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

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

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

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.

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

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

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.

Behavior of critical structures and equipment at 100-N plant during a postulated tornado

International Nuclear Information System (INIS)

Davis, H.S.

1978-01-01

The objective of this report is to document the results of an analytical study for determining the effects of tornadic wind pressures and missiles on critical structures and equipment at 100-N reactor. These particular structures and equipment are essential for maintaining the reactor in a safe, shutdown condition. The analyses show that structures, systems and components required for safe reactor shutdown and operations of the Emergency Core Cooling System (ECCS) would not be damaged by a 175 mph tornado, nor associated missiles, to the extent that the ECCS would not be able to function adequately

Technology development on the assessment of structural integrity of nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Seo, Jeong Moon; Choun, Y. S.; Choi, I. K. and others

1999-04-01

Nuclear power plants in Korea show drop off in their performance and safety margin as the age of plants increase. The reevaluation of Kori-1 Unit on its performance and safety for life extension is expected in the near future. However, technologies and information related are insufficient to quantitatively estimate them. The final goal of this study is to develop the basic testing and evaluation techniques related with structural integrity of important nuclear equipment and structures. A part of the study includes development of equipment qualification technique. To ensure the structural integrity of structures, systems, and equipment in nuclear power plants, the following 5 research tasks were performed in the first year. - Analysis of dynamic characteristics of reactor internals - Analysis of engineering characteristics of instrumental earthquakes recorded in Korea - Analysis of ultimate pressure capacity and failure mode of containments building - Development of advanced NDE techniques using ultrasonic resonance scattering - Development of equipment qualification technique against vibration aging. These technologies developed in this study can be used to ensure the structural safety of operational nuclear power plants, and for the long-term life management. (author)

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...

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.)

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

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)

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...

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

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.)

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

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

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

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)

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)

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...

Localization of equipment for digital plant protection system

Energy Technology Data Exchange (ETDEWEB)

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

2000-10-01

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

Localization of equipment for digital plant protection system

International Nuclear Information System (INIS)

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

2000-10-01

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

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

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.

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

Effects of supporting structures on dynamic response of nuclear power plant equipment and piping systems

International Nuclear Information System (INIS)

Stoykovich, M.

1982-01-01

This paper presents the evaluation of the effects of supporting structures in dynamic analysis of equipment or piping systems, which involves formulations for determining reduced stiffness and mass matrices associated with the number of degrees of freedom corresponding to the support nodal points of a finite element model. Also, evaluation of a composite damping matrix associated with different damping properties of supporting structures, equipment, and piping systems is considered. Determination of spring constants, effective masses and mass moments of inertia, and damping values as fractions of critical damping on the basis of the theory of rigid bases on the surfaces of an elastic halfspace is demonstrated

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

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

Developing Predictive Maintenance Expertise to Improve Plant Equipment Reliability

International Nuclear Information System (INIS)

Wurzbach, Richard N.

2002-01-01

key technologies such as vibration analysis, infrared thermography, and oil analysis not as singular entities, but as a toolbox resource from which to address overall equipment and plant reliability in a structured program and decision environment. (authors)

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

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.)

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...

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.)

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)

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)

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

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)

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)

Generation of floor response spectra for a model structure of nuclear power plant

International Nuclear Information System (INIS)

Vaidyanathan, C.V.; Kamatchi, P.; Ravichandran, R.; Lakshmanan, N.

2003-01-01

The importance of Nuclear power plants and the consequences of a nuclear accident require that the nuclear structures be designed for the most severe environmental conditions. Earthquakes constitutes major design consideration for the system, structures and equipment of a nuclear power plant. The design of structures on ground is based on the ground response spectra. Many important parts 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 rather than by the general ground motion to which the structure is supported. Hence the seismic response of equipment is generally based on the response spectrum of the floor on which it is mounted. In this paper such floor response spectra have been generated at different nodes of a chosen model structure of a nuclear power plant. In the present study a detailed nonlinear time history analysis has been carried out on the mathematical model of the chosen Nuclear Power Plant model structure with the spectrum compatible time history. The acceleration response results of the time history analysis has been used in the spectral analysis and the response spectra are generated. Further peak broadening has been done to account for uncertainties in the material properties and soil characteristics. (author)

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

Seismic response analysis of Wolsung NPP structure and equipment subjected to scenario earthquakes

Energy Technology Data Exchange (ETDEWEB)

Choi, In Kil; Ahn, Seong Moon; Choun, Young Sun; Seo, Jeong Moon

2005-03-15

The standard response spectrum proposed by US NRC has been used as a design earthquake for the design of Korean nuclear power plant structures. However, it does not reflect the characteristic of seismological and geological of Korea. In this study, the seismic response analysis of Wolsung NPP structure and equipment were performed. Three types of input motions, artificial time histories that envelop the US NRC Regulatory Guide 1.60 spectrum and the probability based scenario earthquake spectra developed for the Korean NPP site and a typical near-fault earthquake recorded at thirty sites, were used as input motions. The acceleration, displacement and shear force responses of Wolsung containment structure due to the design earthquake were larger than those due to the other input earthquakes. But, considering displacement response increases abruptly as Wolsung NPP structure does nonlinear behavior, the reassessment of the seismic safety margin based on the displacement is necessary if the structure does nonlinear behavior; although it has adequate the seismic safety margin within elastic limit. Among the main safety-related devices, electrical cabinet and pump showed the large responses on the scenario earthquake which has the high frequency characteristic. This has great effects of the seismic capacity of the main devices installed inside of the building. This means that the design earthquake is not so conservative for the safety of the safety related nuclear power plant equipments.

Seismic response analysis of Wolsung NPP structure and equipment subjected to scenario earthquakes

International Nuclear Information System (INIS)

Choi, In Kil; Ahn, Seong Moon; Choun, Young Sun; Seo, Jeong Moon

2005-03-01

The standard response spectrum proposed by US NRC has been used as a design earthquake for the design of Korean nuclear power plant structures. However, it does not reflect the characteristic of seismological and geological of Korea. In this study, the seismic response analysis of Wolsung NPP structure and equipment were performed. Three types of input motions, artificial time histories that envelop the US NRC Regulatory Guide 1.60 spectrum and the probability based scenario earthquake spectra developed for the Korean NPP site and a typical near-fault earthquake recorded at thirty sites, were used as input motions. The acceleration, displacement and shear force responses of Wolsung containment structure due to the design earthquake were larger than those due to the other input earthquakes. But, considering displacement response increases abruptly as Wolsung NPP structure does nonlinear behavior, the reassessment of the seismic safety margin based on the displacement is necessary if the structure does nonlinear behavior; although it has adequate the seismic safety margin within elastic limit. Among the main safety-related devices, electrical cabinet and pump showed the large responses on the scenario earthquake which has the high frequency characteristic. This has great effects of the seismic capacity of the main devices installed inside of the building. This means that the design earthquake is not so conservative for the safety of the safety related nuclear power plant equipments

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

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.

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...

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

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)

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

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.

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)

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

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)

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.

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

Lifetime of Mechanical Equipment

Energy Technology Data Exchange (ETDEWEB)

Leland, K.

1999-07-01

The gas plant at Kaarstoe was built as part of the Statpipe gas transport system and went on stream in 1985. In 1993 another line was routed from the Sleipner field to carry condensate, and the plant was extended accordingly. Today heavy additional supply- and export lines are under construction, and the plant is extended more than ever. The main role of the factory is to separate the raw gas into commercial products and to pump or ship it to the markets. The site covers a large number of well-known mechanical equipment. This presentation deals with piping, mechanical and structural disciplines. The lifetime of mechanical equipment is often difficult to predict as it depends on many factors, and the subject is complex. Mechanical equipment has been kept in-house, which provides detailed knowledge of the stages from a new to a 14 years old plant. The production regularity has always been very high, as required. The standard of the equipment is well kept, support systems are efficient, and human improvisation is extremely valuable.

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)

Earthquake protection of essential civil and industrial equipments

International Nuclear Information System (INIS)

Bourrier, P.; Le Breton, F.; Thevenot, A.

1986-01-01

This document presents the principal reflexions concerning seismic engineering applications for equipment and the difference of the non-employment towards these structures. The notion of essential equipment is then pointed out as well as the main particularities of equipment considered as structures. Finally, this document illustrates a few pathological examples encountered after an earthquake, and presents some equipments of a nuclear power plant which to resist an increased safety seism [fr

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.

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

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.

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.

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.

Complex modal properties of coupled moderately light equipment-structure systems

International Nuclear Information System (INIS)

Gupta, A.K.; Jaw Jingwen

1986-01-01

A new improved perturbation method for evaluating complex modal properties of coupled equipment-structure systems is presented. The method is applicable even when the equipment is not very light, and when the secondary system (equipment) introduces static constraint on the primary system (structure). The new method is applied to nine 8DOF coupled multiply connected equipment-structure systems. It is shown that the new method yields results which are in excellent agreement with the corresponding exact results. (orig.)

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.)

(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...

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.)

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

A review of procedures available to seismically requalify operating nuclear plant structures, equipment and distribution systems

International Nuclear Information System (INIS)

Stevenson, J.D.

1985-01-01

It is well known that the loads and procedures used to seismically qualify nuclear power plant structures and components have changed dramatically during the past 15 to 20 years. In this paper, the various methods available to seismically qualify or requalify structures and components in operating nuclear power plants are identified and the advantages and disadvantages of each briefly summarized. (orig.)

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

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...

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)

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

Construction of a model of the process of accumulation of radionuclides of corrosion products on the equipment in nuclear power plants with boiling-water reactors

International Nuclear Information System (INIS)

Tevlin, S.A.

1985-01-01

This paper addresses the problem of corrosion of the structural materials of the reactor loop. This problem can be solved by constructing physical models of the process of accumulation of radionuclides on the equipment at nuclear power plants and by constructing the analytical apparatus for describing them. These models are presented here, and allow the analyzing of the effect of separate states and thermophysical factors, determination of the basic factors, and the ability to foresee in timely fashion the water state and structural measures required to lower the rate of growth and to decrease the amount of radionuclides deposited on the equipment in the nuclear power plant

Structural mechanics in nuclear power plant

International Nuclear Information System (INIS)

Han Liangbi

1998-01-01

The main research works in structural mechanics in reactor technology are emphatically introduced. It is completed by structural mechanics engineers at Shanghai Nuclear Research and Design Institute associated with the design and construction problems for Qinshan NPP Unit 1 and Pakistani CHASNUPP. About structural mechanics problem for the containment, the rock and soft soil two different bases are considered. For the later the interaction between soil and structure is carefully studied. About the structural mechanics problem for the equipment and pipings, the three dimensional stress and fracture analyses are studied. For the structural dynamics problem, including flow induced vibration, the response analyses under earthquake and loss coolant accident loadings are studied. For pipings, the leak before break technique has been emphatically introduced. A lot of mathematical models, the used computer codes, analytical calculations and experimental results are also introduced. This is a comprehensive description about structural mechanics problem in pressurized water reactor nuclear power plant

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.

Structuring free form verbal descriptions in equipment failure reports

International Nuclear Information System (INIS)

Huzdovich, J.

1983-01-01

Information is encoded for convenience in computer sort/search routines used to manage a large number of records. The codes in use for equipment failure reports are limited due to practical considerations, and this limitation forces the reporter to leave out information to satisfy the coding requirements. The free form verbal descriptions, as found in the Generating Availability Data System (GADS) and the Nuclear Plant Reliability Data System (NPRDS), allow for reporting of this non-codable information. A systematic approach to constructing the verbal description based on rules of grammar, especially syntax, results in a structured narrative suitable for computer data management schemes. In addition, the reporter has a full range of descriptive terminology and does not have to select subjectively from a predetermined, limited vocabulary to describe the event. This paper introduces a concept that places in perspective the integration of structured, formal reporting and free form verbal description. A second benefit of this structured narrative is the systematic development of failure mode/failure cause relationships in the event

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

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...

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)

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

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.

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.

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.

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

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

Constructing a User Interface for Cellular Phones Using Equipment and its Relations

Directory of Open Access Journals (Sweden)

Misayo Kitamura

2003-04-01

Full Text Available In a domain of SCADA (Supervisory Control And Data Acquisition systems, it is necessary to obtain information about plants such as water plants in remote places using a cellular phone in order to ascertain plant status in case of emergency.T o utilize the small screen of a cellular phone and to eliminate the engineering cost of creating de.nition data to show plant status, a method of constructing user interface using equipment in the plant and its relations is proposed. In this method, some equipment is selected from all supervised equipment using the relations between the equipment, and then the content to be displayed is generated dynamically using the selected equipment. The equipment in plants is organized as a graph structure, which involves the equipment and the relations between the equipment.T he relations adopted in this method are both the physical connections between the equipment and the conceptual relationships.The result of the selection depends on the relations and their parameter values called the context dependent weight, which changes dynamically by viewpoints.

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

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

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

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...

Structural load inventory database for the Kansas City Plant

International Nuclear Information System (INIS)

Hashimoto, P.S.; Johnson, M.W.; Nakaki, D.K.; Wilson, J.J.; Lynch, D.T.; Drury, M.A.

1993-01-01

A structural load inventory database (LID) has been developed to support configuration management at the DOE Kansas City Plant (KCP). The objective of the LID is to record loads supported by the plant structures and to provide rapid assessments of the impact of future facility modifications on structural adequacy. Development of the LID was initiated for the KCP's Main Manufacturing Building. Field walkdowns were performed to determine all significant loads supported by the structure, including the weight of piping, service equipment, etc. These loads were compiled in the LID. Structural analyses for natural phenomena hazards were performed in accordance with UCRL-15910. Software to calculate demands on the structural members due to gravity loads, total demands including both gravity and seismic loads, and structural member demand-to-capacity ratios were also developed and integrated into the LID. Operation of the LID is menu-driven. The LID user has options to review and print existing loads and corresponding demand-to-capacity ratios, and to update the supported loads and demand-to-capacity ratios for any future facility modifications

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)

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)

Full-scale dynamic structural testing of Paks nuclear power plant

International Nuclear Information System (INIS)

Da Rin, E.M.; Muzzi, F.P.

1995-01-01

Within the framework of the IAEA coordinated 'Benchmark Study for the seismic analysis and testing of WWER-type NPPs', in-situ dynamic structural testing activities have been performed at the Paks Nuclear Power Plant in Hungary. The specific objective of the investigation was to obtain experimental data on the actual dynamic structural behaviour of the plant's major constructions and equipment under normal operating conditions, for enabling a valid seismic safety review to be made. This paper gives a synthetic description of the conducted experiments and presents some results, regarding in particular the free-field excitations produced during the earthquake-simulation experiments and an experiment of the dynamic soil-structure interaction global effects at the base of the reactor containment structure. Moreover, a method which can be used for inferring dynamic structural characteristics from the recorded time-histories is briefly described and a simple illustrative example given. (author)

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...

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)

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)

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…

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...

Equipment response spectra for base-isolated shear beam structures

International Nuclear Information System (INIS)

Ahmadi, G.; Su, L.

1992-01-01

Equipment response spectra in base-isolated structure under seismic ground excitations are studied. The equipment is treated as a single-degree-of-freedom system attached to a nonuniform elastic beam structural model. Several leading base isolation systems, including the laminated rubber bearing, the resilient-friction base isolator with and without a sliding upper plate, and the EDF system are considered. Deflection and acceleration response spectra for the equipment and the shear beam structure subject to a sinusoidal and the accelerogram of the N00W component of El Centro 1940 earthquake are evaluated. Primary-secondary interaction effects are included in the analysis. Several numerical parametric studies are carried out and the effectiveness of different base isolation systems in protecting the nonstructural components is studied. It is shown that use of properly designed base isolation systems provides considerable protection for secondary systems, as well as, the structure against severe seismic loadings. (orig.)

Modernization of the automation control system of technological processes at the preparation plant in the conditions of technical re-equipment

Science.gov (United States)

Lyakhovets, M. V.; Wenger, K. G.; Myshlyaev, L. P.; Shipunov, M. V.; Grachev, V. V.; Melkozerov, M. Yu; Fairoshin, Sh A.

2018-05-01

The experience of modernization of the automation control system of technological processes at the preparation plant under the conditions of technical re-equipment of the preparation plant “Barzasskoye Tovarischestvo” LLC (Berezovsky) is considered. The automated process control systems (APCS), the modernization goals and the ways to achieve them are indicated, the main subsystems of the integrated APCS are presented, the enlarged functional and technical structure of the upgraded system is given. The procedure for commissioning an upgraded system is described.

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.)

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)

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)

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

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

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.)

Development of generic floor response spectra for equipment qualification for seismic loads

International Nuclear Information System (INIS)

Curren, J.R.; Costantino, C.J.

1984-01-01

A generic floor response spectra has been developed for use in the qualification of electrical and mechanical equipment in operating nuclear power plants. Actual PWR and BWR - Mark I structural models were used as representative of a class of structures. For each model, the stiffness properties were varied, with the same mass, so as to extend the fundamental base structure natural frequency from 2 cps to 36 cps. This resulted in fundamental mode coupled natural frequencies as low as 0.86 cps and as high as 30 cps. The characteristics of 1000 floor response spectra were studied to determine the generic spectra. A procedure for its application to any operating plant has been established. The procedure uses as much or as little information that currently exists at the plant relating to the question of equipment qualification. A generic floor response spectra is proposed for the top level of a generic structure. Reduction factors are applied to the peak acceleration for equipment at lower levels

Development of generic floor response spectra for equipment qualification for seismic loads

International Nuclear Information System (INIS)

Curreri, J.R.; Costantino, C.J.

1984-10-01

A generic floor response spectra has been developed for use in the qualification of electrical and mechanical equipment in operating nuclear power plants. Actual PWR and BWR - Mark I structural models were used as representative of a class of structures. For each model, the stiffness properties were varied, with the same mass, so as to extend the fundamental base structure natural frequency from 2 cps to 36 cps. This resulted in fundamental mode coupled natural frequencies as low as 0.86 cps and as high as 30 cps. The characteristics of 1000 floor response spectra were studied to determine the generic spectra. A procedure for its application to any operating plant has been established. The procedure uses as much or as little information that currently exists at the plant relating to the question of equipment qualification. A generic floor response spectra is proposed for the top level of a generic structure. Reduction factors are applied to the peak acceleration for equipment at lower levels

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.

Equipment installation structure of roof slab for tank type FBR and method of equipment installation

International Nuclear Information System (INIS)

Sakai, Takao; Yamakawa, Masanori; Otsuka, Masaya; Sekine, Katsuhisa

1986-01-01

Purpose: To reduce equipment thermal stress and deformation by eliminating uneven temperature distribution caused at the equipment through section of the roof slab for the tank FBR, and at the same time, simplify the structure installation. Method: Multiple number of vertical fin projects are fit on the equipment through-section inside wall for the roof slab and the cylindrical equipment peripheral wall, and with these projected fins, the ring space of the through section is vertically divided into multiple sections in the circumferential direction. The vertical fins on the through-section inside wall and the fins on the equipment peripheral wall are contacted with each other by revolving them in the lateral direction. As a result, the natural convection caused by the difference of temperatures in the vertical direction of the ring space becomes a convection within each sector divided, and never generates circumferential circulation, which reduce uneven temperature distribution caused at the equipment through section. (Kawakami, Y.)

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.

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

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.

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.)

Seismic and dynamic qualification of safety-related electrical and mechanical equipment in operating nuclear power plants: development of a method to generate generic floor-response spectra

International Nuclear Information System (INIS)

Curreri, J.; Costantino, C.; Subudhi, M.; Reich, M.

1983-09-01

Generic floor response spectra were developed for use in the qualification of electrical and mechanical equipment in operating nuclear power plants. The characteristics of 1000 floor response spectra were studied to determine the generic spectra. The procedure developed uses as much or as little information that currently exists at the plant relating to the question of equipment qualification. The general approach was to study the effects on the dynamic characteristics of each of the elements in the chain of events that goes between the loads and the responses. This includes the loads, the soils and the structures. A free-field earthquake response spectra was used to generate horizontal earthquake time histories. The excitation was applied through the soil and into the various structures to produce responses in equipment. An entire range of soil conditions was used with each structure. Actual PWR and BWR - Mark I structural models were used. For each model, the stiffness properties were varied, with the same mass, so as to extend the fundamental base structure natural frequency from 2 cps to 36 cps. The natural frequencies of the structures were varied to obtain maximum response conditions. The actual properties were first used to locate the natural frequencies. The stiffness properties were than varied, with the same mass, to extend the range of the fundamental base structure natural frequency. The intention was to have the coupled structural material frequencies in the vicinity of the peak amplitude frequency content of the excitation spectrum. Particular attention was therefore given to the frequency band between 2 Hz and 4 Hz. A horizontal generic floor response spectra is proposed for the top level of a generic structure. Reduction factors are applied to the peak acceleration for equipment at lower levels

Approximate seismic analysis of piping or equipment mounted on elastoplastic structures

International Nuclear Information System (INIS)

Villaverde, R.

1990-01-01

A simple approximate procedure is presented to estimate the maximum response of equipment, piping, or any other light secondary system mounted on nonlinear structures subjected to earthquake ground motions. The procedure is based on the consideration of structure and equipment as an integrated combined system, and on a response spectrum method for the analysis of nonlinear multistory structures. It is formulated in terms of the initial dynamic properties of the independent structure and equipment components, and the nonlinear response spectrum of a specified earthquake ground motion. It may be applied to any linear multiple-degree-of-freedom secondary system connected at one or two arbitrary points of a multistory structure. It fully takes into account the interaction between primary and secondary systems and the nonclassical damping character of structure-equipment systems. It is restricted, however, to structures with elastoplastic load-deformation behaviour and to those cases in which the mass of the secondary system is small in comparison with the mass of the structure. Its accuracy is evaluated by means of a comparative study with the numerical integration solutions of a number of idealized systems. In this comparative study, the proposed procedure estimates the numerical integration solutions with an average error of about 2 per cent. (author)

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)

Study on integrity evaluation of structures associated with nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

The 3.11 Tohoku District -off the Pacific Ocean Earthquake and tsunami made us observations of tsunami height in large exceedance of the design, and besides it gave the most damages to several nuclear power plants facing the Pacific Ocean at source area of the earthquake. Particularly, at Fukushima-Daiichi Nuclear Power Plant, the great tsunami caused the simultaneous failure on several plant's equipment and components, which escalated into the core damage. Considering these background, the objective of this research is to enhance fundamental technology relative to integrity evaluation of SSC's (System, Structure, components) targeting external events such as earthquakes and tsunamis. Specifically, it is performed to develop structure evaluation methods against tsunami, to develop seismic isolation system, and to enhance non-liner analysis methods for building and so on. In viewpoint of the other external events except earthquake and tsunami, it was performed to develop impact analysis methods on building and outdoor structure against swept things caused by tornadoes. After that on the basis of these developments, it is performed to draw up guidelines such as the base isolation structure review guide, and the structure design and risk evaluation guide against tsunami, which are to be used in cross-check analysis targeting integrity evaluation of nuclear power plant's structures against external events such as earthquakes and tsunamis. (author)

Study on integrity evaluation of structures associated with nuclear power plants

International Nuclear Information System (INIS)

2013-01-01

The 3.11 Tohoku District -off the Pacific Ocean Earthquake and tsunami made us observations of tsunami height in large exceedance of the design, and besides it gave the most damages to several nuclear power plants facing the Pacific Ocean at source area of the earthquake. Particularly, at Fukushima-Daiichi Nuclear Power Plant, the great tsunami caused the simultaneous failure on several plant's equipment and components, which escalated into the core damage. Considering these background, the objective of this research is to enhance fundamental technology relative to integrity evaluation of SSC's (System, Structure, components) targeting external events such as earthquakes and tsunamis. Specifically, it is performed to develop structure evaluation methods against tsunami, to develop seismic isolation system, and to enhance non-liner analysis methods for building and so on. In viewpoint of the other external events except earthquake and tsunami, it was performed to develop impact analysis methods on building and outdoor structure against swept things caused by tornadoes. After that on the basis of these developments, it is performed to draw up guidelines such as the base isolation structure review guide, and the structure design and risk evaluation guide against tsunami, which are to be used in cross-check analysis targeting integrity evaluation of nuclear power plant's structures against external events such as earthquakes and tsunamis. (author)

Common floor system vertical earthquake-proof structure for reactor equipment

International Nuclear Information System (INIS)

Morishita, Masaki.

