematically used for industrial products. The difficulty comes from the number of information to handle. The laws of mechanics are the same for composite materials than for steel. The authors want to show that in design by analysis, the composite material approach is only more complete than the metallic approach. Mechanics is more general but not more complicated. A multi-material approach is a natural evolution of design by analysis of composite equipments. The presentation is illustrated by several industrial cases - composite vessel: analogy with metallic calculations; - composite pipes and fittings; - welding and bounding of thermoplastic equipments. (authors)

  1. Electrical System Design Application on Japanese BWR Plants in the Light of the Fukushima Accident and Hitachi Experience of the Solid State Power Equipment in Japanese BWR

    International Nuclear Information System (INIS)

    Sugiyama, Masashi

    2015-01-01

    After the Fukushima Daiichi accident (Loss of all AC and DC power sources and the distribution panels), several design enhancements have been incorporated or are under consideration to Japanese BWRs. Especially, there are several important enhancements in the area of the electrical system design. In this paper, the design enhancements related to the following systems will be introduced. Supplemental AC power supply system Enhancement on DC Battery system In addition, this paper will show our practice of the solid state equipment in Japanese BWRs which have some special specifications, considering the special condition in the NPP's auxiliary electrical power system. (author)

  2. Design quality assurance for nuclear power plants

    International Nuclear Information System (INIS)

    1986-07-01

    This Standard contains the requirements for the quality assurance program applicable to the design phase of a nuclear plant, and is applicable to the design of safety-related equipment, systems, and structures, as identified by the owner. 1 fig

  3. Design quality assurance for nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1986-07-01

    This Standard contains the requirements for the quality assurance program applicable to the design phase of a nuclear plant, and is applicable to the design of safety-related equipment, systems, and structures, as identified by the owner. 1 fig.

  4. Design of a uranium recovery pilot plant

    International Nuclear Information System (INIS)

    1984-01-01

    The engineering design of a pilot plant of uranium recover, is presented. The diagrams and specifications of the equipments such as pipelines, pumps, values tanks, filters, engines, etc... as well as metallic structure and architetonic design is also presented. (author)

  5. Assessment of electrical equipment aging for nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    The electrical and instrumentation equipments, especially whose parts are made of polymer material, are gradually degraded by thermal and radiation environment in the normal operation, and the degradation is thought to progress rapidly when they are exposed to the environment of the design basis event (DBE). The integrity of the equipments is evaluated by the environmental qualification (EQ) test simulating the environment of the normal operation and the DBE. The project of 'Assessment of Cable Aging for Nuclear Power Plants' (ACA, 2002-2008) indicated the importance of applying simultaneous thermal and radiation aging for simulating the aging in normal operation. The project of 'Assessment of Electrical Equipment Aging for Nuclear Power Plants' (AEA) was initiated in FY2008 to apply the outcome of ACA to the other electrical and instrumentation equipment and to establish an advanced EQ test method that can appropriately simulate the environment in actual plants. In FY2012, aging characteristics of thermal aging and simultaneous aging were obtained for the epoxy resin of electrical penetrations and the O-ring of connectors. Physical property measurement was carried out for epoxy resin of electrical penetration subject to the type testing in FY2010. (author)

  6. Assessment of electrical equipment aging for nuclear power plant

    International Nuclear Information System (INIS)

    2013-01-01

    The electrical and instrumentation equipments, especially whose parts are made of polymer material, are gradually degraded by thermal and radiation environment in the normal operation, and the degradation is thought to progress rapidly when they are exposed to the environment of the design basis event (DBE). The integrity of the equipments is evaluated by the environmental qualification (EQ) test simulating the environment of the normal operation and the DBE. The project of 'Assessment of Cable Aging for Nuclear Power Plants' (ACA, 2002-2008) indicated the importance of applying simultaneous thermal and radiation aging for simulating the aging in normal operation. The project of 'Assessment of Electrical Equipment Aging for Nuclear Power Plants' (AEA) was initiated in FY2008 to apply the outcome of ACA to the other electrical and instrumentation equipment and to establish an advanced EQ test method that can appropriately simulate the environment in actual plants. In FY2012, aging characteristics of thermal aging and simultaneous aging were obtained for the epoxy resin of electrical penetrations and the O-ring of connectors. Physical property measurement was carried out for epoxy resin of electrical penetration subject to the type testing in FY2010. (author)

  7. Advanced plant design recommendations from Cook Nuclear Plant experience

    International Nuclear Information System (INIS)

    Zimmerman, W.L.

    1993-01-01

    A project in the American Electric Power Service Corporation to review operating and maintenance experience at Cook Nuclear Plant to identify recommendations for advanced nuclear plant design is described. Recommendations so gathered in the areas of plant fluid systems, instrument and control, testing and surveillance provisions, plant layout of equipment, provisions to enhance effective maintenance, ventilation systems, radiological protection, and construction, are presented accordingly. An example for a design review checklist for effective plant operations and maintenance is suggested

  8. Increasing efficiency and optimizing thermoelectric power plant equipment. Povyshenie effektivnosti i optimizatsiia teploenergeticheskikh ustanovok

    Energy Technology Data Exchange (ETDEWEB)

    Andriushchenko, A.I.

    1981-01-01

    The problems of increasing the efficiency and optimizing the operational conditions of a thermoelectric power plant and providing efficient operational conditions of the primary and auxillary equipment at a thermoelectric power plant are examined. Methodologies and designs for optimizing the primary parameters of the power-generating equipment based on economic factors are given. A number of recommendations for designing equipment based on the research results are given.

  9. Design of chemical plant

    International Nuclear Information System (INIS)

    Lee, Dong Il; Kim, Seung Jae; Yang, Jae Ho; Ryu, Hwa Won

    1993-01-01

    This book describes design of chemical plant, which includes chemical engineer and plan for chemical plant, development of chemical process, cost engineering pattern, design and process development, general plant construction plan, project engineering, foundation for economy on assets and depreciation, estimation for cost on capital investment and manufacturing cost, design with computers optimal design and method like fluid mechanics design chemical device and estimation for cost, such as dispatch of material and device writing on design report and appendixes.

  10. Prototype equipment status monitor for plant operational configuration management

    International Nuclear Information System (INIS)

    DeVerno, M.; Trask, D.; Groom, S.

    1998-01-01

    CANDU plants, such as the Point Lepreau GS, have tens of thousands of operable devices. The status of each operable device must be immediately available to plan and execute future changes to the plant. Historically, changes to the plant's operational configuration have been controlled using manual and administrative methods where the status of each operable device is maintained on operational flowsheets located in the work control area of the main control room. The operational flowsheets are used to plan and develop Operating Orders (OOs) or Order-to-Operate (OTOs) and the control centre work processes are used to manage their execution. After performing each OO procedure, the operational flowsheets are updated to reflect the new plant configuration. This process can be very time consuming, and due to the manual processes, can lead to the potential for time lags and errors in the recording of the current plant configuration. Through a cooperative research and development program, Canadian CANDU utilities and Atomic Energy of Canada Limited, the design organization, have applied modern information technologies to develop a prototype Equipment Status Monitor (ESM) to address processes and information flow for efficient operational configuration management. The ESM integrates electronic operational flowsheets, equipment databases, engineering and work management systems, and computerized procedures to assess, plan, execute, track, and record changes to the plant's operational configuration. This directly leads to improved change control, more timely and accurate plant status information, fewer errors, and better decision making regarding future changes. These improvements to managing the plant's operational configuration are essential to increasing plant safety, achieving a high plant availability, and maintaining high capability and capacity factors. (author)

  11. The establish and application of equipment reliability database in Nuclear Power Plant

    International Nuclear Information System (INIS)

    Zheng Wei; Li He

    2006-03-01

    Take the case of Daya Bay Nuclear Power Plant, the collecting and handling of equipment reliability data, the calculation method of reliability parameters and the establish and application of reliability databases, etc. are discussed. The data source involved the design information of the equipment, the operation information, the maintenance information and periodically test record, etc. Equipment reliability database built on a base of the operation experience. It provided the valid tool for thoroughly and objectively recording the operation history and the present condition of various equipment of the plant; supervising the appearance of the equipment, especially the safety-related equipment, provided the very practical worth information for enhancing the safety and availability management of the equipment and insuring the safety and economic operation of the plant; and provided the essential data for the research and applications in safety management, reliability analysis, probabilistic safety assessment, reliability centered maintenance and economic management in nuclear power plant. (authors)

  12. Broader utilization of programmable automation equipment in French nuclear power plants: Reflections on the choices made by Electricite de France and French designers

    International Nuclear Information System (INIS)

    Baudry, Y.; Varaldi, G.

    1983-01-01

    More than 1000 microprocessors and more than 10,000 data memories in each of the twenty or so 1300 MW units in the French nuclear programme: that was the decision taken by Electricite de France (EDF) in conjunction with the designers in 1974, with the intention of introducing programmable automata on a wide scale in French nuclear power plants. This programme was carried out with the assistance of advanced research services such as the universities, the Commissariat a l'energie atomique (CEA), EDF's design and research service and the designers, most of whom were already EDF suppliers for the 900 MW range. Having used computers for linking sequences (themselves carried out with electromagnetic technology) for its latest natural-uranium gas-cooled graphite-moderated power plants, EDF decided to call a temporary halt, in the case of its 900 MW light-water range, to the use of digital techniques for the control and automation of power plants although it continued to employ such techniques widely in data processing. Thus, the widespread introduction of programmable automata, which was decided upon in 1974/75 at a time when no equivalent existed at the international level, led EDF and French designers to undertake a major development effort in order to meet the requirements - particularly safety and reliability requirements - for such automata to be incorporated into the nuclear field. How does this choice fit in with the logical evolution of the digitalization of French nuclear power plants. What problems has it caused for EDF and French industry. How have these problems been tackled. How have they been overcome. These are the questions dealt with in this paper. (author)

  13. Field vibration test of principal equipment of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Shiraki, Kazuhiro; Fujita, Katsuhisa; Kajimura, Motohiko; Ikegami, Yasuhiko; Hanzawa, Katsumi; Sakai, Yoshiyuki; Kokubo, Eiji; Igarashi, Shigeru

    1984-09-01

    Japan is one of the most earthquake-stricken countries in the world, and demands for aseismic design have become severer recently. In a nuclear power plant in particular, consisting of a reactor vessel and other facilities dealing with a radioactive substance in some form or other, it is essential from the standpoint of safety to eliminate any possibility of radioactive hazards for the local public, and the employees at the plant as well, if these facilities are struck by an earthquake. This paper is related to the reactor vessel, reactor primary cooling equipment and piping system and important general piping as examples of important facilities of a nuclear power plant, and discusses vibration tests of an actual plant in the field from the standpoint of enhancing the aseismic safety of the Mitsubishi PWR nuclear power plant. Especially concerning vibration test technology, the effects in the evaluation of aseismic safety and its limits are studied to prove how it contributes to the enhancement of the reliability of aseismic design of nuclear power plants.

  14. Performance test of nutrient control equipment for hydroponic plants

    Science.gov (United States)

    Rahman, Nurhaidar; Kuala, S. I.; Tribowo, R. I.; Anggara, C. E. W.; Susanti, N. D.

    2017-11-01

    Automatic control equipment has been made for the nutrient content in irrigation water for hydroponic plants. Automatic control equipment with CCT53200E conductivity controller to nutrient content in irrigation water for hydroponic plants, can be used to control the amount of TDS of nutrient solution in the range of TDS numbers that can be set according to the range of TDS requirements for the growth of hydroponically cultivated crops. This equipment can minimize the work time of hydroponic crop cultivators. The equipment measurement range is set between 1260 ppm up to 1610 ppm for spinach plants. Caisim plants were included in this experiment along with spinach plants with a spinach plants TDS range. The average of TDS device is 1450 ppm, while manual (conventional) is 1610 ppm. Nutrient solution in TDS controller has pH 5,5 and temperature 29,2 °C, while manual is pH 5,6 and temperature 31,3 °C. Manually treatment to hydroponic plant crop, yields in an average of 39.6 grams/plant, greater than the yield of spinach plants with TDS control equipment, which is in an average of 24.6 grams / plant. The yield of caisim plants by manual treatment is in an average of 32.3 grams/crop, less than caisim crop yields with TDS control equipment, which is in an average of 49.4 grams/plant.

  15. Response of equipment in nuclear power plants to airplane crash

    International Nuclear Information System (INIS)

    Schalk, M.; Woelfel, H.

    1976-01-01

    Nuclear power plants in Germany are to be designed against airplane crash. Two problems arise: first, the local problem of penetration as well as local destruction of the building and secondly the airplane induced vibrations of the whole building which cause loadings for secondary systems (equipment). This paper deals especially with the second problem. Floor response spectra due to airplane crash are presented for two different power plant buildings. The influence of various parameters (time history of excitation, direction and location of impact, mathematical model, soil, damping, etc.) are discussed. A comparison with the results of earthquake loading is given. Suggestions are made for developing suitable floor design spectra and using them to analyse multidegree-of-freedom systems. However, the paper gives only a partial answer to the questions arising because of some important restrictions which had to be made. Studies concerning these restrictions are still being conducted and will be presented in a separate paper. (Auth.)

  16. Plutonium finishing plant safety systems and equipment list

    International Nuclear Information System (INIS)

    Bergquist, G.G.

    1995-01-01

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

  17. Low activation diagnostic equipment design studies

    International Nuclear Information System (INIS)

    Hopkins, G.R.; Cheng, E.T.; Fisher, R.K.

    1985-01-01

    The low activation fusion concept has been applied to the diagnostic equipment in a fusion reactor. The components where fabrication from low activation materials is feasible have been identified. Other systems where higher activation elements are required can have their activation reduced by design approaches which include shielding and operation only in low flux regions of the reactor. Some components will not operate in a high flux so activation is not a major concern. This low activation diagnostic equipment study completes a series of low activation studies where all the components in a fusion power reactor have now been evaluated. It is concluded that a completely low activation fusion reactor is feasible with all components meeting the functional requirements. This provides an environmentally benign energy source with a high confidence level in meeting safety criteria in operation, maintenance and waste disposal

  18. Localization of equipment for digital plant protection system

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-10-01

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

  19. Localization of equipment for digital plant protection system

    International Nuclear Information System (INIS)

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

    2000-10-01

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

  20. Planning of maintenance of electrical equipment in nuclear plants/laboratories [Paper No.: VB-3

    International Nuclear Information System (INIS)

    Narasinga Rao, S.N.; Bhattacharyya, A.K.

    1981-01-01

    Satisfactory operating performance of electrical systems ensures continuous availability of power to the various plants and machinery in nuclear plant and laboratories. For effective optimal functioning of the electrical equipment and to reduce their down time, scheduled planning of maintenance to the equipment is essential. Maintenance of power plant, nuclear or fossil, and industrial plant and research laboratories demands essential ingredients such as right type of trained and motivated technical personnel, adoption of standard procedures for maintenance, adequate safety and protection for equipment, safety procedures adopted in the installation to prevent hazards to the workers, provision of adequate stores and inventories, facilities for quick repairs and testing of equipment and effective planning of procedures for their maintenance. While breakdown maintenance allows equipment to operate before it is repaired or replaced, preventive maintenance makes use of scheduled inspection and periodical equipment overhaul and has little value for predicting future continuous performances of equipment. The engineered maintenance is most advantageous and offers maximum operating time to reduce down time of the equipment while adding predictive testing technique to aid in determining the frequency of overhaul of equipment. The important checks to be conducted and preventive maintenance programme to be scheduled are discussed in this paper. The safety and reliable functioning of the electrical equipment depend on proper optimal design, selection of equipment, their installation, subsequent maintenance and strict compliance with safety regulations. (author)

  1. Plant design and beam utilization

    International Nuclear Information System (INIS)

    Svendsen, E.B.

    1983-01-01

    Plant design and beam utilization are two things closely tied together: without a proper plant design, one can never get good beam utilization. When a company decides to build an irradiation facility, there are some major decisions to be made right in the beginning. These decisions can be most important for the long-term success or failure of the irradiation facility, because the company normally will have to live with these decisions during the whole life-time of the irradiation equipment. To start with the decision has to be made whether to select a cobalt-60 irradiation plant or an accelerator irradiation plant. This decision can only be reached after a careful study of the products and the 'weight' and the material of the products the company wants to irradiate. As an old accelerator-man, I tend to personally favor accelerators, although I am very impressed by the newer cobalt-60 pallet irradiation plants from A.E.C.L. I believe that they have a great future in the emerging field of food irradiation. As I have primarily been involved with accelerators during the last 14 years, this paper is only dealing with different design approaches and utilizations of accelerator-plants. (author)

  2. 3. General principles of assessing seismic resistance of technological equipment of nuclear power plants

    International Nuclear Information System (INIS)

    1983-01-01

    The evaluation of the seismic resistance of technological equipment is performed by computation, experimental trial, possibly by combining both methods. Existing and prepared standards in the field of seismic resistance of nuclear power plants are mentioned. Accelerograms and response spectra of design-basis earhtquake and maximum credible earthquake serve as the basic data for evaluating seismic resistance. The nuclear power plant in Mochovce will be the first Czechoslovak nuclear power plant with so-called partially seismic design. The problem of dynamic interaction of technological equipment and nuclear power plant systems with a bearing structure is discussed. (E.F.)

  3. The gravitational plant physiology facility-Description of equipment developed for biological research in spacelab

    Science.gov (United States)

    Heathcote, D. G.; Chapman, D. K.; Brown, A. H.; Lewis, R. F.

    1994-01-01

    In January 1992, the NASA Suttle mission STS 42 carried a facility designed to perform experiments on plant gravi- and photo-tropic responses. This equipment, the Gravitational Plant Physiology Facility (GPPF) was made up of a number of interconnected units mounted within a Spacelab double rack. The details of these units and the plant growth containers designed for use in GPPF are described. The equipment functioned well during the mission and returned a substantial body of time-lapse video data on plant responses to tropistic stimuli under conditions of orbital microgravity. GPPF is maintained by NASA Ames Research Center, and is flight qualifiable for future spacelab missions.

  4. Development of fragility descriptions of equipment for seismic risk assessment of nuclear power plants

    International Nuclear Information System (INIS)

    Hardy, G.S.; Campbell, R.D.

    1983-01-01

    Probabilistic risk assessment (PRA) of a nuclear power plant for postulated hazard requires the development of fragility relationships for the plants' safety related equipment. The objective of this paper is to present some general results and conclusions concerning the development of these seismic fragility levels. Participation in fragility-related research and experience gained from the completion of several PRA studies of a variety of nuclear power plants have provided much insight as to the most vulnerable equipment and the most efficient use of resources for development of fragilities. Plants studied had seismic design bases ranging from very simple equivalent static analysis for some of the earlier plants to state-of-the-art complex multimode dyanamic analyses for plants currently under construction. Increased sophistication and rigor in seismic qualification of equipment has resulted for the most part in increased seismic resistance. The majority of equipment has been found, however, to possess more than adequate resistance to seismic loading regardless of the degree of sophistication utilized in design as long as seismic loading was included in the design process. This paper presents conclusions of the authors as to which items of equipment typically require an individual ''plant-specific'' fragility analysis and which can be treated in a generic fashion. In addition, general conclusions on the relative seismic capacity levels and most frequent failure modes are summarized for generic equipment groups

  5. IAS 16 Property, Plant and Equipment - A Closer Look

    OpenAIRE

    Muthupandian, K S

    2009-01-01

    The International Accounting Standards Committee issued the the International Accounting Standard 16 Property, Plant and Equipment. The objective of IAS 16 is to prescribe the accounting treatment for Property, Plant and Equipment (PPE) so that users of the financial statements can discern information about an entity's investment in its PPE and the changes in such investment. The principal issues in accounting for PPE are the recognition of the assets, the determination of their carrying amou...

  6. Diagnostic technology of PWR plant equipment failures

    International Nuclear Information System (INIS)

    Nakamura, Tetsuo; Tanaka, Mamoru; Okamachi, Masao; Taguchi, Shozo; Nagashima, Kazuhiro; Ishikawa, Satoshi

    1985-01-01

    To confirm the soundness of the important facilities in a nuclear power plant contributes to the reliability of the plant operations and improvement of its operation rate. For this purpose, the following diagnostic techniques have been developed. (1). Vibration and loose parts monitoring: Detection of abnormal structural vibrations in the reactor, estimation of its mode, detection of loose parts in the primary system, and estimation of the position and energy of their collisions against the reactor vessel or the like. (2). Valve leak monitoring: Detection of leaks from primary valves in the primary cooling boundary, such as the pressurizer relief valve and safety valve, and estimation of the form of the leaks. (3). Detector noise response diagnosis: Diagnosis of degradation of principal process detectors during plant operation. Furthermore, a diagnostic system incorporating the above diagnostic technology applicable to actual plants has been experimentally manufactured and successfully verified. (author)

  7. On fundamental concept of anti-earthquake design of equipment and pipings

    International Nuclear Information System (INIS)

    Shibata, H.; Kato, M.

    1979-01-01

    This paper deals with a new concept of anti-earthquake design of equipment and pipings in nuclear power plants. Usual anti-earthquake design of such items starts from the design basis ground motions, via floor responses and ends at the stress analysis of each structural element. However, the same type of equipment are used for plants under various site conditions. The ordinarily used method obliges the repetition of such design procedure on each plant. This new design method has been developed to avoid such time-consuming repetitions. (orig.)

  8. The application of GIS equipment in nuclear power plant

    International Nuclear Information System (INIS)

    Ji Lin; Huang Pengbo; Chang Xin'ai

    2012-01-01

    In this paper, the advantage and disadvantage of gas insulated switchgear (GIS) in environmental adaptability, operation safety and economic benefit are analyzed. Issues concerning the manufacture, transportation, on-site installation, operation, maintenance and extension of GIS equipment are discussed. Comparing those characteristics with air insulated switchgear (AIS), GIS is characterized by better aseismic ability, less occupied area and installation process, lower fault rate, longer maintenance period, easier for extension and higher economic benefit, SF6 gas insures the operation safety and reliability of GIS equipment, modular transport and re-assembling improves the installation flexibility. Therefore, GIS equipment may be the first choice for the primary equipment of nuclear power plant. (authors)

  9. Design and construction rules for electrical equipments of nuclear islands. 3. ed.

    International Nuclear Information System (INIS)

    1993-01-01

    The French design and construction for electrical equipments of nuclear islands (RCC-E) deals with equipments of which the failure may have consequences on the safety of persons or appreciable effects on the availability of the power plant. This book presents the rules concerning (1) qualification procedure, (2) the design of functional units, (3) the installation (environmental conditions, protection rules), (4) the constituent elements of equipments, (5) the control and testing methods

  10. Equipment specifications for an electrochemical fuel reprocessing plant

    International Nuclear Information System (INIS)

    Hemphill, Kevin P.

    2010-01-01

    Electrochemical reprocessing is a technique used to chemically separate and dissolve the components of spent nuclear fuel, in order to produce new metal fuel. There are several different variations to electrochemical reprocessing. These variations are accounted for by both the production of different types of spent nuclear fuel, as well as different states and organizations doing research in the field. For this electrochemical reprocessing plant, the spent fuel will be in the metallurgical form, a product of fast breeder reactors, which are used in many nuclear power plants. The equipment line for this process is divided into two main categories, the fuel refining equipment and the fuel fabrication equipment. The fuel refining equipment is responsible for separating out the plutonium and uranium together, while getting rid of the minor transuranic elements and fission products. The fuel fabrication equipment will then convert this plutonium and uranium mixture into readily usable metal fuel.

  11. Studies for aseismatic design of atomic power equipment

    International Nuclear Information System (INIS)

    Uchiyama, Yuichi; Takayanagi, Masaaki; Moriyama, Takeo; Mizuno, Sadao

    1976-01-01

    The social requirements for the safety of nuclear power stations are very severe, and especially in Japan where earthquakes occur frequently, the safety against earthquakes is regarded as an important matter. The numerous machinery and equipments composing nuclear power plants have largely different roles in the safety, accordingly it is important to maintain effectively the safety as a whole. The design is carried out by classifying buildings, machinery and equipments, pipings and electric appliances according to the importance, and so that the aseismatic property corresponding to the importance is given to each. In most cases, the aseismatic design is carried out with computing codes, but it is necessary to forward the design by proving the propriety of analysis models and calculated results. Efforts are exerted in Hitachi Ltd. to improve the accuracy of the aseismatic analysis by carrying out basic experiment, the demonstration test in actual plants and the study on analytical method as many as possible. The measurement of the vibration of actual pipings in order to confirm the supporting conditions, the comparison of multi-input analysis and SRSS method for piping analysis, and the vibration tests of a model containment vessel and a fuel assembly were carried out, and the results are reported in this paper. It is important to improve the accuracy of earthquake response analysis further. (Kako, I.)

  12. 48 CFR 245.608-71 - Screening industrial plant equipment.

    Science.gov (United States)

    2010-10-01

    ... 48 Federal Acquisition Regulations System 3 2010-10-01 2010-10-01 false Screening industrial plant..., and Disposal of Contractor Inventory 245.608-71 Screening industrial plant equipment. (a) Reporting...) After 90th day. If DoD requirement is identified, and item is available, ship item against the...

  13. Method for keeping equipment and pipeline of nuclear power plant

    International Nuclear Information System (INIS)

    Okubo, Osamu.

    1990-01-01

    The present invention intends to suppress corrosion of equipments and pipelines in condensate, feedwater and feedwater heater drain systems during operation of a nuclear power plant. That is, condensate, feedwater and drain remained in equipments and pipelines just after the stopping of operation are passed through pipelines comprising only conduits, or they are introduced to a condensator passing through the pipelines and condensate pipes. Further, the remaining droplets on the inner surfaces are evaporated by the remaining heat of the equipments and the pipelines themselves. Then, the equipments and pipelines are isolated from other regions and kept. In view of the above, since condensate, feedwater and water feeder drains are introduced directly to the condensator passing through the conduits in which other equipments such as tanks and pumps are not present and are isolated and kept, corrosion of the equipments and the pipelines is suppressed and radioactive contamination is suppressed from prevailing by way of cruds. (I.S.)

  14. Earthquake protection of nuclear power plant equipment

    Energy Technology Data Exchange (ETDEWEB)

    Nawrotzki, Peter [GERB Vibration Control Systems, Berlin (Germany)

    2010-05-15

    Power plant machinery can be dynamically decoupled from the substructure by the effective use of helical steel springs and viscous dampers. Turbine foundations, boiler feed pumps and other machine foundations benefit from this type of elastic support systems to mitigate the transmission of operational vibration. The application of these devices may also be used to protect against earthquakes and other catastrophic events, i.e. airplane crash, of particular importance in nuclear facilities. This article illustrates basic principles of elastic support systems and applications on power plant buildings in medium and high seismic areas. Spring-damper combinations with special stiffness properties are used to reduce seismic acceleration levels of turbine components and other safety or non-safety related structures. For turbine buildings, the integration of the turbine substructure into the machine building can further reduce stress levels in all structural members. (orig.)

  15. Earthquake protection of nuclear power plant equipment

    International Nuclear Information System (INIS)

    Nawrotzki, Peter

    2010-01-01

    Power plant machinery can be dynamically decoupled from the substructure by the effective use of helical steel springs and viscous dampers. Turbine foundations, boiler feed pumps and other machine foundations benefit from this type of elastic support systems to mitigate the transmission of operational vibration. The application of these devices may also be used to protect against earthquakes and other catastrophic events, i.e. airplane crash, of particular importance in nuclear facilities. This article illustrates basic principles of elastic support systems and applications on power plant buildings in medium and high seismic areas. Spring-damper combinations with special stiffness properties are used to reduce seismic acceleration levels of turbine components and other safety or non-safety related structures. For turbine buildings, the integration of the turbine substructure into the machine building can further reduce stress levels in all structural members. (orig.)

  16. Unresolved Safety Issue A-46 - seismic qualification of equipment in operating plants

    International Nuclear Information System (INIS)

    Anderson, N.

    1985-01-01

    Seismic Qualification of Equipment in Operating Plants was designated as an Unresolved Safety Issue (USI) in December, 1980. The USI A-46 program was developed in early 1981 to investigate the adequacy of mechanical and electrical equipment in operating plants to withstand a safe shutdown earthquake. The approach taken was to develop viable, cost effective alternatives to current seismic qualification licensing requirements which could be applied to operating nuclear power plants. The tasks investigated include: (1) identification of seismic sensitive systems and equipment; (2) assessment of adequacy of existing seismic qualification methods; (3) development and assessment of in-situ test procedures to assist in qualification of equipment; (4) seismic qualification of equipment using seismic experience data; and (5) development of methods to generate generic floor response spectra. Progress to date and plans for completion of resolution are reported

  17. Reliability of nuclear power plants and equipment

    International Nuclear Information System (INIS)

    1985-01-01

    The standard sets the general principles, a list of reliability indexes and demands on their selection. Reliability indexes of nuclear power plants include the simple indexes of fail-safe operation, life and maintainability, and of storage capability. All terms and notions are explained and methods of evaluating the indexes briefly listed - statistical, and calculation experimental. The dates when the standard comes in force in the individual CMEA countries are given. (M.D.)

  18. ITER L 7 duct remote handling equipment design report

    International Nuclear Information System (INIS)

    Millard, J.

    1996-09-01

    The operation, design and interfaces of the 'Duct Vehicle' and it's associated remote handling equipment are briefly described in this document. This equipment is being designed by Spar Aerospace Ltd. for the Divertor Test Platform as part of ITER Research and Development Project L-7. Canadian Fusion Fuels Technology Project funds this work as part of the Canadian Contribution to ITER. This document describes the equipment design status at the September 1996 design review. 23 figs

  19. 1E Qualification of Electrical Equipment - Requirement for Safety Nuclear Power Plants

    International Nuclear Information System (INIS)

    Geambasu, C.; Segarceanu, D.; Albu, J.

    2002-01-01

    The paper presents the qualification methods of the safety related equipment according to the safety class 1E. There are presented the qualification principles, procedure and documents, emphasis being laid on the qualification approach by type tests. This approach assumes the equipment test under both normal and accident conditions (design basis events) simulating the operational conditions and covers the largest part of electrical equipment from a nuclear power plant.The safety related equipment is to be qualified is subjected to a sequential test that will be detailed in the paper. (author)

  20. Application of Equipment Monitoring Technology in Nuclear Power Plants

    International Nuclear Information System (INIS)

    Kang, H. T.; Lee, J. K.; Lee, K. D.; Jo, S. H.

    2012-01-01

    The major goal of nuclear power industries during the past 10 years is to increase reliability and utility capacity factor. As the capacitor factor, however, crept upward. it became harder to attain next percentage of improvement. Therefore other innovative technologies are required. By the technologies applied to the fossil power plants, equipment health monitoring was performed on equipment to maintain it in operable condition and contributed on improving their reliability a lot. But the equipment monitoring may be limited to the observation of current system states in nuclear power plant. Monitoring of current system states is being augmented with prediction of future operating states and predictive diagnosis of future failure states. Such predictive diagnosis is motivated by the need for nuclear power plants to optimize equipment performance and reduce costs and unscheduled downtime. This paper reviews the application of techniques that focus on improving reliability in nuclear power plant by monitoring and predicting equipment health and suggests how possible to support on-line monitoring

  1. Impact of power uprate on environmental qualification of equipment in nuclear power plants

    International Nuclear Information System (INIS)

    Raheja, R.D.; Mohiuddin, A.; Alsammarae, A.

    1996-01-01

    Many nuclear power facilities are finding it economically beneficial to increase reactor output, from operating plants, by resorting to power uprates. A power uprate implies that a utility can increase the reactor output, or the megawatts generated, by increasing steam pressure without adding or changing any plant systems. This is perhaps one of the least expensive options for increasing the generating capacity of a power plant. However, a nuclear plant requires a comprehensive review of the plant systems, structures and components to assure their capability to withstand the resulting increased normal and accident plant conditions. A power uprate will typically result in a plant operating at higher than the originally designed environmental conditions. Safety related equipment in nuclear plants is presently qualified to the UFSAR Chapter 15 accident events and the resulting temperatures, pressures, radiation levels etc. These values will increase when the reactor is producing a higher MWe output. Components that are sensitive to the environment must be re-evaluated and assessed to determine their acceptability and operability under the revised environmental conditions. Most safety-related mechanical and electrical equipment will require an assessment from an environmental qualification standpoint. Utilities must perform this task in a systematic, auditable and cost effective manner to optimize their resources and minimize plant costs associated with modifications, replacements or equipment testing. This paper discusses various approaches and provides recommendations to achieve equipment qualification while satisfying the plant's objective of a power uprate

  2. Recommendations for managing equipment aging in nuclear power plants

    International Nuclear Information System (INIS)

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

    1992-01-01

    Research conducted under the auspices of the US NRC's Nuclear Plant Aging Research (NPAR) Program has resulted in a large database of component and system operating, maintenance, and testing information. This database has been used to determine the susceptibility to aging of selected components, and the potential for equipment aging to impact plant safety and availability. it has also identified methods for detecting and mitigating component and system aging. This paper describes the research recommendations on electrical components which could be applied to maintenance, testing, and inspection activities to detect and mitigate the effects of aging prior to equipment failures

  3. Virtual environments for nuclear power plant design

    International Nuclear Information System (INIS)

    Brown-VanHoozer, S.A.; Singleterry, R.C. Jr.; King, R.W.

    1996-01-01

    In the design and operation of nuclear power plants, the visualization process inherent in virtual environments (VE) allows for abstract design concepts to be made concrete and simulated without using a physical mock-up. This helps reduce the time and effort required to design and understand the system, thus providing the design team with a less complicated arrangement. Also, the outcome of human interactions with the components and system can be minimized through various testing of scenarios in real-time without the threat of injury to the user or damage to the equipment. If implemented, this will lead to a minimal total design and construction effort for nuclear power plants (NPP)

  4. Conceptual design report, CEBAF basic experimental equipment

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1990-04-13

    The Continuous Electron Beam Accelerator Facility (CEBAF) will be dedicated to basic research in Nuclear Physics using electrons and photons as projectiles. The accelerator configuration allows three nearly continuous beams to be delivered simultaneously in three experimental halls, which will be equipped with complementary sets of instruments: Hall A--two high resolution magnetic spectrometers; Hall B--a large acceptance magnetic spectrometer; Hall C--a high-momentum, moderate resolution, magnetic spectrometer and a variety of more dedicated instruments. This report contains a short description of the initial complement of experimental equipment to be installed in each of the three halls.

  5. Identification and Assessment of Human Error Due to design in damagingto the Sour Water Equipment and SRP Unit of Control Room in A Refinery Plant using SHERPA Technique

    Directory of Open Access Journals (Sweden)

    2013-02-01

    Conclusion: To prevent and control occurring each of the identified errors and to limit the consequences of them, appropriate counter measures such as proper control measures in the form of changes in design, including install the appropriate colored tag, digital indicator and warning lights which must be used base on the kind of system consequently, of this study showed that SHEPA can be an efficientmethod to study humanness in operational site.

  6. 48 CFR 945.407 - Non-Government use of plant equipment.

    Science.gov (United States)

    2010-10-01

    ... plant equipment. 945.407 Section 945.407 Federal Acquisition Regulations System DEPARTMENT OF ENERGY...-Government use of plant equipment. The type of plant equipment and dollar threshold for non-Government use of DOE plant equipment will be determined by the Head of the Contracting Activity which awarded the...

  7. Seismic fragility levels of nuclear power plant equipment

    International Nuclear Information System (INIS)

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

    1987-01-01

    Seismic fragility levels of safety-related electrical and mechanical equipment used in nuclear power plants are discussed. The fragility level is defined as the vibration level corresponding to initiation of equipment malfunctions. The test response spectrum is used as a measure of this vibration level. The fragility phenomenon of an equipment is represented by a number of response spectra corresponding to various failure modes. Analysis methods are described for determination of the fragility level by use of existing test data. Useful conversion factors are tabulated to transform test response spectra from one damping value to another. Results are presented for switch-gears and motor control centers. The capacity levels of these equipment assemblies are observed to be limited by malfunctioning of contactors, motor starters, relays and/or switches. The applicability of the fragility levels, determined in terms of test response spectra, to Seismic Margin Studies and Probabilistic Risk Assessments is discussed and specific recommendations are provided

  8. Implementation of the project of equipment reliability in the nuclear power plant of Laguna Verde

    International Nuclear Information System (INIS)

    Rios O, J. E.; Martinez L, A. G.

    2008-01-01

    A equipment is reliable if it fulfills the function for which was designed and when it is required. To implement a project of reliability in a nuclear power plant this associate to a process of continuous analysis of the operation, of the conditions and faults of the equipment. The analysis of the operation of a system, of the equipment of the same faults and the parts that integrate to equipment take to identify the potential causes of faults. The predictive analysis on components and equipment allow to rectify and to establish guides to optimize the maintenance and to guarantee the reliability and function of the same ones. The reliability in the equipment is without place to doubts a wide project that embraces from the more small component of the equipment going by the proof of the parts of reserve, the operation conditions until the operative techniques of analysis. Without place of doubt for a nuclear power plant the taking of decisions based on the reliability of their systems and equipment will be the appropriate for to assure the operation and reliability of the same one. In this work would appear the project of reliability its processes, criteria, indicators action of improvement and the interaction of the different disciplines from the Nuclear Power Plant of Laguna Verde like a fundamental point for it put in operation. (Author)

  9. Priorities in the design of chemical shops at coke plants

    Energy Technology Data Exchange (ETDEWEB)

    V.I. Rudyka; Y.E. Zingerman; V.V. Grabko; L.A. Kazak [Giprokoks, the State Institute for the Design of Coke-Industry Enterprises, Kharkov (Ukraine)

    2009-07-15

    Recent trends in the design of chemical equipment at coke plants are described, through the lens of experience at Giprokoks. The main priorities were to improve the removal of impurities from coke oven gas; to improve equipment design on the basis of new materials; to reduce reagent consumption; to reduce the materials and energy consumed in the construction of new equipment; and to minimize impacts on the environment and worker health. Some technological equipment is briefly characterized.

  10. Electrical and control equipment in nuclear power plants. Problems when replacing aging equipment

    International Nuclear Information System (INIS)

    Nordling, Anna; Haakansson, Goeran

    2012-01-01

    Interoperability between different technical systems is more complicated when old and new technology meet, such as between analog and digital technology. New electrical and I and C equipment is selected with consideration to simplify and improve the compatibility and interoperability. The original construction of nuclear power plants with electricity and I and C equipment had more natural interfaces. Generally experienced guidance, to the management of interoperability and interfaces, feels insufficient. Skills transfer programs are identified as a major need, as more and more important personnel are retiring and important information is lost with them. Lack of appropriate skills directly affects the ability to produce accurate and complete requirements specification. Failure modes of newer electrical and I and C equipment are perceived as more complex than the older equipment. When choosing equipment, attempts are made to minimize unnecessary features, to reduce the number of potential failure modes. There is a lack of consistent understanding of the meaning of robustness in electrical technology and I and C technology, in the nuclear plant engineering departments. The overall picture is that the robustness has worsened since the facilities were built. The Swedish nuclear power plants have an internal organizational structure with separated client and support organization. This splits the nuclear organization into two distinct parts which threaten to separate the two entities focus. Engineering departments at the Swedish nuclear power plants express a need for increased expertise in the client organization (blocks). Competence requested is for example, system knowledge to facilitate and enhance the quality of the initial analysis performed in the blocks. Suppliers receive more recently larger turnkey projects, both to minimize costs but also to minimize the interfaces and co-function problems. This, however, heightens demands for knowledge transfer between

  11. DU-AGG pilot plant design study

    International Nuclear Information System (INIS)

    Lessing, P.A.; Gillman, H.

    1996-07-01

    The Idaho National Engineering Laboratory (INEL) is developing new methods to produce high-density aggregate (artificial rock) primarily consisting of depleted uranium oxide. The objective is to develop a low-cost method whereby uranium oxide powder (UO[sub 2], U[sub 3]O[sub ]8, or UO[sub 3]) can be processed to produce high-density aggregate pieces (DU-AGG) having physical properties suitable for disposal in low-level radioactive disposal facilities or for use as a component of high-density concrete used as shielding for radioactive materials. A commercial company, G-M Systems, conducted a design study for a manufacturing pilot plant to process DU-AGG. The results of that study are included and summarized in this report. Also explained are design considerations, equipment capacities, the equipment list, system operation, layout of equipment in the plant, cost estimates, and the proposed plan and schedule

  12. Maintainability design criteria for packaging of spacecraft replaceable electronic equipment.

    Science.gov (United States)

    Kappler, J. R.; Folsom, A. B.

    1972-01-01

    Maintainability must be designed into long-duration spacecraft and equipment to provide the required high probability of mission success with the least cost and weight. The ability to perform repairs quickly and easily in a space environment can be achieved by imposing specific maintainability design criteria on spacecraft equipment design and installation. A study was funded to investigate and define design criteria for electronic equipment that would permit rapid removal and replacement in a space environment. The results of the study are discussed together with subsequent simulated zero-g demonstration tests of a mockup with new concepts for packaging.

  13. Plant equipment integrity monitoring and diagnosing method and device therefor, plant equipment maintenance and inspection time determining method and device therefor, as well as nuclear power plant

    International Nuclear Information System (INIS)

    Kato, Takahiko; Ando, Masashi; Osumi, Katsumi; Horiuchi, Tetsuo; Asakura, Yamato; Akamine, Kazuhiko.

    1995-01-01

    The present invention can accurately forecast a time for occurrence of troubles of plant equipments in contact with recycling water, to conduct its maintenance and inspection before occurrence of the troubles. Namely, change of water quality in plant equipments caused by corrosion of recycling water occurred in constitutional parts of the plant equipments is measured. The time upon occurrence of the troubles of the plant equipments to corrosion of the recycling water is forecast based on the measured value. A time till the occurrence of the change of water quality after starting the use of the plant equipments is calculated based on the measured value. The calculated time is compared with a correlation between the time of occurrence of the troubles after starting the use of the plant equipments and the time of occurrence of change of the water quality, to forecast the time of occurrence of the troubles. Preferably, electroconductivity and pH of recycling water in the inside or at the exit of the plant equipments are measured as an object for the measurement of change of water quality. (I.S.)

  14. Inelastic design of nuclear reactor structures and its implications on design of critical equipment

    International Nuclear Information System (INIS)

    Newmark, N.M.

    1977-01-01

    In considering the response of a nuclear reactor structure to seismic motions, one must take account of the implications of various levels of damage, short of impairment of safety, and definitely short of collapse, of the structure. Some structural elements of nuclear power plants must perforce remain elastic or nearly elastic in order to perform their allocated safety function. However, in many instances, a purely linear elastic analysis may be unreasonably conservative when one considers that even up to the near yield point range, there are nonlinearities of sufficient amount to reduce required design levels considerably. Moreover, limited yielding of a structure may reduce the response of equipment located in the structure below those levels of response that would be excited were the structure to remain elastic. Energy absorption in the inelastic range is most conveniently treated by use of the so-called 'ductility factor' introduced by the author for design of structures and equipment to resist explosion and blast forces. In general, for small excursions into the inelastic range, especially when the latter can be approximated by an elasto-plastic resistance curve, the design response spectrum is decreased by a simply determined factor that is related to the ductility factor. Many important parts of equipment of a nuclear power plant facility are attached to the principal parts of the structure and respond in a manner determined by the structural response as well as by the general ground motion to which the structure is subjected. This matter involves some difficulty in analysis, but appropriate calculational techniques and design methods are available. A suitable design simplification is one in which the response of the attachment is related to the modal responses of the structure. This equipment response is affected by the relative mass of the attachment and the structure

  15. Decommissioning and equipment replacement of nuclear power plants under uncertainty

    International Nuclear Information System (INIS)

    Takashima, Ryuta; Naito, Yuta; Kimura, Hiroshi; Madarame, Haruki

    2007-01-01

    This study examines the optimal timing for the decommissioning and equipment replacement of nuclear power plants. We consider that the firm has two options of decommissioning and equipment replacement, and determines to exercise these options under electricity price uncertainty. This problem is formulated as two optimal stopping problems. The solution of this model provides the value of the nuclear power plant and the threshold values for decommissioning and replacement. The dependence of decommissioning and replacement strategies on uncertainty and each cost is shown. In order to investigate the probability of events for decommissioning and replacement, Monte Carlo calculations are performed. We also show the probability distribution and the conditional expected time for each event. (author)

  16. Overview of nuclear power plant equipment qualification issues and practices

    International Nuclear Information System (INIS)

    Torr, K.G.

    1989-01-01

    This report presents a view of and commentary on the current status of equipment qualification (EQ) in nuclear industries of the major western nations. The introductory chapters discuss the concepts of EQ, the elements of EQ process and highlight some of the key issues in EQ. A brief review of industry practices and some of the prevalent industrial standards is presented, followed by an overview of current regulatory positions in the USA, France, Germany and Sweden. A summary and commentary on the latest research findings on issues relating to accident simulation, to aging simulation and some special topics related to EQ, has been contributed by Franklin Research Centre of Philadelphia. The last part of the report deals with equipment qualification in Canada and gives recommendations on EQ for new plants as well as currently operational CANDU nuclear power plants

  17. Equipment and special tool design for remote maintenance

    International Nuclear Information System (INIS)

    Northey, L.M.; Thomson, J.D.

    1985-01-01

    Maintenance tasks performed in locations with hostile environments and/or limited space accesses often require equipment that is operated remotely. This paper discusses considerations that should be addressed in the design of remote maintenance equipment. Some of the topics include proper material selection, interface identifications, operational feedback devices and cost limitations. These considerations add ''human engineering'' to the equipment design to assure protection of the tool and the operating personnel. Examples of remote maintenance and inspection systems that were developed by the Westinghouse Hanford Company and that utilize many of these design considerations are included

  18. Equipment design guidance document for flammable gas waste storage tank new equipment

    International Nuclear Information System (INIS)

    Smet, D.B.

    1996-01-01

    This document is intended to be used as guidance for design engineers who are involved in design of new equipment slated for use in Flammable Gas Waste Storage Tanks. The purpose of this document is to provide design guidance for all new equipment intended for application into those Hanford storage tanks in which flammable gas controls are required to be addressed as part of the equipment design. These design criteria are to be used as guidance. The design of each specific piece of new equipment shall be required, as a minimum to be reviewed by qualified Unreviewed Safety Question evaluators as an integral part of the final design approval. Further Safety Assessment may be also needed. This guidance is intended to be used in conjunction with the Operating Specifications Documents (OSDs) established for defining work controls in the waste storage tanks. The criteria set forth should be reviewed for applicability if the equipment will be required to operate in locations containing unacceptable concentrations of flammable gas

  19. Optimal selection of major equipment in dual purpose plants

    International Nuclear Information System (INIS)

    Gabbrielli, E.

    1981-01-01

    Simulation of different operational conditions with the aid of a computer program is one of the best ways of assisting decision-makers in the selection of the most economic mix of equipment for a dual purpose plant. Using this approach this paper deals with the economic comparison of plants consisting of MSF desalinators and combustion gas or back pressure steam turbines coupled to low capacity electric power generators. The comparison is performed on the basis of the data made available by the OPTDIS computer program and the results are given in terms of yearly cost of production as the sum of capital, manpower, maintenance, fuel and chemical costs. (orig.)

  20. Nuclear power plants. Electrical equipment of the safety system. Qualification

    International Nuclear Information System (INIS)

    2001-01-01

    This International Standard applies to electrical parts of safety systems employed at nuclear power plants, including components and equipment of any interface whose failure could affect unfavourably properties of the safety system. The standard also applies to non-electrical safety-related interfaces. Furthermore, the standard describes the generic process of qualification certification procedures and methods of qualification testing and related documentation. (P.A.)

  1. Equipment and physical plant changes in response to the Fukushima event

    International Nuclear Information System (INIS)

    Newman, G.

    2013-01-01

    The Fukushima event led the international nuclear industry and regulatory bodies to challenge the ability of existing nuclear power plants to prevent and mitigate the effects of a severe external event leading to a total loss of AC power and resultant loss of cooling. Canadian Nuclear Industry's immediate response was to provide a high level of assurance that the existing plant is in a high state of readiness to deal with design basis and beyond design basis events, verify the capability of the existing plant to deal with beyond design basis events (equipment, procedures, staff qualification, external support agreements, etc), verify capability to mitigate station black out events, verify capability to cope with internal and external floods and address vulnerabilities to seismically induced damage to mitigation equipment.

  2. New design system for nuclear power plant

    International Nuclear Information System (INIS)

    Kakuta, Masataka; Yoshinaga, Toshiaki; Yoshida, Ikuzo; Tokumasu, Shinji.

    1980-01-01

    As for the machine and equipment layout and the piping design for nuclear power plants, the multilateral coordination and study on such factors as functions, installation, radiation exposure and maintenance are required, and the high reliability is demanded. On the other hand, the quantity of things handled is enormous, therefore it is difficult to satisfy completely the above described requirements and to make plant planning which is completely free from the mutual interference of machines, equipments and pipings by the ordinary design with drawings only. Thereupon, the following new device was adopted to the design method for the purposes of improving the quality and shortening the construction period. Namely at the time of designing new plants, the rationalization of plant planning method was attempted by introducing color composite drawings and the technique of model engineering, at the same time, the newly developed design system for pipings was applied with a computer, thus the large accomplishment was able to be obtained regarding the improvement of reliability and others by making the check-up of the propriety. The design procedures of layout and piping, the layout design and general coordination in nuclear power stations with models and color composite drawings and the design system are explained. (Kako, I.)

  3. Development of remote automatic equipment for BWR power plants

    International Nuclear Information System (INIS)

    Sasaki, Masayoshi

    1984-01-01

    The development of remote control, automatic equipment for nuclear power stations has been promoted to raise the rate of operation of plants by shortening regular inspection period, to improve the safety and reliability of inspection and maintenance works by mechanization, to reduce the radiation exposure dose of workers and to reduce the manpower required for works. The taking-off of control rod drives from reactors and fixing again have been mechanized, but the disassembling, cleaning, inspection and assembling of control rod drives are manually carried out. Therefore, Hitachi Ltd. has exerted effort to develop the automatic equipment for this purpose. The target of development, investigation, the construction and function of the equipment, the performance and the effect of adopting it are reported. The equipment for the volume reduction of spent fuel channel boxes and spent control rods is developed since these are major high level radioactive solid wastes, and their apparent volume is large. Also the target of development, investigated things, the construction and function of the equipment, the performance and the effect of adopting it are reported. (Kako, I.)