1996-01-01

In an LMFBR type reactor, a reactor container, a recycling pump and a heat exchanger are disposed on a common floor. Vertical earthquake-proof devices which can be stretched only in vertical direction formed by laminating large-sized bellevilles are disposed on a concrete wall at the circumference of each of reactor equipments. A common floor is placed on all of the vertical earthquake-proof devices to support the entire earthquake-proof structure simultaneously. If each of reactor equipments is loaded on the common floor and the common floor is entirely supported against earthquakes altogether, since the movement of each of the reactor equipments loaded on the common floor is identical, relative dislocation is not exerted on the main pipelines which connect the equipments. In addition, since the entire earthquake structure has a flat common floor and each of the reactor equipments is suspended to minimize the distance between a gravitational center and a support point, locking vibration is less caused to the horizontal earthquake. (N.H.)

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

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.

Kozloduy Nuclear Power Plant (Unit 1 and 2). Proposed modifications to increase the seismic capability of equipment and main structures; Central Nuclear de Kozloduy (Unidades 1 y 2). Modificaciones propuestas a los equipos y estructuras principales para incrementar su capacidad sismica

Energy Technology Data Exchange (ETDEWEB)

Ordonez Villalobos, A [Empresarios Agrupados, A.I.E., Madrid (Spain); Monette, P R [Westinghouse Energy Systems International (WESI), Bussels (Belgium)

1993-12-15

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)

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)

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.)

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

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.

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

Seismic soil-structure-equipment interaction analysis of unit 5/6, Kozloduy NPP

Energy Technology Data Exchange (ETDEWEB)

Kostov, M [Bulgarian Academy of Sciences, Central Laboratory for Seismic Mechanics and Earthquake Engineering, Sofia (Bulgaria)

1995-07-01

This research project is aimed to analyse problems of soil-structure-equipment interaction under seismic excitation in case of Kozloduy NPP. Reevaluation and upgrading of Kozloduy NPP has started after 1977 Vrancea earthquake. New Safe Shutdown Earthquake (SSE) level was defined, upgrading most of structural equipment was performed, seismic instrumentation was installed. New investigations were initiated after 1990 IAEA mission visited the site. A comprehensive site confirmation project was started with a subsequent structural and equipment reevaluation and upgrading. This work deals with Units 5 and 6 of WWER-1000 type only.

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.

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.

Equipment reliability process improvement and preventive maintenance optimization

International Nuclear Information System (INIS)

Darragi, M.; Georges, A.; Vaillancourt, R.; Komljenovic, D.; Croteau, M.

2004-01-01

The Gentilly-2 Nuclear Power Plant wants to optimize its preventive maintenance program through an Integrated Equipment Reliability Process. All equipment reliability related activities should be reviewed and optimized in a systematic approach especially for aging plants such as G2. This new approach has to be founded on best practices methods with the purpose of the rationalization of the preventive maintenance program and the performance monitoring of on-site systems, structures and components (SSC). A rational preventive maintenance strategy is based on optimized task scopes and frequencies depending on their applicability, critical effects on system safety and plant availability as well as cost-effectiveness. Preventive maintenance strategy efficiency is systematically monitored through degradation indicators. (author)

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

9 CFR 590.502 - Equipment and utensils; PCB-containing equipment.

Science.gov (United States)

2010-01-01

... Sanitary Standards and accepted practices currently in effect for such equipment. (c) New or replacement equipment or machinery (including any replacement parts) brought onto the premises of any official plant... equipment and machinery, and any replacement parts for such equipment and machinery. Totally enclosed...

40 CFR 265.114 - Disposal or decontamination of equipment, structures and soils.

Science.gov (United States)

2010-07-01

... decontamination of equipment, structures and soils. During the partial and final closure periods, all contaminated... 40 Protection of Environment 25 2010-07-01 2010-07-01 false Disposal or decontamination of equipment, structures and soils. 265.114 Section 265.114 Protection of Environment ENVIRONMENTAL PROTECTION...

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.)

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

Making Plant-Support Structures From Waste Plant Fiber

Science.gov (United States)

Morrow, Robert C.; < oscjmocl. < attjew K/; {ertzbprm. A,amda; Ej (e. Cjad); Hunt, John

2006-01-01

Environmentally benign, biodegradable structures for supporting growing plants can be made in a process based on recycling of such waste plant fiber materials as wheat straw or of such derivative materials as paper and cardboard. Examples of structures that can be made in this way include plant plugs, pots, planter-lining mats, plant fences, and root and shoot barriers. No chemical binders are used in the process. First, the plant material is chopped into smaller particles. The particles are leached with water or steam to remove material that can inhibit plant growth, yielding a fibrous slurry. If the desired structures are plugs or sheets, then the slurry is formed into the desired shapes in a pulp molding subprocess. If the desired structures are root and shoot barriers, pots, or fences, then the slurry is compression-molded to the desired shapes in a heated press. The processed materials in these structures have properties similar to those of commercial pressboard, but unlike pressboard, these materials contain no additives. These structures have been found to withstand one growth cycle, even when wet

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.)

Data base EQDB - data base of the qualified equipment's for NPP

International Nuclear Information System (INIS)

Rovny, K.

2009-01-01

In the contribution, there is presented the project of the data base for qualified equipment's for nuclear power plants. The data base is operated by the 'Certification body which certified the products - the chosen equipment's for nuclear power plants', reg. No. P-028, at VUJE, Inc. Trnava. Data base will serve to the designers, the operators of the nuclear power plants and the workers from Nuclear regulatory authority of the Slovak Republic as a source of information about the state of concrete type equipment's qualification. In the first part of the contribution, there is information about the legislation and technical requirements for equipment's qualification, the way of demonstration and importance of the qualification for the operator. In the next part, there is presented the own structure of data base and the works with own data base regarding the examples of concrete equipment's. The data base will be accessible after the free registration on address WWW.EQDB.sk from 1.5.2009

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)

Automatization of welding for nuclear power equipments and facilities

International Nuclear Information System (INIS)

Tamai, Yasumasa; Matsumoto, Teruo; Koyama, Takaichi

1980-01-01

For the requirement of high reliability in the construction of nuclear power plants and the reduction of radiation exposure in the modefying works of existing plants, the automation and remote operation of welding have increased their necessity. In this paper, the present state of the automation of welding for making machines, equipments and pipings for nuclear power plants in Hitachi Ltd. is described, and the aim of developing the automation, the features of the equipments and the state of application to actual plants are introduced, centering around the automation of welding for large structures such as reactor containment vessels and the remote type automatic welding system for pipings. By these automations, the large outcomes were obtained in the improvement of welding quality required for the machines and equipments for atomic energy. Moreover, the conspicuous results were also obtained in case of the peculiar works to nuclear power plants, in which the reduction of the radiation exposure related to human bodies and the welding of high quality are demanded. The present state of the automation of welding for nuclear installations in Hitachi Ltd., the development of automatic welding equipments and the present state of application to actual plants, and the development and application of the automatic pipe working machine for reducing radiation exposure are explained. (Kako, I.)

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

Role of closely spaced modes in the seismic response of equipment and structures

International Nuclear Information System (INIS)

Nelson, F.C.

1975-01-01

The genesis for this study was a need for an explanation of how to combine the peak accelerations, displacements, stresses, etc., which result from a response spectrum analysis of nuclear power plant structures and equipment. The explanation was intended to be comprehensive in the sense that it would be based on a general principle from which the absolute sum, square root of the sum of the squares (SRSS), and double-sum modal recombination rules could be extracted as special cases. The purpose of such a comprehensive explanation was to convince design engineers that these rules, in particular the recently proposed double-sum rule, were logical in their formulation and therefore sensible in their application

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.

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

Similarity principles for seismic qualification of equipment by experience data

International Nuclear Information System (INIS)

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

1989-01-01

A methodology is developed for seismic qualification of nuclear plant equipment by applying similarity principles to existing experience data. Experience data is that available from previous qualifications by analysis or testing, or from actual earthquake events. Similarity principles are defined in terms of excitation, equipment physical characteristics, and equipment response. Physical similarity is further defined in terms of a critical transfer function for response at a location on a primary structure, whose response can be assumed directly related to fragility of the item under elevated levels of excitation. Procedures are developed for combining experience data into composite specifications for qualification of equipment that can be shown to be physically similar to the reference equipment. Other procedures are developed for extending qualifications beyond the original specifications under certain conditions. Some examples for application and verification of the procedures are given for actual test data available from previous qualifications. The developments are intended to elaborate on the rather broad guidelines by the IEEE 344 Standards Committee for equipment qualification in new nuclear plants. The results also contribute to filling a gap that exists between the IEEE 344 methodology and that which was previously developed for equipment in existing plants by the Seismic Qualification Utilities Group. 10 refs., 9 figs., 1 tab

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)

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

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.)

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.)

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...

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

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

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)

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.)

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.)

Investgation concerning the structure, critical analysis and improvement of the organization of quality assurance in the qualification of electrical equipment for nuclear power plants with new parts and in-service inspections

International Nuclear Information System (INIS)

1985-01-01

The study gives a survey of methods for the qualification of electrical equipment for nuclear power plants presently applied in the Federal Republic of Germany full consideration being given to the organization of quality assurance on the premises of the organizations involved (operators, plant suppliers, manufacturers). The organization of the qualification is to be kept distinct from the technical execution of the qualification. The qualification procedures are compared to those applied in France and in the USA. Aspects of future developments are shown with are promising with respect to: an increase in safety of electrical equipment for nuclear power plants, a simplification of the licencing procedure under Atomic Law. (orig./HP) [de

40 CFR 267.116 - What must I do with contaminated equipment, structure, and soils?

Science.gov (United States)

2010-07-01

... equipment, structure, and soils? 267.116 Section 267.116 Protection of Environment ENVIRONMENTAL PROTECTION..., structure, and soils? You must properly dispose of or decontaminate all contaminated equipment, structures, and soils during the partial and final closure periods. By removing any hazardous wastes or hazardous...

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.

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.

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)

Consideration of higher seismic loads at existing plants

Energy Technology Data Exchange (ETDEWEB)

Liebig, J.; Pellissetti, M.

2015-07-01

Because of advancement of methods in probabilistic seismic hazard analysis, plenty of existing plants face higher seismic loads as an obligation from the national authorities. In case of such obligations safety related structures and equipment have to be reevaluated or requalified for the increased seismic loads. The paper provides solutions for different kinds of structures and equipment inside the plant, avoiding cost intensive hardware exchange. Due to higher seismic loads different kinds of structures and equipment inside a plant have to be reevaluated. For civil structures, primary components, mechanical components, distribution lines and electrical and I&C equipment different innovative concepts will be applied to keep structures and equipment qualified for the higher seismic loads. Detailed analysis, including the modeling of non-linear phenomena, or minor structural upgrades are cost competitive, compared to cost intensive hardware exchanges. Several case studies regarding the re-evaluation and requalification of structures and equipment due to higher seismic loads are presented. It is shown how the creation of coupled finite element models and the consistent propagation of acceleration time histories through the soil, building and primary circuit lead to a significant load reduction Electrical and I&C equipment is reinforced by smart upgrades which increase the natural equipment frequencies. Therefore for all devices inside the cabinets the local acceleration will not increase and the seismic qualification will be maintained. The case studies cover both classical deterministic and probabilistic re-evaluations (fragility analysis). Furthermore, the substantial benefits of non-linear limit load evaluation, such as push-over analysis of buildings and limit load analysis of fuel assemblies, are demonstrated. (Author)

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

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

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

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.

Maintenance and fabrication of electronic equipment

International Nuclear Information System (INIS)

Chung, Chong Eun; Moon, Byung Soo; Hong, Suk Boong; Kim, Yong Keun; Kim, Jung Bok

2003-12-01

Development of radiation monitoring equipment could be the base of domestic development of RMS. And the technique could be adapted to development of other radiation equipment of KAERI as well as hospitals and nuclear power plants. The RMS technology could be adapted to the development of precision instruments related to nuclear radiation and be the base of fundamental technology such as protein structure analysis of bio technology, development of nano advanced material and aircraft material. The technology of multi-channel readout ASIC for nuclear radiation detector, which has been imported from abroad, could be adapted to development of radiation equipment for image processing, position of detection, NDT etc., and also the technique will be expected to contribute to increase the use of radiation technology to industrial applications

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)

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

Application of new control technology during the maintenance of equipment in the Laguna Verde nuclear power plant

International Nuclear Information System (INIS)

Ojeda R, M. A.

2008-01-01

In the nuclear power plant of Laguna Verde, in normal operation and recharges are carried out activities of preventive maintenance and corrective to different equipment, due to the one displacement of radioactive materials from the vessel of the reactor until the one system of vapor, different radiation levels are generated (from low until very high) in the circuits of vapor and water, the particles can be incrusted on those interior surfaces of the pipes and equipment, creating this way a potential risk of contamination and exhibition during the maintenance of equipment. To help to optimize the dose to the personnel the use of new technology the has been implemented which besides contributing an absolute control of the work, it offers bigger comfort to the one worker during the development of their work, also contributing a supervision more effective of the same one. Using the captured and processed information of the work developed you can use for the personnel's capacitation and feedback of the work for the continuous improvement of the same one. During a reduction of programmed power and normal operation are carried out maintenance correctives and specific works to preserve the readiness and ability of the equipment and with this to maintain the security of the nuclear power plant. The development of the theme it is showing the advances and commitments of personnel to take to excellence to the nuclear power plant of Laguna Verde showing to the obtained results of the dose and benefits of 2 works carried out in the nuclear power plant where tools ALARA were applied as well as the use of the new technology (Video Equipment of Tele dosimetry and Audio 'VETA') in works carried out in the building of purification level 10.15, change and cuts of filter of the prefilters of system G16, as well as,the retirement and transfer for its decay of High Integrity Container (HIC) of the building of purification level -0.55 to the Temporary Warehouse in Site. Works of high

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 ...

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.)

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)

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

Classification of methods and equipment recovery secondary waters

Directory of Open Access Journals (Sweden)

G. V. Kalashnikov

2017-01-01

Full Text Available The issues of purification of secondary waters of industrial production have an important place and are relevant in the environmental activities of all food and chemical industries. For cleaning the transporter-washing water of beet-sugar production the key role is played by the equipment of treatment plants. A wide variety of wastewater treatment equipment is classified according to various methods. Typical structures used are sedimentation tanks, hydrocyclones, separators, centrifuges. In turn, they have a different degree of purification, productivity through the incoming suspension and purified secondary water. This is equipment is divided into designs, depending on the range of particles to be removed. A general classification of methods for cleaning the transporter-washing water, as well as the corresponding equipment, is made. Based on the analysis of processes and instrumentation, the main methods of wastewater treatment are identified: mechanical, physicochemical, combined, biological and disinfection. To increase the degree of purification and reduce technical and economic costs, a combined method is widely used. The main task of the site for cleaning the transporter-washing waters of sugar beet production is to provide the enterprise with water in the required quantity and quality, with economical use of water resources, taking into account the absence of pollution of surface and groundwater by industrial wastewater. In the sugar industry is currently new types of washing equipment of foreign production are widely used, which require high quality and a large amount of purified transporter-washing water for normal operation. The proposed classification makes it possible to carry out a comparative technical and economic analysis when choosing the methods and equipment for recuperation of secondary waters. The main equipment secondary water recovery used at the beet-sugar plant is considered. The most common beet processing plant is a

Database structure and file layout of Nuclear Power Plant Database. Database for design information on Light Water Reactors in Japan

International Nuclear Information System (INIS)

Yamamoto, Nobuo; Izumi, Fumio.

1995-12-01

The Nuclear Power Plant Database (PPD) has been developed at the Japan Atomic Energy Research Institute (JAERI) to provide plant design information on domestic Light Water Reactors (LWRs) to be used for nuclear safety research and so forth. This database can run on the main frame computer in the JAERI Tokai Establishment. The PPD contains the information on the plant design concepts, the numbers, capacities, materials, structures and types of equipment and components, etc, based on the safety analysis reports of the domestic LWRs. This report describes the details of the PPD focusing on the database structure and layout of data files so that the users can utilize it efficiently. (author)

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

29 CFR 1926.1000 - Rollover protective structures (ROPS) for material handling equipment.

Science.gov (United States)

2010-07-01

... 29 Labor 8 2010-07-01 2010-07-01 false Rollover protective structures (ROPS) for material handling equipment. 1926.1000 Section 1926.1000 Labor Regulations Relating to Labor (Continued) OCCUPATIONAL SAFETY... CONSTRUCTION Rollover Protective Structures; Overhead Protection § 1926.1000 Rollover protective structures...

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

Management of ageing of I and C equipment in nuclear power plants. Report prepared within the framework of the International Working Group on Nuclear Power Plant Control and Instrumentation

International Nuclear Information System (INIS)

2000-06-01

Experience has shown that ageing and obsolescence have the potential to cause the maintainability and operability of many instrumentation and control (I and C) systems to deteriorate well before the end of plant life. An I and C ageing management strategy is therefore required to control and minimize this threat. This report gives guidance on how to develop such a strategy and provides examples and supporting information on how established and recently developed maintenance, surveillance, and testing techniques may be employed to support the strategy. In some cases, equipment refurbishment may be necessary and guidance on this subject is given in a companion publication (IAEA-TECDOC-1016, Modernization of Instrumentation and Control in Nuclear Power Plants, IAEA, Vienna, 1998). The International Working Group on Nuclear Power Plant Control and Instrumentation (IWG-NPPCI) of the IAEA proposed in 1995 that a technical report be prepared to provide general guidelines on the management of ageing of important I and C equipment in nuclear power plants. The purpose of the report would be to guide the worldwide nuclear industry on potential effects of I and C ageing on plant safety and economy, and the means that are available to help minimize or eliminate any detrimental consequences of ageing. In response, a consultants meeting of five experts from Finland, France, Germany, the United Kingdom and the USA was held by the IAEA in Vienna in September 1997 to exchange national experience on the subject and to discuss the possible content of the report. The group of experts was tasked with bringing together all the information that is available on I and C ageing and ageing management methods. After a thorough discussion and analysis of the available information, an extended outline of the report on the subject was produced. The purpose of the extended outline was to identify a structure of the report, bring together information available at the moment and to provide guidance

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)

Risk exposures for human ornithosis in a poultry processing plant modified by use of personal protective equipment: an analytical outbreak study.

Science.gov (United States)

Williams, C J; Sillis, M; Fearne, V; Pezzoli, L; Beasley, G; Bracebridge, S; Reacher, M; Nair, P

2013-09-01

Ornithosis outbreaks in poultry processing plants are well-described, but evidence for preventive measures is currently lacking. This study describes a case-control study into an outbreak of ornithosis at a poultry processing plant in the East of England, identified following three employees being admitted to hospital. Workers at the affected plant were recruited via their employer, with exposures assessed using a self-completed questionnaire. Cases were ascertained using serological methods or direct antigen detection in sputum. 63/225 (28%) staff participated, with 10% of participants showing evidence of recent infection. Exposure to the killing/defeathering and automated evisceration areas, and contact with viscera or blood were the main risk factors for infection. Personal protective equipment (goggles and FFP3 masks) reduced the effect of exposure to risk areas and to self-contamination with potentially infectious material. Our study provides some evidence of effectiveness for respiratory protective equipment in poultry processing plants where there is a known and current risk of ornithosis. Further studies are required to confirm this tentative finding, but in the meantime respiratory protective equipment is recommended as a precautionary measure in plants where outbreaks of ornithosis occur.

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

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.)

Seismic fragilities for nuclear power plant risk studies

International Nuclear Information System (INIS)

Kennedy, R.P.; Ravindra, M.K.

1983-01-01

Seismic fragilities of critical structures and equipment are developed as families of conditional failure frequency curves plotted against peak ground acceleration. The procedure is based on available data combined with judicious extrapolation of design information on plant structures and equipment. Representative values of fragility parameters for typical modern nuclear power plants are provided. Based on the fragility evaluation for about a dozen nuclear power plants, it is proposed that unnecessary conservatism existing in current seismic design practice could be removed by properly accounting for inelastic energy absorption capabilities of structures. The paper discusses the key contributors to seismic risk and the significance of possible correlation between component failures and potential design and construction errors

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.

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)

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

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

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.

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)

Catalog of physical protection equipment. Book 1: Volume I. Barriers and structural components

International Nuclear Information System (INIS)

Haberman, W.

1977-06-01

A catalog of commercially available physical protection equipment has been prepared for use by the U.S. Nuclear Regulatory Commission (NRC). Included is information on barrier structures and equipment, interior and exterior intrusion detection sensors, entry (access) control devices, surveillance and alarm assessment equipment, contraband detection sensors, automated response equipment, general purpose displays and general purpose communications, with one volume devoted to each of these eight areas. For each item of equipment the information included consists of performance, physical, cost and supply/logistics data. The entire catalog is contained in three notebooks for ease in its use by licensing and inspection staff at NRC

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

Similarity principles for equipment qualification by experience

International Nuclear Information System (INIS)

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

1988-07-01

A methodology is developed for seismic qualification of nuclear plant equipment by applying similarity principles to existing experience data. Experience data are available from previous qualifications by analysis or testing, or from actual earthquake events. Similarity principles are defined in terms of excitation, equipment physical characteristics, and equipment response. Physical similarity is further defined in terms of a critical transfer function for response at a location on a primary structure, whose response can be assumed directly related to ultimate fragility of the item under elevated levels of excitation. Procedures are developed for combining experience data into composite specifications for qualification of equipment that can be shown to be physically similar to the reference equipment. Other procedures are developed for extending qualifications beyond the original specifications under certain conditions. Some examples for application of the procedures and verification of them are given for certain cases that can be approximated by a two degree of freedom simple primary/secondary system. Other examples are based on use of actual test data available from previous qualifications. Relationships of the developments with other previously-published methods are discussed. The developments are intended to elaborate on the rather broad revised guidelines developed by the IEEE 344 Standards Committee for equipment qualification in new nuclear plants. However, the results also contribute to filling a gap that exists between the IEEE 344 methodology and that previously developed by the Seismic Qualification Utilities Group. The relationship of the results to safety margin methodology is also discussed. (author)

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.

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

Modification of evaluation response spectrum by ductility of equipment anchorage

International Nuclear Information System (INIS)

Choi, I. G.; Jun, Y. S.; Su, J. M.

2003-01-01

The failure mode of welded anchorage is assumed as brittle in the seismic capacity evaluation of nuclear power plant equipments. But the welded anchorage has some ductile capacity. This limited displacement capacity can cause the reduction of the effective frequency of high frequency equipments and the increase of the inelastic energy absorption capacity due to the nonlinear behavior. In this study, the uniform hazard spectrum for Korean nuclear power plant site was modified using the response spectrum reduction factor developed by EPRI. The spectral acceleration for various damping ratio was determined by the theoretical method based on the random vibration theory. In conclusion, the high frequency components of evaluation response spectra were greatly reduced due to the consideration of welded anchorage ductility. This reduced response spectra can be used for the development of in-structure response spectra used in the seismic capacity evaluation of high frequency equipments

Progress report of the critical equipment monitoring system

International Nuclear Information System (INIS)

Pantis, M.J.

1984-01-01

The Philadelphia Electric Company has contracted with Energy Data Systems to develop a Critical Equipment Monitoring System for its Peach Bottom Nuclear Plant. This computerized system is designed to acquire and maintain accurate and timely status information on plant equipment. It will provide auditable record of plant and equipment transactions. Positive equipment identification and location will be provided. Errors in complex logical checking will be minimized. This system should reduce operator loading and improve operator communicatin with the plant personnel. Phase I of this system was installed at Peach Bottom Nuclear Station May 1982. It provides the necessary hardware and software to do check-off lists on critical plant systems. This paper describes some of the start-up and operational problems encountered

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.)

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.

Review of structure damping values for elastic seismic analysis of nuclear power plants

International Nuclear Information System (INIS)

Hashimoto, P.S.; Steele, L.K.; Johnson, J.J.; Mensing, R.W.

1993-03-01

Current US Nuclear Regulatory Commission guidance on structure damping values for elastic seismic design analysis of nuclear power plants are contained in Regulatory Guide 1.61 (R.G. 1.61). The objectives of the study described in this report are to investigate the adequacy of R.G1.61 structure damping values based on currently available data, and to recommend revisions to R.G. 1.61 as appropriate. Measured structure damping values, and associated structure, foundation, excitation, and input/response parameters, were collected and compiled. These data were analyzed to identify the parameters that significantly influence structure damping and to quantify structure damping in terms of these parameters. Based on this study, current R.G. 1.61 damping values for structure design are either adequate, or require only minor revision, depending on the structure material. More explicit guidance on structure damping values for seismic analysis to determine input to equipment has been prepared, along with other recommendations to improve the applicability of R.G. 1.61

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

Examining work structure in nuclear power plants

International Nuclear Information System (INIS)

Bauman, M.B.; Boulette, M.D.; Van Cott, H.P.

1985-01-01

This paper describes the assessment of the work structure of ten nuclear power plants. Work structure factors are those factors that relate to the way in which work at all levels in a plant is organized, staffed, managed, rewarded, and perceived by plant personnel. Questionnaires given to a cross-section of personnel at the plants were the primary source of data collection. Structured ''critical incident'' interviews were conducted to verify the questionnaire results. The study revealed that a variety of work structure factor problem areas do exist in nuclear power plants. The paper highlights a prioritized set of candidate research themes to be considered in EPRI's Work Structure and Performance Research Program

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

Seismic analysis and structure capacity evaluation of the Belene nuclear power plant

International Nuclear Information System (INIS)

Johnson, J.J.; Hashimoto, P.S.; Campbell, R.D.; Baltus, R.S.

1993-01-01

The seismic analysis and structure capacity evaluation of the Belene Nuclear Power Plant, a two-unit WWER 1000, was performed. The principal objective of the study was to review the major aspects of the seismic design including ground motion specification, foundation concept and materials, and the Unit I main reactor building structure response and capacity. The main reactor building structure /foundation/soil were modeled and analyzed by a substructure approach to soil-structure interaction (SSI) analysis. The elements of the substructure approach, implemented in the family of computer programs CLASSI, are: Specification of the free-field ground motion; Modeling the soil profile; SSI parameters; Modeling the structure; SSI-response analyses. Each of these aspects is discussed. The Belene Unit 1 main reactor building structure was evaluated to verify the seismic design with respect to current western criteria. The structural capacity evaluation included criteria development, element load distribution analysis, structural element selection, and structural element capacity evaluation. Equipment and commodity design criteria were similarly reviewed and evaluated. Methodology results and recommendations are presented. (author)

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

International Nuclear Information System (INIS)

1987-09-01

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

Re-qualification of switchgear equipment following field modification

International Nuclear Information System (INIS)

Tulk, J.D.; El Bestawi, M.A.

1984-01-01

A set of 4.16 kV switchgear was seismically qualified by shaker table test for use in a nuclear power plant. Later, the equipment was modified by the addition of two extra switches and enclosures. This paper describes the methods used to extend the qualification achieved in the original proof test to the new configuration. Qualification of the modified switchgear depended on demonstrating that the intensity of the shaking that would be experienced by equipment mounted in the extended enclosure during a design basis earthquake (DBE) would be no stronger than the shaking that occurred during the original qualification testing. This was accomplished by a combination of analysis and experiment. Natural frequencies and mode-shapes of the original and extended enclosure structures were determined through in-situ modal tests on the equipment. Mathematical models based on the experimental results were then used for dynamic analyses which generated two sets of in-equipment response spectra: one set representing the original qualification test; and one set representing the response of the modified equipment to a hypothetical DBE. Qualification was confirmed by demonstrating that the test-based response spectra for the original configuration exceeded the response spectra developed for the extended structure subjected to DBE level excitation

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

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

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.)

Synergistic behaviour of nuclear radiation, temperature-humidity extremes and LOCA situation on safety and safety-related equipment in Indian nuclear power plants

International Nuclear Information System (INIS)

Kulkarni, R.D.; Bora, J.S.; Prakash, Ravi; Agarwal, Vivek; Sundersingh, V.P.

2002-01-01

Full text: The general philosophy for the instrumentation in nuclear power plants is based on the use of equipment/instruments which are capable of continuous satisfactory operation over a long period of time with minimum attention. Long term reliability under varying service conditions is of prime importance. The reliability of nuclear power plant depends on the reliability of safety and safety-related electronic instruments/ equipment used for performing the crucial tasks. The electrical and electronic systems/ circuits/ components of the equipment used in reactor safety systems (e.g. reactor protection system, emergency core cooling system, etc.) and reactor safety-related systems (e.g. reactor containment isolation and cooling system, reactor shutdown system, etc.) are responsible for safe and reliable operation of a nuclear power plant. The performance of reactor safety and safety-related equipment/instruments viz. pressure and differential pressure transmitter, amplifier for ion chamber, etc. has been evaluated under synergistic atmosphere including LOCA to find out the critical link in the circuits and subsequent modifications are suggested. The mathematical representation of the generated database has been done to estimate the life span of the instruments and accordingly the guidelines has been prepared for the operational staff to avoid the forced outage of the plant. All the details are included and mathematical models are presented to predict the future performances

Guidelines for nuclear plant response to an earthquake

International Nuclear Information System (INIS)

1989-12-01

Guidelines have been developed to assist nuclear plant personnel in the preparation of earthquake response procedures for nuclear power plants. The objectives of the earthquake response procedures are to determine (1) the immediate effects of an earthquake on the physical condition of the nuclear power plant, (2) if shutdown of the plant is appropriate based on the observed damage to the plant or because the OBE has been exceeded, and (3) the readiness of the plant to resume operation following shutdown due to an earthquake. Readiness of a nuclear power plant to restart is determined on the basis of visual inspections of nuclear plant equipment and structures, and the successful completion of surveillance tests which demonstrate that the limiting conditions for operation as defined in the plant Technical Specifications are met. The guidelines are based on information obtained from a review of earthquake response procedures from numerous US and foreign nuclear power plants, interviews with nuclear plant operations personnel, and a review of reports of damage to industrial equipment and structures in actual earthquakes. 7 refs., 4 figs., 4 tabs

Statistical analysis of the behaviour of the mechanical equipment of EDFs power plants - evaluation of the availability and safety of thermal and nuclear units

International Nuclear Information System (INIS)

Procaccia, H.; Brillon, A.; Cravero, M.; Lucenet, G.

1975-01-01

The investigation and research directorate of EDF has undertaken a statistical analysis of the behaviour of large mechanical equipment at conventional power stations during the ten years following the operating reports of these stations. It has thus been possible to determine the intrinsic reliability, the failure rate, the mean repair time, and the mean good operating time of feed water reheating points, power turbines, pumps and boilers of the various EDF plants (125 and 250 MW) leading to a consideration of the feasibility of an extrapolation to present and future plants. Based on these elementary investigation two methods of calculation have been developed. One is used to assess the overall availability of a thermal or nuclear power station based on the knowledge of the failure rates of the equipment, each piece of equipment being associated with an idea of its technical importance in the functioning of the equipment. A numerical application is given for 125 and 250 MW conventional plants. The purpose of the other method is to estimate the operational safety of the safety equipment of nuclear power stations, based on the development of tree diagrams for faults in basic equipment. A numerical example is given for the cooling systems for Phenix and for one of the Super Phenix versions. (author)

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)

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

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.

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

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

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

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.

On-line testing of nuclear plant temperature and pressure instrumentation and other critical plant equipment. IAEA regional workshop. Working material

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-12-31

Under European regional TC project RER/4/011, IAEA and VUJE Training centre organized a workshop on On-line Testing of Nuclear Power Plant Temperature and Pressure Instrumentation and Other Critical Plant Equipment in Trnava, Slovak Republic, from 25 to 29 May 1998. The objective of the workshop was to review the state-of-the-art in NPP instrumentation, cover typical instrumentation problems and solutions, describe technical and regulatory requirements for verifying the performance of nuclear power plant instrumentation, describe new methods developed and applied in NPPs for on-line verification and performance of instrumentation and present new techniques using existing instrumentation to identify the on-set problems in the plant electrical, mechanical and thermal hydraulic systems. Particular emphasis was placed on temperature measurements by Resistance Temperature Detectors (RTDs) and thermocouples and pressure measurements using motion-balanced and forced-balanced pressure transmitters. This proceedings includes papers presented by the invited speakers and the participants each with an abstract as wells as a summary of the Round-Table discussions Refs, figs, tabs

On-line testing of nuclear plant temperature and pressure instrumentation and other critical plant equipment. IAEA regional workshop. Working material

International Nuclear Information System (INIS)

1998-01-01

Under European regional TC project RER/4/011, IAEA and VUJE Training centre organized a workshop on On-line Testing of Nuclear Power Plant Temperature and Pressure Instrumentation and Other Critical Plant Equipment in Trnava, Slovak Republic, from 25 to 29 May 1998. The objective of the workshop was to review the state-of-the-art in NPP instrumentation, cover typical instrumentation problems and solutions, describe technical and regulatory requirements for verifying the performance of nuclear power plant instrumentation, describe new methods developed and applied in NPPs for on-line verification and performance of instrumentation and present new techniques using existing instrumentation to identify the on-set problems in the plant electrical, mechanical and thermal hydraulic systems. Particular emphasis was placed on temperature measurements by Resistance Temperature Detectors (RTDs) and thermocouples and pressure measurements using motion-balanced and forced-balanced pressure transmitters. This proceedings includes papers presented by the invited speakers and the participants each with an abstract as wells as a summary of the Round-Table discussions

Equipment Reliability Program in NPP Krsko

International Nuclear Information System (INIS)

Skaler, F.; Djetelic, N.

2006-01-01

Operation that is safe, reliable, effective and acceptable to public is the common message in a mission statement of commercial nuclear power plants (NPPs). To fulfill these goals, nuclear industry, among other areas, has to focus on: 1 Human Performance (HU) and 2 Equipment Reliability (EQ). The performance objective of HU is as follows: The behaviors of all personnel result in safe and reliable station operation. While unwanted human behaviors in operations mostly result directly in the event, the behavior flaws either in the area of maintenance or engineering usually cause decreased equipment reliability. Unsatisfied Human performance leads even the best designed power plants into significant operating events, which can be found as well-known examples in nuclear industry. Equipment reliability is today recognized as the key to success. While the human performance at most NPPs has been improving since the start of WANO / INPO / IAEA evaluations, the open energy market has forced the nuclear plants to reduce production costs and operate more reliably and effectively. The balance between these two (opposite) goals has made equipment reliability even more important for safe, reliable and efficient production. Insisting on on-line operation by ignoring some principles of safety could nowadays in a well-developed safety culture and human performance environment exceed the cost of electricity losses. In last decade the leading USA nuclear companies put a lot of effort to improve equipment reliability primarily based on INPO Equipment Reliability Program AP-913 at their NPP stations. The Equipment Reliability Program is the key program not only for safe and reliable operation, but also for the Life Cycle Management and Aging Management on the way to the nuclear power plant life extension. The purpose of Equipment Reliability process is to identify, organize, integrate and coordinate equipment reliability activities (preventive and predictive maintenance, maintenance

Large-component handling equipment and its use

International Nuclear Information System (INIS)

Krieg, S.A.; Swannack, D.L.

1983-01-01

The Fast Flux Test Facility (FFTF) reactor systems have special requirements for component replacements during maintenance servicing. Replacement operations must address handling of equipment within shielded metal containers while maintaining an inert atmosphere to prevent reaction of sodium with air. Plant identification of a failed component results in selecting and assembling the maintenance cask and equipment transport system for transfer from the storage facility to the Reactor Containment Building (RCB). This includes a proper diameter and length cask, inert atmosphere control consoles, component lift fixture and support structure for interface with the facility area surrounding the component. This equipment is staged in modular groups in the Reactor Service Building for transfer through the equipment airlock to the containment interior. The failed component is generally prepared for replacement by installation of the special lifting fixture attachment. Assembly of the cask support structure is performed over the component position on the containment building operating floor. The cask and shroud from the reactor interface are inerted after all manual service connections and handling attachments are completed. The component is lifted from the reactor and into the cask interior through a floor valve which is then closed to isolate the component reactor port. The cask with sodium wetted component is transferred to a service/repair location, either within containment or outside, to the Maintenance Facility cleaning and repair area. The complete equipment and handling operations for replacement of a large reactor component are described

Development of nuclear equipment qualification technology

Energy Technology Data Exchange (ETDEWEB)

Choi, Heon O; Kim, Wu Hyun; Kim, Jin Wuk; Kim, Jeong Hyun; Lee, Jeong Kyu; Kim, Yong Han; Jeong, Hang Keun [Korea Institute of Machinery and Materials, Taejon (Korea)

1999-03-01

In order to enhance testing and evaluation technologies, which is one of the main works of the Chanwon branch of KIMM(Korea Institute of Machinery and Materials), in addition to the present work scope of the testing and evaluation in the industrial facilities such as petroleum and chemical, plants, the qualification technologies of the equipments important to safety used in the key industrial facilities such as nuclear power plants should be localized: Equipments for testing and evaluation is to be set up and the related technologies must be developed. In the first year of this study, of vibration aging qualification technologies of equipments important to safety used in nuclear power plants have been performed. (author). 27 refs., 81 figs., 17 tabs.

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)

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)

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

Response of equipment to aircraft impact

International Nuclear Information System (INIS)

Wolf, J.P.; Bucher, K.M.; Skrikerud, P.E.

1977-01-01

The loading case of an aircraft crashing onto certain safety-relevant buildings of a nuclear-power plant has recently become, in certain countries, as important as that of the safe-shutdown earthquake. Although its probability of occurrence is substantially smaller than that of the SSE, the analysis is justified as an aircraft can also be regarded as representative of other shock loading cases. For the purpose of design it is convenient to distinguish between the response of the actual structure and that of the equipment. The former, which consists of global stress resultants and of local effects such as spalling, scabbing, penetration and perforation of the concrete of the structure, is quite well understood and is not examined in this paper. The latter, caused by propagating shock waves, has, rather surprisingly, received little attention, although, for earthquake excitation, the analysis of equipment is routinely performed. The high accelerations experienced by the equipment are calculated and the corresponding floor-response spectra show that the airplane crash is dominant in the high-frequency range, when compared to the effect of earthquakes. As the aircraft impacts on a small area of the structure only, and the corresponding function of the load versus time exhibits a substantial content of high frequency, the modelling of the structure has to be rather detailed. Several thousand dynamic degrees of freedom in a finite-element idealization result, when analyzing e.g. a typical reactor auxiliary building, the familiar lumped-beam models used successfully for earthquake analysis can obviously not be chosen. (Auth.)

Equipment Obsolescence Management Program

Energy Technology Data Exchange (ETDEWEB)

Redmond, J.

2014-07-01

Nuclear Power Plant (NPP) Operators are challenged with securing reliable supply channels for safety related equipment due to equipment obsolescence. Many Original Equipment Manufacturers (OEMs) have terminated production of spare parts and product life-cycle support. The average component life cycles are much shorter than the NPP design life, which means that replacement components and parts for the original NPP systems are not available for the complete design life of the NPPs. The lack or scarcity of replacement parts adversely affects plant reliability and ultimately the profitability of the affected NPPs. This problem is further compounded when NPPs pursue license renewal and approval for plant-life extension. A reliable and predictable supply of replacement co components is necessary for NPPs to remain economically competitive and meet regulatory requirements and guidelines. Electrical and I and C components, in particular, have short product life cycles and obsolescence issues must be managed pro actively and not reactively in order to mitigate the risk to the NPP to ensure reliable and economic NPP operation. (Author)

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

Estimation of the cost of electro-mechanical equipment for small hydropower plants – review and comparison of methods

Directory of Open Access Journals (Sweden)

Lipiński Seweryn

2017-01-01

Full Text Available The estimate of the cost of electro-mechanical equipment for new small hydropower plants most often amounts to about 30-40% of the total budget. In case of modernization of existing installations, this estimation represents the main cost. This matter constitutes a research problem for at least few decades. Many models have been developed for that purpose. The aim of our work was to collect and analyse formulas that allow estimation of the cost of investment in electro-mechanical equipment for small hydropower plants. Over a dozen functions were analysed. To achieve the aim of our work, these functions were converted into the form allowing their comparison. Then the costs were simulated with respect to plants’ powers and net heads; such approach is novel and allows deeper discussion of the problem, as well as drawing broader conclusions. The following conclusions can be drawn: significant differences in results obtained by using various formulas were observed; there is a need for a wide study based on national investments in small hydropower plants that would allow to develop equations based on local data; the obtained formulas would let to determinate the costs of modernization or a new construction of small hydropower plant more precisely; special attention should be payed to formulas considering turbine type.

Equipment inspection function; A funcao inspecao de equipamentos

Energy Technology Data Exchange (ETDEWEB)

Bressan, Edemir [PETROBRAS (Brazil). Refinaria Alberto Pasqualini

1994-04-01

Development of the safety engineering and equipment inspection policy in petroleum industry, with emphasis to the refineries equipment, are discussed. Presentation of the accidents percentages in petroleum plants under international researches, with details of involved equipment, kind of accident, plant and causes are reported. The start and actual stage of the safety inspection program in the PETROBRAS are also presented 5 refs., 3 figs.

The Ural Electrochemical Integrated Plant Process for Managing Equipment Intended for Nuclear Material Protection, Control and Accounting System Upgrades

International Nuclear Information System (INIS)

Yuldashev, Rashid; Nosov, Andrei; Carroll, Michael F.; Garrett, Albert G.; Dabbs, Richard D.; Ku, Esther M.

2008-01-01

Since 1996, the Ural Electrochemical Integrated Plant (UEIP) located in the town of Novouralsk, Russia, (previously known as Sverdlovsk-44) and the United States Department of Energy (U.S. DOE) have been cooperating under the Nuclear Material Protection, Control and Accounting (MPC and A) Program. Because UEIP is involved in the processing of highly enriched uranium (HEU) into low enriched uranium (LEU), and there are highly enriched nuclear materials on its territory, the main goal of the MPC and A cooperation is to upgrade those systems that ensure secure storage, processing and transportation of nuclear materials at the plant. UEIP has completed key upgrades (equipment procurement and installation) aimed at improving MPC and A systems through significant investments made by both the U.S. DOE and UEIP. These joint cooperative efforts resulted in bringing MPC and A systems into compliance with current regulations, which led to nuclear material (NM) theft risk reduction and prevention from other unlawful actions with respect to them. Upon the U.S. MPC and A project team's suggestion, UEIP has developed an equipment inventory control process to track all the property provided through the MPC and A Program. The UEIP process and system for managing equipment provides many benefits including: greater ease and efficiency in determining the quantities, location, maintenance and repair schedule for equipment; greater assurance that MPC and A equipment is in continued satisfactory operation; and improved control in the development of a site sustainability program. While emphasizing UEIP's equipment inventory control processes, this paper will present process requirements and a methodology that may have practical and helpful applications at other sites.

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.

Comparison of EMI/RFI requirements to qualify the equipment for nuclear power plant. (RG 1.180, EPRI TR 102323, IEC 62003 and GB/T 11684)

Energy Technology Data Exchange (ETDEWEB)

Jang, Tae Heon [Korea Testing Laboratory, Ansan (Korea, Republic of); Kim, Jong Seog [Research Institute of Korea Electric Power Corporation, Daejeon (Korea, Republic of); Seo, Jeong Ho; Cho, Kyoung Youn [Korea Electric Association, Seoul (Korea, Republic of)

2011-05-15

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). In some reports for nuclear power plant (NPP), electromagnetic interference (EMI), radio frequency interference (RFI), and power surges have been identified as environmental conditions that can affect the performance of safety-related electrical equipment. There are mainly two reference guides for applying to qualify EMI/RFI requirements of the equipment used in a NPP: US NRC RG 1.180 and EPRI TR 102323. Recently, IEC published the standard for the equipment in the NPP, IEC 62003. This paper has compared the requirements of these, including comparing of the requirement of the Chinese national standard, GB/T 11684

Comparison of EMI/RFI requirements to qualify the equipment for nuclear power plant. (RG 1.180, EPRI TR 102323, IEC 62003 and GB/T 11684)

International Nuclear Information System (INIS)

Jang, Tae Heon; Kim, Jong Seog; Seo, Jeong Ho; Cho, Kyoung Youn

2011-01-01

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). In some reports for nuclear power plant (NPP), electromagnetic interference (EMI), radio frequency interference (RFI), and power surges have been identified as environmental conditions that can affect the performance of safety-related electrical equipment. There are mainly two reference guides for applying to qualify EMI/RFI requirements of the equipment used in a NPP: US NRC RG 1.180 and EPRI TR 102323. Recently, IEC published the standard for the equipment in the NPP, IEC 62003. This paper has compared the requirements of these, including comparing of the requirement of the Chinese national standard, GB/T 11684

30 CFR 77.403-1 - Mobile equipment; rollover protective structures (ROPS).

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Mobile equipment; rollover protective structures (ROPS). 77.403-1 Section 77.403-1 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR COAL MINE SAFETY AND HEALTH MANDATORY SAFETY STANDARDS, SURFACE COAL MINES AND SURFACE...

30 CFR 77.403 - Mobile equipment; falling object protective structures (FOPS).

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Mobile equipment; falling object protective structures (FOPS). 77.403 Section 77.403 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR COAL MINE SAFETY AND HEALTH MANDATORY SAFETY STANDARDS, SURFACE COAL MINES AND SURFACE WORK...

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.

Influence of Cast Iron Structure on the Glassmold Equipment Operational Defects

Directory of Open Access Journals (Sweden)

I. O. Leushin

2015-01-01

Full Text Available The growing demand for glass packaging contributes to the increase in production capacity of glass-container plants. Their equipment (cast iron glass-forming sets operates in continuous mode under complex cyclic thermal loads, which lead to the formation of operational defects on the working surfaces of details: graphite falling, cracks, oxidation, etc. Particular influence on the formation of these defects renders the microstructure of the material at the time of installation of details on the line.The article identifies the causes for formation of operational defects, formulates the ways to remedy them and prevent their occurrence.The authors studied details made from grey cast iron with flake and spherical forms of graphite. It is found that in the process of exploitation of the material is greatly reducing its hardness, strength, resistance to oxidation through of graphitization processes, chemical interaction of glass and iron, shock loads working edges. It is proved that the choice of initial microstructure of cast iron (the metal base, the graphite form, the presence of structural-free cementite exercises a determining influence on the durability of the mold tooling. The article proposes differential (layered arrangement of the graphite phase of cast iron in the alloy matrix (ferrite. This arrangement of high-carbon phase can simultaneously increase the thermal and oxidation resistance of the material. The formation of a layered structure of iron is produced by the intensification of the processes of alloying, modifying and directional freezing the melt.These data can be used to select the material of details by manufacturers glass-molds tooling.

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

Equipment Reliability Process in Krsko NPP

International Nuclear Information System (INIS)

Gluhak, M.

2016-01-01

To ensure long-term safe and reliable plant operation, equipment operability and availability must also be ensured by setting a group of processes to be established within the nuclear power plant. Equipment reliability process represents the integration and coordination of important equipment reliability activities into one process, which enables equipment performance and condition monitoring, preventive maintenance activities development, implementation and optimization, continuous improvement of the processes and long term planning. The initiative for introducing systematic approach for equipment reliability assuring came from US nuclear industry guided by INPO (Institute of Nuclear Power Operations) and by participation of several US nuclear utilities. As a result of the initiative, first edition of INPO document AP-913, 'Equipment Reliability Process Description' was issued and it became a basic document for implementation of equipment reliability process for the whole nuclear industry. The scope of equipment reliability process in Krsko NPP consists of following programs: equipment criticality classification, preventive maintenance program, corrective action program, system health reports and long-term investment plan. By implementation, supervision and continuous improvement of those programs, guided by more than thirty years of operating experience, Krsko NPP will continue to be on a track of safe and reliable operation until the end of prolonged life time. (author).