  4. 核电厂核岛设备设计管理策划研究%Research on the Planning of Nuclear Island Equipment Design Management of Nuclear Power Plant

    Institute of Scientific and Technical Information of China (English)

    张国平

    2015-01-01

    本文对核电厂核岛设备设计管理策划的重要性进行了阐述,提出了设备设计策划过程中应遵循的一般原则以及策划的主要内容,为编制核电厂核岛设备设计策划文件提供借鉴和参考。%This paper presents the importance of the planning of nuclear island equipment design management, and puts forward the general principles in the planning process of the equipment design management and the main content of the planning. This paper provides the reference for writing the nuclear island equipment design management planning documents.

  5. Modular Design in Treaty Verification Equipment

    Energy Technology Data Exchange (ETDEWEB)

    Macarthur, Duncan Whittemore [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Benz, Jacob [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Tolk, Keith [Milagro Consulting, Albuquerque, NM (United States); Weber, Tom [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-01-27

    It is widely believed that modular design is a good thing. However, there are often few explicit arguments, or even an agreed range of definitions, to back up this belief. In this paper, we examine the potential range of design modularity, the implications of various amounts of modularity, and the advantages and disadvantages of each level of modular construction. We conclude with a comparison of the advantages and disadvantages of each type, as well as discuss many caveats that should be observed to take advantage of the positive features of modularity and minimize the effects of the negative. The tradeoffs described in this paper will be evaluated during the conceptual design to determine what amount of modularity should be included.

  6. Design considerations, tooling, and equipment for remote in-service inspection of radioactive piping and pressure-vessel systems

    International Nuclear Information System (INIS)

    Swannack, D.L.; Schmoker, D.S.

    1983-01-01

    This paper summarizes results obtained in use of remotely-operated nondestructive testing (NDT) equipment for inspection of reactor-system components. Experience obtained in operating the Fast Flux Test Facility (FFTF) has provided a basis for field verification of remote NDT equipment designs and has suggested development improvements. Remote Viewing and data gathering systems used include periscopes, borescopes, fiberscopes, hybrid borescopes/fiberscopes, and closed circuit television. A summary of design consideration for inspection equipment and power plant design is presented to achieve improved equipment operation and reduction of plant maintenance downtime

  7. REQUIREMENTS FOR DESIGN, EQUIPMENT AND OPERATION MODE OF TAXI VEHICLES

    Directory of Open Access Journals (Sweden)

    Norayr Oganesovich Bludyan

    2015-09-01

    Full Text Available The analysis of international experience in application of requirements for the taxi vehicles design and equipment. The approaches to improvement of cabbing have been defined by determination of requirements for taxi vehicles.

  8. Different types of superconductors: their effect on equipment design

    International Nuclear Information System (INIS)

    Ansart, A.; Manuel, P.; Fevrier, A.

    1989-01-01

    Development of superconductors for AC current and of higher critical temperature superconducting materials enlarged the field of research in cryoelectrotechnic. The paper discusses the materials characteristics in relation with the functional needs in the equipments. Are reviewed the main points governing conductor design for a given application, the characteristics and manufacturing technics for different superconducting materials. It is shown how lighter, smaller equipments can be designed as well as new ones for functions not covered with present apparatus. 17 refs [fr

  9. Equipment designs for the spent LWR fuel dry storage demonstration

    International Nuclear Information System (INIS)

    Steffen, R.J.; Kurasch, D.H.; Hardin, R.T.; Schmitten, P.F.

    1980-01-01

    In conjunction with the Spent Fuel Handling and Packaging Program (SFHPP) equipment has been designed, fabricated and successfully utilized to demonstrate the packaging and interim dry storage of spent LWR fuel. Surface and near surface storage configurations containing PWR fuel assemblies are currently on test and generating baseline data. Specific areas of hardware design focused upon include storage cell components and the support related equipment associated with encapsulation, leak testing, lag storage, and emplacement operations

  10. Initial acceptance test experience with FFTF plant equipment

    International Nuclear Information System (INIS)

    Brown, R.K.; Coleman, K.A.; Mahaffey, M.K.; McCargar, C.G.; Young, M.W.

    1978-09-01

    The purpose of this paper is to examine the initial acceptance test experience of certain pieces of auxiliary equipment of the Fast Flux Test Facility (FFTF). The scope focuses on the DHX blowers and drive train, inert gas blowers, H and V containment isolation valves, and the Surveillance and In-service Inspection (SISI) transporter and trolley. For each type of equipment, the discussion includes a summary of the design and system function, installation history, preoperational acceptance testing procedures and results, and unusual events and resolutions

  11. Conceptual design and equipment of visitor centres

    International Nuclear Information System (INIS)

    Huettinger, K.

    1993-01-01

    Discussing the needs to be met by a visitor centre, the author develops the strategies to be adopted and defines the items to be included in the information program. The procedure in preparing the layout and design are explained and the media available to provide various levels of information are listed. Principles of selecting and mixing the media are discussed and the functions of the various sections of the centre described. Also included are examples of the costs and time requirements for the establishment of a typical centre. The importance of regular maintenance and updating is emphasized. (author)

  12. Remote-automated inspection and maintenance of nuclear power plant equipment

    Energy Technology Data Exchange (ETDEWEB)

    Sasaki, Masayoshi; Nakano, Yoshiyuki

    1984-12-01

    Employing remote-control inspection and maintenance equipment in nuclear power plants increases the plant availability by decreasing the annual shutdown time (outage), as well as radiation exposure and man-power. This paper presents an outline of the latest designs for an automatic refueling machine, a control rod drive handling machine, a fuel preparation machine, and a main steam line plug, which were supplied to the Fukushima Dai-Ni No. 2 Plant of the Tokyo Electric Power Co., Inc. (Fukushima 2-2). Also, the up-to-date developments of other new automatic machines, such as a CRD disassembly and cleaning system, spent fuel channel box volume reduction equipment, and robotics for nuclear plant use are presented.

  13. Arctic pipeline planning design, construction, and equipment

    CERN Document Server

    Singh, Ramesh

    2013-01-01

    Utilize the most recent developments to combat challenges such as ice mechanics. The perfect companion for engineers wishing to learn state-of-the-art methods or further develop their knowledge of best practice techniques, Arctic Pipeline Planning provides a working knowledge of the technology and techniques for laying pipelines in the coldest regions of the world. Arctic Pipeline Planning provides must-have elements that can be utilized through all phases of arctic pipeline planning and construction. This includes information on how to: Solve challenges in designing arctic pipelines Protect pipelines from everyday threats such as ice gouging and permafrost Maintain safety and communication for construction workers while supporting typical codes and standards Covers such issues as land survey, trenching or above ground, environmental impact of construction Provides on-site problem-solving techniques utilized through all phases of arctic pipeline planning and construction Is packed with easy-to-read and under...

  14. Optimization on replacement and inspection period of plant equipment

    International Nuclear Information System (INIS)

    Takase, Kentaro; Kasai, Masao

    2004-01-01

    Rationalization of the plant maintenance is one of the main topics being investigated in Japanese nuclear power industries. Optimization of the inspection and replacement period of equipments is effective for the maintenance cost reduction. The more realistic model of the replacement policy is proposed in this study. It is based on the classical replacement policy model and its cost is estimated. Then, to consider the inspection for the maintenance, the formulation that includes the risk concept is discussed. Based on it, two variations of the combination of the inspection and the replacement are discussed and the costs are estimated. In this study the effect of the degradation of the equipment is important. The optimized maintenance policy depends on the existence of significant degradation. (author)

  15. Urgent reconstruction and re-equipping of coking plants

    Energy Technology Data Exchange (ETDEWEB)

    Kvitkin, I.A.; Martynenko, V.M.; Rozenfel' d, M.S.; Svyatogorov, A.A.; Shvartsman, I.G.

    1986-03-01

    This paper discusses the various options involved: complete or partial reconstruction of existing buildings and equipment or new construction with new equipment and new underground and surface communications. It explains that reconstruction work is divided into three phases: initial phase (clearance, dismantling, closing down coking batteries); basic phase (fitting heat-resistant materials, prestart-up assembly work); final phase (drying out, heating up, adjustments, start-up). A structured scheme for a typical initial phase is described and a method of calculating the durations of the various phases is discussed. Conclusion is that there is an urgent requirement for a document to be produced for the control of reconstruction work; it should contain standard durations and could serve as a standard for coking plant reconstruction work.

  16. Overview of Mobile Equipment Used in Case of Beyond Design Basis Accident at NPP Krsko

    International Nuclear Information System (INIS)

    Lukacevic, H.; Kopinc, D.; Ivanjko, M.

    2016-01-01

    Terrorist attack in USA in the September 11, 2001 and accident at the Fukushima - Daiichi Nuclear Power Station in the March 11, 2011 highlight the importance of mitigating strategies in responding to Beyond Design Basis Accident (BDBA), while ensuring cooling of reactor core, containment and spent fuel pool. Nuclear Power Plant Krsko (NEK) has acquired additional mobile equipment and made necessary modifications on existing systems for the connection of this equipment (fast couplers). Usage of mobile equipment is not only limited to design basis accident (DBA), but, also to prevent and mitigate the consequences in case of BDBA, when other plant systems are not available. NEK also decided to take steps for upgrade of safety measures and prepared Safety Upgrade Program (SUP), which is consistent with the nuclear industry response to the Fukushima accident and is implementing main projects and modifications related to SUP. NEK mobile equipment is not required to operate under normal reactor plant operation except for periodic surveillance testing and is incorporated into the normal training process. Equipment is dislocated from the reactor building and most of the equipment is located in the new building, able to withstand extreme natural events, including earthquakes and tornadoes. The usage of all mobile equipment is prescribed as an additional option in NEK operating procedures in following cases and enables following options: filling various tanks, filling the steam generators, filling the containment, additional compressed air source, spent fuel pool refilling and spraying, alternative power supply. This document provides an overview of NEK mobile equipment, which consists of various mobile fire protection pumps, air compressors, protective equipment, fire trucks, diesel generators. Sufficient fuel supply for the equipment is provided on site for a minimum three days of operation. (author).

  17. Developing Predictive Maintenance Expertise to Improve Plant Equipment Reliability

    International Nuclear Information System (INIS)

    Wurzbach, Richard N.

    2002-01-01

    On-line equipment condition monitoring is a critical component of the world-class production and safety histories of many successful nuclear plant operators. From addressing availability and operability concerns of nuclear safety-related equipment to increasing profitability through support system reliability and reduced maintenance costs, Predictive Maintenance programs have increasingly become a vital contribution to the maintenance and operation decisions of nuclear facilities. In recent years, significant advancements have been made in the quality and portability of many of the instruments being used, and software improvements have been made as well. However, the single most influential component of the success of these programs is the impact of a trained and experienced team of personnel putting this technology to work. Changes in the nature of the power generation industry brought on by competition, mergers, and acquisitions, has taken the historically stable personnel environment of power generation and created a very dynamic situation. As a result, many facilities have seen a significant turnover in personnel in key positions, including predictive maintenance personnel. It has become the challenge for many nuclear operators to maintain the consistent contribution of quality data and information from predictive maintenance that has become important in the overall equipment decision process. These challenges can be met through the implementation of quality training to predictive maintenance personnel and regular updating and re-certification of key technology holders. The use of data management tools and services aid in the sharing of information across sites within an operating company, and with experts who can contribute value-added data management and analysis. The overall effectiveness of predictive maintenance programs can be improved through the incorporation of newly developed comprehensive technology training courses. These courses address the use of

  18. 48 CFR 445.407 - Non-Government use of plant equipment.

    Science.gov (United States)

    2010-10-01

    ... plant equipment. 445.407 Section 445.407 Federal Acquisition Regulations System DEPARTMENT OF AGRICULTURE CONTRACT MANAGEMENT GOVERNMENT PROPERTY Contractor Use and Rental of Government Property 445.407 Non-Government use of plant equipment. Requests for non-Government use of plant equipment as...

  19. 48 CFR 945.505-11 - Records of transportation and installation costs of plant equipment.

    Science.gov (United States)

    2010-10-01

    ... and installation costs of plant equipment. 945.505-11 Section 945.505-11 Federal Acquisition Regulations System DEPARTMENT OF ENERGY CONTRACT MANAGEMENT GOVERNMENT PROPERTY Management of Government... plant equipment. The requirements of FAR 45.505-11 apply to plant equipment having a unit cost of $1,000...

  20. Predicting the residual life of plant equipment - Why worry

    International Nuclear Information System (INIS)

    Jaske, C.E.

    1985-01-01

    Predicting the residual life of plant equipment that has been in service for 20 to 30 years or more is a major concern of many industries. This paper reviews the reasons for increased concern for residual-life assessment and the general procedures used in performing such assessments. Some examples and case histories illustrating procedures for assessing remaining service life are discussed. Areas where developments are needed to improve the technology for remaining-life estimation are pointed out. Then, some of the critical issues involved in residual-life assessment are identified. Finally, the future role of residual-life prediction is addressed

  1. Reliable design of electronic equipment an engineering guide

    CERN Document Server

    Natarajan, Dhanasekharan

    2014-01-01

    This book explains reliability techniques with examples from electronics design for the benefit of engineers. It presents the application of de-rating, FMEA, overstress analyses and reliability improvement tests for designing reliable electronic equipment. Adequate information is provided for designing computerized reliability database system to support the application of the techniques by designers. Pedantic terms and the associated mathematics of reliability engineering discipline are excluded for the benefit of comprehensiveness and practical applications. This book offers excellent support

  2. Design review report for modifications to RMCS safety class equipment

    International Nuclear Information System (INIS)

    Corbett, J.E.

    1997-01-01

    This report documents the completion of the formal design review for modifications to the Rotary Mode Core Sampling (RMCS) safety class equipment. These modifications are intended to support core sampling operations in waste tanks requiring flammable gas controls. The objective of this review was to approve the Engineering Change Notices affecting safety class equipment used in the RMCS system. The conclusion reached by the review committee was that these changes are acceptable

  3. Design review report for modifications to RMCS safety class equipment

    Energy Technology Data Exchange (ETDEWEB)

    Corbett, J.E.

    1997-05-30

    This report documents the completion of the formal design review for modifications to the Rotary Mode Core Sampling (RMCS) safety class equipment. These modifications are intended to support core sampling operations in waste tanks requiring flammable gas controls. The objective of this review was to approve the Engineering Change Notices affecting safety class equipment used in the RMCS system. The conclusion reached by the review committee was that these changes are acceptable.

  4. UOE Pipe Manufacturing Process Simulation: Equipment Designing and Construction

    Science.gov (United States)

    Delistoian, Dmitri; Chirchor, Mihael

    2017-12-01

    UOE pipe manufacturing process influence directly on pipeline resilience and operation capacity. At present most spreaded pipe manufacturing method is UOE. This method is based on cold forming. After each technological step appears a certain stress and strain level. For pipe stress strain study is designed and constructed special equipment that simulate entire technological process.UOE pipe equipment is dedicated for manufacturing of longitudinally submerged arc welded DN 400 (16 inch) steel pipe.

  5. A Perspective on Equipment Design for Fusion Remote Handling

    International Nuclear Information System (INIS)

    Mills, S.; Haist, B.; Hamilton, D.

    2006-01-01

    For 8 years, JET remote operations have become more capable and confident. Many tasks have been successfully completed, even those never intended to be remote maintenance activities. The general approach to the provision of remote handling equipment at JET has been the preferred use of commercially-off-the-shelf equipment. In the areas of electrical, electronic, software and control this approach has been generally achievable. However, in the area of mechanical equipment it has been more difficult. In particular the RH tooling has been almost entirely bespoke as its requirements are highly sensitive to the design of the JET component being handled and there are many design variations. Hence, JET has required the design and manufacture of over 700 types of bespoke RH equipment. This paper will discuss the experience of introducing and developing remote handling mechanical equipment for JET. The paper will cover the relationship between the remote handling equipment and the JET component design and the potential for improving the design function. A major lesson from the introduction of remote handling to JET has been demonstration of the very close interdependency of the design of JET components with design of remote handling tooling. The JET remote handling manual was originally introduced as the vehicle to ensure remote handling compatibility by the introduction of standards. Experience has shown that in general the remote handling manual approach has been insufficient. Future fusion machines will be much more complex than JET and will demand even greater remote handling compatibility. This paper will discuss possible methods for improving this process. Equipment operating in a high radiation environment must be dependable It may spend part of its time in areas that would be extremely difficult to recover from in the case of failure. The equipment may also have a high duty cycle to minimise shutdown times and probably cannot be manually inspected on a frequent

  6. System design description for master equipment list, phase I

    International Nuclear Information System (INIS)

    Sandoval, J.D.

    1997-01-01

    This System Design Description (SDD) is for the Master Equipment List Phase I (MEL). It has been prepared following the WI-IC-CM-3-10, ''Software Practices,'' (Ref. 6). This SDD describes the internal design for implementation of the MEL Phase I

  7. Verification of the thermal design of electronic equipment

    Energy Technology Data Exchange (ETDEWEB)

    Hienonen, R.; Karjalainen, M.; Lankinen, R. [VTT Automation, Espoo (Finland). ProTechno

    1997-12-31

    The project `Verification of the thermal design of electronic equipment` studied the methodology to be followed in the verification of thermal design of electronic equipment. This project forms part of the `Cool Electronics` research programme funded by TEKES, the Finnish Technology Development Centre. This project was carried out jointly by VTT Automation, Lappeenranta University of Technology, Nokia Research Center and ABB Industry Oy VSD-Technology. The thermal design of electronic equipment has a significant impact on the cost, reliability, tolerance to different environments, selection of components and materials, and ergonomics of the product. This report describes the method for verification of thermal design. It assesses the goals set for thermal design, environmental requirements, technical implementation of the design, thermal simulation and modelling, and design qualification testing and the measurements needed. The verification method covers all packaging levels of electronic equipment from the system level to the electronic component level. The method described in this report can be used as part of the quality system of a corporation. The report includes information about the measurement and test methods needed in the verification process. Some measurement methods for the temperature, flow and pressure of air are described. (orig.) Published in Finnish VTT Julkaisuja 824. 22 refs.

  8. US GCFR demonstration plant design

    International Nuclear Information System (INIS)

    Hunt, P.S.; Snyder, H.J.

    1980-05-01

    A general description of the US GCFR demonstration plant conceptual design is given to provide a context for more detailed papers to follow. The parameters selected for use in the design are presented and the basis for parameter selection is discussed. Nuclear steam supply system (NSSS) and balance of plant (BOP) component arrangements and systems are briefly discussed

  9. Design, manufacture and installation of measuring and control equipments for the advanced thermal prototype reactor 'Fugen'

    International Nuclear Information System (INIS)

    Hirota, Shigeo; Kawabata, Yoshinori

    1979-01-01

    The advanced thermal prototype reactor ''Fugen'' attained the criticality on March 20, 1978, and 100% output operation on November 13, 1978. On March 20, 1979, it passed the final inspection, and since then, it has continued the smooth operation up to now. The measuring and control equipments are provided for grasping the operational conditions of the plant and operating it safely and efficiently. At the time of designing, manufacturing and installing the measuring and control equipments for Fugen, it was required to clarify the requirements of the plant design, to secure the sufficient functions, and to improve the operational process, maintainability and the reliability and accuracy of the equipments. Many design guidelines and criteria were decided in order to coordinate the conditions among five manufacturers and give the unified state as one plant. The outline of the instrumentations for neutrons, radiation monitoring and process data, the control systems for reactivity, reactor output, pressure and water supply, the safety protection system, and the process computer are described. Finally, the installations and tests of the measuring and control equipments are explained. The aseismatic capability of the equipments was confirmed. (Kako, I.)

  10. High Temperature Calcination - MACT Upgrade Equipment Pilot Plant Test

    Energy Technology Data Exchange (ETDEWEB)

    Richard D. Boardman; B. H. O& #39; Brien; N. R. Soelberg; S. O. Bates; R. A. Wood; C. St. Michel

    2004-02-01

    About one million gallons of acidic, hazardous, and radioactive sodium-bearing waste are stored in stainless steel tanks at the Idaho Nuclear Technology and Engineering Center (INTEC), which is a major operating facility of the Idaho National Engineering and Environmental Laboratory. Calcination at high-temperature conditions (600 C, with alumina nitrate and calcium nitrate chemical addition to the feed) is one of four options currently being considered by the Department of Energy for treatment of the remaining tank wastes. If calcination is selected for future processing of the sodium-bearing waste, it will be necessary to install new off-gas control equipment in the New Waste Calcining Facility (NWCF) to comply with the Maximum Achievable Control Technology (MACT) standards for hazardous waste combustors and incinerators. This will require, as a minimum, installing a carbon bed to reduce mercury emissions from their current level of up to 7,500 to <45 {micro}g/dscm, and a staged combustor to reduce unburned kerosene fuel in the off-gas discharge to <100 ppm CO and <10 ppm hydrocarbons. The staged combustor will also reduce NOx concentrations of about 35,000 ppm by 90-95%. A pilot-plant calcination test was completed in a newly constructed 15-cm diameter calciner vessel. The pilot-plant facility was equipped with a prototype MACT off-gas control system, including a highly efficient cyclone separator and off-gas quench/venturi scrubber for particulate removal, a staged combustor for unburned hydrocarbon and NOx destruction, and a packed activated carbon bed for mercury removal and residual chloride capture. Pilot-plant testing was performed during a 50-hour system operability test January 14-16, followed by a 100-hour high-temperature calcination pilot-plant calcination run January 19-23. Two flowsheet blends were tested: a 50-hour test with an aluminum-to-alkali metal molar ratio (AAR) of 2.25, and a 50-hour test with an AAR of 1.75. Results of the testing

  11. Response of equipment in nuclear power plants to airplane crash

    International Nuclear Information System (INIS)

    Schalk, M.; Woelfel, H.

    1975-01-01

    The question has been posed concerning the effect of airplane crash on the safety of the equipment (pipes, vessels, etc.) mounted on the floors and walls inside the outer structure. This equipment is set into vibration by the crash-induced shaking of the outer building; the resulting stresses may be quite appreciable. The following questions arise: a) how large are these stresses. Can they, for example, be larger than the stresses produced by earthquake loading. b) What are the significant response parameters. c) Which methods of analysis and design criteria are reasonable. To what extent is it possible to apply the techniques which have become generally accepted for earthquake loading. The paper presents a preliminary answer to these questions pending further work in this area. (orig./HP) [de

  12. Engineering electrodynamics electric machine, transformer, and power equipment design

    CERN Document Server

    Turowski, Janusz

    2013-01-01

    Due to a huge concentration of electromagnetic fields and eddy currents, large power equipment and systems are prone to crushing forces, overheating, and overloading. Luckily, power failures due to disturbances like these can be predicted and/or prevented.Based on the success of internationally acclaimed computer programs, such as the authors' own RNM-3D, Engineering Electrodynamics: Electric Machine, Transformer, and Power Equipment Design explains how to implement industry-proven modeling and design techniques to solve complex electromagnetic phenomena. Considering recent progress in magneti

  13. A case study in the use of cancelled plant equipment in nuclear plant modifications

    International Nuclear Information System (INIS)

    Anders, D.A.

    1986-01-01

    The nuclear industry has suffered several blows in the recent past in the form of generating plant cancellations. Upon cancellation, the utility must find a way of minimizing its loss on investment already incurred - consisting of purchased property, partially completed plant, and unused equipment. In many cases, the utility has no practical choice but to dispose of its unused equipment at extremely low prices. While this certainly represents an unfortunate situation for the seller, it does present a significant opportunity for other utilities to procure equipment to use in modifications to their own plants. This paper presents a case study in the use of such cancelled plant equipment in modifications at two nuclear generating facilities. In particular, modifications to replace the refueling platforms at each of the two units at Philadelphia Electric Company's (PECo) Peach Bottom Atomic Power Station and Installation of additional Standby Liquid Control equipment at Limerick Generating Station will be examined. The purpose of the paper is to show the applicability of this information to other utilities

  14. Program outline of seismic fragility capacity tests on nuclear power plant equipment

    International Nuclear Information System (INIS)

    Lijima, T.; Abe, H.; Fujita, T.

    2004-01-01

    A seismic probabilistic safety assessment (PSA) is an available method to evaluate residual risk of nuclear plant that is designed with definitive seismic design condition. Seismic fragility capacity data are necessary for seismic PSA, but we don't have sufficient data of active components of nuclear plants in Japan. This paper describes a plan of seismic fragility capacity tests on nuclear power plant equipment. The purpose of those tests is to obtain seismic fragility capacity of important equipment from a safety design point of view. And the equipment for the fragility capacity tests were selected considering effect on core damage frequency (CDF) that was evaluated by our preliminary seismic PSA. Consequently horizontal shaft pump, electric cabinets, Control Rod Drive system (CRD system) of BWR and PWR plant and vertical shaft pump were selected. The seismic fragility capacity tests are conducted from phase-1 to phase-3, and horizontal shaft pump and electric cabinets are tested on phase-1. The fragility capacity test consists of two types of tests. One is actual equipment test and another is element test. On actual equipment test, a real size model is tested with high-level seismic motion, and critical acceleration and failure mode are investigated. Regarding fragility test phase-1, we selected typical type horizontal shaft pump and electric cabinets for the actual equipment test. Those were Reactor Building Closed Cooling Water (RCW) Pump and eight kinds of electric cabinets such as relay cabinet, motor control center. On the test phase-1, maximum input acceleration for the actual equipment test is intended to be 6-G-force. Since the shaking table of TADOTSU facility did not have capability for high acceleration, we made vibration amplifying system. In this system, amplifying device is mounted on original shaking table and it moves in synchronization with original table. The element test is conducted with many samples and critical acceleration, median and

  15. Emplacement and retrieval equipment design considerations for a repository in salt

    International Nuclear Information System (INIS)

    Nair, B.R.; Bahorich, R.J.

    1987-01-01

    The current design concept for the disposal of nuclear high level waste packages in a repository in salt is based on the emplacement of individual packages in vertical boreholes in the underground mine floor. A key requirement is that the waste packages be capable of being retrieved during the last 26 years of the 76-year repository operating period. The unique design considerations relating to the retrieval of waste packages emplaced in bedded salt are presented in this paper. The information is based on the experience developed during the design of vertical emplacement and retrieval equipment in support of the Sandia Defense High Level Waste experiments at the Waste Isolation Pilot Plant. Also included are the impact of retrievability on the design of the equipment, the special salt cutting technology that was developed for this application, and a description of the equipment

  16. Quality of care and investment in property, plant, and equipment in hospitals.

    OpenAIRE

    Levitt, S W

    1994-01-01

    OBJECTIVE. This study explores the relationship between quality of care and investment in property, plant, and equipment (PPE) in hospitals. DATA SOURCES. Hospitals' investment in PPE was derived from audited financial statements for the fiscal years 1984-1989. Peer Review Organization (PRO) Generic Quality Screen (GQS) reviews and confirmed failures between April 1989 and September 1990 were obtained from the Massachusetts PRO. STUDY DESIGN. Weighted least squares regression models used PRO ...

  17. Design considerations, tooling and equipment for remote in-service inspection of radioactive piping and pressure vessel systems

    International Nuclear Information System (INIS)

    Schmoker, D.S.; Swannack, D.L.

    1983-01-01

    In-Service Inspection programs are performed to monitor and verify the integrity of a nuclear power plant's primary pressure boundaries. Early detection of abnormal structural or material degradation could preclude serious damage to plant systems. This paper summarizes results obtained in use of remotely-operated nondestructive testing (NDT) equipment for inspection of reactor system components. Experience obtained in operating the Fast Flux Test Facility (FFTF) has provided a basis for field verification of remote NDT equipment designs and has suggested development improvements. Remote Viewing and data gathering systems used include periscopes, borescopes, fiberscopes, hybrid borescopes/fiberscopes, and closed circuit television. A summary of design consideration for inspection equipment and power plant design is presented to achieve improved equipment operation and reduction of plant maintenance downtime

  18. Conceptual design of a telecommunications equipment container for humanitarian logistics

    Directory of Open Access Journals (Sweden)

    Stella Parisi

    2017-05-01

    Full Text Available Preparedness addresses the strategy in disaster management that allows the implementation of successful operational response immediately after a disaster. With speed as the main driver, product design for humanitarian aid purposes is a key factor of success in situations of high uncertainty and urgency. Within this context, a telecommunications container (TC has been designed that belongs to a group of containers that serve the purpose of immediate response to global disasters. The TC includes all the necessary equipment to establish a telecommunication centre in the destroyed area within the first 72 hours of humanitarian operations. The design focuses on defining the topology of the various parts of equipment by taking into consideration factors of serviceability, functionality, human-product interaction, universal design language, energy consumption, sustainability and the interrelationship with the other containers. The concept parametric design has been implemented with SolidWorks® CAD system.

  19. Seismic qualification of equipment in operating nuclear power plants: Unresolved Safety Issue A-46

    International Nuclear Information System (INIS)

    Chang, T.Y.

    1987-02-01

    The margin of safety provided in existing nuclear power plant equipment to resist seismically induced loads and perform their intended safety functions may vary considerably, because of significant changes in design criteria and methods for the seismic qualification of equipment over the years. Therefore, the seismic qualification of equipment in operating plants must be reassessed to determine whether requalification is necessary. The objective of technical studies performed under the Task Action Plan A-46 was to establish an explicit set of guidelines and acceptance criteria to judge the adequacy of equipment under seismic loading at all operating plants, in lieu of requiring qualification to the current criteria that are applied to new plants. This report summarizes the work accomplished on USI A-46. In addition, the collection and review of seismic experience data and existing seismic test data are presented. Staff assessment of work accomplished under USI A-46 leads to the conclusion that the use of seismic experience data provides the most reasonable alternative to current qualification criteria. Consideration of seismic qualification by use of experience data was a specific task in USI A-46. Several other A-46 tasks serve to support the use of an experienced data base. The principal technical finding of USI A-46 is that seismic experience data, supplemented by existing seismic test data, applied in accordance with the guidelines developed, can be used to verify the seismic adequacy of mechanical and electrical equipment in operating nuclear plants. Explicit seismic qualification should be required only if seismic experience data or existing test data on similar components cannot be shown to apply

  20. Specific issues for seismic performance of power plant equipment

    Energy Technology Data Exchange (ETDEWEB)

    Nawrotzki, Peter [GERB Vibration Control Systems, Berlin (Germany)

    2010-01-15

    Power plant machinery can be dynamically decoupled from the substructure by the effective use of helical steel springs and viscous dampers. Turbine foundations, coal mills, boiler feed pumps and other machine foundations benefit from this type of elastic support systems to mitigate the transmission of operational vibration. The application of these devices may also be used to protect against earthquakes and other catastrophic events, i.e. airplane crash, of particular importance in nuclear facilities. This article illustrates basic principles of elastic support systems and applications on power plant equipment and buildings in medium and high seismic areas. Spring damper combinations with special stiffness properties are used to reduce seismic acceleration levels of turbine components and other safety or non-safety related structures. For turbine buildings, the integration of the turbine sub-structure into the machine building can further reduce stress levels in all structural members. The application of this seismic protection strategy for a spent fuel storage tank in a high seismic area is also discussed. Safety in nuclear facilities is of particular importance and recent seismic events and the resulting damage in these facilities again brings up the discussion. One of the latest events is the 2007 Chuetsu earthquake in Japan. The resulting damage in the Kashiwazaki Kariwa Nuclear Power Plant can be found in several reports, e.g. in Yamashita. (orig.)

  1. Structural design of nuclear reactor machinery and equipment

    International Nuclear Information System (INIS)

    Hara, Hideki

    1992-01-01

    Since the machinery, equipment and piping which compose nuclear power station facilities are diverse, when those are designed, consideration is given sufficiently to the objective of use and the importance of the object machinery and equipment so that those can maintain the soundness over the design life. In this report, on the contents and the design standard in the design techniques for nuclear reactor machinery and equipment, the way of thinking is shown, taking an example of reactor pressure vessel which is stipulated as the vessel kind 1 in the 'Technical standard of structures and others regarding nuclear facilities for electric power generation', Notice No. 501 of the Ministry of International Trade and Industry. The reactor pressure vessel of 1350 MWe improved type BWR (ABWR) is used under the condition of 87.9 kg/cm 2 and 302 degC, and the inside diameter is about 7.2 m, the inside height is about 21 m, and the wall thickness is about 170 mm. The design standard for reactor pressure vessels and its way of thinking, breakdown prevention design and the design techniques for reactor pressure vessels are described. (K.I.)

  2. Remote handling equipment design for the HEDL fuel supply program

    International Nuclear Information System (INIS)

    Metcalf, I.L.

    1984-09-01

    A process line is currently being developed for fabrication of high exposure mixed uranium-plutonium core assemblies. This paper describes the design philosophy, process flow, equipment, and the handling and radiation shielding techniques used for inspection of Fast Flux Test Facility (FFTF) fuel pins and assembly of Driver Fuel Assemblies (DFAs) 6 figures

  3. Equipment Specific Optimum Blast-Design Using Genetic Algorithm

    Directory of Open Access Journals (Sweden)

    Rahul Upadhyay

    2015-08-01

    Full Text Available Design of blasting parameters plays an important role in the optimization of mining cost as well as cost of subsequent processing of ore. Drilling and handling costs are the major mining cost. This work presents an indirect optimization model for mining cost through optimization of blasting parameters for a particular set of drilling and loading equipment.

  4. System Definition and Analysis: Power Plant Design and Layout

    International Nuclear Information System (INIS)

    1996-01-01

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

  5. Design guides for radioactive-material-handling facilities and equipment

    International Nuclear Information System (INIS)

    Doman, D.R.; Barker, R.E.

    1980-01-01

    Fourteen key areas relating to facilities and equipment for handling radioactive materials involved in examination, reprocessing, fusion fuel handling and remote maintenance have been defined and writing groups established to prepare design guides for each areas. The guides will give guidance applicable to design, construction, operation, maintenance and safety, together with examples and checklists. Each guide will be reviewed by an independent review group. The guides are expected to be compiled and published as a single document

  6. Monitoring equipment environment during nuclear plant operation at Salem and Hope Creek generating stations

    International Nuclear Information System (INIS)

    Blum, A.; Smith, R.J.

    1991-01-01

    Monitoring of environmental parameters has become a significant issue for operating nuclear power plants. While the long-term benefits of plant life extension programs are being pursued with comprehensive environmental monitoring programs, the potential effect of local hot spots at various plant locations needs to be evaluated for its effect on equipment degradation and shortening of equipment qualified life. A significant benefit can be experienced from temperature monitoring when a margin exists between the design versus actual operating temperature. This margin can be translated into longer equipment qualified life and significant reduction in maintenance activities. At PSE and G, the immediate need for monitoring environmental parameters is being accomplished via the use of a Logic Beach Bitlogger. The Bitlogger is a portable data loggings system consisting of a system base, input modules and a communication software package. Thermocouples are installed on selected electrical equipment and cables are run from the thermocouples to the input module of the Bitlogger. Temperature readings are taken at selected intervals, stored in memory, and downloaded periodically to a PC software program, i.e., Lotus. The data is formatted into tabular or graphical documents. Because of their versatility, Bitloggers are being used differently at the authors Nuclear facility. At the Salem Station (2 Units-4 loop Westinghouse PWR), a battery powered, fully portable, calibrated Bitlogger is located in an accessible area inside Containment where it monitors the temperature of various electrical equipment within the Pressurizer Enclosure. It is planned that close monitoring of the local hot spot temperatures in this area will allow them to adjust and reconcile the environmental qualification of the equipment

  7. The process control and management on equipment qualification of nuclear power plant

    International Nuclear Information System (INIS)

    Liu Dong; Wang Hongyin; Zhang Yong

    2013-01-01

    The equipment qualification (EQ) to the safety class equipment is an important safety measure for the nuclear power plants (NPP), and also reflects the nuclear safety culture. Along with the continuous constructions of NPP in China, it has become an important issue for NPP engineering company and equipment suppliers how to effectively establish standard EQ process control and management, and provide sufficient technical arrangements to maintain this EQ management system. This paper summarizes three process of EQ including Design Input, EQ Establishment and EQ Maintenance, proposes the measures and key points for EQ process control and management in phase of NPP construction, and introduces the documents management during the whole process of EQ. (authors)

  8. Evolution of design concepts for remotely maintainable equipment racks

    International Nuclear Information System (INIS)

    Peishel, F.L.; Mouring, R.W.; Schrock, S.L.

    1986-01-01

    Equipment racks have been used to support process equipment in radioactive facilities for many years. Improvements in the design of these racks have evolved relatively slowly primarily as a result of limitations in the capabilities of maintenance equipment; that is, tasks could only be approached from above using bridge cranes with viewing primarily through periscopes. In recent years, however, technological advances have been made by the Consolidated Fuel Reprocessing Program (CFRP) at Oak Ridge National Laboratory (ORNL) in bridge-mounted servomanipulators with onboard auxiliary hoists and television viewing systems. These advances permit full cell coverage by the manipulator arms which, in turn, allow maintenance tasks to be approached horizontally as well as from above. Maintainable equipment items can be stacked vertically on a rack because total overhead access is less important and maintenance tasks that would not have been attempted in the past can now be performed. These advances permit greater flexibility in the design and cell layout of the racks and lead to concepts that could significantly increase the availability of a facility. The evolution of rack design and a description of the alternative concepts based on present maintenance systems capabilities are presented in this paper. 13 refs., 11 figs

  9. Aging management guidelines for commercial nuclear power plant equipment

    International Nuclear Information System (INIS)

    Nakos, J.T.; Gazdzinski, R.F.; Toman, G.J.

    1994-01-01

    The US Department of Energy, in cooperation with the Electric Power Research Institute and nuclear power plant utilities, has prepared ''Aging Management Guidelines'' (AMGs) for commodity types of equipment (e.g., pumps, electrical switchgear) important to license renewal. For the most part, this is also consistent with the Maintenance Rule, 10 CFR 50.65 (1991). AMGs concentrate on technical, (not licensing) issues and are directed toward systems engineers and plant maintenance staff. AMGs include a detailed summary of operating history, stressors, aging mechanisms, and various types of maintenance practices that can be combined to create effective programs that manage aging. All aging mechanisms were addressed; no attempt was made to limit the evaluation to aging mechanisms ''unique to license renewal,'' as defined in the License Renewal Rule, 10 CFR 54 (1991). The first AMG on Electrical Switchgear was published in July 1993. Six (6) additional AMGs will be published by the first quarter of calendar year 1994. It is anticipated that two more AMGs will be started in 1994. The seven ongoing AMG topics are as follows: (1) battery chargers, inverters and uninterruptible power supplies; (2) batteries, stationary; (3) heat exchangers; (4) motor control centers; (5) pumps; (6) switchgear, electric; (7) transformers, power and distribution. In Section 7, industry feedback regarding AMGs is discussed. Overall, the response has been very positive

  10. IFR fuel cycle process equipment design environment and objectives

    International Nuclear Information System (INIS)

    Rigg, R.H.

    1993-01-01

    Argonne National laboratory (ANL) is refurbishing the hot cell facility originally constructed with the EBR-II reactor. When refurbishment is complete, the facility win demonstrate the complete fuel cycle for current generation high burnup metallic fuel elements. These are sodium bonded, stainless steel clad fuel pins of U-Zr or U-Pu-Zr composition typical of the fuel type proposed for a future Integral Fast Reactor (IFR) design. To the extent possible, the process equipment is being built at full commercial scale, and the facility is being modified to incorporate current DOE facility design requirements and modem remote maintenance principles. The current regulatory and safety environment has affected the design of the fuel fabrication equipment, most of which will be described in greater detail in subsequent papers in this session

  11. Rehabilitation of heat exchange equipment a key to power plant life extension and performance improvement

    Energy Technology Data Exchange (ETDEWEB)

    Taveau, F.; Huiban, A.M. [Alstom Power Heat Exchange, 78 - Velizy Villacoublay (France)

    2001-07-01

    With the current evolutions of the energy market and the life extension of the power plants, all the equipment initially supplied need one day or another partial or total rehabilitation. For heat exchange equipment, this includes the condensers, feed water heaters and various heat exchangers. Modernization is in particular necessary when in-service monitoring and periodic inspections show significant deteriorations of the tubes and cooling water leakages leading to forced outages or when tube and tube plate materials are no longer suited to cooling water characteristics or to updated specifications of the secondary system. Feedwater heaters and heat exchangers damaged by erosion/corrosion, vibrations, etc. can be re-designed, manufactured and replaced easily. The operation is more complex on condensers and requires technical surveys, study of alternative solutions and has a more direct impact on the global output of the power plant. That is why our conference will focus on the condenser refurbishment. (author)

  12. Rehabilitation of heat exchange equipment a key to power plant life extension and performance improvement

    International Nuclear Information System (INIS)

    Taveau, F.; Huiban, A.M.

    2001-01-01

    With the current evolutions of the energy market and the life extension of the power plants, all the equipment initially supplied need one day or another partial or total rehabilitation. For heat exchange equipment, this includes the condensers, feed water heaters and various heat exchangers. Modernization is in particular necessary when in-service monitoring and periodic inspections show significant deteriorations of the tubes and cooling water leakages leading to forced outages or when tube and tube plate materials are no longer suited to cooling water characteristics or to updated specifications of the secondary system. Feedwater heaters and heat exchangers damaged by erosion/corrosion, vibrations, etc. can be re-designed, manufactured and replaced easily. The operation is more complex on condensers and requires technical surveys, study of alternative solutions and has a more direct impact on the global output of the power plant. That is why our conference will focus on the condenser refurbishment. (author)

  13. 48 CFR 1845.407 - Non-Government use of plant equipment. (NASA supplements paragraph (a)).

    Science.gov (United States)

    2010-10-01

    ... Use and Rental of Government Property 1845.407 Non-Government use of plant equipment. (NASA supplements paragraph (a)). For NASA, the coverage in FAR 45.407, applies to all equipment, not just plant... 48 Federal Acquisition Regulations System 6 2010-10-01 2010-10-01 true Non-Government use of plant...

  14. On the technical superiority of domestic power plant equipment and its development

    International Nuclear Information System (INIS)

    Zhao Zhiyi

    1993-01-01

    Under the Presumption of affirmed superiority of domestic power plant equipment, some existing deficiencies are pointed out. The scientific and technical development of domestic power equipment can be impelled through catching up with advanced technologies. The necessity of optimal matching of plant equipment from the engineering point of view is emphasized by the authors in association with a prospective outlook of key power equipment and development suggestions

  15. The design and equipments of hospital pharmacies in Isfahan, Iran

    Directory of Open Access Journals (Sweden)

    Ali Mohammad Sabzghabaee

    2010-01-01

    Full Text Available Background: Nowadays pharmaceutical care departments located in hospitals are amongst the important pillars of the healthcare system. The aim of this study was to evaluate designing features and equipments of hospital drugstores affiliated with Isfahan University of Medical Sciences. Methods: In this cross-sectional study a self-defined and validated questionnaire was used which included all the necessary and standard needed spaces and equipments of an ideal hospital pharmacy. The questionnaire was filled in by one of the researchers in all twelve hospital drugstores located in the teaching and non-teaching hospitals affiliated with Isfahan University of Medical Sciences. Data analysis was done using SPSS (version 14. Results: Results showed that 56% of drugstore space allocations were unsuitable. Used pharmaceutical equipments in 75% of surveyed hospitals were not according to the standards. Almost all of these pharmacies had rather an enough space for storage, but cold storages were not designed in 58% of them. In 66% of perused hospitals, pharmaceutical services disposal level was admissible. The structural engineering parameters like size and dimensions, available spaces, availability of structural planes, existence of air conditioning systems and brightness controllers, adequate stores for drugs and safe places for narcotics were observed in 55% of pharmacies. Conclusions: There are apparent out of standard space allocations and shortages of needed equipments for offering drug services in studied drugstores that may probably lead to a waste of time and money. These issues may reduce the efficiency and safety of pharmaceutical services and drug administration in hospitals.

  16. Equipment design for reliability testing of protection system

    International Nuclear Information System (INIS)

    Situmorang, Johnny; Tjahjono, H.; Santosa, A. Z.; Tjahjani, S.DT.; Ismu, P.H; Haryanto, D.; Mulyanto, D.; Kusmono, S

    1999-01-01

    The equipment for reliability testing of cable of protection system has been designed as a a furnace with the electric heater have a 4 kW power, and need time 10 minute to reach the designed maximum temperature 3000C. The dimension of furnace is 800 mm diameter and 2000 mm length is isolated use rockwool isolator and coated by aluminium. For the designed maximum temperature the surface temperature is 78 0c. Assemble of specimens is arranged horizontally in the furnace. The failure criteria will be defined based on the behaviour of the load circuit in each line of cable specimens

  17. A Digital Power Quality Monitoring Equipment Designed for Digital Substation

    Science.gov (United States)

    Li, Wei; Wang, Xin; Geng, Jiewen

    2018-01-01

    Taking into account both current status and development trend of digital substation, this paper proposed a design of a new multi-channelled digital power quality monitoring equipment with high compatibility. The overall functional structure, hardware architecture, software architecture, interface architecture and some key techniques such as IEC 61850 modelling of transient event and harmonic measurement method under the condition of non-synchronous sampling are described in this paper.

  18. Design for a Remote Monitoring Equipment of Krypton-85

    Institute of Scientific and Technical Information of China (English)

    LUXue-sheng; LIUGuo-rong; YEFeng; JINHui-min; TANGPei-jia

    2003-01-01

    A prototype equipment for monitoring Krypton-85 in situ is designed and set up. A series of relevant software is also developed for remote control, data acquirement and data analysis. Weight of the system is about 300 kg, which is composed of NaI(T1) detector, digi DART, Marillin measurement container and lead shield. The whole system is placed on a homemade go-cart.

  19. Military Traffic Management Command Financial Reporting of Property, Plant, and Equipment

    National Research Council Canada - National Science Library

    1998-01-01

    The overall audit objective was to determine whether the property, plant, and equipment accounts in the FY 1996 Defense Business Operations Fund consolidated financial statements were presented fairly...

  20. Design of EPON far-end equipment based on FTTH

    Science.gov (United States)

    Feng, Xiancheng; Yun, Xiang

    2008-12-01

    Now, most favors fiber access is mainly the EPON fiber access system. Inheriting from the low cost of Ethernet, usability and bandwidth of optical network, EPON technology is one of the best technologies in fiber access and is adopted by the carriers all over the world widely. According to the scheme analysis to FTTH fan-end equipment, hardware design of ONU is proposed in this paper. The FTTH far-end equipment software design deference modulation design concept, it divides the software designment into 5 function modules: the module of low-layer driver, the module of system management, the module of master/slave communication, and the module of main/Standby switch and the module of command line. The software flow of the host computer is also analyzed. Finally, test is made for Ethernet service performance of FTTH far-end equipment, E1 service performance and the optical path protection switching, and so on. The results of test indicates that all the items are accordance with technical request of far-end ONU equipment and possess good quality and fully reach the requirement of telecommunication level equipment. The far-end equipment of FTTH divides into several parts based on the function: the control module, the exchange module, the UNI interface module, the ONU module, the EPON interface module, the network management debugging module, the voice processing module, the circuit simulation module, the CATV module. In the downstream direction, under the protect condition, we design 2 optical modules. The system can set one group optical module working and another group optical module closure when it is initialized. When the optical fiber line is cut off, the LOS warning comes out. It will cause MUX to replace another group optical module, simultaneously will reset module 3701/3711 and will make it again test the distance, and will give the plug board MPC850 report through the GPIO port. During normal mode, the downstream optical signal is transformed into the

  1. Cost determination of the electro-mechanical equipment of a small hydro-power plant

    Energy Technology Data Exchange (ETDEWEB)

    Ogayar, B.; Vidal, P.G. [Grupo de Investigacion IDEA, Escuela Politecnica Superior, University of Jaen, Campus de Las Lagunillas, s/n. 23071-Jaen (Spain)

    2009-01-15

    One of the most important elements on the recovery of a small hydro-power plant is the electro-mechanical equipment (turbine-alternator), since the cost of the equipment means a high percentage of the total budget of the plant. The present paper intends to develop a series of equations which determine its cost from basic parameters such as power and net head. These calculations are focused at a level of previous study, so it will be necessary to carry out the engineering project and request a budget to companies specialized on the construction of electro-mechanical equipment to know its cost more accurately. Although there is a great diversity in the typology of turbines and alternators, data from manufacturers which cover all the considered range have been used. The above equations have been developed for the most common of turbines: Pelton, Francis, Kaplan and semiKaplan for a power range below 2 MW. The obtained equations have been validated with data from real installations which have been subject to analysis by engineering companies working on the assembly and design of small plants. (author)

  2. Designing equipment for use in gamma radiation environments

    International Nuclear Information System (INIS)

    Vandergriff, K.U.

    1990-05-01

    High levels of gamma radiation are known to cause degradation in a variety of materials and components. When designing systems to operate in a high radiation environment, special precautions and procedures should be followed. This report (1) outlines steps that should be followed in designing equipment and (2) explains the general effects of radiation on various engineering materials and components. Much information exists in the literature on radiation effects upon materials. However, very little information is available to give the designer a step-by-step process for designing systems that will be subject to high levels of gamma radiation, such as those found in a nuclear fuel reprocessing facility. In this report, many radiation effect references are relied upon to aid in the design of components and systems. 11 refs., 4 tabs

  3. Conference on Manned Systems Design : New Methods and Equipment

    CERN Document Server

    Kraiss, K-F

    1981-01-01

    This volume contains the proceedings of a conference held in Freiburg, West Germany, September 22-25, 1980, entitled "Manned Systems Design, New Methods and Equipment". The conference was sponsored by the Special Programme Panel on Human Factors of the Scientific Affairs Division of NATO, and supported by Panel VIII, AC/243, on "Human and Biomedical Sciences". Their sponsorship and support are gratefully acknowledged. The contributions in the book are grouped according to the main themes of the conference with special emphasis on analytical approaches, measurement of performance, and simulator design and evaluat ion. The design of manned systems covers many and highly diversified areas. Therefore, a conference under the general title of "Manned Systems Design" is rather ambitious in itself. However, scientists and engineers engaged in the design of manned systems very often are confronted with problems that can be solved only by having several disciplines working together. So it was felt that knowledge about ...