Mechanical (turbines and auxiliary equipment)

CERN Document Server

Sherry, A; Cruddace, AE

2013-01-01

Modern Power Station Practice, Volume 3: Mechanical (Turbines and Auxiliary Equipment) focuses on the development of turbines and auxiliary equipment used in power stations in Great Britain. Topics covered include thermodynamics and steam turbine theory; turbine auxiliary systems such as lubrication systems, feed water heating systems, and the condenser and cooling water plants. Miscellaneous station services, and pipework in power plants are also described. This book is comprised of five chapters and begins with an overview of thermodynamics and steam turbine theory, paying particular attenti

Life Cycle Management Managing the Aging of Critical Nuclear Plant Components

International Nuclear Information System (INIS)

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

2002-01-01

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

Changing nature of equipment and parts qualification

International Nuclear Information System (INIS)

Bucci, R.M.

1988-01-01

Ideally, the original supplier of a piece of nuclear safety-related equipment has performed a qualification program and will continue to support that equipment throughout the lifetime of the nuclear power plants in which in equipment is installed. The supplier's nuclear quality assurance program will be maintained and he will continue to offer all necessary replacement parts. These parts will be identical to the original parts, certified to the original purchase order requirements, and the parts will be offered at competitive prices. Due to the changing nature of the nuclear plant equipment market, however, one or more of those ideal features are frequently unavailable when safety-related replacement equipment or parts are required. Thus, the process of equipment and parts qualification has had to adjust in order to ensure obtaining qualified replacements when needed. This paper presents some new directions taken in the qualification of replacement equipment and parts to meet changes in the marketplace

Development priorities for non-destructive examination of concrete structures in nuclear plant

International Nuclear Information System (INIS)

1998-01-01

The objective of this report is to provide a basis for assessing development priorities for NDE of safety related concrete structures in nuclear plants, taking account of both the benefit and the cost of potential developments in NDE techniques. An OECD/NEA Workshop which considered the requirements for NDE of safety related concrete structures was held in the UK on 12 November 97. NDE techniques have the potential to satisfy at least some of the needs of the nuclear industry. NDE techniques have been used successfully on a variety of reinforced and post-tensioned concrete structures, notably highway and reservoir structures. However, there is limited experience of their use to evaluate typical nuclear safety related structures having thick sections, steel liners or access to one side only. There is a general lack of confidence in the techniques because there is very little independent advice on their applicability, capability, accuracy and reliability. The information obtained by techniques such as RADAR, ultrasonics, stress wave and radiography appears qualitative rather than quantitative and there is concern that NDE procedures lack the necessary qualification to permit their use on safety critical structures. There is no authoritative international guidance or standard for NDE of concrete structures. NDE of concrete structures is often based upon equipment developed for other materials and technologies, eg. examination of steel, evaluation of ground conditions. Other industries are developing equipment specifically for civil engineering applications and at the recent OECD workshop a number of relevant national and European programmes were identified. The nuclear industry maintain its awareness of developments and should seek to influence the development of equipment. The quantification of the capabilities of NDE techniques is seen as a priority area for development. The provision of authoritative documentation in the form of reports and Standards is desirable

Experiences in seismic upgrading of equipment and structures in Kozloduy nuclear power plant (440 WWER-PWR)

International Nuclear Information System (INIS)

Ordonez Villalobos, A.

1993-01-01

Within the framework of the 'Emergency programme for Nuclear Safety of Kozloduy NPP' it has been concluded that the increase in seismic safety of a NPP can be achieved by upgrading the key equipment in a cost effective way. Essential and vulnerable equipment has to be identified. Seismic capacity should be evaluated base don realistic state of the art criteria. Seismic review teams ef experienced engineers should conduct planned walk-downs in order to propose effective upgrading solutions. Team work of plan engineers and construction engineers would enhance the effectiveness of the solutions. It is recommended that all the participants be motivated and have a clear understanding of the objectives of the upgrading

On results of tests of thermal insulation structural fragments for in-vessel equipment and pipelines of the VG-400 plant on vibrational and acoustic loads

International Nuclear Information System (INIS)

Ledenko, S.A.; Andreev, V.A.; Mirenkov, A.F.; Zakharov, V.A.; Suvorov, V.E.; Prokimnov, V.V.

1990-01-01

Results of vibrostrength and acoustic fatigue tests of the fragments of thermal insulation for in-vessel equipment and pipelines of the VG-400 reactor are presented. The insulation structure is based on the insulation layer made of steel foil and carbon materials. Weak points in the insulation structure, namely - the welded joints of stiffening ribs - are detected in the course of testing. A conclusion is made on the possibility of vibrational test substitution for the acoustic ones

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)

Analysis of the differences between the accounting and tax treatment for items of property, plant and equipment: The Peruvian case

Directory of Open Access Journals (Sweden)

Oscar Alfredo Díaz Becerra

2012-12-01

Full Text Available This research work aims principally to make an analysis showing differences between the measurement and the recognition of items of property, plant and equipment. It focuses on the differences caused by existing differences between the treatment settled in the accounting standards and the one settled in the tax regulations related to Corporate Income Tax, for Peruvian case.A review of the related accounting standards and the standards established in the Peruvian Income Tax Law and its regulations have been considered in the current work. Thus, we are going to identify the main differences arising from the application of both standards regarding items of property, plant and equipment.Finally, we present the main conclusions drawn from this research.

Remote maintenance lessons learned on prototypical reprocessing equipment

International Nuclear Information System (INIS)

Kring, C.T.; Schrock, S.L.

1990-01-01

A major objective of the Consolidated Fuel Reprocessing Program at the Oak Ridge National Laboratory is to develop and demonstrate the technology required to reprocess spent nuclear fuel. The Fuel Recycle Division, over the past 16 years, has undertaken this objective by designing and testing prototypical hardware representing essentially every major equipment item currently included in most fuel reprocessing plant conceptual designs. These designs are based on total remote maintenance to increase plant availability and reduce radiation exposure to plant operators. The designs include modular equipment to facilitate maintainability and the remote manipulation necessary to accomplish maintenance tasks. Prototypic equipment has been installed and tested in a cold mock-up of a reprocessing hot cell, called the remote operations and maintenance demonstration facility. The applied maintenance concept utilizes the dexterity and mobility of bridge-mounted, force-reflecting servomanipulators. Prototypic processing equipment includes a remote disassembly system, a remote shear system, a rotary dissolver, a remote automated sampler system, removable equipment racks to support chemical process equipment items, and the advanced servomanipulators. Each of these systems and a brief description of functions are discussed

Shaking table testing of electrical equipment in Argentina

International Nuclear Information System (INIS)

Carmona, J.S.; Zabala, F.; Santalucia, J.; Sisterna, C.; Magrini, M.; Oldecop, L.

1995-01-01

This paper describes the testing facility, the methodology applied and the results obtained in the seismic qualification tests of different types of electric equipment. These tests were carried out on a shaking table that was developed and built at the Earthquake Research Institute of the National University of San Juan, Argentine. The equipment tested consist of 500 KV and 132 KV current transformers, a 500 KV voltage transformer, a 145 KV disconnecter and a relay cabinet. The acceleration response of the tested equipment was measured at several locations distributed along its height, and strains were measured at critical points by strain gauges cemented on the base of the porcelain insulator. All the information was recorded with a data acquisition system at a sampling rate of 200 times per second in each channel. The facility developed at this Institute is the largest one in operation in Argentina at present and the equipment tested is the highest, heaviest and more slender one which has been seismically qualified on a shaking table in this country. These tests have been a valuable experience in the field of structural dynamic testing applied to equipment of hydroelectric and nuclear power plants. (author)

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

Application of ELD and load forecast in optimal operation of industrial boiler plants equipped with thermal stores

International Nuclear Information System (INIS)

Cao Jiacong

2007-01-01

Optimal operation of industrial boiler plants with objects of high energy efficiency and low fuel cost is still well worth investigating when energy problem becomes a world's concern, for there are a great number of boiler plants serving industries. The optimization of operation is a measure that is less expensive and easier to carry out than many other measures. Economic load dispatch (ELD) is an effective approach to optimal operation of industrial boiler plants. In the paper a newly developed method referred to as the method of minimum-departure model (MDM) is used in the ELD for boiler plants. It is more convenient for carrying out ELD when boiler plants are equipped with thermal energy stores that usually adopt the working mode of optimal segmentation of a daily load curve. In the case of industrial boiler plants, ELD needs a prerequisite, viz., the accurate load forecast, which is performed using artificial neural networks in this paper. A computer program for the optimal operation was completed and applied to an example, which results the minimum daily fuel cost of the whole boiler plant

Engineerig of structural modifications for operating nuclear plants

International Nuclear Information System (INIS)

Duffy, T.J.; Gazda, P.A.

1983-01-01

The engineering of structural modifications for operating nuclear plants offers many challenges in the areas of scheduling of work, field adjustments, and engineering staff planning. The scheduling of structural modification work for operating nuclear plants is normally closely tied to planned or unplanned outages of the plant. Coordination between the structural engineering effort, the operating plant staff, and the contractor who will be performing the modifications is essential to ensure that all work can be completed within the allotted time. Due to the inaccessibility of areas in operating plants or the short time available to perform the structural engineering in the case of an unscheduled outage, field verification of a design is not always possible prior to initiating the construction of the modification. This requires the structural engineer to work closely with the contractor to promptly resolve problems due to unanticipated interferences or material procurement that may arise during the course of construction. The engineering staff planning for structural modifications at an operating nuclear plant must be flexible enough to permit rapid response to the common 'fire drills', but controlled enough to assure technically correct designs and minimize the expenditure of man-hours and resulting engineering cost. (orig.)

Loosening and damage mechanism of thread-joined structures in nuclear power equipment

International Nuclear Information System (INIS)

Tang Hui

1999-01-01

The author proposes a loosening mechanism of thread-joined structures under vibrate environments in the nuclear power equipment and structures, which is on the base of the macro and imperceptible-mechanics analysis. It has answered the problems on the seizing, the adhesive wearing, the generation of cracks, the thread-tooth fracture. So it has a conclusion that the loosening of thread-joined structures is essential trend, in other words, the locking property of thread-pair is failure under vibrate environments

Plant lessons: exploring ABCB functionality through structural modeling

Directory of Open Access Journals (Sweden)

Aurélien eBailly

2012-01-01

Full Text Available In contrast to mammalian ABCB1 proteins, narrow substrate specificity has been extensively documented for plant orthologs shown to catalyze the transport of the plant hormone, auxin. Using the crystal structures of the multidrug exporters Sav1866 and MmABCB1 as templates, we have developed structural models of plant ABCB proteins with a common architecture. Comparisons of these structures identified kingdom-specific candidate substrate-binding regions within the translocation chamber formed by the transmembrane domains of ABCBs from the model plant Arabidopsis. These results suggest an early evolutionary divergence of plant and mammalian ABCBs. Validation of these models becomes a priority for efforts to elucidate ABCB function and manipulate this class of transporters to enhance plant productivity and quality.

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)

Management methodology for pressure equipment

Science.gov (United States)

Bletchly, P. J.

Pressure equipment constitutes a significant investment in capital and a major proportion of potential high-risk plant in many operations and this is particularly so in an alumina refinery. In many jurisdictions pressure equipment is also subject to statutory regulation that imposes obligations on Owners of the equipment with respect to workplace safety. Most modern technical standards and industry codes of practice employ a risk-based approach to support better decision making with respect to pressure equipment. For a management system to be effective it must demonstrate that risk is being managed within acceptable limits.

Equipment reliability improvement process; implementation in Almaraz NPP and Trillo NPP

International Nuclear Information System (INIS)

Risquez Bailon, Aranzazu; Gutierrez Fernandez, Eduardo

2010-01-01

The Equipment Reliability Improvement Process (INPO AP-913) is a non-regulatory process developed by the US Nuclear Industry for improving Plants Availability. This Process integrates and coordinates a broad range of equipment reliability activities into one process, performed by the Plant in a non-centralized way. The integration and coordination of these activities will allow plant personnel to evaluate the trends of important station equipment, develop and implement long-term equipment health plans, monitor equipment performance and condition, and make adjustments to preventive maintenance tasks and frequencies based on equipment operating experience, if necessary, arbitrating operational and design improvements, to reach a Failure-free Operation. This paper describes the methodology of Equipment Reliability Improvement Process, being focused on main aspects of the implementation process, relating to the scope and establishment of an Equipment Reliability Monitoring Plan, which should include and complement the existing mechanisms and organizations in the Plant to monitor the condition and performance of the equipments, with the common aim of achieving an operation free of failures. The paper will describe the tools that Iberdrola Ingenieria has developed to support the implementation and monitoring of the Equipment Reliability Improvement Process, as well as the results and lessons learned from its implementation in Almaraz NPP and Trillo NPP. (authors)

SPECIFIC DEGRADATION STRUCTURE FEATURES AND MECHANICAL PROPERTIES OF FURNACE AND HEAT POWER EQUIPMENT ELEMENTS AFTER LONG-TERM OPERATION

Directory of Open Access Journals (Sweden)

F. I. Panteleenko

2012-01-01

Full Text Available The paper presents results of investigations on structure and mechanical properties of technological equipment elements made of heat-resistant steels. A scale of chrome and molybdenum steel microstructure degradation based on evaluation of  coagulated carbide size and material mechanical properties (a point from 0-operation without time limits, up to 4-operation prohibition has been proposed in the paper. It has been  established that an analysis of  steel microstructure directly on equipment elements by means of a portable microscope is an efficient express method for evaluation of equipment condition and structures due to control of material structure degradation rate of a diagnosed object.

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)

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

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

18 CFR 367.9350 - Account 935, Maintenance of structures and equipment.

Science.gov (United States)

2010-04-01

... POWER ACT AND NATURAL GAS ACT Operation and Maintenance Expense Chart of Accounts § 367.9350 Account 935, Maintenance of structures and equipment. This account must include materials used and expenses incurred in the maintenance of property owned, the cost of which is included in accounts 390 through 399 (§§ 367.3900 through...

Seismic review of existing nuclear power plants

International Nuclear Information System (INIS)

Yanev, P.I.; Mayes, R.L.; Jones, L.R.

1975-01-01

Because of developments in the fields of earthquake and structural engineering over the last two decades, the codes, standards and design criteria for Nuclear Power Plants and other critical structures have changed substantially. As a result, plants designed only a few years ago do not satisfy the requirements for new plants. Accordingly, the Regulatory Agencies are requiring owners of older Nuclear Power Plants to re-qualify the plants seismically, using codes, standards, analytical techniques and knowledge developed in recent years. Seismic review consists of three major phases: establishing the design and performance criteria, re-qualifying the structures, and re-qualifying the equipment. The authors of the paper have been recently involved in the seismic review of existing nuclear power plants in the United States. This paper is a brief summary of their experiences

47 CFR 36.126 - Circuit equipment-Category 4.

Science.gov (United States)

2010-10-01

... separating property associated with special services, circuit equipment included in Categories 4.12 (other... Equipment Excluding Wideband—Category 4.13—The cost of Circuit Equipment associated with exchange line plant... 47 Telecommunication 2 2010-10-01 2010-10-01 false Circuit equipment-Category 4. 36.126 Section 36...

Progress on EPRI electrical equipment qualification research

International Nuclear Information System (INIS)

Sliter, G.E.

1983-01-01

The objective of EPRI's electrical equipment qualification research program is to provide technical assistance to utilities in meeting nuclear plant safety requirements in a manner consistent with the state of the art. This paper reports progress on several research projects including: radiation effects studies, which compile data on degradation of organic materials in electrical equipment exposed to operational and accident radiation doses; the Equipment Qualification Data Bank, which is a remotely accessible computer system for disseminating qualification information on in-plant equipment, seismic data, and materials data; an aging/seismic correlation program, which is providing test data showing that, in many cases, age degradation has a negligibly small effect on the performance of electrical components under seismic excitation; a review of condition monitoring techniques, which has identified surveillance methods for measuring key performance parameters that have the potential for predicting remaining equipment life; and large-scale hydrogen burn equipment response tests, which are providing data to assess the ability of equipment to remain functional during and after hydrogen burning in postulated degraded core accidents

40 CFR 264.114 - Disposal or decontamination of equipment, structures and soils.

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 25 2010-07-01 2010-07-01 false Disposal or decontamination of equipment, structures and soils. 264.114 Section 264.114 Protection of Environment ENVIRONMENTAL PROTECTION... TREATMENT, STORAGE, AND DISPOSAL FACILITIES Closure and Post-Closure § 264.114 Disposal or decontamination...

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)

Study on intelligence fault diagnosis method for nuclear power plant equipment based on rough set and fuzzy neural network

International Nuclear Information System (INIS)

Liu Yongkuo; Xia Hong; Xie Chunli; Chen Zhihui; Chen Hongxia

2007-01-01

Rough set theory and fuzzy neural network are combined, to take full advantages of the two of them. Based on the reduction technology to knowledge of Rough set method, and by drawing the simple rule from a large number of initial data, the fuzzy neural network was set up, which was with better topological structure, improved study speed, accurate judgment, strong fault-tolerant ability, and more practical. In order to test the validity of the method, the inverted U-tubes break accident of Steam Generator and etc are used as examples, and many simulation experiments are performed. The test result shows that it is feasible to incorporate the fault intelligence diagnosis method based on rough set and fuzzy neural network in the nuclear power plant equipment, and the method is simple and convenience, with small calculation amount and reliable result. (authors)

Plutonium Immobilization Can Loading Equipment Review

International Nuclear Information System (INIS)

Kriikku, E.; Ward, C.; Stokes, M.; Randall, B.; Steed, J.; Jones, R.; Hamilton, L.

1998-05-01

This report lists the operations required to complete the Can Loading steps on the Pu Immobilization Plant Flow Sheets and evaluates the equipment options to complete each operation. This report recommends the most appropriate equipment to support Plutonium Immobilization Can Loading operations

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)

Principles of commercially available pretreatment and feeding equipment for baled biomass

Energy Technology Data Exchange (ETDEWEB)

Koch, T. [Thomas Koch Energi, Vanloese (Denmark); Hummelshoej, R.M. [COWIconsult, Lyngby (Denmark)

1993-12-31

During the last 15 years, there has been a growing interest in utilizing waste biomass for energy production in Denmark. Since 1990, it has been unlawful to burn surplus straw on open land. Before the year 2000, it is intended to utilize most of the 2--3 million tons of surplus straw as an energy resource. The type of plants that were built in the beginning were combustion plants for district heating. The feeding equipment for these plants has been developed to an acceptable standard. Later, combustion plants for combined heat and power production based on a steam turbine were introduced. This type of plant demands a much greater continuity in the fuel flow, and the consequences of minor discontinuities are to be dropped from the grid. Gasification and pyrolysis demands a high sealing ability of the feeding equipment, because of the explosive and poisonous gas in the plant and a need for a very high continuity in the fuel feed. The first plants were built with the equipment and experiences from the farming industries, which have a long tradition in working with biomass-handling. The experiences gained with this type of equipment were not very promising, and in the early eighties, a more industrial type of biomass-handling equipment was developed. This paper presents the principles of the heavy-duty biomass pretreatment and feeding equipment that was commercially available in Denmark in May, 1993.

Human performance for the success of equipment reliability programs

International Nuclear Information System (INIS)

Woodcock, J.

2007-01-01

Human performance is a critical element of programs directed at equipment reliability. Reliable equipment performance requires broad support from all levels of plant management and throughout all plant departments. Experience at both nuclear power plants and fuel manufacturing plants shows that human performance must be addressed during all phases of program implementation from the beginning through the establishment of a living, on-going process. At the beginning, certain organizational and management actions during the initiation of the program set the stage for successful adoption by station personnel, leading to more rapid benefits. For the long term, equipment reliability is a living process needed throughout the lifetime of a station, a program which must be motivated and measured. Sustained acceptance and participation by the plant personnel is a requirement, and culture is a key ingredient. This paper will provide an overview of key human performance issues to be considered, using the application of the INPO AP-913 Equipment Reliability Guideline as a basis and gives some best practices for training, communicating and implementing programs. The very last part includes ways to tell if the program is effective

Structural experiences at the Kewaunee Nuclear Power Plant

International Nuclear Information System (INIS)

Setlur, A.V.

1983-01-01

This paper discusses the original structural and geotechnical design and subsequent structural experience at the Kewaunee Nuclear Power Plant. The original design of the 535 MWe Westinghouse two loop PWR nuclear plant operated by Wisconsin Public Service Corporation, was started in 1967 and was completed in 1974 when the unit was put into commercial operation. Since 1974 a number of changes in the regulations and additional requirements have been imposed on operating reactors. The paper traces the influence of the original plant criteria on the backfit evaluations and the minimal physical changes required in the plant's structures and components to comply with the new requirements. In addition, the unique design features and construction challenges of the original design are discussed. Kewaunee Nuclear Power Plant has had one of the best operating performance records in the world. Also, the exposure to radiation for plant personnel and radioactive waste generation has been significantly lower than the average. This has been achieved by a conscientious team effort of all parties involved. Some of the more significant structural design features contributing to the excellent performance is detailed in this paper. (orig.)

Equipment qualification testing - a practical approach

International Nuclear Information System (INIS)

Davies, G.A.; McDougall, R.I.; Poirier, M.P.

1996-01-01

When nuclear safety equipment is credited with a Required Safety Function it must properly perform that function to facilitate safe control and/or shutdown of the plant during a design basis accident. When such equipment is required to be environmentally (EQ) and/or seismically qualified (SQ) for safety related use in CANDU nuclear power plants, the preferred method of qualification is by type testing. The qualification testing process requires that the test specimen equipment be subjected to the aging stressors associated with the normal service conditions that it would experience during it's required qualified (or service) life. Following the aging process, the test specimen is in a condition representative of that in which it would be at the end of its service life in the plant. The test specimen is then subjected to a simulated accident during which it must satisfy performance requirements thereby demonstrating that it can perform its required safety function. The performance requirements specified for the qualification testing must be designed to ensure that satisfactory performance of the safety function is demonstrated during the qualification program. This paper provides descriptions of practical methods used in the deriving and satisfying of relevant performance requirements during the qualification testing of safety related equipment. (author)

The point of view of thermal equipment users; Le point de vue des gestionnaires d`equipements thermiques

Energy Technology Data Exchange (ETDEWEB)

Barroyer, P. [Compagnie Generale de Chauffe, 59 - Saint Andre Lez Lille (France)

1997-12-31

The influence of new pollution regulations in France on the operation of thermal equipment for central heating systems or industrial heat process systems, is examined. The main French regulations concerning air pollution control and energy rational consumption are reviewed, and their effects on the design, equipment, operation and costs of heat plants are discussed: impacts of the decree on upgrading and disposal of fossil fuel ashes, the decree on special protection zone (large cities), the clean air law, the compulsory declaration for classified combustion plants and limit air pollution emission levels

Evaluation of high frequency ground motion effects on the seismic capacity of NPP equipments

International Nuclear Information System (INIS)

Choi, In Kil; Seo, Jeong Moon; Choun, Young Sun

2003-04-01

In this study, the uniform hazard spectrum for the example Korean nuclear power plants sites were developed and compared with various response spectra used in past seismic PRA and SMA. It shows that the high frequency ground motion effects should be considered in seismic safety evaluations. The floor response spectra were developed using the direct generation method that can develop the floor response spectra from the input response spectrum directly with only the dynamic properties of structures obtained from the design calculation. Most attachment of the equipments to the structure has a minimum distortion capacity. This makes it possible to drop the effective frequency of equipment to low frequency before it is severely damaged. The results of this study show that the high frequency ground motion effects on the floor response spectra were significant, and the effects should be considered in the SPRA and SMA for the equipments installed in a building. The high frequency ground motion effects are more important for the seismic capacity evaluation of functional failure modes. The high frequency ground motion effects on the structural failure of equipments that attached to the floor by welding can be reduced by the distortion capacity of welded anchorage

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

Commissioning quality assurance for nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

NONE

1986-09-01

This standard contains the requirements for the quality assurance program applicable to the commissioning phase of a nuclear power plant. This standard embodies the relevant quality assurance requirements of CSA Standard CAN3-N286.0, and is the governing Standard for commissioning quality assurance activities in the event of any conflicting requirements. This Standard applies to the commissioning of safety-related equipment, systems, and structures as identified by the owner. It may be applied to other equipment, systems, and structures at the discretion of the owner. 1 fig.