  4. Considering plant life management influences on new plant design

    International Nuclear Information System (INIS)

    Dam, R.F.; Choy, E.; Soulard, M.; Nickerson, J.H.; Hopwood, J.

    2003-01-01

    After operating successfully for more than half their design life, owners of CANDU reactors are now engaging in Plant Life Management (PLiM) activities to ensure not only life attainment, but also life extension. For several years, Atomic Energy of Canada Ltd. (AECL) has been working with domestic and offshore CANDU utilities on a comprehensive and integrated CANDU PLiM program that will see existing CANDU plants successfully and reliably operate through their design life and beyond. To support the PLiM program development, a significant level of infrastructure has been, and continues to be, developed at AECL. This includes the development of databases that document relevant knowledge and background to allow for a more accessible and complete understanding of degradation issues and the strategies needed to deal with these issues. As the level of integration with various project, services and R and D activities in AECL increases, this infrastructure is growing to encompass a wider range of design, operations and maintenance details to support comprehensive and quantitative assessment of CANDU stations. With the maturation of the PLiM program, these processes were adapted for application to newer plants. In particular, a fully integrated program was developed that interrelates the design basis, operations, safety, and reliability and maintenance strategies, as applied to meet plant design goals. This has led to the development of the maintenance-based design concept. The various PLiM technologies, developed and applied in the above programs with operating stations, are being modified and tailored to assist with the new plant design processes to assure that ACR- Advanced CANDU Reactor meets its targets for operation, maintenance, and lifetime performance. Currently, the ACR, developed by Atomic Energy of Canada Ltd. (AECL), is being designed with features to increase capacity factors, to reduce the risk of major equipment failures, to improve access to key components

  5. Investigation of smoke corrosivity in nuclear power plant equipment

    International Nuclear Information System (INIS)

    Nowlen, S.P.

    1987-01-01

    This paper presents certain results of fire safety research at Sandia National Laboratories (SNL). The work presented here is related to the issue of the development of standardized tests for determining the corrosive potential of materials when burned. This effort is associated with the investigation of the effects of fire on the operability of a nuclear power generating station and involves a number of programs. This paper will focus on information about five specific aspects of the corrosivity issue that has been gathered as a part of several individual experimental and analytical studies. These five topics are (1) the current perception of fire risk for nuclear power plants and the roll of corrosivity in that risk, (2) the composition of smoke particulate from large-scale enclosure cable fire tests, (3) the aging behavior of smoke particulate, (4) the effect of fire size on the physical characteristics of generated smoke particulate, and (5) electrical equipment fire exposure test results. 4 refs., 2 figs., 1 tab

  6. Manual on quality assurance for installation and commissioning of instrumentation, control and electrical equipment in nuclear power plants

    International Nuclear Information System (INIS)

    1989-01-01

    The present Manual on Quality Assurance (QA) for Installation and Commissioning of Instrumentation, Control and Electrical (ICE) Equipment of Nuclear Power Plants contains supporting material and illustrative examples for implementing basic requirements of the quality assurance programme in procurement, receiving, installation and commissioning of this equipment. The Manual on Quality Assurance for Installation and Commissioning of ICE Equipment is designed to supplement and be consistent with the Guidebook as well as with the IAEA Code and Safety Guides on Quality Assurance. It is intended for the use of managerial staff and QA personnel of nuclear power plant owners or the organizations respectively responsible for the legal, technical, administrative and financial aspects of a nuclear power plant. The information provided in the Manual will also be useful to the inspection staff of the regulatory organization in the planning and performance of regulatory inspections at nuclear power plants

  7. Designing a power supply for Nim-bin formatted equipment

    International Nuclear Information System (INIS)

    Banuelos G, L. E.; Hernandez D, V. M.; Vega C, H. R.

    2016-09-01

    From an old Nuclear Chicago power supply that was practically in the trash, was able to recover the 19 inches casing, rear connectors and the housing where the circuits were. From here all mechanical parts were cleaned and the electronic design was started to replace the original voltage and current functions of this equipment. The cards for the ±6, ±12 and ±24 voltages were designed, simulated and tested with circuitry that does not rely on specialized components or that is sold only by the equipment manufacturer. In the handling of the current by each voltage to operate, was possible to tie with the specifications of the manufacturers like Ortec or Canberra where a model of power supply gives a power of 160 Watts. Basic tests were performed to show that the behavior is very similar to commercial equipment; such as the full load regulation index and the noise level in the supply voltages. So our Nim-bin voltage source is viable for use in our institution laboratories. (Author)

  8. Chemical process and plant design bibliography 1959-1989

    International Nuclear Information System (INIS)

    Ray, M.S.

    1991-01-01

    This book is concerned specifically with chemical process in formation and plant equipment design data. It is a source for chemical engineers, students and academics involved in process and design evaluation. Over 500 chemical categories are included, from Acetaldehyde to zirconium Dioxide, with cross-referencing within the book to appropriate associated chemicals

  9. NATO ARI on Ergonomic Data for Equipment Design

    CERN Document Server

    1984-01-01

    For many years ergonomists and human engineering specialists have made significant contributions to the solution of problems faced in the area of human labour and to the introduction of their research results and field experience into the process of equipment design. However, the rapid increase in complexity of equipment in use as well as in development demonstrates the necessity of broaden­ ing the point of view continuously. The workshop held in Munich from March 22nd to March 26th, 1982, was an excellent opportunity for the participants to discuss their respective interests and their interpretation of needs for future research. The workshop was sponsored by the Human Factors Special Programme Panel of the Scientific Affairs Division of NATO. This sponsorship, together with the helpful assistance rendered by Drs. Bayraktar, Kroemer, and Sanders, is gratefully acknowledged. This volume contains the papers presented during the workshop. All these papers are directly related to the general aim: the ex­ chang...

  10. Advanced Neutron Sources: Plant Design Requirements

    International Nuclear Information System (INIS)

    1990-07-01

    The Advanced Neutron Source (ANS) is a new, world class facility for research using hot, thermal, cold, and ultra-cold neutrons. At the heart of the facility is a 350-MW th , heavy water cooled and moderated reactor. The reactor is housed in a central reactor building, with supporting equipment located in an adjoining reactor support building. An array of cold neutron guides fans out into a large guide hall, housing about 30 neutron research stations. Office, laboratory, and shop facilities are included to provide a complete users facility. The ANS is scheduled to begin operation at the Oak Ridge National Laboratory at the end of the decade. This Plant Design Requirements document defines the plant-level requirements for the design, construction, and operation of the ANS. This document also defines and provides input to the individual System Design Description (SDD) documents. Together, this Plant Design Requirements document and the set of SDD documents will define and control the baseline configuration of the ANS

  11. Electric equipment for Koto Refuse Incineration Plant; Tokyoto Koto seiso kojo muke denki setsubi

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-03-10

    Meidensha Corporation, intending to enter into refuse disposal business, delivered electric equipment to a Koto Refuse Incineration Plant, Koto Ward, Tokyo, and the facilities came into operation in October, 1998. The plant is the largest in Japan in terms of refuse processing capacity (1800t/day), and efforts are exerted to harmonize the plant with the surroundings, which involve pollution measures and a building that images a cruising yacht. The power receiving facility consists of a 66kV nominal two-circuit gas insulated switch and gas insulated transformer arranged in a space saving design. Heat from refuse incineration is fed to a steam turbine generator (yielding 50MW, the largest in Japan, with the surplus offered for sale after 15MW fed to loads in the site) and to neighboring facilities. For the suppression of fluctuations in voltage at the power receiving point, reactive power is subjected to control which is done by controlling the generator magnetic field system. An 11kV distribution system is provided to match the steam turbine generator voltage, and the voltage is stepped down to 6.6kV with the intermediary of a 23MVA gas insulated transformer. The power is fed to high voltage motors such as the one used for the induced draft fan, electric equipment in the buildings, power facilities in the plant, etc. A power monitoring board is provided in the central control room for general supervision over the power related facilities. (NEDO)

  12. Implementing plant design changes economically

    International Nuclear Information System (INIS)

    Seamans, L.D.

    1994-01-01

    Consumers Power Company, as is the case with most utilities, is challenged by the necessity to control non-modification design changes. At the Consumers' Palisades Nuclear Plant, the challenge was met by the establishment of a project team that developed an innovative comprehensive design control process called Functional Equivalent Substitution (FES). This paper describes: (1) the project team composition; (2) the process development methodology; and (3) the implementation results

  13. General Atomic HTGR fuel reprocessing pilot plant: results of initial sequential equipment operation

    International Nuclear Information System (INIS)

    1978-09-01

    In September 1977, the processing of 20 large high-temperature gas-cooled reactor (LHTGR) fuel elements was completed sequentially through the head-end cold pilot plant equipment. This report gives a brief description of the equipment and summarizes the results of the sequential operation of the pilot plant. 32 figures, 15 tables

  14. 47 CFR 32.6510 - Other property, plant and equipment expenses.

    Science.gov (United States)

    2010-10-01

    ... 47 Telecommunication 2 2010-10-01 2010-10-01 false Other property, plant and equipment expenses. 32.6510 Section 32.6510 Telecommunication FEDERAL COMMUNICATIONS COMMISSION (CONTINUED) COMMON... Accounts § 32.6510 Other property, plant and equipment expenses. Class B telephone companies shall use this...

  15. Application of ultrasonic inspection data in strength calculations for nuclear power plant equipment

    International Nuclear Information System (INIS)

    Ovchinnikov, A.V.; Rivkin, E.Yu.; Vasilchenko, G.S.; Zvezdin, Yu.I.

    1991-01-01

    Several kinds of test specimens were produced with three types of defects of defined sizes and positions in the particular localities of weld joints. Such specimens have been used for defect parameter characterization by ultrasonic testing. The principles for schematization of such defects and the formulae for the stress intensity factor calculations for elliptical and semielliptical cracks have been worked out. Methods for defining the sizes of defect which are acceptable have been designed for use for use on operational nuclear power plant equipment and take account of the mutual effects of the force, thermal and residual stresses. The method can be used in the brittle, transitional and tough material state. (author)

  16. The design and development of Project Warrior equipment

    International Nuclear Information System (INIS)

    Thompson, V.R.; Jerram, K.

    1987-01-01

    The project Warrior standing for welding and repair robot in Oldbury reactors is the culmination of more than 7 years of conceptual thinking, research, development design and manufacture. The tile encompasses the whole project which includes a work performing manipulator, serving manipulator, the manipulators control system and consoles, a welding head and rangefinder with its associated control system and consoles and a large amount of auxiliary equipment. The system at the forefront of today's technology has been manufactured to carry out remote welding in nuclear reactors, on plate with an oxide covering of up to 200 microns

  17. Design Modification of Kori Unit 1 for the Equipment Qualification

    International Nuclear Information System (INIS)

    Park, J. W.; Kim, M. Y.; Han, K. T.; Park, J. D.

    2007-01-01

    There has not been a strict regulatory requirements for the Equipment Qualification(EQ) in 1970's when Kori Unit 1 had begun the construction and the commercial operation. The Korean regulatory body requested the EQ on the various safety-related components, as a result of Periodic Safety Review. However, the EQ itself is impossible in some areas, due to the high pressure/temperature and flooding environment conditions from the pipe breaks. Design modification is being considered in the Auxiliary Building, the Intermediate Building, the Component Cooling Water Heat Exchanger Building and the Turbine Building, in order to mitigate the environmental conditions for the EQ

  18. Guidelines for Electromagnetic Interference Testing of Power Plant Equipment: Revision 3 to TR-102323

    Energy Technology Data Exchange (ETDEWEB)

    J. Cunningham and J. Shank

    2004-11-01

    To continue meeting safety and reliability requirements while controlling costs, operators of nuclear power plants must be able to replace and upgrade equipment in a cost-effective manner. One issue that has been problematic for new plant equipment and especially for digital instrumentation and control (I&C) systems in recent years is electromagnetic compatibility (EMC). The EMC issue usually involves testing to show that critical equipment will not be adversely affected by electromagnetic interference (EMI) in the plant environment. This guide will help nuclear plant engineers address EMC issues and qualification testing in a consistent, comprehensive manner.

  19. Guidelines for Electromagnetic Interference Testing of Power Plant Equipment: Revision 3 to TR-102323

    International Nuclear Information System (INIS)

    Cunningham, J.; Shank, J.

    2004-01-01

    To continue meeting safety and reliability requirements while controlling costs, operators of nuclear power plants must be able to replace and upgrade equipment in a cost-effective manner. One issue that has been problematic for new plant equipment and especially for digital instrumentation and control (I and C) systems in recent years is electromagnetic compatibility (EMC). The EMC issue usually involves testing to show that critical equipment will not be adversely affected by electromagnetic interference (EMI) in the plant environment. This guide will help nuclear plant engineers address EMC issues and qualification testing in a consistent, comprehensive manner

  20. A Proactive Aging/Asset Management Model to Optimize Equipment Maintenance Resources Over Plant Lifetime

    International Nuclear Information System (INIS)

    Meyer, Theodore A.; Perdue, Robert K.; Woodcock, Joel; Elder, G. Gary

    2002-01-01

    Experience has shown that proactive aging/asset management can best be defined as an ongoing process. Station goals directly supported by such a process include reducing Unplanned Capability Loss Factor and gaining the optimum value from maintenance and aging management budgets. An effective aging/asset management process must meet evolving and sometimes conflicting requirements for efficient and reliable nuclear power plant operation. The process should identify most likely contributors before they fail, and develop cost-effective contingencies. Current trends indicate the need for focused tools that give quantitative input to decision-making. Opposing goals, such as increasing availability while optimizing aging management budgets, must be balanced. Recognizing the importance of experience in reducing the uncertainty inherent in predicting equipment degradation rates, nuclear industry demographics suggest the need to capture existing expert knowledge in a usable form. The Proactive Aging/Asset Management Process has been developed to address these needs. The proactive approach is a process supported by tools. The process identifies goals and develops criteria - including safety, costs, and power production - that are used to prioritize systems and equipment across the plant. The process then draws upon tools to most effectively meet the plant's goals. The Proactive Aging/Asset Management Model TM is one software-enabled tool designed for mathematical optimization. Results assist a plant in developing a plant-wide plan of aging management activities. This paper describes the proactive aging/asset management process and provides an overview of the methodology that has been incorporated in a model to perform a plant-wide optimization of aging management activities. (authors)

  1. Design of remote handling equipment for the ITER NBI

    Energy Technology Data Exchange (ETDEWEB)

    Oka, Kiyoshi; Tada, Eisuke [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-08-01

    The ITER machine has three Neutral Beam Injectors (NBIs) placed tangential to the plasma at a minimum radius of 6.25 m. During operation, neutrons produced by the D-T reactions will irradiate the NBI structure and it will become radioactive. Radiation levels will be such that all subsequent maintenance of the NBIs must be carried out remotely. The presence of tritium and possibly radioactive dust requires that precautions be taken during maintenance to prevent the escape of these contaminants beyond the prescribed boundaries. The scope of this task is both the development of remote maintenance procedures and the design of the remote handling equipment to handle the NBIs. This report describes the design of remote handling tools for the ion source and its filaments, transfer cask, maintenance time, manufacturing schedule and cost estimation. (author)

  2. Conceptual design of the alcohol waste treatment equipment

    International Nuclear Information System (INIS)

    Fujisawa, Morio; Nitta, Kazuhiko; Morita, Yasuhiro; Nakada, Eiju

    2001-01-01

    This report describes the result of Conceptual Design of the Alcohol Waste Treatment Equipment. The experimental fast Reactor, JOYO, saves the radioactive alcohol waste at storage tank. As this alcohol waste is not able to treat with existing equipment, it is stored about 5 m 3 . And the amount of this is increasing every year. So it is necessary to treat the alcohol waste by chemical resolution for example. On account of this, the investigative test about filtration and dialyzer, and conceptual design about catalyst oxidation process, which is composed from head end process to resolution, are done. The results of investigation show as follows. 1. Investigative Test about filtration and dialyzer. (1) The electric conduction is suitable for the judgement of alkyl sodium hydrolysis Alkyl sodium hydrolysis is completed below 39% alcohol concentration. (2) The microfiltration is likely to separate the solid in alcohol waste. (3) From laboratory test, the electrodialyzer is effective for sodium separation in alcohol waste. And sodium remove rate, 96-99%, is confirmed. 2. Conceptual Design. The candidate process is as follows. (1) The head end process is electrodialyzer, and chemical resolution process is catalyst oxidation. (2) The head end process is not installed, and chemical resolution process is catalyst oxidation. (3) The head end process is electrodialyzer, and alcohol extracted by pervaporation. In this Conceptual Design, as far these process, the components, treatment ability, properties of waste, chemical mass balance, safety for fire and explosion, and the plot plan are investigated. As a result, remodeling the existing facility into catalyst oxidation process is effective to treat the alcohol waste, and treatment ability is about 1.25 l/h. (author)

  3. Control of hydrogen sulfide emission from geothermal power plants. Volume III. Final report: demonstration plant equipment descriptions, test plan, and operating instructions

    Energy Technology Data Exchange (ETDEWEB)

    Brown, F.C.; Harvey, W.W.; Warren, R.B.

    1977-01-01

    The elements of the final, detailed design of the demonstration plant for the copper sulfate process for the removal of hydrogen sulfide from geothermal steam are summarized. Descriptions are given of all items of equipment in sufficient detail that they can serve as purchase specifications. The process and mechanical design criteria which were used to develop the specifications, and the process descriptions and material and energy balance bases to which the design criteria were applied are included. (MHR)

  4. Regulatory analysis for resolution of Unresolved Safety Issue A-46, seismic qualification of equipment in operating plants

    International Nuclear Information System (INIS)

    Chang, T.Y.; Anderson, N.R.

    1987-02-01

    The margin of safety provided in existing nuclear power plant equipment to resist seismically induced loads and perform required safety functions may vary considerably, because of significant changes in design criteria and methods for the seismic qualification of equipment over the years. Therefore, the seismic qualification of equipment in operating plants must be reassessed to determine whether requalification is necessary. The objective of technical studies performed under Task Action Plan A-46 was to establish an explicit set of guidelines and acceptance criteria to judge the adequacy of equipment under seismic loading at all operating plants, in lieu of requiring these plants to meet the criteria that are applied to new plants. This report presents the regulatory analysis for Unresolved Safety Issue (USI) A-46. It includes: Statement of the Problem; the Objective of USI A-46; a Summary of A-46 Tasks; a Proposed Implementation Procedure; a Value-Impact Analysis; Application of the Backfit Rule; 10 CFR 50.109; Implementation; and Operating Plants To Be Reviewed to USI A-46 Requirements

  5. Overall plant design of PWRs

    International Nuclear Information System (INIS)

    Meyer, P.J.

    1980-01-01

    In the present lecture the main components and safety related systems are described, to get a general overview about the safety measures in a PWR. The idea to introduce safety systems is to protect the nuclear reactor core against the so-called design accidents and to prevent the release of activity to the environment. Furthermore the operation personnel has to be protected against radioactive contamination. All redundant and diversified safety measures used in a nuclear power station ensure reliable and safe operation of the plant in all modes of operation. To minimize the operational risk to an extended minimum besides active safety systems a lot of passive safety barriers are foreseen. With the design and construction, tests and quality assurance measures are performed to assure a safe plant operation. (orig.)

  6. Design of nuclear power plants

    International Nuclear Information System (INIS)

    Lobo, C.G.

    1987-01-01

    The criteria of design and safety, applied internationally to systems and components of PWR type reactors, are described. The main criteria of the design analysed are: thermohydraulic optimization; optimized arrangement of buildings and components; low costs of energy generation; high level of standardization; application of specific safety criteria for nuclear power plants. The safety criteria aim to: assure the safe reactor shutdown; remove the residual heat and; avoid the release of radioactive elements for environment. Some exemples of safety criteria are given for Angra-2 and Angra-3 reactors. (M.C.K.) [pt

  7. Survey and analysis on environmental and electromagnetic effect on instrumentation and control equipment of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, In Koo; Lee, Dong Young; Cha, Kyung Ho

    2001-03-01

    As the instrumentation and control (I and C) equipment suppliers tend to provide digital components rather than conventional analog type components for instrumentation and control systems of nuclear power plants(NPPs), it is unavoidable to adopt digital equipment for safety I and C systems as well as non-safety systems. However, the full introduction of digital equipment for I and C systems of nuclear power plants raises several concerns which have not been considered in conventional analog I and C equipment. The two major examples of the issues of digital systems are environmental/electromagnetic compatibility (EMC) and software reliability. This report presents the survey and research results on environmental and electromagnetic effect on I and C equipment of nuclear power plants to give a guideline for aging management and design process. Electromagnetic site surveys were conducted to be used as a part of technical basis to demonstrate that I and C systems are compatible with the ambient electromagnetic noise in Korean nuclear power plants.

  8. Technical diagnostics - equipment monitoring for increasing safety and availability of nuclear power plants

    International Nuclear Information System (INIS)

    Sturm, A.; Foerster, R.

    1977-01-01

    Utilization of technical diagnostics in equipment monitoring of nuclear power plants for ensuring nuclear safety, economic availability, and for decision making on necessary maintenance is reviewed. Technical diagnostics is subdivided into inspection and early detection of malfunctions. Moreover, combination of technical diagnostics and equipment monitoring, integration of technical diagnostics into maintenance strategy, and problems of introducing early detection of malfunctions into maintenance management of nuclear power plants are also discussed. In addition, a compilation of measuring techniques used in technical diagnostics has been made. The international state of the art of equipment monitoring in PWR nuclear power plants is illustrated by description of the sound and vibration measuring techniques. (author)

  9. On anti-earthquake design procedure of equipment and pipings in near future

    International Nuclear Information System (INIS)

    Shibata, H.

    1981-01-01

    The requirement of anti-earthquake design of nuclear power plants is getting severe year by year. The author will try to discuss how to control its severity and how to find a proper design procedure for licensing of new plants under such severe requirements. On the other hand we suffered from the enormous volumes of documents. To decrease such volumes, the format of documents should be standardized as well as the design procedure standardization. Starting from this point, we need the research and development on the following subjects: i) Standardization of design procedure. ii) Standardization of document. iii) Establishment of standard review procedure using computer. iv) Standardization of earthquake-resistant designed equipment. v) Standardization of anti-earthquake design procedure of piping systems. vi) Introducing margin evaluation procedure to design procedure. vii) Introducing proving test procedure of active component to design procedure. viii) Establishment of evaluation of human reliability in design, fabrication, inspection procedures. ix) Establishment of the proper relation of seismic trigger level and post-earthquake design procedures. (orig./HP)

  10. Design guide for heat transfer equipment in water-cooled nuclear reactor systems

    International Nuclear Information System (INIS)

    1975-07-01

    Information pertaining to design methods, material selection, fabrication, quality assurance, and performance tests for heat transfer equipment in water-cooled nuclear reactor systems is given in this design guide. This information is intended to assist those concerned with the design, specification, and evaluation of heat transfer equipment for nuclear service and the systems in which this equipment is required. (U.S.)

  11. DNV launches new recommended practice for design of subsea equipment

    Energy Technology Data Exchange (ETDEWEB)

    Toerstad, Elisabeth Heggelund

    2005-07-01

    The application of modern corrosion-resistant alloys to subsea pipelines and components has led to serious problems of hydrogen embrittlement from the corrosion protection system, resulting in cracks and leaks. The most commonly known incidents of this kind are the duplex hubs on Voinhaven, UK and Garn West and the 13% Cr pipelines on Tune and Aasgard, all at the Norwegian shelf. Understanding the nature of this problem has called for a broad multidisciplinary approach, involving expertise on corrosion and metallurgy as well as structural design and integrity. Competence in probabilistic analyses and risk assessments has also been necessary in order to set acceptance criteria and evaluate probability of failure for existing installations. (Det Norske Veritas (DNV) is now about to issues a Recommended Practise (RP) for design of subsea equipment in duplex steels under Cathodic Protection. The RP is based on knowledge gained through failure investigations, qualification pr grams and research work, and a broad industry consortium has contributed with experience and knowledge. Hydrogen Induced Stress Cracking (HISC) due to hydrogen from the Cathodic Protection (CP) system has been one of the most frequent and costly failure modes for subsea equipment and pipelines in the North Sea over the recent years. Today, the international oil business is looking to Norway for assistance in establishing detailed knowledge about and preventive measures against this failure mode. A significant research work is currently being carried out jointly by DNV, SINTEF and TWI (The Welding Institute), supported by a broad industry consortium - aiming at a better understanding of the HISC (Hydrogen Induced Stress Cracking) mechanism and establishment of test methods which reflect the sensitivity to cracking. DNV has also established a Recommended Practice, which will be issued late 2005. It reflects the experience gained from the failure examinations, fitness for purpose evaluations, research

  12. Process and equipment design optimising product properties and attributes

    NARCIS (Netherlands)

    Bongers, P.M.M.; Thullie, J.

    2009-01-01

    Classically, when products have been developed at the bench, process engineers will search for equipment to manufacture the product at large scale. More than often, this search is constraint to the existing equipment base, or a catalog search for standard equipment. It is then not surprising, that

  13. Investigating factors that influence level and dynamics of capital productivity in plants manufacturing equipment for mines

    Energy Technology Data Exchange (ETDEWEB)

    Karenov, R.S. (Karagandinskii Politekhnicheskii Institut (USSR))

    1990-10-01

    Analyzes productivity of capital in plants manufacturing equipment for underground coal mining in the USSR. Effects of the following factors are evaluated: working time, investment, mechanization of manufacturing processes, power of motors used to drive the manufacturing equipment, duration of a manufacturing cycle, cooperation degree, equipment service life. Effects of insufficient specialization of manufacturing plants and the manufacturing of mining equipment by repair shops of individual mines which should rather specialize in equipment repair and maintenance are evaluated. Analysis shows that specialization of the manufacturing plants could increase productivity of capital by 1.5-2.0 times, reduce labor consumption by 3-5 times and consumption of materials by 1.5-1.7 times. 4 refs.

  14. Sterilization plants equipped with the isotopic gamma radiation sources

    International Nuclear Information System (INIS)

    Mehta, K.; Chmielewski, A.G.

    2007-01-01

    Presentation describes different isotopic gamma radiation sources applicable for sterilization of food and medical materials. Certain gamma pallet irradiators, mini gamma irradiators and different scale gamma tote irradiators are presented. It is concluded, that about two hundreds plants with gamma radiation sources operates in different countries. However, industrially developed countries must construct much more plants than operates now

  15. The seismic assessment of wheeled vehicle type equipment (e.g. emergency power supply vehicle) against severe accident for nuclear power plant in Japan

    International Nuclear Information System (INIS)

    Ikeda, Takuya; Mitsuzawa, Daisuke; Yamaguchi, Yoshikazu; Hasebe, Motohiko; Imamura, Ryutaro; Tomitani, Yuji; Ueyama, Ippei; Kawamoto, Takahiro

    2017-01-01

    After the events at the Fukushima Dai-ichi Nuclear Power Plant, the equipment to mitigate the effects of severe accidents has been installed in the domestic nuclear power plants. From the viewpoint of convenience for installation, etc., a number of industry standard-based wheeled vehicle type equipment has been placed. On the other hand, the new regulations require the equipment for severe accidents to withstand the Design Basis Earthquake. Therefore, the seismic qualification is essential item for wheeled vehicle type equipment according to the regulatory requirement. At that time, compared to the traditional safety-related equipment, there was not enough knowledge of seismic evaluation for vehicle type equipment. This paper reports the overview of wheeled vehicle type equipment and the seismic qualification by test. (author)

  16. Development of Information Datasheets of Nuclear Power Plant (NPP) Equipment using cfiXLM schema

    International Nuclear Information System (INIS)

    Lee, Jaiho; Song, Eunhye

    2014-01-01

    In 2009, EPRI (Electrical Power Research Institute) published a new NPP information handover guide to provide NPP owners and operators with data handover templates in consistent format for effective delivery of information during all stages of the handover process. Another difficult concern for NPP data information management is to exchange the data information among many organizations such as NPP owners, operators, engineering companies, suppliers, and vendors. As a matter of fact, the improperly formatted handover of information sometimes occurs due to the discrepancy of data format (e. g., data description language type). This improper delivery can make negative effects on NPP integrity and safety. Thus, the lack of proper exchange for different data information systems of organizations should be resolved by using an international standard data format. The standard data format can reduce the cost and time for data exchange in each phase for design, procurement, delivery, installation, operation and maintenance of equipment. The AEX(automating equipment information exchange) pilot implementation project team under EPRI advanced nuclear technology (ANT) program has been conducted a research for the use of XML equipment schemas for electronic data exchange(EDE). They applied XML equipment schema for the design, selection, quotation, purchase and mock install of a safety injection centrifugal pump using EDE standard HI(hydraulic institute) 50.7. For data exchange, FIATECH, an industry consortium, has equally developed library of templates and reference data for ISO-15926, which is an international standard capable of reducing data-error and delivery time for exchanging data among different organizations. KHNP as an only owner/operator company has not experienced much difficulty in data interoperability with other organizations, but continued its unremitting exertions to develop a robust system capable of managing data information generated in all the stages of NPP

  17. Development of Information Datasheets of Nuclear Power Plant (NPP) Equipment using cfiXLM schema

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jaiho; Song, Eunhye [KHNP-Central Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    In 2009, EPRI (Electrical Power Research Institute) published a new NPP information handover guide to provide NPP owners and operators with data handover templates in consistent format for effective delivery of information during all stages of the handover process. Another difficult concern for NPP data information management is to exchange the data information among many organizations such as NPP owners, operators, engineering companies, suppliers, and vendors. As a matter of fact, the improperly formatted handover of information sometimes occurs due to the discrepancy of data format (e. g., data description language type). This improper delivery can make negative effects on NPP integrity and safety. Thus, the lack of proper exchange for different data information systems of organizations should be resolved by using an international standard data format. The standard data format can reduce the cost and time for data exchange in each phase for design, procurement, delivery, installation, operation and maintenance of equipment. The AEX(automating equipment information exchange) pilot implementation project team under EPRI advanced nuclear technology (ANT) program has been conducted a research for the use of XML equipment schemas for electronic data exchange(EDE). They applied XML equipment schema for the design, selection, quotation, purchase and mock install of a safety injection centrifugal pump using EDE standard HI(hydraulic institute) 50.7. For data exchange, FIATECH, an industry consortium, has equally developed library of templates and reference data for ISO-15926, which is an international standard capable of reducing data-error and delivery time for exchanging data among different organizations. KHNP as an only owner/operator company has not experienced much difficulty in data interoperability with other organizations, but continued its unremitting exertions to develop a robust system capable of managing data information generated in all the stages of NPP

  18. Military Traffic Management Command Financial Reporting of Property, Plant, and Equipment

    National Research Council Canada - National Science Library

    1998-01-01

    .... We also assessed management controls as they applied to the overall audit objective. The MTMC attempted to improve its reporting of property, plant, and equipment values for the FY 1996 Defense Business Operations Fund financial statements...

  19. A Study on Estimating the Next Failure Time of Compressor Equipment in an Offshore Plant

    Directory of Open Access Journals (Sweden)

    SangJe Cho

    2016-01-01

    Full Text Available The offshore plant equipment usually has a long life cycle. During its O&M (Operation and Maintenance phase, since the accidental occurrence of offshore plant equipment causes catastrophic damage, it is necessary to make more efforts for managing critical offshore equipment. Nowadays, due to the emerging ICTs (Information Communication Technologies, it is possible to send health monitoring information to administrator of an offshore plant, which leads to much concern on CBM (Condition-Based Maintenance. This study introduces three approaches for predicting the next failure time of offshore plant equipment (gas compressor with case studies, which are based on finite state continuous time Markov model, linear regression method, and their hybrid model.

  20. 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)

  1. The use of ultrasound for decontamination of tools and equipment in nuclear power plant Krshko

    International Nuclear Information System (INIS)

    Erman, R.

    1987-01-01

    This paper describes the main principles of the ultrasonic generator functioning and the use of ultrasound for decontamination of tools and equipment in nuclear power plant Krshko. The paper gives the operating procedure and presents decontamination results of tools and equipment fabricated from various materials. (author) 3 refs.; 1 tab

  2. The design of Chp plants

    International Nuclear Information System (INIS)

    Tomassetti, G.

    2001-01-01

    Chp is considered with a bottom-up view, as the most efficient way to satisfy the needs of the users. In order to achieve optimal results a particular care must be used in analyzing the thermal and electrical loads and their interactions. On this basis and taking into account the relationships among the user and the suppliers of electricity, fuels and heat, the energy market structure, the cost of energy and the tax assessment it is possible to properly design Chp plants with benefits for the users [it

  3. Designing of Roaming Protocol for Bluetooth Equipped Multi Agent Systems

    Science.gov (United States)

    Subhan, Fazli; Hasbullah, Halabi B.

    Bluetooth is an established standard for low cost, low power, wireless personal area network. Currently, Bluetooth does not support any roaming protocol in which handoff occurs dynamically when a Bluetooth device is moving out of the piconet. If a device is losing its connection to the master device, no provision is made to transfer it to another master. Handoff is not possible in a piconet, as in order to stay within the network, a slave would have to keep the same master. So, by definition intra-handoff is not possible within a piconet. This research mainly focuses on Bluetooth technology and designing a roaming protocol for Bluetooth equipped multi agent systems. A mathematical model is derived for an agent. The idea behind the mathematical model is to know when to initiate the roaming process for an agent. A desired trajectory for the agent is calculated using its x and y coordinates system, and is simulated in SIMULINK. Various roaming techniques are also studied and discussed. The advantage of designing a roaming protocol is to ensure the Bluetooth enabled roaming devices can freely move inside the network coverage without losing its connection or break of service in case of changing the base stations.

  4. Property, Plant and Equipment disclosure requirements and firm characteristics: the Portuguese Accounting Standardization System

    OpenAIRE

    Botelho, Rafaela; Azevedo, Graça; Costa, Alberto J.; Oliveira, Jonas

    2015-01-01

    In the new Portuguese accounting frame of reference (Portuguese Accounting Standardization System – Sistema de Normalização Contabilística), the issues related to Property, Plant and Equipment assets are dealt with in the Accounting and Financial Reporting Standard (Norma Contabilística de Relato Financeiro – NCRF) 7 (Property, Plant & Equipment). The present study intends to assess the degree of compliance with the disclosure requirements of this accounting standard by Portuguese unlisted co...

  5. Report on assessment of electrical equipment aging for nuclear power plant (AEA), FY2011

    International Nuclear Information System (INIS)

    Minakawa, T.

    2012-11-01

    Electrical components with safety function used in nuclear power plants, such as cables, medium voltage motors, low voltage motors, electrical penetration of reactor containment vessel, motor operated valve, pressure transmitter, temperature detector, etc, are required to be operational under the environment of design basis event (DBE) to shut down a reactor safely and to prevent radioactive materials from being leaked to outside. Polymer materials used as parts of these equipments are gradually degraded by thermal and radiation environment in the normal operation. In addition, the degradation is thought to progress rapidly when they are exposed to the DBE environment and a decrease in performance of the equipment may be caused. From these reason, electrical components with safety function are tested for long-term integrity in accordance with IEEE standard. However, conventional method of accelerated aging which assumes thermal and radiation aging during normal operation is said to have uncertainty in simulating the degradation given in actual operating environment. To address this issue, the project of 'Assessment of Cable Aging for Nuclear Power Plants' (ACA, 2002-2008) was carried out and 'Guide for Cable Environmental Qualification Test for Nuclear Power Plant' was developed. The need for developing an aging evaluation method for other electrical and I and C components was pointed out in the 'Strategy maps 2007', prepared by the cooperation among government, industry and academia. Under the circumstance, the project of 'Assessment of Electrical Equipment Aging for Nuclear Power Plants' (AEA) was initiated in FY2008. In this study, parts of electrical and I and C component with safety function used in nuclear power plant whose aging needs to be considered are employed as specimens, and their aging characteristics under the thermal environment and the combined radiation and thermal environment are obtained (herein after referred to as 'critical part test

  6. New designs of medium power WWER reactor plants

    International Nuclear Information System (INIS)

    Ryzhov, S.B.; Mokhov, V.A.; Nikitenko, M.P.; Chetverikov, A.E.; Veselov, D.O.; Shchekin, I.G.; Petrov, V.V.

    2010-01-01

    The task of constructing NPPs as the objects of regional power industry is included into the Federal Target Program on nuclear power technologies of new generation for the period till 2020. Such NPPs are considered as perspective sources of energy for solution of the problems concerning provision of electric energy, household and industrial heat to the regions with limited capabilities of the power grid. OKB 'GIDROPRESS' present the conceptual study of RP design for the Unit of 600 MW (el.) power, taking into account their long-term experience in the field of development and operation of WWER reactor plants. Practical implementation of WWER-600 and WWER-300 RP designs seems to be feasible: practice in manufacturing the main equipment is available; cooperation of design, scientific organizations and manufacturers of equipment; is established; basic design solutions for equipment are of reference character

  7. Design and construction of a air pollutant gases sampler equipment

    International Nuclear Information System (INIS)

    Gomez S, R.A.; Rodriguez, T.J.

    1996-01-01

    This thesis is about the sketch and construction of an equipment which samples contaminated gases in the air. The topic of this work, is to propose a solution for imported and national equipment. The solution consist on lower prices of imported and national equipment without loosing the accuracy and the precision of those now available. The investigation shows all process to sample gases and theirs measurement for which all the mechanical, electric and electronic equipment, and the necessary software for giving the results in a computerized way were outlined. With this work it was able to succeed in measurements with a national low price, accurate, reliable, programmable, completely automatic and easy to use. This equipment exceed in accuracy the Japanese and the american equipment

  8. DESIGNING FEATURES OF POWER OPTICAL UNITS FOR TECHNOLOGICAL EQUIPMENT

    Directory of Open Access Journals (Sweden)

    M. Y. Afanasiev

    2016-03-01

    Full Text Available This paper considers the question of an optical unit designing for transmitting power laser radiation through an optical fiber. The aim of this work is designing a simple construction unit with minimized reflection losses. The source of radiation in the optical unit described below is an ultraviolet laser with diode pumping. We present the general functioning scheme and designing features for the three main parts: laser beam deflecting system, laser beam dump and optical unit control system. The described laser beam deflection system is composed of a moving flat mirror and a spherical scattering mirror. Comparative analysis of the production technology for such mirrors was carried out, and, as a result, the decision was made to produce both mirrors of 99.99 % pure molybdenum without coating. A moving mirror deflects laser emission from a source through a fiber or deflects it on a spherical mirror and into the laser beam dump, moreover, switching from one position to another occurs almost immediately. It is shown that a scattering mirror is necessary, otherwise, the absorbing surface of the beam dump is being worn out irregularly. The laser beam dump is an open conical cavity, in which the conical element with its spire turned to the emission source is placed. Special microgeometry of the internal surface of the beam dump is suggested for the better absorption effect. An optical unit control system consists of a laser beam deflection system, laser temperature sensor, deflection system solenoid temperature sensor, and deflection mirror position sensor. The signal processing algorithm for signals coming from the sensors to the controller is described. The optical unit will be used in special technological equipment.

  9. Study for process and equipment design of wet gelation stages in vibropacking process

    International Nuclear Information System (INIS)

    Tanimoto, Ryoji; Kikuchi, Toshiaki; Tanaka, Hirokazu; Amino, Masaki; Yanai, Minoru

    2004-02-01

    Process and layout design of external wet gelation stages in vibropacking process was examined for the feasibility study of commercialized FBR cycle system. In this study, following process stages for the oxide core fuel production line were covered, that is, solidification, washing, drying, calcination, reduction, sintering stages including interim storage of sintering particles and reagent recovery stage. The main results obtained by this study are as follows: (1) Based on the process examination results conducted previously, process-flow, mass-balance and number of production line/equipment were clarified. The process is covered from the receive tank of feed solution to the interim storage equipment. Reagent recovery process-flow, mass-balance were also clarified. And preliminary design of the main equipment was reexamined. (2) Normal operation procedure and the procedure after process failure were summarized along with a remote automated operation procedure. Operation sequence of each production line was mapped out by using a time-chart. (3) Design outline of reagent recovery equipments, installed to recover waste liquid from the wet gelation stages in the view of environmental impact were examined. Effective techniques such as collection of solvent, residue waste treatment method were examined its applicability and selected. Schematic block diagram was presented. (4) Analytical items and analyzing apparatus were extracted taking into account of quality control and process management. Analytical sample taking position and frequency of sampling were also examined. (5) A schematic layout drawing of main manufacturing process and reagent recovery process was presented taking into account of material handling. (6) A feature of the operating rate at each process stage was examined by analyzing failure rate reliability of each component. applying the reliability-centred method. (RCM), the operating rate was evaluated and annual maintenance period was estimated using

  10. Materials selection for process equipment in the Hanford waste vitrification plant

    Energy Technology Data Exchange (ETDEWEB)

    Elmore, M R; Jensen, G A

    1991-07-01

    The Hanford Waste Vitrification Plant (HWVP) is being designed to vitrify defense liquid high-level wastes and transuranic wastes stored at Hanford. The HWVP Functional Design Criteria (FDC) requires that materials used for fabrication of remote process equipment and piping in the facility be compatible with the expected waste stream compositions and process conditions. To satisfy FDC requirements, corrosion-resistant materials have been evaluated under simulated HWVP-specific conditions and recommendations have been made for HWVP applications. The materials recommendations provide to the project architect/engineer the best available corrosion rate information for the materials under the expected HWVP process conditions. Existing data and sound engineering judgement must be used and a solid technical basis must be developed to define an approach to selecting suitable construction materials for the HWVP. This report contains the strategy, approach, criteria, and technical basis developed for selecting materials of construction. Based on materials testing specific to HWVP and on related outside testing, this report recommends for constructing specific process equipment and identifies future testing needs to complete verification of the performance of the selected materials. 30 refs., 7 figs., 11 tabs.

  11. Plant designer's view of the operator's role in nuclear plant safety

    International Nuclear Information System (INIS)

    Corcoran, W.R.; Church, J.F.; Cross, M.T.; Porter, N.J.

    1981-01-01

    The nuclear plant operator's role supports the design assumptions and equipment with four functional tasks. He must set up th plant for predictable response to disturbances, operate the plant so as to minimize the likelihood and severity of event initiators, assist in accomplishing the safety functions, and feed back operating experiences to reinforce or redefine the safety analyses' assumptions. The latter role enhances the operator effectiveness in the former three roles. The Safety Level Concept offers a different perspective that enables the operator to view his roles in nuclear plant safety. This paper outlines the operator's role in nuclear safety and classifies his tasks using the Safety Level Concept

  12. Seismic qualification of equipment in operating nuclear power plants. Unresolved safety issue A-46, draft report for comment

    International Nuclear Information System (INIS)

    Chang, T.Y.

    1985-08-01

    The margin of safety provided in existing nuclear power plant equipment to resist seismically induced loads and perform their intended safety functions may vary considerably, because of significant changes in design criteria and methods for the seismic qualification of equipment over the years. Therefore, the seismic qualification of equipment in operating plants should be reassessed to determine whether requalification is necessary. The objective of technical studies performed under the Task Action Plan A-46 was to establish an explicit set of guidelines and acceptance criteria to judge the adequacy of equipment under seismic loading at all operating plants, in lieu of requiring qualification to the current criteria that are applied to new plants. This report summarizes the work accomplished on USI A-46 by the Nuclear Regulatory Commission staff and its contractors, Idaho National Engineering Laboratory, Southwest Research Institute, Brookhaven National Laboratory, and Lawrence Livermore National Laboratory. In addition, the collection and review of seismic experience data by the Seismic Qualification Utility Group and the review and recommendations of a group of seismic consultants, the Senior Seismic Review Advisory Panel, are presented. Staff assessment of work accomplished under USI A-46 leads to the conclusion that the use of seismic experience data provides the most reasonable alternative to current qualification criteria. Consideration of seismic qualification by use of experience data was a specific task in USI A-46. Several other A-46 tasks serve to support the use of an experience data base

  13. Electromagnetic compatibility for the control and command equipments in nuclear power plants

    International Nuclear Information System (INIS)

    Buisson, J.

    1985-06-01

    Different kinds of electrical interference produce some disturbance on electronic sub-assemblies used to assume the control and the command of nuclear reactors. Following interferences are described: power supply lines perturbations, potential difference between grounding connections, electromagnetic fields. A method is described for testing the EMC of different equipments. The advantages of this method are: no destructive method, usable for testing equipment ''in situ'' in operating conditions on nuclear power plant, usable for testing equipment before operating conditions (acceptance test), level of the testing signals similar to the electrical interference level induced by the electromagnetic environment in normal operating conditions, no particular equipment and installation for test are required [fr

  14. PWR heavy equipments manufacture for nuclear power plants

    International Nuclear Information System (INIS)

    Boury, C.; Terrien, J.F.

    1983-10-01

    The manufacture of boilers has been imported by the French nuclear program to the societe FRAMATOME. FRAMATOME, because of the size of this market, has constructed two special plants for manufacturing of nuclear components (vapor generators, reactor tanks, pressurizers); these two high technical facilities are presented: production, staff training, technical overseas assistance, and technical and economical repercussions on the industrial vicinity [fr

  15. Electrical systems design applications on Japanese PWR plants in light of the Fukushima Daiichi Accident

    International Nuclear Information System (INIS)

    Nomoto, Tsutomu

    2015-01-01

    After the Fukushima Daiichi nuclear power plant (1F-NPP) accident (i.e. Station Blackout), several design enhancements have been incorporated or are under considering to Mitsubishi PWR plants' design of not only operational plants' design but also new plants' design. Especially, there are several important enhancements in the area of the electrical system design. In this presentation, design enhancements related to following electrical systems/equipment are introduced; - Offsite Power System; - Emergency Power Source; - Safety-related Battery; - Alternative AC Power Supply Systems. In addition, relevant design requirements/conditions which are or will be considered in Mitsubishi PWR plants are introduced. (authors)

  16. A preliminary design of mechanical device on industrial digital radiography equipment design

    International Nuclear Information System (INIS)

    Nur Khasan; Samuel Praptoyo

    2015-01-01

    A preliminary design of mechanical device on industrial digital radiography equipment has been done. this design is intended as a basis for the manufacture of complete facilities for the realization a prototype on industrial digital radiography equipment. the design and construction were carried out by paying attention to the general configuration of the basic design in which its mechanical design has several components with specific dimensions and heavy mass. this design consist of a main frame holder, flat panel detector support and hydraulic hand stacker for mounting the x-ray machine. this mechanical device design will then be fabricated to facilitate and assist work of digital radiographic retrieval. computer application programs sketch-up is used to draw this design and the analysis stress of autodesk inventor to analysis the strength construction design. the results of this design are the configuration drawing, sketch drawings of construction and the safety factor of construction design with a minimum value of 2.39 as well as a maximum value of 15 when to be simulated by the load 500 Kg which is 4 times of total workload. (author)

  17. Optimization of power take-off equipment for an oscillating water column wave energy plant

    Energy Technology Data Exchange (ETDEWEB)

    Gato, L.M.C.; Falcao, Antonio de F.O. [Dept. de Engenharia Mecanica do IST, Lisboa (Portugal); Paulo Alexandre Justino [INETI/DER, Lisboa (Portugal)

    2005-07-01

    The paper reports the optimization study of the electro-mechanical power take-off equipment for the OWC plant whose structure is a caisson forming the head of the new Douro breakwater. The stochastic approach is employed to model the wave-to-wire energy conversion. The optimization includes rotational speed (for each sea state), turbine geometry and size, and generator rated power. The procedure is implemented into a fully integrated computer code, that yields numerical results for the multi-variable optimization process and for the electrical power output (annual average and for different sea states) with modest computing time (much less than if a time-domain model were used instead). Although focused into a particular real case, the paper is intended to outline a design method that can be applied to a wider class of wave energy converters.

  18. Safety review for seismic qualification on nuclear power plant equipment

    International Nuclear Information System (INIS)

    Fang Qingxian

    1995-01-01

    The standards and requirements for seismic qualification of nuclear power plant's component have been fully addressed, including the scope of seismic qualification, the approach and the method of common seismic qualification, the procedure of the seismic tests, and the criteria for the seismic qualification review. The problems discovered in the safety review and the solution for these problems and some other issues are also discussed

  19. Seismic effects on technological equipment and systems of nuclear power plants

    International Nuclear Information System (INIS)

    Masopust, R.; Pecinka, L.; Podrouzek, J.

    1983-01-01

    A survey is given of problems related to the construction of nuclear power plants with regard to seismic resistance. Sei--smic resistance of technological equipment is evaluated by experimental trials, calculation or the combination of both. Existing and future standards are given for the given field. The Czechoslovak situation is discussed as related to the construction of the Mochovce nuclear power plant. Procedures for testing seismic resistance, types of tests and methods of simulating seismic excitation are described. Antiseismic measures together with structural elements for limiting the seismic effects on technological equipment and nuclear power plant systems are summed up on the basis of foreign experience. (E.F.)

  20. Confirmation of the seismic resistance of nuclear power plant equipment after assembly

    International Nuclear Information System (INIS)

    Kaznovsky, P. S.; Kaznovsky, A. P.; Saakov, E. S.; Ryasnyj, S. I.

    2013-01-01

    It is shown that the natural frequencies and damping decrements of nuclear power plant equipment can only be determined experimentally and directly at the power generation units (reactors) of nuclear power plants under real disassembly conditions for the equipment, piping network, thermal insulation, etc. A computational experimental method is described in which the natural frequencies and damping decrements are determined in the field and the seismic resistance is reevaluated using these values. This method is the basis of the standards document “Methods for confirming the dynamic characteristics of systems and components of the generating units of nuclear power plants which are important for safety” prepared and introduced in 2012.

  1. Study on a quantitative evaluation method of equipment maintenance level and plant safety level for giant complex plant system

    International Nuclear Information System (INIS)

    Aoki, Takayuki

    2010-01-01

    In this study, a quantitative method on maintenance level which is determined by the two factors, maintenance plan and field work implementation ability by maintenance crew is discussed. And also a quantitative evaluation method on safety level for giant complex plant system is discussed. As a result of consideration, the following results were obtained. (1) It was considered that equipment condition after maintenance work was determined by the two factors, maintenance plan and field work implementation ability possessed by maintenance crew. The equipment condition determined by the two factors was named as 'equipment maintenance level' and its quantitative evaluation method was clarified. (2) It was considered that CDF in a nuclear power plant, evaluated by using a failure rate counting the above maintenance level was quite different from CDF evaluated by using existing failure rates including a safety margin. Then, the former CDF was named as 'plant safety level' of plant system and its quantitative evaluation method was clarified. (3) Enhancing equipment maintenance level means an improvement of maintenance quality. That results in the enhancement of plant safety level. Therefore, plant safety level should be always watched as a plant performance indicator. (author)

  2. Technical notes for the conceptual design for an atmospheric fluidized-bed direct combustion power generating plant. [570 MWe plant

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-04-01

    The design, arrangement, thermodynamics, and economics of a 592 MW(e) (nominal gross) electric power generating plant equipped with a Babcock and Wilcox Company (B and W) atmospheric fluidized bed (AFB) boiler are described. Information is included on capital and operating costs, process systems, electrical systems, control and instrumentation, and environmental systems. This document represents a portion of an overall report describing the conceptual designs of two atmospheric fluidized bed boilers and balance of plants for the generation of electric power and the analysis and comparison of these conceptual designs to a conventional pulverized coal-fired electric power generation plant equipped with a wet limestone flue gas desulfurization system.

  3. Technical notes for the conceptual design for an atmospheric fluidized-bed direct combustion power generating plant. [570 MWe plant

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-04-01

    The design, arrangement, thermodynamics, and economics of a 578 MW(e) (nominal gross) electric power generating plant equipped with a Foster Wheeler Energy Corporation (FWEC) atmospheric fluidized bed (AFB) boiler are described. Information is included on capital and operating costs, process systems, electrical systems, control and instrumentation, and environmental systems. This document represents a portion of an overall report describing the conceptual designs of two atmospheric fluidized bed boilers and balance of plants for the generation of electric power and the analysis and comparison of these conceptual designs to a conventional pulverized coal-fired electric power generation plant equipped with a wet limestone flue gas desulfurization system.

  4. Influence of different outdoor design conditions on design cooling load and design capacities of air conditioning equipments

    International Nuclear Information System (INIS)

    Aktacir, Mehmet Azmi; Bueyuekalaca, Orhan; Bulut, Huesamettin; Yilmaz, Tuncay

    2008-01-01

    Outdoor design conditions are important parameters for energy efficiency of buildings. The result of incorrect selection of outdoor design conditions can be dramatic in view of comfort and energy consumption. In this study, the influence of different outdoor design conditions on air conditioning systems is investigated. For this purpose, cooling loads and capacities of air conditioning equipments for a sample building located in Adana, Turkey are calculated using different outdoor design conditions recommended by ASHRAE, the current design data used in Turkey and the daily maximum dry and wet bulb temperatures of July 21st, which is generally accepted as the design day. The cooling coil capacities obtained from the different outdoor design conditions considered in this study are compared with each other. The cost analysis of air conditioning systems is also performed. It is seen that the selection of outdoor design conditions is a very critical step in calculation of the building cooling loads and design capacities of air conditioning equipments

  5. Reality testing a plant design 'virtually' anywhere

    International Nuclear Information System (INIS)

    Anon.

    1997-01-01

    The development of a new world-wide-web compatible information system known as HyperPlant will allow users to navigate real-time three-dimensional plant design and contraction software. It is anticipated that corporate Intranets will be created to facilitate computer-aided design of industrial plants such as piping routes, process schematics, fabrication drawings, and allow use of PDMS (the Plant Design Management System). HyperPlant can also assist in plant commissioning and operation as well as for planning operation and maintenance procedures. (UK)

  6. Design, construction and testing of replacement nuclear coolant pump stators to meet today's equipment reliability expectations

    International Nuclear Information System (INIS)

    Fostier, L.; Howell, D.

    2005-01-01

    The reliability expectations of equipment and components in today's nuclear power plant are much greater than three or more decades ago when nuclear plants were first constructed due to economic impact of a failure. Very few components in a pressurized water reactor plant can have as much impact of the plants capacity factor as a catastrophic failure of a reactor coolant pump winding. This paper describes the maintenance approach taken by one North American utility in attempt to preclude such failures. The paper will discuss the challenges of the reactor coolant pump application and the enhancements made in the winding design and construction by the supplier to address failure mechanisms so as to better meet present reliability expectations in accordance with dedicated specifications. The paper will also present the in-process and final testing requirements and limits imposed in an attempt to ensure quality of the machine windings, along with selected test results from the stators that have been designed and constructed to these specifications to date. (author)

  7. Design of a lunar oxygen production plant

    Science.gov (United States)

    Radhakrishnan, Ramalingam

    1990-01-01

    To achieve permanent human presence and activity on the moon, oxygen is required for both life support and propulsion. Lunar oxygen production using resources existing on the moon will reduce or eliminate the need to transport liquid oxygen from earth. In addition, the co-products of oxygen production will provide metals, structural ceramics, and other volatile compounds. This will enable development of even greater self-sufficiency as the lunar outpost evolves. Ilmenite is the most abundant metal-oxide mineral in the lunar regolith. A process involving the reaction of ilmenite with hydrogen at 1000 C to produce water, followed by the electrolysis of this water to provide oxygen and recycle the hydrogen has been explored. The objective of this 1990 Summer Faculty Project was to design a lunar oxygen-production plant to provide 5 metric tons of liquid oxygen per year from lunar soil. The results of this study describe the size and mass of the equipment, the power needs, feedstock quantity and the engineering details of the plant.

  8. Reliability analysis techniques in power plant design

    International Nuclear Information System (INIS)

    Chang, N.E.

    1981-01-01

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

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

    International Nuclear Information System (INIS)

    1997-01-01

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

  10. Condition monitoring of electrical equipment in nuclear power plants

    International Nuclear Information System (INIS)

    Sugarman, A.

    1986-01-01

    Condition monitoring (CM) is a subset of maintenance testing. It is a quantitative, predictive technique for assessing the effects of all types of aging (environmental, cyclic, operational, etc) on the ''health'' of the equipment. A difference between CM and maintenance testing is that the latter is neither quantitative (i.e., measures the relative condition of the component or material as opposed to merely verifying that its condition is acceptable) nor predictive (i.e., makes judgments, on the ability of the component to perform at a future time). A common example of the principle of CM can be illustrated with the automobile which has a lifetime that is small enough to observe all the periods (break in, random failure, wear out) that occur throughout aging. There are several weak link components in the car (e.g., water hoses, contacts in the distributor, generator, spark plug cables, solenoid, etc) which if they fail will cause failure of the automobile to either start or run. From the day the car is put on the road and is subjected to heat and vibration, significant aging of these components occurs. Degradation in the water hoses, for example is manifested by the elastomeric casing becoming brittle and cracking

  11. Control room design and human engineering in power plants

    International Nuclear Information System (INIS)

    Herbst, L.; Hinz, W.

    1981-01-01

    Automation reduces the human work load. Employment of functional areas permits optimization of operational sequences. Computer based information processing makes it possible to output information in accordance with operating requirements. Design based on human engineering principles assures the quality of the interaction between the operator and the equipment. The degree to which these conceptional features play a role in design of power plant control rooms depends on the unit rating, the mode of operation and on the requirements respecting safety and availability of the plant. (orig./RW)

  12. Diagnosis device for abnormality of power plant equipment

    International Nuclear Information System (INIS)

    Matono, Chiemi; Okamachi, Masao.

    1995-01-01

    In a power plant, if contained water leaks in a condensate system, water is automatically supplied from a feedwater system in accordance with the leakage. Since the diameter of the pipeline of the supplementary feedwater system is generally small, the influence of the leakage of the contained water appears remarkably as a change of the amount of the feed water in the supplementary feed water system. The change of the supplementary state of water to the condensate system by the supplementary feed water system is monitored, to estimate the presence or absence of the leakage of water contained in the condensate system depending on the result of the monitoring, and then the results are informed. In addition, when leakage of water contained in the condensate system is estimated, guidance information is prepared and outputted for coping with the result of the estimation, and the result of the estimation and the guidance information are displayed auditory and visually. As a result, a plant operator can recognize the abnormality of water leakage in the condensate system rapidly and accurately without observing various states of the condensate system and with no knowledge or experiences at high levels, thereby enabling to conduct appropriate processing rapidly. (N.H.)

  13. 10 CFR Appendix N to Part 110 - Illustrative List of Lithium Isotope Separation Facilities, Plants and Equipment Under NRC's...

    Science.gov (United States)

    2010-01-01

    ..., Plants and Equipment Under NRC's Export Licensing Authority N Appendix N to Part 110 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) EXPORT AND IMPORT OF NUCLEAR EQUIPMENT AND MATERIAL Pt. 110, App. N Appendix N to Part 110—Illustrative List of Lithium Isotope Separation Facilities, Plants and Equipment...

  14. Shielding design for better plant availability

    International Nuclear Information System (INIS)

    Biro, G.G.

    1975-01-01

    Design methods are described for providing a shield system for nuclear power plants that will facilitate maintenance and inspection, increase overall plant availability, and ensure that man-rem exposures are as low as practicable

  15. Design of equipment used for high-level waste vitrification at the West Valley Demonstration Project

    International Nuclear Information System (INIS)

    Vance, R.F.; Brill, B.A.; Carl, D.E.

    1997-06-01

    The equipment as designed, started, and operated for high-level radioactive waste vitrification at the West Valley Demonstration Project in western New York State is described. Equipment for the processes of melter feed make-up, vitrification, canister handling, and off-gas treatment are included. For each item of equipment the functional requirements, process description, and hardware descriptions are presented

  16. Advanced Neutron Source: Plant Design Requirements

    International Nuclear Information System (INIS)

    1990-07-01

    The Advanced Neutron Source will be a new world-class facility for research using hot, thermal, cold, and ultra-cold neutrons. The heart of the facility will be a 330-MW (fission), heavy-water cooled and heavy-water moderated reactor. The reactor will be housed in a central reactor building, with supporting equipment located in an adjoining reactor support building. An array of cold neutron guides will fan out into a large guide hall, housing about 30 neutron research stations. Appropriate office, laboratory, and shop facilities will be included to provide a complete facility for users. The ANS is scheduled to begin operation at the Oak Ridge National Laboratory early in the next decade. This PDR document defines the plant-level requirements for the design, construction, and operation of ANS. It also defines and provides input to the individual System Design Description (SDD) documents. Together, this PDR document and the set of SDD documents will define and control the baseline configuration of ANS

  17. Design of a chemical batch plant : a study of dedicated parallel lines with intermediate storage and the plant performance

    OpenAIRE

    Verbiest, Floor; Cornelissens, Trijntje; Springael, Johan

    2016-01-01

    Abstract: Production plants worldwide face huge challenges in satisfying high service levels and outperforming competition. These challenges require appropriate strategic decisions on plant design and production strategies. In this paper, we focus on multiproduct chemical batch plants, which are typically equipped with multiple production lines and intermediate storage tanks. First we extend the existing MI(N) LP design models with the concept of parallel production lines, and optimise the as...

  18. The design and qualification of radiation tolerant equipment for the nuclear power industry

    International Nuclear Information System (INIS)

    Sharp, R.; Pater, L.

    1995-01-01

    The nuclear power industry has many demands for equipment tolerant to the damaging effects of radiation. The wide variety of applications, including components handling, tooling, monitoring and communications, means that a systematic evaluation of the effects of radiation on materials and components used for equipment in radioactive facilities is often required. This paper describes the various effects of radiation on equipment, and discusses how to manage them when using and designing equipment. (Author)

  19. IEEE standard for design qualification of safety systems equipment used in nuclear power generating stations

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

    This standard is written to serve as a general standard for qualification of all types of safety systems equipment, mechanical and instrumentation as well as electrical. It also establishes principles and procedures to be followed in preparing specific safety systems equipment standards. Guidance for qualifying specific safety systems equipment may be found in various specific equipment qualification standards that are now available or are being prepared. It is required that safety systems equipment in nuclear power generating stations meet or exceed its performance requirements throughout its installed life. This is accomplished by a disciplined program of design qualification and quality assurance of design, production, installation, maintenance and surveillance. This standard is for the design qualification section of the program only. Design qualification is intended to demonstrate the capability of the equipment design to perform its safety function(s) over the expected range of normal, abnormal, design basis event, post design basis event, and in-service test conditions. Inherent to design qualification is the requirement for demonstration, within limitations afforded by established technical state-of-the-art, that in-service aging throughout the qualified life established for the equipment will not degrade safety systems equipment from its original design condition to the point where it cannot perform its required safety function(s), upon demand. The above requirement reflects the primary role of design qualification to provide reasonable assurance that design- and age-related common failure modes will not occur during performance of safety function(s) under postulated service conditions

  20. Design of special purpose equipment - remote control dozer

    International Nuclear Information System (INIS)

    Aprameyan, K.

    1990-01-01

    Operation environment in handling hot slag, radio active material, clearing/dismantling buildings and loose rocky zones pose hazards with the operation of heavy duty vehicles. Under such hazardous environment conditions, elimination of operator becomes the prime criteria. Remote control of heavy vehicles is resorted to operate the equipment in various working conditions. Radio control systems coupled with penumatic/hydraulic actuators and proportional control logics aim total control of the equipment from a distance using hand pendants. Bharat Earth Moovers Limited has successfully developed remote control system for dozers of 200hp and 300hp. (author). 3 figs

  1. Hydroelectric plants: economical and ecological consequences of equipment and exploitation variants

    International Nuclear Information System (INIS)

    Maire, P.; Bansard, J.F.; Do, T.

    1995-01-01

    The increasing number of renewal demands for hydroelectric plants authorizations has raised the question of the pertinency and efficiency of the equipments used. Choices are rarely clearly justified by the petitioners. After reminding the reasons and consequences of a given choice and equipment, the necessary steps of an authorization demand are illustrated by a concrete case. It shows that some equipment-management combinations can lead to a more satisfying economical and ecological balance-sheet than those generally proposed. The popularization of computer use allows the examining services to dispose of clear and pedagogical elements to select the regular choices. (J.S.). 10 refs., 11 figs., 2 tabs

  2. Investigation of potential fire-related damage to safety-related equipment in nuclear power plants

    International Nuclear Information System (INIS)

    Wanless, J.

    1985-11-01

    Based on a review of vendor information, fire damage reports, equipment qualification and hydrogen burn test results, and material properties, thirty-three types of equipment found in nuclear power plants were ranked in terms of their potential sensitivity to fire environments. The ranking considered both the functional requirements and damage proneness of each component. A further review of the seven top-ranked components was performed, considering the relative prevalence and potential safety significance of each. From this, relays and hand switches dominate as first choices for fire damage testing with logic equipment, power supplies, transmitters, and motor control centers as future candidates

  3. Effects of foundation conditions on plant design

    International Nuclear Information System (INIS)

    Ehasz, J.L.

    1975-01-01

    Design considerations for nuclear plant foundations are examined including site stability, bearing capacity and settlement, dynamic response, and structural modeling techniques for dynamic analysis. (U.S.)

  4. Zirconium-made equipment for the new La Hague reprocessing plants

    International Nuclear Information System (INIS)

    Decours, J.; Demay, R.; Bernard, C.; Mouroux, J.P.; Simonnet, J.

    1991-01-01

    The use of zirconium was developed to solve some problems of severe corrosion in boiling nitric medium, and to guarantee the service life of the equipment concerned. The paper presents the experience gained since the early 1970s, when the first units made of zirconium were used in French reprocessing plants. For the new La Hague UP3 and UP2 800 plants, it was decided to extend the use of zirconium to make large-scale equipment and, to do so, a major R and D program was implemented, of which the main results are presented

  5. Long-term preventive maintenance of instrumentation control equipment for PWR plants

    International Nuclear Information System (INIS)

    Sugitani, S.; Nanba, M.

    2006-01-01

    Since the PWR plants in Japan have been operated more than 30 years, main instrumentation control equipment of analog systems has been renewed to digital control systems. Renewal works had to be done in short period within periodical inspection term and for several facilities. The Mitsubishi LTD group had been provided with these market needs by its digital control system (MELTAC-NplusR 3) applicable to main instrumentation control equipment for primary and secondary systems and had already finished the renewal for practical plants. (T. Tanaka)

  6. Quality control of repair of equipment for coal preparation plants. Upravlenie kachestvom remonta oborudovaniya ugleobogatitel'nykh fabrik

    Energy Technology Data Exchange (ETDEWEB)

    Okonishnikov, A I; Neskoromnykh, V M; Surzhenko, V S; Sirichenko, R P; Pavlyuchenko, S G; Lesikov, A V

    1984-01-01

    The Ukrniiugleobogashchenie, Kalininsk and Sukhodol'sk coal preparation plants have developed the SUKRO system for control of repair quality of coal preparation equipment in the USSR. The system is based on a system of standards used in coal preparation plants. The following systems of standards used by the SUKRO system are analyzed: organization standards (order of repair in a coal preparation plant, repair planning, spare part systems, methods for determining equipment wear, analysis of equipment failures), standards for maintenance and repair (methods for equipment maintenance, service life of each equipment component or system, structure of preventive repair or repair, organizational models of repair operations, lubrication systems), standards for assessment of labor quality during repair operations. Use of the SUKRO system in the Sukhodol'sk coal preparation plant is evaluated. The SUKRO forms a system of standards for repair and maintenance of equipment considering operation conditions in coal preparation plants, requirements for equipment reliability and service life. (4 refs.)

  7. 10 CFR Appendix J to Part 110 - Illustrative List of Uranium Conversion Plant Equipment and Plutonium Conversion Plant Equipment...

    Science.gov (United States)

    2010-01-01

    .... (2) Especially designed or prepared systems for plutonium metal production. This process usually... or UF6, conversion of UF4 to UF6, conversion of UF6 to UF4, conversion of UF4 to uranium metal, and... several segments of the chemical process industry, including furnaces, rotary kilns, fluidized bed...

  8. 47 CFR 36.352 - Other property plant and equipment expenses-Account 6510 (Class B telephone companies); Accounts...

    Science.gov (United States)

    2010-10-01

    ... 47 Telecommunication 2 2010-10-01 2010-10-01 false Other property plant and equipment expenses... Plant Expenses-Other § 36.352 Other property plant and equipment expenses—Account 6510 (Class B... JURISDICTIONAL SEPARATIONS PROCEDURES; STANDARD PROCEDURES FOR SEPARATING TELECOMMUNICATIONS PROPERTY COSTS...

  9. New technologies for lower-cost design and construction of new nuclear power plants. Annex 20

    International Nuclear Information System (INIS)

    Ritterbusch, S.E.; Bryan, R.E.; Harmon, D.L.

    2002-01-01

    Electric Power Research Institute studies indicate that in order to be competitive with gas-fired electric power plant capital costs, new nuclear plant capital cost in the USA must be decreased by at least 35% to 40% relative to costs of some Advanced Light Water Reactors designed in the early 1990s. To address this need, the U. S. Department of Energy is sponsoring three separate projects under its Nuclear Energy Research Initiative. These projects are the Risk-Informed Assessment of Regulatory and Design Requirements for Future Nuclear Power Plants, the Smart Equipment Nuclear Power Plant Program, and the Design, Procure, Construct, Install and Test Program. The goal of the Design-Construction program is reduction of the complete nuclear plant design-procure-construct-install-test cycle schedule and cost. A 3D plant model was combined with a construction schedule to produce a 4D visualization of plant construction, which was then used to analyze plant construction methods. Insights include the need for concurrent engineering, a plant-wide central database, and use of the World-Wide WEB. The goal of Smart Equipment program is to design, develop, and evaluate the methods for implementing smart equipment and predictive maintenance technology. 'Smart' equipment means components and systems that are instrumented and monitored to detect incipient failures in order to improve their reliability. The resulting smart equipment methods will be combined with a more risk-informed regulatory approach to allow plant designers to (1) simplify designs without compromising overall reliability and safety and (2) maintain more reliable plants at lower cost. Initial results show that rotating equipment such as charging pumps would benefit most from smart instrumentation and that the technique of Bayesian Belief Networks would be most appropriate for providing input to a health monitoring system. (author)

  10. ESBWR-An economical passive plant design

    International Nuclear Information System (INIS)

    Gonzalez Lopez, A.; Rao, A.

    1996-01-01

    This paper provides an overview of the design features of the European Simplified Boiling Water Reactor (ESBWR) design. The ESBWR is a plant design that builds on the Simplified Boiling Water Reactor (SBWR) design described in Reference 1 and 2. The major objective of the ESBWR programme is to develop a plant design that utilizes the basic simplicity of the SBWR design features to improve overall economics as discussed in Reference 3. The design is being developed by an international team of utilities, designers and researchers, with the objective of applying it to the utility and regulatory requirements of Europe. (Author)

  11. Operational planning optimization of steam power plants considering equipment failure in petrochemical complex

    International Nuclear Information System (INIS)

    Luo, Xianglong; Zhang, Bingjian; Chen, Ying; Mo, Songping

    2013-01-01

    Highlights: ► We develop a systematic programming methodology to address equipment failure. ► We classify different operation conditions into real periods and virtual periods. ► The formulated MILP models guarantee cost reduction and enough operation safety. ► The consideration of reserving operation redundancy is effective. - Abstract: One or more interconnected steam power plants (SPPs) are constructed in a petrochemical complex to supply utility energy to the process. To avoid large economic penalties or process shutdowns, these SPPs should be flexible and reliable enough to meet the process energy requirement under varying conditions. Unexpected utility equipment failure is inevitable and difficult to be predicted. Most of the conventional methods are based on the assumption that SPPs do not experience any kind of equipment failure. Unfortunately, a process shutdown cannot be avoided when equipment fails unexpectedly. In this paper, a systematic methodology is presented to minimize the total cost under normal conditions while reserving enough flexibility and safety for unexpected equipment failure conditions. The proposed method transforms the different conditions into real periods to indicate normal scenarios and virtual periods to indicate unexpected equipment failure scenarios. The optimization strategy incorporating various operation redundancy scheduling, the transition constraints from equipment failure conditions to normal conditions, and the boiler load increase behavior modeling are presented to save cost and guarantee operation safety. A detailed industrial case study shows that the proposed systematic methodology is effective and practical in coping with equipment failure conditions with only few additional cost penalties

  12. Computer aided design of piping for a radiochemical plant

    Energy Technology Data Exchange (ETDEWEB)

    Selvaraj, P G; Chandrasekhar, A; Chandrasekar, A V [Reprocessing Group, Indira Gandhi Centre for Atomic Research, Kalpakkam (India); Raju, R P; Mahudeeswaran, K V; Kumar, S V [Reprocessing Group, Bhabha Atomic Research Centre, Mumbai (India)

    1994-06-01

    In a radiochemical plant such as reprocessing plants, process equipment, storage tanks, liquid transfer systems and the associated pipe lines etc. are housed in series of concrete cells. Availability of limited cell space/volume, provision of various modes of liquid transfers with associated redundancies and instrumentation lines with standby alternatives increase the overall piping density. Designing such high density piping layout without interference is quite complex and needs lot of human efforts. This paper briefly describes development of computer codes for the entire scheme of design, drafting and fabrication of piping for nuclear fuel reprocessing plant. The general organisation of various programs, their functions, the complete sequence of the scheme and the flow of data are presented. High degree of reliability of each routine, considerable error checking facilities, marking legends on the drawings, provision for scaling in drafting and accuracy to the extent of one mm in layout design are some of the important features of this scheme. (author). 1 fig.

  13. Control room design and human engineering in power plants

    International Nuclear Information System (INIS)

    Herbst, L.; Hinz, W.

    1982-01-01

    The concept for modern plant control rooms is primary influenced by: The automation of protection, binary control and closed loop control functions; organization employing functional areas; computer based information processing; human engineered design. Automation reduces the human work load. Employment of functional areas permits optimization of operational sequences. Computer based information processing makes it possible to output information in accordance with operating requirements. Design based on human engineering principles assures the quality of the interaction between the operator and the equipment. The degree to which these conceptional features play a role in design of power plant control rooms depends on the unit rating, the mode of operation and on the requirements respecting safety and availability of the plant. (orig.)

  14. HTGR gas turbine power plant preliminary design

    International Nuclear Information System (INIS)

    Koutz, S.L.; Krase, J.M.; Meyer, L.

    1973-01-01

    The preliminary reference design of the HTGR gas turbine power plant is presented. Economic and practical problems and incentives related to the development and introduction of this type of power plant are evaluated. The plant features and major components are described, and a discussion of its performance, economics, development, safety, control, and maintenance is presented. 4 references

  15. Technical guidelines for aseismic design of nuclear power plants

    International Nuclear Information System (INIS)

    Park, Y.J.; Hofmayer, C.H.

    1994-06-01

    This document is a translation, in its entirety, of the Japan Electric Association (JEA) publication entitled open-quotes Technical Guidelines for Aseismic Design of Nuclear Power Plants - JEAG 4601-1987.close quotes This guideline describes in detail the aseismic design techniques used in Japan for nuclear power plants. It contains chapters dealing with: (a)the selection of earthquake ground motions for a site, (b) the investigation of foundation and bedrock conditions, (c) the evaluation of ground stability and the effects of ground movement on buried piping and structures, (d) the analysis and design of structures, and (e) the analysis and design of equipment and distribution systems (piping, electrical raceways, instrumentation, tubing and HVAC duct). The guideline also includes appendices which summarize data, information and references related to aseismic design technology

  16. Anchorage of equipment - requirements and verification methods with emphasis on equipment of existing and constructed VVER-type nuclear power plants

    International Nuclear Information System (INIS)

    Masopust, R.

    1999-01-01

    Criteria and verification methods which are recommended for use in the capacity evaluation of anchorage of safety-related equipment at WWER-type nuclear power plants are presented. Developed in compliance with the relevant basic standards documents specifically for anchorage of WWER-type equipment components, the criteria and methods cover different types of anchor bolts and other anchorage elements which are typical of existing, constructed, or reconstructed WWER-type nuclear power plants

  17. Design of plant safety model in plant enterprise engineering environment

    International Nuclear Information System (INIS)

    Gabbar, Hossam A.; Suzuki, Kazuhiko; Shimada, Yukiyasu

    2001-01-01

    Plant enterprise engineering environment (PEEE) is an approach aiming to manage the plant through its lifecycle. In such environment, safety is considered as the common objective for all activities throughout the plant lifecycle. One approach to achieve plant safety is to embed safety aspects within each function and activity within such environment. One ideal way to enable safety aspects within each automated function is through modeling. This paper proposes a theoretical approach to design plant safety model as integrated with the plant lifecycle model within such environment. Object-oriented modeling approach is used to construct the plant safety model using OO CASE tool on the basis of unified modeling language (UML). Multiple views are defined for plant objects to express static, dynamic, and functional semantics of these objects. Process safety aspects are mapped to each model element and inherited from design to operation stage, as it is naturally embedded within plant's objects. By developing and realizing the plant safety model, safer plant operation can be achieved and plant safety can be assured

  18. An approach to nuclear plant design and modification support for Russian-designed plants in Eastern Europe

    International Nuclear Information System (INIS)

    Ioannidi, J.; Akins, M.J.

    2002-01-01

    projects and providing plant personnel the continuity of assistance necessary to gain sufficient experience in these areas. It further helps assure that the plant's expectations for modifications are met. The Project Management Consultant would establish a modification planning and scheduling protocol for use by plant management on all tasks. The Consultant would focus on development of the modification package process from the problem definition phase through detailed design to the new as-built documentation phase. Additionally, the Consultant would work with the plant staff to develop those managerial programs required to support the modification package process. At project completion, plant management would have these new processes in place and the experience of working with these processes on all plant equipment changes performed during the period, enabling them to work effectively within traditional Western-style contracting schemes.(author)

  19. [Equipment design for magnetic therapy and "Polus" devices].

    Science.gov (United States)

    Viktorov, V A; Malkov, Iu V

    1994-01-01

    Low-frequency magnetotherapy is among the most original therapeutical techniques since it is effective in treating most diseases and has practically no contraindications. Due to atraumaticity of a low-frequency electromagnetic field, the application of magnetotherapy is indicated in geriatric care, in particular. The VNIIMP-VITA Joint-Stock Company has developed a family of the Polyus system equipment for magnetotherapy, whereby all the well-known therapeutical procedures of low-frequency magnetotherapy can be performed. The magnetotherapeutic room equipped with Polyus-2D, Polyus-3, and Polyus-4 ensures a complete set of magneto-therapeutical exposure. The rooms are supplied by the VNIIMP-VITA Joint-Stock Company.

  20. (According to TAS 16, Measurement After Recognition on Property, Plant andEquipment and Accounting İmplementation)

    OpenAIRE

    Örten, Remzi; Bayırlı, Rıdvan

    2007-01-01

    The objective of TAS 16 is to prescribe the accounting treatment for property, plant and equipment so that users of the financial statements can determine information about an entity’s investment in its property, plant and equipment and the changes in such investment. Property, plant and equipment have major importance within the total assets of companies and thus they are very significant in the determination of financial analysis. According to this standard, there are two methods cencerning...

  1. Simplified nuclear plant design for tomorrow's energy needs

    International Nuclear Information System (INIS)

    Slember, R.

    1989-09-01

    Commercial nuclear powered plants play an important role in the strategic energy plans of many countries throughout the world. Many energy planners agree that nuclear plants will have to supply an increasing amount of electrical energy in the 1990s and beyond. Just as other major industries are continually taking steps to update and improve existing products, the United States' nuclear industry has embarked on a program to simplify plant systems, shorten construction time and improve economics for new plant models. One of the models being developed by Westinghouse Electric Corporation and Burns and Roe Company is the Advanced Passive 600 MWe design which incorporates safety features that passively protect the reactor during assumed abnormal operating events. These passive safety systems utilize natural circulation/cooling for mitigating abnormal events and simplify plant design and operation. This type of system eliminates the need for costly active safety grade components, results in a reduction of ancillary equipment and assists in shortening construction time. The use of passive safety systems also permits design simplification of the auxiliary systems effectively reducing operating and maintenance requirements. Collectively, the AP600 design features result in a safe plant that addresses and alleviates the critical industry issues that developed in the 1980s. Further, the design addresses utility and regulatory requirements for safety, reliability, maintainability, operations and economics. Program results to date give confidence that the objectives of the Advanced Passive 600 design are achievable through overall plant simplification. The report will include timely results from the work being performed on the salient technical features of the design, plant construction and operation. Other required institutional changes, such as the prerequisite for a design which is complete and licensed prior to start of construction, will also be presented

  2. Dry Ice Blast Decontamination to in-service equipment in Japanese PWR plant

    International Nuclear Information System (INIS)

    2016-01-01

    MHI had developed several mechanical decontamination methods. Mechanical decontamination is beneficial when it is applied to equipment whose surface is narrow. Especially in terms of secondary waste reduction, MHI started the study of application of Dry Ice Blast Decontamination to actual PWR plant. This paper provides an introduction to Dry Ice Blast Decontamination principle, its system and actual application result to PWR plant. (J.P.N.)

  3. 46 CFR 130.220 - Design of equipment for cooking and heating.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 4 2010-10-01 2010-10-01 false Design of equipment for cooking and heating. 130.220... Design of equipment for cooking and heating. (a) Doors on each cooking appliance must be provided with heavy-duty hinges and locking-devices to prevent accidental opening in heavy weather. (b) Each cooking...

  4. Thermal power plant design and operation

    CERN Document Server

    Sarkar, Dipak

    2015-01-01

    Thermal Power Plant: Design and Operation deals with various aspects of a thermal power plant, providing a new dimension to the subject, with focus on operating practices and troubleshooting, as well as technology and design. Its author has a 40-long association with thermal power plants in design as well as field engineering, sharing his experience with professional engineers under various training capacities, such as training programs for graduate engineers and operating personnel. Thermal Power Plant presents practical content on coal-, gas-, oil-, peat- and biomass-fueled thermal power

  5. Design Of Feedforward Controllers For Multivariable Plants

    Science.gov (United States)

    Seraji, Homayoun

    1989-01-01

    Controllers based on simple low-order transfer functions. Mathematical criteria derived for design of feedforward controllers for class of multiple-input/multiple-output linear plants. Represented by simple low-order transfer functions, obtained without reconstruction of states of commands and disturbances. Enables plant to track command while remaining unresponsive to disturbance in steady state. Feedback controller added independently to stabilize plant or to make control system less susceptible to variations in parameters of plant.

  6. NSS design and plant construction interfaces

    International Nuclear Information System (INIS)

    Stewart, J.J.; Cobb, W.A.

    1976-01-01

    Interface management between NSS design, balance-of-plant design, and plant construction may have a significant effect on schedule stretchout and total plant costs. The paper discusses the importance of the NSS supplier's interface management role, the favorable and unfavorable influencing factors, and examples of interface areas in which experience has demonstrated that problems may arise. Where appropriate, actions are defined to avoid the problems or mitigate the consequences

  7. ESBWR-an economical passive plant design

    International Nuclear Information System (INIS)

    Arnold, H.; Stoop, P.M.; Gonzales, A.; Rao, A.

    1996-01-01

    The ESBWR is a plant design that builds on the GKN Dodewaard natural-circulation reactor and the simplified boiling water reactor (SBWR) design. The major objective of the ESBWR program, which has been in place for the past 3 yr, is to develop a plant design with proven technology that improves the overall plant economics. It utilizes the experience and basic simplicity of the Dodewaard plant and 670-MW(electric) SBWR design features. The design is being developed by an international team of utilities, designers, and researchers. It is being designed to meet the utility and regulatory requirements of Europe. It also addresses the key economic challenges for future nuclear power stations

  8. Accounting Issues: An Essay Series Part IV--Property, Plant, & Equipment

    Science.gov (United States)

    Laux, Judy

    2007-01-01

    This fourth article in a series of theoretical essays intended to supplement the introductory financial accounting course is dedicated to the topic of property, plant, and equipment (PP&E), including both the accounting treatment and its related conceptual connections. The paper also addresses the measurement dilemmas, scandalous accounting…

  9. Production capacity of equipment for medium and large hydroelectric power plant in China

    Energy Technology Data Exchange (ETDEWEB)

    Huang Shenyang [Ministry of Electric Power, Beijing (China). Bureau of Electric Power Machinery

    1995-07-01

    This document presents an overview on the production capacity of equipment for medium and large hydroelectric power plant in China. The document approaches general aspects, production capability and testing facilities related to Francis, Kaplan, tubular and impulse hydroelectric generating sets, and the introduction of main manufacturers as well.

  10. Motives for property, plant and equipment revaluation according to positive accounting theory

    OpenAIRE

    Katarzyna Bareja; Magdalena Giedroyć

    2016-01-01

    The paper identifies motives for property, plant and equipment (PPE) revaluations according to the three main hypotheses proposed by Watts and Zimmerman. Attempt to lower debt-equity ratio is the main motive for PPE revaluation. The method of inductive inference was applied.

  11. Concise aspects regarding the accounting treatment for property, plant and equipment in according with IAS 16

    OpenAIRE

    Ecobici, N

    2007-01-01

    The objective of this paper is to describe the accounting treatment for property, plant and equipment, in according with the IAS 16, including: timing of the recognition of assets, determination of asset carrying amounts using both the cost model and a reevaluation model, depreciation charges and impairment losses to be recognized in relation to these values.

  12. Motives for property, plant and equipment revaluation according to positive accounting theory

    Directory of Open Access Journals (Sweden)

    Katarzyna Bareja

    2016-07-01

    Full Text Available The paper identifies motives for property, plant and equipment (PPE revaluations according to the three main hypotheses proposed by Watts and Zimmerman. Attempt to lower debt-equity ratio is the main motive for PPE revaluation. The method of inductive inference was applied.

  13. Controller design for interval plants

    International Nuclear Information System (INIS)

    Al-Sunni, F.M.

    2003-01-01

    We make use of celebrated Kharitoniv theorem to come up with a design procedure for the stabilization of uncertain systems in the parameters using low order controllers. The proposed design is based on classical design methods. A Non-linear programming (NLP) approach for the design of higher order controllers is also presented. We present our results and give illustrating examples. (author)

  14. Development of support system for maintenance and administration of reprocessing plant equipment

    International Nuclear Information System (INIS)

    Iwasaki, Syogo; Taniguchi, Takayuki; Shiraishi, Yoshihiko; Isaka, Kazuo

    1998-01-01

    Each year, maintenance work is carried out for about 10,000 pieces of equipment, including mechanical devices, electric equipment and instruments, at the Tokai Reprocessing Plant. Ninety percent of such maintenance work is preventive maintenance. In order to manage the information about the maintenance work, a computer support system was developed between 1985 and 1992. Twenty-seven thousand pieces of equipment and 180,000 maintenance histories have already been registered in the system. The system has been used for planning inspections and replacement of equipment as well as checking their maintenance histories. Actual usage of the system has shown that some auxiliary functions need to be added. The system will therefore be improved and extended. (author)

  15. Implementation of the project of equipment reliability in the nuclear power plant of Laguna Verde; Implementacion del proyecto de confiabilidad de equipo en la Central Nucleoelectrica Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

    Rios O, J. E.; Martinez L, A. G. [CFE, Central Laguna Verde, Subgerencia General de Operacion, Veracruz (Mexico)]. e-mail: jrios@cfe.gob.mx

    2008-07-01

    A equipment is reliable if it fulfills the function for which was designed and when it is required. To implement a project of reliability in a nuclear power plant this associate to a process of continuous analysis of the operation, of the conditions and faults of the equipment. The analysis of the operation of a system, of the equipment of the same faults and the parts that integrate to equipment take to identify the potential causes of faults. The predictive analysis on components and equipment allow to rectify and to establish guides to optimize the maintenance and to guarantee the reliability and function of the same ones. The reliability in the equipment is without place to doubts a wide project that embraces from the more small component of the equipment going by the proof of the parts of reserve, the operation conditions until the operative techniques of analysis. Without place of doubt for a nuclear power plant the taking of decisions based on the reliability of their systems and equipment will be the appropriate for to assure the operation and reliability of the same one. In this work would appear the project of reliability its processes, criteria, indicators action of improvement and the interaction of the different disciplines from the Nuclear Power Plant of Laguna Verde like a fundamental point for it put in operation. (Author)

  16. Plant aging and design bases documentation

    International Nuclear Information System (INIS)

    Kelly, J.

    1985-01-01

    As interest in plant aging and lifetime extension continues to grow, the need to identify and capture the original design bases for the plant becomes more urgent. Decisions on lifetime extension and availability must be based on a rational understanding of design input, assumptions, and objectives. As operating plant time accumulates, the history of the early design begins to fade. The longer the utility waits, the harder it will be to re-establish the original design bases. Therefore, the time to develop this foundation is now. This paper demonstrates the impact that collecting and maintaining the original design bases of the plant can have on a utility's lifetime extension program. This impact becomes apparent when considering the technical, regulatory and financial aspects of lifetime extension. It is not good enough to know that the design information is buried somewhere in the corporate archives, and that given enough time, it could be retrieved. To be useful to the lifetime extension program, plant design information must be concise, readily available (i.e., retrievable), and easy to use. These objectives can only be met through a systematic program for collecting and presenting plant design documentation. To get the maximum benefit from a lifetime extension program, usable design bases documentation should be available as early in the plant life as possible. It will help identify areas that require monitoring today so that data is available to make rational decisions in the future

  17. Safety design of Qinshan Nuclear Power Plant

    International Nuclear Information System (INIS)

    Ouyang Yu; Zhang Lian; Du Shenghua; Zhao Jiayu

    1984-01-01

    Safety issues have been greatly emphasized through the design of the Qinshan Nuclear Power Plant. Reasonable safety margine has been taken into account in the plant design parameters, the design incorporated various safeguard systems, such as engineering safety feature systems, safety protection systems and the features to resist natural catastrophes, e. g. earthquake, hurricanes, tide and so on. Preliminary safety analysis and environmental effect assessment have been done and anti-accident provisions and emergency policy were carefully considered. Qinshan Nuclear Power Plant safety related systems are designed in accordance with the common international standards established in the late 70's, as well as the existing engineering standard of China

  18. High-temperature gas-cooled reactor steam-cycle/cogeneration lead plant. Plant Protection and Instrumentation System design description

    International Nuclear Information System (INIS)

    1983-01-01

    The Plant Protection and Instrumentation System provides plant safety system sense and command features, actuation of plant safety system execute features, preventive features which maintain safety system integrity, and safety-related instrumentation which monitors the plant and its safety systems. The primary function of the Plant Protection and Instrumentation system is to sense plant process variables to detect abnormal plant conditions and to provide input to actuation devices directly controlling equipment required to mitigate the consequences of design basis events to protect the public health and safety. The secondary functions of the Plant Protection and Instrumentation System are to provide plant preventive features, sybsystems that monitor plant safety systems status, subsystems that monitor the plant under normal operating and accident conditions, safety-related controls which allow control of reactor shutdown and cooling from a remote shutdown area

  19. Plant Design for the Production of DUAGG

    International Nuclear Information System (INIS)

    Ferrada, J.J.

    2003-01-01

    The cost of producing DUAGG is an important consideration for any interested private firm in determining whether DUCRETE is economically viable as a material of construction in next-generation spent nuclear fuel casks. This study analyzed this project as if it was a stand-alone project. The capital cost includes engineering design, equipment costs and installation, start up, and management; the study is not intended to be a life-cycle cost analysis. The costs estimated by this study are shown in Table ES.1, and the conclusions of this study are listed in Table ES.2. The development of DUAGG and DUCRETE is a major thrust of the Depleted Uranium Uses Research and Development Project. An obvious use of depleted uranium is as a shielding material (e.g., DUCRETE). DUCRETE is made by replacing the conventional stone aggregate in concrete with DUAGG. One objective of this project is to bring the development of DUCRETE to a point at which a demonstrated basis exists for its commercial deployment. The estimation of the costs to manufacture DUAGG is an important part of this effort. Paul Lessing and William Quapp developed DUAGG and DUCRETE as part of an Idaho National Engineering and Environmental Laboratory (INEEL) program to find beneficial uses for depleted uranium (DU). Subsequently, this technology was licensed to Teton Technologies, Inc. The DUAGG process mixes DUO 2 with sintering materials and additives to form pressed briquettes. These briquettes are sintered at 1300 C, and the very dense sintered briquettes are then crushed and classified into gap-graded size fractions. The graded DUAGG is then ready to be used to make high-strength heavy DUCRETE. The DUCRETE shielding will be placed into an annular steel cask-shell mold, which has internal steel reinforcing bars. The objectives of this study are to (1) use previous DUAGG process developments to design a plant that will produce DUAGG at a baseline rate, (2) determine the size of the equipment required to meet the

  20. The basic discussion on nuclear power safety improvement based on nuclear equipment design

    International Nuclear Information System (INIS)

    Zhao Feiyun; Yao Yangui; Yu Hao; He Yinbiao; Gao Lei; Yao Weida

    2013-01-01

    The safety of strengthening nuclear power design was described based on nuclear equipment design after Fukushima nuclear accident. From these aspects, such as advanced standard system, advanced design method, suitable test means, consideration of beyond design basis event, and nuclear safety culture construction, the importance of nuclear safety improvement was emphatically presented. The enlightenment was given to nuclear power designer. (authors)

  1. Military electronic equipment shelter electrical wiring design of electromagnetic compatibility

    International Nuclear Information System (INIS)

    Yang Xuemei

    2012-01-01

    Electromagnetic compatibility is the military electronics shelter design is an important indicator of the shelter's electrical wiring is the key to the design of electromagnetic compatibility. Introduces the basic concepts of electromagnetic compatibility, and focusing on the shelter layout design problems that need attention, and to solve these problems. (authors)

  2. Kozloduy Nuclear Power Plant (Unit 1 and 2). Proposed modifications to increase the seismic capability of equipment and main structures

    International Nuclear Information System (INIS)

    Ordonez Villalobos, A.; Monette, P.R.

    1993-01-01

    Within the framework of the European Community's PHARE Programme of improvement to facilities, their operating systems, equipment and buildings of the Kozloduy NPP in Bulgaria, plant safety during seismic events is considered to be an issue of overriding importance, especially in view of the earthquakes the region suffered during the last decade. Westinghouse Energy Systems International (WESI) and Empresarios Agrupados (EA) have initiated an intensive programme for physical upgrading of equipment with a view to augmenting its seismic capability and, at the same time, to studying design modifications in the diesel-generator buildings, pump house and main building structures (turbines, electrical building). The implementation of these modifications requires an in situ inspection of the real conditions of the various elements, analyses, conceptual design and detail engineering, all of which has to be done in short periods of time using resources available at the plant. This activity is performed by the companies mentioned above, with the collaboration of two engineering companies, Energoproekt of Bulgaria and INITEC of Spain. This paper describes the activities developed and the treatment given to the various aspects of improvement of the seismic capability of equipment and structures. (author)

  3. Prototypical spent nuclear fuel rod consolidation equipment: Phase 2, Final design report: Volume 1, Detailed design

    International Nuclear Information System (INIS)

    Blissell, W.H.; Ciez, A.P.; Goedicke, F.E.; Bessko, C.

    1987-01-01

    This document describes the Westinghouse Final Design for the Prototypical Spent Fuel Consolidation Equipment Demonstration Project. This design represents a fully qualified, licensable, cost effective spent fuel rod consolidation system. As a result of significant concerns raised by DOE and its Technical Review Committee during the 30% Design Review, significant changes were made to the original Preliminary Design resulting from Phase I activities. These changes focused on increased automation, end fitting removal, the rod pulling process and the need to maintain the consolidation canisters as clean as possible. As a result of these changes, the new system is greatly enhanced with a much greater probability of meeting or exceeding the project functional requirements. As a result of delays in resolving cost and contractual differences, additional bench testing was not conducted during Phase II. It is however our belief that the current design exceeds the 90% confidence level required by DOE because of the confidence gained from the Phase I tests, the additional engineering detail completed and the fact that our rod pulling tool has been demonstrated in a similar application at Oconee while our ID tube cutter is a modified (mounting method only) off-the-shelf design. 7 refs., 49 figs., 36 tabs

  4. A new method of knowledge processing for equipment diagnosis of nuclear power plants

    International Nuclear Information System (INIS)

    Fujii, M.; Fukumoto, A.; Tai, I.; Morioka, T.

    1987-01-01

    In this work, the authors complete the development of a new knowledge processing method and representation for equipment diagnosis of nuclear power plants and evaluate its functions by applying to the maintenance and diagnosis support system of the reactor instrumentation. This knowledge processing method system is based on the Cause Generation and Checking concept and has sufficient performance not only in the diagnosis function but also in the man-machine interfacing function. The maintenance and diagnosis support system based on this method leads to the possibility for users to diagnose various phenomena occurred in an objective equipment to the considerable extent by consulting with the system, even if they don't have enough knowledge. With this system, it becomes easy for operators or plant engineers to take immediate actions to counteract against the abnormality. The maintainability of the equipments is improved and MTTR (Mean Time To Repair) is expected to be shorter. This new knowledge processing method is proved to be suited for fault diagnosis of the equipments of nuclear power plants

  5. 10 CFR Appendix H to Part 110 - Illustrative List of Electromagnetic Enrichment Plant Equipment and Components Under NRC Export...

    Science.gov (United States)

    2010-01-01

    ... Equipment and Components Under NRC Export Licensing Authority H Appendix H to Part 110 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) EXPORT AND IMPORT OF NUCLEAR EQUIPMENT AND MATERIAL Pt. 110, App. H Appendix H to Part 110—Illustrative List of Electromagnetic Enrichment Plant Equipment and Components Under...

  6. Information management systems improve advanced plant design

    International Nuclear Information System (INIS)

    Turk, R.S.; Serafin, S.A.; Leckley, J.B.

    1994-01-01

    Computer-aided engineering tools are proving invaluable in both the design and operation of nuclear power plants. ABB Combustion Engineering's Advanced Light Water Reactor (ALWR) features a computerized Information Management System (IMS) as an integral part of the design. The System 80+IMS represents the most powerful information management tool for Nuclear Power Plants commercially available today. Developed by Duke Power Company specifically for use by nuclear power plant owner operators, the IMS consists of appropriate hardware and software to manage and control information flow for all plant related work or tasks in a systematic, consistent, coordinated and informative manner. A significant feature of this IMS is that it is primarily based on plant data. The principal design tool, PASCE (Plant Application and Systems from Combustion Engineering), is comprised of intelligent databases that describe the design and from which accurate plant drawings are created. Additionally the IMS includes, at its hub, a relational database management system and an associated document management system. The data-based approach and applications associated with the IMS were developed, and have proven highly effective, for plant modifications, configuration management, and operations and maintenance applications at Duke Power Company's operating nuclear plants. This paper presents its major features and benefits. 4 refs

  7. Study concerning the power plant control and safety equipment by integrated distributed systems

    International Nuclear Information System (INIS)

    Optea, I.; Oprea, M.; Stanescu, P.

    1995-01-01

    The paper deals with the trends existing in the field of nuclear control and safety equipment and systems, proposing a high-efficiency integrated system. In order to enhance the safety of the plant and reliability of the structure system and components, we present a concept based on the latest computer technology with an open, distributed system, connected by a local area network with high redundancy. A modern conception for the control and safety system is to integrate all the information related to the reactor protection, active engineered safeguard and auxiliary systems parameters, offering a fast flow of information between all the agencies concerned so that situations can be quickly assessed. The integrated distributed control is based on a high performance operating system for realtime applications, flexible enough for transparent networking and modular for demanding configurations. The general design considerations for nuclear reactors instrumentation reliability and testing methods for real-time functions under dynamic regime are presented. Taking into account the fast progress in information technology, we consider the replacement of the old instrumentation of Cernavoda-1 NPP by a modern integrated system as an economical and efficient solution for the next units. (Author) 20 Refs

  8. Integrated equipment for increasing and maintaining coolant pressure in primary circuit of PWR nuclear power plant

    International Nuclear Information System (INIS)

    Sykora, D.