Commissioning quality assurance for nuclear power plants

International Nuclear Information System (INIS)

1986-09-01

This standard contains the requirements for the quality assurance program applicable to the commissioning phase of a nuclear power plant. This standard embodies the relevant quality assurance requirements of CSA Standard CAN3-N286.0, and is the governing Standard for commissioning quality assurance activities in the event of any conflicting requirements. This Standard applies to the commissioning of safety-related equipment, systems, and structures as identified by the owner. It may be applied to other equipment, systems, and structures at the discretion of the owner. 1 fig

AP1000{sup TM} plant modularization

Energy Technology Data Exchange (ETDEWEB)

Cantarero L, C.; Demetri, K. J. [Westinghouse Electric Co., 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States); Quintero C, F. P., E-mail: cantarc@westinghouse.com [Westinghouse Electric Spain, Padilla 17, 28006 Madrid (Spain)

2016-09-15

The AP1000{sup TM} plant is an 1100 M We pressurized water reactor (PWR) with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance and safety. Modules are used extensively in the design of the AP1000 plant nuclear island. The AP1000 plant uses modern, modular-construction techniques for plant construction. The design incorporates vendor-designed skids and equipment packages, as well as large, multi-ton structural modules and special equipment modules. Modularization allows traditionally sequential construction tasks to be completed simultaneously. Factory-built modules can be installed at the site in a planned construction schedule. The modularized AP1000 plant allows many more construction activities to proceed in parallel. This reduces plant construction calendar time, thus lowering the costs of plant financing. Furthermore, performing less work onsite significantly reduces the amount of skilled field-craft labor, which costs more than shop labor. In addition to labor cost savings, doing more welding and fabrication in a factory environment raises the quality of work, allowing more scheduling flexibility and reducing the amount of specialized tools required onsite. The site layout for the AP1000 plant has been established to support modular construction and efficient operations during construction. The plant layout is compact, using less space than previous conventional plant layouts. This paper provides and overview of the AP1000 plant modules with an emphasis on structural modules. Currently the Westinghouse AP1000 plant has four units under construction in China and four units under construction in the United States. All have shown successful fabrication and installation of various AP1000 plant modules. (Author)

AP1000"T"M plant modularization

International Nuclear Information System (INIS)

Cantarero L, C.; Demetri, K. J.; Quintero C, F. P.

2016-09-01

The AP1000"T"M plant is an 1100 M We pressurized water reactor (PWR) with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance and safety. Modules are used extensively in the design of the AP1000 plant nuclear island. The AP1000 plant uses modern, modular-construction techniques for plant construction. The design incorporates vendor-designed skids and equipment packages, as well as large, multi-ton structural modules and special equipment modules. Modularization allows traditionally sequential construction tasks to be completed simultaneously. Factory-built modules can be installed at the site in a planned construction schedule. The modularized AP1000 plant allows many more construction activities to proceed in parallel. This reduces plant construction calendar time, thus lowering the costs of plant financing. Furthermore, performing less work onsite significantly reduces the amount of skilled field-craft labor, which costs more than shop labor. In addition to labor cost savings, doing more welding and fabrication in a factory environment raises the quality of work, allowing more scheduling flexibility and reducing the amount of specialized tools required onsite. The site layout for the AP1000 plant has been established to support modular construction and efficient operations during construction. The plant layout is compact, using less space than previous conventional plant layouts. This paper provides and overview of the AP1000 plant modules with an emphasis on structural modules. Currently the Westinghouse AP1000 plant has four units under construction in China and four units under construction in the United States. All have shown successful fabrication and installation of various AP1000 plant modules. (Author)

Seismic re-evaluation of Heavy Water Plant, Kota

International Nuclear Information System (INIS)

Parulekar, Y.M.; Reddy, G.R.; Vaze, K.K.; Kushwaha, H.S.

2003-10-01

This report deals with seismic re-evaluation of Heavy Water Plant, Kota. Heavy Water Plant, Kota handles considerable amount of H 2 S gas, which is very toxic. During the original design stage as per IS 1893-1966 seismic coefficient for zone-I was zero. Therefore earthquake and its effects were not considered while designing the heavy water plant structures. However as per IS 1893 (1984) the seismic coefficient for zone-I is 0.01 g. Hence seismic re-evaluation of various structures of the heavy water plant is carried out. Analysis of the heavy water plant structures was carried out for self weight, equipment load and earthquake load. Pressure loading was also considered in case of H 2 S storage tanks. Soil structure interaction effect was considered in the analysis. The combined stresses in the structures due to earthquake and dead load were checked with the allowable stresses. (author)

Role of seismic PRA in seismic safety decisions of nuclear power plants

International Nuclear Information System (INIS)

Ravindra, M.K.; Kennedy, R.P.; Sues, R.H.

1985-01-01

This paper highlights the important roles that seismic probabilistic risk assessments (PRAs) can play in the seismic safety decisions of nuclear power plants. If a seismic PRA has been performed for a plant, its results can be utilized to evaluate the seismic capability beyond the safe shutdown event (SSE). Seismic fragilities of key structures and equipment, fragilities of dominant plant damage states and the frequencies of occurrence of these plant damage states are reviewed to establish the seismic safety of the plant beyond the SSE level. Guidelines for seismic margin reviews and upgrading may be developed by first identifying the generic classes of structures and equipment that have been shown to be dominant risk contributors in the completed seismic PRAs, studying the underlying causes for their contribution and examining why certain other items (e.g., piping) have not proved to be high-risk-contributors

Commissioning and maintenance experience on mechanical equipment in steam generators of captive power plant at HWP, Manuguru (Paper No. 5.3)

International Nuclear Information System (INIS)

Bhatnagar, R.; Sinha, Ashok; Mohan Rao, A.C.

1992-01-01

Heavy Water Project (Manuguru) is having a captive power plant to cater to the demands of steam and power for the main plant. During the commissioning and initial run of the steam generators and their auxiliaries, teething/initial problems were encountered in nearly all the equipment of the steam generators. This paper briefly describes some of the major problems faced during the commissioning of the steam generators. (author). 4 figs

Phylogenetic composition of host plant communities drives plant-herbivore food web structure.

Science.gov (United States)

Volf, Martin; Pyszko, Petr; Abe, Tomokazu; Libra, Martin; Kotásková, Nela; Šigut, Martin; Kumar, Rajesh; Kaman, Ondřej; Butterill, Philip T; Šipoš, Jan; Abe, Haruka; Fukushima, Hiroaki; Drozd, Pavel; Kamata, Naoto; Murakami, Masashi; Novotny, Vojtech

2017-05-01

Insects tend to feed on related hosts. The phylogenetic composition of host plant communities thus plays a prominent role in determining insect specialization, food web structure, and diversity. Previous studies showed a high preference of insect herbivores for congeneric and confamilial hosts suggesting that some levels of host plant relationships may play more prominent role that others. We aim to quantify the effects of host phylogeny on the structure of quantitative plant-herbivore food webs. Further, we identify specific patterns in three insect guilds with different life histories and discuss the role of host plant phylogeny in maintaining their diversity. We studied herbivore assemblages in three temperate forests in Japan and the Czech Republic. Sampling from a canopy crane, a cherry picker and felled trees allowed a complete census of plant-herbivore interactions within three 0·1 ha plots for leaf chewing larvae, miners, and gallers. We analyzed the effects of host phylogeny by comparing the observed food webs with randomized models of host selection. Larval leaf chewers exhibited high generality at all three sites, whereas gallers and miners were almost exclusively monophagous. Leaf chewer generality dropped rapidly when older host lineages (5-80 myr) were collated into a single lineage but only decreased slightly when the most closely related congeneric hosts were collated. This shows that leaf chewer generality has been maintained by feeding on confamilial hosts while only a few herbivores were shared between more distant plant lineages and, surprisingly, between some congeneric hosts. In contrast, miner and galler generality was maintained mainly by the terminal nodes of the host phylogeny and dropped immediately after collating congeneric hosts into single lineages. We show that not all levels of host plant phylogeny are equal in their effect on structuring plant-herbivore food webs. In the case of generalist guilds, it is the phylogeny of deeper

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)

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

U.S. Transportation Command's Reporting of Property, Plant, and Equipment Assets on the FY 2000 DoD Agency-Wide Financial Statements

National Research Council Canada - National Science Library

Granetto, Paul

2001-01-01

.... The accuracy of the $78 billion of property, plant, and equipment reported on the FY 2000 DoD Agency-Wide Financial Statements is essential to DoD receiving favorable audit opinions on its financial statements. The U.S...

RATU - Nuclear power plant structural safety research programme

International Nuclear Information System (INIS)

Rintamaa, R.

1992-07-01

Studies on the structural materials in nuclear power plants create the experimental data and background information necessary for the structural integrity assessments of mechanical components. The research is carried out by developing experimental fracture mechanics methods including statistical analysis methods of materials property data, and by studying material ageing and, in particular, mechanisms of material deterioration due to neutron irradiation, corrosion and water chemistry. Besides material studies, new testing methods and sensors for measurement of loading and water chemistry parameters have been developed. The monitoring data obtained in real power plants has been used to simulate more precisely the real environment during laboratory tests. The research on structural analysis has focused on extending and verifying the analysis capabilities for structural assessments of nuclear power plants. A widely applicable system including various computational fracture assessment methods has been created with which different structural problems can be solved reliably and effectively. Research on reliability assessment of maintenance in nuclear power plants is directed to practical case studies on components and structures of safety importance, and to the development of models for maintenance related decision support. A systematic analysis of motor-operated valve has been performed

A Computerized Procedure linked to Virtual Equipment

International Nuclear Information System (INIS)

Jung, Yeon Sub; Song, Tae Young

2011-01-01

Digital, information, and communication technologies have change human's behavior. This is because human has limitation to memorize and process information. Human has to access other information and real time information for important decisions. Those technologies are playing important roles. Nuclear power plants cannot be exception. Many accidents in nuclear power plants result from absent or incorrect information. The information for nuclear personnel is context sensitive. They don't have enough time to verify the context sensitive information. Therefore they skip the information, as resulting in incident. Nuclear personnel are usually carrying static paper procedures during local task performance. The procedure guides them steps to follow. There is, however, no dynamic and context sensitive information in the paper. The effect of the work is evaluated once while getting permission of the work. Afterward they are not informed. The static paper is generally simplified, so that it does not show detail of equipment being manipulated. Particularly novice workers feel difficult to understand the procedure due to lack of detail. Pictures of equipment inserted in the procedure are not enough for comprehension. A computerized procedure linked with virtual equipment is one of the best solutions to increase the detail of procedure. Virtual equipment, however, has still limitation not to provide real time information, because the virtual equipment is not synchronized with real plants

Investigating Effects of Invasive Species on Plant Community Structure

Science.gov (United States)

Franklin, Wilfred

2008-01-01

In this article, the author presents a field study project that explores factors influencing forest community structure and lifts the veil off of "plant blindness." This ecological study consists of three laboratories: (1) preliminary field trip to the study site; (2) plant survey; and (3) analyzing plant community structure with descriptive…

Plant retroviruses: structure, evolution and future applications | Zaki ...

African Journals Online (AJOL)

Until recently, retroviruses were thought to be restricted to vertebrates. Plant sequencing projects revealed that plant genomes contain retroviral-like sequences. This review aims to address the structure and evolution of plant retroviruses. In addition, it proposes future applications for these important key components of plant ...

Characteristics and application study of AP1000 NPPs equipment reliability classification method

International Nuclear Information System (INIS)

Guan Gao

2013-01-01

AP1000 nuclear power plant applies an integrated approach to establish equipment reliability classification, which includes probabilistic risk assessment technique, maintenance rule administrative, power production reliability classification and functional equipment group bounding method, and eventually classify equipment reliability into 4 levels. This classification process and result are very different from classical RCM and streamlined RCM. It studied the characteristic of AP1000 equipment reliability classification approach, considered that equipment reliability classification should effectively support maintenance strategy development and work process control, recommended to use a combined RCM method to establish the future equipment reliability program of AP1000 nuclear power plants. (authors)

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.

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

Domestic manufacturing and reliability improvement of reactor water recirculation equipment

International Nuclear Information System (INIS)

Kobayashi, Hidekazu; Oi, Masao; Shida, Toichi; Yokomori, Takashi

1982-01-01

The reactor coolant recirculation system is one of the important systems to control the reactor output in BWR nuclear power plants. Its components require high reliability and maintainability as well as controllability. For many Japanese nuclear power plants, recirculation pumps, fluid couplings and others have been imported so far. Hitachi Ltd. has established a domestic manufacturing organization through the development and test of these equipment. The fundamental design conditions for these equipment are the improvement of the rate of utilization of plant facility, the capability to follow load, and output power stability. In this paper, the specifications, the investigation of moment of inertia and the design features of recirculation pumps, driving motors and variable frequency power supply systems are described. The paper also reports on the combination test implemented to evaluate the recirculation system. The combination test includes the test using water rheostat for the power source facility and the loading test for a recirculation pump. The application of those system equipment to an actual plant was analyzed and evaluated on a basis of the test data obtained. The result showed that the equipment can achieve the rate of change of reactor power of 30%/min. Those equipment have been employed for No. 2 reactor plant of the Fukushima No. 2 Nuclear Power Station, the Tokyo Electric Power Co., Inc. (Wakatsuki, Y.)

Structural divergence of Plant TCTPs

Directory of Open Access Journals (Sweden)

Diego eGutiérrez-Galeano

2014-07-01

Full Text Available The Translationally Controlled Tumor Protein (TCTP is a highly conserved protein at the level of sequence, considered to play an essential role in the regulation of growth and development in eukaryotes. However, this function has been inferred from studies in a few model systems, such as mice and mammalian cell lines, Drosophila and Arabidopsis. Thus, the knowledge regarding this protein is far from complete. In the present study bioinformatic analysis showed the presence of one or more TCTP genes per genome in plants with highly conserved signatures and subtle variations at the level of primary structure but with more noticeable differences at the level of predicted three-dimensional structures. These structures show differences in the pocket region close to the center of the protein and in its flexible loop domain. In fact, all predictive TCTP structures can be divided into two groups: 1 AtTCTP1-like and 2 CmTCTP-like, based on the predicted structures of an Arabidopsis TCTP and a Cucurbita maxima TCTP; according to this classification we propose that their probable function in plants may be inferred in principle. Thus different TCTP genes in a single organism may have different functions; additionally, in those species harboring a single TCTP gene this could carry multiple functions. On the other hand, in silico analysis of AtTCTP1-like and CmTCTP-like promoters suggest that these share common motifs but with different abundance, which may underscore differences in their gene expression patterns. Finally, the absence of TCTP genes in most chlorophytes with the exception of Coccomyxa subellipsoidea, indicates that other proteins perform the roles played by TCTP or the pathways regulated by TCTP occur through alternative routes. These findings provide insight into the evolution of this gene family in plants.

Conduct of Operations at Nuclear Power Plants. Safety Guide (Spanish Edition)

International Nuclear Information System (INIS)

2012-01-01

This Safety Guide identifies the main responsibilities and practices of nuclear power plant (NPP) operations departments in relation to their responsibility for the safe functioning of the plant. The guide presents the factors to be considered in structuring the operations department of an NPP; setting high standards of performance; making safety related decisions in an effective manner; conducting control room and field activities in a thorough and professional manner; and maintaining an NPP within established operational limits and conditions. Contents: 1. Introduction; 2. Management and organization of plant operations; 3. Shift complement and functions; 4. Shift routines and operating practices; 5. Control of equipment and plant status; 6. Operations equipment and operator aids; 7. Work control and authorization.

Timing criteria for supplemental BWR emergency response equipment

International Nuclear Information System (INIS)

Bickel, John H.

2015-01-01

The Great Tohuku Earthquake and subsequent Tsunami represented a double failure event which destroyed offsite power connections to Fukushima-Daiichi site and then destroyed on-site electrical systems needed to run decay heat removal systems. The accident could have been mitigated had there been supplemental portable battery chargers, supplemental pumps, and in-place piping connections to provide alternate decay heat removal. In response to this event in the USA, two national response centers, one in Memphis, Tennessee, and another in Phoenix, Arizona, will begin operation. They will be able to dispatch supplemental emergency response equipment to any nuclear plant in the U.S. within 24 hours. In order to define requirements for supplemental nuclear power plant emergency response equipment maintained onsite vs. in a regional support center it is necessary to confirm: (a) the earliest time such equipment might be needed depending on the specific scenario, (b) the nominal time to move the equipment from a storage location either on-site or within the region of a nuclear power plant, and (c) the time required to connect in the supplemental equipment to use it. This paper describes an evaluation process for a BWR-4 with a Mark I Containment starting with: (a) severe accident simulation to define best estimate times available for recovery based on the specific scenario, (b) identify the key supplemental response equipment needed at specific times to accomplish recovery of key safety functions, and (c) evaluate what types of equipment should be warehoused on-site vs. in regional response centers. (authors)

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)

Towards aspect-oriented functional–structural plant modelling

Science.gov (United States)

Cieslak, Mikolaj; Seleznyova, Alla N.; Prusinkiewicz, Przemyslaw; Hanan, Jim

2011-01-01

Background and Aims Functional–structural plant models (FSPMs) are used to integrate knowledge and test hypotheses of plant behaviour, and to aid in the development of decision support systems. A significant amount of effort is being put into providing a sound methodology for building them. Standard techniques, such as procedural or object-oriented programming, are not suited for clearly separating aspects of plant function that criss-cross between different components of plant structure, which makes it difficult to reuse and share their implementations. The aim of this paper is to present an aspect-oriented programming approach that helps to overcome this difficulty. Methods The L-system-based plant modelling language L+C was used to develop an aspect-oriented approach to plant modelling based on multi-modules. Each element of the plant structure was represented by a sequence of L-system modules (rather than a single module), with each module representing an aspect of the element's function. Separate sets of productions were used for modelling each aspect, with context-sensitive rules facilitated by local lists of modules to consider/ignore. Aspect weaving or communication between aspects was made possible through the use of pseudo-L-systems, where the strict-predecessor of a production rule was specified as a multi-module. Key Results The new approach was used to integrate previously modelled aspects of carbon dynamics, apical dominance and biomechanics with a model of a developing kiwifruit shoot. These aspects were specified independently and their implementation was based on source code provided by the original authors without major changes. Conclusions This new aspect-oriented approach to plant modelling is well suited for studying complex phenomena in plant science, because it can be used to integrate separate models of individual aspects of plant development and function, both previously constructed and new, into clearly organized, comprehensive FSPMs. In

Towards aspect-oriented functional--structural plant modelling.

Science.gov (United States)

Cieslak, Mikolaj; Seleznyova, Alla N; Prusinkiewicz, Przemyslaw; Hanan, Jim

2011-10-01

Functional-structural plant models (FSPMs) are used to integrate knowledge and test hypotheses of plant behaviour, and to aid in the development of decision support systems. A significant amount of effort is being put into providing a sound methodology for building them. Standard techniques, such as procedural or object-oriented programming, are not suited for clearly separating aspects of plant function that criss-cross between different components of plant structure, which makes it difficult to reuse and share their implementations. The aim of this paper is to present an aspect-oriented programming approach that helps to overcome this difficulty. The L-system-based plant modelling language L+C was used to develop an aspect-oriented approach to plant modelling based on multi-modules. Each element of the plant structure was represented by a sequence of L-system modules (rather than a single module), with each module representing an aspect of the element's function. Separate sets of productions were used for modelling each aspect, with context-sensitive rules facilitated by local lists of modules to consider/ignore. Aspect weaving or communication between aspects was made possible through the use of pseudo-L-systems, where the strict-predecessor of a production rule was specified as a multi-module. The new approach was used to integrate previously modelled aspects of carbon dynamics, apical dominance and biomechanics with a model of a developing kiwifruit shoot. These aspects were specified independently and their implementation was based on source code provided by the original authors without major changes. This new aspect-oriented approach to plant modelling is well suited for studying complex phenomena in plant science, because it can be used to integrate separate models of individual aspects of plant development and function, both previously constructed and new, into clearly organized, comprehensive FSPMs. In a future work, this approach could be further

ITER plant layout and site services

International Nuclear Information System (INIS)

Chuyanov, V.A.

2000-01-01

The ITER site has not yet been determined. Nevertheless, to develop a construction plan and a cost estimate, it is necessary to have a detailed layout of the buildings, structures and outdoor equipment integrated with the balance of plant service systems prototypical of large fusion power plants. These services include electrical power for magnet feeds and plasma heating systems, cryogenic and conventional cooling systems, compressed air, gas supplies, demineralized water, steam and drainage. Nuclear grade facilities are provided to handle tritium fuel and activated waste, as well as to prevent radiation exposure of workers and the public. To prevent interference between services of different types and for efficient arrangement of buildings, structures and equipment within the site area, a plan was developed which segregated different classes of services to four quadrants surrounding the tokamak building, placed at the approximate geographical centre of the site. The locations of the buildings on the generic site were selected to meet all design requirements at minimum total project cost. A similar approach was used to determine the locations of services above, at and below grade. The generic site plan can be adapted to the site selected for ITER without significant changes to the buildings or equipment. Some rearrangements may be required by site topography, resulting primarily in changes to the length of services that link the buildings and equipment. (author)

Aging of concrete structures in nuclear power plants

International Nuclear Information System (INIS)

Naus, D.J.; Pland, C.B.; Arndt, E.G.

1991-01-01

The Structural Aging (SAG) Program, sponsored by the US Nuclear Regulatory Commission (USNRC) and conducted by the Oak Ridge National Laboratory (ORNL), had the overall objective of providing the USNRC with an improved basis for evaluating nuclear power plant structures for continued service. The program consists of three technical tasks: materials property data base, structural component assessment/repair technology, and quantitative methodology for continued service determinations. Major accomplishments under the SAG Program during the first two years of its planned five-year duration have included: development of a Structural Materials Information Center and formulation of a Structural Aging Assessment Methodology for Concrete Structures in Nuclear Power Plants. 9 refs

Evaluation of the kinematic structure of indicators key elements of sports equipment exercise by postural orientation movements

Directory of Open Access Journals (Sweden)

Y.V. Litvinenko

2014-12-01

Full Text Available Purpose : Examine the kinematic structure of indicators key elements of sports equipment exercise (difficult to coordinate. The method of postural orientation movements. Material : The study involved acrobats jumpers on the path of high qualification (n = 7. The method used video - computer recording the movements of the athlete. Results : Identified nodal elements of sports equipment double back somersault tuck. Exercise performed after rondat and double back flip and stretch after rondat - flick (coup ago. In the preparatory phase of motor actions acrobatic exercises isolated and studied central element of sports equipment - starting posture of the body; in the phase of the main motor action - animation poses of the body; in the final phase - the final body posture (stable landing. Conclusions : The method of video - computer registration allowed to perform a biomechanical analysis and evaluation of key elements of sports equipment double back somersault tuck and a double back flip and stretch. Also gain new knowledge about the mechanism of the phase structure of movements when performing double somersaults.

Equipment Qualification Data Base user manual

International Nuclear Information System (INIS)

Decker, Q.R.; Fackrell, L.J.; Fitch, L.R.; Meeky, O.B.