    1986-01-01

    An open heat pump circuit is claimed connected to the primary circuit. The pump circuit consists of a steam pressurizer with a built-in steam distributor, a compressor, an expander, a reducing valve, an auxiliary pump, and of water and steam pipes. The operation is described and a block diagram is shown of integrated equipment for increasing and maintaining pressure in the nuclear power plant primary circuit. The appropriate entropy diagram is also shown. The advantage of the open pump circuit consists in reducing the electric power input and electric power consumption for the steam pressurizers, removing entropy loss in heat transfer with high temperature gradient, in the possibility of inserting, between the expander and the auxiliary pump, a primary circuit coolant treatment station, in simplified design and manufacture of the high-pressure steam pressurizer vessel, reducing the weight of the steam pressurizer by changing its shape from cylindrical to spherical, increasing the rate of pressure growth in the primary circuit. (E.S.)

  9. Quality of care and investment in property, plant, and equipment in hospitals.

    Science.gov (United States)

    Levitt, S W

    1994-01-01

    OBJECTIVE. This study explores the relationship between quality of care and investment in property, plant, and equipment (PPE) in hospitals. DATA SOURCES. Hospitals' investment in PPE was derived from audited financial statements for the fiscal years 1984-1989. Peer Review Organization (PRO) Generic Quality Screen (GQS) reviews and confirmed failures between April 1989 and September 1990 were obtained from the Massachusetts PRO. STUDY DESIGN. Weighted least squares regression models used PRO GQS confirmed failure rates as the dependent variable, and investment in PPE as the key explanatory variable. DATA EXTRACTION. Investment in PPE was standardized, summed by the hospital over the six years, and divided by the hospital's average number of beds in that period. The number of PRO reviewed cases with one or more GQS confirmed failures was divided by the total number of cases reviewed to create confirmed failure rates. PRINCIPAL FINDINGS. Investment in PPE in Massachusetts hospitals is correlated with GQS confirmed failure rates. CONCLUSIONS. A financial variable, investment in PPE, predicts certain dimensions of quality of care in hospitals. PMID:8113054

  10. Research on U.S. nuclear power plant major equipment aging

    International Nuclear Information System (INIS)

    Nakos, J.T.; Rosinski, S.T.

    1994-01-01

    The U.S. Department of Energy (DOE) and the Electric Power Research Institute (EPRI), in cooperation with nuclear power plant utilities and the Nuclear Energy Institute, have prepared equipment aging evaluations of nuclear power plant equipment for life extension considerations. Specifically, these evaluations focused on equipment considered important for plant license renewal (U.S. Code of Federal Regulations 10CFR54). open-quotes Industry Reportsclose quotes (IRs), jointly funded by DOE and EPRI, evaluated the aging of major systems, structures, and components (e.g., reactor pressure vessels, Class I structures, PWR and BWR containments, etc.) and contain a mixture of technical and licensing information. open-quotes Aging Management Guidelinesclose quotes (AMGs), funded by DOE, evaluate aging for commodity types of equipment (e.g., pumps, electrical switchgear, heat exchangers, etc.) and concentrate on technical issues only. AMGs are intended for systems engineers and plant maintenance staff. A significant number of technical issues were resolved during IR interactions with the U.S. Nuclear Regulatory Commission (NRC). However, certain technical issues have not been resolved and are considered open-quotes openclose quotes. Examples include certain issues related to fatigue, neutron irradiation embrittlement, intergranular stress corrosion cracking (IGSCC) and electrical cable equipment qualification. Direct NRC interaction did not take place during preparation of individual AMGs due to their purely technical nature. The eventual use of AMGs in a future license renewal application will likely require NRC interaction at that time. With a few noted exceptions, the AMG process indicated that current aging management practices of U.S. utilities were effective in preventing age-related degradation. This paper briefly describes the IR and AMG processes and summarizes the unresolved technical issues identified through preparation of the documents

  11. MHI - Westinghouse joint FBR tank plant design

    International Nuclear Information System (INIS)

    Arnold, W.H.; Vijuk, R.M.; Aoki, I.; Messhil, T.

    1988-01-01

    Mitsubishi Heavy Industries and Westinghouse Advanced Energy Systems Division have combined their experience and capabilities to design a tank type fast breeder reactor plant. This tank type reactor has been refined and improved during the last three years to better compete in cost, safety, and operation with alternative power plants. This Mitsubishi/Westinghouse joint design offers economic advantages due to the use of steel structures, modular construction, nitrogen cells for the intermediate loops, reactor cavity air cooling and the use of the guard vessel as the containment vessel. Inherent characteristics in the reactor design provide protection to the public and the plant investment

  12. Challenges associated with the design of underground grinding plant at McArthur River project

    International Nuclear Information System (INIS)

    Jamrozek, J.S.

    2000-01-01

    McArthur River is an unique high grade uranium underground mine. Ore grinding and thickening are part of the underground operation. The grinding circuit is designed to operate in conditions different from conventional plant environments. Design of the grinding plant was a collective effort of a multi-disciplinary engineering team closely cooperating with project operating personnel. The equipment had to be selected to reflect widely varying ore properties. A user-friendly plant layout provides access to equipment inspections, services, and the delivery of necessary components. The size of the grinding chamber was limited in order to keep the rock stress levels within allowable values. All underground equipment brought to the construction site was restricted in size and weight. Plant construction faced limited storage space underground, tight erection sequencing, and schedule. Plant ventilation is a critical design feature. It efficiently removes radioactive dust from work areas, eliminates stagnant air pockets, and separates clean air from contaminated air areas. Radiation shielding on the equipment is designed to correspond with operational and maintenance functions. Plant operation is remotely controlled and requires little attendance. Video cameras are used on critical equipment and in controlled access areas. An extensive program of preventive and predictive maintenance allows highly reliable plant operation. (author)

  13. Applying technology to operator requirements in medical equipment design.

    Science.gov (United States)

    Woodring, P L

    1986-01-01

    The methodology used by the author consists of the following elements: Expose the design team to the user environment, followed by question and answer periods with users while still in the use environment. Place biomedical engineers in the leading teaching institutions where they will have day-to-day exposure to the use of products similar to the one being designed. Bring biomedical engineers back into the company as part of the design team. Expose concepts to focus groups while the product is in the definition stage. Bring a select group of users into the design review process. Evaluate the ease of use of the device as part of clinical trials. Establish a means of monitoring product performance after the product has been released. How well such a methodology will work in any particular environment is a function of management's recognition of the concept that ease of operator use is a vital element to the overall success of the product.

  14. Current fusion power plant design concepts

    International Nuclear Information System (INIS)

    Gore, B.F.; Murphy, E.S.

    1976-09-01

    Nine current U.S. designs for fusion power plants are described in this document. Summary tabulations include a tenth concept, for which the design document was unavailable during preparation of the descriptions. The information contained in the descriptions was used to define an envelope of fusion power plant characteristics which formed the basis for definition of reference first commercial fusion power plant design. A brief prose summary of primary plant features introduces each of the descriptions contained in the body of this document. In addition, summary tables are presented. These tables summarize in side-by-side fashion, plant parameters, processes, combinations of materials used, requirements for construction materials, requirements for replacement materials during operation, and production of wastes

  15. Manual on Safety Aspects of the Design and Equipment of Hot Laboratories

    International Nuclear Information System (INIS)

    1969-01-01

    With the development of atomic energy application and research, hot laboratories are now being constructed in a number of countries. The present publication describes and discusses experience in several countries in designing equipment for these laboratories. The safe handling of highly radioactive substances is the main purpose of hot laboratory design and equipment. The manual aims at helping those persons, particularly in the developing countries, who plan to design and construct a new hot laboratory or modify an existing one. It does not deal in great detail with the engineering design of protective and handling equipment; these matters can be found in the comprehensive list of references. The manual itself covers only basic ideas and different approaches in the design of laboratory building, hot cells, shielded and glove boxes, fume cupboards, and handling and viewing equipment. Systems for transferring materials and main services are also discussed.

  16. Upgrading of seismic design of nuclear power plant building

    Energy Technology Data Exchange (ETDEWEB)

    Akiyama, Hiroshi [Tokyo Univ. (Japan). Faculty of Engineering; Kitada, Yoshio

    1997-03-01

    In Japan seismic design methodology of nuclear power plant (NPP) structures has been established as introduced in the previous session. And yet efforts have been continued to date to upgrade the methodology, because of conservative nature given to the methodology in regard to unknown phenomena and technically-limited modeling involved in design analyses. The conservative nature tends to produce excessive safety margins, and inevitably send NPP construction cost up. Moreover, excessive seismic design can increase the burden on normal plant operation, though not necessarily contributing to overall plant safety. Therefore, seismic engineering has put to many tests and simulation analyses in hopes to rationalize seismic design and enhance reliability of seismic safety of NPPs. In this paper, we describe some studies on structural seismic design of NPP underway as part of Japan`s effort to upgrade existing seismic design methodology. Most studies described here are carried out by NUPEC (Nuclear Power Engineering Company) funded by MITI (the Ministry of International Trade and Industry Japan), though, similar studies with the same motive are also carrying out by nuclear industries such as utilities, NPP equipment and system manufacturers and building constructors. This paper consists of three sections, each introducing studies relating to NPP structural seismic design, new siting technology, and upgrading of the methodology of structural design analyses. (J.P.N.)

  17. Upgrading of seismic design of nuclear power plant building

    International Nuclear Information System (INIS)

    Akiyama, Hiroshi; Kitada, Yoshio.

    1997-01-01

    In Japan seismic design methodology of nuclear power plant (NPP) structures has been established as introduced in the previous session. And yet efforts have been continued to date to upgrade the methodology, because of conservative nature given to the methodology in regard to unknown phenomena and technically-limited modeling involved in design analyses. The conservative nature tends to produce excessive safety margins, and inevitably send NPP construction cost up. Moreover, excessive seismic design can increase the burden on normal plant operation, though not necessarily contributing to overall plant safety. Therefore, seismic engineering has put to many tests and simulation analyses in hopes to rationalize seismic design and enhance reliability of seismic safety of NPPs. In this paper, we describe some studies on structural seismic design of NPP underway as part of Japan's effort to upgrade existing seismic design methodology. Most studies described here are carried out by NUPEC (Nuclear Power Engineering Company) funded by MITI (the Ministry of International Trade and Industry Japan), though, similar studies with the same motive are also carrying out by nuclear industries such as utilities, NPP equipment and system manufacturers and building constructors. This paper consists of three sections, each introducing studies relating to NPP structural seismic design, new siting technology, and upgrading of the methodology of structural design analyses. (J.P.N.)

  18. How to Design and Equip a Mentalization-Based Play Therapy Room.

    Science.gov (United States)

    Rüth, Ulrich; Holch, Astrid

    2018-01-01

    Designing and equipping a play therapy room as a differentiated tool in a psychodynamic approach to child psychotherapy is seldom discussed. This article sketches out the equipment and furnishing of a play therapy room to be used for mentalization-based psychodynamic psychotherapy and gives examples of the use of such a room in practice.

  19. Designing efficient logging systems for northern hardwoods using equipment production capabilities and costs.

    Science.gov (United States)

    R.B. Gardner

    1966-01-01

    Describes a typical logging system used in the Lake and Northeastern States, discusses each step in the operation, and presents a simple method for designing and efficient logging system for such an operation. Points out that a system should always be built around the key piece of equipment, which is usually the skidder. Specific equipment types and their production...

  20. System design and equipment reliability for wide web working at Hem Heath Colliery

    Energy Technology Data Exchange (ETDEWEB)

    Hall, L

    1982-02-01

    This paper outlines the challenge to mining engineers on system design and reliability of equipment and how Hem Heath Colliery in the UK, is meeting this challenge, by describing current systems of work and equipment on the faces, summarising experience gained with succeeding faces, the results achieved, and posing considerations for the future.

  1. Design of mobile receiving and treatment equipment for radioactive liquid waste

    International Nuclear Information System (INIS)

    Kong Jinsong; Guo Weiqun; Lu Jingbin

    2012-01-01

    The advantage and disadvantage of radioactive liquid waste treatment technology are analyzed in this paper. The experimental disposal equipment for radioactive liquid waste with complicated sources is designed by combining the far infrared calcification technology with evaporation technology. It has advantages of low energy consuming and high decontamination efficiency. The frothy and dirt appear rarely in this equipment. (authors)

  2. RCC-E: Design and construction rules for electrical equipment of PWR nuclear islands

    International Nuclear Information System (INIS)

    2016-01-01

    RCC-E describes the rules for designing, building and installing electrical and I and C systems and equipment for pressurized water reactors. The code was drafted in partnership with industry, engineering firms, manufacturers, building control firms and operators, and represents a collection of best practices in accordance with IAEA requirements and IEC standards. The code's scope covers: architecture and the associated systems, materials engineering and the qualification procedure for normal and accidental environmental conditions, facility engineering and management of common cause failures (electrical and I and C) and electromagnetic interference, testing and inspecting electrical characteristics, quality assurance requirements supplementing ISO 9001 and activity monitoring. Use: RCC-E has been used to build the following power plants: France's last 12 nuclear units (1,300 MWe (8) and 1,450 MWe (4)), 2 M310 reactors in 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, 1 EPR reactor in France. RCC-E is used for maintenance operations in French power plants (58 units) and Chinese M310 and CPR-1000 power plants. RCC-E has been chosen for the construction of the EPR plants in Hinkley Point, UK. Contents of the 2016 edition of the RCC-E code: Volume 1 - General requirements and quality assurance; Volume 2 - Specification of requirements; Volume 3 - I and C systems; Volume 4 - Electrical systems; Volume 5 - Materials engineering; Volume 6 - Installation of electrical and I and C systems; Volume 7 - Inspection and test methods

  3. Stiffness Parameter Design of Suspension Element of Under-Chassis-Equipment for A Rail Vehicle

    Science.gov (United States)

    Ma, Menglin; Wang, Chengqiang; Deng, Hai

    2017-06-01

    According to the frequency configuration requirements of the vibration of railway under-chassis-equipment, the three- dimension stiffness of the suspension elements of under-chassis-equipment is designed based on the static principle and dynamics principle. The design results of the concrete engineering case show that, compared with the design method based on the static principle, the three- dimension stiffness of the suspension elements designed by the dynamic principle design method is more uniform. The frequency and decoupling degree analysis show that the calculation frequency of under-chassis-equipment under the two design methods is basically the same as the predetermined frequency. Compared with the design method based on the static principle, the design method based on the dynamic principle is adopted. The decoupling degree can be kept high, and the coupling vibration of the corresponding vibration mode can be reduced effectively, which can effectively reduce the fatigue damage of the key parts of the hanging element.

  4. Over facility design description for the CPDF [Centrifuge Plant Demonstration Facility]: SDD-1 [System Design Description

    International Nuclear Information System (INIS)

    1987-04-01

    The Centrifuge Plant Demonstration Facility (CPDF) is an essential part of the continuing development of first-production-plant centrifuge technology that will integrate centrifuge machines into a process and enrichment plant design. The CPDF will provide facilities for testing and continued development of a unit cascade in direct support of the commercial Gas Centrifuge Enrichment Plant (GCEP). The basic cascade-oriented equipment, feed, withdrawal, drive system, process piping, utility piping, and other auxiliary and support equipment will be tested in an operating configuration that represents, to the extent possible, GCEP arrangement and operating conditions. The objective will be to demonstrate procedures for production cascade installation, start-up, operation, and maintenance, and to provide proof of overall cascade and associated system design, construction, and operating and maintenance concepts. To the maximum possible extent, all equipment for the CPDF will be procured from commercial sources. Centrifuges will be procured from industry using government-supplied specifications and drawings. The existing Component Preparation Laboratory (CPL) located near the CPDF site will be used for centrifuge component receiving, inspection, assembly, and qualification testing of pre-production test machines. Later in the test program, samples of production machines planned for use in the GCEP will be tested in the CPDF

  5. IRIS Nuclear Power Plant design

    International Nuclear Information System (INIS)

    Carelli, M. D.; Cobian, J.

    2002-01-01

    IRIS(International Reactor Innovative and Secure) is a novel light water reactor with a modular, integral primary system configuration. This concept, initially developed in response to the first NERI solicitation, is now being pursued by an international consortium of 20 participants from seven countries. IRIS is designed to satisfy the four key requirements for Generation IV systems: enhanced safety, improved economics, proliferation resistance and waste minimization. Its main features are: small-to-medium power (100-335 MWe/module); long life core 5 to 10 years) without shuffling or refueling; optimized maintenance with repair shutdown intervals of a least four years; simplified compact design with the primary vessel housing steam generators, pressurizer and pumps; safety by design where accidents are positively eliminated by design rather than engineering to cope with their consequences; loss of coolant accidents of any size and loss of low accidents are eliminated as major safety concerns; estimated power generation total cost is projected to be competitive with other power options. IRIS is one of four new reactor designs currently under NRC review. Projected schedule calls for design certification by 2008 and being ready for deployment by 2001 or later. This rather short schedule is made possible by the fact that IRIS is based on proven light water technology and new technology development is not required. (Author)

  6. Emergency Mitigating Equipments - Post Fukushima Actions at Canadian Nuclear Power Plants - Portable AC Power Sources

    International Nuclear Information System (INIS)

    Vucetic, Jasmina; Kameswaran, R.

    2015-01-01

    In response to the Fukushima Daiichi NPP accident in 2011, the Canadian Nuclear Safety Commission set up a Task Force to evaluate operational, technical and regulatory implications on Canadian NPPs. While accepting that the risk from beyond-design-basis accidents (BDBA) at Canadian NPPs is very low, the Task Force identified a number of areas where additional improvements or confirmatory assessments would further enhance safety. As a result, a set of 36 Fukushima Action Items (FAIs) were assigned to the licensees. This paper focuses on the FAI related to electrical power system enhancements to address a total loss of all AC Power leading to a possibility of loss of heat sinks (i.e. Station Blackout). This required the licensees to implement the following: - Additional back up power supplied by portable diesel generator(s) to allow key instrumentation and control equipment and key electrical loads to remain operable; - Provisions for a storage and timely transportation and connection of the portable generator(s) to the applicable units; - Provisions for testing of the portable generator; - Provisions for fuelling of portable generators; - Provisions such as panels, receptacles, and connectors to quickly deploy the portable generators to plant system, and separate feeder cables route to avoid a common mode failure; - Load shedding strategy to extend the existing station's battery life to ensure that the connection of portable generators can be completed before the batteries are depleted; - Provisions to supply water to steam generators and Irradiated Fuel Bay using portable pumps; The paper will also provide a brief description of Electrical power systems of the Canadian NPPs designed to satisfy the high safety and reliability requirements for nuclear systems, which are based on the following: - 2 group design philosophy (Group 1 and Group 2 Electrical Power Systems) - 2 separate groups of onsite emergency generators (Class III Standby generators and Emergency

  7. Reliability analysis of mining equipment: A case study of a crushing plant at Jajarm Bauxite Mine in Iran

    International Nuclear Information System (INIS)

    Barabady, Javad; Kumar, Uday

    2008-01-01

    The performance of mining machines depends on the reliability of the equipment used, the operating environment, the maintenance efficiency, the operation process, the technical expertise of the miners, etc. As the size and complexity of mining equipments continue to increase, the implications of equipment failure become ever more critical. Therefore, reliability analysis is required to identify the bottlenecks in the system and to find the components or subsystems with low reliability for a given designed performance. It is important to select a suitable method for data collection as well as for reliability analysis. This paper presents a case study describing reliability and availability analysis of the crushing plant number 3 at Jajarm Bauxite Mine in Iran. In this study, the crushing plant number 3 is divided into six subsystems. The parameters of some probability distributions, such as Weibull, Exponential, and Lognormal distributions have been estimated by using ReliaSoft's Weibull++6 software. The results of the analysis show that the conveyer subsystem and secondary screen subsystem are critical from a reliability point of view, and the secondary crusher subsystem and conveyer subsystem are critical from an availability point of view. The study also shows that the reliability analysis is very useful for deciding maintenance intervals

  8. Quality assurance of analytical, scientific, and design computer programs for nuclear power plants

    International Nuclear Information System (INIS)

    1994-06-01

    This Standard applies to the design and development, modification, documentation, execution, and configuration management of computer programs used to perform analytical, scientific, and design computations during the design and analysis of safety-related nuclear power plant equipment, systems, structures, and components as identified by the owner. 2 figs

  9. Quality assurance of analytical, scientific, and design computer programs for nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-06-01

    This Standard applies to the design and development, modification, documentation, execution, and configuration management of computer programs used to perform analytical, scientific, and design computations during the design and analysis of safety-related nuclear power plant equipment, systems, structures, and components as identified by the owner. 2 figs.

  10. The optimization design of nuclear measurement teaching equipment

    International Nuclear Information System (INIS)

    Tang Rulong; Qiu Xiaoping

    2008-01-01

    So far domestic student-oriented experimental nuclear measuring instruments are used only to measure object density, thickness or material level, and in the choice of sources activity is mostly about 10 mCi. this design will proposed a optimization program dealing with domestic situation. It discussed the radioactive sources activity, the structural design of sealed sources, such as the choice of the tested material in order to get a program optimization. The program used 1 mCi activity radioactive sources 137 Cs to reduce the radiation dose, and the measurement function was improved. So that the apparatus can measure density, thickness nad material level. (authors)

  11. Strategy and implementation of resources control of key equipments in nuclear power plant

    International Nuclear Information System (INIS)

    Zha Qing

    2014-01-01

    The strategic resources of the construction of nuclear power plant, which include the main equipment of nuclear island, heavy forgings, the bottleneck equipment and strategic materials, is one of the key issues in the construction of nuclear power projects. The control of these strategic resources has become the focus of competition in industry and the major nuclear power groups are willing to fight for this huge advantages. The resource control strategies of key equipment of nuclear power projects are analyzed in this paper. This paper put forward specific measures and methods for the strategic resources control. By the application to a plurality of nuclear power engineering construction projects, these specific measures and methods achieved good results and will be with important guidance and reference for the construction of future nuclear power projects in China. (author)

  12. Computer methods in designing tourist equipment for people with disabilities

    Science.gov (United States)

    Zuzda, Jolanta GraŻyna; Borkowski, Piotr; Popławska, Justyna; Latosiewicz, Robert; Moska, Eleonora

    2017-11-01

    Modern technologies enable disabled people to enjoy physical activity every day. Many new structures are matched individually and created for people who fancy active tourism, giving them wider opportunities for active pastime. The process of creating this type of devices in every stage, from initial design through assessment to validation, is assisted by various types of computer support software.

  13. Gas Reactor International Cooperative program. Pebble bed reactor plant: screening evaluation. Volume 3. Appendix A. Equipment list

    International Nuclear Information System (INIS)

    1979-11-01

    This report consists of three volumes which describe the design concepts and screening evaluation for a 3000 MW(t) Pebble Bed Reactor Multiplex Plant (PBR-MX). The Multiplex plant produces both electricity and transportable chemical energy via the thermochemical pipeline (TCP). The evaluation was limited to a direct cycle plant which has the steam generators and steam reformers in the primary circuit. Volume 1 reports the overall plant and reactor system and was prepared by the General Electric Company. Core scoping studies were performed which evaluated the effects of annular and cylindrical core configurations, radial blanket zones, burnup, and ball heavy metal loadings. The reactor system, including the PCRV, was investigated for both the annular and cylindrical core configurations. Volume 3 is an Appendix containing the equipment list for the plant and was also prepared by United Engineers and Constructors, Inc. It tabulates the major components of the plant and describes each in terms of quantity, type, orientation, etc., to provide a basis for cost estimation

  14. Research program for seismic qualification of nuclear plant electrical and mechanical equipment. Task 3. Recommendations for improvement of equipment qualification methodology and criteria. Volume 3

    International Nuclear Information System (INIS)

    Kana, D.D.; Pomerening, D.J.

    1984-08-01

    The Research Program for Seismic Qualification of Nuclear Plant Electrical and Mechanical Equipment has spanned a period of three years and resulted in seven technical summary reports, each of which covered in detail the findings of different tasks and subtasks, and have been combined into five NUREG/CR volumes. Volume 3 presents recommendations for improvement of equipment qualification methodology and procedural clarification/modification. The fifth category identifies issues where adequate information does not exist to allow a recommendation to be made

  15. Comparison of the incidence of Listeria on equipment versus environmental sites within dairy processing plants.

    Science.gov (United States)

    Pritchard, T J; Flanders, K J; Donnelly, C W

    1995-08-01

    This study was undertaken to compare the incidence of Listeria contamination of processing equipment with that of the general dairy processing environment. A total of 378 sponge samples obtained from 21 dairy plants were analyzed for Listeria using three different enrichment media. Use of extended microbiological analysis allowed us to identify 26 Listeria positive sites which would have not been identified had a single test format been employed. Eighty (80) of 378 sites (21.2%) were identified as Listeria positive. Listeria innocua was isolated from 59 of the 80 (73.8%) positive samples, L. monocytogenes was identified in 35 (43.8%) of the positive samples, and L. seeligeri was isolated from 5 (6.3%) of the Listeria positive samples. Positive equipment samples were obtained from 6 of the 21 (28.6%) plants and 19 of the 21 (90.5%) plants had positive environmental sites. Seventeen of the 215 (7.9%) samples from equipment were positive for Listeria species. Eleven of these sites, including 3 holding tanks, 2 table tops, 3 conveyor/chain systems, a pasta filata wheel, a pint milk filler and a brine pre-filter machine, were positive for L. monocytogenes. Nineteen of the 21 (90.5%) plants had positive environmental sites. Sixty-three of the 163 (41.1%) samples from environmental sites were Listeria positive and 24 were positive for L. monocytogenes. Two-tailed student t-test analysis of the mean frequencies indicated that the level of contamination was significantly higher (p plant, and that greater emphasis needs to be placed on the cleaning and sanitizing of the plant environment.

  16. Development of a standard equipment management model for nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Hee Seung; Ju, Tae Young; Kim, Jung Wun [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    Most utilities that have achieved high performance have introduced a management model to improve performance and operate plants safely. The Nuclear Energy Institute has developed and updated its Standard Nuclear Performance Model (SNPM) in order to provide a summary of nuclear processes, cost definitions, and key business performance measures for business performance comparison and benchmarking. Over the past decade, Korea Hydro and Nuclear Power Co. (KHNP) has introduced and implemented many engineering processes such as Equipment Reliability (ER), Maintenance Rule (MR), Single Point Vulnerability (SPV), Corrective Action Program (CAP), and Self Assessment (SA) to improve plant performance and to sustain high performance. Some processes, however, are not well interfaced with other processes, because they were developed separately and were focused on the process itself. KHNP is developing a Standard Equipment Management Model (SEMM) to integrate these engineering processes and to improve the interrelation among the processes. In this paper, a draft model and attributes of the SEMM are discussed.

  17. Challenges and Prospects of Equipment Health Monitoring with Wireless Sensor Network in Nuclear Power Plants

    International Nuclear Information System (INIS)

    Chen, Dongyi; Jiang, Jin; Bari, Ataul; Wang, Quan; Hashemian, Hash-M.

    2014-01-01

    A wireless sensor network (WSN) system can offer tremendous benefits to equipment condition monitoring in newly-constructed and/or refurbished nuclear power plants (NPPs). However, it has not been widely accepted so far because of the following requirements by the NPP operators ectromagnetic (EM) emissions from the wireless transceivers must not interfere with the functionalities of the sensitive safety and protection systems in the plant, WSN must perform reliably in the presence of high levels of EM interference from devices such as relays and motor driven pumps, and ionizing radiation sources, dependable WSN performance in harsh industrial environments that are cluttered with cable trays, piping, valves, pumps, motors, and concrete and steel structures, and trict compliance with nuclear regulatory guidelines on EM emissions by the wireless devices. This paper will review the key issues associated with the deployment of WSN for equipment condition monitoring in NPPs. Some promising WSN technologies that can be used in NPP applications are also discussed

  18. Challenges and Prospects of Equipment Health Monitoring with Wireless Sensor Network in Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Dongyi [University of Electronic Science and Technology of China, Chengdu (China); Jiang, Jin; Bari, Ataul; Wang, Quan [University of Western Ontario, Ontario (Canada); Hashemian, Hash-M. [AMS Technology Center Knoxville (United States)

    2014-08-15

    A wireless sensor network (WSN) system can offer tremendous benefits to equipment condition monitoring in newly-constructed and/or refurbished nuclear power plants (NPPs). However, it has not been widely accepted so far because of the following requirements by the NPP operators ectromagnetic (EM) emissions from the wireless transceivers must not interfere with the functionalities of the sensitive safety and protection systems in the plant, WSN must perform reliably in the presence of high levels of EM interference from devices such as relays and motor driven pumps, and ionizing radiation sources, dependable WSN performance in harsh industrial environments that are cluttered with cable trays, piping, valves, pumps, motors, and concrete and steel structures, and trict compliance with nuclear regulatory guidelines on EM emissions by the wireless devices. This paper will review the key issues associated with the deployment of WSN for equipment condition monitoring in NPPs. Some promising WSN technologies that can be used in NPP applications are also discussed.

  19. Natural phenomena hazards evaluation of equipment and piping of Gaseous Diffusion Plant Uranium Enrichment Facility

    International Nuclear Information System (INIS)

    Singhal, M.K.; Kincaid, J.H.; Hammond, C.R.; Stockdale, B.I.; Walls, J.C.

    1995-01-01

    In support of the Gaseous Diffusion Plant Safety Analysis Report Upgrade program (GDP SARUP), a natural phenomena hazards evaluation was performed for the main process equipment and piping in the uranium enrichment buildings at Paducah and Portsmouth gaseous diffusion plants. In order to reduce the cost of rigorous analyses, the evaluation methodology utilized a graded approach based on an experience data base collected by SQUG/EPRI that contains information on the performance of industrial equipment and piping during past earthquakes. This method consisted of a screening walkthrough of the facility in combination with the use of engineering judgment and simple calculations. By using these screenings combined with evaluations that contain decreasing conservatism, reductions in the time and cost of the analyses were significant. A team of experienced seismic engineers who were trained in the use of the DOE SQUG/EPRI Walkdown Screening Material was essential to the success of this natural phenomena hazards evaluation

  20. Development of a standard equipment management model for nuclear power plants

    International Nuclear Information System (INIS)

    Chang, Hee Seung; Ju, Tae Young; Kim, Jung Wun

    2012-01-01

    Most utilities that have achieved high performance have introduced a management model to improve performance and operate plants safely. The Nuclear Energy Institute has developed and updated its Standard Nuclear Performance Model (SNPM) in order to provide a summary of nuclear processes, cost definitions, and key business performance measures for business performance comparison and benchmarking. Over the past decade, Korea Hydro and Nuclear Power Co. (KHNP) has introduced and implemented many engineering processes such as Equipment Reliability (ER), Maintenance Rule (MR), Single Point Vulnerability (SPV), Corrective Action Program (CAP), and Self Assessment (SA) to improve plant performance and to sustain high performance. Some processes, however, are not well interfaced with other processes, because they were developed separately and were focused on the process itself. KHNP is developing a Standard Equipment Management Model (SEMM) to integrate these engineering processes and to improve the interrelation among the processes. In this paper, a draft model and attributes of the SEMM are discussed

  1. Natural phenomena hazards evaluation of equipment and piping of Gaseous Diffusion Plant Uranium Enrichment Facility

    Energy Technology Data Exchange (ETDEWEB)

    Singhal, M.K.; Kincaid, J.H.; Hammond, C.R.; Stockdale, B.I.; Walls, J.C. [Oak Ridge National Lab., TN (United States). Technical Programs and Services; Brock, W.R.; Denton, D.R. [Lockheed Martin Energy Systems, Inc., Oak Ridge, TN (United States)

    1995-12-31

    In support of the Gaseous Diffusion Plant Safety Analysis Report Upgrade program (GDP SARUP), a natural phenomena hazards evaluation was performed for the main process equipment and piping in the uranium enrichment buildings at Paducah and Portsmouth gaseous diffusion plants. In order to reduce the cost of rigorous analyses, the evaluation methodology utilized a graded approach based on an experience data base collected by SQUG/EPRI that contains information on the performance of industrial equipment and piping during past earthquakes. This method consisted of a screening walkthrough of the facility in combination with the use of engineering judgment and simple calculations. By using these screenings combined with evaluations that contain decreasing conservatism, reductions in the time and cost of the analyses were significant. A team of experienced seismic engineers who were trained in the use of the DOE SQUG/EPRI Walkdown Screening Material was essential to the success of this natural phenomena hazards evaluation.

  2. Seismic design for Monju FBR power plant

    International Nuclear Information System (INIS)

    1982-01-01

    This technical report introduces the basic concept on the aseismatic design of the FBR ''Monju'' power station, of which the construction in Tsuruga is planned by the Power Reactor and Nuclear Fuel Development Corp. The safety design of Monju has been performed according to ''The concept of evaluating the safety of fast breeder reactors'', and the thought concerning the aseismatic design also is written in it. According to it, ''The guide for the examination of aseismatic design regarding power reactor facilities'' should be referred to, and the classification according to the importance in aseismatic design must be made, taking the features in the design of liquid metal-cooled FBRs fully in consideration. In the aseismatic design of Monju performed according to these basic concept, the following two points were examined. In the aseismatic design of the equipment and piping, the difference of construction from LWRs such as low pressure, thin walled and high temperature construction is taken in consideration. The classification according to the aseismatic importance of the system and equipment is made on the basis of the features in the design of Monju. The classification according to aseismatic importance, the method of calculating earthquake power, the combination of loads and the allowable limit, and the aseismatic construction of the main facilities are reported. (Kako, I.)

  3. Selection of construction materials for equipment in an experimental reprocessing plant

    International Nuclear Information System (INIS)

    Mizrahi, R.; Cragnolino, G.A.

    1994-01-01

    A review is made of the most significant corrosion problems that may be present in different stages of the process in a spent fuel reprocessing plant. The influence of different variables is analyzed: concentration of nitric acid and other oxidizing species, temperature, etc., in corrosion of materials of most frequent use in pipings and equipment. The materials are austenitic stainless steels and refractory metals, especially zirconium and its alloys. Both general and localized corrosion phenomena are analyzed for these materials. Selection criteria for the use of adequate material in different components of the plant are also discussed. (author). 32 refs., 20 figs., 3 tabs

  4. Design requirements, criteria and methods for seismic qualification of CANDU power plants

    International Nuclear Information System (INIS)

    Singh, N.; Duff, C.G.

    1979-10-01

    This report describes the requirements and criteria for the seismic design and qualification of systems and equipment in CANDU nuclear power plants. Acceptable methods and techniques for seismic qualification of CANDU nuclear power plants to mitigate the effects or the consequences of earthquakes are also described. (auth)

  5. Comparative research of finite element methods for perforated structures of nuclear power plant primary equipment

    International Nuclear Information System (INIS)

    Xiong Guangming; Deng Xiaoyun; Jin Ting

    2013-01-01

    Many perforated structures are used for nuclear power plant primary equipment, and they are complex, and have various forms. In order to explore the analysis and evaluation method, this paper used finite element method and equivalent analytic method to do the comparative analysis of perforated structures. The paper considered the main influence factors (including perforated forms, arrangements, and etc.), obtaining the systematic analysis methods of perforated structures. (authors)

  6. Process pump operating problems and equipment failures, F-Canyon Reprocessing Facility, Savannah River Plant

    International Nuclear Information System (INIS)

    Durant, W.S.; Starks, J.B.; Galloway, W.D.

    1987-02-01

    A compilation of operating problems and equipment failures associated with the process pumps in the Savannah River Plant F-Canyon Fuel Reprocessing Facility is presented. These data have been collected over the 30-year operation of the facility. An analysis of the failure rates of the pumps is also presented. A brief description of the pumps and the data bank from which the information was sorted is also included

  7. Use of complex electronic equipment within radiative areas of PWR power plants: feability study

    International Nuclear Information System (INIS)

    Fremont, P.; Carquet, M.

    1988-01-01

    EDF has undertaken a study in order to evaluate the technical and economical feasibility of using complex electronic equipment within radiative areas of PWR power plants. This study lies on tests of VLSI components (Random Access Memories) under gamma rays irradiations, which aims are to evaluate the radiation dose that they can withstand and to develop a selection method. 125 rad/h and 16 rad/h tests results are given [fr

  8. Design of a synthesizer for magnetic resonance equipment using FPGA

    International Nuclear Information System (INIS)

    Sonora A

    2006-01-01

    This paper exposes the design of a direct digital synthesizer in FPGA. This desing can generate a sine wave output up to 4MHZ with 3,33 mHz of precision. The frequency is set by 32bit word of phase increment in 350ns. The desing was made for Magnetic Resonance scanners and uses a 97% of logic resources of device. Functions for the synthesizer control are implemented in the same chip

  9. St. Louis demonstration final report: refuse processing plant equipment, facilities, and environmental evaluations

    Energy Technology Data Exchange (ETDEWEB)

    Fiscus, D.E.; Gorman, P.G.; Schrag, M.P.; Shannon, L.J.

    1977-09-01

    The results are presented of processing plant evaluations of the St. Louis-Union Electric Refuse Fuel Project, including equipment and facilities as well as assessment of environmental emissions at both the processing and the power plants. Data on plant material flows and operating parameters, plant operating costs, characteristics of plant material flows, and emissions from various processing operations were obtained during a testing program encompassing 53 calendar weeks. Refuse derived fuel (RDF) is the major product (80.6% by weight) of the refuse processing plant, the other being ferrous metal scrap, a marketable by-product. Average operating costs for the entire evaluation period were $8.26/Mg ($7.49/ton). The average overall processing rate for the period was 168 Mg/8-h day (185.5 tons/8-h day) at 31.0 Mg/h (34.2 tons/h). Future plants using an air classification system of the type used at the St. Louis demonstration plant will need an emissions control device for particulates from the large de-entrainment cyclone. Also in the air exhaust from the cyclone were total counts of bacteria and viruses several times higher than those of suburban ambient air. No water effluent or noise exposure problems were encountered, although landfill leachate mixed with ground water could result in contamination, given low dilution rates.

  10. Application and issues of online maintenance for equipment of nuclear power plants

    International Nuclear Information System (INIS)

    Higasa, Hisakazu

    2011-01-01

    The maintenance systems for long-term safety and repair costs reduction of equipment of nuclear power plants are stated. Planned maintenance contained the breakdown maintenance (BM) and the preventive maintenance, which consists of the time based maintenance (MBM) and the condition based maintenance (CBM). Explained are the characteristics of equipments, maintenance methods, maintenance solutions and the self-evaluation maintenance power, damage mechanism and solutions, and monitoring tools and application. Stated are the maintenance system and application of monitoring technology, periodical maintenance, application of diagnosis, vibration monitoring techniques, decision of vibration monitoring, and application of monitoring techniques for improvement of maintenance. Illustrated are realization of planned maintenance by reorganization of maintenance, a trend of maintenance of equipments, table of classified maintenance systems, change of maintenance program, maintenance data and investigation of damage mechanism, examples of self-evaluation maintenance power, examples of analysis of damage of parts of equipments, evaluation of rotating machines by vibration method, examples of results of diagnosis of bearing of rotating machines, online maintenance system of Asahi Kasei Engineering Corporation, degradation pattern of pomp, estimation of lifetime by total vibration and vibration on acceleration, and improvement of equipments. (S.Y.)

  11. Rules and procedures for the design and operation of hazardous research equipment

    International Nuclear Information System (INIS)

    1978-12-01

    The manual has been prepared for use by research personnel involved in experiments at the Lawrence Berkeley Laboratory. It contains rules and procedures for the design, test, installation, and operation of hazardous research equipment. Sect. I contains such information as responsibility of experimenters for safety, descriptions of the various Laboratory safety organizations, and enumeration of various services available to experimenters at the Laboratory. Sect. II describes specific rules for the setup and operation of experimental equipment at the Laboratory. Sect. III gives detailed design criteria and procedures for equipment frequently encountered in the high energy physics laboratory

  12. LBB considerations for a new plant design

    Energy Technology Data Exchange (ETDEWEB)

    Swamy, S.A.; Mandava, P.R.; Bhowmick, D.C.; Prager, D.E. [Westinghouse Electric Corp., Pittsburgh, PA (United States)

    1997-04-01

    The leak-before-break (LBB) methodology is accepted as a technically justifiable approach for eliminating postulation of Double-Ended Guillotine Breaks (DEGB) in high energy piping systems. This is the result of extensive research, development, and rigorous evaluations by the NRC and the commercial nuclear power industry since the early 1970s. The DEGB postulation is responsible for the many hundreds of pipe whip restraints and jet shields found in commercial nuclear plants. These restraints and jet shields not only cost many millions of dollars, but also cause plant congestion leading to reduced reliability in inservice inspection and increased man-rem exposure. While use of leak-before-break technology saved hundreds of millions of dollars in backfit costs to many operating Westinghouse plants, value-impacts resulting from the application of this technology for future plants are greater on a per plant basis. These benefits will be highlighted in this paper. The LBB technology has been applied extensively to high energy piping systems in operating plants. However, there are differences between the application of LBB technology to an operating plant and to a new plant design. In this paper an approach is proposed which is suitable for application of LBB to a new plant design such as the Westinghouse AP600. The approach is based on generating Bounding Analyses Curves (BAC) for the candidate piping systems. The general methodology and criteria used for developing the BACs are based on modified GDC-4 and Standard Review Plan (SRP) 3.6.3. The BAC allows advance evaluation of the piping system from the LBB standpoint thereby assuring LBB conformance for the piping system. The piping designer can use the results of the BACs to determine acceptability of design loads and make modifications (in terms of piping layout and support configurations) as necessary at the design stage to assure LBB for the, piping systems under consideration.

  13. Sustainable hydrogen - A challenge for materials science and equipment design

    International Nuclear Information System (INIS)

    Duta, Anca; Enesca, Alexandru Ioan; Perniu, Dana

    2006-01-01

    Full text: Hydrogen is the ideal fuel, considering its fully non-polluting by-products. Still, in discussions on 'sustainable hydrogen', there must be considered all the steps implied in hydrogen production, storage and use and the overall energy balance represents the real starting point of evaluating the sustainability. So far, hydrogen production is related to rather energy-consuming processes; extended research is devoted to develop high efficiency processes, but the industrial hydrogen production makes use of either large electrical or thermal energy amounts. Hydrogen production via water photolysis represents, consequently a viable alternative although many steps have to be elaborated to reached the industrial scale of these processes. Hydrogen storing represents another problem that affects its application; a safe storage way, in metal hydrides, is still under intensive research all over the world. The group of the Centre of Product Design for Sustainable Development is engaged in research for developing a laboratory photolyser, able to produce hydrogen and to offer an efficient storage alternative. The photolyser is a photo-electrochemical cell, and the efficiency of the photolysis process depends on several factors: - the photo-electrodes: thin films of wide band gap semiconductors with tailored properties; - the aqueous environment, with effect on the electrode materials properties and stability; - the external bias; - the cell design. The paper focuses mainly on the photo-electrode materials that were tested. The influence of the composition, crystalline and defect structure, of the morphology and of the interfaces on the photolysis process are reviewed. The effect of the pH in the aqueous media is discussed along with the stability of the materials and the reversibility of the adsorption/desorption processes. The design criteria that must be fulfilled in developing the photolyser are also discussed. (authors)

  14. Designing an object-oriented scheme for accelerator equipment control

    International Nuclear Information System (INIS)

    Trofimov, N.N.

    1994-01-01

    The object-oriented analysis (OOA) and design (OOD) formalisms developed by Project Technology Inc. is presented in this report. The method was originated in 1979 with the analysis phase of a large real-time project at Lawrence Berkeley Laboratory and since that time has been used in several large-scale industrial projects. It is supported by the Teamwork CASE package from Cadre Technologies Inc. For introducing basic concepts and definitions of the OOA/OOD, a typical accelerator device example: remotely controllable power supply which is used to power magnets in an accelerator beam transfer line is used through out this report. 18 refs., 18 figs

  15. Analog design of wireless control for home equipment

    Science.gov (United States)

    Zheng, Shiyong; Li, Zhao; Li, Biqing; Jiang, Suping

    2018-04-01

    This design consists of a STC89C52 microcontroller, a serial Bluetooth module and the Android system. Production of STC89C52 controlled by single-chip computer telephone systems. The system is composed of mobile phone Android system as a master in the family centre,via serial Bluetooth module pass instructions and information to implement wireless transceiver using STC89C52 MCU wireless Bluetooth transmission to control homedevices. System high reliability, low cost easy to use, stong applicability and other characerristics, can be used in single-user family, has great significance.

  16. Development of floor design response spectra for seismic design of floor-supported equipment or components, Revision 1, February 1978

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

    This guide presents an acceptable method for developing two horizontal and one vertical floor design response spectra at various floor or other equipment-support locations from the time-history motions resulting from the dynamic analysis of the supporting structure. These floor design response spectra are used in the dynamic analysis of systems or equipment supported at various locations of the supporting structure. Consulation has been provided by the Advisory Committee on Reactor Safeguards

  17. Supporting industrial equipment development through a set of design-for-maintenance guidelines

    NARCIS (Netherlands)

    Mulder, W.; Basten, Robertus Johannes Ida; Jauregui Becker, Juan Manuel; Blok, Jeroen; Hoekstra, Sipke; Kokkeler, Frans; Marjanovic, Dorian; Storga, Mario; Pavkovic, Neven; Bojcetic, Nenad

    2014-01-01

    This paper presents a leaflet with design-for-maintenance guidelines. It aims at supporting developers of industrial equipment in their design-for-maintenance practices. The use of this leaflet should lead to increased attention for design-for-maintenance aspects and to improve idea generation. The

  18. Production equipment development needs for a 700 metric ton/year light water reactor mixed oxide fuel manufacturing plant

    International Nuclear Information System (INIS)

    Blahnik, D.E.

    1977-09-01

    A literature search and survey of fuel suppliers was conducted to determine how much development of production equipment is needed for a 700 metric tons/y LWR mixed-oxide (UO 2 --PuO 2 ) fuel fabrication plant. Results indicate that moderate to major production equipment development is needed in the powder and pellet processing areas. The equipment in the rod and assembly processing areas need only minor development effort. Required equipment development for a 700 MT/y plant is not anticipated to delay startup of the plant. The development, whether major or minor, can be done well within the time frame for licensing and construction of the plant as long as conventional production equipment is used

  19. Design of 100 MW LNG Floating Barge Power Plant

    Directory of Open Access Journals (Sweden)

    I Made Ariana

    2017-06-01

    Full Text Available Floating bargepower plant able to supply amount of electricity to undeveloped island in Indonesia. In this research, the generator will be use in the power plant is dual-fuel engine. The process was determine the engine and every equipment along with its configuration then arrange the equipment. The result, MAN18V51/60DF selected along with its system configuration and its general arrangement. The final design enable 7.06 days of operation with daily average load (64.76 MW or 4.57 days with continues 100 MW load. In the end, the mobile power plant can be built on Damen B32SPo9832 Barge and comply with the regulation floating bargepower plant able to supply amount of electricity to undeveloped island in Indonesia. In this research, the generator will be use in the power plant is dual-fuel engine. The process was determine the engine and every equipment along with its configuration then arrange the equipment. The result, MAN18V51/60DF selected along with its system configuration and its general arrangement. The final design enable 7.06 days of operation with daily average load (64.76 MW or 4.57 days with continues 100 MW load. In the end, the mobile power plant can be built on Damen B32SPo9832 Barge and comply with the regulation Normal 0 false false false EN-US X-NONE X-NONE /* Style Definitions */ table.MsoNormalTable {mso-style-name:"Table Normal"; mso-tstyle-rowband-size:0; mso-tstyle-colband-size:0; mso-style-noshow:yes; mso-style-priority:99; mso-style-qformat:yes; mso-style-parent:""; mso-padding-alt:0in 5.4pt 0in 5.4pt; mso-para-margin:0in; mso-para-margin-bottom:.0001pt; mso-pagination:widow-orphan; font-size:11.0pt; font-family:"Calibri","sans-serif"; mso-ascii-font-family:Calibri; mso-ascii-theme-font:minor-latin; mso-fareast-font-family:"Times New Roman"; mso-fareast-theme-font:minor-fareast; mso-hansi-font-family:Calibri; mso-hansi-theme-font:minor-latin; mso-bidi-font-family:"Times New Roman"; mso-bidi-theme-font:minor-bidi;}

  20. Evaluation of Design Models of Process Equipment for Use in PRIDE: Remote Operability and Maintainability

    International Nuclear Information System (INIS)

    Kim, Ki Ho; Kim, Sung Hyun; Yu, Seung Nam; Lee, Jong Kwang; Park, Byung Suk; Han, Jong Hui; Cho, Il Je; Lee, Han Soo

    2012-01-01

    Process equipment for pyroprocessing are being developed at KAERI (Korea Atomic Energy Research Institute). Those equipment should be operated and maintained in a fully remote manner in the argon gas filled cell of PRIDE (PyRoprocess Integrated inactive DEmonstration facility) at KAERI because direct human access to the in-cell is not possible during an operation due to the high toxicity of the argon gas. To make such process equipment remotely operable and maintainable, their design developments have been tested and evaluated in a simulator before they are constructed. A simulator as a means of evaluating the remote operability and maintainability of the design models of process equipment for pyroprocessing is described, and results of the design models tested and evaluated in a simulator are presented

  1. SEPI an expert system for plant design

    International Nuclear Information System (INIS)

    Carotenuto, M.; Corleto, P.; Landeyro, P.

    1988-01-01

    The availability and suitability of technological information is of great importance in every kind of design task, especially when safety and reliability considerations are involved. In this paper an ''expert system for plant design'' (SEPI), is presented, together with its first application to nuclear back-end plants. This system is available on ENEA computer network. It is thought to be used both to collect know-how developed in the field and to assist unskilled designers during selection, evaluation and dimensioning tasks. It attemps to reproduce the normal way of ''reasoning'' and acting, and provides some graphic facilities

  2. Conceptual designs of pressurized fluidized bed and pulverized coal fired power plants

    International Nuclear Information System (INIS)

    Doss, H.S.; Bezella, W.A.; Hamm, J.R.; Pietruszkiewicz, J.