1985-09-01

This manual details the Equipment Qualification Data Base (EQDB), its usage, and contents. The EQDB consists of two files; the Plant Qualification File (PQF) and the Equipment Qualification File (EQF). The PQF contains plant specific environmental data and the EQF contains summaries of various test results. Two data management systems are used to manipulate the data and are discussed in this manual. SAS Institute System 2000 (S2K) is the management system for the PQF and Query Update (QU) is the operating system for the EQF. Each management system contains report writers. These writers and how to use them are discussed in detail in this manual

A Tsunami PSA for Nuclear Power Plants in Korea

International Nuclear Information System (INIS)

Kim, Min Kyu; Choi, In Kil; Park, Jin Hee; Seo, Kyung Suk; Seo, Jeong Moon; Yang, Joon Eon

2010-06-01

For the evaluation of safety of NPP caused by Tsunami event, probabilistic safety assessment (PSA) method was applied in this study. At first, an empirical tsunami hazard analysis performed for an evaluation of tsunami return period. A procedure for tsunami fragility methodology was established, and target equipment and structures for investigation of Tsunami Hazard assessment were selected. A several fragility calculations were performed for equipment in Nuclear Power Plant and finally accident scenario of tsunami event in NPP was presented. Finally, a system analysis performed in the case of tsunami event for an evaluation of a CDF of Ulchin 56 NPP site. For the evaluation of safety of NPP caused by Tsunami event, probabilistic safety assessment (PSA) method was applied. A procedure for tsunami fragility methodology was established, and target equipment and structures for investigation of Tsunami Hazard assessment were selected. A several fragility calculations were performed for equipment in Nuclear Power Plant and finally accident scenario of tsunami event in NPP was presented. As a result, in the case of tsunami event, functional failure is mostly governed total failure probability of facilities in NPP site

Methods for seismic analysis of nuclear power plants

International Nuclear Information System (INIS)

Gantenbein, F.

1990-01-01

The seismic analysis of a complex structure, such as a nuclear power plant, is done in various steps. An overview of the methods, used in each of these steps will be given in the following chapters: Seismic analysis of the buildings taking into account structures with important mass or stiffness. The input to the building analysis, called ground motion, is described by an accelerogram or a response spectra. In this step, soil structure interaction has to be taken into account. Various methods are available: Impedance, finite element. The response of the structure can be calculated by spectral method or by time history analysis; advantages and limitations of each method will be shown. Calculation of floor response spectrum which are the data for the equipment analysis. Methods to calculate this spectrum will be described. Seismic analysis of the equipments. Presentation of the methods for both monosupported and multisupported equipment will be given. In addition methods to analyse equipments which present non-linearities associated to the boundary conditions such as impacts, sliding will be presented. (author). 30 refs, 15 figs

Investigation of the effects of a seismic event on accelerated aged components and benefits in equipment life extension

International Nuclear Information System (INIS)

Rygg, D.E.; Epstein, J.L.

1985-01-01

Westinghouse has performed extensive testing to determine the effects of aging on a wide range of components. Additionally, Westinghouse has an extensive data base of nuclear plant equipment and components. This paper presents how the data base of information on plant parts can be analyzed, modified, and managed or tracked to reflect in-plant parts life extension based on actual tests on aging. Such an approach can benefit utility programs for parts inventory, plant operations and plant availability, and can also reduce the costs of parts reordering. Rather than weigh the merits of the positions in this debate, this paper presents the results of a component aging program which simulates from five to twenty years of operation followed by a seismic event and identifies the possible incorporation of this data into a plant data base which offers quick reference and provides other relevant information on the component and equipment. The use of this data as part of a well structured maintenance and surveillance program offers an avenue to resolve this debate in a cost effective manner

Structural analysis of the equipment removal system for tank 241SY101

International Nuclear Information System (INIS)

Mackey, T.C.

1995-01-01

The calculations documented in this report show that the ERS major components are structurally qualified to complete the objective, i.e., to install the removed equipment into a shipping container and transport and store the container at the Central Waste Complex (CWC). The analysis for the structural members of the ERS components considers live load with an impact factor of 125 % added to dead load. An allowable stress of one-third yield is used for all structural components carrying the load based on DOE-RL-92-36. Adherence to DOE-RL-92-36 is not a code requirement. However, the loads considered make this factor of safety appropriate. The calculations meet the strength requirements of the American Institute for Steel Construction (ASIC 1989) for all non-critical structural elements

Condition monitoring and maintenance of nuclear power plant concrete structures

International Nuclear Information System (INIS)

Orr, R.; Prasad, N.

1988-01-01

Nuclear power plant concrete structures are potentially subject to deterioration due to several environmental conditions, including weather exposure, ground water exposure, and sustained high temperature and radiation levels. The nuclear power plant are generally licensed for a term of 40 years. In order to maximize the return from the existing plants, feasibility studies are in progress for continued operation of many of these plants beyond the original licensed life span. This paper describes a study that was performed with an objective to define appropriate condition monitoring and maintenance procedures. A timely implementation of a condition monitoring and maintenance program would provide a valuable database and would provide justification for extension of the plant's design life. The study included concrete structures such as the containment buildings, interior structures, basemats, intake structures and cooling towers. Age-related deterioration at several operating power plants was surveyed and the potential degradation mechanisms have been identified

Laser microbeams for the manipulation of plant cells and subcellular structures

International Nuclear Information System (INIS)

Hoffmann, F.

1996-01-01

Laser microsurgery has been used in plants to study physiological, cell biological and genetical questions for over 10 years. More recently, the optical trap became available as an additional tool. Specific areas of research include membrane physiology, motility, transformation and protoplast fusion. Compared to the data reported in animal systems, the contributions of laser microbeam manipulations in plant biology are rather limited. However, with increased awareness of the enormous potential of the technology and better accessibility to less expensive and more user-friendly equipment, the next decade should be more productive. (author)

Safety classification of nuclear power plant systems, structures and components

International Nuclear Information System (INIS)

1992-01-01

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

Seismic qualification of non-safety class equipment whose failure would damage safety class equipment

International Nuclear Information System (INIS)

LaSalle, F.R.

1991-01-01

Both Code of Federal Regulations, Title 10, Part 50, and US Department of Energy Order 6340.1A have requirements to assess the interaction of non-safety and safety class structures and equipment during a seismic event to maintain the safety function. At the Hanford Site, a cost effective program has been developed to perform the evaluation of non-safety class equipment. Seismic qualification is performed by analysis, test, or upgrading of the equipment to ensure the integrity of safety class structures and equipment. This paper gives a brief overview and synopsis that address design analysis guidelines including applied loading, damping values, component anchorage, allowable loads, and stresses. Test qualification of equipment and walkdown acceptance criteria for heating ampersand ventilation (H ampersand V) ducting, conduit, cable tray, missile zone of influence, as well as energy criteria are presented

Qualification of Electrical Equipment in Nuclear Power Plants - Management of ageing

International Nuclear Information System (INIS)

Spaang, Kjell; Staahl, Gunnar

2013-02-01

The purpose of this report is to describe programs and tools for assessment of accomplished and documented qualification with respect to ageing of electrical equipment and for development of complimentary ageing management programs. In addition to description of complete programs for management of ageing, tools for validation of the status with regard to ageing of installed ('old') equipment and, where needed, for complementation of their qualification are also included. The report is restricted to safety related equipment containing ageing sensitive parts, mainly organic materials. To this category belong cables and cable joints and a number of equipment containing oils, seals (o-rings), etc. For equipment located in the containment, the possibilities of continuous supervision are limited. The accessibility for regular inspections is also limited in many cases. The main part of this report deals with the qualification of such equipment. Some safety related equipment outside the containment can be located in areas where they are subjected to high temperature and other excessive environmental stresses during normal operation and in areas affected by an accident. Therefore, some material is given also on qualification of equipment located outside containment with better possibilities for frequent inspection and supervision. Part 1 of the report is an executive summary with a general review of the methodologies and their application. The more detailed description of the programs and underlying material, useful data, etc. is given in Part 2

Equipment Maintenance management support system based on statistical analysis of maintenance history data

International Nuclear Information System (INIS)

Shimizu, S.; Ando, Y.; Morioka, T.

1990-01-01

Plant maintenance is recently becoming important with the increase in the number of nuclear power stations and in plant operating time. Various kinds of requirements for plant maintenance, such as countermeasures for equipment degradation and saving maintenance costs while keeping up plant reliability and productivity, are proposed. For this purpose, plant maintenance programs should be improved based on equipment reliability estimated by field data. In order to meet these requirements, it is planned to develop an equipment maintenance management support system for nuclear power plants based on statistical analysis of equipment maintenance history data. The large difference between this proposed new method and current similar methods is to evaluate not only failure data but maintenance data, which includes normal termination data and some degree of degradation or functional disorder data for equipment and parts. So, it is possible to utilize these field data for improving maintenance schedules and to evaluate actual equipment and parts reliability under the current maintenance schedule. In the present paper, the authors show the objectives of this system, an outline of this system and its functions, and the basic technique for collecting and managing of maintenance history data on statistical analysis. It is shown, from the results of feasibility tests using simulation data of maintenance history, that this system has the ability to provide useful information for maintenance and the design enhancement

Equipment abnormality monitoring device

International Nuclear Information System (INIS)

Ando, Yasumasa

1991-01-01

When an operator hears sounds in a plantsite, the operator compares normal sounds of equipment which he previously heard and remembered with sounds he actually hears, to judge if they are normal or abnormal. According to the method, there is a worry that abnormal conditions can not be appropriately judged in a case where the number of objective equipments is increased and in a case that the sounds are changed gradually slightly. Then, the device of the present invention comprises a plurality of monitors for monitoring the operation sound of equipments, a recording/reproducing device for recording and reproducing the signals, a selection device for selecting the reproducing signals among the recorded signals, an acoustic device for converting the signals to sounds, a switching device for switching the signals to be transmitted to the acoustic device between to signals of the monitor and the recording/reproducing signals. The abnormality of the equipments can be determined easily by comparing the sounds representing the operation conditions of equipments for controlling the plant operation and the sounds recorded in their normal conditions. (N.H.)

How plant architecture affects light absorption and photosynthesis in tomato: towards an ideotype for plant architecture using a functional-structural plant model

NARCIS (Netherlands)

Sarlikioti, V.; Visser, de P.H.B.; Buck-Sorlin, G.H.; Marcelis, L.F.M.

2011-01-01

Background and Aims - Manipulation of plant structure can strongly affect light distribution in the canopy and photosynthesis. The aim of this paper is to find a plant ideotype for optimization of light absorption and canopy photosynthesis. Using a static functional structural plant model (FSPM), a

Development of ultrasonic testing equipment incorporating electromagnetic acoustic transducer

International Nuclear Information System (INIS)

Sato, Michio; Kimura, Motohiko; Okano, Hideharu; Miyazawa, Tatsuo; Nagase, Koichi; Ishikawa, Masaaki

1989-01-01

An ultrasonic testing equipment for use in in-service inspection of nuclear power plant piping has been developed, which comprises an angle-beam electromagnetic acoustic transducer mounted on a vehicle for scanning the piping surface to be inspected. The transducer functions without direct contact with the piping surface through couplant, and the vehicle does not require a guide track installed on the piping surface, being equipped with magnetic wheels that adhere to the piping material, permitting it to travel along the circumferential weld joint of a carbon steel pipe. The equipment thus dispenses with the laborious manual work involved in preparing the piping for inspection, such as removal of protective coating, surface polishing and installation of guide track and thereby considerably reduces the duration of inspection. The functioning principle and structural features of the transducer and vehicle are described, together with the results of trial operation of a prototype unit, which proved a 1mm deep notch cut on a test piece of 25mm thick carbon steel plate to be locatable with an accuracy of ±2mm. (author)

Structural building response review

International Nuclear Information System (INIS)

1980-01-01

The integrity of a nuclear power plant during a postulated seismic event is required to protect the public against radiation. Therefore, a detailed set of seismic analyses of various structures and equipment is performed while designing a nuclear power plant. This report describes the structural response analysis method, including the structural model, soil-structure interaction as it relates to structural models, methods for seismic structural analysis, numerical integration methods, methods for non-seismic response analysis approaches for various response combinations, structural damping values, nonlinear response, uncertainties in structural properties, and structural response analysis using random properties. The report describes the state-of-the-art in these areas for nuclear power plants. It also details the past studies made at Sargent and Lundy to evaluate different alternatives and the conclusions reached for the specific purposes that those studies were intended. These results were incorporated here because they fall into the general scope of this report. The scope of the present task does not include performing new calculations

Underwater nuclear power plant structure

International Nuclear Information System (INIS)

Severs, S.; Toll, H.V.

1982-01-01

A structure for an underwater nuclear power generating plant comprising a triangular platform formed of tubular leg and truss members upon which are attached one or more large spherical pressure vessels and one or more small cylindrical auxiliary pressure vessels. (author)

Problems and future outlook in the nuclear equipment manufacturing industry

International Nuclear Information System (INIS)

Suenaga, Soichiro

1984-01-01

The energy policy in Japan is based on a balance between the energy security and the energy cost for the purpose of realizing optimal supply/demand structure. In this field, nuclear equipment manufacturers should cooperate in the settlement of LWR power generation through plant safety and reliability and through high economical efficiency, all involving the advancement of technology. As a new concept being developed, there is an APWR (advanced PWR) which has the electric output of 1,350 MWe. The export of nuclear power plants, though there are various problems, should be enhanced in the high-technology export area. The following matters are described: the settlement of and the heightening of technology in nuclear power generation, the development of the advanced PWR, and the measures for the export of nuclear power plants and components. (Mori, K.)

Wire rope isolators for vibration isolation of equipment and structures – A review

International Nuclear Information System (INIS)

Balaji, P S; Rahman, M E; Lau, H H; Moussa, Leblouba

2015-01-01

Vibrations and shocks are studied using various techniques and analyzed to predict their detrimental effect on the equipment and structures. In cases, where the effects of vibration become unacceptable, it may cause structural damage and affect the operation of the equipment. Hence, adding a discrete system to isolate the vibration from source becomes necessary. The Wire Rope Isolator (WRI) can be used to effectively isolate the system from disturbing vibrations. The WRI is a type of passive isolator that exhibits nonlinear behavior. It consists of stranded wire rope held between two metal retainer bars and the metal wire rope is made up of individual wire strands that are in frictional contact with each other, hence, it is a kind of friction-type isolator. This paper compiles the research work on wire rope isolators. This paper presents the research work under two categories, namely monotonic and cyclic loading behaviors of WRI. The review also discusses the different terminologies associated with vibration isolation system and highlights the comparison between various isolation systems. (paper)

Remote Sensing of plant functional types: Relative importance of biochemical and structural plant traits

Science.gov (United States)

Kattenborn, Teja; Schmidtlein, Sebastian

2017-04-01

Monitoring ecosystems is a key priority in order to understand vegetation patterns, underlying resource cycles and changes their off. Driven by biotic and abiotic factors, plant species within an ecosystem are likely to share similar structural, physiological or phenological traits and can therefore be grouped into plant functional types (PFT). It can be assumed that plants which share similar traits also share similar optical characteristics. Therefore optical remote sensing was identified as a valuable tool for differentiating PFT. Although several authors list structural and biochemical plant traits which are important for differentiating PFT using hyperspectral remote sensing, there is no quantitative or qualitative information on the relative importance of these traits. Thus, little is known about the explicit role of plant traits for an optical discrimination of PFT. One of the main reasons for this is that various optical traits affect the same wavelength regions and it is therefore difficult to isolate the discriminative power of a single trait. A way to determine the effect of single plant traits on the optical reflectance of plant canopies is given by radiative transfer models. The most established radiative transfer model is PROSAIL, which incorporates biochemical and structural plant traits, such as pigment contents or leaf area index. In the present study 25 grassland species of different PFT were cultivated and traits relevant for PROSAIL were measured for the entire vegetation season of 2016. The information content of each trait for differentiating PFTs was determined by applying a Multi-response Permutation Procedure on the actual traits, as well as on simulated canopy spectra derived from PROSAIL. According to our results some traits, especially biochemical traits, show a weaker separability of PFT on a spectral level than compared to the actual trait measurements. Overall structural traits (leaf angle and leaf area index) are more important for

Impact of structural aging on seismic risk assessment of reinforced concrete structures in nuclear power plants

International Nuclear Information System (INIS)

Ellingwood, B.; Song, J.

1996-03-01

The Structural Aging Program is addressing the potential for degradation of concrete structural components and systems in nuclear power plants over time due to aging and aggressive environmental stressors. Structures are passive under normal operating conditions but play a key role in mitigating design-basis events, particularly those arising from external challenges such as earthquakes, extreme winds, fires and floods. Structures are plant-specific and unique, often are difficult to inspect, and are virtually impossible to replace. The importance of structural failures in accident mitigation is amplified because such failures may lead to common-cause failures of other components. Structural condition assessment and service life prediction must focus on a few critical components and systems within the plant. Components and systems that are dominant contributors to risk and that require particular attention can be identified through the mathematical formalism of a probabilistic risk assessment, or PRA. To illustrate, the role of structural degradation due to aging on plant risk is examined through the framework of a Level 1 seismic PRA of a nuclear power plant. Plausible mechanisms of structural degradation are found to increase the core damage probability by approximately a factor of two

RPC industries - UV and EB equipment manufacturers

International Nuclear Information System (INIS)

Rodrigues, A.M.

1987-01-01

RPC Industries has been manufacturing electron beam and ultraviolet equipment for the industrial processing of materials for more than 15 years. RPC maintains its headquarters and electron processor manufacturing plant in Hayward, California. UV equipment is made in the company's plant near Chicago. Sales offices are maintained in New York, Illinois, and California in the USA, and in Germany, Japan, Australia, Italy, Israel, and Sweden. Complete testing and pilot facilities are available in Hayward (EB) and near Chicago (UV). Described below are the basic system components, applications and advantages of RPC's UV and EB systems. (orig.)

Review of nuclear power plant systems

International Nuclear Information System (INIS)

Doehler

1980-01-01

This presentation starts with a brief description of the Technischer Ueberwachungs-Verein (TUeV) and its main activities in the field of technical assessments. The TUeV-organisation is in general the assessor who performs the review if nuclear power plant systems, structures and equipment. All aspects relating to the safe operation of nuclear power plants are assessed by the TUeV. This paper stresses the review of the design of nuclear power plant systems and structures. It gives an outline on the procedure of an assessment, starting with the regulatory requirements, going into the papers of the applicant and finally ending with the TUeV-appraisal. This procedure is shown using settlement measuring requirements as an example. The review of the design of mechanical structures such as pipes, valves, pump and vessels is shown in detail. (RW)

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)

Improved control rod drive handling equipment for BWRs [boiling-water reactors]: Final report

International Nuclear Information System (INIS)

Turner, A.P.L.; Gorman, J.A.

1987-08-01

Improved equipment for removing and replacing control rod drives (CRDs) in BWR plants has been designed, built and tested. Control rod drives must be removed from the reactor periodically for servicing. Removal and replacement of CRDs using equipment originally supplied with the plant has long been recognized as one of the more difficult and highest radiation exposure maintenance operations that must be performed at BWR plants. The improved equipment was used for the first time at Quad Cities, Unit 2, during a Fall 1986 outage. The trial of the equipment was highly successful, and it was shown that the new equipment significantly improves CRD handling operations. The new equipment significantly simplifies the sequence of operations required to lower a CRD from its housing, upend it to a horizontal orientation, and transport it out of the reactor containment. All operations of the new equipment are performed from the undervessel equipment handling platform, thus, eliminating the requirement for a person to work on the lower level of the undervessel gallery which is often highly contaminated. Typically, one less person is required to operate the equipment than were used with the older equipment. The new equipment incorporates a number of redundant and fail safe features that improve operations and reduce the chances for accidents

The problem of ensuring the seismic stability of atomic electric power plant equipment and ways of solving it

International Nuclear Information System (INIS)

Kaznovskii; Filippov, G.A.

1983-01-01

By seismic stability the authors mean the ability of the equipment and buildings to retain certain properties when subjected to seismic loads: leakproofness, strength, the absence of any residual changes of shape, which interfere with normal operation, ability to be repaired, nuclear and radiation safety. The latter requirement is the main thing which differentiates atomic electric power plants from other constructions, including other power-generation plants. Whereas, for example, an accident in the event of an earthquake in a thermal electric power plant can be regarded as a local accident, and the measures to ensure seismic stability are determined by economic factors and safety requirements for the operating staff, to ensure the seismic stability of an AES it is essential to take account in the first instance of the possibility of dangerous radiation effects both in the AES and in the vast area around it

A Study on Salt Attack Protection of Structural and Finishing Materials in Power Plant Structures

Energy Technology Data Exchange (ETDEWEB)

Kim, W.B.; Kweon, K.J.; Suh, Y.P.; Nah, H.S.; Lee, K.J.; Park, D.S.; Jo, Y.K. [Korea Electric Power Research Institute, Taejeon (Korea, Republic of)

1997-12-31

This is a final report written by both KEPRI and KICT as a co-operative research titled {sup A} study on Salt Protection of Structural and Finishings in Power Plant Structures{sup .} This study presented the methods to prevent the chloride-induced corrosion of power plant structures through collection and analysis of research datum relating to design, construction and maintenance for the prevention of structural and finishing materials, thru material performance tests for anti-corrosion under many kinds of chloride-induced corrosion environments. As a result, this study proposed the guidelines for design, construction and maintenance of power plant structures due to chloride-induced corrosion. (author). 257 refs., 111 figs., 86 tabs.

A Study on Salt Attack Protection of Structural and Finishing Materials in Power Plant Structures

Energy Technology Data Exchange (ETDEWEB)

Kim, W B; Kweon, K J; Suh, Y P; Nah, H S; Lee, K J; Park, D S; Jo, Y K [Korea Electric Power Research Institute, Taejeon (Korea, Republic of)

1998-12-31

This is a final report written by both KEPRI and KICT as a co-operative research titled {sup A} study on Salt Protection of Structural and Finishings in Power Plant Structures{sup .} This study presented the methods to prevent the chloride-induced corrosion of power plant structures through collection and analysis of research datum relating to design, construction and maintenance for the prevention of structural and finishing materials, thru material performance tests for anti-corrosion under many kinds of chloride-induced corrosion environments. As a result, this study proposed the guidelines for design, construction and maintenance of power plant structures due to chloride-induced corrosion. (author). 257 refs., 111 figs., 86 tabs.

Concept of diagnostic monitoring of condition of selected equipment for V-1 nuclear power plant

International Nuclear Information System (INIS)

Jaros, I.

1981-01-01

The vibroacoustic method based on picking up and processing vibrations, shocks and structural noise from the outer surface of equipment was chosen for testing the mechanical conditions of the reactor and of the main circulating pumps. The location of vibration pickups on the primary circuit components, their specifications, signal processing and evaluation are described. (M.D.)

Environmental qualification program of electric equipment for Angra 1

Energy Technology Data Exchange (ETDEWEB)

Muzitano, Grazielle F.; Justino, Marcelo C.; Silva, Marcos C., E-mail: grazi@eletronuclear.gov.br, E-mail: justino@eletronuclear.gov.br, E-mail: candeia@eletronuclear.gov.br [Eletrobras Termonuclear S.A. (ELETRONUCLEAR), Rio de Janeiro, RJ (Brazil)

2017-11-01

This paper describes the development of the environmental qualification program for important electrical equipment (EQPEE) used for safety in Angra 1 Nuclear Power Plant (NPP). The environmental qualification program started in United States of America by the Nuclear Regulatory Commission (NRC) after the Three Mile Island (TMI) Nuclear Power Plant Accident in 1979. In that event, some equipment installed inside the reactor containment of TMI failed due the harsh conditions that occurred after the accident. Because of this fact, the NRC issued the Regulation 50.49 'Environmental qualification of electric equipment important to safety for nuclear power plants'. The Brazilian regulatory commission CNEN also asked Angra 1 to follow this regulation and to implement an EQPEE similar to the programs adopted by American and other NPPs around the world. Due to the importance to maintain the critical equipment operating in normal and abnormal environment conditions, the program aims to assure that this equipment remains qualified to work under the harsh conditions found inside the reactor containment. The aging of these components are also analyzed in this program that is important in the process to extend the operating life of Angra 1 for more 20 years, which is normally referred as Long Term Operation (LTO). (author)

Environmental qualification program of electric equipment for Angra 1

International Nuclear Information System (INIS)

Muzitano, Grazielle F.; Justino, Marcelo C.; Silva, Marcos C.