    1984-01-01

    This paper presents the major technical and economic characteristics of steam and air-cooled pressurized fluidized bed (PFB) power plant concepts, along with the characteristics of a pulverized coal fired power plant equipped with an adipic acid enhanced wet-limestone flue gas desulfurization system. Conceptual designs for the three plants were prepared to satisfy a set of common groundrules developed for the study. Grassroots plants, located on a generic plant site were assumed. The designs incorporate technologies projected to be commercial in the 1990 time frame. Power outputs, heat rates, and costs are presented

  3. The method of executing the vibration tendency management of the intermittent driving equipment in the nuclear plant

    International Nuclear Information System (INIS)

    Yonekawa, Yutaka; Fukunaga, Tatsuya

    2008-01-01

    The main rotary machine is often an intermittent driving machine in the nuclear plant. On the other hand, it was a problem for the vibration method to detect the vibration when rotating, and very to achieve the vibration tendency management for the equipment that did not rotate though it positively worked on the introduction of the equipment diagnosis technology by the vibration method of the rotation equipment in the nuclear plant. This time, because the tendency management system of the intermittent driving equipment is developed, and the tendency management was achieved, it introduces the outline and an actual case. (author)

  4. Nuclear power plant C and I design verification by simulation

    International Nuclear Information System (INIS)

    Storm, Joachim; Yu, Kim; Lee, D.Y

    2003-01-01

    An important part of the Advanced Boiling Water Reactor (ABWR) in the Taiwan NPP Lungmen Units no.1 and no.2 is the Full Scope Simulator (FSS). The simulator was to be built according to design data and therefore, apart from the training aspect, a major part of the development is to apply a simulation based test bed for the verification, validation and improvement of plant design in the control and instrumentation (C and I) areas of unit control room equipment, operator Man Machine Interface (MMI), process computer functions and plant procedures. Furthermore the Full Scope Simulator will be used after that to allow proper training of the plant operators two years before Unit no.1 fuel load. The article describes scope, methods and results of the advanced verification and validation process and highlights the advantages of test bed simulation for real power plant design and implementation. Subsequent application of advanced simulation software tools like instrumentation and control translators, graphical model builders, process models, graphical on-line test tools and screen based or projected soft panels, allowed a team to fulfil the task of C and I verification in time before the implementation of the Distributed Control and Information System (DCIS) started. An additional area of activity was the Human Factors Engineering (HFE) for the operator MMI. Due to the fact that the ABWR design incorporates a display-based operation with most of the plant components, a dedicated verification and validation process is required by NUREG-0711. In order to support this activity an engineering test system had been installed for all the necessary HFE investigations. All detected improvements had been properly documented and used to update the plant design documentation by a defined process. The Full Scope Simulator (FSS) with hard panels and stimulated digital control and information system are in the final acceptance test process with the end customer, Taiwan Power Company

  5. Hot cell design in the vitrification plant China

    International Nuclear Information System (INIS)

    Jiang Yubo; Wang Guangkai; Zhang Wei; Liang Runan; Dou Yuan

    2015-01-01

    In the area of reprocessing and radioactive waste management, gloveboxes and cells are a kind of non-standard equipments providing an isolated room to operate radioactive material inside, while the operator outside with essential biological shield and protection. The hot cell is a typical one, which could handle high radioactive material with various operating means and tight enclosure. The dissertation is based on Vitrification Plant China, a cooperation project between China and Germany. For the sino-western difference in design philosophy, it was presented how to draft an acceptable design proposal of applicable huge hot cells by analysing the design requirements, such as radioprotection, observation, illumination, remote handling, transportation, maintenance and decontamination. The construction feasibility of hot cells was also approved. Thanks to 3D software Autodesk Inventor, digital hot cell was built to integrate all the interfaces inside, which validated the design by checking the mechanical interference. (author)

  6. Enhanced design, operation and maintenance practices for a longer plant service life

    International Nuclear Information System (INIS)

    Raimondo, E.; Courcoux, A.

    2004-01-01

    Plant service life problems have been under detailed investigation in France and the experience acquired by our company over the past 25 years in the design, construction and maintenance of Pressurized Water Reactors has contributed to develop skills, equipment and capabilities available for efficient plant aging management and component service life extension. The service life of a nuclear power plant is deeply dependant of the provisions made during the design stage, directly linked to good operating conditions and adequate maintenance practices. This paper presents the importance of these three steps (design, operation and maintenance) for plant service life concern. (author)

  7. Upgrading instrumentation and control in nuclear power plants. Design criteria

    International Nuclear Information System (INIS)

    Rodriguez Rodriguez, M.C.; Alvarez Menendez, A.

    1997-01-01

    The use of programmed digital technology in Protection, Control, Monitoring and Information Systems in new generation nuclear power plants, or the use of this technology to replace or upgrade existing systems based on wired analog instrumentation and electromechanical relays, has led to new international standards which establish new design requirements or adapt existing requirements to this technology. Additionally, both regulatory organisations and the industry are discussing the reliability of this technology, regarding common mode failures that may occur in redundant protection channels, due to the use of equipment and software with the same characteristics. The first part of this paper addresses the most important aspects of new international standards regarding classification criteria for I and C systems, equipment and functions, depending on their importance to safety and the design criteria applicable to each category. Special attention is drawn to requirements concerning software quality assurance and the design of new control rooms. The paper then goes on to discuss the different technical solutions being implemented, using equipment and software diversification, in order to prevent the possibility of common mode failures affecting the protection function. (Author)

  8. The effect of plant aging on equipment qualification and human performance issues related to license renewal

    International Nuclear Information System (INIS)

    Gunther, W.E.; Higgins, J.C.; Aggarwal, S.K.

    1991-01-01

    The aging of nuclear power plants is one of the most important issues facing the nuclear industry worldwide. Aging encompasses as forms of degradation to nuclear power plant components, systems, and structures that result from exposure to environmental conditions or from operational stresses. Both the degradation from aging and actions taken to address the aging, such as increased maintenance and testing, can significantly impact human performance in the plant. Research into the causes and effects of aging as obtained through the assessment of operating experience and testing have raised questions regarding the adequacy of existing industry standards for addressing the concerns raised by this research. This paper discusses these issues, with particular emphasis in the area of equipment qualification and human performance

  9. The effect of plant aging on equipment qualification and human performance issues related to license renewal

    Energy Technology Data Exchange (ETDEWEB)

    Gunther, W.E.; Higgins, J.C. [Brookhaven National Lab., Upton, NY (United States); Aggarwal, S.K. [Nuclear Regulatory Commission, Washington, DC (United States)

    1991-12-31

    The aging of nuclear power plants is one of the most important issues facing the nuclear industry worldwide. Aging encompasses as forms of degradation to nuclear power plant components, systems, and structures that result from exposure to environmental conditions or from operational stresses. Both the degradation from aging and actions taken to address the aging, such as increased maintenance and testing, can significantly impact human performance in the plant. Research into the causes and effects of aging as obtained through the assessment of operating experience and testing have raised questions regarding the adequacy of existing industry standards for addressing the concerns raised by this research. This paper discusses these issues, with particular emphasis in the area of equipment qualification and human performance.

  10. The effect of plant aging on equipment qualification and human performance issues related to license renewal

    Energy Technology Data Exchange (ETDEWEB)

    Gunther, W.E.; Higgins, J.C. (Brookhaven National Lab., Upton, NY (United States)); Aggarwal, S.K. (Nuclear Regulatory Commission, Washington, DC (United States))

    1991-01-01

    The aging of nuclear power plants is one of the most important issues facing the nuclear industry worldwide. Aging encompasses as forms of degradation to nuclear power plant components, systems, and structures that result from exposure to environmental conditions or from operational stresses. Both the degradation from aging and actions taken to address the aging, such as increased maintenance and testing, can significantly impact human performance in the plant. Research into the causes and effects of aging as obtained through the assessment of operating experience and testing have raised questions regarding the adequacy of existing industry standards for addressing the concerns raised by this research. This paper discusses these issues, with particular emphasis in the area of equipment qualification and human performance.

  11. Increasing reliability of nuclear energy equipment and at nuclear power plants

    International Nuclear Information System (INIS)

    Ochrana, L.

    1997-01-01

    The Institute of Nuclear Energy at the Technical University in Brno cooperates with nuclear power plants in increasing their reliability. The teaching programme is briefly described. The scientific research programme of the Department of Heat and Nuclear Power Energy Equipment in the field of reliability is based on a complex systematic concept securing a high level of reliability. In 1996 the Department prepared a study dealing with the evaluation of the maintenance system in a nuclear power plant. The proposed techniques make it possible to evaluate the reliability and maintenance characteristics of any individual component in a nuclear power plant, and to monitor, record and evaluate data at any given time intervals. (M.D.)

  12. Advanced power plant materials, design and technology

    Energy Technology Data Exchange (ETDEWEB)

    Roddy, D. (ed.) [Newcastle University (United Kingdom). Sir Joseph Swan Institute

    2010-07-01

    The book is a comprehensive reference on the state of the art of gas-fired and coal-fired power plants, their major components and performance improvement options. Selected chapters are: Integrated gasification combined cycle (IGCC) power plant design and technology by Y. Zhu, and H. C. Frey; Improving thermal cycle efficiency in advanced power plants: water and steam chemistry and materials performance by B. Dooley; Advanced carbon dioxide (CO{sub 2}) gas separation membrane development for power plants by A. Basile, F. Gallucci, and P. Morrone; Advanced flue gas cleaning systems for sulphur oxides (SOx), nitrogen oxides (NOx) and mercury emissions control in power plants by S. Miller and B.G. Miller; Advanced flue gas dedusting systems and filters for ash and particulate emissions control in power plants by B.G. Miller; Advanced sensors for combustion monitoring in power plants: towards smart high-density sensor networks by M. Yu and A.K. Gupta; Advanced monitoring and process control technology for coal-fired power plants by Y. Yan; Low-rank coal properties, upgrading and utilisation for improving the fuel flexibility of advanced power plants by T. Dlouhy; Development and integration of underground coal gasification (UCG) for improving the environmental impact of advanced power plants by M. Green; Development and application of carbon dioxide (CO{sub 2}) storage for improving the environmental impact of advanced power plants by B. McPherson; and Advanced technologies for syngas and hydrogen (H{sub 2}) production from fossil-fuel feedstocks in power plants by P. Chiesa.

  13. Design of a nuclear steam reforming plant

    International Nuclear Information System (INIS)

    Malherbe, J.

    1980-01-01

    The design of a plant for the steam reforming of methane using a High Temperature Reactor has been studied by CEA in connection with the G.E.G.N. This group of companies (CEA, GAZ DE FRANCE, CHARBONNAGES DE FRANCE, CREUSOT-LOIRE, NOVATOME) is in charge of studying the feasibility of the coal gasification process by using a nuclear reactor. The process is based on the hydrogenation of the coal in liquid phase with hydrogen produced by a methane steam reformer. The reformer plant is fed by a pipe of natural gas or SNG. The produced hydrogen feeds the gasification plant which could not be located on the same site. An intermediate hydrogen storage between the two plants could make the coupling more flexible. The gasification plant does not need a great deal of heat and this heat can be satisfied mostly by internal heat exchanges

  14. Designing for nuclear power plant maintainability and operability

    International Nuclear Information System (INIS)

    Pedersen, T.J.

    1998-01-01

    Experience has shown that maintenance and operability aspects must be addressed in the design work. ABB Atom has since long an ambition of achieving optimised, overall plant designs, and efficient feedback of growing operating experience has stepwise eliminated shortcomings, and yielded better and better plant operating performances. The records of the plants of the latest design versions are very good; four units in Sweden have operated at an energy availability of 90.1%, and the two Olkiluoto units in Finland at a load factor of 92.7%, over the last decade. The occupational radiation exposures have also been at a low level. The possibilities for implementing 'lessons learned' in existing plants are obviously limited by practical constraints. In Finland and Sweden, significant modernisations are still underway, however, involving replacement of mechanical equipment, and upgrading and backfitting of I and C systems on a large scale, in most of the plants. The BWR 90 design focuses on meeting requirements from utilities as well as new regulatory requirements, with a particular emphasis on the consequences of severe accidents; there shall be no large releases to the environment. Other design improvements involve: all-digital I and C systems and enhanced human factors engineering to improve work environment for operators, optimisation of buildings and containment to decrease construction time and costs, and selection of materials as well as maintenance of operating procedures to reduce radiation exposures even further. The BWR 90 design was offered to Finland in the early 1990s, but development work continues. It has been selected by a number of European utilities for assessing its conformance with the European Utility Requirements (EUR), aiming at a specific EUR Volume 3 for the BWR 90. Some characteristics of the ABB BWRs, with emphasis on features of importance for achieving improved economy and enhanced safety, are described below. (author)

  15. Human-centered design of the human-system interfaces of medical equipment: thyroid uptake system

    International Nuclear Information System (INIS)

    Monteiro, Jonathan K.R.; Farias, Marcos S.; Santos, Isaac J.A. Luquetti; Monteiro, Beany G.

    2013-01-01

    Technology plays an important role in modern medical centers, making healthcare increasingly complex, relying on complex technical equipment. This technical complexity is particularly noticeable in the nuclear medicine. Poorly design human-system interfaces can increase the risks for human error. The human-centered approach emphasizes the development of the equipment with a deep understanding of the users activities, current work practices, needs and abilities of the users. An important concept of human-centered design is that the ease-of-use of the equipment can be ensured only if users are actively incorporated in all phases of the life cycle of design process. Representative groups of users are exposed to the equipment at various stages in development, in a variety of testing, evaluation and interviewing situations. The users feedback obtained is then used to refine the design, with the result serving as input to the next interaction of design process. The limits of the approach are that the users cannot address any particular future needs without prior experience or knowledge about the equipment operation. The aim of this paper is to present a methodological framework that contributes to the design of the human-system interfaces, through an approach related to the users and their activities. A case study is described in which the methodological framework is being applied in development of new human-system interfaces of the thyroid uptake system. (author)

  16. Evaluation formulas of manpower needs for dismantling of equipments in uranium refining and conversion plant

    International Nuclear Information System (INIS)

    Izumo, Sari; Usui, Hideo; Kubota, Shintaro; Tachibana, Mitsuo; Kawagoshi, Hiroshi; Tokuyasu, Takashi; Takahashi, Nobuo; Morimoto, Yasuyuki; Tanaka, Yoshio; Sugitsue, Noritake

    2014-07-01

    Japan Atomic Energy Agency has developed PROject management data evaluation code for DIsmantling Activities (PRODIA) to make an efficient decommissioning for nuclear facilities. PRODIA is a source code which provides estimated value such as manpower needs, costs, etc., for dismantling by evaluation formulas according to the type of nuclear facility. Evaluation formulas of manpower needs for dismantling of equipments about reprocessed uranium conversion in Uranium Refining and Conversion Plant (URCP) have been developed in this report. In the result, evaluation formulas of manpower needs for dismantling of equipment were derived based on the classifications of equipment's functions or work items. These evaluation formulas are widely applicable to the estimation of the manpower needs for dismantling the other nuclear facilities, in particular uranium handling facilities. It was confirmed that some of these evaluation formulas with the same applicable condition could be unified to some inclusive evaluation formulas. It turned out that all steel equipment contaminated by uranium could be evaluated by one evaluation formula. (author)

  17. In situ measurement of dynamic characteristics of atomic power plant equipment

    International Nuclear Information System (INIS)

    Arya, A.S.; Gupta, S.P.; Shrivastava, S.K.

    1977-01-01

    For the realistic assessment of stiffness and damping, full scale free vibration tests have been carried out on various pieces of equipment located in plant buildings both during the construction stage and after they are erected. Initial displacement or initial velocity was used to excite the free vibrations. Initial displacement was imparted by means of steel rope pulled with chain pulley block. The sudden release was achieved by means of a clutch system. Acceleration transducer with amplifier and ink writting oscillograph was used for recording the vibrations. Frequency and damping was evaluated from the acceleration records. Observed values for some equipment are given. For some equipment, it has been possible to obtain the values with and without pipe connections. The frequency of L.P. Heater in longitudinal and transverse directions without pipe connection were 17.86 and 10.04 Hz but with pipe connections the values increased to 26.74 and 17.85 Hz. Similarly there has been increase in the damping values too. Thus both the frequency and damping increases substantially with the addition of pipe connections. Moreover, their values are quite different in the two principal directions, pointing out to the importance of in situ measurements on prototype equipment

  18. Underwater fuel handling equipment maintenance. Verification of design assumptions, specific problems and tools, case study

    International Nuclear Information System (INIS)

    Kurek, J.B.

    1995-01-01

    The majority of CANDU Fuel Transfer System equipment at Pickering is located under fourteen feet of water, as dictated by the containment and shielding requirements. Such arrangement, however, creates specific problems with equipment maintenance. Each single piece of equipment serves two generating units, which means in case of defect- double losses on production, or two units shut down simultaneously for planned maintenance. The requirement for underwater maintenance was not anticipated at the design stage, which multiples the level of difficulty, and creates requirement for developing special tools for each work. Removal of the damaged fuel from the receiving bays and decontamination of submerged equipment is also part of the problem. The purpose of this presentation is to share our experience with the designers, operators, maintenance mechanics and technical personnel of the other CANDU generating stations

  19. Overview of seismic resistant design of Indian Nuclear Power Plants

    International Nuclear Information System (INIS)

    Sharma, G.K.; Hawaldar, R.V.K.P.; Vinod Kumar

    2007-01-01

    Safe operation of a Nuclear Power Plant (NPP) is of utmost importance. NPPs consist of various Structure, System and Equipment (SS and E) that are designed to resist the forces generated due to a natural phenomenon like earthquake. An earthquake causes severe oscillatory ground motion of short duration. Seismic resistant design of SS and E calls for evaluation of effect of severe ground shaking for assuring the structural integrity and operability during and after the occurrence of earthquake event. Overall exercise is a multi-disciplinary approach. First of standardized 220 MWe design reactor is Narora Atomic Power Station. Seismic design was carried out as per state of art then, for the first time. The twelve 220 MWe reactors and two 540 MWe reactors designed since 1975 have been seismically qualified for the earthquake loads expected in the region. Seismic design of 700 MWe reactor is under advanced stage of finalization. Seismic re-evaluation of six numbers of old plants has been completed as per latest state of art. Over the years, expertise have been developed at Nuclear Power Corporation of India Limited, Bhabha Atomic Research Centre, prominent educational institutes, research laboratories and engineering consultants in the country in the area of seismic design, analysis and shake table testing. (author)

  20. Design scheme of automatic feeding equipment of domestic uranium chemical concentrate

    International Nuclear Information System (INIS)

    Hu Jinming; Wang Chao; Peng Jinhui; Zhang Libo

    2014-01-01

    In order to solve problems by artificial feeding mode with low work efficiency, large intensity manual labor and environmental pollution in domestic uranium concentrate purification process, the design scheme of automatic feeding device was set up, including work flow sheet, composition of automatic equipment and operation. By application of automatic feeding equipment, the feeding speed can be greatly increased, labor force can be reduced, and harm to workman health can be decreased. (authors)

  1. Development of automated equipment for reduction of personnel radiation exposure in nuclear power plants

    International Nuclear Information System (INIS)

    Ogushi, Akira; Fujii, Masaaki; Mizuno, Katsuhiro.

    1976-01-01

    Described are a mobile remote inspection system and an automatic analyzer for radioactive nuclides in reactor coolant now being developed as a means of reducing personnel radiation exposure in nuclear power plants. In the mobile remote inspection system ''TELEPAT'', a self-propelled vehicle equipped with a thermometer, accelerometer, microphone, ionization chamber, etc. is remote operated from the main control room to inspect the equipment in the reactor building. The automatic analyzer for radioactive nuclides in reactor coolant automates the series of operations ranging from sampling of reactor coolant to measurement of radioactivity and analyses of measured data, with a view to saving labor in radioactivity analysis work while reducing exposure of personnel to radiation. (auth.)

  2. Recent developments in methodology for dynamic qualification of nuclear plant equipment

    International Nuclear Information System (INIS)

    Kana, D.D.; Pomerening, D.J.

    1984-01-01

    Dynamic qualification of nuclear plant electrical and mechanical equipment is performed basically under guidelines given in IEEE Standards 323 and 344, and a variety of NRC regulatory guides. Over the last fifteen years qualification methodology prescribed by these documents has changed significantly as interpretations, equipment capability, and imagination of the qualification engineers have progressed. This progress has been sparked by concurrent NRC and industry sponsored research programs that have identified anomalies and developed new methodologies for resolving them. Revisions of the standards have only resulted after a lengthy debate of all such new information and subsequent judgment of its validity. The purpose of this paper is to review a variety of procedural improvements and developments in qualification methodology that are under current consideration as revisions to the standards. Many of the improvements and developments have resulted from recent research programs. All are very likely to appear in one type of standard or another in the near future

  3. Design criteria for the new waste calcining facility at the Idaho Chemical Processing Plant

    International Nuclear Information System (INIS)

    Anderson, F.H.; Bingham, G.E.; Buckham, J.A.; Dickey, B.R.; Slansky, C.M.; Wheeler, B.R.

    1976-01-01

    The New Waste Calcining Facility (NWCF) at the Idaho Chemical Processing Plant (ICPP) is being built to replace the existing fluidized-bed, high-level waste calcining facility (WCF). Performance of the WCF is reviewed, equipment failures in WCF operation are examined, and pilot-plant studies on calciner improvements are given in relation to NWCF design. Design features of the NWCF are given with emphasis on process and equipment improvements. A major feature of the NWCF is the use of remote maintenance facilities for equipment with high maintenance requirements, thereby reducing personnel exposures during maintenance and reducing downtime resulting from plant decontamination. The NWCF will have a design net processing rate of 11.36 m 3 of high-level waste per day, and will incorporate in-bed combustion of kerosene for heating the fluidized bed calciner. The off-gas cleaning system will be similar to that for the WCF

  4. Tolerance-based Structural Design of Tubular-Structure Loading Equipments

    Directory of Open Access Journals (Sweden)

    Jiping Lu

    2011-05-01

    is worked out under different ball screws, trapezoidal screw threads, worm and worm gears. To meet the requirement of tolerance in tubular-structure assembly, mechanisms for all motions are defined. The design of loading equipment is tested and assessed by experiments, and the result shows the design is highly qualified for its assembly.

  5. Design/Operations review of core sampling trucks and associated equipment

    International Nuclear Information System (INIS)

    Shrivastava, H.P.

    1996-01-01

    A systematic review of the design and operations of the core sampling trucks was commissioned by Characterization Equipment Engineering of the Westinghouse Hanford Company in October 1995. The review team reviewed the design documents, specifications, operating procedure, training manuals and safety analysis reports. The review process, findings and corrective actions are summarized in this supporting document

  6. Prediction of Maintenance Period of Equipment Through Risk Assessment of Thermal Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Song, Gee Wook; Kim, Bum Shin; Choi, Woo Song; Park, Myung Soo [KEPCO Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    Risk-based inspection (RBI) is a well-known method that is used to optimize inspection activities based on risk analysis in order to identify the high-risk components of major facilities such as power plants. RBI, when implemented and maintained properly, improves plant reliability and safety while reducing unplanned outages and repair costs. Risk is given by the product of the probability of failure (Pof) and the consequence of failure (COF). A semi-quantitative method is generally used for risk assessment. Semi-quantitative risk assessment complements the low accuracy of qualitative risk assessment and the high expense and long calculation time of quantitative risk assessment. The first step of RB I is to identify important failure modes and causes in the equipment. Once these are defined, the Pof and COF can be assessed for each failure. During Pof and COF assessment, an effective inspection method and range can be easily found. In this paper, the calculation of the Pof is improved for accurate risk assessment. A modified semi-quantitative risk assessment was carried out for boiler facilities of thermal power plants, and the next maintenance schedules for the equipment were decided.

  7. Prediction of Maintenance Period of Equipment Through Risk Assessment of Thermal Power Plants

    International Nuclear Information System (INIS)

    Song, Gee Wook; Kim, Bum Shin; Choi, Woo Song; Park, Myung Soo

    2013-01-01

    Risk-based inspection (RBI) is a well-known method that is used to optimize inspection activities based on risk analysis in order to identify the high-risk components of major facilities such as power plants. RBI, when implemented and maintained properly, improves plant reliability and safety while reducing unplanned outages and repair costs. Risk is given by the product of the probability of failure (Pof) and the consequence of failure (COF). A semi-quantitative method is generally used for risk assessment. Semi-quantitative risk assessment complements the low accuracy of qualitative risk assessment and the high expense and long calculation time of quantitative risk assessment. The first step of RB I is to identify important failure modes and causes in the equipment. Once these are defined, the Pof and COF can be assessed for each failure. During Pof and COF assessment, an effective inspection method and range can be easily found. In this paper, the calculation of the Pof is improved for accurate risk assessment. A modified semi-quantitative risk assessment was carried out for boiler facilities of thermal power plants, and the next maintenance schedules for the equipment were decided

  8. Ageing of polymers in electrical equipment used in nuclear power plants

    International Nuclear Information System (INIS)

    Clavreul, R.

    1999-01-01

    Ageing of polymers in electrical equipment used in nuclear power plants has been studied in (Electricite de France) EDF for several years. The objective of such studies is to predict the polymers lifetime in normal and accidental conditions. The prediction of polymers behaviour in normal conditions requires accelerated tests in order to get rapidly experimental results. Experimental conditions must carefully be chosen and representative of real ageing. Accelerated ageing is usually done by applying higher temperature, (dose) or dose rate. When such experiments are done, the effects of temperature, (dose) or dose rate are first determined. In a second step, experimental results are extrapolated to real conditions. To predict lifetime of polymers, the following recommendations have to be checked: in order to assume that accelerated tests are representative of normal ageing, the observed mechanisms in experiments must be the same as those in real conditions. For accidental conditions, the same tests as those described in standards can be applied to polymers. The simulation of any accident occurring just after the installation of electrical equipment in nuclear power plants is easy to manage: only the accidental test can be carried out on the electrical equipment. To determine whether polymers in electrical equipment would have a good behaviour or not when an accident would occur after a period of several years or decades in normal conditions in a nuclear power plant, the accidental test must be done on aged materials; their physical, mechanical and electrical characteristics must be relevant to aged polymers in normal conditions. In order to detect any evolution of properties during ageing, the electrical, mechanical or chemical tests have to be proceeded on polymers samples. The characterisation tests which are applied on non-aged and aged samples depend on the nature of the polymers, their application in electrical equipment and their environment. The IEC 544

  9. Design concepts to enhance nuclear power plant protection

    International Nuclear Information System (INIS)

    Ericson, D.M. Jr.; Varnado, G.B.

    1980-01-01

    Using a modern design for a nuclear power plant as a point of departure, this study examines the enhancement of protection which may be achieved by changes to the design. These changes include concepts such as complete physical separation of redundant trains of safety equipment, hardened enclosures for water storage tanks, and hardened shutdown heat removal systems. The degree of enhancement (value) is examined in terms such as the potential reduction in the number of vital areas and the increase in probability of adversary sequence interruption. The impacts considered include constraints imposed upon operations and maintenance personnel and increased capital and operating costs. The study concludes that structural design changes alone do not provide significant increases in protection

  10. Reliability data of fire protection equipment and features in German nuclear power plants

    International Nuclear Information System (INIS)

    Roewekamp, M.; Riekert, T.; Sehrbrock, W.

    1997-01-01

    In order to perform probabilistic fire safety analyses, a comprehensive data base is needed including physical characteristics of fire compartments and their inventory, fire occurrence frequencies, technical reliability data for all fire-related equipment, human actions and human error probabilities, etc. In order to provide updated and realistic reliability data, the operational behaviour of different fire protection features in two German nuclear power plants was analysed in the framework of the study presented here. The analyses are based on the examination of reported results of the regular inspection and maintenance programs for nuclear power plants. Besides a plant specific assessment of the reliability data a generic assessment for an application as input data for fault tree analyses in the framework of probabilistic risk studies for other German plants was carried out. The analyses of failures and unavailabilities gave the impression that most of them are single failures without relevance for the plant safety. The data gained from NPPs were compared to reliability data of the German insurance companies for the same protection features installed in non-nuclear installations and to older nuclear specific reliability data. This comparison showed up a higher reliability. (orig.) [de

  11. Computer aided process control equipment at the Karlsruhe reprocessing pilot plant, WAK

    International Nuclear Information System (INIS)

    Winter, R.; Finsterwalder, L.; Gutzeit, G.; Reif, J.; Stollenwerk, A.H.; Weinbrecht, E.; Weishaupt, M.

    1991-01-01

    A computer aided process control system has been installed at the Karlsruhe Spent Fuel Reprocessing Plant, WAK. All necessary process control data of the first extraction cycle is collected via a data collection system and is displayed in suitable ways on a screen for the operator in charge of the unit. To aid verification of displayed data, various measurements are associated to each other using balance type process modeling. Thus, deviation of flowsheet conditions and malfunctioning of measuring equipment are easily detected. (orig.) [de

  12. Improvement of nuclear power plant monitor and control equipment. Computer application backfitting

    International Nuclear Information System (INIS)

    Hayakawa, H.; Kawamura, A.; Suto, O.; Kinoshita, Y.; Toda, Y.

    1985-01-01

    This paper describes the application of advanced computer technology to existing Japanese Boiling Water Reactor (BWR) nuclear power plants for backfitting. First we review the background of the backfitting and the objectives of backfitting. A feature of backfitting such as restrictions and constraints imposed by the existing equipment are discussed and how to overcome these restrictions by introduction of new technology such as highly efficient data transmission using multiplexing, and compact space saving computer systems are described. Role of the computer system in reliable NPS are described with a wide spectrum of TOSHIBA backfitting computer system application experiences. (author)

  13. Spare-parts and perpetuity of equipment in French PWR plants

    International Nuclear Information System (INIS)

    Briolat, R.

    1993-01-01

    Supply of plants with new or repaired parts in strict quality conditions aids maintaining safety in operation and energy availability. Taking into account their expected life-time, a process of perpetuity in partnership with suppliers is necessary to ensure operation for the medium and long term. At EDF, the method involves a classification of mechanical and electrical spare parts in two levels of quality, responding to safety and availability imperatives and current available industrial practices. A diagram is presented to define optimal strategy for each equipment component, which gives choice between spare part storage, longevity agreement with the supplier, or a technology transfer agreement. 1 tab

  14. HTGR fuel reprocessing pilot plant: results of the sequential equipment operation

    International Nuclear Information System (INIS)

    Strand, J.B.; Fields, D.E.; Kergis, C.A.

    1979-05-01

    The second sequential operation of the HTGR fuel reprocessing cold-dry head-end pilot plant equipment has been successfully completed. Twenty standard LHGTR fuel elements were crushed to a size suitable for combustion in a fluid bed burner. The graphite was combusted leaving a product of fissile and fertile fuel particles. These particles were separated in a pneumatic classifier. The fissile particles were fractured and reburned in a fluid bed to remove the inner carbon coatings. The remaining products are ready for dissolution and solvent extraction fuel recovery

  15. Methods and equipment for diagnosis of components of Novovoronezh nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Prokop, K [Energoinvest, Dukovany (Czechoslovakia). Zavod Jaderna Elektrarna

    1981-12-01

    The results are reported obtained in applying diagnostic techniques adn diagnostic equipment in the Novovoronezh nuclear power plant. Vibroacoustic, neutron and hydrodynamic noiose of the installation was monitored. The test level method and the mean value comparison method were used for assessing the installation condition. Dispersion analysis methods are used for predicting the propagation of anomalies while for determining specific defects leading to the formation of anomalies the method is used based on the correlation analysis of vibroacoustic signals and other technological noise. The flow charts and descriptions are given of the systems of acoustic emission testing, reactor internals testing using neutron noise, pump testing, and the spectral analyzer.

  16. Active seismic response control systems for nuclear power plant equipment facilities

    International Nuclear Information System (INIS)

    Kobori, Takuji; Kanayama, Hiroo; Kamagata, Shuichi

    1989-01-01

    To sustain severe earthquake ground motion, a new type of anti-seismic structure is proposed, called a Dynamic Intelligent Building (DIB) system, which is positioned as an active seismic response controlled the structure. The structural concept starts from a new recognition of earthquake ground motion, and the structural natural frequency is actively adjusted to avoid resonant vibration, and similarly the external counter-force cancels the resonant force which comes from the dynamic structural motion energy. These concepts are verified using an analytical simulator program. The advanced application of the DIB system, is the Active Supporting system and the Active Stabilizer system for nuclear power plant equipment facilities. (orig.)

  17. The plant design analyser and its applications

    International Nuclear Information System (INIS)

    Whitmarsh-Everiss, M.J.

    1992-01-01

    Consideration is given to the history of computational methods for the non-linear dynamic analysis of plant behaviour. This is traced from analogue to hybrid computers. When these were phased out simulation languages were used in the batch mode and the interactive computational capabilities were lost. These have subsequently been recovered using mainframe computing architecture in the context of small models using the Prototype Plant Design Analyser. Given the development of parallel processing architectures, the restriction on model size can be lifted. This capability and the use of advanced Work Stations and graphics software has enabled an advanced interactive design environment to be developed. This system is generic and can be used, with suitable graphics development, to study the dynamics and control behaviour of any plant or system for minimum cost. Examples of past and possible future uses are identified. (author)

  18. A Method of Numerical Control Equipment Appearance Design Based on Product Identity

    Science.gov (United States)

    Zhu, Zhijuan; Zhou, Qi; Li, Bin; Visser, Steve

    Research on numerical control (NC) equipment has been more and more abundant; however, there are few existing studies in the field of appearance design for NC equipments. This paper provided a method to generate new appearance design of NC equipments based on product identity (PI). For the purpose of providing guidelines to generate new concept of NC equipment design, this paper, therefore, took the DMG Company (a Germen NC equipment company) as a case, examined the total products of this company from two aspects: Product Image and Product Family. Task 1 was an evaluate task about the Product Image by using the semantic differential (SD) evaluation method; Task 2 was a study task about Product Family to find out features of the products and classify these features. During the Task 2, several features have been found out and summarized, and these features were classified into 3 different levels according to their frequency and importance. In the end, two appearance design samples have been generated based on the analysis above to prove the application of the research.

  19. Quality planning for major plant design modifications

    International Nuclear Information System (INIS)

    Dulee, R.J.

    1988-01-01

    This paper reviews the approach and activities undertaken by Public Service Electric and Gas Company's (PSE and G's) nuclear quality assurance (QA) department to support major plant design modifications conducted during refueling outages at Salem Generating Station. It includes the planning and implementation of quality plans developed to provide both QA and quality control (QC) coverage of modification performed by contracted service organizations

  20. Fiscal 1981 Sunshine Project research report. Development of hydrothermal power plant. Development of binary cycle power plant. Conceptual plant design; 1981 nendo nessui riyo hatsuden plant no kaihatsu / binary cycle hatsuden plant no kaihatsu seika hokokusho . Plant gainen sekkei

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1982-03-01

    Conceptual design was made on a 10MW class binary cycle power plant for a demonstration plant superior in reliability and profitability, under most realistic current geothermal field conditions. In the design, study was made on heat balance, main pipe system, equipment allocation, and electric system for a plant system configuration, and study was also made on preheater, evaporator, condenser, turbine and others for plant component equipment. Further study was made on optimization of mist cooling condenser, instrumentation, control, utility, and environmental measures. The following basic data were obtained through the conceptual design: plant inlet hot water temperature: 130 degrees C, plant outlet hot water temperature: 70 degrees C, hot water flow rate: 1,415t/h, working fluid: R-114, R-114 pressure in evaporator: 11.98kg/cm{sup 2} abs, R-114 evaporation temperature: 91.1 degrees C, R-114 condensation temperature: 31.0 degrees C, R- 114 flow rate: 2,265t/h, site area: 106.5m x 102.4m, building area: 48.7m x 16.8m, and building height: 13.0m. (NEDO)

  1. Preliminary design report for prototypical spent nuclear fuel rod consolidation equipment

    International Nuclear Information System (INIS)

    Judson, B.F.; Maillet, J.; O'Neill, G.L.; Tsitsichvili, J.; Tucoulat, D.

    1986-12-01

    The purpose of the Prototypical Consolidation Demonstration Project (PCDP) is to develop and demonstrate the equipment system that will be used to consolidate the bulk of the spent nuclear fuel generated in the United States prior to its placement in a geological repository. The equipment must thus be capable of operating on a routine production basis over a long period of time with stringent requirements for safety, reliability, productivity and cost-effectiveness. Four phases are planned for the PCDP. Phase 1 is the Preliminary Design of generic consolidation equipment that could be installed at a Monitored Retrievable Storage (MRS) facility or in the Receiving ampersand Handling Facility at a geologic repository site. Phase 2 will be the Final Design and preparation of procurement packages for the equipment in a configuration capable of being installed and tested in a special enclosure within the TAN Hot Shop at DOE's Idaho National Engineering Laboratory. In Phase 3 the equipment will be fabricated and then tested with mock fuel elements in a contractor's facility. Finally, in Phase 4 the equipment will be moved to the TAN facility for demonstration operation with irradiated spent fuel elements. 55 figs., 15 tabs

  2. Design and construction of the Portable Industrial X-Ray Equipment

    International Nuclear Information System (INIS)

    Nguyen Phuc; Nguyen Van Si; Le Tien Quan; Trinh Anh Tuan; Nguyen Manh Hung; Trinh Dinh Tuong

    2011-01-01

    The main purposes of the project are the supporting to design and construction of the Portable Industrial X-Ray Equipment; with the accuracy ±2% of Output High Voltage 200 kV and Tube current 5 mA. The Equipment is composed of control unit, X-ray generator, and power cable, connection cable. X-ray generator is assembling construction X-ray tube, H.V transformer together with gas insulation (SF6) are sealed up in aluminum bucked cabinet, fan and heat-sink are mounted in the end of X-ray generator as cooling, SF6 is a gas electrical performance to H.V. Alarm lamp is used to warn, flashing, show generating X-ray. Control unit is box construction. Four printed circuit boards (PCB) and electronic device are mounted in it. All operating buttons switches and displays are equipped on the panel. We have completed to design and construct the Portable Industrial X-Ray Equipment; and have tested the electronic parameters of all test points and the main parameters of equipment (the accuracy ±2% of Output High Voltage 200 kV and Tube current 5 mA). We also have successful used the Portable Industrial X-Ray Equipment to evaluate the welds in industry. (author)

  3. Investigation of human system interface design in nuclear power plant

    International Nuclear Information System (INIS)

    Feng Yan; Zhang Yunbo; Wang Zhongqiu

    2012-01-01

    The paper introduces the importance of HFE in designing nuclear power plant, and introduces briefly the content and scope of HFE, discusses human system interface design of new built nuclear power plants. This paper also describes human system interface design of foreign nuclear power plant, and describes in detail human system interface design of domestic nuclear power plant. (authors)

  4. Study on the Management for the Nuclear Power Plant Maintenance and Equipment Reliability

    International Nuclear Information System (INIS)

    Yoon, Kyeongseop; Lee, Sangheon; Kim, Myungjin; Lee, Unjang

    2015-01-01

    In our country, many studies on the regulatory policy of the plant maintenance have ever been performed since 1998, but the relevant regulatory requirements were not established yet. These background mentioned above request us to study on the regulation policy and maintenance plan to improve the safety, reliability and efficiency of NPP. To solve these problems, in this study, we deduct the management methodology for the improvement of NPP maintenance and equipment reliability that is essential to secure the safety and efficiency of the commercial NPP. For analysis the maintenance and equipment reliability management methodology in overseas NPP. We studied maintenance and equipment reliability of USA, Canada and Europe(France, England, German). We also studied status and application condition of Korean NPP maintenance management technical development. We deducted an effective maintenance methodology that is needed to Korean NPP, as a result of comparison on the technical trend of the maintenance management between overseas and Korean, such like following. - Regulation form ·Specific provision of regulation requirement and application of form that is clarifying application standard - Maintenance management methodology, Maintenance management program. This results of study could be applied for regulation policy, law and guideline establishment of NPP maintenance, operation, supervision and a system establishment for maintenance management, education data about maintenance for NPP employees

  5. Electrical and control equipment in nuclear power plants. Problems when replacing aging equipment; El och kontrollutrustning i kaernkraftverk - Problematik vid utbyte av aaldrad utrustning

    Energy Technology Data Exchange (ETDEWEB)

    Nordling, Anna; Haakansson, Goeran

    2012-11-01

    Interoperability between different technical systems is more complicated when old and new technology meet, such as between analog and digital technology. New electrical and I and C equipment is selected with consideration to simplify and improve the compatibility and interoperability. The original construction of nuclear power plants with electricity and I and C equipment had more natural interfaces. Generally experienced guidance, to the management of interoperability and interfaces, feels insufficient. Skills transfer programs are identified as a major need, as more and more important personnel are retiring and important information is lost with them. Lack of appropriate skills directly affects the ability to produce accurate and complete requirements specification. Failure modes of newer electrical and I and C equipment are perceived as more complex than the older equipment. When choosing equipment, attempts are made to minimize unnecessary features, to reduce the number of potential failure modes. There is a lack of consistent understanding of the meaning of robustness in electrical technology and I and C technology, in the nuclear plant engineering departments. The overall picture is that the robustness has worsened since the facilities were built. The Swedish nuclear power plants have an internal organizational structure with separated client and support organization. This splits the nuclear organization into two distinct parts which threaten to separate the two entities focus. Engineering departments at the Swedish nuclear power plants express a need for increased expertise in the client organization (blocks). Competence requested is for example, system knowledge to facilitate and enhance the quality of the initial analysis performed in the blocks. Suppliers receive more recently larger turnkey projects, both to minimize costs but also to minimize the interfaces and co-function problems. This, however, heightens demands for knowledge transfer between

  6. Maintenance of the Photovolatic Plants Using UAV Equipped with Low-cost GNSS RTK Receiver

    DEFF Research Database (Denmark)

    Bilal, Muhammad; Prasad, Ramjee; Nisi, Marco

    2017-01-01

    of low-cost GNSS RTK receiver, is the maintenance of photovoltaic (PV) plants using Unmanned Aerial Vehicle (UAV). This paper proposes a solution that aims at automating the maintenance of PV plant with enhanced reliability in a time and cost effective manner, which otherwise requires intermediate human...... intervention. The paper presents the architectural concept, system design, and end-to-end algorithm that plays a pivotal role in enabling the automatic report generation of PV plant status. Preliminary results of the proof-of-concept shows the feasibility of the proposed solution....

  7. Structural modules in AP1000 plant design

    International Nuclear Information System (INIS)

    Prasad, N.; Tunon-Sanjur, L.

    2007-01-01

    Structural modules are extensively used in AP1000 plant design. The shop manufacturing of modules components improves the quality and reliability of plant structures. The application of modules has a positive impact on construction schedules, and results in substantial savings in the construction cost. This paper describes various types of structural modules used for AP1000 plant structures. CA structural wall modules are steel plate modules with concrete placed, on or within the module, after module installation. The layout and design of the largest CA wall modules, CA01 and CA20, is described in detail. General discussion of structural floor modules, such as the composite and finned floors, is also included. Steel form CB modules (liners) consist of plate reinforced with angle stiffeners and tee sections. The angles and the tee sections are on the concrete side of the plate. Design of CB20 has been included as an example of CB type modules. Design codes and structural concepts related to module designs are discussed. (authors)

  8. Design of 90-mgd wastewater reclamation plant

    International Nuclear Information System (INIS)

    Cain, C.B.; Kluesener, J.W.; Lazarus, E.

    1981-01-01

    Ninety MGD of municipal secondary effluent is to be reclaimed for use as cooling water in the Palo Verde Nuclear Generating Station in Arizona. The water reclamation plant design criteria are presented for nitrification, lime-soda softening, filtration, and sludge processing. In-pipe storage is used for flow equalization. Forced air down-draft trickling filters with computer-controlled recycle are used for nitrification. Effluent recirculation is used to maintain constant sludge blanket dept in the upflow solids contact clarifier softening units. An influent flow-splitting, gravity-backwash, dual-media filter is used for final suspended solids removal. Rheological data was used in the hydraulic design of sludge piping. Computerized automatic flushing also aids in preventing sludge pumping problems. Reclamation plant design and construction are integrated in one Engineering-Procurement-Construction project conducted by a single company. 5 refs

  9. Application of the system engineering approach for reactor plants design

    International Nuclear Information System (INIS)

    Sitskiy, S. B.

    2010-01-01

    The main activities planned for to be implemented are: developing a data model of the reactor plant plus integration with the information model of the plant (3D model + P & ID); reengineering of processes, developing of electronic documents; description of the equipment for information management of the reactor plant lifecycle – according ISO15926

  10. Injury survey of a non-traditional 'soft-edged' trampoline designed to lower equipment hazards.

    Science.gov (United States)

    Eager, David B; Scarrott, Carl; Nixon, Jim; Alexander, Keith

    2013-01-01

    In Australia trampolines contribute one quarter of all childhood play equipment injuries. The objective of this study was to gather and evaluate injury data from a non-traditional, 'soft-edged', consumer trampoline, where the design aimed to minimise injuries from the equipment and from falling off. The manufacturer of the non-traditional trampoline provided the University of Technology Sydney with their Australian customer database. The study involved surveys in Queensland and New South Wales, between May 2007 and March 2010. Initially injury data was gathered by a phone interview pilot study, then in the full study, through an email survey. The 3817 respondents were the carers of child users of the 'soft-edge' trampolines. Responses were compared with Australian and US emergency department data. In both countries the proportion of injuries caused by the equipment and falling off was compared with the proportion caused by the jumpers to themselves or each other. The comparisons showed a significantly lower proportion resulted from falling-off or hitting the equipment for this design when compared to traditional trampolines, both in Australia and the US. This research concludes that equipment-induced and falling-off injuries, the more severe injuries on traditional trampolines, can be significantly reduced with appropriate trampoline design.

  11. Automation and mechanization of in-service inspection of selected equipment in FRG's nuclear power plants

    International Nuclear Information System (INIS)

    Metke, E.

    1988-01-01

    The procedures and equipment are described for the automation and mechanization of in-service inspection in nuclear power plants in the FRG, used by the KWU company. Checks of the pressure vessel are done by visual means using a colour tv camera, the method of eddy currents and the ultrasonic method. An analysis is made of the time schedule of ultrasonic inspections, and the central column manipulator is described which allows to check all internal regions of the pressure vessel. Attention is also devoted to other devices, e.g., those for prestressing shanks, cleaning shanks, cleaning thread apertures, etc. A combined probe using the ultrasonic method and the eddy current method serves the inspection of heat exchange tubes in the steam generator. For inspecting the primary circuit the KWU company uses devices for checking and working the inner surface of pipes. Briefly described are examples of using KWU equipment in nuclear power plants in CMEA countries. (Z.M.). 11 figs., 6 refs

  12. Technical design considerations in the provision of a commercial MOX plant

    International Nuclear Information System (INIS)

    Elliott, M.F.

    1997-01-01

    The Sellafield MOX Plant (SMP) has a design production target of 120 t/year Heavy Metal of mixed uranium dioxide and plutonium dioxided (MOX) fuel. It will have the capability to produce fuel with fissile enrichments up to 10%. The feed materials are those arising from reprocessing operations on the Sellafield site, although the plant also has the capability to receive and process plutonium from overseas reprocessing plants. The ability to produce 10% enriched fuels, together with the requirement to use high burn-up feed has posed a number of design challenges to prevent excessive powder temperatures within the plant. As no stimulants are available to represent the heat generating nature of plutonium powders, it is difficult to prove equipment design by experiment. Extensive use has therefore been made of finite element analysis techniques. The requirement to process material of low burn-up (i.e. high fissile enrichment) has also impacted on equipment design in order to ensure that criticality limits are not exceeded. This has been achieved where possible by 'safe by geometry' design and, where appropriate, by high integrity protection systems. SMP has been designed with a high plant availability but at minimum cost. The requirement to minimize cost has meant that high availability must be obtained with the minimum of equipment. This had led to major challenges for equipment designers in terms of both the reliability and also the maintainability of equipment. Extensive use has been made of theoretical modelling techniques which have given confidence that plant throughput can be achieved. (author). 1 fig

  13. Integrated design of SIGMA uranium enrichment plants

    International Nuclear Information System (INIS)

    Rivarola, Martin E.; Brasnarof, Daniel O.

    1999-01-01

    In the present work, we describe a preliminary analysis of the design feedbacks in a Uranium Enrichment Plant, using the SIGMA concept. Starting from the result of this analysis, a computer code has been generated, which allows finding the optimal configurations of plants, for a fixed production rate. The computer code developed includes the model of the Thermohydraulic loop of a SIGMA module. The model contains numerical calculations of the main components of the circuit. During the calculations, the main components are dimensioned, for a posterior cost compute. The program also makes an estimation of the enrichment gain of the porous membrane, for each separation stage. Once the dimensions of the main components are known, using the enrichment cascade calculation, the capital and operation costs of the plant could be determined. At this point it is simple to calculate a leveled cost of the Separative Work Unit (SWU). A numerical optimizer is also included in the program. This optimizer finds the optimal cascade configuration, for a given set of design parameters. The whole-integrated program permits to investigate in detail the feedback in the component design. Therefore, the sensibility of the more relevant parameters can be computed, with respect of the economical variables of the plant. (author)

  14. Design and fabrication of stainless steel components for long life of spent fuel reprocessing plants

    International Nuclear Information System (INIS)

    Natarajan, R.; Ramkumar, P.; Sundararaman, V.; Kamachi Mudali, U.; Baldev Raj; Shanmugam, K.

    2010-01-01

    Reprocessing of spent nuclear fuels based on the PUREX process is the proven process with many commercial plants operating satisfactorily worldwide. The process medium being nitric acid, austenitic stainless steel is the material of construction as it is the best commercially available material for meeting the conditions in the reprocessing plants. Because of the high radiation fields, contact maintenance of equipment and systems of these plants are very time consuming and costly unlike other chemical process plants. Though the plants constructed in the early years required extensive shut downs for replacement of equipment and systems within the first fifteen years of operation itself, development in the field of stainless steel metallurgy and fabrication techniques have made it possible to design the present day plants for an operating life period of forty years. A review of the operational experience of the PUREX process based aqueous reprocessing plants has been made in this paper and reveals that life limiting failures of equipment and systems are mainly due to corrosion while a few are due to stresses. Presently there are no standards for design specification of materials and fabrication of reprocessing plants like the nuclear power plants, where well laid down ASTM and ASME codes and standards are available which are based on the large scale operational feedbacks on pressure vessels for conventional and nuclear industries. (author)

  15. Design and Manufacturing of Composite Tower Structure for Wind Turbine Equipment

    Science.gov (United States)

    Park, Hyunbum

    2018-02-01

    This study proposes the composite tower design process for large wind turbine equipment. In this work, structural design of tower and analysis using finite element method was performed. After structural design, prototype blade manufacturing and test was performed. The used material is a glass fiber and epoxy resin composite. And also, sand was used in the middle part. The optimized structural design and analysis was performed. The parameter for optimized structural design is weight reduction and safety of structure. Finally, structure of tower will be confirmed by structural test.

  16. Heat exchanger design for desalination plants

    International Nuclear Information System (INIS)

    1979-03-01

    The Office of Saline Water (OSW) accomplished a very large amount of significant work related to the design and performance of large heat exchanger bundles and enhanced heat transfer surfaces. This work was undertaken to provide basic technical and economic data for the design of distillation plants for the desalination of seawater, and should be of value to other industrial applications as well. The OSW work covers almost every aspect of heat exchanger design, and ranges academic research to data gathering on commercial desalting plants. Exchanger design configurations include multistage flash plant condensers, vertical tube falling film and upflow evaporators, and horizontal tube spray film evaporators. Unfortunately, the data is scattered through a large number of reports of which many are concerned primarily with factors other than heat transfer, and the quality of reporting and the quality of the data are far from consistent. This report catalogues and organizes the heat exchanger data developed by the OSW. Some analysis as to the validity of the data is made and ranges of performance that can be expected are given. Emphasis is placed on the vertical tube, falling film evaporators. A thorough analysis of the large literature file that was surveyed was not possible. No analysis was made of the quality of original data, but apparent data discrepancies are pointed out where such discrepancies happen to be found

  17. Risk informed life cycle plant design

    International Nuclear Information System (INIS)

    Hill, Ralph S. III; Nutt, Mark M.

    2003-01-01

    Many facility life cycle activities including design, construction, fabrication, inspection and maintenance are evolving from a deterministic to a risk-informed basis. The risk informed approach uses probabilistic methods to evaluate the contribution of individual system components to total system performance. Total system performance considers both safety and cost considerations including system failure, reliability, and availability. By necessity, a risk-informed approach considers both the component's life cycle and the life cycle of the system. In the nuclear industry, risk-informed approaches, namely probabilistic risk assessment (PRA) or probabilistic safety assessment (PSA), have become a standard tool used to evaluate the safety of nuclear power plants. Recent studies pertaining to advanced reactor development have indicated that these new power plants must provide enhanced safety over existing nuclear facilities and be cost-competitive with other energy sources. Risk-informed approaches, beyond traditional PRA, offer the opportunity to optimize design while considering the total life cycle of the plant in order to realize these goals. The use of risk-informed design approaches in the nuclear industry is only beginning, with recent promulgation of risk-informed regulations and proposals for risk-informed codes. This paper briefly summarizes the current state of affairs regarding the use of risk-informed approaches in design. Key points to fully realize the benefit of applying a risk-informed approach to nuclear power plant design are then presented. These points are equally applicable to non-nuclear facilities where optimization for cost competitiveness and/or safety is desired. (author)

  18. Westinghouse AP600 advanced nuclear plant design

    International Nuclear Information System (INIS)

    Gangloff, W.

    1999-01-01

    As part of the cooperative US Department of Energy (DOE) Advanced Light Water Reactor (ALWR) Program and the Electric Power Research Institute (EPRI), the Westinghouse AP600 team has developed a simplified, safe, and economic 600-megawatt plant to enter into a new era of nuclear power generation. Designed to satisfy the standards set by DOE and defined in the ALWR Utility Requirements Document (URD), the Westinghouse AP600 is an elegant combination of innovative safety systems that rely on dependable natural forces and proven technologies. The Westinghouse AP600 design simplifies plant systems and significant operation, inspections, maintenance, and quality assurance requirements by greatly reducing the amount of valves, pumps, piping, HVAC ducting, and other complex components. The AP600 safety systems are predominantly passive, depending on the reliable natural forces of gravity, circulation, convection, evaporation, and condensation, instead of AC power supplies and motor-driven components. The AP600 provides a high degree of public safety and licensing certainty. It draws upon 40 years of experience in light water reactor components and technology, so no demonstration plant is required. During the AP600 design program, a comprehensive test program was carried out to verify plant components, passive safety systems components, and containment behavior. When the test program was completed at the end of 1994, the AP600 became the most thoroughly tested advanced reactor design ever reviewed by the US Nuclear Regulatory Commission (NRC). The test results confirmed the exceptional behavior of the passive systems and have been instrumental in facilitating code validations. Westinghouse received Final Design Approval from the NRC in September 1998. (author)

  19. Application of Pressure Equipment Standard at nuclear power plants; Aplicacion del Reglamento de Equipos a Presion a las centrales nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Mostaza, J. M.

    2011-07-01

    Regarding with the paper presented on 9{sup t}h June 2011 referred to the Industrial Security standard in Nuclear Plants, it was about the application of Pressure Equipment standard to mentioned Nuclear Plants, this article is an extract of the paper going to be exposed. (Author)

  20. Modern technologies and equipment for environment and sustainable development at ROMAG-PROD Heavy Water Plant

    International Nuclear Information System (INIS)

    Preda, Marius Cristian; Patrascu, Mihai; Pop, Artimisia; Chilom, Rodica

    2004-01-01

    At ROMAG-PROD Heavy Water Plant, the sustainable development concept incorporates as a priority the environmental protection through the production process technology. Norway's Prime Minister, Mr. Gro Harlem Brundtland used the concept of 'sustainable development' in 1987, when as President of International Commission for Environment and Sustainable Development, he presented his report 'Our common future'. Sustainable development means that development that allows satisfying our present needs without spoiling the next generation capacity to meet their own needs. Any technology has both advantages and disadvantages; when considering the concept of sustainable development we have to take into account all the aspects, namely: - causes identification and review; - results evaluation; - corrective and preventive actions. Thus, ROMAG-PROD Heavy Water Plant has implemented a typical environment management system by means of what the general and specific objectives have been established, these objectives being stated in an Environment Policy Declaration: - Environment Management System as per SR EN ISO 14001/1997; - Quality Management System as per SR EN ISO 9001/2000; - IQNet- The International Certification Network. The paper presents the modern equipment for emissions and in-missions management with real time data transmission, for air and water as environment elements. Section two deals with trial of modern technology for industrial discharged wastewater treatment using the method of controlled batching of surface-active materials. Investigations on method application and laboratory testing as well as findings are given. As a conclusion, one can state that ROMAG-PROD Heavy Water Plant, has as one of its main concern keeping on high standards the safety of its equipment operation, sustainable development and risk eliminating so that neither environment or the population in vicinity is affected. (authors)

  1. Power system engineering planning, design, and operation of power systems and equipment

    CERN Document Server

    Schlabbach, Juergen

    2014-01-01

    With its focus on the requirements and procedures of tendering and project contracting, this book enables the reader to adapt the basics of power systems and equipment design to special tasks and engineering projects, e.g. the integration of renewable energy sources.

  2. Prototypical spent fuel rod consolidation equipment preliminary design report: Volume 2, Drawings

    International Nuclear Information System (INIS)

    1986-01-01

    This volume consists of 65 E size drawings and 4 sketches of the NUS spent fuel rod consolidation equipment. The drawings have been grouped into categories; a detailed list of the drawings is included. The sketches prepared during the preliminary design process have been included

  3. Housing, Equipment, and Design Research and Scholarship: A Family and Consumer Sciences Perspective.

    Science.gov (United States)

    Beamish, Julia O.; Ahn, Mira; Seiling, Sharon

    2001-01-01

    Analysis of research on housing, equipment, and design (n=333) in the Journal of Home Economics/Journal of Family and Consumer Sciences (1985-2000), Home Economics Research Journal/Family and Consumer Sciences Research Journal (1985- 2000), and Housing and Society (1985-1999) found that articles declined by more than 50% and behavior theories were…

  4. Design and fabrication of equipment and devices for a Radiotherapy department

    International Nuclear Information System (INIS)

    Picon, C.L.; Zaharia B, M.

    1998-01-01

    The objective of this work is to present the possibility to design and construct utilizing local technology, a series of equipment and devices necessaries for a routine in a Radiotherapy Department with typical budget in the Latin-American hospitals. (Author)

  5. Review of the application of ergonomics design of trackless mining equipment (TME) - lessons and challenges

    CSIR Research Space (South Africa)

    James, JP

    2007-06-01

    Full Text Available the design of trackless mining equipment (TME) is poor, with improvised seating, poor cabin layouts and sub-standard display instrumentation. This paper will present the key findings of two studies assessing ergonomics risk factors associated with automated...

  6. Design of channel experiment equipment for measuring coolant velocity of innovative research reactor

    International Nuclear Information System (INIS)

    Muhammad Subekti; Endiah Puji Hastuti; Dedi Heriyanto

    2014-01-01

    The design of innovative high flux research reactor (RRI) requires high power so that the capability core cooling requires to be improved by designing the faster core coolant velocity near to the critical velocity limit. Hence, the critical coolant velocity as the one of the important parameter of the reactor safety shall be measured by special equipment to the velocity limit that may induce fuel element degradation. The research aims is to calculate theoretically the critical coolant velocity and to design the special experiment equipment namely EXNal for measuring the critical coolant velocity in fuel element subchannel of the RRI. EXNal design considers the critical velocity calculation result of 20.52 m/s to determine the variation of flow rate of 4.5-29.2 m 3 /h, in which the experiment could simulate the 1-4X standard coolant velocity of RSG-GAS as well as destructive test of RRI's fuel plate. (author)

  7. Exergy analysis of an IGCC design configuration for Plant Wansley

    International Nuclear Information System (INIS)

    Tsatsaronis, G.; Tawfik, T.; Lin, L.; Gallaspy, D.T.

    1989-01-01

    An integrated gasification-combined-cycle power plant design was developed for Georgia Power Company's Plant Wansley. This paper discusses the plant configuration and presents the most important results obtained from a detailed exergy analysis of the plant design. This analysis will be completed in a subsequent paper through an exergoeconomic analysis to identify design improvements for reducing the electricity cost

  8. Analysis and design recommendation on rabbeted capping plate of equipment cell in nuclear chemical facility

    International Nuclear Information System (INIS)

    Zhang Jingyu; Yin Xiaozhan

    2013-01-01

    Rabbeted capping plates are widely used in the roof of equipment cells in order to meet the requirements of nuclear radiation protection. The key considerations in the design include vertical load, seismic load and repair load. This article establishes T shaped and Z-shaped plate model via FEM software (ANSYS), analyzes the bearing capacity and displacement distribution in different load cases, and provides recommendations to the design and construction accordingly. (authors)

  9. Seismic stress of plants and equipment in nuclear power station construction

    International Nuclear Information System (INIS)

    Hampe, E.; Schwarz, J.

    1984-01-01

    The applicability of floor spectra for designing components of nuclear power plants taking into account seismic effects is discussed. Methods for the determination of seismic floor excitation and various kinds of floor spectra are presented. As an example the floor spectra method is applied to containment buildings

  10. Absolute air filtering equipment in the nuclear industrie. Design - Safety - Experience

    International Nuclear Information System (INIS)

    Lucas, J.C.

    1977-01-01

    The problems encountered in the design of absolute filters (HEPA FILTERS) are presented: glass-fibre filter papers; standards and characteristics: efficiency, fire-resistance, humidity-resistance, radiation-resistance, etc; various types of paper folding: deep folds and small folds, dihedrally mounted; filtering elements; designs; characteristics and quality control; The design of filtration equipment is also analysed: mounting in metal or concrete casings. French and American designs (Regulatory Guide 1-52); and gas-tight casings allowing contaminated filters to be renewed without breaking the gas-tight seal

  11. Design Methodology of Process Layout considering Various Equipment Types for Large scale Pyro processing Facility

    International Nuclear Information System (INIS)

    Yu, Seung Nam; Lee, Jong Kwang; Lee, Hyo Jik

    2016-01-01

    At present, each item of process equipment required for integrated processing is being examined, based on experience acquired during the Pyropocess Integrated Inactive Demonstration Facility (PRIDE) project, and considering the requirements and desired performance enhancement of KAPF as a new facility beyond PRIDE. Essentially, KAPF will be required to handle hazardous materials such as spent nuclear fuel, which must be processed in an isolated and shielded area separate from the operator location. Moreover, an inert-gas atmosphere must be maintained, because of the radiation and deliquescence of the materials. KAPF must also achieve the goal of significantly increased yearly production beyond that of the previous facility; therefore, several parts of the production line must be automated. This article presents the method considered for the conceptual design of both the production line and the overall layout of the KAPF process equipment. This study has proposed a design methodology that can be utilized as a preliminary step for the design of a hot-cell-type, large-scale facility, in which the various types of processing equipment operated by the remote handling system are integrated. The proposed methodology applies to part of the overall design procedure and contains various weaknesses. However, if the designer is required to maximize the efficiency of the installed material-handling system while considering operation restrictions and maintenance conditions, this kind of design process can accommodate the essential components that must be employed simultaneously in a general hot-cell system

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

    International Nuclear Information System (INIS)

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

    1980-02-01

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

  13. Life Cycle Management for Equipments in Nuclear Plants Based on Reliability

    International Nuclear Information System (INIS)

    Wang Dalin; Sun Jinlong

    2012-01-01

    Majority of equipments in nuclear power stations are not un-repairable, but repaired after failures or maintained periodically and then returned to running. Along with aging or fatigue causes, duration between failures may be variable such as larger or shorter. Equipments failure occurrences are consistent with non-homogeneous Poisson process (NHPP), in which their failure rates are variable. We chose Weibull process to simplify the actual circumstances to obtain the optimal life cycle management plans approximately. In application we need to estimate parameters of Weibull process with existing data. Before that, trend test is necessary. Total time test (TTT) is an alternative. Trend test can determine whether the failure rate changes along operating time and whether the devices operated under imperfect or minimal repair. Parameters of Weibull process λ, β can be estimated by maximum likelihood estimation method. LCM model is designed to found an optimizing model for those aging equipments under minimal repair or imperfect maintenance with maximum economic or availability as objective function. The purpose is to establish the optimal replacement strategies. Periodical replacement is the most regular choice and so LCM model aims to calculate the optimal replacement periodical duration.

  14. Aseismatic design of electrical equipments and instruments for nuclear power stations

    International Nuclear Information System (INIS)

    Suzuki, Yasuharu; Nishizawa, Kazuo; Miyazaki, Yoshio; Miura, Takumi

    1977-01-01

    The aseismatic design of electrical instruments is carried out according to IEEE Standard 344-1971 in the USA. In Japan also, the method of aseismatic design of electrical instruments has been investigated by the representatives of electric power companies and electric machine makers since 1972. In Hitachi Ltd., the statical method of confirming aseismatic property was established on the basis of the rigid design for electrical instruments. It is convenient to examine the aseismatic property of electrical equipments by classifying them into control and switch boards, electrical appliances, equipments and circuits. It is possible to use the static method treating earthquake force as static load by avoiding resonance with the electrical equipments which have the higher natural frequency than that of buildings. The purposes of the vibration test are to prove the validity of the theoretical analysis, to clarify the vibration characteristics, and to confirm the maintenance of functions and the strength of the equipments. The vibration tests of control boards, the switch boards of enclosed type, motor control centers, the racks for instrumentation, storage batteries and electrical appliances are explained. Moreover, the vibration analysis with a computer according to finite element method is described. (Kako, I.)

  15. Designing electronic equipment on the basis of standard mechanical structures using internet re­sour­ces

    Directory of Open Access Journals (Sweden)

    Karlangach A. P.

    2016-12-01

    Full Text Available The author proposes a method to design electronic equipment based on functional-node design method that involves the use of 2D- and 3D- models mechanical structures for electronic equipment as a way to reduce development time and errors when creating design documentation for electronic equipment. At present, most areas of science and technology are computerized, more problems in designing electronic equipment are dealt with using computer-aided design (CAD and Computer-aided manufacturing (CAM to reduce the time required for development and manufacturing of electronic equipment. Development of design documentation also requires a more effective approach, because the less the time for development of the design documentation is, the faster the developed device will go into production. The aim of the study is to develop a method of designing electronic equipment using 2D and 3D models of standard mechanical structures for electronic equipment using Internet resources. Based on the presented methods is an example of designing a device from standard bearing structures. Compared with traditional technology, the method of designing electronic equipment using standard parts has the following advantages: - reduces time and improves quality of development through the use of existing design documentation; - accelerates the implementation and introducing into production processes; - increases unification of design solutions.

  16. Generation of floor spectra compatible time histories for equipment seismic qualification in nuclear power plants

    International Nuclear Information System (INIS)

    Shyu, Y.-S.; Luh, Gary G.; Blum, Arie

    2004-01-01

    This paper proposes a procedure for generating floor response spectra compatible time histories used for equipment seismic qualification in nuclear power plants. From the 84th percentile power spectrum density function of an earthquake ensemble of four randomly generated time history motions, a statistically equivalent time history can be obtained by converting the power spectrum density function from the frequency domain into the time domain. With minor modification, if needed, the converted time history will satisfy both the spectral and the power spectrum density enveloping criteria, as required by the USNRC per Revision 2 of the Standard Review Plan, Section 3.7.1. Step-by-step generating procedures and two numerical examples are presented to illustrate the applications of the methodology. (author)

  17. Design and safety data of commercial nuclear power plants in Japan, 1977 edition

    International Nuclear Information System (INIS)

    Izumi, Fumio; Harayama, Yasuo

    1977-09-01

    Following on previous JAERI-M 5959(1975) and JAERI-M 6732(1976), which contained the data for design parameters, performance, components and equipments of Japanese nuclear power plants, the 1977 updated edition is compiled as of June 1977. The data are arranged and tabulated by computer processing. (auth.)

  18. Design and safety data of commercial nuclear power plants in Japan

    International Nuclear Information System (INIS)

    Izumi, Fumio; Harayama, Yasuo; Nakazima, Tetuo

    1979-02-01

    Following the previous JAERI-M 6732(1976) and JAERI-M 7261(1977), the 1978 edition as of the December is presented, which contains the data of design parameters, performance, components and equipments in nuclear power plants of Japan. Data are given in tables by computer processing. (author)

  19. Project designing of Temelin nuclear power plant

    International Nuclear Information System (INIS)

    Krychtalek, Z.; Linek, V.

    1989-01-01

    The geological and seismic parameters are listed of the Temelin nuclear power plant. The division of the site in building zones is described. The main zones consist of the power generation unit zone with the related auxiliary buildings of hot plants and of the auxiliary buildings of the nonactive part with industrial buildings. The important buildings are interconnected with communication and technology bridges. Cooling towers and spray pools and the entrance area are part of the urbanistic design. The architectonic design of the buildings uses standard building elements and materials. The design of the buildings is based on the requirements on their function and on structural load and on the demands of maximal utilization of the type of the reinforced concrete prefab structure system. The structure is made of concrete or steel cells. The project design is based on Soviet projects. The layout is shown of the main power generation units and a section is presented of a 1,000 MW unit. (J.B.). 2 figs

  20. Design of a compact low-power human-computer interaction equipment for hand motion

    Science.gov (United States)

    Wu, Xianwei; Jin, Wenguang

    2017-01-01

    Human-Computer Interaction (HCI) raises demand of convenience, endurance, responsiveness and naturalness. This paper describes a design of a compact wearable low-power HCI equipment applied to gesture recognition. System combines multi-mode sense signals: the vision sense signal and the motion sense signal, and the equipment is equipped with the depth camera and the motion sensor. The dimension (40 mm × 30 mm) and structure is compact and portable after tight integration. System is built on a module layered framework, which contributes to real-time collection (60 fps), process and transmission via synchronous confusion with asynchronous concurrent collection and wireless Blue 4.0 transmission. To minimize equipment's energy consumption, system makes use of low-power components, managing peripheral state dynamically, switching into idle mode intelligently, pulse-width modulation (PWM) of the NIR LEDs of the depth camera and algorithm optimization by the motion sensor. To test this equipment's function and performance, a gesture recognition algorithm is applied to system. As the result presents, general energy consumption could be as low as 0.5 W.

  1. Modelling and design of undercarriage components of large-scale earthmoving equipment in tar sand operations

    Energy Technology Data Exchange (ETDEWEB)

    Szymanski, J.; Frimpong, S.; Sobieski, R. [Alberta Univ., Edmonton, AB (Canada). Centre for Advanced Energy and Minerals Research

    2004-07-01

    This presentation described the fundamental and applied research work which has been carried out at the University of Alberta's Centre for Advanced Energy and Minerals Research to improve the undercarriage elements of large scale earthmoving equipment used in oil sands mining operations. A new method has been developed to predict the optimum curvature and blade geometry of earth moving equipment such as bulldozers and motor graders. A mathematical relationship has been found to approximate the optimum blade shape for reducing cutting resistance and fill resistance. The equation is a function of blade geometry and soil properties. It is the first model that can mathematically optimize the shape of a blade on earth moving equipment. A significant saving in undercarriage components can be achieved from reducing the amount of cutting and filling resistance for this type of equipment working on different soils. A Sprocket Carrier Roller for a Tracked Vehicle was also invented to replace the conventional cylindrical carrier roller. The new sprocket type carrier roller offers greater support for the drive track and other components of the undercarriage assembly. A unique retaining pin assembly has also been designed to detach connecting disposable wear parts from earthmoving equipment. The retaining pin assembly is easy to assemble and disassemble and includes reusable parts. 13 figs.

  2. Multi-criteria evaluation and priority analysis for localization equipment in a thermal power plant using the AHP (analytic hierarchy process)

    International Nuclear Information System (INIS)

    Yagmur, Levent

    2016-01-01

    Ensuring the safety of its energy supply is one of the main issues for newly industrialized/developing countries when utilizing domestic sources for electricity generation. Turkey depends heavily on imported gas to generate electricity, and the ratio of natural gas power generation to total electricity production is nearly 50%. Coal-fired thermal power plants using domestic resources are considered a good option to decrease the large amount of imported natural gas, and to supply a secure energy demand. However, electricity generation from coal-fired power plants using local lignite reserves is not adequate to maintain a secure energy mix and provide sustainable development, as Turkey does not have indigenous energy sector technology. Therefore, technology transfer and its localization are crucial for newly industrialized/developing countries such as Turkey. The aim of this study is to use the analytic hierarchy process to determine a priority analysis in relation to localization equipment for a thermal power plant. Parameters involved, such as readiness of both infrastructure and human resources, manpower as skilled labor, market potential for equipment developed by transferred technology, and competition in global/internal market, are related to localization in thermal power plant technologies, and are considered in relation to the country's technological capability, design ability, possession of materials/equipment, and ability to erect a plant. Results of analysis show that the boiler is the most important piece of equipment in this respect, and that heaters and fans are ranked after the boiler with respect to local conditions. - Highlights: • Localization of foreign technology was determined for developing countries. • An evaluation and priority analysis were performed for parts of a thermal power plant. • Analytic hierarchy process was applied for the hierarchical ordering of parts when transferring technology.

  3. International Economic Association on organization of co-operative production and development of equipment and providing technical assistance in construction of nuclear power plants - ''INTERATOMENERGO''

    International Nuclear Information System (INIS)

    Mal'tsev, N.D.

    1979-01-01

    History is stated of foundation of the International Economic Association ''Interatomenergo''. Structure is given of the Association and the list of main problems to be solved by it. Project is given of the programm of co-operation in the field of scientific and technical works as well as of design and projecting works in creation of new types of equipment for nuclear power plants, in particular, creation of serial power units with improved WWER-1000 reactor. Directions are stated of activity of the Association in the field of providing assistance in construction and exploitation of nuclear power plants as well as in training of operational personnel [ru

  4. Technical basis for environmental qualification of microprocessor-based safety-related equipment in nuclear power plants

    International Nuclear Information System (INIS)

    Korsah, K.; Wood, R.T.; Hassan, M.; Tanaka, T.J.

    1998-01-01

    This document presents the results of studies sponsored by the Nuclear Regulatory Commission (NRC) to provide the technical basis for environmental qualification of computer-based safety equipment in nuclear power plants. The studies were conducted by Oak Ridge National Laboratory (ORNL), Sandia National Laboratories (SNL), and Brookhaven National Laboratory (BNL). The studies address the following: (1) adequacy of the present test methods for qualification of digital I and C systems; (2) preferred (i.e., Regulatory Guide-endorsed) standards; (3) recommended stressors to be included in the qualification process during type testing; (4) resolution of need for accelerated aging for equipment to be located in a benign environment; and (5) determination of an appropriate approach for addressing the impact of smoke in digital equipment qualification programs. Significant findings from the studies form the technical basis for a recommended approach to the environmental qualification of microprocessor-based safety-related equipment in nuclear power plants

  5. Technical basis for environmental qualification of microprocessor-based safety-related equipment in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Korsah, K.; Wood, R.T. [Oak Ridge National Lab., TN (United States); Hassan, M. [Brookhaven National Lab., Upton, NY (United States); Tanaka, T.J. [Sandia National Labs., Albuquerque, NM (United States)

    1998-01-01

    This document presents the results of studies sponsored by the Nuclear Regulatory Commission (NRC) to provide the technical basis for environmental qualification of computer-based safety equipment in nuclear power plants. The studies were conducted by Oak Ridge National Laboratory (ORNL), Sandia National Laboratories (SNL), and Brookhaven National Laboratory (BNL). The studies address the following: (1) adequacy of the present test methods for qualification of digital I and C systems; (2) preferred (i.e., Regulatory Guide-endorsed) standards; (3) recommended stressors to be included in the qualification process during type testing; (4) resolution of need for accelerated aging for equipment to be located in a benign environment; and (5) determination of an appropriate approach for addressing the impact of smoke in digital equipment qualification programs. Significant findings from the studies form the technical basis for a recommended approach to the environmental qualification of microprocessor-based safety-related equipment in nuclear power plants.

  6. Integrated plant information technology design support functionality

    International Nuclear Information System (INIS)

    Kim, Yeon Seung; Kim, Dae Jin; Barber, P. W.; Goland, D.

    1996-06-01

    This technical report was written as a result of Integrated Plant Information System (IPIS) feasibility study on CANDU 9 project which had been carried out from January, 1994 to March, 1994 at AECL (Atomic Energy Canada Limited) in Canada. From 1987, AECL had done endeavour to change engineering work process from paper based work process to computer based work process through CANDU 3 project. Even though AECL had a lot of good results form computerizing the Process Engineering, Instrumentation Control and Electrical Engineering, Mechanical Engineering, Computer Aided Design and Drafting, and Document Management System, but there remains the problem of information isolation and integration. On this feasibility study, IPIS design support functionality guideline was suggested by evaluating current AECL CAE tools, analyzing computer aided engineering task and work flow, investigating request for implementing integrated computer aided engineering and describing Korean request for future CANDU design including CANDU 9. 6 figs. (Author)

  7. Integrated plant information technology design support functionality

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeon Seung; Kim, Dae Jin [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Barber, P W; Goland, D [Atomic Energy Canada Ltd., (Canada)

    1996-06-01

    This technical report was written as a result of Integrated Plant Information System (IPIS) feasibility study on CANDU 9 project which had been carried out from January, 1994 to March, 1994 at AECL (Atomic Energy Canada Limited) in Canada. From 1987, AECL had done endeavour to change engineering work process from paper based work process to computer based work process through CANDU 3 project. Even though AECL had a lot of good results form computerizing the Process Engineering, Instrumentation Control and Electrical Engineering, Mechanical Engineering, Computer Aided Design and Drafting, and Document Management System, but there remains the problem of information isolation and integration. On this feasibility study, IPIS design support functionality guideline was suggested by evaluating current AECL CAE tools, analyzing computer aided engineering task and work flow, investigating request for implementing integrated computer aided engineering and describing Korean request for future CANDU design including CANDU 9. 6 figs. (Author).

  8. Economic comparison of fusion power plant designs

    International Nuclear Information System (INIS)

    O'Neill, J.E.

    1986-01-01

    Over the past 10 yr, a number of studies have been developed for fusion power plants of various types (tokamaks, mirrors, etc.) complete with figures of merit such as cost estimates and estimates of the cost of generating electricity (COE). Each of these designs involves unresolved physics and engineering problems which, it is assumed, will eventually be worked out. Because of such uncertainties the figures of merit associated with such designs are not to be compared as absolute measures of worth but as relative indicators of progress within a given concept type. As part of Grumman's involvement in fusion energy development, an effort has been undertaken to compare economic indicators from the referenced studies in order to determine the cost trend in recent reactor design activities

  9. Design and operation of equipment used to develop remote coating capability for HTGR fuel particles

    International Nuclear Information System (INIS)

    Suchomel, R.R.; Stinton, D.P.; Preston, M.K.; Heck, J.L.; Bolfing, B.J.; Lackey, W.J.

    1978-12-01

    Refabrication of HTGR fuels is a manufacturing process that consists of preparation of fuel kernels, application of multiple layers of pyrolytic carbon and silicon carbide, preparation of fuel rods, and assembly of fuel rods into fuel elements. All the equipment for refabrication of 233 U-containing fuel must be designed for completely remote operation and maintenance in hot-cell facilities. Equipment to remotely coated HTGR fuel particles has been designed and operated. Although not all of the equipment development needed for a fully remote coating system has been completed, significant progress has been made. The most important component of the coating furnace is the gas distributor, which must be simple, reliable, and easily maintainable. Techniques for loading and unloading the coater and handling microspheres have been developed. An engineering-scale system, currently in operation, is being used to verify the workability of these concepts. Coating crucible handling components are used to remove the crucible from the furnace, remove coated particles, and exchange the crucible, if necessary. After the batch of particles has been unloaded, it is transferred, weighed, and sampled. The components used in these processes have been tested to ensure that no particle breakage or holdup occurs. Tests of the particle handling system have been very encouraging because no major problems have been encountered. Instrumentation that controls the equipment performed very smoothly and reliably and can be operated remotely

  10. A mechanical design for positioning of gm detector for system of avian flu virus detection equipment

    International Nuclear Information System (INIS)

    Rahmat; Budi Santoso; Krismawan; Abdul Jalil

    2010-01-01

    Mechanical design for positioning of GM detector system has been done. It is used for avian flu detection equipment. The requirements for the design are to protect detection system against shock, portable, and easy to maintain. The mechanical system consists of connectors, cable assemblies, holders, casing, housing and detectors cover. The selected material should have small gamma radiation absorption property in order to give optimum counts for the detector. The design result should give a system that is easy to operate, cheap and easy to assemble. (author)

  11. Mechanical design for positioning of GM detector for system of avian flu virus detection equipment

    International Nuclear Information System (INIS)

    Rahmat; Budi Santoso; Krismawan; Abdul Jalil

    2010-01-01

    Mechanical design for positioning of GM detector system has been done. It is used for avian flu detection equipment. The requirements for the design are to protect detection system against shock, portable, and easy to maintain. The mechanical system consists of connectors, cable assemblies, holders, casing, housing and detectors cover. The selected material should have small gamma radiation absorption property in order to give optimum counts for the detector. The design result should give a system that is easy to operate, cheap and easy to assemble. (author)

  12. Design-Only Conceptual Design Report: Plutonium Immobilization Plant

    International Nuclear Information System (INIS)

    DiSabatino, A.; Loftus, D.

    1999-01-01

    This design-only conceptual design report was prepared to support a funding request by the Department of Energy Office of Fissile Materials Disposition for engineering and design of the Plutonium Immobilization Plant, which will be used to immobilize up to 50 tonnes of surplus plutonium. The siting for the Plutonium Immobilization Plant will be determined pursuant to the site-specific Surplus Plutonium Disposition Environmental Impact Statement in a Plutonium Deposition Record of Decision in early 1999. This document reflects a new facility using the preferred technology (ceramic immobilization using the can-in-canister approach) and the preferred site (at Savannah River). The Plutonium Immobilization Plant accepts plutonium from pit conversion and from non-pit sources and, through a ceramic immobilization process, converts the plutonium into mineral-like forms that are subsequently encapsulated within a large canister of high-level waste glass. The final immobilized product must make the plutonium as inherently unattractive and inaccessible for use in nuclear weapons as the plutonium in spent fuel from commercial reactors and must be suitable for geologic disposal. Plutonium immobilization at the Savannah River Site uses: (1) A new building, the Plutonium Immobilization Plant, which will convert non-pit surplus plutonium to an oxide form suitable for the immobilization process, immobilize plutonium in a titanate-based ceramic form, place cans of the plutonium-ceramic forms into magazines, and load the magazines into a canister; (2) The existing Defense Waste Processing Facility for the pouring of high-level waste glass into the canisters; and (3) The Actinide Packaging and Storage Facility to receive and store feed materials. The Plutonium Immobilization Plant uses existing Savannah River Site infra-structure for analytical laboratory services, waste handling, fire protection, training, and other support utilities and services. The Plutonium Immobilization Plant

  13. An AI-based layout design system for nuclear power plants

    International Nuclear Information System (INIS)

    Fujita, Kikuo; Akagi, Shinsuke; Nakatogawa, Tetsundo; Tanaka, Kazuo; Takeuchi, Makoto.

    1991-01-01

    An AI-based layout design system for nuclear power plants has been developed. The design of the layout of nuclear power plants is a time-consuming task requiring expertise, in which a lot of machinery and equipment must be arranged in a plant building considering various kinds of design constraints, i.e. spatial, functional, economical etc. Computer aided layout design systems have been widely expected and the application of AI technology is expected as a promising approach for the synthesis phase of this task. In this paper, we present an approach to the layout design of nuclear power plants based on a constraint-directed search; one of the AI techniques. In addition, we show how it was implemented with an object-oriented programming technique and give an example of its application. (author)

  14. New nuclear plant design and licensing process

    International Nuclear Information System (INIS)

    Luangdilok, W.

    1996-01-01

    This paper describes latest developments in the nuclear power reactor technology with emphasis on three areas: (1) the US technology of advanced passive light water reactors (AP600 and S BWR), (2) regulatory processes that certify their safety, and (3) current engineering concerns. The goal is to provide and insight of how the government's regulatory agency guarantees public safety by looking into how new passive safety features were designed and tested by vendors and how they were re-evaluated and retested by the US NRC. The paper then discusses the US 1989 nuclear licensing reform (10 CFR Part 52) whose objectives are to promote the standardization of nuclear power plants and provide for the early and definitive resolution of site and design issues before plants are built. The new licensing process avoids the unpredictability nd escalated construction cost under the old licensing process. Finally, the paper summarizes engineering concerns found in current light water reactors that may not go away in the new design. The concerns are related the material and water chemistry technology in dealing with corrosion problems in water-cooled nuclear reactor systems (PWRs and BWRs). These engineering concerns include core shroud cracking (BWRs), jet pump hold-down beam cracking (BWRs), steam generator tube stress corrosion cracking (PWR)

  15. OLP embedment design method research for AP1000 nuclear plant

    International Nuclear Information System (INIS)

    Li Cheng; Li Shaoping; Liu Jianwei

    2013-01-01

    Background: One of the most advanced nuclear power technology, the first AP1000 reactor is under construction in China. Modularization is one of the main characteristics for AP1000 nuclear plant building. Module wall with steel face plate is used instead of reinforced concrete structure wall. A number of OLP embedments need to be installed into the module wall to connect other structures such as pipes, equipment, operation platforms and any other component attached to the module wall. Therefore, the design of embedment is very important in AP1000 structural design. Purpose: A finite element analysis method and tool for embedment design is needed for convenience. Methods: This paper applies the self-developed GTStrudl command template and VBA macro program for embedment capacity calculation and evaluation based on Microsoft Excel to the embedment design. Results: A Microsoft Excel template for embedment design is developed. Conclusions: The analysis method and template brings reasonable results and may provide some help and use for reference for the engineering practice. (authors)

  16. Remote-automation of nuclear power plant equipment inspection and maintenance

    International Nuclear Information System (INIS)

    Sasaki, Masayoshi; Kawamura, Hironobu; Nakano, Yoshiyuki; Izumi, Shigeru.

    1984-01-01

    The remotely operated automation of the checkup and maintenance of nuclear power generation facilities has largely contributed to the rise of capacity ratio of plants due to the shortening of regular inspection period and to the reduction of radiation exposure dose during working, the labor saving in working and so on. In this paper, the new technologies adopted in an automatic fuel exchanger, a remotely operated automatic CRD exchanger, a new type channel handling machine, pressure-withstanding main steam line plugs and so on for No.2 plant in the Fukushima No.2 Nuclear Power Station, Tokyo Electric Power Co., Inc., are reported. Besides, the state of development of new remotely operated automatic machines for nuclear power use, such as CRD disassembling and cleaning device, volume reduction equipment for spent fuel channel boxes and control rods, multi-functional robots for use under high radiation and so on is described. Also the trend of development of latest robot technology which will be put in practical use in near future is outlined, such as a running manipulator for checkup and inspection, a variable form crawler vehicle and a five-leg movable manipulator. (Kako, I.)

  17. Impact of Auxiliary Equipments Consumption on Electricity Generation Cost in Selected Power Plants of Pakistan

    Directory of Open Access Journals (Sweden)

    DILEEP KUMAR

    2017-04-01

    Full Text Available This study focuses on higher generation cost of electricity in selected TPPs (Thermal Power Plants in Sindh, Pakistan. It also investigates the energy consumed by the auxiliary equipment of the selected TPPs in Sindh, Pakistan. The AC (Auxiliary Consumption of selected TPPs is compared with that in UK and other developed countries. Results show that the AC in selected TPPs in Sindh, Pakistan exceeds the average AC of the TPPs situated in developed countries. Many energy conservation measures such as impeller trimming and de-staging, boiler feed pump, high voltage inverter, variable frequency drive, and upgrading the existing cooling tower fan blades with fiber reinforced plastic are discussed to overcome higher AC. This study shows that harnessing various available energy conservative measures the AC and unit cost can be reduced by 4.13 and 8.8%; also adverse environmental impacts can be mitigated. Results show that the unit cost of electricity can be reduced from Rs.20 to19/kWh in JTPP (Jamshoro Thermal Power Plant, Rs.9 to 8.8/kWh in GTPS (Gas Turbine Power Station Kotri and Rs. 11 to 10.27/ kWh in LPS (Lakhara Power Station. Thus, electricity production can be improved with the existing capacity, which will eventually assist to manage the current energy crisis and ensure its conservation

  18. Prototypical spent fuel rod consolidation equipment preliminary design report: Volume 1, Report

    International Nuclear Information System (INIS)

    1986-01-01

    This design report describes the NUS Preliminary Design of the Prototype Spent Nuclear Fuel Rod Consolidation Equipment for the Department of Energy. The sections of the report elaborate on each facet of the preliminary design. A concept summary is provided to assist the reader in rapidly understanding the complete design. The NUS Prototype Spent Fuel Rod Consolidation System is an automatically controlled system to consolidate a minimum of 750 MT (heavy metal)/year of US commercial nuclear reactor fuel, at 75% availability. The system is designed with replaceable components utilizing the latest state-of-the-art technology. This approach gives the system the flexibility to be developed without costly development programs, yet accept new technology as it evolves over the next ten years. Capability is also provided in the system design to accommodate a wide variety of fuel conditions and to recover from any situation which may arise

  19. Design and implementation of equipment for monitoring the salinity in the subsoil

    Directory of Open Access Journals (Sweden)

    Mariano Norzagaray Campos

    2012-09-01

    Full Text Available Geochemical exploration equipments to explore the contaminants and structures geological in the subsoil come from abroad are expensive and sometimes parts for replacement are not available in the market. So it is necessary design apparatus that meet cover these needs. To the monitoring of variations in salinity there is semi-automatic equipment, but it always has difficult to manage. However, is not equipment for the indirect study of salinity in the subsoil. In this work was design equipment for measurement the apparent resistivity in the subsoil, at same time allow know the salinity, as well as the detection of any pollutant in groundwater. For make it, was selected a design of earthing systems, with electronic hardware which were jointed for apply to subsoil a direct current (DC through an array dipole-dipole and vertical electric sounding, with brass and stainless steel electrodes. In the earthing systems the electrodes were collocated in the equidistant line between the detectors of potential and current. A geometric factor (K, that depend on theelectrodes distance and direct current (I injected in the electrodes A and B, was used for measure the potential difference between the electrodes M and N; after was calculate the resistivity point to point for obtain a subsoil tomography geoelectrical. The equipment was calibrated with minimum error (rms < 2% whit respect to curves obtained in similar commercial equipment. On this situation, in this work was modernized and automated an equipment to determine thesalinity of the subsoil. The instrument was tasted in the micro basin Texcoco, State of Mexico, to define the environment geometry formed by alluvial or lake sediments from igneous rocks (andesites, rhyolites and tuffs vitreous or “tepetates” which by its mineralogical composition allowed to the lateral resistivity be associate with free components trace from the aquifer: Cd, Cu, Cr, Co, Ni, Pb or Zn and others. This method constitutes

  20. Biorefinery plant design, engineering and process optimisation

    DEFF Research Database (Denmark)

    Holm-Nielsen, Jens Bo; Ehimen, Ehiazesebhor Augustine

    2014-01-01

    Before new biorefinery systems can be implemented, or the modification of existing single product biomass processing units into biorefineries can be carried out, proper planning of the intended biorefinery scheme must be performed initially. This chapter outlines design and synthesis approaches...... applicable for the planning and upgrading of intended biorefinery systems, and includes discussions on the operation of an existing lignocellulosic-based biorefinery platform. Furthermore, technical considerations and tools (i.e., process analytical tools) which could be applied to optimise the operations...... of existing and potential biorefinery plants are elucidated....

  1. Biogas plants: Design, construction and operation

    International Nuclear Information System (INIS)

    2001-01-01

    At the big readiness of waste coming from the agricultural activities are looked for the production of Energy and Payments, the biogas like product of the organic decomposition under anaerobic conditions, their composition and characteristic. The elements that conform the design as the digester, the storage, the load tanks and it discharges and the conduction is described and analyzed. They are given a series of elements to obtain the characteristics of the system possible to place as: planning, calculations, evaluation, execution and operation. Lastly the steps are indicated that should be continued in the construction of the plant including planning for the work

  2. Designs of new plants of high capacity

    International Nuclear Information System (INIS)

    Borges R, Diego A

    1999-01-01

    The Caracas electricity in their desire to lend the best service to the community is doing the necessary projects of generation expansion to fulfill and requirements of demand of next decade in to the metropolitan area. The projects of the new plants of Recifes and El Sitio, have been conceptualized in way of executing the engineering and patterns construction of highest quality and once setting in service, to reach the highest indexes of operative. To reach these goals it is planned to use the most advanced technological designs that are in the market at the moment to world in generation and transmission of power

  3. Design of free-space optical transmission system in computer tomography equipment

    Science.gov (United States)

    Liu, Min; Fu, Weiwei; Zhang, Tao

    2018-04-01

    Traditional computer tomography (CT) based on capacitive coupling cannot satisfy the high data rate transmission requirement. We design and experimentally demonstrate a free-space optical transmission system for CT equipment at a data rate of 10 Gb / s. Two interchangeable sections of 12 pieces of fiber with equal length is fabricated and tested by our designed laser phase distance measurement system. By locating the 12 collimators in the edge of the circle wheel evenly, the optical propagation characteristics for the 12 wired and wireless paths are similar, which can satisfy the requirement of high-speed CT transmission system. After bit error rate (BER) measurement in several conditions, the BER performances are below the value of 10 - 11, which has the potential in the future application scenario of CT equipment.

  4. Experience in the use of teleoperated robotic equipment in the Spanish Nuclear Power Plants (S.R.T. Project)

    International Nuclear Information System (INIS)

    Bielza, M.; Gomez, J.; Izquierdo, J.A.; Bercedo, A.; Espallardo, J.A.; Martinez, A.; Carmena, P.; Pascual, J.L.

    1998-01-01

    The main objective of the Teleoperated Robotic Systems project (SRT) is the development of teleoperated robotic systems for use in the inspection, surveillance and maintenance operations in nuclear and radioactive installations. These systems make possible the reduction of the individual and collective doses of the workers that undertake these operations, as well as an increase of plant availability as it is possible to carry out specific tasks of inspection and surveillance in high radiation dose areas without to reduce the power of the installation. This project started in 1995 deciding to priorize the inspection equipment in a first phase. As a result of this work currently several semi-industrial products are available: ROBCAR, ROBICEN y ANAES. These have been installed for testing and commercial operation in Spanish nuclear installations. In this paper the experience of using these prototypes in the S.M. de Garona and Cofrentes nuclear power plants is described, as well as the improvements in the design of these systems derived from these experiences that have been incorporated to the industrial prototypes that are currently in the last phase of development. (Author)

  5. Application of Microprocessor-Based Equipment in Nuclear Power Plants - Technical Basis for a Qualification Methodology

    International Nuclear Information System (INIS)

    Korsah, K.

    2001-01-01

    This document (1) summarizes the most significant findings of the ''Qualification of Advanced Instrumentation and Control (I and C) Systems'' program initiated by the Nuclear Regulatory Commission (NRC); (2) documents a comparative analysis of U.S. and European qualification standards; and (3) provides recommendations for enhancing regulatory guidance for environmental qualification of microprocessor-based safety-related systems. Safety-related I and C system upgrades of present-day nuclear power plants, as well as I and C systems of Advanced Light-Water Reactors (ALWRs), are expected to make increasing use of microprocessor-based technology. The Nuclear Regulatory Commission (NRC) recognized that the use of such technology may pose environmental qualification challenges different from current, analog-based I and C systems. Hence, it initiated the ''Qualification of Advanced Instrumentation and Control Systems'' program. The objectives of this confirmatory research project are to (1) identify any unique environmental-stress-related failure modes posed by digital technologies and their potential impact on the safety systems and (2) develop the technical basis for regulatory guidance using these findings. Previous findings from this study have been documented in several technical reports. This final report in the series documents a comparative analysis of two environmental qualification standards--Institute of Electrical and Electronics Engineers (IEEE) Std 323-1983 and International Electrotechnical Commission (IEC) 60780 (1998)--and provides recommendations for environmental qualification of microprocessor-based systems based on this analysis as well as on the findings documented in the previous reports. The two standards were chosen for this analysis because IEEE 323 is the standard used in the U.S. for the qualification of safety-related equipment in nuclear power plants, and IEC 60780 is its European counterpart. In addition, the IEC document was published in

  6. Design of inspection equipment vehicle of mobile cobalt-60 container scanner

    International Nuclear Information System (INIS)

    Zhang Yanmin; An Jigang; Zhou Liye; Liu Yisi; Wu Zhifang; Xiang Xincheng; Huang Songling

    2000-01-01

    The author introduced the design principle of Inspection Equipment Vehicle of TCM-SCAN Mobile 60 Co container inspection system. Three schemes had been compared. The scheme of scan gate in form of pucker-able rails standing on the vehicle was determined. Prototype was prepared in September 1999. Mobile 60 Co container inspection system had the same inspection performance of fixed 60 Co container inspection system

  7. Cyber and physical equipment digital control system in Industry 4.0 item designing company

    Science.gov (United States)

    Gurjanov, A. V.; Zakoldaev, D. A.; Shukalov, A. V.; Zharinov, I. O.

    2018-05-01

    The problem of organization of digital control of the item designing company equipped with cyber and physical systems is being studied. A scheme of cyber and physical systems and personnel interaction in the Industry 4.0 smart factory company is presented. A scheme of assembly units transportation in the Industry 4.0 smart factory company is provided. A scheme of digital control system in the Industry 4.0 smart factory company is given.

  8. Design and construction of stepper motor controller for brachy-therapy equipment

    International Nuclear Information System (INIS)

    Ahmad Rifai; Usep Setia Gunawan; Indarzah Masbatin Putra

    2011-01-01

    Based on a microcontroller, a stepper motor controller for brachy-therapy equipment has been designed and prototyped. The embedded control program routine and its other associated routines enable the microcontroller to generate required pulses via external commands for stepper motor based position control. The controller receives ASCII text command via a serial port. The constructed algorithm implements the widely used method that allows motor to rotate in three phases, i.e: acceleration, constant speed, and deceleration. (author)

  9. Preparation of plant and system design description documents

    International Nuclear Information System (INIS)

    1989-01-01

    This standard prescribes the purpose, scope, organization, and content of plant design requirements (PDR) documents and system design descriptions (SDDs), to provide a unified approach to their preparation and use by a project as the principal means to establish the plant design requirements and to establish, describe, and control the individual system designs from conception and throughout the lifetime of the plant. The Electric Power Research Institute's Advanced Light Water Reactor (LWR) Requirements Document should be considered for LWR plants

  10. Survey of injury sources for a trampoline with equipment hazards designed out.

    Science.gov (United States)

    Eager, David; Scarrott, Carl; Nixon, Jim; Alexander, Keith

    2012-07-01

    In Australia, trampolines contribute approximately one-quarter of all childhood play-equipment injuries. The purpose of this study was to gather and evaluate injury data from a nontraditional, 'soft-edged', consumer trampoline in which the equipment injury sources have been designed out. A survey was undertaken in Queensland and New South Wales. The manufacturer of the nontraditional trampoline provided the University of Technology, Sydney, with their Australian customer database. Injury data were gathered in a pilot study by phone interview, then in a full study through an email survey. Results from 3817 respondents were compared with earlier Australian and US data from traditional trampolines gathered from emergency departments.   A significantly lower proportion of the injuries caused by falling off or striking the equipment was found for this new design when compared with traditional trampolines both in Australia and in the USA. The age of children being injured on trampolines in Australia was found to be markedly lower than in North America. This research indicates that with appropriate design the more severe injuries on traditional trampolines can be significantly reduced. © 2012 The Authors. Journal of Paediatrics and Child Health © 2012 Paediatrics and Child Health Division (Royal Australasian College of Physicians).