2017-01-01

This paper describes the development of the environmental qualification program for important electrical equipment (EQPEE) used for safety in Angra 1 Nuclear Power Plant (NPP). The environmental qualification program started in United States of America by the Nuclear Regulatory Commission (NRC) after the Three Mile Island (TMI) Nuclear Power Plant Accident in 1979. In that event, some equipment installed inside the reactor containment of TMI failed due the harsh conditions that occurred after the accident. Because of this fact, the NRC issued the Regulation 50.49 'Environmental qualification of electric equipment important to safety for nuclear power plants'. The Brazilian regulatory commission CNEN also asked Angra 1 to follow this regulation and to implement an EQPEE similar to the programs adopted by American and other NPPs around the world. Due to the importance to maintain the critical equipment operating in normal and abnormal environment conditions, the program aims to assure that this equipment remains qualified to work under the harsh conditions found inside the reactor containment. The aging of these components are also analyzed in this program that is important in the process to extend the operating life of Angra 1 for more 20 years, which is normally referred as Long Term Operation (LTO). (author)

Report on aging of nuclear power plant reinforced concrete structures

International Nuclear Information System (INIS)

Naus, D.J.; Oland, C.B.; Ellingwood, B.R.

1996-03-01

The Structural Aging Program provides the US Nuclear Regulatory Commission with potential structural safety issues and acceptance criteria for use in continued service assessments of nuclear power plant safety-related concrete structures. The program was organized under four task areas: Program Management, Materials Property Data Base, Structural Component Assessment/Repair Technology, and Quantitative Methodology for Continued Service Determinations. Under these tasks, over 90 papers and reports were prepared addressing pertinent aspects associated with aging management of nuclear power plant reinforced concrete structures. Contained in this report is a summary of program results in the form of information related to longevity of nuclear power plant reinforced concrete structures, a Structural Materials Information Center presenting data and information on the time variation of concrete materials under the influence of environmental stressors and aging factors, in-service inspection and condition assessments techniques, repair materials and methods, evaluation of nuclear power plant reinforced concrete structures, and a reliability-based methodology for current and future condition assessments. Recommendations for future activities are also provided. 308 refs., 61 figs., 50 tabs

Report on aging of nuclear power plant reinforced concrete structures

Energy Technology Data Exchange (ETDEWEB)

Naus, D.J.; Oland, C.B. [Oak Ridge National Lab., TN (United States); Ellingwood, B.R. [Johns Hopkins Univ., Baltimore, MD (United States). Dept. of Civil Engineering

1996-03-01

The Structural Aging Program provides the US Nuclear Regulatory Commission with potential structural safety issues and acceptance criteria for use in continued service assessments of nuclear power plant safety-related concrete structures. The program was organized under four task areas: Program Management, Materials Property Data Base, Structural Component Assessment/Repair Technology, and Quantitative Methodology for Continued Service Determinations. Under these tasks, over 90 papers and reports were prepared addressing pertinent aspects associated with aging management of nuclear power plant reinforced concrete structures. Contained in this report is a summary of program results in the form of information related to longevity of nuclear power plant reinforced concrete structures, a Structural Materials Information Center presenting data and information on the time variation of concrete materials under the influence of environmental stressors and aging factors, in-service inspection and condition assessments techniques, repair materials and methods, evaluation of nuclear power plant reinforced concrete structures, and a reliability-based methodology for current and future condition assessments. Recommendations for future activities are also provided. 308 refs., 61 figs., 50 tabs.

Potentially damaging failure modes of high- and medium-voltage electrical equipment

International Nuclear Information System (INIS)

Hoy, H.C.

1984-01-01

The high- and medium-voltage electrical equipment failures of both nuclear and nonnuclear electric utilities have been reviewed for possible disruptive failure modes that would be of special concern in a nuclear power plant. The resulting emphasis was on the electrical faults of transformers, switchgear (circuit breakers), lightning (surge) arrestors, high-voltage cabling and buswork, control boards, and other electrical equipment that, through failure, can be the initiating event that may expand the original fault to nearby or associated equipment. Many failures of such equipment were found and documented, although the failure rate of electrical equipment in utilities is historically quite low. Nuclear plants record too few failures to be statistically valid, but failures that have been recorded show that good design usually restricts the failure to a single piece of equipment. Conclusions and recommendations pertaining to the design, maintenance, and operation of the affected electrical equipment are presented

AVLIS Production Plant work breakdown structure and Dictionary

International Nuclear Information System (INIS)

1984-01-01

The work breakdown structure has been prepared for the AVLIS Production Plant to define, organize, and identify the work efforts and is summarized in Fig. 1-1 for the top three project levels. The work breakdown structure itself is intended to be the primary organizational tool of the AVLIS Production Plant and is consistent with the overall AVLIS Program Work Breakdown Structure. It is designed to provide a framework for definition and accounting of all of the elements that are required for the eventual design, procurement, and construction of the AVLIS Production Plant. During the present phase of the AVLIS Project, the conceptual engineering phase, the work breakdown structure is intended to be the master structure and project organizer of documents, designs, and cost estimates. As the master project organizer, the key role of the work breakdown structure is to provide the mechanism for developing completeness in AVLIS cost estimates and design development of all hardware and systems. The work breakdown structure provides the framework for tracking, on a one-to-one basis, the component design criteria, systems requirements, design concepts, design drawings, performance projections, and conceptual cost estimates. It also serves as a vehicle for contract reporting. 12 figures, 2 tables

Damages of industrial equipments in the 1995 Hyougoken-Nanbu Earthquake

International Nuclear Information System (INIS)

Iwatsubo, Takuzo

1997-01-01

Hanshin-Awaji area has a population of approximately 3 million and many industries, including heavy industry, harbor facilities and international trading companies. The 1995 Hyougoken-Nanbu Earthquake occurred just in this area which is 25kmx2km oblong containing Kobe city. About 5,500 people were killed and 250,000 people lost their houses. Japan society of mechanical engineers organized the investigative committee of earthquake disaster of industrial equipments after the earthquake in order to investigate the disaster damages of industrial equipments and to give data for a design manual for mechanical equipments against earthquake excitation. This is an investigation report of the disaster damages of industrial machine equipments. Damages to machine equipment of industries in the high intensity region of the earthquake are illustrated. The mechanisms of the damages and measures against earthquake and safety of nuclear power plant design are discussed. Then it is known that the design of nuclear power plant is different from the general industrial facilities and the damage which was suffered in the general industrial facilities does not occur in the nuclear power plant. (J.P.N.)

Damages of industrial equipments in the 1995 Hyougoken-Nanbu Earthquake

Energy Technology Data Exchange (ETDEWEB)

Iwatsubo, Takuzo [Kobe Univ. (Japan). Faculty of Engineering

1997-03-01

Hanshin-Awaji area has a population of approximately 3 million and many industries, including heavy industry, harbor facilities and international trading companies. The 1995 Hyougoken-Nanbu Earthquake occurred just in this area which is 25kmx2km oblong containing Kobe city. About 5,500 people were killed and 250,000 people lost their houses. Japan society of mechanical engineers organized the investigative committee of earthquake disaster of industrial equipments after the earthquake in order to investigate the disaster damages of industrial equipments and to give data for a design manual for mechanical equipments against earthquake excitation. This is an investigation report of the disaster damages of industrial machine equipments. Damages to machine equipment of industries in the high intensity region of the earthquake are illustrated. The mechanisms of the damages and measures against earthquake and safety of nuclear power plant design are discussed. Then it is known that the design of nuclear power plant is different from the general industrial facilities and the damage which was suffered in the general industrial facilities does not occur in the nuclear power plant. (J.P.N.)

Equipment Obsolescence Analysis and Management Software

Energy Technology Data Exchange (ETDEWEB)

Redmond, J.; Carret, L.; Shaon, S.; Schultz, C.

2015-07-01

The procurement engineering resources at Nuclear Power Plants (NPPs) are experiencing increasing backlog for procurement items primarily due to the inability to order the original replacement parts. The level of effort and time required to prepare procurement packages is increasing since the number of obsolete parts are increasing exponentially. Procurement packages for obsolete components and parts are much more complex and take more time to prepare because of the need to perform equivalency evaluations, testing requirements and test acceptance criteria development, commercial grade dedication or equipment qualification, and increasing efforts to verify that no fraudulent or counterfeit parts are procured. This problem will be further compounded when NPPs pursue license renewal and approval for plant-life extension. Advanced planning and advanced knowledge of equipment obsolescence is required to allow for sufficient time to properly procure replacement parts for obsolete items. The uncertain supply chain capability due to obsolescence is a real problem and can cause a risk to reliable plant operations due to the potential for a lack of available spare parts and replacement components to support outages and unplanned component failures. Advanced notification of obsolescence is increasingly more important to ensure that adequate time and planning is scheduled to procure the proper replacement parts. A thorough analysis of Original Equipment Manufacturer (OEM) availability and inventory as well as an analysis of failure rates and usage rates is required to predict critical part needs to allow for early identification of obsolescence issues so that a planned and controlled strategy to qualify replacement equipment can be implemented. (Author)

Remotely operated replaceable process equipment

International Nuclear Information System (INIS)

Westendorf, H.

1987-01-01

The coupling process of pneumatic and electrical auxiliary lines of a pneumatic control pressure line in a large cell of the reprocessing plant is carried out, together with the coupling process of the connecting flange of the process equipment. The coupling places of the auxiliary lines, such as control or supply lines, are laid in the flange parts of the flanges to be connected. The pipe flange on the frame side remains flush with the connecting flange of the process equipment. (DG) [de

The development of web based power plant maintenance management system

International Nuclear Information System (INIS)

Kim, Bum Shin; Kim, Eui Hyun; Jang, Dong Sik; Cho, Jae Min; Chae, Gil Seok; Jung, Gyu Chol

2004-01-01

Most power plants have operated many independent computerize systems for maintenance. Independence of systems have caused complexity of business process and inconvenience of computer system management. Because the equipment and material master data is not standardize and structurize, it is difficult to manage equipment maintenance history and material delivery. Especially equipment classification criterion is important for standardization of every maintenance information. It is necessary to integrate function of independent systems for business process simplification and rapid work flow. This paper provides equipment classification criterion design and system integration method with the case of live system development

PCB transformer fires: the risk in nuclear power plants

International Nuclear Information System (INIS)

Blackmon, K.

1988-01-01

It is estimated that 1/2 of the present nuclear power plants operate with PCB-filled transformer equipment. In an attempt to obtain better estimates of clean-up costs in a nuclear power plant under reasonable-loss scenarios, a study was commissioned. This study was a joint venture between Blackmon-Mooring Steamatic Technologies, Inc., (BMS-TECH) and M and M Protection Consultants. This joint study was conducted at a typical pressurized-water reactor plant consisting of two 1000-MW units. Three specific scenarios were selected and analyzed for this typical power plant. These scenarios were: (1) an electrical failure of a transformer in an isolated switch gear room; (2) a transformer exposed to a 55-gallon transient combustion oil fire in the auxiliary building; and (3) a PCB transformer involved in a major turbine lube fire in the turbine building. Based on results of this study, the insurance carriers for this industry implemented an adjustment in their rate structures for nuclear power plants that have PCB equipment

The development of in-cell remote inspection system in Tokai reprocessing plant

International Nuclear Information System (INIS)

Ishibashi, Yuzo

1985-01-01

In the Tokai fuel reprocessing plant, the containment is triple, i.e. the vessel containing radioactive material, then the concrete cell structure and finally the housing building. The fuel reprocessing plant is now proceeding with the development of an in-cell remote inspection system. The inspection system is for inspection of the cell itself and the equipment etc. in the cell, concerning the integrity. Described are the following: the course taken and problems in development of the remote inspection system; development of the floor rambling type remote inspection equipment and the multiple armed type, both for inspection of in-cell ''drip trays''; in-cell equipment inspection devices in specifications etc.; problems in its future development. (Mori, K.)

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.

Developing ''SMART'' equipment and systems through collaborative NERI research and development

International Nuclear Information System (INIS)

Harmon, Daryl L.; Chapman, Leon D.; Golay, Michael W.; Maynard, Kenneth P.; SpencerR, Joseph W.

2000-01-01

The United States Department of Energy initiated the Nuclear Energy Research Initiative (NERI) to conduct research and development with the objectives of: (1) overcoming the principal technical obstacles to expanded nuclear energy use, (2) advancing the state of nuclear technology to maintain its competitive position in domestic and world markets, and (3) improving the performance, efficiency, reliability, and economics of nuclear energy. Fiscal Year 1999 program funding is $19 Million, with increased finding expected for subsequent years, emphasizing international cooperation. Among the programs selected for funding is the ''Smart Equipment and Systems to Improve Reliability and Safety in Future Nuclear Power Plant Operations''. This program is a 30 month collaborative effort bringing together the technical capabilities of ABB C-E Nuclear Power, Inc. (ABB CENP), Sandia National Laboratories, Duke Engineering and Services (DE and S), Massachusetts Institute of Technology (MIT) and Pennsylvania State University (PSU). The program's goal is to design, develop and evaluate an integrated set of smart equipment and predictive maintenance tools and methodologies that will significantly reduce nuclear plant construction, operation and maintenance costs. To accomplish this goal the Smart Equipment program will: (1) Identify and prioritize nuclear plant equipment that would most likely benefit from adding smart features; (2) Develop a methodology for systematically monitoring the health of individual pieces of equipment implemented with smart features (i.e. smart equipment); (3) Develop a methodology to provide plant operators with real-time information through smart equipment Man-Machine Interfaces (MMI) to support their decision making; (4) Demonstrate the methodology on a targeted component and/or system; (5) Expand the concept to system and plant levels that allow communication and integration of data among smart equipment. This paper will discuss (1) detailed subtask

Horizontal and vertical seismic isolation of a nuclear power plant

International Nuclear Information System (INIS)

Ikonomou, A.S.

1983-01-01

This paper presents a study for the horizontal and vertical seismic isolation of a nuclear power plant with a base isolation system, developed by the author, called the Alexisismon. This system -- which comprises different schemes for horizontal or vertical or both horizontal and vertical isolation -- is a linear system based on the principle of separation of functions. That is, horizontal and vertical isolation are realized through different components and act independently from each other. As far as horizontal isolation is concerned, the role of transmitting vertical loads is uncoupled from the role of inducing horizontal restoring forces so that both functions can be performed without instability. It is possible either to provide both horizontal and vertical isolation to the whole nuclear plant or to isolate the whole plant horizontally and to provide vertical isolation to sensitive and costly equipment only. When the fundamental period of the plant or equipment is 2 seconds and when the vertical displacements are of the order of + or - 20 inches, the structure or equipment are protected against earthquakes up to 1.10 and 1.30 g for actual and 0.60 and 1.50 g for artificial accelerograms. In both cases all the isolation elements behave elastically up to these acceleration limits as well as the superstructure and equipment

Procurement strategic analysis of nuclear safety equipment

International Nuclear Information System (INIS)

Wu Caixia; Yang Haifeng; Li Xiaoyang; Li Shixin

2013-01-01

The nuclear power development plan in China puts forward a challenge on procurement of nuclear safety equipment. Based on the characteristics of the procurement of nuclear safety equipment, requirements are raised for procurement process, including further clarification of equipment technical specification, establishment and improvement of the expert database of the nuclear power industry, adoption of more reasonable evaluation method and establishment of a unified platform for nuclear power plants to procure nuclear safety equipment. This paper makes recommendation of procurement strategy for nuclear power production enterprises from following aspects, making a plan of procurement progress, dividing procurement packages rationally, establishing supplier database through qualification review and implementing classified management, promoting localization process of key equipment continually and further improving the system and mechanism of procurement of nuclear safety equipment. (authors)

Prophylactic and thermovision measurements of electric machines and equipment

International Nuclear Information System (INIS)

Jedlicka, R.; Brestovansky, L.

1996-01-01

High-voltage measurements of generators, unit and service transformers and some significant motor drives used at a nuclear power plant are described in this paper. Thermovision measurements of electric machines and distribution systems are dealt with in the second part of the paper. Power electric equipment represent one of the most significant components of a nuclear power plant. Turbine mechanical energy is converted into the electrical energy within these equipment. Power generated by generators is transformed by transformers so that it can achieve appropriate parameters for both the transmission over the distribution system and the power plant service power supply. The service power supply switchboards and cables provide supply to motors and other consumers necessary for the nuclear power plant technological process. The whole complex of equipment has to be maintained in good technical conditions. It is necessary to make thermovision and prophylactic measurements to identify and verify the electric equipment technical condition. The mentioned measurements warn the operation staff in advance against both gradual deterioration of power connection contact resistances, i.e. power connections overheating, and the machine insulation systems condition deterioration. The operation staff try to prevent the electric equipment operation accidents by early removing the detected failures, thus, improving the nuclear safety. In order to provide the above-mentioned activities a special prophylactic measurement group was established at the NPP Bohunice in 1983. The group specialists make following types of measurements. 1. Prophylactic measurements of electric machines. Prophylactics of 220 MW generators and 6 MW service power generators; Prophylactics of both unit and service transformers and VHV bushings; Prophylactics of major 6 kV motor drives. 2. Thermovision measurements of current connections. Measurements enumarated in paragraph 1 are made on disconnected electric

Benefits of Digital Equipment Generic Qualification Activities

International Nuclear Information System (INIS)

Thomas, James E.; Steiman, Samuel C.

2002-01-01

As a result of nuclear power plant instrumentation and control obsolescence issues, there have been numerous activities during recent years relating to the qualification of digital equipment. Some of these activities have been 'generic' in nature in that the qualification was not limited to plant specific applications, but was intended to cover a broad base of potential applications of the digital equipment. These generic qualifications have been funded by equipment manufacturers and by utility groups and organizations. The generic activities sponsored by the Electric Power Research Institute (EPRI) have been pilot projects for an overall generic qualification approach. The primary benefit resulting from the generic qualification work to date is that a number of digital platforms and digital devices are now available for use in various nuclear safety-related applications. Many of the tests and evaluations necessary to support plant specific applications have been completed. The amount of data and documentation that each utility must develop on a case by case basis has been significantly reduced. There are also a number of additional benefits resulting from these industry efforts. The challenges and difficulties in qualifying digital equipment for safety-related applications are now more clearly understood. EPRI has published a lessons learned document (EPRI Report 1001452, Generic Qualification of Commercial Grade Digital Devices: Lessons Learned from Initial Pilots, which covers several different qualification areas, including device selection, project planning, vendor surveys and design reviews, and electromagnetic compatibility (EMC) qualification. Application of the experience and lessons learned from the EPRI pilot activities should help reduce the effort and cost required for future qualification work. Most generic qualification activities for commercial equipment have been conducted using the approach of EPRI TR-106439, Guideline on Evaluation and Acceptance

Upgrading Planning and Executive Strategy for Reactor Protection System and Relative Equipment in Qinshan Nuclear Power Plant

International Nuclear Information System (INIS)

Jiang Zuyue

2010-01-01

Qinshan Nuclear Power Plant (QNPP) is the first nuclear power plant in China which completed the reactor protection system (RPS) upgrading with new digital safety instrumentation and control (I and C) platform instead of original analog system. At the same time,the nuclear instrumentation system (NIS) was upgraded with the same digital I and C platform. For adapting QNPP's actual engineering situation,the upgrading planning was taken by comprehensively investigating current development and application of digital safety I and C platform in the worldwide scope and by reviewing plant's original systems operation history. The project executive strategy-QNPP's leading role with necessary overseas cooperation and internal technical supports as great as possible, was determined. Some significant factors might influence and restrict the RPS and relative equipment upgrading executive actions in an operating NPP were analyzed.Finally, the engineering feasibility was briefly assessed to recognize the anticipated issues and difficulties and to prepare the relative solutions in advance for the purpose of ensuring the RPS upgrading objectives completely realized. (authors)

Assets optimization at Heavy Water Plants

International Nuclear Information System (INIS)

Hiremath, S.C.

2006-01-01

In the world where the fittest can only survive, manufacturing and production enterprises are under intense pressure to achieve maximum efficiency in each and every field related to the ultimate production of plant. The winners will be those that use their assets, i.e men, material, machinery and money most effectively. The objective is to optimize the utilization of all plant assets-from entire process lines to individual pressure vessels, piping, process machinery, and vital machine components. Assets of Heavy Water Plants mainly consist of Civil Structures, Equipment and Systems (Mechanical, Electrical) and Resources like Water, Energy and Environment

2011 Plant Lipids: Structure, Metabolism, & Function Gordon Research Conference

Energy Technology Data Exchange (ETDEWEB)

Christopher Benning

2011-02-04

This is the second Gordon Research Conference on 'Plant Lipids: Structure, Metabolism & Function'. It covers current topics in lipid structure, metabolism and function in eukaryotic photosynthetic organisms including seed plants, algae, mosses and ferns. Work in photosynthetic bacteria is considered as well as it serves the understanding of specific aspects of lipid metabolism in plants. Breakthroughs are discussed in research on plant lipids as diverse as glycerolipids, sphingolipids, lipids of the cell surface, isoprenoids, fatty acids and their derivatives. The program covers nine concepts at the forefront of research under which afore mentioned plant lipid classes are discussed. The goal is to integrate areas such as lipid signaling, basic lipid metabolism, membrane function, lipid analysis, and lipid engineering to achieve a high level of stimulating interaction among diverse researchers with interests in plant lipids. One Emphasis is on the dynamics and regulation of lipid metabolism during plant cell development and in response to environmental factors.