  11. Aseismic design of Hamaoka Nuclear Power Plant

    International Nuclear Information System (INIS)

    Mizuno, Norihiro

    1975-01-01

    The Hamaoka Nuclear Power Plant of Chubu Electric Power Co. is designed so as to maintain structural safety against an earthquake of 300 gal. For the purpose, a compound-type reactor-housing building is employed, which contains a reactor, operation control and waste disposal facilities. The merits accruing from this scheme are as follows. (1) The shielding walls of the waste disposal facility can be utilized effectively in aseismatic design, leading to the increased rigidity of the building and the uniform distribution of resistance. (2) Due to the large area of the foundation, the load in earthquake can be mitigated, and it resulted in the higher structural stability. Moreover, seismic energy can be dissipated into ground. After the description of the compound building structure, it is explained how the structural resistance and the ground dissipation of seismic energy contribute to potential earthquake resistance. (Mori, K.)

  12. Spherical tokamak power plant design issues

    International Nuclear Information System (INIS)

    Hender, T.C.; Bond, A.; Edwards, J.; Karditsas, P.J.; McClements, K.G.; Mustoe, J.; Sherwood, D.V.; Voss, G.M.; Wilson, H.R.

    2000-01-01

    The very high β potential of the spherical tokamak has been demonstrated in the START experiment. Systems code studies show the cost of electricity from spherical tokamak power plants, operating at high β in second ballooning mode stable regime, is comparable with fossil fuels and fission. Outline engineering designs are presented based on two concepts for the central rod of the toroidal field (TF) circuit - a room temperature water cooled copper rod or a helium cooled cryogenic aluminium rod. For the copper rod case the TF return limbs are supported by the vacuum vessel, while for the aluminium rod the TF coils form an independent structure. In both cases thermohydraulic and stress calculations indicate the viability of the design. Two-dimensional neutronics calculations show the feasibility of tritium self-sufficiency without an inboard blanket. The spherical tokamak has unique maintenance possibilities based on lowering major component structures into a hot cell beneath the device and these are discussed

  13. Safety of Nuclear Power Plants: Design. Specific Safety Requirements

    International Nuclear Information System (INIS)

    2012-01-01

    On the basis of the principles included in the Fundamental Safety Principles, IAEA Safety Standards Series No. SF-1, this Safety Requirements publication establishes requirements applicable to the design of nuclear power plants. It covers the design phase and provides input for the safe operation of the power plant. It elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

  14. Safeguards by Design at the Encapsulation Plant in Finland

    International Nuclear Information System (INIS)

    Ingegneri, M.; Baird, K.; Park, W.-S.; Coyne, J.M.; Enkhjin, L.; Chew, L.S.; Plenteda, R.; Sprinkle, J.; Yudin, Y.; Ciuculescu, C.; Koutsoyannopoulos, C.; Murtezi, M.; Schwalbach, P.; Vaccaro, S.; Pekkarinen, J.; Thomas, M.; Zein, A.; Honkamaa, T.; Hamalainen, M.; Martikka, E.; Moring, M.; Okko, O.

    2015-01-01

    Finland has launched a spent fuel disposition project to encapsulate all of its spent fuel assemblies and confine the disposal canisters in a deep geological repository. The construction of the underground premises started several years ago with the drilling, blasting and reinforcement of tunnels and shafts to ensure the safe deep underground construction and disposal techniques in the repository, while the design of the encapsulation plant (EP) enters the licencing phase preliminary to its construction. The spent fuel assemblies, which have been safeguarded for decades at the nuclear power plants, are going to be transported to the EP, loaded into copper canisters and stored in underground tunnels where they become inaccessible after backfilling. Safeguards measures are needed to ensure that final spent fuel verification is performed before its encapsulation and that no nuclear material is diverted during the process. This is an opportunity for the inspectorates to have the infrastructure necessary for the safeguards equipment incorporated in the design of the encapsulation plant before licencing for construction occurs. The peculiarity of this project is that it is going to run for more than a century. Therefore, significant changes are to be expected in the technical capabilities available for implementing safeguards (e.g., verification techniques and instruments), as well as in the process itself, e.g., redesign for the encapsulation of future fuel types. For these reasons a high degree of flexibility is required in order to be able to shift to different solutions at a later stage while minimizing the interference with the licencing process and facility operations. This paper describes the process leading to the definition of the technical requirements by IAEA and Euratom to be incorporated in the facility's design. (author)

  15. DC systems design and research of Hainan Changjiang nuclear power plant

    International Nuclear Information System (INIS)

    Jiang Qingshui; Wang Yuhan

    2014-01-01

    Hainan Changjiang nuclear power plant is different from the referent power plant, the DC and 220 V AC uninterrupted systems of the nuclear island have been changed since the control system use DCS. It has different design on DC systems, power supply, selectivity of breakers, capacity of equipments and layout. We optimize the design of DC systems at the basement of Fuqing and Fangjiashan project. These are good experiments for the three generation nuclear power project about DC systems design of ACP1000. (authors)

  16. 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)

  17. The Design and Experiment of the Track-Type Equipment for Feeding Dairy Cows

    Directory of Open Access Journals (Sweden)

    Yaping LI

    2014-06-01

    Full Text Available As one of the effective ways of reducing the cost and increase the revenue, feeding dairy cows with individual precision in quantity could further develop the production potential of high- yield dairy cows and improve the milk yield of every dairy cow. Therefore, based on three kinds of technologies of radio frequency identification technology, wireless transmission technology and infrared technology and grounded on the foundation of previous studies, this thesis has completed the track- type equipment for feeding dairy cows with individual precision in quantity. The equipment takes the computer as the information management platform, adopts the singly chip microcomputer as the control platform, uses the wireless module and U disk to conduct the transmission of individual feeding data of dairy cows and guides the whole process with the track. This kind of equipment could advance automatically, identify and feed dairy cows precisely during the feeding process. This thesis also designs the matching feeding technology of the equipment, and accomplishes the test of feeding outside the cowshed, which takes “TMR + precise feeding with concentrated feed” as the feeding model, so that the feeding process could be more effective and further steady.

  18. Mercury Control for Plants Firing Texas Lignite and Equipped with ESP-wet FGD

    Energy Technology Data Exchange (ETDEWEB)

    Katherine Dombrowski

    2009-12-31

    This report presents the results of a multi-year test program conducted as part of Cooperative Agreement DE-FC26-06NT42779, 'Mercury Control for Plants Firing Texas Lignite and Equipped with ESP-wet FGD.' The objective of this program was to determine the level of mercury removal achievable using sorbent injection for a plant firing Texas lignite fuel and equipped with an ESP and wet FGD. The project was primarily funded by the U.S. DOE National Energy Technology Laboratory. EPRI, NRG Texas, Luminant (formerly TXU), and AEP were project co-funders. URS Group was the prime contractor, and Apogee Scientific and ADA-ES were subcontractors. The host site for this program was NRG Texas Limestone Electric Generating Station (LMS) Units 1 and 2, located in Jewett, Texas. The plant fires a blend of Texas lignite and Powder River Basin (PRB) coal. Full-scale tests were conducted to evaluate the mercury removal performance of powdered sorbents injected into the flue gas upstream of the ESP (traditional configuration), upstream of the air preheater, and/or between electric fields within the ESP (Toxecon{trademark} II configuration). Phases I through III of the test program, conducted on Unit 1 in 2006-2007, consisted of three short-term parametric test phases followed by a 60-day continuous operation test. Selected mercury sorbents were injected to treat one quarter of the flue gas (e.g., approximately 225 MW equivalence) produced by Limestone Unit 1. Six sorbents and three injection configurations were evaluated and results were used to select the best combination of sorbent (Norit Americas DARCO Hg-LH at 2 lb/Macf) and injection location (upstream of the ESP) for a two-month performance evaluation. A mercury removal rate of 50-70% was targeted for the long-term test. During this continuous-injection test, mercury removal performance and variability were evaluated as the plant operated under normal conditions. Additional evaluations were made to determine any

  19. Innovative Offshore Wind Plant Design Study

    Energy Technology Data Exchange (ETDEWEB)

    Hurley, William L. [Glosten Associates, Inc., Seattle, WA (United States); Nordstrom, Charles J. [Glosten Associates, Inc., Seattle, WA (United States); Morrison, Brent J. [Glosten Associates, Inc., Seattle, WA (United States)

    2013-12-18

    Technological advancements in the Glosten PelaStar floating wind turbine system have led to projected cost of energy (COE) reductions from today’s best-in-class offshore wind systems. The PelaStar system is projected to deliver a COE that is 35% lower than that delivered by the current offshore wind plants. Several technology developments have been achieved that directly target significant cost of energy reductions. These include: Application of state-of-the-art steel construction materials and methods, including fatigue-resistant welding techniques and technologies, to reduce hull steel weight; Advancements in synthetic fiber tendon design for the mooring system, which are made possible by laboratory analysis of full-scale sub-rope specimens; Investigations into selected anchor technologies to improve anchor installation methods; Refinement of the installation method, specifically through development of the PelaStar Support Barge design. Together, these technology developments drive down the capital cost and operating cost of offshore wind plants and enable access to superb wind resources in deep water locations. These technology developments also reduce the uncertainty of the PelaStar system costs, which increases confidence in the projected COE reductions.

  20. Radio frequency sputtering equipment: design considerations for the disc and annulus system

    International Nuclear Information System (INIS)

    Usher, D.M.; Cox, G.A.

    1981-01-01

    The basic equations important in the design of rf sputtering equipment are presented. Attention is restricted to the 'disc and annulus' system and simple techniques in lead-through fabrication and electrode shielding are described. Also rf power oscillator theory is used to derive the important parameters of power generation. Load coupling and power matching techniques are discussed. Problems of the magnetic field coil design are analysed and a value for the thermal conductivity of coils wound from polyurethane coated wire is given, namely k = 0.64 W m -1 / 0 C -1 , permitting the steady state temperature distribution inside such coils to be predicted. (author)

  1. Research, Calculation for Designing the Technological Equipment Line of Cementation with 3,000 Tonnes of Solid Waste/Year Capacity for ZOC Plan

    International Nuclear Information System (INIS)

    Tran Van Hoa; Ngo Van Tuyen; Nguyen Ba Tien; Le Ba Thuan; Phung Quoc Khanh; Mai Duc Lanh; Vuong Huu Anh; Luu Cao Nguyen; Duong Dinh Tho; Phung Vu Phong; Nguyen An Thai; Pham Quang Minh

    2013-01-01

    Technological process and equipment line for cementation with 3,000 tons of solid wastes/year capacity for ZOC plant was studied according to the quantity, the composition and the property of solid wastes from ZOC plant. The influence of several parameters with cementation process such as cement mixing ratio, size, moisture and NaCl content in residue, which affected to mechanical and chemical strength of waste block after cementation was evaluated. The calculation and designing for cementation equipment line were thus conducted and total investment cost of construction and installation as well as operation of this equipment line was also estimated. Radiation safety and environmental protection for waste treatment facility was calculated and the design on radiation safety and environmental protection for equipment line was proposed. Technological process and design document of this production line seemed to be reasonable because it is consistent with properties of wastes as well as it has been commonly used in the world to stabilize solid residues in nuclear industry. The advantages of this propose were simple structure of the devices, locally made, easy operation, locally available and inexpensive materials resulting in low cost of investment and operation. (author)

  2. Study on designing a complete pilot plant for processing sandstone ores in Palua-Parong area

    International Nuclear Information System (INIS)

    Le Quang Thai; Tran Van Son; Tran The Dinh; Trinh Nguyen Quynh; Vu Khac Tuan

    2015-01-01

    Design work is the first step of the construction and operation of pilot plant. Thus, the project Study on designing a complete pilot plant for processing sandstone ores in Palua - Parong area was conducted to design a pilot plant for testing entire technological process to obtain yellowcake. Based on a literature review of uranium ore processing technology in the world, information of ore and previous research results of uranium ore in PaLua - PaRong area at the Institute for Technology of Radioactive and Rare Elements, a suitable technological flowsheet for processing this ore has been selected. The size, location of the pilot plant and planed experiments has been selected during the implementation of this project, in which basic parameters, designed system of equipment, buildings, ect. were also calculated. (author)

  3. Development of preventive maintenance technology and advanced service equipment for operating nuclear power plants

    International Nuclear Information System (INIS)

    Abe, Kazuhiro; Sumikawa, Masaharu; Hirakawa, Hiromasa; Arakawa, Tadao; Hasegawa, Kunio; Kato, Kanji.

    1990-01-01

    Hitachi Ltd. as a manufacturer of nuclear power plants has carried out the consistent general preventive maintenance activities from the planning of the plants and the design of maintainability in the construction phase to the planning and working of preventive maintenance in the operation and maintenance phase, and exerted efforts to heighten the capacity ratio and reliability of the plants. For the purpose, the steady activities of reliability improvement have been carried out throughout the whole company, and the rationalization of the planning and management of the preventive maintenance with a computer and the development of the robots to which Al is applied have been promoted. As the technology of upgrading the facilities, boron racks, the control rods having long life and so on were developed, and their practical use is advanced. Moreover for the future, the development of the diagnostic technology on material deterioration using superconductive quantum interference devices (SQUID) is in progress. The preventive maintenance activities in Hitachi Ltd., the technical development for the purpose and the upgrading of the plant facilities are reported. (K.I.)

  4. Preliminary design report: Prototypical Spent Fuel Consolidation Equipment Demonstration Project: Phase 1

    International Nuclear Information System (INIS)

    Blissell, W.H.; Ciez, A.P.; Mitchell, J.L.; Winkler, C.J.

    1986-12-01

    This document describes the Westinghouse Preliminary Design for the Prototypical Consolidation Demonstration Project per Department of Energy (DOE) Contract No. DE-AC07-86ID12649 and under direction of the DOE Idaho Operations Office. The preliminary design is the first step to providing the Department of Energy with a fully qualified, licensable, cost-effective spent fuel rod consolidation system. The design was developed using proven technologies and equipment to create an innovative approach to previous rod consolidation concepts. These innovations will better enable the Westinghouse system to: consolidate fuel rods in a precise, fully-controlled, accountable manner; package all rods from two PWR fuel assemblies or from four BWR fuel assemblies in one 8.5 inch square consolidated rods canister; meet all functional requirements; operate with all fuel types common to the US commercial nuclear industry with minimal tooling changeouts; and meet consolidation production process rates, while maintaining operator and public health and safety. This Preliminary Design Report provides both detailed descriptions of the equipment required to perform the rod consolidation process and the supporting analyses to validate the design

  5. Design concepts of nuclear desalination plants

    International Nuclear Information System (INIS)

    2002-11-01

    Interest in using nuclear energy for producing potable water has been growing worldwide in the past decade. This has been motivated by a variety of factors, including economic competitiveness of nuclear energy, the growing need for worldwide energy supply diversification, the need to conserve limited supplies of fossil fuels, protecting the environment from greenhouse gas emissions, and potentially advantageous spin-off effects of nuclear technology for industrial development. Various studies, and at least one demonstration project, have been considered by Member States with the aim of assessing the feasibility of using nuclear energy for desalination applications under specific conditions. In order to facilitate information exchange on the subject area, the IAEA has been active for a number of years in compiling related technical publications. In 1999, an inter regional technical co-operation project on Integrated Nuclear Power and desalination System Design was launched to facilitate international collaboration for the joint development by technology holders and potential end users of an integrated nuclear desalination system. This publication presents material on the current status of nuclear desalination activities and preliminary design concepts of nuclear desalination plants, as made available to the IAEA by various Member States. It is aimed at planners, designers and potential end-users in those Member States interested in further assessment of nuclear desalination. Interested readers are also referred to two related and recent IAEA publications, which contain useful information in this area: Introduction of Nuclear Desalination: A Guidebook, Technical Report Series No. 400 (2000) and Safety Aspects of Nuclear Plants Coupled with Seawater Desalination Units, IAEA-TECDOC-1235 (2001)

  6. Methodological framework for the ergonomic design of children's playground equipment: a Serbian experience.

    Science.gov (United States)

    Grozdanovic, Miroljub; Jekic, Savko; Stojiljkovic, Evica

    2014-01-01

    Adequate application of the static and dynamic anthropometric measures of pre-school children in ergonometric design of children's playground equipment should eliminate all dangers and difficulties in their use. Possibilities of injuries, insecure movements, discomfort able positions and some other dangerous actions may be minimized; and safety and health protection of pre-school children will be increased. Children's playground represents a significant space of activity for pre-school children. Therefore, it is necessary to apply ergonomic principles which contribute to the adjustment of the playground elements to children's anatomic features. Based on the results presented in this paper, new constructions were designed and new playgrounds were installed in Serbia. Participants were children from three pre-school age groups: Junior age group (3-4 years of age, 17 children), Medium age group (4-5 years of age, 22 children), and Senior age group (5-6 years of age, 26 children). Thirty-one static anthropometric measures (12 in standing position, 11 in sitting position, 7 related to dimensions of hand, foot an head, with body weight and shoe size) and 15 dynamic anthropometric measures (7 in standing position, 6 in sitting position and 2 dimension of foot and hand) were defined for the study. Measurements were taken using an anthrop-meter, a flexible measuring tape. Equations for ergonomic design of children's playground elements were also defined. Basic statistical data of static and dynamic anthropometric measurements of the pre-school children are presented in this paper, as well as the statistical calculation of the corrective anthropometric measurements. Measurements were performed in "Poletarac" kindergarten, part of the pre-school institution "Radost" in Cacak. Elements of playground equipment in "Bambi" kindergarten in Kragujevac (the Indian tent "wigwam", gate-house, swing and carousel) were designed and built using these parameters. Based on the obtained

  7. Seismic design margin evaluation of systems and equipment required for safe shutdown of North Anna, Units 1 and 2, following an SSE (safe-shutdown earthquake) event. Technical report

    International Nuclear Information System (INIS)

    Desai, K.D.

    1981-06-01

    The Advisory Committee on Reactor Safeguards recommended that the NRC staff review in detail the capability and available seismic design margin of fluid systems and equipment used in North Anna, Units 1 and 2 to achieve safe shutdown following a site-design safe-shutdown earthquake (SSE). The staff conducted a series of plant visits and meetings with the licensee to view and discuss the seismic design methodology used for systems, equipment and their supports. The report is a description and evaluation of the seismic design criteria, design conservatisms and seismic design margin for North Anna, Units 1 and 2

  8. Study on compact design of remote handling equipment for ITER blanket maintenance

    International Nuclear Information System (INIS)

    Takeda, Nobukazu; Kakudate, Satoshi; Nakahira, Masataka; Shibanuma, Kiyoshi

    2006-03-01

    In the ITER, the neutrons created by D-T reactions activate structural materials, and thereby, the circumstance in the vacuum vessel is under intense gamma radiation field. Thus, the in-vessel components such as blanket are handled and replaced by remote handling equipment. The objective of this report is to study the compactness of the remote handling equipment (a vehicle/manipulator) for the ITER blanket maintenance. In order to avoid the interferences between the blanket and the equipment during blanket replacement in the restricted vacuum vessel, a compact design of the equipment is required. Therefore, the compact design is performed, including kinematic analyses aiming at the reduction of the sizes of the vehicle equipped with a manipulator handling the blanket and the rail for the vehicle traveling in the vacuum vessel. Major results are as follows: 1. The compact vehicle/manipulator is designed concentration on the reduction of the rail size and simplification of the guide roller mechanism as well as the reduction of the gear diameter for vehicle rotation around the rail. Height of the rail is reduced from 500 mm to 400 mm by a parameter survey for weight, stiffness and stress of the rail. The roller mechanism is divided into two simple functional mechanisms composed of rollers and a pad, that is, the rollers support relatively light loads during rail deployment and vehicle traveling while a pad supports heavy loads during blanket replacement. Regarding the rotation mechanism, the double helical gear is adopted, because it has higher contact ratio than the normal spur gear and consequently can transfer higher force. The smaller double helical gear, 996 mm in diameter, can achieve 26% higher output torque, 123.5 kN·m, than that of the original spur gear of 1,460 mm in diameter, 98 kN·m. As a result, the manipulator becomes about 30% lighter, 8 tons, than the original weight, 11.2 tons. 2. Based on the compact design of the vehicle/manipulator, the

  9. On the evolution of the regulatory guidance for seismic qualification of electric and active mechanical equipment for nuclear power plants

    International Nuclear Information System (INIS)

    Ng, Ching Hang; Chen, Pei-Ying

    2009-01-01

    All electric and active mechanical equipment important to safety for nuclear power plants must be seismically qualified by testing, analysis, or combined analysis and testing. The general requirements for seismic qualification of electric and active mechanical equipment in nuclear power plants are delineated in Appendix S, 'Earthquake Engineering Criteria for Nuclear Power Plants,' to Title 10, Part 50, 'Domestic Licensing of Production and Utilization Facilities,' of the Code of Federal Regulations (10 CFR Part 50), item 52.47(20) of 10 CFR 52.47, 'Contents of Applications; Technical Information,' and Appendix A, 'Seismic and Geologic Siting Criteria for Nuclear Power Plants,' to 10 CFR Part 100, 'Reactor Site Criteria.' The United States Nuclear Regulatory Commission (NRC) issued Revision 2 of Regulatory Guide (RG) 1.100, 'Seismic Qualification of Electric and Mechanical for Nuclear Power Plants' in 1988, which endorsed, with restrictions, exceptions, and clarifications, Institute of Electrical and Electronics Engineers (IEEE) Standard 344-1987 'IEEE Recommended Practice for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Stations,' for use in seismic qualification of both electric and mechanical equipment. In 2008, the staff at the NRC drafted Revision 3 of RG 1.100 to endorse, with restrictions, exceptions, and clarifications, the IEEE Std 344-2004 and the American Society of Mechanical Engineers (ASME) QME-1-2007 'Qualification of Active Mechanical Equipment Used in Nuclear Power Plants.' IEEE Std 344-2004 was an update of Std 344-1987 and ASME QME-1-2007 was an update of QME-1-2002. The major changes in IEEE Std 344-2004 and ASME QME-1-2007 include the update and expansion of criteria and procedures describing the use of experience data as a method for seismic qualification of Class 1E electric equipment (including I and C components) as well as active mechanical equipment. In this paper, the staff will compare the draft Revision 3 to

  10. The Architecture Design of Detection and Calibration System for High-voltage Electrical Equipment

    Science.gov (United States)

    Ma, Y.; Lin, Y.; Yang, Y.; Gu, Ch; Yang, F.; Zou, L. D.

    2018-01-01

    With the construction of Material Quality Inspection Center of Shandong electric power company, Electric Power Research Institute takes on more jobs on quality analysis and laboratory calibration for high-voltage electrical equipment, and informationization construction becomes urgent. In the paper we design a consolidated system, which implements the electronic management and online automation process for material sampling, test apparatus detection and field test. In the three jobs we use QR code scanning, online Word editing and electronic signature. These techniques simplify the complex process of warehouse management and testing report transferring, and largely reduce the manual procedure. The construction of the standardized detection information platform realizes the integrated management of high-voltage electrical equipment from their networking, running to periodic detection. According to system operation evaluation, the speed of transferring report is doubled, and querying data is also easier and faster.

  11. Height, fun and safety in the design of children's playground equipment.

    Science.gov (United States)

    Wakes, Sarah; Beukes, Amanda

    2012-01-01

    The study reported in this paper adopted a holistic design approach to investigate issues associated with height related playground injuries from a users' perspective. The main objective was to gain an understanding of the relationship between height and fun so as to establish practical guidelines for addressing the causes of height related injuries whilst maintaining the attributes of playground equipment that children find fun and challenging. Results show that, on the one hand, the risk of injury increases when height is coupled with the use of upper body strength and, on the other hand, that coordination is a greater source of fun and challenge than height for children. Accordingly, it is suggested that the level of risk of injury attached to children's playground equipment can be reduced when the use of lower body strength and coordination are combined with lower free fall heights.

  12. Repository Planning, Design, and Engineering: Part II-Equipment and Costing.

    Science.gov (United States)

    Baird, Phillip M; Gunter, Elaine W

    2016-08-01

    Part II of this article discusses and provides guidance on the equipment and systems necessary to operate a repository. The various types of storage equipment and monitoring and support systems are presented in detail. While the material focuses on the large repository, the requirements for a small-scale startup are also presented. Cost estimates and a cost model for establishing a repository are presented. The cost model presents an expected range of acquisition costs for the large capital items in developing a repository. A range of 5,000-7,000 ft(2) constructed has been assumed, with 50 frozen storage units, to reflect a successful operation with growth potential. No design or engineering costs, permit or regulatory costs, or smaller items such as the computers, software, furniture, phones, and barcode readers required for operations have been included.

  13. Effect of Designated Pollutants on Plants

    Science.gov (United States)

    1977-11-01

    6 Materials and Methods ...... ................... .......... ..... 7 Exposure equipment...levels, and other studies on the interaction of pollutants and biological systems. 6 MATERIALS AND METHODS EXPOSURE EQUIPMENT Lexan chamber A rectilinear...citrus, and petunia are likely to be found com- mercially in the Lompoc area. A source of Bishop pine seeds enabled us to begin experiments with

  14. Guideline on dependability management for the power industry: detailed description of international power plant equipment dependability indicators

    International Nuclear Information System (INIS)

    Procaccia, H.; Silberberg, S.

    1997-01-01

    Dependability Management involves the management of reliability, availability maintainability and maintenance support, and in the power industry is necessary to ensure that plant meets the Reliability, Availability and Maintainability (RAM) targets set by the Utilities. In 1993, a joint Standard on Dependability Programme Management - Part 1: Dependability Programme Management), ISO 9000-': 1993 (Quality Management and Quality Assurance Standards - Part 4: Guide to Dependability Programme Management). UNIPEDE established a group of experts (Nulethermaint) to produce guidelines on its implementation specifically for use in the power industry. The present document comprises Part 2 OF THE UNIPEDE plant performance indicators and can be applied to both nuclear and fossil plant. There are five different equipment dependability indicators, all relating to equipment maintenance activities and the impact that these activities have on the loss of both system function and unit capability. Per year, each of the indicators can be applied separately to both preventive maintenance and corrective maintenance, giving rise to as many as ten indicator values for each item of equipment. Used in this way, the indicators provide a comprehensive picture of the maintenance strategy employed for key pieces of equipment, and its effectiveness. They are, therefore, a valuable managerial tool for improving maintenance activities at the unit level within a utility. This document provides guidance on the division of both nuclear and fossil power plant into their component parts and in each case the types of equipment having the most dominant effect on dependability are identified. These are the items which merit the greatest attention with regard to the equipment dependability indicators. (authors)

  15. Improved servicing equipment for steam generators

    International Nuclear Information System (INIS)

    Hedtke, James C.

    1998-01-01

    To help keep personnel exposure as low as reasonably achievable and reduce critical path outage time, most nuclear plants of PWR design in the USA are now using improved equipment to service their steam generators (SGs) during outages. Because of the success of this equipment in the USA, two Belgian plants and one English plant have purchased this equipment, and other nuclear plants in Europe are also considering procurement. The improved SG servicing equipment discussed in this paper discusses consists of nozzle dams, segmented multi-stud tensioner, primary manway cover handling tool set, shield door and fastener cleaner. This equipment is specifically designed for the individual plant application and can also be specified for replacement SG projects. All of the equipment can be used without modification of the existing SGs. (author)

  16. Quality of care and investment in property, plant, and equipment in hospitals.

    Science.gov (United States)

    Levitt, S W

    1994-02-01

    This study explores the relationship between quality of care and investment in property, plant, and equipment (PPE) in hospitals. Hospitals' investment in PPE was derived from audited financial statements for the fiscal years 1984-1989. Peer Review Organization (PRO) Generic Quality Screen (GQS) reviews and confirmed failures between April 1989 and September 1990 were obtained from the Massachusetts PRO. Weighted least squares regression models used PRO GQS confirmed failure rates as the dependent variable, and investment in PPE as the key explanatory variable. Investment in PPE was standardized, summed by the hospital over the six years, and divided by the hospital's average number of beds in that period. The number of PRO reviewed cases with one or more GQS confirmed failures was divided by the total number of cases reviewed to create confirmed failure rates. Investment in PPE in Massachusetts hospitals is correlated with GQS confirmed failure rates. A financial variable, investment in PPE, predicts certain dimensions of quality of care in hospitals.

  17. Electrical equipment performance under severe accident conditions (BWR/Mark 1 plant analysis): Summary report

    International Nuclear Information System (INIS)

    Bennett, P.R.; Kolaczkowski, A.M.; Medford, G.T.

    1986-09-01

    The purpose of the Performance Evaluation of Electrical Equipment during Severe Accident States Program is to determine the performance of electrical equipment, important to safety, under severe accident conditions. In FY85, a method was devised to identify important electrical equipment and the severe accident environments in which the equipment was likely to fail. This method was used to evaluate the equipment and severe accident environments for Browns Ferry Unit 1, a BWR/Mark I. Following this work, a test plan was written in FY86 to experimentally determine the performance of one selected component to two severe accident environments

  18. Integral design small nuclear power plant UNITHERM

    International Nuclear Information System (INIS)

    Adamovich, L. A.; Grechko, G. I.; Ulasevich, V. K.; Shishkin, V. A.

    1995-01-01

    The need to erect expensive energy transmission lines to these places demands to use independent local energy sources. Therefore, a reasonable alternative to the plants fired fossil fuel, mostly hydrocarbon fuel, may come from the nuclear power plants (NPP) of relatively small capacity which are nonattended, shipped to the site by large-assembled modules and completely withdrawable from the site during decommissioning. Application of NPPs for power and heat supply may prove to be cost-efficient and rather positive from social and ecological point of view. UNITHERM NPP belongs to such energy sources and may be used for heat and power supply. Heat can be provided both as hot water and superheated steam. The consumers are able to specify heat/energy supply ratio. NPP design provides for independent energy supply to the consumers and the possibility to disconnect each of them without disruption of operation of the others. Thermal hydraulic diagram of UNITHERM NPP provides for the use of three interconnected, process circuits. The consumers of thermal energy (turbogenerator unit and boilers of the central heating unit) are arranged in the last circuit

  19. The development of automated fuel dismantling equipment for a future head-end plant

    International Nuclear Information System (INIS)

    Haberlin, M.M.

    1987-01-01

    For future reprocessing plants, practicable methods for dismantling fuel elements are being examined at Springfields Nuclear Power Development Laboratories which would meet the requirements of a high throughput facility. This paper contains the initial results of an experimental programme undertaken to develop and evaluate an automated high speed single/multiple pin extraction unit. Concomitant parts of the programme include the design and evaluation of single and multi-pin extraction chucks. Dummy fuel elements, a 325 pin gridded LMFBR assembly and a 17 x 17 pin gridded LWR assembly were used to assess process efficacy

  20. Equipment Design and Cost Estimation for Small Modular Biomass Systems, Synthesis Gas Cleanup, and Oxygen Separation Equipment; Task 9: Mixed Alcohols From Syngas -- State of Technology

    Energy Technology Data Exchange (ETDEWEB)

    Nexant Inc.

    2006-05-01

    This deliverable is for Task 9, Mixed Alcohols from Syngas: State of Technology, as part of National Renewable Energy Laboratory (NREL) Award ACO-5-44027, ''Equipment Design and Cost Estimation for Small Modular Biomass Systems, Synthesis Gas Cleanup and Oxygen Separation Equipment''. Task 9 supplements the work previously done by NREL in the mixed alcohols section of the 2003 technical report Preliminary Screening--Technical and Economic Assessment of Synthesis Gas to Fuels and Chemicals with Emphasis on the Potential for Biomass-Derived Syngas.

  1. New technology for optimized I and C maintenance and management of ageing of critical equipment in nuclear power plants

    International Nuclear Information System (INIS)

    Hashemian, H.M.

    2000-01-01

    Advanced sensors and new testing and maintenance technologies have become available over the last ten years for nuclear power plants (NPPs) to replace outdated, obsolete, and troublesome instruments, provide for management of ageing of critical plant equipment, optimize maintenance activities, reduce maintenance costs and personnel radiation exposure, and at the same time, improve plant safety and availability. These new developments are reviewed in this TECDOC. The material covered here has been summarized from NUREG/CR-5501, a 1998 report written by H.M. Hashemian and his co-authors for the US Nuclear Regulatory Commission. (author)

  2. Reactor building design of nuclear power plant ATUCHA II, Argentina

    International Nuclear Information System (INIS)

    Rufino, R.E.; Hermann, E.R.; Richter, E.

    1984-01-01

    It is presented the civil engineering project carried out by the joint venture Hochtief - Techint-Bignoli (HTB) for the reactor building at the Atucha II power plant (PHWR of 745 MWe) in Buenos Aires. All the other civil projects at Atucha II are also being carried out by HTB. This building has the same general characteristics of the PWR plants developed by KWU in Germany, known for the spherical steel containment 56m in diameter. Nevertheless, it differs from those principally in the equipment lay-out and the remarkable foundation depth. From the basic engineering provided by ENACE, the joint venture has had to face the challenge of designing a tridimensional structure of large size. This has necessitated using simplified models which had to be superimposed, since the use of only one spatial mode would be highly inadequate, lacking the flexibility necessary to absorb the numerous modifications that this type of project undergoes during construction. In addition, this procedure has eliminated resorting to numerous and costly computer processings. (Author) [pt

  3. Quality assurance in the design of nuclear power plants

    International Nuclear Information System (INIS)

    1981-01-01

    This Safety Guide provides the requirements and recommendations related to the establishment and implementation of quality assurance for design of items for a nuclear power plant. The requirements of this Guide shall be applied to the extent necessary during all constituent activities of the nuclear power plant project, such as design, manufacture, construction, commissioning and operations. Its requirements and recommendations shall be implemented, as appropriate, by the responsible organization or by its designated representatives: by plant designers, architect-engineers or manufacturers, when involved in performing design activities related to items to be manufactured; by site constructors, when involved in field engineering activities; by plant operators and other organizations, when involved in design activities related to plant modifications or to selection of spare or replacement parts; and by design consultants and other technical organizations, when performing any engineering activity that affects the work of other design organizations during various stages of nuclear power plant projects

  4. Designing nuclear power plants for improved operation and maintenance

    International Nuclear Information System (INIS)

    1996-09-01

    The purpose of this publication is to compile demonstrated, experience based design guidelines for improving the operability and maintainability of nuclear power plants. The guidelines are for use principally in the design of new nuclear power plants, but should also be useful in upgrading existing designs. The guidelines derive from the experience of operating and maintaining existing nuclear power plants as well as from the design of recent plants. In particular these guidelines are based on and consistent with both the EPRI advanced Light Water Reactor Utility Requirements Document, Volume 1, and the European Utility Requirements for LWR Nuclear Power Plants. 6 refs, 1 fig

  5. Designing nuclear power plants for improved operation and maintenance

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-09-01

    The purpose of this publication is to compile demonstrated, experience based design guidelines for improving the operability and maintainability of nuclear power plants. The guidelines are for use principally in the design of new nuclear power plants, but should also be useful in upgrading existing designs. The guidelines derive from the experience of operating and maintaining existing nuclear power plants as well as from the design of recent plants. In particular these guidelines are based on and consistent with both the EPRI advanced Light Water Reactor Utility Requirements Document, Volume 1, and the European Utility Requirements for LWR Nuclear Power Plants. 6 refs, 1 fig.

  6. Design and Construction of Equipment for Applying the Geophysical Prospecting Method Electric Tomography

    Directory of Open Access Journals (Sweden)

    Fabio Héctor Giraldo Sánchez

    2013-06-01

    Full Text Available Outlines the procedure for the design and construction of electric equipment for geophysical prospecting through electrical tomography method. The team is of average power, ensuring exploration depths quite suitable for applications and commercial and geotechnical studies. The device is essentially a DC voltage source of 500 volts that is able to provide a maximum current of 1 amp. It also contains a small charge current source of electrical currents counteract naturally found in the subsoil and are manifested as a difference in the surface potential. A general explanation of the geophysical method in question, helps to understand the basic principles of operation of the equipment and functions to be fulfilled. After building the team, we conducted a field data acquisition, in area near the town of Gachancipa Cundinamarca. The data from this equipment are processed with specialized software. The images obtained with the software presents the distributions of subsurface resistivity can be associated with the possible structures and geology of the study area.

  7. Design and Evaluation of a Secure Virtual Power Plant.

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, Jay Tillay [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-09-01

    For three years, Sandia National Laboratories, Georgia Institute of Technology, and University of Illinois at Urbana-Champaign investigated a smart grid vision in which renewable-centric Virtual Power Plants (VPPs) provided ancillary services with interoperable distributed energy resources (DER). This team researched, designed, built, and evaluated real-time VPP designs incorporating DER forecasting, stochastic optimization, controls, and cyber security to construct a system capable of delivering reliable ancillary services, which have been traditionally provided by large power plants or other dedicated equipment. VPPs have become possible through an evolving landscape of state and national interconnection standards, which now require DER to include grid-support functionality and communications capabilities. This makes it possible for third party aggregators to provide a range of critical grid services such as voltage regulation, frequency regulation, and contingency reserves to grid operators. This paradigm (a) enables renewable energy, demand response, and energy storage to participate in grid operations and provide grid services, (b) improves grid reliability by providing additional operating reserves for utilities, independent system operators (ISOs), and regional transmission organization (RTOs), and (c) removes renewable energy high-penetration barriers by providing services with photovoltaics and wind resources that traditionally were the jobs of thermal generators. Therefore, it is believed VPP deployment will have far-reaching positive consequences for grid operations and may provide a robust pathway to high penetrations of renewables on US power systems. In this report, we design VPPs to provide a range of grid-support services and demonstrate one VPP which simultaneously provides bulk-system energy and ancillary reserves.

  8. Aseismic design of turbine houses of nuclear power plants

    International Nuclear Information System (INIS)

    Danisch, R.; Labes, M.

    1975-01-01

    The turbine house does not belong to the safety-related parts of equipment of a nuclear power plant. A special protection against earthquakes is not demanded by the authorities as long as it is proven that safety-related parts of equipment will not be restricted in their function by a collaps of the turbine house. The degree of an aseismic design is largely up to the customer, who has to weigh the risk of costs and availability against the additional costs, that are necessary for the earthquake calculation and for constructive hardening. In comparison to the high-tuned turbine foundations as they are in use in the USA today, low-tuned turbine foundations as a result of helical-spring-support, which are constructed by the KWU exclusively, pose special problems with the aseismic design. This is discussed in the present report. The spring-supported mass constitutes about a quarter of the building-mass. For mechanical reasons the spring elements are chosen in such a way, that the turbine foundation has a natural frequency of approximately 3 Hz. Thus it remains within the same frequency range as the turbine house and within that very range which is particularly amplificated by an earthquake. It is therefore likely that resonance effects as well as oscillation annulment effects may occur. The standardized calculation methods for conventional buildings without safety function such as DIN 4149 (Germany) or SIA 162 (Switzerland) do not cover the oscillation conduct of such a complicate structure. One receives informations about possible relative displacements between the building and the turbine foundation (hammering-effect) and about the stresses on the turbine and other components only by dynamic calculation methods such as the time-history or the response-spectrum method

  9. Development of processing procedure preparing for digital computer controlled equipment on modular design base

    International Nuclear Information System (INIS)

    Starosel'tsev, O.P.; Khrundin, V.I.

    1982-01-01

    In order to reduce labour consumption of technological preparation of production for digital computer controlled machines during the treatment of steam turbines articles created is a system of modular design of technological processes and controlling programs. A set of typical modulas-transitions, being a number of surfaces of an articles treated with one cutting tool in optimum sequence, and a library of cutting tools are the base of the system. Introduction of such a system sharply enhaneces the efficiency of the equipment utilization [ru

  10. Radiation effects on electronic equipment: a designers'/users' guide for the nuclear power industry

    International Nuclear Information System (INIS)

    Sharp, R.E.; Garlick, D.R.

    1994-01-01

    The Designers'/Users' Guide to the effects of radiation on electronics is published by the Radiation Testing Service of AEA Technology. The aim of the Guide is to document the available information that we have generated and collected over some ten years whilst operating as a radiation effects and design consultancy to the nuclear power industry. We hope that this will enable workers within the industry better to understand the likely effects of radiation on the system or plant being designed and so minimise the problems that can arise. (Author)

  11. How to optimize hydrogen plant designs

    Energy Technology Data Exchange (ETDEWEB)

    van Weenen, W F; Tielrooy, J

    1983-01-01

    In a typical hydrogen plant of the type which will be discussed, methane or higher hydrocarbons are reformed with steam in a steam hydrocarbon reformer operating at a pressure of 250 to 400 psig, a temperature of 1500 to 1600/sup 0/F, and with a ratio of steam to carbon in the feed of about 3.0. Following the reformer and cooling, there is a single stage of high temperature carbon monoxide shift conversion. Optionally, after further cooling, this may be followed by a second stage of carbon monoxide shift conversion operating at a lower temperature to obtain a more favourable equilibrium; this is called low temperature shift conversion. After cooling to ambient temperature, and separation of the condensate, the gas is passed through a Pressure Swing Adsorption (PSA)l unit which removes all the impurities along with a small amount of hydrogen. The waste gas from the PSA unit containing all the impurities is used as fuel to the reformer. Heat is recovered from the reformer flue gas, reformer product, high temperature shift converter product and low temperature shift converter product. This paper discusses some of the process variables and design variables which must be considered in arriving at an optimized design. Seven different flow schemes are discussed in the light of the objectives they are designed for. The seven schemes and their objectives are: Flow Scheme 1 - lowest first cost; moderate efficiency, Flow Scheme 2 - high efficiency, low cost; Flow Scheme 3 - low feed plus fuel, moderately high efficiency; Flow Scheme 4 - lowest feed plus fuel; Flow Scheme 5 - lowest feed, low fuel; Flow Scheme 6 -lowest feed, highest efficiency; and Flow Scheme 7 - lowest feed plus fuel, export electric power instead of export electric power instead of export steam. 15 figures, 1 table.

  12. Treatment of decontamination liquid waste of nuclear power plant components by heterogeneous photo catalysis with a continuous recirculation equipment

    International Nuclear Information System (INIS)

    Litter, Marta I.; La Gamma, Ana M.; Chocron, Mauricio; Blesa, Miguel A.; Repetto, Pablo

    1999-01-01

    It has been designed a bench scale, recirculation device, for testing the degradation of solutions of ethylendiamine tetraacetic acid (EDTA) by heterogeneous photo catalysis under irradiation with UV and titanium dioxide (TiO 2 ). Solutions of EDTA have been employed at concentrations and pH values similar to those used when a decontamination of nuclear power plant equipment is carried out. The circuit is composed of a photo reactor, a heat exchanger, a reservoir tank and a peristaltic pump. In the present paper, the results of the experiments of photo catalytic degradation of aqueous suspensions of TiO 2 (Degussa P-25) 1 g/L with EDTA (10 g/L) at pH 3.7 and 25 degree C and two irradiation wavelengths (366 and 254 nm) have been presented. At 366 nm the full degradation of EDTA has occurred in 10 hours. The 95% degradation of total organic carbon (TOC) has been achieved after 39 hours of irradiation. The irradiation at 254 nm in the same conditions has been much less effective (EDTA and TOC reduction of approximately 1%), due to a screening effect produced by the semiconductor. (author)

  13. Cycle chemistry monitoring system as means of improving the reliability of the equipment at the power plants

    Science.gov (United States)

    Yegoshina, O. V.; Voronov, V. N.; Yarovoy, V. O.; Bolshakova, N. A.

    2017-11-01

    There are many problems in domestic energy at the present that require urgent solutions in the near future. One of these problems - the aging of the main and auxiliary equipment. Wear of equipment is the cause of decrease reliability and efficiency of power plants. Reliability of the equipment are associated with the introduction of cycle chemistry monitoring system. The most damageable equipment’s are boilers (52.2 %), turbines (12.6 %) and heating systems (12.3 %) according to the review of failure rate on the power plants. The most part of the damageability of the boiler is heated surfaces (73.2 %). According to the Russian technical requirements, the monitoring systems are responsible to reduce damageability the boiler heating surfaces and to increase the reliability of the equipment. All power units capacity of over 50 MW are equipped with cycle chemistry monitoring systems in order to maintain water chemistry within operating limits. The main idea of cycle chemistry monitoring systems is to improve water chemistry at power plants. According to the guidelines, cycle chemistry monitoring systems of a single unit depends on its type (drum or once-through boiler) and consists of: 20…50 parameters of on-line chemical analyzers; 20…30 «grab» sample analyses (daily) and about 15…20 on-line monitored operating parameters. The operator of modern power plant uses with many data at different points of steam/water cycle. Operators do not can estimate quality of the cycle chemistry due to the large volume of daily and every shift information and dispersion of data, lack of systematization. In this paper, an algorithm for calculating the quality index developed for improving control the water chemistry of the condensate, feed water and prevent scaling and corrosion in the steam/water cycle.

  14. Calculation and design of natural gas preheater equipments. Berechnung und Auslegung von Erdgas-Vorwaermeanlagen

    Energy Technology Data Exchange (ETDEWEB)

    Fasold, H G [Ruhrgas AG, Essen (Germany); Wahle, H N [Ruhrgas AG, Essen (Germany)

    1994-04-01

    A greatly simplified model of a regulating station - consisting of the station components ''preheater'' and ''control unit'' - is used for the calculation and design of natural gas preheating plants. It is hereby possible to calculate the Joule-Thomson effect which occurs on the expansion of natural gas in the controller, the resulting drop in temperature and the thermal output required to compensate this which is to be supplied to the gas flow by the preheating plant. The calculation method and procedure are explained using a programming flowchart. The computational model presented was converted into a personal computer program, whose functioning is elucidated using a numerical example. (orig.)

  15. State of technology assessment for life extension of electrical and I and C equipment in nuclear power plants

    International Nuclear Information System (INIS)

    DuCharme, A.R.; Boger, R.M.; Meyer, L.C.; Beament, P.R.

    1988-01-01

    As part of the IEEE Working Group 3.4 on Nuclear Plant Life Extension, an assessment is made of the current state of technology for the life extension of certain classes of electrical and IandC equipment. The classes investigated include motors, cables, emergency diesel generators, penetrations, inverters/chargers, switchgear, and reactor protection systems. The work is focussed on assessment of current or recently completed RandD efforts to resolve issues affecting life extension of the equipment. Aspects discussed include the degree of resolution of these issues, potentially affected standards, and technical aspects requiring further research. 15 refs., 2 tabs

  16. State of technology assessment for life extension of electrical and I and C equipment in nuclear power plants

    International Nuclear Information System (INIS)

    Du Charme, A.R.; Boger, R.M.; Meyer, L.C.; Beament, P.R.

    1988-01-01

    As part of the IEEE Working Group 3.4 on Nuclear Plant Life Extension, an assessment is made of the current state of technology for the life extension of certain classes of electrical and I and C equipment. The classes investigated include motors, cables, emergency diesel generators, penetrations, inverters/charges, switchgear, and reactor protection systems. The work is focussed on assessment of current or recently completed R and D efforts to resolve issues affecting life extension of the equipment. Aspects discussed include the degree of resolution of these issues, potentially affected standards, and technical aspects requiring further research

  17. Revision of nuclear power plants safety systems' routine testing assigned periodicity during the design extension period

    International Nuclear Information System (INIS)

    Skalozubov, V.I.; Kozlov, Yi.L.; Chulkyin, O.O.

    2017-01-01

    When nuclear power plants safety systems thermal equipment operation extending, a necessary requirement shall rely on revising the scheduled equipment tests frequency to optimize those tests schedule taking into account the equipment remained lifespan. On the one hand, there exists a need for tests frequency increase to detect ''hidden'' failures, and on the another, frequent tests cause a premature wear of the equipment. Proposed is an original method for optimizing the frequency of NPPs safety systems thermal engineering equipment testing. Essential in the proposed method is the optimization criterion chosen: index of security system failure probability non-exceedance during the beyond-design operating period as referred to the failure probability expected considering the equipment residual resource during the design operating period. The developed method implementation when applied to NPPs safety systems operated beyond the design service life at nuclear power plants with WWER-1000 series reactors, allowed to establish that the optimal tests frequency makes half the designed one when the equipment service life is extended by five years and three times less that the designed frequency when subject lifespan extended by 10 years.

  18. Thermal Integration of CO{sub 2} Compression Processes with Coal-Fired Power Plants Equipped with Carbon Capture

    Energy Technology Data Exchange (ETDEWEB)

    Edward Levy

    2012-06-29

    Coal-fired power plants, equipped either with oxycombustion or post-combustion CO{sub 2} capture, will require a CO{sub 2} compression system to increase the pressure of the CO{sub 2} to the level needed for sequestration. Most analyses show that CO{sub 2} compression will have a significant effect on parasitic load, will be a major capital cost, and will contribute significantly to reduced unit efficiency. This project used first principle engineering analyses and computer simulations to determine the effects of utilizing compressor waste heat to improve power plant efficiency and increase net power output of coal-fired power plants with carbon capture. This was done for units with post combustion solvent-based CO{sub 2} capture systems and for oxyfired power plants, firing bituminous, PRB and lignite coals. The thermal integration opportunities analyzed for oxycombustion capture are use of compressor waste heat to reheat recirculated flue gas, preheat boiler feedwater and predry high-moisture coals prior to pulverizing the coal. Among the thermal integration opportunities analyzed for post combustion capture systems are use of compressor waste heat and heat recovered from the stripper condenser to regenerate post-combustion CO{sub 2} capture solvent, preheat boiler feedwater and predry high-moisture coals. The overall conclusion from the oxyfuel simulations is that thermal integration of compressor heat has the potential to improve net unit heat rate by up to 8.4 percent, but the actual magnitude of the improvement will depend on the type of heat sink used and to a lesser extent, compressor design and coal rank. The simulations of a unit with a MEA post combustion capture system showed that thermal integration of either compressor heat or stripper condenser heat to preheat boiler feedwater would result in heat rate improvements from 1.20 percent to 4.19 percent. The MEA capture simulations further showed that partial drying of low rank coals, done in combination

  19. Osiris and SOMBRERO inertial confinement fusion power plant designs

    International Nuclear Information System (INIS)

    Meier, W.R.; Bieri, R.L.; Monsler, M.J.

    1992-03-01

    Conceptual designs and assessments have been completed for two inertial fusion energy (IFE) electric power plants. The detailed designs and results of the assessment studies are presented in this report. Osiris is a heavy-ion-beam (HIB) driven power plant and SOMBRERO is a Krypton-Fluoride (KrF) laser-driven power plant. Both plants are sized for a net electric power of 1000 MWe

  20. 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