Plant Life Management of the EC6 Concrete Containment Structure

Energy Technology Data Exchange (ETDEWEB)

Abrishami, Homayoun; Ricciuti, Rick; Khan, Azhar [CANDU Energy Inc., Mississauga (Canada)

2012-03-15

Aging of reinforced concrete structures due to service conditions, aggressive environments, or accidents may cause their strength, serviceability and durability to decrease over time. Due to the complex nature of safety-related structures in nuclear power plants in comparison to other structures, they possess a number of characteristics that make them comparison to other structures, they possess a number of characteristics that make them unique. These characteristics are: thick concrete cross-sections, heavy reinforcement, often one-side access only, subjected to such ageing stresses as irradiation and elevated temperature, in addition to other typical ageing mechanisms (i. e., exposure to freeze/thaw cycles, aggressive chemicals, etc.) that typically affects other types of non-nuclear structures. For a new plant, the Plant Life Management Program (PLiM) should start in the design process and then continues through construction, plant operation and decommissioning. Hence PLiM must provide not only Ageing Management program (AMP) but also provide requirements on material characteristic and the design criteria as well. The purpose of this paper is to present the Plant Life Management (PLiM) strategy for the concrete containment structure of EC6 (Enhanced CANDU 6) Nuclear Power Plant designed by CANDU Energy Inc. The EC6 is designed for 100-year plant life including a 60-year operating life and an additional 40-year decommissioning period of time. The approach adopted for the PLiM strategy of the concrete containment structure is a preventive one, key areas being: 1) design methodology, 2) material performance and 3) life cycle management and ageing management program. In addition to strength and serviceability, durability is a major consideration during the design phase, service life and up to the completion of decommissioning. Factors affecting durability design include: a) concrete performance, b) structural application, and c) consideration of environmental

Plant Life Management of the EC6 Concrete Containment Structure

International Nuclear Information System (INIS)

Abrishami, Homayoun; Ricciuti, Rick; Khan, Azhar

2012-01-01

Aging of reinforced concrete structures due to service conditions, aggressive environments, or accidents may cause their strength, serviceability and durability to decrease over time. Due to the complex nature of safety-related structures in nuclear power plants in comparison to other structures, they possess a number of characteristics that make them comparison to other structures, they possess a number of characteristics that make them unique. These characteristics are: thick concrete cross-sections, heavy reinforcement, often one-side access only, subjected to such ageing stresses as irradiation and elevated temperature, in addition to other typical ageing mechanisms (i. e., exposure to freeze/thaw cycles, aggressive chemicals, etc.) that typically affects other types of non-nuclear structures. For a new plant, the Plant Life Management Program (PLiM) should start in the design process and then continues through construction, plant operation and decommissioning. Hence PLiM must provide not only Ageing Management program (AMP) but also provide requirements on material characteristic and the design criteria as well. The purpose of this paper is to present the Plant Life Management (PLiM) strategy for the concrete containment structure of EC6 (Enhanced CANDU 6) Nuclear Power Plant designed by CANDU Energy Inc. The EC6 is designed for 100-year plant life including a 60-year operating life and an additional 40-year decommissioning period of time. The approach adopted for the PLiM strategy of the concrete containment structure is a preventive one, key areas being: 1) design methodology, 2) material performance and 3) life cycle management and ageing management program. In addition to strength and serviceability, durability is a major consideration during the design phase, service life and up to the completion of decommissioning. Factors affecting durability design include: a) concrete performance, b) structural application, and c) consideration of environmental

FY 1990 report on the results of the development of the entrained bed coal gasification power plant. Part 3. Fabrication/installation of pilot plant (Fabrication/installation drawings and fabrication/installation pictures - 2/2); 1990 nendo seika hokokusho. Funryusho sekitan gaska hatsuden plant kaihatsu - Sono 3. Pilot plant seisaku suetsuke hen (Seisaku suetsukezu oyobi seisaku suetsuke shashin) (2/2)

Energy Technology Data Exchange (ETDEWEB)

NONE

1991-03-01

For the purpose of establishing the technology of the integrated coal gasification combined cycle power generation, the fabrication, installation work, etc. were conducted of a 200t/d entrained bed coal gasification pilot plant, and fabrication/installation drawings and fabrication/installation pictures were summarized. In fabrication/installation drawings, drawings of the following were included: actual-pressure/actual-size combustor test equipment (structural drawing of exhaust temperature reducing device, structural drawing of hot air device, system diagram of piping, etc.), safety environmental equipment (total system diagram, layout of electric room equipment, layout of control equipment room, etc.), total control system (structural drawing of the total control system, front view of auxiliary panel of safety environment equipment, etc.), 66kV/6.9kV indoor switching station facilities (layout of equipment of indoor switching station facilities, outline drawing of the main transformer, outline drawing of gas circuit breaker, etc.), common facilities (total layout, diagram of the nitrogen gas pipe system, system diagram of utility equipment, etc.) In pictures of fabrication/installation, pictures of the following were included: state of the construction work, gasifier equipment, gas refining facilities, gas turbine facilities, actual-pressure/actual-size combustor test equipment, safety environmental equipment, total control system, 66kV/6.9kV indoor switching station facilities, common facilities. (NEDO)

Behavior of Equipment Support Beam Joint Directly Connected to A Steel-plate Concrete(SC) Wall

International Nuclear Information System (INIS)

Kim, K. S.; Kwon, K. J.

2008-01-01

To decrease the time for building nuclear power plants, a modular construction method, 'Steel-plate Concrete(SC)', has been investigated for over a decade. To construct a SC wall, a pair of steel plates are placed in parallel similar to a form-work in conventional reinforced concrete (RC) structures, and concrete is filled between the steel plates. Instead of removing the steel plates after the concrete has cured, the steel plates serve as components of the structural member. The exposed steel plate of SC structures serves as the base plate for the equipment support, and the headed studs welded to the steel plates are used as anchor bolts. Then, a support beam can be directly welded to the surface of the steel plate in any preferred position. In this study, we discuss the behavior and evaluation method of the equipment support joint directly connected to exposed steel plate of SC wall

Implications of NUREG-1150 for equipment qualification

International Nuclear Information System (INIS)

Siu, R.P.; Sarram, M.

1988-01-01

Current USNRC regulations in 10CFR50 requires that structures, systems and components important to safety in a nuclear power plant are to be designed to accommodate the effects of harsh environments associated with potential accidents. Detailed regulatory requirements are as discussed in Regulatory Guide 1.89. Some of these regulatory requirements have been based upon rather outdated understanding of the phenomenology of severe accidents -- e.g., the radiation source terms for equipment qualification can be traced to TID-14844 which was originally published in 1962. Furthermore, the definition of important-to-safety in these regulatory requirements is rather broad, usually resulting in a long Q-list of equipment to be qualified for the design-basis Large LOCA. In the past few years, extensive research/analysis on severe accidents have been conducted -- from the IREP/NREP activities, the IDCOR programs, to the NUREG-1150 efforts. This paper reports that as a result of all these PRA-related activities, a much-improved understanding of the phenomenology and environmental conditions associated with severe accidents has been gained

Assessment of inservice conditions of safety-related nuclear plant structures

International Nuclear Information System (INIS)

Ashar, H.; Bagchi, G.

1995-06-01

The report is a compilation from a number of sources of information related to the condition Of structures and civil engineering features at operating nuclear power plants in the United States. The most significant information came from the hands-on inspection of the six old plants (licensed prior to 1977) performed by the staff of the Civil Engineering and Geosciences Branch (ECGB) in the Division of Engineering of the Office of Nuclear Reactor Regulation. For the containment structures, most of the information related to the degraded conditions came from the licensees as part of the Licensing Event Report System (10 CFR 50.73), or as part of the requirement under limiting condition of operation of the plant-specific Technical Specifications. Most of the information related to the degradation of other Structures and civil engineering features was extracted from the industry survey, the reported incidents, and the plant visits. The report discusses the condition of the structures and civil engineering features at operating nuclear power plants and provides information that would help detect, alleviate, and correct the degraded conditions of the structures and civil engineering features

Assessment of the performance of containment and surveillance equipment part 1: methodology

International Nuclear Information System (INIS)

Rezniczek, A.; Richter, B.

2009-01-01

Equipment performance aims at the creation of relevant data. As Containment and Surveillance (C/S) is playing an ever increasing role in safeguards systems, the issue of how to assess the performance of C/S equipment is being addressed by the ESARDA Working Group on C/S. The issue is important not only for the development of appropriate safeguards approaches but also for the review of existing approaches with regard to the implementation of the Additional Protocol (A P) and Integrated Safeguards. It is expected that the selection process of appropriate equipment, especially for unattended operation, is facilitated by the availability of methods to determine the performance of such equipment. Apart from EURATOM, the users of assessment methodologies would be the International Atomic Energy Agency (IAEA), plant operators, and instrument developers. The paper describes a non-quantitative performance assessment methodology. A structured procedure is outlined that allows assessing the suitability of different C/S instrumentation to comply with the objectives of its application. The principle to determine the performance of C/S equipment is to define, based on safeguards requirements, a task profile and to check the performance profile against the task profile. The performance profile of C/S equipment can be derived from the functional specifications and design basis tolerances provided by the equipment manufacturers.

Chlorine dioxide as biocide to prevent biofouling in the hydro technical structures at KKNPP

International Nuclear Information System (INIS)

Ganesh, S.; Selvaraj, S.; Balasubramanian, M.R.; Selvavinayagam, P.; Sundar, R.S.

2008-01-01

Chlorination is envisaged in the sea water systems of KKNPP to control macro and micro bio-fouling of underwater structures and equipments. KKNPP intake and the fore bay structures are shown in detail. The sodium hypo chlorite required for chlorination is produced in the electro chlorination plant at site by the electrolysis of sea water. It is added in the sea water at the intake structure, tunnels and fore bay on continuous as well as periodic basis. The sea water to chlorination plant is supplied by the pumps located at the main pump house. Chlorination of sea water system by electro-chlorination is possible only after pump house flooding and commissioning of electro-chlorination plant. So for the period from breach of temporary dyke till commissioning of electro chlorination plant, chlorination by temporary method has to be done to prevent the bio-fouling of underwater structures and equipments. The flooding of the pump house subsequent to breach of temporary dyke is done

Non-structural carbohydrates in woody plants compared among laboratories

NARCIS (Netherlands)

Quentin, Audrey G.; Pinkard, Elizabeth A.; Ryan, Michael G.; Tissue, David T.; Baggett, Scott L.; Adams, Henry D.; Maillard, Pascale; Marchand, Jacqueline; Landhäusser, Simon M.; Lacointe, André; Gibon, Yves; Anderegg, William R.L.; Asao, Shinichi; Atkin, Owen K.; Bonhomme, Marc; Claye, Caroline; Chow, Pak S.; Clément-Vidal, Anne; Davies, Noel W.; Dickman, Turin L.; Dumbur, Rita; Ellsworth, David S.; Falk, Kristen; Galiano, Lucía; Grünzweig, José M.; Hartmann, Henrik; Hoch, Günter; Hood, Sharon; Jones, Joanna E.; Koike, Takayoshi; Kuhlmann, Iris; Lloret, Francisco; Maestro, Melchor; Mansfield, Shawn D.; Martínez-Vilalta, Jordi; Maucourt, Mickael; McDowell, Nathan G.; Moing, Annick; Muller, Bertrand; Nebauer, Sergio G.; Niinemets, Ülo; Palacio, Sara; Piper, Frida; Raveh, Eran; Richter, Andreas; Rolland, Gaëlle; Rosas, Teresa; Joanis, Brigitte Saint; Sala, Anna; Smith, Renee A.; Sterck, Frank; Stinziano, Joseph R.; Tobias, Mari; Unda, Faride; Watanabe, Makoto; Way, Danielle A.; Weerasinghe, Lasantha K.; Wild, Birgit; Wiley, Erin; Woodruff, David R.

2015-01-01

Non-structural carbohydrates (NSC) in plant tissue are frequently quantified to make inferences about plant responses to environmental conditions. Laboratories publishing estimates of NSC of woody plants use many different methods to evaluate NSC. We asked whether NSC estimates in the recent

Seismic reevaluation of existing nuclear power plants

International Nuclear Information System (INIS)

Hennart, J.C.

1978-01-01

The codes and regulations governing Nuclear Power Plant seismic analysis are continuously becoming more stringent. In addition, design ground accelerations of existing plants must sometimes be increased as a result of discovery of faulting zones or recording of recent earthquakes near the plant location after plant design. These new factors can result in augmented seismic design criteria. Seismic reanalysius of the existing Nuclear Power Plant structures and equipments is necessary to prevent the consequences of newly postulated accidents that could cause undue risk to the health or safety of the public. This paper reviews the developments of seismic analysis as applied to Nuclear Power Plants and the methods used by Westinghouse to requalify existing plants to the most recent safety requirements. (author)

Aging management of containment structures in nuclear power plants

International Nuclear Information System (INIS)

Naus, D.J.; Oland, C.B.; Ellingwood, B.R.; Graves, H.L. III; Norris, W.E.

1996-01-01

Research is being conducted by Oak Ridge National Laboratory under US nuclear regulatory commission (USNRC) sponsorship to address aging management of nuclear power plant containment and other safety-related structures. Documentation is being prepared to provide the USNRC with potential structural safety issues and acceptance criteria for use in continued service evaluations of nuclear power plants. Accomplishments include development of a structural materials information center containing data and information on the time variation of 144 material properties under the influence of pertinent environmental stressors or aging factors, evaluation of models for potential concrete containment degradation factors, development of a procedure to identify critical structures and degradation factors important to aging management, evaluations of non-destructive evaluation techniques, assessments of European and North American repair practices for concrete, review of parameters affecting corrosion of metals embedded in concrete, and development of methodologies for making current condition assessments and service life predictions of new or existing reinforced concrete structures in nuclear power plants. (orig.)

Aging management of containment structures in nuclear power plants

International Nuclear Information System (INIS)

Naus, D.J.; Oland, C.B.; Ellingwood, B.R.

1994-01-01

Research is being conducted by Oak Ridge National Laboratory under U.S. Nuclear Regulatory Commission sponsorship to address aging management of nuclear power plant containment and other safety-related structures. Documentation is being prepared to provide the US-NRC with potential structural safety issues and acceptance criteria for use in continued service evaluations of nuclear power plants. Accomplishments include development of a Structural Materials Information Center containing data and information on the time variation of 144 material properties under the influence of pertinent environmental stressors or aging factors, evaluation of models for potential concrete containment degradation factors, development of a procedure to identify critical structures and degradation factors important to aging management, evaluations of nondestructive evaluation techniques, assessments of European and North American repair practices for concrete, review of parameters affecting corrosion of metals embedded in concrete, and development of methodologies for making current condition assessments and service life predictions of new or existing reinforced concrete structures in nuclear power plants. (author). 29 refs., 2 figs

Aging management of containment structures in nuclear power plants

International Nuclear Information System (INIS)

Naus, D.J.; Oland, C.B.; Ellingwood, B.R.; Graves, H.L. III; Norris, W.E.

1994-01-01

Research is being conducted by ORNL under US Nuclear Regulatory Commission (USNRC) sponsorship to address aging management of nuclear power plant containment and other safety-related structures. Documentation is being prepared to provide the USNRC with potential structural safety issues and acceptance criteria for use in continued service evaluations of nuclear power plants. Accomplishments include development of a Structural Materials Information Center containing data and information on the time variation of 144 material properties under the influence of pertinent environmental stressors or aging factors, evaluation of models for potential concrete containment degradation factors, development of a procedure to identify critical structures and degradation factors important to aging management, evaluations of nondestructive evaluation techniques. assessments of European and North American repair practices for concrete, review of parameters affecting corrosion of metals embedded in concrete, and development of methodologies for making current condition assessments and service life predictions of new or existing reinforced concrete structures in nuclear power plants

Dynamic testing of nuclear power plant structures: an evaluation

International Nuclear Information System (INIS)

Weaver, H.J.

1980-02-01

Lawrence Livermore Laboratory (LLL) evaluated the applications of system identification techniques to the dynamic testing of nuclear power plant structures and subsystems. These experimental techniques involve exciting a structure and measuring, digitizing, and processing the time-history motions that result. The data can be compared to parameters calculated using finite element or other models of the test systems to validate the model and to verify the seismic analysis. This report summarizes work in three main areas: (1) analytical qualification of a set of computer programs developed at LLL to extract model parameters from the time histories; (2) examination of the feasibility of safely exciting nuclear power plant structures and accurately recording the resulting time-history motions; (3) study of how the model parameters that are extracted from the data be used best to evaluate structural integrity and analyze nuclear power plants

Paradigms of structural safety of AGED plants. Lessons learned from Russian activities

International Nuclear Information System (INIS)

Saji, Genn; Timofeev, Boris

2007-01-01

The study of the effects behind the degradation of components and material is becoming increasingly important for the safe operation of aged plants especially when it comes to life-extension. Since the Russian nuclear community began to examine life extension issues nearly fifteen years ago, there is much to learn from these Russian pioneering studies, a portion of which were performed under the TACIS (Technical Assistance for Commonwealth of Independent States) international collaboration program with EU countries. At the Ninth International Conference, recent data were introduced regarding the ageing effects of mechanical properties of various kinds of steel and the welding joints of Russian NPP components. The full title of the conference was Material Issues in Design, Manufacturing and Operation of Nuclear Power Plants Equipment. The meeting was organized by the Central Research Institute of Structural Materials (CRISM) 'Prometey' in cooperation with the IAEA and JRC-EU. In reviewing the recent data presented at the Ninth Conference, the authors think that the paradigms of structural integrity issues in aged plants are now reasonably well established in (1) fracture mechanics and irradiation hardening of reactor vessels and core internals and (2) thermal ageing and annealing effects. Yet even when considering these well established paradigms, the current direction of study is not adequately addressing the effects of a corrosive environment. The first author believes that the current approach of low cycle fatigue is far from able to prevent and predict environmentally assisted cracks. This fundamental flaw stems from design codes, which do not incorporate the basic knowledge of corrosion mechanisms. Our focus in researching aged plants should be re-directed toward environmentally assisted cracking, typically the film rupture-slip dissolution mechanism of crack propagation under the effect of long cell action on local cells, as discussed by the first author in

Seismic verification methods for structures and equipment of VVER-type and RBMK-type NPPs (summary of experiences)

International Nuclear Information System (INIS)

Masopust, R.

2003-01-01

The main verification methods for structures and equipment of already existing VVER-type and RBMK-type NPPs are briefly described. The following aspects are discussed: fundamental seismic safety assessment principles for VVER/RBMK-type NPPs (seismic safety assessment procedure, typical work plan for seismic safety assessment of existing NPPs, SMA (HCLPF) calculations, modified GIP (GIP-VVER) procedure, similarity of VVER/RBMK equipment to that included in the SQUG databases and seismic interactions

Aging of concrete containment structures in nuclear power plants

International Nuclear Information System (INIS)

Naus, D.J.; Oland, C.B.; Ellingwood, B.; Mori, Yasuhiro; Arndt, E.G.

1992-01-01

Concrete structures play a vital role in the safe operation of all light-water reactor plants in the US Pertinent concrete structures are described in terms of their importance design, considerations, and materials of construction. Degradation factors which can potentially impact the ability of these structures to meet their functional and performance requirements are identified. Current inservice inspection requirements for concrete containments are summarized. A review of the performance history of the concrete components in nuclear power plants is provided. A summary is presented. A summary is presented of the Structural Aging (SAG) Program being conducted at the Oak Ridge National Laboratory for the US Nuclear Regulatory Commission. The SAG Program is addressing the aging management of safety-related concrete structures in nuclear power plants for the purpose of providing improved bases for their continued service. The program consists of a management task and three technical tasks: materials property data base, structural component assessment/repair technologies, and quantitiative methodology for continued service conditions. Objectives and a summary of accomplishments under each of these tasks are presented

Development of Advanced Concept for Shortening Construction Period of ABWR Plant

International Nuclear Information System (INIS)

Hiroshi Ijichi; Toshio Yamashita; Masahiro Tsutagawa; Hiroya Mori; Nobuaki Ooshima; Jun Miura; Minoru Kanechika; Nobuaki Miura

2002-01-01

Construction of a nuclear power plant (NPP) requires a very long period because of large amount of construction materials and many issues for negotiation among multiple sections. Shortening the construction period advances the date of return on an investment, and can also result in reduced construction cost. Therefore, the study of this subject has a very high priority for utilities. We achieved a construction period of 37 months from the first concrete work to fuel loading (F/L) (51.5 months from the inspection of the foundation (I/F) to the start of commercial operation (C/O)) at the Kashiwazaki-Kariwa NPPs No. 6 and 7 (KK-6/7), which are the first ABWR plants in the world. At TEPCO's next plant, we think that a construction period of less than 36 months (45 months from I/F to C/O) can be realized based on conventional methods such as early start of equipment installation and blocking of equipment to be brought in advance. Furthermore, we are studying the feasibility of a 21.5-month construction period (30 months from I/F to C/O) with advanced ideas and methods. The important concepts for a 21.5-month construction period are adoption of a new building structure that is the steel plate reinforced concrete (SC) structure and promotion of extensive modularization of equipment and building structure. With introducing these new concepts, we are planning the master schedule (M/S) and finding solutions to conflicts in the schedule of area release from building construction work to equipment installation work (schedule-conflicts.) In this report, we present the shortest construction period and an effective method to put it into practice for the conventional general arrangement (GA) of ABWR. In the future, we will continue the study on the improvement of building configuration and arrangements, and make clear of the concept for large composite modules of building structures and equipment. (authors)

Experience in the manufacture of nuclear equipment in India

International Nuclear Information System (INIS)

Challappa, S.; Murthy, G.S.K.; Mehta, S.K.; Kakodkar, A.; Natarajan, A.

1977-01-01

Department of Atomic Energy with its programme for achieving self-sufficiency was involved in engineering, manufacture, inspection, performance testing and quality surveillance of major precision and critical equipment such as reactor vessels, shields, fuelling machines, coolant channel components etc. etc. high pressure equipment for Heavy Water Plants, specialized components for Fuel Complex, major equipment for Cyclotron Project and various research projects. These had to be manufactured at various shops in the country depending upon the availability of machines. The relative importance of various important parameters associated with the manufacture of this equipment were assessed in a separate R and D programme. This has helped in re-designing in some areas to suit the manufacture under Indian conditions. Assessment of any marginal variations that take place during manufacture was also possible because of the availability of data of this kind. Critical components and equipment are tested for their performance under simulated conditions before shipments. B.A.R.C. has contributed immensely in achieving the self-sufficiency and also for designs for future plants

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)

Prophylactic and thermovision measurements of electric machines and equipment

Energy Technology Data Exchange (ETDEWEB)

Jedlicka, R; Brestovansky, L [Atomova Elektraren Bohunice, Jaslovske Bohunice (Slovakia)

1997-12-31

High-voltage measurements of generators, unit and service transformers and some significant motor drives used at a nuclear power plant are described in this paper. Thermovision measurements of electric machines and distribution systems are dealt with in the second part of the paper. Power electric equipment represent one of the most significant components of a nuclear power plant. Turbine mechanical energy is converted into the electrical energy within these equipment. Power generated by generators is transformed by transformers so that it can achieve appropriate parameters for both the transmission over the distribution system and the power plant service power supply. The service power supply switchboards and cables provide supply to motors and other consumers necessary for the nuclear power plant technological process. The whole complex of equipment has to be maintained in good technical conditions. It is necessary to make thermovision and prophylactic measurements to identify and verify the electric equipment technical condition. The mentioned measurements warn the operation staff in advance against both gradual deterioration of power connection contact resistances, i.e. power connections overheating, and the machine insulation systems condition deterioration. The operation staff try to prevent the electric equipment operation accidents by early removing the detected failures, thus, improving the nuclear safety. In order to provide the above-mentioned activities a special prophylactic measurement group was established at the NPP Bohunice in 1983. The group specialists make following types of measurements. 1. Prophylactic measurements of electric machines. Prophylactics of 220 MW generators and 6 MW service power generators; Prophylactics of both unit and service transformers and VHV bushings; Prophylactics of major 6 kV motor drives. 2. Thermovision measurements of current connections. (Abstract Truncated)

Plant control device

International Nuclear Information System (INIS)

Sato, Masuo; Ono, Makoto.

1995-01-01

A plant control device comprises an intellectual instrumentation group for measuring a predetermined process amount, an intellectual equipment group operating in accordance with a self-countermeasure, a system information space for outputting system information, a system level monitoring and diagnosing information generalization section for outputting system information, a system level maintenance information generalization section for outputting information concerning maintenance, a plant level information space and a plant level information generalization section. Each of them determines a state of the plant autonomously, and when abnormality is detected, each of the intellectual instrumentation, equipments and systems exchange information with each other, to conduct required operations including operations of intellectual robots, as required. Appropriate countermeasures for gauges, equipments and systems can be conducted autonomously at a place where operators can not access to improve reliability of complicate operations in the working site, as well as improve plant safety and reliability. (N.H.)

Esau's Plant anatomy: meristems, cells, and tissues of the plant body : their structure, function, and development

National Research Council Canada - National Science Library

Evert, Ray Franklin; Esau, Katherine; Eichhorn, Susan E

2006-01-01

... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . xix Chapter 1 Structure and Development of the Plant Body- An Overview . . . . . . . . . . . . . . . . . . . . . . . . 1 Internal Organization of the Plant Body...

Let a sewage plant running smart

Science.gov (United States)

Yang, Shan-Shan; Pang, Ji-Wei; Jin, Xiao-Man; Wu, Zhong-Yang; Yang, Xiao-Yin; Guo, Wan-Qian; Zhao, Zhi-Qing; Ren, Nan-Qi

2018-03-01

Out-dated technical equipment, occlusive information communication, inadequate sanitation, low management level and some irrational distribution structures in the existing sewage plants bring about lower sewage treatment efficiency and poorer water quality, thereby permanently harming human health and severely damaging the environment. With the rapid development of scientific-technological progress and the vigorous support of the entire international community, the existing sewage plants call for more and more intelligent operation and management in the future. This review for the first time proposes the novel concept of the “smart” sewage plant, and gives a through interpretation of its special functions and attributes. We envision that the future smart sewage plant will became an “ambient intelligence” in all aspects in the sewage plants.

Equipment nonconformance and degradation: Promptly determining operability and establishing corrective action plans

International Nuclear Information System (INIS)

Hoxie, C.L.; Cotton, K.R.; Emch, R.L. Jr.